Text

Appendices a, B and C
	ML11175A207
Person / Time
Site:	Salem
Issue date:	06/30/2011
From:	; Office of Nuclear Reactor Regulation
To:	;
	CuadradoDeJesus S
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ML11136A134	List: ML11136A064; ML11139A343; ML11168A073; ML11168A074; ML11175A207;
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V LICENSE AUTHORITY FILE OR D0 ioT REMOVE afar c~~~~, g7-/.3S 74 SALEM NUCLEAR GENERATING STATION UNIT 1 TECHNICAL SPECIFICATIONS APPENDIX "A" TO LICENSE NO. DPR-70 K)

I

INDEX DEFINITIONS SECTION PAGE 1.0 DEFINITIONS DEFINED TERMS................................................... 1-1 ACTION .......................................................... 1-1 AXIAL FLUX DIFFERENCE ........................................... 1-1 CHANNEL CALIBRATION............................................. 1-1 CHANNEL CHECK................................................... 1-1 CHANNEL FUNCTIONAL TEST ......................................... 1-1 CONTAINMENT INTEGRITY .......................................... 1-2 CORE ALTERATION ................................................. 1-2 CORE OPERATING LIMITS REPORT ................................... 1-2 DOSE EQUIVALENT I-131 ............................... 1-2 E-AVERAGE DISINTEGRATION ENERGY ................................. 1-3 ENGINEERED SAFETY FEATURE RESPONSE TIME ......................... 1-3 FREQUENCY NOTATION. 1-3 FULLY WITHDRAWN ............................... 1-3 GASEOUS RADWASTE TREATMENT SYSTEM............................... 1-3 IDENTIFIED LEAKAGE .............................................. 1-3 MEMBER(S) OF THE PUBLIC ......................................... 1-4 OFFSITE DOSE CALCULATION MANUAL (ODCM) .......................... 1-4 OPERABLE - OPERABILITY .......................................... 1-4 OPERATIONAL MODE - MODE .........................................

PHYSICS TESTS ...................................................

1-4

.1-5 I

PRESSURE BOUNDARY LEAKAGE ....................................... 1-5 PROCESS CONTROL PROGRAM (PCP) ................................... 1-5 PURGE-PURGING................................................... 1-5 QUADRANT POWER TILT RATIO ....................................... 1-5 RATED THERMAL POWER............................................. 1-5 REACTOR TRIP SYSTEM RESPONSE TIME ............................... 1-6 REPORTABLE EVENT ................................................ 1-6 SHUTDOWN MARGIN ................................................. 1-6 SITE BOUNDARY ................................................... 1-6 SOLIDIFICATION .................................................. 1-6 SOURCE CHECK.................................................... 1-6 STAGGERED TEST BASIS............................................ 1-6 THERMAL POWER ................................................... 1-7 UNIDENTIFIED LEAKAGE ............................................ 1-7 UNRESTRICTED AREA ....................; 1-7 VENTILATION EXHAUST TREATMENT SYSTEM............................ 1-7 VENTING ......................................................... 1-7 225 SALEM - UNIT 1 I Amendment No.

NOV 0 2 1999

INDEX SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SECTION PAGE 2.1 SAFETY LIMITS Reactor Core . . . . . . . . . . . . . . . . . . . . . . . 2-1 Reactor Coolant System Pressure . . . . . . . . . . . . . 2-3 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Trip System Instrumentation Setpoints . . . . . . 2-4 BASES SECTION PAGE 2.1 SAFETY LIMITS Reactor Core . . . . . . . . . . . . . . . . . . . . . . . B 2-1 Reactor Coolant System Pressure . . . . . . . . . . . . . B 2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Trip System Instrumentation Setpoints . . . . . . B 2-3 SALEM - UNIT I II Amendment No. 201 N0V 2 6 1997-ye6 -1)l~ ,40 6C

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.0 APPLICABILITY 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL Shutdown Margin - Tavg > 200OF 3/4 1-1 Shutdown Margin - Tavg < 2001F 3/4 1-3 I

Moderator Temperature Coefficient 3/4 1-5 Minimum Temperature for Criticality 3/4 1-6 3/4.1.2 BORATION SYSTEMS Flow Paths - Shutdown 3/4 1-7 Flow Paths - Operating 3/4 1-8 Charging Pump - Shutdown 3/4 1-10 Charging Pump - Operating 3/4 1-11 I Borated Water Sources - Shutdown 3/4 1-14 Borated Water Sources - Operating 3/4 1-16 3/4.1.3 MOVABLE CONTROL ASSEMBLIES Group Height 3/4 1-18 Position Indicating System - Operating 3/4 1-19 Position Indicating Systems - Shutdown 3/4 1-20 Rod Drop Time 3/4 1-21 Shutdown Rod Insertion Limit 3/4 1-22 Control Rod Insertion Limits 3/4 1-23 SALEM - UNIT 1 III Amendment No. 225 NOV 0 2 1399

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2:1 AXIAL FLUX DIFFERENCE 3/4 2-1 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR 3/4 2-5 3/4.2.3 NUCLEAR ENTHALPY HOT CHANNEL FACTOR 3/4 2-9 3/4.2.4 QUADRANT POWER TILT RATIO 3/4 2-11 3/4.2.5 DNB PARAMETERS 3/4 2-13 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP.SYSTEM INSTRUMENTATION 3/4 3-1 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION 3/4 3-14 3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation 3/4 3-35 Remote Shutdown Instrumentation 3/4 3-46 Accident Monitoring Instrumentation 3/4 3-53 Radioactive Liquid Effluent Monitoring Instrumentation 3/4 3-58 Power Distribution Monitoring System 3/4 3-70 3/4.3.4 DELETED SALEM - UNIT 1 IV Amendment No. 282

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS Normal Operation 3/44-1 Hot Standby 3/44-2 Hot Shutdown 3/4 4-3 Cold Shutdown 3/4 4-3b 3/4.4.2.1 SAFETY VALVES - SHUTDOWN 3/4 4-4 3/4.4.2.2 SAFETY VALVES - OPERATING 3/4 4-4a 3/4.4.3 RELIEF VALVES 3/44-5 3/4.4.4 PRESSURIZER 3/44-6 3/4.4.5 STEAM GENERATOR (SG) TUBE INTEGRITy 3/4 4-7 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection System 3/4 4-14 Operational Leakage 3/4 4-15 Primary Coolant System Pressure Isolation Valves 3/4 4-16a 3/4.4.7 DELETED 3/4.4.8 SPECIFIC ACTiViTy 3/4 4-20 3/4.4.9 PRESSURE/TEMPERATURE LIMITS Reactor Coolant System 3/4 4-24 Pressurizer 3/4 4-29 Overpressure Protection Systems 3/4 4-30 3/4.4.10 DELETED 3/4 4-32 3/4.4.11 INTENTIONALLY BLANK 3/44-34 3/4.4.12 HEAD VENTS 3/4 4-35 SALEM - UNIT 1 V Amendment No. 298

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 ACCUMULATORS .................. ......................... 3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - T avg Ž 350cF... ......................... 3/4 5-3 3/4.5.3 ECCS SUBSYSTEMS - Tavg < 350FF... ......................... 3/4 5-6 3/4.5.4 SEAL INJECTION FLOW ..................................... 3/4 5-6b 3/4.5.5 REFUELING WATER STORAGE TANK ............................ 3/4 5-7 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Containment Integrity ............... 3/4 6-1 Containment Leakage ................. 3/4 6-2 Containment Air Locks ............... .3/4 6-5 Internal Pressure ................... .3/4 6-6 Air Temperature ..................... .3/4 6-7 Containment Structural Integrity .... 3/4 6-8 Containment Ventilation System ...... 3/4 6-8a 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS Containment Spray System ............ . 3/4 6-9 Spray Additive System ............... .3/4 6-10 Containment Cooling System .......... .3/4 6-11 3/4.6.3 CONTAINMENT ISOLATION VALVES ............................. 3/4 6-12 SALEM - UNIT 1 VI Amendment No.281

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE Safety Valves . . . . . . . . . . . . . 7-1 .3/

Auxiliary Feedwater System . . . . . . 7-5 .3/4 Auxiliary Feed Storage Tank . . . . . . 3/4 7-7 ..

Activity . . . . . . . . . . . . . . . 3/4 7-8 ..

Main Steam Line Isolation Valves . . . 7-10 .3/

3/4.7.2 STEAM GENERATOR PRESSURE/TEMPERATURE LIMITATION . . . . . . . . . . . . . . 7-14 .3/4 3/4.7.3 COMPONENT COOLING WATER SYSTEM . . . . 7-15 .3/4 3/4.7.4 SERVICE WATER SYSTEM . . . . . . . . . 3/4 7-16 ..

3/4.7.5 FLOOD PROTECTION . . . . . . . . . . . 7-17 .3/4 3/4.7.6 CONTROL ROOM EMERGENCY AIR CONDITIONING SYSTEM . . . . . . . . . . 3/4 7-18 ..

3/4.7.7 AUXILIARY BUILDING VENTILATION SYSTEM 7-22 .3/4 I

3/4.7.8 SEALED SOURCE CONTAMINATION . . . . . . 3/4 7-26 ..

3/4.7.9 SNUBBERS . . . . . . . . . . . . . . . 7-28 .3/4 3/4.7.10 CHILLED WATER SYSTEM -

AUXILIARY BUILDING SUBSYSTEM. . . . . . 3/4 7-33 ..

3/4.7.11 FUEL STORAGE POOL BORON CONCENTRATION 3/4 7-35 ..

3/4.7.12 FUEL ASSEMBLY STORAGE IN THE SPENT FUEL POOL. . . .3/4 7-36 SALEM - UNIT 1 VII Amendment No. 271

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A. C. SOURCES Operating .......................................... 3/4 8-1 Shutdown .......................................... 3/4 8-5c I 3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEMS A.C. Distribution - Operating ..... ................ 3/4 8-6 A.C. Distribution - Shutdown ..... ................. 3/4 8-7 125-Volt D.C. Distribution - Operating .... ........ 3/4 8-8 125-Volt D.C. Distribution - Shutdown .... .......... 3/4 8-10 28-Volt D.C. Distribution - Operating .... ......... 3/4 8-11 28-Volt D.C. Distribution - Shutdown .... ......... 3/4 8-13 3/4.8.3 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES .... ....... 3/4 8-14 SALEM - UNIT 1 VIII Amendment No. 225 tlv 0 C) ,.39

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION 3/4 9-1 3/4.9.2 INSTRUMENTATION 3/4 9-2 3/4.9.3 DECAY TIME 3/4 9-3 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS 3/4 9-4 3/4.9.5 DELETED 3/4 9-5 3/4.9.6 DELETED 3/4 9-6 3/4.9.7 DELETED 3/4 9-7 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION All Water Levels 3/4 9-8 Low Water Level 3/4 9-8a 3/4.9.9 DELETED 3/4.9.10 WATER LEVEL - REACTOR VESSEL 3/4 9-10 3/4.9.11 STORAGE POOL WATER LEVEL 3/4 9-11 3/4.9.12 FUEL HANDLING AREA VENTILATION SYSTEM 3/4 9-12 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN . . . . 3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS 3/4 10-2 3/4.10.3 PHYSICS TESTS 3/4 10-3 3/4.10.4 NO FLOW TESTS 3/4 10-4 SALEM - UNIT 1 IX Amendment No. 293

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS Liquid Holdup Tanks 3/4 11-7 3/4.11.2 GASEOUS EFFLUENTS Explosive Gas Mixture 3/4 11-15 3/4.11.3 Deleted 3/4 11-17 3/4.11.4 Deleted 3/4 11-18 3/4.12 Deleted SALEM - UNIT 1 X Amendment No. 234

INDEX BASES SECTIDON PAGE 3/4.0 APPLICABILITY. ............. B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1 .1 BORATION CONTROL . . . . . . . . . . . . . . . B 3/4 1-1 3/4.1 .2 BORATION SYSTEMS . . . . . . . . . . . . . . . B 3/4 1-3 3/4.1 .3 MOVABLE CONTROL ASSEMBLIES . . . . . . . . . . B 3/4 1-4 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2 .1 AXIAL FLUX DIFFERENCE . . . . . . . ... . . . . B 3/4 2-1 1

3/4.2 .2 HEAT FLUX and AND 3/4.2 .3 NUCLEAR ENTHALPY HOT CHANNEL AND RADIAL PEAKING FACTORS . . . . . . . . . . . . .B 3/4 2-4 3/4.2, .4 QUADRANT POWER TILT RATIO . . . . . . . . . . . B 3/4 2-5 3/4.2. .5 DNB PARAMETERS . . . . . . . . . . . . . . . . B 3/4 2-6 I

SALEM - UNIT 1 XI Amendment No. 225 1'OV 0 2 12399

INDEX 8ASES SECTION PAGE 3/4.3 INSTRUMENTATION 3/4.3.1 PROTECTIVE AND 3/4.3.2 ENGINEERED SAFETY FEATURES (ESF)

INSTRUMENTATION 83/43-1 3/4.3.3 MONITORING INSTRUMENTATION 83/4 3-1a 3/4.3.4 TUR81NE OVERSPEED PROTECTION 83/43-4 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION 83/44-1 3/4.4.2 SAFETY VALVES 83/4 4-1a 3/4.4.3 RELIEF VALVES 83/44-1 a 3/4.4.4 PRESSURiZER 8 3/4 4-2 3/4.4.5 STEAM GENERATOR (SG) TU8E INTEGRITy 83/44-2 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 8 3/44-4a 3/4.4.7 DELETED 3/4.4.8 SPECIFIC ACTiViTy 83/44-5 3/4.4.9 PRESSUREITEMPERATURE LIMITS 8 3/44-6 3/4.4.10 DELETED 83/44-17 3/4.4.11 8LANK 83/44-17 3/4.4.12 REACTOR VESSEL HEAD VENTS 83/44-17 SALEM - UNIT 1 XII Amendment No. 298

INDEX BASES SECTION PAGE 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4'5.1 ACCUMULATORS ............ ................... B 3/4 5-1 3/4.5.2 and ECCS SUBSYSTEMS ........... ................ B 3/4 5-1 3/4.5.3 3/4.5.4 SEAL INJECTION FLOW ........... .............. B 3/4 5-2 3/4.5.5 REFUELING WATER STORAGE TANK (RWST) .... ...... B 3/4 5-2 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT ........... .............. B 3/4 6-1 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS ... ...... B 3/4 6-3 3/4.6.3 CONTAINMENT ISOLATION VALVES ...... .......... B 3/4 6-3 28 1 SALEM - UNIT 1 XiII Amendment No.

INDEX BASES SECTION PAGE 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE .B 3/4 7-1 3/4.7.2 STEAM GENERATOR PRESSURE/TEMPERATURE LIMITATION B 3/4 7-4 3/4.7.3 COMPONENT COOLING WATER SYSTEM . . . . . . . . B 3/4 7-4 3/4.7.4 SERVICE WATER SYSTEM . . . . . . . . . . . . . B 3/4 7-4 3/4.7.5 FLOOD PROTECTION . . . . . . . . . . . . . . . B 3/4 7-5 3/4.7.6 CONTROL ROOM EMERGENCY AIR CONDITIONING SYSTEM . . . . . . . . . . . . . . B 3/4 7-5 3/4.7.7 AUXILIARY BUILDING EXHAUST AIR FILTRATION SYSTEM . . . . . . . . . . . . . . . B 3/4 7-5c 3/4.7.8 SEALED SOURCE CONTAMINATION . . . . . . . . . . B 3/4 7-5c 3/4.7.9 SNUBBERS ................... B 3/4 7-6 3/4.7.10 CHILLED WATER SYSTEM -

AUXILIARY BUILDING SUBSYSTEM . . . . . . . . . .B 3/4 7-8 3/4.7.11 FUEL STORAGE POOL BORON CONCENTRATION . . . . . B 3/4 7-9 3/4.7.12 FUEL ASSEMBLY STORAGE IN THE SPENT FUEL POOL. . B 3/4 7-12 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A. C. SOURCES. . . . . . . . . . . . . . . . . B 3/4 8-1 3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEMS . . . . . . . B 3/4 8-1 3/4.8.3 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES . . . . B 3/4 8-4 SALEM - UNIT 1 XIV Amendment No. 262

INDEX BASES

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SECTIO:N PAGE 3/4.9 REFUELING OPERATIONS 3/4.9. 1 BORON CONCENTRATION . . . . . . . . . . . . . . B 3/4 9-1 3/4.9.:2 INSTRUMENTATION . . . . . . . . . . . . . . . . B 3/4 9-lb 3/4.9.:3 DECAY TIME . . . . . . . . . . . . . . . . . . B 3/4 9-lb 3/4.9.,4 CONTAINMENT BUILDING PENETRATIONS . . . . . . . B 3/4 9-ic 3/4.9.!5 COMMUNICATIONS . . . . . . . . . . . . . . . . B 3/4 9-3 3/4.9.'6 MANIPULATOR CRANE . . . . . . . . . . . . . . . B 3/4 9-3 3/4.9.'7 CRANE TRAVEL - SPENT FUEL STORAGE BUILDING . . B 3/4 9-3 3/4.9.4B RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION . B 3/4 9-3 3/4.9.!9 CONTAINMENT PURGE AND PRESSURE-VACUUM RELIEF ISOLATION SYSTEM . . . . . . . . . . . . . . . B 3/4 9-4 3/4.9.:10 WATER LEVEL - REACTOR VESSEL and AND 3/4.9.:11 STORAGE POOL . . . . . . . . . . . . . . . . . B 3/4 9-4 3/4.9.:12 FUEL HANDLING AREA VENTILATION SYSTEM . . . . . B 3/4 9-4 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10 .1 SHUTDOWN MARGIN . . . . . . . . . . . . . . . . B 3/4 10-1 3/4.10 .2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS . . . . . . . . . . . . . . B 3/4 10-1 3/4.10 .3 PHYSICS TESTS . . . . . . . . . . . . . . . . . B 3/4 10-1 3/4.10 .4 NO FLOW TESTS . . . . . . . . . . . . . . . . . B 3/4 10-1 SALEM - UNIT I XV Amendment No. 262

INDEX BASES SECTION PAGE 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS . . . . . . . . . . . . . . . .B 3/4 11-1 3/4.11.2 GASEOUS EFFLUENTS . .............. .B 3/4 11-3 3/4.11.3 Deleted . . . . . . . . . . . . . . . . . . . .B 3/4 11-6 3/4.11.4 Deleted . . . . . . . . . . . . . . . . . . . .B 3/4 11-7 3/4.12 Deleted SALEM - UNIT I XVI Amendment No.234

INDEX DESIGN FEATURES SECTION PAGE 5.1 SITE Exclusion Area ................................... .............. .

Low Population Zone .............................. .............. .

Unrestricted Areas for Radioactive Gaseous and Liquid Effluents ................................. , 1.

5.2 CONTAINMENT Configuration ............................................ 5-1 .

Design Pressure and Temperature .......................... 5-4 .

5.3 REACTOR CORE Fuel Assemblies .......................... ...................... 5-4 Control Rod Assemblies ................... ...................... 5-4 5.4 REACTOR COOLANT SYSTEM Design Pressure and Temperature . ................................ 5-4 5.5 METEOROLOGICAL TOWER LOCATION ................................... 5-5 I 5.6 FUEL STORAGE Criticality ................................... 5-5 Drainage ................................... 5-6<

a Capacity....................................................... 5-6e 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT .......... 5-6a .

SALEM - UNIT 1 XVII Amendment No. 269

INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6 .1 R E S PO N S IB I L I T Y ................................................................................................................................................ 6- 1 6.2 ORGANIZATION Onsite and Office Organizations ............................................................................................ 6-1 Fac i li t y S t a f f ......... .............................. ............. ... ............................................... 6- 2 Shift Techn ica l Adviso r ..................................................................................................................... 6-7 6.3 FACILITY STAFF QUALIFICATIONS ................................................................................................... 6-7

6. 4 T RA I NI N G .................................................................................................................................................................. 6 -7 6.5 REVIEW AND AUDIT. (THIS SECTION DELETED) ..................................................................... 6-8 6.6 REPORTA BLE EV ENT ACTION ...................................................................................................................... 6-16 6.7 SA FETY L IM IT V IO LA TION ........................................................................................................................ 6- 16 6.8 PROCEDURES AN D PROGRAM S .................................................................................................................... 6-17 6.9 REPORTING REQUIREMENTS 6.9.1 ROUTINE REPORTS ............................................................. 6-20 6 .9 .2 S PEC IA L R E PO R T S ............................................................................. .................................................... 6-2 4 6 .10 R EC OR D R E T EN T ION .......................................................................................................................................... 6-2 5 6.11 RADIATION PROTECTION PROGRAM ................................................................................................... 6-26 6.12 HIGH RADIATION AREA .................................................................................................. 6-27 6.13 PROCESS CONTROL PROGRAM (PCP) .................................................................................................... 6-28 6.14 OFFSITE DOSE CALCULATION MANUAL (ODCM) ........................................................................ 6-29 6.15 MAJOR CHANGES TO RADIOACTIVE LIQUID, GASEOUS AND SOLID WASTE TREATMENT SYSTEMS ....................................................................................... 6-29 6.16 ENVIRONMENTAL QUALIFICATION ............................................................... ......................................... 6-30 6.17 TECHNICAL SPECIFICATION (TS) BASES CONTROL PROGRAM .................................... 6-30 6.18 CONTROL ROOM ENVELOPE HABITABILITY PROGRAM ............................................................ 6-31 SALEM - UNIT 1 XVIII Amendment No. 286

SECTION 1.0 DEFINITIONS

1.0 DEFINITIONS DEFINED TERMS 1.1 The DEFINED TERMS of this section appear in capitalized type and are applicable throughout these Technical Specifications.

ACTION 1.2 ACTION shall be that part of a specification which prescribes remedial measures required under designated conditions.

AXIAL FLUX DIFFERENCE 1.3 AXIAL FLUX DIFFERENCE shall be the difference in normalized flux signals between the top and bottom halves of a two section excore neutron detector.

CHANNEL CALIBRATION 1.4 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds within the necessary range and accuracy to known values of the parameter that the channel monitors. The CHANNEL CALIBRATION shall encompass the entire channel, including the required sensor, alarm, display, and trip functions, and shall include the CHANNEL FUNCTIONAL TEST. Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel. Whenever an RTD or thermocouple sensing element is replaced, the next required CHANNEL CALIBRATION shall include an inplace cross calibration that compares the other sensing elements with the recently installed sensing element. The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping, or total channel steps so that the entire channel is calibrated.

CHANNEL CHECK 1.5 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.

CHANNEL FUNCTIONAL TEST 1.6 A CHANNEL FUNCTIONAL TEST shall be the injection of a simulated signal into the channel as close to the primary sensor as practicable to verify OPERABILITY including alarm and/or trip functions.

SALEM - UNIT 1 1-1 Amendment No. 191

FEB 6 1997

DEFINITIONS CONTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when:

1.7.1 All penetrations required to be closed during accident conditions are either:

a. Capable of being closed by an OPERABLE containment automatic isolation valve system, or

b. Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except for valves that are open under administrative control as permitted by Specification 3.6.3.1.

1.7.2 All equipment hatches are closed and sealed, 1.7.3 Each air lock is OPERABLE pursuant to Specification 3.6.1.3, 1.7.4 The containment leakage rates are within the limits of Specification 3.6.1.2, and 1.7.5 The sealing mechanism associated with each penetration (e.g.,

welds, bellows or 0-rings) is OPERABLE.

CORE ALTERATION 1.8 CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe conservative position.

CORE OPERATING LIMITS REPORT 1.9 The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific document that provides core operating limits for the current operating reload cycle.

These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9.1.9. Unit operation within these operating limits is addressed in individual specifications.

DOSE EQUIVALENT I-131 1.10 DOSE EQUIVALENT I-131 shall be that concentration of 1-131 (microcuries per gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present.

The thyroid dose conversion factors used for this calculation shall be those listed in Federal Guidance Report No. 11 (FGR 11), "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion and Ingestion'.

SALEM - UNIT 1 1-2 Amendment No.271

DEFINITIONS E - AVERAGE DISINTEGRATION ENERGY 1.11 E shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half-lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.

ENGINEERED SAFETY FEATURE RESPONSE TIME 1.12 The ENGINEERED SAFETY FEATURE RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable.

FREQUENCY NOTATION 1.13 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2.

FULLY WITHDRAWN 1.13a FULLY WITHDRAWN shall be the condition where control and/or shutdown banks are at a position which is within the interval of 222 to 230 steps withdrawn, inclusive. FULLY WITHDRAWN will be specified in the current reload analysis.

GASEOUS RADWASTE TREATMENT SYSTEM 1.14 A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

IDENTIFIED LEAKAGE 1.15 IDENTIFIED LEAKAGE shall be:

a. Leakage (except Reactor Coolant Pump Seal Water Injection) into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or SALEM - UNIT 1 1-3 Amendment No. 292

DEFINITIONS

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 BOUNDARY LEAKAGE, or

c. Reactor coolant system leakage through a steam generator to the secondary system (primary-to-secondary leakage).

MEMBER(S) OF THE PUBLIC 1.16 MEMBER(S) OF THE PUBLIC shall be all those persons who are not occupationally associated with the plant. This category does not include employees of PSE&G, its contractors, or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational, or other purposes not associated with the plant.

OFFSITE DOSE CALCULATION MANUAL (ODCM) 1.17 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm/Trip setpoints, and in the conduct of the Environmental Radiological Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluent controls and Radiological Environmental Monitoring programs required by Section 6.8.4 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Annual Radioactive Effluent Release Reports required by Specifications 6.9.1.7 and 6.9.1.8 respectively.

OPERABLE - OPERABILITY 1.18 A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

OPERATIONAL MODE - MODE 1.19 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive combination of core reactivity condition, power level and average reactor coolant temperature specified in Table 1.1.

SALEM - UNIT 1 1-4 Amendment No. 268

DEFINITIONS PHYSICS TESTS 1.20 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and 1) described in Chapter 14 of the Updated FSAR, 2) authorized under the provisions of 10CFR50.59, or 3) otherwise by the Commission.

PRESSURE BOUNDARY LEAKAGE 1.21 PRESSURE BOUNDARY LEAKAGE shall be leakage (except primary-to-secondary leakage) through a non-isolable fault in a Reactor Coolant System component body, pipe wall or vessel wall.

PROCESS CONTROL PROGRAM (PCP) 1.22 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, test, and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71, State regulations, burial ground requirements, and other requirements governing the disposal of radioactive waste.

PURGE - PURGING 1.23 PURGE or PURGING shall be 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.

QUADRANT POWER TILT RATIO 1.24 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater. With one excore detector inoperable, the remaining three detectors shall be used for computing the average.

RATED THERMAL POWER 1.25 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3459 MWt.

SALEM - UNIT 1 1-5 Amendment No. 268

DEFINITIONS REACTOR TRIP SYSTEM RESPONSE TIME 1.26 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until loss of stationary gripper coil voltage.

REPORTABLE EVENT 1.27 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10CFR Part 50.

SHUTDOWN MARGIN 1.28 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical br would be subcritical from its present condition assuming all full length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be FULLY WITHDRAWN.

SITE BOUNDARY 1.29 The SITE BOUNDARY shall be that line beyond which the land is not owned, leased, or otherwise controlled by the licensee, as shown in Figure 5.1-3, and which defines the exclusion area as shown in Figure 5.1-1.

SOLIDIFICATION 1.30 Not Used SOURCE CHECK 1.31 SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to either (a) an external source of increased radioactivity, or(b) an internal source of radioactivity (keep-alive source), or (c) an equivalent electronic source check.

STAGGERED TEST BASIS 1.32 A STAGGERED TEST BASIS shall consist of:

a. A test schedule for (n) systems, subsystems, trains, or other designated components obtained by dividing the specified test interval into (n) equal subintervals.

SALEM - UNIT 1 1-6 Amendment No. 280

DEFINITIONS

b. The testing of one system, subsystem, train, or other designated component at the beginning of each subinterval.

THERMAL POWER 1.33 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

UNIDENTIFIED LEAKAGE 1.34 UNIDENTIFIED LEAKAGE shall be all leakage (except Reactor Coolant Pump seal Water Injection) which is not IDENTIFIED LEAKAGE.

UNRESTRICTED AREA 1.35 An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY, access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials, or any area within the SITE BOUNDARY used-for residential quarters or industrial, commercial, institutional, and/or recreational purposes.

VENTILATION EXHAUST TREATMENT SYSTEM 1.36 A VENTILATION EXHAUST TREATMENT SYSTEM shall be any system designed and installed to reduce gaseous radioiodine and radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines 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 TREATMENT SYSTEM components.

VENTING 1.37 VENTING shall be 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 VENTING. Vent, used in system names, does not imply a VENTING process.

SALEM - UNIT 1 1-7 Amendment No. 178 l OCT 30 1995

DEFINITIONS TABLE 1.1 OPERATIONAL MODES REACTIVITY AVERAGE COOLANT MODE CONDITION, Keff THERMAL POWER* TEMPERATURE

1. POWER OPERATION > 0.99 > 5X > 3500 F

2. STARTUP > 0.99 < 5% > 3500 F

3. HOT STANDBY < 0.99 O0 > 3500 F 4.- HOT SHUTDOWN < 0.99 0 3500 F > Tavq

> 2000 F

5. COLD SHUTDOWN < 0.99 0 < 2000 F

6. REFUELING** < 0.95 0 < 140 0 F

Excluding decay heat.

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

SALEM - UNIT 1 1-8 Amendment No. 59

. DEFINITIONS TABLE 1.2 FREQUENCY NOTATION NOTATION FREQUENCY 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 .

w At least once per 7 days.

M At least once per 31 days.

Q At least once per 92 days.

SA At least once per 6 months.

,,R At least once per 18 months.

S/U Prior to each reactor startup.

P Prior to each release.

N.A. Not applicable.

SALEM - UNIT 1 1-9 Amendment No. 59

SECTION 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temperature (Tvz) shall not exceed the limits shown in Figure 2.1-1 for 4 loop operation.

APPLICABILITY: MODES l.and 2.

ACTION:

Whenever the point defined by the combination of the highest operating loop average temperature and THERMAL POWER has exceeded the appropriate pressurizer pressure line, be in HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

SALEM - UNIT 1 2-1 Ameno 201NO

,,zv be, 1c~t AdV jaIru Ae e L 0 O6i.

680 - ..

(0.02, 663.84)

UNACCEPTABLE x

660 - OPERATION OLL 2440 PSIA 0-inZ 0 0 640

- 0 C)

L- W 0

6 (0.02,623.23) 2000 PSIA DNB it Oa>-

~- 620 c) (1.07, 618.56) 0r 1840 PSIA - - \ \

G00 W

'-~ (1.15, 597.51)

,594.31) gOU"

-c:j D,586.10) f 580 - ACCEPTABLE (1.20,582.81)

OPERATION I,572.90) w I I 0V 560 0.2 .4 .6.8 1 1 2 4 FRACTION OF RATED THERMAL POWER FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION SALEM - UNIT 1 2-2 Amendment No. 243 t',AY 2 5 2]i

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig.

APPLICABILITY: MODES 1, 2, 3, 4 and 5.

ACTION:

MODES 1 and 2 Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

MODES 3, 4 and 5 Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes.

I' SALEM - UNIT 1 2-3 Amendment No. 201 Z@5 be at,& ~ H e At i NOV 26 19976 7b o,1MV11

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The reactor trip system instrumentation setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2-1.

APPLICABILITY: As shown for each channel in Table 3.3-1.

ACTION:

With a reactor trip system instrumentation setpoint less conservative than the value shown in the Allowable Values column of Table 2.2-1, declare the channel inoperable and apply the applicable ACTION state-ment requirement of Specification 3.3.1.1 until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.

Ii SALEM - UNIT 1 2-4

TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES

1. Manual Reactor Trip Not-Applicable Not Applicable

2. Power Range, Neutron Fluix Low Setpoint -

25% of RATED Low Setpoint -

26% of RATED THERMAL POWER THERMAL POWER High Setpoint -

109% of RATED High Setpoint -

110% of RATED THERMAL POWER THERMAL POWER

3. Power Range, Neutron Fluix,

5% of RATED THERMAL POWER with

5.5% of RATED THERMAL POWER High Positive Rate a time constant Ž 2 seconds with a time constant Ž 2 seconds

4. Deleted

5. Intermediate Range, Neut iron 25% of RATED THERMAL POWER

30% of RATED THERMAL POWER Flux

6. Source Range, Neutron F!.ux 105 counts per second 1.3 x 105 counts per second

7. Overtemperature AT See Note 1 See Note 3

8. Overpower AT See Note 2 See Note 4

9. Pressurizer Pressure--Lo 1865 psig Ž 1855 psig

10. Pressurizer Pressure--Hi gh 2385 psig 5 2395 psig

11. Pressurizer Water Level- -High 5 92% of instrument span .* 93% of instrument span

12. Loss of Flow 90% of design flow per loop* 89% of design flow per loop*

Design flow is 82,500 gpm p er io<

SALEM - UNIT 1 2-5 Amendment No 278

I TABLE 2.2-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES

13. Steam Generator Water 2 14.0% of narrow range instrument > 13.0% of narrow range instrument I Level--Low-Low span-each steam generator span-each steam generator

14. Deleted

15. Undervoltage-Reactor 2 2900 volts-each bus 2 2850 volts-each bus Coolant Pumps

16. Underfrequency-Reactor 2 56.5 Hz - each bus 2 56.4 Hz - each bus Coolant Pumps

17. Turbine Trip A. Low Trip System 2 45 psig 2 45 psig Pressure B. Turbine Stop Valve

15% off full open 5 15% off full open Closure

18. Safety Injection Input Not Applicable Not Applicable from ESF

19. Reactor Coolant Pump Not Applicable Not Applicable Breaker Position Trip SALEM - UNIT 1 2-6 Amendment No.257 l

)

( c TABLE 2.2-1 (Continued) c REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS NOTATION NOTE 1: Overtemperature AT ATo [(Kj-K 2 ( 1+I1S (T-TI)+K3 (P-P')-f l (AI)]

1+T2S where: ATo = Indicated AT at RATED THERMAL POWER T = Average temperature, OF T' = Indicated Tavg at RATED THERMAL POWER 5 577.9 0 F P = Pressurizer pressure, psig P' = 2235 psig (indicated RCS nominal operating pressure) l+rlS The function generated by the lead-lag controller for Tavg dynamic compensation 1+T 2 S T1 & '2 = Time constants utilized in the lead-lag controller for Tavg ti = 30 secs +/-10%,

'2 = 4 secs. +/-10%

S = Laplace transform operator, Sec -

SALEM - UNIT 1 2-7 Amendment No. 239 DEC 79 200

TABLE 2.2-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS NOTATION (Continued)

Operation with 4 Loops K1 1.22 K2 0.02037 K3 0.001020 and f, (AI) is a function of the indicated difference between top and bottom detectors of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:

(i) for q, - qb between -33 percent and +11 percent, f1 (AI) = 0 (where qe and qb are percent RATED THERMAL POWER in the top and bottom I halves of the core respectively, and qt + qb is total THERMAL POWER in percent of RATED THERMAL POWER).

(ii) for each percent that the magnitude of (qt - q.) exceeds -33 I percent, the AT trip setpoint shall be automatically reduced by 2.34 percent of its value at RATED THERMAL POWER. I (iii) for each percent that the magnitude of (qt - q.) exceeds +11 I percent, the AT trip setpoint shall be automatically reduced by 2.37 percent of its value at RATED THERMAL POWER. I SALEM - UNIT 1 2-8 Amendment No. 243 MAYf 2 5 26'31

TABLE 2.2-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS NOTATTION' (rnnt-iniuri NOTE 2: Overpower: AT

AT. [K4 - K 5[ 3] T - K 6 (T - TV) - f 2 (AI)]

where: AT. Indicated AT at RATED THERMAL POWER T = Average temperature, OF T" = Indicated TV, at RATED THERMAL POWER < 577.9 0 F K4 = 1.09 K5 = 0.02/ 0 F for increasing average temperature and 0 for decreasing average temperature K6 = 0.00149/0 F for T > T"; K6 = 0 for T < T" T3S = The function generated by the rate lag controller 1+T3S for T.,, dynamic compensation T3 = Time constant utilized in the rate lag controller for Tavg T3 = 10 secs. +/-10%

S = Laplace transform operator, Sec1 f2 (AI) = 0 for all Al NOTE 3: The channel's maximum trip point shall not exceed its compute(I trip point by more than 1.1 percent.

NOTE 4: The channel's maximum trip point shall not exceed its computecI trip point by more than 2.1 percent.

SALEM - UNIT 1 2-9 Amendment No. 239 DEC 1 9 200D

SECTIONS 3.0 AND 4.0 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.0 APPLICABILITY LIMITING CONDITION FOR OPERATION 3.0.1 Compliance with the limiting Conditions for Operation contained in the succeeding specifications is required during the OPERATIONAL MODES or other conditions specified therein; except that upon failure to meet the Limiting Conditions for Operation, the associated ACTION requirements shall be met.

3.0.2 Noncompliance with a specification shall exist when the requirements of the Limiting Condition for Operation and associated ACTION requirements are not met within the specified time intervals. If the Limiting Condition for Operation is restored prior to expiration of the specified time intervals, completion of the ACTION requirements is not required.

3.0.3 When a Limiting Condition for Operation is not met except as provided in the associated ACTION requirements, within one hour action shall be initiated to place the unit in a MODE in which the specification does not apply by placing it, as applicable, in:

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 SHUTDOWN 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 .

Where corrective measures are completed that permit operation under the ACTION requirements, the ACTION may be taken in accordance with the specified time limits as measured from the time of failure to meet the Limiting Condition for Operation. Exceptions to these requirements are stated in the individual specifications.

3.0.4 When an LCO is not met, entry into a MODE or other specified condition in the Applicability shall only be made:

a. When the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time; or

b. After performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering the MODE or other specified condition in the Applicability, and establishment of risk management actions, if appropriate; exceptions to this Specification are stated in the individual Specifications, or

c. When an allowance is stated in the individual value, parameter, or other Specification.

This Specification shall not prevent changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit.

SALEM - UNIT 1 3/4 0-1 Amendment No. 27 6

APPLICABILITY LIMITING CONDITION FOR OPERATION 3.0.5 DELETED 3.0.6 Equipment removed from service or declared inoperable to comply with I

ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other equipment. This is an exception to LCO 3.0.2 for the system returned to service under administrative control to perform the testing required to demonstrate OPERABILITY.

SALEM - UNIT 1 3/4 0-2 Amendment No.253 OCT 1 1 2002

APPLICABILITY SURVEILLANCE REQUIREMENTS 4.0.1 Surveillance Requirements shall be met during the OPERATIONAL MODES or other specified conditions in the Applicability for individual Limiting Conditions for Operation, unless otherwise stated in the Surveillance Requirement. Failure to meet a Surveillance, whether such failure is experienced during the performance of the Surveillance or between performances of the Surveillance, shall be failure to meet the Limiting Condition for Operation. Failure to perform a Surveillance within the specified frequency shall be failure to meet the Limiting Condition for Operation, except as provided in Specification 4.0.3. Surveillances do not have to be performed on inoperable equipment or variables outside specified limits.

4.0.2 Each Surveillance Requirement shall be performed within the specified surveillance interval with a maximum allowable extension not to exceed 25 percent of the specified surveillance interval.

4.0.3 If it is discovered that a Surveillance was not performed within its specified frequency, then compliance with the requirement to declare the Limiting Condition for Operation not met may be delayed, from the time of discovery, up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or up to the limit of the specified frequency, whichever is greater. This delay period is permitted to allow performance of the Surveillance. A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and the risk impact shall be managed.

If the Surveillance is not performed within the delay period, the Limiting Condition for Operation must immediately be declared not met and the applicable Actions must be entered.

When the Surveillance is performed within the delay period and the Surveillance is not met, the Limiting Condition for Operation must immediately be declared not met and the applicable Actions must be entered.

4.0.4 Entry into a MODE or other specified condition in the Applicability of an LCO shall only be made when the LCO's Surveillances have been met within their specified Frequency, except as provided by SR 4.0.3. When an LCO is not met due to Surveillances not having been met, entry into a MODE or other specified condition in the Applicability shall only be made in accordance with LCO 3.0.4.

This provision shall not prevent entry into MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit.

SALEM - UNIT 1 3/4 0-3 Amendment No. 297

3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUTDOWN MARGIN - T §Y9 > 200°F LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be ~ 1.3% llklk.

APPLICABILITY: MODES 1, 2*, 3, and 4.

ACTION:

With the SHUTDOWN MARGIN < 1.3% llklk, immediately initiate and continue boration at

~ 33 gpm of a solution containing ~ 6,560 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored.

SURVEILLANCE REQUIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be ~ 1.3% llklk:

a. Within one hour after detection of an inoperable control rod(s) and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the rod(s) is inoperable. If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be increased by an amount at least equal to the withdrawn worth of the immovable or untrippable control rod(s).

b. When in MODES 1 or 2#, in accordance with the Surveillance Frequency Control Program by verifying that control banks are within the limits in the COLR per Specification 3.1.3.5.

c. When in MODE 2##, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to achieving reactor criticality by verifying that the predicted critical control rod position is within the limits in the COLR per specification 3.1.3.5.

See Special Test Exception 3.10.1

With Keff ~ 1.0

1. With Keff < 1.0 SALEM - UNIT 1 3/4 1-1 Amendment No. 299

REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

d. Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors of e below, with the control banks at the maximum insertion limit in the COLR per Specification 3.1.3.5.

e. When in MODES 3 or 4, in accordance with the Surveillance Frequency Control Program by consideration of the following factors:

1. Reactor coolant system boron concentration,

2. Control rod position,

3. Reactor coolant system average temperature,

4. Fuel burnup based on gross thermal energy generation,

5. Xenon concentration, and

6. Samarium concentration.

4.1.1.1.2 The overall core reactivity balance shall be compared to predicted values to demonstrate agreement within +/- 1% Llklk in accordance with the Surveillance Frequency Control Program. This comparison shall consider at least those factors stated in Specification 4.1.1.1.1.e, above. The predicted reactivity values shall be adjusted (normalized) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 Effective Full Power Days after each fuel loading.

SALEM - UNIT 1 3/4 1-2 Amendment No. 299

REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN - T ml < 200°F LIMITING CONDITION FOR OPERATION 3.1.1.2 The SHUTDOWN MARGIN shall be 2: 1.0% Llk/k.

APPLICABILITY: MODE 5.

ACTION:

With the SHUTDOWN MARGIN < 1.0% Llklk, immediately initiate and continue boration at 2: 33 gpm of a solution containing 2: 6,560 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored.

4.1.1.2 The SHUTDOWN MARGIN shall be determined to be 2: 1.0% Llklk:

a. Within one hour after detection of an inoperable control rod(s) and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the rod(s) is inoperable. If the inoperable control rod is immovable or untrippable, the SHUTDOWN MARGIN shall be increased by an amount at least equal to the withdrawn worth of the immovable or untrippable control rod(s).

b. In accordance with the Surveillance Frequency Control Program by consideration of the following factors:

1. Reactor coolant system boron concentration,

2. Control rod position,

3. Reactor coolant system average temperature,

4. Fuel burnup based on gross thermal energy generation,

5. Xenon concentration, and

6. Samarium concentration.

SALEM - UNIT 1 3/4 1-3 Amendment No. 299

THIS PAGE LEFT INTENTIONALLY BLANK SALEM - UNIT 1 3/4 1-4 Amendment No. 109

- JAN 3 1990 I

REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION 3.1.1.4 The moderator temperature coefficient (MTC) shall be within the limits specified in the CORE OPERATING LIMITS REPORT (COLR). The maximum upper limit shall be less positive than or equal to 0 Ak/k/OF.

APPLICABILITY: Beginning of Cycle Life (BOL) Limit - MODES 1 and 2* only#

End of Cycle Life (EOL) Limit - MODES 1, 2 and 3 only#

ACTION:

a. With the MTC more positive than the BOL limit specified in the COLR, operations in MODES 1 and 2 may proceed provided:

1. Control rod withdrawal limits are established and maintained sufficient to restore the MTC to less positive than the BOL limit specified in the COLR within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . These withdrawal limits shall be in addition to the insertion limits in the COLR per Specification 3.1.3.5.

2. The control rods are maintained within the withdrawal limits established above until a subsequent calculation verifies that the MTC has been restored to within its limit for the all rods withdrawn condition.

3. A Special Report is prepared and submitted to the Commission pursuant to Specification 6.9.2 within 10 days, describing the value of the measured MTC, the interim control rod withdrawal limits and the predicted average core burnup necessary for restoring the positive MTC to within its limit for the all rods withdrawn condition.

b. With the MTC more negative than the EOL limit specified in the COLR, be in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

With Kff greater than or equal to 1.0

See Special Test Exception 3.10.3 SALEM - UNIT 1 3/4 1-5 Amendment No- 201 bA- be jn p a r e tv6 2k 1997

REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT SURVEILLANCE REQUIREMENTS 4.1.1.4 The MTC shall be determined to be within its limits during each fuel cycle as follows:

a. The MTC shall be measured and compared to the BOL limit specified in the COLR, prior to initial operation above 5t of RATED THERMAL POWER, after each fuel loading.

b. The MTC shall be measured at any THERMAL POWER and compared to the 300 ppm surveillance limit specified in the COLR (all rods withdrawn, RATED THERMAL POWER condition) within 7 EFPD after reaching an equilibrium boron concentration of .300 ppm. In the event this comparison indicates the MTC is more negative than the 300 ppm surveillance limit specified in the COLR, the MTC shall be remeasured, and compared to the EOL MTC limit specified in the COLR, at least once per 14 EFPD during the remainder of the fuel cycle.

SALEM - UNIT 1 3/4 1-Sa Amendment No. 201 Apt)~d~ , b Af a. rNOV 26 1997

REACTIVITY CONTROL SYSTEMS MINIMUM TEMPERATURE FOR CRITICALITY LIMITING CONDITION FOR OPERATION 3.1.1.5 The Reactor Coolant System lowest operating loop temperature (Tavg) shall be > 541 0F.

APPLICABILITY: MODES 1 and 2#.

ACTION:

With a Reactor Coolant System operating loop temperature (T ) c 541 0F, restore (T ) to within its limit within 15 minutes or be 1XHOT STANDBY within thea xt 15 minutes.

SURVEILLANCE REQUIREMENTS 4.1'.,1 .5 The Reactor Coolant System temperature (Tavg) shall be determined to be > 541'F:av

a. Within 15 minutes prior to achieving reactor criticality, and

b. At least once per 30 minutes when the reactor is critical and the Reactor Coolant System T is less than 5511F with the Deviation Alarm AR reset.

Tavg - TrefDeitoAlr es.

With Keff > 1.0.

SALEM - UNIT 1 3/4 1-6

REACTIVITY CONTROL SYSTEMS 3/4.1.2 BORATION SYSTEMS FLOW PATHS - SHUTDOWN 3.1.2.1 As a minimum, one of the following boron injection flow paths shall be OPERABLE:

a. A flow path from the boric acid tanks via a boric acid transfer pump and a charging pump to the Reactor Coolant System if the boric acid storage system is OPERABLE, per Specification 3.1.2.6a while in MODE 4, or per Specification 3.1.2.Sa while in MODE S or 6, or

b. A flow path from the refueling water storage tank via a charging purnp to the Reactor Coolant System if the refueling water storage tank is OPERABLE per Specification 3.1.2.6b while in MODE 4, or per Specification 3.1.2.Sb while in MODE S or 6.

APPLICABILITY: MODES 4, Sand 6.

ACTION:

With none of the above flow paths OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes until at least one injection path is restored to OPERABLE status.

4.1.2.1 At least one of the above required flow paths shall be demonstrated OPERABLE:

a. When the boric acid tank is a required water source, by verifying in accordance with the Surveillance Frequency Control Program that:

(1) The flow path from the boric acid tank to the boric acid transfer pump, the boric acid transfer pump, and the recirculation path from the boric acid transfer pump to the boriC acid tank is ?! 63°F, and (2) The flow path between the boric acid transfer pump recirculation line to the charging pump suction line is ?! SO°F,

b. In accordance with the Surveillance Frequency Control Program by verifying that each valve (manual, power operated or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.

SALEM - UNIT 1 3/4 1-7 Amendment No. 299

REACTIVITY CONTROL SYSTEMS FLOW PATHS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.2 At least two of the following three boron injection flow paths shall be OPERABLE:

a. A flow path from the boric acid tanks via a boric acid transfer pump and a charging pump to the Reactor Coolant System.

b. Two flow paths from the refueling water storage tank via charging pumps to the Reactor Coolant System.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

With only one of the above required boron injection flow paths to the Reactor Coolant System OPERABLE, restore at least two boron injection flow paths to the Reactor Coolant System to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least 1% delta k/k at 200°F within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore at least two flow paths to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.1.2.2 Each of the above required flow paths shall be demonstrated OPERABLE:

a. By verifying in accordance with the Surveillance Frequency Control Program that (1) The flow path from the boric acid tank to the boric acid transfer pump and from the recirculation line back to the boric acid tank is ;;:: 63°F, and (2) the flow path between the boric acid tank recirculation line to the charging pump suction line is ;;:: 50°F,

b. In accordance with the Surveillance Frequency Control Program by verifying that each valve (manual, power operated or automatiC) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.

SALEM - UNIT 1 3/4 1-8 Amendment No, 299

REACTIVITY CONTROL SYSTEMS SUB\LEILLANCE REQUIREMENTS (Continued)

c. In accordance with the Surveillance Frequency Control Program during shutdown by verifying that each automatic valve in the flow path actuates to its correct position on a safety injection test signal.

d. In accordance with the Surveillance Frequency Control Program by verifying that the flow path required by specification 3.1.2.2.a delivers at least 33 gpm to the Reactor Coolant System.

SALEM - UNIT 1 3/4 1-9 Amendment No. 299

REACTIVITY CONTROL SYSTEMS CHARGING PUMP - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.3 At least one charging pump in the boron injection flow path required by Specification 3.1.2.1 shall be OPERABLE.#

APPLICABILITY: MODES 4, 5 and 6.

ACTION:

With no charging pump OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes until one charging pump is restored to OPERABLE status.

SURVEILLANCE REQUIREMENTS 4.1.2.3 No additional Surveillance Requirements other than those required by the Inservice Testing Program.

A maximum of one centrifugal charging pump shall be OPERABLE while in MODE 4 when the temperature of one or more of the RCS cold legs is less than or equal to 312°F, MODE 5, or MODE 6 when the head is on the reactor vessel.

SALEM - UNIT 1 3/41-10 Amendment No. 297

REACTIVITY CONTROL SYSTEMS CHARGING PUMPS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.4 At least two charging pumps shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

With only one charging pump OPERABLE, restore at least two charging pumps to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least 1% 8k/k at 200°F within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore at least two charging pumps to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.1.2.4 No additional Surveillance Requirements other than those required by the Inservice Testing Program.

SALEM - UNIT 1 3/4 1-11 Amendment No. 297

THIS PAGE LEFT INTENTIONALLY BLANK SALEM - UNIT 1 3/4 1-12 Amendment No. 1 6 9 I JJUC 1 2 1995

THIS PAGE LEFT INTENTIONALLY BLANK SALEM - UNIT 1 3/4 1-13 Amendment No.169 JUN 1 2 1995

REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.5 As a minimum, one of the following borated water sources shall be OPERABLE:

a. A boric acid storage system with:

1. A minimum contained volume of 2,600 gallons,

2. Between 6,560 and 6,990 ppm of boron, and,

3. A minimum solution temperature of 63°F.

b. The refueling water storage tank with:

1. A minimum contained volume of 37,000 gallons,

2. A minimum boron concentration of 2300 ppm, and

3. A minimum solution temperature of 35°F.

APPLICABILITY: MODES 5 and 6.

ACTION:

With no borated water source OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes until at least one borated water source is restored to OPERABLE status.

4.1.2.5 The above required borated water source shall be demonstrated OPERABLE:

a. For the boric acid storage system, when it is the source of borated water in accordance with the Surveillance Frequency Control Program by:

1. Verifying the boron concentration of the water,

2. Verifying the water level of the tank, and

3. Verifying the boric acid storage tank solution temperature when it is the source of borated water.

SALEM - UNIT 1 3/4 1-14 Amendment No. 299

REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

b. For the refueling water storage tank by:

1. Verifying the boron concentration in accordance with the Surveillance Frequency Control Program,

2. Verifying the borated water volume in accordance with the Surveillance Frequency Control Program, and

3. Verifying the solution temperature in accordance with the Surveillance Frequency Control Program, when it is the source of borated water and the outside air temperature is less than 35°F.

SALEM - UNIT 1 3/4 1-15 Amendment No. 299

REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.6 As a minimum, the following borated water source(s) shall be OPERABLE as required by specifications 3.1.2.1 and 3.1.2.2:

a. A boric acid storage system with:

1. A contained volume of borated water in accordance with figure 3.1.2,

2. A boron concentration in accordance with figure 3.1-2, and

3. A minimum solution temperature of 63"F.

b. The refueling water storage tank with:

1. A contained volume of between 364,500 and 400,000 gallons of water,

2. A boron concentration of between 2,300 and 2,500 ppm, and

3. A minimum solution temperature of 35"F.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

a. With the boric acid storage system inoperable and being used as one of the above required boration water systems, restore the storage system to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and borated to a SHUTDOWN MARGIN equivalent to at least 1%

delta Klk at 200°F; restore the boric acid storage system to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With the refueling water storage tank inoperable, restore the tank to OPERABLE status within one hour or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SALEM - UNIT 1 3/4 1-16 Amendment No 299

REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS 4.1.2.6 Each borated water source shall be demonstrated OPERABLE:

a. For the boric acid storage system, when it is the source of borated water in accordance with the Surveillance Frequency Control Program by:

1. Verifying the boron concentration in each water source,

2. Verifying the water level of each water source, and

3. Verifying the boric acid storage system solution temperature.

b. For the refueling water storage tank by:

1. Verifying the boron concentration in accordance with the Surveillance Frequency Control Program,

2. Verifying the borated water volume in accordance with the Surveillance Frequency Control Program, and

3. Verifying the solution temperature in accordance with the Surveillance Frequency Control Program when the outside air temperature is less than 35°F.

SALEM - UNIT 1 3/41-17 Amendment No. 299

BORIC ACID TANK CONTENTS BASED ON RWST CONCENTRATION C)

I I I I I I I z

100% REGION OF 7,8000 z -\

CL 0 \ ACCEPTABLE OPERATION - - - - - - - -

-J I ' a LU I I I I I I I a: 98% ( 7,600 as. a a a 1 H

z I

L, : %% a, Na ,

95% 7 ,40 0 Z LU ,5 ,,.4, LLJ I £ :2

-j

- - -' - -* - '.' - - _' -s _!_ _ ,___!____@

D 0M 92% 7,200 -I 0

0 c 0 90% 7,000 l LU 0

,. , a . . . .... a,s_

F-

-j 87% 23500ppma 6,800 2 2z a a a a 2 _a-; ppm a 66 D i

0 z I-1 -:6560 ppm (minimum) a a a a .

. a 85% -REGI!_ l OF-- , , - 6,600 ZE PNACPEPTABLE OPERAT!ON 690 ppm (maximum)_,G 6,600 6,700 6,800 6,900 7,000 CONCENTRATION IN BAT - ppm BORON SALEM - UNIT 1 Figure 3.1-2 3/4 1-17(a) Amendment No. 14 5 OCT 1 5 1993

REACTIVITY CONTROL SYSTEMS 3/4.1.3 MOVABLE CONTROL ASSEMBLIES GROUP HEIGHT LIMITING CONDITION FOR OPERATION 3.1.3.1 All full length (shutdown and control) rods, shall be OPERABLE and positioned within +/- 18 steps(indicated position) when reactor power is

85 i, RATED THERMAL POWER, or +/- 12 steps (indicated position) when reactor power is

> 85% RATED THERMAL POWER, of their group step counter demand position within one hour after rod motion.

APPLICABILITY: MODES 1* and 2*

ACTION:

a. With one or more full length rods inoperable due to being immovable as a result of excessive friction or mechanical interference or known to be untrippable, determine that the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. With more than one full length rod inoperable or mis-aligned from the group step counter demand position by more than +/- 18 steps(indicated position) at

85 e RATED THERMAL POWER or +/- 12 steps (indicated position) at > 85% RATED THERMAL POWER, be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c. With one full length rod inoperable due to causes other than addressed by ACTION a, above, or mis-aligned from its group step counter demand position by more than +/- 18 steps(indicated position) at S 85 % RATED THERMAL POWER or +/- 12 steps (indicated position) at > 85% RATED THERMAL POWER, POWER OPERATION may continue provided that within one hour either:

1. The rod is restored to OPERABLE status within the above alignment requirements, or

2. The remainder of the rods in the bank with the inoperable rod are aligned to within + 18 steps(indicated position) at S 85 % RATED THERMAL POWER or +/- 12 steps (indicated position) at > 85% RATED THERMAL POWER of the inoperable rod while maintaining the rod sequence and insertion limits in the COLR per specification 3.1.3.5. The THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.5 during subsequent operation, or

3. The rod is declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied. POWER OPERATION may then continue provided that:

See Special Test Exceptions 3.10.2 and 3.10.3.

SALEM - UNIT 1 3/4 1-18 Amendment No. 202 Revised by letter dtd February vq 1QQ8

(, C Qu ) '

BoAs T PM MMIWIfLLY LM

REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION (Continued) a) A reevaluation of each accident analysis of table 3.1-1 is performed within 5 days; this reevaluation shall confirm that the previously analyzed results of these accidents remain valid for the duration of operation under these conditions.

b) The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is determined at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

c) A core power distribution measurement is obtained and Fo(Z) FNllH are verified to be within their limits within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

d) The THERMAL POWER level is reduced to less than or equal to 75% of RATED THERMAL POWER within one hour and within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months the high neutron flux trip setpoint is reduced to less than or equal to 85%

of RATED THERMAL POWER. THERMAL POWER shall be maintained less than or equal to 75% of RATED THERMAL POWER until compliance with ACTIONS 3.1.3.1.c.3.a and 3.1.3.1.c.3.c above are demonstrated.

SURVEILLANCE REQUIREMENTS 4.1.3.1.1 The position of each full length rod shall be determined to be within the limits established in the limiting condition for operation in accordance with the Surveillance Frequency Control Program (allowing for one hour thermal soak after rod motion) except during time intervals when the Rod Position Deviation Monitor is inoperable, then verify the group positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

4.1.3.1.2 Each full length rod not fully inserted in the core shall be determined to be OPERABLE by movement of at least 10 steps in anyone direction in accordance with the Surveillance Frequency Control Program.

SALEM - UNIT 1 3/41-18a Amendment No. 299

TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION INTHTE EVENT OF AN INOPERABLE FULL LENGTH ROD Rod Cluster Control Assembly Insertion Characteristics Rod Cluster Control Assembly Mis-alignment Loss Of Reactor Coolant From Ruptured Pipes Or From Cracks In Large Pipes Which Actuates The Emergency Core Cooling System Single Rod Cluster Control Assembly Withdrawal At Full F wer Mijor Reactor Coolant System Pipe Ruptures (Loss Of Coolant Accident)

Major Secondary System Pipe Rupture Rupture of a Control Rod Drive Mechanism Housing (Rod Cluster Control Assembly Ejection)

SALEM - UNIT 1 3/4 .1-18b Amendment No. 73

REACTIVITY CONTROL SYSTEMS POSITION INDICATION SYSTEMS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.3.2.1 The shutdown and control rod position indication systems shall be OPERABLE and capable of determining the actual and demanded rod positions as follows:

a. Analog rod position indicators, within one hour after rod motion (allowance for thermal soak);

All Shutdown Banks: +/- 18 steps at s 85% reactor power or if reactor power is

> 85% RATED THERMAL POWER +/- 12 steps of the group demand counters for withdrawal ranges of 0-30 steps and 200-230 steps.

Control Bank A: +/- 18 steps at s 85% reactor power or if reactor power is > 85%

RATED THERMAL POWER +/- 12 steps of the group demand counters for withdrawal ranges of 0-30 steps and 200-230 steps.

Control Bank B: +/- 18 steps at s 85% reactor power or if reactor power is > 85%

RATED THERMAL POWER +/- 12 steps of the group demand counters for withdrawal ranges of 0-30 steps and 160-230 steps.

Control Bank C and D: +/- 18 steps at s 85% reactor power or if reactor power is

> 85% RATED THERMAL POWER +/- 12 steps of the group demand counters for withdrawal ranges of 0-230 steps.

b. Group demand counters; +/- 2 steps of the pulsed output of the Slave Cycler Circuit over the withdrawal range of 0-230 steps.

APPLICABILITY: MODES 1 and 2.

ACTION:

a. With a maximum of one analog rod position indicator per bank inoperable either:

1. Determine the position of the non-indicating rod(s) indirectly using the power distribution monitoring system (if power is above 25% RTP) or using the movable incore detectors (if power is less than 25% RTP or the power distribution monitoring system is inoperable) at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and within one hour after any motion of the non-indicating rod which exceeds 24 steps in one direction since the last determination of the rod's position, or

2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

b. With two or more analog rod position indicators per bank inoperable, within one hour restore the inoperable rod position indicator(s) to OPERABLE status or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . A maximum of one rod position indicator per bank may remain inoperable following the hour, with Action (a) above being applicable from the original entry time into the LCO.

SALEM - UNIT 1 3/4 1-19 Amendment No. 295

REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)

b. With two or more analog rod position indicators per bank inoperable, within one hour restore the inoperable rod position indicator(s) to OPERABLE status or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . A maximum of one rod position indicator per bank may remain inoperable following the hour, with Action (a) above being applicable from the original entry time into the LCO.

c. With a maximum of one group demand position indicator per bank inoperable either:

1. Verify that all analog rod position indicators for the affected bank are OPERABLE and that the most withdrawn rod and the least withdrawn rod of the bank are within a maximum of 18 steps when reactor power is s; 85% RATED THERMAL POWER or if reactor power is> 85% RATED THERMAL POWER, 12 steps of each other at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , or

2. Reduce THERMAL POWER to less than 50% of RATED POWER within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

SURVEILLANCE REQUIREMENTS 4.1.3.2.1.1 Each analog rod position indicator shall be determined to be OPERABLE by verifying that the demand position indication system and the rod position indication system agree within 18 steps when reactor power is s; 85% RATED THERMAL POWER or if reactor power is> 85% RATED THERMAL POWER, 12 steps (allowing for one hour thermal soak after rod motion) in accordance with the Surveillance Frequency Control Program except during time intervals when the Rod Position Deviation Monitor is inoperable, then compare the demand position indication system and the rod position indication system at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

4.1.3.2.1.2 Each of the above required rod position indicator(s) shall be determined to be OPERABLE by performance of a CHANNEL calibration in accordance with the Surveillance Frequency Control Program.

SALEM - UNIT 1 3/41-19a Amendment No. 299

THIS PAGE INTENTIONALLY LEFT BLANK I

SALEM - UNIT 1 3/4 1-20 Amendment No. 232 AUG 1 7 2000

REACTIVITY CONTROL SYSTEMS ROD DROP TIME LIMITING CONDITION FOR OPERATION 3.1.3.3 The individual full length (shutdown and control) rod drop time from 230 steps withdrawn shall be less than or equal to 2.7 seconds from beginning of decay of stationary gripper coil voltage to dashpot entry with:

a. T avg greater than or equal to 541°F, and

b. All reactor coolant pumps operating.

APPLICABILITY: MODES 1 & 2.

ACTION:

a. With the drop time of any full length rod determined to exceed the above limit, restore the rod drop time to within the above limit prior to proceeding to MODE 1 or 2.

SURVEILLANCE REQUIREMENTS 4.1.3.3 The rod drop time of full length rods shall be demonstrated through measurement prior to reactor criticality:

a. For all rods following each removal of the reactor vessel head,

b. For specifically affected individual rods following any maintenance on or modification to the control rod drive system which could affect the drop time of those specific rods, and

c. In accordance with the Surveillance Frequency Control Program.

SALEM - UNIT 1 3/4 1-21 Amendment No. 299

REACTIVITY CONTROL SYSTEMS SHUTDOWN ROD INSERTION LIMIT LIMITING CONDITION FOR OPERATION 3.1.3.4 All shutdown rods shall be FULLY WITHDRAWN.

APPLICABILITY: MODES 1*, and 2*#@

ACTION:

With a maximum of one shutdown rod not FULLY WITHDRAWN, except for surveillance testing pursuant to Specification 4.1.3.1.2, within one hour either:

a. FULLY WITHDRAW the rod, or,

b. Declare the rod to be inoperable and apply Specification 3.1.3.1.

4.1.3.4 Each shutdown rod shall be determined to be FULLY WITHDRAWN by use of the group demand counters, and verified by the analog rod position indicators**:

a. Within 15 minutes prior to withdrawal of any rods in control banks A, B, C, or D during an approach to reactor criticality. and

b. In accordance with the Surveillance Frequency Control Program thereafter.

See Special Test Exceptions 3.10.2 and 3.10.3

- For power levels below 50% one hour thermal "soak time" is permitted. During this soak time, the absolute value of rod motion is limited to six steps.

With Keff greater than or equal to 1.0

@ Surveillance 4.1.3.4.a is applicable prior to withdrawing control banks in preparation for startup (Mode 2).

SALEM - UNIT 1 3/4 1-22 Amendment No. 299

REACTIVITY CONTROL SYSTEMS CONTROL ROD INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.5 The control banks shall be limited in physical insertion as specified in the CORE OPERATING LIMITS REPORT (COLR).

APPLICABILITY: MODES 1*, and 2*#

ACTION:

With the control banks inserted beyond the above insertion limits, except for surveillance testing pursuant to Specification 4.1.3.1.2, either:

a. Restore the control banks to within the limits within two hours, or

b. Reduce THERMAL POWER within two hours to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the bank position using the insertion limits specified in the COLR, or

c. Be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.1.3.5 The position of each control bank shall be determined to be within the insertion limits in accordance with the Surveillance Frequency Control Program by use of the group demand counters and verified by the analog rod position indicators** except during time intervals when the Rod Insertion Limit Monitor is inoperable, then verify the individual rod positions at least once per 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months s**.

See Special Test Exceptions 3.10.2 and 3.10.3

- For power levels below 50% one hour thermal "soak time" is permitted. During this soak time, the absolute value of rod motion is limited to six steps.

With Keff greater than or equal to 1.0 SALEM - UNIT 1 3/4 1-23 Amendment No. 299

THIS PAGE INTENTIONALLY LEFT BLANK SALEM - UNIT I

-- ,totter-Jw tuk , fAt6 to maft 3/4 :1-24 Amendment No. 201 NOV 2 6 1997 I q Wut QAtcZt em9

THIS PAGE INTENTIONALLY LEFT BLANK SALEM - UNIT 1 3/4 1-25 Amendment No. 201 NOV 2 6 1397

- kfat aW mu e

THIS PAGE INTENTIONALLY LEFT BLANK

-N I SALEM - UNIT 1 3/4 1-26 Amendment No. 16 JUN 1 1979

3/4.2 POWER DISTRIBUTION LIMITS AXIAL FLUX DIFFERENCE (AFD)

LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within the target band about the target flux difference as specified in the CORE OPERATING LIMITS REPORT (COLR).

APPLICABILITY: MODE 1 ABOVE 50% RATED THERMAL POWER*

ACTION:

a. With the indicated AXIAL FLUX DIFFERENCE outside of the limits specified in the COLR and with THERMAL POWER: l

1. Above 90% of RATED THERMAL POWER, within 15 minutes:

a) Either restore the indicated AFD to within the target band limits, or b) Reduce THERMAL POWER to less than 90% of RATED THERMAL POWER.

2. Between 50% and 90% of RATED THERMAL POWER:

) a) POWER OPERATION may continue provided:

1) The indicated AFD has not been outside of the limits specified in the COLR for more than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months penalty l deviation cumulative during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and

2) The indicated AFD is within the limits specified in the COLR. Otherwise, reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux-High Trip Setpoints to s 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b) Surveillance testing of the Power Range Neutron Flux Channels may be performed pursuant to Specification 4.3.1.1.1 provided the indicated AFD is maintained within the limits specified in the COLR. A total of 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months operation may be accumulated with the AFD outside of the target band during this testing without penalty deviation.

See Special Test Exception 3.10.2 SALEM - UNIT 1 3/4 2-1 Amendment No. 201 P(WY,o<pz XA + m 47 NOV 2 6 1997 "L12 tLb go

POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION (Continued)

b. THERMAL POWER shall not be increased above 90% of RATED THERMAL POWER unless the indicated AFD is within the limits specified in the COLR and ACTION a.2.a)1), above has been satisfied.

c. THERMAL POWER shall not be increased above 50% of RATED THERMAL POWER unless the indicated AFD has not been outside of the limits specified in the COLR for more than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months penalty deviation cumulative during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REQUIREMENTS 4.2.1.1 The indicated AXIAL FLUX DIFFERENCE shall be determined to be within its limits during POWER OPERATION above 15% of RATED THERMAL POWER by:

a. Monitoring the indicated AFD for each OPERABLE excore channel:

1. In accordance with the Surveillance Frequency Control Program when the AFD Monitor Alarm is OPERABLE, and

2. In accordance with the Surveillance Frequency Control Program for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after restoring the AFD Monitor Alarm to OPERABLE status.

b. Monitoring and logging the indicated AXIAL FLUX DIFFERENCE for each OPERABLE excore channel in accordance with the Surveillance Frequency Control Program for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and in accordance with the Surveillance Frequency Control Program thereafter, when the AXIAL FLUX DIFFERENCE Monitor Alarm is inoperable. The logged values of the indicated AXIAL FLUX DIFFERENCE shall be assumed to exist during the interval preceding each logging.

4.2.1.2 The indicated AFD shall be considered outside of its limits when at least 2 of 4 or 2 of 3 OPERABLE excore channels are indicating the AFD to be outside of the target band. Penalty deviation outside of the target band shall be accumulated on a time basis of:

a. One minute penalty deviation for each one minute of POWER OPERATION outside of the limits at THERMAL POWER levels equal to or above 50% of RATED THERMAL POWER, and

b. One-half minute penalty deviation for each one minute of POWER OPERATION outside of the limits at THERMAL POWER levels below 50% of RATED THERMAL POWER.

SALEM - UNIT 1 3/4 2-2 Amendment No. 299

POWER DISTRIBUTION LIMITS SURVEIL~ANCE REQUIREMENTS (Continued) 4.2.1.3 The target flux difference of each OPERABLE excore channel shall be determined by measurement in accordance with the Surveillance Frequency Control Program. The provisions of Specification 4.0.4 are not applicable.

4.2.1.4 The target flux difference shall be updated in accordance with the Surveillance Frequency Control Program by either determining the target flux difference pursuant to 4.2.1.3 above or by linear interpolation between the most recently measured value and 0 percent at the end of the cycle life. The provisions of Specification 4.0.4 are not applicable.

SALEM - UNIT 1 3/4 2-3 Amendment No. 299

THIS PAGE INTENTIONALLY LEFT BLANK SALEM - UNIT 1 3/4 2-4 Amendment No- 201

-A) bC is p2 HW As NOV 2 6 1997 40, & oagl

POWER DISTRIBUTION LIMITS HEAT FLUX HOT CHANNEL FACTOR-FQ(Z)

LIMITING CONDITION FOR OPERATION 3.2.2 FQ(z) shall be limited by the following relationships:

FQ(Z)

FRTPQ

K(z) for P > 0.5, and P

FQ(z) < FRTPQ

K(z) for P

0.5, 0.5 RTP Where: F P = the FQ limit at RATED THERMAL POWER (RTP) specified in the CORE OPERATING LIMITS REPORT (COLR),

P = THERMAL POWER and RATED THERMAL POWER K(z) = the normalized FQ(z) as a function of core height as specified in the COLR.

APPLICABILITY: MODE 1 ACTION:

With FQ(z) exceeding its limit:

a. Reduce THERMAL POWER at least 1% for each 1% FQ(Z) exceeds the limit within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; subsequent POWER OPERATION may proceed provided the Overpower AT Trip Setpoints have been reduced at least 1% for each 1% FQ(Z) exceeds the limit. The Overpower AT Trip Setpoint reduction shall be performed with the reactor in at least HOT STANDBY.

b. Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER above the reduced limit required by a.

above; THERMAL POWER may then be increased provided FQ(Z) is demonstrated through a core power distribution measurement to be within its limit.

SALEM - UNIT 1 3 /4 2-5 Amendment No. 237 NOV 0 6 2G00

POWER DISTRIBUTION LIMITS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

4.2.2.2 Fxy shall be evaluated to determine if Fo(Z) is within its limit by:

a. Using the movable incore detectors to obtain a power distribution map:

1. When THERMAL POWER is s 25%, but> 5% of RATED THERMAL POWER, or

2. When the Power Distribution Monitoring System (PDMS) is inoperable; and increasing the Measured Fo(Z) by the applicable manufacturing and measurement uncertainties as specified in the COLR.

b. Using the PDMS or the moveable incore detectors when THERMAL POWER is> 25% of RATED THERMAL POWER, and increasing the measured Fo(Z) by the applicable manufacturing and measurement uncertainties as specified in the COLR.

c. Comparing the Fxy computed (FXYc) obtained in b, above to:

1. The Fxy limits for RATED THERMAL POWER (FXyRTP) for the appropriate measured core planes given in e and f below, and

2. The relationship:

TP FXyL = Fxt [1 + PF xy (1-P)]

where Fx/ is the limit for fractional THERMAL POWER operation expressed as a function of FXyRTP, PF xy is the power factor multiplier for Fxy in the COLR, and P is the fraction of RATED THERMAL POWER at which Fxy was measured.

SALEM - UNIT 1 3/42-6 Amendment No. 299

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENT&JContinued)

d. Remeasuring Fxy according to the following schedule:

1. When Fxyc is greater than the FxyRTP limit for the appropriate measured core plane but less than the FXyL relationship, additional core power distribution measurements shall be taken and FxyC compared to FxyRTP and FxyL..

a) Either within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding by 20% of RATED THERMAL POWER or greater, the THERMAL POWER at which Fxyc was last determined, or b) In accordance with the Surveillance Frequency Control Program, whichever occurs first.

TP

2. When the Fxyc is less than or equal to the Fx/ limit for the appropriate measured core plane, additional core power distribution measurements shall be taken and Fxyc compared to FxyRTP and FxyL in accordance with the Surveillance Frequency Control Program.

TP

e. The Fxy limit for Rated Thermal Power (F x/ ) shall be provided for all core planes containing bank "0" control rods and all unrodded core planes in the COLR per specification 6.9.1.9.

f. The Fxy limits of e, above, are not applicable in the following core plane regions as measured in percent of core height from the bottom of the fuel:

1. Lower core region from 0 to 15% inclusive.

2. Upper core region from 85 to 100% inclusive.

3. Grid plane regions at 17.8 +/- 2%, 32.1 +/- 2%,46.4 +/- 2%,60.6 +/- 2%, and 74.9 +/- 2% inclusive.

4. Core plane regions within +/- 2% of core height (+/- 2.88 inches) about the bank demand position of the bank "0" control rods.

g. Evaluating the effects of Fx( on Fa(Z) to determine if Fa(Z) is within its limit whenever FxyC exceeds Fxy .

SALEM - UNIT 1 3/4 2-7 Amendment No. 299

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) 4.2.2.3 When FQ(Z) is measured pursuant to specification 4.10.2.2, an overall measured FQ(Z) shall be obtained from a core power distribution measurement and increased by the applicable manufacturing and measurement uncertainties as specified in the COLR.

SALEM - UNIT 1 3/4 2-7a Amendment No. 237 UPO 0 6 2KO

THIS PAGE INTENTIONALLY LEFT BLANK.

SALEM - UNIT 1 3/4 2-8 Amendment No. 201 lpZ4& pidb 06 fAl4o NOV 2 6 1997 4 -A AL I

POWER DISTRIBUTION LIMITS NUCLEAR ENTHALPY HOT CHANNEL FACTOR - FNAH LIMITING CONDITION FOR OPERATION 3.2.3 FAH shall be limited by the following relationship:

FAH = F AH [1.0 + PFAH (1.0 - P))

Where: FRTPAH is the limit at RATED THERMAL POWER (RTP) specified in the CORE OPERATING LIMITS REPORT (COLR).

PFAH is the Power Factor Multiplier for FNAH specified in the COLR, and P is THERMAL POWER RATED THERMAL POWER APPLICABILITY: MODE 1 ACTION:

With FNAH exceeding its limit:

a. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High Trip Setpoints to < 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ,

b. Demonstrate thru a core power distribution measurement that FNAH is I within its limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding the limit or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , and

c. Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER above the reduced limit required by a.

or b. above; subsequent POWER OPERATION may proceed provided that FNaH is demonstrated through a core power distribution measurement I to be within its limit at a nominal 50% of RATED THERMAL POWER prior to exceeding this THERMAL POWER, at a nominal 75% of RATED THERMAL POWER prior to exceeding this THERMAL power and within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after attaining 95% or greater RATED THERMAL POWER.

SALEM - UNIT 1 3/4 2-9 Amendment No. 237 t11V 0 6 200D

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS 4.2.3.1 FN~H shall be determined to be within its limit by obtaining a core power distribution measurement:

a. Prior to operation above 75% of RATED THERMAL POWER after each fuel loading, and

b. In accordance with the Surveillance Frequency Control Program.

c. The provisions of Specification 4.0.4 are not applicable.

4.2.3.2 The measured FN~H of 4.2.3.1 above, shall be increased by the applicable FN~H uncertainties specified in the COLR.

SALEM - UNIT 1 Amendment No. 299

POWER DISTRIBUTION LIMITS QUADRANT POWER TILT RATIO LIMITING CONDITION FOR OPERATION 3.2.4 THE QUADRANT POWER TILT RATIO shall not exceed 1.02.

APPLICABILITY: MODE 1 ABOVE 50% OF RATED THERMAL POWER*

ACTION:

a. With the QUADRANT POWER TILT RATIO determined to exceed 1.02 but < 1.09:

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 RATIO to within its limit, or b) Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1.0 and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

2. Verify that the QUADRANT POWER TILT RATIO is within. its limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding the limit or reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High Trip setpoints to < 55% of RATED THERMAL POWER 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 con-dition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL power may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit 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.

b. With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to misalignment of either a shutdown or control rod:

1. Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1.0, within 30 minutes.

2. Verify that the QUADRANT POWER TILT RATIO is within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after exceeding the limit or

See Special Test Exception 3.10.2.

SALEM - UNIT 1 3/4 2-11 Amendment No. A, 16 JUN 1 1979

POWER DISTRIBUTION LIMITING CONDITION FOR OPERATION (Continued) reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High trip Setpoints to :s; 55% of RATED THERMAL POWER 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 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit 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.

c. With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to causes other than the misalignment of either a shutdown or control rod:

1. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High Trip Setpoints to:s; 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

2. Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit 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 at 95% or greater RATED THERMAL POWER.

SURVEILLANCE REQUIREMENTS 4.2.4 The QUADRANT POWER TILT RATIO shall be determined to be within the limit above 50% of RATED THERMAL POWER by:

a. Calculating the ratio in accordance with the Surveillance Frequency Control Program when the alarm is OPERABLE.

b. Calculating the ratio in accordance with the Surveillance Frequency Control Program during steady state operation when the alarm is inoperable.

c. Obtaining a core power distribution measurement to determine the QUADRANT POWER TILT RATIO in accordance with the Surveillance Frequency Control Program when one Power Range Channel is inoperable and THERMAL POWER is > 75 percent of RATED THERMAL POWER.

SALEM - UNIT 1 3/42-12 Amendment No. 299

POWER DISTRIBUTION LIMITS DNB PARAMETERS LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB related parameters shall be maintained within the limits shown on Table 3.2-1:

a. Reactor Coolant System T avg.

b. Pressurizer Pressure

c. Reactor Coolant System Total Flow Rate APPLICABILITY: MODE 1 ACTION:

With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

4.2.5.1 Each of the parameters of Table 3.2-1 shall be verified to be within their limits in accordance with the Surveillance Frequency Control Program.

4.2.5.2 The Reactor Coolant System total flow rate shall be determined to be within the limits of Table 3.2~1 by performing a precision heat balance within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after achieving steady state conditions 2: 90% RATED THERMAL POWER in accordance with the Surveillance Frequency Control Program. The provisions of Specification 4.0.4 are not applicable.

SALEM - UNIT 1 3/42-13 Amendment No. 299

TABLE 3.2-1 DNB PARAMETERS PARAMETER LIMITS 4 Loops In Operation Reactor Coolant System T avg Pressurizer Pressure ~ 2200 psia*

Reactor Coolant System Flow ~ 341,000 gpm#

Limit not applicable during either THERMAL POWER ramp in excess of 5% RATED THERMAL POWER per minute or a THERMAL POWER step in excess of 10% RATED THERMAL POWER.

Includes a 2.4% flow measurement uncertainty plus a 0.1% measurement uncertainty due to feedwater venturi fouling.

SALEM - UNIT 1 3/42-14 Amendment No. 295

3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1.1 As a minimum, the reactor trip system instrumentation channels and interlocks of Table 3.3-1 shall be OPERABLE.

APPLICABILITY: As shown in Table 3.3-1.

ACTION:

As shown in Table 3.3-1.

SURVEILLANCE REQUIREMENTS 4.3.1.1.1 Each reactor trip system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies specified in the Surveillance Frequency Control Program unless otherwise noted in Table 4.3-1.

4.3.1.1.2 The logic for the interlocks shall be demonstrated OPERABLE prior to each reactor startup unless performed during the preceding 92 days. The total interlock function shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program during CHANNEL CALIBRATION testing of each channel affected by interlock operation.

4.3.1.1.3 The REACTOR TRIP SYSTEM RESPONSE TIME of each reactor trip function shall be verified to be within its limit in accordance with the Surveillance Frequency Control Program.

Neutron detectors are exempt from response time testing.

SALEM - UNIT 1 3/43-1 Amendment No. 299

TABLE 3.3-1 R EACTOR TRIP SYSTEM INSTRUMENTATION MINIMUM TOTAL NUMBER CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

1. Manual Reactor Trip 2 i. 2 1,2 and

12

2. Power Range, Neutron Flux 4 2 3 1,2, and 3* 2

3. Power Range, Neutron Flux 4 2 3 1,2 2 High Positive Rate

4. Deleted

5. Intermediate Range, Neutron Flux 2 1 1,2 and

3

6. Source Range, Neutron Flux A. Startup 2 1 2 2## and

4 B. Shutdown 2 0 1 3,4, and 5 5

7. Overtemperature AT 4 2 3 1,2 6

8. Overpower AT 4 2 3 1,2 6

9. Pressurizer Pressure-Low 4 2 3 1,2 6

10. Pressurizer Pressure--High 4 2 3 1,2 6 278 SALEM - UNIT 1 3/4 3-2 Amendment No.

TABLE 3.3-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION MINIMUM TOTAL NUMBER CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

11. Pressurizer Water Level--High 3 2 1, 2 6

12. Loss of Flow - Single Loop 3/loop 2/loop in 2/loop in 1 6 (Above P-8) any oper- each oper-ating loop ating loop 2 2

13. Loss of Flow - Two Loops 3/loop /loop in /loop in 1 6 (Above P-7 and below P-8) two oper- each oper-ating loops ating loop

14. Steam Generator Water Level-- 3/loop 2/loop in 2/loop in 1, 2 6 Low-Low any oper- each oper-ating loops ating loop

15. Deleted

16. Undervoltage-Reactor Coolant Pumps 4-1/bus 1/2 twice 3 6

17. Underfrequency-Reactor Coolant 4-1/bus 1/2 twice 3 1 6 Pumps SALEM - UNIT 1 3/4 3-3 Amendment No. 276

TABLE 3.3-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION INSTRUMENTATION MINIMUM TOTAL NUMBER CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

18. Turbine Trip

a. Low Autostop Oil Pressure 3 2 2 1 6

b. Turbine Stop Valve Closure 4 4 3 1 6

19. Safety Injection Input from ESF 2 1 2 1,2 10

20. Reactor Coolant Pump Breaker 1/breaker 2 1/breaker 1 11 Position Trip (above P-7) per opera-ting loop

21. Reactor Trip Breakers 2 1 2 1, 2 1###, 14 3*,4*,5* 13

22. Automatic Trip Logic 2 1 2. 1, 2 10 3*,4*, 5* 13 SALEM - UNIT 1 3/4 3-4 Amendment No. 276

TABLE 3.3-1 (Continued)

TABLE NOTATION With the reactor trip system breakers in the closed position and the control rod drive system capable of rod withdrawal.

1. High voltage to detector may be de-energized above P-6.

1. 1. If ACTION Statement 1 is entered as a result of Reactor Trip Breaker (RTB) or Reactor Trip Bypass Breakers (RTBB) maintenance testing results exceeding the following acceptance criteria, NRC reporting shall be made within 30 days in accordance with Specification 6.9.2:

1. A RTB or RTBB trip failure during any surveillance test with less than or equal to 300 grams of weight added to the breaker trip bar.

2. A RTB or RTBB time response failure that results in the overall reactor trip system time response exceeding the Technical Specification limit.

ACTION STATEMENTS ACTION 1 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.1.1.1 provided the other channel is OPERABLE.

ACTION 2 With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a. The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing of other channels per Specification 4.3.1.1.1.

c. Either, THERMAL POWER is restricted to _*75% of RATED THERMAL POWER and the Power Range, Neutron Flux trip setpoint is reduced to *< 85% of RATED THERMAL POWER within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; or, the QUADRANT POWER TILT RATIO is monitored at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SALEM U-NIT 1 3/4 3-5 Amendment No 2 7 6

TABLE 3.3-1 (Continued)

ACTION 3 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement and with the THERMAL POWER level:

a. Below P-6, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the P-6 Setpoint.

b. Above P-6 but below 5% of RATED THERMAL POWER, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above 5% of RATED THERMAL POWER.

c. Above 5% of RATED THERMAL POWER, POWER OPERATION may continue.

ACTION 4 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement and with the THERMAL POWER level:

a. Below P-6, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the P-6 Setpoint.

b. Above P-6, operation may continue.

ACTION 5 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, verify compliance with the SHUTDOWN MARGIN requirements of Specification 3.1.1.1 or 3.1.1.2, as applicable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter.

ACTION 6 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a. The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing of other channels per Specification 4.3.1.1.1.

ACTION 7 - NOT USED I

ACTION 8 - NOT USED ACTION 9 - NOT USED I SALEM - UNIT 1 3/4 3-6 Amendment No. 142 AUG 4 1993

TABLE 3.3-1 (Continued)

ACTION 10 - With the number of OPERABLE Channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in at least HOT STANDBY in the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing per Specification 4.3.1.1.1, provided the other channel is OPERABLE.

ACTION 11 - With less than the Minimum Number of Channels OPERABLE, operation may continue provided the inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 12 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and/or open the reactor trip breakers.

ACTION 13 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or open the reactor trip breakers within the next hour.

ACTION 14 - With one of the diverse trip features (Undervoltage or shunt trip attachment) inoperable, restore it to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or declare the breaker inoperable and be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . The breaker shall not be bypassed while one of the diverse trip features is inoperable except for the time required for performing maintenance to restore the breaker to OPERABLE status.

REACTOR TRIP SYSTEM INTERLOCKS DESIGNATION CONDITION AND SETPOINT FUNCTION P-6 With 2 of 2 Intermediate Range P-6 prevents or defeats Neutron Flux Channels < 6x10-1 the manual block of source amps. range reactor trip.

P-7 With 2 of 4 Power Range Neutron P-7 prevents or defeats Flux Channels 2 11% of RATED the automatic block of THERMAL POWER or 1 of 2 Turbine reactor trip on: Low flow steam line inlet pressure channels 2 a pressure equivalent to 11% of in more than one primary coolant loop, reactor I RATED THERMAL POWER. coolant pump undervoltage and under-frequency, pressurizer low pressure, pressurizer high level, and the opening of more than one reactor coolant pump breaker.

SALEM - UNIT 1 3/4 3-7 Amendment No. 259

TABLE 3.3-1 'Continued)

IES IGNAT ION CONDITION AND SETPO INT FUNCTION P-8 With 2 of 4 Power Range Neutron P-8 prevents or defeats Flux channels > 36% of RATED the automatic block of THERMAL POWER. reactor trip on low coolant flow in a single loop.

P-9 With 2 of 4 Power Range neutron P-9 prevents or defeats flux channels > 50% RATED the automatic block of THERMAL POWER.- reactor trip on turbine trip.

P-10 With 3 of 4 Power range neutron P-10 prevents or defeats flux channels < 9% of RATED the manual block of:

THERMAL POWER. Power range low setpoint reactor trip, Intermediate range reactor trip, and intermediate range rod stops.

Provides input to P-7.

K)

K) SALEM - UNIT 1 3/4 3-8 Amendment No. 85 I M 27 E

1G,

%W p0Ad& w6ww AW&WtcL IL &Zd0 I 8Mazy we ze dv k 4uee /9

Page Left Blank Intentionally SALEM - UNIT 1 3/4 3-9 Amendment No283

Page Left Blank Intentionally SALEM - UNIT I 3/4 3-10 Amendment No. 283

TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS MODES IN CHANNEL WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK(15) CALIBRATION(15) TEST(15) REQUIRED

1. Manual Reactor Trip Switch N.A N.A (9) 1,2,and*

2. Power Range, Neutron Flux (2), (3) (6) 1,2, and 3*

3. Power Range, Neutron Flux, N.A (6) 1,2 High Positive Rate

4. Deleted

5. Intermediate Range, Neutron (6) S/U(1) 1,2 and ..

Flux

6. Source Range, Neutron Flux (7) (6) (16)and 2,3,4,5 S/U(l) and ..

7. Overtemperature liT 1,2

8. Overpower liT 1,2

9. Pressurizer Pressure--Low 1,2

10. Pressurizer Pressure--High 1,2

11. Pressurizer Water 1,2 Level--High

12. Loss of Flow - Single Loop 1 SALEM - UNIT 1 3/43-11 Amendment No. 299

TABLE 4.3-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS MODES IN CHANNEL WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK(15) CALI BRATION(15) TEST(15) REQUIRED

13. Loss of Flow Two Loops NA

14. Steam Generator Water 1,2 Level--Low-Low

15. DELETED

16. Undervoltage - Reactor NA 1 Coolant Pumps

17. Underfrequency - Reactor NA 1 Coolant Pumps

18. Turbine Trip

a. Low Autostop Oil N.A. NA S/U(1) 1,2 Pressure

b. Turbine Stop Valve NA N.A. S/U(1) 1,2 Closure

19. Safety Illjection Input from NA NA (4)(5) 1,2 ESF

20. Reactor Coolant Pump NA NA Breaker Position Trip

21. Reactor Trip Breaker NA NA (5)(11 )(13) 1,2 and *

(14)

22. Automatic Trip Logic N.A. NA (5) 1,2 and

SALEM - UNIT 1 3/43-12 Amendment No. 299

TABLE 4.3-1 (Continued)

NOTATION

With the reactor trip system breakers closed and the control rod drive system capable of rod withdrawal.

(1) If not performed in previous 31 days.

(2) Heat balance only, above 15% of RATED THERMAL POWER.

(3) Compare incore to excore axial offset above 15% of RATED THERMAL POWER.

Recalibrate if absolute difference 2: 3 percent.

(4) Manual SSPS functional input check in accordance with the Surveillance Frequency Control Program.

(5) Each train or logic channel shall be tested in accordance with the Surveillance Frequency Control Program.

(6) - Neutron detectors may be excluded from CHANNEL CALIBRATION.

(7) Below P-6 (Block of Source Range Reactor Trip) setpoint.

(8) - Deleted (9) The CHANNEL FUNCTIONAL TEST shall independently verify the OPERABILITY of the Undervoltage and Shunt Trip mechanism for the Manual Reactor Trip Function.

The Test shall also verify OPERABILITY of the Bypass Breaker Trip circuits.

(10) DELETED (11) The CHANNEL FUNCTIONAL TEST shall independently verify the OPERABILITY of the Reactor Trip Breaker Undervoltage and Shunt Trip mechanisms.

(12) DELETED (13) Verify operation of Bypass Breakers Shunt Trip function from local pushbutton while breaker is in the test position prior to placing breaker in service.

(14) Perform a functional test of the Bypass Breakers UV. Attachment via the SSPS.

(15) Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table.

(16) At the frequency specified in the Surveillance Frequency Control Program.

SALEM - UNIT 1 3/43-13 Amendment No. 299

INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2.1 The Engineered Safety Feature Actuation System (ESFAS) instrumentation channels and interlocks shown in Table 3.3-3 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4.

APPLICABILITY: As shown in Table 3.3-3.

ACTION:

a. With an ESFAS instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3-4, declare the channel inoperable and apply the applicable ACTION requirement of Table 3.3-3 until the channel is restored to OPERABLE status with the trip setpoint adjusted consistent with the Trip Setpoint value.

b. With an ESFAS instrumentation channel inoperable, take the ACTION shown in Table 3.3-3.

SURVEILLANCE REQUIREMENTS 4.3.2.1.1 Each ESFAS instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies specified in the Surveillance Frequency Control Program unless otherwise noted in Table 4.3-2.

4.3.2.1.2 The logic for the interlocks shall be demonstrated OPERABLE during the automatic actuation logic test. The total interlock function shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program during CHANNEL CALIBRATION testing of each channel affected by interlock operation.

4.3.2.1.3 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function shall be verified to be within the limit in accordance with the Surveillance Frequency Control Program. The provisions of Specification 4.0.4 are not applicable to MSIV closure time testing.

The provisions of Specification 4.0.4 are not applicable to the turbine driven auxiliary feedwater pump provided the surveillance is performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the secondary steam generator pressure is greater than 680 psig.

SALEM - UNIT 1 3/4 3-14 Amendment No. 299

TABLE 3.3-3 ENGI]NEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLIICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MO[ DES ACTION

1. SAFETY INJECTION, TURBINE TRIP AND FEEDWATER ISOLATION

a. Manual Initiation 2 2 1,2,3,4 18

b. Automatic Actuation Logic 2 2 1,2,3,4 13

c. Containment Pressure-High 3 2 2 1,2,3 19

d. Pressurizer Pressure-Low 3 2 2 1,2, 3# 19

e. Differential Pressure Between 3/steam line 2/steam line 2/steam 1,2, 3## 19 Steam Lines - High any steam line line

f. Steam Flow in Two Steam 2/steam line 1/steam line 1/steam 1,2, 3## 19 Lines-High any2 steam line lines COINCIDENT WITH EITHER Tavg -- Low-Low I Tavg/loop 1 Tavg in 1 Tavg 1,2, 3## 19 any 2 loops in any 3 loops OR, COINCIDENT WITH Steam Line Pressure-Low 1 pressure/ 1 pressure 1 pressure 1,2, 3## 19 loop any 2 loops any 3 loops 27 6 SALEM - UNIT 1 3/4 3-15. Amendment No.

This page is intentionally left blank.

I SALEM - UNIT 1 3/4 3-16 Amendment No. 138 I JAN 5 1993

C c--/

TABLE 3.3-3 (Continued)

C C: ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM c

1-43 TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 1-J

2. CONTAINMENT SPRAY

a. Manual 2 sets of 2 1 set of 2 2 sets of 2 1,2,3,4 18

b. Automatic Actuation Logic 2 1 2 1,2,3,4 13

c. Containment Pressure--High-High 4 2 3 1,2,3 16

3. CONTAINMENT ISOLATION

a. Phase "A" Isolation w

1) Manual 2 1 2 1,2,3,4 18 w

2) From Safety Injection 2 1 2 1,2,3,4 13 I-A

-J Automatic Actuation Logic

rn TABLE 3.3-3 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE H FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

b. Phase "B" Isolation

1) Manual 2 sets of 2 1 set of 2 2 sets of 2 1,2,3,4 18

2) Automatic Actuation Logic 2 1 2 1,2,3,4 13

3) Containment Pressure--High-High 4 2 3 1,2,3 16

c. Containment Ventilation Isolation I

1) Manual 2 1 2 1,2,3,4 17

2) Automatic Actuation Logic 2 1 2 1,2,3,4 13 1Aj co l

3) Containment Atmosphere Gaseous per table 3. 3-6 Radioactivity-High

TABLE 3.3-3 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS A]PPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

4. STEAM LINE ISOLATION

a. Manual 2/steam line 1/steam line 1/operating 1,2,3 23 steam line

b. Automatic Actuation Logic 1 2 1,2,3 20 C. Containment Pressure--High-High 4 2 3 1,2,3 16

d. Steam Flow in Two Steam 2/steam line 1/steam line I/steam line 1,2, 3## 19 Lines--High any 2 steam lines COINCIDENT WITH EITHER Tavg--Low-Low 1 Tavg/loop 1 Tavg in 1 Tavg in 1,2, 3## 19 any 2 loops any 3 loops OR, COINCIDENT WITH Steam Line Pressure-Low 1 pressure/ 1 pressure 1 pressure 1,2, 3## 19 loop any 2 loops any 3 loops 7

SALEM - UNIT 1 3/4 3-19 Amendment No. 2 6

TABLE 3.3-3 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

5. TURBINE TRIP & FEEDWATER ISOLATION 3

a. Steam Generator Water level-- /loop 2/loop in 2/loop in 1,2,3 19 High-High any operating each operating loop loop

6. SAFEGUARDS EQUIPMENT 3 2 3 1,2,3,4 13 CONTROL SYSTEM (SEC)

7. UNDERVOLTAGE, VITAL BUS

a. Loss of Voltage 1/bus 2 3 1,2,3 14

b. Sustained Degraded Voltage 3/bus 2/bus 3/bus 1,2,3 14 SALEM - UNIT 1 3/4 3-20 , Amendment No. 276

TABLE 3.3-3 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

8. AUXILIARY FEEDWATER

a. Automatic Actuation Logic **2 1 2 1,2,3 20

b. NOT USED

c. Steam Generator Water Level--Low-Low

i. Start Motor Driven Pumps 3/stm. gen. 2/stm. gen. 2/stm.gen. 1,2,3 19 any stm.gen.

ii.Start Turbine Driven Pumps 3/stm. gen. 2/stm. gen. 2/stm.gen. 1,2,3 19 any 2 stm.gen.

d. Undervoltage - RCP Start 4-1/bus 1/2 x 2 3 1,2 19 Turbine - Driven Pump

e. S.I. Start Motor-Driven Pumps See 1 above (All S.I. ini tiating functions and requirements)

f. Trip of Main Feedwater Pumps 2/pump 1/pump l/pump 1,2 21 Start Motor Driven Pumps

g. Station Blackout See 6 and 7 above (SEC and U/V Vital Bus)

SALEM - UNIT 1 3/4 3-20a Amendment No. 276

TABLE 3.3-3 (Continued)

TABLE NOTATION

Trip function may be bypassed in this MODE below P-1i.

1. Trip function may .be bypassed in this MODE below P-12.

- Applies to Functional Unit 8 items c and d.

- - The automatic actuation logic includes two redundant solenoid operated vent valves for each Main Steam Isolation Valve,. One vent valve on any one Main Steam Isolation Valve may be isolated without affecting the function of the automatic actuation logic provided the remaining seven solenoid vent valves remain OPERABLE. The isolated MSIV vent valve shall be returned to OPERABLE status upon the first entry into MODE 5 following determination that the vent valve is inoperable. For any condition where more than one of the eight solenoid vent valves are inoperable, entry into ACTION 20 is required.

ACTION STATEMENTS ACTION 13 - With the number of OPERABLE Channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or, be in 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 ; however, one channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing per Specification 4.3.2.1.1 provided the other channel is OPERABLE.

ACTION 14 - With the number of OPERABLE Channels one less than the Total Number of Channels, operation may proceed until performance of the next required CHANNEL FUNCTIONAL TEST, provided the inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

ACTION 15 - NOT USED ACTION 16 - With the number of OPERABLE Channels one less than the Total Number of Channels, operation may proceed provided the inoperable channel is placed in the bypassed condition and the Minimum Channels OPERABLE requirement is demonstrated by CHANNEL CHECK within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; one additional channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing per Specification 4.3.2.1.1.

ACTION 17 - With less than the Minimum Channels OPERABLE, operations may continue provided the containment purge and exhaust valves are maintained closed.

ACTION 18 - With the number of OPERABLE Channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SALEM - UNIT 1 3/4 3-21 Amendment No. 276

TABLE 3.3-3 (Continued)

ACTION 19 - With the number of OPERABLE Channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a. The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. The Minimum Channels OPERABLE requirements is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing of other channels per Specification 4.3.2.1.1.

ACTION 20 - With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing per Specification 4.3.2.1.1 provided the other channel is OPERABLE.

ACTION 21 - With the number of OPERABLE channels one less than the Minimum Number of Channels, operation may proceed provided that the inoperable channel is restored to OPERABLE within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

ACTION 22 - NOT USED ACTION 23 - With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . I SALEM - UNIT 1 3/4 3-22 Amendment No.225 NtoV 0 2 133

TABLE 3.3-3 (Continued)

ENGINEERED SAFETY FEATURES INTERLOCKS DESIGNATION CONDITION AND SETPOINT FUNCTION P-li With 2 of 3 pressurizer P-li prevents or defeats pressure channels manual block of safety 2 1925 psig. injection actuation on low pressurizer pressure.

P-12 With 3 of 4 Tavz channels P-12 prevents or defeats at a setpoint of 5430 F and manual block of safety TaV increasing injection actuation high (with an allowable setpoint steam line flow and low value z545 0 F) steam line pressure.

With 2 of 4 TaVg channels Allows manual block of at a setpoint of 5430 F and safety injection actuation TV decreasing on high steam line flow and (with an allowable setpoint low steam line pressure.

value k541 0F) Causes steam line isolation on high steam flow. Affects steam dump blocks.

K)

SALEM - UNIT 1 3/4 3-22a Amendment No.185 NOV 4 M9

A CI TABLE 3.3-4 c

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES

1. SAFETY INJECTION, TURBINE TRIP AND FEEDWATER ISOLATION

a. Manual Initiation Not Applicable Not Applicable

b. Automatic Actuation Logic Not Applicable Not Applicable

c. Containment Pressure--High s4.0 psig s4.5 psig

d. Pressurizer Pressure--Low 21765 psig z1755 psig

e. Differential Pressure slOO psi .112 psi Between Steam Lines--High

f. Steam Flow in Two Steam Lines-- sA function defined as sA function defined as High Coincident with T 0,g--Low-Low -follows: A Ap corresponding follows: A Ap corresponding or Steam Line Pressure--Low to 40% of full steam flow to 44% of full steam flow between 0% and 20% load between 0% and 20% load and

and then a Ap increasing then a Ap increasing linearly linearly to a Ap correspond- to a Ap corresponding to ing to 110% of full steam 111.5% of full steam flow flow at full load at full load Tayg z 543 0F Tavg a 541°F Z 600 psig steam line > 579 psig steam line I pressure pressure SALEM - UNIT I 3/4 3-23 Amendment No. 187

- e napd Q 1cb - Qt 0 lt a np~tM~ pa*, ~JAN 1 7 1997 14W , ' 7LC a 3 '+7 A ff Sa t v-4,L.

C C C TABLE 3.3-4 (Continued)

ENdrNEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES

2. CONTAINMENT SPRAY

a. Manual Initiation Not Applicable Not Applicable

b. Automatic Actuation Logic Not Applicable Not Applicable

c. Containment Presuure--High-High S 15.0 pslg S 16.0 psig I

3. CONTAINMENT ISOLATION

a. Phase "A Isolation

1. Manual Not Applicable Not Applicable

2. From Safety Injection Not Applicable Not Applicable Automatic Actuation Logic

b. Phase "BO Isolation

1. Manual Not Applicable Not Applicable

2. Automatic Actuation Logic Not Applicable Not Applicable

3. Containment Pressure--High-High S 15.0 psig s 16.0 paig I

c. Containment Ventilation Isolation

1. Manual Not Applicable Not Applicable

2. Automatic Actuation Logic Not Applicable Not Applicable

C C C TABLE 3.3-4 (Continued)

VuNi.Ykrn cVrpTY What1DP ACTUATION SYSTEM INSTRUHENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES

3. Containment Atmosphere Per Table 3.3-6 Gaseous Radioactivity

4. STEAM LINE ISOLATION

a. Manual Not Applicable Not Applicable

b. Automatic Actuation Logic Not Applicable Not Applicable

c. Containment Pressure--High-High S 15.0 psig 5 16.0 psig I

d. Steam Flow In Two Steam Lines-- S A function defined as S A function defined as High Coincident with follows: A Ap corresponding follows: A Ap corresponding Tavg -- Low-Low or to 40% of full steam flow to 44% of full steam flow Steam Line Pressure -- Low between OS and 20% load and between O and 20% load and then a Ap increasing linearly then a Ap increasing linearly to a Ap corresponding to to a Ap corresponding to 110% of full steam flow at 111.5% of full steam flow at full load. full load.

T avg 2 543@F T avg 2 541OF 2 600 pslg steam line 2 579 psig steam line pressure pressure

TABLE 3.3-4 (continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES

5. TURBINE TRIP AND FEEDWATER ISOLATION A. Steam Generator Water Level -- 5 67% of narrow range S 68% of narrow range High-High instrument span each instrument span each steam generator steam generator

6. SAFEGUARDS EQUIPMENT CONTROL Not Applicable Not Applicable SYSTEM (SEC)

7. UNDERVOLTAGE, VITAL BUS

a. Loss of Voltage 2 70% of bus voltage 2 65% of bus voltage

b. Sustained Degraded Voltage 2 94.6% of bus voltage for 2 94% of bus voltage for 5 13 seconds 5 15 seconds

8. AUXILIARY FEEDWATER

a. Automatic Actuation Logic Not Applicable Not Applicable

b. NOT USED

c. Steam Generator Water Level-- 2 14.0% of narrow range 2 13.0% of narrow range I Low-Low instrument span each instrument span each steam generator steam generator

d. Undervoltage - RCP 2 70% RCP bus voltage 2 65% RCP bus voltage

e. S.I. See 1 above (All S.I. setpoints)

f. Trip of Main Feedwater Pumps Not Applicable Not Applicable

g. Station Blackout See 6 and 7 above (SEC and Undervoltage, Vital Bus)

S - UNIT 1 3/4 -26 Amendment No2; 7 I 9J

Page Left Blank Intentionally SALEM - UNIT 1 3/4 3-27 Amendment No. 283

Page Left Blank Intentionally SALEM - UNIT 1 3/4 3-28 Amendment No.283

Page Left Blank Intentionally SALEM - UNIT 1 3/4 3-29 Amendment No.283

Page Left Blank Intentionally SALEM - UNIT 1 3/4 3-30 Amendment No.283

Page Left Blank Intentionally SALEM - UNIT 1 3/4 3-31 Amendment No. 283

TABLE 4.3-2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS MODES IN CHANNEL WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK(7) CALIBRATION(7) TEST(7) REQUIRED

1. SAFETY INJECTION.

TURBINE TRIP AND FEEDWATER ISOLATION

a. Manual Initiation NA NA 1,2,3,4

b. Automatic Actuation Logic NA NA (2) 1,2,3,4

c. Containment Pressure-High (3) 1,2,3

d. Pressurizer Pressure--Low 1,2,3

e. Differential Pressure 1,2,3 Between Steam Unes--High

f. Steam Flow in Two Steam 1,2,3 Lines--High coincident with Tavg--Low-Low or Steam Line Pressure-Low

2. CONTAINMENT SPRAY

a. Manual Initiation NA NA 1,2,3,4

b. Automatic Actuation Logic NA N.A. (2) 1,2,3,4

c. Containment (3) 1,2,3 Pressu re--H i9 h-H ig h SALEM - UNIT 1 3/43-31a Amendment No. 299

TABLE 4.3-2 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS MODES IN CHANNEL WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK(7) CALIBRATION(?) TEST(?) REQUIRED

3. CONTAINMENT ISOLATION

a. Phase "A" Isolation

1. Manual NA NA 1,2,3,4

2. From Safety Injection NA NA (2) 1,2,3,4 Automatic Actuation Logic

b. Phase "8" Isolation

1. Manual NA NA 1,2,3,4

2. Automatic Actuation NA NA (2) 1,2,3,4 Logic

3. Containment Pressure- (3) 1,2,3

-High-High

c. Containment Ventilation Isolation

1. Manual NA NA 1,2,3,4

2. Automatic Actuation NA NA (2) 1,2,3,4 Logic

3. Containment Per Surveillance Requirement 4.3.3.1 Atmosphere Gaseous Radioactivity - High SALEM - UNIT 1 3/43-32 Amendment No. 299

TABLE 4.3-2 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS MODES IN CHANNEL WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK(7) CALI BRATION(7) TEST(7) REQUIRED

4. STEAM LINE ISOLATION

a. Manual N.A. N.A. 1,2,3**

b. Automatic Actuation Logic N.A. N.A. (2) 1,2,3

c. Containment (3) 1,2,3 Pressure--High-High

d. Steam Flow in Two Steam 1,2,3 Lines--High Coincident with Tavg--Low-Low or Steam Line Pressure--Low

5. TURBINE TRIP AND FEEDWATER ISOLATION

a. Steam Generator Water 1,2,3 Level--High-High

6. SAFEGUARDS EQUIPMENT CONTROL SYSTEM (SEC)

LOGIC

a. Inputs N.A. N.A. (6) 1,2,3,4

b. Logic, Timing and Outputs

N.A. N.A. (1 ) 1,2,3,4

7. UNDERVOLTAGE, VITAL BUS

a. Loss of Voltage 1,2,3

b. Sustained Degraded 1,2,3 Voltage SALEM - UNIT 1 3/43-32a Amendment No. 299

TABLE 4.3-2 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS MODES IN CHANNEL WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK(7) CALIBRATION(7) TEST(7) REQUIRED

8. AUXILIARY FEEDWATER

a. Automatic Actuation Logic NA NA (2) 1,2,3

b. NOT USED

c. Steam Generator Water 1,2,3 Level--Low-Low

d. Undervoltage - RCP 1,2 e, S.1. See 1 above (All S.1. surveillance requirements)

f. Trip of Main Feedwater NA NA Pumps

g. Station Blackout See 6b and 7 above (SEC and UN Vital Bus)

SALEM - UNIT 1 3/43-33 Amendment No. 299

TABLE 4.3-2 (Continued)

TABLE NOTATION

Outputs are up to, but not including, the output relays.

- The provisions of Specification 4.0.4 are not applicable.

(1 ) Each logic channel shall be tested in accordance with the Surveillance Frequency Control Program. The CHANNEL FUNCTION TEST of each logic channel shall verify that its associated diesel generator automatic load sequence timer is OPERABLE with the interval between each load block within 1 second of its design interval.

(2) Each train or logic channel shall be tested in accordance with the Surveillance Frequency Control Program.

(3) The CHANNEL FUNCTIONAL TEST shall include exercising the transmitter by applying either a vacuum or pressure to the appropriate side of the transmitter.

(4) NOT USED (5) NOT USED (6) Inputs from Undervoltage, Vital Bus, shall be tested in accordance with the Surveillance Frequency Control Program. Inputs from Solid State Protection System shall be tested in accordance with the Surveillance Frequency Control Program.

(7) Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table.

SALEM - UNIT 1 3/43-34 Amendment No. 299

INSTRUMENTATION 3/4.3.3 MONITORING INSTRUMENTATION RADIATION MONITORING INSTRUMENTATION 3.3.3.1 The radiation monitoring instrumentation channels shown in Table 3.3-6 shall be OPERABLE with their alarm/trip setpoints within the specified limits.

APPLICABILITY: As shown in Table 3.3-6.

ACTION:

a. With a radiation monitoring channel alarm/trip setpoint exceeding the value shown in Table 3.3-6, adjust the setpoint to within the limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or declare the channel inoperable.

b. With one or more radiation monitoring channels inoperable, take the ACTION shown in Table 3.3-6.

c. The provisions of Specification 3.0.3 are not applicable.

4.3.3.1 Each radiation monitoring instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations in accordance with the Surveillance Frequency Control Program.

SALEM - UNIT 1 3/43-35 Amendment No. 299

TABLE 3.3-6 RADIATION MONITORING INSTRUMENTATION MINIMUM CHANNELS APPLICABLE ALARM/TRIP MEASUREMENT INSTRUMENT OPERABLE MODES SETPOINT RANGE ACTION

1. AREA MONITORS

a. Fuel Storage Area *15 mR/hr 10-1-104 mR/hr 19 I

2. PROCESS MONITORS

a. Containment

1) Gaseous Activity a) Purge & Pressure - 1# 1,2,3,4&5 per ODCM 101_106 cpm 23 Vacuum Relief Control 3.3.3.9 Isolation b) RCS Leakage 1 1,2,3&4 N/A 101-106 cpm 20 Detection

2) Air Particulate Activity a) (NOT USED) b) RCS Leakage 1 1,2, 3&4 N/A 101_106 CpM 20 Detection With fuel in the storage pool or building.

The plant vent noble gas monitor may also function in this capacity when the purge/pressure-vacuum relief isolation valves are open.

SALEM - UNIT 1 3/4 3-36 Amendment No.272

TABLE 3.3-6 (Continued)

RADIATION MONITORING INSTRUMENTATION MINIMUM CHANNELS APPLICABLE ALARMITRIP MEASUREMENT INSTRUMENT OPERABLE MODES SETPOINT RANGE ACTION

2. PROCESS MONITORS

b. Noble Gas Effluent Monitors

1) Medium Range Auxiliary Building 1 1,2,3&4 s 3.Ox1 0-2 IJCi/cm3 10-3_10 1 IJCi/cm3 23 Exhaust System (Plant Vent) (Alarm only)

2) High Range Auxiliary Building 1 1,2,3&4 s 1.Ox1 02 IJCi/cm3 10-1_105 IJ Ci/cm 3 23 Exhaust System (Plant Vent) (Alarm only)

3) Condenser Exhaust System 1 1,2,3&4 s 1.27x1 04 cpm 1-10 6 cpm 23 (Alarm only)

3. CONTROL ROOM

a. Air Intake - Radiation Level 2/lntake## ** s 2.48x10 3 cpm 101-107 cpm 24, 25

1. Control Room air intakes shared between Unit 1 and 2.

- ALL MODES and during movement of irradiated fuel assemblies and during CORE ALTERATIONS.

SALEM - UNIT 1 3/43-36a Amendment No. 295

TABLE 3.3-6 (Continued)

TABLE NOTATION ACTION 19- With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, perform area surveys of the monitored area with portable monitoring instrumentation at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 20- With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.4.6.1.

ACTION 22 - (Not Used)

ACTION 23 - With the number of OPERABLE Channels less than required by the Minimum Channels OPERABLE requirements, initiate the preplanned alternate method of monitoring the appropriate parameter(s),

within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , and:

1) either restore the inoperable Channel(s) to OPERABLE status within 7 days of the event, or

2) prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 14 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

ACTION 24- With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel(s) to OPERABLE status within 7 days or initiate and maintain operation of the Control Room Emergency Air Conditioning System (CREACS) in the pressurization or recirculation mode of operation. CORE ALTERATIONS and movement of irradiated fuel assemblies will be suspended during operation in the recirculation mode.

ACTION 25 - With no channels OPERABLE In a Control Room air intake, immediately initiate and maintain operation of the CREACS in the pressurization or recirculation mode of operation. CORE ALTERATIONS and movement of irradiated fuel assemblies will be suspended during operation in the recirculation mode.

SALEM - UNIT 1 314 3-37 Amendment No. 263

TABLE 4.3-3 DELETED SALEM - UNIT 1 3/43-38 Amendment No. 299

PAGES 3/4 3-39 THROUGH 3/4 3-45 ARE DELETED SALEM - UNIT 1 3/4 3-39 through 3/4 3-45 Amendment No .282

INSTRUMENTATION REMOTE SHUTDOWN INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.5 The remote shutdown monitoring instrumentation channels shown in Table 3.3-9 shall be OPERABLE with readouts displayed external to the control room.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

a. With the number of OPERABLE remote shutdown monitoring channels less than required by Table 3.3-9, either restore the inoperable channel to OPERABLE status within 30 days or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE REQUIREMENTS 4.3.3.5 Each remote shutdown monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, and CHANNEL CALIBRATION operations in accordance with the Surveillance Frequency Control Program.

SALEM - UNIT 1 3/43-46 Amendment No. 299

-r.

( .d CA I-TABLE 3.3-9 X3.

REMOTE SHUTDOWN 1MONITORING INSTRUMENTATION.

MINIMUM

-1 READOUT MEASURE MENT CHANNELS INSTRUMENT LOCATION RANG E OPERABLE 1.. Pressurizer Pressurre Hot Shutdown Panel 213 1700-25 00 psig 1

2. Pressurizer Level Hot Shutdown Panel 213 0 - 100%

3.. Steam Generator PrE!ssure Hot Shutdown Panel 213 0 - 1200 psig 1/steam generator

4. Steam Generator Le%rel Hot Shutdown Panel 213 0 - 100% 1/steam generator (A.Y CD C2 co

-o -

0 I

TABLE 4.3-6 DELETED SALEM - UNIT 1 3/43-48 Amendment No. 299

PAGES 3/4 3-49 'IffPUG 3/4 3-52 ARE WrECNIONALTI BLANK I

SAIM - UNIT 1 3/4 3-49 through 3-52 Ameilnent No. 139 JAN 5 1993

INSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.7 The accident monitoring instrumentation channels shown in Table 3.3-11 shall be operable.

APPLICABILITY: MODES 1,2, and 3.

ACTION:

a. As shown in Table 3.3-11.

SURVEILLANCE REQUIREMENTS 4.3.3.7 Each accident monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies specified in the Surveillance Frequency Control Program unless otherwise noted in Table 4.3-11.

SALEM - UNIT 1 3/43-53 Amendment No. 299

C C.

TABLE 3.3-11 AlCCIDENT MONITORING INSTRUMENTATION REQUIRED MINIMUM NO. OF No. OF I NSTRUMENT CHANNELS CHANNELS ACT ION

1. Reactor Coolant Outlet Temperature 2 1 1 2 TNOT (Wide Range)

2. Reactor Coolant Inlet Temperature - 2 I 1. 2 TCOLD (Wide Range)

3. Reactor Coolant Pressure (Wide Range) 2 1 I1* 2

4. Pressurizer water Level 2 I1* 2

5. Steam Line Pressure 2/Steam Generator 1/Steam Generator 2

6. Steam Generator Water Level (Narrow 2/Steam Generator 1/Steam Generator I, 2 Range)

7. Steam Generator Uater Level (Wide 4 (1/Steam Generator) 3 (1/Steam Generator) 1, 2 Range)

S. Refueling Water Storage Tank Water 2 I 1, 2 Leve I

9. deleted I

10. Auxiliary Feedwater Flow Rate 4 (1/Steam Generator) 3 (1/Steam Generator) 4, 6

11. Reactor Coolant System Subcooling 2 1 1* 2 Margin Monitor

12. PORV Position Indicator 2/valIve** I 1, 2 1 47 SALEM - UNIT 1 3/4 3-54 Amendment No.

NOV 1 6 1993

TABLE 3.3-11 (CONTINUED)

ACCIDENT MONITORING INSTRUMENTATION REQUIRED MINIMUM NO. OF NO. OF INSTRUMENT CHANNELS CHANNELS ACTION

13. PORV Block Valve Position Indicator 2/valve** 1 1, 2

14. Pressurizer Safety Valve Position 2/valve** 1 1, 2 Indicator

15. Containment Pressure - Narrow Range 2 1 1, 2

16. Containment Pressure - Wide Range 2 1 7, 2

17. Containment Water Level - 2 1 7, 2 Wide Range

18. Core Exit Thermocouples 4/core quadrant 2/core quadrant 1, 2

19. Reactor Vessel Level Instrumentation 2 1 8, 9 System (RVLIS)

20. Containment High Range Accident 2 2 10 Radiation Monitor

21. Main Steamline M/ 1/ 10 Discharge (Safety MS Line MS Line Valves and Atmospheric Steam Dumps) Monitor

(**'I Total number of channels is considered to be two (2) with one (1) of the channels being any one (1) of the following alternate means of determining PORV, PORV Block, or Safety Valve position: Tailpipe Temperatures for the valves, Pressurizer Relief Tank Temperature Pressurizer Relief Tank Level OPERABLE.

280 SALEM - UNIT 1 3/4 3-55 Amendment No.

TABLE 3.3-11 (continued)

TABLE NOTATION ACTION 1 With the number of OPERABLE accident monitoring channels less than the Required Number of Channels shown in Table 3.3-11, restore the inoperable channel(s) to OPERABLE status within 7 days, or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 2 With the number of OPERABLE accident monitoring channels less than the MINIMUM Number of Channels shown in Table 3.3-11, restore the inoperable channel(s) to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 3 deleted ACTION 4 With the number of OPERABLE channels one less than the Required Number of Channels shown in Table 3.3-11, operation may proceed I provided that an OPERABLE Steam Generator Wide Range Level channel is available as an alternate means of indication for the Steam Generator with no OPERATABLE Auxiliary Feedwater Flow Rate channel.

ACTION 5 deleted SALEM - UNIT 1 3/4 3-56 Amendment No. 225 NOV 0 2 19q9

TABLE 3.3-11 (continued)

TABLE NOTATION ACTION 6 With the number of OPERABLE channels less than the Minimum Number of channels shown in Table 3.3-11, restore the inoperable channel(s) to OPERABLE status within 7 days, or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 7 With the number of OPERABLE channels one less than the Required Number of Channels shown in Table 3.3-11, operation may proceed until the next CHANNEL CALIBRATION (which shall be performed upon the next entry into MODE 5, COLD SHUTDOWN).

ACTION 8 With one RVLIS channel inoperable, restore the RVLIS channel to OPERABLE status within 30 days, or submit a special report in accordance with Specification 6.9.4.

ACTION 9 With both RVLIS channels inoperable, restore one channel to OPERABLE status within 7 days or submit a special report in accordance with Specification 6.9.4.

ACTION 10 With the number of OPERABLE Channels less than required by the minimum channels OPERABLE requirements, initiate the preplanned alternate method of monitoring the appropriate parameter within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , and:

1) either restore the inoperable Channel(s) to OPERABLE status within 7 days of the event, or

2) prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 14 days following the event outlining the actions taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

SALEM - UNIT 1 3/4 3-56a Amendment No.272

TABLE 4.3-11 SURVEILLANCE REQUIREMENTS FOR ACCIDENT MONITORING INSTRUMENTATION CHANNEL CHANNEL CHANNEL FUNCTIONAL INSTRUMENT CHECK(1) CALI BRA TION(1) TEST(1)

1. Reactor Coolant Outlet Temperature - N.A.

T HOT (Wide Range)

2. Reactor Coolant Inlet Temperature - N.A.

T COLD (Wide Range)

3. Reactor Coolant Pressure (Wide Range) N.A.

4. Pressurizer Water Level N.A.

5. Steam Line Pressure N.A.

6. Steam Generator Water Level N.A.

(Narrow Range)

7. Steam Generator Water Level N.A.

(Wide Range)

8. Refueling Water Storage Tank Water Level N.A.

9. deleted

10. Auxiliary Feedwater Flow Rate S/U# N.A.

11. Reactor Coolant System Subcooling N.A* N.A.

Margin Monitor

Auxiliary Feedwater System is used on each startup and flow rate indication is verified at that time.

The instruments used to develop RCS subcooling margin are calibrated in accordance with the Surveillance Frequency Control Program; the monitor will be compared with calculated subcooling margin for known input values in accordance with the Surveillance Frequency Control Program.

SALEM - UNIT 1 3/43-57 Amendment No. 299

TABLE 4.3-11 (Continued)

SURVEILLANCE REQUIREMENTS FOR ACCIDENT MONITORING INSTRUMENTATION CHANNEL CHANNEL CHANNEL FUNCTIONAL INSTRUMENT CHECK(1) CALIBRATION(1) TEST(1)

12. PORV Position Indicator N.A.

13. PORV Block Valve Position Indicator N.A. *

14. Pressurizer Safety Valve Position N.A.

Indicator

15. Containment Pressure - Narrow Range N.A.

16. Containment Pressure - Wide Range N.A.

17. Containment Water Level - Wide Range N.A.

18. Core Exit Thermocouples I'J.A.

19. Reactor Vessel Level Instrumentation N.A.

System (RVLlS)

20. Containment High Range Accident Radiation Monitor

21. Main Steamline Discharge (Safety Valves and Atmospheric Steam Dumps) Monitor Table Notation (1 ) Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table.

Unless the block valve is closed in order to meet the requirements of Action b, or c in specification 3.4.3.

SALEM - UNIT 1 3/43-57a Amendment No. 299

INSTRUMENTATION RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION 3.3.3.8 The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.3-12 shall be operable to ensure that the limits of ODCM Control 3.11.1.1 are not exceeded.

APPLICABILITY: At all times.

ACTION:

a. Not Used

b. With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3-12.

Exert best efforts to return the instrument to OPERABLE status within 30 days and, if unsuccessful, explain in the next annual radioactive effluent release report why the inoperability was not corrected in a timely manner.

c. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.8 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations at the frequencies specified in the Surveillance Frequency Control Program unless otherwise noted in Table 4.3-12.

SALEM - UNIT 1 3/43-58 Amendment No. 299

C TABLE 3.3-12 C

RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS INSTRUMENT OPERABLE ACTION

1. Not Used

2. Not Used

3. Not Used

4. TANK LEVEL INDICATING DEVICES

a. Temporary Outsi-'e Storage Tanks as Required 1 30 TABLE NOTATION ACTION 26 - Not Used ACTION 27 - Not Used ACTION 28 - Not Used ACTION 29 - Not Used ACTION 30 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, liquid additions to this tank may continue for up to 30 days provided the tank liquid level is estimated during all liquid additions to the tank.

SALEM - UNIT 1 3/4 3 -59 Amendment No. 234

TABLE 4.3-12 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL SOURCE CHANNEL FUNCTIONAL INSTRUMENT CHECK(1) CHECK(1) CALIBRATION(1) TEST(1)

1. Not Used

2. Not Used

3. Not Used

4. TANK LEVEL INDICATING DEVICES**

a. Temporary Outside Storage D* N.A.

Tanks as Required TABLE NOTATION

During liquid additions to the tank.

- If tank level indication is not provided, verification will be done by visual inspection.

(1 ) Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table.

SALEM - UNIT 1 3/43-60 Amendment No. 299

Pages 3/4 3-62 Through 3/4 3-69 Deleted SALEM - UNIT I 3/4 3-61 Amendment No. 282

INSTRUMENTATION POWER DISTRIBUTION MONITORING SYSTEM LIMITING CONDITION FOR OPERATION 3.3.3.14 The Power Distribution Monitoring System (PDMS) shall be OPERABLE with:

a. A minimum of the following inputs from the plant available for use by the PDMS as defined in Table 3.3-14;

1. Control Bank Position

2. Tcold

3. Reactor Power Level

4. NIS Power Range Detector Section Signals

b. Core Exit Thermocouples (T/C) meeting the criteria:

1. At least 25% operable T/C with at least 2 T/C per quadrant, and

2. The T/C pattern has coverage of all interior fuel assemblies (no face along the baffle), within a chess knight's move, radially, from a responding, calibrated T/C, or

3. At least 25%, operable T/C with at least 2 T/C per quadrant, and the installed PDMS calibration was determined within the last 31 Effective Full Power Days (EFPD).

4. The T/C temperatures used by the PDMS are calibrated via cross calibration with the loop temperature measurement RTDs, and using the T/C flow mixing factors determined during installed PDMS calibration.

c. An installed PDMS calibration satisfying the criteria:

1. The initial calibration in each operating cycle is determined using measurements from at least 75% of the incore movable detector thimbles obtained at a THERMAL POWER greater than 25% of RATED THERMAL POWER.

2. The calibration is determined using measurements from at least 50% of the incore movable detector thimbles at any time except as specified in 3.3.3.14.c.1, and

3. The calibration is determined using a minimum of 2 detector thimbles per core quadrant.

SALEM - UNIT 1 3/4 3-70 Amendment No. 237 NOV 06 2MUD

INSTRUMENTATION POWER DISTRIBUTION MONITORING SYSTEM LIMITING CONDITION FOR OPERATION (Continued)

APPLICABIL TY: MODE 1, above 25% RATED THERMAL POWER (RTP)

ACTION:

With any of the operability criteria listed in 3.3.3.14.a, 3.3.3.14.b, or 3.3.3.14.c not met, either correct the deficient operability condition, or declare the PDMS inoperable and use the incore movable detector system to obtain any required core power distribution measurements.

Increase the measured core peaking factors using the values listed in the COLR for the PDMS inoperable condition.

The provisions of Specification 3.0.3 are not applicable.

4.3.3.14.1 The operability criteria listed in 3.3.3.14.a, 3.3.3.14.b, and 3.3.3.14.c shall be verified to be satisfied prior to acceptance of the PDMS core power distribution measurement results.

4.3.3.14.2 Calibration of the PDMS is required:

a. In accordance with the Surveillance Frequency Control Program when the minimum number and core coverage criteria as defined in 3.3.3.14.b.1 and 3.3.3.14.b.2 are satisfied, or

b. In accordance with the Surveillance Frequency Control Program when only the minimum number criterion as defined in 3.3.3.14.b.3 is satisfied.

SALEM - UNIT 1 3/43-71 Amendment No. 299

INSTRUMENTATION TABLE 3.3-14 REQUIRED PDMS PLANT INPUT INFORMATION PLANT INPUT INFORMATION AVAILABLE MINIMUM NO. OF APPLICABLE INPUTS VALID INPUTS MODES Control Bank Position 4 4a 1c Tcold 4 2 ic Reactor Power Level 3 lb Ic NIS Power Range Excore 8 6d c Detector Section Signals TABLE NOTATIONS

a. Determined from either valid Demand Position or the average of the valid individual RCCA position indications for all RCCAs in the Control Bank.

b. Determined from either the reactor THERMAL POWER derived using a valid secondary calorimetric measurement, the average NIS Power Range Detector Power, or the average RCS Loop AT.

c. Greater than 25% RTP.

d. Comprised of an upper and lower detector section signal per Power Range Channel; a minimum of 3 OPERABLE channels required.

SALEM - UNIT 1 3 /4 3-72 Amendment No. 237 NiOv oC -UMID

3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS NORMAL OPERATION 3.4.1.1 All reactor coolant loops shall be in operation.

APPLICABILITY: MODES 1 and 2*

ACTION:

With less than the above required reactor coolant loops in operation, be in at least HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

4.4.1.1 The above required reactor coolant loops shall be verified to be in operation and circulating reactor coolant in accordance with the Surveillance Frequency Control Program.

See Special Test Exception 3.10.4.

SALEM - UNIT 1 3/44-1 Amendment No. 299

REACTOR COOLANT SYSTEM HOT STANDBY LIMITING CONDITION FOR OPERATION 3.4.1.2 a. At least two of the reactor coolant loops listed below shall be OPERABLE:

1. Reactor Coolant Loop 11 and its associated steam generator and reactor coolant pump,

2. Reactor Coolant Loop 12 and its associated steam generator and reactor coolant pump,

3. Reactor Coolant Loop 13 and its associated steam generator and reactor coolant pump,

4. Reactor Coolant Loop 14 and its associated steam generator and reactor coolant pump.

b. At least one of the above coolant loops shall be in operation* when the rod control system is deenergized**.

c. All of the above coolant loops shall be in operation when the rod control system is energized**.

APPLICABILITY: MODE 3 ACTION:

a. With less than the above required reactor coolant loops OPERABLE, restore the required loops 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 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. With no reactor coolant loolp in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required coolant loop to operation.

. SALEM - UNIT 1 3/4 4-2 , Amendment No. 69

REACTOR COOLANT SYSTEM HOT STANDBY SURVEILLANCE REQUIREMENTS 4.4.1.2.1 At least the above required reactor coolant pumps, if not in operation, shall be determined to be OPERABLE in accordance with the Surveillance Frequency Control Program by verifying correct breaker alignments and indicated power availability.

4.4.1.2.2 At least one cooling loop shall be verified to be in operation and circulating reactor coolant in accordance with the Surveillance Frequency Control Program.

4.4.1.2.3 The required steam generator{s) shall be determined OPERABLE by verifying secondary side water level to be greater than or equal to 5% (narrow range) in accordance with the Surveillance Frequency Control Program.

All reactor coolant pumps may be de-energized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided: (1) no operations are permitted that would cause dilution of the reactor coolant system boron concentration (2) core outlet temperature is maintained at least 10°F below saturation temperature, and (3) the rod control system is de-energized**

- The rod control system shall be considered de-energized when one or more of the following conditions exist:

1) Both Rod Drive MG set motor breakers are open.

2) Both Rod Drive MG set generator breakers are open.

3) A combination of at least three of the Reactor Trip and/or Reactor Trip Bypass Breakers are open.

If none of the above conditions for de-energizing the rod control system are met; the system shall be considered energized.

SALEM - UNIT 1 3/44-2a Amendment No. 299

REACTOR COOLANT SYSTEM SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.1.3 a. At least two of the coolant loops listed below shall be OPERABLE:

1. Reactor Coolant Loop (11) and its associated steam generator and reactor coolant pump,*-

2. Reactor Coolant Loop (1,2) and its associated steam generator and reactor coolant pump,*

3. Reactor Coolant Loop (13) and its associated steam generator and reactor coolant pump,*

4. Reactor Coolant' Loop (14) and its associated steam generator and reactor'coolant pump,*

5. Residual Peat Removal Loop (11),

6. -Residual Heat Removal Loop (12).

b. At least one of the above coolant loops shall be in operation.**

APPLICABILITY: MODE 4 ACTION:

a. Vith less than the above required loops nPERABLE, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible; be in COLD SHUTDOWN within 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months .

b. With no coolant loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required coolant loop to operation.

A reactor coolant pump shall not be started with one or more of the RCS cold lea temperatures less than or equal to (312)0 F unless 1) the pressurizer water volume is less than 1650 cubic feet (93.2% of pres-surizer level indication) or 2) the secondary water temperature of each steam aenerator is less than 500 F above each of the RCS cold leg temperatures.

- All reactor coolant pumps and residual heat removal pumps may be de-energized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided 1) no operations are permitted that would cause dilution of the reactor coolant system boron concentration, and 2) core outlet temperature is maintained at least 100 F below satura-tion temperature.

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

REACTOR COOLANT SYSTEM 4.4.1.3.1 The required residual heat removalloop(s) shall be determined OPERABLE per the inservice testing schedule.

4.4.1.3.2 The required reactor coolant pump(s), if not in operation, shall be determined to be OPERABLE in accordance with the Surveillance Frequency Control Program by verifying correct breaker alignments and indicated power availability.

4.4.1.3.3 The required steam generator(s) shall be determined OPERABLE by verifying secondary side level to be greater than or equal to 5% of narrow range in accordance with the Surveillance Frequency Control Program.

4.4.1.3.4 At least one coolant loop shall be verified to be in operation and circulating reactor coolant in accordance with the Surveillance Frequency Control Program.

SALEM - UN IT 1 3/44-3a Amendment No. 299

REACTOR COOLANT SYSTEM COLD SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.1.4 Two# residual heat removal loops shall be OPERABLE* and at least one RHR loop shall be in operation.**

APPLICABILITY: MODE 5.##

ACTION:

a. With less than the above required loops operable, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible.

b. With no RHR loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to operation.

SURVEILLANCE REQUIREMENTS 4.4.1.4 At least one residual heat removal loop shall be verified to be in operation and circulating reactor coolant in accordance with the Surveillance Frequency Control Program.

One RHR loop may be inoperable for up to two hours for surveillance testing, provided the other RHR loop is OPERABLE and in operation. Additionally, four filled reactor coolant loops, with at least two steam generators with their secondary side water levels greater than or equal to 5% (narrow range), may be substituted for one residual heat removal loop.

1. A reactor coolant pump shall not be started with one or more of the RCS cold leg temperatures less than or equal to 312°F unless 1) the pressurizer water volume is less than 1650 cubic feet (equivalent to approximately 93.2% of level), or 2) the secondary water temperature of each steam generator is less than 50°F above each of the RCS cold leg temperatures.

Systems supporting RHR loop operability may be excepted as follows:

a. The normal or emergency power source may be inoperable.

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

SALEM - UNIT 1 3/44-3b Amendment No. 299

THIS PAGE LEFT INTENTIONALLY BLANK SALEM - UNIT 1 3/4 4-3c Amendment No. 214 SEP 0 8 1998

REACTOR COOLANT SYSTEM 3/4.4.2 SAFETY VALVES SAFETY VALVES - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.2.1 A minimum of one pressurizer code safety valve shall be OPERABLE* with a lift setting of 2485 psig +/- 3%.**'***

APPLICABILITY: MODE 4 and 5 ACTION:

With no pressurizer code safety valve OPERABLE, immediately suspend all operations involving positive reactivity changes and place an OPERABLE RHR loop into operation in the shutdown cooling mode.

SURVEILLANCE REQUIREMENTS 4.4.2.1 No additional Surveillance Requirements other than those required by the Inservice Testing Program.

While in Mode 5, an equivalent size vent pathway may be used provided that the vent pathway is not isolated or sealed.

- The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

- - Following testing the lift setting shall be reset to within +/- 1%.

SALEM - UNIT 1 3/4 4-4 Amendment No. 297

3/4.4 REACTOR COOLANT SYSTEM 3/4.4.2 SAFETY VALVES SAFETY VALVES - OPERATING LIMITING CONDITION FOR OPERATION 3.4.2.2 All pressurizer code safety valve shall be OPERABLE with a lift setting of 2485 psig +/- 3%.*,**

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

With one pressurizer code safety valve inoperable, either restore the inoperable valve to OPERABLE status within 15 minutes, or be in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE REQUIREMENTS 4.4.2.2 No additional Surveillance Requirements other than those required by the Inservice Testing Program.

The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

- Following testing the lift setting shall be reset to within +/- 1%.

SALEM - UNIT 1 3/44-4a Amendment No. 297

REACTOR COOLANT SYSTEM 3/4.4.3 RELIEF VALVES SURVEILLANCE REQUIREMENTS 4.4.3.1 In addition to the requirements of the Inservice Testing Program, each PORV shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program by:

a. Operating the PORV through one complete cycle of full travel during MODES 3 or 4, and

b. Operating solenoid valves, air control valves, and check valves on associated air accumulators in PORV control systems through one complete cycle of full travel, and

c. Performing a CHANNEL CALIBRATION of the actuation instrumentation.

4.4.3.2 Each block valve shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program by operating the valve through one complete cycle of full travel unless the block valve is closed in order to meet the requirements of ACTION b, or c in Specification 3.4.3.

SALEM - UNIT 1 3/44-5a Amendment No. 299

REACTOR COOLANT SYSTEM 3/4.4.4 PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.4 The pressurizer shall be OPERABLE with a water volume of less than or equal to 1650 cubic feet (92% indicated level), and at least two groups of pressurizer heaters each having a capacity of ~ 150 kw and capable of being powered from an emergency power supply.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

a. With the pressurizer inoperable due to an inoperable emergency power supply to the pressurizer heaters either restore the inoperable emergency power supply within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. With the pressurizer otherwise inoperable, be in at least HOT STANDBY with the reactor trip breakers open within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

4.4.4.1 The pressurizer water volume shall be determined to be within its limit in accordance with the Surveillance Frequency Control Program.

4.4.4.2 The capacity of each of the above required groups of pressurizer heaters shall be verified by measuring circuit current in accordance with the Surveillance Frequency Control Program.

4.4.4.3 The emergency power supply for the pressurizer heaters shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program by manually transferring power from the normal to the emergency power supply and energizing the heaters.

SALEM - UNIT 1 3/4 4-6 Amendment No. 299

REACTOR COOLANT SYSTEM STEAM GENERATOR (SG) TUBE INTEGRITY I LIMITING CONDITION FOR OPERATION 3.4.5 SG tube integrity shall be maintained and all SG tubes satisfying the tube repair criteria shall be plugged in accordance with the Steam Generator Program.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

a.* With one or more SG tubes satisfying the tube repair criteria and not plugged in accordance with the Steam Generator Program:

1. Verify tube integrity of the affected tube(s) is maintained until the next refueling outage or SG tube inspection within 7 days; and

2. Plug the affected tube(s) in accordance with the Steam Generator Program prior to entering HOT SHUTDOWN following the next refueling outage or SG tube inspection.

b. With SG tube integrity not maintained or the required Action of a.

above not met, 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 at least COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.4.5.1 Verify SG tube integrity in accordance with the Steam Generator Program.

4.4.5.2 Verify that each inspected SG tube that satisfies the tube repair criteria is plugged in accordance with the Steam Generator Program prior to entering HOT SHUTDOWN following a SG tube inspection.

Separate Action is allowed for each SG tube.

SALEM - UNIT 1 3/4 4-7 Amendment No. 268

PAGES 3/4 4-8 THROUGH 3/4 4-13a DELETED SALEM - UNIT I 3/4 4-8 Through 3/4 4-13a Amendment No.- 268

REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.6.1 The following Reactor Coolant System leakage detection systems shall be OPERABLE:

a. The containment atmosphere particulate radioactivity monitoring system,

b. The containment sump level monitoring system, and

c. Either the containment fan cooler condensate flow rate or the containment atmosphere gaseous radioactivity monitoring system.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With only two of the above required leakage detection systems OPERABLE, operation may continue for up to 30 days provided grab samples of the containment atmosphere are obtained and analyzed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the required gaseous and/or particulate radioactivity monitoring system is inoperable; otherwise, 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 .

4.4.6.1 The leakage detection systems shall be demonstrated OPERABLE by:

a. Containment atmosphere particulate and gaseous (if being used) monitoring systems-performance of CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST in accordance with the Surveillance Frequency Control Program.

b. Containment sump level and containment fan cooler condensate flow rate (if being used) monitoring systems-performance of CHANNEL CALIBRATION in accordance with the Surveillance Frequency Control Program.

SALEM - UNIT 1 3/44-14 Amendment No. 299

REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR QPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE,

b. 1 GPM UNIDENTIFIED LEAKAGE,

c. 150 gallons per day primary-to-secondary leakage through anyone steam generator, and

d. 10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System.

APPLICABILITY: MODES 1, 2, 3 and 4 ACTION:

a. With any PRESSURE BOUNDARY LEAKAGE or primary-to-secondary leakage not within limit, 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 any Reactor Coolant System leakage greater than anyone of the above limits, excluding PRESSURE BOUNDARY LEAKAGE and primary-to-secondary leakage, reduce the leakage rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREfYI,~EgN~T~S==================

4.4.6.2 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by;

a. Monitoring the containment atmosphere particulate radioactivity monitor in accordance with the Surveillance Frequency Control Program.

b. Monitoring the containment sump inventory in accordance with the Surveillance Frequency Control Program.

SALEM - UNIT 1 3/44-15 Amendment No. 299

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

c.

Verifying primary-to-secondary leakage is :5 150 gallons per day through anyone steam generator in accordance with the Surveillance Frequency Control Program during steady state operation.

d.

Performance of a Reactor Coolant System water inventory balance** in accordance with the Surveillance Frequency Control Program. The water inventory balance shall be performed with the plant at steady state conditions.

The provisions of speCification 4.0.4 are not applicable for entry into Mode 4, and

e. Monitoring the reactor head flange leakoff system in accordance with the Surveillance Frequency Control Program.

Not required to be completed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

- Not applicable to primary-to-secondary leakage.

SALEM - UNIT 1 3/44-16 Amendment No. 299

REACTOR COOLANT SYSTEM PRIMARY COOLANT SYSTEM PRESSURE ISOLATION VALVES LIMITING QONDITION FQR OPERATION 3.4.6.3 Reactor Coolant System Pressure Isolation Valves specified in table 4.4-3 shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

a. With any Reactor Coolant System Pressure Isolation Valve leakage greater than the specified limit in Table 4.4-3, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least two closed manual or deactivated automatic valves, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.4.6.3 Each Reactor Coolant System Pressure Isolation Valve specified in Table 4.4-3 shall be demonstrated OPERABLE pursuant to the Inservice Testing Program, except that in lieu of any leakage testing required by the Inservice Testing Program, each valve shall be demonstrated OPERABLE by verifying leakage to be within its limit:

a. In accordance with the Surveillance Frequency Control Program.

b. Prior to entering MODE 2 whenever the plant has been in COLD SHUTDOWN for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or more and if leakage testing has not been performed in the previous 9 months.

c. Prior to returning the valve to service following maintenance repair or replacement work on the valve.

d. For the Residual Heat Removal and Safety Injection Systems hot and cold leg injection valves and accumulator valves listed in Table 4.4-3 the testing will be done within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve actuation due to automatic or manual action or flow through the valve. For all other systems testing will be done once per refueling.

The provisions of specification 4.0.4 are not applicable for entry into MODE 3 or 4.

SALEM - UNIT 1 3/44-16a Amendment No. 299

TABLE 4.4-3 J REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES Maxi(muml) (b)

System Valve No. Allowable Leakage Low Pressure Safety Injection 5.0 GPM each valve Loop 11, cold leg 1SJ56 5.0 GPM each valve 11SJ43 5.0 GPM each valve Loop 12, cold leg 12SJ56 5.0 GPM each valve 12SJ43 5.0 GPM each valve Loop 13, cold leg 13SJ56 5.0 GPM each valve 13SJ43 5.0 GPM each valve Loop 13, hot leg 13SJ156 5.0 GPM each valve 13RH27 5.0 GPM each valve Loop 14, cold leg 14SJ56 5.0 GPM each valve 14SJ43 5.0 GPM each valve Loop 14, hot leg 14SJ156 5.0 GPM each valve 14RH27 5.0 GPM each valve Intermediate Pressure Safety Injection 5.0 GPM each valve Loop 11, cold leg 1SJ144 5.0 GPM each valve Loop 11, hot leg llSJ156 5.0 GPM each valve IISJ139 5.0 GPM each valve Loop 12, cold leg 12S3144 5.0 GPM each valve Loop 12, hot leg 12SJ156 5.0 GPM each valve 12SJ139 5.0 GPM each valve Loop 13, cold leg 13SJ144 5.0 GPM each valve S5 Loop 13, hot leg 13SJ156 5.0 GPM each valve 13SJ139 5.0 GPM each valve Loop 14, cold leg 14SJ144 5.0 GPM each valve Loop 14, hot leg 14SJ156 5.0 GPM each valve 14SJ139 5.0 GPM each valve Safety Injection Accumulators 3 to cold leg loop 11, cold leg 11SJ55 5.0 GPM each valve loop 12, cold leg 12SJ55 5.0 GPM each valve loop 13, cold leg 13SJ55 5.0 GPM each valve loop 14, cold leg 14SJ55 5.0 GPM each valve Safety Injection Boron Injection to cold legs loop 11, cold leg IISJ17 5.0 GPM each valve loop 12, cold leg 12SJ17 5.0 GPM each valve loop 13, cold leg 13SJ17 5.0 GPM each valve loop 14, cold leg 14SJ17 5.0 GPM each valve lSJ150 5.0 GPM each valve

' RHR Suction loop 11 lRHl S 5.0 GPM each valve loop 11 1RH2 S 5.0 GPM each valve SALEM - UNIT 1 3/4 4-16b Amendment No. 225 HOV I ,'I9

TABLE 4.4-3 (CONT'D) I (a) 1. Leakage rates less than or equal to 1.0 gpm are considered acceptable. However, for initial tests, or tests following valve repair or replacement, leakage rates less than or equal to 5.0 gpm are considered acceptable.

2. Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm are considered acceptable if the latest measured rate has not exceeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of 5.0 gpm by 50% or greater.

3. Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm are considered unacceptable if the latest measured rate exceeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of 5.0 gpm by 50% or greater.

4. Leakage rates greater than 5.0 gpm are considered unacceptable.

(b) Minimum differential test pressure shall not be less than 150 psid.

SALEM - UNIT 1 3/4 4-16c Amendment No. 225

':v0c 2, *29

PAGES 3/4 4-17 THROUGH 3/4 4-19 DELETED SALEM - UNIT 1 3/4 4-17 Through 3/4 4-19 Amendment No.180 (8 2 2 13936

REACTOR COOLANT SYSTEM SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.8 The spedific activity of the primary coolant shall be limited to:

a.

1.0 pCi/gram DOSE EQUIVALENT 1-131, and

b.

100/EpCi/gram.

APPLICABILITY: MODES 1, 2, 3, 4 and 5 ACTION:

MODES 1, 2 and 3*

a. With the specific activity of the primary coolant > 1.0 pCi/gram DOSE EQUIVALENT 1-131 for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in at least HOT STANDBY with Tavg < 500'F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. With the specific activity of the primary coolant > 100/94iCi/gram, be in at least HOT STANDBY with Tavg < 500'F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c. LCO 3.0.4.c is applicable.

MODES 1, 2, 3, 4 and 5

a. With the specific activity of the primary coolant > 1.0 pCi/gram DOSE EQUIVALENT 1-131 or > 100/EpCi/gram, perform the sampling and analysis reauirements of item 4a of Table 4.4-4 until the specific activity of the primary coolant is restored to within its limits.

SURVEILLANCE REQUIREMENTS 4.4.8 The specific activity of the primary coolant shall be determined to be within the limits by performance of the sampling and analysis program of Table 4.4-4.

With Tavg - 500°F.

SALEM - UNIT 1 -3/4 4-20 Amendment No. 2 7 6

\

i This page intentionally left blank.

I SALEM - UNIT 1 3/4 4-21 Amendment No. 133 l JUN- 8 1992

TABLE 4.4-4 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM MODES IN WHICH SAMPLE AND TYPE OF MEASUREMENT SAMPLE AND ANALYSIS ANALYSIS AND ANALYSIS FREQUENCY REQUIRED

1. Gross Activity Determination In accordance with the 1,2,3,4 Surveillance Frequency Control Program

2. Isotopic Analysis for DOSE In accordance with the 1 EQUIVALENT 1-131 Concentration Surveillance Frequency Control Program

3. Radiochemical for E Determination In accordance with the 1 Surveillance Frequency Control Program*

4. Isotopic Analysis for Iodine a) Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , Including 1-131, 1-133, and 1-135 whenever the specific activity exceeds 1.0 IJCi/gram DOSE EQUIVALENT 1-131 or 1ODIE IJCi/gram, and b) One sample between 1,2,3 2 & 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months following a THERMAL POWER change exceeding 15 percent of the RATED THERMAL POWER within a one hour period.

Until the specific activity of the primary coolant system is restored within its limits.

Sample to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since reactor was last subcritical for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or longer.

SALEM - UNIT 1 3/44-22 Amendment No. 299

250 E

I-5

!oo -NACCE 'TABLE OPERATION C,_ IIJ U.

I oft I cr:

0-0 w

cn.

z iI ACCEPT*AB LE - _t; 11 L \ i@§l g.

-J 50ti OPERATION-5 II I I I I I I I I I I I I I I - - - I It I I I I I I I 11 I I I I

U I

--r I

I

'II i

_lI z --H- !1;1i 20 30 40 50 60 70 80 90 100 PERCENT OF RATED THERMAL POWER FIGURE 3.4-1 DOSE EQUIVALENT 1-131 Primary Coolant Specific Activity Limit Versus Percent of RATED THERMAL POWER with the Primary Coolant Specific Activity> 1.0 jCi/gram Dose Equivalent 1-131 SALEM - UNIT 1 3/4 4-23

I Ii REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE/TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and. pressure shall be limited in accordance with the limit lines shown on Figures 3.4-2 and 3.4-3 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:

a. A maximum heatup of 100 0 F in any one hour period,

b. A maximum cooldown of 1000 F in any one hour period, and

c. A maximum temperature change of < 50 F in any one hour period, during hydrostatic testing operations above system design pressure.

APPLICABILITY: At all times. it ACTION:

With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the fracture toughness properties of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation 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 reduce the RCS T. and pressure to less than 2007F and 500 psig, respectively, within t Ve~following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SALEM - UNIT 1 3/4 4-24

REACTOR COOLANT SYSTEM SURVEILLANCE Rl;QUIREMENTS 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits in accordance with the Surveillance Frequency Control Program during system heatup, cooldown, and inservice leak and hydrostatic testing operations.

4.4.9.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, at the intervals required by 10 CFR 50, Appendix H. The results of these examinations shall be used to update Figures 3.4-2 and 3.4-3.

SALEM - UNIT 1 3/44-25 Amendment No. 299

Limiting Material Property

@1/4T @3/4T Weld 3-042C Plate B2402-1 Initial RTNDT -56°F 45°F RTNDT after 32 EFPY 232°F 171°F 2500 -

2250 .- t /

2000-1750-Unaccaptabe Apctable Operation Operation

~1500 - - - - - -- -

CrFcIUmit

= 100 rHetuip tup7L ltDog. F<d t) IODD:e

.R Rate /

7 1100 Deg. r _

500 - - _ - C_rWcality Limitbased on

_Irwvlca hydrostatic test temperature (312 F)for the period up to 32 EFPY 5_ser.i 250 Plus 2 D 0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Dog. F)

Figure 3.4-2 Salem Unit 1 Reactor Coolant System Heatup Limitations Applicable for Heatup Rates up to 100°F/HR for the Service Period up to 32 EFPY (with uncertainties for instrumentation errors).

SALEM - UNIT 1 3/4 4-26 Amendment No. 243 IMAY 2 5 2Q5

Limiting Material Property

@1/4T @3/4T Weld 3-042C Plate B2402-1 Initial RTNDT -56OF 450 F RTNDT after 32 EFPY 2320 F 1710 F 2500 2250 2000 1750 0

0 50 1

IL 15D0 1250 2

0.

75 500 250 0

0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Dog. F)

Figure 3.4-3 Salem Unit 1 Reactor Coolant System Cooldown Limitations Applicable for Cooldown Rates up to 100 0F/HR for the Service Period up to 32 EFPY (with uncertainties for instrumentation errors)

I SALEM - UNIT 1 3/4 4-27 Amendment No. 243 MAY 2 5 2,301

I This Page Intentionally Left Blank I

I SALEM - UNIT 1 3/4 4-28 Amendment No. 25 APR 2 2 1980

REACTOR COOLANT SYSTEM PRESSURIZER 3.4.9.2 The pressurizer temperature shall be limited to:

a. A maximum heatup of 100°F in anyone hour period,

b. A maximum cooldown of 200°F in anyone hour period, and

c. A maximum spray water temperature differential of 320°F.

APPLICABILITY: At all times.

ACTION:

With the pressurizer temperature limits in excess of any of the above limits, restore the temperature to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the fracture toughness properties of the pressurizer; determine that the pressurizer remains acceptable for continued operation 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 reduce the pressurizer pressure to less than 500 psig within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.4.9.2 The pressurizer temperatures shall be determined to be within the limits in accordance with the Surveillance Frequency Control Program during system heatup or cooldown. The spray water temperature differential shall be determined to be within the limit in accordance with the Surveillance Frequency Control Program during auxiliary spray operation.

SALEM - UNIT 1 3/44-29 Amendment No. 299

REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.9.3 At least one of the following overpressure protection systems shall be OPERABLE:

a. Two Pressurizer Overpressure Protection System relief valves (POPS) with a lift setting of less than or equal to 375 psig, or

b. A reactor coolant system vent of greater than or equal to 3.14 square inches.

APPLICABILITY: When the temperature of one or more of the RCS cold.legs is less than or equal to 312 0 F, except when the reactor vessel head is removed.

ACTION:

a. With one POPS inoperable in MODE 4 and the temperature of one or more of the RCS cold legs is less than or equal to 312 0 F, either restore the inoperable POPS to OPERABLE status within 7 days or depressurize and vent the RCS through a 3.14 square inch vent(s) within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ; maintain the RCS in a vented condition until both POPSs have been restored to OPERABLE status.

b. With one POPS inoperable in MODES 5 or 6 with the Reactor Vessel Head installed, restore the inoperable POPS to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or complete depressurization and venting of the RCS through at least a 3.14 square. inch vent(s) within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ; maintain the RCS in a vented condition until both POPSs have been restored to OPERABLE status.

c. With both POPSs inoperable, depressurize and vent the RCS through a 3.14 square inch vent(s) within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ; maintain the RCS in a vented condition until both POPSs have been restored to OPERABLE status.

d. In the event either the POPS or the RCS vent(s) are used to.

mitigate a RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days. The report shall describe the circumstances initiating the transient, the effect of the POPS or vent(s) on the transient and any corrective action necessary to prevent recurrence.

e. LCO 3.0.4.b is not applicable when entering MODE 4.

SALEM - UNIT 1 3/4 4-30 Amendment No. 276

REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS SURVEILLA~CE REQUI REMENTS 4.4.9.3.1 Each POPS shall be demonstrated OPERABLE by:

a. Performance of a CHANNEL FUNCTIONAL TEST on the POPS actuation channel, but excluding valve operation, within 31 days prior to entering a condition in which the POPS is required OPERABLE, and in accordance with the Surveillance Frequency Control Program thereafter when the POPS is required OPERABLE.

b. Performance of a CHANNEL CALIBRATION on the POPS actuation channel in accordance with the Surveillance Frequency Control Program.

c. Verifying the POPS isolation valve is open in accordance with the Surveillance Frequency Control Program when the POPS is being used for overpressure protection.

d. Testing pursuant to the Inservice Testing Program.

4.4.9.3.2 The RCS vent(s) shall be verified to be open in accordance with the Surveillance Frequency Control Program* when the vents(s) is being used for overpressure protection.

Except when the vent pathway is provided with a valve which is locked, sealed, or otherwise secured in the open position, then verify these valves open in accordance with the Surveillance Frequency Control Program.

SALEM - UNIT 1 3/44-31 Amendment No. 299

REACTOR COOLANT SYSTEM 3.4.10 DELETED SALEM - UNIT 1 3/44-32 Amendment No. 298

THIS PAGE INTENTIONALLY BLANK SALEM - UNIT 1 3/44-33 Amendment No. 298

- . A . . " t - I .,

.. A ; I p

t4 SECTION 3/4.4.11 INTENTIONALLY BLANK ctE ii

.q be I

41 I!

-I 13/4 4-34 Anriogmgn~ i& 101 SALEK - UNIT 1

REACTOR COOLANT SYSTEM 3/4.4.12 HEAD VENTS 3.4.12 Four reactor vessel head vent paths shall be operable with the vent paths closed. A vent path consists of at least two head vent valves in series, powered from vital sources, and associated flowpath.

APPLICABILITY: MODES 1, 2, 3 AND 4.

ACTION:

a. With one, two or three reactor vessel head vent path(s) inoperable, STARTUP and/or POWER OPERATION may continue provided the inoperable vent path(s) is maintained closed with the valve actuators key locked in the closed position; restore the inoperable vent path(s) to OPERABLE status within 30 days, or, be in 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 four reactor vessel head vent paths inoperable; maintain the inoperable vent paths closed with power removed from the valve actuators of all the vent valves in the inoperable vent paths, and restore at least one of the vent paths to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in 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 .

SURVEILLANCE REQUIR~mE~M~E~N~T~S~==================

4.4.12 Reactor vessel head vent system vent paths shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program by:

1. Verifying all manual isolation valves in each vent path are locked in the open position.

2. Cycling each valve in the vent paths through at least one complete cycle of full travel from the control room during COLD SHUTDOWN or REFUELING.

3. Verifying flow through the reactor vessel head vent system vent path during venting.

SALEM - UNIT 1 3/44-35 Amendment No. 299

3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

ACCUMULATORS LIMITING CONDITION FOR OPERATION 3.5.1 Each reactor coolant system accumulator shall be OPERABLE with:

a. The isolation valve open,

b. A contained volume of between 6,223 and 6,500 gallons of borated water,

c. A boron concentration of between 2,200 and 2,500 ppm, and,

d. A nitrogen cover-pressure of between 595.5 and 647.5 psig.

APPLICABILITY: MODES 1,2 and 3*.

ACTION:

a. With one accumulator inoperable, except as a result of a closed isolation valve or boron concentration outside the required limits, restore the inoperable accumulator to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. With one accumulator inoperable due to the isolation valve being closed, either immediately open the isolation valve or be in HOT STANDBY within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

c. With the boron concentration of one accumulator outside the required limits, restore the boron concentration to within the required limits within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce pressurizer pressure to less than or equal to 1000 psig within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.5.1 Each accumulator shall be demonstrated OPERABLE:

a. In accordance with the Surveillance Frequency Control Program by:

1. Verifying the water level and nitrogen cover-pressure in the tanks, and

2. Verifying that each accumulator isolation valve is open.

Pressurizer Pressure above 1000 psig.

SALEM - UNIT 1 3/45-1 Amendment No. 299

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

b. In accordance with the Surveillance Frequency Control Program and within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after each solution volume increase of ~ 1% of tank volume by verifying the boron concentration of the accumulator solution.

c. In accordance with the Surveillance Frequency Control Program when the RCS pressure is greater than 1000 psig by verifying that the power lockout switch is in lockout.

d. In accordance with the Surveillance Frequency Control Program by verifying that each accumulator isolation valve opens automatically upon receipt of a safety injection test signal.

SALEM - UNIT 1 3/4 5-2 Amendment No. 299

EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS - Tayg k 350OF LIMITING CONDITION FOR OPERATION 3.5.2 Two independent ECCS subsystems shall be OPERABLE with each subsystem comprised of the following injection systems:

a. One OPERABLE centrifugal charging pump and associated flow path capable of taking suction from the refueling water storage tank and transferring suction to the residual heat removal pump discharge piping and;

1. Discharging into each Reactor Coolant System (RCS) cold leg.

b. One OPERABLE safety injection pump and associated flow path capable of taking suction from the refueling water storage tank and transferring suction to the residual heat removal pump discharge piping and;

1. Discharging into each RCS cold leg, and; upon manual initiation,

2. Discharging into its two associated RCS hot legs.

C. One OPERABLE residual heat removal pump and associated residual heat removal heat exchanger and flow path capable of taking suction from the refueling water storage tank on a safety injection signal and transferring suction to the containment sump during the recirculation phase of operation and;

1. Discharging into each RCS cold leg.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

a. With one ECCS subsystem inoperable, restore the inoperable subsystem 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 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .- -

b. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date.

c. With both ECCS subsystems inoperable for surveillance testing;

.restore at least one subsystem to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and at least COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SALEM - UNIT 1 3/4 5-3 Amendment No. 200 SEP 1 1 1997

EIVIERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE:

a. In accordance with the Surveillance Frequency Control Program by:

1. Verifying that the following valves are in the indicated positions with power to the valve operators removed:

Valve Number Valve Function Valve Position

a. 1 SJ 69 a. RHR pump suction a. open

b. 1 SJ 30 b. SI pump suction b. open

c. 11 SJ 40 c. SI discharge to hot legs c. closed

d. 12 SJ 40 d. SI discharge to hot legs d. closed

e. 1 RH 26 e. RHR discharge to hot legs e. closed

f. 11 SJ 49 f. RHR discharge to cold legs f. open

g. 12 SJ 49 g. RHR discharge to cold legs g. open

h. 1 CS 14# h. Spray additive tank discharge h. open

i. 1 SJ 135 i. SI discharge to cold legs i. open

j. 1 SJ 67 j. SI recirc. line isolation j. open

k. 1 SJ 68 k. SI recirc. line isolation k. open I. 11 SJ 44 I. Containment sump isolation valve I. closed

m. 12 SJ 44 m. Containment sump isolation valve m. closed

2. Verifying that the following valves are in the indicated positions:

Valve Number Valve Function Valve Position

a. 11 RH 19 a. RHR crosstie valve a. open

b. 12 RH 19 b. RHR crosstie valve b. open

b. In accordance with the Surveillance Frequency Control Program by:

1. Verifying that each valve (manual, power operated or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.

2. Verifying that the ECCS piping is full of water by venting the ECCS pump casings and accessible discharge piping high points.

If inoperable, the applicable Technical Specification is 3.6.2.2.

SALEM - UNIT 1 Amendment No. 299

EMERGENCY CORE COOLING SYSTEMS SURVEI LLANCEREQUI REMENTS (Continued)

c. By a visual inspection which verifies that no loose debris (rags, trash, clothing, etc.) is present in the containment which could be transported to the containment sump and cause restriction of the pump suctions during LOCA conditions. This visual inspection shall be performed:

1. For all accessible areas of the containment prior to establishing CONTAINMENT INTEGRITY, and

2. In accordance with the Surveillance Frequency Control Program the areas affected within containment by containment entry and during the final entry when CONTAINMENT INTEGRITY is established.

d. In accordance with the Surveillance Frequency Control Program by:

1. A visual inspection of the containment sump and verifying that the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens, etc.) show no evidence of structural distress or corrosion.

e. In accordance with the Surveillance Frequency Control Program, during shutdown, by:

1. Verifying that each automatic valve in the flow path actuates to its correct position on a safety injection test signal.

2. Verifying that each of the following pumps start automatically upon receipt of a safety injection test signal:

a) Centrifugal charging pump b) Safety injection pump c) Residual heat removal pump SALEM - UNIT 1 3/45-5 Amendment No. 299

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

f. By verifying that each of the following pumps develops the indicated Total Dynamic Head (TDH) when tested at the test flow point pursuant to the Inservice Testing Program:

1. Centrifugal charging pump ~ 2338 psi TDH

2. Safety Injection Pump ~ 1369 psi TDH

3. Residual heat removal pump ~ 165 psi TDH

g. By verifying the correct position of each of the following ECCS throttle valves:

1. Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months following completion of each valve stroking operation or maintenance on the valve when the ECCS subsystems are required to be OPERABLE.

2. In accordance with the Surveillance Frequency Control Program.

HPSISYSTEM LPSI SYSTEM VALVE NUMBER VALVE NUMBER 11 SJ 16 11 SJ 138 12 SJ 16 12 SJ 138 13 SJ 16 13 SJ 138 14 SJ 16 14 SJ 138 11 SJ 143 12 SJ 143 13 SJ 143 14 SJ 143

h. By performing a flow balance test, during shutdown, following completion of modifications to the ECCS subsystems that alter the subsystem flow characteristics and verifying that:

1. For Safety Injection pumps, with a single pump running:

a) The sum of the injection line flow rates, excluding the highest flow rate, is ~ 453 gpm; and b) The total flow rate through all four injection lines is s 647 gpm, and c) The difference between any pair of injection line flow rates is s 12.0 gpm, and d) The total pump flow rate is ::; 664 gpm in the cold leg alignment, and e) The total pump flow rate is s 654 gpm in the hot leg alignment.

SALEM - UNIT 1 3/45-5a Amendment No. 299

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

2. For Centrifugal Charging pumps, with a single pump running:

a) The sum of the injection line flow rates, excluding the highest flow rate, is 2 306 gpm, and b) The total flow rate through all four injection lines is S 444 gpm, and c) The difference between any pair of injection line flow rates is S 10.5 gpm, and d) The total pump flow rate is 5 554 gpm.

I

i. The automatic interlock function of the PER System shall be verified within the seven (7) days prior to placing the MER System in service for cooling of the Reactor Coolant System. This shall be done by verifying with a test signal corresponding to a reactor coolant pressure of 375 psig or greater, that the iREl and 1UE2 valves cannot be opened.

SALEM - UNIT 1 3/4 5-Sb Amendment No.:W, 208 IAR 1 2 1998

EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS - Tavg <350'F LIMITING CONDITION FOR OPERATION 3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:

a. One OPERABLE centrifugal charging pumpý and associated flow path capable of taking suction from the refueling water storage tank and transferring suction to the residual heat removal pump discharge piping and;

1. Discharging into each Reactor Coolant System (RCS) cold leg.

b. One OPERABLE residual heat removal pump and associated residual heat removal heat exchanger and flow path capable of taking suction from the refueling water storage tank on a safety injection signal and transferring suction to the containment sump during the recirculation phase of operation and;

1. Discharging into each RCS cold leg, and; upon manual initiation,

2. Discharging into two RCS hot legs.

APPLICABILITY: MODE 4.

ACTION:

a. With no ECCS subsystem OPERABLE because of the inoperability of either the centrifugal charging pump or the flow path from the refueling water storage tank, restore at least one .ECCS subsystem to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in COLD SHUTDOWN within the next 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months .

b. With no ECCS subsystem OPERABLE because of the inoperability of either the residual heat removal heat exchanger or residual heat removal pump, restore at least one ECCS subsystem to OPERABLE status or maintain the Reactor Coolant System Tavý less than 350'F by use of alternate heat removal methods.

c. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation.

cycles to date.

d. LCO 3.0.4.b is not applicable to ECCS high head subsystem

A maximum of one safety injection pump or one centrifugal charging pump shall be OPERABLE in MODE 4 when the temperature of one or more of the RCS cold legs is less than or equal to 312'F, Mode 5, or Mode 6 when the head is on the reactor vessel.

SALEM - UNIT ! 3/1 5-6 Amendment No. 276

EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS - Tavg < 350°F SURVEILLANCE REQUIREMENTS 4.5.3.1 The ECCS subsystem shall be demonstrated OPERABLE per applicable Surveillance Requirements of 4.5.2.

4.5.3.2 All safety injection pumps and centrifugal charging pumps, except the above required OPERABLE pump, shall be demonstrated to be inoperable in accordance with the Surveillance Frequency Control Program while in MODE 4 and the temperature of one or more of the RCS cold legs is less than or equal to 312°F, MODE 5, or MODE 6 when the head is on the reactor vessel by either of the following methods:

a. By verifying that the motor circuit breakers have been removed from their electrical power supply circuits or,

b. For verifying that the pump is in a recirculation ftow path and that two independent means of preventing RCS injection are utilized.

SALEM - UNIT 1 3/45-6a Amendment No. 299

EMERGENCY CORE COOLING SYSTEMS SEAL INJECTION FLOW LIMITING CONDITION FOR OPERATION 3.5.4 Reactor coolant pump seal injection flow shall be :s; 40 gpm with centrifugal charging pump discharge header pressure 2!: 2430 psig and the charging flow control valve full open.

APPLICABILITY: MODES 1. 2. and 3 ACTION:

With seal injection flow not within the limit. adjust manual seal injection throttle valves to give a flow within the limit with the charging pump discharge pressure 2!: 2430 psig and the charging flow control valve full open within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-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 HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

4.5.4 In accordance with the Surveillance Frequency Control Program. verify manual seal injection throttle valves are adjusted to give a flow within the limit with centrifugal charging pump discharge header pressure 2!: 2430 psig, and the charging flow control valve full open.

The provisions of Specification 4.0.4 are not applicable for entry into Mode 3. This exemption is allowed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after the Reactor Coolant System pressure stabilizes at 2235 +/- 20 psig.

SALEM - UNIT 1 3/45-6b Amendment No. 299

EMERGENCY CORE COOLING SYSTEMS REFUELING WATER STORAGE TANK

~IMITING CONDIIION FOR OPERATION 3.5.5 The refueling water storage tank (RWST) shall be OPERABLE with:

a. A contained volume of between 364,500 and 400,000 gallons of borated water.

b. A boron concentration of between 2300 and 2500 ppm, and

c. A minimum water temperature of 35°F.

APPLICABILITY: MODES 1.2.3 and 4.

ACTION:

With the refueling water storage tank inoperable. restore the tank to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 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 .

4.5.5 The RWST shall be demonstrated OPERABLE:

a. In accordance with the Surveillance Frequency Control Program by:

1. Verifying the water level in the tank. and

2. Verifying the boron concentration of the water.

b. In accordance with the Surveillance Frequency Control Program by verifying the RWST temperature when the outside air temperature is < 35°F.

SALEM - UNIT 1 3/45-7 Amendment No. 299

3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATIOf'J 3.6.1.1 Primary CONTAINMENT INTEGRITY shall be maintained.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

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

SURVEILLANCE REQUIREMENTS 4.6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated:

a1. In accordance with the Surveillance Frequency Control Program by verifying that each containment manual valve or blind flange that is located outside containment and required to be closed during accident conditions is closed, except for containment isolation valves that are open under administrative controls. Valves and blind flanges in high radiation areas may be verified by use of administrative controls.

a2. Prior to entering Mode 4 from Mode 5 if not performed within the last 92 days by verifying that each containment manual valve or blind flange that is located inside containment and required to be closed during accident conditions is closed, except for containment isolation valves that are open under administrative controls. Valves and blind flanges in high radiation areas may be verified by use of administrative controls.

b. By verifying that each containment air lock is OPERABLE per Specification 3.6.1.3.

c. In accordance with the Surveillance Frequency Control Program by verifying that the surveillance requirements of 4.6.2.3.a are met for penetrations associated with the containment fan coil units.

d. In accordance with the Surveillance Frequency Control Program by verifying that the surveillance requirements of 4.6.2.3.d are met for penetrations associated with the containment fan coil units.

SALEM UNIT 1 3/46-1 Amendment No. 299

CONTAINMENT SYSTEMS CONTAINMENT LEAKAGE LIMITING CONDITION FOR OPERATION

3. 6.1.2 Containment leakage rates shall be limited to:

a. An overall integrated leakage rate (Type A test) in accordance with the Containment Leakage Rate Testing Program.

b. A combined leakage rate in accordance with the Containment Leakage Rate Testing Program for all penetrations and valves subject to Type B and C tests.

APPLICABILITY: MODES 1, 2, 3 and 4.

I ACTION:

With either (a) the measured overall integrated containment leakage (Type A test) not in accordance with the Containment Leakage Rate Testing Program, or (b) with the measured combined leakage rate for all penetrations and valves subject to Types B and C tests not in accordance with the Containment Leakage Rate Testing Program, restore the leakage rate(s) to within the limit(s) prior to increasing the Reactor Coolant System temperature above 200'F.

SURVEILLANCE REQUIREMENTS 4.6.1.2 The containment leakage rates shall be demonstrated as follows:

a. Type A tests shall be in accordance with the Containment Leakage Rate Testing Program.

b. Type B and C tests shall be conducted in accordance with the Containment Leakage Rate Testing Program.

c. Air locks shall be tested and demonstrated OPERABLE in accordance with the Containment Leakage Rate Testing Program.

SALEM - UNIT 1 3/4 6-2 Amendment No. 225 i'OV 0 2 t9gj

PAGES 3/4 6-3 AND 3/4 6-4 ARE INTENTIONALLY BLANK SALEM - UNIT 1 3/4 6-3, 3/4 6-4 Amendment No.18 4 IJUL 1 1995

CONTAINMENT SYSTEMS CONTAINMENT AIR LOCKS LIMITING CONDITION FOR OPERATION 3.6.1.3 Each containment air lock shall be OPERABLE with:

a. Both doors closed except when the air lock is being used for normal transit entry and exit through the containment, then at least one air lock door shall be closed, and:

b. An overall air lock leakage rate in accordance with the Containment Leakage Rate Testing Program.

APPLICABILITY: MODES 1, 2, 3, and 4 ACTION:

Notes (1) Entry and exit is permissible to perform repairs on the affected air lock components.

(2) Separate condition entry is allowed for each air lock.

(3) Required ACTIONS a.1, a.2, and a.3 are not applicable if both doors in the same air lock are inoperable and condition c. is entered.

(4) Required ACTIONS b.1, b.2, and b.3 are not applicable if both doors in the same air lock are inoperable and condition c. is entered.

(5) Enter applicable Conditions and required Actions of LCO 3.6.1, "Primary Containment," when air lock leakage results in exceeding the overall containment leakage rate.

a. One or more containment air locks with one containment airlock door inoperable:

1. Verify the OPERABLE door is closed in the affected air lock within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and:

2. Lock the OPERABLE door closed in the affected air lock within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and:

3. Verify the OPERABLE door is locked closed in the affected air lock once per 31 days. Entry and exit is permissible for 7 days (from initial LCO entry) under administrative controls if one door is inoperable in each air lock. Air lock doors in high radiation areas may be verified locked closed by administrative means.

b. One or more containment air locks with only the containment air lock interlock mechanism inoperable.

1. Verify an OPERABLE door is closed in the affected air lock within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and:

2. Lock an OPERABLE door closed in the affected air lock within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and:

3. Verify an OPERABLE door is locked closed in the affected air lock once per 31 days. Entry and exit of containment is permissible under the control of a dedicated individual for the duration of the entry to ensure only one door is open at a time. Air lock doors in high radiation areas may be verified locked closed by administrative means.

SALEM - UNIT 1 3/4 6-5 Amendment No- 215 DEC EJ/ig

CONTAINMENT SYSTEMS CONTAINMENT AIR LOCKS LIMITING CONDITION FOR OPERATIONS (Continued)

c. One or more containment air locks inoperable for reasons other than condition a.

or b.

1. Immediately initiate action to evaluate overall containment leakage per LCO 3.6.1, and:

2. Verify that at least one door is closed in the affected air lock within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and:

3. Restore the air lock to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

d. If the ACTIONS and associated completion times of a., b., or c. cannot be met, be in 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 .

4.6.1.3 Each containment air lock shall be demonstrated OPERABLE:

a. By verifying seal leakage rate in accordance with the Containment Leakage Rate Testing program.

b. By conducting an overall air lock leakage test in accordance with the Containment Leakage Rate Testing Program.

c. In accordance with the Surveillance Frequency Control Program by verifying that only one door in each air lock can be opened at a time.

SALEM - UNIT 1 3/46-5a Amendment No. 299

CONTAI NMENT SYSTEMS INTERNAL PRESSURE LIMITING CONDITION FOR OPERATION 3.6.1.4 Primary containment internal pressure shall be maintained between -1.5 and +0.3 psig.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With the containment internal pressure outside of the limits above, restore the internal pressure to within the limits within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

4.6.1.4 The primary containment internal pressure shall be determined to be within the limits in accordance with the Surveillance Frequency Control Program.

SALEM - UNIT 1 3/4 6-6 Amendment No. 299

CONTAINMENT SYSTEMS AIR TEMPERATURE LIMITING CONDITION FOR OPERATION 3.6.1.5 Primary containment average air temperature shall not exceed 120°F.

APPLICABILITY: MODES 1,2,3 and 4.

ACTION:

With the containment average air temperature> 120°F, reduce the average air temperature to within the limit within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.6.1.5 Verify the Containment Average Air Temperature is within limit in accordance with the Surveillance Frequency Control Program.

SALEM - UNIT 1 3/4 6-7 Amendment 299

CONTAINMENT SYSTEMS CONTAINMENT STRUCTURAL INTEGRITY LIMITING CONDITIONS FOR OPERATION 3.6.1.6 The structural integrity of the containment shall be maintained at a level consistent with the acceptance criteria in Specification 4.6.1.6. 1.

APPLICABILITY: MODES 1, 2, 3 and 4.

I ACTION:

With the structural integrity of the containment not conforming to the above requirements, restore the structural integrity to within the limits prior to increasing the Reactor Coolant System temperature above 200 0 F.

SURVEILLANCE REQUIREMENTS 4.6.1.6.1 Containment Surfaces The structural integrity of the exposed accessible interior and exterior surfaces of the containment, including the liner plate, shall be determined in accordance with the Containment Leakage Rate Testing Program.

4.6.1.6.2 Reports Any abnormal degradation of the containment structure detected during the above required inspections shall be evaluated for reportability pursuant to 10CFR50.72 and 10CFR50.73. The evaluation shall be documented and shall include a description of the condition of the concrete, the inspection procedure, the tolerances on cracking, and the corrective action taken.

SALEM - UNIT 1 3/4 6-8 Amendment No. 225 NCoV G. i,^

This page left intentionally blank.

Note that the elements of TS 3.6.1.7 and 4.6.1.7 were relocated to TS 3/4.6.3.

277 SALEM - UNIT 1 3/4 .6-8a Amendment No.

CONTAINMENT SYSTEMS 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT SPRAY SYSTEM LIMITING CONDITION FOR.~ObP~EldRA~Tddl~O~N~================

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

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

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

SURVEILLANCE REQUIREMENTS 4.6.2.1 Each containment spray system shall be demonstrated OPERABLE:

a. In accordance with the Surveillance Frequency Control Program by verifying that each valve (manual, power operated or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.

b. By verifying, that on recirculation flow, each pump develops a differential pressure of greater than or equal to 204 psid when tested pursuant to the Inservice Testing Program,

c. In accordance with the Surveillance Frequency Control Program during shutdown, by:

1. Verifying that each automatic valve in the flow path actuates to its correct position on a Containment High-High pressure test signal.

2. Verifying that each spray pump starts automatically on a Containment High-High pressure test signal.

d. Following activities that could result in nozzle blockage, either evaluate the work performed to determine the impact to the containment spray system, or perform an air or smoke flow test through each spray header and verifying each spray nozzle is unobstructed.

SALEM - UNIT 1 3/4 6-9 Amendment No. 299

CONTAINMENT SYSTEMS SPRAY ADDITIVE SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.2 The spray additive system shall be OPERABLE with:

a. A spray additive tank containing a volume of between 2568 and 4000 gallons of between 30 and 32 percent by weight NaOH solution, and

b. Two spray additive eductors each capable of adding NaOH solution from the chemical additive tank to a containment spray system pump flow.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

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

SURVEILLANCE REQUIREMENTS 4.6.2.2 The spray additive system shall be demonstrated OPERABLE:

a. In accordance with the Surveillance Frequency Control Program by verifying that each valve (manual, power operated or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.

b. In accordance with the Surveillance Frequency Control Program by:

1. Verifying the solution level in the tank, and

2. Verifying the concentration of the NaOH solution by chemical analYSis.

c. In accordance with the Surveillance Frequency Control Program during shutdown, by verifying that each automatic valve in the flow path actuates to its correct position on a Containment High-High pressure test signal.

d. In accordance with the Surveillance Frequency Control Program by:

1. Verifying a NaOH solution flow rate of 12 +/- 3 gpm from the spray additive tank through sample valve 1CS61 with the spray additive tank at 2.5 +/- 0.5 psig and SALEM - UNIT 1 3/46-10 Amendment No. 299

CONTAINMENT SYSTEMS SPRAY ADDITIVE SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

2. Verifying that the spray additive tank eductor flow will be 35 +/- 3.5 gpm to each containment spray system. Testing may be performed by measuring the flow of borated water from the RWST through the installed 2" test line and Valve CS31; using this test line up with the spray pump operating in the recirculation mode and the RUST level at 41 feet + 0.5 feet, the measured flow shall be 57 gpm +/-

5.7 gpm.

SALEM - UNIT 1 3/4 6-lOa Amendment No. 122, MAR 2 5 1991

CONTAINMENT SYSTEMS CONTAINMENT COOLING SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.3 Five containment cooling fans shall be OPERABLE.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

a. With one or two of the above required containment cooling fans inoperable, restore the inoperable cooling fan(s) to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With three or more of the above required containment cooling fans inoperable, restore at least three cooling fans to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY WITHIN the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . Restore the remaining inoperable cooling fans to OPERABLE status within 7 days of initial loss or be in at least-HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.6.2.3 Each containment cooling fan shall be demonstrated OPERABLE:

SALEM - UNIT I 3/4 6-11 Amendment No. 2 6 6

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

a. In accordance with the Surveillance Frequency Control Program by:

1. Verifying the water level in each service water accumulator vessel is greater than or equal to 226 inches and less than or equal to 252 inches.

2. Verifying the temperature in each service water accumulator vessel is greater than or equal to 55"F and less than or equal to 95°F.

3. Verifying the nitrogen cover pressure in each service water accumulator vessel is greater than or equal to 135 pSig and less than or equal to 160 pSig.

b. In accordance with the Surveillance Frequency Control Program by:

1. Starting (unless already operating) each fan from the control room in low speed.

2. Verifying that each fan operates for at least 15 minutes in low speed.

3. Verifying a cooling water flow rate of greater than or equal to 1300 gpm to each cooler.

c. In accordance with the Surveillance Frequency Control Program by verifying that on a safety injection test signal:

1. Each fan starts automatically in low speed.

2. The automatic valves and dampers actuate to their correct positions and that the cooling water flow rate to each cooler is greater than or equal to 1300 gpm.

d. In accordance with the Surveillance Frequency Control Program by verifying that on a loss of offsite power test Signal, each service water accumulator vessel discharge valve response time is within limits.

SALEM - UNIT 1 3/46-11a Amendment No. 299

CONTAINMENT SYSTEMS 3/4.6.3 CONTAINMENT ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.6.3.1 Each containment isolation valve shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

NOTE 1 Penetration flow paths, except for the containment purge valves, may be unisolated intermittently under administrative controls.

Note 2 A containment purge valve is not a required containment isolation valve when its flow path is isolated with a testable blind flange tested in accordance with SR 4.6.1.2.b. The required containment purge supply and exhaust isolation valves shall be closed. (Valves immobilized in shut position with control air to valve operators isolated and tagged out of service).

NOTE 3 The containment pressure-vacuum relief isolation valves may be opened on an intermittent basis, under administrative control, as necessary to sati!fy the requirement of Specification 3.6.1.4.

1. With one or more of the isolation valve(s) inoperable, maintain at least one isolation valve OPERABLE in each affected penetration that is open and either:

a. Restore the inoperable valve(s) to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , or

b. Isolate each affected penetration within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least one deactivated automatic valve secured in the isolation position, or

c. Isolate each affected penetration within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least one closed manual valve or blind flange; or

d. 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 .

2. With one required containment purge supply and/or exhaust isolation valve open, close the open valve(s) within one hour or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.6.3.1.1 DELETED SALEM - UNIT 1 3/4 6-12 Amendment No.277

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 4.6.3.1.2 Each containment isolation valve shall be demonstrated OPERABLE during the COLD SHUTDOWN or REFUELING MODE in accordance with the Surveillance Frequency Control Program by:

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

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

c. Not used.

d. Verifying that on a Containment Purge and Pressure-Vacuum Relief isolation test signal, each required Purge and each Pressure-Vacuum Relief valve actuates to its isolation position.

e. Verifying that the Containment Pressure-Vacuum Relief Isolation valves are limited to s 60% opening angle.

4.6.3.1.3 In accordance with the Surveillance Frequency Control Program, verify that on a main steam isolation test signal, each main steam isolation valve actuates to its isolation position.

4.6.3.1.4 The isolation time of each power operated or automatic containment isolation valve shall be determined to be within its limit when tested pursuant to the Inservice Testing Program.

4.6.3.1.5 Each required containment purge isolation valve shall be demonstrated OPERABLE within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after each closing of the valve, except when the valve is being used for multiple cyclings, then in accordance with the Surveillance Frequency Control Program, by verifying that when the measured leakage rate is added to the leakage rates determined pursuant to Specification 4.6.1.2.b for all other Type Band C penetrations, the combined leakage rate is less than or equal to 0.60La.

4.6.3.1.6 A pressure drop test to identify excessive degradation of resilient valve seals shall be conducted on the:

a. Required Containment Purge Supply and Exhaust Isolation Valves in accordance with the Surveillance Frequency Control Program.

b. Deleted.

4.6.3.1.7 The required containment purge supply and exhaust isolation valves shall be determined closed in accordance with the Surveillance Frequency Control Program.

SALEM - UNIT 1 3/46-13 Amendment No. 299

PAGES 3/4 6-14 THROUGH 3/4 6-19 ARE INTENTIONALLY LEFT BLANK SALEM - UNIT 1 3/4 6-14 through 19 Amendment No.281

3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE SAFETY VALVES LIMITING CONDITION FOR OPERATION 3.7.1.1 All main steam line code safety valves (MSSVs) associated with each steam generator shall be OPERABLE.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

a. With one or two main steam line code safety valves inoperable in one or more steam generators, operation in Modes 1, 2 and 3 may proceed provided, that within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , either the inoperable valve is restored to OPERABLE status or reduce power to less than or equal to the applicable percent of RATED THERMAL POWER per Table 3.7-1; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With three main steam line code safety valves inoperable in one or more steam generators, operation in Modes 1, 2 and 3 may proceed provided, that within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , either the inoperable valves are restored to OPERABLE status or reduce power to less than or equal to the applicable percent of RATED THERMAL POWER per Table 3.7-1 and within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months , reduce the Power Range Neutron Flux High trip setpoint to less than or equal to the RATED THERMAL POWER per Table 3.7-1; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.7.1.1 Verify each required MSSV lift setpoint per Table 4.7-1. No additional Surveillance Requirements other than those required by the Inservice Testing Program.

SALEM - UNIT 1 3/47-1 Amendment No. 297

laIt TABLE 3.7-1 C

MAXIMUM ALLOWABLE POWER WITH INOPERABLE STEAM LINE SAFETY VALVES Maximum Number of Inoperable Safety Maximum Allowable Power*

Valves on Any Operating Steam Generator (Percent of RATED THERMAL POWER) 1 87 2 59 3 39

The values do not provide any allowance for calorimetric error.

SALEM - UNIT 1 3/4 7-2 Amendment No. 244

This Page Intentionally Blank SALEM - UNIT 1 3/4 7-3 Amendment No.244

1C C C TABLE 4.7-1 STEAM LINE SAFETY VALVES PER LOOP LIFT VALVE NUMBER SETTING (i 3%)* ORIFICE SIZE I (sq. inches)

Loop A Loop B Loop C Loop D

a. 11MS1l 12MS11 13MS11 14MS11 1125 psig 16.0

b. 1lMS12 12MS12 13MS12 14MS12 1120 psig 16.0 C. llMS13 12MS13 13MS13 14MS13 1110 psig 16.0

d. llMS14 12MS14 13MS14 14MS14 1100 psig 16.0

e. 11MS15 12MS15 13MS15 14MS15 1070 psig 16.0

Following testing the lift setting shall be reset to within +/- 1%.

I SALEM - UNIT I 3/4 7-4 Amendment No. 2 44

PLANT SYSTEMS AUXILIARY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2 At least three independent steam generator auxiliary feedwater pumps and associated manual activation switches in the control room and flow paths shall be OPERABLE with:

a. Two feedwater pumps, each capable of being powered from separate vital busses, and

b. One feedwater pump capable of being powered from an OPERABLE steam supply system.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

a. With one auxiliary feedwater pump inoperable, restore the required auxiliary feedwater pumps to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. With two auxiliary feedwater pumps inoperable 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 HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c. With three auxiliary feedwater pumps inoperable, immediately initiate corrective ACTION to restore at least one auxiliary feedwater pump to OPERABLE status as soon as possible.

d. LCO 3.0A.b is not applicable.

4.7.1.2 Each auxiliary feedwater pump shall be demonstrated OPERABLE:

a. In accordance with the Surveillance Frequency Control Program by:

1. Verifying that each non-automatic valve in the flow path that is not locked, sealed or otherwise secured in position, is in its correct position.

2. Verify the manual maintenance valves in the flow path to each steam generator are locked open.

SALEM - UNIT 1 3/47-5 Amendment No. 299

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (continued)

b. In accordance with the Surveillance Frequency Control Program by:

1. Verify that the developed head of each motor driven pump at the flow test point is greater than or equal to the required developed head.

2. Verify that the developed head of the steam driven pump at the flow test point is greater than or equal to the required developed head when the steam generator pressure is > 680 psig. The provisions of Specification 4.0.4 are not applicable provided the surveillance is performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after secondary side pressure is greater than 680 psig.

c. In accordance with the Surveillance Frequency Control Program by:

1. Verifying that each auxiliary feedwater automatic valve that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.

2. Verifying that each auxiliary feedwater pump starts automatically on an actual or simulated actuation signal.

The provisions of Specification 4.0.4 are not applicable to the turbine driven auxiliary feedwater pump provided the surveillance is performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the secondary steam generator pressure is greater than 680 psig.

SALEM - UNIT 1 3/4 7-6 Amendment No. 299

PLANT SYSTEMS AUXILIARY FEED STORAGE TANK LIMITING CONDITION FOR OPERATION 3.7.1.3 The auxiliary feed storage tank (AFST) shall be OPERABLE with a minimum contained volume of 200,000 gallons of water.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

With the auxiliary feed storage tank inoperable, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either:

a. Restore the AFST to OPERABLE status or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , or

b. Demonstrate the OPERABILITY of a demineralized water or a fire protection/domestic water storage tank as a backup supply to the auxiliary feedwater pumps and restore the auxiliary feed storage tank to OPERABLE status within 7 days or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.7.1.3.1 The auxiliary feed storage tank shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program by verifying the water level is within its limits when the tank is the supply source for the auxiliary feedwater pumps.

4.7.1.3.2 A demineralized water storage tank shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program by verifying the tank contains;:: 200,000 gallons of water and by verifying proper alignment of valves for taking suction from this tank when it is the supply source for the auxiliary feedwater pumps.

4.7.1.3.3 A fire protection/domestic water storage tank shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program by verifying the tank contains

200,000 gallons of water and by verifying proper alignment of valves for taking suction from this tank when it is the supply source for the auxiliary feedwater pumps.

SALEM - UNIT 1 3/4 7-7 Amendment No. 299

I .. ,I PLANT SYSTEMS

,J ACTIVITY LIMITING CONDITION FOR OPERATION 3.7.1.4 The specific activity of the secondary coolant system shall be

<0.10 uCi/gram DOSE EQUIVALENT I-131.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

With the specific activity of the secondary coolant system > 0.10 PCi/

gram DOSE EQUIVALENT 1-131, 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 .

I SURVEILLANCE REQUIREMENTS 4.7.1.4 The specific activity of the secondary coolant system shall be determined to be within the limit by performance of the sampling and analysis program of Table 4.7-2.

SALEM-UNIT 1 3/4 7-8 I I

TABLE 4.7-2 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM TYPE OF MEASUREMENT AND ANALYSIS SAMPLE AND ANALYSIS FREQUENCY

1. Gross Activity Determination In accordance with the Surveillance Frequency Control Program

2. Isotopic Analysis for DOSE a) In accordance with the Surveillance EQUIVALENT 1-131 Concentration Frequency Control Program, whenever the gross activity determination indicates iodine concentrations greater than 10% of the allowable limit.

b) In accordance with the Surveillance Frequency Control Program, whenever the gross activity determination indicates iodine concentrations below 10% of the allowable limit.

SALEM - UNIT 1 3/4 7-9 Amendment No. 299

PLANT SYSTEMS MAIN STEAM LINE ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.7.1.5 Each main steam line isolation valve shall be OPERABLE.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

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

MODES 2 - With one or more main steam line isolation valve(s) inoperable, subsequent and 3 operation in MODES 2 or 3 may proceed provided;

a. The isolation valve(s) is (are) maintained closed, and

b. The isolation valve(s) is (are) verified closed once per 7 days.

Otherwise, be in MODE 3, HOT STANDBY, within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and MODE 4, HOT SHUTDOWN, within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.7.1.5 Each main steam line isolation valve shall be demonstrated OPERABLE by verifying full closure within 5 seconds when tested pursuant to the Inservice Testing Program. The provisions of Specification 4.0.4 are not applicable.

SALEM - UNIT 1 3/47-10 Amendment No. 297

I I PLANT SYSTEMS SECONDARY WATER CHEMISTRY -

LIMITING CONDITION FOR OPERATION This Page Intentionally Left Blank ALEM-UNIT 1 3/4 7-11 Amendment Mo. 25

.1 .

APR 22 1980

'.7.-

.,,Ir This Page Intentionally Left Blank I

SALEM-UNIT 1 3/4 7-12 Amendment No. 25 I I ( .

APR 2 2 1980

v This Page Intentionally Left Blank

. . 1

. . . I

' .* I I SALEM-UNIT 1 3/4 7-13 Amendment No. 25 APR 22 1980

PLANT SYSTEMS 3/4.7.2 STEAM GENERATOR PRESSURE/TEMPERATURE LIMITATION LIMITING CONDITION FOR OPERATION 3.7.2.1 The temperatures of both the primary and secondary coolants in the steam generators shall be> 70°F when the pressure of either coolant in the steam generator is> 200 psig.

APPLICABILITY: At all times.

ACTION:

With the requirements of the above specification not satisfied:

a. Reduce the steam generator pressure of the applicable side to ~ 200 psig within 30 minutes, and

b. Perform an engineering evaluation to determine the effect of the overpressurization on the structural integrity of the steam generator. Determine that the steam generator remains acceptable for continued operation prior to increasing its temperatures above 200°F.

4.7.2.1 The pressure in each side of the steam generator shall be determined to be < 200 psig in accordance with the Surveillance Frequency Control Program when the temperature of either the primary or secondary coolant is < 70°F.

SALEM - UNIT 1 3/47-14 Amendment No. 299

PLANT SYSTEMS 3/4.7.3 COMPONENT COOLING WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.3.1 At least two independent component cooling water loops shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With only one component cooling water loop OPERABLE, restore at least two loops to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.7.3.1 At least two component cooling water loops shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program by verifying that each valve (manual, power operated or automatic) servicing safety related equipment that is not locked, sealed, or otherwise secured in position, is in its correct position.

SALEM - UNIT 1 3/47-15 Amendment No. 299

PLANT SYSTEMS 3/4.7.4 SERVICE WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.4.1 At least two independent service water loops shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With only one service water loop OPERABLE, restore at least two loops to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months

or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.7.4.1 At least two service water loops shall be demonstrated OPERABLE:

a. In accordance with the Surveillance Frequency Control Program by verifying that each valve (manual, power operated or automatic) servicing safety related equipment that is not locked, sealed, or otherwise secured in position, is in its correct position.

b. In accordance with the Surveillance Frequency Control Program during shutdown, by verifying that each automatic valve servicing safety related equipment actuates to its correct position on Safeguards Initiation signal.

Operation with only the 11 service water loop OPERABLE may continue for up to 10 days. This note is applicable for one time use during Salem Unit No.1 Cycle 15.

SALEM - UNIT 1 3/47-16 Amendment No. 299

PLANT SYSTEMS 3/4.7.5 FLOOD PROTECTION LIMITING CONDITION FOR OPERATION 3.7.5.1 Flood protection shall be provided for all safety related systems, components and structures when the water level of the Delaware River exceeds 10.5' Mean Sea Level USGS datum, at the service water intake structure.

APPLICABILITY: At all times.

ACTION:

a. With the water level at the service water intake structure above elevation 10.5' Mean Sea Level USGS datum, close all watertight doors within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

b. With the water level at the service water intake structure above elevation 11.5' Mean Sea Level USGS datum, 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 .

SURVEILLANCE REQUIREMENTS 4.7.5.1 The water level at the service water intake structure shall be determined to be within the limits by:

a. Measurement in accordance with the Surveillance Frequency Control Program when the water level is below elevation 10.5' Mean Sea Level USGS datum, and

b. Measurement in accordance with the Surveillance Frequency Control Program when the water level is equal to or above elevation 10.5' Mean Sea Level USGS datum.

SALEM ~ UNIT 1 3/47-17 Amendment No. 299

PLANT SYSTEMS 3/4.7.6, CONTROL ROOM EMERGENCY AIR CONDITIONING SYSTEM LIMITING CONDITION FOR OPERATION 3.7.6.1 The common control room emergency air conditioning system (CREACS) shall be OPERABLE with:

a. Two independent air conditioning filtration trains (one from each unit) consisting of:

1. Two fans and associated outlet dampers,

2. One cooling coil,

3. One charcoal adsorber and HEPA filter array,

4. Return air isolation damper.

b. All other automatic dampers required for operation in the pressurization or recirculation modes.

c. The control room envelope intact.

NOTE: The control room envelope (CRE) boundary may be opened intermittently under administrative control.

APPLICABILITY: ALL MODES and during movement of irradiated fuel assemblies.

ACTION: MODES 1, 2, 3, and 4

a. With one filtration train inoperable, align CREACS for single filtration train operation within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , and restore the inoperable filtration train to OPERABLE status within 30 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 .

b. With CREACS aligned for single filtration train operation and with one of the two remaining fans or associated outlet damper inoperable, restore the inoperable fan or damper to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c. With the Control Room Envelope boundary inoperable:

1. Immediately, initiate action to implement mitigating actions, and

2. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , verify mitigating actions ensure CRE occupant exposures to radiological, chemical, and smoke hazards will not exceed limits, and

3. Within 90 days, ýrestore the Control Room Envelope boundary to OPERABLE status, 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 .

The CREACS is a shared system with Salem Unit 2 SALEM - UNIT 1 3/4 7-18 Amendment No. 286

PLANT SYSTEMS:

LIMITING CONDITION FOR OPERATION (Continued)

d. With one. or both series isolation damper(s) on a normal Control Area Air Conditioning System (CAACS) outside air intake or exhaust duct inoperable, close the affected duct within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least one isolation damper secured in the closed position 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 . (Refer to ACTION 25 of Table 3.3-6.)

e. With one or both isolation damper(s) on an outside emergency air conditioning air intake duct inoperable, close the affected duct within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least one isolation damper secured in the closed position and restore the damper(s) to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

f. With any isolation damper between the normal CAACS and the CREACS inoperable, secure the damper in the closed position within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

MODES 5 and 6 or during movement of irradiated fuel assemblies

a. With one filtration train inoperable, align CREACS for single filtration train operation within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , or suspend movement of irradiated fuel assemblies.

b. With CREACS aligned for single filtration train operation with one of the two remaining fans or associated outlet damper inoperable, restore the fan or damper to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or suspend movement ofirradiated fuel assemblies.

c. With two filtration trains inoperable, immediately suspend movement of irradiated fuel assemblies.

d. With the Control Room Envelope boundary inoperable, immediately suspend movement of irradiated fuel assemblies.

e. With one or both series isolation damper(s) on a normal CAACS outside air intake or exhaust duct inoperable, immediately suspend movement of irradiated fuel assemblies until the affected duct is closed by use of at least one isolation damper secured in the closed position. (Refer to ACTION 25 of Table 3.3-6.)

f. With one or both series isolation damper(s) on an outside emergency air conditioning air intake duct inoperable,, immediately suspend movement of irradiated fuel assemblies until the affected duct is closed by use of at least one isolation damper secured in the closed position. To resume movement of irradiated fuel assemblies, at least one emergency air intake duct must be operable on each unit.

g. With any isolation damper between the CAACS and the CREACS inoperable, immediately suspend movement of irradiated fuel assemblies until the damper is closed and secured in the closed position.

SALEM - UNIT 1 3/4 7-19 Amendment No. 286

PLANT SYSTEMS 4.7.6.1 Each control room emergency air conditioning system filtration train shall be demonstrated OPERABLE:

a. In accordance with the Surveillance Frequency Control Program by initiating flow through the HEPA filter and charcoal adsorber train(s) and verifying that the train(s) operates with each fan operating for at least 15 minutes.

b. In accordance with the Surveillance Frequency Control Program or prior to return to service (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the system, by:

1. Verifying that the charcoal adsorbers remove 2! 99% of a halogenated hydrocarbon refrigerant test gas when they are tested in-place while operating the ventilation system at a flow rate of 8000 cfm +/- 10%.

2. Verifying that the HEPA filter banks remove 2! 99% of the OOP when they are tested in-place while operating the ventilation system at a flow rate of 8000 cfm +/- 10%.

3. Verifying within 31 days after removal from the CREACS unit, that a laboratory test of a sample of the charcoal adsorber, when obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration less than 2.5% when tested in accordance with ASTM 03803-1989 at a temperature of 30"C and a relative humidity of 95%.

c. After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation by verifying within 31 days after removal from the CREACS unit, that a laboratory analysis of a representative carbon sample, when obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows a methyl iodide penetration less than 2.5% when tested in accordance with ASTIVI 03803-1989 at a temperature of 30"C and a relative humidity of 95%.

d. In accordance with the Surveillance Frequency Control Program by:

1. Verifying that the pressure drop across the combined HEPA filter and charcoal adsorber bank is ::; 3.5 inches water gauge while operating the ventilation system at a flow rate of 8000 cfm +/- 10%.

SALEM - UNIT 1 3/4 7-20 Amendment No. 299

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

2. Verifying that on a safety injection test signal or control room intake high radiation test signal, the system automatically actuates in the pressurization mode by opening the outside air supply and diverting air flow through the HEPA filter and charcoal adsorber bank.

3. Deleted.

4. Verifying that on a manual actuation signal, the system will actuate to the required pressurization or recirculation operating mode.

5. Verify each CREACS train has the capability to remove the assumed heat load.

e. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove ~ 99% of the DOP when they are tested in-place while operating the filter system at a flow rate of 8000 cfm +/- 10%.

f. After each complete or partial replacement of a charcoal absorber bank by verifying that the charcoal absorbers remove ~ 99% of a halogenated hydrocarbon refrigerant test gas when they are tested in-place while operating the filter system at a flow rate of 8000 cfrn +/- 10%.

4.7.6.2 Perform required CRE unfiltered air inleakage testing in accordance with the Control Room Habitability Program (Refer to TS 6.18).

SALEM - UNIT 1 3/47-21 Amendment No. 295

PLANT SYSTEMS 3/4.7.7 AUXILIARY BUILDING VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.7.7.1 At least two supply fans, and three exhaust fans shall be OPERABLE (*)

to maintain the Auxiliary Building at slightly negative pressure.

NOTE-----------------------------------------

The intermittent opening of the Auxiliary Building pressure boundary causing a loss of negative pressure may be performed under administrative controls.

APPLICABILITY: At all times ACTION:

Modes 1 thru 4 a) With one supply fan and/or one exhaust fan inoperable, restore the fan(s) to OPERABLE status within 14 days or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b) With two supply and/or two exhaust fans inoperable restore at least one inoperable supply and two exhaust fans to operable status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c) With the Auxiliary Building pressure not maintained slightly negative, restore the Auxiliary Building to slightly negative pressure within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in 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 .

During CORE ALTERATIONS d) With the Auxiliary Building pressure not maintained slightly negative, restore the Auxiliary Building to slightly negative pressure within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or suspend all operations involving CORE ALTERATIONS.

At all times e) With the Auxiliary Building pressure not maintained slightly negative, suspend all operations involving radioactive gaseous releases via the Auxiliary Building immediately.

(*) One of the supply fans may be considered OPERABLE with its auto start circuit administratively controlled (removed from service) to prevent more than one supply fan from operating at any time.

SALEM - UNIT I 3 /4 7 -22 Amendment No. 271

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS 4.7.7.1 The above required Auxiliary Building Ventilation System shall be demonstrated OPERABLE by:

a. In accordance with the Surveillance Frequency Control Program by verifying negative pressure in the Auxiliary Building.

b. In accordance with the Surveillance Frequency Control Program by starting each fan, from the control room, and verifying that each fan operates for at least 15 minutes.

c. In accordance with the Surveillance Frequency Control Program by verifying that the System starts following a Safety Injection Test Signal.

SALEM - UNIT 1 3/47-23 Amendment No. 299

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS THIS PAGE LEFT BLANK INTENTIONALLY SALEM - UNIT 1 3 /4 7-24 Amendment No.271

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS THIS PAGE LEFT BLANK INTENTIONALLY SALEM - UNIT 1 3 /4 7-25 Amendment No. 271

PLANT SYSTEMS 3/4.7.8 SEALED SOURCE CONTAMINATION LIMITING CONDITlQN FOR OPERATION 3.7.8.1 Each sealed source containing radioactive material either in excess of 100 microcuries of beta and/or gamma emitting material or 5 microcuries of alpha emitting material shall be free of;?; 0.005 microcuries of removable contamination.

APPLICABILITY: At all times.

ACTION:

a. Each sealed source with removable contamination in excess of the above limits shall be immediately withdrawn from use and:

1. Either decontaminated and repaired, or

2. Disposed of in accordance with Commission Regulations.

b. The provisions of Specification 3.0.3 are not applicable.

4.7.8.1.1 Test Requirements - Each sealed source shall be tested for leakage and/or contamination by:

a. The licensee, or

b. Other persons specifically authorized by the Commission or an Agreement State.

The test method shall have a detection sensitivity of at least 0.005 microcuries per test sample.

4.7.8.1.2 Test Frequencies - Each category of sealed sources shall be tested at the frequency described below.

a. Sources in use (excluding startup sources and fission detectors previously subjected to core flux) - In accordance with the Surveillance Frequency Control Program for all sealed sources containing radioactive materials.

SALEM UNIT 1 3/47-26 Amendment No. 299

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

1. With a half-life greater than 30 days (excluding Hydrogen 3),

and

2. In any form other than gas.

b. Stored sources not in use - Each sealed source and fission detector shall be tested prior to use or transfer to another licensee unless tested within the previous six months. Sealed sources and fission detectors transferred without a certificate indicating the last test date shall be tested prior to being placed into use.

c.. Startup sources and fission detectors - Each sealed startup source and fission detector shall be tested within 31 days prior to being subjected to core flux and following repair or maintenance to the source or detector.

4.7.8.1.3 Reports - A report shall be prepared and submitted to the Commission on an annual basis if sealed source or fission detector leakage tests reveal the presence of > 0.005 microcuries of removable contamination.

SALEM-UNIT 1 3/4 7-27

PLANT SYSTEMS 3/4.7.9 SNUBBERS LIMITING CONDITION FOR OPERATION 3.7.9 All snubbers shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4. (MODES 5 and 6 for snubbers located on systems required OPERABLE in those MODES).

ACTION:

With one or more snubbers inoperable, within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , replace or restore the inoperable snubber(s) to OPERABLE status and perform an engineering evaluation per Specification 4.7.9c on the supported component or declare the supported system inoperable and follow appropriate ACTION statement for that system.

SURVEILLANCE REQUIREMENTS 4.7.9 Each snubber shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program, in addition to the regular Inservice Inspection Program requirements.

a. Visual Inspection All snubbers shall be categorized into two groups: those accessible and those inaccessible during reactor operation. The visual inspection interval for each category of snubbers shall be determined based upon the criteria provided in Table 4.7-3.

SALEM - UNIT 1 3/47-28 Amendment No. 297

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

b. Visual Inspection Acceptance Criteria Visual inspections shall verify (1) that there are no visible indications of damage or impaired OPERABILITY, (2) attachments to the foundation or supporting structure are secure, and (3) In those locations where snubber movement can be manually induced without disconnnecting the snubber, that the snubber has freedom of movement and is not frozen up. Snubbers which appear inoperable as a result of visual inspections may be determined OPERABLE for the purpose of establishing the next visual inspection interval, providing that (1) the cause of the rejection is clearly established and remedied for that particular snubber and for other snubbers that may be generically susceptible; and (2) the affected snubber is functionally tested in the as found condition and determined OPERABLE per Specification 4.7.9d or 4.7.9e as applicable. However, when the fluid port of a hydraulic snubber is found to be uncovered, the snubber shall be declared inoperable and cannot be determined OPERABLE via functional testing for the purpose of establishing the next visual inspection interval. All snubbers connected to an inoperable common hydraulic fluid reservoir shall be counted as inoperable snubbers.

c. Functional Tests At least once per 18 months during shutdown, a representative sample of 10% of the total of each type of snubber in use in the plant shall be functionally tested either in place or in a bench test. For each type of snubber that does not meet the functional test acceptance criteria of Specification 4.7.9d or 4.7.9e, an additional 10% of that type of snubber shall be functionally tested.

The representative sample selected for functional testing shall include the various configurations, operating environments and the range of size and capacity of snubbers. At least 25% of the snubbers in the representative sample shall include snubbers from the following three categories:

1. The first snubber away from each reactor vessel nozzle

2. Snubbers within five feet of heavy equipment (valve, pump, turbine, motor, etc.)

3. Snubbers within ten feet of the discharge from a safety relief valve SALEM - UNIT 1 3/4 7-29 Amendment No. 93 MAY 1 1989

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

.'In addition to the regular sample, snubbers which failed the previous functional test shall be retested.during the next test period. If a spare snubber has been instilled in place of a failed snubber, then both the failed snubber (if it is repaired and installed in another position) and the spare snubber shall be retested. Test results of these snubbers may not be included for the re-sampling.

If any snubber selected for functional testing either fails to lockup or fails to move, i.e., frozen in place, the cause will be evaluated and if caused by manufacturer or design deficiency all snubbers of the same design subject to the same defect shall be functionally tested. This testing requirement shall be independent of the requirements stated above for snubbers not meeting the functional test acceptance criteria.

For the snubber(s) found inoperable, an engineering evaluation shall be performed on the components which are supported by the snubber(s). The purpose of this engineering evaluation shall be to determine if the components supported by the snubber(s) were adversely affected by the inoperability of the snubber(s) in order to ensure that the supported component remains capible of meeting the designed service.

d. Hydraulic Snubbers Functional Test Acceptance Criteria The hydraulic snubber functional test shall verify that:

1. Activation (restraining action) is achieved within the specified range of velocity or acceleration in both tension and compression.

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

e. Mechanical Snubbers Functional Acceptance Criteria The mechanical snubber functional test shall verify that:

1. The force that initiates free movement of the snubber rod in either'tension or compression is less than the specified maximum drag force.

2. Activation (restraining action) is achieved within the specified range of velocity or acceleration in both tension and compression.

SALEM - UNIT 1 3/4 7-30 Amendment No. 98 Effective Date:

May 12, 1989 JUN 5 1989

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

3. Snubber release rate, where required, is within the specified range in compression or tension.

For snubbers specifically required not to displace under continuous load, the ability of the snubber to withstand load without displacement shall be verified.

f. Snubber Service Life Monitoring A record of the service life of each snubber, the date at which the designated service life commences and the installation and maintenance records on which the designated service life is based shall be maintained as required by Specification 6.10.2.m.

Concurrent with the first inservice visual inspection and at least once per 18 months thereafter, the installation and maintenance records for each snubber shall be reviewed to verify that the indicated service life has not been exceeded or will not be exceeded prior to the next scheduled snubber service life review. If the indicated service life will be exceeded prior to the next scheduled snubber service life review, the snubber service life shall be reevaluated or the snubber shall be replaced or reconditioned so as to extend its service life beyond the date of the next scheduled service life review. This reevaluation, replacement or reconditioning shall be indicated in the records.

SALEM - UNIT 1 3/4 7-31 Amendment No.225 NOV 0 2 1999

TABLE 4.7-3 SNUBBER VISUAL INSPECTION INTERVAL Number of Unacceptable Snubbers Column A3 '6 Column B4,6 Column C5'6 1

Populations. Extend Repeat Reduce

/Category Interval Interval Interval 1 0 0 1 80 0 0 2 100 0 1 4 150 0 3 8 Notes: 1. The next visual inspection interval for the population of a snubber category shall be determined based upon the most recent inspection interval and the number of unacceptable snubbers found during that interval. Snubbers may be categorized, based upon their accessibility during power operation, as accessible or inaccessible. These categories may be examined separately or jointly. This decision shall be made and documented before any inspection and used as the basis upon which to determine the next inspection interval for that category.

2. Interpolation between population or category sizes and the number of unacceptable snubbers is permissible. Where the limit for unacceptable snubbers in Columns A, B, or C is determined by interpolation and includes a fractional value, the limit may be reduced to the next lower integer.

3. If the number of unacceptable snubbers is equal to or less than the number in Column A, the next inspection interval may be twice the previous interval but not greater than 48 months.

4. If the number of unacceptable snubbers is equal to or less than the number in Column B but greater than the number in Column A, the next inspection interval shall be the same as the current interval;

5. If the number of unacceptable snubbers is equal to or greater than the number in Column C, the next inspection interval shall be two-thirds of the current interval. However, if the number of unacceptable snubbers is less than the number in Column C but greater than the number in Column B, the next interval shall be reduced proportionally by interpolation, that is:

U -B II 10 * /3 C -B where: I1 - next inspection interval 10 - current inspection interval U

number of unacceptable snubbers found during the previous inspection interval B - number in Column B C - number in Column C

6. The provisions of Specification 4.0.2 are applicable for all inspection intervals up to and including 48 months.

SALEM - UNIT 1 3/4 7-32 Amendment No. 161 VC 1 2 1994

PLANT SYSTEMS 3/4.7.10 CHILLED WATER SYSTEM - AUXILIARY BUILDING SUBSYSTEM LIMITING CONDITION FOR OPERATION 3.7.10 The chilled water system loop which services the safety-related loads in the Auxiliary Building shall be OPERABLE with:

a. Three OPERABLE chillers

b. Two OPERABLE chilled water pumps APPLICABILITY: ALL MODES and during movement of irradiated fuel assemblies.

ACTION: MODES 1, 2, 3, and 4

a. With one chiller inoperable:

1. Remove the appropriate non-essential heat loads from the chilled water system within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and;

2. Restore the chiller to OPERABLE status within 14 days or; I

3. Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With two chillers inoperable:

1. Remove the appropriate non-essential heat loads from the chilled water system within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and;

2. Align the control room emergency air conditioning system (CREACs) for single filtration operation using the Salem Unit 2 train within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and;

3. Restore at least one chiller to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or; I

4. Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c. With one chilled water pump inoperable, restore the chilled water pump to OPERABLE status within 7 days or be in at least 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 HOT I

following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SALEM - UNIT I 3/4 7-33 Amendment No. 225 NOV 0 2 1399

PLANT SYSTEMS LIMITING CONPITION FOR OPERATION (Continued)

ACTION: MODES 5 and 6 or during movement of irradiated fuel assemblies. *

a. With one chiller inoperable:

1. Remove the appropriate non-essential heat loads from the chilled water system within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and;

2. Restore the chiller to OPERABLE status within 14 days or;

3. Suspend CORE ALTERATIONS and movement of irradiated fuel assemblies.

b. With two chillers inoperable:

1. Remove the appropriate non-essential heat loads from the chilled water system within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and;

2. Align the control room emergency air conditioning system (CREACs) for single filtration operation using the Salem Unit 2 train within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and;

3. Restore at least one chiller to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or;

4. Suspend CORE ALTERATIONS and movement of irradiated fuel assemblies.

c. With one chilled water pump inoperable, restore the chilled water pump to OPERABLE status within 7 days or suspend CORE ALTERATIONS and movement of irradiated fuel assemblies.

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

a. In accordance with the Surveillance Frequency Control Program by verifying that each manual valve in the chilled water system flow path servicing safety related components that is not locked, sealed, or otherwise secured in position, is in its correct position.

b. In accordance with the Surveillance Frequency Control Program, by verifying that each automatic valve actuates to its correct position on a Safeguards Initiation signal.

c. In accordance with the Surveillance Frequency Control Program by verifying that each chiller starts and runs.

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

SALEM - UNIT 1 3/47-34 Amendment No. 299

PLANT SYSTEMS 3/4.7.11 FUEL STORAGE POOL BORON CONCENTRATION LIMITING CONDITION FOR OPERATION 3.7.11 The fuel storage pool boron concentration shall be ~ 800 ppm.

APPLICABILITY: When fuel assemblies are stored in the fuel storage pool and a fuel storage pool verification has not been performed since the last movement of fuel assemblies in the fuel storage pool.

ACTION:

With fuel storage pool boron concentration not within limit:

a. Immediately suspend movement of fuel assemblies in the fuel storage pool and

b. Initiate action to:

1. immediately restore fuel storage pool boron concentration to within limit or

2. immediately perform a fuel storage pool verification.

c. LCO 3.0.3 is not applicable.

SURVEILLANCE REQUIREMENTS 4.7.11 Verify the fuel storage pool boron concentration is within limit in accordance with the Surveillance Frequency Control Program.

SALEM - UNIT 1 3/47-35 Amendment No. 299

PLANT SYSTEMS 3/4.7.12 FUEL ASSEMBLY STORAGE IN THE SPENT FUEL POOL LIMITING CONDITION FOR OPERATION 3.7.12 The combination of initial enrichment, burnup, and Integral Fuel Burnable Absorber (IFBA) of each fuel assembly stored in Region 1 or Region 2, shall be within the acceptable limits described in the surveillance requirements below.

APPLICABILITY: When any fuel assembly is stored in Region 1 or Region 2 of the spent fuel storage pool.

ACTION:

If the requirements of the LCO are not met:

a. Immediately verify the fuel storage boron concentration meets the requirements of TS 3.7.11 and

b. Immediately initiate action to move the non-complying fuel assembly to a location that complies with the surveillance requirements.

c. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.12.1 Prior to storing fuel assemblies in Region 1, verify by administrative means that the fuel assemblies meet one of the following storage constraints:

a. Unirradiated fuel assemblies with a maximum enrichment of 4.25 wt% U-235 have unrestricted storage.

b. Unirradiated fuel assemblies with enrichments greater than 4.25 wt% U-235 and less than or equal to 5.0 wt% U-235, that do not contain IFBA pins, may only be stored in the peripheral cells facing the concrete wall.

c. Unirradiated fuel assemblies with enrichments (E) greater than 4.25 wt% U-235 and less than or equal to 5.0 wt% U-235, which contain a minimum number of IFBA pins have unrestricted storage. This minimum number of IFBA pins shall have an equivalent reactivity hold-down which is greater than or equal to the reactivity hold-down associated with N IFBA pins, at a nominal 2.35 mg B-10/linear inch loading (l.5x), determined by the equation below:

N 42.67 (E - 4.25)

Salem Unit 1 3/4 7-36 Amendment No.276

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (continued)

d. Irradiated fuel assemblies with enrichments (E) greater than 4.25 wt%

U-235 and less than or equal to 5.0 wt%, that have attained the minimum burnup (BU) as determined by the equation below, have unrestricted storage.

BU (MWD/kg U) = -26.212 + 6.1677E 4.7.12.2 Prior to storing fuel assemblies in Region 2, verify by administrative means that the fuel assemblies meet one of the following storage constraints:

a. Unirradiated fuel assemblies with a maximum enrichment of 5.0 wt% U-235 may be stored in a checkerboard pattern with intermediate cells containing only water or non-fissile bearing material.

b. Unirradiated fuel assemblies with a maximum enrichment (E) of 5.0 wt%

U-235 may be stored in the central cell of any 3x3 array of cells provided the surrounding eight cells are empty or contain fuel assemblies that have attained the minimum burnup (BU) as determined by the equation below.

BU (MWD/kg U) = -15.48 + 17.80E - 0.7038E2 In this configuration, none of the nine cells in any 3x3 array shall be common to cells in any other similar 3x3 array. Along the rack periphery, the concrete wall is equivalent to 3 outer cells in a 3x3 array.

c. Irradiated fuel assemblies with a maximum enrichment (E) of 5.0 wt%

U-235 that have attained the minimum burnup (BU) as determined by the equation below, have unrestricted storage.

BU (MWD/kg U) = -32.06 + 25.21E - 3.723E2 + 0.3535E3

d. Irradiated fuel assemblies with a maximum enrichment (E) of 5.0 wt%

U-235 that have attained the minimum burnup (BU) as determined by the equation below, may be stored in a peripheral cell facing the concrete wall.

BU (MWD/kg U) = -25.56 + 15.14E - 0.602E2 Salem - Unit 1 3/4 7-37 Amendment No. 262

3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES OPERATING LIMITING CONDITION FOR OPERATION 3.8.1.1 As a minimum, the following A.C. electrical power sources shall be OPERABLE:

a. Two physically independent A.C. circuits between the offsite transmission network and the onsite Class 1E distribution system (vital bus system), and

b. Three separate and independent diesel generators with:

1. Separate day tanks containing a minimum volume of 130 gallons of fuel, and

2. A common fuel storage system consisting of two storage tanks, each containing a minimum volume of 23,000 gallons of fuel, and two fuel transfer pumps.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

a. With an independent A.C. circuit of the above required A.C.

electrical power sources inoperable:

1. Demonstrate the OPERABILITY of the remaining independent A.C. circuit by performing Surveillance Requirement 4.8.1.1.1.a within one hour and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter; and

2. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , declare required systems or components with no offsite power available inoperable when a redundant required system or component is inoperable, 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 ; and

3. Restore the inoperable independent A.C. circuit to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 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 one diesel generator of the above required A.C. electrical power sources inoperable:

1. Demonstrate the OPERABILITY of the independent A.C. circuits by performing Surveillance Requirement 4.8.1.1.1.a within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter; and

2. Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , declare required systems or components supported by the inoperable diesel generator inoperable when a required redundant system or component is inoperable, 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 ; and SALEM - UNIT 1 3/4 8-1 Amendment No. 253 OCT 11 2002

ELECTRICAL POWER SYSTEMS ACTION (Continued)

3. Determine the two remaining OPERABLE diesel generators are not inoperable due to common cause failure or perform Surveillance Requirement 4.8.1.1.2.a.2 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . If the diesel generator is inoperable for preventive maintenance, the two remaining OPERABLE diesel generators need not be tested nor the OPERABILITY evaluated; and

4. In any case, restore the inoperable diesel generator to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c. With one independent A.C. circuit and one diesel generator of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining independent A.C.

circuit by performing Surveillance Requirement 4.8.1.1.1.a within one hour and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter; demonstrate the OPERABILITY of the remaining OPERABLE diesel generators by performing Surveillance Requirement 4.8.1.1.2.a.2 within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ; restore at least one of the inoperable sources to OPERABLE status 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 SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . Restore at least two independent A.C. circuits and three diesel generators to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months from the time of initial loss 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 .

d. With two of the above required independent A.C. circuits inoperable:

1. Demonstrate the OPERABILITY of three diesel generators by performing Surveillance Requirement 4.8.1.1.2.a.2 within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , unless the diesel generators are already operating; and

2. Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , declare required systems or components supported by the inoperable offsite circuits inoperable when a required redundant system or component is inoperable, 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 ; and

3. Restore at least one of the inoperable independent A.C.

circuits to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; and

4. With only one of the independent A.C. circuits OPERABLE, restore the other independent A.C. circuit to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months from time of initial loss 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 .

SALEM - UNIT 1 3/4 8-2 Amendment No. 253 OCT 1 1 2OU2

ELECTRICAL POWER SYSTEMS ACTION (Continued)

e. With two or more of the above required diesel generators inoperable, demonstrate the OPERABILITY of two independent A.C. circuits by performing Surveillance Requirement 4.8.1.1.1.a within one hour and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter; restore at least two of the inoperable diesel generators to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . Restore three diesel generators to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months from time of initial loss or be in 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 .

f. With one of the above required fuel transfer pumps inoperable, either restore it to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

g. With one of the above required fuel storage tanks inoperable, either restore it to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

h. LCO 3.0.4.b is not. applicable to DGs.

SALEM - UNIT 1 3/4 8-2a Amendment No. 276

THIS PAGE INTENTIONALLY LEFT BLANK SALEM - UNIT 1 3/4 8-2b Amendment No. 253 OCT 11 2M2

ELECTRICAL POWER SYSTEMS 4.8.1.1.1 Two physically independent AC. circuits between the offsite transmission network and the onsite Class 1E distribution system (vital bus system) shall be:

a. Determined OPERABLE in accordance with the Surveillance Frequency Control Program by verifying correct breaker alignments, power availability, and

b. Demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program during shutdown by transferring (manually and automatically) vital bus supply from one 13/4 kv transformer to the other 13/4 kv transformer.

4.8.1.1.2 Each diesel generator shall be demonstrated OPERABLE:

a. In accordance with the Surveillance Frequency Control Program by:

1. Verifying the fuel level in its day tank.

2. Verifying the diesel generator starts from standby conditions* and achieves ~ 3910 volts and ~ 58.8 Hz in s 13 seconds, and subsequently achieves steady state voltage of ~ 3910 and :S 4400 volts and frequency of 60 +/- 1.2 Hz.

Subsequently, verifying the generator is synchronized with voltage maintained ~ 3910 and s 4580 volts, gradually loaded to 2340-2600 kw**,

and operates at a load of 2340-2600 kw for greater than or equal to 60 minutes.

3. Verifying the diesel generator is aligned to provide standby power to the associated vital bus.

b. In accordance with the Surveillance Frequency Control Program and after each operation of the diesel where the period of operation was greater than or equal to one hour by checking for and removing accumulated water from the day tanks.

c. In accordance with the Surveillance Frequency Control Program by verifying the diesel generator starts from standby conditions* and achieves;::: 3910 volts and

~ 58.8 Hz in S 13 seconds, and subsequently achieves steady state voltage of

~ 3910 and S 4400 volts and frequency of 60 +/- 1.2 Hz.

The generator shall be synchronized to its emergency bus with voltage maintained ~ 3910 and S 4580 volts, loaded to 2340-2600** kw in less than or equal to 60 seconds, and operate at a load of 2340-2600 kw for at least 60 minutes.

This test, if it is performed so it coincides with the testing required by Surveillance Requirement 4.8.1.1.2.a.2, may also serve to concurrently meet those requirements.

SALEM - UNIT 1 3/4 8-3 Amendment No. 299

ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

d. In accordance with the Surveillance Frequency Control Program during shutdown by:

1. DELETED

2. Verifying that, on rejection of a load greater than or equal to 820 kw, the voltage and frequency are restored to 2: 3910 and :s; 4400 volts and 60 +/- 1.2 Hz within 4 seconds, and subsequently achieves a steady state frequency of 2: 58.8 and :s; 60.5 Hz.

3. Simulating a loss of offsite power by itself, and:

a) Verifying de-energization of the vital bus and load shedding from the vital bus.

b) Verifying the diesel starts on the auto-start signal*, energizes the vital bus with permanently connected loads within 13 seconds, energizes the auto-connected shutdown loads through the load sequencer and operates for greater than or equal to 5 minutes while its generator is loaded with the shutdown loads. After energization of these loads, the steady state voltage and frequency of the vital bus shall be maintained at 2: 3910 and

s; 4400 volts and 2: 58.8 and :s; 60.5 Hz during this test.

4. Verifying that on an ESF actuation test signal without loss of offsite power the diesel generator starts on the auto-start signal and operates on standby for greater than or equal to 5 minutes*. The diesel generator shall achieve 2: 3910 volts and 2: 58.8 Hz in :s; 13 seconds, and subsequently achieves steady state voltage of 2: 3910 and :s; 4400 volts and frequency of 2: 58.8 and :s; 60.5 Hz.

5. Not Used.

6. Simulating a loss of offsite power in conjunction with an ESF actuation test signal, and:

a) Verifying de-energization of the vital bus and load shedding from the vital bus.

b) Verifying the diesel starts on the auto-start signal*, energizes the vital bus with permanently connected loads within 13 seconds, energizes the auto-connected emergency (accident) loads through the load sequencer and operates for greater than or equal to 5 minutes while its generator is loaded with the emergency loads.

After energization of these loads, the steady state voltage and frequency of the vital bus shall be maintained at 2: 3910 and

s; 4400 volts and 2: 58.8 and :s; 60.5 Hz during this test.

SALEM - UNIT 1 3/4 8-4 Amendment No. 299

ELECTRICAL POWER SYSTEMS c) Verifying that all nonessential automatic diesel generator trips (Le., other than engine overspeed, lube oil pressure low, 4 KV bus differential and generator differential), are automatically bypassed upon loss of voltage on the vital bus concurrent with a safety injection actuation signal.

7. Deleted

8. Verifying that the auto-connected loads to each diesel generator do not exceed the two hour rating of 2860 kw.

9. Verifying that with the diesel generator operating in a test mode (connected to its bus), a simulated safety injection signal overrides the test mode by (1) returning the diesel generator to standby operation and (2) automatically energizing the emergency loads with offsite power.

e. In accordance with the Surveillance Frequency Control Program or after any modifications which could affect diesel generator interdependence by starting all diesel generators simultaneously*, during shutdown, and verifying that all diesel generators accelerate to at least 58.8 Hz in less than or equal to 13 seconds.

f. In accordance with the Surveillance Frequency Control Program, the following test shall be performed within 5 minutes of diesel shutdown after the diesel has operated for at least two hours at 2340-2600 kw**:

Verifying the diesel generator starts and achieves ~ 3910 volts and ~ 58.8 Hz in s 13 seconds, and subsequently achieves steady state voltage of ~ 3910 and s 4400 volts and frequency of 60 +/- 1.2 Hz.

g. In accordance with the Surveillance Frequency Control Program verifying the diesel generator operates for at least 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months s*. During the first 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of this test, the diesel generators shall be loaded to 2760-2860 Kw**. During the remaining 22 hours2.546296e-4 days 0.00611 hours 3.637566e-5 weeks 8.371e-6 months of this test, the diesel generator shall be loaded to 2500-2600 Kw**. The steady state voltage and frequency shall be maintained at ~ 3910 and s 4580 volts and 60 +/- 1.2 Hz during this test.

4.8.1.1.3 The diesel fuel oil storage and transfer system shall be demonstrated OPERABLE:

a. In accordance with the Surveillance Frequency Control Program by:

1. Verifying the level in each of the above required fuel storage tanks.

2. Verifying that both fuel transfer pumps can be started and transfer fuel from the fuel storage tanks to the day tanks.

SALEM - UNIT 1 3/48-5 Amendment No. 299

ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

b. In accordance with the Surveillance Frequency Control Program by verifying that a sample of diesel fuel from each of the above required fuel storage tanks is within the acceptable limits specified in Table 1 of ASTM 0975-77 when checked for viscosity, water and sediment.

4.8.1.1.4 Reports - NOT USED

Surveillance testing may be conducted in accordance with the manufacturer's recommendations regarding engine prelube, warm-up and loading (unless loading times are specified in the individual Surveillance Requirements).

- This band is meant as guidance to preclude routine exceedances of the diesel generator manufacturer's design ratings. Loads in excess of this band for special testing or momentary variations due to changing bus loads shall not invalidate the test.

SALEM - UNIT 1 3/48-5a Amendment No. 299

ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

TABLE 4.8-1 DIESEL GENERATOR TEST SCHEDULE NOT USED SALEM - UNIT 1 3/4 8-5b Amendment No. Iff,203 JAN 8 1998

I ELECTRICAL POWER SYSTEMS SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.1.2 As a minimum, the following A.C. electrical power sources shall be OPERABLE:

a. One circuit between the offsite transmission network and the onsite Class lE distribution system (vital bus system), and

b. Two separate and independent diesel generators with:

1. Separate day tanks containing a minimum volume of 130 gallons of fuel, and

2. A common fuel storage system containing a minimum volume of 23,000 gallons of fuel, and

3. A fuel transfer pump.

APPLICABILITY: MODES 5 and 6.

During movement of irradiated fuel assemblies.

ACTION:

a. With one of the above minimum required A.C. electrical power sources not OPERABLE, immediately declare the affected required features inoperable, or suspend all operations involving CORE ALTERATIONS, movement of-irradiated fuel, and positive reactivity changes until the minimum required A.C. electrical power sources are restored to OPERABLE status.

b. With two of the required diesel generators not OPERABLE, suspend all operations involving CORE ALTERATIONS, movement of irradiated fuel, and all operations involving positive reactivity additions, and immediately initiate action to restore one required DG to OPERABLE status.

SURVEILLANCE REQUIREMENTS

NOTE----------------------------------

The following surveillances are not required to be performed to maintain operability during Modes 5 and 6. These surveillances are: 4.8.1.1.1.b, 4.8.1.1.2.d.2, 4.8.1.1.2.d.3, 4.8.1.1.2.d.4, 4.8.1.1.2.d.6, 4.8.1.1.2.d.9, 4.8.1.1.2.e, 4.8.1.1.2.f, and 4.8.1.1.2.g.

4.8.1.2 The above required A.C. electrical power sources shall be demonstrated OPERABLE by the performance of each of the Surveillance Requirements of 4.8.1.1.1, 4.8.1.1.2, 4.8.1.1.3 (except for requirement 4.8.1.1.3.a.2) and 4.8.1.1.4.

SALEM - UNIT 1 3/4 8-5c Amendment No. 264

ELECTRICAL POWER SYSTEMS 3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEMS AC. DISTRIBUTION - OPERATING 3.8.2.1 The following AC. electrical busses shall be OPERABLE and energized from sources of power other than the diesel generators:

4 kvolt Vital Bus # 1A 4 kvolt Vital Bus # 1B 4 kvolt Vital Bus # 1C 460 volt Vital Bus # 1A and associated control centers 460 volt Vital Bus # 1B and associated control centers 460 volt Vital Bus # 1C and associated control centers 230 volt Vital Bus # 1A and associated control centers 230 volt Vital Bus # 1B and associated control centers 230 volt Vital Bus # 1C and associated control centers 115 volt Vital Instrument Bus # 1A and Inverter

115 volt Vital Instrument Bus # 1B and Inverter

115 volt Vital Instrument Bus # 1C and Inverter

115 volt Vital Instrument Bus # 1D and Inverter

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

a. With less than the above complement of AC. busses OPERABLE or energized, restore the inoperable bus to OPERABLE and energized status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With one inverter inoperable, energize the associated AC. Vital Bus within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ; restore the inoperable 1A, 1B, or 1C inverter to OPERABLE and energized status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ; restore the inoperable 1D inverter to OPERABLE and energized status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

4.8.2.1 The specified AC. busses shall be determined OPERABLE and energized from AC.

sources other than the diesel generators in accordance with the Surveillance Frequency Control Program by verifying correct breaker alignment and indicated power availability.

(*) An inverter may be disconnected from its DC source for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for the purpose of performing an equalizing charge on its associated battery bank provided (1) its vital bus is OPERABLE and energized, and (2) the vital busses associated with the other battery banks are OPERABLE and energized.

SALEM - UNIT 1 3/4 8-6 Amendment No. 299

ELECTRICAL POWER SYSTEMS AC. DISTRIBUTION - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.2.2 As a minimum, two AC. electrical bus trains shall be OPERABLE and energized from sources of power other than a diesel generator but aligned to an OPERABLE diesel generator with each train consisting of:

1 4 kvolt Vital Bus 1 460 volt Vital Bus and associated control centers 1 230 volt Vital Bus and associated control centers 1 115 volt Instrument Bus energized from its respective inverter connected to its respective D.C. bus train.

APPLICABILITY: MODES 5 and 6.

During movement of irradiated fuel assemblies.

ACTION:

With less than the above complement of AC. busses and inverters OPERABLE and energized, immediately declare the affected required features inoperable, or suspend all operations involving CORE ALTERATIONS, positive reactivity changes, and movement of irradiated fuel assemblies until the minimum required AC. electrical power sources are restored to OPERABLE status.

4.8.2.2 The specified AC. busses and inverters shall be determined OPERABLE and energized from AC. sources other than the diesel generators in accordance with the Surveillance Frequency Control Program by verifying correct breaker alignment and indicated voltage on the busses.

SALEM - UNIT 1 3/4 8-7 Amendment No. 299

ELECTRICAL POWER SYSTEMS 125-VOLT D.C. DISTRIBUTION - OPERATING LIMITING CONDITION FOR OPERATION 3.8.2.3 The following D.C. bus trains shall be OPERABLE and energized:

TRAIN IA consisting of 125-volt D.C. bus No. 1A, 125-volt D.C.

battery No. IA and battery charger lAl.

TRAIN 1B consisting of 125-volt D.C. bus No. 1B, 125-volt D.C.

battery No. 1B and battery charger lB1.

TRAIN IC consisting of 125-volt D.C. bus No. IC, 125-volt D.C.

battery No. iC and battery charger 1C1.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

a. With one 125-volt D.C. bus inoperable or not energized, restore the inoperable bus to OPERABLE and energized status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With one 125-volt D.C. battery charger inoperable, restore the inoperable charger to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or connect the backup charger for no more than 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c. With one or more 125-volt D.C. batteries with one or more battery cell parameters not within the Category A or B limits of Table 4.8.2.3-1:

1. Verify within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , that the electrolyte level and float voltage for the pilot cell meets Table 4.8.2.3-1 Category C limits, and

2. Verify within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , that the battery cell parameters of all connected cells meet Table 4.8.2.3-1 Category C limits, and

3. Restore battery cell parameters to Category A and B limits of Table 4.8.2.3-1 within 31 days, and

4. If any of the above listed requirements cannot be met, comply with the requirements of action f.

d. With one or more 125-volt D.C. batteries with one or more battery cell parameters not within Table 4.8.2.3-1 Category C values, comply with the requirements of action f.

e. With average electrolyte temperature of representative cells less than 650 F, comply with the requirements of action f.

f. Restore the battery to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SALEM - UNIT 1 3/4 8-8 Amendment No. 225 NiOV 0 2 ¶139

ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS 4.8.2.3.1 Each D.C. bus train shall be determined OPERABLE and energized in accordance with the Surveillance Frequency Control Program by verifying correct breaker alignment and power availability.

4.8.2.3.2 Each 125-volt battery and above required charger shall be demonstrated OPERABLE:

a. In accordance with the Surveillance Frequency Control Program by verifying that:

1. The parameters in Table 4.8.2.3-1 meet Category A limits.

2. The overall battery voltage is greater than or equal to 125 volts on float charge.

b. In accordance with the Surveillance Frequency Control Program and once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after a battery discharge < 110 V and once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after a battery overcharge> 150 V by verifying that the parameters in Table 4.8.2.3-1 meet the Category B limits.

c. In accordance with the Surveillance Frequency Control Program by verifying that:

1. There is no visible corrosion at terminals or connectors or the connection resistance is:

s; 150 micro ohms for inter-cell connections, s; 350 micro ohms for inter-rack connections, s; 350 micro ohms for inter-tier connections, s; 70 micro ohms for field cable terminal connections, and s; 2500 micro ohms for the total battery connection reSistance which includes all inter-cell connections (including bus bars), all inter-rack connections (including cable resistance) all inter-tier connections (including cable resistance) and all field terminal connections at the battery.

2. The average electrolyte temperature of the representative cells is above 65°F.

d. In accordance with the Surveillance Frequency Control Program by verifying that:

1. The cells, cell plates and battery racks show no visual indication of physical damage or abnormal deterioration.

2. Remove visible terminal corrosion and verify cell-to-cell and terminal connections are coated with anti-corrosion material.

SALEM - UNIT 1 3/4 8-9 Amendment No. 299

ELECTRICAL POWER SYSTEMS

3. The connection resistance is:

s 150 micro ohms for inter-cell connections, s 350 micro ohms for inter-rack connections, s 350 micro ohms for inter-tier connections, s 70 micro ohms for field cable terminal connections, and s 2500 micro ohms for the total battery connection resistance which includes all inter-cell connections (including bus bars), all inter-rack connections (including cable resistance), all inter-tier connections (including cable resistance), and all field terminal connections at the battery.

e. In accordance with the Surveillance Frequency Control Program by verifying that the battery charger will supply at least 170 amperes at 125 volts for at least 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

f. In accordance with the Surveillance Frequency Control Program, during shutdown, by verifying that the battery capacity is adequate to supply and maintain in OPERABLE status all of the actual or simulated emergency loads for the design duty cycie when the battery is subjected to a battery service test.

g. In accordance with the Surveillance Frequency Control Program, during shutdown, by verifying that the battery capacity is at least 80% of the manufacturer's rating when subjected to a performance discharge test.

Satisfactory completion of this performance discharge test shall also satisfy the requirements of Specification 4.8.2.3.2.f if the performance discharge test is conducted during a shutdown where that test and the battery service test would both be required.

h. At least once per 12 months, during shutdown, if the battery shows signs of degradation OR has reached 85% of the service life with a capacity less than 100% of manufacturers rating, by verifying that the battery capacity is at least 80% of the manufacturer's rating when subjected to a performance discharge test. Degradation is indicated when the battery capacity drops more than 10% of rated capacity from its capacity on the previous performance test, or is below 90% of the manufacturer's rating.

i. At least once per 24 months, during shutdown, if the battery has reached 85% of the service life with capacity greater than or equal to 100% of manufacturers rating, by verifying that the battery capacity is at least 80% of the manufacturer's rating when subjected to a performance discharge test.

SALEM - UNIT 1 3/48-9a Amendment No. 299

ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

Table 4.8.2.3-1 Battery Cell Parameters Requirements CATEGORY A: CATEGORY C:

LIMITS FOR EACH CATEGORY B: ALLOWABLE VALUE DESIGNATED PILOT LIMITS FOR EACH FOR EACH PARAMETER CELL CONNECTED CELL CONNECTED CELL mmmmmum mmmmmmm m mmmm minm mm mmmmmmmmmms m inmmminmmmm minmm~mmmmmmminmmmm mm mmmm Electrolyte > Minimum level > Minimum level Above top of level indication mark, indication mark, plates, and not and : .1/4 inch and 5 1/4 inch overflowing above maximum above maximum level indication level indication mark (a) mark (a)

Float Voltage 2 2.13 V 2 2.13 V 2 2.07 V Specific 2 1.195 2 1.190 Not more than 0.020 Gravity AND below average of all (b) (c) Average of all connected cells connected cells AND 2 1.200 Average of all connected cells 2 1.190 (a) It is acceptable for the electrolyte level to temporarily increase above the specified maximum level during equalizing charge provided it is not overflowing.

(b) Corrected for electrolyte temperature and level. Level correction is not required, however, when battery charging is < 3 amps when on float charge.

(c) Or battery charging current is < 3 amps when on float charge. This is acceptable only during a maximum of 7 days following a battery recharge.

SALEM - UNIT 1 3/4 8-9b Amendment No. 177 SEP 1 9 1995

ELECTRICAL POWER SYSTEMS 125-VOLT D.C. DISTRIBUTION - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.2.4 As a minimum, the following D.C. electrical equipment and bus shall be energized and OPERABLE:

2 125-volt D.C. busses, and 2 125-volt batteries, each with at least one full capacity charger, associated with each of the above D.C. busses.

APPLICABILITY: MODES 5 and 6.

During movement of irradiated fuel assemblies.

ACTION:

With less than the above complement of D.C. equipment and busses OPERABLE, immediately declare the affected required features inoperable, or suspend all operations involving CORE ALTERATIONS, positive reactivity changes, and movement of irradiated fuel assemblies until the minimum required 125 Volt D.C. electrical power sources are restored to OPERABLE status.

SURVEILLANCE REQUIREMENTS 4.8.2.4.1 The above required 125-volt D.C. busses shall be determined OPERABLE and energized in accordance with the Surveillance Frequency Control Program by verifying correct breaker alignment and indicated power availability.

4.8.2.4.2 The above required 125-volt batteries and chargers shall be demonstrated OPERABLE per Surveillance Requirement 4.8.2.3.2.

SALEM - UNIT 1 3/48-10 Amendment No. 299

ELECTRICAL POWER SYSTEMS 28-VOLT D.C. DISTRIBUTION - OPERATING LIMITING CONDITION FOR OPERATION 3.8.2.5 The following D.C. bus trains shall be energized and OPERABLE:

TRAIN 1A consisting of 28-volt D.C. bus No. 1A, 28-volt D.C.

battery No. 1A and battery charger lAl.

TRAIN 1B consisting of 28-volt D.C. bus No. 1B, 28-volt D.C.

battery No. 1B and battery charger lB.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

a. With one 28-volt D.C. bus inoperable or not energized, restore the inoperable bus to OPERABLE and energized status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With one required 28-volt D.C. battery charger inoperable, restore the inoperable battery charger to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or connect the backup charger for no more than 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c. With one or more 28-volt D.C. batteries with one or more battery cell parameters not within the Category A or B limits of Table 4.8.2.5-1:

1. Verify within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , that the electrolyte level and float voltage for the pilot cell meets Table 4.8.2.5-1 Category C limits, and

2. Verify within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , that the battery cell parameters of all connected cells meet Table 4.8.2.5-1 Category C limits, and

3. Restore battery cell parameters to Category A and B limits of Table 4.8.2.5-1 within 31 days, and

4. If any of the above listed requirements cannot be met, comply with the requirements of action f.

d. With one or more 28-volt D.C. batteries with one or more battery cell parameters not within Table 4.8.2.5-1 Category C values, comply with the requirements of action f.

e. With average electrolyte temperature of representative cells less than 65'F, comply with the requirements of action f.

f. Restore the battery to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SALEM - UNIT 1 3/4 8-11 Amendment No249 JAN 0o 4XZ

ELECTRICAL POWER SYSTEMS 4.8.2.5.1 Each D.C. bus train shall be determined OPERABLE and energized in accordance with the Surveillance Frequency Control Program by verifying correct breaker alignment and power availability.

4.8.2.5.2 Each 28-volt battery and above required charger shall be demonstrated OPERABLE:

a. In accordance with the Surveillance Frequency Control Program by verifying that:

1. The parameters in Table 4.8.2.5-1 meet Category A limits.

2. The overall battery voltage is greater than or equal to 27 volts on float charge.

b. In accordance with the Surveillance Frequency Control Program and once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after a battery discharge < 25.7 V and once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after a battery overcharge> 35 V by verifying that the parameters in Table 4.8.2.5-1 meet the Category B limits.

c. In accordance with the Surveillance Frequency Control Program by verifying that:

1. There is no visible corrosion at terminals or connectors or the connection resistance is:

s; 50 micro ohms for inter-cell connections, s; 200 micro ohms for inter-tier connections, s; 70 micro ohms for field cable terminal connections, and s; 500 micro ohms for the total battery connection resistance which includes all inter-cell connections (including bus bars), all inter-tier connections (including cable resistance) and all field terminal connections at the battery.

2. The average electrolyte temperature of the representative cells is ;;:= 65°F.

d. In accordance with the Surveillance Frequency Control Program by verifying that:

1. The cells, cell plates and battery racks show no visual indication of physical damage or abnormal deterioration.

2. Remove visible terminal corrosion and verify cell-to-cell and terminal connections are coated with anti-corrosion material.

SALEM - UNIT 1 3/48-12 Amendment No. 299

ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

3. The connection resistance is:

s 50 micro ohms for inter-cell connections, s 200 micro ohms for inter-tier connections, s 70 micro ohms for field cable terminal connections, and s 500 micro ohms for the total battery connection resistance which includes all inter-cell connections (including bus bars), all inter-tier connections (including cable resistance) and all field terminal connections at the battery.

e. In accordance with the Surveillance Frequency Control Program by verifying that the battery charger will supply;::: 150 amperes at ;::: 28 volts for;::: 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

f. In accordance with the Surveillance Frequency Control Program, during shutdown, by verifying that the battery capacity is adequate to supply and maintain in OPERABLE status all of the actual or simulated emergency loads for the design duty cycle when the battery is subjected to a battery service test.

g. In accordance with the Surveillance Frequency Control Program, during shutdown, by verifying that the battery capacity is at least 80% of the manufacturer's rating when subjected to a performance discharge test.

Satisfactory completion of this performance discharge test shall also satisfy the requirements of Specification 4.8.2.5.2.f if the performance discharge test is conducted during a shutdown where that test and the battery service test would both be required.

h. At least once per 12 months, during shutdown, if the battery shows signs of degradation OR has reached 85% of the service life with a capacity less than 100% of manufacturers rating, by verifying that the battery capacity is at least 80% of the manufacturer's rating when subjected to a performance discharge test. Degradation is indicated when the battery capacity drops more than 10% of rated capacity from its capacity on the previous performance test, or is below 90% of the manufacturer's rating.

i. At least once per 24 months, during shutdown, if the battery has reached 85% of the service life with capacity greater than or equal to 100% of manufacturers rating, by verifying that the battery capacity is at least 80% of the manufacturer's rating when subjected to a performance discharge test.

SALEM - UNIT 1 3/48-12a Amendment No. 299

ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

TABLE 4.8.2.5-1 BATTERY CELL PARAMETER REQUIREMENTS CATEGORY A: CATEGORY B: CATEGORY C:

PARAMETER LIMITS FOR EACH LIMITS FOR EACH ALLOWABLE VALUE DESIGNATED PILOT CELL CONNECTED CELL FOR EACH CONNECTED CELL Electrolyte >Minimum level >Minimum level Above top of Level indication mark and indication mark plates and not 5 1/4 inch above and

1/4 inch overflowing maximum level above maximum indication markia) level indication mark (a)

Float Voltage 22.13 V 22.13 V Ž2.07 V Specific 21.195 21.190 Not more than Gravity(b) (c) 0.020 below the AND average of all connected cells Average of all Connected cells AND

21. 200 Average of all connected cells

Ž1.190 (a) It is acceptable for the electrolyte level to temporarily increase above the specified maximum level during equalizing charge provided it is not overflowing.

(b) Corrected for electrolyte temperature and level. Level correction is not required, however, when battery charging is < 2 amps when on float charge.

(c) Or battery charging current is < 2 amps when on float charge. This is acceptable only during a maximum of 7 days following a battery recharge.

SALEM - UNIT 1 3/4 8-12b Amendment No249 JAN 0 4 2t-32

ELECTRICAL POWER SYSTEIVIS 28*VOL T D.C. DISTRIBUTION* SHUTDOWN 3.8.2.6 As a minimum, the following D.C. electrical equipment and bus shall be energized and OPERABLE:

1 28-volt D.C. bus, and 1 28-volt battery and at least one full capacity charger associated with the above D.C. bus.

APPLICABILITY: MODES 5 and 6.

During movement of irradiated fuel assemblies.

ACTION:

With less than the above complement of D.C. equipment and busses OPERABLE, immediately declare the affected required features inoperable, or suspend all operations involving CORE ALTERATIONS, positive reactivity changes, and movement or irradiated fuel assemblies until the minimum required 28 Volt D.C. electrical power sources are restored to OPERABLE status.

4.8.2.6.1 The above required 28-volt D.C. bus shall be determined OPERABLE and energized in accordance with the Surveillance Frequency Control Program by verifying correct breaker alignment and voltage on the bus.

4.8.2.6.2 The above required 28-volt batteries and charger shall be demonstrated OPERABLE per Surveillance Requirement 4.8.2.5.2.

SALEM - UNIT 1 3/48-13 Amendment No. 299

ELECTRICAL POWER SYSTEMS 3/48.3 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES LIMITING CONDITION FOR OPERATION 3.8.3.1 All containment penetration conductor overcurrent protective devices required to provide thermal protection of penetrations shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With one or more of the required containment penetration conductor overcurrent protective device(s) inoperable:

a. Restore the protective device(s) to OPERABLE status or de-energize the circuit(s) by tripping either the primary or backup protective device, or racking out or removing the primary or backup device within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , declare the affected system or component inoperable, and verify the primary or backup protective device to be tripped, or the primary or backup device racked out or removed at least once per 7 days thereafter; or

b. Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.8.3.1 All required containment penetration conductor overcurrent protective devices shall be demonstrated OPERABLE:

a. In accordance with the Surveillance Frequency Control Program:

1. For 4.16 KV reactor coolant pump circuits, by performance of:

(a) A CHANNEL CALIBRATION of the associated protective relays, and (b) An integrated system functional test which includes simulated automatic actuation of the system and verifying that each relay and associated circuit breakers and control circuits function as designed.

SALEM - UNIT 1 3/48-14 Amendment No. 299

ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

2. By verifying the OPERABILITY of the required molded case and lower voltage circuit breakers, by selecting and functionally testing a representative sample of at least 10% of all the circuit breakers of that type. Circuit breakers selected for functional testing shall be selected on a rotating basis. The functional test shall consist of injecting a current input at the specified setpoint to each selected circuit breaker and verifying that each circuit breaker functions as designed. Circuit breakers found inoperable during functional testing shall be restored to OPERABLE status prior to resuming operation. For each circuit breaker found inoperable during the functional tests, an additional representative sample of at least 10% of all the circuit breakers of the inoperable type shall also be functionally tested until no more failures are found or all circuit breakers of that type have been functionally tested.

b. In accordance with the Surveillance Frequency Control Program by subjecting each circuit breaker to an inspection and preventive maintenance in accordance with procedures prepared in conjunction with its manufacturer's recommendations.

SALEM - UNIT 1 3/48-15 Amendment No. 299

3/4.9 REFUELING OPERATIONS BORON CONCENTRATION LIMITING CONDITION FOR OPERATION 3.9.1 The boron concentration of the Reactor Coolant System, the fuel storage pool, the refueling canal, and the refueling cavity shall be maintained within the limit specified in the CORE OPERATING LIMITS REPORT (COLR).

APPLICABILITY: MODE 6 (Only applicable to the refueling canal, the fuel storage pool and refueling cavity when connected to the Reactor Coolant System)

ACTION:

With the requirements of the above specification not satisfied, immediately

a. Suspend CORE ALTERATIONS and

b. Suspend positive reactivity additions and

c. Initiate action to restore boron concentration to within limit specified in the COLR.

d. The provisions of Specification 3.0.3 are not applicable.

4.9.1. Verify the boron concentration is within the limit of the COLR in accordance with the Surveillance Frequency Control Program.

SALEM - UNIT 1 3/49-1 Amendment No. 299

REFUELING OPERATIONS INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.9.2 As a minimum, two source range neutron flux monitors shall be operating, each with continuous visual indication in the control room and one with audible indication in the containment and control room.

APPLICABILITY: MODE 6.

ACTION:

With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.2 Each source range neutron flux monitor shall be demonstrated OPERABLE by performance of:

a. A CHANNEL FUNCTIONAL TEST in accordance with the Surveillance Frequency Control Program, and

b. A CHANNEL CHECK in accordance with the Surveillance Frequency Control Program during CORE ALTERATIONS.

SALEM - UNIT 1 3/4 9-2 Amendment No. 299

REFUELING OPERATIONS DECAY TIME LIMITING. CONDITION FOR OPERATION 3.9.3 The reactor shall be subcritical for at least:.

a. 80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months

b. 168 hours0.00194 days 0.0467 hours 2.777778e-4 weeks 6.3924e-5 months APPLICABILITY: Specification 3.9.3.a - From October 1 5th through May 1 5th, during movement of irradiated fuel in the reactor pressure vessel.

Specification 3.9.3.b - From May 1 6 th through October 1 4 th, during movement of irradiated fuel in the reactor pressure vessel.

ACTION:

With the reactor subcritical for less than the required time, suspend all operations involving movement of irradiated fuel in the reactor pressure vessel. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.3 The reactor shall be determined to have been subcritical as required by verification of the date and time of subcriticality prior to movement of irradiated fuel in the reactor pressure vessel.

SALEM - UNIT 1 34 9 - 3 DuriendmentN . 8

REFUELING OPERATIONS CONTAINMENT BUILDING PENETRATIONS LIMITING CONDITION FOR OPERATION 3.9.4 The containment building penetrations shall be in the following status:

a. The equipment hatch inside door is capable of being closed and held in place by a minimum of four bolts, or an equivalent closure device installed and capable of being closed,

b. A minimum of one door in each airlock is capable of being closed

c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere shall be either:

1. closed by a manual or automatic isolation valve, blind flange, or equivalent, or

2. capable of being closed by the Containment Purge and Pressure-Vacuum Relief Isolation System.

Note: Penetration flow path(s) providing direct access from the containment atmosphere to the outside atmosphere may be unisolated under administrative controls.

APPLICABILITY: During movement of irradiated fuel within the containment.

ACTION:

With the requirements of the above specification not satisfied, immediately suspend all operations involving movement of irradiated fuel in the containment building. The provisions of Specification 3.0.3 are not applicable.

4.9.4.1 Each of the above required containment building penetrations shall be determined to be either in its required condition or capable of being closed by a manual or automatic containment isolation valve in accordance with the Surveillance Frequency Control Program.

4.9.4.2 Once per refueling prior to the start of movement of irradiated fuel assemblies within the containment building, verify the capability to close, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , the equipment hatch inside door or an equivalent closure device. Applicable only when the equipment hatch is open during movement of irradiated fuel in the containment building.

4.9.4.3 Verify, in accordance with the Surveillance Frequency Control Program, each required containment purge isolation valve actuates to the isolation position on a manual actuation signal.

SALEM - UNIT 1 3/4 9-4 Amendment No. 299

REFUELING OPERATIONS THIS PAGE INTENTIONALLY BLANK SALEM - UNIT 1 3/4 9-5 Amendment No. 293

REFUELING OPERATIONS THIS PAGE INTENTIONALLY BLANK SALEM - UNIT 1 3/4 9- 6 Amendment No. 293

REFUELING OPERATIONS THIS PAGE INTENTIONALLY BLANK SALEM - UNIT 1 3/4 9-7 Amendment No. 293

REFUELING OPERATIONS 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION ALL WATER LEVELS LIMITING CONDITION FOR OPERATION 3.9.8.1 At least one residual heat removal loop shall be in operation.

APPLICABILITY: MODE 6.

ACTION:

a. With less than one residual heat removal loop in operation, except as provided in

b. below. suspend all operations involving an increase in the reactor decay heat load or a reduction in boron concentration of the Reactor Coolant System, Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b, The residual heat removal loop may be removed from operation for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period during the performance of CORE ALTERATIONS in the vicinity of the reactor pressure vessel hot legs, c, The provisions of Specification 3,0,3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.8.1 In accordance with the Surveillance Frequency Control Program one RHR loop shall be verified in operation and circulating coolant at a flow rate of:

a. greater than or equal to 1000 gpm, and

b. sufficient to maintain the RCS temperature at less than or equal to 140°F, SALEM - UNIT 1 3/4 9-8 Amendment No. 299

REFUELING OPERATIONS LOW WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.8.2 Two independent Residual Heat Removal (RHR) loops shall be OPERABLE.*

APPLICABILITY: MODE 6 when water level above the top of the reactor pressure vessel flange is less than 23 feet.

ACTION:

a. With less than the required RHR loops operable, immediately initiate corrective action to return the required RHR loops to OPERABLE status as soon as possible.

b. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.8.2 The required Residual Heat Removal loops shall be determined OPERABLE per the Inservice Testing Program.

Systems supporting RHR loop operability may be excepted as follows:

a. The normal or emergency power source may be inoperable.

SALEM - UNIT 1 3/49-8a Amendment No. 297

REFUELING OPERATIONS THIS PAGE INTENTIONALLY BLANK I SALEM - UNIT 1 3/4 9-9 Amendment No. 263 i

REFUELING OPERATIONS WATER LEVEL - REACTOR VESSEL LIMITING CONDITION FOR OPERATION 3.9.10 At least 23 feet of water shall be maintained over the top of the reactor pressure vessel flange.

APPLICABILITY: During movement of fuel assemblies or control rods within the reactor pressure vessel while in MODE 6.

ACTION:

With the requirements of the above specification not satisfied, suspend all operations involving movement of fuel assemblies or control rods within the pressure vessel. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.10 The water level shall be determined to be at least its minimum required depth within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior to the start of and in accordance with the Surveillance Frequency Control Program thereafter during movements of fuel assemblies or control rods.

SALEM - UNIT 1 3/49-10 Amendment No. 299

REFUELING OPERATIONS STORAGE POOL WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.11 At least 23 feet of water shall be maintained over the top of irradiated fuel assemblies seated in the storage racks.

APPLICABILITY: Whenever irradiated fuel assemblies are in the storage pool.

ACTION:

With the requirements of the specification not satisfied, suspend all movement of fuel assemblies and crane operations with loads in the fuel storage areas and restore water level to within its limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.11 The water level in the storage pool shall be determined to be at least its minimum required depth in accordance with the Surveillance Frequency Control Program when irradiated fuel assemblies are in the fuel storage pool.

SALEM - UNIT 1 3/49-11 Amendment No. 299

REFUELING OPERATIONS FUEL HANDLING AREA VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.9.12 The Fuel Handling Area Ventilation System shall be OPERABLE with:

a. Two exhaust fans and one supply fan OPERABLE and operating, and

b. Capable of maintaining slightly negative pressure in the Fuel Handling Building.

APPLICABILITY: During movement of irradiated fuel within the Fuel Handling Building ACTION:

a. With no Fuel Handling Area Ventilation System OPERABLE, suspend all operations involving movement of fuel within the storage pool until the Fuel Handling Area Ventilation System is restored to OPERABLE status.

b. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.12 The above required ventilation system shall be demonstrated OPERABLE:

a. In accordance with the Surveillance Frequency Control Program by verifying that, the Fuel Handling Building is maintained at a slightly negative pressure with respect to atmospheric pressure.

b. In accordance with the Surveillance Frequency Control Program by verifying both exhaust fans and one supply fan start and operate for at least 15 minutes, if not operating already.

c. In accordance with the Surveillance Frequency Control Program by verifying a system flowrate of 19,490 cfm +/- 10% during system operation.

SALEM - UNIT 1 3/49-12 Amendment No. 299

REFUELING OPERATIONS THIS PAGE INTENTIONALLY BLANK I

SALEM - UNIT 1 3/4 9-1 3 Amendment No.2631

REFUELING OPERATIONS THIS PAGE INTENTIONALLY BLANK I SALEM - UNIT 1 3/4 9-14 Amendment No. 263 l

This page left intentionally blank SALEM - UNIT 1 3/4 9-15 AAmendment No. 231 l

3/4.10 SPECIAL TEST EXCEPTIONS SHUTDOWN MARGIN LIMITING CONDITION FOR OPERATION 3.10.1 The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 may be suspended for measurement of control rod worth and shutdown margin provided the reactivity equivalent to at least the highest estimated control rod worth is available for trip insertion from OPERABLE control rod(s).

APPLICABILITY: MODE 2.

ACTION:

a. With any full length control rod not fully inserted and with less than the above reactivity equivalent available for trip insertion, immediately initiate and continue boration at ~ 33 gpm of a solution containing ~ 6,560 ppm boron or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.

b. With all full length control rods inserted and the reactor subcritical by less than the above reactivity equivalent, immediately initiate and continue boration at ~ 33 gpm of a solution containing ~ 6,560 ppm boron or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.

4.10.1.1 The position of each full length and part length rod either partially or FULLY WITHDRAWN shall be determined in accordance with the Surveillance Frequency Control Program.

4.10.1.2 Each full length rod not fully inserted shall be demonstrated capable of full insertion when tripped from at least the 50% withdrawn position within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to reducing the SHUTDOWN MARGIN to less than the limits of Specification 3.1.1.1.

SALEM - UNIT 1 3/410-1 Amendment No. 299

SPECIAL TEST EXCEPTIONS GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION 3.10.2 The group height, insertion and power distribution limits of Specifications 3.1.3.1, 3.1.3.4, 3.1.3.5, 3.2.1, and 3.2.4 may be suspended during the performance of PHYSICS TESTS provided:

a. The THERMAL POWER is maintained :5 85% of RATED THERMAL POWER, and

b. The limits of Specifications 3.2.2 and 3.2.3 are maintained and determined at the frequencies specified in Specification 4.10.2.2 below.

APPLICABILITY: MODE 1 ACTION:

With any of the limits of Specifications 3.2.2 or 3.2.3 being exceeded while the requirements of Specifications 3.1.3.1, 3.1.3.4, 3.1.3.5, 3.2.1 and 3.2.4 are suspended, either:

a. Reduce THERMAL POWER sufficient to satisfy the ACTION requirements of Specifications 3.2.2 and 3.2.3, or

b. Be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

4.10.2.1 The THERMAL POWER shall be determined to be :5 85% of RATED THERMAL POWER in accordance with the Surveillance Frequency Control Program during PHYSICS TESTS.

4.10.2.2 The below listed surveillance requirements shall be performed in accordance with the Surveillance Frequency Control Program during PHYSICS TESTS:

a. Surveillances 4.2.2.2 and 4.2.2.3.

b. Surveillances 4.2.3.1 and 4.2.3.2.

SALEM - UNIT 1 3/4 10-2 Amendment No. 299

SPECIAL TEST EXCEPTIONS PHYSICS TESTS LIMITING CONDITION FOR OPERATION 3.10.3 The limitations of Specifications 3.1.1.4, 3.1.3.1, 3.1.3.4, and 3.1.3.5 may be suspended during the performance of PHYSICS TESTS provided:

a. The THERMAL POWER does not exceed 5% of RATED THERMAL POWER, and

b. The reactor trip setpoints on the OPERABLE Intermediate and Power Range Nuclear Channels are set at:s; 25% of RATED THERMAL POWER.

APPLICABILITY: MODE 2.

ACTION:

With the THERMAL POWER> 5% of RATED THERMAL POWER, immediately open the reactor trip breakers.

SURVEILLANCE REQUIREMENTS 4.10.3.1 The THERMAL POWER shall be determined to be:s; 5% of RATED THERMAL POWER in accordance with the Surveillance Frequency Control Program during PHYSICS TESTS.

4.10.3.2 Each Intermediate and Power Range Channel shall be subjected to a CHANNEL FUNCTIONAL TEST prior to initiating PHYSICS TESTS.

SALEM - UNIT 1 3/410-3 Amendment No. 299

SPECIAL TEST EXCEPTION NO FLOW TESTS LIMITING CONDITION FOR OPERATION 3.10.4 The limitations of Specification 3.4.1.1 may be suspended during the performance of startup and PHYSICS TESTS, provided:

a. The THERMAL POWER does not exceed the P-7 Interlock Setpoint, and

b. The Reactor Trip Setpoints on the OPERABLE Intermediate and Power Range Channels are set:s; 25% of RATED THERMAL POWER APPLICABILITY: During operation below the P-7 Interlock Setpoint.

ACTION:

With the THERMAL POWER greater than the P-7 Interlock Setpoint, immediately open the reactor trip breakers.

SURVEILLANCE REQUIREMENTS 4.10.4.1 The THERMAL POWER shall be determined to be less than P-7 Interlock Setpoint in accordance with the Surveillance Frequency Control Program during startup and PHYSICS TESTS.

4.10.4.2 Each Intermediate, Power Range Channel and P-7 Interlock shall be subjected to a CHANNEL FUNCTIONAL TEST prior to initiating startup or PHYSICS TESTS.

SALEM - UNIT 1 3/4 10-4 Amendment No. 299

3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 Deleted SALEM - UNIT I 3/4 11-1 Amendment No. 234

Pages 3/4 11-2 through 3/4 11-4 Intentionally Blank SALEM - UNIT I 3/4 11-2 Amendment No. 234

RADIOACTIVE EFFLUENTS 3/4.11.1.2 Deleted SALEM - UNIT I 3/4 11-S Amendment No. 234

RADIOACTIVE EFFLUENTS 3/4.11.1.3 Deleted SALEM - UNIT I 3/4 11-6 Amendment No. 234

RADIOACTIVE EFFLUENTS LIQUID HOLDUP TANKS*

JJMITING CONDITION FOR OPERATION 3.11.1.4 The quantity of radioactive material contained in each outdoor temporary tank shall be limited to less than or equal to 10 curies, excluding tritium and dissolved or entrained noble gases.

APPLICABILITY: At all times.

ACTION:

a. With the quantity of radioactive material in any of the above listed tanks 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.

b. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREM~TS 4.11.1.4 The quantity of radioactive material contained in each outdoor temporary tank shall be determined to be within the above limit by analyzing a representative sample of the tank's contents in accordance with the Surveillance Frequency Control Program when radioactive materials are being added to the tank.

Tanks included in this Specification are those outdoor temporary tanks that are not surrounded by liners, dikes, or walls capable of holding the tank contents and that do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system.

SALEM - UNIT 1 3/4 11-7 Amendment No. 299

RADIOACTIVE EFFLUENTS 3/4.11.2.1 Deleted SALEM - UNIT I 3/4 11-8 Amendment No. 234

Pages 3/4 11-9 through 3/4 11-11 Intentionally Blank SALEM - UNIT 1 3/4 11-9 Amendment No. 234

RADIOACTIVE EFFLUENTS 3/4.11.2.2 Deleted SALEM - UNIT I 3/4 11-12 Amendment No.234

RADIOACTIVE EFFLUENTS 3/4.11.2.3 Deleted SALEM - UNIT 1 3/4 11-13 Amendment No.234

RADIOACTIVE EFFLUENTS 3/4.11.2.4 Deleted SALEM - UNIT 1 3/4 11-14 Amendment No. 234

RADIOACTIVE EFFLUENTS EXPLOSIVE GAS MIXTURE LIMITING CONDITION FOR OPERAIIQN 3.11.2.5 The concentration of oxygen in the waste gas holdup system shall be limited to less than or equal to 2% by volume.

APPLICABILITY: At all times.

ACTION:

a. With the concentration of oxygen in the waste gas holdup system greater than 2% by volume but less than or equal to 4% by volume, reduce the oxygen concentration to the above limits within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

b. With the concentration of oxygen in the waste gas holdup system greater than 4% by volume immediately suspend all additions of waste gases to the system and reduce the concentration of oxygen to less than or equal to 2% by volume without delay.

c. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.5 The concentration of oxygen in the waste gas holdup system shall be determined to be within the above limits by continuously** monitoring the waste gases in the waste gas holdup system with the oxygen monitor. If hydrogen is not measured, the concentration of hydrogen shall be assumed to exceed 4% by volume.

Not applicable to portions of the Waste Gas System removed from service for maintenance provided that, the portions removed for maintenance are isolated, and purged of hydrogen to less than 4% by volume.

- If the oxygen monitoring instrumentation is inoperable, operation of the waste gas holdup system may continue provided grab samples are collected in accordance with the Surveillance Frequency Control Program and analyzed within the following 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SALEM - UNIT 1 3/4 11-15 Amendment No. 299

THIS PAGE fl IONALLY BLANK SALEM4 - UNIT 1 3/4 11-16 Amendment No. 63

SOLID RADIOACTIVE WASTE 3/4.11.3. Deleted SALEM - UNIT I 3/4 11-17 Amendment No. 234

RADIOACTIVE EFFLUENTS 3/4.11.4 Deleted -

SALEM - UNIT I 3/4 11-18 Amendment No. 234

This Page Intentionally Blank SALEM - UNIT 1 3/4 11-19 Amendment No. 234

3/4.12 Deleted SALEM - UNIT 1 3/4 12-1 Amendment No. 234

This Page Intentionally Blank Pages 3/4 12-3 through 3/4 12-13 Deleted SALEM - UNIT 1 3/4 12-2 Amendment No. 234

SECTION 5.0 DESIGN FEATURES

5.0 DESIGN FEATURES 5.1 SITE EXCLUSION AREA 5.1.1 The exclusion area shall be shown in Figure 5.1.1 LOW POPULATION ZONE 5.1.2 The low population zone shall be as shown in Figure 5.1-2.

UNRESTRICTED AREAS FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS 5.1.3 UNRESTRICTED AREAS within the SITE BOUNDARY that are accessible to MEMBERS OF THE PUBLIC, shall be as shown in Figure 5.1-3.

5.2 CONTAINMENT CONFIGURATION 5.2.1 The reactor containment building is a steel lined, reinforced concrete building of cylindrical shape, with a dome roof and having the following design features:

a. Nominal inside diameter = 140 feet.

b. Nominal inside height = 210 feet.

c. Minimum thickness of concrete walls = 4.5 feet.

d. ,Minimum thickness of concrete roof = 3.5 feet.

e. Minimum thickness of concrete floor mat = 16 feet.

f. Nominal thickness of steel liner = 1/4 to 1/2 inch.

g. Net free volume = 2.62 x 106 cubic feet.

SALEM - UNIT 1 5-1 Amendment No. 59

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LOW POPUL-40iNH ZONE FILGURE i.1-2 SALEUH - UN17T 1 :-i3 Amendment No. 59

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DESIGN FEATURES DESIGN PRESSURE AND TEMPERATURE 5.2.2 The reactor containment is designed and shall be maintained for a maximum internal pressure of 47 psig.

Containment air temperatures up to 351.3 F are acceptable providing the containment pressure is in accordance with that described in the UFSAR.

5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor shall contain 193 fuel assemblies. Each assembly shall consist of a matrix of zircaloy or ZIRLO clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide as fuel material. Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with NRC-approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.

CONTROL ROD ASSEMBLIES 5.3.2 The reactor core shall contain 53 full length and no part length control rod assemblies. The full length control rod assemblies shall contain a nominal 142 inches of absorber material. The nominal values of absorber material shall be 80 percent silver, 15 percent indium and 5 percent cadmium. All control rods shall be clad with stainless steel tubing.

5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE j 5.4.1 The reactor coolant system is designed and shall be maintained:

SALEM - UNIT 1 5-4 Amendment No. 225 NWV 02239

DESIGN FEATURES

a. In accordance with the code requirements specified in Section 4.1 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements,

b. For a pressure of 2485 psig, and

c. For a temperature of 650'F, except for the pressurizer which is 680 0F.

5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure 5.1-1.

5.6 FUEL STORAGE CRITICALITY 5.6.1.1 The new fuel storage racks are designed and shall be maintained with:

a. A maximum Keff equivalent of 0.95 with the storage racks flooded with unborated water.

b. A nominal 21.0 inch center-to-center distance between fuel assemblies.

c. Unirradiated fuel assemblies with enrichments less than or equal to 4.25 weight percent (w/o) U-235 with no requirements for Integral Fuel Burnable Absorber (IFBA) pins.

d. Unirradiated fuel assemblies with enrichments (E) greater than 4.25 w/o U-235 and less than or equal to 5.0 w/o U-235 which contain a minimum number of Integral Fuel Burnable Absorber (IFBA) pins. This minimum number of IFBA pins shall have an equivalent reactivity hold-down which is greater than or equal to the reactivity hold down associated with N IFBA pins, at a nominal 2.35 mg B-10/linear inch loading (1.5X), determined by the equation below:

N = 42.67 ( E - 4.25 SALEM - UNIT 1 5-5 Amendment No. 269

DESIGN FEATURES 5.6.1.2 The spent fuel storage racks are designed and shall be maintained with:

a. A maximum 4K ff equivalent of 0.95 with the storage racks filled with unborated water.

b. A nominal 10.5 inch center-to-center distance between fuel assemblies stored in Region 1 (flux trap type) racks.

c. A nominal 9.05 inch center-to-center distance between fuel assemblies stored in Region 2 (non-flux trap) racks.

SALEM - UNIT 1 5-5a Amendment No. 262

DESIGN FEATURES This Page Is Intentionally Blank SALEM - UNIT I 5-6 Amendment No. 262

DESIGN FEATURES DRAINAGE 5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 124'8".

CAPACITY 5.6.3 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1632 fuel assemblies.

5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT 5.7.1 The components identified in Table 5.7-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.7-1.

SALEM - UNIT 1 5-6a Amendment No. 262 I

C' (r-(I V,

TABLE 5.7-1 I-m COMPONENT CYCLIC OR TRANSIENT LIMITS r-6-4 CYCLIC OR DESIGN CYCLE COMPONENT TRANSIENT LIMIT OR TRANSIENT Reactor Coolant System 200 heatup cycles at < 100'F/hr Heatup cycle - Tavg from < 200'F and 200 cooldown cycles at to > 5420F.

< 100'F/hr (pressurizer cooldown CooTdown cycle - Tavg from > 5420F at < 2000F/hr). to < 200 0F.

80 loss of load cycles. Without immediate turbine or reactor trip.

' 40 cycles of loss of offsite Loss of offsite A.C. electrical A.C. electrical power. power source supplying the onsite Class lE distribution system.

80 cycles of loss of flow in one Loss of only one reactor reactor coolant loop. coolant pump.

400 reactor trip cycles. 100% to 0% of RATED THERMAL POWER.

200 large step decreases 50% of RATED THERMAL POWER in load. step load decrease with steam dump.

Vi)

TABLE 5.7-1 3

I-COMPONENT CYCLIC OR TRANSIENT LIMITS C

0-I CYCLIC OR DESIGN CYCLE COMPONENT TRANSIENT LIMIT OR TRANSIENT Reactor Coolant System 1 main reactor coolant pipe break. Break in a reactor coolant pipe

> 13.5 inches equivalent diameter.

Operating Basis Earthquake 50 cycles Design Basis Earthquake 10 cycles; 0.20g horizontal, 0.136g vertical.

50 leak tests. Pressurized to > 2485 psig.

so 5 hydrostatic pressure tests Pressurized to > 3107 psig.

co Secondary System 1 steam line break Break in a steam line-> 6 inches equivalent diameter.

5 hydrostatic pressure tests Pressurized to > 1356 psig.

10 turbine roll tests Turbine roll on pump heat resulting in plant cooldown > 1000 F/hr.

Q CI . .

I

.1

SECTION 6.0 ADMINISTRATIVE CONTROLS

ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The plant manager shall be responsible for overall facility operation and shall delegate, in writing, the succession to this responsibility during his absence.

6.1.2 The Senior Nuclear Shift Supervisor or, during his absence from the control room, a designated individual, shall be responsible for the control room command function. A management directive to this effect, signed by the senior corporate nuclear officer, shall be reissued to all station personnel on an annual basis.

6.2 ORGANIZATION 6.2.1 ONSITE AND OFFSITE ORGANIZATIONS Onsite and offsite organizations shall be established for unit operation and corporate management, respectively. The onsite and offsite organizations shall include positions for activities affecting the safety of the nuclear power plant.

a. Lines of authority, responsibility, and communication shall be established and defined from the highest management levels through intermediate levels to and including all operating organization positions. These relationships shall be documented and updated, as appropriate, in the form of organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of-documentation. These requirements shall be documented in the Salem Updated Final Safety Analysis Report and updated in accordance with 10 CFR 50.71(e).

b. The plant manager shall be responsible for overall unit safe operation and shall have control over those onsite activities necessary for safe operation and maintenance of the plant.

C. The senior corporate nuclear officer shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety.

d. The individuals who train the operating staff and those who carry out health physics and quality assurance functions may report to the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their independence from operating pressures.

SALEM - UNIT 1 6-1 Amendment No.192 MAR 2 1 i997

ADMINISTRATIVE CONTROLS 6.2.2 FACILITY STAFF The facility organization shall be subject to the following:

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 Reactor Operator shall be in the control room when fuel is in the reactor. In addition, at least one licensed Senior Reactor Operator shall be in the Control Room area at all times.

c. All CORE ALTERATIONS shall be observed and directly supervised by a licensed Senior Reactor Operator who has no other concurrent responsibilities during this operation.

SALEM - UNIT 1 6-2 Amendment No. 290

Figure 6.2-1 CORPORATE HEADQUARTERS AND OFF-SITE ORGANIZATION FOR MANAGEMENT AND TECHNICAL SUPPORT (Deleted)

SALEM - UNIT 1 6-3 Amendment No. 133 I

JUN-8 1992

FIGURE 6.2 - 2 FACILITY ORGANIZATION i (Deleted)

SALEM - UNIT 1 6-4 Amendment No. 133 l JUN-8 1992

TABLE 6.2-1 MINIMUM SHIFT CREW COMPOSITION SALEM UNIT 1 WITH UNIT 2 IN MODES 5 OR 6 OR DE-FUELED POSITION NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION MODES 1, 2, 3 & 4 MODES 5 & 6 SNSS 1a a SRO Ib none STA 1b none NCO 2 1 EO/UO 3 2 Maintenance Electrician 1 none Rad. Pro. Technician a 1 ae WITH UNIT 2 IN MODES 1, 2, 3 OR 4 POSITION NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION MODES 1, 2, 3 & 4 MODES 5 & 6 SNSS 1a 1a SRO 2b none STA 1 none NCO 2d 1 EO/UO 3a 1 Maintenance Electrician 2 none Rad. Pro. Technician 1a Ia a/ Individual may fill the same position on Unit 2.

b/ Individual who fulfills the STA requirement may fill the same position on Unit 2. The STA, if a licensed SRO, may concurrently fill the SRO position on one unit; the other unit also requires a qualified SRO on shift.

c/ One of the two required individuals may fill the position on Unit 2, such that there are a total of three EO/UO's for both units.

d/ One of the three required individuals may fill the same position of Unit 2, such that there are a total of five EO/UO's for both units.

e/ Not needed if both reactors are de-fueled. I SALEM - UNIT 1 6-5 Amendment. No. 133 JUN- 8 1992

TABLE 6.2-1 (Continued)

SNSS - Senior Nuclear Shift Supervisor with a Senior Reactor Operator License on both units.

SRO - Individual with a Senior Reactor Operator License on both units (normally, a Nuclear Shift Supervisor ).

NCO - Nuclear Control Operator with a Reactor Operator License on both units.

STA - Shift Technical Advisor (if licensed as SRO, may be assigned duties as a Nuclear shift Supervisor).

EO/UO - Equipment Operator or Utility Operator.

Except for the Senior Nuclear Shift Supervisor, the Shift Crew Composition may

be one less than the minimum requirements of Table 6.2-1 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 the unexpected absence of on-duty shift crew members provided that immediate action is taken to restore the Shift Crew Composition to within the minimum requirements of Table 6.2-1.

This provision does not permit any shift crew position to be unmanned upon shift change due to an oncoming shift crewperson's being late or absent.

During any absence of the Senior Nuclear Shift Supervisor from the Control Room area while the unit is in any MODE, an individual with a valid SRO License shall be designated to assume the Control Room command function.

SALEM - UNIT 1 6-6 Amendment No. 133 JUN- 8 .1992.

I

ADMINISTRATIVE CONTROLS 6.2.3 SHIFT TECHNICAL ADVISOR 6.2.3.1 The Shift Technical Advisor shall serve in an advisory capacity to the Senior Nuclear Shift Supervisor on matters pertaining to the engineering aspects assuring safe operation of the unit.

6.2.3.2 The Shift Technical Advisor shall have a Bachelor's Degree or equivalent in a scientific or engineering discipline with specific training in plant design and response and analysis of the plant for transients and accidents.

6.3 FACILITY STAFF QUALIFICATIONS 6.3.1 Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except for the individual designated as the Radiation Protection Manager who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975, the individual designated as the Operations Manager who shall meet or exceed the minimum qualifications of ANSI N18.1-1971 except as modified by Specification 6.3.3, and the licensed operators who shall comply with the requirements of 10CFR55.

6.3.2 The Operations Manager or Assistant Operations Manager shall hold an SRO license. The Senior Nuclear Shift Supervisors and Nuclear Shift Supervisors shall each hold a senior reactor operator license. The Nuclear Control Operators shall hold reactor operator licenses.

6.3.3 The Operations Manager shall meet one of the following:

1) Hold an SRO license, or

2) Have held an SRO license for a similar unit (PWR), or

3) Have been certified at an appropriate simulator for equivalent senior operator knowledge.

6.4 TRAINING 6.4.1 A retraining and replacement training program for the facility staff shall:

1) be coordinated by each functional level manager (Department Head) at the facility and maintained under the direction of the Director -

Nuclear Training

2) meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.2-1971 for all affect.ed positions except licensed operators and

3) comply with the requirements of 10CFR55 for licensed operators.

SALEM - UNIT 1 6-7 Amendment No.192 MAR 2 1 1997

ADMINISTRATIVE CONTROLS 6.5 REVIEW AND AUDIT (THIS SECTION DELETED) I SALEM - UNIT 1 6-8 Amendment No. 192 MAUR2 1 IG07

ADMINISTRATIVE CONTROLS PAGES 6-9 THROUGH 6-14 ARE DELETED.

I SALEM - UNIT 1 6-9 through 6-14 Amendment No. 192 MAR 2 1 1'i'37

ADMINISTRATIVE CONTROLS THIS PAGE IS INTENTIONALLY BLANK.

I SALEM - UNIT 1 6-15 Amendment No. 192 MAR 2 1 1997

ADMINISTRATIVE CONTROLS 6.6 REPORTABLE EVENT ACTION 6.6.1 The following actions shall be taken for REPORTABLE EVENTS:

a. The Commission shall be notified and/or a report submitted pursuant to the requirements of Section 50.73 to 10CFR Part 50, and

b. Each REPORTABLE EVENT shall be reviewed by the Station Operations Review Committee (SORC) and the resultant Licensee Event Report submitted to the Nuclear Review Board and the senior corporate nuclear officer.

6.7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:

a. The unit shall be placed in at least HOT STANDBY within one hour.

b. The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within one hour. The senior corporate nuclear officer and senior management position with responsibility for independent nuclear safety assessment activities and quality program oversight 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 be prepared. The report shall be reviewed by the SORC. 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 senior management position with responsibility for independent nuclear safety assessment activities and quality program oversight and the senior corporate nuclear officer within 14 days of the violation.

SALEM - UNIT 1 6-16 Amendment No. 192

?vlAR 2 1 1S97

ADMINISTRATIVE CONTROLS 6.8 PROCEDURES AND PROGRAMS 6.8.1 Written procedures shall be established, implemented and maintained covering the activities referenced below:

a. The applicable procedures recommended in Appendix "A" of Regulatory Guide 1.33, Revision 2, February 1978.

b. Refueling operations.

c. Surveillance and test activities of safety related equipment.

d. Security Plan implementation.

e. Emergency Plan implementation.

f. Fire Protection Program implementation.

g. PROCESS CONTROL PROGRAM implementation.

h. OFFSITE DOSE CALCULATION MANUAL implementation.

i. Quality Assurance Program for effluent and environmental monitoring.

6.8.2 Each procedure and administrative policy of 6.8.1 above, except 6.8.1.d and 6.8.1.e, and changes thereto, shall be reviewed and approved in accordance with requirements in Updated Final Safety Analysis Report (UFSAR) section 17.2 for SORC or for Technical Review and Control, as appropriate, prior to implementation and reviewed periodically as set forth in administrative procedures. Procedures of 6.8.1.d and 6.8.1.e shall be reviewed and approved in accordance with the Facility's Security and Emergency Plans or requirements in Updated Final Safety Analysis Report (UFSAR) section 17.2 for Technical Review and Control, as appropriate, prior to implementation and reviewed periodically as set forth in administrative procedures.

6.8.3 On-the-spot changes to procedures of 6.8.1 above may be made provided:

a. The intent of the original procedure is not altered.

b. The change is approved by two members of the plant management staff, at least one of whom holds a Senior Reactor Operator's License on the unit affected.

c. The change is documented and receives the same level of review and approval as the original procedure within 14 days of implementation.

SALEM - UNIT 1 6-17 Amendment No.192

LMAR 2 1 1037

ADMINISTRATIVE CONTROLS

. 6.8.4 'The following programs shall be maintained:

a. Primary Coolant Sources Outside Containment A program to reduce leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. The systems include (recirculation spray, safety injection, chemical and volume control, gas stripper, recombiners, ... ). The program shall include the following:

(i) Preventative maintenance and periodic visual inspection requirements, and (ii) Integrated leak test requirements for each system at refueling cycle intervals or less.

b. In-Plant Radiation Monitorina A program which will ensure the capability to accurately determine the airborne iodine concentration in areas under accident conditions. This program shall include the following:

(i) Training of personnel, (ii) Procedures for monitoring, and (iii) Provisions for maintenance of sampling and analyses equipment.

c. Secondary Water Chemistry A program for monitoring of secondary water chemistry to inhibit steam generator tube degradation. This program shall include:

(i) Identification of a sampling schedule for the critical variables and the control points for these variables, (ii) Identification of the procedures used to measure the values of the critical variables, (iii) Identification of process sampling points, including monitoring at the discharge of the condensate pumps for evidence of condenser in-leakage.

(iv) Procedures for the recording and management of data, (v) Procedures defining corrective actions for off-control-point chemistry conditions, SALEM - UNIT 1 6-18 Amendment No. 133 JUN 8 1992 -

ADMINISTRATIVE CONTROLS following testing in accordance with this program, the leakage rate acceptance criteria are less than or equal to 0.6 L. for Type B and Type C tests and less than or equal to 0.75 L. for Type A tests;

b. Air lock testing acceptance criteria are:

1) Overall air lock leakage rate is less than or equal to 0.05 L.

when tested at greater than or equal to P.,

2) Seal leakage rate less than or equal to 0.01 L. per hour when the gap between the door seals is pressurized to 10.0 psig.

Test frequencies and applicable extensions will be controlled by the Primary Containment Leakage Rate Testing Program.

The provisions of Specification 4.0.3 will be applied to-the Primary Containment Leakage Rate Testing Program.

6.8.4.g Radioactive Effluent Controls Program A program shall be provided conforming with 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to the MEMBERS OF THE PUBLIC from radioactive effluents as low as reasonably achievable. The program (1) shall be contained in the ODCM, (2) shall be implemented by operating procedures, and (3) shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:

1) Limitations on the operability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM,

2) Limitations on the concentrations of radioactive material released in liquid effluents to UNRESTRICTED AREAS conforming to 10 CFR 20, Appendix B, Table II, Column 2,

3) Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.106 and with the methodology and parameters in the ODCM,

4) Limitations on the annual and quarterly doses or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released from each unit to UNRESTRICTED AREAS conforming to Appendix I to 10 CFR Part 50,

5) Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days.

6) Limitations on the operability and use of the liquid and gaseous effluent treatment systems to ensure that the appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a 92-day period would exceed a suitable fraction of the guidelines for the annual dose or dose commitment conforming to Appendix I to 10 CFR Part 50, SALEM - UNIT I 6-19a Amendment No. 234

ADMINISTRATIVE CONTROLS

7) Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the SITE BOUNDARY conforming to the doses associated with 10 CFR Part 20, Appendix B, Table II, Column 1,

8) Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50,

9) Limitations on the annual and quarterly doses to a MEMBER OF THE PUBLIC from Iodine-131, Iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released from each unit to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50,

10) Limitations on the annual dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources conforming to 40 CFR Part 190.

6.8.4.h Radiological Environmental Monitoring Program A program shall be provided to monitor the radiation and radionuclides in the environs of the plant. The program shall provide (1) representative measurements of radioactivity in the highest potential exposure pathways, and (2) verification of the accuracy of the effluent monitoring program and modeling of environmental exposure pathways. The program shall (1) be contained in the ODCM, (2) conform to the guidance of Appendix I to 10 CFR Part 50, and (3) include the following:

1) Monitoring, sampling, analysis, and reporting of radiation and radionuclides in the environment in accordance with the methodology and parameters in the ODCM,

2) A Land Use Census to ensure that changes in the use of areas at and beyond the SITE BOUNDARY are identified and that modifications to the monitoring program are made if required by the results of the census, and

3) Participation in a Interlaboratory Comparison Program to ensure that independent checks on the precision and accuracy of the measurements of radioactive materials in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring.

6.8.4.i Steam Generator (SG) Program A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following provisions:

a. Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the "as found, condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found, condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes.

Condition monitoring assessments shall be conducted during each SALEM - UNIT I 6-19b Amendment No. 268

ADMINISTRATIVE CONTROLS outage during which the SG tubes are inspected or plugged to confirm that the performance criteria are being met.

b. Performance criteria for SG tube integrity. SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational leakage.

1. Structural integrity performance criterion: All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down and all anticipated transients included in the design specification) and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.

2. Accident induced leakage performance criterion: The primary-to-secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG.

Leakage is not to exceed 1 gallon per minute per SG.

3. The operational leakage performance criterion is specified in LCO 3.4.6.2, "Reactor Coolant System Operational Leakage."

c. Provisions for SG tube repair criteria. Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.

The following alternate repair criteria shall be applied as an alternative to the 40% depth based criteria:

For Refuel Outage 'IR20 through the subsequent operating cycles until the next scheduled SG tube inspection, tubes with service-induced flaws located greater than 13.1 inches below the top of the tubesheet do not require plugging. Tubes with service induced flaws located in the portion of the tube from the top of the tubesheet to 13.1 inches below the top of the tubesheet shall be plugged upon detection.

d. Provisions for SG tube inspections. Periodic SG tube inspections shall be performed.

The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., VOlumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria. For Refuel Outage 1R20 through SALEM - UNIT 1 6-19c Amendment No. 294

ADMINISTRATIVE CONTROLS the subsequent operating cycles until the next scheduled SG tube inspection, portions of the tube below 13.1 inches from the top of the tubesheet are excluded from this requirement.

The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. An assessment of degradation shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

1. Inspect 100% of the tubes in each SG during the first refueling outage following SG replacement.

2. Inspect 100% of the tubes at sequential periods of 120, 90, and thereafter, 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period. No SG shall operate for more than 48 effective full power months or two refueling outages (whichever is less) without being inspected.

3. If crack indications are found in portions of the SG tube not excluded above, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.

e. Provisions for monitoring operational primary-to-secondary leakage.

SALEM - UNIT 1 6-19d Amendment No. 294

ADMINISTRATIVE CONTROLS 6.8A.j Inservice Testing Program This Program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components. The program shall include the following:

a. Testing frequencies applicable to the ASME Code for Operations and Maintenance of Nuclear Power Plants (ASME OM Code) and applicable Addenda as follows:

ASME OM Code and applicable Required Frequencies for Addenda terminology for performing inservice inservice testing activities testing activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 731 days

b. The provisions of Specification 4.0.2 are applicable to the above required frequencies and to other normal and accelerated frequencies specified as 2 years or less in the Inservice Testing Program for performing inservice testing activities,

c. The provisions of Specification 4.0.3 are applicable to inservice testing activities, and

d. Nothing in the ASME OM Code shall be construed to supersede the requirements of any Technical Specification.

6.8A.k Reactor Coolant Pump Flywheel Inspection Program In addition to the requirements of the lSI Program, each Reactor Coolant Pump flywheel shall be inspected per the recommendations of Regulatory Position C.4.b of Regulatory Guide 1.14, Revision 1, August 1975. In lieu of Position CA.b(1) and CA.b(2), a qualified in-place UT examination over the volume from the inner bore of the flywheel to the circle one-half of the outer radius or a surface examination (MT and/or PT) of exposed surfaces of the removed flywheels may be conducted at 20 year intervals.

SALEM - UNIT 1 6-1ge Amendment No. 298

ADMINISTRATIVE CONTROLS 6.8.4.1 Surveillance Frequency Control Program This program provides controls for Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met.

a. The Surveillance Frequency Control Program shall contain a list of Frequencies of those Surveillance Requirements for which the Frequency is controlled by the program.

b. Changes to the Frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI 04-10, "Risk-Informed Method for Control of Surveillance Frequencies," Revision 1.

c. The provisions of Surveillance Requirements 4.0.2 and 4.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program.

SALEM - UNIT 1 6-19f Amendment No. 299

ADMINISTRATIVE CONTROLS 6.9 REPORTING REQUIREMENTS ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the U.S.

Nuclear Regulatory Commission, Document Control Desk, Washington, D.C. 20555, with a copy to the USNRC Administrator, Region I, unless otherwise noted.

STARTUP REPORT 6.9.1.1 A summary report of plant startup and power escalation testing shall be submitted following (1) receipt of an operating license, (2) amendment to the license involving a planned increase in power level, (3) installation of fuel that has different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant.

6.9.1.2 The startup report shall address each of the tests identified in the FSAR and shall include a description of measured values of the'operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license conditions based on other commitments shall be included in this report.

6.9.1.3 Startup reports shall be submitted within (l) 90 days following completion of the startup test program, (2) 90 days following resumption or commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest. If the Startup Report does not cover all three events (i.e., initial criticality, completion of startup test program, and resumption or commencement of commercial power operation), supplementary reports shall be submitted at least every three months until all three events have been completed.

ANNUAL REPORTS*

6.9.1.4 Annual reports covering the activities of the unit as described below for the previous calendar year shall be submitted prior to March 1 of each year.

A single submittal may be made for a multiple unit station. The submittal should combine those sections that are common to all units at the station.

SALEM - UNIT 1 6-20 Amendment No. 133 JUN -8 1992

ADMINISTRATIVE CONTROLS 6.9.1.5 Reports required on an annual basis shall include:

a. DELETED

b. DELETED

c. The results of any specific activity analyses in which the primary coolant exceeded the limits of Specification 3.4.8. The following information shall be included: (1) Reactor power history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded; (2) Results of the last isotopic analysis for radioiodine performed prior to exceeding the limit, results of analysis while the limit was exceeded and results of one analysis after the radioiodine activity was reduced to less than the limit. Each result should include date and time of sampling and the radioiodine concentrations; (3) Clean-up system flow history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded; (4) Graph of the 1-131 concentration and one other radioiodine isotope concentration in microcuries per gram as a function of time for the duration of the specific activity above the steady state level; and (5) The time duration when the specific activity of the primary coolant exceeded the radioiodine limit.

MONTHLY OPERATING REPORT 6,9.1.6 DELETED SALEM - UNIT 1 6-21 Amendment No. 278

ADMINISTRATIVE CONTROLS ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT*

6.9.1.7 The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted before May 1 of each year. The report shall include summaries, interpretations, and analysis of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the objectives outlined in (1) the ODCM and (2) Sections IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR 50.

ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT*

6.9.1.8 The Annual Radioactive Effluent Release Report covering the operation of the unit during the previous 12 months of operation shall be submitted before May 1 of each year. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be (1) consistent with the objective outlined in the ODCM and PCP and (2) in conformance with 10 CFR 50.36a and Section IV.B.1 of Appendix I to 10 CFR 50.

A single submittal may be made for a multiple unit station. The submittal should combine those sections that are common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.

SALEM - UNIT I 6-22 Amendment No. 234

This Page Intentionally Blank SALEM - UNIT I 6-23 Amendment No. 234

ADMINISTRATIVE CONTROLS 6.9.1.9 CORE OPERATING LIMITS REPORT (COLR)

a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

1. Moderator Temperature Coefficient Beginning of Life (BOL) and End of Life (EOL) limits and 300 ppm surveillance limit for Specification 3/4.1.1.4,

2. Control Bank Insertion Limits for Specification 3/4.1.3.5,

3. Axial Flux Difference Limits and target band for Specification 3/4.2.1,

4. Heat Flux Hot Channel Factor, FQ, its variation with core height, K(z), and Power Factor Multiplier PF.,Y Specification 3/4.2.2, and

5. Nuclear Enthalpy Hot Channel Factor, and Power Factor Multiplier, PF8H for Specification 3/4.2.3.

6. Refueling boron concentration per Specification 3.9.1

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1. WCAP-9272-P-A, Westinghouse Reload Safety Evaluation Methodology, (W Proprietary), Methodology for Specifications listed in 6.9.f.9.a.

SALEM - UNIT 1 6-24 Amendment No.284

ADMINISTRATIVE CONTROLS

2. WCAP-8385, Power Distribution Control and Load Following Procedures - Topical Report, (W Proprietary) Methodology for Specification 3/4.2.1 Axial Flux Difference.

3. WCAP-10054-P-A, Westinghouse Small Break ECCS Evaluation Model Using NOTRUMP Code (W Proprietary), Methodology for Specification 3/4.2.2 Heat Flux Hot Channel Factor.

4. WCAP-10266-P-A, The 1981 Version of Westinghouse Evaluation Model Using BASH Code, (W Proprietary) Methodology for Specification 3/4.2.2 Heat Flux Hot Channel Factor.

5. WCAP-12472-P-A, BEACON - Core Monitoring and Operations Support System, (W Proprietary).

6. CENPD-397-P-A, Improved Flow Measurement Accuracy Using Crossflow Ultrasonic Flow Measurement Technology.

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

6.9.1.10 STEAM GENERATOR TUBE INSPECTION REPORT A report shall be submitted within 180 days after the initial entry into HOT SHUTDOWN following completion of an inspection performed in accordance with the Specification 6.8.4.i, "Steam Generator (SG) Program." The report shall include:

a. The scope of inspections performed on each SG,

b. Active degradation mechanisms found,

c. Nondestructive examination techniques utilized for each degradation mechanism,

d. Location, 'orientation (if linear), and measured sizes (if available) of service induced indications,

e. Number of tubes plugged during the inspection outage for each active degradation mechanism,

f. Total number and percentage of tubes plugged to date, and

g. The results of condition monitoring, including the results of tube pulls and in-situ testing.

SALEM - UNIT 1 6- 24 a Amendment No.284

ADMINISTRATIVE CONTROLS Reporting requirements h, i and j are applicable for Refuel Outage 1R20 through the subsequent operating cycles until the next scheduled SG tube inspection.

h. The primary to secondary leakage rate observed in each SG (if it is not practical to assign the leakage to an individual SG, the entire primary to secondary leakage should be conservatively assumed to be from one SG) during the cycle preceding the inspection which is the subject of the report,

i. The calculated accident induced leakage rate from the portion of the tubes below 13.1 inches from the top of the tubesheet for the most limiting accident in the most limiting SG. In addition, if the calculated accident induced leakage rate from the most limiting accident is less than 2.16 times the maximum operational primary to secondary leakage rate, the report should describe how it was determined,

j. The results of monitoring for tube axial displacement (slippage). If slippage is discovered, the implications of the discovery and corrective action shall be provided.

SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the U.S. Nuclear Regulatory Commission, Document Control Desk, Washington, D.C. 20555, with a copy to the Administrator, USNRC Region I within the time period specified for each report.

6.9.3 DELETED 6.9.4 When a report is required by ACTION 8 or 9 of Table 3.3-11 "Accident Monitoring Instrumentation", a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring for inadequate core cooling, the cause of the inoperability, and the plans and schedule for restoring the instrument channels to OPERABLE status.

SALEM - UNIT 1 6-24b Amendment No. 294

AEMINISTATIV COUNMIS 6.10 RECRD RETE 0N In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at least the mnnirmn period indicated.

6.10.1 The following records shall be retained for at least five years:

a. Records and logs of unit operation covering time interval at each poer level.

b. Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety.

c. All REEOSrABLE EVENS submitted to the Commission.

d. Records of surveillance activities, inspections and calibrations required by these Technical Specifications.

e. DELETED

f. Records of changes made to Operating Procedures required by Specification 6.8.1.

g. Records of radioactive shipments.

h. Records of sealed source and fission detector leak tests and results.

i. Records of annual physical inventory of all sealed source material of record.

j. Records of reviews performed for changes made to procedures or reviews of tests and experiments, pursuant to 10CFR50.59.

6.10.2 The following records shall be retained for the duration of the Unit Operating license:

a. Records and drawing changes reflecting unit design modifications made to systems and equipment described in the Final Safety Analysis Report, pursuant to IOCFR50.59.

b. Records of new and irradiated fuel inventory, fuel transfers and assembly burnmp histories.

SALEK - UNIT 1 6-25 Aexndment No. 139 JANS5 1993 J

ADMINISTRATIVE CONTROLS

c. Records of radiation exposure for all individuals entering radiation control areas.

d. Records of gaseous and liquid radioactive material released to the environs.

e. Records of transient or operational cycles for those facility components identified in Table 5.7-1.

f. Records of reactor tests and experiments.

g. Records of training and qualification for current members of the plant staff.

h. Records of in-service inspections.

i. Records of Quality Assurance activities required by the QA Manual.

j. DELETED

k. Records of SORC meetings and activities of the Nuclear Review Board (and activities of its predecessor, the Offsite Safety Review (OSR) staff.)

I. Records for Environmental Qualification which are covered under the provisions of Paragraph 6.16.

m. Records of the service lives of all hydraulic and mechanical snubbers including the date at which the service life commences and associated installation and maintenance records.

n. Records of secondary water sampling and water quality.

o. Records of analyses required by the radiological environmental monitoring program which would permit evaluation of the accuracy of the analysis at a later date. This should include procedures effective at specified times and QA records showing that these procedures were followed.

p. Records of reviews performed for changes made to the OFFSITE DOSE CALCULATION MANUAL and the PROCESS CONTROL PROGRAM.

6.11 RADIATION PROTECTION PROGRAM Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for all operations involVing personnel radiation exposure.

SALEM - UNIT 1 6-26 Amendment No. 297

ADMINISTRATIVE CONTROLS 6.12 HIGH RADIATION AREA As provided in paragraph 20.1601(c) of 10 CFR Part 20, the following controls shall be applied to high radiation areas in place of the controls required by paragraph 20.1601(a) and (b) of 10 CFR Part 20:

6.12.1 High Radiation Areas with Dose Rates Not Exceeding 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation

a. Each entryway to such an area shall be barricaded and conspicuously posted as a high radiation area. Such barricades may be opened as necessary to permit entry or exit of personnel or equipment.

b. Access to, and activities in, each such area shall be controlled by means of Radiation Work Permit (RWP) or equivalent that includes specification of radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures.

c. Individuals qualified in radiation protection procedures and personnel continuously escorted by such individuals may be exempted from the requirement for an RWIP or equivalent while performing their assigned duties provided that they are otherwise following plant radiation protection procedures for entry to, exit from, and work in such areas.

d. Each individual or group entering such an area shall possess:

1. A radiation monitoring device that continuously displays radiation dose rates in the area; or

2. A radiation monitoring device that continuously integrates the radiation dose rates in the area and alarms when the device's dose alarm setpoint is reached, with an appropriate alarm setpoint, or

3. A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area, or

4. A self-reading dosimeter (e.g., pocket ionization chamber or electronic dosimeter) and, (i) Be under the surveillance, as specified in the RWP or equivalent, while in the area, of an individual qualified in radiation protection procedures, equipped with a radiation monitoring device that continuously displays radiation dose rates in the area; who is responsible for controlling personnel exposure within the area, or SALEM - UNIT 1 6-27 Amendment No. 255 APR 1 0 MD

(ii) Be under the surveillance as specified in the RWP or equivalent, while in the area, by means of closed circuit television, of personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and with the means to communicate with individuals in the area who are covered by such surveillance.

e. Except for individuals qualified in radiation protection procedures, or personnel continuously escorted by such individuals, entry into such areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them. These continuously escorted personnel will receive a pre-job briefing prior to entry into such areas. This dose rate determination, knowledge, and pre-job briefing does not require documentation prior to initial entry.

6.12.2 High Radiation Areas with Dose Rates Greater than 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation, but less than 500 rads/hour at 1 Meter.

from the Radiation Source or from any Surface Penetrated by the Radiation

a. Each entryway to such an area shall be conspicuously posted as a high radiation area and shall be provided with a locked or continuously guarded door or gate that prevents unauthorized entry, and, in addition:

1. All such door and gate keys shall be maintained under the administrative control of the shift supervisor, radiation protection manager, or his or her designee.

2. Doors and gates shall remain locked except during periods of personnel or equipment entry or exit.

b. Access to, and activities in, each such area shall be controlled by means of an RWP or equivalent that includes specification of radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures.

c. Individuals qualified in radiation protection procedures may be exempted from the requirement for an RWP or equivalent while performing radiation surveys in such areas provided that they are otherwise following plant radiation protection procedures for entry to, exit from, and work in such areas.

d. Each individual or group entering such an area shall possess:

1. A radiation monitoring device that continuously integrates the radiation dose rates in the area and alarms when the device's dose alarm setpoint is reached, with an appropriate alarm setpoint, or SALEM - UNIT 1 6-27a Amendment No.255 APR 10 2003

2. A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area with the means to communicate with and control every individual in the area, or

3. A self-reading dosimeter (e.g., pocket ionization chamber or electronic dosimeter) and, (i) Be under the surveillance, as specified in the RWP or equivalent, while in the area, of an individual qualified in radiation protection procedures, equipped with a radiation monitoring device that continuously displays radiation dose rates in the area; who is responsible for controlling personnel exposure within the area, or (ii) Be under the surveillance as specified in the RWP or equivalent, while in the area, by means of closed circuit television, of personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and with the means to communicate with and control every individual in the area.

4. In those cases where options (2) and (3), above, are impractical or determined to be inconsistent with the "As Low As is Reasonably Achievable' principle, a radiation monitoring device that continuously displays radiation dose rates in the area.

e. Except for individuals qualified in radiation protection procedures, or personnel continuously escorted by such individuals, entry into such areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them. These continuously escorted personnel will receive a pre-job briefing prior to entry into such areas. This dose rate determination, knowledge, and pre-job briefing does not require documentation prior to initial entry.

f. Such individual areas that are within a larger area where no enclosure exists for the purpose of locking and where no enclosure can reasonably be constructed around the individual area need not be controlled by a locked door or gate, nor continuously guarded, but shall be barricaded, conspicuously posted, and a clearly visible flashing light shall be activated at the area as a warning device.

SALEM - UNIT 1 6-27b Amendment No.255 APR 1 0 2D3

ADMINISTRATIVE CONTROLS 6.13 PROCESS CONTROL PROGRAM (PCP) 6.13.1 The PCP shall be approved by the Commission prior to implementation.

6.13.2 Changes to the PCP:

1. Shall be documented and records of review performed shall be retained as required by Specification 6.10.3p. This documentation shall contain:

a) Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change(s) and b) A determination that the change will maintain the overall conformance of the solidified waste product to existing requirements of Federal, State, or other applicable regulations.

2. Shall become-effective after review and acceptance by the SORC and the approval of the Plant Manager.

SALEM - UNIT 1 6-28 Amendment No. 255 I

APR 1 0 2003

ADMINISTRATIVE CONTROLS 6.14 OFFSITE DOSE CALCULATION MANUAL (ODCM) 6.14.1 The ODCM shall be approved by the Commission prior to implementation.

6.14.2 Changes to the ODCM:

1. Shall be documented and records of reviews performed shall be retained as required by Specification 6.10.3p. This documentation shall contain:

a) Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change(s) and b) A determination that the change will maintain the level of radioactive effluent control required by 10 CFR 20.106, 40 CFR Part 190, 10 CFR 50.36a, and Appendix I to 10 CFR Part 50 and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations.

2. Shall become effective after review and acceptance by the SORC and the approval of the Plant Manager.

3. Shall be submitted to the Commission in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Annual Radioactive Effluent Release Report for the period of the report in which any change to the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g. month/year) the change was implemented.

6.15 MAJOR CHANGES TO RADIOACTIVE LIQUID, GASEOUS AND SOLID WASTE TREATMENT SYSTEMS 6.15.1 Licensee initiated major changes to the radioactive waste system (liquid, gaseous and solid):

1. Shall be reported to the Commission in the UFSAR for the period in which the evaluation was reviewed by (SORC). The discussion of each change shall contain:

a. A summary of the evaluation that led to the determination that the change could be made in accordance with 10CFR50.59;

b. Sufficient detailed information to totally support the reason for the change without benefit of additional or supplemental information; SALEM - UNIT l 6-29 Amendment No. 234

ADMINISTRATIVE CONTROLS C. A detailed description of the equipment, components and processes involved and the interfaces with other plant systems;

d. An evaluation of the change, which shows the predicted releases of radioactive materials in liquid and gaseous effluents and/or quantity of solid waste that differ from those previously predicted in the license application and amendments thereto;

e. An evaluation of the change, which shows the expected maximum exposures to individual in the unrestricted area and to the general population that differ from those previously estimated in the license application and amendments thereto;

f. A comparison of the predicted releases of radioactive materials, in liquid and gaseous effluents and in solid waste, to the actual releases for the period prior to when the changes are to be made;

g. An estimate of the exposure to plant operating personnel as a result of the change; and

h. Documentation of the fact that the change was reviewed and found acceptable by the (SORC).

2. Shall become effective upon review and acceptance by the SORC.

6.16 ENVIRONMENTAL QUALIFICATION 6.16.1 All safety-related electrical equipment in the facility shall be qualified in accordance with the provisions of: Division of Operating Reactors OGuidelines for Evaluating Environmental Qualification of Class 1E Electrical Equipment in Operating Reactors' (DOR Guidelines); or, NUREG-0588

'Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment,* December 1979. Copies of these documents are attached to Order for Modification of License No. DPR-70 dated October 24, 1980.

6.16.2 Complete and auditable records shall be available and maintained at a-central location which describe the environmental qualification method used for all safety related electrical equipment in sufficient detail to document the degree of compliance with the DOR Guidelines or NUREG-0588. Such records should be updated and maintained current as equipment is replaced, further tested, or otherwise further qualified.

6.17 TECHNICAL SPECIFICATION (TS) BASES CONTROL PROGRAM This program provides a means for processing changes to the Bases of these Technical Specifications.

a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

b. PSEG may make changes to the Bases without prior NRC approval provided the changes do not require either of the following:

1. A change in the TS incorporated in the License, or

2. A change to the updated FSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.

SALEM - UNIT I 6-30 Amendment No. 256

ADMINISTRATIVE CONTROLS

c. Proposed changes to the Bases that require either condition of Specification 6.17.b above shall be reviewed and approved by the NRC

d. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).

e. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the UFSAR.

6.18 CONTROL ROOM ENVELOPE HABITABILITY PROGRAM A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Emergency Air Conditioning System (CREACS), CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of 5 rem total effective dose equivalent (TEDE) for the duration of the accident.

The program shall include the following elements:

a. The definition of the CRE and the CRE boundary.

b. Requirements for maintaining the CRE boundary in its design condition including configuration control and preventive maintenance.

c. Requirements for (i) determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors," Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and .C.2 of Regulatory Guide 1.197, Revision 0.

d. Measurement, at designated locations, of the CRE pressure relative to all external areas adjacent to the CRE boundary during the pressurization mode of operation by one train of the CREACS, operating at the flow rate required by the Surveillance Requirements, at a frequency of 36 months on a STAGGERED TEST BASIS. The results shall be trended and used as part of the 18 month assessment of the CRE boundary.

e. The quantitative limits on unfiltered air inleakage into the CRE. These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph

c. The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of DBA consequences. Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.

f. The provisions of Surveillance Requirements 4.0.2 are applicable to the Frequencies for assessing CRE habitability, determining CRE unfiltered inleakage, and measuring CRE pressure and assessing the CRE boundary as required by paragraphs c and d, respectively.

SALEM - UNIT 1 6-31 Amendment No. 286

APPENDIX B TO FACILITY OPERATING LICENSE NO. DPR-70 SALEM GENERATING STATION UNIT 1 DOCKET NO. 50-272 AND FACILITY OPERATING LICENSE NO. DPR-75 SALEM GENERATING STATION UNIT 2 DOCKET NO. 50-311 PSEG NUCLEAR LLC I

ENVIRONMENTAL PROTECTION PLAN (NONRADIOLOGICAL)

Amendment No. 100, 233 I AUG 21 2D00

SALEU GENETIM SrTIC UNIT NS. I AND 2 EVIRONME N , PRII ON PLAN (NRADIOLMICIL)

TABLE OF CONTENTS Section LIM 1.0 Cbjectives of the Enviromental Protection Plan ............. 1-1 2.0 Envirsrnental Protection Issues ......................................

2-1 3.0 Consistency Requirements .. ..3-..3 3.1 Plant Design and Operation .................................................3-1 3.2 Reportinq Related to the NJPMS Permit and State certificaticn ....*....... *.... *................... # ...........

.. 3-2 3.3 Changes Required for Ompliarc with Other Envircnmental Regulations ....................................................... 3-3 4.0 Environmental Ccnditions ........................................... 4-1 4.1 Unusual or Important Environmental Events .................. . 4-1 4.2 Environmental Mbnitoring . . .4-1 5.0 Adninistrative Procedures ............. ...................... 5-1 5.1 Review .................... o............................ . . 5-l 5.2 Records Retention. ............... ............. *............... 5-1 5.3 Changes in Environmental Protection Plan .................... 5-2 5.4 Plant Repcstinq Requirements .............. Ant.................

5-2 Amendment No. 100 AUG ° 1 IN9

1.0 Cbjectives of the nvirrmental Protetion Plan The Environental Protection Plan (EPP) is to provide for protection of noradioloical envirental values during cperation of the nuclear facility. The principal ojectives of the EPP are as follows:

(1) 'Verify that the facility is operated in an enviroxnntally aceptable manner, as establishei by the Final Envirormental

Statement - Cperating Licersirg Stage (FES-OL) arxl other NRC environmental impact assessnts.

(2) Coordinate NFC requirements and maintain consisterncy with other Federal, State and local reuyirements for environmental protection.

(3) Keep NRC informed of the ernvironmental effects of facility crdnstzction and operation and of actions taken to control those effects. \

Environnental corns identified in the FES-OL vhidc relate to water quality matters are regulated by way of the licensee's NJES permit.

1-1 Amendment No. 100 AUG ° 1 1989

2.0 EnVircrnztal Protwicn Issues In the FES, dated April 1973, the staff cridersd the enviromenta:l impacts associated with the cperation of Salem Generating Station Unit Nos. 1 and 2. Certain envirormntal issues wre identified whici required study or license caditicns to resolve aid to assure adequate protection of the enviroesnt. The Appendix B Environmental Technical Specifications (EIS) issued with the operating license included discharge restrictions and monitoring programs related to aquatic and terrestrial rescurcs.

1. Protection of the aquatic envirtoxnnt by limiting the thermal characteristics of the discharge.

2. Protection of the aquatic environment from biocide used in plant operations.

3. Protection of the aquatic enviomesnt from suspe~ed solids and changes in pH in releases from the non-radioactive liquid waste dispsal ystem.

4. Surveillance progrm for dissolved gases, suspernded solids, chemical releases, and the general aquatic ecological surveys to establish impact of plant operation on the biotic environment.

2-1 Amendment No. 100 AUG z 1 1989

2.1 Aquatic Issue Raeqiireents for stuxy of station intake and disdharges effects were removed fry the ErS by License Amrendmets 51 (Unit 1) and 18 Unit 2, dated March 14, 1983 and Mardi 11, 1983, respectively. These ises no are addressed by the effluent limitatirs and snitoring rwaireits c rtairne in the effective NjPESb Permit No. NJOO5622 issued by the State of New Jersey, and by the determination of the State of New Jersey on the Section 316(a) and (b) demonstration submitted by licensee. 'he NRC will rely on the State for regulation of matters involving water quality and aquatic biota.

2.2 Terrestrial Issues Requirements for study of station effects on terrapins and raptors have been met.

2-2 Amendment No. 100 AUG ' 1 1989

3.0 CcUisteroy Reyireiets 3.1 Plant Design and Cperation The licensee may make dcwhes in station design or Coration or perform tests or experients affecting the envirint provided such activities do not involve an unreviewed envirormental question and do not involve a change in the EPP*. Charnes in station design or operation performance of tests or experiments which do not affect the envirorment are not subject to the requirents of this EPP.

Activities governed by Section 3.3 are not subject to the reuirements of this Section.

Before enraging in additional ctruction or operational activities which may significantly affect the envirement, the licensee shall prepare and record an environmental evaluation of such activity.

Activities are excluded fram this reurmnt if all measurable nonradiological environmental effects are confined to the on-site areas previously disturbed during site preparation and plarA contruction.

When the evaluation indicates that such activity involves an unreviewed environmental question, the licensee shall provide a written evaluation of sudh activity and obtain prior NC approval.

When such activity involves a change in the EPP, sudh activity and change to the EPP may be implemented only in accordance with an appropriate license amenmbent as set forth in Section 5.3 of this EPP.

This provision does not relieve the licensee of the requirements of 10 CFR 50.59.

3-1 Amendment No. 100 AUG 2 3 1989

A pripoed ctarge, test or experirent shall be deemed to involve an ureiewe enviromental questicn if it cr ns: (1) a matter which may result in a significant increase in any adverse enviroimental i=pact previously evaluated in the FES-OL, envirrruental upact appraisals, or in any decision s of the Atomic Safety ard Uicnsing Board; or (2) as significant change in effluents or power level; or (3) a matter, not previously reviewed and evaluated in the documents specified in (1) of this Subsection, which may have a significant adverse environmental impact.

The licensee shall maintain records of changes in facility design or operation and of tests and experimnts arried out pursuant to this Subsecticr. These records shall include written evaluations which provide bases for the deterinatin that the change, test, or experiment does not involve an unreviewed envircnmental question or constitute a decrease in the effectiveness of this EPP to meet the objectives specified in Section 1.0. The licensee shall include as part of the Annual Environmental Cperating Report (per SubsectionJ .4.1) brief descriptions, analyses, interpretations, and evaluations of such changes, tests and experiments.

3.2 Reportinn Related to the NnMM Permit or the State Certification Changes to, or renewals of, the KurPM Permit or the State certification shall be reported to the NRC within 30 days following the date the change or renewal is approved. If a permit or certification, in part or in its entirety, is appealed and stayed, the NRC shall be notified within 30 days following the date the stay is granted.

3-2 Amendment No. ioo AUG n 1199

e licene Ishall notify the NRC of changes to the effective Go Permit pryposed by the linrerie by providing NRC with a coy of the prcposed charne at the sa- tie it is suWaitted to the prmitting agency. The licensee shall provide the NC a copy of the application for renewal of the NJOES Permit at the same time the application is submitted to the permitting agency.

3.3 Charnges RPqired for Cmzplia with Other Erwvironmntal Regulations Changes in plant design or operation and performance of tests or experiments which are required to achieve calliance with other Federal, State, ard lo1l envir=intal-regulatin are not ibjsect to the requiremnts of Sectia 3.1.

3-3 Amendment No. 100 AUG ° 1 1989

1/4'.

4.0 Environmental Conditions 4.1 Unusual or Important Environmental Events Any occurrence of an unusual or important event that indicates or could result in significant environmental impact causally related to plant operation shall be recorded and reported to the NRC within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months followed by a written report per Subsection 5.4.2. The following are examples: excessive bird impaction events; onsite plant or animal disease outbreaks; mortality or unusual occurrence of any species protected by the Endangered Species Act of 1973; fish kills or impingement events on the intake screens; increase in nuisance organisms or conditions; unanticipated or emergency discharge of waste water or chemical substances.

No routine monitoring programs are required to implement this condition.

4.2 Environmental Monitoring 4.2.1 Aquatic Monitoring The certifications and permits required under the Clean Water Act provide mechanisms for protecting water quality and, indirectly, aquatic biota. The Nuclear Regulatory Commission (NRC) will rely on the decisions made by the State of New Jersey under the authority of the Clean Water Act and, in the case of sea turtles and shortnose sturgeon, decisions made by the National Marine Fisheries Service (NMFS) under the authority of the Endangered Species Act, for any requirements pertaining to aquatic monitoring.

In accordance with Section 7(a) of the Endangered Species Act, on May 14, 1993, the National Marine Fisheries Service issued a Section 7 Consultation Biological Opinion related to the operation of Salem Unit 1 and 2 Generating Stations. This Section 7 Consultation entitled, "Reinitiation of a consultation in accordance with Section 7(a) of the Endangered Species Act regarding continued operation of the Salem and Hope Creek Nuclear Generating Stations on the eastern shore of the Delaware River in New Jersey," concluded that "...continued operation is not likely to jeopardize the continued existence of listed species."

PSEG Nuclear LLC shall adhere to the specific requirements within the Incidental Take Statement, to the Biological Opinion. Changes to the incidental take statement must be proceeded by consultation between the NRC, as the authorizing agency, and NMFS.

The Conservation Recommendation, to the Biological Opinion suggests conservation recommendations for implementation by Salem Generating Station.

The Station shall implement these recommendations to the satisfaction of the NRC and National Marine Fisheries Service.

SALEM UNIT 1 4-1 Amendment No. 3 r,233 AUG 2 1 20O0

I 4.2.2 Terrestrial Monitoring Terrestrial monitoring is not required.

SALEM UNIT 1 4-2 Amendment No. 216 DEC 1 'SW

5.0 Aninistrative Procdres 5.1 Review me licensee shall provide for review of omplianc with the EPP. The review shall be conducted irveperdently of the irdividual or group responsible for performirg the specific activity. A description of the organization structure utilized to achieve the irdeperdent review funrtion and results of the review activities shall be mintailed and made available for inspection.

5.2 Records Retention Records and logs relative to the enviromental aspects of station cperation shall be mae and retained in a manner convenient for review and inspection. These rourds and logs shall be made available to NRC on request.

Records of rmdificatians to station structures, sXsel and =cnrents determdnd to potentially affect the oxtinued'protection of the envirnmental shall be retained for the life of the station. All other records, data and logs relating to this EPP shall be retained for five years or, Where applicable, in ardance with the requirements of other agencies.

5-1 Amendment No. 100 RUG n 1 that

5.3 Changes in Envirormental Protection Plan Requests for changes in the EPP shall include an assessment of the environmental impact of the prcposed change and a supporting justification. Implementation of such changes in the EPP shall not cammenoe prior to NRC approval of the proposed changes in the form of a license amendment incorporating the appropriate revision to the EPP.

5.4 Plant Reporting Requireients 5.4.1 Routine Reports An Annual Environmental Operating Report d ibing implmentation of this EPP for the previous year shall be submitted to the NRC prior to May 1 of each year.

2e report shall include summaries and analyses of the results of the envircnmental protecticn activities required by Subsection 4.2 (if any) of this EPP for the report period, irnluding a cAparison with related preoperational studies, operational cotrols (as apprcpriate),

and previcus nrzradiological envirciimntal

_ monitoring reports, and an assessnt of the dbserved impacts of the plant operation on the environment. If harmful effects or evidence of trends toward irreversible damage to the environmet are observed, the licensee shall provide a detailed analysis of the data and a prpcosed course of mitigating action.

5-2 Amendment No.100 AU1G ~ I ASS

The Annual Environmental Operating Report shall also include:

(1) A list of EPP noncompliances and the corrective actions taken to remedy them.

(2) A list of all changes in station design or operation, tests, and experiments made in accordance with Subsection 3.1 which involved a potentially significant unreviewed environmental question.

(3) A list of nonroutine reports submitted in accordance with Subsection 5.4.2.

In the event that some results are not available by the report due date, the report shall be submitted noting and explaining the missing results. The missing results shall be submitted as soon as possible in a supplementary report.

5.4.2 Nonroutine Reports A written report shall be submitted to the NRC within 30 days of occurrence of a nonroutine event. The report shall: (a) describe, analyze, and evaluate the event, including extent and magnitude of the impact, and plant operating characteristics; (b) describe the probable cause of the event; (c) indicate the action taken to correct the reported event; (d) indicate the corrective action taken to 5-3 Amendment No. 100 AJG2 1.

I %

preclude repetition of the event and to prevent similar occurrences involving similar components or systems; and (e) indicate the agencies notified and their preliminary responses.

Events reportable under this subsection which also require reports to other Federal, State or local agencies shall be reported in accordance with those reporting requirements in lieu of the requirements of this subsection. The NRC shall be provided with a copy of such report at the same time it is submitted to the other agency.

5-4 Amendment No. 100 AUG 2 1 1989

APPENDIX C ADDITIONAL CONDITIONS OPERATING LICENSE NO. DPR-70 PSEG Nuclear LLC, and the Exelon Generation Company, LLC, shall comply with the following conditions on the schedules noted below:

Amendment Ntmber Additional Condition Implementation Date 192 The licensee is authorized to relocate certain The amendment shall be Technical Specification requirements to licensee- implemented within 60 controlled documents. Implementation of this days from March 21, amendment shall include the relocation of these 1997.

technical specification requirements to the appropriate documents, as described in the licensee's application dated January 11, 1996, as supplemented by letters dated February 26, May 22, June 27, July 12, December 23, 1996, and March 17, 1997, and evaluated in the staff's safety evaluation attached to this amendment.

194 The licensee is authorized to upgrade the initiation The amendment shall be circuitry for the power operated relief valves, as implemented prior to described in the licensee's application dated January entry into Mode 3 from 31, 1997, as supplemented by letters dated March 14, the current outage for April 8. and April 28, 1997, and evaluated in the Salem Unit 1.

staff's safety evaluation attached to this amendment.

196 Containment Fan Cooler Units The amendment shall be implemented prior to The licensee shall complete all modifications entry into Mode 3 from associated with the amendment request concerning the current outage for Containment Fan Cooler Units (CFCU) response time Salem Unit 1.

dated October 25, 1996. as described in the letters supplementing the amendment request dated December 11, 1996. January 28, March 27, April 24, June 3, and June 12, 1997, prior to entry into Mode 3 following refueling outage 12. All modifications made in support of this amendment request and described in the referenced sumittals shall be in conformance with the existing design basis for Salem Unit 1, and programmatic controls for tank monitoring instrumentation shall be as described in the letter dated April 24. 1997. Post modification testing and confirmatory analyses shall be as described in the letter dated March 27, 1997. Future changes to the design described in these submittals may be made in accordance with the provisions of 10 CFR 50.59.

Further, the administrative controls associated with CFCU operability and containment integrity described in the letters dated March 27, and April 24, 1997 shall not be relaxed or changed without prior staff review until such time as the license has been amended to include the administrative controls as technical specification requirements.

196 The licensee shall perform an evaluation of the The amcncment shall be containment liner anchorage by November 30, 1997, for implemented within 30 the loading induced on the containment liner during a days from July 17, 1997.

Main Steam Line Break event to confirm the assumptions provided in the Preliminary Safety Analysis Report and Updated Final Safety Analysis Report.

Amendment No. 246 l

v t e Letter Sequence Request
TAC:ME1834, Control Room Habitability (Approved, Closed)
Initiation Request, Request, Request, Request, Request, Request Acceptance Administration Questions 14 April 2010 24 May 2010 17 June 2010 10 June 2010 7 June 2010 10 June 2010 14 June 2010 30 June 2010 11 June 2010 6 August 2010 11 June 2010 25 June 2010 29 June 2010 12 July 2010 30 July 2010 3 August 2010 19 July 2010 23 July 2010 3 August 2010 9 August 2010 19 August 2010 29 September 2010 12 October 2010 22 November 2010 25 October 2010 4 November 2010 10 December 2010 20 December 2010 20 August 2012 20 August 2012 Answers 28 May 2010 13 July 2010 10 August 2010 26 August 2010 7 September 2010 14 December 2010 21 December 2010 Results Approval, Approval, Approval Other: LR-N16-0144, Completion of Activities to Support Entry Into the Period of Extended Operation, ML092400531, ML092400532, ML092430231, ML092460442, ML092470039, ML092520185, ML092650382, ML092660174, ML092660447, ML100850459, ML102280211, ML102430586, ML11125A024
MONTHYEAR2010-06-11 [Table View]

ML11175A207

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