Text

Application for Amend to Licenses DPR-70 & DPR-75,revising Amend Requests 79-01 & 81-19 Re Radiological Effluent & Administrative Controls & Tech Specs Re Surveillance Testing & Reporting Requirements for Reactor Trip Breakers
	ML18089A198
Person / Time
Site:	Salem
Issue date:	06/17/1983
From:	Liden E; Public Service Enterprise Group
To:	Varga S; Office of Nuclear Reactor Regulation
Shared Package
ML18089A199	List: ML18089A198; ML18089A200;
References
	LCR-79-01, LCR-79-1, LCR-81-19, LCR-83-08, LCR-83-8, NUDOCS 8306270299
	Download: ML18089A198 (246)
	v • d • e

,/

. P~G Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038 Nuclear Department Ref: LCR 79-01 LCR 81-19 LCR 83-08 June 17, 1983 Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention: Mr. Steven Varga, Chief Operations Reactors Branch 1 Division of Licensing Gentlemen:

REQUEST FOR AMENDMENT FACILITY OPERATING LICENSES DPR-70 AND DPR-75 UNIT NOS. 1 AND 2 SALEM GENERATING STATION DOCKET NOS. 50-272 AND 50-311 In accordance with the Atomic Energy Act of 1954, as amended and the regulations thereunder, we hereby transmit copies of our request for amendment and our analyses of the changes to Facility Operating Licenses DPR-70 and DPR-75 for Salem Generating Station, Unit Nos. 1 and 2.

This three part request consists of revisions to two previously submitted requests for Amendment (79-01 and 81-19) pertaining to Radiological Effluent Controls and Administrative Controls, and to one new request (83-08).

Amendment request (83-08) includes proposed changes to the Safety Technical Specifications, Appendix A, pertaining to surveillance testing and reporting requirements for reactor trip breakers, which are being submitted as part of our Corrective Action Program.

This change involves a single safety issue and is, therefore, determined to be a Class III Amendment for one unit and a Class I Amendment for the other unit as defined by 10 CFR170.22. A check in the amount of $4,400 is enclosed.

The Energy People

,~/

~.

Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission 6/17/83 Pursuant to the requirements of 10 CFR50.9l(b)(l), a copy of this request for amendment has been sent to the State of New Jersey as indicated below.

This submittal includes three (3) signed originals and forty (40) copies.

Very truly yours, Manager - Nuclear Licensing and Regulation Enclosure CC: Mr. Donald C. Fischer Licensing Project Manager Mr. Leif Norrholm Senior Resident Inspector Mr. Frank Cosolito, Acting Chief Bureau of Radiation Protection Department of Environmental Protection 380 Scotch Road Trenton, New Jersey 08628

Ref: LCR 79-01 LCR 81-19 LCR 83-08 STATE OF NEW JERSEY )

) SS.

COUNTY OF SALEM )

RICHARD A. UDERITZ, being duly sworn according to law deposes and says:

I am a Vice President of Public Service Electric and Gas Company, and as such, I find the matters set forth in our letter dated June 17, 1983, concerning, Request for Amendment to Facility Operating Licenses DPR-70 and DPR-75, are true to the best of my knowledge, information and belief.

RICHARD Subscribed and sworn to before me this RUDOLPH L. von FISCHER JA.

Notary Public of New Jers,ay My Commission expires on My Commission Expires Sept. 10, 1986

-* INDEX l

DEFINITIONS SECTION PAGE 1.0 DEFINITIONS DEFINED TERMS ..... .... 1-1 ACTION * ...... . . .... 1-1 AXIAL FLUX DIFFERENCE ..... 1-1 CHANNEL CALIBRATION CHANNEL CHECK .. . .

.. *. .. .. 1-1 1-1 CHANNEL FUNCTIONAL TEST ...... 1-1 CONTAINMENT INTEGRITY . .... 1-2 CONTROLLED LEAKAGE ... 1-2 CORE ALTERATION ....

DOSE EQUIVALENT I-131 ****

. .. .. .. .. .. .. . . . . . 1-2 1-2

.E-AVERAGE DISINTEGRATION ENERGY ... .. 1-3 ENGINEERED SAFETY FEATURE RESPONSE TIME 1-3 FREQUENCY NOTATION ... .. ....

. . 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 OPERATI6NAL MODE PHYSICS TESTS . .. .. .. .. .. .. .. . . . . . . .. .. . . . . .

PRESSURE BOUNDARY LEAKAGE ***

1-4 1-5 1-5 PROCESS CONTROL PROGRAM (PCP). 1-5 PURGE-PURGING ............. ... 1-5 QUADRANT POWER TILT RATIO RATED THERMAL POWER . . . . . . . . . . . . . . .. .. . .

REACTOR TRIP SYSTEM RESPONSE TIME 1-5 1-5 1-6 REPORTABLE OCCURRENCE .... 1-6 SHUTDOWN MARGIN. .... 1..,6 SITE BOUNDARY ..... ........ ....... 1-6 SOLIDIFICATION * .... . 1-6 SOURCE CHECK * ..... .... ..

1-6 STAGGERED TEST BASIS

1-6 THERMAL POWER . . . ..

UNIDENTIFIED LEAKAGE * . ......

1-7 1-7 UNRESTRICTED AREA . ... ......

. 1-7 VENTILATION EXHAUST TREATMENT SYSTEM ........ 1-7 VENTING . ..... .. .. .

. . . . 1-7

~

SALEM - UNIT 1 I

. INDEX SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SECTION PAGE 2.1. SAFETY LIMITS Reactor Core * .......... ........ 2-1 Reactor Coolant System Pressure 2-1 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 SETTTINGS .

Reactor Trip System Instrumentation Setpoints B 2-3 SALEM - UNIT 1 II

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.0 APPLICABILITY * * * * * * * * * * * * * * * * * * * *

3/4 0-1 3/4.l REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL Shutdown Margin - Tavg > 200°F ** 3/4 1-1 Shutdown Margin - Tavg i 200°F ** 3/4 1-3 Boron Dilution * * * * * * * *

3/4 1-4 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 ..

Boric Acid Transfer Pumps - Shutdown ..

... 3/4 3/4 1-ll 1-12 Boric Acid Transfer Pumps - Operating * ... 3/4 1-13 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 Systems Rod Drop Time * * * * * *

Shutdown Rod Insertion Limit 3/4 3/4 3/4 1-20 1-21 1-22 Control Rod Insertion Limits 3/4 1-23

SALEM - UNIT 1 I II

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 Movable Incore Detectors * * * * * * *

3/4 3-39 Seismic Instrumentation * * * * * * * * ...... 3/4 3/4 3-40 3-43 Meteorological Instrumentation **

Remote Shutdown Instrumentation ** 3/4 3-46 Fire Detection Instrumentation *** 3/4 3-49 Accident Monitoring Instrumentation * * * * * * * * *

3/4 3-53 Radfoactive Liquid Effluent Monitoring Instrumentation. 3/4 3-58 Radioactive Gaseous Effluent Monitoring Instrumentation 3/4 3-64 SALEM - UNIT 1 IV'

--- --------- -------- ----*-~------~-

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION Startup and Power Operation Hot Standby 3/4 3/4 4-1 4-2 Hot Shutdown Cold Shutdown *

.. . .. . . . .. 3/4 3/4 4-3 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 * ..... .. .... 3/4 4-5 3/4.4.4 PRESSURIZER * . . . . ... 3/4 4-6 3/4.4.5 STEAM GENERATORS ... 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 Pressure Isolation Valves ........* . 3/4 4-16a 3/4.4. 7 3/4.4.8 CHEMISTRY * . . . . . . .

SPECIFIC ACTIVITY * . . .

3/4 4-17

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 STRUCTURAL INTEGRITY ASME Code Class 1, 2, and 3 Components ........ 3/4 4-32 SALEM - UNIT 1 v

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 - Tavg ~ 350°F 3/4 5-3 3/4.5.3 ECCS SUBSYSTEMS - Tavg < 350°F 3/4 5-6 3/4.5.4 BORON INJECTION SYSTEM Boron Injection Tank . .... 3/4 5:..7 Heat Tracing ..... 3/4 5-8 3/4.5.5 REFUELING WATER STORAGE TANK ......... 3/4 5-9 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 Containment Ventilation System

3/4 3/4 6-8 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 3/4.6.4 COMBUSTIBLE GAS CONTROL Hydrogen Analyzers ** 3/4 6-18 Electric Hydrogen Recombiners 3/4 6-19 SALEM - UNIT 1 VI

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

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREM~NTS SECTION PAGE 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE Safety Va,l ves .............. 3/4 7-1 Auxiliary Feedwater System 3/4 7-5 Activity . . . . . . . . . . .. . . . .. . . .. .. .

Auxiliary Feed Storage Tank

Main Steam Liner Isolation Valves

3/4 3/4 3/4 7-7 7-8 7-10 3/4.7.2 STEAM GENERATOR PRESSURE/TEMPERATURE LIMITATION * . . . 3/4 7-14 3/4.7.3 COMPONENT COOLING WATER SYSTEM .... 3/4 7-15 3/4.7.4 SERVlCE WATER SYSTEM .. 3/4 7-16 3/4.7.5 FLOOD PROTECTION . . .... 3/4 7-17 3/4.7.6 CONTROL ROOM EMERGENCY AIR CONDITIONING SYSTEM 3/4 7-18 3/4.7.7 AUXILIARY BUILDING EXHAUST AIR FILTRATION SYSTEM 3/4 7-22 3/4.7.8 SEALED SOURCE CONTAMINATION * ... ... 3/4 7-26 3/4.7.9 SNUB BERS ........ ..... 3/4 7-28 3/4.7.10 FIRE SUPPRESSION SYSTEMS Fire Suppression Water System

Spray and/or Sprinkler System

. . . .. . . .. .. .. . 3/4 3/4 7-34 7-37 Low Pressure C02 Systems Fire Hose Stations ... ...

3/4 3/4 7-39 7-40 3/4.7.11 PENETRATI.ON FIRE BARRIERS .... 3/4 7-42

SALEM - UNIT 1 VII

- --- -- - ---- - - - ' - -- ~- --- - *- *- - ~ *- - *-- - ~- ~---* *----;------ ----- -----

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 8-5 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. Di~tribution - Shutdown 3/4 8-13

SALEM - UNIT 1 VIII

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 COMMUNICATIONS

3/4 9-5 3/4.9.6 MANIPULATOR CRANE * ........ 3/4 9-6 3/4.9.7 CRANE TRAVEL - FUEL HANDLING AREA .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 CONTAINMENT PURGE AND PRESSURE-VACUUM RELIEF ISOLATION SYSTEM ****** .... 3/4 9-9 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

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS Concentration * .. ..

. 3/4 11-1 Dose ....... ... . .... 3/4 11-5 Liquid Radwaste Treatment 3/4 11-6 Liquid Holdup Tanks .... 3/4 11-7

3/4.11.2 GASEOUS EFFLUENTS Dose. Rate * ..... ................

. 3/4 11-8 Dose-Noble Gases . ........ .... 3/4 11-12 Dose-Iodine-131, Tritium, and Radionuclides in Particulate Form ........ 3/4 11-13 Gaseous Radwaste Treatment .... .... 3/4 11-14 Explosive Gas Mixture

3/4 11-15 Gas Storage Tanks * .. ........ 3/4 11-16 3/4.11.3 SOLID RADIOACTIVE WASTE 3/4 11-17 3/4.11.4 TOTAL DOSE ...... .. .... 3/4 11-18 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM ..... ..... 3/4 12-1 3/4.12.2 LAND USE CENSUS * . ........ 3/4 12-11 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM ....... 3/4 12-13

- SALEM - UNIT 1 x

BASES INDEX SECTION 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 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR

and AND 3/4.2.3 NUCLEAR ENTHALPY HOT CHANNEL FACTOR

B 3/4 2-~

3/4.2.4 QUADRANT POWER TILT RATIO ** B 3/4 2-5 3/4.2.5 DNB PARAMETERS ***** B 3/4 2-5

SALEM - UNIT 1 XI

INDEX BASES SECTION PAGE 3/4.3 INSTRUMENTATION 3/4.3.1 PROTECTIVE INSTRUMENTATION ~ 3/4 3-)

3/4.3.2 ENGINEERED SAFETY FEATURES (ESF)

INSTRUMENTATION ****** . .. B 3/4 3-1 3/4.3.3 MONITORING INSTRUMENTATION B 3/4 3-1 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION B-3/4 4-1 3/4.4.2 SAFETY VALVES .

\

. B 3-/4 4-la 3/4.4.3 RELIEF VALVES . ...

,. B 3/4 4-la 3/4.4.4 PRESSURIZER

B 3/4 4-2 3/4.4.5 STEAM GENERATORS

B 3/4 4-2 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE B 3/4 4-3 3/4.4.7 CHEMISTRY * * * ** B 3/4 4-4 3/4.4.8 SPECIFIC ACTIVITY

B 3/4 4-5 3/4.4.9 PRESSURE/TEMPERATURE LIMITS ** B 3/4 4-6 3/4.4.10 . STRUCTURAL INTEGRITY B 3/4 4-12 SALEM - UNIT 1. XII

- . ---- **- - ---- . -- --- ~--- - - -*. - . '" . *-. - ----*-* - - - .. -- ..

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 BORON INJECTION SYSTEM ......... B 3/4 5-2 3/4.5.5 REFUELING WATER STORAGE TANK (RWST)

B 3/4 5-2

3/4.6 3/4.6.1 3/4.6.2 3/4.6.3 CONTAINMENT SYSTEMS PRIMARY-CONTAINMENT DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT ISOLATION VALVES *

B 3/4 6:..1 B 3/4 6-3 B 3/4 6-3

. 3/4.6.4 COMBUSTIBLE GAS CONTROL B 3/4 6-4 SALEM - UNIT 1 XIII

INDEX BASES SECTION PAGE 3/4.7 PtANT 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 SYSTE~ *

B 3/4 7-5

3/4.7.7 3/4.7.8 3/4.7.9 AUXILIARY BUILDING EXHAUST AIR FILTRATION SYSTEM SEALED SOURCE CONTAMINATION

SNUBBERS **

B 3/4 7-5 B 3/4 7-5 B 3/4 7-6 3/4.7.10 FIRE SUPPRESSION SYSTEMS B 3/4 7-7 3/4.7.11 PENETRATION FIRE BARRIERS B 3/4 7-7 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A. C. SOURCES -

and AND 3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEMS B 3/4 8-1

SALEM - UNIT 1 XIV

INDEX BASES SECTION 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 9-1 3/4.9.3 DECAY TIME * * .... B 3/4 9-1 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS B 3/4 9-1 3/4.9.5 COMMUNICATIONS ** . .. . . .......

B 3/4 9-1 3/4.9.6 MANIPULATOR CRANE

B 3/4 9-2 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE BUILDING B 3/4 9-2 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION

B 3/4 9-2 3/4.9.9 CONTAINMENT PURGE AND PRESSURE-VACUUM RELIEF ISOLATION SYSTEM * * * * * * * * * * * * * * . .. . . .

. B 3/4 9-2 3/4.9.10 WATER LEVEL - REACTOR VESSEL and AND 3/4.9 .11 STORAGE POOL ********* .... B 3/4 9-3 3/4.9.12 FUEL HANDLING AREA VENTILATION SYSTEM ** B 3/4 9-3 '

3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN **** ................ B 3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS * * &I * * .. 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 1 xv

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 SOLID RADIOACTIVE WASTE

B 3/4 11-6 3/4.11.4 TOTAL*DOSE ******* ... B 3/4 11-7 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM B 3/4 12-1 3/4.12.2 LAND USE CENSUS *

B 3/4 12-1 3/4.12 .3 . INTERLABORATORY COMPARISON PROGRAM B 3/4 12-2,

SALEM - UNIT 1 XVI

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

... 5-1 5-1 Low Population Zone 5.2 CONTAINMENT Configuration . . . . . . . .. . . ..... .... 5-1 5-4 Design Pressure and Temperature 5.3 REACTOR CORE Fuel Assemblies Control Rod Assemblies 5-4 5-4 5.4 REACTOR COOLANT SYSTEM Design Pressure and Temperature ........ 5-::4 Volume * ........... ........ 5-4 5.5 METEOROLOGICAL TOWER LOCATION 5-5 5.6 FUEL STORAGE Criti ca 1ity

5-5 Drainage* ...... 5-5 Capacity ...... 5-5 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT ... .. 5-6 SALEM - UNIT 1 XVII

INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.1 RESPONSIBILITY 6-1 6.c ORGANIZATION Off site 6-1 Facility Staff

6-1 Shift Technical Advisor 6-6 6.3 FACILITY STAFF QUALIFICATIONS 6-6 6.4 TRAINING

6-6 6.5 REVIEW AND AUDIT 6.5.1 STATION OPERATIONS REVIEW COMMITTEE Function ... 6-6 Composition

6-7 Alternates 6-7 Meeting Frequency 6-7 Quorum 6-7 Responsibilities 6-7 Authority

Records 6-8 6..;9 -*

6.5.2 NUCLEAR REVIEW BOARD Function 6-9 Composition 6-9 Alternates 6-9 Consultants

6-9 Meeting Frequency 6-10 Quorum 6-10 Re.view 6-10 Audits 6-10 Authority 6-11 Records

6-11 SALEM - UNIT 1 XVIII

ADMINISTRATIVE CONTROLS INDEX SECTION PAGE 6.5.3 SAFETY REVIEW GROUP .................... 6-12 6.6- -

- REPORTABLE

- - - -OCCURRENCE

- - - -ACTION

-- ................ 6-12 6.7 SAFETY LIMIT VIOLATION * .................. 6-12 6.8 PROCEDURES AND PROGRAMS . . . . . . . . . . . . .* . . . . . 6-13 6.9 REPORTING REQUIREMENTS 6.9.1 ROUTINE REPORTS AND REPORTABLE OCCURRENCES ... 6-15 6.9.2 SPECIAL REPORTS .. .*. 6-21 6.10 RECORD RETENTION ................ 6-21 6.11 RADIATION PROTECTION PROGRAM . . . . . . . . . . . . . . 6-23 6.12 HIGH RADIATION AREA * . . . . . . . . . . . . . . . .

~ 6-23 -*

6.13 PROCESS CONTROL PROGRAM . . . . . . . . . . . . . . . . . 6-24 6.14 OFFSITE DOSE CALCULATION MANUAL . . . . . . . . . . . . . . 6-24 6.15 MAJOR CHANGES TO RADIOACTIVE LIQUID, GASEOUS AND SOLID WASTE TREATMENT SYSTEMS ***** 6-25 6.16 ENVIRONMENTAL QUALIFICATION 6-26

SALEM - UNIT 1 XIX

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 with the necessary range and accuracy to known values of the parameter which the channel monitors. The CHANNEL CALIBRATION shall encompass the entire channel including the senso~ and alarm

and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST. The

.CHANNEL CALIBRATION may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is calibrated.

CHANNEL CHECK 1.5 . A CHANNEL CHECK shall be the qualitative asse$sment 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

,.~*_.,

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 ob being closed by an OPERABLE containment automatic isolation valve system, or

b. Closed by manual valves, blind flanges, or deactivated automatic vlaves secured in their closed positions, except as provided in Table 3.6-1 of 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.

CONTROLLED LEAKAGE 1.8 CONTROLLED LEAKAGE shall be that seal water flow from the reactor coolant pump seals.

CORE ALTERATION 1.9 CORE ALTERATION shall be the movement or manipulation of any component within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position.

DOSE EQUIVALENT I-131 1.10 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries per gram) which altine would produce th~ same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The SALEM - UNIT 1 1-2

DEFINITIONS thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844 "Calculation of Distance Factors for Power and Test Reactor Sites."

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 *

. GASEOUS RADWASTE TREATMENT SYSTEM 1.14 A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed a.nd 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 CONTROLLED LEAKAGE) 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

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.

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 shall be that manual which contains the current methodology and parameters used in the calculation of offsite doses due to 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 mohitoring program.

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 function(s), and when aJl necessary attendant instrumentation, controls, a normal and an emergency electrical power source, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its*function(s) are also capable of performing their related support function(s).

OPERATIONAL MODE - MODE 1.19 An OPERATIONAL MODE (ie., 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

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 steam generator tube 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 shall be that program which contains the current formula, sampling, analyses, test, and determinations to be made to ensure that the 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 Part 20, 10 CFR Part 71 and Federal and State regulations and other requirements governing the

disposal of the 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 3411 MWt *

SALEM - UNIT 1 1-5

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 th~ channel sensor until loss of stationary gripper coil voltage.

REPORTABLE OCCURRENCE 1.27 A REPORTABLE OCCURRENCE sh*all be any of those conditions specified in Specifications 6.9.1.8 and 6.9.1.9.

SHUTDOWN MARGIN 1.28 . SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritica.l or* 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. ass urned to be fully withdrawn *.

SITE BOUNDARY

. L. 29** The: ,SITE BO.UNDARY shal 1 . be that 1ine beyond ~hi ch* the- 1and is-_ not owned~

1eased, or otherwise controlled -by the 1 icensee, as, shown i*n Figure 5.1-3", and which de.fines the exclusion area as* shown in Figure* 5.1-L

SOLIDIFICATION*

1.30 SOLIDIFICATION shall be* the conversion of wet radioactive wastes into a f6rm that meets shipping and burial ground requirements~

SOURCE CHECK 1.31 SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a. source of increased radioactivity.

STAGGERED TEST BASIS 1.32 A STAGGERED TEST BASIS shal 1 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

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 which is not IDENTIFIED LEAKAGE or CONTROLLED 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 instalied 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 confiilement 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

~-~~~~--- ---- --

DEFINITIONS

)

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

1. POWER OPERATION > 0.99 > 5% > 350°F

2. STARTUP > 0.99 < 5% > 350°F

3. HOT STANDBY < 0.99 0 > 350°F

4. HOT SHUTDOWN < 0.99 0 350°F > Tavg

> 200°F

5. COLD SHUTDOWN < 0.99 0 < 200°F

6. REFUELING** < 0.95 0 < 140°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

DEFINITIONS TABLE 1.2 FREQUENCY NOTATION NOTATION FREQUENCY s At 1east once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

D At 1east once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

w At 1east once per 7 days.

M At least once per 31 days.

Ill Q At least once per 92 days.

SA At 1east once per 6 months.

R At 1east 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

INSTRUMENTATION RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.8 The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.3-12 shall be OPERABLE ~ith their alarm/trip setpoints set to ensure that the limits of Specification 3.11.1.1 are not*.exceeded. The alarm/trip setpoints of these channels shall be determined in accordance with the OFFSITE DOSE CALCULATION MANUAL (ODCM).

APPLICABILITY: At all times.

ACTION:

a. With a radioactive liquid effluent monitoring instrumentation channel alarm/trip setpoint less conservative than required by the above specification, without delay suspend the release of radioactive liquid effluents monitored by the affected channel or declare the channel inoperable or change the setpoint so it is acceptably conservative.

b. With less than the minimum number of* radioactive liquid effluent

c.

monitoring instrumentation channels* OPERABLE, take the ACTION shown in Table 3.3-12. Exert best efforts tb return the instrument to operable status within 30 days and, if unsuccessful, explain in the next semi-annual radioactive effluent release report why the inoperability was not corrected in a timely manner.

The provisions of Specifications 3.0.3, 3.0.4, and 6.9.1.9.b 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, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3-12.

SALEM - UNIT 1 3/4 3-58

---*. -- --- --- - -- -; - - -- . - --- ------- *- *- -- ~-- --

TABLE 3.3-12 c

z RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION

-I MINIMUM CHANNELS INSTRUMENT OPERABLE ACTION

1. GROSS RADIOACTIVITY MONITORS. PROVIDING AUTOMATIC TERMINATION OF RELEASE

a. Liquid Radwaste Effluent Line (1-R18) 1 26

b. Steam Generator Blowdown Line 4 27 (1-R19 A, B, C, and D)

-w

..j::>

w I

2. GROSS RADIOACTIVITY MONITORS NOT PROVIDING AUTOMATIC TERMINATION dF RELEASE Ul

\.0 a. Containment Fan Coolers - Service Water Line 5 28 (1-R13 A, B, C, D, E) Discharge

b. Chemical Waste Basin Line (R37) 1 28

3. FLOW RATE MEASUREMENT DEVICES

a. Liquid Radwaste Effluent Line 1 29 .

b. Steam Generator Blowdown Line 4 29

4. TANK LEVEL INDICATING DEVICES .

a. Temporary Outside Storage Tanks as Required 1 30 I, ' I I

ACTION 26 -

TABLE 3.3-12 (Continued)

TABLE NOTATION With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases may continue provided that prior to initiating a release:

a. At least two independent samples are analyzed in accordance with Specification 4.11.1.1.3, and

b. At least two technically qualified members of the Facility Staff independently verify the release rate calculations and discharge line valving; Otherwise, suspend release of radioactive effluents via this pathway.

ACTION 27 - With the number of channe 1s OPERABLE 1ess than required by the Minimum Channels OPERABLE requirement, efflue~t releases via this pathway may continue provided grab samples are analyzed for gross radioactivity (beta or gamma) at a limit of detection of at least l0-7 microcuries/gram:

a. At least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months when the specific activity of the secondary coolant is greater than 0.01 microcuries/gram DOSE EQUIVALENT I-131.

b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the specific activity of the secondary coolant is less than or equal to 0.01 microcuries/

gram DOSE EQUIVALENT I-131.

ACTION 28 - With the number of channels OPERABLE less than required by the Minimum Channels QPERABLE requirement, effluent releases via this pathway may continue provided that, at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , grab samples are collected and analyzed for gross radioactivity (beta or gamma) at a limit of detection of at least 10-7 microcuries/gram.

SALEM - UNIT 1 3/4 3-60

TABLE 3.3-12 (Continued)

TABLE NOTATION ACTION 29 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months duririg actual releases. Pump performance curves may be used to estimate flow.

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 p*rovided the tank liquid level is estimated during all liquid additions to the tank.

SALEM - UNIT 1 3/4 3-61

r

(;')

);:.

['Tl

s: :

TABLE 4.3-12 c

z

-I RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL SOURCE CHANNEL FUNCTIONAL INSTRUMENT CHECK CHECK CALIBRATION TEST

1. GROSS RADIOACTIVITY MONITORS PROVIDING ALARM AND AUTOMATIC TERMINATION OF RELEASE

a. Liquid Radwaste Effluent Line (l-Rl8) D P# R(3) Q(l)

b. Steam Generator Slowdown Line D M R(3) Q(l)

(l-Rl9 A, B, C, and D) w

2. GROSS RADIOACTIVITY MONITORS PROVIDING ALARM

+:> BUT NOT PROVIDING AUTOMATIC TERMINATION OF w

I RELEASE O'l N

a. Containment Fan Coolers - Service Water Line D M R(3) Q(2)

(l-Rl3 A, B, C, D, E) Discharge

b. Chemical Waste Basin Line (R37) D M R(3) Q(2)

3. FLOW RATE MEASUREMENT DEVICES

a. Liquid Radwaste Effluent Line D(4) N.A. R N.A.

b. Steam Generator Slowdown Line D( 4)

N.A. R N.A.

4. TANK LEVEL INDICATING DEVICES**

a. Temporary Outside Storage Tanks as Required . D* N.A

R Q I

'* I '

TABLE 4.3-12 (Continued)

TABLE NOTATION (1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if any of the following conditions exist:

1. Instrument indicates measured levels above the alarm/trip setpoint.

2. Circuit failure. (Loss of Power)

3. Instrument indicates a downscale failure. (Alarm Only)

(2) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exist:

1. Instrument indicates measured levels above the alarm/trip setpoint.

2.

Circuit failure. (Loss of Power)

3. Instrument indicates a downscale failure.

4. Instrument controls not set in operate mode.

(3) The initial CHANNEL CALIBRATION was performed using appropriate liquid or gaseous calibration sources obtained from reputable suppliers. The activity of the calibration sources were reconfirmed using a multi-channel analyzer which was calibrated using one or more NBS standards.

(4) CHANNEL CHECK shall consist of verifying indication of flow during periods of release. CHANNEL CHECK shall be made at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months on days on which continuous, periodic, or batch releases are made.

During liquid additions. to the tank.

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

The R18 channel is an in-line channel which requires periodic decontamination. Any count rate indication above 10,000 cpm constitutes a CHANNEL CHECK for compliance purposes.

SALEM - UNIT 1 3/4 3-63

INSTRUMENTATION RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.9 The radioactive gaseous effluent monitoring instrumentation channels shown in Table 3.3-13 shall be OPERABLE with their alarm/trip setpoints set to ensure that the limits of Specification 3.11.2.1 are not exceeded. The alarm/trip setpoints of these channels shall be determined in accordance with the (ODCM). _)

APPLICABILITY: As shown in Table 3.3-13 ACTION:

a. With a radioactive gaseous effluent monitoring instrumentation channel alarm/trip setpoint less conservative than required by the above

. specification, without delay suspend the release of radioactive gaseous effluents monitored by the affected channel or declare the channel inoperable or change the setpoint so it is acceptably conservative.

b. With less than the minimum number of_ radioactive gaseous effluent monitoring instrumentation channels OPERABLE, take the*ACTION shown in Table 3.3-13. Exert best efforts to return the instrument to ORerable status within 30 days and, if unsuccessful, explain in the next semi-annual radioactive effluent* release report why the inoperability was not corrected in a timely.manner.

c. The provisions of Specifications 3.0.3, 3.0.4, and 6.9.1.9.b are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.9 Each radioactive gaseous effluent monitoring instrumentation channel shal 1 be demonstrated OPERABLE by performance of the CHANNEL CHECK, SO.URGE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3-13 *

- SALEM - UNIT 1 3/4 3-64

/

(/)

)>

r m TABLE 3.3-13

<= RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION

z

--i MINIMUM t--'

CHANNELS INSTRUMENT OPERABLE APPLICABILITY ACTION

1. WASTE GAS HOLDUP SYSTEM

!* a. Noble Gas Activity Monitor - Providing

' Alarm and Automatic Termination of Release 1

31 (1-R41C)

b. Oxygen Monitor 1 ** 35

2. PLANT VENT HEADER SYSTEM#

-w

~

w I

O'I a.

b.

Noble Gas Activity Monitor (1-R16 or 1-R41C)

Iodine Sampler 1

1 33 &34 36 U1

c. Particulate Sampler l

36

d. Flow Rate Monitor 1

32

e. Sampler Flow Rate Monitor 1

32

The following process streams are routed to the plant vent where they are effectively monitored by the instruments described:

(a) Condenser Air Removal System (b) Auxiliary Buildi°ng Ventilation System (c) Fuel Handling. Building Ventilation System (d) Radwaste Area Ventilation System (e) Containment Purges Action item #34 applies to the purging of the containment only.

.1

. I I,

'* I

TABLE 3.3-13 (Continued)

TABLE NOTATION ACTION 31 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, the contents of the tank(s) may be released to the environment provided that prior to initiating the release:

- I

a. At least two independent samples of the tank's contents are I i

analyzed, and

b. At.least two technically qualified members of the Facility Staff independently verify the release rate calculations and discharge valving lineup; Otherwise, suspend release of radioactive effluents via this pathway.

ACTION 32 - With the number of channels OPERABLE less than required by the ,

Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

ACTION 33 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided grab samples are taken at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and these samples are analyzed for gross activity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 34 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, immediately suspend PURGING of radioactive effluents via thi.s pathway.

At all times, other than when the line is valved out and locked.

- During waste gas holdup system operation.

SALEM - UNIT 1 3/4 3-66

TABLE 3.3-13 {Continued)

TABLE NOTATION ACTION 35 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, operation of the waste gas holdup system may continue provided grab samples are collected at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and analyzed within the following 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

ACTION* 36 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via the effected pathway may continue provided samples are continuously collected with auxi.lia~ sampling equipment as required in Table 4.11-Z. .

SALEM - UNIT 1 3/4 3-67

(/)

i: :.

r-IT1 3: TABLE 3.3-13 c:

z RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS

...... CHANNEL MODES IN WHICH

CHANNEL SOURCE CHANNEL FUNCTIONAL SURVEILLANCE INSTRUMENT CHECK CHECK CALIBRATION TEST REQUIRED

1. WASTE GAS HOLDUP SYSTEM

a. Noble Gas Activity Monitor -

Providing Alarm and Automatic Termination of Release (1-R41C) p p R(3) Q(l) *

b. Oxygen Monitor D N.A. Q(4) M **

2. PLANT VENT HEADER SYSTEM#

w

~ a. Noble Gas Activity Monitor (1-R16 D M R(3) Q(2)

w I

or 1-R41C)

CJ) co

b. Iodine Sampler w N.A. N.A. N.A. *

c. Particulate Sampler w N.A. N.A. N.A. *

d. Flow Rate Monitor* D N.A. R N.A. *

e. Sampler Fl ow Rate Monitor w N.A. R N.A. *

The following process streams are routed to.the plant.vent where they are effectively monitored by the instruments described:

(a) Condenser Air Removal System (b) Auxiliary Building Ventilation Sy.stem (c) Fuel Handling Building Ventilation System (d) Radwaste Area Ventilation System (e) Containment Purges I, . I I

TABLE 4.3-13 (Continued)

TABLE NOTATION (1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if any of the following conditions exist:

1. Instrument indicates measured levels above the alarm/trip setpoint.

2. Circuit failure. (Loss of Power)

3. Instrument tndicates a downscale failure. {Alarm Only)

(2) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exist:

1. Instrument indicates measured levels above the alarm/trip setpoint.

2. Circuit failure. (Loss of Power)

3. Instrument indicates'a downscale failure.

4. Instrument controls not set in operate mode *

. (3) The initi~l CHANNEL CALIBRATION was performed using appropriate liquid or gaseous calibration sources obtained from reputable suppliers. The activity of the calibration sources were reconfirmed using a multi-channel analyzer which was calibrated using one or more NBS standards.

(4) The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal:

1. One volume percent oxygen, balance nitrogen, and

2. Four volume percent oxygen, balance nitrogen.

At all times

- During waste gas holdup system operation *

- SALEM - UNIT 1 3/4 3-69

3/4.11 3/4.11~1 RADIOACTIVE EFFLUENTS LIQUID EFFLUENTS CONCENTRATION LIMITING CONDITION FOR OPERATION 3.11.1.1 The toncentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS (See Figure 5.1-3) shall be limited to the concentrations specified in 10 CFR Part 20, Appendix B, Table II, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 2 x 10-4 microcuries/ml.

APPLICABILITY: At all times.

ACTION:

With the concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS exceeding the above limit~, without delay restore the concentration to within the above limits.

SURVEILLANCE REQUIREMENTS 4.11.1.1.1 Radioactive liquid wastes shall be sampled and analyzed according to the sampling and analyses program in Table 4.11-1.

4.11.1.1.2 The results of the radioactivity analyses shall be used in a~cordance with the ODCM to assure that the concentrations at the point of rel~ase are maintained within the limits of Specification 3.11.1.1

SALEM - UNIT 1 3/4 11-1

I TABLE 4.11-1 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM Lower Limit Minimum of Detection Liquid Release Sampling An~lysis Type of Activity (LLD)a Type Frequency Frequency Analysis (uCi /ml)

A. Batch Waste p p Release Each Batch Each Batch Principal Gamma 5xl0-7 Tanksb Emittersc I-131 lxl0-6 p M Dissolve and lxl0-5 One Batch/M Entrained Gases (Gamma Emitters)

- p M H-3 1x1*0-5 Each Batch Composited Gross Alpha lx 10-7 0 p Each Batch Q

Composited Sr-89, Sr-90 Fe-55 5xl0-8 lxl0-6 B. Continuous w Principal Gamma 5x10-7 Releasese Weekly Composite Emittersc

1. Steam I-131 lxl0-6 Gener.

Slowdown M Grab Sample M Dissolved and Entrained Gases lxlQ-5 Weekly M H-3 lxl0-5 Composite Gross Alpha lxl0-7 Weeklyd Q Sr-89, Sr-90 5xl0-8 Composite Fe-55 lxl0-6 SALEM - UNIT 1 3/4 11-2

TABLE 4.11-1 (Continued)

TABLE NOTATION

a. The LLD is defined, for purposes of these specifications as the smallest concentration of radioactive material in a sample that will yield a net count {above system background) that will be detected with 95% probability with only 5% probability* of falsely con*c1 udi ng that a blank observation represents a "real" signal~

For a particular measurement system (which may include radiochemical separation):

4.66 Sb LLD =

E

V

2.22x106

t,;

e (-A.tit)

Where:

LLD is the "a priori 11 lower limit of detection as defined above (as microcuries per unit mass or volume),

Sb is the st~ndard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute),

E is the counting efficiency (as counts per disintegration),

V is the sample size (in units of mass or volume),

2.22 x 106 is the number of disintegrations per minute per microcurie, t,;is the fractional radiochemical yield (when applicable),

A.is the radioactive decay constant for the particular radionuclide, and tit for environmental samples is the elapsed time between sample collection (or end of the sample collection period) and time of counting.

Typical values of E-, V, t,;, and tit should be used in the calculation.

It should be recognized that the LLD is defined as an ~priori (before the fact) limit representing the capability of a measurement system and not as an~ posteriori (after the fact) limit for a particular measurement.

SALEM - UNIT 1 3/4 11-3

TABLE 4.11-1 (Continued)

TABLE NOTATION

b. A batch release is the discharge of liquid wastes of a discrete volume.

Prior to sampling for analyses, each batch shall be isolated, and then thoroughly mixed to assure representative sampling.

c. The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Mn-54, Fe~59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144. This list does not mean that only these nuclides are to be.detected and reported. Other peaks that are measurable and identifiable, together with the above nuclides, shall also be identified and reported.

d. A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen that is representative of the liquids released.

e. A continuous release is the discharge of liquid wastes of a nondiscrete volume, e.g., from a volume of a system that has an input flow during the continuous release.

SALEM - UNIT 1 3/4 11-4

*-*- - -1 ** *-*- --------

RADIOACTIVE EFFLUENTS DOSE LIMITING CONDITION FOR OPERATION 3.11.1.2 The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents rele~sed, from each reactor unit, to UNRESTRICTED AREAS (see Figure 5.1-3) shall be limited:

a. During any calendar quarter to less than or equal to 1.5 mrems to the total body and to less than or equal to 5 mrems to any organ, and

b. During any calendar year to less than or equal to 3 mrems to the totai body and to less than or equal to 10 mrems to any organ.

APPLICABILITY: At all times.

ACTION:

a. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, in lieu of a Licensee Event Report, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit(s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

b. The provisions of Specification 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.1.2 Cumulative dose contributions from liquid effluents shall be determined in accordance with the ODCM at least once per 31 days

SALEM - UNIT 1 3/4 11-5

RADIOACTIVE EFFLUENTS LIQUID RAOWASTE TREATMENT LIMITING CONDITION FOR OPERATION 3.11.1.3 The liquid radwaste treatment system shall be used to reduce the radioactive materials liquid wastes prior to their discharge when the projected cumulative doses due to the liquid effluent from each reactor to UNRESTRICTED

AREAS (see Figure 5.1-3) exceed 0.375 mrem to the total body or 1.25 mrem to any*

organ during any calendar quarter.

APPLICABILITY: At all times.

ACTION:

a. With the radioactive liquid waste being discharged without treatment and in excess of the above limits, in lieu of a Licensee Event Report,

.prepare and submit to the Commission within 30 days pursuant to Specification 6.9.2 a Special Report that includes the following information:

1. Explanation of why liquid radwaste was being discharged without treatment, identification of any inoperable equipment or subsystems, and the reason for the inoperability.

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

3. Summary description of action(s) taken to prevent a recurrence.

b. The provisions of Specification 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS . -

4.11.1.3 Doses due to liquid releases shall be projected at least once per 31 days in accordance with the OOCM.

SALEM - UNIT 1 3/4 11-6

RADIOACTIVE EFFLUENTS LIQUID HOLDUP TANKS*

LIMITING CONDITION FOR OPERATION 3.11.1.4 The quantity of radioactive material contained in each outdoor te~porary 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 materia1 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 and 3.0.4 are not applicable

SURVEILLANCE REQUIREMENTS 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 at least once per 7 days 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

I.

RADIOACTIVE EFFLUENTS 3/4.11.2 GASEOUS EFFLUENTS DOSE RATE LIMITING CONDITION FOR OPERATION 3.11.2.1 The dose rate due to radioactive materials released in gaseous effluents from the site to areas at and beyond the SITE BOUNDARY (see Figure 5.1-3) shall be limited to the following:

a. For noble gases: Less than or equal to.500 mrems/yr to the total body and less than or equal to 3000 mrems/yr to the skin, and

b. For iodine-131, for tritium, and for all radionuclides in particulate form with half lives greater than 8 days: Less than or equal to 1500 mrems/yr to any organ.

APPLICABILITY: At all times.

ACTION:

. With the dose rate(s)- exceeding the above limits, without delay restore the release rate to within the above limit(s).

SURVEILLANCE REQUIREMENTS 4.11.2.1.1 The dose rate due to noble gases in gaseous effluents shall be determined continuously to be within the above limits in accordance with the ODCM.

4.11.2.1.2 The dose rate due. to iodine-131, tritium, and all radionuclides in particulate form with half lives greater than 8 days in gaseous effluents shall be determined to be within the above limits in accordance with the ODCM by obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified in Table 4.11-2.

. SALEM - UNIT 1 3/4 11-8

TABLE 4.11-2 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM Lower Limit Minimum of Detection Gaseous Release Sampling Analysis Type of Activity {LLD)

Type Frequency Frequency Analysis (uCi /ml)

A. Waste Gas p* p Storage Each Tank Each Tank Principal Gamma lxl0-4

i. Tank Grab Sample Emittersb B. Containment p p Principal Gamma lxl0-4 PURGE Each PURGE Each PURGE Emittersb Grab Sample H-3 lxl0-6 Principal Gamma lxl0-4 C. Plant Vent MC d e MC Emitterb

' ' Sample Grab H-3 lxl0-6 D. All Release Continuousf wg I-131 lxlQ-12 Types as Charcoal Listed in A, Sample B, and c Above Continuousf wg Principal Gamma 1-10-11 Particulate Emittersb Sample ( I-131, Others)

Continuousf M Gross Alpha lxl0-11 Composite Pa rt i cu 1ate .

Sample Continuousf Q Sr-89, Sr-90 lxl0-11 Composite -

Particulate Sample Continuousf Noble Gas Noble Gasses lxl0-6 Monitor Gross Beta or Gamma SALEM - UNIT 1 3/4 11-9

TABLE 4.11-2 {Continued)

TABLE NOTATION

a. The LLD is defined in Table 4.11.1

b. The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58.,

Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144 for particulate emissions: This list does not mean that only these nuclides are to be detected and reported. Other peaks that are measurable *and identifiable, together with the above nuclides, shall also be identified and reported.

c. Sampling and analysis shall also be performed following shutdown, startup or a THERMAL POWER change exceeding 15 percent of RATED THERMAL POWER within one hour unless:

1. Analysis shown that the DOSE EQUIVALENT I-131 concentrations in the primary coolant has not increased more than a factor of three.

2. The noble gas activity monitor shows that effluent activity has not increased by more than a factor of three above the monitor setpoint.

d. Tritium grab samples shall be taken at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the refueling canal is flooded.

e. Tritium grab samples shall be taken at least once per 7 days from the ventilation exhaust from the spent fuel pool area whenever spent fuel is in the spent fuel pool.

f *. The ratio of the sample flow rate to the sampled stream flow rate shall be known fo~ the time period covered by each dose or dose rate calculation made in accordance with Specifications 3.11.2.1, 3.11.2.2 and 3.11.2.3.

SALEM - UNIT 1 3/4 11-10

. j-TABLE 4.11-2 (Continued)

TABLE NOTATION

g. Samples shall be changed at least once per .7 days and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after changing (or after removal from sampler).

Sampling shall also be performed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for at least 7 days following each shutdown, startup or THERMAL POWER change exceedi.ng 15 percent of 'RATED THERMAL POWER in one hour and analyses shall be completed for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months are analyzed, the corresponding LLDs may be increased by a factor of 10. This requirement does not apply if (1) analysis shown that the DOSE EQUIVALENT I-131 concentration in the primary coolant has not increased more than a factor of 3; and (2) the noble gas monitor shows that effluent activity has not increased more than a factor of 3.

SALEM - UNIT 1 3/4 11-11

RADIOACTIVE EFFLUENTS DOSE - NOBLE GASES LIMITING CONDITION FOR OPERATION 3.11.2.2 The air dose due to noble gases released in gaseous effluents, from each reactor unit, from the site areas at and beyond the SITE BOUNDARY (see Figure 5.1-3) shall be* limited to the following:

a. During any calendar quarter: Less than or.equal to 5 mrad for gamma radiation and less than or equal to 10 mrad for beta radiation and,

b. During any calendar year:* Less than or equal to 10 mrad for*gamma

radiation and less than or equal to 20 mrad for beta radiation.

APPLICABILITY: At all times.

ACTION:

a. With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, in lieu of a Licensee Event Report, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit(s) and defines the corrective actions that have been taken to reduce the release and the proposed corrective actions to be taken to assure that subsequent releases will be in

. compliance with the above limits.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.2 Cumulative dose contributions for the current calendar quarter and current calendar year shall be determined in accordance with the ODCM at least once per 31 days.

SALEM - UNIT 1 3/4 11-12

RADIOACTIVE EFFLUENTS DOSE - IODINE-131, TRITIUM, AND RADIONUCLIDES IN PARTICULATE FORM LIMITING CONDITION FOR OPERATION J.11.2.3 The dose to a MEMBER OF THE PUBLIC from iodine-131, from tritium, and from all radionuclides in particulate form with half-l*ives greater than 8 days in gase*ous effluents released, from each reactor unit, from the site to areas at and beyond the SITE BOUNDARY (see Figure 5.1-3) shall be limited to the following:

a. During any* calendar quarter: Less than or equal to 1:s'mrems to any organ and,*

b. During any calendar year: Less than or equal to 15 mrems to any organ.

APPLICABILITY: At all times.

ACTION:

a. With the calculated are dose from the release of iodine-131, tritium, and radionuclides in particulate form with half~lives greater than 8 days, in gaseous effluents exceeding any of the above limits, in lieu of a Licensee Event Report, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit and defines the corrective actions that have been taken to reduce the release and the proposed corrective actions to be take~ to assure that subsequent releases will be in compliance with the above limits.

b. The provisions of Specifications 3.0.3 and 3.0~4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.3 Cumulative dose contributiQns for the current calendar quarter and current calendar year for iodine-131, tritium, and radionuclides in particulate form with half-lives greater than 8 days shall be determined in accordance with the ODCM at least once per 31 day.s.

SALEM - UNIT 1 3/4 11-13

RADIOACTIVE EFFLUENTS GASEOUS RADWASTE TREATMENT LIMITING CONDITION FOR OPERATION 3.11.2.4 The GASEOUS RADWASTE TREATMENT SYSTEM and the VENTILIATION EXHAUST TREATMENT SYSTEM shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected gaseous* effluent air doses due to gaseous effluent releases, from the site to areas at and beyond the SITE BOUNDARY (see Figure 5.1-3) exceed 0.625 mrad for gamma radiation and 1.25 mrad for beta radiation in any calendar quarter. The VENTILIATION EXHAUST TREATMENT SYSTEM shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected doses due to gaseous effluent releases, from each reactor unit, from the site to areas at and beyond the SITE BOUNDARY (see Figure 5.1-3) would exceed 1.875 mrem to any organ in any calendar quarter.

APPLICABILITY: At all times.

ACTION:

a. With gaseous waste being discharged without treatment and in excess of the above limits, in lieu of a Licensee Event Report, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that includes the following information:

1. Explanation of why gaseous radwaste was being discharged without treatment, identification of any inoperable equipment or subsystems, and the reason for the inoperability,

2. Action(s) taken to restore the inoperable equipment equipment to OPERABLE status, and 3.* Summary description of action(s) taken to prevent a recurrence.

b. The provisions of Specifications 3.0.4 are not applicable. -

SURVEILLANCE REQUIREMENTS 4.11.2.4 Doses due to gaseous releases from the site shall be projected at least once per 31 days in accordance with the ODCM.

~.': .

SALEM - UNIT 1 3/4 11-14

RADIOACTIVE EFFLUENTS EXPLOSIVE GAS MIXTURE LIMITING CONDITION FOR OPERATION 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 whenever the hydrogen concentration exceeds 4% 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 4% by volume but less than or equal 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 and the hydrogen concentration greater than 2% by volume, immediately suspend all additions of waste gases to the system and reduce the concentration*of oxygen to less than or equal to 23 by volume without delay.

c. The provision of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.5 The concentrations of hydrogen and 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 required OPERABLE by Table 3.3-13.

SALEM - UNIT 1 3/4 11-15

RADIOACTIVE EFFLUENTS GAS STORAGE TANKS LIMITING CONDITION FOR OPERATION 3.11.2.6 The quantity of radioactivity contained in each gas storage tank shall be limited to 36,000 curies noble gases (considered as Xe-133).

APPLICABILITY: At all times.

ACTION:

a. With the quantity of radioactive material in any gas storage tank exceeding the above limit, immediately suspend all additions of radioactive material to the tank and within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduce the tank contents to within the limit.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS

- 4.11.2.6 The,quantity of radioactive material contained in each gas storage tank shall be deter.mined to be within the above limit at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when radioactive materials are being added to the tank.

SALEM - UNIT 1 3/4 11-16

RADIOACTIVE EFFLUENTS SOLID RADIOACTIVE WASTE LIMITING CONDITION FOR OPERATION 3.11.3. The solid radwaste system shall be used in accordance with a PROCESS 60NTROL PROGRAM to process wet radioactive waste to meet shipping and burial ground requirements.

APPLICABILITY: At all times.

ACTION:

a. With the prov1s1ons of the PROCESS CONTROL PROGRAM not satisfied, suspend shipments of defectively processed or defectively packaged solid radioactive wastes from the site. *

b. The provisions of Specifications 3.0.3 and 3.0~4, and 6.9.1.9.b are not applicable.

SURVEILLANCE.REQUIREMENTS 4.11.3. The PROCESS CONTROL PROGRAM shall be used to verify the SOLIDIFICATION of at least one representative test specimen from at least every tenth batch of each type of wet radioactive waste*(e.g., filter sludges, spend resins, evaporator bottoms, boric acid solutions, and sodium sulfate solutions).

a. If any test specimen fails to verify SOLIDIFICATION, the SOLIDIFICATION of the batch under test shall be suspended until such time as additional test specimens can be obtained, alternative SOLIDIFICATION parameters can be determined in accordance with the PROCESS CONTROL PROGRAM, and a subsequent test verifies SOLIDIFICATION. SOLIDIFICATION of the batch may then be resumed using the alternative SOLIDIFCATION parameters determined by the PROCESS CONTROL PROGRAM.

b. If the initial test specimen from a batch of waste fails to verify SOLIDIFICATION, the PROCESS CONTROL PROGRAM shall provide for the collection and testing of representative test specimens from each consecutive batch of the same type of wet waste until at least three consecutive initial test specimens demonstrate SOLIDIFICATION. The PROCESS CONTROL PROGRAM shall be modified as required, as provided in Specification 6.13, to assure SOLIDIFICATION of subsequent batches of

. waste.

SALEM - UNIT 1 3/4 11-17

RADIOACTIVE EFFLUENTS 3/4.11.4 TOTAL DOSE LIMITING CONDITION FOR OPERATION 3.11.4 The annual (calendar year) dose or dose commitment to any MEMBER OF THE PUBLIC, due to releases of radioactivity and radiation, from urnaium fuel cycle sources shall be 1imited to 1ess than or equal to 25 mrems to the total body or any organ (except the thyroid, which shall be limited to less than or equal to 75 mrems).

APPLICABILITY: At all times ACTION:

a. With the calculated*dcises from the release of radioactive mate~ials in liquid or gaseous effluents exceeding twice the limits of Specification 3.ll.1.2a, 3.11.1.2b, 3.11.2.2a, 3.11.2.2b, 3.11.2.3a, or 3.ll.2~3b, calculations should be made including direct radiation contributions from the reactor units and from outside storage tanks to determine whether the limits of this Specification have been exceeded. If such is the case in liel.i of a Licensee*Event Report, prepare and submit to the Commission with-in- 30 days, pursuant to Specification 6.9.2, a Special Report that defines the corrective action to be taken to redu.ce subsequent releases to prevent recurrence of exceeding the above limits and includes the schedule for achieving conformance with the above limits. This-Specical Report, as defined in 10 CFR Part 20.405c, shall include ~n analysis. that estimates the radiation exposure (dose) to a MEMBER OF THE PUBLIC from uranium fuel cycle sources, including all

effluent pathways and direct radiation, for the calendar year that includes the release(s) covered by this report. It shall also describe levels of .radiation and concentrations of radioactive material involved, and the cause of the exposure levels or concentrations. ,If the estimated dose(s) exceeds the above limits, and if the release condtion resulting in violation of 40 CFR Part 190 has not already been corrected, the Special Report shall include a request for a variance in accordance with the provisions of 40 CFR Part 190. Submittal of the report is considered a timely request, and a* variance is granted until staff action on the request is complete.

SALEM - UNIT 1 3/4 11-18

RADIOACTIVE EFFLUENTS 3/4.11.4 TOTAL DOSE LIMITING CONDITION FOR OPERATION ACTION: (Cont 1 d)

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.4.1 Cumulative dose contributions from liquid and gaseous effluents shall be determined in accordance with Specifications 4.11.1.2, 4.11.2.2, 4.11.2.3, and in accordance with the ODCM.

4.11.4.2 Cumulative dose contributions from direct radiation from the reactor units and from radwaste storage shall be determined in accordance with the ODCM.

- I SALEM - UNIT 1 3/4 11-19

3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM LIMITING CONDITION FOR OPERATION 3.12.1. The radiological environmental monitoring program shall be conducted as specified in Table 3.12-1.

APPLICABILITY: At all times.

ACTION:

a. With the radiological environmental monitoring program not being conducted as specified in Table 3.12-1, in lieu of a Licensee Event Report, prepare and submit to the Commission, in the Annual Radiological Environmental Operating Report required by Specification 6.9.1.11, a description of the reasons for not conducting the program as r~quired and the plans for preventing a recurrence.

b. With the-level of* radioactivity as the r_esult of plant effluents in an

environmental sampling medium at a specified location exceeding the reporting levels of Table 3.12-2 when averaged over any calendar

quarter; in lieu of a Licensee Event Report, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit(s) and defines the corrective actions to be taken to reduce radioactive effluents so that the potential annual dose to a MEMBER OF THE PUBLIC is less than the calendar year limits of Specificat~ons 3.11.1.2, 3.11.2.2, and 3.11.2.3. When more than one of the radionuclides in Table 3.12-2 are detected in the sampling medium, this report shall be submitted if:

concentration (1) concentration (2).

+ + *** >1.0 reporting level (1) reporting level (2)

When radionuclides other than those in Table 3.12-2 are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose to a member of the public is equal to or greater SALEM - UNIT 1 3/4. 12-1 .

3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM LIMITING CONDITION FOR OPERATION ACTION: (Cont'd) than the calendar year limits of Specifications 3.11.1.2, 3.11.2.2, and 3.11.2.3. This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report.

c. With milk or fresh leafy vegetable samples unavailable from one or more of the sample locations required by Table 3.12-1, identify locations for obtaining replacement samples and add them to the radiological environmental monitoring program within 30 days. The specific locations from which samples were unavailable may then be deleted from the monitoring program.

In lieu of a Licensee Event Report and pursuant to Specification .

6.9.1.11, identify the cause of the unavailab1lity of samples and the new location(s) for obtaining replacement samples in the next Semiannual Radioactive Effluent Release Report. Include in the report a revised figure(s) and table for the ODCM reflecting the new location(s).

d. The provisions of Specifications.3~0.3 and 3.0.4 ar~ not applicable.

SURVEILLANCE REQUIREMENTS 4.12.1 The radiological environmental monitoring samples shall be collected pursuant to Table 3.12-1 from the locations specified in the ODCM and shall be analyzed pursuant to the requirements of Table 4.12-1.

SALEM - UNIT 1 3/4 12-2

TABLE 3.12-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM c:: Number of Samples

z and Sampling and Type and frequency

-I Pathway Sample Locationsa Collection Frequencya of Analysis

1. DIRECT RADIATIONb About 40 routine monitoring Monthly, Quarterly, Gamma dose monthly, stations with two or more or semi-annually quarterly, or semi-dosimeters for measuring annually dose continuously to be placed as follows: 1) an inner ring of stations in the general area of the Site and an outer ring in the 2- to 8-km range from the Site with a station in each sector of each ring.

The balance of the stations

-w

~

N I

should be placed in special interest areas such as population centers, nearby residences, schools, and in 2 or 3 areas to w

serve as control stations.

2. AIRBORNE Radioiodine and Samples from 5 locations: Continuous sampler Particulates operation with sample

a. 3 samples from close to collection weekly or the Site locations. as required by dust Radioiodine Cannister:

loading which ever is I-131 analysis weekly.

I ~ b. 1. sample from the vicinity more frequentc of a community. Particulate Sampler Gross beta radioacti-

c. 1 sample from a control vity analysis follow-location 15-30 km distance. ing filter change; Gamma isotopic analysis of complete (by location)

~ ua rte r I y *

. I

(/)

);:.

I . I TABLE 3.12-1 (Cont'd) rr1 3:

c:

z

,_.. Num~er of Samples

-I and Sampling and Type and frequency I-'

Pathway Sample Locationsa Collection Frequencya of Analysis

3. WATERBORNE

- ' Gamma isotopic analy-

a. Surfaceg a. 1 sample upstreaming and Two gallon grab sample

b. 1 sample downstream collected monthly. sis quarterly. Compo-

c. 1 sample outfall site for tritium

d. 1 sample cross-stream analysis quarterly.

b. Ground Samples from 1 or 2 sources Two gallon grab sample Gamma isotopic analy-only if likely to be affectedj. collected quarterly. sis quarterly. Compo-site for tritium

- w

~

I-'

N I

c. Drinking a. 1 sample of the nearest

water supply.

50 ml allquot taken daily and composited to analysis quarterly.

I-131 analysis on each composite when the

~

a monthly sample of two dose calculated for gallons. the consumption of the water is greater than 1 mrem per year.

Composite for tritium analysis quarterly.

d. Sediment a. 1 sample downstream Samples taken semi- Gamma isotopic analy-in river b. 1 sample cross-stream annually. sis semiannually.

c. 1 sample outfall I, ' I I I

(./l

):>

r rn 3:

TABLE 3.12-1 (Cont'd) c z

...... Number of Samples

--i

...... and Sampling and Type and frequency Pathway Sample Locationsa Collection Frequencya of Analysis

4. INGESTION

a. Milk a. Samples from milking animals Semimonthly when Gamma isotopic and in 3 locations within animals are on pasture, I-131 analysis semi-5 km distance having the monthly at other monthly when animals highest dose potential. time. are on pasture; If there are none, then, 1 monthly at other sample from milking animals times.

5 to 8 km distant where doses are calculated to be w greater than 1 mrem per

+>- yrk.

N U1 I

b. 1 sample from milking animals at a control location.

b. Fish and a. 1 sample of each commerci- Sample in season, or Gamma isotopic analy-Invertebrates ally and recreationally semiannual if they sis on edible.

important species in vicinity are not seasonal portions.

of discharge point.

c. Food a. 1 sample of each principal At time of harvestl Gamma isotopic analy-class of food products from sis on edible any area that is irrigated portion.

by water in which liquid plant wastes have been dis-charged.

I, I I I

TABLE 3.12.1 (Cont'd)

TABLE NOTATION a In the event that it is not possible or practicable to obtain samples of the media of choice at the most desired location or time, suitable alternative media and locations may be chosen for the particular pathway. Actual locations (distance and direction) from the site shall be provided for all sample. locations in the ODCM.

b One or more ins~ruments, such as pressurized ion chamber, for measuring and .

recording dose rate continuously may be used in place of, or in addition to, integrating dosimeters. For the purposes of this table, a thermoluminescent dosimeter may be considered to be one phospor, and two or more phosphors in a packet may be considered as two or more dosimeters. Film badges shall not be used for measuring direct radiation. The 40 stations is not an absolute number. This number may be reduced according to geographical limitations; e.g., at an ocean site, some sectors will be over water so that the number of dosimeters may be reduced accordingly. An option open to licensees is to place some of .the 40 routine TLD monitoring stations inside the SITE BOUNDARY. Such stations could provide useful information under both routine and accident conditions, and would be particularly*valuable for the larger sites. The frequency of analysis or readout will depend upon the characteristics of the specific TLD system used and should be selected to obtain optimum dose information with minimal fading.

c Canisters for the collection of radioiodine in air shall be checked for channeling before operation in the field to insure complete recovery of iodine.

d Particulate sample filters shall be analyzed for gross beta radioactivity 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or more after sampling to all ow for radon and thoron daughter product decay. If gross beta activity in air or water is greater than ten times the yearly mean of control samples for any medium, gamma isotopic analysis shall be performed on the individual samples.

SALEM - UNIT 1 3/4 12-6

TABLE 3.12-1 (Cont 1 d)

TABLE NOTATION e Gamma isotopic analysis means the identification and quantification of gamma-emitting radionuclides that may be attributable to the effulents from the facility.

f The purpose of this sample is to obtain background information. If it is not practical *to establish control locations in accordance with the distance and wind direction criteria, other sites that provide valid background data may be substituted.

g The 11 upstream sample 11 shall be taken at a distance beyond significant influence of the discharge. The 11 downstream 11 sample shall be taken in an area beyond but near the mixing zone. 11 Upstream 11 samples in an estuary must be taken far enough upstream to be beyond the plant influence.

h Salt water shall be sampled only when the receiving water is utilized for recreational activities*.

i Composite samples shall be collected with equipment that is capable of collecting an aliquot at time intervals that are very short (e.g., hourly) relative to the compositing period (e.g., monthly) in order to assure obtaining a representative sample

j Groundwater samples shall be taken when this source is tapped for drinking or irrigation purposes in areas where the hydraulic gradient or recharge properties are suitable for contamination.

k The dose shall be calculated for the maximum organ and age group, using the methodology contained in Regulatory Guide 1.109, Rev. 1, October 1977, and the_

actual parameters particular to the site.

1 If harvest occurs more than once a year, sampling shall be performed during each discrete harvest. If harvest occurs continuously, sampling shall be monthly. Attention shall be paid to including samples of tuberous and root food products.

SALEM - UNIT 1 3/4 12-7

TABLE 3.12-2 REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES c

z

,__. Reporting Levels

--I Water Airborne Particulate Fish Milk Food Products Analysis {pCi/1) or Gases {pCi/m3) {pCi/Kg, wet) {pCi/1) {pCi/Kg, wet)

H-3 2 x 104( a)

Mn-54 1 x 103 3 x 104 Fe-59 4 x 102 1 x 104 Co-58 1 x 103 3 x 104 Co-60 3 x 102 1 x 104 N Zn-65 3 x 102 2 x 104 I

00 Zr-Nb-95 4 x 102 I-131 2 0.9 3 1 x 102 Cs-134 30 10 1 x. 103 60 1 x 103 Cs-137 50 20 2 x 103 70 2 x 103 Ba-La-140 2 x 102 3 x 102 (a) For drinking water samples. This is 40 CFR Part 141 value.

~

I, ' I I I

i TABLE 4.12-1 MAXIMUM VALUES FOR THE LOWER LIMITS OF DETECTION (LLD) c 2

-l Water Airborne Particulate Fish Milk Food Products Sediment Analysis (pCi/1) or Gases (pCi/m3) (pCi/Kg, wet) (pCi/1) (pCi/Kg, wet) (pCi/Kg, dry) gross beta 4 1 x 10-2 H-3 2000 Mn-54 15 130 Fe-59 30 260 Co-58, 60 15 130 I-'

N Zn-65 30 260 I

l.O Zr-Nb_95 15 I-131 1 7 x 10-2 1 60 Cs-136 15 5 x 10-2 130 15 60 150 Cs-137 18 6 x 10-2 150 18 80 180 Ba-La-140 15 15

' I I

'* I

TABLE 4.12-1 (Cont'd)

TABLE NOTATION a Required detection capabilities for thermoluminescent dosimeters used for environmental measurements are given in Regulatory Guide 4.13, Rev. 1, July 1977.

b The LLD is defined in Table 4.11-1.

SALEM - UNIT 1 3/4 12-10

RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.2 *LAND USE CENSUS LIMITING CONDITION FOR OPERATIOK 3.12.2. A land use census shall be conducted and shall identify within a distance of 8 km (5 miles) the location in each of the 16 meteorological sectors of the nearest milk animal,,, the nearest residence and the nearest garden* of greater than 50 m2 (500 ft~) producing broad leaf vegetation. (For elevated releases as defined in Regulatory Guide 1.111, Revision 1, July 1977, the land use census shall also identify within a distance of 5 km (3 miles) the locations in each of the 16 meteorological sectors of all milk animals and all* gardens of greater than 50 m2 producing broad leaf vegetation. -

APPLICABILITY: At all times.

ACTION:

a. With a land use fensus identifying a location(s) that yields a calculated dose or dose commitment greater than the values currently being calculated in Specification 4.11.2.3, in lieu of a Licensee Event Report, identify the new 1ocati on (s) in the next Semi annua 1 Radioactive Effluent Release Report, pursuant to Specification 6.9.1.12.

b. With a land use census identifying a location(s) that yields a calculated dose or dose commitment (via) the same exposure pathway) 20 percent greater than at a location from which samples are currently being obtained in accordance with Specification 3.12.1, add the new loc~tion(s) to the radiological environmental monitoring program within 30 days. The sampling location(s), excluding the control station location, having the lowest calculated dose or dose commitment(s) (via the same exposure pathway) may be deleted from this monitoring program after (Obtober 31) of the year in which this land use census was conducted. In lieu. of a Licensee Event Report and pursuant to Specification 6.9.1.12, identify the new location(s) in the next Semiannual Radioactive Effluent Release Report and also include in the report a revised figure(s) and table for the ODCM reflecting the new location(s).

c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

Broad leaf vegetation sampling of at least three different kinds of vegetation may be performed at the SITE BOUNDARY in each of two different direction sectors with the highest predicted D/Qa vegetation sampling lin Table *3.12-1.4c shall be followed, including analysis of control samples.

SALEM - UNIT 1 3/4 12-11

RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.2 LAND USE CENSUS (Cont'd)

SURVEILLANCE REQUIREMENTS 4.12.2 The land use census shall be conducted during the growing season at least once per 12 months using that information that will provide the best results, such as by a door-to-door survey, aerial survey, or by consulting local, agriculture authorities. The results of the land use cen~us shall be included in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.10 SALEM - UNIT 1 3/4 12-12

RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.3. INTERLABORATORY COMPARISON PROGRAM LIMITING CONDITION FOR OPERATION 3.12.3 Analyses shall be performed on radioactive materials supplied as part of an Interlaboratory Comparison Program which has been approved by the Commission.

APPLICABILITY: At an times.

ACTION:

a. With analyses not being performed as required above, report the corrective actions taken to prevent a recurrence to the Commission in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.10.

b. The provisions of Specifications 3.0.3 and 3.0.4. are not applicable.

SURVEILLANCE REQUIREMENTS 4.12.3 A summary of the results obtained as part of the above required Interlaboratory Comparison Program and in accordance with the ODCM shall be included in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1~10.

SALEM - UNIT 1 3/4 12-13

INSTRUMENTATION BASES 3/4.3.3.6 FIRE DETECTION INSTRUMENTATION OPERABILITY of the fire detection instrumentation ensures that adequate warning capability is available for the prompt detection of fires. This capability is required in order to detect and locate fires in their early stages. Prompt de-tection of fires will reduce the potential for damage to safety related equip-ment and is an integral element in the overall facility fire protection program.

In the event that a portion of the fire detection instrumentation is inoperable, the establishment of frequent fire patrols in the affected areas is required to provide detection capability until the inoperable instrumentation is restored to OPERABILITY.

3/4.3.3.7 ACCIDENT MONITORING INSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures that suffi-cient information is available on selected plant parameters to monitor and assess these variables following an accident. This capability is consistent with the Recommendatio*ns of Regulatory Guide 1.97, "Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident, December 1975.

11 3/4.3.3.8 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION The radioactive liquid. effluent instrumentation is provided to monitor and con-trol, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of .liquid effluents. The alarm/trip set-points for these instruments shall be calculated and adjusted in accordance with the procedures in the ODCM to ensure that the alarm/trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this instru-mentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50. The purpose of tank level indicating devices is to assure the detection and control of leaks that if not controlled --

could potentially result. in the transport of radioactive materials to UNRESTRICTED AREAS.

3/4.3.3.9 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION The radioactive gaseous effluent instrumentation is provided to monitor and con-trol, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The alarm/trip set-points for these instruments shall be calculated and adjusted in accordance with the procedures in the ODCM to ensure that the alarm/trip will occur prior to exceeding the limits of 10 CFR Part 20. This instrumentation also includes pro-visions for monitoring (and controlling) the concentrations of potentially explosive gas mixtures in the waste gas holdup system. The OPERABILITY and use of this instrumentatio~ is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50 *

SALEM - UNIT 1 B 3/4 3-3

3/4.11 RADfOACTIVE EFFLUENTS BASES 3/4.11.1 LIQUID EFFLUENTS 3/4.11.1.1 CONCENTRATION The specification is provided to ensure that the concentration of radioactive materials released in liquid waste effluents will be less than the concentration levels specified in 10 CFR Part 20, Appendix B Table II, Column 2. This limitation provides additional assurance that the levels of radioactive materials in bodies of water in UNRESTRICTED AREAS will result in exposures within (1) the Section II.A design objectives of Appendix I, 10 CFR Part 50, to a MEMBER OF THE 'PUBLIC and (2) the limits of 10 CFR Part 20.106(a) to the population. The concentration limit for dissolved or entrained noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2.

The required detection capabilities for radioactive materials in liquid waste samples are tabulated in terms of the lower limits of detection (LLDs) 3/4.11.1.2 DOSE This specification is provided to implement the requirements of Sections II.A, III.A, and IV.A of Appendix I. 10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section II.A of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth *in Section IV.A of Appendix I to assure that the releases of radioactive material in liquid effluents will be kept 11 as low as is reasonably achievable." Also, for freshwater sites with drinking water supplies that can be potentially affe.cted by pl ant ope rat i ans, there *is reasonable assurance that the operation of the facility will not result in radionuclide concentrations in the finished drinking water that are in excess of SALEM - UNIT 1 B 3/4 11-1

RADIOACTIVE EFFLUENTS BASES the requirements of 40 CFR Part 141. The dose calculations in the ODCM

implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the ODCM for calculating the*doses due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual does to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Complianc~ with 10 CFR Part 50, Appendix I, 11 Revision 1, October 1977 and Regulatory Guide 1.113, "Estimating Aquatic Dispersion of Effluents from Accidential and Routine Reactor Releases for the Purposes of Implementing Appendix I, 11 April 1977.

The specification applies to the release of liquid effluents from each reactor

at the site. For units with shared radwaste treatment systems, the liquid effluents from the shared system are proportioned amoung the units sharing that system.

3/4.11.1.3 LIQUID RADWASTE TREATMENT The requirement that the appropriate portions of this system be used, when specified, provides assurance that the releases of radioactive materiali in liquid effluents will be kept 11 as low as is reasonably achievable". This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to to CFR Part 50 and the design objective given in Section II.O of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the liquid radwaste treatment system were specified as. a suitable fraction of the dose design objectives set forth the Section II.A of Appendix I, 10 CFR Part 50, for liquid effluents.

3/4.11.1.4 LIQUID HOLDUP TANKS The tanks listed in this specification include all those outdoor 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 B 3/4 11-2

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

RADIOACTIVE EFFLUENTS BASES Restricting the quantity of raqioactive meterial contained in the specified tanks provides assurance that in the event of an uncontrolled release of the tanks' contents, the resulting concentrations would be less than the limits of 10 CFR Part 20, Appendix B, Table II, Column 2, at the nearest potable water supply and the nearest surface water supply in an UNRESTRICTED AREA.

3/4.11.2 GASEOUS EFFLUENTS 3/4.11.2.1 DOSE RATE This specification is provided to ensure that the dose at any time at and beyond the SITE BOUNDARY from gaseous effluents from all units on the site will be within the annual dose limits of 10 CFR Part 20. The annual dose limits are the doses associated with the concentrations of 10 CFR Part 20, Appendix B, Table II, Column 1. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of a MEMBER OF THE PUBLIC either within or outside the SITE BOUNDARY, to annual average concentrations exceeding the limits specified in Appendix B, Table II of 10 CFR Part 20 [13 CFR Part 20.106(b)]. For MEMBERS OF THE PUBLIC who may at times be within the SITE BOUNDARY, the occupancy of the individual will usually be. sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the SITE BOUNDARY. Examples of calculations for such MEMBERS OF THE PUBLIC with the appropriate occupancy factors shall be given in the ODCM. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to a MEMBER OF THE PUBLIC at or beyond the SITE BOUNDARY to less than or equal to 500 mrems/year to the skin. These release rate limits also restrict, at all times, the corresponding thyroid dose rate above background to a child via the inhalation pathway to less than or equal to 1500 mrems/year. --

This specification applies to the release of gaseous effluents from.all reactors at the site.

SALEM - UNIT 1 B 3/4 11-3

RADIOACTIVE EFFLUENTS BASES 3/4.11.2.2 DOSE - NOBLE GASES This specification is provided to implement the requirements of Section II.B, III.A and IV.A of Appendix I, 10 CFR Part 50. The limiting condition for operation implements the guides set forth in Section II.8 of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in gaseous effluents will be kept 11 as low .

as is reasonably achievable. 11 The Surveillance Requirements implement the

requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The dose calculations established in the ODCM for calculating the doses due to the actual release rutes of radioactive noble gases in gaseous effluents are consistent with the methodology provided in Regulatory Guide 1.109, 11 Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR 50, Appendix I, 11 Revision I, October 1977 and Regulatory Guide 1.111, 11 Methods for Estimating Atmospheric Transport and Dispersin of Gaseous Effluents in Routine Rel eases from Light-Water Cooled Reactors, 11 Revision 1, July 1977. The ODCM equations provided for determining the air~

doses at and beyond the SITE BOUNDARY are based upon the historical average atmospheric conditions.

3/4.11.2.3 DOSE - IODINE-131, TRITIUM, AND RADIONUCLIDES IN PARTICULATE FORM This specification is provided to implement the requirements of Section II.C, III.A and IV.A of Appendix I, 10 CFR Part 50. The limiting condition for operation are the guides set forth in Section II.C of Appendix I. The ACTION statements provide the required operating flexibility and*at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the SALEM - UNIT 1 B 3/4 11-4

RADIOACTIVE EFFLUENTS BASES releases of radioactive material in gaseous effluents will be kept "as low as is reasonably achievable." The ODCM calculational methods specified in Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown. by calculational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The ODCM calculational methods for calculating the doses due to the actual release rates of the subject materials are consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR 50, Appendix I," Revision 1, October 1977 and Regulatory G.ui de 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors," Revision 1, July 1977. These equations also provide for determining the actual dose based upon the historical average atmospheric conditions. The release rate specifications for iodine-131, tritium, and radionuclides in particu1ate form with half-life greater than 8 days are dependent on the existing radionuclide pathways to man in the areas at and beyond the SITE BOUNDARY. The pathways that were examined in the development of these calculations were: 1) individual inhalation of airborne radionuclides, 2) deposition of radionuclides onto green leafy v.egetation with subsequent consumption by man, 3) deposition onto grassy areas where milk animals and meat producing animals graze with consumption of the milk and meat by man, and 4) deposition on the ground with subsequent exposure of man.

The requirement that the appropriate portions of this system be used, when specified, provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonable achievable." This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objectives given in Section II.O of Appendix I to 10 CFR Part*50. The specified limits governing the use of appropriate portions of the systems were specified as a suitable fraction of the dose design objectives set forth in Section Il.B and II.C of Appendix I, 10 CFR Part 50, for gaseous effluents.

SALEM - UNIT 1 B 3/4 11-5

RADIOACTIVE EFFLUENTS BASES 3/4.11.2.5 EXPLOSIVE GAS MIXTURE This specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the waste gas holdup system is maintained below the flammability limits of hydrogen and oxygen. (Automatic control features are included in the system to prevent the hydrogen and oxygen concentrations from reaching these flammability limits. These automatic control features include isolation of the source of hydrogen and/or oxygen, automatic diversion to recombiners; or injection of dilutants to reduce the concentration below the flammability limits). Maintaining the concentrationof hydrogen and oxygen below their flammability limits proves assurance that the releases of radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50.

3/4.11.2.6 GAS STORAGE TANKS ..

The tanks included in this specification are those tanks for which the quantity of radioactivity contained is not limited directly or indirectly by another Technical ~pecification to a quantity that is less than the quantity which *-"

provides assurance that in the event of a uncontrolled release of the tank's contents, the resulting total body exposure to an individual at the nearest exclusion a.rea boundary will not exceed 0.5 rem in an event of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

Restricting the quan~ity of radioactivity contained in each gas storage tank provides assurance that in the even_t of an uncontrolled release of the nearest exclusion area boundary will not exceed 0.5 rem. This is conistent with Branch Technical Position ETSB 11-5 in NUREG-0800, July 1981.

3/4.11.3 SOLID RADIOACTIVE WASTE This specification implements the requirements of 10 CFR Part 50.36a and General Design Criterion 60 of Appendix A to 10 CFR Part 50. The process parameters included in establishing the PROCESS CONTROL PROGRAM may include, but are not limited to waste type, waste pH, waste/liquid/solidification agent/catalyst ratios, waste oil content, waste principal chemical cohstituents, mixing and curing times.

SALEM - UNIT 1 . B 3/4 11-6

RADIOACTIVE EFFLUENTS BASES 3/4.11.4 TOTAL DOSE This specification is provided to meet the dose limitations of 40 CFR Part 190 that have now.been incorporated into 10 CFR Part 10 to 46 Fr 18525. The specification requires the preparation and submittal of a Special Report whenever the c~lculated doses from plant radioactive effluents exceed twice the design objective dose~ of Appendix I. For sites containing up to 4 reactors, it is highly unlikely that the resultant dose to a MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR Part 190 if the individual reactors remain within the reporting requirement level. The Special Report will describe a course of action that should result in the limitation of the annual dose to a MEMBER OF THE PUBLIC to within the 40 CFR Part 190 limits. For purposes of the Special Report, it may be assumed that the dose commitment to the MEMBER-OF THE PUBLIC from other uranium fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of 8 km must be considered. If the soe to any MEMBER OF THE PUBLIC is estimated to exceed the requirements of 40 CFR Part 190, the Special Report with a request for a variance (provided the release conditions resulting in violation of 40 CFR Part 190 have not already been corrected), in accordance with the provisions of 40 CFR Part 190 and 10 CFR Part 20.405c, is considered to be a timely request and fulfills the requirements of 40 CFR Part 190 until NRC staff action is completed. The variance only relates to the limits of 40 CFR Part 190, and does not apply in any way to the other requirements for dose limitation of 10 CFR Part 20, as addressed in Specifications 3.11.1 and 3.11.2.

An individual is not considered a MEMBER OF THE PUBLIC during any period in which he/she i's engaged in carrying out any opera ti on that is part of the nuclear fuel cycle.

SALEM - UNIT 1 B 3/4 11-7

RAD*IOACTIVE EFFLUENTS BASES 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM The radiological environmental monitoring program required by this specification provides measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides that lead to the highest potential radiation exposures of MEM~ERS OF THE PUBLIC resulting from the station operation. This monitoring* program implementsSection IV.B.2 of Appendix I to 10 CFR Part 50 and thereby supplements the radiological effluent monitoring program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and the modeling,of the environmental exposure pathways.

The initial specified monitoring program will be effective for at .least the first three years of commercial operation. Following this period, program changes may be initiated based on operational experience.

The LLDs required by_ Table 4.12-1 are considered optimum for routine environmental measurements in industrial laboratories. It should be recognized the the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for a particular measureme~t. -

3/4.12.2 LAND USE CENSUS This specification is provided to ensure that changes in the use of areas at and beyond the SITE BOUNDARY are identified and that modificatins to the radial ogi cal. environmental m<;>nitori ng program are made if required by the results of his census. The best information from the door-to-door survey,

aerial survey or consulting with local agricultural auth9rities shall be used.

This census satisfies the requirements of Section IV.8.3 of Appendix I to 10 CFR Part 50. Restricting the census to gardens of greater than 50m2 provides .

assurance that significant exposure pathways via leafy vegetables will be identified and monitored since a garden of this size is the minimum required to produce the quantity (26 kg/year) of leafy vegetables assumed in Regulatory Guide 1.109 for consumption by a child. To determine this minimum garden size, the following assumptions were made: 1) 203 of the garden was used for growing broad leaf vegetation (i.e., similar to lettuce and cabbage), and 2) yield of 2 kg/m2.

SALEM - UNIT 1 B 3/4 12-1

RADIOACTIVE EFFLUENTS

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BASES 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM This requirement for participation in an Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices are

performed as part of the quality assurance program for environmental monitoring in order to demonstrate that the results are reasonably valid for the purposes of Section IV.B.2 of Appendix I to 10 CFR Part 50.

SALEM - UNIT 1 B 3/4 12-2

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

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DESIGN FEATURES DESIGN PRESSURE AND TEMPERATURE 5.2.2 The reactor containment building is designed and shall be maintained for a m~ximum internal pressure of 47 psig and an air temperature of 271°F.

5~3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 193 fuel assemblies with each fuel assembly containing 264 fuel rods clad with Zircaloy-4. Each fuel rod shall have a nominal active fuel length of 143.7 inches and contain a maximum total weight of 1766 grams uranium. The initial core loading shall have a maximum enrichment of 3.35 weight percent U-235. Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum enrichment of 3.5 weight percent U-235.

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 per~ent cadmium. All control rods shall be clad with stainless steel tubing.

5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The reactor coolant system is designed and shall be maintained:

SALEM - UNIT 1 5-4

ADMINiSTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The General Manager - Salem Operations shall be responsible for overall facility operation and shall delegate in writing the succession to this respon-sibility during his absence.

6.1.2 The Senior Shift Supervisor or during his absence from the Control Room, a designated individual shall be responsible for the Control Room command func-tion. A management directive to this effect, signed by the Vice President -

Nuclear shall be reissued to ~11 station personnel on an annual basis.

6.2 ORGANIZATION OFF SITE 6.2.1 The offsite organization for facility management and technical support shall be as shown on Figure 6.2-1.

FACILITY STAFF 6.2.2 The Facility organization shall be as shown on Figure 6.2-2 and:

a. Each on duty shift shall be composed of at least the minimum shift crew composition shown in Table 6.2-1.

b. At least one licensed Operator shall be in the control room when fuel is in the reactor. In addition, at least one licensed Senior Reactor Operator shall be in the Control Room while the unit is in MODE 1, 2, 3 or 4.

c. A health physics technician# shall be on site when fuel is in the reactor.

d. ALL CORE ALTERATIONS shall be observed and directly supervised by a 1i c*ensed Senior Reactor Operator who has no other concurrent responsi-bilities during this _operation.

e. A site Fire Brigade of at least 5 members shall be maintained onsite at all times#. The Fire Brigade shall not include 4 members of the minimum shift crew necessary for.safe shutdown of.the unit or any personnel required for other essential functions during a fire emergency.

.- f. The amount of overtime worked by plant staff members performing

< safety-related functions must be limited in accordance with the NRC Policy Statement on working hours (Generic Letter No. 82-12).

#The health physics technical and Fire Brigade composition may be less than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accommodate unexpected absence of the health physics technician and/or Fire Brigade members provided immediate action is taken to restore the health physics technician and/or Fire-Brigade to within the minimum requirements.

SALEM - UNIT 1 6-1

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I FIGURE a *2 - 2 ;FACILITY ORGANIZATION
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
ss 1a 1a SRO 1a none RO 2 1 AO 3 2b STA 1 none Maintenance 1 none
Electrician 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 SS 1a 1a SRO 1a none RO 2, 1 AO 3c 1 STA 1a none Maintenance 1a none Electrician 2-_/ Individual may fill the same position on Unit 2 EJ One of the two required individuals may fill the same position on Unit 2, such that there are a total of three AOs for both units.
S:./ One of the three required individuals may fill the same position on Unit 2, such that there are a total of five AOs for both units.
SALEM - UNIT 1 6-4
TABLE 6.2-1 (Continued)
SS - Shift Supervisor with a Senior Reactor Operators License on Unit 1 SRO - Individual with a Senior Reactor Operators License on Unit 1 RO Individual with a Reactor Operators License on Unit 1 AO Auxiliary Operator STA - Shift Technical Advisor Except for the 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 exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accommodate unexpected absence of on-duty shift crew.
crew members provided immediate action is taken to restore the Shift Cre~
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 crewman being late or absent.
During any absence of the Shift Supervisor from the Control Room while the unit is in MODE 1, 2, 3 or 4, an individual (other than the Shift Technical Advisor) with a v~lid SRO license shall be designated to assume the Control Room command function. During any absence. of the Shift Supervisor from the Control Room while the unit is in MODE 5 or 6, an individual with a valid RO license (other than the.Shift Technical Advisor) shall be designated to assume the Control Room command function.
SALEM - UNIT 1 6-5
ADMINISTRATIVE CONTROLS 6.2.3 SHIFT TECHNICAL ADVISOR 6.2.3.1 The Shift Technical Advisor shall serve in an advisory capacity to the Shift Supervisor on matters pertaining to the engineering aspects assuring safe operation of the unit.
6.2.3.2. The Shift Techni*cal Advisor shall have a Bachelor's Degree or equivalent in a sci~ntific or engineering discipline with speciffc 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 m1n1mum qualifications of ANSI Nl8.1-1971 for comparable positions and the supplemental requirements specified in Sections A and C of Enclosure 1 of the March 28, 1980 NRC letter to all licensees, except for the Radiation Protection Engineer who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975.
6.4 TRAINING 6.4.1 A retraining and replacement training 'program for the facility staff shall be coordinated by each functional level manager (Department Head) at the fad 1 i ty and maintained under the* direction of the Manager - Nuclear Training and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI Nl8.1-1971 and Appendix A of 10 CFR Part 55 and the supplemental 11 11 requirements specified in Sections A and C of Enclosure 1 of the March 28, 1980 NRC letter to all licensees, and shall include familiarization with relevant industry operational experience.
6.4.2 A training program for the Fire Brigade shall be maintained under the direction of the Manager - Nuclear Training and shall meet or exceed the requirements of Section 27 of NFPA Code-1975, except for Fire Brigade training sessions which shall be held at least quarterly.
6.5 REVIEW AND AUDIT 6.5.1 STATION OPERATIONS REVIEW COMMITTEE (SORC)
FUNCTION 6.5.1.1 The Station Operations Review Committee* shall function to advise the General Manager - Salem Operations on all matters related to nuclear safety.
SALEM .- UN IT 1 6-6
ADMINISTRATIVE CONTROLS COMPOSITION 6.5.1.2 The Station Operations Review Committee shall be composed of the:
Chairman: Assistant General Manager Salem Operations Vice Chairman: Operations Manager Vice Chairman: Technical Manager Vice Chairman: Maintenance Manager Member: Operating Engineer Member: I&C Engineer Member: Senior Shift Supervisor Member: Technical Engineer Member: Maintenance Engineer Member: Radiation Protection Engineer Member: Senior Radiation Protection Supervisor Member: Chemistry Engineer ALTERNATES 6.5.1.3 All alternate members shall be appointed in writing by_ the SORC Chairman to serve on a temporary basis; however, no more than two alternates shall participate as voting members in SORC activities at an:Y one time.
MEETING FREQUENCY 6.5.1.4 The SORC shall meet at least once per calendar month and as convened by the SORC Chairman or his designated alternate.
QUORUM 6.5.1.5. The minimum quorum *of the SORC necessary for the performance of the SORC responsibility and authority provisions of these technical specifications shall consist of the Chairman or his designated alternate and four members including alternates. -
RESPONSIBILITIES 6.5.1.6 The Station Operations Review Committee shall be responsible for:
a. Review of 1) all procedures required by Specification 6.8 and changes thereto, 2) all programs required by Specification 6.8 and changes thereto, and 3) any other proposed procedures or changes thereto as determined by the General *Manager - Salem Op.erations to affect nuclear safety.
b. Review of all proposed tests and experiments that affect nuclear safety.
c. Review of all proposed changes to Appendix "A" Technical Speci fi cations.
SALEM - UNIT 1 6-7
ADMINISTRATIVE CONTROLS
d. Review of all proposed changes or modifications to plant systems or equipment that affect nuclear safety.
e. Investigation of all violations of the Technical Specifications including the preparation and forwarding of reports covering evaluation and recommendations to prevent recurrence to the Vice President - Nuclear and to the Chairman of the Nuclear Review Board.
f. Review of events requiring 24 *hour written notification to the Commission.
g. Review of facility operations to detect potential nuclear safety hazards.
h. Performance of special reviews, investigations or analyses and reports thereon as :requested by the Chairman of the Nuclear Review Board.
i. Review of the Plant Security Plan and implementing procedures and shall submit recommended changes to the Chairman of the Nuclear Review Board.
j. Review of the Emergency Plan and implementing procedures and shall submit recommended changes to the Chairman of the Nuclear Review Board.
k. Review of every unplanned onsite release of radioactive material to the environs including the preparation of reports covering evaluation, recommendations, and disposition of the corrective action to prevent recurrence and the forwarding of these reports to the Vice President - )
Nuclear and to the Nuclear Review Board.
l. Review of changes to the PROCESS CONTROL PROGRAM and the OFFSITE DOSE I CALCULATION MANUAL. I AUTHORITY 6.5.1.7 The Station Operations Review Committee shall:
a. Recommend to the General Manager - Salem Operatio(ns) written a(pp)roval ~
or disapproval of items considered under 6.5.1.6 a through d above.
b. Render determinations in writing with regard to whether or not each item considered under 6:5.1.6(a) through (e) above constitutes an unreviewed safety question.
c. Provide written notification within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to the Vice President Nuclear and the Nuclear Review Board of disagreement between the SORC and the General Manager ~ Salem Operations; however, the General Manager - Salem Operations shall have responsibility for resolution of such disagreements pursuant to 6.1.1 above.
SALEM - UNIT 1 6-8
ADMINISTRATIVE CONTROLS RECORDS 6.5.1.8 The Station Operations Review Committee shall maintain written minutes of each meeting that, at a minimum, document. the results of all SORC activities performed under the responsibility and authority provisions of these technical specifications. Copies shall be provided to the Vice President - Nuclear and Chairman of the Nuclear Review Board.
6.5.2 NUCLEAR REVIEW BOARD (NRB)
FUNCTION 6.5.2.1 The Nuclear Review Board shall function to provide independent review and audit of designated activities in the areas of:
a. nuclear power plant operatioris
b. nuclear engineering
c. chemistry and radiochemistry
d. .metal 1urgy
e. instrumentation and control
f. radiological safety
g. mechanical engineering
h. electrical engineering
i. quality assurance
j. nondestructive testing
k. emergency prepardness COMPOSITION 6.5.2.2 The Vice President - Nuclear shall, appoint at least nine members to the Nuclear Review Board and shall designate from this membership a chairman and at least one vice chairman. The membership shall collectiv~ly possess experi-ence and competence to provide independent review and audit in the areas listed in Section 6.5.2.1. The chai'rman and vice chairman shall have nuclear back-ground in engineering or operations and shall be capable of determining when to -*
call in experts to assist the NRB review of complex problems. All members shall have at least a Bachelor Degree in Engineering or related sciences. The chairman shall have at.least six years of professional level managerial experi-ence in the power field and all other members shall have at least five years of cumulative professional level experience in the fields listed in.Section 6.5.2.1. ,
ALTERNATES 6.5.2.3 All alternate members shall be appointed in writing by the NRB Chairman to serve on a temporary basis; however, no*more than two alternates shall par-ticipate as voting members in NRB activities at any one time. Educational and experience ~ualifications required of members are also applicable to alternates *
CONSULTANTS 6.5.2.4 Consultants shall be utilized as determined by the NRB Chairman to provide expert advice to the NRB.
SALEM - UNIT 1 6-9
ADMINISTRATIVE CONTROLS MEETING FREQUENCY 6.5.2.5 The NRB shall meet at least once per calendar quarter during the .
initial year of unit operation following fuel loading and at least once per six months thereafter.
QUORUM 6.5.2.6 The m1n1mum quorum of NRB necessary for the performance of the NRB review and audit functions of these technical specifications shall consist of the Chairman or his designated alternate and at least 4 NRB members including alternates. No more than a minority of the quorum shall have li.ne responsi-bility for operation of the facility.
REVIEW 6.5.2.7 The NRB shall review:
a. The safety evaluations for 1) changes to procedures, equipment or systems and 2) tests or experiments completed under the provision of Section 50.59, 10 CFR, to verify that such actions did not constitute an unreviewed safety question.
b. Proposed changes to procedures, equipment or systems which involve an unreviewed safety question as defined 1n Section 50.59, 10 CFR.
c. Proposed tests or experiments which involve an unreviewed safety question as defined in Section 50.59, 10 CFR.
d. Proposed changes to Technicai .Specifications or this operating license.
e. Violations of codes, regulations, orders, Technical Specifications, license requirements, or of internal procedures or instruction having nuclear safety significance. *-
f. Significant operating abnormalities or deviations from normal and expected performance of plant equipment that affect nuclear safe4y.
~* Events requiring 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months written notification to the Commission.
h. All recognized indications of an unanticipated deficiency in some aspect of design or operation of safety related structures, systems, or components.
i. Reports and meetings minutes of the Station Opera~ions Review Committee.
AUDITS 6.5.2.8 Audits of facility activities shall be performed under the cognizance of the NRB. These audits shall encompass:
SALEM - UNIT 1 6-10
ADMINISTRATIVE CONTROLS
a. The conformance of facility operation to prov1s1ons contained within the Technical Specifications and applicable license conditions at least on~e per 12 months.
b. The performance, training and qualifications of the entire facility staff at least once per 12 months.
c. The results of actions taken to correct deficiencies occurrring in facility equipment, structures, systems or method of operation that affect nu.clear safety at least once per 6 months.
d. The performance of activities required by the Operational Quality Assurance Program to meet the criteria of Appendix 11 B11 , 10 CFR 50, at least once per 24 months.
e. The Facility Emergency Plan and implementing procedures at least once per 12 months.
f. The Facility Security Plan and_ implementing procedures at least* once per 12 months.
g. Any other area of facility operation considered appropriate by the NRB or the Vice President - Nuclear.
h. The Facility Fire Protection Program and implementing procedures at least once per 24 months.
i. An independent fire protection and loss prevention program inspection and audit shall.be performed at least once per 12 months utilizing either qualified offsite licensee personnel or an outside fire protection firm.
j. An inspection and audit of the fire protection and loss prevention program shall be performed by a qualified outside fire consultant at least once per 36 months *.
k. The radiological environmental monitoring program and the results thereof at least once per 12 months.
AUTHORITY 6.5.2.9 The NRB shall report to and advise the Vice President - Nuclear on those areas of responsibility specified in Sections 6.5.2.7 and 6.5.2.8.
RECORDS 6.5.2.10 Records of NRB activities shall be prepared, approved and distributed as indicated below:
a. Minutes of each NRB meeting shall be prepared, approved and forwarded to the Vice President - Nuclear within 14 days following each meeting.
SALEM - UNIT 1 . 6-11
ADMINISTRATIVE CONTROLS
b. Reports of reviews encompassed by Section 6.5.2.7 above, shall be prepared, approved and forwarded to the Vice President - Nuclear within 14 days following completion of the review.
c. Audit reports encompassed. by Section 6.5.2.8 above, shall be forwarded to the Vice President - Nuclear and to the management positions responsible for the areas audited within 30 days after completion of the *audit.
6.5.3 SAFETY REVIEW GROUP (SRG)
FUNCTION 6.5.3.1 The SRG shall perform independent reviews of plant operations and to advise appropriate station/corporate management on the overa 11 safety of pl ant operations.
COMPOSITION 6.5.3.2 The SRG shall be composed of at least five dedicated, full-time engineers, located on Artificial Island and shall report to the General Manager -
Nuclear Support *
RESPONSIBILITIES 6.5.3.3 a.
The SRG shall be responsible for:
Review of selected plant operating characteristics, NRC*issuances, industry advisories, and othe sour*ces of plant design/operating experience data that may indicate areas for improving plant safety.
b. Reviews of selected facility features, equipment and systems.
c. Reviews of selected procedures, and plant activities including main- --
tenarice, modifications, operational problems and operational analysis.
d. Surveillance of selected plant operations and maintenance activities to provide independent verification* that they are performed correctly and that human errors are reduced to as low as reasonably achievable.
AUTHORITY 6.5.3.4 The SRG shall make detailed recommendations for improving plant safety to the appropriate station/corporate management through the General Manager -
Nuclear Support.
Not responsible for sign-off function *
SALEM - UNIT 1 6-12
ADMINISTRATIVE CONTROLS 6,6 REPORTABLE OCCURRENCE ACTION 6.6.1 The following actions shall be taken for REPORTABLE OCCURRENCES:
a. The Commission shall be notified and/or a report submitted pursuant to the requirements of Specification 6.9.
b. Each REPORTABLE OCCURRENCE requiring 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months notification to the Commission shall be reviewed by the SORC and submitted to the NRB and the Vice President - Nuclear.
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 Vice President - Nuclear I and the NRB 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 Commis-sion, the NRB and the Vice President - Nuclear within 14 days of the violation.
SALEM - UN IT 1 6-12a
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 Reguiatory 11 11 Guide 1.33, Revision 2, February 1978.
b. Refueling ope~ations *.
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. Qua1ity Assurance Program for effluent and environmental monitoring.
6.8.2 Each procedure and administrative policy of 6.8.1 above, and changes thereto, shall be reviewed by the SORC and approved by the General Manager -
Salem Operations prior to implementation and reviewed periodically as set forth in administrative procedures.
6.. 8."3 a.
Temporary changes to procedures of 6.8.1 above may be made provided:
The intent of the original procedure is not altered.
b. The change is approved by two members of .the plant management staff,
~t least one of who~ holds a Senior Reactor Operator's License on the unit affected. *
c. The change is documented, reviewed by the SORC and approved by the General Manager - Salem Operations within 14 days of implementation.
SALEM - UNIT 1 6-13
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 transie.nt or accident to as low as practical levels. The systems include (recircul.ation spray, safety injection, chemical and volume control, gas stripper, recombiners, *** )~ The program shall include the following:
(i). Preventative maintenance and periodic visual inspection require-ments, and
(ii) Integrated leak* test requirements for each system at refueling cycle interval.s or less.
b. In-Plant Radiation Monitoring A program which wi 11 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 th~ 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, *
(vi) A procedure identifying (a) the authority responsible for the interpretation of the data, and (b) the sequence and timing of administrative events required to initiate corrective action.
SALEM - UNIT 1 6-14
ADMINISTRATIVE CONTROLS
d. Backup Method for Determining Subcooling Margin A program which will ensure the capability to accurately monitor the Reactor Coolant System Subcooling Margin. This program shall include
.the following:
(i) Training of personnel, and (ii) Procedures for monitoring.
6.9 REPORTING REQUIREMENTS ROUTINE REPORTS AND REPORTABLE OCCURRENCES 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 Administra-tor of the Regional Office of Inspection and Enforcement unless otherwise noted.
STARTUP REPORT
6.9.1.1 A summary report of plant startup and power escalation testing shall be submitted follpwing (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 differ,ent 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 the 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 (1) 90 days fo~lowing 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 REPORTSl/
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. The initial report shall be submitted prior to March 1 of the year
following initial criticality.
1/ A single submittal may be made for a multiple unit station. The submittal should combine th6se sections that are common to all units at the* station.
SALEM - UNIT -1 6-15
\
ADMINISTRATIVE CONTROLS 6.9.1.5 Reports required on an annual basis shall include:
a. A tabulation on an annual basis of the number of station, utility and other personnel (including contractors) receiving exposures greater than 100 mrem/yr and their associated man rem exposure according to work and job functions,2/ e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling. The dose assignment to various duty functions may be estimates based on pocket dosimeter, TLD, or film badge measurements. Small exposures totalling less than 20% of the individual total dose need not be accounted for._
In the aggregate, at least 80% of the total whole body dose received from external sources shall be assigned to specific major work functions.
b. The complete results of steam generator tube inservice inspections performed during the report period (reference Specificat~on
4.4.5~5.b).
MONTHLY OPERATING REPORT 6.9.l.6 ~outine reports of operating statistics and shutdown experience, including documentation of all challenges to the PORVs or safety valves, shall be submitted on a monthly_ basis to the Director, Office of Management Informa-tion and Program Analysis, U.S. Nuclear Regulate~ Commission, Washington, D.C.
20555, with a copy of the Regional Office of OI&E, no later than the 15th of each month following the calendar month covered by the report.
REPORTABLE OCCURRENCES 6.9.1.7 The REPORTABLE OCCURRENCES of Specifications 6.9.1.8 and 6.9.1.9 below, including corrective actions and measures to prevent recurrence, shall be reported to the NRC. Supplemental reports may be required to fully describe final resolution of occurrence. In case of corrected or supplemental reports, a --.
licensee event report shall be completed and reference shall be made to the original report date.
PROMPT NOTIFICATION WITH WRITTEN FOLLOWUP 6.9.1.8 The types of events listed below shall be reported within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by telephone and confirmed by telegraph, mailgram, or facsmile transmission to the Adm1nistrator of the Regional Office, or his designate no later than the first working day following the event, with a written followup report within 14 days.
The wirtten followup report shall include, as a minimum, a completed copy of a licensee event report form. Information provided on the licensee event report form shall be supplemented, as needed, by additional narrative material to provide complete explanation of the circumstances surrounding the event.
2; This tabulation supplements the requirements of section 20.407 of 10 CFR Part 20.
SALEM - UNIT 1 6-16
ADMINISTRATIVE CONTROLS
a. Failure of the reactor protection system or other systems subject to limiting safety system settings to initiate the required protective function by the time a monitored parameter reaches the setpoint speci-fied as the limiting safety system setting in the technical specifica-tions or failure to complete the required protective function.
b. Operation of the unit or affected systems when any parameter or opera-tion subject to a limiting condition for operation is less conserva-tive than the least conservati~e aspect of the limiting conditio~ for operation established in the technical specifications *.
c. Abnormal degradation discovered in fuel cladding, reactor coolant pressure boundary, or primary containment.
d. Reactivity anomalies involving disagreement with the predicted value of reactivity balance under steady state conditions during power operation greater than or equal to 1% delta k/k; a calculated reac-tivity balance indicating a SHUTDOWN MARGIN less conservative than specified in the technical specifications; short-term reactivity increases that correspond to a reactor period of less than 5 seconds or, if subcritical, an unplanned reactivity insertion of more than 0.5% delta k/k; or occurrence of any *unplanned criticality.
e. Failure or malfunction of one or more components which prevents or could prevent, by itself, the fulfillment of the functional require-ments of system(s) used to cope with accidents analyzed in the SAR.
f. Personnel error or procedural inadequacy which prevents or could prevent, by itself, the fulfillment of the functional requirements of systems required to cope with accidents analyzed in the SAR.
g. Conditions arising from natural or man-made events that, as a direct result of the event require unit shutdown, operation of safety systems, or other protective measures required by technical specifications.
h. Errors discovered in the transient or accident analyses or in the methods .used for such analyses as described in the safety analysis report or in the bases for the technical specifications that have or could have permitted reactor operation in a manner less conservative than assumed in the analyses.
i. Performance of structures, systems, or components that requires remedial action or corrective measures to prevent operation in a manner less conservative than assumed in the accident analyses in the safety analysis report or technical speci fi cations bases; or discovery during unit life of conditions not specifically considered in the safety analysis report or technical specifications that require .
remedial action or corrective measures to prevent the existence or development of an unsafe condition.
j
Offsite releases of radioactive materials in liquid and gaseous efflu-ents that exceed the limits of Specification 3.Il.1.1 or 3.11.2.1.
SALEM - UNIT 1 6-17
ADMINISTRATIVE CONTROLS k~ Exceeding the limits in Specification 3.11.1.4 or 3.11.2.6 for the storage of radioactive materials in the listed tanks. The written follow-up report shall include a schedule and a description of activities planned and/or taken to reduce the contents to within the specified limits.
THIRTY DAY WRITTEN REPORTS 6.9.1:9 The types of events listed below shall be the subject of written reports to the Administrator of the Regional Office within thirty days of occurrence of the event. The written report shall include, as a minimum, a completed copy of a licensee event report form. Information provided on the licensee event report form shall be supplemented, *as needed, by additional narrative material to provide complete explanation of the circumstances surrounding the event~ *
a. Reactor protection system or engineered safety feature instrument settings which are found to be less conservative than those estab-1 ished by the technical specifications but which do not prevent the fulfillment of the functional requirements of affected systems.
b. Conditions leading to operation in a degraded mode permitted by a limiting condition for operation or plant shutdown required by a limiting condition for ope,ration.
c. Observed inadequacies in the implementation of administrative or procedural controls which threaten to cause reduction of degree of redundancy provided in reactor protection systems or engineered safety feature systems.
d. Abnormal degradation of systems other than those specified. in 6.9.1.8.c above designed to contain radioactive material resulting from the fission process.
e. An unplanned offsite releases of 1). more than 1 curie of radioactive material in liquid effluents, 2) more than 150 curies of noble gas in gaseous effluents. The report of an unplanned offsite release of radioactive material shall include the following information.
1. A description of .the event and equipment involved. *
2. Cause(s) for the unplanned release.
3. Actions taken to prevent recurrence.
4. Consequences of the unplanned release.
SALEM - UNIT 1 6-18
ADMINISTRATIVE CONTROLS ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT*
6.9.1.10 Routine Radiological Environmental Operating Reports covering the operation of the unit during the previous calendar year shall be submitted prior to May .1 of each year.
The Annual Radiological Environmental Operating Reports shall include summaries, interpretations, and an analysis of trends of the results of the radiological environmental surveillance activities for the r*eport period, including a com-parison with preoperational studies with operational controls (as appropriate),
and with previous environmental surveilance reports, and an a~sessment of the observed impacts of the plant operation on the environment. The reports shall also include the results of land use censuses required by Specification 3.12.2.
The Annual Radiological Environmental Operating Reports shall include the results of analysis of all radiological environmental samples and of all meas-urements taken during the period pursuant to the Table and Figures in the envi-ronmental radiation section of the ODCM; as well as summarized and tabulated re-sults of locations specified in these analyses and measurements in the format of the table in the Radi-ological Assessment Branch Technical Position, Revision 1, November 197~. In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted as soon as possible in a supplementary report.
The reports shall also include the following: a summary description of the radiological environmental monitoring program; at least two legible maps** cov-ering all sampling locations keyed to a table giving distances and directions from the centerline of one reactor; the results of licensee participation in the Interlaboratory Comparison Program, required by SpecificatiOn 3.12.1; and. dis-cussion of all analyses in which the LLD required by Table 4.12-1 was not achievable.
SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT*
6.9.1.11 Routine Radioactive Effluent Release Reports covering the operation of -
the unit during the previous six months of operation shall be submitted within 60 days after January 1 and July 1 of each year.
The Radioactive Effluent Release Reports shall include a summary of the quanti-ties of radioactive liquid and gaseous effluents and solid waste released from the unit as outlined in Regulatory Guide 1.21. "Measuring, Evaluating, and
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.
In lieu of submission with the Annual Radiological Environmental Operating Report, the licensee has the option of retaining thfa summary of required meteorological data on site in a file that shall be provided to the NRC upon request.
SALEM - UNIT 1 6-19
ADMINISTRATIVE CONTROLS Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plant, 11 Revision 1, June 1974, with data summarized on a quarterly basis following the format of Appendix B thereof.
The Radioactive Effluent Release Report to be submitted within 60 days after January 1 of each year shall include an annual summary of hourly meteorological data collected over the previous year. This annual summary may be efther in the form of an hour-by-hour listing of magnetic tape of wind speed, wind direction, atmospheric stability, and precipitation (if measured), or in the form of joint frequency distributions of wind speed, wind direction, and atmospheric
stability.*** This same report shall include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the unit or station during the previous calendar year. This same report shall also include an assessment of the radiation doses from radioactive liquid and gaseous effluents to MEMBERS OF THE PUBLIC due to their activities inside the SITE BOUNDARY (Figure 5.1-3) during the report period. All assumptions used in making these assessments (i.e., specific activity, exposure time and location) shall be included in these reports. The meteorological conditions concurrent with the time of release of radioactive materials in gaseous efflu~nts (as determined by sampling frequency and measurement) shall be used for determining the gaseous pathway doses. The assessment of radiation doses shall be performed in accordance with the OFFSITE DOSE CALCULATION MANUAL.
The Radioactive Effluent Release Report to be submitted 60 days after January 1 of each year shall also include an assessment of radiation doses to the likely most exposed MEMBER OF THE PUBLIC from reactor releases and other nearby uranium fuel cycle sources (including doses from primary effluent pathways and direct .
radiation) for the previous calendar year to show conformance with 40 CFR Part 190, Environmental Radiation Protection Standards for Nuclear Power Operation.
Acceptable methods for calculating the dose contribution from liquid and gaseous effluents are given in Regul a,tory Gui de 1.109, Rev. 1, October 1977.
The Radioactive Effluent Release Reports shall include the following information -
for each class of solid waste (as defined by 10 CRF Part 61) shipped offsite during the report period:
a. Container volume,
b. Total curie quantity (specify whether determined by measurement or estimate),
c. Principal d Source of waste and processing employed (e.g., dewatered spent resin, compacted dry waste, evaporator bottoms),
In lieu of submission with the first half year Radioactive Effluent Release Report, the licensee has the option of retaining this summary of required meteorological data on site in a file tha.t shall be provided to the NRG *upon request.
SALEM - UNIT 1 6-20
ADMINISTRATIVE CONTROLS
e. Type of container (e.g., LSA, Type A, Type B, Large Quantity), and
f. Solidification agent or absorbent (e~g., cement, urea formaldahyde).
The Radioactive Effluent Release Reports shall include a list of description of unplanned releases from the site to UNRESTRICTED AREAS of radioactive materials in gaseous and liquid effluents made during the reporting period.
The Radioactive Effluent Rel ease Reports sha 11 i nc'l ude any changes made during the reporting period to the PROCESS CONTROL PROGRAM (PCP) and to the OFFSITE DOSE CALCULATION MANUAL (ODCM), as well as a listing of new locations for dose calculations and/or enyironmental monitoring identified by the land use census pursuant to Specification 3.12.2.
SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Administrator of the Regional Office within the time p~riod specified for each report.
6.10 RECORD RETENTION In addition to -the applfcable record retention r-equirements of Title 10, Code of Federal Regulations, the following records shall be retined for at least the minimum period indicated.
6.10.1 The following ~ecords shall be retained for at least five years:
a. Records and logs of unit operation covering time interval at each power level * *
b. Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety. *--
c. ALL REPORTABLE OCCURRENCES submitted to the Commission.
d. Records of surveillance activities, inspections and calibrations required by these Technical Specifications.
e. Records of reactor tests and experiments.
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.
SALEM - UNIT 1 6-21
ADMINISTRATIVE CONTROLS 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.
b. Records of new and irradiated fuel inventory, fuel transfers and assembly burnup histories.
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 performed pursuant to these Technical Specifications.
i. Records of Quality Assurance activities required by the QA Manual.
j. Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59.
k. Records of meetings of the SORC and the NRB. -*
l. 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 listed on Tables 3.7-4a and 3.7-4b 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.
SALEM - UNIT 1 6-22
ADMINISTRATIVE CONTROLS 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.
6.12 HIGH RADIATION AREA 6.12.1 In lieu of the 11 control device 11 or 11 alarm* signal 11 required by paragraph 20.203(c)(2) of 10 CFR Part 20, each high radiation area in which the intensity of radiation is greater than 100 mrem/hr but less than 1000 mrem/hr shall be barricaded and conspicuously _posted as a High Radiation Area and entrance thereto shall be controlled by issuance of a Radiation Exposure Permit*. Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:
a. A radiation monitoring device which continuously indicates the radiation dose rate in the area.
b. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate level in the area has been established and personnel have been made knowledgeable of them.
c. A health physics qualifi~d individual (i.e., qualified in radiation protection procedures) with a radiation dose rate monitoring device who* is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the facility Health Physicist in the Radiation Work Permit.
6.12.2 In addition to the requirements of 6.12.1, areas accessible to personnel with radiation levels such that a major portion of the body could receive in one hour a dose greater than 1000 mrem shall be provided with locked doors to prevent unauthorized entry, and t~e keys shall be maintained under the administrative control of the Senior Shift Supervisor on duty and/or Senior Supervisor - Radiation Protection. Doors shall remain locked except during periods of access by personnel under an approved Radiation Work Permit which shall specify the dose rate levels in the immediate work area and the maximum allowable stay time for individuals in that area. For individual areas accessible to personnel with radiation levels such that a major portion of the body could receive in one hour
Health Physics Personnel or personnel escorted by Health Physics personnel shall be exempt from the RWP issuance requirement during the performance of their assigned radiation protection duties, provided they are otherwise following plant radiation protection procedures for entry into high radiation areas.
SALEM - UNIT 1 6-23
ADMINISTRATIVE CONTROLS a dose in excess of 1000 mrem** that are located within large areas, such as PWR containment, where no enclosure exists for purposes of locking, and no enclosure can be reasonable constructed around the individual areas, then that area shall be roped off, conspicuously posted and a flashing light shall be activated as a warning device. In lieu of the stay time specification of the RWP, direct or remote (such as use of closed circuit TV cameras) continuous surveillance may be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities w.ithin the area.
6.13 PROCESS CONTROL PROGRAM (PCP) 6.13.1. The PCP shall be approved by the Commission prior to implementation.
6.13.2 Licensee initiated changes to the PCP:
1. Shall be submitted to the Commission in the Semiannual Radioactive Effluent Re 1ease Report for the peri ad in which the change ( s) was
made. This submittal shall contain:
a. Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information;
b. A determination that the change did not reduce the overall conformance of the solidified waste product to existing criteria for solid wastes; and
c. Documentation of the fact that the change has been reviewed and found acceptable by the SORC.
2. Shall become effective.upon review and acceptance by the SORC.
6.14 OFFSITE DOSE CALCULATION MANUAL (ODCM) 6.14.1 The ODCM shall be approved by the Commission prior to implementation.
6.14.2 Licensee initiated changes to the ODCM:
1. Shall be submitted to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the change(s) was made. This submittal shall contain:
a. Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information. Information submitted should consist of a package of those pages of the ODCM to be changed with each page numbered and provided with an approval and date box, together with appropriate analyses or evaluations justifying the change(s);
Measurement made at 18 11 from source of radi oacti_vity.
SALEM - UNIT 1 6-24
ADMINISTRATIVE CONTROLS
.. b. A determination .that the change will not reduce the accuracy or reliability of dose calculations or setpoin~ determination; and
c. Documentation of the.fact that the change has been reviewed and found acceptab.le by the SORC.
2. Shall become effective upon review and acceptance by the SORC.
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 FSAR for the period in which the evaluation was reviewed by (SORC). The discussion of each changes shall contain:
a. A summary of the evaluation that led to the determination could be made in accordance with 10 CFR Part .50.59;
b. Sufficient detailed information to totally support the reason for the change without benefit of additional or supplemental information;
c. A detailed description of the equipment, components and processes involved and th~ 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.
SALEM - UNIT 1 6-25
ADMINISTRATIVE CONTROLS 6.16 ENVIRONMENTAL QUALIFICATION 6.16.1 By no later than June 30, 1982 all safety-related electrical equipment in the facility shall be qualified in accordance with the provisions of:
Division of Operating Reactors "Guidelines for Evaluating Environmental Qualification of Class lE 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 By no later than December 1, 1980, complete and auditible records must 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. Thereafter, such records should be updated and maintained current as equipment is replaced, further tested, or otherwise further qualified.
~
SALEM - UNIT 1 6-26
INDEX DEFINITIONS SECTION PAGE 1.0 DEFINITIONS DEFINED TERMS ***** 1-1 ACTION CD * * * * * * * * * * * * * * * * * * *
1-1 AXIAL FLUX DIFFERENCE ***** 1-1 CHANNEL CALIBRATION
1-1 CHANNEL CHECK * * * * * * *
1-1 CHANNEL FUNCTIONAL TEST **** .... 1-1 1-2 CO.NTAINMENT INTEGRITY * * * * * * *
CONTROLLED LEAKAGE * * * *
1-2 CORE ALTERATION * * * . * * * .... 1-2 1-2 DOSE EQUIVALENT I-131 *****
E-AVERAGE DISINTEGRATION ENERGY *** 1-3 ENGINEERED SAFETY FEATURE RESPONSE TIME 1-3 .
FREQUENCY NOTATION * * * * * * * * * * ... 1-3 1-3 GASEOUS RADWASTE TREATMENT SYSTEM ***
IDENTIFIED LEAKAGE ****** * **** 1-3 MEMBER(S) OF THE PUBLIC * * * * * * * * * * * * ..... 1-4 OFFSITE DOSE CALCULATION MANUAL (ODCM)
1-4 OPERABLE - OPERABILITY ***** ... 1-4 1-4 OPERATIONAL MODE * * * * * * * * * ***
PHYSICS TESTS * * * * * * * * * * * * * * *
1-5 PRES~URE BOUNDARY LEAKAGE~ * * * * * * * * *
1-5 PROCESS CONTROL PROGRAM (PCP) ** . .. . . 1-5 PURGE-PURGING * * * * * * * -. * *** .... 1-5 QUADRANT POWER TILT RATIO * ~ * * * * * * *
RATED THERMAL POWER * * * * *
REACTOR TRIP SYSTEM RESPONSE TIME 1-5 1-5 1-6 REPORTABLE OCCURRENCE * * * *.* 1-6 SHUTDOWN MARGIN ***** . . .. 1-6 I
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 * * * * * * * * *
VENTILATION EXHAUST TREATMENT SYSTEM ** ~
VENT! NG. * *********** . ..
1-7 1-7 1-7 I SALEM - UNIT 2 I
INDEX SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SECTION PAGE 2.1 SAFETY LIMITS Reactor Core. * .****** 2-1 Reactor Coolant System Pressure ... 2-1 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Trip System Instrumentation Setpoints 2-4 BASES SECTION PAGE 2.1 SAFETY LIMITS Reactor Core * * * * * * * * *
Reactor Coolant System Pressure .... .
B 2-1 B 2-2 2.2 LIMITING SAFETY SYSTEM SETTTIN.GS Reactor Trip System Instrumentation Setpoints B 2-3 SALEM - UNIT 2 II
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 Marg~n - Tavg > 2oo:F ** 3/4 1-1 Shutdown Margin - Tavg -5_ 200 F **** ........ 3/4 1-3 Moderator Temperature Coefficient
3/4 1-4 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 3/4 1-8 1-9 Charging.Pump - Shutdown Charging Pump - Operating * * *** 3/4 1-10 Borated Water Sources - Shutdown **** 3/4 1-11 Borated Water Sources - Operating
3/4 1-12 3/4.1.3 MOVABLE CONTROL ASSEMBLIES Group Height * * * * * * * * * * * * *
3/4 1-13 Position Indicating Systems - Operating
3/4 1-16 Position Indicating Systems - Shutdown 3/4 1-17 Rod Drop Time * * * * * *
3/4 1-18 Shutdown Rod Insertion Limit 3/4 1-19 Control Rod Insertion Limits ***** 3/4 1-20
SALEM - UNIT 2 III
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 RCS FLOW RATE.AND R **** 3/4 2-9 3/4.2.4 QUADRANT POWER TILT RATIO
3/4 2-13 3/4.2.5 DNB PARAMETERS ........ 3/4 2-16 3/4.3 INSTRUMENTATlON 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 3-38 Movable Incore Detectors * * * * * * *
3/4 3-42 Remote Shutdown Instrumentation **** 3/4 3-43 Fire Detection Instrumentation * * * * * **** 3/4 3-46 Accident Monitoring Instrumentation * * * * * * * * *
3/4 3-50 Rad~oactive Liquid Effluent Monitoring Instrumentation. 3/4 3-53 Radioactive Gaseous Effluent Monitoring Instrumentation 3/4 3-59 3/4.3.4 TURBINE OVERSPEED PROTECTION * * * * * * * * * * * *
3/4 3-65 SALEM - UNIT 2 IV
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE .
3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION
~
Startup and Power Operation * ... 3/4 4-1 Hot Standby
Hot Shutdown
... 3/4 3/4 4-2 4-3 Cold Shutdown * ...... .... 3/4 4-4a 3/4.4.2 SAFETY VALVES - SHUTDOWN . . . . 3/4 4-5 3/4.4.3 SAFETY VALVES - OPERATING
3/4 4-6 3/4.4.4 PRESSURIZER ....
. 3/4 4-7 3/4.4.5 RELIEF VALVES * . 3/4 4-8 3/4.4.6 STEAM GENERATORS . . . . . ... 3/4 4-9 3/4.4.7 REACTOR COOLANT SYSTEM LEAKAGE Operational Leakage * . . .. .. . . .
Leakage Detection System 3/4 4-16 3/4 4-17 3/4.4.8 CHEMISTRY * . . . . . . . . . . . . . . . .... 3/4 4-20 3/4.4.9 SPECIFIC ACTIVITY .... 3/4 4-23 3/4 ."4 .10 PRESSURE/TEMPERATURE LIMITS Reactor Coolant System 3/4 4-27 Pressurizer * .....
Overpressure Protection Systems 3/4 4-30 3/4 4-31 3/4.4.11 STRUCTURAL INTEGRITY ASME Code Class 1~ 2, and 3 Components ........ 3/4 4-3.3
SALEM - UNIT 2 v
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 - Tavg 350°F . . . .
~ .... 3/4 5-3 3/4.5.3 ECCS SUBSYSTEMS - Tavg < 350°F . . . . 3/4 5-7 3/4.5.4 BORON INJECTION SYSTEM Heat Tracing . . . . .. . . . .
Boron Injection Tank .... 3/4 5-9 3/4 5-10 3/4.5.5 REFUELING WATER STORAGE TANK . . . . . .... 3/4 5-11 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Containment Integrity . 3/4 6-1 Containment Leakage
Containment Air Locks 3/4 3/4 6-2 6-4 Internal Pressure * .. ...... 3/4 6-6 Air Temperature * ............ .... 3/4 6-7 Containment Structural Integrity Containment Ventilation System . .....
3/4 3/4 6-8 6-9 -
3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS Containment Spray System ...
Spray Additive System * . *
Containment Cooling System
.. . . .. ... ... ... . . . . . 3/4 6-10 3/4 6-11 3/4 6-12 3/4.6.3 CONTAINMENT ISOLATION VALVES .... 3/4 6-14 3/4.6.4 COMBUSTIBLE GAS CONTROL Hydrogen Analyzers 3/4 6-21 Electric Hydrogen Recombiners * ..... ..... 3/4 6-22
SALEM - UNIT 2 VI
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE Safety Valves ........ .... 3/4 7-1 Auxiliary Feedwater System 3/4 7-5 Activity . . . . . . . . . .. .. . . .. .. ..
Auxiliary Feed Storage Tank
Main Steam Liner Isolation Valves . ..
3/4 3/4 3/4 7-7 7-8 7-10 3/4.7.2 STEAM GENERATOR PRESSURE/TEMPERATURE LIMITATION 3/4 7-11 3/4.7.3 COMPONENT COOLING WATER SYSTEM .... 3/4 7-12 3/4.7.4 SERVICE WATER SYSTEM ...... 3/4 7-13 3/4.7.5 FLOOD PROTECTION . . . .... 3/4 7-14 3/4.7.6 CONTROL ROOM EMERGENCY AIR CONDITIONING SYSTEM . . . . 3/4 7-15 3/4.7.7 AUXILIARY BUILDING EXHAUST AIR FILTRATION SYSTEM 3/4 7-18 3/4.7.8 SEALED SOURCE CONTAMINATION
3/4 7-21 3/4.7.9 SNUB BERS ......... ...... 3/4 7-23 3/4.7.10 FIRE SUPPRESSION SYSTEMS Fire Suppression Water System 3/4 7-31 Spray and/or Spri nk 1er System * . 3/4 7-34 Low Pressure C02 Systems 3/4 7-36 Fire Hose Stations .. .. .... ..
~ 3/4 7-37 3/4.7.11 PENETRATION FIRE BARRIERS
3/4 7-39 SALEM - UNIT 2 VII
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 8-7 3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEMS A.C. Distribution - Operating
3/4 8-8 A.C. Distribution - Shutdown ** 3/4 8-9 125-Volt D.C. Distribution - Operating 3/4 8-10 125-Volt D.C. Distribution - Shutdown ** 3/4 8-12 28-Volt D.C. Distribution - Operating
3/4 8-13 28-Volt D.C. Distribution - Shutdown ** 3/4 8-15 3/4.8.3 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES Containment Penetration Conductor Overcurrent Protective Devices * * * * * * * * * * * *
3/4 8-16 SALEM - UNIT 2 VIII
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 PEN.ETRATIONS
3/4 9-4 3/4.9.5 COMMUNICATIONS
3/4 9-5 3/4.9.6 MANIPULATOR CRANE
3/4 9-6 3/4.9.7 CRANE TRAVEL - FUEL HANDLING AREA
3/4 9-7.
3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION All Water Levels * * * * * * * * * * * * *
3/4 9-8 Low Water Leve 1 * * * * * * * .* * * * * * *. *
3/4 9-9 3/4.9.9 CONTAINMENT PURGE AND PRESSURE-VACUUM RELIEF ISOLATION SYSTEM * * * * * * * *** 3/4 9-10 3/4.9.10 WATER LEVEL - REACTOR VESSEL 3/4 9-11 3/4.9.11 STORAGE POOL WATER LEVEL ** 3/4 9-12 3/4.9.12 FUEL HANDLING AREA VENTILATION SYSTEM 3/4 9-13 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-.4 3/4.10.4 NO FLOW TESTS 3/4 10-5 3/4.10.5 POSITION INDICATION SYSTEM - SHUTDOWN
3/4 10-6 SALEM - UNIT 2 IX
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS Concentration ** ...... ..... 3/4 3/4 11-1 11-5 Dose * * * * * * * * *
Liquid Radwaste Treatment ..... 3/4 3/4 11-6 11-7 Liquid Holdup Tanks *****
3/4.11.2 GASEOUS EFFLUENTS Dose Rate * * * * * * * . * * * *
3/4 11-8 Dose-Noble Gases * * * * * * * *
3/4 11-12 Dose-Iodine-131, Tritium, and Radionuclides in Particulate Form ** 3/4 11-13 Gaseous Radwaste Treatment * * *
3/4 11-14 Explosive Gas Mixture
3/4 11-15 Gas Storage Tanks *** 3/4 11-16 3/4.11.3 SOLID RADIOACTIVE WASTE
3/4 11-17 3/4.11.4 TOTAL DOSE ****** 3/4 11-18 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM 3/4 12-1 3/4.12.2 LAND USE CENSUS ** 3/4 12-11 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM ** 3/4 12-13
SALEM - UNIT 2 x
INDEX BASES SECTION 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 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR and 3/4.2.3 RCS FLOW RATE AND R * * * *
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 2-5
~-
SALEM - UNIT 2 . XI
INDEX BASES SECTION PAGE 3/4.3 INSTRUMENTATION 3/4.3.1 PROTECTIVE INSTRUMENTATION B.3/4 3-1 3/4.3.2 ENGINEERED SAFETY FEATURES (ESF)
INSTRUMENTATION * * * *
B 3/4 3-1 3/4.3.3 MONITORING INSTRUMENTATION B 3/4 3-1 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION ..... B 3/4 4-1 3/4.4.2 and SAFETY VALVES ................ ... .. B 3/4 4-2 3/4.4.3 3/4.4.4 PRESSURIZER .... .... B 3/4 4-2 3/4.4.5 RELIEF VALVES .... .... ..... B 3/4 4-2 3/4.4.6 STEAM GENERATORS .... B 3/4 4-3 3/4.4.7 REACTOR COOLANT SYSTEM LEAKAGE .... ..... B 3/4 4-4 3/4.4.8 CHEMISTRY * . . . . .... B 3/4 4-5 3/4.4.9 SPECIFIC ACTIVITY * . . . . . ........... B 3/4 4-6 3/4.4.10 PRESSURE/TEMPERATURE LIMITS * .... B 3/4 4-7 3/4.4.11 STRUCTURAL INTEGRITY . . . . ... B 3/4 4-17
SALEM - UNIT 2 XII
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 BORON INJECTION SYSTEM B 3/4 5-2 3/4.5.5 REFUELING WATER STORAGE TANK (RWST) *
B 3/4 5-3 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT B 3/4 6-1 3/4.6.2 DEPRESSURIZATION AND COOLLNG SYSTEMS B 3/4 6-3 3/4.6.3 CONTAINMENT ISOLATION VALVES B 3/4 6-4 3/4.6.4 COMBUSTIBLE GAS CONTROL B 3/4 6-4 SALEM - UNIT 2 XIII
~~~-~~--=---=--=--=*=---=--**~-~---=---=----'!'::_=~_:_".::._ __ -:_-_-~--=~----*------------------- --
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. p CONTROL ROOM EMERGENCY AIR CONDITIONING SYSTEM B 3/4.7-5 3/4.7.7 AUXILIARY BUILDING *EXHAUST AIR FILTRATION SYSTEM B 3/4 7-5 3/4.7.8 SEALED SOURCE CONTAMINATION B 3/4 7-5 3/4.7.9 SNUBBERS * * * * * * *
B 3/4 7-6 3/4.7 .10 FIRE SUPPRESSION SYSTEMS B 3/4 7-7 3/4.7.11 PENETRATION FIRE BARRIERS
B 3/4 7-8 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A. C. SOURCES and AND 3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEMS B 3/4 8-1 3/4.8.3 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES
B 3/4 8-1 SALEM - UNIT 2 *XIV
INDEX BASES SECTION PAGE 3/4.9 REFUELING OPERATIONS 3/4.9.l BORON CONCENTRATION ........ B 3/4 9-1
3/4.9.2 INSTRUMENTATION
B 3/4 9-1 3/4.9.3 DECAY TIME ** ** B 3/4 9-1 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS B 3/4 9-1 3/4.9.5 COMMUNICATIONS ** .... B 3/4 9-1 3/4.9.6 MANIPULATOR CRANE
B 3/4 9-2 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE BUILDING B 3/4 9-2 3/4.9.8 RESIDUAL. HEAT REMOVAL AND COOLANT CIRCULATION B 3/4 9-2 3/4.9.9 CONTAINMENT PURGE AND PRESSURE-VACUUM RELIEF ISOLATION SYSTEM * * * * * * * * * * * * *
B 3/4 9-2 3/4.9.10 WATER LEVEL - REACTOR VESSEL and AND 3/4.9.11 STORAGE POOL ********* B 3/4 9-3 3/4.9.12 FUEL HANDLING AREA VENTILATION SYSTEM
B 3/4 9-3 3/4.10 . SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN **** -B 3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS **** 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 3/4.10.5 POSITION INDICATION - SHUTDOWN B 3/4 10-1 SALEM - UNIT 2 xv
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 SOLID RADIOACTIVE WASTE
B 3/4 11-6 3/4.11.4 TOTAL DOSE ****** B 3/4 11-7 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM B 3/4 12-1 3/4.12.2 LAND USE CENSUS *** B 3/4 12-1 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM B 3/4 12-2
SALEM - UNIT 2 XVI
INDEX DESIGN FEATURES SECTION PAGE 5.1 SITE Exclusion Area ......... 5-1 Low Population Zone .......... .. 5-1 5.2 CONTAINMENT Con f i gu ration . ........ ..... 5-1 Destgn Pressure and Temperature .......... 5-4 5.3 REACTOR CORE Fuel Assemblies Control Rod Assemblies .....
. . .. . . . . . . . . . 5-4 5-4 5.4 REACTOR COOLANT SYSTEM Design Pressure and Temperature
. . .. .. . . . . . . . 5-4 Volume
5-4 5.5 METEOROLOGICAL TOWER LOCATION ......... 5-5 5.6 FUEL STORAGE Cr'iticality * ..... ........ 5-5 Drainage .. .... 5-5 Capacity . . . . . . . . .... .... ..... 5-5 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT 5-5
SALEM - UNIT 2 XVII
ADMINISTRATIVE CONTROLS INDEX SECTION PAGE 6.1 RESPONSIBILITY 6-1 6.2 ORGANIZATION Offsite ****** 6-1 Facility Staff * ** *.* 6-1 Shift Technical Advi.sor 6-6 I.
6.3 FACILITY STAFF QUALIFICATIONS 6-6 6.4 TRAINING * * * * * * * * * * * ...... .....
. 6-6 6.5 REVIEW AND AUDIT 6.5.1 STATION OPERATIONS REVIEW COMMITTEE Function ****** 6-6.
Composition ***** 6-7 Alternates * * * * * * * * *
6-7 Meeting Frequency 6-7 Quorum .***** 6-7 Responsibilities **** 6-7 Authority **** 6-8 Records * * * * * . ..
~ 6-9 6.5.2 NUCLEAR REVIEW BOARD Function 6-9 Composition Alternates
. 6-9 6-9 Consultants **** 6-9 Meeting Frequency
6-10 Quorum 6-10 Review 6-10 Audits 6-10 Authority
6-11 Records *
6-11 SALEM - UNIT 2 XVIII
INDEX ADMINISTRATIVE CONTROLS

~--------------------------------------------------------------------

SECTION PAGE 6.5.3 SAFETY REVIEW GROUP .................... 6-12 6.6. REPORTABLE OCCURRENCE ACTION 6-12a 6.7 SAFETY LIMIT VIOLATION
6-12a 6.8 PROCEDURES AND PROGRAMS . ................. 6-13 6.9 REPORTING REQUIREMENTS 6.9.l ROUTINE REPORTS AND REPORTABLE OCCURRENCES ... 6-15 6.9.2 SPECIAL REPORTS ...................... 6-21 6.10 RECORD RETENTION 6-21 6.11 RADlATION PROTECTION PROGRAM .... 6-23 6.12 HIGH RADIATION AREA ** 6-23 -
6.13 PROCESS CONTROL PROGRAM .................. . 6-24 6.14 OFFSITE DOSE CALCULATION MANUAL 6-24 6.15 MAJOR CHANGES TO RADIOACTIVE LIQUID, GASEOUS AND SOLID WASTE TREATMENT SYSTEMS 6-25
SALEM - UNIT 2 XIX
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 with the necessary range and accuracy to known values of the parameter which the channel monitors. The CHANNEL CALIBRATION shall encompass the entire channel including the sensor and alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST. The CHANNEL CALIBRATION may be performed by any series of sequential, overlapping or total channel steps such 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 i'nclude, 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 2 1-1
DEFINITIONS CONTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when:
1.7.1 All penetrations required to be closed durin[ accident conditions are either:
a~ Capable ob being closed by an OPERABLE containment automatic isolation valve system~ or
b. Closed by manual valves, blind flanges, or deactivated automatic vlaves secured in their closed positions, except as provided in Table 3.6-1 of 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) i~ OPERABLE. *
. CONTROLLED LEAKAGE 1.8 CONTROLLED LEAKAGE shall .be that seal water flow from the reactor coolant pump seals.
CORE ALTERATION 1.9 CORE ALTERATION shall be the movement or man1pulation of any component within the reactor pressure vessel with the vessel head removed and fuel in the
.vessel. Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position.
DOSE EQUIVALENT I-131 1.10 DOSE EQUIVALENT I-131 shall be that concentration of I-i31 (microcuries per gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The SALEM - UNIT 2 1-2
"DEF IN IT IONS thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844 Calculation of Distance Factors for Power and Test Reactor Sites. 11 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 pe~ 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.
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 CONTROLLED LEAKAGE) 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 2 1-3
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.
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 shall be that manual which contains the current methodology and parameters used in the calculation. of offsite doses due to radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm/trip setpoints, and in the conduct of the env~ronmental radiological monitoring program.
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 function(s), and when*
all necessary attendant instrumentation, controls, a normal and an emergency electrical power source, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function(s) are also capable of performing their related support function(s).
OPERATIONAL MODE - MODE 1.19 An OPERATIONAL MODE (ie., 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 2 1-4
DEFINITIONS PHYSICS TESTS 1.20 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the re~ctor 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 steam generator tube 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 shall be that program which contains the current formula, s~mpling, analyses, test, and determinations to be made to ensure that the 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 Part 20, 10 CFR Part 71 and Federal and State regulations and other requirements governing the disposal of the 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, cir 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 sha*ll be the ratio of the maxi.mum 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 3411 MWt.
SALEM - UNIT 2 1-5
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 OCCURRENCE 1.27 A REPORTABLE OCCURRENCE shall be any* of those conditions specified in Specifications 6.9.1.8 and 6.9.1.9.
SHUTDOWN MARGIN 1.28 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming all full length rod cluster assemblies (shutdown arid 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, 1eased, or otherwise cont ro l1 ed by the 1i censee, as shown in Figure 5.1-3, and which defines the exclusion* area as shown in Figure 5.1-1.
SOLIDIFICATION 1.30 SOLIDIFICATION shall be the conversion of wet radioactive wastes into a form t~at meets shipping and burial ground requirements.
SOURCE CHECK 1.31 SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity.
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 2 1-6
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 sh~ll be all leakage which is not IDENTIFIED LEAKAGE or CONTROLLED 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 p~rposes.
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 adsorb~rs and/br 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 2 1-7
DEFINITIONS TABLE 1.1 OPERATIONAL MODES REACTIVITY AVERAGE COOLANT MODE CONDITION, Keff THERMAL POWER* TEMPERATURE
1. POWER OPERATION > 0.99 > 5% > 350°F
2. STARTUP > 0.99 < 5% > 350°F
3. HOT STANDBY < 0.99 0 > 350°F
4. HOT SHUTDOWN < 0.99 0 350°F > Tavg
> 200°F 5
COLD SHUTDOWN < 0.99 0 < 200°F
6. REFUELING** < 0.95 0 < 140°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 2 1-8
DEFINITIONS TABLE 1.2 FREQUENCY NOTATION NOTATION FREQUENCY s At 1east once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
D At 1east once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
w At 1east once per 7 days.
M At 1east once per 31 days.
Q At 1east once per 92 days.
SA At 1east once per 6 months.
R At 1east once per 18 months.
S/U Prior to each reactor startup.
p Prior to each release.
N.A. Not applicable.
SALEM - UNIT 2 1-9
INSTRUMENTATION RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.8 The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.3-12 shall be OPERABLE with their alarm/trip setpoints set to ensure that the limits of Specification 3.11.1.1 are not exceeded. The alarm/trip setpoints of these channels shall be determined in accordance with the OFFSITE DOSE CALCULATION MANUAL (ODCM).
APPLICABILITY: At all times.
ACTION:
a. With a radioactive liquid effluent monitoring instrumentation channel alarm/trip setpoint less conservative than required by the above specification, without delay suspend the release of radioactive liquid effluents monitored by the affected channel or declare the channel inoperable or change the setpoint so it is acceptably conservative.
b. With less than the minimum number of radioactive 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 semi-annual radioactive effluent release report why the inoperability was not corrected in a timely manner.
c. The provisions of Specifications 3.0.3, 3.0.4, and 6.9.1.9.b are not applicable.
SURVEILLANCE REQUIREMENTS 4.3.3.8 Each radioactive liquid effluent monitoring instrumentation channel shall be demor:istrated OPERABLE* by performance ofJ:he CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3-12.
SALEM - UNIT 2 3/4 3-53
"TABLE 3.3-12 c
z
....... RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION
-i N
MINIMUM CHANNELS INSTRUMENT OPERABLE ACTION
1. GROSS RADIOACTIVITY MONITORS PROVIDING AUTOMATIC TERMINATION OF RELEASE
a. Liquid Radwaste Effluent Line (2-R18) . 1 26
b. Steam Generator Slowdown Line 4 27 (2-R19 A, B, C, and D)
-w
. p.
w U1 I
2. GROSS RADIOACTIVITY MONITORS NOT PROVIDING AUTOMATIC TERMINATION OF RELEASE
.p.
a. Containment Fan Coolers - Service Water Line 5 28 (2-R13 A, B, C, D, E) Discharge
b. Chemical Waste Basin Line "(R37) 1 28
3. FLOW RATE MEASUREMENT DEVICES
a. Liquid Radwaste Effluent Line 1 29
b. Steam Generator Blowdown Line 4 29
4. TANK LEVEL INDICATING DEVICES
a. Temporary Outside Storage Tanks as Required 1 30 1,
I '
TABLE.3.3-12 (Continued)
TABLE NOTATION ACTION 26 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases may continue provided that prior to initiating a release:
a. At least two independent samples are analyzed in accordance with Specification 4.11.1.1.3, and
b. At least two technically qualified members of the Facility Staff independently verify the release rate calculations and discharge line valving; Otherwise, suspend release of radioactive effluents via this pathway.
ACTION 27 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided grab samples are analyzed for gross radioactivity (beta or gamma) at a limit of detection of at least 10-7 microcuries/gram:
a. At leasl once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months when the specific activity of the secondary coolant is greater than 0.01 microcuries/gram DOSE EQUIVALENT I-131.
b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the specific activity of the secondary coolant is less than or equal to 0.01 microcuries/
gram DOSE EQUIVALENT I-131.
ACTION 28 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that, at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , grab samples are collected and analyzed for gross radioactivity (beta or gamma) at a limit of detection of at least 10-7 microcuries/gram.
SALEM - UNIT 2 3/4 3-55
TABLE 3.3-12 (Continued)
TABLE NOTATION ACTION 29 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during actual releases. Pump performance curves may be used to estimate flow.
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 2 3/4 3-56
(/')
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TABLE 4.3-12
z
--t RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS N
CHANNEL CHANNEL SOURCE CHANNEL FUNCTIONAL INSTRUMENT CHECK CHECK CALIBRATION TEST
1. GROSS RADIOACTIVITY MONITORS PROVIDING ALARM AND AUTOMATIC TERMINATION OF RELEASE
a. Liquid Radwaste Effluent Line {2-R18) D P# R{3) Q{l)
b. Steam Generator Blowdown Line D M R(3) Q(l)
(2-R19 A, B, C, and D) w I* ........
~
2* GROSS RADIOACTIVITY MONITORS PROVIDING ALARM w
BUT NOT PROVIDING AUTOMATIC TERMINATION OF I
U'1 RELEASE
a. Containment Fan Coolers - Service Water Line D M R(3) Q(2)
(2-Rl3 A, B, C, D, E) Discharge
b. Chemical Waste Basin Line (R37) D M R(3) Q(2)
3. FLOW RATE MEASUREMENT DEVICES
a. Liquid Radwaste Effluent Line D(4) N.A. R N.A.
b. Steam Generator Blowdown Line D(4} N.A. R N.A.
4. TANK LEVEL INDICATING DEVICES**
a. Temporary Outside Storage Tanks as Required D* N.A. R Q I, . I I I
TABLE" 4.3~12 {Continued)
TABLE NOTATION (1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if any of the following conditions exist:
. 1. Instrument. indicates measured levels above the alarm/trip setpoint.
2. Circuit failure. (Loss of Power)
3. Instrument indicates a downscale failure. (Alarm Only)
(2) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if. any of the following conditions exist: "
1. Instrument indicates measured levels above the alarm/trip setpoint.
2. Circuit failure. (Loss of Power)
3. Instrument indicates a downscale failure *
(3)
4. Instrument controls not set in operate mode.
The initial CHANNEL CALIBRATION was performed using appropriate liquid or gaseous calibration sources obtained from reputable suppliers. The activity of the calibration sources were reconfirmed using a multi-channel analyzer which was calibrated using one or more NBS standards.
(4) CHANNEL ,.CHECK shall consist of verifying indication of flow during periods of release. CHANNEL CHECK shall be made at least on*ce per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months on days on which continuous, periodic, or batch releases are made. -
During "liquid additions to the tank.
If tank level indication is not provided, vertification will be done by visual inspection.
The R18 channel is an in-line channel which requires periodic decontamination. Any count rate i ndi cation above 10 ,000 cp~ constitutes a CHANNEL CHECK for compliance purposes.
SALEM - UNIT 2 3/4 3-58
INSTRUMENTATION RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1.9 The radioactive gaseous effluent monitoring instrumentation channels shown in Table 3.3-13 shall be OPERABLE with their alarm/trip setpoints set to ensure that the limits of Specification 3.11.2:1 are not exceeded. The alarm/trip setpoints of these channels shall be determined in accordance with the {ODCM).
APPLICABILITY: As shown in Table 3.3-13 ACTION:
a. With a radioactive gaseous effluent monitoring instrumentation channel alarm/trip setpoint less conservative than required by the above specificatiGn, without delay suspend the release of radioactive gaseous effluents monitored by the affected channel or declare the channel inoperable or change the setpoint so it is acceptably conservative.
b. With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3-13. Exert best efforts to return the instrument to operable status within 30 days and, if unsuccessful, explain in the next semi-annual radioactive effluent release report why the inoperability was not corrected in a timely manner.
c. The provisions of Specifications 3.0.3, 3.0.4, and 6.9.1.9.b are not applicable.
SURVEILLANCE REQ~IREMENTS 4.3.3.9 Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3-13.
SALEM - UNIT 2 3/4 3-59
TABLE 3.3-13 c
z
....... RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION
--i N MINIMUM CHANNELS INSTRUMENT OPERABLE APPLICABILITY ACTION
1. WASTE GAS HOLDUP SYSTEM
a. Noble Gas Activity Monitor - Providing Alarm and Automatic Termination,of Release 1
31 (2-R41C)
b. Oxygen Monitor 1 ** 35 w 2. PLANT VENT HEADER SYSTEM#
.p.
w a. Noble Gas Activity Monitor (2-R16 or 2-R41C) 1
33 &34 I
())
a b. Iodine Sampler 1
36
c. Particulate Sampler 1
36
d. Flow Rate Monitor 1
32
e. Sampler Flow Rate Monitor 1
32
The following process streams are routed to the plant vent where they are effectively monitored by the instruments described:
(a) Condenser Air Removal System (b) Auxiliary Building Ventilation System (c) Fuel Handling Building Ventilation System (d) Radwaste Area Ventilation System (e) Containment Purges Action item #34 applies to the purging of the containment only.
I, I , I
TABLE 3.3-13 (Conti~ued)
TABLE NOTATION ACTION 31 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement~ the contents of the tank(s) may be released to the environment provided that prior to initiating the release:
a. *At least two independent samples of the tank's contents are analyzed, and *
b. At least two technically qualified members of the Facility Staff independently verify the release rate calculations and discharge valving lineup; Otherwise, suspend release of radioactive effluents via this pathway.
ACTION 32 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .
ACTION 33 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided grab samples are taken at least once
per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and these samples are analyzed for gross activity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
ACTION 34 - With the number of channels OPERABLE 1ess than required by the Minimum Channels OPERABLE requirement, immediately suspend PURGING of radioactive effluents via this pathway.
At all times, other than when the li~e is v~lved out and locked.
During waste gas holdup system operation.
SALEM - UNIT 2 3/4 3-61
TABLE 3.3-13 (Continued)
TABLE NOTATION ACTION 35 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, operation of the waste gas holdup system may continue provided grab samples are collected at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and analyzed within the following 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .
ACTION 36 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via the effected pathway may continue provided samples are continuously collected with auxiliary sampling equipment as required in Table 4.11-2.
SALEM - UNIT 2 3/4 3-62
TABLE 4.3-13 c
z
-i RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS N
CHANNEL MODES IN WHICH CHANNEL SOURCE CHANNEL FUNCTIONAL SURVEILLANCE INSTRUMENT CHECK CHECK CALIBRATION TEST REQUIRED
1. WASTE GAS HOLDUP SYSTEM
a. Noble Gas Activity Monitor -
Providing Alarm and Automatic Termination of Release (2-R41C) p p R(3) Q(l) *
b. Oxygen Monitor D N.A. Q(4) M **
-w
~
w I
O'I
2. PLANT VENT HEADER SYSTEM#
a. Noble Gas Activity Monitor (2-R16 D M R(3) Q(2)
w or 2-R41C)
b. Iodine Sampler w N.A. N.A. N.A. *
c. Particulate Sampler w N.A. N.A. N.A. *
d. Flow Rate Monitor D- N.A. R N.A. *
e. Sampler Flow R~te Monitor w N.A. R N.A. *
The following process streams are routed to the plant vent where they are effectively monitored by the instruments described:
(a) Condenser Air Removal System (b) Auxiliary Building Ventilation System (c) Fuel *Handling Building Ventilation System (d) Radwaste Area Ventilation System (e) Containment Purges 1, I I I.
v TABLE 4.3-13 (Continued)
TABLE NOTATION (1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if any of the following conditions exist:
1. Instrument indicates*measured levels above the alarm/trip setpoint.
2. Circuit failure. (Loss of Power)
3. Instrument indicates a downscale failure. (Alarm Only)
(2) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exist:
1. Instrument indicates measured levels above the alarm/trip setpoint *
2. Circuit failure. (Loss of Power)
3. Instrument indicates a downscale failure.
4. Instrument controls not set in. operate mode.
(3) _The initial CHANNEL CALIBRATION was performed using appropriate liquid or gaseous calibration sources obtained from reputable suppliers. The activity of the calibration sources were reconfirmed using a multi-channel analyzer which was calibrated using one or more NBS standards.
(4) The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal:
1. One volume percent oxygen, balance nitrogen, and 2.. Four volume percent oxygen, balance nitrogen.
At all times
During waste gas holdup system operation.
SALEM - UNIT 2 3/4 3-64
RADIOACTIVE EFFLUENTS LIQUID HOLDUP TANKS*
LIMITING 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 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 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 at least once per 7 days 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 cont~nts and that do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system.
SALEM - UNIT 2 3/4 11-7
INSTRUMENTATION 3/4.3.4 TURBINE OVERSPEED PROTECTION LIMITING CONDITION FOR OPERATION 3.3.4 At least one turbine overspeed protection system shall be OPERABLE.
APPLICABILITY: MODES 1, 2 and 3.
ACTION:
a. With one stop valve or one control v~lve per high pressure turbine steam lead inoperable and/or with one reheat stop valve or one reheat intercept valve per low pressure turbine steam lead inoperable, restore the inoperable valves(s) to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or close at least one valve in the affected steam lead; otherwise, isolate the turbine from the steam supply within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
b. With the above required turbine overspeed protection system otherwise inoperable, within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months either restore the system to OPERABLE status or isolate the turbine from the steam supply.
SURVEILLANCE REQUIREMENTS 4.3.4.1 The provisions of Specification 4.0.4 are not applicable.
4.3.4.2 The above required turbine overspeed protection system shall be demonstrated OPERABLE; (1) prior to admitting steam to the turbine during each startup unless performed within the past 7 days, (2) within 24* hours of attaining greater than or equal to 85% of RATED THERMAL POWER, and (3) at least once per 7 days while ope~ating at greater than.or equal to 85% of RATED THERMAL POWER by cycling each of the following valves through at least one complete cycle from the running position.
a. Four high pressure turbine stop valves.
b. Four high pressure turbine control valves.
c. Six low pressure hot reheat stop valves_.
d. Six low pressure hot reheat intercept valves.
SALEM - UNIT 2 3/4 3-65
INSTRUMENTATION SURVEILLANCE.REQUIREMENTS (Continued) 4.3.4.3 The above required turbine overspeed protection system shall be demonstrated OPERABLE:
a. At least on~e per 31 days, while operating at greater than or equal to 85% of RATED THERMAL POWER, by directing observatio~ of the movement of each of the above valves through one complete cycle from the running position.
b. At least once per 18 months by performance of a CHANNEL CALIBRATION on the turbine overspeed protection systems.
c. At least one per 40 months by disassembling at least one of each of the above valves and performing a visual and surface inspection of valve seats, disks and stems and verifying no unacceptable flaws or corrosion.
I l'*
SALEM - UNIT 2 3/4 3-66
3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS CONCENTRATION LIMITING CONDITION FOR OPERATION 3.11.1.1 The concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS (See Figure 5.1-3) shall be 1imited to the concentrations specified in 10 CFR Part 20; Appendix B, Table II, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 2 x 10-4 microcuries/ml.
APPLICABILITY: At all times.
ACTION:
With the concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS exceeding the above limits, without delay restore the concentration to within the above limits.
SURVEILLANCE REQUIREMENTS 4.11.1.1.1 Radioactive liquid wastes shall be samples and analyzed according to the sampling and analyses program in Table 4.11-1.
4.11.1.1.2 The results of the radioactivity analyses shall be used in accordance with in the ODCM to assure that the concentrati o*ns at the point of release are maintained within the limits of.Specification 3.11.1.1 SALEM - UNIT 2 3/4 11-1
TABLE 4.11-1 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM Lower Limit Minimum of Detection Liquid Release Sampling Analysis Type of Activity (LLD)a Type Frequency Frequency Analysis {uCi /ml)
A. Batch Waste p p Release Each Batch Each Batch Principal Gamma 5xl0-7 Tanksb Emittersc I-131 lxl0-6 p M Dissolve and lxlQ-5 One Batch/M Entrained Gases (Gamma Emitters) p M H-3 lxl0-5 Each Batch Composited Gross Alpha lxl0-7
B. Continuous p
Each Batch Q
Composited w
Sr-89, Sr-90 Fe-55 Principal Gamma 5xlo.,.8 lxl0-6 5x10- 7 Releasese Weekly Composite Emittersc
1. Steam I-131 lxl0-6 Generator Blowdciwn M M Dissolved and lxl0-5 Grab Sample Entrained Gases Weekly M H-3 lxl0-5 Composite Gross Alpha lxl0-7 Weeklyd Q Sr-89, Sr-90 5xl0-8 Composite Fe-55 lxl0-6 SALEM - UNIT 2 3/4 11-2 l --
TABLE 4.11-1 (Continued)
TABLE NOTATION
a. The LLD is defined, for purposes of these specifications as the smallest concentration of radioactive material in a sample that will yield a net count (above system background) that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a 11 real" signal.
For a particular measurement system (which may include radiochemical sep*aration):
4.66 Sb LLD ~ ~~~~~~~~~~~~~~~~~~~
E
V .2.22x106
s;
e (-A.11 t)
Where:
LLD is the 11 a priori 11 lower limit of detection as defined above (as microcuries per unit mass or volume),
Sb is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute),
E is the counting efficiency (as counts per disintegration),
Y is the sample size (in units of mass or volume),
2.22 x 106 is the number of disintegrations per minute per microcurie, s; is the fractional radiochemical yield (when applicable),
A. is the radioactive decay constant for the particular radionuclide, and 11 t for environmental samples is the elapsed time between sample collection (or end of the sample collection period) and time of counting.
Typical values of E, V, s;, and 11t should be used in the calculation.
It should be recognized that the LLD is defined as an ~priori (before the fact) limit representing the capability of a measurement system and not as an~ posteriori (after the fact) limit for a particular measurement.
SALEM - UNIT 2 3/4 11-3
TABLE 4.11-1 (tontinued)
TABLE NOTATION
b. A bat~h release is the discharge of liquid wastes* of a discrete volume.
Prior to sampling for analyses, each batch shall be isolated, and then thoroughly mixed to assure representative sampling. *
c. The principal gamma emitters for which the LLD specification applies
exclusively are the following radionuclides: Mn-54, Fe-59, *Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144. This list.does not mean that only these nµclides are to be detected and reported. Other peaks that are measurable and identifiable, together with the above nuclides, shall
also be identified and reported.
d. A composite sample is one in which the quantity of liquid sampled is.
proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen that is representative of the liquids rel_eased.
e. A continuous release is the discharge of liquid wastes of a nondiscrete volume, e.g., from a volume of a system that has an input flow during 'the continuous release.
SALEM - UNIT 2 3/4 11-4
RADIOACTIVE EFFLUENTS DOSE LIMITING CONDITION FOR OPERATION 3.11.1.2 The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released, from each reactor unit, to UNRESTRICTED AREAS (see Figure 5.1-3) shall be limited:
a. During any calendar quarter to less than or equal to 1.5 mrems to the total body and to less than or equal to 5 mrems to any organ, and
b. During any calendar year to less than or equal to 3 mrems to the total body and to less than or equal to 10 mrems to any organ.
APPLICABILITY: At all times.
ACTION:
a. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, in lieu of a Licensee Event Report, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit(s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.
b. The provisions of Specification 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.1.2 Cumulative dose contributions from liquid effluents shall be determined in accordance with the ODCM at least once per 31 days.
SALEM - UNIT 2 3/4 11-5
'RADIOACTIVE EFFLUENTS LIQUID RADWASTE TREATMENT LIMITING CONDITION FOR OPERATION 3.11.1.3 The liquid radwaste treatment system shall be used to reduce the radioactive materials liquid wastes prior to their discharge when the projected cumulative doses due to the liquid effluent from each reactor to UNRESTRICTED AREAS (see Figure 5.1-3) exceed 0.375 mrem to the total body or 1.25 mrem to any organ* during any calendar quarter.
APPLICABILITY: At all times.
I ACTION:
I.
a. With the radioactive liquid waste being discharged without* treatment and in excess of the above limits, in lieu of a Licensee Event Report, prepare and submit to the Commission within 30 days pursuant to Specification 6.9.2 a Special Report that includes the following information: -
1. Explanation of why liquid radwaste was being discharged without treatment, identification of any inoperable equipment or subsystems, and .the reason for the inoperability.
2. Action(s) taken to restore the inoperable equipment to OPERABLE status, and
3. Summary description of action(s) taken to prevent a recurrence.
b. The provisions of Specification 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS -*
4.11.1.3 Doses due to liquid releases shall be projected at least once per 31 days in accordance with the ODCM.
SALEM - UNIT 2 3/4 11-6
- ---------~-- -------~~~~~~~~-~-~
RADIOACTIVE EFFLUENTS LIQUID HO~DUP TANKS*
LIMITING 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 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS*
4.11.1.4 The quantity of radioactive material contained in each outdoor temporary tank shall be determined to be within the above 1imi t by analyzing a representative sample of the tank's contents.at least once per 7 days 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 2 3/4 11-7
RADIOACTIVE EFFLUENTS 3/4.11.2 GASEOUS EFFLUENTS DOSE RATE LIMITING CONDITION FOR OPERATION 3.11.2.1 The dose rate due to radioactive materials released in gaseous effluents from the site to areas at and beyond the SITE BOUNDARY (see Figure 5.1-3) shall be limited to tha following:
a. For noble gases: Less than or equal to 500 mrems/yr to the total body and less than or equal to 3000 mrems/yr to the skin, and
b. For iodine-131, for tritium, and for all radionuclides in particulate form with half lives greater than 8 days: Less than or equal to 1500 mrems/yr to any organ.
APPLICABILITY: At all times.
ACTION:
With the dose rate(s) exceeding the above limits, without delay restore the release rate to within the above limit(s).
SURVEILLANCE REQUIREMENTS 4.11.2.1.1 The dose rate due to noble gases in gaseous effluents shall be determined continuously to be within the above limits in accordance with the ODCM.
4.11.2.1.2 The dose rate due to iodine-131, tritium, and all radionuclides in particulate form with half lives greater than 8 days in gaseous effluents. shall be determined to be within the above limits in accordance with the ODCM by obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified in Table 4.11-2.
SALEM - UNIT 2 3/4 11-8
TABLE 4.11-2 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM Lower Limit Minimum of Detection Gaseous Release Sampling Analysis Type of Activity (LLD)
Type
Frequency Frequency Analysis (uCi /ml)
A. Waste Gas p p Storage Each Tank Each Tank ~rincipal Gamma lxl0-4 Tank Grab Sample Emittersb B. Containment p p Principal Gamma lxl0-4 PURGE Each PURGE Each PURGE Emittersb Grab Sample H-3 lxl0-6 Principal Gamma lxl0-4 C. Pl ant Vent MC d e MC Emitterb Grab' ' Sample H-3 lxl0-6 D. All Release Continuousf wg I-131 lxlo-12 Types as Charcoal Listed in A, Sample B, and C Above Continuousf wg Principal Gamma 1-10-11 Particulate Emittersb
... Sample (I-131, Others)
Continuousf M Gross Alpha lxl0-11 Composite
. Particulate Sample Continuousf Q Sr-89, Sr-90 lxl0-11 Composite Particulate Samp-1 e Continuousf Noble Gas Noble Gasses lxl0-6 Monitor Gross Beta or Gamma SALEM - UNIT 2 3/4 11-9

-*----*- ------ ----- ------~ -- ------*--------

TABLE 4.11-2 (Continued)
- TABLE NOTATION
a. The LLD is defined in Table 4.11.1.
b. The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Kr-87, kr-88, Xe-133, Xe-133m, Xe-135, Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144 for particulate emissions. This list does not mean that only these nuclides are to be detected and reported. Other peaks that ~re measurable and identifiable, together with the above nuclides, shall also be identified and reported.
c. Sampling and analysis shall also be performed following shutdown, startup or a THERMAL POWER change exceeding 15 percent of RATED THERMAL POWER within one hour unless:
1. Analysis shown that the DOSE EQUIVALENT I-131 concentrations in the primary coolant has not increased more than a factor of three.
2. The noble gas activity monitor shows that effluent activity has not increased by more than a factor of three above the monitor setpoint.
d. Tritium grab samples shall be taken at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the refueling canal is flooded.
e. Tritium grab samples shall be taken at least once per 7 days from the ventilation exhaust from the spent fuel pool area whenever spent fuel is in the spent fuel pool. --
f. The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time period covered by each dose or dose rate calculation made in accordance with Specifications 3.11.2.1, 3.11.2.2 and 3.11.2.3.
SALEM - UNIT 2 3/4 11-10
TABLE 4.11-2 (Continued)
TABLE NOTATION
g. Samples shall be changed at least once per 7 days and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after changtng (or after removal from sampler).
Sampling shall also be performed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for at least 7 days following each shutdown, startup or THERMAL POWER change exceeding 15 percent of RATED THERMAL POWER in one hour and analyses shall be completed for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months are analyzed, the corresponding LLDs may be increased by a factor of 10. This requirement does not apply if (1) analysis shown that the DOSE EQUIVALENT I-131 concentration in the primary coolant has not increased more than a factor of 3; and (2) the noble gas monitor shows that effluent activity has not increased more than a factor of 3 *
SALEM - UNIT 2 3/4 11-11
RADIOACTIVE EFFLUENTS DOSE - NOBLE GASES LIMITING CONDITION FOR OPERATION 3.11.2.2 The air dose due to noble gases released in gaseous effluents, from each reactor unit, from the site areas at and beyond the SITE BOUNDARY (see Figure 5.1-3) shall be limited to the. following:
a. During any calendar quarter: Less than or equal to 5 mrad for gamma radiation and less than or equal to 10 mrad for beta radiation and,
b. During any calendar year: Less than or equal to 10 mrad for gamma radiation and less than or equal to 20 mrad for beta radiation.
APPLICABILITY: At all times.
ACTION:
a. With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, in lieu of a Licensee Event Report, prepare and submit to the Commission within 30.days, pursuant to Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit(s) and defines the corrective actions that have been taken to reduce the release and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.
b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.2.2 Cumulative dose contributions for the current calendar quarter and current calendar year shall be determined in accordance with the ODCM at least once per 31 days
SALEM - UNIT 2 3/4 11-12
RADIOACTIVE EFFLUENTS DOSE - IODINE-131, TRITIUM, AND RADJONUCLIDES IN PARTICULATE FORM LIMITING CONDITION FOR OPERATION 3.11.2.3 The dose to a MEMBER OF THE PUBLIC from iodine-131, from tritium, and from all radionuclides in particulate form with half-lives greater than 8 days in gaseous eff.luents released, from each reactor unit, from the site to areas at and beyond the SITE BOUNDARY (see Figure 5.1-3) shall be limited to the following: *
a. During any calendar quarter: Less than or equal to 7.5 mrems to any organ and,
b. During any calendar year: Less than or equal to 15 mrems to any organ.
APPLICABILITY: At all times.
ACTION:
a. With the calculated are dose from the release of iodine-131, tritium, and radionuclides in particulate form with half-lives greater than 8 days, in gaseous effluents exceeding any of the above limits, in lieu of a Licensee Event Report, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit and defines the corrective actions that have been taken to reduce the release and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.
b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS -
4.11.2.3 Cumulative dose contributions for the current calendar quarter and current calendar year for iodine-131, tritium, and radionuclides in particulate form with half-lives greater than 8 days shall be determined in accordance with the ODCM at least once per 31 days *
SALEM - UNIT 2 3/4 11-13
RADIOACTIVE EFFLUENTS GASEOUS RADWASTE TREATMENT LIMITING CONDITION FOR OPERATION 3.11.2.4 The GASEOUS RADWASTE TREATMENT SYSTEM and the VENTILIATION EXHAUST TREATMENT SYSTEM shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected gaseous effluent air doses due to gaseous effluent releases, from the site to areas at and beyond the SITE .
BOUNDARY (see Figure 5.1-3) exceed 0.625 mrad for gamma radiation and 1.25 mrad for beta radiation in any calendar quarter. The VENTILIATION EXHAUST TREATMENT SYSTEM shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected doses due to gaseous effluent re 1eases, from each reactor unit, from the site to areas at and beyond the SITE BOUNDARY (see Figure 5.1-3) would exceed 1.875 mrem to any organ .in any calendar quarter.
APPLICABILITY: At all times.
ACTION:
a. With gaseous waste being discharged without treatment and in excess of the above limits, in lieu of a Licensee Event Report, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that includes the following information:
1. Explanation of why gaseous radwaste was being discharged without treatment, identification of any inoperable equipment or subsystems, and the* reason for the i noperabi 1ity,
2. Action(s) taken to restore the inoperable equipment equipment to OPERABLE status, and
3. Summary description of action(s) taken to prevent a recurrence.
b. The provisions of Specifications 3.0.4 are not applicable. -*
SURVEILLANCE REQUIREMENTS 4.11.2.4 Doses due to gaseous releases from the site shall be project~d at least once per 31 days in accordance with the ODCM *
SALEM - UNIT 2 3/4 11-14
RADIOACTIVE EFFLUENTS
. EXPLOSIVE GAS MIXTURE LIMITING CONDITION FOR OPERATION 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 whenever the hydrogen concentration exceeds 4% 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 4% by volume but less than or equal 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 and the hydrogen concentration greater than 2% by volume, immediately suspend all additions of waste gases to the system and reduce the concentration of oxygen to less than or equal to*23 by volume without delay.
c. The provision of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.2.5 The concentrations of hydrogen and oxyg~n 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 required OPERABLE by Table 3.3-13.
SALEM - UNIT 2 3/4 11-15
RADIOACTIVE EFFLUENTS GAS STORAGE TANKS LIMITING CONDITION FOR OPERATION 3.11.2.6 The quantity of radioactivity contained in each gas storage tank shall be limited to 36,000 curies noble gases (considered as Xe-133).
APPLICABILITY: At all times.
ACTION:
. a. With the quantity of radioactive material in any gas storage tank exceeding the above limit, immediately suspend all additions of radioactive material to the tank and within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduce the tank contents to within the limit.
b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.2.6 The quantity of radioactive material contained in each gas storage tank shall be determined to be within the above limit at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when radioactive materials are being added to the tank.
SALEM - UNIT 2 3/4 11-16
RADIOACTIVE EFFLUENTS 3/4.11.3 SOLID RADIOACTIVE WASTE LIMITING CONDITION FOR OPERATION 3.11.3 *. The sol id radwaste system shall be used in accordance with a PROCESS CONTROL PROGRAM to process wet radioactive waste to meet shipping and burial ground requirements.
APPLICABILITY: At all times.
ACTION:
a. With the prov1s1ons of the PROCESS CONTROL PROGRAM not satisfied, suspend shipments of defectively processed or defectively packaged solid radioactive wastes from the site.
b. The provisions of Specifications 3.0.3 and 3.0.4, and 6.9.1.9.b are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.3. The PROCESS CONTROL PROGRAM shall be used to verify the SOLIDIFICATION of at least one representat.ive test specimen from at leavy every tenth batch of each type of wet radioactive waste (e.g., filter sludges, spend resins, evaporator bottoms, boric acid solutions, and sodium sulfate solutions) *
a. If any test specimen fai 1s. to verify SOLIDIFICATION, the SOLIDIFICATION of the batch_ under test shall be suspended until such time as additional test specimens can be obtained, alternative SOLIDIFICATION
. parameters can be determined in accordance with the PROCESS CONTROL PROGRAM, and a subsequent test verifies SOLIDIFICATION. SOLIDIFICATION of the batch may then be resumed using the alternative SOLIDIFCATION parameters determined by the PROCESS CONTROL PROGRAM.
b. If the initial test specimen from a batch of waste fails to verify SOLIDIFICATION, the PROCESS CONTROL PROGRAM shall provide for the collection and testing of representative test specimens from each consec_utive batch of the same type of wet waste until at least three consecutive initial test specimens demonstrate SOLIDIFICATION. The PROCESS CONTROL PROGRAM shall be modified as required, as provided in Specification 6.13, to assure SOLIDIFICATION of subsequent batches of waste.
SALEM - UNIT 2 3/4 11-17
RADIOACTIVE EFFLUENTS 3/4.11.4 TOT.AL DOSE LIMITING CONDITION FOR OPERATION 3.11.4 The annual (calendar year) dose or dose commitment to any MEMBER OF THE PUBLIC, due to releases of radioactivity and radiation, from urnaium fuel cycle sources shall be 1imited to *1 ess than or equal to 25 mrems to the total body or any organ (except the thyroid, which shall be limited to less than or equal to 75 mrems).
APPLICABILITY: At all times ACTION:
a. With the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Specification 3.ll.l.2al 3.11.1.2b, 3.11.2.2a, 3.11.2.2b, 3.ll.2.3a, or 3.ll.2.3b, calculations should be made including direct radiation contributions from the reactor units arid from outside storage tanks to determine whether the limits of this Specification have been exceeded. If such is the caie in lieu of a Licensee Event Report, prepare and submit to the c*ommission within 30 days, pursuant to Specification 6.9.2, a **-* I Special Report that defines the corrective action to be taken to reduce subsequent releases to prevent recurrence of exceeding the above limits and includes the schedule for achieving confo~mance with the above limits. This Specical Report, as defined in 10 CFR Part 2Q.405c, shall include an analysis that estimates the radiation exposure (dose) to a MEMBER OF THE PUBLIC from uranium fuel cycle sources, including all effluent pathways and direct radiation, for the calendar year that includes the release(s) covered by this report. It shall also describe levels of radiation and concentrations of radioactive material involved, and the cause of the exposure levels or concentrations. If the estimated dose ( s) exceeds the above 1 i mits, *and if the re 1ease
condtion resulting in violation of 40 CFR Part 190 has not al ready been corrected, the Special Report shall include a request for a variance in accordance with the provisions of 40 CFR Part 190. Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete
SALEM - UNIT 2 . 3/4 11-18
RADIOACTIVE EFFLUENTS 3/4.11.4 TOTAL DOSE LIMITING CONDITION FOR OPERATION ACTION: (Cont'd)
b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.4.1 Cumulative dose contributions from liquid and gaseous effluents shall be determined in accordance with Specifications 4.11.1.2, 4.11.2.2, 4.11.2.3, and in accordance with the ODCM.
4.11.4.2 Cumulative dose contributions from direct radiation from the reactor units and from radwaste storage shall be determined in accordance with the ODCM.
SALEM - UNIT 2 3/4 11-19
3/4.12 RADIOLOGICAL ENVIRONMENTAL MONlTORING 3/4.12.1 MONITORING PROGRAM LIMITING CONDITION FOR OPERATION 3.12.1. The radiological environmental monitoring program shall be conducted as specified in Table 3.12-1.
APPLICABILITY: At all times.
ACTION:
a. With the radiological environmental monitoring program not being conducted as specified in Table 3.12-1, in lieu of a Licensee Event Report, prepare and submit to the Commission, in the Annual Radiological Environmental Operating Report required by Specification 6.9.1.11, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence.
b. With the level of radioactivity as the result of plant effluents in an environmental sampling medium at a specified location exceeding the reporting levels of Table 3.12-2 when averaged over any calendar quarter, in lieu of a Licensee Event Report, prepar~ and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit(s) and defines the corrective actions to* be taken to reduce radioactive effluents so that the potential annual dose to a MEMBER OF THE PUBLIC is less than the calendar year limits of Specifications 3.11.1.2, 3.11.2.2, and 3.11.2.3. When more than one of the radionuclides in Table 3.12-2 are detected in the sampling medium, this report shall be submitted if:
concentration (1) concentratJon (2) reporting level (1)
+
reporting level (2)
+*** >1.0 When radionuclides other than those in Table 3.12-2 are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose to a member of the public is equal to or greater SALEM - UNIT 2 3/4 12-1
3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM LIMITING CONDITION FOR OPERATION ACTION: (Cont'd) than the calendar year limits of Specifications 3.11.1.2, 3.11.2.2, and 3.11.2.3. This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual
Radiological Environmental Operating Report.
c. With milk or fresh leafy vegetable samples unavailable from one or more of the sample locations required by Table 3.12-1, identify locations for obtaining replacement samples and add them to the radiological environmental monitoring program within 30 days. The specific locations from which samples were unavailable may then be deleted from the monitoring program.
In lieu of a Licensee Event Report and pursuant to Specification 6.9.1.11, identify the cause of the unavailability of samples and the new location(s) for obtaining repl~cement samples in the next Semiannual Radioactive Effluent Release Report. Include in the report a revised figure(s) and table for the ODCM reflecting the new 1ocat ion ( s).
d. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.12.1 The radiological environmental monitoring samples shall be collected pursuant to Table 3.12-1 from the locations specified in the ODCM and shall be analyzed pursuant to the requirements of Table 4.12-1.
SALEM - UNIT 2 3/4 12-2
TABLE 3.12-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM c
z Number of Samples
--i and Sampling and Type and frequency N Pathway Sample Locationsa Collection Frequencya of Analysis
1. DIRECT RADIATIONb About 40 routine monitoring Monthly, Quarterly Gamma dose monthly, stations with two or more or semi-annually quarterly, or semi-dosimeters for measuring annually dose continuously to be placed as follows: 1) an inner ring of stations in the general area of the Site and an outer ring in the 2- to 8-km range from the Site with a station in each sector of each ring.
The balance of the stations
-N w
I should be placed in special interest areas such as population centers, nearby residences, schools, and in 2 or 3 areas to serve as control stations.
2. AIRBORNE Radioiodine and Samples from 5 locations: Continuous sampler Particulates operation with sample
a. 3 samples from. close to collection weekly or the Site locations. as required by dust Radioiodine Cannister:
loading which ever is I-131 analysis weekly.
b. 1 sample from the vicinity more frequentc of a community. Particulate Sampler Gross beta radioacti-
c. 1 sample from a control vity analysis follow-location 15-30 km distance. ing filter change; Gamma isotopic analysis of complete (bv location)
(! ua rte r 1y
I, I '
I
TABLE 3.12-1 (Cont 1 d) c z...... Number of Samples
-; and Sampling and - Type and frequency N Pathway Sample Locationsa Collection Frequencya of Analysis
3. WATERBORNE
a. Surfaceg a. 1 sample upstreaming and Two gallon grab sample Gamma isotopic analy-
b. 1 sample downstream collected monthly. sis quarterly. Compo-
c. 1 sample outfall site for tritium
d. 1 sample cross-stream analysis quarterly.
b. Ground Samples from 1 or 2 sources Two gallon grab sample Gamma isotopic analy-w only if likely to be affectedj. collected quarterly. sis quarterly. Compo-
~ site for tritium N
analysis quarterly
I
~
c *. Drink_ing a. 1 sampl~ of the nearest 50 ml allquot taken I-131 analysis on each water supply. daily and composited to composite when the a monthly sample of two dose calculated for gallons. the consumption of the water is greater than 1 mrem per year.
Composite for gross beta and gamma isotopic analyses monthly.
Composite for tritium analysis quarterly.
d. Sediment a. 1 sample downstream Samples taken semi- Gamma isotopic analy-in river b. 1 sample cross-stream annually. sis semiannually.
c. 1 sample outfall I, ' I I
TABLE 3.12-1 (Cont'd)
.C
z Number of Samples
~
--i .and Sampling and Type and frequency N .Pathway Sample Locationsa Collection Frequencya of Analysis
4. INGESTION
a. Milk a. Samples from milking animals Semimonthly when Gamma isotopic and
-in 3 locations within animals are on pasture, I-131 analysis semi-5 km distance having the monthly at other monthly when animals highest dose potential. time. are on pasture; If there are none, then, 1 monthly at other sample from milking animals times.
5 to 8 km distant where doses are calculated to be greater than 1 mrem per
........ yrk
N I
C.TI b. 1 sample from milking animals at a control location.
b. Fish and a. 1 sample of each commerci- Sample in season, or Gamma isotopic analy-Invertebrates ally and recreationally sem*i annual if they sis on edible.
important species in vicinity are not seasonal portions.
of discharge point.
c. Food a. 1 sample of each principal At time of harvestl Gamma isotopic analy-class of food products from sis on edible any area that is irrigated portion.
by water in which liquid plant wastes have been dis-charged.
1, I I
TABLE 3.12.1 (Cont'd)
TABLE NOTATION a In the event that it is not possible or practicable to obtain samples of the media of choice at the most desired location or time, suitable alternative media and locations may be chosen for the particular pathway. Actual locations (distance and direction) from the site shall be provided for all sample locations in the ODCM.
b One or more instruments, such as pressurized ion chamber, for measuring and recording dose rate continuously may be used in place of, or in addition to, integrating dosimeters. For the purposes of this table, a thermoluminescent dosimeter may be considered to* be one phospor, and two- or more phosphors in a packet may be.considered as two or more dosimeters. Film badges shall not be used for measuring direct radiation. (The 40 stations is not an absolute number. Th.is number may be reduced according to geographical limitations; e.g., at an ocean site, some sectors will be over water so that the number of dosimeters may be reduced accordingly. An option open to licensees is to place some of the 40 routine TLD monitoring stations inside the SITE BOUNDARY. Such stations could provide useful information under both routine and accident conditions, and would be particularly valuable for the larger sites. The frequency of analysis or readout will depend upon the characteristics of the specific TLD system used and should be selected to obtain optimum dose information with minimal fading.)
c Canisters for the collection of radioiodine in air shall be checked for channeling before operation in the field to insure complete recovery of iodine.
d Particulate sample filters shall be analyzed for gross beta radioactivity 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or more after sampling to allow for radon and thoron daughter product decay. If gross beta activity in air or water is greater than ten times the yearly mean of control samples for any medium, gamma isotopic analysis shall be performed on the individual samples.
SALEM - UNIT 2 3/4 12-6
TABLE 3.12-1 (Cont'd)
TABLE NOTATION e Gamma isotopic analysis means the identification and quantification of gamma-emitting radi onucl ides that may* be attributable to the efful ents from the facility.
f The purpose of this sample is to obtain background information. If it is not practica~ to establish control locations in ~ccordance with the distance and wind direction criteria, other sites that provide valid background data may be substituted.
g The "upstream sample" shall be taken at a distance beyond significant influence of the discharge. The "downstream" sample shall be taken in an area beyond but near* the mixing zone. "Upstream" samples in an estuary must be taken far enough upstream to be beyond the plant influence.
h Salt water shall be sampled only when the receiving water is utilized for recreational activities.
i Composite samples shall be collected with equipment that is capable of collecting an aliquot at time intervals that are very short (e.g., hourly) relatjve to the compositing period (e.g., monthly) in order to assure obtaining a representafi ve sampi e. .
j Groundwater samples *shall be taken when this source is tapped for drinking or irrigation purposes in areas where the hydraulic gradient or recharge properties are suitable for contamination.
k The dose sha}l be calculated for the maximum organ and age group, using the methodofogy contained in Regulatory Guide 1.109, Rev. 1, October 1977, and the actual parameters particular to the site.
1 If harvest occurs more than once a year, sampling shall be performed during each discrete harvest. If harvest occurs continuously, sampling shall be
monthly. Attention shall be paid to including samples of tuberous and root food products.
SALEM - UNIT 2 3/4 12-7

*------*-- --~------

I,.
i TABLE 3.12-2 REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES c Reporting Levels
z
-I N
Water Airborne Particulate Fish Milk Food Products Analysis (pCi/1) or Gases (pCi/m3) (pCi/Kg, wet) (pCi/1) (pCi/Kg, wet)
H-3 2 x 104(a)
Mn-54 1 x 103 3 x 104 Fe-59 4 x 102 1 x 104 Co-58 1 x 103 3 x 104 Co-60 3 x 102 1 x 104
-N I
(X)
Zn-65 Zr-Nb-95 3 x 102 4 x 102 2 x 104 I-131 2 0.9 3 1 x 102 Cs-134 30 10 1 x 103 60 1 x 103 Cs-137 50. 20 2 x 103 70 2 x 103 Ba-La-140 2 x 102. 3 x 102 (a) For drinking water samples. This is 40 CFR Part 141 value.
TABLE 4.12-1 MAXIMUM VALUES FOR THE LOWER LIMITS OF DETECTION {LLD) c:
z
-I N
Water Airborne Particulate Fish Milk Food Products Sediment Analysis {pCi/1) or Gases {pCi/m3) (pCi/Kg, wet) (pCi/1) (pCi/Kg, wet) {pCi/Kg, dry) gross beta ~
4 1 x 10-2 H-3 2000 Mn-54 15 130 Fe-59 30 260 Co-58, 60 15 130 I-'
N I
Zn-65 30 260 l.O
. 15 ~'
Zr-Nb-95 I-131 1 7 x 10-2 1 60 Cs-136 15 5 x 10-2 130 15 60 150 Cs-137 18 6 x 10-2 150 18 80 180 Ba-La-140 15 15
TABLE 4.12-1 (Cont'd)
TABLE NOTATION a Required detection capabilities for thermoluminescent dosimeters used for environmental measurements are given in Regulatory Guide 4.13, Rev. 1, July 1977.
b The LLD is defined in Table 4.11-1.
SALEM - UNIT 2 3/4 12-10
r RADIOLOGICAL ENVIRONMENTAL MONITORING
. 3/4.12.2 LAND USE CENSUS LIMITING CONDITION FOR OPERATION 3.12.2. A land use census shall be conducted and shall identify within a distance of 8 km (5 miles) the location in each of the 16 meteorological sectors of the.nearest milk animalA the nearest residence and the nearest garden* of greater than 50 m2 (500 ft~) producing broad l~af vegetation. (For elevated releases as defined in Regulatory Guide 1.111, Revision 1, July 1977, the land use census shall also identify within a distance of 5 km (3 miles) the locations in each of the 16 meteorological sectors of all milk animals and all gardens of greater than 50 m2 producing broad leaf vegetation. ~
APPLICABILITY: At all times.
ACTION:
a. With a land use census identifying a location(s) that yields a calculated dose or dose commitment greater thari the values currently being calculated in Specification 4.11.2.3, in lieu of a Licensee Event Report, identify the new location(s)in the .next Semiannual Radioactive Effluent Release Report, pursuant to Specification 6.9.1.12.
b. With a land use census identifying a location(s) that yields a calculated dose or dose commitment (via) the same exposure pathway) 20 percent greater than at a location from whi~h samples are currently being obtained in accordance with Specification 3.12.1, add the new location(s) to the radiological environmental monitoring program Within 30 days. The sampling location(s), excluding the control station location, having the lowest calculated dose or dose commitment(s) (via the same exposure pathway) may be deleted from this monitoring program after October 31 of the year in which this land use census was conducted. In lieu of a Licensee Event Report and pursuant to Specification 6.9.1.12, identify the new location(s) in the next Semiannual Radioactive Effluent Release Report and also include in the report a revised figure(s) and table for the ODCM reflecting the new location(s).
c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
Broad leaf vegetation sampling of at least three different kinds of vegetation may be performed at the SITE BOUNDARY in each of two different direction sectors with the highest predicted D/Qa vegetation sampling lin Table 3.12-1.4c shall .be followed, including analysis of control samples *
SALEM - UNIT 2 3/4 12-11
)
RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.2 LAND USE CENSUS (Cont 1 d)
SURVEILLANCE REQUIREMENTS 4.12.2 The land use census shall be conducted during the growing season at least once per 12 months using that information that will provide the best results, such as ~Y a door-to-door survey, aerial survey, or by consulting local
agriculture authorities. The results of the land use census shall be included in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.10
SALEM - UNIT 2 3/4 12-12
RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.3. INTERLABORATORY COMPARISON PROGRAM LIMITING CONDITION FOR OPERATION 3.12.3 Analyses shall be performed on radioactive materials supplied as part of an Interlaboratory Comparison Program which has been approved by the Commission.
APPLICABILITY: At all times.
ACTION:
a. With analyses not being performed as required above, report the corrective actions taken to prevent a recurrence to the Commission in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.10.
b. The provisions of Specifications 3.0.3 and 3.0.4. are not applicable.
SURVEILLANCE REQUIREMENTS 4.12.3 A summary of the results obtained as part of the above required Interlaboratory Comparison Program and in accordance with the ODCM shall be included in the Annual Radiological. Environmental Operating Report pursuant to Specification 6.9.1.10.
SALEM - UNIT 2 . 3/4 12-13
INSTRUMENTATION BASES 3/4.3.3.8 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The alarm/trip setpoints for these instruments shall be calculated and adjusted in accordance with the procedures in the ODCM to ensure that the alarm/trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50. The purpose of tank level indicating devices is to assure the detection and control of leaks that if not controlled could potentially result in the transport of radioactive materials to UNRESTRICTED AREAS.
1.
3/4.3.3.9 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of_radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The alarm/trip setpoints for these instruments shall be calculated and adjusted in accordance with the procedures in the ODCM to ensure that the alarm/trip will occur prior to exceeding the limits of 10 CFR Part 20. This instrumentation also includes provisions for monitoring (and controlling) the concentrations of potentially explosive gas mixtures in the waste gas holdup system. The OPERABILITY and* use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50 *.
3/4.3.4 TURBINE OVERSPEED PROTECTION This specification is provided to ensure that the turbine overspeed protection instrumentation and the turbine speed control valves are OPERABLE and will protect the turbine from excessive overspeed. Protection from turbine e~cessive overspeed is required since excessive overspeed of the turbine could generate potentially damaging missiles* which could impact and damage safety-related components, equipment or structures.
To prevent double shocking the turbine, valve testing is not required when steam is being admitted to the turbine and THERMAL POWER is less than 85% of RATED THERMAL POWER, provided the valves are tested prior to startup and within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of attaining and once per 7 days while operating at greater than or equal to 85% of RATED THERMAL POWER *
SALEM - UNIT 2 B 3/4 3-3
3/4.11 RADIOACTIVE EFFLUENTS BASES 3/4.11.1 LIQUID EFFLUENTS 3/4.11.1.1 CONCENTRATION The specification is provided to ensure that the concentration of radioactive materials released in liquid waste efflu~nts will be less than the concentration levels specified in 10 CFR Part 20, Appendix B Table II, Column 2. This limitation provides additional assurance that the levels of radioactive materials in bodies of water in UNRESTRICTED AREAS will result in exposures within (1) the Section II.A design objectives of Appendix I, 10 CFR Part 50, to a MEMBER OF THE PUBLIC and (2) the limits of 10 CFR Part 20.106(a) to the population. The concentration limit for dissolved or entrained noble gases is based upon the assumption that Xe-135 is the controlling radioisotope.and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2.
The required detection capabilities -for radioactive materials in liquid waste samples are tabulated in terms of the lower limits of detection (LLDs) 3/4.11.1.2 DOSE This specification is provided to implement the requirements of Sections II.A, III.A, and IV.A of Appendix I. 10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section II.A of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement-the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in liquid effluents will .be kept "as low as is reasonably achievable." Also, for freshwater sites with drinking water supplies that can be potentially affected by plant operations, there is reasonable assurance that the operation of the facility wi 11 *not result in radionuclide concentrations in the finished drinking water that are in excess of SALEM - UNIT 2 B 3/4 11-1
RADIOACTIVE EFFLUENTS BASES the requirements of 40 CFR Part 141. The dose calculations in the ODCM implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropria~e pathways is unlikely to be substantially underestimated. The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual does to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.113, "Estimating Aquatic Dispersion of Effluents from Accidential and Routine Reactor Releases for the Purposes of Implementing Appendix I," April 1977.
The specification applies to the release of liquid effluents from each reactor at the site. For units with shared radwaste treatment systems, the liquid effluents from the shared system are proportioned amoung the units sharing that system.
3/4.11.1.3 . LIQUID RADWASTE TREATMENT The requirement that the appropriate portions of this system be used, when specified, provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as is reasonably achievable". This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objective given in Section II.O of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the liquid radwaste treatment system were specified as a suitable fraction of the dose design objectives set forth the Section II.A of Appendix I, 10 CFR Part 50, for liquid effluents.
3/.4.11.1.4 LIQUID HOLDUP TANKS The tanks listed in this specification include all those outdoor 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 l i.qui d radwaste treatment, system.
SALEM - UN IT 2 B 3/4 11-2
RADIOACTIVE EFFLUENTS BASES Restricting the quantity of radi.oactive meterial contained in the specified tanks provides assurance that in the event of an uncontrolled release of the tanks' contents, the resulting concentrations wouJd be less than the limits of 10 CFR Part 20, Appendix B, Table II, Column 2, at the nearest potable water supply and the nearest surface water supply in an UNRESTRICTED AREA.
3/4.11.2 GASEOUS EFFLUENTS 3/4.11.2.1 DOSE RATE This specification is provided to ensure that the dose at any time at and beyond the SITE BOUNDARY from gaseous effluents from all units on the site will be within the annual dose limits of 10 CFR Part 20. The annual dose limits are the doses associated with the concentrations of 10 CFR Part 20~ Appendix B, Table II, Column 1. These limits provide reasonable assurance that radioactive material discharged in gaseou~ effluents will not result in the exposure of a MEMBER OF THE PUBLIC either within or outside the SITE BOUNDARY, to annual average concentrations exceeding the limits specified in Appendix B, Table II. of 10 CFR Part 20 [13 CFR Part 20.106(b)]. For MEMBERS OF THE PUBLIC who may at times be within the SITE BOUNDARY, the occupancy of the individual will usually be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for th~ SITE BOUNDARY. Examples of calculations for such MEMBERS OF THE PUBLIC with the appropriate occupancy factors shall be given .in the ODCM. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to a MEMBER OF THE PUBLIC at or beyond the SITE BOUNDARY to less than or equal to 500 mrems/year to the skin. These release rate limits also restrict, at all times, the corresponding thyroid dose rate above background to a child via the inhalation pathway to less than or equal to 1500 mrems/year.
This specification applies to the release of gaseous effluents from a1*1 reactors at the site *
SALEM - UNIT 2 B 3/4 11-3
RADIOACTIVE. EFFLUENTS BASES 3/4.11.2.2 DOSE - NOBLE GASES This specification is provided to implement the requirements of Section II.B, III.A and IV.A of Appendix I, 10 CFR Part 50. The limiting condition for operation implements the guides set forth in Section II.8 of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in gaseous effluents will be kept 11 as low as is reasonably achievable. 11 The Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The dose calculations established in the ODCM for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents are consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR 50, Appendix I, 11 Revision I, October 1977 and Regulatory Guide 1.111, "Methods for Estimating Atmospheric Transport and Dispersin of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors, 11 Revision 1, July 1977. The ODCM equations provided for determining the air doses at and beyond the SITE BOUNDARY are based upon the historical average atmospheric conditions.
3/4.11.2.3 DOSE - IODINE-131, TRITIUM, AND RADIONUCLIDES IN PARTICULATE FORM This specification is provided to implement the requirements of Section II.C, III.A and IV.A of Appendix I, 10 CFR Part 50. The limiting condition for operation are the guides set forth in Section II.C of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the SALEM - UNIT 2 B 3/4 11-4
RADIOACTIVE EFFLUENTS BASES releases of radioactive material in gaseous effluents will be kept "as low as is reasonably achievable." The ODCM calculational methods specified in Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be
substantially underestimated. The ODCM calculational methods for calculating the doses due to the actual release rates of the subject materials are consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the
. Purpose of Evaluating Compliance with 10 CFR 50, Appendix I," Revi~ion 1, October 1977 and Regulatory Guide 1.111, "Methods for Estimating AtmospheriC Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors," Revision 1, July 1977. These equations also.
provide for determining the actual dose based upon the historical average atmospheric conditions. The release rate specifications for iodine-131, tritium, and radionuclides in particulate form with half-life greater than 8 days are dependent on the existing radionuclide pathways to man in the areas at and beyond the SITE BOUNDARY. The pathways that were examined in the development of these calculations were: 1) individual inhalation of airborne radionuclides, 2) deposition of radionuclides onto green leafy vegetation with subsequent consumpt1on by man, 3) deposition onto grassy areas where milk animals and meat producing animals graze with Gonsumption of the milk and meat by ma~, and 4) deposition on the ground with subsequent exposure of man.
The requirement that the appropriate portions of this system be used, when specified, provides reasonable assurance that the releases of radioactive materials in gaseous effluent~ will be kept "as low as is reasonable achievable." This specification implements the requirements of 10 CFR Part*
50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objectives given in Section II.O of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the systems were specified as a suitable fraction of the dose design objectives set forth in Section II.B and II.C of Appendix I, 10 CFR Part 50, for gaseous effluents.
SALEM - UNIT 2 B 3/4 11-5
RADIOACTIVE EFFLUENTS BASES 3/4.11.2.5 EXPLOSIVE GAS MIXTURE This specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the waste ga*s holdup system is maintained below the flammability limits of hydrogen and oxygen. (Automatic control features are included in the system to prevent the hydrogen and oxygen concentrations from reaching these flammability limits. These automatic control features include isolation of the source of hydrogen and/or oxygen, automatic dive~sion to recombiners; or injection of dilutants to reduce the concentration below the flammability limits). Maintaining the concentrationof hydrogen and oxygen below their flammability limits proves assurance that the releases of radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50.
3/4.11.2.6 GAS STORAGE TANKS The tanks insluded in this specification are those tanks for which the quantity of radioactivity contained is not limited directly or indirectly by another
- * * *- Technical Specification to a quantity that is le.ss than the quantity which provides assurance that in the event of a uncontrolled release of the tank's
contents, the resulting total body exposure to an individual at the nearest exclusion area boundary will not exceed 0.5 rem in an event of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .
Restricting the quantity of radioactivity contained in each gas storage tank provides assurance that in the event of an uncontrolled release of the nearest exclusion area boundary will not exceed 0.5 rem. This is conistent with Branch Technical Position ETSB 11-5 in NUREG-0800, July 1981.
3/4.11.3 SOLID RADIOACTIVE WASTE This specification implements the requirements of 10 CFR Part 50.36a and Gener.al Design Criterion 60 of Appendix A to 10 CFR Part 50. The process parameters included in establishing the PROCESS CONTROL PROGRAM may include, but are not limited to waste type, waste pH, waste/liquid/solidification agent/catalyst ratios, waste oil content, waste principal chemical constitOents, mixing and curing times.
SALEM - UNIT 2 B 3/4. 11-6
RADIOACTIVE EFFLUENTS BASES 3/4.11.4 TOTAL DOSE This specification is provided to meet the dose limitations of 40 CFR Part 190 that have now been incorporated into 10 CFR Part 10 to 46 Fr 18525. The specification requires the preparation and submittal of a Special Report whenever the calculated doses from plant radioactive effluents exceed twice the design objective doses of Appendix I. For sites containing up to 4 reactors, it is highly unlikely that the resultant dose to a MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR Part 190 if the individual reactors remain within the reporting requirement level. The Special Report will describe a course of action that should result in the limitation of the annual dose to a MEMBER OF THE PUBLIC to within the 40 CFR Part 190 limits. For purposes of the Special Report, it may be assumed that the dose commitment to the MEMBER OF THE PUBLIC from other uranium fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of 8 km must be considered. If the soe to any MEMBER OF THE PUBLIC is estimated to exceed the requirements of 40 CFR Part 190, the. Special Report with a request for a variance (provided the release conditions resulting in violation of 40 CFR Part 190 have not already been corrected), in accordance with the provisions of 40 CFR Part 190 and 10 CFR Part 20.405c, is considered to be a timely request and fulfills the requirements of 40 CFR Part 190 until NRC staff action is completed. The variance only relates to the limits of 40 CFR Part 190, and does not apply in any way to the other requirements for dose limitation of 10 CFR Part 20, as addressed in Specifications 3.11.1 and 3.11.2.
An indfvidual is not considered a MEMBER OF THE PUBLIC during any period in which he/she is engaged in carrying out any operation that is part of the nuclear fuel cycle.
SALEM - UNIT 2 B 3/4 11-7
RADIOACTIVE EFFLUENTS BASES 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM The radiological environmental monitoring program required by this specification provides measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides that lead to the highest potential radiation exposures of MEMBERS OF THE PUBLIC resulting from the station operation. This monitoring program implementsSection IV.B.2 of Appendix I to 10 CFR Part 50 and thereby supplements the radiological effluent monitoring program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and the modeling of the environmental exposure pathways.
The initial specified monitoring program will be effective for at least the first three years of commercial operation. Following this period, program changes may be initiated based on operational experience.
The LLDs required by Table 4.12-1 are considered optimum for routine environmental measurements in industrial laboratories. It should be recognized the the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for a particular measurement. -
3/4.12.2 LAND USE CENSUS This specification is provided to ensure that changes in the use of areas at and beyond the SITE BOUNDARY are identified and that modificatins to the radiological environmental monitoring program are made if required by the results of his census. *The best information from the door-to-door survey, aerial survey or consulting with local agricultural auth9rities shall be used.
This census satisfies the requirements of Section IV.8.3 of Appendix I to 10 CFR Part 50. Restricting the census to gardens of greater than 50m2 provides assurance* that significant exposure pathways via leafy vegetables will be identified and ~onitored since a garden of this size is the minimum required to produce the quantity (26 kg/year) of leafy vegetables assumed in Regulatory Guide 1.109 for consumption by a child. To determine this minimum garden size, the following assumptions were made: 1) 20% of the garden was used for growing broad leaf vegetation (i.e., similar to lettuce and cabbage), and 2) yield of 2 kg/m2.
SALEM - UNIT 2 B 3/4 12-1
RADIOACTIVE EFFLUENTS BASES 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM This requirement for participation in an Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices are performed as part of the quali~ assurance pro~ram for environmental monitoring in order to demonstrate that the results are reasonably valid for the purposes of Section IV.B.2 of Appendix I to 10 CFR Part 50.
SALEM - UNIT 2 B 3/4 12-2
5.0 5.1 DESIGN FEATURES 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 concr~te 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 2 5-1
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DESIGN FEATURES DESIGN PRESSURE AND TEMPERATURE 5.2.2 The reactor containment building is designed and shall be maintained* for a maximum internal pressure of 47 psig and an air temperature of 271°F.
5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 193 fuel assemblies with each fuel assembly containing 264 fuel rods cl ad with Zi real oy-4. Each fuel rod shall have a nominal active fuel length of 143.7 inches and contain a maximum total weight of 1766 grams uranium. The initial core loading shall have a maximum enrichment of 3.35 weight percent U-235. Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum enrichment of 3.5 weight percent U-235.
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 5.4.1 The reactor coolant system is designed and shall be maintained:
a. In accordance with the *code requirements specified in Section 4.1 of the FSAR, with alowance for normal degradation pursuant to the applicable Surveillance Requirements,
b. For a pressure of 2485 psig, and
c. For a temperature of 650°F, except for the pressurizer which is 680°F.
VOLUME 5.4.2 The total. water and steam volume of the reactor cool ant system is 12,811 + 100 cubic feet at a nominal Tavg of 581.0°F *
SALEM - UNIT 2 5-4
ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The General Manager - Salem Operations shall be responsible for overall facility operation and shall delegate in writing the succession to this respon-sibility during his absence.
6~1.2 The Senior Shift Supervisor or during his absence from the Control Room, a designated individual shall be responsible for the Control Room command func-tion. A management directive to this effect, signed by the Vice President -
Nuclear shall be reissued to all station personnel on an annual basis.
6.2 ORGANIZATION OFF SITE 6.2.1 The offsite organization for facility management and technical support shall be as shown on Figure 6.2-1.
FACILITY STAFF 6.2.2 The Facility organization shall be as shown on Figure 6.2-2 and:*
a. Each on duty shift shall be composed of at least the minimum shift crew composition shown in Table 6.2-1.
b. At least one licensed Operator shall be in the control room when fuel is in the.reactor. In addition, at least one licensed Senior Reactor Operator shall be in the Control Room while the unit is in MODE 1, 2, 3 or 4.
c. A health physics technician# shall be on site when fuel is in the reactor.
d. ALL CORE ALTERATIONS shall be observed and directly supe*rvi sed by a licensed Senior Reactor Operator who has no other concurrent responsi-bilities during this operation.
e. A site Fire Brigade of at least 5 members shall be maintained onsite at all times#. The Fire Brigade shall not include 4 members of the minimum shift crew necessary for safe shutdown of the unit or any personnel required for other essential functions during a fire emergency.
f. The amount of overtime worked by plant staff members performing safety-related functions must be limited in accordance with the NRC Policy Statement on working hours (Generic Letter No* 82-12).
The health physics technical and Fire Brigade composition may be less than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accommodate unexpected absence of the health physics technician and/or Fire Brigade members provided immediate action is taken to restore the health physics technician and/or Fire Brigade to within the minimum requirements.
SALEM - UNIT 2 6-1
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& mHER THE lll'EltATIIJ6 IMH<<al DA THE llPERATING ENt:lta:R EDUIPHENT llill loET 1llE OUl'l..lFn:ATilJlS Ill' ANSI NIS.Hq7J FIJI THE POSIT((Jj DPERATIJ1S II' lll'ERATIDllS MANAGER.AU. SENIOR 511If1 SUPERYl5a!S SHAU. REPORT
'TO OH IMIJVIDUAL ~ SHAU. HOLD A SENJCll REACTUR OPERATOR'S UCENSE ANO HAVE SIX VA*RS CF RESPONSIBLE PO\/Ell PLANT EXPERIENtE CF llHlDI A MINIMl.H Of T'w'O YEARS SHALL BE Nl.U..fAR roWER PLANT ElCPEJUEN:E. A MAXJHUH OF Tt£ REHA(HfNll TWO IF THE F£HAIMNG RJJR
'![MS OF PO'tl?R Pl.ANT EXPERIENCE tt'\Y BE f\Jl.FIUEO BY SATJSFACTOOY UTILITY TEOlNICAL TECHNICAL IXH'l.ET!DN l1F l'CAOEHJC OR RELATED TECHNICAL TRAININO ON A ONE TJHE BASIS. OPERl\lr.A!I WffiKERS WORKERS AESPOllS!BLE FOR FlRE PflJTECTJON PROIJIA~ fl'Pl.EHENTATlrtl.

STA lti'tllflED ~N) fU.PlLL9 STA REOLDRl;HENT9 OF TABLE B..2-1. tF SRO l.U:£NSBl WILL BE. AS9Jlll£D SHIFT Sll'ERVJSDRY DUTIES. POSITION HAY BE fJLLED SY !>TA llHO lt'OOLD llEPOIU DIRECTLY TO THE OPEAATJ1£ ENGKER.
- - Sllllflf!ES COORDINATION BETlt'EEN 1-W!AGER- NUCLEAR TRAININO NIU llEPARTl'EHT MAH*CEAS AT THE STATION.
FIGURE 6
2 - 2 FACILITY ORGANIZATION
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TABLE 6.2-1 MINIMUM SHIFT CREW COMPOSITION SALEM UNIT 2 WITH UNIT 1 IN MODES 5 OR 6 OR DE-FUELED POSITION NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION MODES 1, 2, 3 & 4 MODES 5 & 6 SS 1a 1a SRO 1a none RO 2 1 AO 3 2b STA 1 none Maintenance 1 none Electrician
WITH UNIT 1 IN MODES 1, 2, 3 OR 4 POSITION NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION
. MODES 1, 2, 3 &4 MODES 5 & 6 SS 1a 1a SRO 1a none RO 2 1 AO 3c 1 STA 1a none Maintenance 1a none El ectri ci an
~/ Individual may fill the same position on Unit 1
'p_/ One of the two required individuals may fill the same position on Unit 1, such that there are a total of three AOs for both units.
!::._/ One of the three required individuals may fill the same position on Unit 1, such that there are a total of five AOs for both units.
SALEM - UN IT 2 6-4
-- - -------- ----- ~- ----*-.--- -----------
TABLE 6.2-1 (Continued)
SS Shift Supervisor with a Senior Reactor Operators License on Unit 2 SRO - Individual with a Senior Reactor Operators License on Unit 2 RO Individual with a Reactor Operators License on Unit 2 AO - Auxiliary Operator STA - Shift Technical Advisor Except for the Shift Supervi~or, 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 exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order* to accommodate unexpected absence of on-duty shift crew crew members provided 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 crewman being lat~ or absent.
During any absence of-* the Shift Supervisor from the Control Room whi 1e .the unit
is in MODE 1, 2, 3 or 4, an individual (other than the Shift Technical Advisor) with a valid SRO license shall be designated to assume the Control Room command function.' During any absence of the Shift Supervisor from the Control Room while the unit is in MODE 5 or 6, an individual with a valid RO license (other than the Shift Technical Advisor) shall be designated to assume the Control Room command function *
SALEM - UNIT 2 6-5
ADMINISTRATIVE CONTROLS 6.2.3 SHIFT TECHNICAL ADVISOR 6.2.3.1 - The Shift Technical Advisor shall serve in an advisory capacity to the Shift Supervisor on matters pertaining to the engineering aspects assuring safe operation of the unit.
6.2.3.2 The Shfft 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 m1n1mum qualifications of ANSI *N18.l-1971 for comparable positions and the supplemental requirements specified in Section~ A and C of Enclosure 1 of the March 28, 1980 NRC letter to all licensees, except for the Radiation Protection Engineer who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975.
6.4 TRAINING 6.4.1 A retraining and replacement training program for the facility staff shall be coordinated by each functional level manager (Department Head) at the facility and maintained under the direction of the Manager - Nuclear Training and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI Nl8.1-1971 and Appendix '1 A" of 10 CFR Part 55 and the supplemental requirements specified in Sections A and C of Enclosure 1 of the March 28, 1980 NRC letter to all licensees, and shall include familiarization with relevant industry operational experience.
6.4.2 A training program for the Fire Brigade shall be maintained under the direction of the Manager - Nuclear Training. and shall meet or exceed the requirements of Section 27 of NFPA Code-1975~ except for Fire Brigade training sessions which shall be held at least quarterly.
6.5 REVIEW AND AUDIT 6.5.1 STATION OPERATIONS REVIEW COMMITTEE (SORC)
FUNCTION 6.5.1.1 The Station Operations Review Committee shall function to advise the General Manager - Salem Operations on all matters related to nuclear safety.
SALEM - UNIT 2 6-6
-** *--*-**-*--*----*---*-----!..*------*-----~------- ----*-*--- .. ~.-*----*-**-- -** -* ~--**----*---*** . *****-**-*-----* .. -*--*-*-* -* ...... -*** -* . --*- ----
ADMINISTRATIVE CONTROLS COMPOSITION 6.5.1.2 The Station Operations Review Committee shall be composed of the:
Chairman: Assistant General Manager - Salem Operations Vice Chairman: Operations Manager Vice Chairman: Technical Manager Vice Chairman: Maintenance Manager Member: Operating Engineer Member: I&C Engineer Member: Senior Shift Supervisor Member: Technical Engineer Member: Maintenance Engineer Member: Radiation Protection Engineer Member: Senior Radiation Protection Supervisor Member: Chemistry Engineer ALTERNATES 6.5.1.3 All alternate members shall be appointed in writing by the SORG Chairman to serve on a temporary basis; however, no more than two alternates shall _participate as voting members in SORC activities at any one time.
MEETING FREQUENCY 6.5.1.4 The SORC shall meet at least once per calendar month and as convened by the SORC Chairman or his designated alternate.
QUORUM 6.5.1.5 The minimum quorum of the SORC necessary for the performance of the SORC responsibility and authority provisions of these technical specifications shall consist of the Chairman or his designated alternate and four members including alternates.
RESPONSIBILITIES 6.5.1.6 The Station Operations Review Committee shall be responsible for:
a. Review of 1) all procedures required by Specification 6.8 and changes thereto, 2) all programs required by Specification 6.8 and changes thereto, and 3) any other proposed procedures or changes thereto as determined by the General Manager - Salem Operations to affect nuclear safety.
b. Review of alJ proposed tests and experiments that affect nuclear safety.
c. Review of all proposed changes to Appendix "A" Technical Speci fi cations.
SALEM - UNIT 2 6-7

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ADMINISTRATIVE CONTROLS
d. Review of all proposed changes or modifications to plant systems or equipment that affect nuclear safety.
e. Investigation of all violations of the Technical Specifications including the preparation and forwarding of reports covering evaluation and recommendations to prevent recurrence to the Vice President - Nuclear and to the Chairman of the Nuclear Review Board. 1
f. Review of events requiring 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months written notification to the Commission.
g. Review of facility operations to detect potential nuclear safety hazards.
h. Performance of special reviews, investigations or analyses and reports thereon as requested by the Chairman of the Nuclear Review Board.
i. Review of the Plant Security Pl~n and implementing procedures and shall submit recommended changes to the Chairman of the Nuclear Review Board.
j. Review of the Emergency Plan and implementing procedures and shall submit recommended changes to the Chairman of the Nuclear Review Board.
k. Review of every unplanned onsite release of radioactive material to the environs including the preparation of reports covering evaluation, recommendations, ana disposition of the corrective action to prevent recurrence and the forwarding of these reports to the Vice President -
Nuclear and to the Nuclear Review Board.
1. Review of changes to the PROCESS CONTROL PROGRAM and the OFFSITE DOSE CALCULATION MANUAL.
AUTHORITY 6.5.1.7 The Station Operations Review Committee shall:
a. Recommend to the General Manager - Salem Operations written approval or disapproval of items considered under 6.5.1.6(a) through (d) above.
b. Render determinations in writing with regard to whether or not each item considered under 6.5.l.6(a) through (e) above constitutes an
_unreviewed safety question.
c. Provide written notification within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to the Vice President Nuclear and the Nuclear Review Board of disagreement between the SORC and the General Manager - Salem Operations; however, the General Manager - Salem Operations shall have responsibility for resolution of such disagreements pursuant to 6.1.1 above.
SALEM - UNIT 2 6-8
ADMINISTRATIVE CONTROLS RECORDS 6.5.1.8 The Station Operations Review Committee shall maintain written minutes of each meeting that, at a minimum, document the results of all SORC activities performed under the responsibility and authority provisions of these technical specifications. Copies shall be provided to the Vice President - Nuclear and Chairman of the Nuclear Review Board.
6.5.2 NUCLEAR REVIEW BOARD (NRB)
FUNCTION 6.5.2.1 The Nuclear Review Board shall function to provide indepen~ent review and audit of designated activities in the areas of:
a. nuclear power plant operations
b. nuclear engineering
c. chemistry and radiochemistry
d. metallurgy
e. instrumentation and control
f. radiological safety
g. mechanical engineering
h. electrical engineering
i. quality assurance
j. nondestructive testing
k. emergency prepardness COMPOSITION 6.5.2.2 The Vice President - Nuclear shall, appoint at least nine members to the Nuclear Review Board and shall designate from this membership a chairman and at least one vice chairman. The membership shall collectively possess experi-ence and competence to provide independent review and audit in the areas listed in Section 6.5.2.1. The chairman and vice chairman shall have nuclear back-ground in engineering or operations and shall be capable of determining when to call in experts to assist the NRB review of complex problems. All members shall have at least a Bachelor Degree in Engineering or related sciences. The chai~man shall have at least six years of professional level managerial experi-ence in the power field and all other members shall have at least five years of cumulative professional level experience in the fields listed in Section 6.5.2.1.
ALTERNATES 6.5.2.3 All alternate members shall be appointed in writing by the NRB Chairman to serve on a temporary basis; however, no more than two alternates shall par-ticipate as voting members in NRB activities at any one time. Educational and experience qualifications required of members are also applicable to alternates *
CONSULTANTS 6.5.2.4 Consultants shall be utilized as provide expert advice to the NRB.
SALEM - UNIT 2 6-9 d~termined by the NRB Chairman to
ADMINISTRATIVE CONTROLS MEETING FREQUENCY 6~5.2.5 The NRB shall meet at least once per calendar quarter during the initial year of unit operation following fuel loading and at least once per six months thereafter.
QUORUM 6.5.2.6 The' minimum quorum of NRB necessary for the performance of the NRB review and audit functions of these technical specifications shall consist of the Chairman or his designated alternate and at least 4 NRB members including alternates. No more than a minority of the quorum shall have line responsi-bility for operation of the facility.
REVIEW 6.5.2.7 The NRB shall revie~:
a. The safety evaluations for 1) changes to procedures,_equipment or systems and 2) tests or experiments completed under the provision of Section 50.59, 10 CFR, to verify that such actions did not constitute an unreviewed safety question.
b. Proposed changes to procedures, equipment or systems which involve an unreviewed safety question as defined in Section 50.59, 10 CFR.
c. Proposed tests or experiments which involve an unreviewed safety question as defined in Section 50.59, 10 CFR.
d. Proposed changes to Technical Specifications or this operating license.
e. Violations of codes, regulations, orders, Technical Specifications, license requirements, or of internal procedures or instruction having*
nuclear safety significance.
f. .Significant operating abnormalities or deviations from normal and expected performance of plant equipment that affect nuclear safety.
g. Events requiring 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months written notification to the Commission.
h. All recognized indications of an unanticipated deficiency in some aspect of design or operation of safety related structures, systems, or components.
i. Reports and meetings minutes of the Station Operations Review Committee.
AUDITS 6.5.2.8 Audits of facility activities shall be performed under the cognizance of the NRB. These audits shall encompass:
SALEM - UNIT 2 6-10
ADMINISTRATIVE CONTROLS
a. The conformance of facility operation to prov1s1ons contained within the Technical Specifications and applicable license conditions at.
least once per 12 months.
b. The performance, training and qualifications of the entire facility staff at least once per 12 months.
c. The results of actions taken to correct deficiencies occurrring in facility equipment, structures, systems or method of operation that affect nuclear safety at least once per 6 months.
d. The performance of activities required by the Operational Quality Assurance Program to meet the criteria of Appendix 11 B11 , 10 CFR 50, at least once per 24 months.
e. The Facility Emergency Plan and implementing procedures at least once per 12 months.
f. The Facility Security Plan and implementing procedures at least once per 12 months.
g. Any other area of facility operation considered appropriate by the NRB or the Vice President - Nuclear.
h. The Facility Fire Protection Program and implementing procedures at least once per 24 months.
i. An independent fire protection and loss prevention program inspection and audit shall be performed at least once per 12 months utilizing either qualified offsite licensee personnel or an outside fire protection firm.*
j. An inspection and audit of the fire protection and loss prevention program shall be performed by a qualified outside fire consultant at least once per 36 months.
k. The radiological environmental monitoring program* and the results thereof at least once per 12 months.
AUTHORITY 6.5.2.9 The NRB shall report to and advise the Vice President - Nuclear on those areas of responsibility specified in Sections 6.5.2.7 and 6.5.2.8.
RECORDS 6.5.2.10 Records of NRB activities shall be prepared, approved and distributed as indicated below:
a. Minutes of each NRB meeting shall be prepared, approved and forwarded to the Vice President - Nuclear within 14 days *following_each meeting.
SALEM - UNIT 2 6-11
ADMINISTRATIVE CONTROLS
b. Reports of reviews encompassed by Section 6.5.2.7 above, shall be prepared, approved and forwarded to the Vice President - Nuclear within 14 days following completion of the review. l
c. Audit reports encompassed by Section 6.5.2.8 above, shall be forwarded to the Vice.President - Nuclear and to the management positions responsible for the areas audited within 30 days after completion of the audit.
6.5.3 SAFETY REVIEW GROUP {SRG)
FUNCTION 6.5.3.1 The SRG shall perform independent reviews of plant operations and to advise appropriate station/corporate management on the overall safety of plant operations.
COMPOSITION 6.5.3.2 The SRG shall be composed* of at least five dedicated, full-time engineers, located on Artificial Island and shall report to the General Manager -
Nuclear Support.
RESPONSIBILITIES 6.5.3.3 The SRG shall be responsible for:
a. Review of selected plant operating characteristics, NRC issuances, industry advi-sories, and othe sources of plant design/operating*
experience data that may indicate areas for improving plant safety.
b. Reviews of selected facility features, equipment and systems.
c. Reviews of selected procedures, and plant activities including main-tenance, modifications, operational problems and operational analysis.
d. Surveillance of selected plant operations and maintenance activities to provide independent verification* that they are performed correctly and t~at human errors are reduced to as low as reasonably achievable.
AUTHORITY 6.5.3.4 The SRG shall make detailed recommendations for improving plant safety to the appropriate station/corporate management through the General Manager -
Nuclear Support.
Not responsible for sign-off function *
SALEM - UNIT 2 6-12
ADMINISTRATIVE CONTROLS 6.6 REPORTABLE OCCURRENCE ACTION 6.6.1 The follpwing actions shall be taken for REPORTABLE OCCURRtNCES:
a. The Commission shall be notified and/or a report submitted pursuant to.
the requirements of Specification 6.9.
b. Each REPORTABLE OCCURRENCE requiring 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months notification to the Commission shall be reviewed by the SORC and submitted to the NRB and
. the Vice President - Nuclear.
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 Vice President - Nuclear I
and the NRB 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 Commis-sion, the NRB and the Vice President - Nuclear within 14 days of the l violation.
SALEM - UNIT 2 6-12a
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 11 A11 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, and changes thereto, shall be reviewed -by the SORG and approved by the General Manager - 1*
Salem Operations prior to implementation and reviewed periodically as set forth in administrative procedures.
6.8.3 Temporary 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, reviewed by the SORC and approved by the General Manager - Salem Operations within 14 days of implementation.
SALEM - UN IT 2 6-13
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 require-ments, and (ii) Integrated leak test requirements for each system at refueling cycle intervals or less.
b. In-Plant Radiation Monitoring 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)
(ii)
Training of personnel, Procedures for monitoring, and (iii) Provisions for maintenance o{ 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, (vi) A procedure identifying (a) the authority responsible for the interpretation of the data, and (b) the sequence and timing of administrative events required to initiate corrective action.
SALEM - UNIT 2 6-14
ADMINISTRATIV~
d.
CONTROLS Backup Method for Determining Subcooling Margin A program which will ensure the capability to accurately monitor the Reactor Coolant System Subcooling Margin. This program shall include the following:
(i) Training of personnel, and (ii) Procedures for monitoring.
6.9 REPORTING REQUIREMENTS ROUTINE REPORTS AND REPORTABLE OCCURRENCES 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 Administra-tor of the Regional Office of Inspection and Enforcement 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 the 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 (1) 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, whi,chever 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 REPORTSl/
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. The initial report shall be submitted prior to March 1 of the year following initial criticality.
1;' 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 2 6-15
ADMINISTRATIVE CONTROLS 6.9.1.5 Reports required on an annual basis shall include:
a. A tabulation on an annual basis of the number of station, utility and other personnel (including contractors) receiving exposures greater than 100 mrem/yr and their associated man rem exposure according to work and job f.unctions,2/ e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling. The* dose assignment to various duty functions may be estimates based on pocket dosimeter, TLD, or film badge measurements. Small exposures totalling less than 20% of the individual total dose need not be accounted for.
In the aggregate, at least 80% of the total whole body dose received
from external sources shall be assigned to specific major work functions.
b. The complete results of steam generator tube inservice inspections performed during the report period (reference Specification 4.4.5.5.b). _}
MONTHLY OPERATING REPORT 6.9.1.6 Routine reports of operating statistics and shutdown experience,
including documentation of all challenges to the PORVs or safety valves; shall be submitted on a monthly basis to the Director, Office of Management Informa-tion and Program Analysis, U.S. Nuclear Regulatory Commission, Washington, D.C.
20555, with a copy of the Regional Office of OI&E, no later than the 15th of
each month following the calendar month covered by the report.
REPORTABLE OCCURRENCES 6.9.1.7 The REPORTABLE OCCURRENCES of Specifications 6.9.1~8 and 6.9.1.9 below, including corrective actions and measures to prevent recurrence, shall be reported to the NRC. Supplemental reports may be required to fully describe final resolution of occurrence. In case of corrected or supplemental reports, a licensee event report shall be completed and reference shall be made to the original report date.
PROMPT NOTIFICATION WITH WRITTEN FOLLOWUP 6.9.1.8 The' types of events listed below shall be reported within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by telephone and confirmed by telegraph, mailgram, or facsmile transmission to the Administrator of the Regional Office, or his designate no later than the first working day following the event, with a written followup report within 14 days.
The wirtten followup report shall include, as a minimum, a completed copy of a licensee event report form. Information provided on the licensee event report form shall be supplemented, as needed, by additional narrative material to provide complete explanation of the circumstances surrounding the event.
2; This tabulation supplements the req~irements of section 20.407 of 10 CFR Part 20.
SALEM - UNIT 2 6-16
ADMINISTRATIVE CONTROLS
a. Failure of the reactor protection system or other systems subject to limiting safety system settings to initiate the required protective function by the time a monitored parameter reaches the setpoi nt speci-fied as the limiting safety system setting in the technical specifica-tions or failure to complete the required protective function.
b. Operation of the unit or affected .systems when any parameter or opera-tion subject to a limiting condition for operation is less conserva-tive than the least conservative aspect of the limiting condition for operation established in the technical specifications.
c. Abnormal degradation discovered in fuel cladding, reactor coolant pressure boundary, or primary containment.
d. Reactivity anomalies involving disagreement with the predicted value of reactivity ba 1ance under steady state conditions du ring power operation greater than or equal to 13 delta k/k; a calculated reac-tivity balance indicating a SHUTDOWN MARGIN less conservative than specified in the technical specifications; short-term reactivity increases that correspond to a reactor period of less than 5 seconds or, if subcritical, an unplanned reactivity insertion of more than 0.5% delta k/k; or occurrence of any unplanned criticality *
e.
f.
Failure or malfunction of one*or more components which prevents or could prevent, by itself, the fulfillment of the functional require-ments of system(s) used to cope with accidents analyzed in the SAR.
Personnel error or procedural inadequacy which prevents or could prevent, by itself, the fulfillment of the functional requirements of systems requi~ed to cope with accidents analyzed in the SAR.
g. Conditions arising from natural or man-made events that, as a direct result of the event require unit shutdown, operation of safety systems, or other protective measures required by technical speci fi cations.
h. Errors discovered in the transient or accident analyses or in the methods used for such analyses as described in the safety analysis report or in the bases. for the technical specifications that have or could have permitted reactor operation in a manner less conservative than assumed in the analyses.
i. Performance of structures, systems, or components that requires remedial action or corrective measures to prevent operation in a manner less conservative than assumed in the accident analyses in the safety analysis report or technical specifications bases; or discovery during unit life of conditions not specifically considered in the safety analysis report or technical specifications that require remedial action or corrective measures to prevent the existence or development of a~ unsafe conditioh.
j. Offsite releases of radioactive materials in liquid and gaseous efflu-ents that exceed the limits of Specification 3.11.1.1 or 3.11.2.1.
SALEM - UNIT 2 6-17
ADMINISTRATIVE CONTROLS
k. Exceeding the limits in Specification 3.11.1.4 or 3.11.2.6 for the storage of radioactive materials in the listed tanks. The written follow-up report shall include a schedule and a description of.
activities planned and/or taken to reduce the contents to within the specified limits.
THIRTY DAY WRITTEN REPORTS 6.9.1.9 The types of events listed below shall be the subject of written reports to the Administrator of the Regional Office within thirty days of occurrence of the event. The written report shall include, as a minimum, a completed copy of a licensee event report form. Information provided on the licensee event report form shall be supplemented, as needed, by additional narrative material to provide complete explanation of the circumstances surrounding the event.
a. Reactor protection system or. engineered safety feature instrument settings which are found to be less conservative than those estab-
-*- b.
lished by the technical specifitations but which d.o not prevent the-fulf.i.llment_of the functional requirements of affected syst~ms.
Conditions leading to operation in a degraded mode permitted by a limiting condition for operation or plant shutdown requir:-ed by a limiting condition for operation.
c.
Observed inadequacies in the implementation of administrative or procedural controls which threaten to cause reduction of degree of redundancy provided in reactor protection systems or engineered safety feature systems. *
d. Abnormal degradation of systems other than thos~ specified in 6.9.1.8.c above designed to contain radioact1ve material resulting from the fission process.
e. An unplanned offsite releases of 1) more than 1 curie of radioactive material in liquid effluents, 2) more than 150 curies of noble gas in gaseous effluents. The report of an unplanned offsite release of radioactive material shall inclu~e the following information.
1. A description of the event and equipment involved.
2. Cause(s) for the unplanned release.
3. Actions taken to prevent recurrence ..
4
Consequences of the unplanned release.
SALEM - UNIT 2 6-18
--** **-*- ~-*--- -----*-- :-'---*--- - **-** -- _: _____ .
ADMINISTRATIVE CONTROLS ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT*
6.9.1.10 Routine Radiological Environmental Operating Reports covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year.
The A~nual Radiological Environmental Operating Reports shall .include summaries, interpretations, and an analysis of trends of the results of the radiological environmental surveillance activities for the report-period, including a com-parison with preoperational studies with operational controls (as appropriate),
and with previous environmental surveilance reports, and an assessment of the observed impacts of the plant operation on the environment. The reports shall also include the results of land use censuses required by Specification 3.12.2.
The Annual Radiological Environmental Operating Reports shall include the results of analysis of all radiological environmental samples and of all meas-urements taken during the period pursuant to the Table and Figures in the envi-ronmental radiation section of the ODCM; as well as summarized and tabulated re-*
sults of locations specified in these analyses and measurements in the format of the table in the Radiological Assessment Branch' Technical Position, Revision 1, November 1979. In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted as soon as possible in a supplementary report.
The reports shall also include the following: a summary description of the radiological environmental monitoring program; at least two legible maps** cov-ering ~11 sampling locations keyed to a table giving distances and directions from the* centerline of one reactor; the results of licensee participation in the.
Interlaboratory Comparison Program, required by Specification 3.12.1; and dis-cussion of all analyses in which the LLD required by Table 4.12-1.was not achievable.
SEMIANNUAL RADIOACTIVE EFFLUENT.RELEASE REPORT*
6.9.1.11 Routine Radioactive Effluent Release Reports covering the operation of the unit during the previous six months of operation shall be submitted within 60 days after January 1 and July 1 of each year.
The Radioactive Effluent Release Reports shall include a summary of the quanti-ties of radioactive liquid and gaseous effluents and solid waste released from the unit as outlined in Regulatory Guide 1.21. "Measuring, Evaluating, and
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.
In lieu of submission with the Annual Radiological Environmental Operating Report, the licensee has the option of retaining this summary of required meteorological data on site in a file that shall be provided to the NRC upon request.
SALEM - UNIT" 2 6-19
ADMINISTRATIVE CONTROLS Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plant,"
Revision 1, June 1974, with data summarized on a quarterly basis following the format of Appendix B thereof.
The Radioactive Effluent Release Report to be submitted within 60 days after January 1 of each year shall include an annual summary of hourly meteorological data collected over the previous year. This annual summary may be either in the form of an hour-by-hour listing of magnetic tape of wind speed, wind direction, atmospheric stability, and precipitation (if measured), or in the form of joint frequency distributions of wind speed, wind direction, and atmospheric .
stability.*** This same report shall include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the unit or station during the previous calendar year. This same report shall also include an assessment of the radiation doses from radioactive liquid and gaseous effluents to MEMBERS OF THE PUBLIC due to their activities inside the SITE BOUNDARY (Figure 5.1-3) during the report period. All assumptions used in making these assessments (i.e., specific activity, exposure time and location) shall be included in these reports. The meteorological conditions concurrent with the time of release of radioactive materials in gaseous effluents (as determined by sampling frequency and measurement) shall be used for determining the gaseous pathway doses. The assessment of radiation doses shall be performed in accordance with the OFFSITE DOSE CALCULATION MANUAL.*
The Radioactive Effluent Release Report to be submitted 60 days after January 1 of each year shall also include an assessment of radiation doses to the likely most exposed MEMBER OF THE PUBLIC from reactor rel eases and other nearby uranium fuel cycle sources (including doses from primary effluent pathways and direct radiation) for the previous calendar year to show conformance with 40 CFR Part 190, Environmental Radiation Protection Standards'for Nuclear Power Operation.
Acceptable methods for calculating the dose contribution from liquid and gaseous effluents are given in Regulatory Guide 1.109, Rev. 1, October 1977.
The Radioactive Effluent Release Reports shall include the following information for each class of solid waste (as defined by 10 CRF Part 61) shipped offsite during the report period:
a. Container volume,
b. *Total curie quantity (specify whether determined by measurement or estimate),
c. Principal radionuclides (specify whether determine by measurement or estimate),
d. Source of waste and processing employed (e.g., dewatered spent resin, compacted dry waste, evaporator bottoms),
In lieu of submission with the first half year Radioactive Effluent Release Report, the licensee has the option of retaining this summary of required meteorological data on site in a file that shall be provided to the NRC upon request.
SALEM - UN IT 2 6-20
ADMINISTRATIVE CONTROLS
e. Type of container (e.g., LSA, Type A, Type B, Large Quantity), and
f. Solidification agent or absorbent (e.g., cement, urea formaldahyde).
The Radioactive Effluent Release Reports shall include a list of description of unplanned releases from the site to UNRESTRICTED AREAS of radioactive materials in gaseous and liquid effluents made during the reporting period.
The Radioactive Effluent Release Reports shall include any changes made during the* reporting period to the PROCESS CONTROL PROGRAM (PCP) and to the OFFSITE DOSE CALCULATION MANUAL (ODCM), as well as a listing of new locations for dose calculations and/or environmental monitoring identified by the land use census pursuant to Specification 3.12.2.
SPECIAL REPORTS 6.9~2 Special reports shall be submitted to the Administrator of the Regional Office within the time period specified for each report.
6.10 RECORD RETENTION In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retined for at least the minimum 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 iTiterv~l at each power 1evel *
b. Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety.
c. ALL REPORTABLE OCCURRENCES submitted to the Commission.
d. Records of surveillance activities, inspections and calibrations required by these Technical Specifications.
e. Records of reactor ~ests and experiments.
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.
SALEM - UNIT 2 6-21
ADMINISTRATIVE CONTROLS 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.
b. Records of new and irradiated fuel inventory, fuel transfers and assembly burnup histories.
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 performed pursuant to these Technical Specifications.
i. Records of Quality Assurance activities required by the QA Manual.
j. Records of reviews performed for changes made to procedures or equi~ment or reviews of te~ts and experiments pursuant to 10 CFR 50.59.
k. Records of meetings of the SORC and the NRB.
l. Records for Environmental Qualification which are covered under the provisions of Paragraphs 2.C(7) and 2.C(8) of Facility Operating License DPR-75.
m. Records of the service lives of all hydraulic and mechanical snubbers listed on Tables 3.7-4a and 3.7-4b 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.
SALEM - UNIT 2 6-22
ADMINISTRATIVE CONTROLS 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.
6.12 . HIGH RADIATION AREA 6.12.1 In lieu of the control device or alarm signal required by paragraph 11 11 11 11 20.203(c)(2) of 10 CFR Part 20, each high radiation area in which the intensity of radiation is greater than 100 *mrem/hr but less* than 1000 mrem/hr shall be barricaded and conspicuously posted as a High Radiation Area and entrance thereto shall be controlled by issuance of a Radiation Exposure Permit*. Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:
a. A radiation monitoring device which continuously indicates the radiation dose rate in the area.
b. A radiation monitoring device which continuously integrates the.
radiation dose *rate in the area ~nd alarms when~ preset integrated dose is received. Entry into- such areas with this monitoring device may be made after the dose rate level in the area has been established and personnel have been made knowledgeable of them.
c. A health physics qualified individual (i.e., qualified in radiation .
protection procedures) *with a radiation dose rate monitoring device who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the facility Health Physicist in the Radiation Work Permit.
6.12.2 In addition to the requirements of 6.12.1, areas accessible to personnel with radiation levels such that a major portion of the body could receive in one hour a dose greater than 1000 mrem shall be provided with locked doors to prevent unauthorized entry, and the keys shall be maintained under the administrative control of the Senior Shift Supervisor on duty and/or*Senior Supervisor - Radiation Protection. Doors shall remain locked except during periods of access by personnel under an approved Radiation Work Permit which shall specify the dose rate levels in the immediate work area and the maximum allowable stay time for individuals in that area. For individual areas accessible to personnel with radiation levels such that a major portion of the body could receive in one hour
Health Physics Personnel or personnel escorted by Health Physics personnel shall be exempt from the RWP issuance requirement during the performance of their assigned radiation protection duties, provided they are otherwise following plant radiation protection procedures for entry into high radiation areas.
SALEM - UNIT 2 6-23
ADMINISTRATIVE CONTROLS a dose in excess of 1000 mrem** that are located within large areas, such as PWR containment, where no enclosure exists for purposes of locking, and no enclosure can be reasonable constructed around the indi~idual areas, then that area shall be roped off, conspicuously posted and a flashing light shall be activated as a warning device. In lieu of the.stay time specification of the RWP, direct or remote (such as use of closed circuit TV cameras) continuous*
surveillance may be made by personnel qualHied in radiation protection procedures to provide positive exposure control over the activities within the area.
6.13 PROCESS CONTROL PROGRAM {PCP) 6.13.1 The PCP shall be approved by the Commission prior to implementation.
6.13.2 Licensee initiated changes to the PCP:
1. Shall be submitted to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the change(s) was made. This submittal shall contain:
a. Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information;
b. A determination that the change did not reduce the overall conformance of the solidified waste product to existing criteria for solid wastes; and c *. Documentation of the fact that the change has been reviewed and found acceptable by the SORC.
2. Shall become effective upon review and acceptance by the SORC.
6.14 OFFSITE DOSE CALCULATION MANUAL (ODCM) 6.i4.1 The.ODCM shall be approved by the Commission prior to implementation.
6.14.2 Licensee initiated changes to the ODCM:
1. Shall be* submitted to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the change(s) was made. This submittal shall contain:
a. Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information. Information submitted should consist of a package of those pages of the ODCM to be changed with each page numbered and provided with an approval and date box, together with appropriate analyses or evaluations justifying the change(s);
Measurement made at 18 11 from source of radi oacti vi-ty.
SALEM - UNIT .2 6-24
- ----------~------~----*--*---
.......... *------~--.---~------------. ****---*---- --*****-**-- --*---------*----- **----------*------**- *--*------...--------------
ADMINISTRATIVE CONTROLS
\
b. A determination that the change will not reduce the accuracy or reliability of dose calculations or setpoint determination; and
c. Documentation of the fact that the change has been reviewed and found acceptable by the SORC.
2. Shall become effective upon review and acceptance* by the SORC.
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 FSAR for the period in which the evaluation was reviewed by (SORC). The discussion of each changes shall contain:
a. A summary of the_ evaluation that led to the determination could be made in accordance with 10 CFR Part 50.59;
b. Sufficient detajled informa~ion to totally support the reason for the change without benefit of additional or supplemental information; *
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.
SALEM - UNIT 2 6-25
LCR 83-08 DESCRIPTION OF CHANGE This change incorporates NRC notification requirements for Reactor Trip Breakers' and Reactor Trip Bypass Breakers' maintenance testing results that exceed acceptance criteria and for measured trip forces that exceed the acceptable upper limit. The proposed change also incorporates additional surveillance requirements committed to the NRC as part of PSE&G's corrective action program associated with the Reactor Trip and Bypass Breakers.
REASON FOR CHANGE This proposed change is the result of a PSE&G's commitment to submit, within 30 days of Startup of Unit No. 1, Technical Specifications covering the maintenance and surveillance testing for the Reactor Trip and Bypass Breakers and reporting requirments associated with that testing .
SAFETY EVALUATION This proposed change adds testing and reporting requirements which were reviewed and accepted by the NRC in their SER (NUREG-0995). No margin of safety is reduced by the change which ensures that existing equipment is more thoroughly tested and that any testing results that exceed acceptable limits are immediately reported to the NRC. This additional testing will identify any degradation of the trip breakers and, in conjunction with the long term operability verification program, will further enhance safety. This proposed change, therefore, would not:
(1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.
(/)
)>
r rn 3:
TABLE 3.3-1 (Continued) c
z
-I REACTOR TRIP SYSTEM INSTRUMENTATION I-'
MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION
18. Turbine Trip
a. Low Autostop Oil Pressure 3 2 2 1 7#
b. Turbine Stop Valve Closure 4 4 3 1 6#
19. Safety Injection Input from w
+::>
w I
+::>
20.
SSPS Reactor Coolant Pump Breaker 2 1 2 1,2 1 Position Trip
a. Above P-8 l/breaker 1 l/breaker 1 10
b. Above P-7 l/breaker 2 l/breaker 1 11 per oper-ating loop
21. Reactor Trip Breakers 2 1 2 1, 2 and
1###
22. Automatic Trip Logic 2 1 2 1, 2 and
1
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.
The channel (s) associated with the protective functions derived from the out of service Reactor Coolant Loop shall be placed in the tripped condition.
The provisions of Specification 3.0.4 are not applicable.
High voltage to detector may be de-energized above P-6.
If ACTION Statement 1 is entered due to the results of maintenance testing on Reactor Trip Breakers (RTB) or Reactor Trip Bypass Breakers (RTBB) exceeding any procedural acceptance criteria, or RTB/RTBB trip forces exceeding the recommended upper limit, immediate NRC notification [within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and prior to any repair or adjustment to the affected breaker(s)] is required in accordance with Specification 6.9.2 ACTION STATEMENTS ACTION 1 - With the number of channeis 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 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months for
.surveillance testing per Specification 4.3.1.1.1.
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 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .
b. The Minimum Channels OPERABLE requirement is met; however, one additional 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.
c. Either, THERMAL POWER is restricted to < 75% of RATED THERMAL 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 .
ACTION 3 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement and with the THERMAL POWER level :
~ SALEM - UNIT 1 3/4 3-5
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 7 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~ 3 percent.
(4) - Manual SSPS functional input check every 18 months.
( 5) Each train or logic channel shall be tested at least every 62 days on a STAGGERED TEST BASIS.
( 6) Neutron detectors may be excluded from CHANNEL CALIBRATION.
(7) Below P-6 (Block of Source Range Reactor Trip) setpoint.
(8) Logic only, if not performed in previous 92 days.
(9) If not performed in the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , conduct a functional test of the Manual Reactor Trip Switches (using voltmeters).
(10) - If. not performed in the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , conduct a functional test of:
Reactor Trip Breaker UV Trip (via SSPS)
Reactor Trip Breaker Shunt Trip (via manual pushbutton controls)
(11) - Perform a functional test of:
Reactor Trip Breaker UV Trip (via SSPS) and conduct response time testing of UV/Breakers (event recorders)
Reactor Trip Breaker Shunt Trip (via manual pushbutton controls)
(12) - Perform periodic maintenance on Reactor Trip Breakers and Reactor Trip Bypass Breakers semiannually as follows:
a. response time testing, (3 times) (visicorder) trend data
b. trip bar lift force measurements
c. UV output force measurement
d. dropout voltage check
e. servicing/lubrication/adjustments (See Table 3.3-1 Notation ###)
f. repeat testing steps (a-d) following any necessary actions at step (e)
~ SALEM - UNIT 1 3/4 3-13 I -* ..
~
(./)
J;:>
I rl1 3:
TABLE 3.3-1 (Continued) c=
z 1---i REACTOR TRIP SYSTEM INSTRUMENTATION
--l N
MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION
18. Turbine Trip
a. Low Autostop Oil Pressure 3 2 2 1 7#
b. Turbine Stop Valve Closure 4 4 4 1 711
19. Safety Injection Input from SSPS 2 1 2 1,2 1
- w
. p.
w I
.p.
20. Reactor Coolant Pump Breaker Position Trip
a. Above P-8 1/breaker 1 1/breaker 1 10
b. Above P-7 1/breaker 2 1/breaker 1 11 per oper-ating loop
21. Reactor Trip Breakers 2 1 2 1, 2 and
1111111
22. Automatic Trip Logic 2 1 2 1, 2 and
1
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.
The channel (s) associated with the protective functions derived from the out of service Reactor Coolant Loop shall be placed in the tripped condition.
The provisions of Specification 3.0.4 are not applicable.
High voltage to detector may be de-energized above P-6.
If ACTION Statement 1 is entered due to the results of maintenance testing on Reactor Trip Breakers (RTB) or Reactor Tr.ip Bypass Breakers (RTBB) exceeding any procedural acceptance criteria, or RTB/RTBB trip forces exceeding the recommended upper limit, immediate NRC notification [within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and prior to any repair or adjustment to the affected breaker(s)] is required in accordance with Specification 6.9.2 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 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 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .
b. The Minimum Channels OPERABLE requirement is met; however, one additional 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.
c. Either, THERMAL-POWER is restricted to less than or equal to 75%
of RATED THERMAL and the Power Range, Neutron Flux trip setpoint is reduced to less than or equal 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 .
d. The QUADRANT POWER TILT RATIO, as indicated by the rema1n1ng three detectors, is verified consistent with the normalized symmetric power distribution obtained by using the movable in-core detectors in the four pairs of symmetric thimble locations at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when THERMAL POWER is greater than 75% of RATED THERMAL POWER
SALEM - UNIT 2 3/4 3-5
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 the P-6 (Block of Source Range Reactor Trip) setpoint, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the P-6 Setpoint.
b. Above the P-6 (Block of Source Reactor Trip) setpoint 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.
d. Above 10% of RATED THERMAL POWER, the provisions of Specification 3.0.3 are not applicable.
ACTION 4 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement and with the THERMAL POWER 1eve l :
a. Below the P-6 (Block of Source Range Reactor Trip) setpoint, restore the inoperable channel to OPERABLE status prior to
b.
increasing THERMAL POWER above the P-6 Setpoint.
Above the P-6 (Block of Source Range Reactor Trip) setpoint, operation may continue ..
ACTION 5 - With the number of channe 1s 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 ~hannel is placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .
b. The Minimum Channel OPERABLE requirement is met; however, one additional 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.
ACTION 7 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed until performance of the next required CHANNEL FUNCTIONAL TEST provided the inoperable channel is placed in the tripped condition with 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months *
SALEM - UNIT 2 3/4 3-6
TABLE 4.3-1 (Continued)
REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS c:
z:
--i CHANNEL MODES lN WHICH N CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED
13. Loss of Flow - Two Loops s R N.A. 1
14. Steam Generator Water Level--
Low-Low s R M 1, 2
15. Steam/Feedwater Flow Mismatch &
Low Steam Generator Water Level s R M 1, 2
16. Undervoltage - Reactor Coolant Pumps N.A. R M 1
17. Underfrequency - Reactor Coolant w N.A. R M 1 I
I-'
Pumps N
18. Turbine Trip
a. Low Autostop Oil Pressure N.A. N.A. S/U(l) N.A.
b. Turbine Stop Valve Closure N.A. N.A. S/U(l) N.A.
19. Safety Injection Input from SSPS N.A. N.A. M(4) 1, 2
20. Reactor Coolant Pump Breaker Position Trip N.A. R S/U(B) 1
21. Reactor Trip Breaker N.A. N.A. S/U ( 10) , M( 11) 1, 2 and
and SA(12)
22. Automatic Trip Logic N.A. N.A. M(5) 1, 2 and
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 7 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~ 3 percent.
( 4) Manual SSPS functional input check every 18 months.
( 5) Each train or logic channel shall be tested at least every 62 days on a STAGGERED TEST BASIS.
( 6) Neutron detectors may be excluded from CHANNEL CALIBRATION.
(7) Below P-6 (Block of Source Range Reactor Trip) setpoint.
(8) Logic only, if not performed i~ previous 92 days.
(9) If not performed in the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , conduct a functional test of
( 10) -
the Manual Reactor Trip Switches (using voltmeters).
If not performed in the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , conduct a functional test of:
Reactor Trip Breaker UV Trip (via SSPS)
Reactor Trip Breaker Shunt Trip (via manual pushbutton controls)
(11) - Perform a functiona1 test of:
Reactor Trip Breaker UV Trip (via SSPS) and conduct response time testing of UV/Breakers (event recorders)
Reactor Trip Breaker Shunt Trip (via manual pushbutton controls)
(12) - Perform periodic maintenance on Reactor Trip Breakers and Reactor Trip Bypass Breakers semiannually as follows:
a. response time testing, (3 times) (visicorder) trend data
b. trip bar lift force measurements
c. UV output force measurement
d. dropout voltage check
e. servicing/lubrication/adjustments (See Table 3.3-1 Notation ###)
f. repeat testing steps (a-d) following any necessary actions at step (e)
SALEM - UNIT 2 3/4 3-13
'\

ML18089A198

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