Text

Forwards First Draft Tech Specs,Description of SER Items Requiring Util to Propose Tech Specs & Current Schedule for Tech Spec Review Showing Major Util Milestones
	ML20147D900
Person / Time
Site:	Comanche Peak
Issue date:	01/06/1988
From:	Charemagne Grimes; NRC OFFICE OF SPECIAL PROJECTS
To:	Counsil W; TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC)
References
	NUDOCS 8801200361
	Download: ML20147D900 (486)
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+f org[o UNITED STATES 8 R' go NUCLEAR REGULATORY COMMISSION

{ ; WASHINGTON, D. C. 20665

% January 6, 1988

. t Docket No. 50-445 Mr. William G. Counsil -

Executive Vice President Texas Utilities Electric Company

. 400 North Olive Street, L.B. 81 Dallas, Texas 75201 '

Dear Mr. Counsil:

SUBJECT:

TECHNICAL SPECIFICATIONS FOR COMANCHE PEAK STEAM ELECTRIC STATION (CPSES), UNIT 1 i

( Enclosed for your review is a typed copy of the first draft technical specifications (TS) for CPSES, Unit 1, (Enclosure 1). which were forwarded to you by letter-on.Decembe'r. 3, 1987. This first draft was prepared from -

tha current Westinghouse Standard Technical Specifications (STS), forwarded to you on August 14, 1987, sups!emented by information from the Safety Evaluation Report related to t1e operation of the CPSES, Units 1 and 2 t I (NUREG-0797), the Final Draft Comanche Peak TS dated October 1984, and the.

l marked-up version of the STS you submitted on October 30,1987 (TX)(-6905).

Enclosure 1 is being placed in NRC's Public Document Room and the Local Public Document Room. The final Comanche Peak TS', to be issued with the operating ;

license, will be provided to the service list.

~ '

Enclosura 2 is a description of the Safety Evaluation Re) ort (SER) items for ;

which technical specifications are required' and need to se proposed by t TV Electric. Your October 30, 1987 mark-up of the STS did not include these plant-specific TS nor did you provide justifications for not including them. :

All other SER-required TS have been incorporated by TV Electric or by the staff in Enclos r e 1. p Provided in Inclosure 3 is our current schedule for T3 review showing major TV Electric.41estones. TU Electric should make every effort to meet its dates to ensure completion of the TS review, i.e., completion of Appendix A '

to the CPSES, Unit 1 operating license, on June 30, 1988 as requested in your August 28, 1987 letter (TXX-6691). The next scheduled milestone is the site '

visit by the NRC staff reviewers during the week of Janury 11, 1988 to resolve differences between TV Electric's mark-up of the STS and the first draft TS and i to develop a list of TV Electric and NRC action items, i i

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January 6, 1988 Mr. W. G. Counsil Should you have any questions, please contact either one of our Project Managers, Melinda Halloy at (301)492-0738 orAnnetteVietti-Cookat(301)492-0737.

Sincerely, (original signed by)

Christopher I. Grimes, Director Comanche Peak Project Division Office of Special Projects

Enclosures:

1. First draf t TS

2. SER Items Requiring TS

3. TS Review Milestones cc: R. D. Walker TV Electric cc: w/o enclosure 1: See next page DISTR BJTION (all enclosures) DISTRIBUTION (w/o enclosure 1) f Dec m 711e .- UlF Reading

'" 'NRC'PDR SEbneter/JAxelrad Local PDR CGrimes CPPD Reading PMcKee RWarnick JLyons MMalloy AYietti-Cook 0Kelley OGC-Bethesda FMiraglia EJordan JPartlow CRossi EButcher REmch RGiardina 1 r ACRS(10) 7 IlNde As MP DD:CPP .

CPPD: t 0:CPPD:05 MMally[.cm AVi tti-Cook PfMcKeeg CIGrimes 01/(f/88 01/4/88 ,0) 68 01/b/68 CPP :05)

JHW afA 9 01/ /fd8 J

Mr. W. G. Counsil .

Should you have any questions, please contact either one of our Project Managers Melinda Malloy at (301)492-0738 or Annette Vietti-Cook at (301)492-0737.

Sincerely,

(\f wk 'M Christopher 1. Grimes, Director Comanche Peak Project Division Office of Special Projects

Enclosures:

1. First draft TS

2. SER Items Requiring TS

3. TS Review Milestones

~

cc: R. D. Walker TU Electric i

cc: w/o enclosure 1: See next page .

9

W. G. Counsil Comanche Peak Steam Electric Station Texas Utilities Electric Company Units 1 and 2 CC:

Jack R. Newman, Esq. Asst. Director for Inspec. Programs Newman & Holtzinger, P.C. Comanche Peak' Project Division Suite 1000 U.S. Nuclear Regulatory Commission 1615 L Street, N.W. P. O. Box 1029 Washington, D.C. 20036 Granbury, Texas 76048 Robert A. Wooldridge, Esq. Regional Administrator, Region lY Worsham Forsythe, Sampels & U.S. Nuclear Regulatory Commission Wooldridge 611 Ryan Plaza Drive, Suite 1000 2001 Bryan Tower, Suite 2500 Arlington, Texas 76011 Dallas, Texas 75201 Lanny'A. Sinkin Mr.' Homer C. Schmidt Christic Institute Director of Nuclear Services 1324 North Capitol. Street Texas Utilities Electric Company Washington. 0.C. 20002 Skyway Tower 400 North Olive Street, L.B. 81 Ms. Billie Pirner Garde, Esq.

Dallas, Texas 75201

Government Accountability Project Midwest' Office Mr. Robert E. Ba1. lard, Jr. 104 East Wisconsin Avenue Director of Projects . .Appleton, Wisconsin 54911 Gibbs and Hill, Inc.

11 Penn Plaza New York, New York 10001 David R. Pigott, Esq.

Orrick, Herrington & Sutcliffe 4

600 Montgomery Street Mr. R. S. Howard San Francisco, California 94111 Westinghouse Electric Corporation P. O. Box 355 Anthony Z. Roisman, Esq.

Pittsburgh, Pennsylvania 15230 Suite 600 1401 New York Avenue, NW Renea Hicks. Esq. Washington, D.C. 20005 Assistant Attorney General Environmental Protection Division Robert Jablon P. O. Box 12548, Capitol Station Bonnie S. Blair Spiegel & McDiarmid l Austin, Texas 78711 1350 New York Avenue, NW Mrs. Juanita Ellis, President Washington, D.C. 20005-4798 Citizens Association for Sound Energy 1426 South Polk George A. Parker, Chairman Dallas, Texas 75224 Public Utility Committee Senior Citizens Alliance Of Ms. Nancy H. Williams Tarrant County, Inc. '

CYGNA Energy Services 6048 Wonder Drive

2121 N. California Blvd., Suite 390

Fort Worth, Texas 76133 '

i Walnut Creek, CA 94596 1

W. G. Counsil Comanche Peak Electric Station Texas Utilities Electric Company Units 1 and 2 CC:

Joseph'F. Fulbright '

Fulbright 8 Jaworski 1301 McKinney Street -

Houston, Texas 77010 Roger D. Walker Manager, Nuclear Licensing Texas Utilities Electric Company Skyway Tower 400 North Olive Street, L.B. 81 Dallas, Texas 75201 Mr. Jack.Redding c/o Bethesda Licensing Te'xas Utilities Electric Company 3 Metro Center, Suite 610 Bethesda, Maryland 20814

. William A. Burchette, Esq.

Counsel for Tex-La Electric Cooperative of Texas Heron, Burchette, Ruckert &'Rothwell Suite 700 1025 Thomas Jefferson Street, NW

Washington, D.C. 20007 James P. McGaughy, Jr.

GDS Associates, Inc.

Suite 720 1850 Parkway Place Marietta, Georgia 30067 Administrative Judge Peter Bloch U.S. Nuclear Regulatory Commission -

Washington, D.C. 20555 Elizabeth B. Johnson Administrative Judge Oak Ridge National Laboratory P. O. Box X Building 3500 Oak Ridge Tennessee 37830 .

Dr. Kenneth A. McCollom ,

1107 West Knapp Stillwater, Oklahoma 74075 .

Dr. Walter H. Jordan c/o Carib Terrace Hotel 522 N. Ocean Boulevard Pompano Beach, Florida 33062

ENCLOSURE 1 First Draft Technical Specifications Comanche Peak Steam Electric Station, Unit 1 .

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COMANCHE PEAK - UNIT 1

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INDEX ch.;

DEFINITIONS SECTION PAGE

1. 0 DEFINITIONS................................................... 1-0

'1,1 ACTI0N...........................,.......'...........'.....'..... ' 1-1 1.2 ACTUATION LOGIC TEST..............................'............ 1-1 1.3 ANALOG CHANNEL OPERATIONAL TEST............................... 1-1 1.4 AXIAL FLUX DIFFERENCE......................................... 1-1 1.5 CHANNEL CALIBRATION........................................... 1-1 1.6 CHANNEL CHECK................................................. 1-1 1.7 CONTAINMENT INTEGRITY......................................... 1-2 1.8 CONTROLLED LEAKAGE............................................ 1-2 1.9 CORE ALTERATION........................... ................... 1-2 1.10 DIGITAL CHANNEL OPERATIONAL TEST.~...........c...;............. 1-2 1.11 DOSE EQUIVALENT I-131........................................ 1-2 1.12 5-AVERAGE DISINTEGRATION ENERGY.............................. 1-3 1.13 ENGINEERED SAFETY FEATURES RESPONSE TIME......<............. . 1-3 1.14 EXCLUSION AREA B0VNDAhY....................................... 1-3 1.15 FREQUENCY N'0TATION........................................... 1-3 1.16 IDENTIFIED LEAKAGE........................................... 1-3 1.17 MASTER RELAY TEST............................................ 1-3 1.18 MEMBER (S) 0F THE PUBLIC...................................... 1-4 1.19 0FFSITE DOSE CALCULATION MANUAL.............................. 1-4 1.20 OPERABLE - OPERABILITY....................................... 1-4 1.21 OPERATIONAL MODE - M0DE...................................... 1-4 1.22 PHYSICS TESTS................................................ 1-4 1.23 PRESSURE BOUNDARY LEAKAGE.................................... 1-4 1.24 PRIMAR) PLANT VENTILATION SYSTEM.............................. 1-5 1.25 PROCESS CONTROL PR0 GRAM...................................... 1-5 1.26 PURGE - PURGING...................... ....................... 1-5 1.27 QUADRANT POWER TILT RATI0.................................... 1-5 1.28 RATED THERMAL P0WER.......................................... 1-5 1.29 REACTOR TRIP SYSTF.M RESPONSE TIME............................ 1-5 1.3Q REPORTABLE EVENT................................'............. 1-5 1.31 SHUTDOWN MARGIN.............................................. 1-6 COMANCHE PEAK - UNIT 1 I

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INDEX DEFINITIONS SECTION PAGE 1.32 SITE B0VNDARY. .............................................. 1-6 1.33 SLAVE RELAY TEST...............................'.............. 1-6 1.34 SOLIDIFICATION...................e........................... 1-6 1.35 SOURCE CHECK................................................. 1-6 1.36 STAGGEREDTESTBASIS......................................... 1-6 1.37 THERMAL P0WER................................................ 1-6 1.38 TRIP ACTUATING DEVICE OPERATIONAL TEST....................... 1-6 1,39 UNIDENTIFIED LEAKAGE......................................... 1-7 1.40 UNRESTRICTED AREA............................................ 1-7 1.41 VENTING...................................................... 1-7 1.42 WASTE GAS HOLDUP SYSTEM...................................... 1-7 TABLE 1.1 FREQUENCY N0TATION...................................... 1-8 TABLE 1.2 OPERATIONAL M0 DES....................................... 1-9 b

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INDEX SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SECTION

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2.1 SAFETY LIMITS 2.1.1 REACTOR C0RE....................!........................... 2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE..........................,.. 2-1 FIGURE'2.1-1 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION.. 2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETP0!NTS............... 2-3 TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS.... 2-4 BASES SECTION PAGE 2.1 SAFETY LIMITS 2.1.1 REACTOR C0RE................................................ B 2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE............................. B 2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETP0!NTS......./....... B 2-3 i

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SECTION PAGE 3/4.0 APPLICABILITY...............................................

3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL Shutdown Margin - T avg Greater Than 200*F................ 3/4 1-1 Shutdown Margin - T Less Than or Equal to 200 F....... 3/4 1-3 avg Moderator Temperature Coefficient........................ 3/4 1-4 Minimum Temperature for Criticality...................... 3/4 1-6 3/4.1.2 B0 RATION SYSTEMS F l ew Pa th - S h u td own . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-7 Flow Paths - Operating................................... 3/4 1-8 Charging Pump - Shutdown................................. 3/4 1-9 Charging Pumps '0perating............................... 3/4 1-10 Borated Water Source - Shutdown.......................... 3/4 1-11 Borated Water Sources - Operating......................... 3/4 1-12 3/4.1.3 MOVABLE CONTROL ASSEMBLIES .

Group Height............................................. 3/4 1-14 TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE -

EVENT OF AN INOPERABLE FULL-LENGTH R0D................... 3/4 1-16 Position Indication Systems - Operating.................. 3/4 1-17 Position Indication System - Shutdown.................... 3/4 1-18 Rod Prop Tine............................................ 3/4 1-19 l Shutdown Rod Insertion Limit............................. 3/4 1-20 i Control Rod Insertion Limits............................. 3/4 1-21

! FIGURE 3.1-1 ROD BANK INSERIION LIMITS VERSUS THERMAL POWER l FOUR-LOOP 0PERATION....................................., 3/4 1-22 l

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INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 DIFFERENCE....................................

AXIAL FLUX 3/4 2-1 '

FIGURE 3.2-1 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL P0WER...................................... 3/4 2-3 3/4.2.2 HEAT FLUX HOT CHANNEL FACT 0R............................. 3/4 2-4 FIGURE 3.2-2 K(Z) - NORMALIZED F (Z) q AS A FUNCTION OF CORE HEIGHT. 3/4 2-5 3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACT 0R.................................................... 3/4 2-8 FIGURE 3.2-3 RCS TOTAL FLOW RATE VERSUS R - FOUR LOOPS IN 0PERATION............................................. 3/4 2-9 3/4.2.4 QUADRANT POWER TILT RATI0................................ 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 TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION....... 3.......... 3/4 3-2 TABLE 3.3-2' REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES.... 3/4 3-8 TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS............................................. 3/4 3-10 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION.......................................... 3/4 3-15 TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION.......................................... 3/4 3-17 TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETP0lNTS........................... 3/4 3-27 TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES............. 3/4 3-34

TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM l INSTRUMENTATION SURVEILLANCE REQUIREMENTS................ 3/4 3-39 3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitoring For Plant Operations................ 3/4 3-46 l

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LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION -

PAGE TABLE 3-3-6 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS..................................;.. 3/4 3-47 Movable Incore Detectors................................. 3/4 3-49 Seismic Instrumentation.................................. 3/4 3-50 TABLE 3.3-7 SEISMIC MONITORING INSTRUMENTATION.................... 3/4 3-51 TABLE 4.3-3 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS............................................. 3/4 3-52 Meteorological Instrumentation........................... 3/4 3-53 TABLE 3.3-8 METEOROLOGICAL MONITORING INSTRUMENTATION............. 3/4 3-54 Remote Shutdown System Instrumentation................... 3/4 3-55 TABLE 3.3-9 REMOTE SHUTDOWN SYSTEM HONITORING I.NSTRUMENTATION..... 3/4.3-56 TABLE.4.3-4 REMOTE SHU'TDOWN MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS................................ 3/4 3-57 Accident Monitoring Instrumentation...................... 3/4 3-58 TABLE 3.3-10 ACCIDENT MONITORING INSTRUMENTATION.................. 3/4 3-60 Chlorine Detection Systems............................... 3/4 3-62 Loose-Part Detection System /............................. 3/4 3-63 Radioactive Liquid Effluent Monitoring Instrumentation... 3/4 3-64 TABLE 3.3-11 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION 3/4 3-55 TABLE 4.3-5 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS................ 3/4 3-67 Radioactive Gaseous Effluent Monitoring Instrumentation.. 3/4 3-70 TABLE 3.3-12 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION....../................................... 3/4 3-71 TABLE 4.3-6 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS................ 3/4 3-73 3/4.3.4 TURBINE OVERSPEED PROTECTION.............................. 3/4 3-75 i

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LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.4 REACTCR COOLANT SYSTEM 3/4.4.1 R'EACTOR COOLANT LOOPS AND COOLANT CIRCULATION Startup,and Power Operation.............................. 3/4 4-1 Hot Standby.............................................. 3/4 4-2 Hot Shutdown............................................. 3/4 4-4 Cold Shutdown - Loops Fi11ed............................. 3/4 4-6 Cold Shutdown - Loops Not Fi 11ed. . . . . . . . . . . . . . . . . . . . . . . . . 3/4.4-7 3/4.4.2 SAFETY VALVES Shutdown............................................... 3/4 4-8 ;

0perating............................................. 3/4 4-9

'3/4. 4.'3 PRESSURIZER.................................... ;........ ' 3/44-10 ,

3/4.4.4 RELIEF VALVES..........................................., 3/4 4-11 3/4.4.5 STEAM GENERATORS......................................... 3/4 4-13 TABLE 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE. INSPECTION............................. 3/4 4-18 TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION....................... 3/4 4-19 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE l Leakage Detection Systems................................ 3/4 4-20 j Operational Leakage...................................... 3/4 4-21 TABLE 3.4-1 REACTOR C0'0LANT SYSTEM PRESSURE ISOLATION VALVES...... 3/4 4-23 :

3/4.4.7 CHEMISTRY................................................ 3/4 4-24 TABLE 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS............... 3/4 4-25 -

l 3/4.4.8 SPECIFIC ACTIVITY........................................ 3/4 4-26 FIGURE 3.4-1 00SE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENi 0F RATED THERMAL POWER l WITH THE REACTOR COOLANT SPECIFIC ACTIVITY > 1pCi/ gram DOSE EQUIVALENT I-131.................................... 3/4 4-27 TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM.................................................. 3/4 4-28 i

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INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System............................<........ 3/4 4-30 FIGURE'3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS -

APPLICABLE UP TO EFPY.............................. . 3/4 4-31 FIGURE 3.4-3 REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS -

APPLICABLE UP TO EFPY.............................. 3/4 4-32 TABLE 4.4-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM -

WITHDRAWAL SCHEDULE...................................... 3/4 4-33 Pressurizer............................... .............. 3/4 4-34 Overpressure Protection Systems.......................... 3/4 4-35 3/4.4.10 STRUCTURAL INTEGRITY..................................... 3/4 4-37 3/4.4.11 REACTOR COOLANT SYSTEM VENTS............................. . . 3/4 4-38 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS .

4 Cold Leg Injection........................................ 3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - T,yg ' GREATER THAN OR EQJAL TO 3503/4 5-3 F....

3/4.5.3 ECCS SUBSYSTEMS - T,yg LESS THAN 350 F ECCS Subsystem........................................... 3/4 5-7 j Safety Injection Pumps................................... 3/4 5-9

3/4. 5' . 4 REFUELING WATER STORAGE T NK............................. 3/4 5-10 1

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SECTION PAGE, 3/4.6 CONTAINMENT SYSTEMS .

3/4.6.1 PRIMARY CONTAINMENT Containment Integrity.................................... 3/4 6-1

.s Containment Leakage....................................... 3/4 6-2 A Containment Air Locks...................... ............. 3/4 6-4 Internal Pressure........................................ 3/4 6-6 e Air Temperature.......................................... 3/4 6-7 f ' Containment Structural Integrity......................... 3/4 6-8 Containment Ventilation System........................... 3/4 6-9

'3/4.6 7. DEPRESSURIZATION AND COOLING SYSTEMS t, IontainmentSpraySystem................................ 3/4 6-11

(# Spray Additive System.................................... 3/4 6-12 .

3/4L6.3 . CONTAINMENT ISOLATION VALVES............................. 3/4 6-13 TABLE 3. 6-1 CONTAINMENT ISOLATION VALVES. . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6-15 y ,

3/CS. 4 COM3USTIBLE GAS CONTROL 4 Hydrogen Monitors........................l............... 3/4 6-33

a. ',

/ Electri c }!ydrogen Recombi ners. . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6-34

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LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE .

3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE Safety Va1ves............................................ 3/4 7-1 TABLE 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING FOUR LOOP 0PERATION...................................... 3/4 7-2 TABLE 3.7-2 STEAM LINE SAFETY VALVES PER L00P..................... 3/4 7-2 Aux.iliary Feedwater System............................... 3/4 7-3 Condensate Storage Tank.................................. 3/4 7-5 Specific Activity........................................ 3/4 7-6 TABLE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM...........'.......................... 3/4 7-7 Main Steam Line Isolation Va1ves......................... 3/4 7-8 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION.......... 3/4 7-9 3/4.7.3 COMPONENT COOLING WA ER SYSTEM........................... 3/4 7-10 3/4.7.4 STATION SERVICE WATER SYSTEM............................. 3/4 7-11 .

3/4.7.5 ULTIMATE HEAT SINK............................ .......... '3/4 7-12 3/4.7.. PROTECTION.............'............................

FLOOD 3/4 7-13 3/4.7.7 CONTROL ROOM HVAC SY5 TEM................................. 3/4 7-14 3/4.7.8 PRIMARY PLANT VENTILATION SYSTEM - ESF FILTRATION UNITS.. 3/4 7-17 3/4.7.9 SNUBBERS....................... ....................... 3/4 7-19 FIGURE 4.7-1 SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST........... 3/4 7-24 3/4.7.10 SEALED SOURCE CONTAMINATI0h.............................. 3/4 7-25 4

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INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.7.11 ~ AREA TEMPERATURE M0NITORING.............................. 3/4 7-27 TABLE 3.7-3 AREA TEMPERATURE M0NITORING........................... 3/4 7-28 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES 0perating...................,............................. 3/4 8-1 TABLE 4.8-1 DIESEL GENERATOR TEST SCHECULE........................ 3/4 8-8 TABLE 4.8-2 ADDITIONAL RELIABILITY ACTI0NS........................ 3/4 8-8 Shutdown................................................. 3/4 8-11 3/4.8.2, D.C. SOURCES 0perating................................................ 3/4 8-12 TABLE 4.8-3 BATTERY SURVEILLANCE REQUIREM'ENTS..................... 3/4 8-14 Shutdown............... ................................. 3/4 8-15 3/4.8.3 ONSITE POWER DISTRIBUTION 0perating................................................ 3/4 8-16 Shutdown................................................. 3/4 8-18 l

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O0 l {1[id n LIMITING CONDITI0'NS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES A.C. Circuits Insid'e Primary Containment................. 3/4 8-19 Containment Penetration Conductor Overcurrent Protective Devices..................................... 3/4 8 20 TABLE 3.8-1 CONTA.INMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES........ .............................. 3/4 8-22 Motor-Operated Valves Thermal Overload Protection........ 3/4 8-38 TABLE 3.8-2 M'OTOR-0PERATED VALVES THERMAL OVERLOAD PROTECTION AND/0R BYPASS DEVICES.................................... 3/4 8-39 3/4.9 REFUELING OPERATIONS 3/4.9.1 , BORON C0NC'ENTRATION...................................... 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 REFUELING MACHINE........................................ 3/4 9-6 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE POOL BUILDING.......... 3/4 9-7 l 3/4.9.8 RE,IDUAL HEAT REMOVAL AND COOLANT CIRCULATION l

High Water Level........ ................................ 3/4 9-8 l

l Low Water Leve1.......................................... 3/4 9-9 l

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5 INDEX LI_MITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.9.9 CONTAINMENT VENTILATION ISOLATION SYSTEM................. 3/4 9-10 ,

3/4.9.10 WAT ER LEVE L - REACTOR VESS E L. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 9-11 3/4.9.11 WATER LEVEL - IRRADIATED FUEL STORAGE ................... 3/4 9-12 3/4.9.12 FUEL STORAGE P00L AIR CLEANUP SYSTEM..................... 3/4 9-13 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN..........,............................... 3/4 10-l' 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 . REACTOR COOLANT L00PS.................................... 3/4 10-4 3/4.10.5 POSITION INDICATION SYSTEM - SHUTD0WN.................. . 3/4 10-5 3/4.11 RADI0ACTIVt EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS Concentration............................................ 3/4 11-1 TABLE 4.11-1 EADI0 ACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PR03 RAM.................................................. 3/4 11-2 Dose.................................... ................ 3/4 11-5 l

l Liquid Radwaste Treatment System............. ..... ..... 3/4 11-6 Liquid Holdup Tanks.............................. .......

3/4 11-7 s

1 COMANCHE PEAK - UNIT 1 XIII l

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$l$

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS . _ _

SECTION PAGE 3/4.11.2 GASE0US EFFLUENTS Dose Rate......................... ............ ...... 3/4 11-8 TABLE 4.11-? RADI0 ACTIVE GASE0US WASTE SAMPLING AND ANALYSIS PR0 GRAM................................................ 3/4 11-9 '

Dose - Noble Gases....................................... 3/4 11-12 Dose - Iodine-131, Iodine-133, Tritium, and Radioactive

, Ma'terial in Particulate Form............................. 3/4 11-13 Gaseous Radwaste Treatment System........................ 3/4 11-14 Explosive Gas Mixture.................................... 3/4 11-15 Gas Storaga Tanks........................................ 3/4 11-16.

3/4.11.3 SOLID RADI0 ACTIVE WASTES........ ........................ 3/4 11-17 3/4.11.4 TOTAL 00SE...............s............................... 3/4 11-18 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING .

i 3/4.12.1 MONITORING PR0 GRAM....................................... 3/4 12-1 l

TABLE 3.12-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM........ 3/4 12-3 TABLE 3.12-2 REPORTING LEVELS FOR RADI0 ACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES................................. 3/4 12-10 TABLE 4.12-1 OETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSIS................................................. 3/4 12-11 3/4.12.2 LAND USE CENSUS.......................................... 3/4 12-14 3/4.12.3 INTERLABORATORY COMPARISON PR0 GRAM....................... 3/4 12-16 COMANCHE PEAK - UNIT 1 XIV l

INDEX el BASES SECTION PAGE APPLICABILITY,..........................'.....................

3/4.0 B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTR0L.......................................... B 3/4 1-1 3/4.1.2 B0 RATION SYSTEMS.....,.................................... B 3/4 1-2 3/4.1.3 MOVABLE CONTROL ASSEMBLIES.........'....................... B 3/4 1-3 3/4.2 POWER DISTRT.BUTION LIMITS................................... B 3/4 2-1

~

3/4.2.1 AXIAL FLUX D1FFERENCE................... ................. B 3/4 2-1 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR......... B 3/4 2-2 .

FIGURE B 3/4.2-1 TYPICAL INDICATED AXIAL FLUX DIFFERENCE VERSUS POWER.............................................

~

THERMAL. 'B 3/4 2-3 3/4.2.4 QUADRANT POWER TILT RATI0................................. B 3/4 2-5 3/4.2.5 DNB PARAMETERS............................................ B 3/4 2-6 l 3/4.3 INSTRUMENTATIO_N L 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM and ENGINEERED SAFETY l FEATURES ACTUATION SYSTER INSTRUMENTATION.................

. B 3/4 3-1 l

l l

3/4.3.3 MONITORING INSTRUMENTATION.............. ................. B 3/4 3-3 3/4.3.4 TURBINE OVERSPFED PROTECTION.............. . .. .......... C 3/4 3-6 l

P COMANCHC PEAK - UNIT I XV

. . . _ _ _ . _ . _. __ .- - r~ l

I*

INDEX .

pLu-BASES SECTION PAGE 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR C'00LANT LOOPS AND COOLANT CIRCule. TION.........'... . B 3/4 4-1 3/4.4.2 SAFETY VALVES........................................... B 3/4 4-1 3/4.4.3 PRESSURIZER............................................... B 3/4 4-2 3/4.4.4 RELIEF VALVES............................................. B 3/4 4-3 3/4.4.5 STEAM GENERATORS.......................................... B 3/4 4-3 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE............................ B 3/4 4-4 3/4.4.7 CHEMISTRY................................................. ,B 3/4 4-5 3/4.4.8 SPECIFIC ACTIVITY......................................... 3 3/4 4-6 3/4.4.9 PRESSURE / TEMPERATURE LIMITS.................... ......... B .3/4 4-7 TABLE B 3/4.4-1 REACTOR VESSEL T0VGHNESS.......................... B 3/4 4'9 FIGURE B 3/4.4-1 FAST NEUTRON FLUENCE (E>1MeV)'AS A FUNCT FULL POWER SERVICE LIFE.................. ...... ION OF

......... B 3/4 4-10 FIGURE'B 3/4.4-2 EFFECT OF FLUENCE AND COPPER CONTENT ON SHIFT OF RT FOR REACTOR VESSELS EXPOSED TO 550 F............ B 3/4 4-11 NDT 3/4.4.10 STRUCTURAL INTEGRITY................................. ... B 3/4 4-16 l 3/4.4.11 REACTOR COO LANT SYSTEM VENTS. . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 4-16 l

l 3/4.5 EMERGENCY CORE COOLING SYSTEMS

. 3/4.5.1 ACCUMULATORS.......... ................................... B 3/4 5-1

- 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS............................... B 3/4 5-1 i

l 3/4.5.4 REFUELING WATER STORAGE TANK. . . . . . . . . . . . . ................ B 3/4 5-2 j

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1 D.

c TNDEX BASES SECTION PAGE

' 3/4.'6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT.......................... ............ B 3/4 6-1 -

3/4.6.2 DEPRES.SURIZATION AND COOLING SYSTEMS.......... . ......... B 3/4 6-3 i 3/4.6.3 CONTAINMENT ISOLATION VALVES.............................. B 3/4 6-3 3/4.6.4 COMBUSTIBLE GAS CONTR0L................................... B 3/4 6-4 f.>MANCHE PE/V - UNIT 1 XVII t

'n1 . f 'v- - -< - - ,,,,rc

m n r ~~

INDEX 16 E3 i .

BASES SECTION PAGE

.i/4.7 PLANT SYSTEMS 3/4.7.1. TURBINE CYCLE............................................. B 3/4 7-1 3/4.7.2 STEAM GENER/10R PRESSURE / TEMPERATURE LIMITATION........... B 3/4 7-3 3/4.7.3 COMPONENT COOLING WATER SYSTEM............................ B 3/4 7-3 3/4.7.4 STATION SERVICE WATER SYSTEM.............................. B 3/4 7-3 3/4.7.S ULTIMATE HEAT SINK........................................ B 3/4 7-3 3/4.7.6 FLOOD PROTECTION......... ................................ B 3/4 7-4 3/4.7.7 CONTROL ROOM HVAC 5YSTEM.................................. B 3/4 7-4 3/4.7.9 PPIMARY PL."AT VENTILATION SYSTEM - ECF FILTRATION UNITS... B 3/4 7-5 3/4.7.9 SNUBBERS.................................................. B'3/4 7-5

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3/4.7.10 SEALED SOURCE CONTAMINATION.............. ............... B 3/4 7-7 3/4.7.11 AREA TEMPERATURE MONITORING................... .......... B 3/4 7-7 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1, 3/4.8.2, and 3/4.8.3 A.C. SOURCES, D.C. SOURCES, and ONSITE POWER DISTRIBUTION................................. B 3/4 8-1 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES................... B 3/4 8-3 -

t i

6 COMANCHE PEAK - UNIT 1 XVIII

l I DEX 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.S.4 CONTAINMENT BUILDING PENETRATIONS......................... B 3/4 9-1 3/4.9.5 C0'MUNICATIONS.......................................

M .... B 3/4 9-1 3/4.9.6 REFUELING MACHINE......................................... B 3/4 9-2 3/4.9.7 CRANE TRAVtt - SPENT FUEL STORAGE BUILDING................ B 3/4 9-2 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULAT. ION............. B 3/4 9-2 3/4.9.9 CONTAINMENT VENTILATI'ON ISOLATION SYSTEM............ ... B 3/4 9-2' 3/4.9.10 and 3/4.9.11 WATER'tEVEL - REACTOR VESSEL and -

I R R ADI ATE'; F U E L STO P 5 G E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 9-3 3/4.9.12 STORAGE POOL VENTILATION SY5 TEM........................... B 3/4 9-3 3/4.30 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUT 00WN MnP. GIN........... ................... ........... B 3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION, ANC POWER DISTRIBUTION LIMITS.... B 3/4 10-1 3/4.10.3 PHYSICS TESTS............................................ B 3/4 10-1 3/4.10.4 REACTOR COOLANT LOOPS.. ............ ..................... B 3/4 10-1 3/4.10.5 POSITION INDICATION SYSTEM - SHUT 00WN..................... B 3/4 10-1 COMANCHE PEAK - UNIT 1 XIX a- -", - , - ,, ,

^J

INDEX Js 1 BASES 3/4.11 RADI0 ACTIVE 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 SOLIO RADI0 ACTIVE WASTES................................ B 3/4 11-6 3/4.11.4 TOTAL D0SE.............................................. B 3/4 11-6 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PRrGRA;1...................................... B 3/4 12-1 3/4.12.2 LAND USE CENSUS......................................... B 3/4 12-1 3/4.M.3 INTERLABORATORY COMPARISON PR0 GRAM...................... B 3/4 12-2 I

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COMANCHE PEAK - UNIT 1 XX l

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.--,m..--.y-.... . . ., . , . _ . _ . - . . , -. -.- .-. , ,,. - -

y

INDEX e -'"

DESIGN FEATURES U S --

SECTION PAGE 5.1 SITE 5.1.1 EXCLUSION AREA......................r....................... 5-1 5.1.2 LOW POPULATION Z0NE......................................... 5-1 5.1.3 MAPS DEFINING UNRESTRICTED AREAS AND EXCLUSION AREA B0UNDARY FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS................ 5-1 FIGURE 5.1-1 EXCLUSION AREA....................................... 5-2 FIGURE 5.1-2 LOW POPULATION Z0NE.................................. 5-3 FIGURE 5.1-3 UNRESTRICTED AREA AND EXCLUSION AREA B0UNDARY FOR RADI0 ACTIVE GASE0US EFFLUENTS........................ 5-4 5.2 CONTAINMENT 5.2.1 CONFIGURATION..........................,,................... 5-1 5.2.2 DESIGN PRESSURE AND TEkPERATURE............................. 5-1 5.3 REACTOR CORE

- 5.3.1 FUEL ASSEMBLIES............................................. 5-5 5.3.2 CONTROL ROD ASSEMBLIES....................................... 5 .5 5.4 REACTOR COOLANT SYSTEM 5.4.1 DESIGN PRESSURE AND TEMPERATURE............................. 5-5 5.4.2 V0LUME...................................................... 5-5

5. 5 METEOROLOGICAL TOWER LOCATION................................. 5-5

5. 6 FUEL STORAGE 5.6.1 CRITICALITY................................................. 5-6 5.6.2 DRAINAGE.................................................. . 5-6 5.6.3 CAPACITY.................................................... 5-6 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT........................... 5-6 TABLE 5.7-1 COMPONENT CYCLIC OR TRANSIENT LIMITS.................. 5-7 l

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INDEX P E8" ADMINISTRATIVE CONTROLS SECTION PAGE 6.1 RESPONSIBILITY..............................................

6-1 6.2 ORGANIZATION................................................ 6-1 6.2.1 0FFSITE................................................... 6-1 6.2.2 UNIT STAFF................................................ 6-1 FIGURE 6.2-1 0FFSITE ORGANIZATION............................... 6-3 FIGURE 6.2-2 UNIT ORGANIZATION.................................. 6-4 TABLE 6.2-1 HINIMUM SHIFT CREW COMPOSITION...................... 6-5 6.2.3 INDEPENDENT SAFETY ENGINEERING GROUP Function.................................................. 6-6 .

Composition............................................... 6-6 Responsibilities.......................................... 6-6 Records...................................................

6-6 6.2.4. SHIFT TECHNICAL ADVIS0R................................... 6-6 6.3 UNIT STAFF QUALIFICATIONS................................... 6-6 6.4 TRAINING..................................'.................. 6-7 6.5 REVIEW AND AUDIT................................... ........ 6-7 6.5.1 STATIONS OPERATIONS REVIEW COMMITTEE (SORC)

Function.................................................. 6-7 Composition............................................... 6-7 Alternates................................................ 6-7 Meeting Frequency......................................... 6-8 Quorum.................................................... 6-8 Responsibilities.......................................... 6-8 Records................................................... 6-9 ,

COMANCHE PEAK - UNIT 1 XXII

_ m

.e INDEX .I y1 ou ADMINISTRATIVE CONTROLS SECTION 6.5.2. OPERATIONS REVIEW COMMITTEE (ORC) ,

Fun'ction.................................................. 6-10 Composition............................................... 6-10 Alternates................................................ 6-10 Consultants............................................... 6-11 Meeting Frequency......................................... 6-11 Quorum.................................................... .6-11 Review.................................................... 6-11

~

Audits.................................................... 6-12 Records................................................... 6-13 6.5.3 TECHNICAL REVIEW AND CONTR0LS.................,........... 6-13 6.6 REPORTABLE EVENT ACTI0N..................................... 6-14 6.7 SAFETY LIMIT VIOLATION...................................... .

6.'15 6.8 PROCEDURES AND PR0 GRAMS..................................... .

6-15 6.9 REPORTING REQUIREMENTS...................................... 6-18 6.9.1 ROUTINE REP 0RTS........................................... 6-18 l Startup Report............................................ 6-18 l Annual Reports............................................ 6-18 l Annual Radiological Environmental Operating Report........ 6-19 Semiannual Radioactive Effluent Release Report............ 6-20 Monthly Operating Report.................................. 6-22 Radial Peaking Factor Limit Report........................ 6-22 l 6.9.2 SPECIAL REP 0RTS........................................... 6-23

6.10 RECORD RETENTION........................................... 6-23 l

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l COMANCHE PEAK - UNIT 1 XXIII 1

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A INDEX hh 3 pIMjJ 'J ADMINISTRATIVE CONTROLS SECTION 6.11 RADIATION PROTECTION PR0 GRAM............................... 6-24 6.12 HIGH' RADIATION AREA........................................ 6-25 6.13 PROCESS CONTROL- PROGRAM (PCP). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-26 6.14 0FFSITE DOSE CALCULATION MANUAL (0DCM)..................... 6-26 6.15 MAJOR CHANGES TO LIQUID, GASEOUS, AND SOLID RADWASTE TREATMENT SYSTEMS................................. 6-27 4

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COMANCHE PEAK - UNIT 1 XXIV

4 e

SECTION 1.0 DEFINITIONS i

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O COMANCHE PEAK - UNIT 1 1-0

4 hfab)'

E t. h d 1.0 DEFINITIONS The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications.

ACTION -

i.1 ACTION shall be that part of a Technical Specification which prescribes remedial measures required under designated conditions.

ACTUATION LOGIC TEST 1.2 An ACTUATION LOGIC TEST shall be the application of various s'imulated input combinations in conjunction with each possible interlock logic state and verification of the required logic output. The ACTUATION LOGIC TEST shall include a continuity check, as a minimum, of output devices. -

ANALOG CHANNEL OPERATIONAL TEST 1.3 An ANALOG CHANNEL.0PERATIONAL TEST shall be the. injection of a simulated signal into the channel as close to the sensor as practicable to' verify OPERABILITY of alarm, interlock and/or trip functions. The ANALOG CHANNEL OPERATIONAL TEST shall include adjustments, as necessary, of the alarm, inter-lock and/or Trip Setpoints such that the setpoints are within the required range and accuracy.

AXIAL FLUX DIFFERENCE f 1. 4' AXIAL FLUX DIFFERENCE;shall be the difference in normalized f' lux signals between the top and bottom halves o'f a four section excore neutron detector.

CHANNEL Ct.LIBRATION 1.5 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel such that it responds within the required range and accuracy to known values of input. The CHANNEL CALIBRATION shall encompass the entire channel i;icluding the sensors and alarm, interlock and/or trip functions and may be performed by any series of sequential, overlapping, or total channel steps such that the entire channel is calibrated.

CHANNEL CHECK .

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

9 C0HANCHE PEAK - UNIT 1 1-1

. n h QP DEFINITIONS W a u- .

CONTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when:

a. All penetrations required to be closed during accident conditions are eithe.r: .

1) Capable of being closed by an OPERABLE containment automatic isolation valve system, or

2) Closed by manual valves, blind flanges, or deactivated automatic valves secured'in their closed positions, except as provided in Table 3.6-1 of Specification 3.6.3.

b. All. equipment hatches are closed and sealed,

c. Each air lock is in compliance with the requirements of Specification 3.6.1.3, d.. The containment leakage rates are within the limits ofcSpecification 3.6.1.2, and '

e. 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 supplied to the reactor coolant pump seals.

CORE ALTERATIONS

1. 9 CORE ALTERATIONS 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 ALTERATIONS shall not preclude completion of t

movement of a component to a safe conservative position.

DIGITAL CHANNEL OPERATIONAL TEST 1.10 A DIGITAL CHANNEL OPERATIONAL TEST shall consist of exercising the digital computer hardware using data base manipulation and injecting simulated '

process data to verify OPERABILITY of alarm and/or trip functions.

DOSE EQUIVALENT I-131 ,

1.11 DOSE EQUIVALEAT I-131 shall be that concentration of I-131 (microcurie / gram) which alone would' produce the same thyroid dose as the quantity and' isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, "Calculation of Distance Factors for Power and Test .

Reactor Sites" or Table E-7 of NRC Regulatory Guide 1.109, Revision 1, October 1977.

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COMANCHE PEAK - UNIT 1 1-2 l

t

h Kip DEFINITIONS Mbli )

E - AVERAGE DISINTEGRATION ENERGY 1.12 E shall be the average (weighted in~ proportion to the concentration of each radionuclide in the sample) of the sum of the average beta and gamma energies per disintegration (MeV/d) for the radionuclides in the sample.

ENGINEERED SAFETY FEATURES RESPONSE TIME 1.13 The ENGINEERED SAFETY FEATURES (ESF) RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF Actuation Setpoint a; 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.

EXCLUSION AREA BOUNDARY 1.14 The Exclusion Area Boundary, used for establishing effluent release limits is shown in Figure 5.1-1.

FREQUENCY NOTATION 1.15 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.

IDENTIFIED LEAKAGE 1.16 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

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 l LEAKAGE, or

c. Reactor Coolant System leakage through a steam generator to the Secondary Coolant System.

MASTER RELAY TEST 1.17 A MASTER RELAY TEST shall be the energization of each master relay and

, verification of OPERABILITY of each relay. The MASTER RELAY TEST shall* include a continuity check of each associated slave relay.

l COMANCHE PEAK - UNIT 1 1-3 L

DEFINITIONS # #.

, MEMBER (S) 0F THE PUBLIC 1.115 MEMBER (S) 0F THE PUBLIC shall include all persons who are not occupa-tionally associated with the plant. This category does not include employees of the licensee, its contractors, or vendors. Also excluded from this category are persons who enter the site to service equipmen't or to make deliveries.

This category does include persons who use portions of ,the site for recre-ational, occupational, or other purposes not associated with the plant.

OFFSITE DOSE CALCULATION MANUAL 1.19 The OFFSITE DOSE CALCULATION MANUAL (0DCM) shall contain the 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 Environ-mental Radiological Monitoring Program. j OPERABLE - OPERABILITY

' 1.20 A system, subsystem, train,-component or' device sh'all be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s),

and when all necessary attendarit instrumentation, controls, electrical power, 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 1.21 An OPERATIONAs MODE (i.e., MODE) shall correspond to any one inclusive combination of core reactivity condition, power level, and average reactor coolant temperature specified in Table 1.2.

PHYSICS TESTS 1.22 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation:

(1) described in Chapter 14.0 of the FSAL', (2) authorized under the provisions of 10 CFR 50.59, or (3) otherwise approved by the Commission.

PRESSURE BOUNDARY LEAKAGE 1.23 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube leakage) through a nonisolable fault in a Reactor Coolant System component body, pipe wall, or vessel wall.

e COMANCHE PEAK - UNIT 1 1-4

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,J . u DEFINITIONS PRIMARY PLANT VENTILATION SYSTEM ,

1.24 A PRIMARY PLANT VENTILATION SYSTEM shall be any system designed and installed to reduce gaseous radioiodine or 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 td the. release to the environment. Such a system is not considered to have any effect on noble gas effluents. Engineered Safety. Features Atmospheric Cleanup Systems are not considered to be PRIMARY PLANT VENTILATION SYSTEM components.

PROCESS CONTROL PROGRAM 1.25 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, tests, and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71 and Federal ana State regulations, burial ground requi,rements, and other require-ments governing the disposal of radioactive waste.

PURGE - PURGING 1.26 PURGE or PURGING shall be.any controlled process of discharging air or gas from a confinement to maintain temperature, press'ure, humiditp, concentration or other operating condition, in such a manner that replacement air or gas is required to pu'rify'the confinement.

QUADRANT POWER TILT RATIO 1.27 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper half excore detector calibrated output to the average of the upper half excore detector calibrated outputs, or the ratio of the maximum lower half excore detector calibrated output to the average of the lower half 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 1.28 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3411 MWt.

REACTOR TRIP SYSTEM RESPONSE TIME 1.29 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when.the monitored parameter exceeds its . Trip Setpoint at the channel sensor until loss of stationary gripper coil voltage.

' REPORTABLE EVENT ,

'1.30 A REPORTABLE EVENT shall be any of those conditions specified in 10 CFR 50.73.

COMANCHE PEAK - UNIT 1 1-5

su..a DEFINITIONS SHUTOOWN MARGIN 1.31 SHUTOOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming all rod cluster assemblies (shutdown and, control) are fully inserted except for the single rod cluster assembly'of highest reactivity worth which is a.ssumed to be fully withdrawn. ,

SITE BOUNDARY 1.32 The SITE BOUNDARY shall be that line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee.

SLAVE RELAY TEST 1.33 A SLAVE RELAY TEST shall be the energization of each slave relay and verification of OPERABILITY of each relay. The SLAVE RELAY TEST shall include a continuity check, as a minimum, of associated testable actuation devices.

SOLIDIFICATION 1.34 SOLIDIFICATICN shall be the conversion of wet wastes into a form that meets shipping and burial ground requirements.

SOURCE CHECK l'.35 A SOURCE CHECK shall be the qualitative assessment of channel response -

when the channel ~ sensor is exposed to a source of increased radioact'ivity.

STAGGERED TEST BASIS 1.36 'A STAGGERED TEST BASIS shall consist of:

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

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

THERMAL POWER 1.37 THERMAL POWER shall be the total core heat transfer rate to the reactor coolant.

TR?P ACTUATING DEVICE OPERATIONAL. TEST 1.38 A TRIP ACTUATING DEVICE OPERATIONAL TEST shall. consist of operating the Trip Actuating Device and verifying OPERABILITY of alarm, interlock and/or trip functions. The TRIP ACTUATING DEVICE OPERATIONAL TEST shall include adjustment, as necessary, of the Trip Actuating Device such that it actuates at the required setpoint within the required accuracy.

COMANCHE PEAK - UNIT 1 1-6

T DEFINITIONS UNIDENTIFIED LEAKAGE 1.39 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE or CONTROLLED LEAKAGE.

UNRESTRICTED AR'EA .

1.40 An UNRESTRICTED AREA shall be any area at or beyond the EXCLUSION AREA BOUNDARY for purposes of protection of individuals from exposure to radiation and radioactive materials, or any area within the EXCLUSION AREA B0UNDARY'used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes.

VENTING 1.41 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 pro-vided or required during VENTING. Vent, used in system names, does not imply a VENTING process.

WASTE GAS HOLOUP SYSTEM

' 1.42 A WASTE GAS HOLDUP SYSTEM shall be any system designed and installed to reduce radioactive gaseous effluents by collecting. Reactor Coola'nt System offgases from the Reactor Coolant System-and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

i 4

i L

F COMANCHE PEAK - UNIT 1 1-7

TABLE 1.'1' FREQUENCY NOTATION NOTATION FREQUENCY S

At.least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . .

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

W At least once per 7 days.

M At least once per 31 days.

Q At least once per 92 days.

SA At least once per 184 days.

R At least once per 18 months.

S/U Prior to each reactor startup.

N.A. Not applicable.

P Completed prior to each release.

O COMANCHE PEAK - UNIT 1 1-8 )

TABLE 1.2 3 p.

OPERATIONAL MODES ld1[

. REACTIVITY % RATED AVERAGE COOLANT MODE CONDITION, K,7f THERMAL POWER

_ TEMPERATURE

1. POWER OPERATION > 0.99 ,

> 5% > 350*F .

2. STARTUP > 0.99 5 5% > 350 F

3. HOT STANDBY < 0.99 0 > 350*F

4. HOT SHUTDOWN < 0.99 0 350 F > T

> 200 F avg

5. COLD SHUTDOWN < 0.99 0 5 200 F

~

6. REFUELING ** 5 0.95 0 5 140 F

Excluding decay heat.

- Fuel:in the reactor vessel w'ith the vessel head closure bolts less than fully tensioned or with the head removed. ,

O COMANCHE PEAK - UNIT 1 1-9

w- --

1 l

DRAFT I

SECTION 2.0 SAFETY LIMITS AND .

LIMITING SAFETY SYSTEM SETTINGS 4 .

n G

COMANCHE PEAK - UNIT 1 2-0

[.,

_2. 0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORE s 2.1.1 ,The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temperature (T,yg) shall not exceed the limits shown in Figures 2.1-1. for four loop operation.

/

APPLICABILITY: MUDES 1 and 2.

ACTION:

Whenever the point defined by the combination of the highest operating loop average temperature and THERMAL POWER has exceeded the appropriate pressurizer pressure line, be in HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and comply with the require-ments of Saccification 6.7.1.

REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig.

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

ACTI_O_N: ,

MODES 1 and 2:

Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ., and comply with the requirements of Specification 6.7.1.

s t MODES 3, 4 and 5:

Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes, and comply with the requirements of Specification 6.7.1.

4 e

COMANCHE PEAK - UNIT 1 2-1 e

r --

I k FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION COi4ANCHE PEAK - UNIT 1 g.g

' Einc!'

!!s t) {

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFETY SYSTEM C"TTINGS REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The Reactor Trip System Instrumentation an.d Interlock Setpoints shall

- be set consistent with the Trip Sitpoint values shown in-Table 2.2-1.

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

A,CTION:

a. With a Reactor Trip System Instrumentation o: InterlocP Setpoint less conservative than the value shown in the Trip Setpoint column but more conservative than the value shown in the Allowable Value column of Table 2.2-1, adjust the setpoint consistent with the Trip setpoint value.

b. With the Reactor Trip System Instrumentati)n or . Interlock Setpoint lest. conservative than the value shown in the Allowable Values column of Table 2.2-1, either:

1. Adjust the setpoint consister.t with the Trip Setpoint value of Table 2.2-1 and determine within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that Equation 2.2 I was satisfied for the affected channel, or

2. Declare the channel inoperable and apply the spplicable ACTION -

statement requirement of Specification 3.3.1 until the channel is restored to OPERABLE status with its setroint adjusted .

consistent with the -Trip Setpoir.t value.

Equation 2.2-1 Z + R + 5 5 TA Where:

Z = The value from Column Z of Table 2.2-1 for the affected channel, R = The "as mecsured" value (in percent span) of rack error for the ,

affected channel, S = Either the "a* reasured" value (in percent span) of the sensor error, or the value from Column S (Sensor T>ror) of Tcble 2.2-1 for the affected channel, and TA = The value from Column TA (Total Allowance) cf Table 2.2-1 for the affected channel.

e C0KANCHE PEAK - UNIT 1 2-3

- - . ~ .

TABLE 2.2-1 n

o REACTOR TRIP SYSTEM INSTRUMu(TATION TRIP SiTPOINTS 3

z g SENSOR m TOTAL ERROR g FUNCTIONAL UNIT ALLOWANCE (?A) Z (S) TRIP SETPOINI ALLOWABLE VALUE-

1. Manual Reactor Trip N.A. N.A. N.A. N.A. . N.A. .

g 2. Power Range, Neutron Flux Z

g a. High Setpoint 7.5 4.56 0 1109% of RTP* 1111.2%'of RTP*

b. Low Setpoint 8.3 4.56 0 125% of RTP* 127.2% of RTP*

3. Power Range, Neutron Flux, 1. 6 0.5 0 <5% of RTP* with <6.3% of RTP* with High Positive Rate i time constant i time constant 12 seconds 12 seconds ,

4. Power Range, Neutron Flux, 1. 6 0". 5 0 <5% of RTP* with <6.3% of RTP* with y High Negative Rate .i time constant i time constant 12 seconds 12 seconds

5. Intermediate Range, 17.0 8.4 0 ' $25% of RTP* $31% of RTP*

Neutcon Flux

6. Source Range, Neutron Flux 17.0 10 0 $105 cps $1.4 x 105 cps

7. Overtemperature N-16 6.4 4.71 1. 8 See Note 1 Se.e Note 2

8. Overpower N-16 4.0 1.91 1. 3 1112% 1114.5%

9. Pressuri7er Pressure-Low 8.8 2.81 1. 5 11910 psig 11896 psig

10. Pressurizer'Presture-High 7.5 4.96 0.5 $2385 psig $2399 psig E" 3 s

?TP = RATED THERMAL ?0WER Pt3

iw e

3w - . - - . w -

o

TABLE 2.2-1-(Continued)

o REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 3

z g SENSOR ni TOTAL ERROR g FUNCTIDNAL UNIT ALLOWANCE (TA) Z (5) TRIP SETPOINT ALLOWABLE VALUE 7 11. Pressur~.zer Water i.evei .uigh 5.0 '2.18 1. 5 $92% of instrument

' $93.8% of instrument span span C

h 12. Reactor Coolant flow-Low 2. 5 1.31 0.6 >90% ef loop >88.8% of loop y Besign flow ** - design flow **

13. Steam Generctor Water 8.8 7.08 1. 5 .>43.4% of narrow >42.1% of narrow Leve: Low-Low range instrument range instrument span span

14. Undervoltage - Reactor 7.7 0 0 )4781~ volts-Coolant. Pumps 14830 volt's- ,

each bus' each bus -

y 15. Underfrequene - Raactor 4.4 - 0 0 >57.2 Hz '

>57.1 Hz m Coolant Pompe

16. Turbine M ,

. a. Low Trip System Pressure N.A. N.A. N.A. >45 psig >43 psig

b. Tu' hine Stop Va?ve N.A. N.A. N.A. >1% open >1% open C1ris;re

17. ' Safety Injection Input N.A. N.A. N.A. N.A. N.A.

free ESF

- Loop desiga tsow = 95,700 gpm.

C"3

~ PC

'c 3

..cn c:q y g .+

TT.BLE 2.2-1 (Ccntinued)

REACTOR TRIP 3YSTEM INSTRUMENTATION TRID SETPOINTS .

z o

SENSOR 55 TOTAL ERROR v FUNCTIOfiAL UNIT m ALLOWANCE (TA) { (S) TRIP SETPOINT ALLOWABLE VALUE

$E 18. Reactor Trip System

. Interlocks C

a. Intermediate Range N.A. N.A. N.A. ->l x 10 10 amps-Neutron Flux, P-6 ->6 x 10 11 amps

]

b. Low Power Reactor Trips #

Block, P-7

1) P-10 input N.A. N. A. N.A. <10% of RTP* <12.2% of RTP*

2) P-13 ir.put N.A. N.A. N.A. <10% RTP* Turbine <12.2% RTP* Turbine

.First Stage Pres- First Stage Pressure y sure Equivalent Equivalent cn

c. Power Range Neutron N.A. N.A. N.A.

Flux, P-8 -<48% of RTP* <50.2% of RTP*

d. Power Range Neutron N.A. N.A. N. A. ~>10% of RTP*

Flux, P-10 - <7.8% of RTP*

e. Turbine IOpalse Chamber N.A. N.A. N.A <10% RTP* Turbine Pressure, P-13 <12.2% RTP* Turbine First Stage First Stage Power Pressure Equivalent

, Equivalent

19. Reactor Trip Breakers ,

N.A. .N.A. N.A N.A. N.A.

20. Automatic Trip and Interlock N.A. 'N.A. N.A. N.A. N.A.

Logic C

' N l *RTP = RATED THERMAL POWER

==

]

9 -_ - _

TABLE 2.2-1 (Continued) n -

l TABLE NOTATIONS M

g -

, NOTE 1: Overtemperature N-16

' m 3

3E Ki-K2

' N = [** Tc -Tc]+K3 (P-P ) - f i (aq) 1+12S C

(

Where: N = Measured N-16 Power by ion chambers, T

c

= Cold leg temperatuce, 'F, ,

T* =

559.6*F, Reference T ;at RATED '"* M L POWER, C C Kt = 1.069 .

K2 = 0.00948/ F, y f =

The function generated by the lead-lag compensator for rieasured T ,

u c ti,12 =

Time constants utilized in thc lead-lag co.:v osator for Tc , t's = 10 s, and T2=3S.

K3 = 0.000494/psig, b

e 9

t3

??:n '

ri 9

. +

-w-e-. tr s gm r-- + m ~tY -

ev m s- n -

TABLE 2.2-1 (Continued) .

[$ TABLE NOTATIONS (Continued) m NOTE 1: (Continued) 2 P = Pressurizer pressure, psig,

[' P2 =

2235 psig (Nominal RCS operating pressure),

C g5 S = laplace transform operator, s-1,

-4

>=

and f (aq) is a function of the indicated difference between top and bottom.halvec of i -

detectors of the power-range neutron ion chambers; with gains to be salected based on-

measured instrument response during plant STARTUP tests such that: -

(i) for q t

~

Ab between -3S% aad +10%,3 f (Aq) = 0,, whereg q and q are pertent RATED THERMAL POWER in the t'op and bottom halves of the core respective 5y, and qt*9b is otal THERMAL POWER in percent of RA1ED THERMAL POWER, =

y (ii) for each percent that the magnitude of q g qgexceeds -35%, the N-16 Trip Setpoint shall be automatically reduced by 1.2S% of its'value at RATED THERMAL- ,

POWER, and j (iii) for each percer,t tha't the magnitude of q gbexceeds +10%, the N 13 Trip Setpoint shall be automatically reduced by 1.5S% of its value at RATED THERMAL POWER.

NOTE 2: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 1.4%

of span. .

23 ,

- ~

%"J * :

s.azl k

n 7 "YN-- *1 T T 'M T $ T h ' n %' 'T N' C 7 T- M

gg . . .. ..

' G, ,

4)f

1 m .

,.; .;n v n ' '

o 4-BASES FOR SECTION 2.0 4

SAFETY LIMITS AND

' . LIMITING SA/ETY SYSTEM SETTINGS NOTE The BASES contained in succeeding pages summari;.o '

the reasons for the Specifications in Section 2.0, but in accordance with 10 CFR 50.36 are not part ,

of these Technica1' Specifications.

8 6 F

l l

I l- ?

l I c 1

I h

I COMANCHE PEAK - UNIT 2. B 2-0

2.1 SAFETY LIMITS BA$ tis 2.1.1 REACTOR CORE . .

The restrictions of this Safety Limit prevent overheating of the fuel a'nd possible cladding perforation which would result in the release of fission products to the reactor coolant', Overheating of the fuel cladding is pre-vented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficied is large and the cladding surface tempera-ture is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures becaure of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and reactor coolant temparature and pressure have been

. related to DNB through the W-3 correlation. The~W-3 DNB correlation has been developed to pre' dict the DNB flux and the location of DN8 for axially uniform and nonuniform heat flux distributions. The local DN8 heat flux ratio (DNBR) -

is defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux and is indicative of the margin to ON8.

The minimum value of the DNBR during steady-state operation, normal

, operational transients, and anticipated transients is limited to 1.30.. This value corresponds to a 95% probability at a 95% confidence level that DNB will not occur and is chosen as an appropriate margin,to DNB for all operating c6nditions.

The curves of Figure 2.1-1 show tt.e loci of points of THERMAL POWER, -

Rea: tor Coolant System pressure and average temperature for which the minimum DNBR is no less than 1.30, or the average enthalpy at the ves: 1 exit is equal to the enthalpy of saturated liquid.

N These curves are based on an enthalpy hot :hannel ctor, F f ". 55 and a referencc cosine with a peak of 1.55 for axial power shape. A$H,llosanceisa included for an increase in F H at reduced power based on the expression:

F g - 1.5- [1+ 0.2 (1-P))

Where P is the fraction of RATED THERMAL POWER.

These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance is within tne limts of the

f t (AI) function of the 0vertemperature N-16 trip. When the axial power imbalance is not within the tolerance, the axial power imbalance effect on the Overtemperature N-16 trips will reduce the Setpoints to provide protection consistent with' core Safety Limits.

(

COMANCHE PEAK - UNIT 1 B 2-1

. SAFETY LIMITS 7;.y>

1 BASES - - - -

l L

?.1.2

-- REACTO.R COOLANT SYSTEM P' 4ESSURE

'The restriction of this Safety Limit protects the integrity of the Reactor Cr,olant System (RCS)~from overpressurization and thereby prevents the release

of radionuclides contained in the reactor coolant from reaching the containment ,

atmosphere. .

The reactor vessel, pressurizer, and the RCS piping, valves and fittings are designed to Section III of the ASME Code for Nuclear Power Plaats which permits a maximum transient pressure of 110% 2735 psig of design pressure.

The Safety Limit of 2735 psig is therefore consistent with the design criteria and associated Code requirements.

The.entive RCS is hydrotested at 125% (3110 psig) of design pressure, to

-demonstrate integrity prior to initial operation.

i e

4 e

4 COMANCHE PEAK - UNIT 1 B 2-2

22 LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS

'The Reactor Trip Setpoint Limits specified in TabTe 2.2-1 are tLe nominal values at which the Reactor trips are set for each functional unit. The Trip '

Setpoints have been selected to ensure that the core and Reactor Coolant System are provented from exceeding their safety limits during normal operation and design basis anticipated operational occurrences and to assist the Engi-neered Safety Features Actuation System in mitigating the consequences of accidents. The setpoint for a Reactor Trip System or interlock function is considered to be adjusted consistent with the nominal value when the "as measured" setpoint is within the band allow M for calibration accuracy and instrument drift.

To accommodate the instrument drift assumed to occur between operational

. tests and the accuracy to which setpoints can be measured and calibrated, Allowable Values for the Reactor Trip Setpoints ha e'been specified in Table 2.2-1. Operation with setpoints less conse sative than the Trip Set-point but within the Allowable Value is acceptatsa since an allowance has been

. made in the safety analysis to accommodate this error. An optional provision has been included for determining the OPERABILITY of a channel when.its Trip Setpoint is found to exceed the Allowable Value. The methodology of this option utilizes the "as measured" deviation from the specified calibration point for rack and sensor components in conjunction with a statistical combin '

Ltion of the other uncertainties of the instrumentation to measure the process variable and the' uncertainties in calibrating the instrumentation. In Equa-tion 2.2-1, 2 + R + S < TA, the interactive effects of the errors in the rack and the sensor, and the "as measured" values of the errors are considered. Z, as specified in Table 2.1-1, in percent span, is the statistical summation of errors assumed in the analysis excluding those associated with the sensor and rack drift and the accuracy of their measurement. TA or Total Allowance is the difference, in percent span, between the Trip Setpoint and the value used in the analysis for Reactor trip. R or Rack Error is the "as measured" devia-tion, in percent span, for the affected channel from the specified Trip Set- -

point. S or Sensor Error is either the "as measured" deviation of the sensor from its calibration r11nt or the value specified in Table 2.2-1, in percent span, from the analysts assumptions. Use of Equation 2.2-1 allows for a sensor drift factor, an increased rack drift factor, and provides a threshold value *or REPORTABLE EVENTS.

The methodclogy to derive the Trip Setpoints is based upon combining all of the uncertainties in the channels. Inherent to the determination of the Trip Setpuints are the magnitudes of these channel uncertainties. Sensors and -

other instrumentation utilized in these channels are expected to be capable of operating within the allowances of these uncertainty magnitudes. Rack drift in excess of the Allowable Value exhibits the behavior that the rack has not met its allowance. Being that,there is a small statistical chance that this will happen, an infrequent excessive drift is expected. Rack or sensor drift, in excess of the allowance,that is more than occasional, may be indicative of ;

more serious problems and should warrant further investigation.

COMANCHE PEAK - UNIT 1 B 2-0

, i LIMITING SAFETY SYSTEM SETTINGS BASES REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS (Continued)

The various Reactor trip circuits automatically open the Reactor trip breakers whenever a condition monitored by the Reactor Trip System reache's a preset or calculated level. In addition to redundant channels and trains, the design approach provides a Reactor Trip System which monitors numerous system variables, therefore providing Trip System functional-diversity. The functional capability at the specified trip setting is required for those anticipatory or diverbe Reactor trips for which no direct credit was assumed in the safety analysis to enhance the overall reliability of the Reactor Trip System. The Reactor Trip System initiates a Turbine trip signal whenever Reactor trip is initiated. This prevents the reactivity insertion that would otherwise result from excessive Reactor Coolant System cooldown and thus avoids unnecessary actuation of the Engineered Safety Features Actuation System.

Manual Reactor Trip The Reactor Trip System includes manual Reactor trip capability. .

Power Range, Neutron Flux In each o.f the Power Range Neutron Flux channels there are two independent bistables, each with its own trip setting used for a High and Low Range trip-setting. The Low Setpoint trip provides protection during suberitical and low

, power operations to mitigate the consequences of a power excursion beginning from low power, and the High Setpoint trip provides protection during power operations to mitigate the consequences of a reactivity excursion from all power levels.

The Low Setpoint trip may be manually blocked above P-10 (a power level of approxiinately 10% of RATED THERMAL POWER) and is automatically reinstated below the P-10 Setpoint.

Power Range, Neutron Flux, High Rates The Power Range Positive Rate trip provides protection against rapid flux increr.ses which are characteristic of a rupture of a control rod-drive housing.

Specifically, this trip complements the Power Range Neutron Flux High and Low trips to ensure that the criteria are met for rod ejection from mid power.

The Power Range Negative Rate trip provides protection for control rod drop accidents. At high power a single or multiple rod drop accident could cause local flux peaking which could cause an ur. conservative local DNBR to exist, The .

Power Range Negative Rate trip will prevent this from occurrirg by tripping the rebetor. No credit is taken for operation of the Power Range Negative Rate trip for those control rod drop accidents for which DNBRs will be greater than 1.30.

COMANCHE PEAK - UNIT 1 B 2-4

LIMITING SAFETY SYSTEM SETTINGS BASES Intermediate and Source Range, Neutron Flux

The Intermediat.e and Source Range, Neutron Flux trips provide core protection during reactor startup to mitigate the consequences of an uncon-trolled rod cluster control assembly bank withdrawal from a suberitical condition. These trips provide redundant protection to the Low Setpoint trip of the Power Range, Neutron Flux channels. In addition, the 5ource Range Neutron Flux trip provides similar protection during shutdown operations with the reactor trip breakers closed and the rod control system capable of control rod withdrawal. The Source Range channels will initiate a Reactor trip at about 105 counts per second unless manually blocked when P-6 becomes active. The Intermediate Range channele. will initiate a Reactor trip at a current level equivalent to approximately 25% of RATED THERMAL POWER unless manually blocked when P-10 becomes active.

Overtemperature N-16 The Overtemperature N-16 trip provides core protection to prevent DNB for~

all combinations of pressure, power, coolant temperature, and axial power dis-tribution, provided that the transient is slow with respect to piping transit delays from the core.to.the N-16 detectors, and pressure is within the range between the Pressurizer High and Low Pressure trips. The setpoint is auto-matica11y' varied with': (1) coolant temperature to correct for temperature induced changes in density and heat capacity of water and includes dynamic compensation for piping delays from the core to the cold leg temperature detectors, (2) pressurizer pressure, and (3) axial power distribution. With normal axial power distribution, this Reactor trip limit is always below the core Safety Limit as shown in Figure 2.1-1. If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the Reactor trip is automatically reduced according to the notations in Table 2.2-1.

' Overpower N-16 The Overpower N-16 trip provides assurance of fuel integrity (e.g., no fuel pellet melting and less than 1% cladding strain) under all possible over-power conditions, limits the required range for Overtemperature trip, and pro-vides a backup to the High Neutron Flux trip. The Overpower N-16 trip provides protection to mitigate the consequences of various size steam breaks as reported in WCAP-9226, "Reactor Core Response to Excessive Secondary Steam Releases."

fressurizer Pressure In each of the pressurizer pressure channels, there are two independent bistables, each with its own trip setting to provid.e for a High and Low Pres-sure trip thus limiting the pressure capge in which reactor operation is permitted. The Low Setpoint trip protects against low pressure which could lead to DNB by tripping the reactor in the event of a loss of reactor coolant pressure.

COMANCHE PEAK - UNIT 1 B 2-5

= .

LIMITING SAFETY SYSTEM SETTINGS BASES Pressurizer Pressure (Continued)

On decreasing power the Low Setpoiit trip is automatically blocked by P-7 (a power level of approximately 10% of RATED THERMAL POWER with turbine first stage chamber pressure at approximately 10% of full power equivalent); and on in,reasing power, automatically reinstated by P-7.

The High Setpoint trip functions in conjunction with the pressurizer relief and safety valves to protect the Reactor Coolant System against system overpressure.

Pressurizer Water Level The Pressurizer Water Level-High trip is provided to prevent water relief through the pressurizer safety valves. On decreasing power the Pres-surizer High Water Level trip is automatically blocked by P-7 (a power level of approximately 10% of RATED THERMAL POWER with a turbine first stage chamber pressure at approximately 10% of full power equivalent); and'on increasing power, automatically reinstated by P-7.

Reactor Coolant Flow The Reactor Coolant. Flow-Low trip provides core protec' tion to prevent DNB by mitigating the consequences of a loss of flow resulting from the loss of one or more reactor coolant pumps.

On increasing power above P-7 (a power level of approximately 10% of RATED THERMAL POWER or a turbine first stage chamber pressure at approximately 10% of full power equivalent), en automatic Reactor trip will occur if the flow in more than one loop drops below 90% of nominal full loop flow. Above l P-8 (a power level of approximately 48% of RATED THERMAL POWER) an automatic l Reactor trip will occur if the flow in any single loop drops below 90% of nominal full loop flow. Conversely, on decreasing power between P-8 and the P-7 ai) automatic Reactor trip will occur on low reactor coolant flow in more than one loop and below P-7 the trip function is automatically blocked.

I i Steam Generator Water Level l

i The Steam Generator Water Level Low-Low trio protects the reactor from l loss of heat sink in the event of a sustained steam /feedwater flow mismatch resulting from loss of normal feedwater. The specified setpoint provides '

allowances for starting delays of the Auxiliary Feedwater System.

Steam /Feedwater Flow Mismatch and Low Steam Generator Water Level l .

The Steam /Feedwater Flow Mismatch in coincidence with a Steam Generator -

I Water Level-Low-Low trip is not used in the transient and accident analyses but is included in Table 2.2-1 to ensure the functional capability of the l

l COMANCHE PEAK ' UNIT 1 B 2-6 9

e +

LIMITING SAFETY SYSTEM SETTINGS Wi BASES Steam /Feedwater Flow Mismatch and Low Steam Generator Water Level (Continued) specified trip settings and thereby enhance the overall reliability of the Reactor Trip System. This trip is redundant to the Steam Generator Water Level Low-Low trip. The Steam /Feedwater Flow Mismatch portion of this trip is activated when the steam flow exceeds the feedwater flow by greater than or equal to [1.42 x 106] lbs/ hour. The Steam Generator Water Level-Low-Low portion of the trip is activated when the water level drops below [25]%, as indicated by the narrow range inttrument. These trip values include sufficient allowance in excess of normal operating values to preclude spurious trips but will initiate a Reactor trip before the steam generators are dry. Therefore, the required capacity and starting time requirements of the auxiliary feedwater pumps are reduced and the resulting thermal transient on the Reactor Coolant System and steam generators is minimized.

Undervoltage and Underfrequency - Reactor Coolant Pump Busses .

The Undervoltage and Underfrequency Reactor Coolant Pump Bus trips pro-vide core protection against DNB as a result of complete loss of forced coolant flow. The specified setpoints assure a Reactor trip signal is generated before -t e Low Flow Trip Setpoint is reached. -Time delays are incorporated in the Underfreque. hey and Undervoltage trips to prevent spurious Reactor trips from momentary electrical power transients. For undervoltage, the delay'is set so that the time required for a signal to reach the Reactor trip breakers following the simultaneous trip of two or more reactor coolant pump bus circuit breakers shall not exceed 1.2 seconds. For underfrequency, the delay is set so that the time required for a signal to reach the Reactor trip breakers after the Underfrequency Trip Setpoint is reached shall not exceed 0.3 second.

On decreasing power the Undervoltage and Un'derfrequency Reactor Coolant Pump Bus trips are automatically blocked by P-7 (a power level of approximately 10%

of RATED THERMAL POWER with a turbine first stage chamber pressure at approximately 10% of full power equivalent); and on increasing power, reinstated automatically by P-7.

Turbine Trip A Turbine trip initiates a Reactor trip. On decreasing power the Reactor trip from the Turbine trip is automatically blocked by P-7 (a power level of approximately 10% of RATED THERMAL POWER); and on increasing power, reinstated automatically by P-7.

Safety Injection Input from ESF If a Reactor trip has not already been generated by the Reactor Trip System instrumentation, the ESF automatic actuation logic channels will initiate ,

a Reactor. trip upon any signal which initiates a Safety Injection. The ESF ,

instrumentation channels which initiate a Safety Injection si Cnal are shown in Table 3.3-3.

COMANCHE PEAK - UNIT 1 8 2-7

LIMITING SAFETY SYSTEM SETTINGS BASES Undervoltage and 'Jnderfrequency - Reactor Coolant Pump Busses (Continued)

Reactoi Trip System Ihterlocks The Reactor Trip System interlocks perform the following functions:

P-6 On increasing power P-6 allows the manual block of the Source Range trip (i.e., prevents premature block of Source Range trip), provides a backup block for Source Range Neutron Flux doubling, and deener-gizes the high voltage to the detectors. On decreasing power, Source Range Level trips are automatically reactivated and high voltage restored.

P-7 On increasing power P-7 automatically enables Reactor trips on low ,

flow in more than one reactor coolant loop, reactor coolant pump bus undervoltage an.d u'nderfrequency, Turbine trip, pressurizer low pressure and pressurizer high level. On decreasing power, the above ,

listed trips are automatically blocked.

P-8 On increasing power, P-8 automatically enables'the Reactor trips on

, low flow in one reactor coolant loop. On decreasing power, the P 3 -

automatically blocks the reactor trip on low flow in one reactor

. coolant loop. .

P-10 On increasing power, P-10 allows the manual block of the Intermediate Range trip and the Low Setpoint Power Range trip; and automatically blocks the Source Range trip and deenergizes the Source Range high voltage power. On decreasing power, the Intermediate Range trip and the Low Setpoint Power Range trip are automatically reactivated.

Provides input to P-7.

P-13 Turbine first stege chamber pressure provides input to P-7.

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COMANCHE PEAK - UNIT 1 B 2-8 1 '

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SECTIONS 3.0 AND 4.0 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 9

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COMANCHE PEAK - UNIT 1 3/4 0-0 t

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3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCC REQUIREMENTS 3/4.0 APPLICABILITY LIMITING CONDITION FOR OPERATION 3.0.1 Compliance with the Limiting Conditions for Operation contained in..the succ'eeding spec 1fications is required during the OPERATIONAL MODES or other conditions specified therein; except that upon failure to meet the Limiting

Conditions for Operation, the associated ACTION requirements shall be met.

3.0.'2 Noncompliance with a specification shall exist when the requirem'ents of the Limiting Condition for Operation and associated ACTION requirements are not met within the specified time intervals. If the Limiting Condition for Operation is restored prior to expiration of the specified time intervals, completion of the ACTION requirements is not. required.

3.0.3 When a Limiting Condition for Operation is not met, except as provided in the associated ACTION requirements, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months action shall be initiated to place the unit in a MODE in which the specification does not apply by placing.it, as applicable, in: . -

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

b. At least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , cnd

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

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

This specification is not applicable in MODE b or 6.

3.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made when the conditions for the Limiting Conditions for Operation are not met and the associated ACTION requires a shutdown if they are not met within a specified time interval. Entry into an OPERATIONAL MODE or specified condition may be made 'n accordance with ACTION requirements when conformance

- to them permits continued operation of the facility for an unlimited period of time. This provision shall not prevent passage through >r to OPERATIONAL i MODES as required to comply with ACTION requirements. Exceptions to these requirements are stated in the individual specifications.

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COMANCHE PEAK - UNIT 1 3/4 0-1

APPLICABILITY SURVEILLANCE REQUIREMENTS 4.0.1 Surveillance Requirements shall be met during the OPERATIONAL MODES or other conditions specified for individual Limiting Conditions for Operation '

unless otherwise stated in an individual Surveillance Requirement.

4.0.2 Each Surveillance Requirement shall be performed within the specified time interval with:

a. A maximum allowable extension not t'o exceed 25% of the surveillance interval, but

b. The combined time interval for any three consecutive surveillance intervals shall not exceed 3.25 times the specified surveillance

, interval. ,

4.0.3 Failure to perform a Surveillance Requirement within the allowed sur-veillance interval, defined by Specification 4.0.2, shall constitute noncom-pliance with the OPERABILITY requirements for a Limiting Concition for .

Operation. The time limits of the ACTION requirements are applicable at the time it is identified that a Surveillance Requirement has not been performed.

The' ACTION requirements may,be delayed for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to permit the co.m-plation of the surveillance when the allowable outage time limits of the ACTION requirements are less than. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Surveillance Requirements do not

.have to be performed on inoperable equipment. ,

4.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made unless the Surveillanct Requ~irement(s) associated with the Limiting Condition for Operation has been performed within the stated surveillance interval or as otherwise specified. This provision shall not prevent passage through or to OPERATIONAL MODES as required to comply with ACTION requirements.

4.0.5 Surveillance Requirements for inservice inspection and testing of ASME Code Class 1, 2, and 3 components shall be applicable as follows:

a. Inservice inspection of ASME Code Class 1, 2, and 3 components and inservice testing of ASME Code Class 1, 2, and 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i);

COMANCHE PEAK - UNIT 1 3/4 0-2

APPLICABILITY SURVEILLANCE REQUIREMENTS (Continued)

b. Surveillance intervals specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda for the inservice inspection and testing activities require.d by the ASME Boiler and Prcssure Vessel Code and applicable Addenda shall be applicable as follows in these Technical Specifications:

ASME Boiler and Pressure Vessel Required frequencies for Code and applicable Addenda performing inservice terminology for inservice inspection and testing inspection and testing activities activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per.276. days Yearly or annually At least once per 366 days .

c. The provisions of Specification 4.0.2 are applicable to the above required frequencies for performing' inservice inspection and testing activities;

- d. Petformance of the above inservice it,spection and testing activities shall be in addition to~other specified Surveillance Requirements; and

e. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any Technical Specification.

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COMANCHE PEAK - UNIT 1 3/4 0-3 i

3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUTDOWN MARGIN - T,y GREATER THAN 200 F LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to 1.6% ak/k for four loop operation.

2*, 3, and 4.

APPLICABILITY: MODES 1, -

ACTION:

With the SHUTDOWN MARGIN less than 1.6% ok/k, immediately initiate and con-tinue boration at greater'than or equal to 30 gpm of a solution containing greater than or equal to 7,000 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored.

SURVEILLANCE REQUIREMENTS 4.1.'1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1.6% Ak/k: -

a. Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months a'fter d'etection of an inoperable control rod (s) and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the rod (s) is inoperable.

If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod (s);

b. When in MODE 1 or MODE 2 with K,ff greater than or equal to 1 at least onct per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that control bank withdrawal is within the limits of Specification 3.1.3.6;

c. When in MODE 2 with K,ff less than 1, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to achieving reactor criticality by verifying that the predicted critical control rod position is within the limits of Specification 3.1.3.6;

d. Prior to initial operation above 5% RATED TtiERMAL POWER after each fuel loading, by consideration of.the factors of Specifica-tion 4.1.1.1.le, below, with tne contro' banks at the maximum inser-tion limit of Specification 3.1.3.6; and

See Special Test Exceptions Specification 3.10.1.

COMANCHE PEAK - UNIT 1 3/4 1-1 l

REACTIVITY CONTROL SYSTEMS' '

SURVEILLANCE REQUIREMENTS (Continued)

e. -When-in MODE 3 or 4, at least once par 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by consideration of the following factors:

1) Reactor Coolant System boron concentration,

2) Control rod position,

3) Reactor Coolant System average temperature, 4),

Fuel burnup based on gross thermal energy generation,

5) Xenon concentration, and

6) Samarium concentration.

4.1.1.1.2 The overall core reactivity balance shall be compared to predicted values to demonstrate agreement within i 1% ak/k at least once per 31 Effective Full Power Days (EFPD). This comparison shall consider at least.those factors stated in Specification 4.1.1.1.le. , above. The predicted reactivity values shall be adjusted (normalized) to correspond to the actual core conditions. prior to exceeding a fuel burnup of 60 EFPD after each fuel loading. -

COMANCHE PEAK - UNIT 1 3/4 1-2

m REACTIVITY CONTROL SYSTEMS k

SHUTDOWN MARGIN - T,y LESS THAN OR EQUAL TO 200*F LIMITING CONDITION FOR OPERATION 3.1.1.2 The SHUTDOWN MARGIN shall be greater than or equal to 1% Ak/k.

APPLICABILITY: MODE 5.

ACTION:

With the SHUTOOWN MARGIN less than 1% Ak/k, immediately initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to.7,000 ppm boron or eq'!ivalent until the required SHUT 00WN' MARGIN is restored.

SURVEILLANCE REQUIREMENTS 4.1.1.2 The SHUTDOWN MARGIN shall be determinad to be greater than or equal to 1% Ak/k:

a. ' Within l' hour after detection of an inoperable control rod (s) and at.

least.once per la hours thereafter while the rod (s) is inoperable.

If the inoperable control rod is immovable or untrippable, the SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable i

control rod (s); and

b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by consideration of the following factors:

1) Reactor Coolant System boron concentration,

2) Control rod position,

3) Reactor Coolant System average temperature,

4) Fuel burnup based on gross thermal energy generation.

l 5) Xenon concentration, and

6) Samarium concentration.

l COMANCHE PEAK - UNIT 1 3/4 1-3 i _ _

REACTIVITY CONTROL SYSTEMS K

MODERATOR TEMPERATURE COEFFICIENT .

LIMITING CONDITION FOR OPERATION 3.l.1.3 The moderator temperature coefficient (MTC) shall be:

a. Less positive than 0 ak/k/'F for the all rods withdrawn, beginning

. of cycle life (BOL), hot zero THERMAL POWER condition; and

b. Less negative than -4.0 x 10 4 ok/k/ F for the all rods witndrawn, end of cycle life (EOL), RATED THERMAL POWER condition.

APPLICABILITY: Specification 3.1.1.3a. - MODES 1 and 2* only**.

Specification 3.1.1.3b. - MODES 1, 2, and 3 only**.

ACTION:

a. With the MTC more positive than the limit of Specification 3.1.1.3a. '

above, operation in MODES 1 and 2 may proceed provided:

1. Control rod withdrawal limits are established and maintained sufficient to restore the MTC to less positive than 0 ok/k/'F within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , These withdrawal . limits.shall'be in addition to the insertio~n limits of Specification 3.1.3.6;

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

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

b. With the MTC more negative than the limit of Specification 3.1.1.3b.

above, be in HOT SHUT 00WN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

With Kg 7 greater than or equal to 1.

- See Special Test Exceptions Specification 3.10.3.

COMANCHE PEAK - UNIT 1 3/4 1-4

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. REACTIVITY CONTROL SYSTEMS pg' t

SURVEILLANCE REQUIREMENTS 4.1.1.3 The MTC shall be determined to be within its limits during each, fuel .

cycle as'follows:

a. The MTC shall be measured and compared to the BOL limit of Specifi-cation 3.1.1.3a., above, prior to initial operation above 5% of RATED THERMAL POWER, after each fuel loading; and . r

.- b. The t'TC shall be measured at any THERMAL POWER and compared to -3.1 x 10 4 ok/k/'F (all rods withdrawn, RATED THERMAL POWER condition)

. within 7 EFPD after reaching an equilibrium boron concentration of 300 ppm. In the event this comparison indicates the MTC is more negative than -3.1 x 10 4 ok/k/ F, the MTC shall be remeasured, and compared to the E0L MTC limit of Specification 3.1.1.3b., at least -

once per 14 EFPD during the remainder of the fuel cycle.

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COMANCHE PEAK - UNIT 1 3/4 1-5 i

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{'h*f MINIMUMTjMPERATUREFORCRITICALITY LIMITING CONDITION FOR OPERATION 3.1.1.4 The Reactor Coolant System lowest operating loop temperature (Tavg) -

shall be greater than or equal to 551 F.

APPLICABILITY: MODES 1 and 2* #.

ACTION: ,

With a Reactor Coolant System operating loop temperature (T,yg) less than 551'F, restore T,yg to within its limit within 15 minutes or be in HOT STANDBY within the next 15 minutes.

SURVEILLANCE REQUIREMENTS

~

4.1.1.4 The Reactor Coolant System temperature (Tavg) shall be determined to j be greater than or equal to 551*F:

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

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b. At lea'st once per 30 minutes when the reactor is critical and the Reactor Coolant System T,yg is less than 561*F with the T,yg-Tref Deviation Alarm not reset.

Nith K,ff greater than or eoual to 1.

See Special Test Exception,s Specification 3.10.3.

COMANCHE PEAK - UNIT 1 3/4 1-6

REACTIVITY CONTROL SYSTEMS j g 3/4.1.2 BORATION SYSTEMS FLOW-PATH - SHUTOOWN -

LIMITING CONDITI'ON'FOR OPERATION 3.1-2.1 As a minimum, one of the.following boron injection flow paths shall be OPERABLE and capable of being powered from an OPERABLE emergensy. power source: _

a. A flow path from the boric acid tanks via either a boric acid trans--

fer pump or a gravity feed connection and a charging pump to the.

Reactor Coolant System if the boric acid storage tank in Specification 3.1.2.5a. is OPERABLE, or

b. The flow path from the refueling water storage tank via a centrifug:1 charging pump to.the Reactor Coolant System if.the refueling water J

storage tank in Specification.3.1.2.5b. is OPERABLE.

APPLICABILITY: MODES 5 and 6.

ACTION:

With none of-the above flow paths OPERABLE'or capable of being powered from an' OPERABLE emergency power source, suspend all operations involving CORE, ALTERATIONS or positive reactivity changes.

SURVEILLANCE REQUIREMENTS t

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

a. At least once per 7 days by verifying that the temperature of the flow path is greater than or equal to 65 F when a flow path from the boric acid storage tanks is used, and

b. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or oth';-wise secured in position, is in its correct pcsition.

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COMANCHE PEAK - UNIT 1 3/4 1-7 l

REACTIVITY CONTROL SYSTEMS gpy FLOW PATHS'- OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.2 At least two* of the following three boron injection flow path., :5all be OPERABLE:

a. The flow path from the boric acid tanks via either a boric acid transfer pump or a gravity feed connection and a charging pump to the Reactor Coolant System (RCS), and

. b. Two flow paths from the refueling water storage tank via centrifugal charging pumps to the RCS, APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

Wilh only o'ne of the above required boron injection flow path; to the RCS OPERABLE, restore at least two boron injection flow paths to the RCS to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY and borated to a SHUT 00WN MARGIN equivalent to at least 1% Ak/k'at 200 F within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore at least two flow paths to OPERABLE. statu~s within the next 7 days or be in COLD SHUT 00WN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.1.2.2 At least two of the above required flow paths shall be demonstrated OPERABLE:

a. At least once per 7 days by verifying that the temperature of the flow path from the boric acid storage tanks is greater than or equal to 65'F when it is a required water source;

b. At least once per 31 days by veaifying that each valve (manual, power-operated, or automatic) in th1t flow path that is not locked, sealed, or otherwise secured in position, is in its correct position; -

c. At least once per 18 months by verifying that the flow path required by Specification 3.1.2.2a. delivers at least 30 gpm to the RCS.

d. At least once per 18 months by verifying that the flow path required by Specificatlon 3.1.2.2b. is capable of delivering at least 120 gpm to the RCS.

0nly one boron injection flow path is required to be OPERABLE whenever the temperature of one or more of the RCS cold legs is less than or equal to

[275) F.

COMANCHE PEAK - UNIT 1 3/4 1-8

REACTIVITY CONTROL SYSTEMS ,

CHARGING PUMP - SHUTOOWN LIMITING CONDITION FOR OPERATION 3.1.2.3 One charging pump in the boron injection flow path required by -

Specification 3.1.2.1 shall be OPERABLE and capable of being powered from an OPERABLE emergency power source.

APPLICABILITY: MODES 5 and 6. -

ACTION:

With no charging pump OPERABLE or ce .ble of being powered from an OPERABLE emergency power source, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

SURVEILLANCE REQUIREMENTS 4.1.2.3.1 The above required charging pump shall be demonstrate'd OPERABLE by

, verifying that the flow path required by Specification 3.1.2.la is capable of delivering at least 30 gpm to the RCS when the pump is tested pursuant to Specification 4.0.5; or 4.1.2.3.2 At least once per 18 months by verifying that the flow path required by Specification 3.1.2.lb is capable of delivering at least 120 gpm to the RCS when the pump is tested pursuant to Specification 4.0.5.

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4.1.2.3.3 All charging pumps, excluding the ab>ve required OPERABLE pump, shall be demonstrated inoperable at least once per 31 days, except when the

, reactor vessel head is removed, by verifying that the motor circuit breakers are secured in the open position.

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COMANCHE PEAK - UNIT 1 3/4 1-9

REACTIVITY CONTROL SYSTEMS CHARGING PUMPS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2,4 At least two* charging pumps shall be OPERABLE.

APPLICABILITY: H0 DES 1, 2, 3, and 4.

ACTION:

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

SURVEILLANCE REQUIREMENTS 4.1.2.4.1 The required centrifugal char .

OPERABLE by testing pursuant to Specific'ging pump ations 4.0.5 and (s) shall be demonstrated 4.1.2.2.d. -

4.1.2.4.2 The required positive displacement charging pump shall be demonstrated OPERABLE by testing pursuant to Specifications 4.0.5 and 4.1.2.2.c.

4.1.2.4.3 All charging pumps, except the above allowed OPERABLE pump, shall be demonstrated inoperable at least once per 31 days whenever the temperature of one or more of the Reactor Coolant System (RCS) cold legs is less than or equal to 308.7*F by verifying that the motor circuit breakers are secured in the open

position.

"A maximum of one centrifugal charging pump shall be OPERABLE whenever the temperature of one or more of the RCS cold legs is less than or equal to 275'F.

COMANCHE PEAK - UNIT 1 3/4 1-10

REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCE - SHUT 00WN y LIMITING CONDITION FOR OPERATION 3.1.2.5 As a minimum, one of the following borated water. sources'shall be.

OPERABLE: .

a. A Boric Acid Storage Tank with:

1) A minimum contained borated water volume of 6385 gallons,

([Later]% of span), when using the boric acid transfer pump.

2) A minimum contained borated water volume of 15,123 gallons

([Later]% of span), when using the gravity feed connection,

3) A minimum boron concentration of 7000 ppm and i

4) A minimum solution temperature of 65*F.

b. The refueling water storage tank (RWST) with:

1) A minim'um contained borated wate.r volume of 101,120 gallons,

([Later]% of, span),.

2) A minimum boron concentration of 2000 ppm and

3) A~ minimum solution' temperature of 40*F.

APPLICABILITY: MODES 5 and 6.

ACTION:

r With nc borated water source OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

SURVEILLANCE REQUIREMENTS 4.1.2.5 The above rtquired borated water source shall be demonstrated OPERABLE:

a. At least once per 7 days by:

1) Verifying the boron concentration of the water,

2) Verifying the contained borated water. volume, and

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

b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature when it is the. source of borated water and the outside air temperature is

~

less than 40 F.

COMANCHE PEAK - Ui4IT 1 3/4 1-11

-REACTIVITY CONTROL SYSTEMS Q B0 RATED WATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.6 As a minimum, the following borated water source (s) shall be OPERABLE as required by Specification 3.1.2.2: -

a. A Boric Acid Storage Tank with:

1) A minimum contained borated water volume of 22,870 gallons,

'([Later]% of span),

2) A minimum boron concentration of 7000 ppm, and

3) A minimum solution temperature of 65 F.

b. The refueling water storage tank (RWST) with:

1) A minimum contained borated water volume of 479,900 gallons,

([Later]% of span), ,

2) A boron concentration between 2000 ppm and 2200 ppm',

-3) A minimum solution temperature of 40 F, and

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4 A maximum solution temperature of 120 F. .

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

ACTION:

a. With the Beric Acid Storage Tank inoperable and being used as one of the above required borated water sources, restore the tank to .

OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within

. the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and borated to a SHUTDOWN MARGIN equivalent to at least 1% MJk at 200 F; restore the Boric Acid Storage Tank to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With the RWST inoperable, restore the tank to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

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COMANCHE PEAK - UNIT 1 3/4 1-12

REACTIVITY CONTROL SYSTEMS r-==

i

- I SURVEILLANCE REQUIREMENTS 4.1.2.6 Each borated water source shall be demonstrated OPERABLE:

4

a. At least once per'7 days by: .

-1) ' Verifying the boron concentration in the water,

2) Verifying the contained borated water volume of the water source, and

3) Verifying the Boric Acid Storage Tank solution temperature when it is the source of borated water,

b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature when the outside air temperature is either less than 40 F or greater than 120*F. . .

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C0HANCHE PEAK - UNIT 1 3/4 1-13

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REACTIVITY C0'NTROL SYSTEMS 3/4.1.3 MOVABLE CONTROL ASSEMBLIES ,[

GROUP HEIGHT LIMITING CONDITION FOR OPERATION 3.1.3.1 All full-length shutdown an'd~ control rods shall be OPERABLE and positioned within i 12 steps (indicated position) of their group step counter demand position.

APPLICABILITY: MODES 1* and 2*.

ACTION:

a. With one or more full-length rods inoperable due to being immovable as a result of excessive friction or mechanical interference or known to be untrippable, dtcormine that the SHUTOOWN MARGIN require-ment of Specification 3.1.1.1 is satisfied within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and be in HOT STAN0BY-within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,

b. With more than one full-length rod inoperable or misaligned from the group step counter demand position by more than i 12 steps (indicated position), be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c. With one full-le'ngth rod trippable but inoperable d0e to causes other than addressed by ACTION a., above,~or misaligned from' its group step counter demand height _by more than 12 steps (indicated position), POWER OPERATION may continue provided that within 1 hour:

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

2. The rod is declared inoperable and the remainder of the rods in the group with the inoperable rod are aligned to within i 12 steps of the inoperable rod while maintaining the rod sequence and insertion limits of Figure 3.1-1. The THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during l subsequent operation, or

3. The rod is declared inoperable and the SHUTDOWN PtARGIN

, requirement of Specification 3.1.1.1 is satisfied. POWER l

OPERATION may then continue provided that:

a) A reevaluation of each accident analysis of Table 3.1-1 is performed within 5 days; this reevaluation shall confirm that the previously anal.yzed results of these accidents l remain valid for the duration of operation under these conditions; b) The SHUT 00WN MARGIN requirement of Specification 3.1.1.1 is determined at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ;

See Special Test Exceptions. Specifications 3 10.2 and 3.10.3.

COMANCHE PEAK - UNIT 1 3/4 1-14

J _ da$ N ,

REACTIVITY CONTROL SYSTEMS -

LIMITING CONDITION FOR OPERATION w

' ACTION (Continued)

~ '

c) A power distribution map is obtained from the movable incore detectors and F9(Z) and F g are verified to be within their limits within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; and j d) The THERMAL POWER level is reduced to less than or equal to 75% of RATED THERMAL F3WER within the next hour and within'the following 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months the High Neutron Flux !

Trip Setpoint is reduced to less than or equal to 85%

of RATED THERHAL POWER.

. SURVEILLANCE' REQUIREMENTS 4.1.3.1.1 The position of each full-length rod shall be determined to be within the group demand limit by verifying the individual rod positions at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the rod position deviation monitor is inoperable, then verify the group positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

4.1.3.1.2 Each full-length rod not fully inserted in the core shall be [

determined to be OPERABLE by movement of at least 10 steps in any one *

, direction at least once per 31 days.

c e-1 i

G L

h C0HANCHE PEAK - UNIT.1 3/4 1-15 I

. _ .? -

.q'

' ' ^ - '

l$. .

TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT OF AN INOFERABLE FULL-LENGTH ROD Rod Cluster Control Assembly Insertion Characterist.ics 1 Rod Cluster Control Assembly Misalignment Decrease in Reactor Coolant Inventory Inadvertent opening of a pressurizer safety or relief valve Break in instrument line or other lines from reactor coolant pressure boundary that penetrate containment Steam generator tube rupture

' loss of coolant accidents resulting from a. spectrum of postulated piping

, breaks within the ' reactor coolant pressere boundary kncreases in Heat Removal by the Secondary System (steam syst em piping failure)

Spectrum of Rod Cluster Control Assembly Ejection Accidents 9

6 e

1 4

e i .

t COMANCHE PEAK - UNIT 1 3/4 1-16

yd P.EACTIVITY CONTROL SYSTEMS POSITION INDICATION SYSTEMS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.3.2 The Digital Rod Position Indication System and'the Oemand Position Indication System shall be OPERABLE and capable of determin'.ng the control rod positions within i 12 steps.

APPLICABILITY: MODES 1 and 2.

ACTION:

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

1. Determine the position of the nonindicating rod (s) indirectly by the mnvable incore detectors at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and immer.iately after any motion of the no'nind.icating rod'which- ,

exceeds 24 steps in one direction since the last determination r of the rod's position, or

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

b. With a maximum of one demand position indicator per bank inoperable eithar:

1. Verify that all digital rod position indt:ators for the affected bank are OPERABLE and that the most withdrawn rod and the least ^

withdrawn rod of the bank are within a maximum of 12 steps of each other at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , or ,

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

SURVEILLANCE REQUIREMENTS 4.1.3.2 Each digital rod position indicator shall be determined to be OPERABLE i by verifying that the Demand Position Indication System and the Digital Rod Position Indication System agree within 12 steps at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months r except during time intervals when the rod position deviation monitor is inoperable, then compare the Demand Position Indication System and the Digital Rod Position Indication System at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

. , i i

COMANCHE PEAK - UNIT 1 3/4 1-17

--. ,, . . _ . . _ . _ . . , y ,

REACTIVITY CONTROL SYSTEMS f POSITION INDICATION SYSTEM - SHUTDOWN LIMITING CONDITION FOR OPERATION

'3.1.3.3 One digital rod position indicator (excluding demand position indication) shall be OPERABLE and capable of determining the control rod position within t 12 steps-for each shutdown or control rod not fully

inserted. ,

APPLICABILITY: MODES 3* **, 4* **, and 5* **. .

ACTION:

With less than the above required position indicator (s) OPERABLE, immediately open the Reactor Trip System breakers.

J SURVEILLANCE REQUIREMENTS 4.1.3.3 Each of the above required digital rod position indicator (s) shall be

, determined to be OPERABLE by verifying that the digital rod position indicators agree with the demand position indicators within 12 steps when exercised over the full-range of rod travel at least once per 18 months.

I

With the Reactor Trip System breakers in the closed position.

- See Special Test Exceptions Specificatio'n 3.10.5.

COMANCHE PEAK - UNIT 1 3/4 1 18

c-P REACTIVITY CONTROL SYSTEMS ROD DROP TIMC

LIMITING CONDITION FOR OPERATION 3.1.3.4 The individual full-lenqth (shutdown and control) rod drop time from the f'ully withdrawn position sha,1 be less than or equal to 2.4 seconds from t beginning of decay of stationary gripper coil voltage to dashpot entry with:

a. T,yg greater than or equal to 551'F, and

b. All reactor coolant pumps operating.

APPLICABILITY: MODES 1 and 2.

ACTION:

a. With the drop time of any full-length rod determined to exceed the above limit, restore the rod drop time to within the above limit l

prior to proceeding to MODE 1 or 2.

SURVEILLANCE REQUIREMENTS

. 4.1.3.4 The rod drop time of full-length rods shall be demonstrated through measurement prior to reactor criticality:

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

b. For specifically affected individual rods following any maintenance on or modification to the Control Rod Drive System which could .

affect the drop time of those specific rods, and '

c. At least once per 18 months. ,

O COMANCHE PEAK - UNIT 1 3/4 1-19

r- 9 REACTIVITY CONTROL SYSTEMS 1r

\

SHUT 00WN R00 INSERTION LIMIT. N

LIMITING CONDITION FOR OPERATION -

3.1.3.5 All shutdown rods shall be' fully withdrawn. ,

APPLICABILITY: MODES 1* and 2* **.

ACTION:

With a maximum oY one shutdown rod not fully withdrawn, except.for surveillance testing pursuant to Specification 4.1.3.1.2, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either:

a. Fully withdraw the rod, or

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

SURVEILLANCE REQUIREMENTS -

i 4.1.3.5 Each shutdown rod shall be determined to be fully withdrawn:

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

, b. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter.

4 d

1 "See Special Test Exceptions Specifications 3.10.k snd 3.10.3.

- With K,ff greater than or equal to 1. ,

CDMANCHE PEAK - UNIT 1 3/4 1-20

REACTIVITY CONTROL SYSTEMS RI ,

ki e CONTROL ROD INSERTION LIMITS LIMITING CONDITION FOR OPERATION s

3.1.3.6 The control banks shall be limited in physical insertion as shown in Figure 3.1-1.

~

APPLICABILITY: MODES la and 2* **.

ACTION:

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

a. Restore the control banks to within the limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or l

b. Reduce THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to less than or equal to that fraction of RATED THERMALzPOWER which is allowed by the bank posi-tion using the above figure, or

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

~

SURVEILLANCE REQUIREMENTS 4.1.3.6 The position of each. control bank shall be~ determined to be within the insertion limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the rod insertion limit monitor is inoperable, then verify the individual rod positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

"See Special Test Exceptions Specifications 3.10.2 and 3.10.3.

- With K,ff greater than or equal to 1.

COMANCHE PEAK - UNIT 1 3/4 1-21

BRAFI I

l FIGURE 3.1-1 R00 BANK INSERTION LIMITS VERSUS THERMAL POWER FOUR LOOP OPERATION COMANCHE PEAK - UNIT 1 3/4 1-22 i

4 3/4.2 POWER D2STRIBUTION LfMITS )\ ,

3/4.2.1 AXIAL FLUX DIFFERENCE ,

LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX OIFFERENCE (AFD) shall be maintained within the following target band (flux difference units) about the target flux difference: ;

a. i 5% for core average accumulated burnup of less than or equa? to 3000 MWD /MTU; and

b. + 3%, -12% for core average accumulated burnup of greater than 3000 MWD /MTU.

The indicated AFD may deviate outside the above required target band at greater than or equal to 50% but less than 90% of RATED THERMAL POWER provided the indi-cated AFD is within the Acceptable Operation Limits of Figure 3.2-1 and the cuma-lative penalty deviation time does not exceed I hour during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

The indicated AFD may deviate outside the above required target band at greater

than 15% but less than 50% of RATED THERMAL POWER provided the cumulative penalty deviation time does not exceed I hour during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

APPLICABILITY: MODE 1, above 15% of RATED THERMAL POWER.*

i j ACTION:

a. With the indicated AFD outside of the above required' target band and '

with THERMAL POWER greater.than or equal to'90% of RATED THERMAL

POWER, within 15 minutes either:

. r

1. Restore the indicated AFD to within the target band limits, or

2. Reduce THERMAL POWER to less than 90% of RATED THERMAL POWER.

b. With the indicated AFD outside of the above required target band for i more than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of cumulative penalty deviation time during the

previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or outside the Acceptable Operation Limits of .

l Figure 3.2-1 and with THERMAL POWER less than 90% but equal to or greater than 50% of RATED THERMAL POWER:

i

1. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes, and

2. Reduce the Power Range Neutron Flux - High Trip ** Setpoints to less than or equal to 55% of RATED THERMAL POWER within,the next j 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . ,

See Special Test Exc6ptions Specification 3.10.2.

AA Surveillance testing of the Power Range Neutron Flux Channels may be performed l pursuant to Specification 4.3.1.1 provided the indicated AFD is maintained within l

the Acceptable Operation Limits of Figure 3.2-1. A total of 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months operation '

may be accumulated with the AFD outside of the above required target band during i testing without penalty deviation, i

f COMANCHE PEAK - UNIT 1 3/4 2-1 i

l ,,- _ _ _ _ _

1T pCNER OIST_RIBUTION LIMITS

?

LIMITING CONDITION FOR OPERATION E, ACTION (Continued)

c. With the indicated A outside of the above required target band for more than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of cumulative penalty deviation time during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and with THERMAL POWER less than 50% but greater

. than 15% of RATED THERMAL POWER, the THERMAL POWER shall not be increased equal to or greater than 50% of RATED THERMAL POWER until the indicated AFD is within the above required target band.

SURVEILLANCE REQUIREMENTS .

4.2.1.1 The indir.ated AFD shall be determined to be within its limits during POWER OPERATION above 15% of RATED THERMAL POWER by:

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

1) At least once per 7 days when the AFD Monitor Alarm is OPERABLE, and

2) At least once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after restoring the AFD Monitor Alarm to OPERABLE status.

b. Monitoring and logging the indicat04 AFD for each OPERABLE excore -

channel at least once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and at least onco per 30 minutes thereafter, when the AFD Monitor

Alarm is inoperable. The logged values of the indicated AFD shall be assumed to exist during the interval preceding each logging.

4.2.1.2 The indicated AFD shall be considered outside of its target band when two or more OPERABLE excore channels are indicating the AFD to be outside the target band. Penalty deviation outside of the above required target band shall be accumulated on a time basis of:

a. One minute penalty deviation for each 1 minute of POWER OPERATION i

outside of the target band at THERMAL POWER levels equal to or above

50% of RATED THERMAL POWER, and

b. One-half minute penalty deviation for each 1 minute of POWER OPEPATION outside of the target band at THERMAL POWER levels between 15% and 50% of RATED THERMAL POWER.

4.2.1.3 The target flux difference of each OPERABLE excore channel shall be determined by measurement at least once per 92 Effective Full Power D,ays.

The' provisions of Specification 4.0.4 are not applicable.

I 4.2.1.4 The target flux difference shall be updated at least once per 31 Effective Full Power Days by either determining the target flux difference pursuant to Specification 4.2.1.3 above or by linear interpolation between the most recently measured value and 0% at the end of the cycle life. The provi-sions of Specification 4.0.4 are not applicable.

~

COMANCHE PEAK - UNIT 1 3/4 2-2

0 G

1 9

FIGURE 3.2-1 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER COMANCHE PEAK - UNIT 1 3/4 2-3

}

s. ,

POWER DISTRIBUTION LfMITS 3/4.2.2 HEATFLUXHdTCHANNELFACTOR-Fg ,

LIMITING CONDITION FOR OPERATION 3.2.2 F 9(Z) shall be limited by the following relationships: .

FO (2) $ [2.32) [K(Z)] for P > 0.5 P

F9 (Z) 1 [(4.64)] (X(Z)] for P $ 0.5

, and Where: P = THERMAL POWER RATED THERMAL POWER X(Z) = the function obtained from Figure 3.2-2 for a given core height location.

APPLICABILITY: MODE 1.

ACTION:

With Fg (Z) exceeding its limit:

a. Reduce THERMAL POWER at least 1% for each 1%9F (Z) exceeds the limit within 15 minutes and'similarly reduc'e the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; subsequent POWER OPERATION may proceed provided the Overpower N-16 Trip Setpoints have been reduced at hast 1% for each 1% F (Z) exceeds the limit; and 0

b. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced limit re-quired by ACTION a., above; THERMAL POWER may then be increased provided F (Z) is demonstrated through incore mapping to be 9

. within its limit.

COMANCHE PEAK - UNIT 1 3/4 2-4

l 6

FIGURE 3.2-2 K(2) - NORMALIZED F g('Z) AS A FUNCTION OF CORE HEIGHT -

COMANCHE PEAK - UNIT 1 3/4 2-5

POWER DISTRfBUTION LIMITS I SURVEILLANCE REQUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

4.'2.2.2 F xj shall be evaluated to determine ifqF (Z) is within its limit by: ,

, a. Using the movable incore detectors to obtain a power distribution map at any THERMAL POWER greater than 5% of RATED THERMAL POWER, '

Increasing the measured F xy component of the power distrib'ution map b.

by 3% to account for manufacturing tolerances and further increasing the value by 5% to account for measurement uncertainties, C

c. Comparing the F xy computed (FxY) obtained in Specification 4.2.2.2b.,

above to:

1) The F xy limitsforRATEDTHERMALPOWER((RTP)fortheappropriate

measured core planes vgi'en in Specification 4.2.2.2e. and f.,

below, and

2) The relationship: .

Fx f' = F P(1+0.2(1-P)),

l is'the limit for fractional THERMAL POWER operation Where F expresshasafunctionofF RTP and P is.the fraction of RATED x

THERMAL F0WER at which F xy was measured.

xy according to the following schedule:

d. Remeasuring F ,

i 1) When F is greater than the F xRTP limit or the appropriate measured core plane but less than the F relationship, additional power distribution maps shall be taken a d F compared to F,RTP ,

and F either:

xy a) Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding by 20% of RATED THERMAL C

POWER or greater, the THERMAL POWER at which F*Y was last determineu, or

)

b) At least once per 31 Effective Full Power Days (EFPD), -

l whichever occurs first. .

1 .

COMANCHE PEAK - UNIT 1 3/4 2-6 '

. _ .__.i

g POWER DISTRIBUTION LIMITS ,

SURVEILLANCE REQUIREMENTS (Continued)

C

2) When the F x is less than or equal to the F xR limit for the .

appropriate measured co,re plane, additional power distribution *

. maps shall be taken and F compared to F and F at least x x once per 31 EFPD.

xy limits for RATED THERMAL POWERx(FRTP) shall be provided for

e. The F all core planes containing Bank "D" control rods and all unrodded core planes in a Radial Peaking Factor Limit Report per Specifica- !

l tion 6.9.1.6;

f. The F xy limits of Specification 4.2.2.2e., above, are not applicable ;

in the following core planes regions as measured in percent of core i height from the bottom of the fuel: ;

1) Lower core region from 0 to 15%, inclusive,. ;

2) Upper core region from 85 to 100%, inclusive,

?

l 3) Grid plane regions at 17.8 1 2%, 32.1 1 2%, 46.4 2 2%, 60.6 2 2%, ;

and 74.9 1 2%, inclusive, and

, [

4) Core' plane regions within 2% of core height (1 2.88 inches] "

about the bank demand position of the Bank "D" control rods.

g. With F exceeding F , the effects of F n Fg (Z) shall be xy evaluated to determine if F g (Z) is wi?.hin its limits.

4.2.2.3 When Fg (Z) fi, measurea for other than F determinations, an overall

]

i measured F (2) shall be obtained from a power distribution map and incre.tsed 9

by 3% to account for manufacturing tolerances and further increased by 5% to account for measurement uncertainty, i 1 .

i

{

P COMANCHE PEAK - UNIT 1 3/4 2-7

ci 1

l l POWER DISTRIBUTION LIMITS !

" i 3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR '

LIMITING CONDITION FOR OPERATION t 3.2.3 The combination of indicated Reactor Coolant System (RCS) total flow -

rate and R shall be maintained within the region of allowable. operation shown ;

on Figure 3.2-3 for four loop operation. -

Where:

N a' R = F" 0 '

1.49 [1.0 + 0.2 (1.0 - P)] i 2 b* P = THERMAL POWER , and !

RATED THERMAL POWER ;

c. F g= Measured values of F H obtained by using the movable incore detectors to obtain a power distribution map. The measured :

values of F H shall be used to calculate R since Figure 3.2-3 [

includes penalties for undetected feedwater venturi fouling of l 0.3% and for measurement uncertainties of 1.8% for flow and 4% for incore measurement of F H'

' APPLICABILITY: MODE 1.

l ACTION: I With the combination of RCS total flow rate and R outside the region of l acceptable operation shown on Figure 3.2-3: i

a. Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either:

1. Restore the combination of RCS total flow rate and R to within L j the above limits, or

2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER i and reduce the Power Range Neutron Flux - High Trip Setpoint to less than or equal to 55% of RATED THERMAL POWER within the I next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , t

! i l

l-f b

I J L I

COMANCHE /EAK - UNIT 1 3/4 2-8

4 t

S FIGURE 3.2-3 RCS TOTAL FLOW RATE VERSUS R - FOUR LOOPS IN OPERATION COMANCHE PEAK - UNIT 1 3/4 2-9 9

POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION ACTION (Continued)

b. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .of initially being outside the above limits, verify ,

through incore' flux ma~pping and RCS total flow rate comparison that the combination of R and RCS total flow rate are restored to within the above limits, or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

c. Identify and correct the cause of the out-of-limit condition prior

to increasing THERMAL POWER above the reduced THERMAL POWER limit required by ACTION a.2- and/or b., above; subsequent POWER OPERATION may proceed provided that the combination of R and indicated RCS

" total flow rate are demonstrated, through incore flux mapping and RCS total flow rate comparison, to be within the region of acceptable operation shown on Figure 3.2-3 prior to exceeding the following THERMAL POWER levels:

1. A nominal'50% of RATED THERMAL POWER,

2. A nominal 75% of RATED THERMAL POWER, and

3. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of attaining greater than or equal to 95% of RATEQ THERMAL POWER.

l SURVEILLANCE REQUIREMENTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicable 4.2.3.2 The combination of indicated RCS total' flow rate and R shall be deter-mined to be within the region of acceptable operation of Figure 3.2-3:

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

b. ' At least once per 31 Effective Full Power Days.

4.2.3.3 The indicated RCS total flow rate shall be verified to be within the region of acceptable operation of Figure 3.2-3 at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the most recently obtained value of R, obtained per Specification 4.2.3.2, is assumed to exist.

4.2.3.4 The RCS total flow rate indicators shall be subjected to a CHANNEL CALIBRATION at least once per 18 months. The measurement instrumentation

shall be calibrated within 7 days prior to the performance of the calorimetric 1 flow measurement.

l 4.2.3.5 The RCS total flow rate shall be determined by precislen heat balance

measurement at least once per 18 months.

j COMANCHE PEAK - UNIT 1 3/4 2-10 i _

POWER DISTRIBUTION. LIMITS

'3/4.2.4 QUADRANT POWER TILT RATIO ,

LIMITING CONDITION FOR OPERATION 3.2.4 The QUADRANT POWER TILT RATIO shall not exceed 1.02.

APPLICABILITY: MODE 1, above 50% of RATED THERMAL POWER

ACTION:

a. With the QUADRANT POWER TILT RATIO determined to exceed 1.32

1. Calculate the QUADRANT POWER TILT RATIO at least once per hour until either:

a) The QUADRANT POWER TILT RATIO is reduced to within its limit, or.

b) THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER.

2. Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either:

a) Reduce the. QUADRANT POWER TILT RATIO to within its limit, or -

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

3. Verify that the QUADRANT POWER TILT RATIO is within its limit l within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding the limit or reduce THERMAL l

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

4. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified acceptable.at 95%

or greater RATED THERMAL POWER.

i

See Special Test Exceptions Specification 3.10.2.

l l

COMANCHE PEAK - UNIT 1 3/4 2-11 E _

' POWER DISTRIBUTION LIMITS s SURVEILLANCE REQUIREMENTS 4.2.4.1 The QUADRANT POWER TILT RATIO shall be determined to be within the limit above 50% of RATED THERMAL POWER by:

a. Calculating the ratio at least once per 7 days when the alarm is OPERABLE, and

( b. Calculating the ratio at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during steady-state operation when the alarm is inoperable.

4.2.4.2 The QUADRANT POWER TILT RATIO shall be determined to be within the limit when above 75% of RATED THERMAL POWER with one Power Range channel inoperable by using the movable incore detectors to confirm indicated QUADRANT POWER TILT RATIO at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by either:

a. Using the four pairs of symmetric thimble locations or i .

b. Using the Movable Incore Detection System to monitor the QUADRANT POWER TILT RATIO subject to the requirement of Specification 3.3.3.2.

9 8

COMANCHE PEAK - UNIT 1 3/4 2-12

POWER DISTRIBUTION LIMITS 3/4.2.5- DNB PARAlETERS LIMITING CONDITION FOR OPERATION 3.2.5 The followilig DNB related phrameters shall be maintained within tile stated limits:

a. Indicated Reactor Coolant System T < 592 F 3yg

b. Indicated Pressurizer Pressure > 2207 PSIG**

APPLICABILITY: MODE 1.

ACTION:

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

SURVEILLANCE REQUIREf1ENTS .

4.2.5 Each of the above parameters shall be verified to be within its limits at'least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . -

1

Limit not epplicable during eithcr a THERMAL POWER ramp in excess of 5% of i

' RATED THERMAL POWER per minute or a THERMAL POWER step in excess of 10% of rated thermal power.

l COMANCHE PEAK - UNIT 1 3/4 2-13 l

3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION LIMITING-CONDITION FOR OPERATION 3.3.1 s a minimum, the Reactor Trip System instrumentation channels and interlocks of Table 3.3-1 shall be OPERABLE-with Reactor Trip System RESPONSE TIMES as shown in Table 3.3-2.

APPLICABILITY: As shown in Table 3.3-1.

ACTION:

As shown in Table 3.3-1. .

l SURVEILLANCE REQUIREMENTS 4.3.1.1 Each Reactor Trip System instrumentation channel and interlock and the automatic trip logic shall be demonstrated OPERABLE by the performance of the Rea'ctor Trip System Instrumentation Surveillance Requirements specified in Table 4.3-1.

4.3<1.2 The REACTOR TRIP SYSTEM RESPONSE TIME of each Reactor trip function shall be demonstrated ta be within its-limit at least once per 18 months.

! Each test shall incluce at least one train such that both trains are tested at j least once per 36 morths and one channel per function such that all channels l

are tested at least once every N times 18 months where N is the total num'oer ;

of redundant channels in a specific Reactor trip function as shown in the l "Total No. of Channels" column of Table 3.3-1.

I I

l 1

l I

l l

COMANCHE PEAK - UNIT 1 3/4 3-1

~

TABLE 3.3-1 .

n o

I REACTOR TRIP SYSTEM INSTRUMENTATION z

o 5 MINIMUM -

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

7 1. Manual Reactor Trip 2 1 2 10 2 a- 1 2 1 a E 2 3,4,S 10 Z 2. Power Range, Neutron Flux

- a. High Satpoint 4 2 3 I

b. Low Setpoint 4 ? d b 3 1,2 2

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

4. Power Range, Neutron Flux, 4 2 3 1, 2 2b High Negative Rate s*

5. Intermediate Range, Neutron Flux 2 d 1 2 I,2 Y

3

6. Source Range, Neutron Flux

a. Startup 2 1 2 2c

b. Shutdown 4 2 1 2 3,4,5 5

7. Overtemperature N-16 '

a. Four Loop Operation 4 2

. 3 1, 2 6b :

8. Overpower N-16

a. Four Loop Operation 4 2 b

3 1, 2 6

9. Pressurizer Pressure--Low (Four Loop 4 2 3 1 6b Plant)

10. Pressurizer Pressure--High (Four Loop Plant) 4 2 3 1, 2 6b d w .

e

- +

TABLE 3.3-1 (Centinued)'

8 g REACTOR TRIP SYSTEM INSTRUMENTATION M ^

g , MINIMUM

, TOTAL NO. CHANNELS- CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

11. Pressurizer Water Level--High 3 2 2 l' 7 b

E q 12. Reactor Coolant Flow--Low g a. Single Loop (Above P-8) 3/ loop 2/ loop in 2/ loop in 1 7 b

any oper- each oper-ating loop ating loop

b. Two Loops (Above P-7 and 3/ loop '2/ loop in 2/ loop 1 7 b

below P-8) two oper- each oper-ating loops ating loop w 13. Steam Generator Water b 4/stm. gen. 2/stm. gen. 3/stm. gen. 1, 2 6

Level--Low-Low w

in any oper- each oper-w

. ating stm. ating stm.

gen. gen.

14. Undervoltage--Reactor Coolant Pumps 4-1/ bus 2 b 3 l' 6

15. Underfrequency--Reactor Coolant . -

Pumps 4-1/ bus 2 b 3 l' 6

16. Turbine Trip

a. Low Fluid Oil Pressure 3 0 2 1*

2 7

b. Turbine Stop Valve Closure 4 4 1 l' 7 b

17. Safety Injection Input from ESFAS 2 1 2 1, 2 9 W

,c0 m

~

TABLE 3.3-1 (Continued)- .

n o REACTOR TRIP SYSTEM INSTRUMENTATION f

h m

MINIMUM -

TOTAL NO. '. CHANNELS CHANNELS APPLICABLE.

A FUNCTIONAL UNIT OF CHANNELS _TO TRIP OPERABLE MODES ACTION R

, 18. Reactor Trip System Interlocks '

e a. Intermediate Range c

3 a

Neutron Flux, P-6 2 1 2 2 8 H b. Low Power Reactor Trips Block, P-7 P-10 Input 4 2 3 1 8 P-13 Input 2 1 2 1 8

c. Power Range Neutron Flux, P-8 4 2 3 1 8

{ d. Power Range Neutron

, Flux, P-10 4 2 3 1,2 8 e-. Turbine Impulse Ct. amber Pressure, P-13 2 1 2 1 . 8

19. Reactor Trip Breakers 2 1 2 10 2 9, 12 a

2 1 2 3,4,S8 10

20. Automatic Trip and Interlock 2 1 2 1 2 9 Logic 2 1 2 3 , 4^, S.a 10 i M t.-y

':-?t A

ed TABLE 3.3-1 (Continued) -

TABLE NOTATIONS a

When the Reactor Trip System breakers are in the closed 90s'ition and the Control Rod Drive System is capable of rod withdrawal.

The provisions of. Specification 3.0.4 are 'not applicable.

c Below the P-6 (Intermediate Range Neutron Flux Interlock) Setpoint, d

Below the P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.

'Above the P-7 (At Power) Setpoint ACTION STATEMENTS -

ACTION 1 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within.48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .or.be in HOT STANDBY within the next'6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

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 tri.pped 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, the inoperable 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 of other channels per Specification 4.3.1.1, and

c. Either, THERMAL POWER is restricted to less than or equal to 75% of RATED THERMAL POWER 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 per Specification 4.2.4.2.

e l COMANCHE PEAK - UNIT 1 3/4 3-5

1 TABLE 3.3-1 (Continued)

ACTION STATEMENTS (Continued)

' ACTION 3 - With the number of channels OPERABLE one less than the Minimum Channels.0PERABLE requirement and with the THERMAL POWER level:

a. Below the P-6 (Intermediate Range Neutron Flux Interlock) '

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

b. Above the P-6 (Intermediate Range Neutron Flux Interlock)

Setpoint but below 10% of RATED THERMAL POWER, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above 10% of RATED THERMAL POWER.

ACTION 4 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, suspend all operations involving positive reactivity changes.

ACTION 5 - With the_ number of OPERABLE channels one les's than the Minimum Channels OPERABLE quirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or within the next hour open i

the Reactor Trip System breakers, suspend all operations'involv-ing positive reactivity changes and verify either valve ICS-8455 or valves 105-8560, FCV-11113, ICS-8439, 105-8441, and 105-8453 are closed and secured in position, and verify this position at

. least once per 14 days thereafter. With no channels'0PERABLE -

complete all the above actions within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and veri.fy the positions of the above valves at least once per 14 days thereafter.

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

a. The inoperable channel is placed in the tripped condition within I hour, and

b. The Minimum Channels OPERABLE requirement is met; however,

. the inoperable 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 of other channels per l 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 ANALOG CHANNEL OPERATIONAL

' TEST provided the inopert51e channel is placed in the tripped l condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

1 ACTION 8 - With less than' the Minimum Number of Channels OPERABLE, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months determine by observation of the associated permissive

! annunciator window (s) that the.interloc'k is in its required state for the existing plant condition, or apply Specification 3.0.3.

1 COMANCHE PEAK - UNIT 1 3/4 3-6 l

TABLE 3.3-1 (Continued) f ACTION STATEMENTS (Continued)

ACTION 9 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY with.in 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 sur.veillance testing per Specification 4.3.1'.1, provided the other channel is OPERABLE. . ,

ACTION 10 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or open the Reactor Trip System breakers within the next hour.

ACTION 11 - With the number of OPERABLE channels less than the Total Number of Channels, operation may continue provided the inoperable channels are placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

ACTION 12 - With one of the diverse trip features (undervoltage or shunt trip attachment) inoperable, restore it to.0PERABLE status ,

within 48 1ours or declare the breaker inoperable and apply ACTION 9. The breaker shall not be bypassed while one of the diverse trip features is inoperable except for the time required for performing maintenance to restore the breaker to OPERABLE status.

l i

l l

F C0HANCHE PEAK - UNIT 1 3/4 3-7 1

TABLE 3.3-2.

8 3E REACTOR TRIP SYSTEM INSTRUMENTATI N RESPONSE TIMES M

E

, FUNCTIONAL bMIT RESPONSE TIME 2, .

3 1. Manual Reactor Trip N.A.

gi 2. Power Range, Neutron Flux ~

$ 0.5 second*

Q

,, 3. Power Range, Neutron Flux, High Positive Rate N.A.

4. Power Range, Neutron Flux, High Negative Rate 5 0.5 second* .

5. Intermediate Range, Neutron Flux '.

N.A.

o 6. Source Range, Neutron Flux

$ 0.5 seconds i> 7. Overtemperature N-16 m 5 7 seconds *#

8. Overpower N-16 <

_ 7 seconds *

9. Pressurizer Pressure--Low <

_ Z seconds

10. Pressurizer Pressure--High '

_ 2 seconds

11. Pressurizer Water Level--High N.A.

Neutron / gamma detectors are exempt from response time testing. Response time of the neutron flux / gamma signal portion of the channel shall be measured from detector output or input of first electronic component in channel.

Response time includes the the'rmal well response time.

C

ws

. ?E33 y"g rM

4 TABLE 3.3-2 (Centinued)

E3 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES 55 EE FUNCTIONAL UNIT '

RESPONSE TIME A

3; 12. Reactor Coolant Flow--Low ,

[. a. Single Loop (Above P-8) $ [1] second me b. Two Loops (Above P-7 and below P-8) $ [1] second

--e sa 13. Steam Generator Water Level--Low-Low $ [2] seconds

14. . Steam Generator Water Level-Low Coincident with Steam /Feedwater Flow Mismatch N.A.

15. Undervoltage - Reactor Coolant Pumps 5 [1.5] seconds

(( 16. Underfrequency - Reactor Coolant Pumps 5 [0.6] second

}l 17. Turbine Trip

a. Low Fluid Oil Pressure N.A.

b. Turbine Stop Valve Closure N.A.

18. Safety Injection Input from ESF N.A.

19. Reactor Trip System Interlocks N.A.

20. Reactor Trip Breakers N.A.

21. ' Automatic Trip and Interlock logic N.A. - -

.a3 9

$NN

n TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 2

m TRIP u ANALOG ACTUATING MODES FOR S

CHANNEL DEVICE WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGICTES] IS REQUIRED U 1. Manual Reactor Trip N.A. N.A. N.A. R(14) .A. a 1, 2, 3 , 4 , S a

e

2. Power Range, Neutron Flux

a. High Setpoint S D(2, 4), M N.A. N.A. 1, 2 M(3, 4),

Q(4, 6),

R(4, 5)

b. Low Setpoint S R(4) M N.A. N.A. c

. I,2 R

3. Power Range, Neutron Flux, N.A. R(4) M N.A. N.A. 1, 2 High Positive Rete Y

Es 4. Power Range, Neutron Flux, N.A. R(4) M N.A. N.A. 1, 2 High Negative Rate

5. Intermeciate Range, S R(4, 5) S/U(1),M N.A. N.A. 1c,2 Neutron Flux

6. Source Range, Neutron Flux S R(4, 13) S/U(1),M(9) R(12) N.A. b 2 , 3, 4, 5

7. Overtemperature N-16 S R M N.A. N.A. 1, 2

8. Overpower N-16 S R M N.A. N.A. 1, 2

9. - Pressurizer Pressure--Low S R M N.A. N.A. I

10. Pressurizer Pressure--High S R M N.A. N.A. 1, 2

11. Pressurizer Water Level-- S R M N.A. N.A. I c .-

High

12. Reactor Coolant Flow--Low S R M N.A. N.A.

39 M

1 W

TABLE 4.3-1 (Centinued)

.. O REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS f

z g TRIP m

ANALOG ACTUATING MODES FOR g CHANNEL DEVICE WHICH g CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE

, FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED

$ 13. Steam Generator Water Level-- S ,

R M N.A. N.A. 1, 2

-4 Low-Low e-

14. Undervoltage - Reactor Coolant N.A. R N.A. M N.A. 1 Pumps

15. Underfrequency - Reactor N.A. R N.A. M N.A. 1 Coolant Pumps

, IG. Turbine Trip 5 a. Low Fluid Oil Pressure N.A. R N.A. S/U(1, 10) N.A. I 4 b. Turbine Stop Valve N.A. R N.A. S/U(1, 10) N.A. 1

! - Closure

17. Safety Injection Input from N.A. N.A. N.A. R N.A. 1, 2 ESFAS

18. Reactor Trip System Interlocks

. a. Intermediate Range Neutron Flux, P-6 N.A. b R(4) M N.A. N.A. 2

b. Low Power Reactor Trips Block, P-7 N.A. R(4) M(8) N.A. N.A. 1

c. Power Range Neutron j Flux, P-8 N.A. R(4) M(8) N. A. N/A. 1 4

.,N .

W

TABLE 4.3-1 (Centinued)'

8 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS f

5 TRIP 5 ANALOG ACTUATING MODES FOR u - -

CHANNEL DEVICE WHICH

$ CHANNEL. CHANNEL OPERATIONAL OPERATIONAL AC'TUATION SURVEILLANCE 7 FUNCTIONAL UNIT CHECK CALIBRATION. TEST TEST LOGIC TEST IS REQUIRED E 20. Reactor Trip System Interlocks (Continued) e e. Power Range ,

Neutron Flux, P-10 N.A. R(4) M(8) N.A. N.A. 1, 2

f. Turbine Impulse. Chamber Pressure, P-13 N.A. R M(8) N.A. N.A. 1

21. Reactor Trip Breaker N.A. N.A. 8 a 6 N. A.' M(7, 11) N.A. 1, 2, 3 , 4 , 5 R 22. Automatic Trip and Interlock N.A. N.A. N.A. N.A. M(7) 1, 2, 3 a, 4 a, Sq Logic

{

h 23. Reactor Trip Bypass Breaker N.A. N.A. N.A. M(15),R(16) ' N.A. 1, 2, 3", 4a, 56 O

. .y N

W

i TABLE 4.3-1 (Continued) %

TABLE NOTATIONS hWU i "When the Reactor Trip System breakers are closed and the Control Rod Drive System is capable of rod withdrawal.

b '

8elow P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.

c Below P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.

(1) If not performed in previous 7 days. '

(2) Comparison of calorimetric to excore power indication above 15% of RATED THERMAL POWER. Adjust excore channel gains consistent with calorimetric power if absolute' difference of the respective channel is greater than 2%. The provisions of Specification 4.0.4 are not applicable to entry into M00E 2 or 1.

(3) Single point comparison of incore to excore AXIAL FLUX DIFFERENCE above 15% of RATED THERMAL POWER. Recalibrate if the absolute difference is greater than'o'r equal to 3%. The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

(4) Neutron detectors may be excluded from CHANNEL CALIBRATION.

(5) Detector plateau curves shall be obtained, and evaluated and compared to manufacturer's data. For the Intermediate Range and Power Range Neutron' Flux channels the provisions of Specification 4.0.4 are not applicable for entry into MODE 1 or 2.

(6) Incore - Excore Calibration, above 75% of RATED :HERMAL POWER. The provisions of Specification 4.0.4 are not applicable for entry into MODE 1 or 2.

(7) Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS.

l (8) With power greater than.or equal to the Interlock Setpoint the required ANALOG CHANNEL OPERATIONAL TEST shall consist of verifying that the .

interlock is in the required state by observing the permissive annun-l ciator window. .

a a a (9) Monthly surveillance in MODES 3 , 4 , and S shall also include verifica-tion that permissives P-6 and P-10 are in their required state for exist-ing plant conditions by observation of the permissive annunciator window.

Monthly surveillance shall include verification of the Bor,on Dilution Alarm Setpoint of less than or equal to (an increase of twice the count rate within a 10-minute period). ,

l l .

COMANCHE PEAK - UNIT 1 3/4 3-13

TABLE 4.3-1 (Continued)

TABLE NOTATIONS (Continued)

(10) Setpoint verification is not applicable.

t- (11) The TRIP ACTUATING DEVICE OPERATIONAL TEST shall independently verify the l OPERABILITY of the undervoltage and shun.t trip attachments of the Reactor I

Trip Breakers.

(12) At least once per 18 months during shutdown, verify that on a simulated Boron Dilution Doubling test signal the normal CVCS discharge valves close and the centrifugal charging pumps suction valves from the RWST open within 30 seconds.

(13) With the high voltage setting varied as recommended by the manufacturer, an initial discriminator bias curve shall be measured for each detector.

' Subsequent discriminator bias curves shall be obtained, evaluated and compared to the initial curves.

(14) The TRIP ACTUATING DEVICE OPER.\TIONAL TEST shall independently verify the OPERABILITY of the undervoltage and shunt trip circuits for the Manual Reactor, Trip Functi'on. The test'shall also verify the OPERABILITY of the Bypass Breaker trip circuit (s).

(15) local manual shunt trip prior to placing breaker in service.- (Or for plants that do not actuate the shunt trip attachment of.the bypass breakers on a manual reactor trip): Remote manual undervoltage trip when breaker placed in, service. .

(16) Automatic undervoltage trip.

COMAP ME PEAK - UNIT 1 3/4 3-14

- _- _ _ _ _ _ _ _ _ _ . _ _ _ . _ _ _ _ . _ _ _ _ _ . _ _ _ A. _ __ _

INSTRUMENTATION d 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM 16:TRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2 The Engineered Safety Features Actuation System (ESFAS) instrumentation channels and interlocks shown in Table 3.3-3 shall be OPERABLE with their Trip Setpoints set consistent with the valves shown in the Trip Setpoint column of Table 3.3-4 and with ESF RESPONSE TIMES as shown in Table 3.3-5.

APPLICABIllTY: As shown in Table 3.3-3.

ACTION:

a. With an ESFAS Instrumentation or Interlock Trip Setpoint trip less conservative than the value shown in the Trip Setpoint column but more conservative than the value shown in the Allowable Value column of Table 3.3-4, adjust the Setpoint consistent with the Trip Setpoint value.

b. With'an ESFA'S'In'strumentation or Interlock Trip Setpoint less conserva-tive than the value shown in the Allowable Value column of Table 3.3-4, either:

1. Adjust the Setpoint consistent with the Trip Setpoint value of Table 3.3-4, and determine within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that Equation 2.2-1 was satisfied for the affected channel,.or ,

2. ' Declare the channel inoperable and apply the applicable ACTION statement requirements of Table 3.3-3 until the channel is restored to OPERABLE status with its Setpoint adjusted consistent with the Trip Setpoint value.

Equation 2.2-1 Z + R + S < TA Where:

Z = The value from Column Z of Table 3.3-4 for the affected channel, R = The "as measured" value (in percent span) of rack error for the affected channel, S = Either the "as measured" value (in percent span) of the sensor 1

error, or the value from Column S (Sensor Error) of Table 3.3-4 for the affected channel, and TA = The value from Column TA (Total Allowance) of Table 3.3-4 for the affected channel.

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

COMANCHE PEAK - UNIT 1 3/4 3-15

INSTRUMENTATION s#

SURVEILLANCE REQUIREMENTS 4.3.2.1 Each ESFAS instrumentation channel and interlodk and the automatic actuation logic and relays shall be demonstrated OPERABLE by performance of the ESFAS Instrumentat, ion Surveillance Requirements.specified in Table 4.3-2.

4.3.2.2 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function shall be demonstrated to be within the limit at least once per 18 months.

Each test shall include at least one train such that both trains are tested at least once per 36 months and one channel per function such that all channels are tested at least once per N times 18 months where N is the total number of redundant channels in a specific ESFAS function es shown in the "Total No. of Channels" column of Table 3.3-3.

e e

O COMANCHE PEAK - UNIT 1 3/4 3-16

7-___-

i TABLE 3.3-3 g ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION 1

m

~~ MINIMUM m

TOTAL NO. CHANNELS QiANNELS APPLICABLE h FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

[ 1. Safety Injection (ECCS, z Reactor Trip, Feedwater Z Isolation, Control Room e Emergency Recirculation, Emergency Diesel Generator Operation, Containment -

Vent Isolation, Station Service Water, Phase A

Isolation, Auxiliary Feed-water-Motor Driven Pump, Turbine Trip, Component R

Cooling water, Essential

" Ventilation Systems, and 4 Containment Spray Pump.

w

a. Manual Initiation 2 1 2 1,2,3,4 17

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

c. Containment 3 2 '2 1,2,3 8 14 Pressure--High-1

.d. Pressurizer 4 2 3 1,2,3 b yg a Pressure--Low

e. Steam Line Pressure--Low 3/ Steam Line 2/ Steam Line '2/ Steam Line 1,2,3 C 14" In any Steam Line ,

j%

'7

9

TABLE 3.3-3 (Continued) 8 g , ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION M

E MINIMUM m TOTAL NO. CHANNELS CHANNELS APPLICABLE: "

$E FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

[ 2. Containment Spray y a. Manual Initiation 2 pair 1. pair 2 pair 1,2,3,4 17 operated simultaneously -

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

c. Containment Pressure-- 4 2 3 1,2,3 15 g High-3 4 .

y 3. Containment Isolation ,

5 a. Phase "A" Isolation Manual Initiation

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

2) Automatic Actuation 2 1 2 1,2,3,4 13 Logic and Actuation Relays '

3) Safety Injection See Item 1. above for all' Safety Injection initiating functions and requirements.

b. Phase "B" Isolation

1) Hanual Initiation See Item 2a above. Phase "B" isola- 1,2,3,4 17 tion is manually initiated when containment spray function is manually initiated.

2) Automatic Actuation 2 1 2 1,2,3,4 13 Logic and Actuation Relays -

3) Containment 4 2 3 1,2,3 15 og Pressure--High-3 ~

%;p

TABLE 3.3-3 (Ccntinued) ~

n o ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION 3 '

M E MINIMUM .

A TOTAL NO. CHANNELS CIIANNELS APPLICABLE g FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

[ c. Containment Vent z: Isolation j

w g 1) Manual Initiation See Item 2ac and 3.a.1 above. , Containment vent isolation is inanually initiated when Phase "A" isolation function or containment spray function is manually initiated.

1, 2, 3, 4 16 ;

l

2) Automatic Actuation 2 '

1 2 1,2,3,4 16 Logic and Actuation ' '

Relays ,

R

3) Safety Injection See Item 1. above for all Safety Injection initiating functions and

~

i y requirements.

U l 4. Steam Line Isolation ,

a. Manual Initiation 1).' Individual 1/ steam line, 1/ steam line 1/ operating 1, 2, 3 ?2 steam line

2) System 2 1 2 1,2,3 21

b. Automatic Actuation 2 1 2 1,2,3 20 Logic and Actuation Relays

c. Containment Pressure-- 3 2 2 1,2,3 14' High-2 D

.;W

TABLE 3.3-3 (Continued) 8 g .

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION M

E

, MINIMUM g TOTAL NO. CHANNELS CHANNELS APPLICABLE x FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERA 3tE MODES ACTION

[ 4. Steam Line Isolation (Continued) h d. Steam Line Pressure--Low 3/ steam line 2/ steam line 2/ steam line 1,2,3 c 74 a

e. Steam Line Pressure - d a 3/ steam line 2/ steam line 2/ steam line 3 14 Negative Rate--High any steam line S. Turbine Trip and Feedwater Isolation

a. Automatic Actuation 2 1 2 1, 2 23 q Logic and Actuation

Relays

b. 3/stm. gen. a 2 Steam Generator ,2/sta. gen. 2/stm. gen. 1, 2 18 o Water Level-- in an'y oper- in each High-High (P-14) ating stm. gen.

operating stm. gen.

c. Safety Injection See Item 1 above for all safety injection initiating functions and requirements.

6. Auxiliary Feedwater

a. Manual Initiation 2 1 2 1,2,3 21

b. Automatic Actuation Logic 2 1 2 1,2,3 . 20 and Actuation Relays

c. Sta. Gen. Water Level--

Low-Ldw g

1) Start Motor- ~EX 3 Driven Pumps 4/stm. gen. 2/sta. gen. a 3/stm. gen. 1,2,3 18 !"":3s in any oper- in each /'t ating st.m. gen. operating M stm. gen.

, _ m ~

V

TABLE 3.3-3 (Centinued) 8 g ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION M

at

" ~ ~

MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE y FUNCTIONAL UNIT OF CHANNELS TO TRIP. OPERABLE MODES ACTION x

=

6. Auxiliary Feedwater (Continued)

E 2) Start Turbine-Z Driven Pump 4/sta. gen. 2/stm' gen.

. 3/stm. gen. 1, 2, 3 18 8

e in any in each 2 operating operating stm. gen. stm. gen.

d. Safety Injection Start Motor-Driven Pumps See Item 1. above for all Safety Injection initiating functions and requirements.

e. Loss-of-Offsite Power Y

Start Motor-Driven Pumps and Turbine-

{ Driven Pump 1/ train 1/ train 1/ train . 1, 2, 3 17

f. Trip of All Main Feedwater Pumps -

Start Motor-Drive'n Pumps and -

Turbine-Driven Pump 2/ pump 1/ pump 1/ pump 1, 2 17

7. Automatic Initiation of ECCS Switchover to Containment Sump

a. Automatic Actuation 2 1 2 1,2,3,4 13 Logic and Actuation Relays g

y

~.N.ca

TABLE 3.3-3 (Continued)

O ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION h

m TOTAL NO. CHANNELS MINIMUM CHANNELS APP (_ICABLE i

FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 7 7. Automatic Initiation of ECCS c Switchover to Containment z Sump (Continued)

^

[ b. RWST Level--Low-Low 4 , 2 3 1,2,3,4 18 Coircident With: Safety See Item 1. above' for all Safety Injection initiating functions Injection and requirements.

8. Loss of Power (6.9 kV Safeguards ,

System Undervoltage)

a. Preferred Offsite Source Undervoltage a

$ 1) 2)

Undervoltage Relay 2/ bus Diesel Start Timer 1/ bus 2/ bus 1/ bus 1/ bus 1/ bus 1, 2, 3, 4 1, 2, 3, 4 25 28 a

[ 3) Source Bkr Trip 1/ bus 1/ bus 1/ bus 1, 2, 3, 4 28 m Timer

b. Bus Undervoltage

1) Diesel Start a) Undervoltage 2/ bus 2/ bus 1/ bus 1, 2, 3, 4 25a Relay b) Timer - 1/ bus 1/ bus 1/ bus 1, 2, 3, 4 28*

2) Initiation of Solid State Safe-guards System -

Sequencer d

a) Undervoltage 4/ bus 2/ bus 3/ bus 1, 2, 3, 4 18a y

Relay b) Timer 4/ bus 2/ bus 3/ bus 1, 2, 3, 4 18 7":}8 j-

h TABLE 3.3-3 (Continu;d)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION E5 MINIMUM Ei TOTAL NO. CHANNELS CHANNELS' APPLICABLE g> FUNCTIONAL UNIT OF CHANNELS TO TRIP MODES ACTION O_PERABLE 2,

9. Control Room Emergency j, Recirculation z

El a. Manual Initiation 2 .1 2 All '25 s*

b. Automatic Actuation 2 1 2 1,2,3 25 Logic and Actuation Relays

c. Safety Injection See Item 1. above for all Safety Injection initiating functions and requirements Ri

10. Engineered Safety Features .

Actuation System Interlocks

a. Pressurizer Pressure, 3 2 2 1,2,3 19 P-11

b. Low-Low T yg, P-12 4 2 3 1,2,3 19

c. Reactor Trip, P-4 2 2 2 1,2,3 21

d. Steam Generator Water 3h,tm. gen. 2/sta. gen. 2/stm. gen. 1,2,3 19 Level--High-High (P-14) in any in each operating operating stm. gen. stm. gen.

11. Solid State Saf - sr !-

Sequencer (5555

a. Safety Inje- 1/ train 1/ train 1/ train 1, 2, 3, 4 13 Sequence g

b. Black Out Sequence 1/ train 1/ train 1/ train 1, 2, 3, 4 27 UE '

P M

TABLE 3.3-3 (Continued)

TABLE NOTATIONS a

The provisions of Specification 3.0.4 are not applicable.

b The [ Safety Injection] logic for this trip function may be blocked in this

MODE below the P-11 (Pressurizer Pressure Interlock) Setpoint.

c The [ Safety Injection) logic for this trip function may be blocked in this MODE below the P-12 (Low-Low T,yg Interlock) Setpoint, d

Trip function automatically blocked above P-11 and may be blocked below P-11 when Safety Injection on low steam line pressure is not blocked.

ACTION STATEMENTS ACTION 13 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTOOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ; however, one channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.2.1, provided the other channel is OPERABLE.

ACTION 14 - With the number of OPERABLE channels one less than the Total Number of Channels, operation rray proceed until performance of ,

.the next required ANALOG CHANNEL OPERATIONAL TEST provided the

. inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

ACTION 15 - With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed provided the,inoper-able channel is placed in the bypassed condition and the Minimum Channels OPERABLE requirement is met. 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 4.3.2.1.

ACTION 16 - With less than the Minimum Channels OPERABLE requirement, opera-tion may continue provided the containment purge supply and exhaust valves are maintained closed.

ACTION 17 - With the number of OPERABLE channels one less than the Minimum j Channels OPERABLE requirement, restore the' inoperable channel c to 0/ERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUT 00WN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . .

I COMANCHE PEAK - UNIT 1 3/4 3-24 1

l

TABLE 3.3-3 (Continued)

ACTION STATEMENTS (Continued) 4 ACTION 18 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provide.d 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 , and

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 of other channels per Specifica-tion 4.3.2.1.

ACTION 19 - With less than the Minimum Numbe'r of Channels OPERABLE, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months determine by observation of the associated permissive annunciator window (s) that the interlock is in its required state for the existing plant condition, or apply Specification 3.0.3.

ACTION 20 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least HOT SHUTOOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.2.1 provided the

. other channel is OPERABLE.

ACTION 21 - With the number of OPERABLE channels one .less. than th'e Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least HOT SHUTOOWN within the 'following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 22 - With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or declare the associated valve inoper-able and take.the ACTION required by Specification [3.7.1.5].

ACTION 23 - With the number of OPERABLE channels nne less than the Minimum Channels OPERABLE requirement, be in at least 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 i for surveillance testing per Specification 4.3.2.1 provided the other channel is OPERABLE.

ACTION 24 - With the number of OPERABLE channels one less than the Total Numbe~r 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 , and COW.NCHE PEAK - UNIT 1 3/' 25

un

b. The Minimum Channels OPERABLE requirement is met.

ACTION 25 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or initiate and maintain operation of the Control Room Emergency Recirculation System.

ACTION 26 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable ~ Channel

to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY -

sithin the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least H0T SkilT"^WN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 27 - With the numLer of OPERABLE Channels on ors or more trains less than the Minimum Channels OPERABLE requirement, declare the diesel generator (s) associated with the affected train (s) inoperable and apply the appropriate ACTION for Specification l 3.8.1.1.

l ACTION 28 - With less than the Minimum Channels OPERABLE, Startup and/or l Power Operation may proceed provided the timer in the affected channel is bypassed and actions are taken immediately to

~

restore the timer to OPERABLE s'.atus.

9 l

l 1

4 COMANCHE PEAK - UNIT 1 3/4 3-26 !

. - i

0 TABLE 3.3-4 8

g ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS.

M A

, SENSOR g TOTAL ERROR .

FUNCTIONAL UNIT ALLOWANCE (TA) Z_ (S) TRIP SETPOINT ALLOWABLE VALUE g 1. Safety Injection (ECCs, Reactor Trip, 3 Feedwater Isolation, Control Room

g Emergency Recirculation, Emergency Diesel Generator Operation, Contain-ment Vent Isolatioa. Station Service Water, Phase A Isosation, Auxiliary Feedwater-Motor Driven Pump, Turbine .

Trip, Component Cooling Water, Essential Ventilation Systems, and ,

Containeerit Spray Pump. .

u, D a. ~ Manual Ini?iation N.A. N.A. N.A. N.A. N.A.

o>

k b. Automatic Actuation Logic and Actuatio" 9elays N.A. .N.A. N.A. N.A. N.A.

j

c. Containmer.t. Pressure--High 1 2.5 0.71 1.5 5 3 35 psig 5 3.9 psig

d. Pressurizer Pressure--Low 16.1 14.41 1. 5 1 1829 psig 1 1823 psig

e. Steam Line Pressure--Low 17.3 14.81 1.5 1 605 psig* 1 586 psig* l l

2. Containment Spray

a. Manual Initiation N.A. N.A. N.A. N.A. N.A.

b. Automatic Actuation Logic N.A. N.A. N.A. N.A. - N. A.

and Actuation Relays Containment Pressure--liigh-3

c. 2.5 0.71 1. 5 5 18.35 psig 1 18.9 psig c.:4 A

9 c, TABLE 3.3-4 (Continued)' i

.i j ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS 5

m '

SENSOR TOTAL ERROR ~

FUNCTIONAL UNIT ALLOWANCE (TA) Z (S) TRIP SETPOINT <

_ ALLOWABLE VALUE

3. Containment Isolation C

$ a. Phase "A" Isolation

1) Manual Initiation N.A. N.A. N.A. N.A. N.A.

2) Automatic Actuation Logic N. A. N.A. N.A. N. A. N.A.

and Actuation Relays

~

3) Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and -

Allowable Values.

w D b,. Phase "B" Isolation '

w A 1) Manual Initiation See Item 2.a above. Phase "B" isolation is manualls 'nitiated when containment spray function is manually initiated. -

2) Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A.

and Actuation Relays

3) Containment Pressure-- 2.5 0.71 1. 5 5 18.35 psig 5 18.9 psig High-3

c. Containment Vent Isolation

1) Manual Initiation See Items 3.a.1'and 2.a above. Containment Vent Isolation is manually initiated when Phase "A" isolation function or containment spray function is, manually initiation. -

2) Automatic Actuation Logic N.A. N.A. N.A. N.A.

and Actuation Relays N. A'. g i

.g

3) Safety Injection See Item'1. above for all Safety Injection Trip Setpoints and Allowable Values.

@ O

TABLE 3.3-4 (Continued)

O ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS E

h m

SENSOR TOTAL ERROR FUNCTIONAL UNIT (S) TRIP SETPOINT ALLOWABLE VALUE ALLOWANCE (TA)' Z_

7 4. Steam Line Isolation E a. Manual Initiation N.A. N.A. N.A. N.A. N.A.

s b. Automatic Actuation Logic N.T N.A. N.A. N.A. N.A.

and Actuation Relays '

c. Containment Pressure--High-2 2.5 0.71 1. 5 16.35 psig 16.9 psig

d. Steam Line Pressure--Low 17.3 14.8.1 1.5 >605 psig* >586 psig*

e. Steam Line Pressure - 8.0 0.5 0] $100 psi ** 1 111.6 psi **

{ Negative Rate--High T 5. Turbine Trip and Feedwater - -

05 Isolation

a. Automatic Actuation Logic M.A.

~

N.A. N.A. N.A. N.A.

and Actuation Relays

b. Steam Generator Water 7'. 6 4.3 1.5 <82.4% of <84.2% of narrow Level--High-High (P-14) iiarrow range range instrument instrument span.

span.

c. Safety Injection See Item 1 above for all Safety Injection setpoints.and allowable values.

t:$

n.D WP

-s

. A

d TABLE 3.3-4 (Continued)-

8 g ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS 5

m SENSOR -.

, TOTAL ERROR g FUNCTIONAL UNIT ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE n

6. Auxiliary Feedwater ~

a. Manual Initiation N. A. N.A. N.A. N.A. N.A.

b. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A.

and. Actuation Relays

c. Steam Generator Water 8.8 7.08 1. 5 > 43.4% of > 42.1% of narrow Level--Low-Low narrow range range instrument instrument . span. -

span.

d. Safety Injection - Start See Item 1. above for all Safety Injection Trip Setpoints and -

y Motor Driven Pumps Allowable Values.

e. Loss-of-Of fsite Power N.A. N.A. N.A. N.A. N.A. ~

~ f. Trip of All Main Feedwater N.A. N.A. N.A. N.A. N.A.

Pumps ,

7. Automatic Initiation of ECCS Switchover to Containment Sump

a. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A.

and Actuation Relays

b. RWST Level--Low-Low [2.1] [0.71]

Coincident With

[1.2] ~> .10.6% of > 34.94% of span span Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and Allowable Values.

W W

W N

TABLE 3.3-4 (Continued) 8 g ENGINEERED SAFETY FEATURES ACTUATIONJSYSTEM INSTRUMENTATION TRIP SETPOINTS li gi SENSOR

,, TOTAL ERROR FUNCTIONAL UNIT ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE 55 8. Loss of Power (6.9 kV Safeguards

q System Undervoltage)

a. Preferred Offsite Source Undervoltage

1) Undervoltage Relay N.A. N.A N.A >4800 V >4692 V

2) Diesel Start Timer N.A. N.A. N.A. 70.75 s 70.825 s

3) Source Bkr. Trip Timer N.A. N.A. N.A. {0.5s {0.55s

~

R2 .b. Bus Undervoltage 4- .

3> 1) Diesel Start w

a) Undervoltage Relay N.A. N.A. N.A. >2100 V >1992 V b) Timer N.A. N.A. N.A. {0.75s {0,.825s

2) Initiation of Solid State Safeguards System Sequence a) Undervoltage Relays N.A. N.A. N.A. >4800 V >4692 V b) Timer N.A. N.A. N.A. 10.5 s 10.55 s ,

9. Control Room Emergency Recirculation

a. Manual Initiation N.A. . N. A. N.A. N.A. N.A.

b. Automatic Actuation Logic and N.A. N.A. N.A. N.A. N.A.

c.

Actuation Relays Safety Injection See Item 1. above for all Safety injection Trip Setpoints and g) '

er?.

15 7 Allowable Values.

'JLe, t

, s#e'l

TABLE 3.3-4 (Continued)

O ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS f ~

M I SENSOR ni TOTAL. ERROR y FUNCTIONAL UNIT ALLOWANCE (TA) Z, (S) TRIP SETPOINT ALLOWABLE VALUE 7 10. Engineered Safety. Features c Actuation System Interlocks x

a. Pressurizer Pressure, P-11 N.A. N.A. N.A. 5.[1985] psig $ [19%] psig

b. Low-Low T,yg, P-12 N.A. N.A. N.A. > [553] F $ [550.6] F &

c. Reactor Trip, P-4 N.A. N.A. N.A. N.A. N.A.

d. Steam Generator Water Level, See Item 5.~above for all Steam Generator Water Level Trip Setpoints

.. P-14 and Allowable Values.

R

11. Solid State Safeguards Sequencer N.A. N.A. N.A. N.A. N.A.

(5555)

?'

M 0

O

>p

~~n

-1 l

{*

8-TABLE 3.3-4 (Continued)

TABLE NOTATIONS

Time constants utilized in the lead-lag controller for Steam Line Pressure-Low are 13 > 50 seconds and T2 5 5 seconds. CHANNEL CALIBRATION shall ensure that these time constants are adjusted to these values.

- The time cdnstant ut'ilized'in the. rate-lag controller for Steam Line Pressure- <

. Negative Rate-High is greater than or equal to 50 seconds. CHANNEL CALIBRATION- i shall ensure that this time constant is adjusted to this value. >

4 ,

e 9

9 9

e e

i I

G COMANCHE PEAK - UNIT 1 3/4 3-33

e .

TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATION SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS

1. Manual Initiation <

i

a. Safety Injection (ECCS) N.A.

b. Containment Spray (Phase "B" Isolation N.A.

i and Containment Ventilation Isolation)

c. Phase "A" Isolation (Containment N.A.

Ventilation Isolation)

d. Steam Line Isolation N.A.

e. Feedwater Isolation (SI) N.A.

f. Auxiliary Feedwater (SI) N.A.

g. Station Service Water (SI) N.A. ,

h. Component Cooling Water (SI) N.A.

~

1. Contro1 Room'Emerg'ncy e Recirculation (SI) N.A.

j. Reactor Trip N.A.

k. Emergency Diesel Generator Operation N. A.

1. [ ] (SI) N.A.

m. Turbine Trip N.A. '

2. Containment Pressure--High-1

a. Safety Injection (ECCS) 1 27(1,5(a))/[12)(4,5b)

b. Reactor Trip i2

c. Feedwater Isolation < 6.5

d. Phase "A" Isolation 17(2)/27(1)

e. Conteinment Ventilation Isolation 1 [25)(1)/[10)(2)

f. Auxiliary Feedwater < 60

g. Station Service Water [32)(1)/[47)(2)

h. Component Cooling Water 1 [55)(1)/[40)(2)

1. [ ] N.A.

j j. Emergency Diesel Generator Operation 1 12

k. Turbine Trip N.A. ;

1. Control Room Emergency Recirculation N.A.

{

i l

COMANCHE PEAK - UNIT 1 3/4 3-34 l

- - - . .. _ _ _ _ _ _ _ _ _ _ - ~

4 TABLE 3.3-5 (Continued)

ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS

3. Pressurizer Pressure--Low ~

a. Safety. Injection (ECCS) 1 27(1,5e)/12(4,5b)

b. Reactor Trip' 12 .

c. Feedwater Isolation <7

d. Phase "A" Isolat'on [17(2)/27(1)

. e. Containment Ventilation Isolation s 5(6)

f. Auxiliary Feedwater -

< 60

g. Station Service Water , ((47](1)/[32)(2)

h. Component Cooling Water 1 [55)(1)/[40)(2)

1. [ . ] N.A.

j. Emergency Diesel Generators Operation 5 12 '

k. Turbihe Trip N.A.

1. Control Room Emergency' Recirculation N.A.

4. Steam Line Pressure--Low

a. Safety Injection (ECCS) i 22(3,5b)/12(4,5b)

b. Reactor Trip 52 c.- Feedwater Isolation < 6.5

. d. Phase "A" Isolation [17(2)/27(1)

e. Containment Ventilation Isolation s [25)(1)/[10)(2)

f. Auxiliary Feedwater < 60

g. Station Service Water [32)(2)fg47)(1)

h. Component Cooling Water 1 [55)(1)/[40)(2) 4 COMANCHE PEAK - UNIT 1 3/4 3-35

s TABLE 3.3-5 (Continued)

ENGINEERED SAFETY FEATURES RESPONSE TIMES -

INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS 4.

Steam Line Pressure--Low (Continued)

.i. [' '] N.A. -

j. Emergency Diesel Generator Operation 1 12

k. Turbine Trip N.A.

1. Control Room Emergency Recirculation ,

N.A. ,

m. Steam Line Isolation 6.5

n. [ ] N.A. '

5. Containment Pressure--High-3

a. Containment Spray Pump i 22(2)/32(1) .

b. Phase "B" Isolation 1,[65]C1)/[75](2)

^

6. Containment Pressure--High-2 Steam Line Isolation 1 6.5

7. . Steam'Line Pressure - Negative. Rate-High Steam Line Isolation. 17 i

8. Steam Generator Water Level-High-High *

a. Turbine Trip 1 [2.5]

! b. Feedwater Isolation < 11

9. Steam Generator Water Level-Low-Low

a. Motor-Driven Auxiliary i Feedwater Pumps 1 60

b. Turbine-Oriven Auxiliary Feedwater Pump 5 60 i 10. Loss-of-Offsite Power Auxiliary Feedwater 1 [60]

11. Trip of All Main Feedwater Nmps All Auxiliary Feedwater Pumps N.A.

12. RWST Level--Low-Low Coincident with Safety Injection

a. Automatic Initiation of ECCS Switchover to Containment Sump $30 ,

G COMANCHE PEAK - UNIT l' 3/4 3-36

TABLE 3.3-5 (Continueg 1

j

);U !$

ti ENGINEERED SAFETY FEATURES RESFONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS

13. Loss of Power (6.3 kV Safeguards System Undervoltage)

a. Preferred Offsite Source Undervoltage $70

b. Bus Undervoltage $10 G

4 COMANCHE PEAK - UNIT 1 3/4 3-37 l

l

k TABLE 3.3-5 (Continued)

BBLENOTATIONS (1) Diesel generator starting and sequence loading delays included.

~

(2) Diesel generator starting delay not included. Offsite power available.

(3) Diesel generator starting delay included. RHR pumps n_ot included.

(4) Diesel generator starting and sequence loading delays.not included.

RHR pumps not included.

(5) Response Time Limit includes opening of injection path valves.

Following additional time is allowed for completion of the transfer of the pump suction from the VCT to the RWST.

a) 10 seconds b) 15 seconds (6) Includes containment pressure relief line isolation only.

O COMANCHE PEAK - UNIT 1 3/4 3-38

TABLE 4.3- -

O ~

ENGINEERED SAFETY FEATURES ACTUATION SY51EM INSTRUMENTATION .

f SURVEILLANCE REQUIkfMENIS .

2 m

TRIP o

ANALOG ACTUATING MODES 5

CHANNEL'. DEVICE MASTER SLAVE FOR WHICH CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION

' RELAY RELAY SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED Z 1. Safety Injection (Reactor *

.. Trip, Feei iter Isolation, '

Control Room Emergency Recirculation, Emergency -

Diesel Generator Opera-tion, Containment Vent Isolation, Station Service '

Water, Phase A Isnlation, Auxiliary Feedwater-Motor R

Driven Pump, Turbine Trip, Component Cooling Water, Y Essential-Ventilation M Systems, and Containment Spray Pump.

a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1, 2, 3, 4

b. Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) 1,2,3,4 Logic and Actuation Q

' Relays

c. Containment Pressure- S R M N.A. N.A. N.A. N.A. 1, 2, 3 High-1

d. Pressurizer Pressure 5 R M -

N.A. N.A. N.A. N.A. 1, 2, 3 Low

e. Steam Line S R M N.A. N.A. N. A. N.A. 1, 2, 3 Pressure-Low

2. Containment Spray

a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1, 2, @

3C3

===={

W

4 TABLE 4.3-2 (Continued) 8 ENGINEERED SAFETY FEATU9ES ACTUATION SYSTEM INSTRUMENTATION j -

SURVEILLANCE REQUIREMENTS 2

m TRIP u

ANALOG ACTUATING MODES 9

CHANNEL DEVICE MASTER SLAVE FOR MfICH CHANNEL CHANNEL CHANNEL

' OPERATIONAL OPERAsiC'84L ACTUATJON RELAY RELAY SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION . TEST TEST LOGIC TEST TEST C TEST IS REQUIRED 5 b. Automatic Actuation N.A. N.A. N.A. N.A.

Logic and Actuation M(1) M(1) Q. 1, 2, 3, 4 Relays

c. Containment Pressure- S R M

~

N.A. N.A. N.A. N.A. 1, 2, 3 H61 h-3

3. Containment Isolation

a. Phase "A" Isolation R

1) Manual Initiation See item 2.a above. Phase "B" isolation manually initiated when contain- 1,2,3,4

{ ment spray function is manually initiated.

o .

2) Automatic Actuation N.A. N.A. N.A. N.A.

Logic and Actuation M(1) M(1). Q ' 1, 2, 3, 4 Relays

3) Safety: Injection See Item 1. above for all Safety Injection Surveillance Requirements,

b. Phase "B" Isolation

~

1) Manual Initiation See Item 2.a. above Phase "B" isolation is manually initiated . 1, 2, 3, 4 when containment spray function is manually initiated.

2) Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q 1,2,3,4 Logic and Actuation Relays .

3) Containment S R M N.A. N.A. N .' A . N.A. 1, 2, 3 Pressure-High-3 Z

3:=

~r m

TABLE 4.3-2 (Continued) n ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 9 - -

TRIP '

ANALOG ACTUATING MODES m CHANNEL DEVICE MASTER SLAVE FOR WHICH 3E CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY ~ RELAY SURVEILLANCE e FUNCTIONAL UNIT CHECK CALIBRATION TEST. TEST LOGIC TEST TEST TEST IS REQUIRED E c. Containment Vent Isolation G .

~ 1) Manual Initiation See Item 3.a.1 and 2.3 above. Containment vent isolation is manually 1,2,3,4 initiated when Phase "A" isolation function or containment spray function is manually initiated.

2) Automatic Actuation N.A. N.A. N.A. N.A.~

M(1) M(1) Q 1,2,3,4 Logic and Actuation Relays -

3) Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.

R*

4. Steam Line Isolation -

a. Manual Initiation N.A. N.A. N.A. R N.A. N. A.' N.A. 1,2,3

b. Automatic Actuation N.A. N.A N.A N.A. M(1) M(1). Q 1,2,3 Logic and Actuation Relays ~

c. Containment Pressure- S R M N.A. N.A. N. A. . N.A. 1, 2, 3 High-2

d. 5"team Line 5 R M N.A. N.A. N.A. N.A. 1,2,3 Pressure-Low

e. Steam Line Pressure- S R M N. A'. N.A. N.A. N.A. 3 Negative Rate-High S. Turbine Trip and Feedwater Isolation

a. Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q 1, 2 Logic and Actuation -

U Re1+ys M

'D r

W s==

TABLE 4.3-2 (Continued) 8 g ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION g SURVEILLANCE REOUIREMENIS 5

, TRIP g -

ANALOG . ACTUATING MODES ,

7 CHANNEL DEVICE MASTER 5 LAVE FOR WHICH.

CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE g FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED '

% 5. Turbine Trip and Feedwater w

)

Isolation (Continued) l b. Steam Gc.7erator Water S- R M N.A. N.A. N.A. .N. A. 1, 2 Level-High-High

c. Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.

6. Auxiliary Feedwater

$ a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1,2,3 y b. Automatic Actuation N.A. N.A N.A. N.A. M(1) M(1)' Q 1,2,3 g logic and Actuation Relays

c. Steam Generator Water 5 R M N.A. N.A. N.A N.A 1, 2, 3 Level-Low-Low

d. Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.

e. Loss-of-Of fsite Power N.A. R N.A. M N.A. N.A. N.A 1,2,3 f.TripofAliMainFeed N . A .' N.A. N.A. R N.A. N .'A . N.A 1. 2 O

. water Pumps N D

7. Automatic Initiation of ECCS Switcleover to Con-tainment Sump

a. Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q 1, 2, 3, 4 Logic and Actuation "

Relays

TABLE 4.3-2 (Continued) 8 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION j SURVEILLANCE REQUIREMENTS o

5 TRIP m ANALOG ACTUATING MODES 9

CHANNEL DEVICE MASTER SLAVE FOR WHICH CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCEi FUNCTIONAL UNIT CHECK CALIBRATION TEST

[ TEST LOGIC TEST TEST TEST IS REQUIRED 5

b. RWST Level-Low-Low S A M N.A. N.A. N.A. N.A 1, 2, 3, 4 w

Coincident With Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.

8. Loss of Power (6.9 kV Safeguards System Undervoltage) -

Preferred Offsite w a.

1 Source Undervoltage w

i

1) UnA rvoltage Relay N.A. R N.A. M N.A. N.A. N.A. 1, 2, 3, 4

2) Diesel Start Timer N.A. R N.A. M N.A. N.A. N.A. 1,2,3,4

3) Source Bkr. Trip Timer N.A. R Pl. A. M N.A. N. A. N.A. 1, 2, 3, 4

b. Bus Undervoltage

1) Diesel Start .

a) Undervoltage .

Relay N.A. R N.A. M N.A. N.A. N.A. 1,2,3,4 b) Timer N.A. R N.A. . M N.A. N.A. N.A. 1, 2, 3, 4 D

WJ 2:a M

TABLE 4.3-2 (Continued)

O ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION h

o SURVEILLANCE REQUIREMENTS E TRIP o ANALOG ACTUATING MODES 9

CHANNEL DEVICE MASTER SLAVE FOR WHICH CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE

[ FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED 5 -

2) Initiation of Solid State Safe-guards System Sequence a) Undervoltage Relay N.A. R N.A. M N.A. N.A. N.A. 1, 2, 3, 4 b) Timer N.A. R N.A. M N.A. N.A. N.A. 1, 2, 3, 4 Y 9. Control Room Emergency Recirculation

[.

E a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. All i b. Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q 1,2,3 l Logic and Actuation l Relays l

c. Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.

10. Engineered Safety Features Actuation System Interlocks

a. Pressurizer N.A. R H N.A. N.A. N.A. N.A. 1, 2,' 3 Pressure, P-11

b. Low-Low T g, P-12 N.A. R H N.A. N.A. N.A. N.A. 1, 2, 3 D

~0-LD m

-I i

TABLE 4.3-2 (Continued) 8 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION

% SURVEILLANCE REQUIREMENIS M

5 -

TRIP m ANALOG ACTUATING -

MODES 9

CHANNEL DEVICE MASTER SLAVE FOR WHICH CHANNEL CHANNEL CHANNEL OPERATIONAL ' OPERATIONAL ACTUATION RELAY

RELAY SURVEILLANCE-FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED

?5

.a

c. Reactor Trip, P-4 N.A. N.A. N.A. R_ N.A. N.A.. N.A. 1, 2, 3

d. Steam Generator 5 R M

~

a.A. M(1) M(1) Q 1, 2, 3 Water Level, P-14

11. Solid State Safeguards Sequencer (5555) ,

a. Safety Injection N.h. R N.A. N.A. M(1,2) N.A. N.A. 1, 2, 3, 4 w Sequence D

b.

~

w Blackout Sequence N.A. R N.A. N.A. M(1,2) N.A. N.A. 1, 2, 3, 4 h

~

TABLE NOTATION .

(1) Each train shall be tested at least every 62 days on a STAGGERED TEST BA;IS. c ~

(2) Performed by Solid State Safeguards Sequencer Automatic Test. g

$D

INSTRUMENTATION M4.3.3 MONITORING INSTRUMENTATION RADIATION MONITORING FOR PLANT OPERATIONS LIMITING CONDITION FOR OPERATION .

3.3.3.1 The radiation monitoring instrumentation channels for plant operations shown in Table 3.3-6 shall be OPERABLE with their Alarm / Trip Setpoints within the specified limits.

APPLICABILITY: As shown in Table 3.3-6.

ACTION:

a. With a radiation monitoring channel Alarm / Trip Setpoint for plant operations exceeding the value shown in Table 3.3-6, adjust the Setpoint to within the limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or declare the channel.

l inoperable, '

l ,

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

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

SURVEILLANCE REQUIREMENTS I

4.3.3.1 Each radiation monitoring instrumentation channel for plant operations '

shall be demonstrated OPERABLE:

a. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by performance of a CHANNEL CHECK,

{ b. At least once per 18 months by performance of a CHANNEL CALIBRATION,

c. At least once per 31 days by performance of a O!GITAL CHANNEL OPERATIONAL TEST.

l l

COMANCHE PEAK - UNIT I 3/4 3-46

TABLE 3.3-6

! RA01ATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS 5

1

'" ~

MINIMUM iE CHANNELS CHANNELS APPLICABLE ALARM / TRIP

$E FUNCTIOM4L UNIT TO TRIP / ALARM OPERABLE : MODES SETPOINT ACTION -

a gg 1. RCS Leakage Detection x

~#

a. Particulate Radioactivity N.A. I 1, 2, 3, 4 N.A. 32

b. Gaseous Radioactivity N.A. I 1,2,3,4 M.A. 32 l 2. Containment Ventilation Isolation

a. Particulate Radioactivity 1 2 All

29 l b. Gaseous Radioactivity 1 -

2 All

29 l w .

l 32 3. Fuel Storage Pool Areas l us -

I a.. Criticality-Radiation Level 1 2 ** < 15 mR/h 31 gh

4. Control Room Emergency Recirculation

a. Air Intake-Radiation Level 1/ intake 2/ intake .All [Later] 30 pCi/ml

b. Plant Vent Radioactivity-High 1 1 All *** 30

. ::E:s czy L =1

<= 4

TABLE 3.3-6 (Continued)

TABLE NOTATIONS Must satisfy Specification 3.11.2.1 requirements.

With fuel in the fuel storage poo) areas or fuel building.

Must satisfy Specification 3.11.2.1 requirements.

ACTION STATEMENTS ACTION 29 - With less than the Minimum Channels OPERABLE requirement, operation may continue provided the containment ventilation valves are maintained closed. The containment pressure relief valves may only be opened in compliance with Specification 3.6.1.7 and 3.3.3.11.

ACTION 30 - With the number of OPERABLE channels one less than the Minimum

.. Channels OPERABLE requirement, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is~olate the Control Room Emergency Ventilation System and initiate operation .

of the Control Room Emergency Ventilation System in the recirculation mode.

ACTION 31 - With less than the Minimum Channels OPERABLE requirement, opera-tion may continue for up to 30 days provided an appropriate portable continuous monitar with the same Alarm Setpoint is provided in the fuel storage pool area. Restore the inoperable monitors to OPERABLE status within 30 days or sussend all .

i operations involving fuel movement in the fuel storage pool areas.

ACTION 32 - With less than the Minimum Channels OPERACLE requirement, comply with the ACTION requirements of Specification 3.4.6.1.

4 i

l l

COMANCHE PEAK - UNIT 1 3/4 3 48 i

INSTRUMENTATION b MOVABLE INCORE DETECTORS LIMITING CONDITION FOR OPERATION 3.3.3.2 The Movable Incore Detection System shall be OPERABLE with:

a. At least 75% of the detector thimbles,

b. A minimum of two detector thimbles per core quadrant, and

c. Sufficient mov'ble detectors, drive, and readout equipment to map these thimbles.

APPLICABILITY: When the Movable Incore Detection System is used for:

a. Recalibration of the Excore Neutron Flux Detection System, or

^

b. MonitoringtheQUADRANTPOWERTILTRNTIO,or 4

N

c. Measurement of F3g, pq(Z) and F,y, ACTION:

J 4 With the Movable Incore Detection System inoperable, do not use the system for the above applicable monitoring or calibration functions. The provisions of -

Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.2 The Movable Incore Detection System shall be demonstrated OPERABLE at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by normalizing each detector output when required for:

a. Recalibration of the Excore Neutron Flux Detection System, or

b. Monitoring the QUADRANT POWER TILT RATIO, or N

c. Measurement of F3g, pq(Z) and F,y, 5

4 j

COMANCHE PEAK - UNIT 1 3/4 3-49

INSTRUMENTATION j SEISMIC INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.3 The seismic monitoring instrumentation shown in Table 3.3-7 shall be OPERABLE.

APPLICABILITY: At all times.

(

. 1 ACTION:

a. With one or more of the above required seismic monitoring instruments I inoperable for more than 30 days, prepare and submit a Special l Report to the Commission pursuant to Specification 6.9.2 within the l next 10 days outlining the cause of the malfunction and the plans for restoring the instrument (s) to OPERABLE status,

b. The provisions of Specifications 3.0.3 erd 3.0.4 =ca not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.3.1 Each of the above required seismic monitoring instruments shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALI-BRATION, and ANALOG CHANNEL OPERATIONAL TEST at the frequencies shown in Table 4.3-4.

4.3.3.3.2 Each of the above required seismic monitoring instruments actuated during a seismic event greater than or equal to 0.019 shall be restored to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and a CHANNEL CALIBRATION performed within 10 days following the seismic event. Data shall be retrieved frnm actuated instruments and analyzed to determine the magnitude of the vibratory ground motion. A Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 14 days describing the magnitude, fre-quency spectrum, and resultant effect upon facility features important to safety.

l l

C0HANCHE PEAK - UNIT 1 3/4 3-50

TABLE ~3.3 7 1 [*

SEISMIC MONITORING INSTRUMENTATION HINIMUM INSTRUMENTS INSTRUMENTS AND SENSOR LOCATIONS '

OPERABLE

1. Triaxial Time-History Accelerographs J

a. Accelerometer-Fuel Building 1

b. Accelerometer-Containment 1 .

c. Accelerometer-Electrical Manhole 1

d. Seismic Trigger-Fuel Building i

e. Recorder Unit, SMA-3 1

f. Playback Unit, SMP-1 1

2. Triaxial Peak Accelerographs

a. Pressurizer Lifting Trunion 1 9

b. Reactor Coolant Piping 1-

c. Component. Cooling Water Heat Exchanger 1

3. Triaxial Seismic Switch Fuel Building 1*

4. Triaxial Response-Spectrum ~ Recorders

a. Fuel Building 1 I

b. Reactor Bldg. Internal Structure 1

c. Safaguards Building 1 1

5. Respense Spectrum Annunciator l' "With control room indication.

COMANCHE PEAK - UNIT 1 3/4 3-51

t TABLE 4.3-3 N SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS ANALOG

CHANNEL CHANNEL CHANNEL OPERATIONAL INSTRUMENTS AND SENSOR LOCATIONS CHECK CALIBRATION TEST

1. Triaxial Time-History Accelerographs

a. Accelerometer-Fuel Building M R SA

b. Accelerometer-Containment M R SA

c. Accelerometer-Eledtrical Manhole M R SA

d. Saismic Trigger-Fuel Building M R SA

e. Recorder Unit, SMA-3 M R SA

f. Playback Joit, SME-1 M R SA

2. Triaxial Peak 4ccelerographs

a. Pressarizer Lifting Trunion- N.A. R N.A.

b. Rer.ctor Coolant Pipin'g N.A. R N.A.

. c. ' Component Cooling Water Heat N.A. R N.A.

Exchanger

3. Triaxial Seismic Switch Fuel Building ** M R SA

4. Triaxial Response-Spectrum Recorders

a. Fuel Building N.A. R N.A.

b. Reactor Bldg. Internal Structure N.A. R N.A.

c. Safeguards Building N.A. R N.A.

5. Response Spectrum Annunciator ** M R SA

Setpoint verification is not applicable.

- With control room in'dication.

COMANCHE PEAK - UNIT 1 3/4 3-52

3; INSTRUMENTATION :

METEOROLOGICAL INSTkUMENTATION LIMITING CONDITION FOR OPERAT'ON '

~

4

~

3.3.3.4 The meteorologit:a1 monitoring instrumentation channels shown in Table 3.3-8 shall be OPERABLE.

APPLICABILITY: At all times.

ACTION:

a. Wit'h one or more required meteorological monitoring channels inoperable for more than 7 days, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 10 days ,

outlining the cause of the malfunction and the plans for restoring '

the channel (s) to OPERABLE status,

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

. k

~

SURVEILLANCE REQUIREMENTS

.i 4.3.3.4 Each of the above meteorological monitoring instrumentation channels [

shall be demonstrated OPERABLE i

~

a. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by performance of a CHANNEL CHECK, and i

b. At least once per 184 days by performance of a CHANNEL CALIBRATION. i l

I i

e I

I COMANCHE PEAK - UNIT 1 3/4 3-53 I i

a TABLE 3.3-8 1 ;

J t .:... ,

METEOROLOGICAL MONITORING INSTRUMENTATION MINIMUM

. INSTRUMENT , . LOCATION OPERABLE -

1. WIND SPl!ED

~

1 of 3

a. Y.-SY-4117 Nominal Elev. 60 m.

's . X-SY-4118 Nominal Elev. 10 m.

c. X-SY-4128* Nominal Elev. 10 m.

. 2. WIND DIRECTION 1 of 3

a. X-ZY-4113 Nominal Elev. 60 m.

b. X-ZY.-4116 Nominal Elev. 10 m. ~

c. X-ZY-4126* Nominal Elev. 10 m.

4

3. AIR TEMPERATURE - AT 1 of 2

a. X-TY-4119 Nominal Elev. 60 m..and

. Nominal Elev. 10 m.

b. X-TY-4120 Nominal Elev. 60 m. and Nominal Elev. 10 m.

Mounted on backup tower.

COMANCHE PEAK - UNIT 1 3/4 3-54

INSTRUMENTATION p, t

.(EMOTE SHUTDOWN SYSTEM INSTRUMENTATION kJ

. LIMITING CONDITION FOR OPERATION _

3.3.3.5 The. Remote Shutdown System transfer switches, power, controls and ~

monitoring instrumentation channels shown in Table 3.3-9 shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

a. With the number of OPERABLE remote shutdown monitoring channels less than the Minimum Channels OPERABLE as required by Table 3.3-9, restore the inoperable channel (s) to OPERABLE status within 7 days, or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. With the ni.cber of OPERABLE remoto shutdown monitoring channels less than the Total Number of Channels as required by Table 3.3-9, within 60 days restore the inoperable chennel(s) to OPERABLE status or, pursuant to Specification 6.9.2, submit a Special Report that defines the corrective action to be taKen.

c. With one or more Remote Shutdown-System transfer switches, p.ower,.

or control circuits inoperable, restore the inop'erable switch (s)/

circuit (s) to OPERABLE status within 7 days, or.5e in HOT STAN0BY within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

d. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.5.1 Each remote shutdown monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-6.

l l 4.3.3.5.2 Each Remote Shutdown System transfer switch, power and control circuit shall be demonstrated OPERABLE at least once per 18 months by verifying its capability to perform its intended function (s).

l l

l COMANCHE PEAK - UNIT 1 3/4 3-55 1' ._ .

j TABLE 3.3-9 DWT REMOTE SHUT 00WN SYSTEM MONITORING INSTRUMENTATION TOTAL N0. MINIMUM READ 0UT OF CHANNELS INSTRUMENT LOCATION CHANNELS OPERABLE

1. Neutron.F1'ux Monitors HSP 2 l'

2. Wide Range RCS Temp. - T ~ HSP 1/ Loop 1/ Loop c

3. Wide D nge RCS Temp. - T HSP 1/ Loop 1/ Loop h

4. Pressurizer Pressure HSP 1 1

5. Pressurizer Level HSP 2 1

6. Steam Generator Pressure HSP 1/SG 1/SG

7. Steam Generator Level HSP 1/SG 1/SG

8. Auxiliary Feedwater Flow HSP 2/SG 1/SG Rate .to. Steam Generator

9. : Condensate. Storage Tank Level HSP 2 1

10. Charging Pump to CVCS HSP 1 1 Charging and RCP Seals -

Flow Indication SWITCH TRANSFER SWITCFiS [Illustrntional only) LOCATION

1. Auxiliary Feedwater Control

2. Safe Shutdown Equipment Power

a. Auxiliary Feedwater

b. Charging

c. Pressurizer Heaters

d. Valves

3. CVCS Makeup Flow Control

4. Diesel Generator Control

5. Electrical Distribution System Control .

SWITCH CONTROL CIRCUITS [Illustrational only] LOCATION

1. Auxiliary Feedwater Flow

2. Pressurizer Heaters

3. CVCS Hakeup Flow

4. Diesel Generator -

5. Electrical Distribution System .

HSP = Hot Shutdown Panel SG = Steam Generator COMANCHE PEAK - UNIT 1 3/4 3-56

TABLE 4.3-4 9 REMOTE SHUTDOWN MONITORING INSTRUMENTATION h SURVEILLANCE REQU2REMENTS -

9 ni CHANNEL CHANNEL m INSTRUMENT CHECK CALIBRATTON 9

7 1. Power Range Neutron Flux M Q E 2. Intermediate Range Neutron Flux M N.A.

e 3. Source Range M2utron Flux M N.A.

4. Reactor Trip Breaker Indication ,

M H.A.

S. Reactor Coolant Temperature - Average M R

6. Re, actor Coolant Flow Rate M- R

{ 7. Pressurizer Pressure M R .

8. Pressurizer Level M. R

9. Steam Generator Pressure M R

10. Steam Generator Water Level M R ,

11. Control Rod Position Limit Switches M R

12. RHR Flow Rate M R

13. RHR Temperature M R 1

14. Auxiliary Feedwater Flow Rate M R C l W

2:=

"M

====3

DD INSTRUMENTATION y aitz \

ACCIDENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.6 The accident monitoring instrumentation channels shown in Table 3.2-10 shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

a.

With the number of OPERABLE accident monitoring instrumentation channels less than the Total Number of Channels shown in

-Table 3.3-10, restore the inoperable channel (s) to OPERABLE status within 7 days, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least HOT SHUTOOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b.

I With the number of.0PERABLE accident monitoring instrumentation I

channels except the unit vent-high range. noble gas monitor, and the -

steam relief-high range radiation monitor, less than the Minimum

' Channels OPERABLE reauirements of Table 3.3-10, restore the inoper-able chan~nel(s) to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be.in at least HOT STANOBY within within th'e following the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least HOT SHUTDOWN 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c.

With the number of OPERABLE channels for the unit vent-high range nobic gas monitor, or the steam relief-high range radiation monitor or the containment atmosphere-high range radiation monitor, or the reactor coolant radiation level monitor less than required by the l

Minimum Channels OPERABLE requirements, initiate an alternate method of monitoring the appropriate parameter (s), within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , and either restore the inoperable channel (s) to OPERABLE status within

! 7 days or prepare and submit a Special Report to the Commission, pur-suant to Specification 6.9.2, within 14 days that provides actions l

taken, cause of the inoperability, and the plans and schedule for restoring the channels to OPERABLE status.

I d.

The provisions of Specification 3.0.4 are not applicable.

l D

C0HANCHE PEAK - UNIT 1 3/4 3-58

INSTRUMENTATION

,. g ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS' 4.3.3.6 Each accident monitoring instrumentation channel shall b'e demonstrated OPERABLE:

. a. At least once per 31 days by performance of a CHANNEL CHECK, and

b. At least once per 18 months by performance of a CHANNEL CALIBRATION.'*

Containment Area Radiation (High Range) CHANNEL CALIBRATION may consist of an electronic calibration of the channel, not including the detector, for range decades above 10 R/h and a one point calibration check of the detect 6r below 10 R/h with an installed or portable gamma source.

COMANCHE PEAK - UNIT 1 3/4 3-59

TABLE 3.3-10 0 A_CCIDENT MONITORING' INSTRUMENTATION E

h ni TOTAL MINIMUM NO. OF CHANNELS INSTRUMENT CHANNELS QPERABLE 7 1. Containment Pressure (Wide Range) 2 1

2. Containment Pressure (Narrow Range) 2 1

'w 3. Reactor Coolant Outlet Temperature - TH0T (Wide Range) 2 1

4. Reactor Coolant Inlet Temperature - TCOLO (Wide Range)~ 2 1 S. Reactor Coolant Pressure - Wide Range 2 1

6. Pressurizer Water Level 2 1 .

M 7. Steam Generator Water Level - Wide Range and Auxiliary 1/ steam generator 1/ steam generator

[ Feedwater Flow (Sec0ndary Coolant Availability) 8 8. Steam Generator Water Level - Narrow Range 1/ steam generator 1/ steam generator

9. Containment Water Level (Wide Range) 2' 1

10. Core Exit Temperature (Thermocouples) 4/ core quadrant 2/ core quadrant

11. Steam Relief Vent - Noble Gas Monitors N.A. 4 .

12. Containment Area Radiation (High Range) 2 1

13. Reactor Vessel Water Level -

2 1 1

P e

A

TABLE 3.3-10 (Centinued) 8 g ACCIDENT MONITORING INSTRUMENTATION M

g -

TOTAL MI:4IMUM NO. OF CHANNELS INSTRUMENT CHANNELS OPERABLE

14. Neutron Flux (Source Rarige) 2 1 E

q 15. Neutron Flux (Intermediate Range) 2. 1

16. Condensate Storage Tank Level 2 (2 Sensors / Channel) 1 (1 Sensor / Channel)

17. Steam Line Pressure 2/ steam generator 1/ steam generator

18. Refueling Water Storage Tank Water Level 2

.. 1

19. Reactor Coolant System Subcooling Margin Monitor. 2 1 R*

~

20. Plant Vent Stack - Mobile Gas Monitors w ^

h a. Intermediate Range N.A. 1

b. High Range N.A. 1 W w M

O

f}s \T INSTRUMENTATION pih g

CHLORINE DETECTION SYSTEMS -

LIMITIN'G CONDITION FOR OPERATI'ON 3.3.3.7 Two independent Chlorine Detection Systems for each fresh air intake, with their Alarm / Trip Setpoints adjusted to actuate at a chlorine concentra-tion of less than or equal to 5 ppm, shall be OPERABLE.

APPLICdBILITY: All MODES.

ACTION:

a. With one Chlorine Detection System at a fresh air intake inoperable, restore the inoperable system to OPERABLE status within 7 days or within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months isolate the affected fresh air intake and comply with the provisions of Specification 3.7.7 and either (1) operate the Control Room HVAC System from the unaffected fresh air intake or (2) initiate and maintain operation of the Control Room HVAC System in the isolation mode of operation.

b. With both Chlorine Detection Systems at a fresh air intake inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months isolate the affected fresh air intake and comply with the' provisions of Specification 3.7.7 anJ either .

. (1) operate the Control Room HVAC System from ths, unaffected fresh air intake or (2) initiate and maintain operation of the Control Room HVAC System in the isolation mode of operati,on.-

c. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.7 Each Chlorine Detection System shall be demonstrated OPERABLE:

l l a. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by performance of a CHANNEL CHECK,

b. At least once per 31 days by verifying alarm and trip relay actuation when each channel is tested using installed test circuitry, and

c. At les;L once per 18 months by performance of a CHANNEL CALIBRATION.

O COMANCHE PEAK - UNIT 1 3/4 3-62

rw n r' INSTRUMENTATION LOOSE-PART DETECTION SYSTEM LIMITING CONDITION FOR OPERATION 3.3.3.8 The Loose-Part Detection System ~shall be OPERABLE.

APPLICABILITY: MODES 1 and 2.

ACTION:

a. With one or more Loose-Part Detection System channels inoperable for more than 30 days, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 10 days outlining the cause of the malfunction and the plans for restoring the channel (s) to OPERABLE status,

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

SURVEILLANCE R QUIREMENTS l

4.3.3.8 Ea'ch channel of'the Loose-Part Detection Systems shall be demonstrated i

I OPERABLE by performance of:

a. A CHANNEL CHECK at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , -

b. An ANALOG CHANNEL OPERATIONAL TEST at least once per 31 days, and l

l

c. A CHANNEL CALIBRATION at least once per 18 months.

1 COMANCHE PEAK - UNIT 1 3/4 3-63 l

l

l t

INSTRUMENTATION id 3 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.9 'The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.3-11 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 Setpnints of these channels shall be determined and adjusted in acco-dance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (0DCM). -

APPLICABILITY: At all times.

ACTION:

a. With a radioactive liquid effluent monitoring instrumentation channel l Alarm / Trip Setpoint less conservative than required by the above l

. specification, immediately suspend the release of radioactive liquid effluents monitored by the affected channel, or declare the channel inoperable,

b. With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, take'the ACTION shown in Table 3.3-11. . Restore the inoperable instrumentatio,n to OPERABLE status within 30 days and, if unsuccessful, explain in the next Semiannual Radioactive Effluent Release Report pursuant to Specifi-cation 6.9.1.4 why this inoperability was not corrected in a timely manner.

c. The provisions of Specifications 3.0.2 and 3.0.4, are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.9 Each radioactive liquid effluent monitoring instrume.ntation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and DIGITAL CHANNEL OPERATIONAL TEST or ANALOG l CHANNEL OPERATIONAL TEST at the frequencies shown in Table 4.3-5.

i I

i C0HANCHE PEAK - UNIT 1 3/4 3-64 1

s TABLE 3.3-11 o

RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION f

M E MINIMUM .

A CHANNELS y INSTRUMENT. . OPERABLE ACTION

[ 1. Radioactivity Monito -s Providing Alarm and 5 Automatic Termination of Release

--e w a. Liquid N deste Effluent Line 1 33

b. Turbine Building (Floor Drains) Sumps Effluent Line 1 - 34

2. Radioactivity Mon'itors Providing Alarm But Mot Providing Automatic Termination of Release Service Water System Effluent Line 1/ train 35 1 3.

, Flow Rate Measurement Devices ,

O Liquid Radwaste Effluent Line 1 36 FC8

=

--f1

,A .

TABLE 3.3-11 (Continued)

ACTION STATEMENTS 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 that prior.tp initiating a -

release:

a. At least two independent samples are analyzed in accordance with Specification 4.11.1.1.1, and

b. At least tw'o technically qualified members of the facility staff independently verify the release rate calculations and discharge valve lineup.

Otherwise, suspend relcase of radioactive effluents via this pathway.

ACTION 34 - 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 radioactivity at a lower limit of detection of no more than 10 7 microcurie /ml:

a. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the specific activity of

. the secondary coolant is greater than 0.01 microcurie /gran DOSE EQUIVALENT I-131, or

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 microcurie / gram 003E EQUIVALENT I-131.

ACTION 35 - With the number of channels OPERABLE less than required by the l Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that, at least once per i

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , grab samples are collected and analyzed for radio-activity at a lower limit of detection of no more than 10 7 microcurie /ml.

ACTION 36 - With the number of channels OPERABLE less than required by the '

Minimum Channels OPERABLE requirement, effluent releases via l 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 perfor-

! mance curves generated in pl. ace may be used to estimate flow.

t 1

l COMANCHE PEAK - UNIl 3/4 3-66

TABLE 4.3-5 .

E3 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS l 3E M

Ri -

DIGITAL ANALOG

,2 CHANNEL CHANNEL S2 CHANNEL SOURCE CHANNEL OPERATIONAL OPERATIONAL

'[ INSTRUMENT CHECK CHECK CALIBRATION TEST TEST EE 1. Radioactivity Monitors Providing 13 Alarm and Automatic Termination e, of Release

a. Liquid Radwaste Effluent Line 'D P R(4) Q(1) N.A.

b. Turbine Building'(Floor Drains) Sumps Effluent Line D M R(4) Q(2) N.A.

2. Radioactivity Monitors Providing Alarm But t' Not Providing Automatic Termination, of Release T ~

C0 Service Water System Effluent Line D M R(4) Q(3) N.A.

3. Flow Rate Measurement Devices Liquid Radwaste Effluent Line 0(S) N.A. R , N.A. Q 6

m3 p

a

1

TABLE 4.3-5 (Continued)

TABLE NOTATIONS (1) The DIGITAL CHANNEL OPERATIONAL TEST shall also demonstrate that automatic

isolation of this pathway and control room alarm annunciation occur if any of the following conditions exists:

a. Instrument indicates measured levels above the Alarm / Trip Setpoint, or

b. Circuit failure (Channel Out of. Service - Loss of Power, Loss of Counts, Loss of Flow, or Check Source Failure), or

c. Instrument indicates a downscale failure, or

d. Instrument controls not set in operate mode.

l (2) The DIGITAL CHANNEL OPERATIONAL TEST shall also demonstrate that automatic i

flow diversion of this pathyway (from the Low Volume Waste Treatment System to the Co-Current Waste Treatment Syste.m) and Control Room alarm annunciation occur if any of the followin'g conditions exist:

a. Instrument indicates measured levels above the Alarm / Trip Setpoint, or

b. ' Circuit failure (Channel Out of Service - Loss of Power, Loss of Counts, Loss of Sample Flow, or Check Source Failed).

c. Instrument indicates a down' scale. failure, or

d. Instrument controls hot set in operate mode.

(3) The DIGITAL CHANNEL OPERATIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exists:

a. Instrument indicates measured levels above the Alarm Setpoint, or

b. Circuit failure (Channel Out of Service - Loss of Power, Loss of Counts, Loss of Flow or Check Source Failure), or

c. Instrument indicates a downscale failure, or

~

d. Instrument controls not set in operate mcds I

(4) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards (NBS) or using standards that have been obtained from suppliers that participate l in measurement assurance activities with NBS. These standards shall permit l calibrating the system over its intended range of energy and measurement

. range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration, reference standards certified by NBS, or standards that have been obtained from suppliers that participate in measurement assurance activities with NBS shall be used.

i COMANCHE PEAK - UNIT 1 3/4 3-68 i

TABLE 4.3-5 (Continued)

TABLE NOTATIONS (Continued) 9 (5) 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. -

e n

1 1

l f

L a

l i

i i

6 e

l-COMANCHE PEAK - UNIT 1 3/4 3-69 l -- _ - . - _ _ _ _ . . _ . . _ ,_ _ _ . _

INSTRUMENTATION ilS RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.10 The radioactive gaseous 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 Specifications 3.11.2.1 and 3.11.2.5 are not exceeded.

The Alarm / Trip Setpoints of these channels meeting Specification 3.11.2.1 shall be determined and adjusted in accordance with the methodology and parameters in the 00CM.

APPLICABILITY: As shown in Table 3.3-12

' ACTION:

a. With a radioactive gaseous effluent monitoring instrumentation channel Alarm / Trip Setpoint less conservative than required by the above specification, inkediately suspend the release of radioactive-gaseous effluents monitored by the affected channel, or declare the channel inoperable.

b. With less than the minimum number of radioactive gase~ous e'ffluent monitoring instrumentation channels OPERABLE, take the ACTIONJshown' in Table 3.3-12. Restore the inoperable instrunentation to OPERABLE status within 30 days and, if unsuccessful explain in the'next' Semi- ,

annual Radioactive Effluent Release Report pursuant to Specifica-tion 6.9.1.4 why this,inoperability was not corrected in a timely manner.

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

SURVEILLANCE REQUIREMENTS 4.3.3.10 Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHELK, SOURCE

. CHECK, CHANNEL CALIBRATION and DIGITAL CHANNEL OPERATIONAL TF5f or ANALOG CHANNEL OPERATIONAL TEST at the frequencies shown in Table 4.3-6.

e C0HANCHE PEAK - UNIT 1 3/4 3-70 t

1 TABLE 3.3-12 O RADI0 ACTIVE GASEOUS EFFLUENT Mt'IT0' RING INSTRUMENTATION f

M 5 ~

MINIMUM CHANNELS INSTRUMENT OPERABLE APPLICABILITY ACTION .

7 1. WASTE GAS HOLOUP SYSTEM ,

E a. Noble Gas Activity Monitor -

Z Providing Alarm and Autematic w Termination of Release 1/ stack *** 37

2. WASTE GAS HOLDUP SYSTEM Explosive Gas Monitoring System

a. Hydrogen Monitorc 2 ** 40, 42

b. Oxygen Monitors 2 **

40

3. Primary Plant Ventilation U a. Noble Gas Activity Monitor 1/ stack-

39

b. Iodine Sampler 1/ stack

41

c. Particulate Sampler 1/ stack

41

d. Flow flate Measuring Devi.ce 1/ stack

38

.e. Sampler Flow Rate Monitor 1/ stack

38 d

W :w.

O e

TABLE 3.3-12 (Continued) r' TABLE NOTATIONS *J'

At all times.

'** During WASTE GAS H0 LOUP. SYSTEM operation.

- - During Batch Radioactive Releases via this pathway.

ACTION STATEMENTS ACTION 37 - With the number of channels OPERABLE less than required by the Minimum Channels GPERABLE requirement, the contents of the tank (s) may be released to the environment provided that prior to initiating the release:

a. The auxiliary building vent duct monitor is confirmed OPERABLE, or

b. 'At least two independent samples of the tank's contents

'are analyzed, and

c. At least two technically qualified membars of the facility '

staff independently verify the release rate calculations and discharge valve lineup. ,

Otherwise, suspend ral?ase of radioactive eff?.!ents via this pathway.

ACTION 38 -

With the number of channels OPERABLE fess than required by the Minimum Channels OPERABLE requirement, effluent releases via this psthway 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 39 - 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 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and these samples are analyzed for radioactivity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 40 - With the number of channnels OPERABLE one less than required.by the Minimum Channels OPERABLE requirement, operation of this system may continue provided grab samples are taken and analyzed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . With both channels inoperable, operation may continue provided grab samples are taken and analyzed j at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during degassing operations and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months during other operations.

ACTION 41 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via the affected pathway may continue provided samples are contin-uously collected with auxiliary sampling equipment as required in Table 4.11-2.

ACTION 42 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, suspend oxygen supply l to the recombiner.

i i COMANCHE PEAK - UNIT 1 3/4 3-72 l l

TABLE ~4.3-6 O

$ RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS M

z r3 DIGITAL . ANALOG E CHANNEL SOURCE CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL e

INSTRUMENT CHECK CHECK c CALIBRATION TEST TEST 3 1. WASTE GAS HOLDUP SYSTEM

a. Noble Gas Activity Monitor -

Providing Alarm and Automatic

. Termination of Release P P R(3). Q(1) N.A.

2. WASTE GAS HOLDUP SYSTEM Explosive Gas Monitoring System

a. Hydrogen Monitors D N.A. Q(4) N.A. M w b. 0xygen Monitors D N.A. Q(4) N.A.

M

3. Primary Plant Ventilation

a. Noble Gas Activity Monitor D M# R(3) Q(2) N.A.

b. Iodine Sampler W'5) N A. N.A. N.A. N.A.

c. Particulate Sampler W(5) N.A. N.A. 'N.A. N.A.

d. Flow Rate Measuring Des; ice D N.A. R N.A. Q

~

e. Sampler Flow Rate Monitor D M.A. R N.A. -Q LW "EQ LP

~

M

'l

. 1 TABLE 4.3-6 (Continued)

TABLE NOTATIONS

Also prior to any release from the waste gas holdup s'ystem or containment purging or vent.

(1) The DIGITAL' CHANNEL OPERATIONAL TEST shall also demonstra'te that automatic isolation of this pathway and control room alarm annunciation occurs if any of the following conditions exists:

a. Instrument indicates measured levels above the Alarm / Trip Setpoint, or

b. Circuit failure (Channel Out of Service - Loss of Power, Loss of Counts, Loss of Sample-Flow, or Check Source Failure), or

c. Instrument indicates a downscale failure, or

d. Instrument controls not set in operate mode.

~

(2) The DIGITAL CHANNEL 0PERATIO4AL TEST sha11 also demonstrate that control room' alarm a'nn'unciation occurs if any of the following conditions exist;.

a. Instrument indicates measured levels above the Alarm Setpoint, or

b. Circuit failure (Channel Out of Service - Loss of Power, Loss of Counts, Loss of Sample-Flow, or Check Source Failure), or

c. InstrumentindicatesadoEnscalefailure,or

d. Instrument controls not set in operate mode.

(3) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards (NBS) or using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS. These standards shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration, reference standards certified by NBS, or standards that have been obtained from suppliers that participate in measurement assurance activities with NBS shall be used.

(4) The CHANNEL CALIBRATION shall include the use of standard gas samples

containing a nominal:

a. One volume percent hydrogen, balance nitrogen, and

b. Four volume percent hydrogen, balance nitrogen.

(5) The Channel Check shall consist of visually verifying that the collection element (i.e., filter or cartridge, etc.) is in place fer sampling.

4 COMANCHE PEAK - UNIT 1 3/4 3-74

- - - , _ . g.--

INSTRUMENTATION .

3/4.3.4 TURBINE OVERSPEED-PROTECTION LIMITING CONDITION FOR OPERATION 3.3.4 At least one Turbine Overspeed Protection System shall be OPER'ABLE.

APPICABILITY: MODES 1, 2, and 3.

ACTION:

a. With one stop valve or one governor valve per high pressure turbine steam line inoperable and/or with one reheat stop valve or one reheat intercept valve per low pressure turbine steam line inoperable, restore the inoperable valve (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 line(s) or 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 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 overspeed protection system shall be demonstrated OPERABLE:

a. At least once per 14 days by cycling each of the following valves through at least one complete cycle from the running position using

the manual test or Automatic Turbine Tester (ATT):

1) Four high pressure turbine stop valves,

2) Four high pressure turbine control valves,

3) Six low pressure turbine stop valves, and

4) Six low pressure turbine control valves,

b. At least once per 14 days by testing of the two mechanical overspeed devices using the Automatic Turbine Tester or manual test.

c. At least once per 31 days by direct observation of the movement of each of the above valves through one complete cycle from the running position.

d. At least once 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 stem's and verifying no unacceptable flaws.

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

COMANCHE PEAK - UNIT 1 3/4 3-75

3/4.4 REACTOR C00L4NT SYSTEM >; t 3/4.4.1- REACTOR. COOLANT LOOPS AND COOLANT CIRCULATION STARTUP r.hD POWER OPERATION LIMITING CONDITION FOR OPERATION 3.4.1.1 All reactor coolant loops shall be in operation.

APPLICABILITY: MODES 1 and 2.

ACTION:

With less than the above required reactor coolant loops in operation, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.4.1.1 The above required reactor coolant loops shall be verified in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . ,

t COMANCHE PEAK - UNIT 1 3/4 4-1

, REACTOR COOLANT SYSTEM HOT STANDBY #-

LIMITING CONDITION FOR OPERATION 3.4.1.2 At least two of the reactor cool, ant loops listed below shall be OPERABLE with at least two reactor c'oolant loops in operation when the reactor trip breakers are closed and at least one reactor coolant loop in operation when the reactor trip breakers are open:*

a .' Reactor Coolant Loop 1 and its associated steam generator and reactor coolant pump,

b. Reactor Coolant Loop 2 and its associated steam generator and reactor coolant pump,

c. Reactor Coolant Loop 3 and its associated steam generator and reactor coolant pump, and

d. Reactor Coolant Loop 4 and its associated steam generator and reactor coolant pump.

APPLICABILITY: MODE 3.**

ACTION:

a. With less than the above required reactor coolant loops OPERABLE, restore the required loops to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . -

b. With only one reactor coolant loop in operation and the reactor trip breakers in the closed position, within.1~ hour restore two loops to '

operation or open the reactor trip breakers.

c. With no reactor coolant loop in operation, open the reactor trip breakers and suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required reactor coolant loop to operation.

SURVEILLANCE REQUIREMENTS .

4.4.1.2.1 At least the above required reactor coolant pumps, if not in operation, shall be determined OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability.

All reactor coolant pumps may be deenergized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided:

(1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10*F below saturation temperature.

- See special test exceptions Specification 3.10.4.

COMANCHE PEAK - UNIT 1 3/4 4-2

REACTOR COOLANT SYSTEM

' HOT STANDBY SURVEILLANCE REQUIREMENTS (Continued)

. 4.4.1.2.2 The' required steam generators shall be determined OPERABLE by .

verifying secondary side water level'to be greater than or equal to 17% at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.4.1.2.3 The required reactor coolant loops shall be verified in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . '

9 4 >

+

1 ,

e e

9 B

1 e

. ?

4 COMANCHE PEAK - UNIT 1 3/4 4-3

REACTOR COOLANT SYSTEM .

HOT SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.1.3 At least two of the loops listed below shall be OPERABLE and at least one of these loops shall be in operation:*

a. Reactor Coolant Loop l'and its associated steam generator and reactor coolant pump,** -

b. Reactor Coolant Loop 2 and its associated steam generator and reactor coolant pump,**

c. Reactor Coolant Loop 3 and its associated steam generator and reactor coolant pump,** .

d. Reactor Coolant Loop 4.and its associated steam generator and reactor coolant pump,**~

e. RHR Loop A, or

.f. RHR Loop B.

APPLICABILITY: 'HODE 4. .

ACTION:

a. With less than the above required loops OPERABLE, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible; if the remaining OPFrtABLE loop is an RHR loop, be in COLD SHUTDOWN within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

All reactor ~ coolant pumps and RHR pumps may be deenergized for up to I hour provided: (1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10'F below saturation temperature.

- A reactor coolant pump shall not be started in Mode 4 unless the secondary water temperature of each steam generator is less than 50'F above each of the Reactor Coolant System cold leg temperatures.

COMANCHE PLAK - UNIT 1 3/4 4-4

REACTOR COOLANT SYSTEM HOT SHUTDOWN LIMITING CONDITION FOR OPERATION

b. With no loop in operation, suspend'all op'erations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required loop to operation.

SURVEILLANCE REQUIREMENTS 4.4.1.3.1 The required reactor coolant pump (s), and/or RHR pump (s) if not in operation, shall be determined OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability.

4.4.1.3.2 The required steam generato'r(s) shall be determined,0PERABLE by verifying secondary side water level to be greater than or equal to 17% at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.4.1.3.3 At least one reactor cool. ant or RHR loop shall be verified in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4 e

G 0

COMANCHE PEAK - UNIT 1 3/4 4-5 l

)

' REACTOR COOLANT SYSTEM .

COLD SHUTDOWN - LOOPS FILLED LIMITING CONDITION FOR OPERATION 3.4.1.4.1 At least one residual heat removal (RHR) loop shall be OPERABLE and in operation *, and either: '

a. One additional RHR loop,shall be OPERABLE **, or

b. The secondary side water level of at least two steam generators shall be greater than 17%. -

APPLICABILITY: MODE'S wi-th reactor coolant loops filled ***.

ACTION:

a. With one of the RHR loops inoperable or with less than.the required steam generator water level, immediately initiate corrective action to return the inoperable RHR loop to OPERABLE status or restore the required steam ge'nerato'r water level as soon as possible,

b. With no RHR loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to operation. ,

SURVEILLANCE REQUIREMENTS 4.4.1,4.1.1 The secondary side water level of at least two steam generators when required shall be determined to be within limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.4.1.4.1.2 At least one RHR loop shall be determined to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

The RHR pump may be deenergized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided: (1) no operations are permitted that would cause dilution of tne Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at 14ast 10*F below satur,ation temperature.

'One RHR loop may De inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing provided the other RHR loop 15 OPERABLE arid in operation.

- - A reactor coolant pump shall not be started in Mode 5 unless the secondary water temperature of each steam generator is less than 50'F above each of the Reactor Coolant System cold leg temperatures.

COMANCHE PEAK - UNIT 1 3/4 4-6

. .. .. . ~ . . . . . ..

, REACTOR COOLANT SYSTEM )

COLD SHUT 00WN - LOOPS NOT FILLED LIMITING CONDITION FOR OPERATION 3.4.1.4.2 Two residual heat.' removal (RHR) loops shall be OPERABLE

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

iPPLICABILITY: MODE 5 with reactor coolant loops not filled.

ACTION:

a. With less than the above required RHR loops OPERABLE, immediately initiate corrective action to return the required RHR loops to OPERABLE status as soon as possible,

b. With no RHR loop in operation suspend all operations involving a reduction .in boron concentration of the Recctor Coolant System and .

immediately initiate corrective action to return the required RHR loop to operation.

SURVEILLANCE REQUIREMENTS .

L -

4.4.1.4.2. At least one RHR loop soall be determined to be i'n operation and circulating reactor coolant at least once per 12' hours.

s

0ne RHR loop may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveil'.ance testing provided the other RHR loop is OPERABLE and in operation.

- The RHR up'mp may be deenergized for up ts 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided: (1) no opera-tions are permitted that would cause dilut'7n of the Reactor Coolant System boron concertration, and (2) core outlet temperature is maintained at least 10'F below saturation temperature.

COMANCHE PEAK - UNIT 1 3/4 4-7

e

. . -l REACTOR COOLANT SYSTEM- D i 3/4.~4.2- SAFETY VALVES SHUTDOWN LIMITING'. CONDITION FOR OPERATION

=3.4.2.1 A minimum of or.e pressurizer Code safety valve shall be OPERABLE with a lift setting of 2485 psig i.VG.*

APPLICABILITY: MODES 4 and 5.

ACTIONi With no pressurizer Code safety valve OPERABLE, immediately suspend all

. operations involving positive. reactivity changes and place an OPERABLE RHR loop into ope, ration i.n the shutdown cooling mode.

SURVEILLANCE REQUIREMENTS a

4.4.2.1 No additional requirements other than those required by i Specification-4.0.5.

The lift setting pressure shall correspond to ambient conditions of the valve.

at nominal operating temperature and pressure.

COMANCHE PEAK - UNIT 1 3/4 4-8

4

' REACTOR COOLANT SYSTEM OPERATING ,

LIMITING CONDITION FOR OPERATION 3.4.2.2 All pressurizer Code safety valves shall be OPERABLE with a lift setting of 2485 psig i 1%.*

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

With one pressurizer Code safety valve inoperable, either restore the inoperable valve to OPERABLE status within 15 minutes or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS -

4.4.2.2 No additional requirements other than those re utred by Specification 4.0.5.

The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

COMANCHE PEAK - UNIT 1 3/4 4-9

REACTOR COOLANT SYSTEM T?y\\i.

i;

'- ~

3/4.4.3 ' PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.3 The pressurizer shall be OPERABLE with a water volume of less than or equal to 1662 cubic feet (92% of span), and at least two groups of pressurizer heaters each having a capacity of at least [150,] kW.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

a. With only one group of pressurizer heaters OPERABLE, restore at least two groups to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTOOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. With the pressurizer otherwise inoperable,'be in at least HOT STANDBY ~

with the Reactor Trip System breakers open within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.4.3.1 The pressurizer water volume shall be determined to be within its limit at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.4.3.2 The capacity of each of the above' required groups of pressurizer heaters shall be verified by energizing the heaters and measuring circuit current at least once per 92 days. .

0 COMANCHE PEAK - UNIT 1 3/4 4-10

REACTOR COOLANT SYSTEM

,3]i.4.4 RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.4.4 All power-operated relief valves (PORVs) and their associated block' -

valves shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

a. With one or more PORV(s) inoperable, because of excessive seat leak-age, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore the PORV(s) to OPERABLE status or close the associated block valve (s); otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the fol-lowing 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ,

b. With one PORV inoperable due to causes ot'h'r e' than excessive seat leakage, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore the PORV to OPERABLE status or close the associated block valve and remove power from the block valve; restore the PORV to OPERABLE status within the following 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be.in HOT' STANDBY within the next 6. hours and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c. With both PORV(s) inoperable due to causes other than excessive seat leakage,.within I hour either restore each of the PORV(s) to OPERABLE status or close their associated block valve (s) and remove i

' power from the block valve (s) and be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ,

d. With one or more block valve (s) inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months (1) restore the block valve (s) to OPERABLE status or close the block valve (s) and remove power from the block valve (s); or close the PORV and remove power from its associated solenoid valve; and (2) apply ACTION b above, as appropriate, for the isolated PORV(s).

e. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.4.4.1 In addition to the requirements of Specification 4.0.5, each PORV shall be demonstrated OPERABLE at least once per 18 months by:

a. Operating the valve through one complete cycle of full travel, and

b. Performing a CHANNEL CALIBRATION of the actuation instrumentation.

COMANCHE PEAK - UNIT 1 3/4 4-11

4 , a < . - - ._.a . .. a . . ~ . . . , ,

4:

iREACTOR COOLANT SYSTEM .TT v' k 3/4.4.4=-RELIEF VALVES UAg'

. SURVEILLANCE REQUIREMENTS 4.4.4.2- Each block valve shall be demonstrated OPERABLE at ieast once per.

92 days:by operating the valve through one complete cycle of full travel unless the' block valve is closed in order to meet the requirements of ACTION a

, . and b in Specification 3.4.4.

s b

1 l~

9 1-I i

l.

- e COMANCHE PEAK - UNIT 1 3/4 4-12

l -

REACTOR COOLANT SYSTEM 3/4.4.5 STEAM GENERATORS LIMITING CONDITION FOR'0PERATION 3.4.5 Each steam generator shall be OPERABLE.

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

ACTION:

With one or more steam generators inoperable, restore the inoperable generator (s) to OPERABLE. status prior to increasing T,yg above 200 F.

SURVEILLANCE REQUIREMENTS 4.4.5.0 Each steam generator shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the requirements of Specification 4.0.5.

4.4.5.1 Steam Generator Sainple Selection and Inspection - Each steam generator shall be determined OPERABLE during shutdown by selecting and inspecting at .

least the minimum number of steam generators specified in Table 4.4-1.

4.4.5.2 Steam Generator Tube Sample Selection and Inspection - The steam generator tube minimum sample size, inspection result classification, and the corresponding action required shall be as specified in Table 4.4-2. The inservice inspection of steam generator tubes shall be performed at the fre-quencies specified in Specification 4.4.5.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 4.4.5.4. The tubes selected for each inservice inspection shall include at least 3% of all the expanded tubes and at least 3% of the remaining number of tubes in all steam generators; the tubes selected for these inspections shall be selected on a random basis except:

a. Where experience in similar plants with similar water chemistry indicates critical areas to be inspected, then at least 50% of the tubes inspected shall bt from these critical areas;

b. The first sample of tubes selected for each inservice inspection (subsequent to the preservice inspection) of each steam generator shall include:

a e

COMANCHE PEAK - UNIT 1 3/4 4-13

-e REACTOR COOLANT SYSTEM STEAM GENERATORS SURVEILLANCE REQUIREMENTS (Continued)

1) All nonplugged tubes that previously had detectable wall' penetrations (greater than 20%),

2) Tubes in those areas where experience has indicated potential problems, and

3) A tube inspection (pursuant to Specification 4.4.5.4a.8) shall be performed on each selected tube. If any selected tube does not permit the passage of the_ eddy current probe.for a tube inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection.

c. The tubes selected as the second and third samples (if required by Table 4.4-2) during each inservi.ce. inspection may be subjected to a

partial tube inspection Qov'ided:

1) The tubes selected for these samples include the tubes from those areas of the tube sheet array where tubes with

, imperfections were previously found, and

2) The inspections include those portions of the. tubes where imperfections were previously found.

The results of each sample inspection shall be classified into one of the following three categories:

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

C-2 One or more tubes, but not more than 1% of the total tubes inspected are defective, or between 5% and 10% of the total tubes inspected are degraded tubes.

C-3 More than 10% of the total tubes inspected are degraded tubes or more than 1% of the inspected tubes are defective.

Note: In all inspections, previously degraded tubes must exhibit significant (greater than 10%) further wall penetrations to be included in the above percentage calculations.

COMANCHE PEAK - UNIT 1 3/4 4-14

REACTOR COOLANT SYSTEM STEAM GENERATORS SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.3 Inspection Frequencies 'The above required inservice inspections of ,

steam generator tubes shall be performed at the following frequencies:

a. The first inservice inspection shall be performed after 6 Effective Full Power Months (EFPM) and before 12 EFPM and shall include a special inspection of all expanded tubes in all steam generators.

Subsequent inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection. If two consecutive inspections, not including the preser-vice inspection, result in all inspection results falling into the C-1 category or if two consecutive inspections demonstrate that previously observed degradation has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of once per 40~ months.; ,

b. If the results of the inservice inspection of a steam generator conducted in accordance with Table 4.4-2 at 40-month intervals fall in Category C-3, the inspection frequency shall be increased to at least once per 20 months. The increase in. inspection frequency shall apply until the subsequent inspections satisfy the criteria of Specification 4.4.5.3a.; the interval may then be extended to a maximum of once per 40 nonths; and c.

Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection ,

specified in Table 4.4-2 during the shutdown subsequent to any of ^

the following conditions: .

1) Primary-to secondary tubes leak (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of Sper.ification 3.4.6.2, or

2) A seismic occurrence greater than the Operating Basis Earthquake, or

'3) A loss-of-coolant accident requiring actuation of the Engineered Safety Features, or

4) A main steam line or feedwater line break.

COMANCHE PEAK - UNIT 1 3/4 4-15 e

l REACTOR COOLANT SYSTEM .

STEAM GENERATOR -

F*

SURVEILLANCE REQUIREMENTS (Continued)

4. 4. 5.'4 e Aece'ptance Criteria

a. As used in this specification:

1) Imperfection means an exception to the dimensions, finish, or contour of a tube from that required by fabrication drawings or specifications. Eddy-current testing indications below 20% of tFe nominal tube wall thickness, if detectable, may be considered as imperfections;

2) Degradation means a service-induced cracking, wastage, wear, or general corrosion occurring on either inside or outside of a tube;

3) Degraded Tube means a tube containing imperfections greater than or equal' to 20% of the no'.ninal wall thickness caused by i

degradation; i

4) % Degradation means the percentage of the tube. wall thickness affected or removed by degradation; .

5) Defect means an imperfection of such severity that it exceedr,

. the plugging limit. A tube containing a, defect is defective;

~

6) Plugging Limit means the imperfection depth at or beyond which the tube shall be removed from service and is equal to 40%*

of the nominal tube vall thickness;

7) Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integ-rity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in Specification 4.4.5.3c., above'

8) Tube Inspection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg; and~

Value to be determined in accordance with the recommendations of Regulatory Guide 1.121, August 1976.

l COMANCHE PEAK - UNIT 1 3/4 4-16

< ~

+ . .

REACTOR COOLANT SYSTEM STEAM GENERATOR ,

SURVEILLANCE REQUIREMENTS (Continued)

9) Preservice Inspection means an inspection of the "fuT1 length.of '

each tube in.each steam generator performed by eddy current techniques prior to service to establish a baseline condition of the tubing. This inspection shall be performed prior to initial POWER OPERATION using the equipme expected to be used during subsequent ervice ins,nt and techniques inspections.

b. The steam generator shall be determined OPERABLE after completing

, the corresponding actions (plug all tubes exceeding the plugging limit and all tubes containing through-wall cracks) required by Table 4.4-2.

4.4.5.5 Reports

a. Within"15 da'ys following the completion of each inservice inspection of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the Commission in a Special Report pursuant to Specification 6.9.2;

b. The complete results of the steam generator tube inservice inspec-tion shall be submitted to the Commission in a Special Report pur-l' suant to Specification 6.9.2 within 12 inonths following the

, completion of the inspection. This Special Report shall include

i l 1) Number and extent of tubes inspected, i

2) Location and percent of wall-thickness penetration for each indication of an imperfection, and

3) Identification of tubes plugged.

c. Results of steam generator. tube inspections which fall into Category C-3 shall be reported to the Commission purcuant to 10 CFR Part 50.72.

within four hours of initial discovery, and pursuant to Specifica-l tion 6.3.2 within 30 days and prior to resumption of plant operation.

l This report shall provide a description of investigations conducted l to determine cause of the tube degradation and corrective measures taken to prevent recurrence, i

e l

I j COMANCHE PEAK - UNIT 1 3/4 4-17 l

l

TABLE 4.'4-1 '

MINIMUM NUMBER OF ' STEAM GENERATORS TO BE INSPECTEDDURINGINSERVICEINSPEUTION ,

S ic

. r

~

c Preservice Inspection . Four z ,

U No. of Steam Generators per Unit Four .

w .

First Inservice Inspection ' '

Two Second & Subsequent Inservice Inspections Onel TABLE NOTATIONS

1. Each of the other two steam generators not inspected during the first inservice inspections

, y shall be inspected during the second and third inspections. For the fourth and subsequent g inspections, the inservice inspection may be limited to one steam generator on a rotating schedule encompassing 12% of the tubes if the results of the first or previous inspections indicate that all steam generators are performing.in a like manner. . Note that under some circumstances, the operating conditions in one or more steam generators may be found to be more severe than those in other steam generators. Ucder such circumstances the sample sequence shall be modified to inspect the most severe conditions.

p rv -

b

.e:- w n- 1r,w---- ,,m, . - - - - - g ----- -e -- - -

- , , w : --

-r. , --.m-- , x ~c; , -m ..

TABLE 4.4-2, STEAM GENERATOR TUBE INSPECTION O IST SAMPLE INSPECTION 2ND SAMPLE INSPECTION f 3RD SAMPLE INSPECTION z

gSample Size Result Action Required Result ' Action Required Result Action Required m

yA minimum of C-1 None -

N.A. N.Ad N.A. N.A. -

RS Tubes per

,S.G. C-2 Plug defective tubes C-1 None N.A. N.A.

e and inspect additional z

25 tubes in this S.G. Plug defective tubes C-1 None C-2 and inspect additional 45 tubes'in this S.G. C-2 Plug defective tubes Perform action for C-3 C-3 result of first sample

~ Perform. action for m C-3 C-3 result of first N.A. N.A.

} . sample

[ C-3 Inspect all tubes in All other e this S.G., plug de- S.G.s are None N.A. N.A.

fective tubes and C-1 inspect 25 tubes in '

each other S.G. Some 5.G.s Perform action for N.A. N.A.

C-2 but no C-2 result of second additional sample Notification to NRC S.G. are pursuant to $50.72 C-3 (b)(2) of 10 CFR Part 50 Additional , Inspect all tubes'in ~

S.G.-is each S.G. and plug ,

d C-3 defective tubes.

' Notification to NRC N.A. N.A.

pursuant to $50.72 -

(b)(2) of 10 CFR Part 50 S = 1_2% Where n is the number of steam generators inspected during an inspection n

___ l

REACTOR COOLANT SYSTEM i 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE M LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.6.1 The following Reactor Coolant System Leakage Detection Systems shall be OPERABLE:

a. The Containment Atmosphere (Gaseous or Particulate] Radioactivity Monitoring System,

b. The Containment Sump Level and Flow Menitoring System, and

c. Either the containment air. cooler condensate flow rate or the Con-tainment Atmosphere (Gaseous or Particulate] Radioactivity Monitoring System.

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

ACTION:

.With only two of the above required Leakage Detection Systems OPERABLE,

. operation may continue for up to 30 days provided grab samples of the contain'-

ment atmosphere are obtained and analyzed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the.

required Gaseous or Particulate Radioactive Monitoring System is inoperable; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.4.6.1 The Leakage Detection Systems shall be demonstrated OPERABLE by:

a. Containment Atmosphere Gaseous and Particulate Monitoring Systems-performance of CHANNEL CHECK, CHANNEL CALIBRATION, and DIGITAL .

CHANNEL OPERATIONAL TEST at the frequencies specified in Table 4.3-3,

b. Containment Sump Level and Flow Monitoring System performance of CHANNEL CALIBRATION at least once per 18 months, and

. c. Containment Air Cooler Condensate flow Rate Monitoring System -

performance of CHANNEL CALIBRATION at least once per 18 months.

I COMANCHE PEAK - UNIT 1 3/4 4-20

REACTOR COOLANT SYSTEM y

OPERATIONAL LEAKACf t i

LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to:

1

a. No PRESSURE BOUNDARY LEAKAGE,

b. 1 GPM UNIDENTIFIED LEAKAGE, *

c. 1 GPM total reactor-to-secondary leakage through all steam generators not isolated from the Reactor Coolant System and 500 gallons per day through any one steam generator not isolated from the Reactor Coolant System,

d. 10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System,

e. 40 GPM CONTROLLED LEAKAGE at a Reactor Coolani. System pressure of 2235 1 20 psig, and

f. 0.5 GPM leakage per nominal inch of valve size up to.a maximum of 5 GPM at a Reactor Coolant System pressure of 2235'i 20 psig from any Reactor Coolant System Pressure Isolation Va.1ve specified in Table 3.4-1. '

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

ACTION:

a. With any PRESSURE BOUNDARY' LEAKAGE, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ,

b. With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE and leakage from Reactor Coolant System Pressure Isolation Valves, reduce the leakage rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following

'30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c. With any Reactor Coolant System Pressure Isolation Valve leakage greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 ho,urs by use of at least two closed manual or deactivated automatic valves, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

4 COMANCHE PEAK - UNIT 1 3/4 4-21

REACTOR COOLANT SYSTEM ,

OPERATIONAL LEAKAGE SURVEILLANCE REQUIREMENTS

' 4. 4. 6. 2.1, React'or Coolant System leakages shall be demonstrated to be within each of the above limits by:

a. Monitoring the containment atmosphere gaseous or particulate radioactivity monitor at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ;

b. Monitoring the containment sump inventory and discharge at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ;

c. Meas'urement of the CONTROLLED LEAKAGE to the reactor coolant pump seals when the Reactor Coolant System pressure is 2235 1 20 psig at least once per 31 days with the modulating valve fully open. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4;

^

d. Performance of a Reactor Coolant System water inventory balance at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; and e,- Monitoring the Rector Head Flange Leakoff Systein at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . ,

4.4.6.2.2 Each Reactor Coolant System Pressure Isolation Valve specified in

Table 3.4-1 shall be demonstrated OPERABLE by verifying leakage to be within its limit:

a. At least once per 18 months,

b. Prior to entering MODE 2 whenever the plant has been in COLD SHUTOOWN for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or more and if leakage testing has not been performed in the previous 9 months, except for valves 8701A, 8701B, 8702A, and 87028.

c. Prior to returning the valve to service following maintenance, repair or replacement work on the valve and ,

d. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve actuation due to automatic or manual action or flow through the valve,

e. As outlined in the ASME Code,Section XI, paragraph IW-3427(b).

The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4.

COMANCHE PEAK - UNIT 1 3/4 4-22

.o y ,

TABLE 3.4-1

. REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES VALVE NUMBER , ,

FUNCTION 8948 A,-B, C, D AccumulatdrTankDischarge 8956 A, B, C, D Accumulator Tank Discharge

. 8905 A, B, C, D SI Hot Leg Injection 8949 A, B, C, D SI Hot Leg Injection a a 8818 A, B, C, D RHR Cold Leg Irijection

", 8819 A, B, C, D SI Cold Leg Injection 8701 A, B RHR Suction Isolation -

8702 A, B RHR Suction Isolation -

8705 A, B RHR Suction Isolation Relief 8841 A, B RHRHotLegInjection l 8815 CCP Cold Leg Injection 8900 A, B, C, D CCP Cold. Leg Injection N

, 9 4 9

l I

e 4

5 6

COMANCHE PEAK - UNIT 1 3/4 4-23

REACTOR COOLANT SYSTEM lg 3/4.4.7 CHEMISTRY LIMITING CONDITION FOR OPERATION 3.4.7 The Reactor Coolant System chemistry shall be maintained within the limits specified in Table 3.4-2.

APPLICABILITY: At'all times.

ACTION:

MODES 1, 2, 3, and 4:

a. With any one or more chemistry parameter in excess of its Steady-State Limit but within its Transient Limit, restore the parameter to within its Steady-State Limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ; and

b. With any one or more chemistry parameter in excess of its Transient Limit, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN

.within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . '

~

.At Al'l Other Times: ,-

With~the concentration of either chloride or fluoride in the Reactor Coolant System in exces's of its Steady-State Limit for more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or in excess of its Transient Limit, reduce the pressurizer pressure to less than or equal to 500 psig, if applicable, and perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation prior to increasing the pressurizer pressure above 500 psig or prior to

, proceeding to MODE 4.

SURVEILLANCE REQUIREMENTS 4.4.7 The Reactor Coolant System chemistry shall be determined to be within the limits by analysis of those parameters specified in Table 3.4-2 at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .*

Sampleandanalgsisfordissolvedoxygenisnotre'quiredwithT,yg or equal to 250 F.

less than COMANCHE PEAK - UNIT 1 3/4 4-24

- e p

TABLE 3.4-2 !

MACTORCOOLANTSYSTEM ,

CHEMISTRY LIMITS f

--r STEADY-STATE ' TRANSIENT !

PARAMETER LIMIT LIMIT ,

Dissolved Oxygen * < 0.10 ppm- 3 1.00 ppm l

Chloride < 0.15 ppm 5 1,50 ppm Fluoride 1 0.15 ppm 5 1.50 ppm

. r 9

j

l

Limit not applicable with T,yg less than or equal to 250'F.

COMANCHE PEAK - UNIT 1 3/4 4-25 t

REACTOR COOLANT SYSTEM g 3/4.4.8 ' SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.8 The specific activity of f.he reactor coolant shall be limited to:

a. Less than or equal to 1 microcurie per gram DOSE EQUIVALENT I-131, and

b. Less than or equal to 100/I microcuries per gram of gross radioactivity.

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

ACTION:

^

MODES 1, 2 and 3*:

a. With the specific activity of the reactor coolant greater than 1 microcurie per gram DOSE EQUIVALFNT I-131 for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval or exceeding the limit line shown on Figure 500"F 3.4-1, within be in at 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; least HOT STANDBY with T"V9 less than and b.

Witt)I 100/ microcuries per gram, be in 'at least HOT STANDBY with less Tthe. spec than 500*F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , avg MODES 1, 2, 3, 4, and 5:

With the specific activity of the reactor coolant greater than 1 microcurie per gram DOSE EQUIVALENT I-131 or greater than 100/E micro-Curies per gram, perform the sampling and analysis requirements of Item 4.a) of Table 4.4-4 until the specific activity of the reactor coolant is restored to within its limits.

SURVEILLANCE REQUIREMENTS i

4.4.8 The specific activity of the reactor coolant shall be determined to be within the limits by performance of the sampling and analysis program of Table 4.4-4.

With T,yg greater than or equal to 500*F.

COMANCHE PEAK - UNIT 1 3/4 4-26

[

i f

i.

s -

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L i.

[

L 1 i f

1

i I

i i

I E -

FIGURE 3.4-1 ;

i

DOSE EQUIVALENT l-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS l PERCENT OF RATED THERMAL POWER WITH THE REACTOR C0OLANT SPECIFIC .

L ACTIVITY

>l pCi/ gram DOSE EQUIVALENT !=131 t

' I

~

COMANCHE PEAK - UNIT 1 3/4 4-27 !

1

.,,.-.._,__..r.

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

o TABLE 4.4-4 .,

i g .

REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM m

E

=

TYPE OF MEASUREMENT i 9

" AND ANALYSIS SAMPLE AND ANALYSI5 t

- FREQUENCY M00ES IN WHICH SAMPLE e

1. Gross Radioactivity .AND ANALYSIS REQUIRED'

) g Determination At least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . 1,2,3,4

{ Z 2. Isotopic Analysis for DOSE EQUIVA-t - 1 per 14 days.

LENT I-131 Concentration I

!, 3. Radiochemical for 5 Determination

1 per 6 months ** 1 i 4. Isotopic Analysis for Iodine Including I-131, I-133, and 1-135 a) Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , 1#, 2#, 3#, 4#, 5#

whr.never the specifir adtivity exceeds 1 w pCi/ gram DOSE -

1 EQUIVALENT I-131 j 9 or 100 d pCi/ gram of Es gross radioactivity, and

] b) One sample between 2 1,2,3 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months following -

a THERMAL POWER change

, exceeding 15% -

} of the RATED THERMAL POWER within a 1-hour 1- period.

e O ,

l l .,

4 M 4

e

. _ _ _ _ _ . _ . . _ _,_._.___.._. .. _ ._. _ _.=._._ .. _ _ _ _ . . - _ _ _ . _ , . _ . . , _ _ - - - . _ . _ . .. _ _ . _ . _ _ _ _ _. . _ .

4 TABLE 4.4-4 (Continued)

TABLE NOTATIONS

~

. *A radiochemical analysis for i shall consist of the quantitative measurement of the specific activity for each radionuclide, except for radionuclides with half-lives less than 10 minutes and all radioiodines, which is identified in the reactor coolant. The specific activities fgr these individual radio-nuclides shall be used in the determination of_E for the reactor coolant sample. Determination of the contributors to E shall be based upon those energy peaks identifiable with a 95% confidence level.

- Sample to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since reactor was last suberitical for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or longer.

Until the specific activity of the Reactor Coolant System is restored within its limits.

4 i

t 4

I L

! . r I

l COMANCHE PEAK - UNIT 1 3/4 4-29 .

l :

L ..

REACTOR COOLANT SYSTEM . (,

3/4.4.9 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM

, LIMITING CONDITION FOR OPERATION

3.4.9.1 The Recctor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figures 3.4-2 and S.4-3 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:

a. A maximum heatup of.100'F in an) 1-hour period,

b. A maximum cooldewn of 100 F in any 1-hour period, and

c. A maximum temperature change of less.than or equal to 10'F in any 1-hour period durirg inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.

APPLICABILITY: At all times.

ACTION: .

With .any of the abNe limits. exceeded, restore the temperature and/or pressure to within the limit within'30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity

. of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce the RCS T,yg and pressure to less than 200*F and ,

500 psig, respectively, within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operations.

4.4.9.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examited, to determine changes in material properties, as required by 10 CFR Part 50, Appendix H, in accordance with the schedule in Table d.4-5. The results of these examinations shall be used to update Figures 3.4-2 and 3.4-3.

6 COMANCHE PEAK - UNIT 3 3/4 4-30

O O

9 9

FIGURE 3.4-2 RENCTORCOOLANTSYSTEMHEATUPLIMITATIONS-APPLICABLEUPTO EFPY COMANCHE PEAK - UNIT 1 3/4 4-31 i

O 4

0 t

t FIGURE 30 4-3 REACTORCOOLANTSYSTEMC00LDOWNLIMITATIONS-APPLICABLENPTO EFPY COMANCHE PEAK - UNIT 1 3/4 4-32

8 4

9 0

4 O

8 t W b M

a t -

g . p E

t E

. ~

. .E

~ ' .

3 '

4 1 4 .

e .

.m g h E

>=

1 d

a W

8 d5 a-8 Ws .

G e

e Siit COMANCHE PEAK - UNIT 1 3/4 4-33

REACTOR COOLANT SYSTEM f\$

PRESSURIZER [.;a LIMITING CONDITION FOR OPERATION _

l3.4.9.2 The pressurizer temperature shall de limited to:

a. A maximes heatup of 100*F in any 1-hour period,

b. A maximum cooldown of 200*F in any 1-hour period, and -

c. ,

A maximum spray water temperature differential of 625'F. -

APPLICABILITY: At all times.

ACTION:

With the pressurizer. temperature limits in excess of' any of the above limits,'

restore the temperature to within the limits within 30 minutes; perform an :

engineering evaluation to determine the effects of the out-of-limit condition ;

on the structural integrity of the pressurizer; determine that the pressurizer '

remains acceptable for continued operation or be in at least HOT STAN0BY within .

the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce the pressurizer pressure to less than 500 psig within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . ,,

SURVEILLANCE REQUIREMENTS y i 4.4.9.2 The pressurizer temperatures shall be determined to be within the ,.

limits at least once per 30 minutes during system heatup or cooldown. The spray water temperature differential shall be determined to be within the limit at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during auxiliary spray operation.

4 4

i

[

i h

i 3/4 4-34

~

COMANCHE PEAK - UNIT 1 t

t

REACTOR' COOLANT' SYSTEM OVERPRESSURE PROTECTION S'rSTEMS a

LIMITING CONDITION FOR' OPERATION 1

3.4.9.3 At'least one of the following Overpressure Protection Systems shall be OPERABLE:

a. Two power-operated relief valves (PORVs) with a lift setting of less than or equal to [450] psig, or

b. The Reactor Coolant System (RCS) depressurized with an RCS vent of greater than or equal to 2.98 square inches.

APPLICABILITY: MODE 4 when the temperature of any RCS cold leg is less than or equal to [275] F, MODE 5 and MODE 6 with the reactor vessel head on.

ACTION:

a. With one PORV inoperable, restore the inoperable PORV to OPERABLE status within 7 days or depret.surize and vent the RCS tnrough at least a 2,98 square inch. vent within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . -

b. With both PORVs inoperable, depressurize and vent.the RCS through at

. least a 2.98 square inch vent witt.*n 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

c. In the event either the PORVs or the RCS vent (s) are used to mitigate an RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days. The report shall describe the circumstances initiating tne transient, the effect of the PORVs or RCS vent (s) on the transient, and any corrective action necessary to prevent recurrence.

l d. The provisions of $pecification 3.0 4 are not applicable.

l l

l .

l COMANCHE PEAK - UNIT 1 3/4 4-35 i

i

REACTOR C00 ANT' SYSTEM l OVERPRESSURE PROTECTION SYSTEM SURVEILLANCE REQUIREMENTS

~4.4.9.3.1 Each PORV shall be demonstrated OPERABLE by:

a. Performance of an ANALOG CHANNEL OPERATIONAL TEST on the PORY

. actuation channel, but excluding valve operation, within 31 days prior to entering a condition in which the PORV is required OPERABLE and at least once per 31 days thereafter when the PORV is required OPERABLE; b.- Performance of a CHANNEL CALIBRATION on the PORV actuation channel at least once per 18 months; and

c. Verifying the PORV isolation valve is open at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months when the PORV is being used for overpressure protection.

4.4.9.3.2 The RCS vent (s) shall be verified to be open at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

when the vent (s) is being used for overpressure protectio.n.

I

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

(

I .

I c

l COMANCHE PEAK - UNIT 1 3/4 4-36

l REACTOR COOLANT SYSTEM n3 3 A 3/4.4.10 STRUCTURAL INTEGRITY W d' LIMITING CONDITION FOR OPERATION 3.4.10' The structural integrity of ASME Code Class 1, 2, and 3 components shall be maintained in acco~rdance with Specification 4.4.10:

APPLICABILITY: All MODES. .

ACTION:

l

a. With the structural integrity of any ASME Code Class 1 component (s) not conforming to.the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature more than 50 F above the minimum temperature required by NDT considerations. .

b. With the structural integrity of any ASME Code Class 2 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected componen.t(s) prior to increasing the Reactor Coolant System temperature above'200'F.

c. With the struc'tural. integrity.of.any'ASME Code Class 3 component (s)'

not c.onforming to the above requirements, restore '.he structural l integrity of the affected component (s) to within its limit or isolate l the affected component (s) from service.

d. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.4.10 In addition to the requirements.of Specification 4.0.5, each reactor coolant pump flywheel shall be inspected per the recommendations of Regulat ry Position C.4.b of Regulatory Guide 1.14, Revision 1, August 1975.

1 e

l l

l I

COMANCHE PEAK - UNIT 1 3/4 4-37

l REACTOR COOLANT SYSTEi!

3/4.4.11 REACTOR COOLANT SYSTEM VENTS LIMITING CONDITION FOR OPERATION 3.4.11 At least one Reactor Coolant System vent path consisting of [two] vent -

valves (s) and [one] block valve powered from emergency uusses shall be OPERABLE and closed at each of the following locations:

a. Reactor vessel head, and

b. Pressurizer steam space.

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

ACTION:

a. With one of the above Reactor Coolant System vent paths inoperable, STARTUP and/or POWER OPERATION may continue provided the inoperable vent path is maintained closed with power removed from the valve actuator of all the vent valves and block valves in the inoperable vent path; restore:the inoperable vent path to OPERABLE status within 30 days, or, be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTOOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ,

b. With both Reactor Coolant System vent paths inoperable; maintain the inoperable vent paths closed with power. removed from the valve actua-tors of all the vent valves and block valves in the inoperable vent paths, and restore at least [two] of the vent paths to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.4.11.1 Each Reactor Coolant System vent path block valve not required to -

be closed by ACTION a. or b., above, shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel from the control room.

l 4.4.11.2 Each Reactor Coolant System vent path shall be demonstrated OPERABLE at least once per 18 months by:

a. Verifying all manual isolation valves in each vent path are locked in the open position, *

b. Cycling each vent valve through at least one complete cycle of

! full travel from the control room, and J

. c. Verifying flow through the Reactor Coolant System vent paths during l venting.

l COMANCHE PEAK - UNIT 1 3/4 4-38 l

EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS *

' COLD LEG INJECTION LIMITING CONDITION FOR OPERATION 3.5.1 ~ Each cold leg injection accumulator shall be OPERABLE with:

a. The discharge isolation valve open with power removed,

b. A contained borated water volume of between 6253 gallons ([Later]%

of span)'and 6465 gallons ([Later]% span)

c. A boron concentration of between [1900] and [2100] ppm, and

d. A nitrogen cover pressure of between 605 and 655 psig. ,

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

l l

ACTION:

a. With one cold leg injection accumulator inoperable, except as a result of a closed isolation valve, restore'the inoperable accumulator to '

' OPERABLE status'within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within.

the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce pressurizer pressure to less than 1000.psig within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. With one cold leg injection accumulator inoperable due to the

! isolation valve being closed, either immediately open the isolation i

valve or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce pressurizer pressure to less than 1000 psig within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.5.1.1 Each cold leg injection accumulator shall be demonstrated OPERABLE:

I a. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by:

l

1) Verifying the contained borated water volume and nitrogen cover pressure in the tanks, and l

2) Verifying that each cold leg injection accumulator isolation valve is open.

Pressurizer pressure above 1000 psig.

COMANCHE PEAK - UNIT 1 3/4 5-1 l

c .

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

b. At least once per 31 days and within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after each indicated .

solution volume increase of greater than or eq'ual. to 101 gallons

([Later]% of span) by verifying the boron concentration of the solution in the water-filled accumulator;

c. At least once por 31 days when the RCS pressure is above 1000 psig by verifying that power to the isolation valve operator is removed.

d. At least once per 18 months by verifying that each accumulator isola-tion valve opens automatically under each of the following conditions:

1) When an actual or a simulated RCS pressure signal exceeds the

, P-11 (Pressurizer Pressure Block of Safety Injection) setpoint, l and

2) Upon receipt of a Safety Injection test signal.

4. 5.1.'2 Each accumulator water level and pressure channel shall be demon-strated OPERABLE:

a. At least once per 31 days be the performance of an ANALOG CHANNEL

. OPERATIONAL TEST, and ,

b. At least once per 18 months by the performance of a CHANNEL ~

CALIBRATION.

l 1

1 I

COMANCHE PEAK - UNIT 1 3/4 5-2

EMERGENCY CORE COOLING SYSTEMS 3/4.5.2 ECCS SUBSYSTEMS - T,yg GREATER THAN OR EQUAL TO 350 F LIMITING CONDITION FOR OPERATION 3.5.2 Two independent Emergency Core Cooling System (ECCS) subsystems shall be OPERABLE with each subsystem comprised of:

a. One OPERABLE centrifugal charging pump,

b. One OPERABLE Safety Injection pump,

c. One OPERABLE RHR heat exchanger,

d. One OPERABLE RHR pump, and

e. An OPERABLE flow path capable.of taking suction from the refueling water storage tank on a S'afety Injection signal and automatically opening thh containment sump suction valves during the recirculation l' phase of operation.

l -

APPLICABILITY: MODES 1, 2, and 3*.

ACTION: .

a. With one ECCS subsystem inoperable, r'estore the ir. operable subsystem to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,

b. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describ-ing the circumstances of the actuation and the total accumulated actuation cycles to date. The current value of the usage factor for each affected Safety Injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.

l l

The provisions of Specification 3.0.4 and 4.0.4 are not applicable for entry into Mode 3 for the centrifugal charging pumps and the safety injection pumps declared inoperable pursuant to Specification 3.5.3 provided the centrifugal charging pumps and the saf,ety injection pumps are restored to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or prior to-the temperature of one or more of the RCS cold leg exceeding 375 F, whichever comes first.

COMANCHE PEAK - UNIT 1 3/4 5-3 4

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE:

a. At least 'once pei' 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months b'y verifying that the following valves are in the indicated positions with power to the valve operators removed:

Valve Number Valve Function Valve Position 8802 A & B SI Pump to Hot Legs Closed 8808 A, B, C, D Accum. Discharge Open*

8809 A & B RHR to Cold Legs Open 8835 SI Pump to Cold Legs Open 8840 RHR to Hot Legs Closed 8806 SI Pump Suction from RWST Open 8813 SI Pump Mini-Flow Valve Open

b. At least'once per 31 days by:

1) Verifying'that the ECCS piping is full of water by venting.the' ECCS pump casings and accessible discharge piping high points, and

2) Verifying that each valve (manual, power-operated, or automatic)

,in the flow path that is not 16cked, sealed, or otherwise secured.in position, is in its correct position.

c. By'a visual inspectiion which verifies that no loose debris (rags, trash, clothing, etc.) is present in the containment which could be transported to the containment sump and cause restriction of the pump suctions during LOCA conditions. This visual inspection shall be performed:

1) For all accessible areas of' the containment prior to establish-ing CONTAINMENT INTEGRITY, and

2) Of the areas affected within containment at the completion of each containment entry when CONTAINMENT INTEGRITY is established.

d. At least once per 18 months by:

! 1) Verifying automatic isolation and interlock action of the RHR system from the Reactor Coolant System to ensure that:

l a) With a simulated or actual Reactor Coolant System pressure signal greater than or equal to [425] psig the interlocks prevent the valves from being opened, and b) With a simulated or actual Reactor Coolant Systen pressure

signal less than or equal to 750 psig the interlocks will j cause the valves to automatically close.

l -

' ~

Surveillance Requirements covered in Specification 4.5.1.1.

l i COMANCHE PEAK - UNTT 1 3/4 5-4 l .

o.

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

2) A visual in,spection of the containment. sump and verifying .that the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens, etc.) show no evidence of structural distress or abnormal corrosion.

e. At least once per 18 months, during shutdown, by:

1) Verifying that each automatic valve in the flow path actuates to its correct position on Safety Injection actuation and test signals, and

2) Verifying that each of the following pumps start automatically upon receipt of a Safety Injection actuation test signal:

a) Centrifugal. charging pumps, b) Safety Injection pumps, and c) RHR pumps.

~f. By verifying that each of the following pumps develops the indicated d.ifferential pressure on recirculation flow when tested pursuant to Specification 4.0.5:

1) Cen,trifugal charging pump 1 2370 psid,

2) Safety Injection pump 1 1440 psid, and

3) RHR pump > 170 psid.

g. By verifying the correct position of each mechanical position stop for the following ECCS throttle valves:

1) Within 4. hours following completion of each valve stroking operation or maintenance on the valve when the ECCS subsystems are required to be OPERABLE, and l

l 2) At least once per 18 months.

CCP/SI System Valve Number SI System Valve Number SI-8810A SI-8822A SI-8816A SI-88108 SI-8822B SI-8816B SI-8810C SI-8822C SI-8816C SI-88100 SI-88220 SI-8816D l

l COMANCHE PEAK - UNIT 1 3/4 5-5 L .

,h EMERGENCY CORE COOLING SYSTEMS W SURVEILLANCE REQUIREMENTS (Continued)

h. By performing a flow balance test, during shutdo.wn, .following com-pletion of modifications to the ECCS subsystems'that alter the' subsystem flow characteristics and verifying that:

1) For centrifugal charging pump lines, with a single pump running:

a) The sum of the injection line flow rates, excluding the highest flow. rate, is greater than or equal to 333 gpm, and b) The total pump flow rate is less than or equal to 555 gpm.

2) For Safety Injection pump lines, with a single pump running:

a) The sum of the injection line flow rates, excluding the highest flow rate, is greater than or equal to 437 gpm, and b) The total pump flow rate is less than or equal to 660 gpm.

3) For RHR' pump lines, wi'th a single pump running, the sum of the injection line flow rates is greater than or equal to 4652 gpm.

l s.

i COMANCHE PEAK - UNIT 1 3/4 5-6

A g,

t EMERGENCY CORE COOLING SYSTEMS 3/4.5.3 ECCS SUBSYSTEMS .T,yg LESS THAN 350 F

_ECCS SUBSYSTEMS LIMITING CONDITI0ri FOR OPERATION 3.5.3.1 ~As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:

a. One OPERABLE centrifugal charging pump,*

b. One '0PERABLE RHR heat exchanger,

c. One OPERABLE RHR pnmp, and

d. An OPERABLE flow path capable of taking suction from the refueling water storage tank upon being manually realigned and trarisferring suction to the containment sump during the recirculation phase of

! operation.

I I

APPLICABILITY: MODE 4.

ACTION: .

a. 'With no ECCS subsystem OPERABLE because of th'e inoperability of either the centrifugal charging pump or the flow path from the' refueling water storage tank, restore at least one ECCS subsystem to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in COLD SHUT 00WN within the next 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months .

b. With no ECCS subsystem OPERABLE because of the inoperability of either the residual heat removal heat exchanger or RHR pump, restore at least one ECCS subsystem to OPERABLE status or uaintain the Reac-tor Coolant System T**9 less than 350'F by use of alternate heat removal methods.

c. In the event the ECCS is actuated and injects water into the Reactor l Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describ-l ing the circumstances of the actuation and the total accumulated actuation cycles to date. The current value of the usage factor for each affected Safety Injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70. '

\

l l

A maximum of one centrifugal charging pump shall be OPERABLE whenever the temperature of one or more of the RCS cold legs is less than or equal to l

350*F.

COMANCHE PEAX - UNIT 1 3/4 5-7 l

l t _ _ _ _ _

EMERGENCY CORE'C00 LING SYSTEMS SURVEILLANCE REQUIREMENTS 4.5.3.1.'1 The.ECCS subsystem shall be demonstrated OPERABLE per the applicable req 0irements of Specification 4.5.2.

4.5.3.2 All charging. pumps, except the above required OPERABLE pumps, shall be demonstrated inoperable

by verifying that the motor circuit breakers are secured in the open position at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months whenever the tempera-ture of one or more of the RCS cold legs is less than or equal-to 350 F.

s I

An inoperable pump may be energized for testing provided the discharge of the pump has been isolated from the RCS by a closed isolation valve with power removed from the valve operator or by a manual isolation valve secured in the closed position.

t l

i COMANCHE PEAK - UNIT 1 3/4 5-8 4 _ , _ _

EMERGENCYeCORE COOLING SYSTEMS j !

3/4.5.3 ECCS SUBSYSTEMS - T < 350 F ava SAFETY INJECTION PUMPS -

LIMITING CONDITION FOR OPERATION 3.5 3.2 All Safety Injection pumps shall be inoperable.

. APPLICABILITY: Modes 4, 5, and 6 with the reactor vessel head on.

ACTION:

With a Safety Injection pump OPERABLE, restore all Safety Injection pumps'to an inoperable status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS 1

~

l 4.5.~3.2 All Safety Injection pumps shall be demonstratea' inoperable

by

. verifying that the motor circuit breakers are secured in the open position.

within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after entering MODE 4 from HODE 3 or pricr to the temperature of one or more of the RCS cold legs decreasing below 325*F, whichever occurs first and at least once per 31 days thereafter. >

1 l

i I -

i

An inoperable pump may be energized fnr testing or for filling occumulators l provided the discharge at the pump has been isolated from the RCS by a closed '

! isolation valve with power removed from the valve operator, or by a manual l

isolation valve secured in the closed position.

l COMANCHE PEAK - UNIT 1 3/4 5-9

') .-'

,. BORON INJECTION SYSTEM -

, {%[$

3/4.5.4 REFUELING WATER STORAGE TANK i

LIMIfING CONDITION FOR OPERATION 3.5.4 The refueling water storage tank (RWST) shall be OPERABLE with:

a. A minimum contained borated water volume of 479,900 gallons

([Later]% of span), '

b. A boron concentration of between 2000 and 2200 ppm of boron, c .' A minimum solution temperature of 40 F, and

d. A maximum solution temperature of 120 F.

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

ACTION.

With the RWST inoperable, restore the tank to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .or be'in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOVN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . ,

SURVEILLANCE REQUIREMENTS 4.5.4 The RWST shall be demonstrated OPERABLE:

a. At least once per 7 days by:

1) Verifying the contained borated water volume in the tank, and

2) Verifying the boron concentration of the water.

b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperatdre when l

the outside air temperature is less than 35 F or greater than i

120*F.

COMANCHE PEAK - UNIT 1 3/4 5-10

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

l

3. 6.1.1 Primary CONTAINMENT INTEGRITY shall be maintained.

JAPPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

Without pri' mary CONT INMENT INTEGRITY, restore CONTAINMENT INTEGRITY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTOOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.,6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated:

a. At least once per 31 days by verifying that all penetrations

not capable.of being closed by OPERABLE containment automatic isolation .

valves and required to be closed during accident conditions are closed by valves, blind flan'ges, or deactivated automatic valves secured in their positions, except as provided in Table 3.6-1 of Specification 3.6.4.1;

b. By verifying that each conta'inment air lock is in compliance with i the requirements of Specification 3.6.1.3; and

c. After each closing of each penetration subject to Type B testing, l except the containment air locks, if opened following a Type A or B i

test, by leak rate testing the seal with gas at a pressure not less than P , 48.3 psig, and verifying that when the measured leakage rate forth$sesealsisaddedtotheleakageratesdeterminedpursuantto Specification 4.6.1.2d. for all other Type B and C penetrations, the combined leakage rate is less than 0.60 L,.

l *Except valves, blind flanges, and deactivated automatic valves which are l located inside the containment and are locked, sealed or otherwise secured i -

in the closed position. These penetrations shall be verified closed during each COLD SHUTOOWN except that such verification need not be performed more

.often than once per 92 days.

l l

l l COMANCHE PEAK - UNIT 1 3/4 6-1 l

CONTAINMENT SYSTEMS CONTAINMENT LEAKAGE .

LIMITING CONDITION FOR OPERATION 3.6.1.2 Containment leakage rates shall' be lim'ited to:

a. An overall integrated leakage rate of:

1) Less than or equal to L,, 0.10% by weight of the containment air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at Pa , 48.3 psig, or

2) Less than or equal to L t, 0.036% by weight of the containment air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at a reduced pressure of Pt , 24.05 psig.

b. A combined leakage rate of less than 0.60 L, for all penetrations '

and valve's subject to Typ B.and'C tdsts, when pressurized to P

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

ACTION: .

With either the measured overall integrated containment leakage rate exceeding 0.75 L, o'r 0.75 L t, as ap'licable, p or the measured combined leakage rate for all penetra'tions and valyes subject to Types B and C tests exceeding 0.60 L '

a restore the overall integrated leakage rate to less than 0.75 L a r less than 0.75 Lt , as applicable, and the combined leakage rate for all penetrations subject to Type B and C tests to less than 0.60 L, prior to increasing the Reactor Coolant System temperature above 200 F.

SURVEILLANCE REQUIREMENTS 4.6.1.2 The containment leakage rates shall be demonstrated at the following l test schedule and shall be~ determined in conformance with the criteria speci-l fied in Appendix J of 10 CFR Part 50 using the methods and provisions of ANSI

! N45.4-1972:

a. Three Type A tests (Overall Integrated Containment Leakage Rate) shall be conducte.d at 40 1 10 month' intervals during shutdown at a pressure not less than either P,, 48.3 psig, or at P t, 24.05 psig, during each 10 year service period. .The th!rd test of each set

. shall be conducted during the shutdown for the 10 year plant inservice inspection; I

COMANCHE PEAK - UNIT 1 3/4 6-2

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

b. If any periodic Type A test fails to meet either 0.75 L r 0.75 Lt '

a the test schedule for .s'ubsequent Ty'pe A tests shall be reviewed and approved by the Commission. If two consecutive Type A tests fail to meet either 0.75 La or 0.75 L t, a Type A test shall be performed at least every 18 months until two consecutive Type A-tests meet either 0.75Lfor0.75L attwhich time the above test schedule may be resumed;

c. The accuracy of each Type A test shall be verified by a supplemental test which:

1) Confirms the accuracy of the test by verifying that the supple-mental test result, L c, is in accordance with the appropriate following equation:

lL e- (L,, + L,) l 5 0.25 La or l Lc -Stm + l )o l 5 0.25 L t where L,, or Ltm is the measured Type A tes.t leakage and t o is the .c.uperimposed leak;

> f

2) .Has a duration sufficient to establish accurately th'e change inl leakage rate between the Type A test and the supp.lemental test;

, and ,

3) Requires that the rate at which gas is injected into the contain-ment or bled from the containmerit during the supplemental test is between 0.75 L a and 1.25 L,; or 0.75 Lt and 1.25 Lt*

d. Type B and C tests shall be conducted with gas at a pressure not less than Pa , 48.3 psig, at intervals no greater than 24 months except for tests involving:

1) Air locks,

2) Containmant ventilation isolation valves with resilient material seals,

e. Air locks shall be tested and demonstrated OPERABLE by the require-ments of Specification 4.6.1.3;

f. Containment ventilation isolation valves with resilient material seals shall be tested and demonstrated OPERABLE by the requirements of Specification 4.6.1.7.3 or 4.6.1.7.4, as applicable;

g. The provisions of Specification 4.0.2 are not applicable.

COMANCHE PEAK - UNIT 1 3/4 6-3

-CONTAINMENT SYSTEMS CONTAINMENT AIR LOCKS J

LIMITING CONDITION FOR OPERATION 3.6.1.3 Each containment air lock shall be OPERABLE with: .

a. Both doors closed except when the air lock is being used for normal transit entry and exit through the containment, then at least one

~

air lock door shall be cl'osed, and

b. An overall air lock leakage rate of less than or equal to 0.05 La at Pj,48.3psig.

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

ACTION:

a. With one containment air lock door inoperable:

1. Maintain at least the OPERABLE air lock door closed and either restore the inoperable air lock door.to OPERABLE status within

. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or . lock the OPERABLE air lock door closed;

, - 2. Operation may then continue until performance of the-~next' .

required overall air lock leakage test p'rovided that the'0PERABLE air lock door is verified to be locked closed at least once per .

31 days;

3. Otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ; and

4. The provisions of Specification 3.0.4 are not applicable.

b. With the containment air lock inoperable,.except as the result of an inoperable air lock door, maintain at least one air lock door closed; restore the inoperable air lock to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . .

I e

O COMANCHE PEAK - UNIT 1 3/4 6-4

CONTAINMENT SYSTEMS 5*

SURVEILLANCE REQUIREMENTS 4.6.1.3 Each containment air lock shall be demonstrated OPERABLE:

a. Within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months foll'owing each closing, except when the air lock is being used for multiple entries, then at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , by verifying seal leakara is less than 0.01 L as determined by precision flowmeasurementswhenmeasuredforatlea$t30secondswiththa '

volume between the seals at a constant pressure of greater than or equal to 48.3 psig;

b. By conducting overall air lock leakage tests at not less than P3, 48.3 psig, and verifying the overall air lock leakage rate is within its limit: -

1) .At least once per 6 months,* and

2) Prior to establishing CONTAINMENT INTEGRITY when maintenance has been performed on the air lock that could affect the air lock sealing capability.**

c. At least once per 6 months by verifying that only one door in each air lock can be opened at.a time.

l l

The provisions of Specification 4.0.2 are not applicable.

- This represents an exemption to 10 CFR 50 Appendix J, paragraph III.D.2(b)(ii).

l COMANCHE PEAK - UNIT 1 3/4 6-5

CONTAINMENT SYSTEMS

~ INTERNAL PRESSURE LIMITING CONDITION FOR OPERATION 3.6.1.4 Primary containment internal pressure shall be maintained between

-0.5 and 1.5 psig.

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

ACTION:

With the. containment internal pressure outside of the limits above, restore the internal pressure to within the limits within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLO SHUT 00WN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS -

4.6.1.4 The primary containment internal pressure shall be determined to be within the limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

e r

0 l

COMANCHE PEAK - UNIT 1 3/4 6-6 t

CONTAINMENT SYSTEMS y 1*

AIR TEMPERATURE LIMITING CONDITION FOR OPERATION 3.6.1.5 Primary containment average air temperature shall not exceed 120 F.

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

ACTION:

With the containment average air temperature greater than 120 F, reduce the average air temperature to within the limit within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN withi,n the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . -

4 SURVEILLANCE REQUIREMENTS 4.6.1.5 The primary containment average air temperature shall be the arith-metical average of the temperatures at the following locations and shall be determined at least once per 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months s:

Location a.

b.

C.

d.

e.

Note: Minimum of three elevations required.

O COMANCHE PEAK - UNIT 1 3/4 6-7

- , - - - - - - - - ~ , . . , -., - ,,

CONTAINMENT SYSTEMS CONTAINMENT STRUCTURAL INTEGRITY F LIMITING CONDITION FOR OPERATION

^

3. 6.1. 6 The structural integrity of the containment shall be maintained at a level consistent with the acceptance criteria in Specification 4.6.1.6.1.

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

ACTION:

With the structural integrity of the containment not conforming to~the above requirements, restore the structural integrity to within the limits within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTOOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.6.1.6.1. Containment Surfaces. The structural integrity of the exposed accessible interior and exterior surfaces of.the containment, including the

~

l.iner plate, shall be determined during the shutdown for each Type A contain-ment leakage rate test (reference Specification 4.6.1.2) by a visual inspection of these surfaces. This inspection shall be performed prior to the Type A containment leakage rate test to verify no apparent changes in appearance or other abnormal degradation.

4.6.1.6.2 Repbrts. Any abnormal degradation of the containment structure detected during the above required inspections shall be reported to the Commis-sion in a Special Report pursuant to Specification 6.9.2 within 15 days.

This report shall include a description of the condition of the concrete, the inspection procedure, the tolerances on cracking, and the corrective actions taken.

l i

COMANCHE PEAK - UNIT 1 3/4 6-8

CONTAINMENT SYSTEMS $1 CONTAINMENT. VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION

3. 6.1. 7 Each containment and hydrogen purge supply and exhaust, isolation valve shall be OPERABLE and:

a. Each 48-inch and 12-inch containment and hydrogen purge supply and exhaust isolation valve shall be locked closed, and

b. The 18-inch contain~ ment pressure relief discharge isolation valve (s) may be open for up to 90 hours0.00104 days 0.025 hours 1.488095e-4 weeks 3.4245e-5 months during a calendar year.

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

ACTION:

a. With any 48-inch or 12-inch containment and hydrogen purge supply and/or exhaust isolation valve open or no.t locked closed, lock close that valve or isolate the penetration (s) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , otherwise be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ,

b. With the 18-inch containment pressure relief discharge isolation valve (s) open for more than 90 hours0.00104 days 0.025 hours 1.488095e-4 weeks 3.4245e-5 months during a calendar year, close the open 18-inch valve (s) or isolate the penetration (s) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , otherwise be in 'at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in COLD SHUT 00WN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c. With a containment pressurc relief discharge isolation valve (s) having a measured leakage rate in excess of the limits of Specifications 4.6.1.7.3 or with the containment and hydrogen purge supply or exhaust isolation valve (s) having a measured leakage rate in excess of the limit of Specification 4.6.1.7.4, restore the inoperable valve (s) to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , otherwise be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in COLD SHUTOOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.6.1.7.1 Each 48-inch and 12-inch containment and hydrogen purge supply and exhaust isolation valve shall be verified to be locked closed at least once per 31 days.

i

! 4.6.1.7.2 The cumulative time that all 18-inch pressure relief discharge isolation valves have been open during a calendar year shall be determined at I

least once per 7 days. .

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l COMANCHE PEAK - UNIT 1 3/4 6-9 l .

l _

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 4.6.1.7.3 At least once per 184 days on a STAGGERED TEST BASIS, the inboard and outboard isolation valves With'risilient material seals in each locked closed 48-inch and 12-inch containment hydrogen purge suoply and exhaust pene-tration shall be demonstrated OPERABLE by verifying that the measured leakage

. rate is less than 0.05 L awhen pressurized to P .

3 4.6.1.7.4 At least once per 92 days each 18-inch containment pressure relief discharge isolation valve with resilient material seals shall be demonstrated OPERABLE by verifying that the measured leakage rate is less than 0.06 L when pressurized to P . a 3

e e

e s

j COMANCHE PEAK - UNIT 1 3/4 6-10

CONTAINMENT SYSTEMS 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS

CONTAINMENT SPRAY SYSTEM .

LIMITING CONDITION FOR OPERATION .

3.6.2.1 Two independent Containment Spray Systems shall be OPERABLE with each Spray System capable of taking suction from the RWST and manually transferring suction to the containment sump. .

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

ACTION:

With one Containment Spray System inoperable, restore the inoperable Containment Spray System to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in-at leas.t HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore the inoperable Containment Spray System to OPERABLE status within the next 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .s or be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.6.2.1 Each Containment S' pray System shall be demonstrated'0PERABLE:

~

a. At least once per 31 days by verifying.that each valve (manual,

' power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct posi~t ion;

b. By verifying that in the test mode each train provides a discharge flow through the test header of greater than or equal.to 5800 gpm with the pump eductor line open when tested pursuant to Specifica-tion 4.0.5;

c. At least once per 18 months during shutdown, by:

1) Verifying that each automatic valve in the flow path actuates to its correct position on a Containment Spray Actuation test signal, and

2) Verifying that each spray pump starts automatically on a Containment Spray Actuation or Safety Injection test signal.

d. At least once per 5 years by performing an air or smoke flow test through each spray header and verifyir,g each s ray nozzle is unobstructed.

O COMANCHE PEAK - UNIT 1 3/4 6-11

L.

CONTAINMENT SYSTEMS , ft,1, t.

SPRAY ADDITIVC S'/ STEM LIMITING CONDITION FOR OPERATION 3.6.2.2 The Spray Additive Syste'm sh31.1 be OPERABLE with:

a. A spray additive tank containing a volume of between 4900 and 5167 gallons of between 28 and 30% by weight Na0H solution, and

b. Two soray additive-eductors each capable of adding Na0H solution from the chemical additive tank to a Containment Spray System pump flow.

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

ACTION: ,

, With the Spray Additi.ve System inoperable, restore the system to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be'in 'at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore'the Spray Additive System to OPERABLE, status within the next 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

'SURVE'ILLANCE REQUIREMENTS 4.6.2.?'TheSprayAdditiveSysteNshallbedemonstratedOPERABLE: .

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

~

b. At least once per 6 months by:

1) Verifying the contained solution volume in the tank, and

2) Verifying the concentration of the NaOH solution by chemical analysis.

c. At least once per 18 months during shutdown, by verifying that each automatic valve in the flow path actuates to its correct position on a Containment Spray Actuation test signal; and

d. At least once per 5 years by verifying:

1) The flow path through the Spray. Additive supply line, and

2) RWST test water flow rates of between 50 GPM and 100 GPM through the eductcr test loop of each of the Spray Additive System.

9 COMANCHE PEAK - UNIT 1 3/4 6-12

CONTAINMENT SYSTEMS p.e 3/4.6.3 CONTAINMENT ISOLATION VALVES -

LIMITING CONDITION FOR OPERATION 3,. 6. 3 The containment isolation valves specified in Table 3.6-1 shall be OPERABLE with isolation times as shown in Table 3.6-1.

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

ACTION:

With one or more of the containment isolation valve (s) specified in Table 3.6-1 inoperable, maintain at least one isolation valve OPERABLE in each '

affected penetration that is open and:

a. Restore the inoperable valve (s) to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , or

b. Isolate each affected penetration within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least one deactivated automatic valve secured in the isolation position, or

c. Isolate each affected pe.netration within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 'by u:;G of at least one closed manual valve or blind flange, or

d. Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 tours.

SURVEILLANCE REQUIREMENTS l

! 4.6.3.1 The containment isolation valves specified in Table 3.6-1 shall be demonstrated OPERABLE prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or its associated actuator, control or power circuit by performance of a cycling test, and verification of isolation time. .

CAUTION: The inocerable isolation valve (s) may be part of a system (s).

Isolating the afi4cted penetration (s) may affect the use'of the system (s).

Consider the technical specification requirements on the affected system (s) and act accordingly.

COMANCHE PEAK - UNIT 1 3/4 6-13

.++

CONTAINMENT SYSTEMS L SURVEILLANCE REQUIREMENTS (Continue'd) 4.6.3.2 Each containment isolation valve specified in Table 3.6-1 shall be . <

demonstrated OPERABLE during the COLD SHUTDOWN or REFUELING MODE at least once-per 18 months by:-

a. Verifying that on a Phase "A" Isolation test signal, each Phase "A" isolation valve actuates to its isolation position;

b. Verifying that on a Phase "B" Isolation test signal, each Phase "B" isolation valve actuates to its isolation position; and

c. Verifying that on a Containr.;ent Ventilation Isolation test signal, each ventilation valve actuates to its isolation position.

4.6.3.3 The isolation time of each power-operated or automatic valve of Table 3.6-1 shall be deter.nined to be within its limit when tested pursuant to Specification 4.0.5.

i

( .

l COMANCHE PEAK - UNIT 1 3/4 6-14

W TABLE 3.6.1 \ _

CONTAINMENT ISOLATION VALVES MAXIMUM ISOLATION TYPE '

FSAR TABLE TIME . LEAK TEST VALVE NO. R'EFERENCE NO.* LINE OR 3ERVICE (Seconds) REQUIREMENTS.

1. Phase "A" Isolction Valves 1HY-2154 20 Fecdwater Sample 5 Nota 1 (FW to Stm Gen #1) .

1HV-2155 22 Sample 5 Note 1 (FW to Stm Gen #2) 1HV-2399 27 B10wdown From Steam 5 Note 1 -

Generator #3 1HV 2398 28 Blowdown From-Steam 5 Note 1 Generator #2 -

1HV-2397 49 Blowdown From Steam 5 Note 1 Generator #1 30.

1 HV-2400 Blowdown From Steam 5 Note 1 Generator.#4 . 3

.1-8152 32 Letdown Line to -

10 C Letdown

Heat Exchanger 1-8160 32 Letdown Line to 10 C -

Letoown Heat Exchanger .

1-8890A 35 RHR to Cold Leg Loops 15 Note 2

1 & #2 Test'Line 1-8890B 36 RHR to Cold Leg loops. 15 Note 2 l

c, #3 & #4 Test Line 1-8047 41 Reactor Makeup Water 10 C to Pressure Relief Tank & RC Ptep Stand Pipe o 1-8843 42 SI to RC System Cold 10 Note 2 i Leg Loops #1, #2, #3, #4 j Test Line 1-8881 43 SI to RC System Hot 10 Note 2 Leg Loops #2 & #3

i. Test Line 1 ,

COMANCHE PEAK - UNIT 1 3/4 6-15

TABLE 3.6.1(Continuedj ,

CONTAINMENT ISOLATION VALVE _S MAXIMUM ISOLATION TYPE FSAR TABLE TIHE LFt% TEST

. VALVE.NO. REFERENCE NO.* LINE OR SERVICE (Seconds) REQUIREMJ H S Phase "A" Isolation Valves (Continued) 1-8824 44 SI to RC System Hot 10 Note 2 Leg loops #1 & #4 Test Line 1-8823 45 SI to RC System Cold 10 Note 2 Leg Loops 1#1, #2, #3,

'& #4 Test Line 1-8100 51 Seal Water Return 10 C and Excess Letdown

~ 1-8112 51 Sea Water Return 10 C and Excess Letdown 1-7136 52 RCDT Heat Exchanger 10 C to Waste Hold Up Tank

LCV-1003 52 RCDT Heat Exchanger 10 C to Waste Hold Up Tank 1HV-53E5 60 Demineralized Water 5 C Supply 1HV-5366 60 Demineralized Water 5 C -

Supply 1HV-5157 61 Containment Sump Pump 5 C :

1 Discharge 1HV-5158 61 Containment Sump Pump 5 C 4

Discharge IHV-3487 62 Instrument Air to 5 C Containment 1-8825 63 RHR to Hot Leg Loops 15 Note 2

2 & #3 Test Line l IHV-2405 73 Sample from Steam 5 Note 1 .

Generator #1 '

1HV-4170 74 RC Sample From Hot 5 C Legs CORNCHE PEAK - UNIT 1 3/4 6-16

TABLE 3.6.1 (Continued) Q%

i"t.4*

CONTAINMENT ISOLATION VALVES -

r:

MAXIMUM ISOLATION TYPE FSAR TABLE TIME LEAK ~ TEST V,ALVE NO. REFERENCE NO.* LINE OR SERVICE .(Seconds) REQUIREMENTS *

1. . Phase "A" Isolation Valves (Continued).

1HV-4168 74 RC Sample From Hot 5 C Leg #1 1HV-4169 74 RC Sample From Hot 5 C Leg #4 1HV-2406 76 Sample from Steam 5 Note 1 Generator #2 1HV-4167 77 Pressurizer Liquid 5 C '

Space Sample -

1HV-4166 77 Pressurizer Liquid 5 C Space Sample 1HV-4176 78 Pressurizer Stea'm 5 C Space Samp1e 1HV-4165 78 Pressurizer Steam 5 C Space Sample

. 1HV-2407 79 Sample from Steam 5 Note 1 Generator #3 1HV-4175 80 Accumulators 5 C l 1HV-4171 80 Sample from 5 C l Accumulator #1 1HV-4172 80 Sample from 5 C Accumulator #2 IHV-4173 80 Sample from 5 C Accumulator #3 1HV-4174 80 Sample from 5 C Accumulator #4 1HV-7311 81 RC PASS Sample 5 C Discharge to RCOT l IHV-7312. 81 RC PASS Sample 5 C

. Discharge to RCDT ,

l COMANCHE PEAK - UNIT 1 3/4 6-17

TABLE 3.6.1 (Continued) py[

CONTAINMENT ISOLATION VALVES -

Ed MAXIMuti FSAR TABLE ISOLATION V_ALVE NO. TYPE REFERENCE NO.* TIME

1. Phase "A" L_INE OR SERVICE . LEAK TEST Isolation Valves (Continued) ISeconds) REQUIREMENT _

1HV-2408

82 Sample from Steam 5

Generator #4 Note 1 8871 83 Accumulator Test and Fill 10 C 1-8888 83 Accumulator Test and Fill 10 C 1-8964 83 Accumulator Test and Fill 10 C 1HV-5556 84 Containment Air PASS Return 5 C 1HV-5557 84 Containment Air PASS Return 5 C 1HV-5544 94~

Radiation Monitoring

Sample 5 C 1HV-5545 94 '

Radiation Monitnring Sample 5 C 1HV-5558 97 Containment Air PASS Inlet 5 C 1HV-5559 97 i

Containment Air PASS Inlet 5 C 1HV-5560 100 Containment Air PASS Inlet 5 C 1HV-5561 100 Containment Air PASS Inlet 5 C i

1HV-5546 102 Radiation Monitoring 5 Sample Return C 1HV-5547 102 '

Radiation Monitoring 5 Sample Return C 1-8880 ,

t 104 ,

i N2 Supply to i 10 C i Accumulators COMAN M PEAK - UNIT 1 3/4 6-18 I

DV TABLE 3.6.1 (Continued) lIG -

CONTAINMENT ISOLATION VALVES MAXIMUM ISOLATION TYPE FSAR TABLE TIME LEAK TEST

. -VALVE NO. REFEREZ E NO.* LINE OR SERVICE , (Seconds) _ REQUIREMENTS

1. Phase "A" Isolation Valves (Continued) -

1-7126 -105 H2 Supply to RC Drain 10 C Tank 1-7150 105 H2 Supply to RC Orain 10 C Tank 1HV-4710 111 CC Supply to Excess 5 Note 1 Letdown & RC Orain Tank Heat Exchanger 1HV-4711 112 CC Return from Excess 5 Note 1 Letdown & RC Orain Tank Heat Exchanger 1HV-3486 113 Service Air to 5 C Containment 1HV-4725 , 114 ' Containment CCW Drain 5 C -

Tank Pumps Discharge 1HV-4726 114 Containment c'CW Drain 5 C Tank Pumps Discharge 1-8027 116 Nitrogen Supply to PRT 10 0 1-8026 116 Nitrogen Supply to PRT 10 C IHV-6084 120 Chilled Water Supply 10 C .

to Containment Coolers

~

1HV-6082 121 Chilled Water Return 10 C From Containment j Coolers 1HV-6083 121 Chilled Water Return 10 C From Containment Coolers .

l 1HV-4075B 124 Fire Protection System 10 C Isolation 1HV-4075C 124 -

Fire Protection System 10 C

. Isolation l

l l

l COMANCHE PEAK - UNIT 1 3/4 6-19 .

TABLE 3.6.1 (Continued) Q,k ,,

CONTAINMENT ISOLATION VALVES MAXIMUM ISOLATION TYPE FSAR TABLE TIME LEAK TEST VALVE NO. REFERENCE NO.* LINE OR SERVICE (Seconds) REQUIREMENTS 2.' ' Phase "B" Isolation Valves 1HV-4708 117 CC Return from RCP's 10 C Motors 1HV-4701 117 CC Return from RCP's 10 C Motors 1HV-4700 118 CC Supply to RCP's 10 C Motors 1HV-4709 119 CC Return From RCP.'s 10 C Thermal Barrier .

1HV-4696 119 CC Return From RCP's 10 C Thermal Sarrier ,

3. Containment Ventilation Isolation Valves 1HV-5542 58 Hydrogen Purge Supply 10 C -

1HV-5543 58 Hydrogen Purge Supply 10 C IHV-5563 58 Hydrogen Purge Supply 10 C 1HV-5540 59 Hydrogen Purge Exhaust 10 0 1HV-L541 59 Hydrogen Purge Exhaust 10 C IHV-5E62 59 Hydrogen Purge Exhaust 10 C 1HV-5536 109 Containment Purge Air 5 C Supply 1HV-E537 109 Containment Purge Air 5 C Supply 1HV-5538 110 Containment Purge Air 5 C Exhaust 1HV-5539 110 Containment Purge Air 5 C Exhaust 1HV-5548 122 Containment P'ressure 3 C Relief COMANCHE PEAK - UNIT 1 3/4 6-20

v

g ,\

A; gi %

. TABLE 3.6.1 (Continued)

CONTAINMENT ISOLATION VALVES

, MAXIMUM ISOLATION TYPE FSAR TABLE TIME LEAK TEST VALVE NO.. REFERENCE NO.* .LINE OR SERVICE (Seconds) REQUIREMENTS

3. Containment Ventilation Isolation Valves (Continued)

IHV-5549 122 Containment Pressure 3 C Relief 4 Manual Valves LMS-711 4a .TDAFW Pump Warm up N.A. Note 1, 11 Valve IMS-390 Sa N2 Supply to Steam N. A. Note 1 Generator #1

^

1MS-387 9a N2 Supply to Steam N.A. Note 1

, Generator #2 1MS-384 13a N2 Supply to Steam N. A. Note 1

-Generator #3 ,

. IMS-712 17a TDAFW Pump Warm-up N.A. Note 1, 11 ~

' Valve 1MS-393 18a N2 Supply to Steam N.A. Note 1 Generator #4 1FW-106 20b li2Supply to Steam N.A. Note 1 '

Generator #1 IFW-104 22b N2 Supply to Steam N.A. Note 1 Generator #2 IFW-110 24 Secondary Sampling N.A. Note 1 IFW-102 24b N2 Supply to Steam N.A. Note 1 Generator #3 1FW-119 25 Secondary Sampling li. A. Note 1 IFW-108 26b N 2 Supply to Steam N.A. Note 1 Generator #4 1-7135 52 RCDT Heat Exchanger to N.A. C Waste Holdup Tank l 1SF-011 56 Refueling Water N.A. C Purification to Refueling Cavity COMANCHE PEAK - UNIT 1 3/4 6-21 l

O TABLE 3.6.1 (Continued)

CONTAINMENT ISOLATION VALVES MAXIMUM .

ISOLATION TYPE FSAR TABLE TIME LEAK TEST VALVE NO. REFERENCE NO.* LINE OR SERVICE (Seconds) REQUIREMENTS

4. Manual Valves (Continued)

ISF-012 56 Refueling Water N.A. C Purification to Refueling Cavity 1SF-021 67 Refueling Cavity to N.A. C Refueling Water Purification Pump ISF-022 67 Refueling Cavity to N.A. C Refueling Water Puri.fication Pump ISF-053 71 Refueling Cavity . N.A. C Skimmer Pump

! Discharge ISF-054 71 Refueling Cavity N. A. C Skimmer Pump .

Discharge 1HV-2333B 2 MSIV Bypass from N.A. Note 1, 6 Steam Generator #1 1HV-2334D 7 MSIV Bypass from N.A. Note 1, 6 Steam Generator #2 1HV-2335B 11 MSIV Bypass from N.A. Note 1, 6 Steam Generator #3 1HV-2336B 15 MSIV Bypass from N.A. Note 1, 6 Steam Generator #4

5. Power-Operated Isolation Valves l 1HV-2452-1 4 Main. Steam to Aux. FPT N.A. Note 1 From Steam Line #1 IPV-2325 5 Steam Generator #1 N.A. Note 1 .

l Atmospharic Relief l IPV-2326 9 Steam Generator #2 N.A. Note 1 Atmospheric Relief i

l COMANCHE PEAK - UNIT 1 3/4 6-22 1

l

TABLE 3.6.1 (Continued)

CONTAINMENT ISOLATION VALVES MAXIMUM ISOLATION TYPE FSAR TABLE TIME LEAK TEST VALVE NO. ' REFERENCE NO.* LINE OR SERVICE (Seconds) REQUIREMENTS

5. Power-0perated Isolation Valves (Continued)

IPV-2327 13 Steam Generator #3 N.A. Note 1 Atmospheric-Relief 1HV-2452-2 17 Main Steam to Aux. FPT N.A. Note 1 From Steam Line #4 1PV-2328 18 Steam Generator #4 N.A. Note 1 Atmospheric Relief 1HV-2491A 20a Auxiliary Feedwater N.A. Note 1 to Steam Generator #1 ,

Auxiliary Feedwater 1HV-2491B 20a N.A. Note 1 to Steam Generator #1 1HV-2492A 22a Auxiliary Feedwater N.A. Note 1

. to Steam Generator #2 1HV-24928 22a Auxiliary Feedwater N.A. Note 1 to Steam Generator #2 1HV-2493A 24a Auxiliary Feedwater N.A. Note 1 to Steam Generator #3 1HV-24938 24a Auxiliary Feedwater N.A. Note 1 to Steam Generator #3 1HV-2494A 26a Auxiliary Feedwater N.A. Note 1 to Steam Generator #4 1HV-2494B 26a Auxiliary Feedwater N.A. Note 1 to Steam Generator #4 1-8701B 33 RHR From Hot Leg N.A. Note 5 Loop #4 1-8701A 34 RHR From Hot Leg N.A. Note 5 Loop #1 1-8809A 35 RHR to Cold Leg Loops N.A. Note 4

1 and #2 1-8809B. 36 RHR to Cold. Leg Loops N.A. Note 4 l #3 and #4 COMANCHE PEAK - UNIT 1 3/4 6-23

~ -~ ~ ~ ~~ ~

TABLE 3.6.1 (Continued)

CONTAINMENT' ISOLATION VALVES

. MAXIMUM ISOLATION TYPE FSAR TABLE TIME LEAK TEST

,. VALVE NO. REFERENCE NO.* LINE OR SERVICE (Seconds) REQUIREMENTS .

5. Power-OperatedIsolai.ionValves(Continued) 1-8801A 42 High Head Safety N.A, Noto 7

. Injection to Cold Leg Loeps #1, #2, #3, & #4 1-8801B 42 High Head Safety N.A. Note 7 Injection to Cold Leg Loops #1, #2, #3, & #4 1-880CA 43 SI Injection to Hot Leg N.A. Note 8 Loops #2 and #3 1-8802B 44 SI Injection co Hot Leg N.A. Note 8 Loops #1 ano #4 1-8835 45 SI Injer' ion to Cold N.A. Note 4 Leg Lo',ps #1, #2, #3,.

and #4 .

1-8351A 47- Seal Injection to RC N.A. 'C Pump (Loop #1) ,

1-8351B 48 Seal Injection to RC N.A. C Pump (Loop #2) 1-8351C 49 Seal Injection to RC N.A. C Pump (Loop #3) l i 1-8351D 50 Seal Injection to RC N.A. C Pump (Loop #4) ,

1HV-4777 54 Containment Spray to N.A. Note 3 Spray Header (Tr. B) l IHV-4776 55 Containment Spray to N.A. Note 3 Spray Header (Tr. A) 1-8840 63 RHR to Hot leg Loops N . A'. Note 8 j #2 and #3 1-8811A 125 Containment Recirc. N.A. Note 1, 10 Sump to RHR Pumps j (Train A) l COMANCHE PEAK - UNIT i 3/4 6-24

TABLE 3.6.1 (Continued) r(-

~

CONTAINMENT ISOLATION VALVES MAXIMUM ISOLATION TYPE FSAR TABLE T,IME LEAK TEST.

VALVE NO. REFERENCE NO.*

LINE OR SERVICE. .-(Seconds)' REQUIREMENTS

5. Power-Operated Isolation Valves (Continued) 1-8811B 126- Containment Recirc. N.A. Note 1, 10 Sump to RHR Pumps (Train B) 1HV-4782 127 Containment Recirc. N.A. Note 1, 10 to Spray Pumps (Train A) 1HV-4783 128 Containment Recire. N.A. Note 1, 10 to Spray Pumps (Train B)

6. Check Valves 1-8818A 35 RHR to Cold Leg N.A. Note 2 Loop #1

'l-8818B 35 RHR ;o Cold. Leg N.A. Note 2

' Loop #2

'l-8818C 36 RHR to Cold Leg N. A. Note 2 Loop #3 1-88180 36 RHR to Cold Leg N.A. Note 2 Loop #4 1-8046 41 Reactor Makeup Water N. A. C to Pressurizer Relief Tank and RC Pump Stand Pipe -

1-8815 42 High Head Safety N. A. Note 2 Injection to Cold Leg loops #1, #2,

3 and #4 ISI-8905A 44 . ' SI to RC System Het N.A. . Note 2 Leg Loop #1 ISI-8905B 43 SI to RC System Hot N.A. Note 2 Leg loop #2 l 151-8905C 43 SI to RC System Hot N. A, Note 2 Leg Loop #3 -

COMANCHE PEAK - UNIT 1 3/4 6-25

i .

TABLE 3.6.1 (Continued)

CONTAINMENT ISOLATION VALVES MAXIMUM ISOLATION TYPE FSAR. TABLE TIME LEAK TEST VALVE NO. REFERENCE NO.* LINE OR SERVICE (Seconds) REQUIREMENTS

6. Check Valves (Continued) '

1SI-8905D 44 SI to RC System Hot N.A. . Note 2 Leg Loop #4 ISI-8819A 45 SI to RC System Cold N.A. Note 2 Leg Loop #1

151-8819B 45 SI to RC System Cold N.A. Note 2 Leg loop #2 ISI-8819C 45 SI to RC System Cold N.A. Note 2 Leg Loop #3 -

151-8819D '45 SI to RC System Cold N.A. Note 2 Leg Loop #4 1-8381 -

46 Charging Line to N.A. C Regenerative Heat .

Exchanger ICS-8368A 47 ' Seal' Injection'.o -

N.A. C RC Pump (Loop #1) 1C5-8368B 48 Seal Injectico to N.A. C RC Pump (Loop #2) 105-8368C 49 Seal Injection to N.A. C RC Pump (Loop #3) 4 1

105-83680 50 Seal Injection to N.A. C RC Pump (Loop #4) 105-8180 51 Seal Water Return and N.A. C '

Excess Letdown ICT-145 54 Containment Spray- to N.A, Note 3 Spray Header (Tr. B)

'1CT-142 55 Containment Spray to N. A. Note 3 Spray Header (Tt*, A) ,

1C1-030 62 Instrument Air to it. A. C Containment 1-8841A -

63 RHR to Hot Leg N.A. Note 2 Loop #2 COMANCHE PEAK - UNIT 1 3/4 6-26

.g i

i TABLE 3.6.1 (Continued) g, CONTAINMENT ISOLATION VALVES

~ '

MAXIMUM ISOLATION TYPE FSAR TABLE . TIME LEAK TEST

. VALVE NO. REFERENCE NO.* LINE OR SERVICE (Seconds) REQUIR g NJ

6. Check Valves (Continued) 1-8841B 63 RHR to Hot leg N.A. Note 2 Loop #3 151-8068 104 N2 Supply To N.A. C Accumulators 1CA-016 113 Service Air to N.A. C Containment ICC-629 117 CC Rtturn from RCP's N.A. C Motors 10C-713 118 CC Supply to RCP's N.A. C Motors ICC-831 11S CC Return from RCP's N.A. C -

Thermal Barrier

ICH-024 120 ' Chilled Water Supply N.A. C to Containment Coolers

7. Steam Line Isolation Signal '

1HV-2333A 1 MSIV #1 5 Note 1, 9, 12

_1HV-2409 3 Drain From Main 5 Note 1 i Steam Line #1 1HV-2334A 6 MSIV #2 5 Note 1, 9,12 IHV-2410 8 Drain From Main 5 Note 1 Steam Line #2 1HV-2335A 10 MSIV #3 5 Note 1, 9, 12 1HV-2411 12 Drain From Main 5 Note 1 Steam Lina #3 ,

1HV-2336A 14 MSIV #4 5 Note 1, 9, 12 IHV-2412 16 '

Drain From Main 5 Note 1 Steam Line #4 COMANCHE PEAK - UNIT 1 3/4 6-27 l

TABLE 3.6.1 (Continued)

CONTAINMENT ISOLATION VALVES MAXIMUM ISOLATION TYPE FSAR TABLE . TIME LEAK TEST VALVE NO. REFERENCE NO.* LINE OR SERVICE (Seconds) REQUIREMENTS

8. Feedwater Line Isolation Signal 1HV-2134 19 Feedwater Isolation 5 Note 1, 12 Steam Generator #1 1FV-2193 2dc .Feedwo er Bypass 5 Note 1, 12

, Line

, 1HV-2185 20d Feedwater Isolation 5 Note 1, 12 Bypass Line 1HV-2135 21 Feedwater Isolation 5 Note 1, 12 Steam Generator #2 1FV-2194 22c Feedwater Bypass 5 Note 1, 12 Line 1HV-2186 -22d Feedwater I. solation 5 Note 1, 12.

Bypass Line 1HV-2136 23 Feedwater Isolation 5 Note 1, 12 Steam Generator #3 IFV-2195 24c Feedwater Bypass 5 Note 1, 12 Line 1HV-2187 24d Feedwater Isolation 5 Note 1, 12 Bypass Line 1HV-2137 25 Feedwater Isolation 5 Note 1, 12 Steam Generator #4 1FV-2196 26d Feedwater Bypas; 5 Note 1, 12

, Line -

1HV-2188 26e' Feedwater Isolation 5 Note 1, 12 Bypass Line

9. Safety Injection Actuation Isolation 1-8105 4G Charging Line to 10 C Regenerative Heat Exchanger COMANCHE PEAK - UNIT 1 3/4 6-28 1 .._ __ _ .-

w.

LABLE 3.6.1 (Continued) .

CONTAINMENT ISOLATION VALVES MAXIMUM ISOLATION TYPE FSAR TABLE VALVE NO. TIME LEAK TEST REFERENCE NO.* LINE OR SERVICE (Seconds) REOUIREMENTS

10. Relief Valves 1-87088 33 RHR From Hot Leg N.A. Note 5 Loop #4 1-8708A 34 RHR From Hot Leg N.A. Note 5 Loop #1 1MS-021 Sb Main Steam Safety [N.A.] Note 1, 12 Valve S.G. #1 IMS-022 5b Main Steam Safety [N.A.] Note 1, 12 Valve S.G. #1 1MS-023 Sb Main Steam Safety [N.A.] Note 1, 12 Valve S.G. #1 1

IMS-024 5b

Main Steam Safety [N.A.] Note 1, 12 Valve S.G. #1.

1MS-025 Sb , Main Steam Safety' [N.A.] Note 1, 12 '

Valve S.G. #1 1MS-058 9b Main Steam Safety [N.A.] Note 1, 12 Valve S.G. #2 1MS-059 9b Main Steam Safety [N.A.] Note 1, 12 Valve S.G. #2 1MS-060 9b Main Steam Safety [N.A.] Note 1, 12 Valve S.G. #2 1MS-061 9b Main Steam Safety [N.A.] Note 1, 12 Valve S.G. #2 1MS-062 9b l Main Steam Safety [N.A.] Note 1, 12 Valve S.G. #2 1MS-09 13b Main Steam Safety [N.A.]* Note 1, 12 Valve S.G. #3 1MS-094 13b Main Steam Safety [N.A.] Note 1, 12

,. Valve S.G. #3 1MS-095 13b Main Steam Safety [N.A.] Note 1, 12 Valve S.G. #3 J COMANCHE PEAK - UNIT 1 3/4 6-29

TABLE 3.6.1 (Continued)

CONTAINMENT ISOLATION VALVES MAXIMdM ISOLATION TYPE .

FSAR TABLE . TIME LEAK TEST VALVE NO. REFERENCE NO.* LINE OR SERVICE (Seconds) REQUIREMENTS

10. Relief Valves (Continued) 1MS-096 13b Main Steam Safety [N.A.] Note 1, 12 Valve S.G. #3 1MS-097 13b Main Steam Safety [N.A.] Note 1, 12 Valve S.G. #3 1MS-129 18b Main Steam Safety [N.A.] Note 1, 12 Valve S.G. #4 1MS-130 18b Main Steam Safety [N.A.] N-te 1; 12 Valve S.G. #4 1M3-131 18b Main Steam Safety [N.A.] Note 1, 12 Valve S.G. #4

'1MS-132 18b Main Steam Safety' [N.A.] Note 1,'12 Valve S.G. #4 1MS-133 18b Msin Steam Safety [N.A.] Note l', 12 Valve S.G. #4 1RC-036 41 RMUW to PRT & RCP N.A. C Standpipe IWP-7176 52 RCDT HX to Waste N.A. C Holdup Tank 100-430 60 Demineralized Water N.A. C Supply IVD-907 61 Cont. Sump Pump N.A. C Discharge IPS-193 80 Sample from N.A. C Accumulators 10C-1067 114 Containment CCW Orain N.A. C Tank Pump / Discharge ICH-271 120 Chilled Water Supply N.A. C to Cont. Coolers ICH-272 121 Chilled W'ater Supply N.A. C from Cont. Coolers COMANCHE PEAK - UNIT 1 3/4 6-30

TABLE 3.6.1 (Continued)

TABLE NOTATIONS

Identification code for containment penetration and associated isolation valves in FSAR Tables 6.2.4-1, 6.2.4-2, and 6.2.4-3.

Note 1: These are closed systems which meet the requirements of NUREG-0800 Section 6.2.4, II.'6, paragraph o. These valves are therefore not required to be leak tested. '

' Note 2: These valves inside containment are part of closed systems outside containment which are in service post accident at a pressure in excess of containment design pressure and satisfy single failure criterion. These valves are therefore not required to be leak tested.

Note 3: These are closed systems outside containment which.are in service post accident and have a water-filled loop seal on the containment side of the valves for a period greater than 30 days following the

. accident. These valves are therefore leak rate tested with water at a pressure of P,.

Note 4: These ESF salves are normally open and remain open during post-accident conditions. Postaccident they are continually pressurized ih excess of containment pressure from an ESF source which meets the single failur.e. criterion. 'These valves are therefore not required to be leak tested.

Note 5:' An effective fluid seal on these' penetrations is provided by the suction sour ~ces to the residual heat removal pumps during'and fol-lowing an accident. In addition, these containment isolation valves

, .are non-automatic, are not required to operate postaccident and are located inside containment. These valves are therefore not required to be leak tested.

Note 6: All four MSIV bypass valves are locked closed in Mode 1. During Modes 2, 3, and 4, one MSIV bypass valve may be opened provided i the other three MSIV bypass valves are locked closed and their associated MSIVs are closed.

Note 7: These are parallel ESF valves that are normally closed, but are

. designed to open during post-accident conditions. Failure of one valve to open will not prevent system pressurization on both sides of both valves in excess of containment pressure. These valves are i therefore not required to be leak tested.

Note 8: These valves located outside containment are normally closed and see a pressure in excess. of containment pressure in post-accident conditions. A valve stem leakacje check will be performed cn a quarterly basis to assure no significant stem leakage would occur in post-accident conditions. ,

Note 9: These valves require steam to be tested are are thus not required to be tested until the plant is in MODE 3.

COMANCHE PEAK - L' NIT 1 3/4 6-31 l

,i

. a, 4

TABLE 3.6.1 (Continued)

TABLE NOTATIONS . ;

Note 10: These valves will have water against them during post-accident conditions to preclude any release of containment atmosphere to the environment. , ,

Note 11: These valves are normally locked closed and are open only to warm-up the steam supply li'nes prior to normal surveillance testing.

Note 12: These valves are included for table completeness, the requirements of Specification 3.6.3 do not apply. Instead, the requirements of Specification 3.7.1.1, 3.7.1.5 and 3.7.1.6 apply for main steam safety valves, main steam isolation valves and feedwater isolation valved, respectively.

l 4

4 COMANCHE PEAK - UNIT 1 3/4 6-32

a CONTAINMENT SYSTEMS 3/4.6.4 COMBUSTIBLE GAS CONTROL HYOROGEN MONITORS LIMITING-CONDITION FOR OPERATION .

'3.6.4.1 Two independent containment hydrogen monitor trains (with at least one channel per train) shall_be OPERABLE.

APPLICABILITY: MODES 1 and 2.

ACTION:

a. With one hydrogen monitor train inoperable, restore the inoperable monitor train to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 6' hours.

b.. With both hydr' ogen monitor trains inoperable, restore at least one ' monitor train to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY i

within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS .

4.6.4.1 Each hydrogen monitor shall be demonstrated OPERABLE by the per-formance of a CHANNEL CHECK at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , an ANALOG CHANNEL OPERATIONAL TEST at least once per 31 days, and at least once per 92 days on 3 a STAGGERED TEST BASIS by performing a CHANNEL CALIBRATION using sample gas containing:

~

a. One volume per. cent hydrogen, balance nitrogen, and

b. Four. volume percent hydrogen, balance nitrogen.

i

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! COMANCHE PEAK - UNIT 1 3/4 6-33 l

4

, _ CONTAINMENT _ SYSTEMS _

ELECTRIC HYOROGEN RECOMBINERS

{.

LIMITING CONDITION FOR OPERATION 3.6.4.2 !

Two independent Hydrogen Recombiner Systems shall be OPE APPLICABILITY: MODES I and 2.

ACTION:

4 With one Hydrogen Recombiner System inoperable, restore the e system to OPERABLE next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . status within 30 days or be in at least nHOT the STANDB

_ SURVEILLANCE REQUIREMENTS t.

L 4.6.4.2 Each Hydrogen Recombiner System shall be demonstrated OPE !

a. -

System functional test, that the minimum r he

{

increases to greater than or equal to 700*F within 90 minutes v

Upon to 60 reaching 700'F, increase the power, s equal kW, and

b. i At least once per 18 months by: i

1) !

tion and control circuits, Performing a CHANNEL CALIB 2)

Verifying through a visual examination that there is no t (i.e. , loose wiring or structural connections, d l 1 foreign materials, etc.), and 4 3) performing required functional test.

a resistance to ground test followin heater phase shall be greater than or equal to 10,000 ohm 4

b COMANCHE PEAK - UNIT 1 3/4 6-34 i

ll

3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE k,

. SAFETY VALVES LIMITING CONDITION FOR OPERATION -

3.7.1.1 All main steam line Code safety valves associated with each steam generator shall be OPERABLE with lift settings as specified in Table 3.7-2.

APPLICABILITY: % DES 1, 2, and 3.

ACTION:

a. With four reactor coolant loops and associated steam generators in operation and with one or more main steam line Code safety valves

inoperable, operation in MODES 1, 2, and 3 may proceed provided, that within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , aither the inoperable valve is restored to OPERABLE status or the Power Range Neutron Flux High Trip Setpoint is reduced par Table 3.7-1; otherwise, be in at'least HOT STANDBY vithin the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUT 00WN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ,

b. The provision ~s of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS , 4.7.1.1 No additional requirements other thar, those required by Specification 4.0.5.

r e

COMANCHE PEAK - UNIT 1 3/4 7-1

t TABLE 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES OURING FOUR LOOP OPERATION MAXIMUM NUMBER OF INOPERABLE MAXIMUM ALLOWABLE POWER RANGE SAFETY VALVES ON ANY ,

NEUTRON FLUX HIGH SETPOINT.

OPERATING STEAM GENERATOR (PERCENT OF RATED THERMAL POWER) .

1 87 2 65 3 43 TABLE 3.7-2

=

I STEAM LINE SAFETY VALVES PER LOOP -

VALVE NUMBER LIFT SETTING (i 1%)* ORIFICE SIZE LOOP 1 LOOP 2 ' LOOP 3 LOOP 4 1MS-021, 058, 093, 029 1185 psig 16 in 2 1MS-022, OS9, 094, 130 1195 psig 16 in 2

IMS-023, 060, 095, 131 1205 psig 16 in 2 1MS-024, 061, 096, 132 1215 psig 16 in 2 ,

IMS-025, 062, 097, 133 1235 psig 16 in 2 l

The lift setting pressure shall correspond to ambient conditions of the. i valve at nominal operating temperature and pressure. ,

4 COMANCHE PEAK - UNIT 1 3/4 7-2

PLANT SYSTEMS AUXILIARY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1 2 At least three independent steam generator auxiliary feedwater pumps and associated flow paths shall be OPERABLE with:

a. Two motor-driven auxiliary feedwater pumps, cach capable of being d powered from separate emerg6ncy busses, and

b. One steam turbine-driven auxiliary feedwatte pump capable of being powered from an OPERABLE steam supply system.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

a. With one auxiliary feedwater pump or associated flow path inoper- .

able, restore the required auxiliary feedwater pumps or associated flow path to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT ,

STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the fo11'owing 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. With two auxiliary feedwater pumps 'or associated flow paths inoper-able, be in at least HOT STANDBY witin 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUT 00WN i within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,

c. With three auxiliary feedwater pumps or associated flow paths inor, etable, immediately initiate corrective action to restore at least one auxiliary feedwater pump to OPERABLE status as soon as possible. 4 SURVEILLANCE REQUIREMENTS

4.7.1.2.1 Each auxiliary feedwater pump and associated flow path shall be demonstrated OPERABLE:

a. At least once per 31 days on a STAGGERED TEST BASIS by:

1) Verifying that each motor-driven pump develops a discharge pressure of greater than or equal to (Later) psig at a flow of greater than or equal to 430 gpm;

2) Verifying tha't the steam turbine-driven pump develops a dis-charge pressure of greater than or equal to [Later) psig at a '

flow of greater than or equal to 860 gpm when the secondary steam supply pressure is greater than 532 psig. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3; ;

I COMANCHE PEAK - UNIT 1 3/4 7-3

[ ' i.

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

3) Verifying that each non-automatic valve in the flow path that -

, is not locked, sealed, or otherwise secured in position is in its correct position; and'

4) Verifying that each automatic valve in the flow path is in the fully open position whenever the Auxiliary Feedwater System is in standby for au.y.iliary feedwater automatic initiation or when above 10% RATED THERMAL POWER.

b. At least once per 18 months during shutdown by:

1) Verifying that each automatic valve in the flow path actuates to its correct position upon receipt of an Auxiliary feedwater Actuation test signal, and

2) Verifying that each auxiliary feedwater pump starts as designed automatically upon receipt of an Auxiliary Feedwater Actuation i test signal. The provisions of Specification 4.0,4 are not applicable to the turbine driven auxiliary feedwater pump for entry into Mode 3. '

1 i

, =

i s

)

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COMANCHE PEAK - UNIT 1 3/4 7-4 I

PLANT SYSTEMS g rt E CONDENSATE STORAGE TANK

4 '"

LIMITING CONDITION FOR' OPERATION

. 3.7.1.3 The condensate storage tank (CST) shall'be OPERABLE with a contained water. volume of at least 282,540 gallons ( % of span) of water.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

With the CST inoperable, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either:

a. Restore the CST to OPERABLE status or be in at least HOT

' STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTOOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , or b.

Demonstrate the OPERABILITY of the Station Service Water (SSW) system as a backup supply to the auxiliary feedwater pumps and restore the CST to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTOOWN within the' following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCEREQUiREMENTS 4.7.1.3.1 The CST shall be demonstrated OPERABLE at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying the contained water volume is within its limits when the tank is the supply source for the auxiliary feedwater pumps.

4.7.1.3.2 The SSW system shall be demonstrated OPERABLE at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months whenever the SSW system is being used as an alternate supply source to the auxiliary fec-dwater pumps by verifying the SSW system operable ar.d each i

motor operated valve between the SSW system and each operable auxiliary feed -

water pump is operable.

4 1

COMANCHE PEAK - UNIT 1 3/4 7-5

i ,

PLANT SYSTEMS '

SPECIFIC ACTIVITY '

LIMITING CONDITION FOR OPERATION 3.7.1.4 The. specific activ'ity of the Secondary Coolant. System shall be less than'or equal to 0.1 microcurie / gram DOSE EQUIVALENT I-131.

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

ACTION:

With the spe.cific activity of the Secondary Coolant System greater than 0.1 microcurie / gram DOSE EQUIVALENT I-131, be in at lear,t HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 ours.

l SURVEILLANCE REQUIREMENTS

4. 7.1. '4 The specific activity of the Secondary Coolant System shall bE

' determined to be within the limit by performance of the sampling end analysis program of Table 4.7-1.

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l

\

e C0KANCHE PEAK - UNIT 1 3/4 7-6

k TABLE 4.7-1

SECONDARY COOLANT SYSTEM. SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM TYPE OF MEASUREMENT. .

' SAMPLE AND ANALYSIS AND ANALYSIS ,

FREQUENCY '

1. Gross Radioactivity .

At least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

Determination *

2. Isotopic Analysis for DOSE a) Once per 31 days, when-EQUIVALENT I-131 Concentration ever the gross radio-activity de. termination indicates concentrations greater than 10% of the

allowable limit for radiciodines.

b) Once per 6 months, when-ever the gross radio-activity determination indicates concentrations less than or equal to 10%

of the allowable limit for radiofodines. '

l L

I i

A gross radioactivity analysis shall consist of the quantitative measurement of the total specific activity of the secondary coolant except for radio-l nuclides with half-lives less than 10 minutes. Determination of the '

contributors to the gross specific activity shall be ' ased c upon those energy i peaks identifiable with a 95% confidence level.

l COMANCHE PEAK - UNIT 1 3/4 7-7 I

1 PLANT SYSTEMS '

MAIN STEAM LINE ISOLATION VALVES ([.1 LIMITING CONDITION FOR OPERATION 3.7.1.5 Each main' steam line isolation valve (MSIV) shall be OPERABLE.

' APPLICABILITY: MODES 1, 2, and 3. !

ACTION:

MODE 1:

With one MSIV inoperable but open, POWER OPERATION may continue n provided the inoperable valve is restored to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; otherwise be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . -

MODES 2 and 3:

With one MSIV inoperable, subsequent operation in MODE 2 or 3 may proceed provided the isolation valve is maintained closed. Otherwise, be in HOT STANDBY'within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS , 4.7.1.5 Each MSIV shall be demonstrated OPERABLE by verifying full closure within 5 seconds when tested pursuant to Specification 4.0.5. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3.

k 4 l' b

8 1

COMANCHE PEAK - UNIT 1 3/4 7-8 l

~

1 ,

.. m j

~

e. p l PLANT SYSTEMS %

3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATIR LIMITING CONDITION FOR OPERATION M

3.7.2 The temperatures of both the primary and secondary coolants in the-steam generators shall be greater than 70*F when the pressure of either coolant in the steam generator is greater than 200 psig.

APPLICABILITY: At all times.

. ACTION: -

With the requirements of the above specification not satisfied:

, a. Reduce the steam generator pressure of the applicable side to

! less than or equal to 200 psig within 30 minutes, and

b. Perform an eng'ineering evaluation to P termine tha effect of the overpressurization on the structural integrity of the steam generator. Detenaias that the steam generator remains acceptable for continumj operation prior to increasing its temperatures abo've'200 F.

SURVEILLANPE REQUIREMENTS

,, 4.7.2 The pressure in each side o.' the steam generator shall be determined to be less than 200 p;ig at least once per hour when the temperature of either the primary or secondary coolant is'less than 70 F. -

l l

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1 l

l COMANCHE PEAK - UNIT 1 3/4 7-9

rhd PLANT SYSTEMS ) 5 3/4.7.3 COMPONENT COOLING WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.3 At least two independent component cooling water loops shall be OPERABLE.

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

ACTION:

With only one component cooling water loop OPERABLE, restore at least two loops to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVElLLANCE REQUIREMENTS 4.7.3 Each component cooling water loop shall be demonstrated GPERABLE:

a. At least once per 31 days by verifying that cach valve (manual, .

power-operated, or automatic) Jervicing safety related equipment that ,

is not' locked,. sealed, or otherwise secured in position is in i.ts correct position; and ,

b '. At least once per 18 months d'uring shutdown,-by verifying that:

1) Each automatic valve sorvicing safety-related equipment actuates to its correct position on its associated engineering safety feature actuation signal, and

2) Each Component Coolina Water System pump starts automatically 65 i Safely injsction test signal.

C COMANCHE PEAK - UNIT 1 3/4 7-10

PLANT SYSTEMS 94 to 3/4.7.4 STATION SERVICE WATER SYSTEM LIMITING CONDITION FOR OPERATION f

'3.7.4 At least two independent service water loops shall be OPERABLE.

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

ACTION:

With only one service water loop OPERABLE, restore at least two loops to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least ' DOT 5TANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and if COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

. SURVEILLANCE REQUIREMENTS 4.7.4 E.ach service water loop shall be demonstrated OPERABLE:

a. At least once p're 31 days by verifying that each valve (manual, power-operated, or automatic) servicing safety-related equipment that

,- is not locked, sealed, or otherwise secured in position is in its correct position; and i ,

b. At least once per 18 months during shutdown, by verifying that:

I i 1) Each automatic volve servicing safety-related equipment actuates to its ccrrect position on a Safety Injection test signal, and l 2) Each station service water system pump starts automatically on i a Safety Injection test signal.

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C0HANCHE PEAK - UNIT 1 3/4 7-11

PLANT SYSTEMS giy 3/4.7.5 bLTIMATEHEATSINK p .

O LIMITING CONDITION FOR OPERATION 3.7.5 The ultimate heat sink (VHS) shall be OPERABLE with: -

a. A minimum water level at or above elevation 770 Mean Sea Level, USGS datum,'and

b. A station service water intake temperature of less than or equal to ,

109. F, and

c. A maximum avc.' age sediment deptn of less than or equal to 1.5 feet in the service w&ter intake channel.

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

ACTION.

a. With be requirements for water level and intake temperature not satisfied, be.in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLJ SHUT 00WN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . .

b- ifith the average sediment de~pth in the service water channel greater than 1.5 feet, the channel shall be cleaned within 30 days to reduce the average sediment depth to less than 0.5 feet.

SURVEILLANCE REQUIREMENTS 4.7.5 The ultimate heat sink shall be determined OPERABLE:

a. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the station service water

,- intake temperature and UHS water level to be within their limits.

l

b. At least once per 12 months by visually inspecting the dam and verifying no abnormal oegradation or erosion, and

c. At least once per 12 months by verifying that the average. sediment cepth in the service water intake channel is less than or equal to

. 1.5 feet.

O COMANCHE PEAK - UNIT 1 3/4 7-12

PLAthSYSTEMS +t 3/4.7.6 FLOOD PROTECTION-LIMITING CONDITION FOR OPERATION 3.7.6 Flood protection shall be provided for all safety related systems,

-components, and structures when the water level of the Squaw Creek Reservoir (SCR) exceeds 777.5 Mean Sea Level, U5GS datum. .

APPLILABILITY: At all times.

ACTION:

i With the water level of SCR above elevation 777.5 Mean Sea level, USGS datum, initiate and complete within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , the flood protection measures verifying that any equipment which is to be opened or is opened for maintenance is isolated from the SCR by isolation. valves, or stop gates, or is at an elevation above 790 feet.

4 SURVEILLANCE REQUIREMENTS 4.7.6 The water level of SCR shall be determined to be within the limits by:

a. Heas'u'rement at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the water level is below elevatioi 776 Mean Sea Level .USGS datum, and

b. Measureme1t at least once per 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months when the water level is equal to or above elevation 776 Mean Sea Level, USGS datum,

c. With the water level of SCR above 777.0 Mean Sea Level, USGS datum, verify flood protection measures are in effect by verifying once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that flow paths from the SCR wttch are open for l maintenance are isolated from the SCR by isol tion valves, or stop I

gates, or are at an elevation above 790 feet.

e C0HANCHE PEAK - UNIT 1 3/4 7-13

. i i

l PLANT SYSTEMS 3/4.7.7 CONTROL ROOM HVAC SYSTEM j LIMITING CONDITION FOR OPERATION 3.7.7 Two independent Control Room HVAC trains shall be OPERABLE' .

APPLICABILITY: All MODES.

ACTION:

MODES 1, 2, 3 and 4:

With one Control Room HVAC train inoperable, restore the inoperabic t~ain r to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD' SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

, MODES 5 and 6:

a. With one Control Room HVAC train inoperable, restore the inoperable system to OPERABLE status within 7 days or initiate and maintain operation of the remaining OPERABLE Control Room HVAC train in .the emergency recirculation mode. .

b. With both Control Room HVAC trains inoperable, or with the OPERABLE Control Room'HVAC trains required to be in the emergency recircula-

. tion mode.by ACTION a , not capable of being powered by an 00ERABLE emergency power source, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

SURVEILLANCE REQUIREMENTS l

4.7.7 Each Control Room HVAC train shall be demonstrated OPERABLE: -

a. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that the control room air '

temperature is less than or equal to 80 F;

b. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters and charcoal absorbers and verifying that the train operates for at least 10 continuous hours with the emergency pressurization unit heaters operating;

COMANCHE PEAK - UNIT 1 3/4 7-14

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

c. At least'once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following paintirig, fire, or chemical release in any ventilation zone communi-cating with the system by:

1) Verifying that the filtration unit satisfies the in place pene-tration and bypass leakage testing acceptance criteria of less than 0.05% by using the test procedure guidance in Regulatory Position C.S.a C.5.c, and C.5.d of Regulatory Guide 1.52, Revisions 2, March 1978, and the emergency filtration unit flow rate is 8000 cfm i 10%, and the emergency pressurization unit flow rate is 800 cfm i 10%;

2) Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accor-I dance with Regulatory Position C.6.b of Regulatory Guide 1.52, l Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position ~C.6.a'of Regulatory Guide 1.52, Revi-sion 2, March 1978, for a methyl iodide penetration of less than 0.2%; and

3) Verifying an emergency filtration unit flow rate of 8000 cfm:

110% and an emergency pressurization unit flow rate of. 800 cfm i 10% during system operation when tested in accordance with ANSI N510-1975. .

'd. Af ter every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation, by verifying, within 31 days after removal, that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Regulatory Position C.6.b of Regu"atory Guide 1.52, Revision 2, March 1978, meets the laboratory tt Mg criteria of Regulatory Position C.6.a of Regalatory Guide 1.52, Revision 2, March 1978, for a methyl iodide penetration of less than 0.2%;

e. At least once per 18 months by:

1) Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 7.7 inches l

Water Gauge while operating the emergency filtration unit at a flow rate of 8000 ch 110%, and is less than 9.25 inches Water Gauge while operating the emergency pressurization unit at a l flow rate of 800 cfm i 10%;

2) Verifying that on a Safety Injection, Loss-of-Of fsite Power, Intake Vent-High Radiation, or Plant Vent-High Radiation test signal, the train automatically switches into the emergency recirculation mode of operation with flow through the HEPA filters and charcoal adsorber banks;

3) Verifying that the emergency pressurization unit maintains the l control room at a positive pressure of greater than or equal COMANCHE PEAK - UNIT 1 3/4 7-15

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) to 1/8 inch Water Gauge relative to the adjacent areas,.

including t:he outside atmosphere, at a flow rate of less than or equal to 800 cfm during' system operation;

4) Verifying that the heaters in the emergency pressurization units dissipate 10 + 1 kW when tested in accordance with ANSI N510-1975; and

5) Verifying that on a High Chlorine test signal, the train auto-matically switches into the isolation mode of operation with flow through the emergency filtration HEPA filters and char-coal adsorber banks within 10 seconds,

f. After each complete or partial replacement of a HEPA filter bank in the emergency filtration unit (s), by verifying that the unit satis-fies the in place penetration and bypass leakage testing acceptance criteria of less than 0.05% in accordance with ANSI N510-1975 for a DOP test aerosol while operating the unit at a flow rate of 8000 cfm i 10%;

g. After each complete or partial replacement of a charcoal adorber bank in the emergency filtration unit (s), by verifying that the unit satisfies the in place penetration an.d bypass leakage testing

, acceptance criteria of less than 0.05% in accordance,with ANSI .

N510-1975 for a halogenated hydrocarbon refrigerant test gas'while operating the unit at a flow rate of 8000 cfm i 1.0%;

h. After each complete or partial replacement of a HEPA filter bank in the emergency pressurization unit (s), by verifying that the unit satisfies the in place penetration and bypass leakage testing acceptance criteria of less than 0.05% in accordance with ANSI

.N510-1975 for a 00P test aerosol while operating the unit at a flow rate of 800 cfm i 10%; and i,

After each complete or partial replacement of a charcoal absorber bank in the emergency pressurization unit (s), by verifying that the unit satisfies the in place penetration and bypass leakage testing acceptance criteria of less than 0.05% in accordance with ANSI N510-1975 for a halogenated hydrocarbon refrigerant test gas while operating the unit at a flow rate of 800 cfm i 10%.

O COMANCHE PEAK - UNIT 1 3/4 7-16 ,

PLANT SYSTdMS ,

3/4.7.8 PRIMARY PLANT VENTILATION SYSTEM - ESF FILTRATION UNITS LIMITING CONDITION FOR OPERATION 3.7.8 Two independent ESF Filtration Units shall be OPERABLE.

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

ACTION:

With one ESF Filtration Unit inoperable, restore the inoperable ESF Filtration Unit to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.7.8 Each ESF Filtration Unit shall be demonstrated OPERABLE:

a. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that each ESF Filt' ration Unit ope' rates for at least 10 continuous hours with the heaters operating;

b. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical rela se in any ventilation zone com-municating with the system by:

1) Verifying that each ESF Filtration Unit satisfies the in place penetration and bypass leakage, testing acceptance criteria of less than 1.0% by using the test procedure guidance in Regula-tory Positions C.S.a, C.5.c, and C.S.d of Regulatory Guide 1.52, Revision 2, March 1978, and verifying the flow rate is 15,000 cfm i 10% per ESF Filtration Unit when tested in accordan'ce with ANSI H510-1975; and

2) Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accor-dance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revi-sion 2, March 1978, for a methyl iodide penetration of less than 1.0%.

COMANCHE' PEAK - UNIT 1 3/4 7-17

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

) c. After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation, by verifying, within 31 days after removal, that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Regulatory Posi' tion C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, -

meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl iodide penetration of less than 1.0%;

d. At least once per 18 months by:

1) Verifying that the pressure drop across the combined HEPA filters and charcoal acsorber banks is less than 8.25 inches Water Gauge while operating each ESF Filtration Unit at a flow rate of 15,000 cfm 1 10%,

2) Verifying that each ESF Filtration Unit starts on a Safety Injection test signal, and 3). Verifying that the heaters dissipate 100 1 5 kW when tested in accordance with ANSI N510-1975.

e. After each complete or partial replacement of a HEPA filt'er bank, by verifying that the associated ESF Filtration Unit satisfies the in place penetration and bypass leakage testing acceptance criteria of less than 1.0% in accordance with ANSI N510-1975 for a 00P test aerosol while operating the associated ESF Filtration Unit at a flow rate of 15,000 cfm i 10%; and

f. After each complete or partial replacement of a charcoal adsorber bank, by verifying that the associated ESF Filtration Unit satis-fies the in-place penetration and bypass leakage testing acceptance criteria of less than 1.0% in accordance with ANSI N510-1975 for a halogenated hydrocarbon refrigerant test gas while operating the system at a flow rate of 15,000 cfm 10%.

COMANCHE PEAK - UNIT 1 3/4 7-18 l

_ - _ - _ _ . . - _ - . - - . - - - '- - - ' - - - ' - -- - ~^

PLANT $YSTEMS 3/4.7.9 SNUBBERS -

LIMITING CONDITION FOR OPERATION 3.7.9 All snubbers shall be OPERABLE. The only snubbers excluded from the requjrements are those installed on nonsafety related systems and then only if'their fail'ure of. failure of the system on which they are installed would have no adverse effect on any safety-related system.

APPLICABILITY: MODES 1, 2, 3, and 4. MODES 5 and 6 for snubbers located on systems required OPERABLE in thosa MODES.

ACTION:

With one or more snubbers inoperable on any system, within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months replace or re-store the inoperable snubber (s) to OPERABLE status ana perform an en'gineering eval-uation per Specification 4.7.9 ,9 on the attached component or declare the attached system inoperable and follow the appropriate ACTION statement for that system.

SURVEILLANCE REQUIREMENTS 4.7.9 fach snubber sh'al'l.be demonstrated OPERABLE by performance'of the following augmented i'nservice inspection program in addition to the require-ments of Specification 4.0.5.

a. Inspection Types As used in this specification, type of snubber shall mean snubbers of the same design and manufacturer, irrespective of capacity.

b. Visual Inspections

Snubbers are categorized as inaccessjble or accessible during reactor operation. Each of these groups (inaccessible and accessible) may be inspected independently according to the schedule below. The first inservice visual inspection of each type of snubber shall be performed after 4 months but within 10 months of commencing POWER OPERATION and shall include all snubbers. If all snubbers of each type are found OPERABLE during the first inservice visual inspection, the second inservice visual inspection shall be performed at the l first refueling outage. Otherwise, subsequent visual inspections shall be performed in accordance with the following schedule:

No. of Inonerable Snubbers of Each Type Subsequent Visual per Inspection Period Inspection Period * **

O 18 months i 25%

1 12 months i 25%

2 6 months i 25%

l 3,4 124 days i 25%

5,6,7 62 days i 25%

8 or more 31 days i 25%

The inspection interval for each type of snubber shall not be lengthened

, more than one step at a time unless a generic problem has been identified l and corrected; in that event the inspection interval may be lengthened one step the first time and two steps thereafter if no inoperable snubbers of that type are found.

- The provisions of Specification 4.0.2 are not applicable.

l COMANCHE PEAK - UNIT 1 3/4 7-19

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

c. Visual Inspection Acceptance Criteria.

Visual inspections shall verify that: (1) there are no visible indications of damage or impaired OPERABILITY,- (2). attachments to the foundation or supporting structure are functional, and (3) fasten-ers for a'ttachment of the snubber to the comptnent and to the snubber anchorage are functional. Snubbers which appear inoperable as a result of visual inspections may be determined OPERABLE for the purpose of establishing the next visual inspection interval, provided that: (1) the cause of the rejection is clearly established and remedied for that particular snubber and for other snubbers irrespec-tive of type that may be generically susceptible; and (2) the affected snubber is functionally tested in the as-found condition and determined OPERABLE per Specification 4.7.9f. All snubbers connected to an inoperable common hydraulic fluid reservoir shall be counted as inoperable snubbers,

d. Trahsient' Event Inspection An inspection shall be performed of all snubbers attached to sections of systerrs that have experienced unexpected, potentially damaging transients as determined from a review of operational data and a visual in:pection of the systems within 6 months following such an

. event. In addition to satisfying the visual inspection acceptance criteria, freedom-of-motion of'mschanical snubbers shall be verified using at least one of the followi.ng: (1) manually induced snubber movement; or (2) evaluation of in place snubber piston setting; or (3) stroking the mechanical snubber through its full range of travel.

e. Functional Tests During the first refueling shutdown and at least once per 18 months thereafter during shutdown, a representative sample of snubberc of each type shall be tested using one of the following sample plans.

The sample plan for each type shall be selected prior to the test period and cannot, be changed during the test period. The NRC Regional Administrator shall be notified in writing of the sample plan selected i

' for each snubber type prior to the test period or the sample plan used in the prior test period shall be implemented:

1) At least 10% of the total of each type of snubber shall be functionally tested either in place or in a bench test. For i each snubber of a type that does not meet the functional test

) acceptance. criteria of Specification 4.7.9f., an. additional 10%

of that type of snubber shall be functionally tested until no more failures are found or until all snubbers of that type have been functionally tested; or l

l e

COMANCHE PEAK - UNIT 1 3/4 7-20

BRMI e

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

e. Functional _ Tests (Continued)

2) A representative sample of each. type of snubber shall be func-tionally tested in accordance with Figure 4.7-1. "C" is the total number of snubbers of a type found not meeting the' accept-ance requirements of Specification 4.7.9f. The cumulative number of snubbers of a type tested is denoted by "N". At the end of each day's testing, the new values of "N" and "C" (pre-vious day's total plus current day's increments) shall be plotted on Figure.4.7-1. If at any time the point plotted falls in the "Reject" region, all snubbers of that type shall be functionally tested. If at any time the point plotted falls in the "Accept" region, testing of snubbers of that type may be terminated. When the point plotted lies in the "Continue Testing" region, additional snubbers of that type shall be tested until the point falls in the "Accept" region or the "Reject" region, or all the snubbers of'that type have been tested; or

3) An initial representative sample of 55 snubbers shall be func-tionally tested. For each snubber type which does not meet the functional test acceptance criteria, another sample'of at least one-half the size of the initial sample shall be tested until the total number tested is equal to the initial sample size multiplied by the factor, 1.+ C/2, where "C" is the number of snubbers found which do not meet the functional test acceptance criteria. The results from this sample plan shall be plotted using an "Accept" line which follows the equation N = 55(1

+ C/2). Each snubber point should be plotted as soon as the snubber is tested. If the point plotted falls on or below the "Accept" line, testing of that type of snubber may be terminated.

If the point plotted falls above the "Accept" line, testing

! must continue until the point falls in the "Accept" region or all the snubbers of that type have been tested.

Testing equipment failure during functional testing may invalidate that day's testing and allow that day's testing to resume anew at a later time provided all snubbers tested with the failed equipment during l the day of equipment failure are retested. The representative sample

' selected for the functional test sample plans shall be randomly selected from the snubbers of each type and reviewed before beginning the testing.

The review shall ensure, as far as practicable, that they are represen-l ' tative of the various configurations, operating environments, range 'of size,'and capacity of snubbers of each type. Snubbers placed in the same location as snubbers which failed the previous functional test shall be retested at the time of the next functional test but shall not be included in the sample plan. If during the functional testing,

additional sampling is required due to failure of only one type of l snubber, the functional test results shall be reviewed at that time

j to determine if additional samples should be limited to the type of snubber which has failed the functional testing.

! COMANCHE PEAK - UNIT 1 3/4 7-21 i

'PLdNT SYSTEMS s

SURVEILLANCE REQUIREMENTS (Continued) -

[ } .-

f. Functional Test Acceptance Criteria The, snubber functional test shall verify that: .

I

1) Activation (restraining action) is achieved within the specified range in both tension and compression;

2) . Snubber bleed, or release rate where required, is present in both tension and compression, within the specified range; _

3) For mechanical snubbers, the force required to initiate or -

maintain motion of the snubber is within the specified range in both directions of travel; and'

4) For snubbers specifically required not to displace under continuous load, the ability of the snubber to withstand. load

(

without displacement.

l l Testing methods,may be used to measure parameters indirectly or l

parameters other than those specified if those results can be correlated to the specified parameters through established methods.

g. Functional Test Failure Analysis l An engineering evaluation shall be made of each failure to meet the functional testfacceptance criteria to determine the cause of the' failure. The results of this evaluation shall be used, if applicable,

! in selecting snubbers to be tested in an effort to determine the OPERABILITY of other snubbers irrespective ef type which may be subject to the same failure mode.

For the snubbers found inoperable, an engineering evaluation shall be performed on the components to which the inoperable snubbers are attached. The purpose of this engineering evaluation shall be to determine if the components to which the inoperable snubbers are attached were adversely affected by the inoperability of the snubbers in order to ensure that the component remains capable of meeting the designed service.

If any ' snubber selected for functional testing either fails to lock up or fails to move, i.e., frozen-in place, the cause will be evaluated and, if caused by manufacturer or design deficiency, all snubbers of the same type subject to the same defect shall be func-tionally tested. This testing requirement shall be independent of the requirements stated in Specification 4.7.9e. for snubbers not meeting the functional test acceptance criteria.

9 COMANCHE PEAK - UNIT 1 3/4 7-22

PLANT SYSTEMS {" II SURVEILLANCE REQUIREMENTS (Continued)

h. Functional Testing of Repaired and Replaced Snubbers Snubbers which fail the visual inspection or'the functional test acceptance criteria shall be repaired or' replaced. Replacement snubbers and snubbers which have repairs which might affect the functional test results shall be tested to meet the functional test criteria before installation in the unit. Mechanical snubbers shall have met the acceptance criteria subsequent to their most recent service, and the freedom-of-motion test must have been performed within 12 months before being installed in the unit.

i. Snubber Service Life Program The service life of hydraulic and mechanical snubbers shall be monitored to ensure that the service life is not exceeded between surveillance inspections. The maximum expected service life for various seals, springs, and other critical. parts shall be deter-mined and established based on engineering information and shall be extended or shortened based on monitored test results and failure history. Critical parts shall be replaced so that the maximum l

service life will not be exceeded durin.g a period when the snubber is required to be OPERABLE. The. parts replacements shall be.docu-

mented and the documentation shall be retaine.d in accordance with Specification 6.10.3.

4 s COMANCHE PEAK - UNIT 1 3/4 7-23

L \ l l l l 71 \ SAMPLE PLAN 2) F0 Syygfg FUNCT10NAL TEST I l COMANCHE PEAK - UNIT 1 3/4 7-24

R F,l' PLANT SYSTEMS

                                                                                                     \;\\k 3/4.7.10 SEALED SOURCE CONTAMINATION LIMITING CONDITION FOR OPERATION 3.7.10 Each sealed source containing radioactive material either in excess of 100 microCuries of beta and/or gamma emitting material or 5 microCuries of alpha emitting material shall be free of greater than or equal to 0.005 microcurie of removable contamination.

APPLICABILITY: At all times. ACTION:

a. With a sealed source having removable contamination in excess of the above limits, immediately withdraw the sealed source from use and either:

1. Decontaminate and repair the sealed source, or

2. Dispose of the sealed source in accordance with Commission Regulations.

b. The provisions of Specifications 3.'O.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS' ( 4.7.10.1 Test Requirements - Each sealed source shall be tested for leakage and/or contamination by:

a. The licensee, or

b. Other persons specifically authorized by the Commission or an Agreement State.

The test method shall have a detection sensitivity of.at least 0.005 microcurie per test sample. 4.7.10.2 Test Frequencies - Each category of sealed sources (excluding startup sources and fission detectors previously subjected to core flux) shall be tested at the frequency described below,

a. Sources in use - At least once per 6 months for all sealed sources containing radioactive materials:

1) With a half-life greater than 30 days (excluding Hydrogen 3),

and

2) In any form other than gas.

COMANCHE PEAK - UNIT 1 3/4 7-25 l

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

b. Stored sources not in use - Each sealed source and fission detector shall be tested prior to use.or transfer to another licensee unless tested within the previous 6 months. Sealed sourdes and fission detectors transferred without a certificate indicating the last test

 ,                   date shall be tested prior to being placed into use; and

c. Startup sources and fission detectors - Each sealed startup source and fission detector shall be tested within 31 days prior to being subjected to core flux or installed in the core and following repair or maintenance to the source. '

4.7.10.3 Reports A report shall be prepared and submitted to the Commission on an annual basis if sealed soLrce or fission detector leakage tests reveal the presence of greater than or equal to 0.005 microcurie of removable contamination. e b l l 1 l l 1 i COMANCHE PEAK - UNIT 1 3/4 7-26

   ,,-            __            . m                -~                               _

PLANT SYSTEMS na 3/4.7.11 AREATEMPERATUREMONITORING WkkId V* ' , LIMITING CONDITION FOR OPERATION 3.7.11 The temperature of each area shown in Table 3.7-3 shall not be exceeded for more than 8 hours or by more than 30 F. APPLICABILITY: Whenever the equipment in an affected area is required to be OPERABLE. ACTION: i

a. With one or more' areas exceeding the temperatu'e limit (s) shown in Table 3.7-3 for more than 8 hours, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that provides a record of the cumulative time and the amount by which the temperature in the affected area (s) exceeded the limit (s) and an analysis to demonstrate the' contin'ued OPERABILITY of the affected equipment. The provisions of Specifi-cationsj3.0.3 and 3.0.4 are not applicable.

b. With one or more areas exceeding the temperature limit (s.) shown in Table.3.7-3 by more than 30 F, prepare and submit a Special Report as , required by ACTION a. above and within 4 hours either restore

                  .the area (s).to within the temperature limit (s)'or declare the equip-ment in the affected area (s) inoperable.

SURVEILLANCE REQUIREMENTS 4.7.11 The temperature in each of the areas shown in Table 3.7-3 shall be determined to be within its limit at least once per 12 hours. O 4 COMANCHE PEAK - UNIT 1 3/4 7-27

kL

 ,                                       TABLE 3.7-3 AREA TEMPERATURE MONITORING
                                                                                 ", b AREA                                      TEMPERATURE LIMIT (*F) 1.

2. - ,

3. .

4. 5. 9 d e f i 9 l l o 9 COMANCHE PEAK - UNIT 1 3/4 7-28

3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES OPERATING LIMITING CONDITION FOR OPERATION 3.8.1.1 As a minimum, the following A.C. electrical power sources ~shall'b'e ' OPERABLE:

a. Two physically independent circuits between the offsite transmission network and the onsite Class 1E Distribution System, and

b. Two separate and independent diesel generators, each with:

1) A separate day fuel tanks containing a minimum volume of 1440 gallons of fuel,

2) A separate Fuel Storage System containing a minimum volume of 88,175 gallons of fuel,

3) A separate fuel transfer pump,

4) Lubricating oil storage containing a. minimum total vohlme of gallons of lubricating oil, and

5) Capability to transfer lubricating oil from storage to the diesel generator unit. -

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

a. With one offsite circuit'of the above-required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C.

sources by performing Surveillance Requirement 4.8.1.1.1.a within l 1 hour and at least once per 8 hours thereafter. If either diesel generator has not been successfully tested within the past 24 hours, demonstrate its OPERABILITY by performing Surveillance Requirements 4.8.1.1.2.a.5 and 4.8.1.1.2.a.6 for each such diesel generator, separately, within 24 hours. Restore the offsite circuit tc OPERABLE status within 72 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTOOWN within the following 24 hours.

b. With either diesel generator inoperable, demonstrate the OPERABILITY of the above required A.C. offsite sources by performing Surveillance Requirement 4.8.1.1.1.a within 1 hour and at least once per 8 hours

thereafter. If the diesel generator became inoperable due to any cause other than preplanned preventive maintenance or testing, demon-l strate the OPERABILITY of the remaining OPERABLE diesel generator by performing Surveillance Requirements 4.8.1.1.2.a.5 and 4.8.1.1.2.a.6 within 24 hours *. Restore the inoperable diesel generator to OPERABLE status within 72 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTOOWN within the following 24 hours.

    *This test is requi.ed to be completed regartless of when the inoper3ble diesel generator is restored to OPERABILITY.

( COMANCHE PEAK - UNIT.1 3/4 8-1

ELECTRICAL POWER SYSTEMS no

                                                                                                   ' (V;s     A
                                                                                                       \

LIMITING CONDITION FOR OPERATION ACTION (Continued)

c. With one diesel generator inoperable in addition to ACTION a. or b.

above, verify that:

1. All required systems, subsystems, trains, components, and devices that depend on the remaining OPERABLE diesel generator as a source of eme,gency power are also OPERABLE, and

2. When in MODE 1, 2, or 3, the steam-driven auxiliary feedwater pump is OPERABLE.

If these conditions are not satisfied within 2 hours be in at least HOT STANDBY within the next 6 hours and in COLO SHUTOOWN within the following 30 hours.

d. With two of the above required offsite A.C. circuits inoperable, demonstrate the OPERABILITY of two diesel generato~rs separately by performing the requirements of Specification 4.8.1.1.2a.5 and 4.8.1.1.2.a 6) within 1 hour and.at least once per 8 hours thereafter, unless the diesel generators are already operating; restore at least one of the inoperable offsite sources to OPERABLE status within 24 hours or be in at least. HOT STANDBY within the next 6 hours. With only one offsite source restored, restore at least two offsite .

circuits to OPERABLE status within 72 hours from time of initial loss or be in at least HOT STANDBY within the next 6 hours'and in COLD SHUTDOWN within the follo'ing w 30 hours.

e. With two of the above required diesel generators inoperable, demonstrate the OPERABILITY of two offsite A.C. circuits by performing the require-ments of Specification 4.8.1.1.la, within 1 hour and at least once per 8 hours thereafter; restore at least one of the inoperable diesel generators to OPERABLE status within 2 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTOOWN within the following 30 hours. Restore at least two diesel generators to OPERABLE status within 72 hours from time of initial loss or be in least HOT STAN0BY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.

l SURVEILLANCE REQUIREMENTS l l 4.8.1.1.1 Each of the above required independent circuits between the offsite j transmission network and the Onsite Class 1E Distribution System shall be: l

          'a.                         Determined OPERABLE at least once per 7 days by verifying correct breaker alignments, indicated power availability, and l           b.                         Demonstrated OPERABLE at least once per 18 months during shutdown by
        ,                             transferring (manually and automatically) the 6.9 kV safeguards bus power supply from the preferred offsite source to the alternate offsite source.

COMANCHE PEAK - UNIT 1 3/4 8-2

ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) F 4.8.1.1.2 Each diesel generator shall be' demonstrated OPERABLE:

a. In accordance with th'e frequency specified in Table 4.8-1 on a STAGGERED TEST BASIS by:

1) Verifying the fuel level in the day and engine mounted fuel tank,

2) Verifying the fuel level in the fuel storage tank,

3) Verifying the fuel transfer pump starts and transfers fuel from the storage system to the day fuel tank,

4) Verifying the lubricating oil inventory 'in storage,

5) Verifying the diesel starts from ambient condition and acceler-ates to at least 441 rpm in less than or equal to 10 seconds.*

The generator voltage and frequency shall be 6.900 1 690 volts and 60 1 1.2 Hz within 10 seconds" after the start signal. The diesel generator shall be started for this test by using one of the following signals: a) Manual, or b) Start-up transformer' secondary winding undervoltage, or c) Sim' u lated loss of preferred offsite power by itself, or d) Simulated safeguards bus undervoltage, or e) Safety Injection Actuation test signal in conjunction with loss'of preferred offsite power, or f) Safety Injection Actuation test s'.gnal by itself, t 6) Verifying the geneyator is synchronized, loaded to between l 6,800 and 7,000 kW in less than or equal to 60 seconds *, and operates at this load condition for at least 60 minutes, and

7) Verifying the diesel generator is aligned to provide standby power to the associated emergency busses.

        *These diesel generator starts from ambient conditions shall be performed only once per 184 days in these surveillance tests and all other engine starts for the purpose of this surveillance testing shall include the acceleration to rated speed in less than or equal to 10 seconds and be preceded by an engine i

pre-lube period and/or other warmup procedures such as gradual loading l (>80 sec) recommended by the manufacturer so that the mechanical stress and i wear on the diesel engine is minimized.

        #This band is meant as guidance to avoid routine overloading of diesel generator.      Loads in excess of the band or momentary variations due to chang-

{ ing bus loads shall not invalidate the test. l COMANCHE PEAK - UNIT 1 3/4 8-3

-v-

ELECTRICAL POWER SYSTEfjS SURVEILLANCE REQUIREMENTS (Continued)

b. At least once per 31 days and after each operation _of the diesel where the period of operation was greater than or equal to I hour by checking for and removing accumulated water from the day fuel tanks;

c. At least once per 31 days by checking for and removing accumulated water from the fuel oil storage tanks;

d. By sampling new fuel oil in accordance with ASTM-04057-1981 prior to addition to storage tanks and:

1) By verifying in accordance with the tests specified in ASTM-0975-1981 prior to addition to the storage tanks that the sample has:

a) An API Gravity of within 0.3 degrees at 60 F, or a speci-fic gravity of within 0.0016 at 60/60 F, when compared to the supplier's certificate, or an absolute specific gra-vity at 60/60 F of greater thar, or~ equal'to 0.83 but less than or equal to 0.89, or an API gravity of greater than or equal to [26] degrees but less than or equal to [38] i degrees; l l - b) A kinematic viscosity at 40*F of greater than or equal to 1.9 centistokes, but less than or equal to 4.1 centistokes (alternatively, Saybolt viscosity, SUS at 100 F of great *er than or equal to 32.6, but less than or equal.to 40.1), if gravity was not determined by comparison with the sup-plier's certification; i c) A flash point equal to or greater than 125 F; d) A clear and bright appearance with proper color when tested in accordance with ASTM-04176-1982; l 2) By verifying within 30 days of obtaining the sample that the l other properties specified in Table 1 of ASTM-D975-1981 are net when tested in accordance with ASTM-0975-1981 except that the I analysis for sulfur may be performed in accordance with ASTM- ! 01552-1979 or ASTM-02622-1982. 1

e. At least once every 31 days by obtaining a sample of fuel oil in accordance with ASTM-D2276-1978, and verifying that total particu-late contamination is less than 10 mg/ liter when checked in accor-dance with ASTM-D2276-1978, Method A; l

l COMANCHE PEAK - UNIT 1 3/4 8-4 e

f*S ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

g. At least once per 18 months *, during shutdown, by:

1) Subjecting the diesel to an inspection in accordance with procedures prepared in conjunction- with its manufacturer's recommendations for this class of standby service;

2) Verifying the generator capability to re' ject a load of greater than or eque. to [783] kW while maintaining voltage at 6900 690 vol.. and frequency at 60 1 1.2 Hz;

3) Verifying the generator capability to reject a load of 7000 kW without tripping. The generator voltage shall not exceed 7590 volts during and following the load rejection;

4) Simulating a loss-of-offsite power by itself, and:

a) Verifying deenergization of the emergency busses and load shedding from the' emergency busses, and b) Verifying the diesel starts on the auto-start signal, energizes the emergency busses tith permanently connected loads within 10 seconds, energizes the auto-connected shutdown loads through the load sequencer and operates for greater than.'or equal to 5 minutes while its generator is loaded with the shutd'own loads. After energization, the steady-state voltage and frequency of the emergency busses

                          , shall be maintained at 6900 1 690 volts and 60 + 1.2 Hz
 ,                          during this test.

5) Verifying that on a Safety Injection Actuation test signal, without loss-of-offsite power, the diesel generator starts on the auto-start signal and operates on standby for greater than or equal to 5 minutes. The generator voltage and frequency shall be 6900 2 690 volts and 60 1 1.2 bz within 10 seconds after the auto-start signal; the steady-state generator voltage and frequency shall be maintained within these limits during this test;

6) Simulatihy a less-of-offsite power in conjunction with a odiety Injection Actuation test signal, and:

a) Verifying deenergization of the emergency busses and load shedding from the emergency busses;

   *For any start of a diesel, the diesel must be operated with a load in accor'h: ice with the manufacturer's recommendations.

COMANCHE PEAK - UNIT 1 3/4 8-5 e

g 't ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) - b) Verifying the diesel starts on the auto-start signal, energizes the emergency busses with permanently connected loads within 10 seconds,. energizes the auto-connected emergency'(accident) loads through' the load sequencer and operates for greater than'or equal to 5 minutes while its generator is loaded with the emergency loads. After energization, the steady-state voltage and frequency of the emergency busses shall be maintained at (4160] 1 6900 1 690 volts and 60 1 1.2 Hz durino this test; and c) Varifying that all automatic diesel generator trips, except engine overspeed and generator differential, are automatically bypassed upon loss of voltage on the emergency bus concurrent with a Safety Injection Actuation signal.

7) Verifying the die.sel generator operates for at least 24 hours.

During the first 2 hours of this test the diesel generator shall be loaded to an indicated 7600 - 7700 kW# and during the remaining 22 hours of this test, the diesel generator shall be loaded to an indicated 6800 - 7000 kW . The generator voltage. and frequency shall be 6900 1 690 volts and 60 1.2 Hz within

                                    ;10 seconds after the. start signal; the steady-state gener.ator voltage and frequency shall'be maintained within these limits
                                    .during this test.                    Within 5 minutes after completing this 24-hour test, perform. Specification 4.8.1.1.2e.6)b);*                                        '

8) Verifying that the auto-connected loads to each diesel generator do not exceed the continuous rating of 7,000 kW;

9) Verifying the diesel generator's capability to:

a) Synchronize with the offsite power source while the ; generator is loaded with its emergency loads upon a simulated restoration of offsite power,

b) Transfer its loads to the offsite power source, and c) 8e restored to its standby status.

       #This band is meant as guidance to avoid routine overloading of the di2sel

generator. Loads in excess of the band or momeritary variations due to changing bus loads shall not invalidate the test.

       *If Specification 4.8.1.1.2e.6)b) is no't satisfactorily completed, it is not neces,sary to repeat the preceriing 24-haur test. Instead, the diesel generator may be operated between 6800 - 7000 kW for 1 hour or until operating tempera-ture has stabilized.

COMANCHE PEAK - UNIT 1 3/4 8-6 4

            - - - - - ,     - - - ,       , . - - - - - - - - - -         .                            a . - - - - . - - - , - - -        - - - - - - - -

   .r          .

h ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

10) Verifying that with the diesel generator operating in a test mode, connected to its bus, a simulated Safety Injection signal overrides the test mode by: (1) returning the diesel generato'r to standby operation, and (2) automatically energizing the emergency loads with offsite power;

11) Verifying that the fuel transfer pump transfers fuel from fuel storage tank to the day tank of its associated diesel via the installed lines;

12) Verifying that the automatic load sequence timers are OPERABLE with the interval between each load clock within + 10% of its design interval; -

l 13) Verifying that the following diesel generator lockout features l prevent diesel generator starting: ' a) Bareing device engaged, or b) Maintenance Lockout Mode.

f. At least once per 10 years or a'fter any mod'ifications which could affect diesel generator interdependence by starting both, diesel generators simultaneously, during shutdown, and verifying that both diesel generators accelerate to at least 440 rpm in less.than or equal to 10 se'conds; and

g. At least once per 10 years by:

1) Pumping out each fuel oil storage tank, removing the accumul-ated sediment and cleaning the tank using a sodium hypochlorite solution or equivalent, and

2) Performing a pressure test of those portions of the diesel fuel oil system designed to Section III, subsection ND of the ASME Code at a test pressure equal to 110% of the system design pressure.

4.8.1.1.3 Reports - All diesel generator failures, valid or non valid, shall be reported to the CSmmission in a Special Report pursuant to Specification 6.9.2 within 30 days. Reports of diesel generator fai, lures shall include the information recommended in Regulatory Position C.,3.b of Regulatory Guide 1.108, Revision 1, August 1977. If the number of failures in the last 100 valid tests on a per nuclear unit basis is greater than or equal to 7, the report shall be supplemented to include the additional information recommended in Regulatory Position C.3 b of Regulatory Guide 1.108, Revision 1, August 1977. COMANCHE PEAK - UNIT 1 3/4 8-7

TABLE 4.8-1 DIESEL GENERATOR TEST SCHEDULE NUMBER OF FAILURES IN NUMBER OF FAILURES IN LAST 20 VALIO TESTS

LAST 100 VALIO TESTS
TEST FREQUENCY

                                  <1                                                                                   <4                 Once per 31 days 1 2**                                                                                15                 Once per 7 days TABLE 4.8-2 ADDITIONAL RELIABILITY ACTIONS NO. OF FAILURES IN~      ,

NO. OF F.AILURES IN

LAST 20 VALID TESTS LAST 100 VALIO TESTS ACTION 3 6 Within 14 days prepare and maintain a report for NRC audit describing the diesel generator reliability improvement program ir.iplemented at the site.

Minim 6m requirements for the report are indicated in Attachment 1 to this table. 5 11 Declare the diesel generator inoperable. Perform a re-qualification test program for the affected diesel generator. Requalification test program requirements are indicated in Attachment 2 to this table. 1.

Criteria for determining numb'er of failures and number of valid tests shall be j in accordance with Regulatory Po ition C.2.e of Regulatory Guide 1.108, but determined on a per diesel generator basis.

             **The associated test frequency shall be maintained until seven consecutive failure free demands have been performed and the number of failures in the last 20 valid demands has been reduced to one, l

l COMANCHE PEAK - UNIT 1 3/4 8-8

t . ATTACHMENT 1 TO TABLE 4.8-2 REPORTING REQUIREMENT 'O As a minimum the Reliability Improvement Program report for NRC audit shall include: a) a summary of all' ' tests (valid ,.nd invalid) that occurred within the time period over which the last 20/100 va'id tests were performed b) analysis of failures and determination of root causes of failures

      *                 ~

c) evaluation of each of the recommendati n s of NUREG/CR-0660, "Enhancement - of Onsite Emergency Diesel Generator Geliability in Operating Reactors," with respect to their application to the plant d) identification of all actions taken or to be taken to 1) correct the root causes of failures defined in b) above and (2) achieve a general improvement of diesel generator reliability e) the schedule for implementation'of sach action from d) above f) an assessment of the existing reliability of electric power to engineered-safety-feature equipment Once a licensee has prepared and maintain an' initial report detailing the diesel generator reliability improvement. program at his site, as defined

           'above, the licensee need. prepare orty a supplemental report within 14 days after each failu e during a valid o: mand for so long as the affected diesel

generator unit continues to violate the criteria (3/20 or 6/100) for the reliability improvement program remedial action. The supplemental report need only update the failure / demand history for the affected diesel generator unit since.the last report for that diese! generator. The supplemental report shall also present an analysis of the failure (s) with a root cause determination, if possible, and shall delineate any further procedural, hardware or operational changes to be incorporated into the site diesel

          . generator improvement program and the schedule for implementation of those changes.

In addition to the above, submit a yearly data report on the diesel generator reliabliity. , t l e e

 'I COMANCHE PEAK - UNIT 1                    3/4 8-9 2

ATTACHMENT 2 TO TABLE 4.8-2 M.' DIESEL GENERATOR RE0VALIFICATION PROGRAM V

1. Perform seven consecutive successful demands without a failure within 30 days of diesel generator being restored to operable status and fourteen c'onsecutive successful demands without a failure within 75 days of diesel generator of being restored te operable status.

2. If a failure occurs during the first seven tests in the reoualification test program, perform seven successful demands without an additional failure within 30 days of diesel genera. tor of being restored to operable status and fourteen consecutive successful demands without a failure within 75 days of being restored to operable status.

3. If a failure occurs during the second seven tests (tests 8 through 14) of l (1) above, perform fourteen consecutive successful demands without an l additional failure within 75 days of the failure which occurred during the requalification testing.

4. Followidg the.second failure during the requalification test program, be in at least h0T STANDBY within the next 6 hours and COLD SHUTDOWN within the following 30 hours.

5. During requalification testing the diesel generator should not be tested more frequently than at 24-hour intervals.

After a diesel generator has been successfully requalified, subsequent. repeated requalification tests will not be required for that diesel generator under the following conditions: (a) The number of failures in the last 20 valid demands is less than 5. (b) The number of failures in the last 100 valid demands is less than 11. (c) In the event that following successful requalification of a diesel generator, the number of failures is still in excess cf the remedial action criteria (a and/or b above) the following exception will be allowed until the diesel generator is no longer in violation of the remedial action criteria (a and/or b above). Requalification testing will not be required provided that after each valid demand the number of failures in the last 20 and/or 100 valid demands has not increased. Once the diesel generator is no longer in violation of the remedial action criteria above the provisions of those criteria alone will prevail. . I COMANCHE PEAK - UNIT 1 3/4 8-10

ELECTRICAL POWER SYSTEMS A.C. SOURCES hk' 'g t V F* SHUT 00WN LIMITING CONDITION FOR OPERATION 4 3.8.1.2 As a minimum, the following A.C. electrical power sources shall be OPERABLE: a. One circuit between the offsite transmission network and the Onsite Class 1E Distribution System, and ,

b. -One diesel generator with:

1) Day fuel tanks containing a minimum volume of 1440 gallons of fuel,

2) A fuel storage system containing a minimum volume of.88,175 gallons of fuel,

3) A fuel transfer pump,

4) Lubricating oil storage containing a minimum total valume of' gallons of lubricating oil, and .

5) Capability to transfer lubricating oil from storage to the {

diesel generator unit. . APPLICABILITY: MODES 5 and 6. ACTION: With less than the above minimum required A.C. electrical power sources OPERABLE, immediately suspend all operations involving CORE ALTERATIONS, positive reactivity changes, movement of irradiated fuel, or crane operation with loads over the fuel storage pool, and within 8 hours, depressurize and vent the Reactor Coolant System through a greater than or equal to 2.98 square inch vent. In addition, when in MODE 5 with the reactor coolant loops not i filled, or in MODE 6 with'the water level less than 23 feet above the reactor vessel flange, immediately initiate corrective action to restore the required sources to OPERABLE status as soon as possible. SURVEILLANCE REQUIREMENTS 4.8.1.2 The,above required A.C. electrical power sources shall be demonstrated OPERABLE by the performance of each of the requirements of Specifications l 4.8.1.1.1, 4.8.1.1.2 (except for Specification 4.8.1.1.2a 6)), and 4.8.1.1.3. COMANCHE PEAK - UNIT 1 3/4 8-11

3/4.8.2 D.C. SOURCES 9 OPERATING LIMITING CONDITION FOR OPERATION

3. 8.2.1 As a minimum, the following D.C. electrical sources shall be. PERABLE:

a. Train A - 125 volt D.C. Station Batteries BTIED1 and BTIE03 and at least one full capacity charger associated with each battery and b. Train B.- 125 volt D.C. Station Batteries BTIE02 and BTIED4 and at least one full capacity charger associated with each battery. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With one of the required battery trains and/or full-capacity chargers inop-

       .erable,. restore the inoperable battery train and/or full-capacity charger to OPERABLE' status within 2 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.

SURVEILLANCE REQUIREME'NTS

 .                                                                                ~~

4.8.2.1 Each 125 V D.C. station battery and charger shall be demonstrated . OPERABLE:

a. At least once per 7 days by verifying that:

1) The parameters in Table 4.8-2 meet the Category A limits, and

2) The total battery terminal voltage is greater than or equal to 128 volts on float charge.

COMANCHE PEAK - UNIT 1 3/4 8-12

O. C. SOURCES y SURVEILLANCE REQUIREMENTS (Continued)

b. At least once per 92 days and within 7 days af ter a battery discharge With battery terminal voltage below 110 ~ volts, 'or battery overcharge '

with battery terminal voltage above 150 volts, by verifying that:

1) The parameters in Table 4.8-2 meet the Category B limits, l

1

2) There is no visible corrosion at either terminals or connectors, l or the connection resistance of these items is less than I 150 x 10 6 chm, and

3) The average electrolyte temperature of 12 of connected cells is

above 70 F.

c. At least once per 18~ months by verifying that:

1) -The cells, cell plates', and bettery racks show no visual indication of physical damage or abnormal deterioration,

2) The cell,-to-c. ell and terminal connections are clean, tight, and coated with anticorrosion material,

3) ..The resistance of.each cell-to-cell and terminal connection is less than or' equal to 150 x 10 8 ohm, and -

4) The battery charger will supply at least 300 amperes at 125 volts for at least 12 hours. >

d, At least once per 18 months, during shutdown, by verifying that the battery capacity is adequate to .<upply and maintain in OPERABLE status all of the actual or simulated emergency loads for the design duty cycle when the battery is subjected to a battery service test; i ~

e. At least once per 60 months, during shutdown, by verifying that the battery capacity is at least 80% of the manufacturer's rating when '

i subjected to a performance discharge test. Once per 60-month interval this performance discharge test may be performed in lieu j of the battery service test required by Specification 4.8.2.1d. ; and i

f. At least once per 18 months, during shutdown, by giving performance discharge tests of battery capacity to any battery that shows signs of degradation or has reached 85% of the service life expected for the application. Degradation is indicated when the battery capacity
drops more than 10% of rated capacity from its average on previous performance tests, or is below 90% of the manufacturer's rating.

l COMANCHE PEAK - UNIT 1 3/4 8-13 E - u- _ _ ._ _

TABLE 4.8-3 1 BATTERY SURVEILLANCE REQUIREMENTS CATEGORY A II) CATEGORY B(2) PARAMETER LIMITS FOR EACH LIMITS FOR EACH ALLOWA3LE(3) DESIGNATED PILOT CONNECTED CELL VALUE FOR EACh CELL CONNFCTED CELL Electrolyte >Hinimum level . > Minimum level Abote top of Level indication mark, indication mark, plates, and < " above and < \" above anc not maximum level maximum level overflowing indication mark indication mark Float Voltage 2 2.13 volts 1 2.13 volts (6) > 2.07 volts Not more than 0.020 below the average of all Specific 1 1.200(5) 1 1.195 connected cells Gravity (4) Average of all Average of all connected cells connected cells -

                                                         > 1.205 1.195(5)

TABLE NOTATIONS (1) For any Category A parameter (s) outside the limit (s) shown, the battery may be considered OPERABLE provided that within 24 hours all the Category 8 measurements are taken and found to be within their allowable values, and provided all Category A an/ B parameter (s) are restored to within limits within the next 6 days. (2) for any Category B Nrameter(s) outside the limit (s) shown, the battery may be considered OPERABLE provided that the Category B parameters are

              'within their allowable values and provided the Category B parameter (s) are restored to within limits' within 7 days.

(3) Any Category B para.seter not within its allowable value indicates an j inoperable battery. (4) Corrected for electrolyte temperature (reference temperature of 77*F) and level. (5) Or battery qharging current is less than 2 amps when on charge. (6) Corrected for average electrolyte temperature. l l COMANCHE PEAK - UNIT 1 3/4 8-14 l l

_ ~ l: ..

          .v                                                                              <<

Dj . SOURCES I*

             -SHUT 00WN LIMITING CONDITION FOR OPERATION 3.8.2.2 As a minimum, two 125V 0'.C. station. batteries of one train and at least one. associated full-capacity charger for each required battery shall be OPERABLE.

APPLICABILITY: M00FS 5 and 6. ACTION: With the required batteries train and/or full-cepacity chargers inoperable, immediately suspend all operations involving CORE ALTERATIONS, positive reactivity changes, or movement of irradiated fuel; initiate corrective action to restore the required battery train and full-capacity charger to OPERABLE-status as soon as possible, and within 8 hours, depressurize and vent the Reactor Coolant System through a'2.98 square inch vent. SURVEILLANCE REQUIREMENTS

4.8.2.2 The above required'125V 0.C. station batteries and full-cap'acity charger shall be demonstrated OPERABLE in accordance with Specification 4.8.2.1. \ G t l l . COMANCHE PEAK - UNIT 1 3/4 8-15

__,_m . - - 3/4.8.3 ONSITE POWER DISTRIBUTION OPERATING LIMITING CONDITION FOR OPERATION 3.8.3.1 The following electrical busses shall be energized in the specified

                  . manner:

a. Train A A.C. Emergency Busses consisting of:

1) 6900-Volt Emergency B n 1EA1,

2) 480-Volt Emergency Bus 1EB1 from transformer T1EB1, and

3) 480-Volt Emergency Bus IEB3 from transformer T1EB3.

b. Train'B A.C. Emergency Busses consisting of:

1) 6900cVolt Emergency Bus 1EA2,

2) 480-Volt Emergency Dus IEB2 from transformer T1EB2, and

3) 480-Volt Emergency Bus 1E24 from transformer T1EB4.

c. 118-Volt A.C. Instrument Bus IPC1, IPC3, and IEC1 energized from its associated inverter connected to 0.C. Bus 1E01*; '

d. 118-Volt A.C. Instrument Bus 1PC2, IPC4, and IEC2 energized from its associated inverter connected to 0.C. Bus 1E02*;

e. 118'. Volt' A.C. Instrument Bus 1EC5 energized'from its associated inverter connected to 0.C. Bus IE03*;

f. 118-Volt A.C. Instrument Bus IEC6 energized from its associated inverter connected to 0.C. Bus 1E04*;

g. Train A 125-Volt 0.C. Busses IE01 and IE03 energized from Station Batteries BT1E01 and BTIED3, respectively; and

h. Train B 125-Volt 0.C. Busses 1E02 and 1E04 energized frcm Station Batteries BTIE02 and BTIE04, respectively.

APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: [

a. With one of the required trains of A.C. emergency busses not fully energized, reenergize the trains within 8 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTOOWN within the ,

following 30 hours. l

                    *Two invertera may be disconnected from their 0.C. bus for up to 24 hours as necessary, for the purpose of performing an equalizing charge on their asso-ciated battery train provided: (1) their vital busses are energized, and (2) the vital b'usses associated with the other battery train are energized j                     from their associated inverters and connected to their associated 0.C. bus.

l l l COMANCHE PEAK - UNIT 1 3/4 8-16 l l

g<e ONSITE POWER DISTRIBUTION hj LIMITING CONDITION FOR OPERATION ACTION (Continued)

b. - With one A.C. vita'l but either not energ.ized from its associated inverter, or with the inverter not connected to its associated 0.C.

bus: (1) reenergize the A.C. Vital bus within 2 hours or be in at least HOT STAND 8Y within the next 6 hours and in COLD SHUTDOWN within the following 30 hours; and (2) reenergize the A.C. vital bus from its associated inverter connected to its associated 0.C. bus within 24 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTOOWN within the following 30 hours,

c. With one O.C. bus not energized from its associated station battery, reenergize the D.C. bus from its associated station battery within l 2 hours or be in at least HOT STANDBY within the next 6 hours and in

COLD SHUTOOWN within the following 30 hours.

     . SURVEILLANCE REQUIREHENTS 4.8.3.1. The specified busses shall be determined energized in the re' quired 3       manner at least once'per.7 days by verifying correct breaker alignment and indicated voltage on the busses.                                                     -

l l l l l S O COMANCHE PEAK - UNIT 1 3/4 8-17

c.4 ONSITE POWER DISTRIBUTION \ I SHUTDOWN LIMITING CONDITION FOR OPERATION ' 3.8.3.2 As a minimum, the fo'llowing electrical busses shal'1 be energized in the specified manner:

a. One train of A.C. emergency busses consisting of one 6900-volt and two 480-volt A.C. emergency bus; t

b. Two 118-volt A.C. instrument busses (channel-oriented) energized from their associated inverters connected to their respective D.C.

busses;

c. One train of A.C. instrument busses consisting of two 118-volt A.C.

instrument busses energized from their associated inverters connected to their respective D.C. busses. Busses shall be of the same train as Specifications 3.8.3.2a. and d.; and

d. One train of D.C. busses consisting of two 125-volt D.C. busses energized from their associated battery banks. Busses shall be of the same train as Specifications 3.8.3.2a. and c.

APPLICABILITY MODES 5 and 6. ACTION: With any of the above required electrical busses not energized in the required manner, immediately suspend all operations involving CORE ALTERATIONS, positive reactivity changes, or movement of irradiated fuel, initiate corrective action to energize the required electrical busses in 'he specified manner as soon as possible, and within 8 hours, depressurize and vent the RCS through at least a 2.98 square inch vent. SURVEILLANCE REQUIREMENTS 4.8.3.2 The specified busses shall be determined energized in the required manner at least once per 7 days by verifying correct breaker alignment and indicated voltage on the busses. l G (CMANCHE PEAK - UNIT 1 3/4 8-18

3/4.8.4 ELECTRICAL EQUIPPENT PROTECTIVE DEVICES ,' A.C. CIRCUITS INSIDE PRIMARY CONTAINMENT l\~ , LIMITING CONDITION FOR OPERATION 3.8.4.1 At least the.following A.C. circuits inside primary.r.ontainment shall . be deenergized: ,

a. Circuit numbers [- , , and ) in panel ( ).

b. Circuit numbers [ , , and ] in panel [. ).

APPLICABILITY: MODES 1, 2, and 3. ACTION: With any of the above required circuits energized, trip the associated circuit breaker (s) in the specified panel (s) within I hour. SURVEILLANCE REQUIREMENTS 4.8.4.1 Each of the above required A.C. circuits shall be determined to be deenergized at least once per 31 dayi by verifying that the associated circuit break.ers are locked in the open position, i l [ l COMANCHE PEAK - UNIT 1 3/4 8-19 l .

ELECTRICAL EQUIPMENT PROTECTIVE DEVICES s CONTAINMENT PENETRATION CONOUCTOR OVERCURRENT PROTECTIVE DEVICES ! LIMITING CONDITIGN r0R OPERATION

            .3.8.4.2 All containment penetration conductor overcurrent protective devices

given in Table 3.8-1 shall be OPERABLE.

j APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: i With one or more of the containment penetration conductor'overcurrent protective device (s) given in Table 3.8-1 inoperable:

a. Restore the protective device (s) to OPERABLE status or deenergize i the circuit (s) by tripping the associated backup circuit breaker I or racking o,ut or removing the inoperable circuit breaker within 72 hours,. declare the affected system or component inoperable, and verify the backup circuit. breaker to be tripped or the'inoper-able circ.it breaker racked out or removed at least once per 7 days thereafter; the provisions of Specification 3.0.4 are not applicable

, to overcurrent devices in circuits which'have their backup. circuit breakers tripped, their inoperable circuit breakers racked out, or f removed, or *

b. Be in at least HOT STANDBY within the next 6 hours and in COLD

SHUTDOWN within the following 30 hours. i SURVEILLANCE REQUIREMENTS -

4.8.4.2 All containment penetration conductor overcurrent protective devices given in Table 3.8-1 shall be demonstrated OPERABLE

i a. At least once per 18 months:

1) By verifying that the medium *;oltage 6.9 kV and low voltage 480V switchgear circuit breakers are OPERABLE by selecting, on a j rotating basis, at least one or 10% of the circuit breakers i whichever is greater of each current rating and performing the

! following: a) A CHANNEL CALIBRATION of the associated protective relays, l i b) An integrated system functional test which includes simulated l automatic actuation of the system and verifying that each

relay and associated circuit breakers and control circuits l

function as designed, and i \ COMANCHE PEAK - UNIT 1 3/4 8-20 1

ELECTRICAL EQUIPMENT PROTECTIVE DEVICES SURVEILLANCE REQUIREMENTS (Continued) c) For each circuit breaker found inoperable during these functional tests, one or an additional representative sample of at least 10% of all'the' circuit breakers.of the inoperable type shall also be functionally tested until no more failures are found or all circuit breakers of that type have been functionally tested; and

2) By selecting and functionally testing a representative sample of at least 10% of each type 480 V molded case circuit breakers and of lower voltage circuit breakers. Circuit breakers selected for functional testing shall be selected on a rotating basis. Testing of these :ircuit breakers shall consist of injecting a current with e value equal to 300% of the pickup of the long-time delay trip element and 150% of the pickup of the short-time delay trip element, and verifying that the circuit breaker operates within the time delay band width for that current specified by the manufacturer. The instantaneous element shall be tested by injecting a current equal to 120% of the pickup value of the element and verifying that the circuit breaker trips instantaneously with no intentional time delay.

Molded case circuit breaker testing shall also follow this procedure except that generally no more than two trip elements, time delay and instantaneous, will be involved. Circuit breakers found inoperable during functional testing shall be

   .             . restored to OPERABLE status prior to' resuming operation. For

' each circuit breaker found inoperable during these functional tests, an additional representative sample of at least 10% of all the circuit breakers of the inoperable type shall also be functionally tested until no more failures are found or all circuit breakers of that type have been functionally tested;

b. At least once per 60 months by subjecting each circuit breaker to an inspection and preventive maintenance in accordance with procedures prepared in conjunction with its manufacturer's recommendations.

COMANCHE PEAK - UNIT 1 3/4 8-21 i

1 *e TABLE 3.8-1 CONTAINMENT PENETRATION CONOUCTOR , OVERCURRENT PROTECTIVE DEVICES

DEVICE NUMBER SYSTEM AND LOCATION -

POWERED

1. 6.9'KVAC from Switchgears

a. Switchgear Bus 1Al RCP #11

1) Primary Brea,ker IPCPX1 a) Relay 50M1-51

       ,                                                b)    Relay 26 c)    Relay 86M

2) Backup Breakers 1Al-1 or 1Al-2

                                                     .a)      Relay 51M2
                                                      .b)     Relay 51 for 1Al-1 c)    Relay 51 for IAl-2 d)    Relay 86/1A1

b. Switchgear Bus 1A2 RCP #12

1) Primary Breaker 1PCPX2 a) Relay 50M1-51 -

b) Relay 26 t c) Relay 86M

2) Bai:kup Breakers 1A2-1 or 1A2-2 a) Relay 51M2 b) Relay 51 for 1A2-1 c) Relay 51 for 1A2-2 d) Relay ^U~.A2

c. Switchgear Bus 1A3 RCP #13

1) Primary Breaker IPCPX3 a) Relay 50H1-51 b) Relay 26

                                        ,              c)     Relay 86M

2) Backup Breaker 1A3-1 or IA3-2 a) Relay 51M2 b) Relay 51 for 1A3-1 c) Relay 51 for 1A3-2

d) Relay 86/1A3 COMANCHE PEAK - UNIT 1 3/4 8-22

s . TABLE 3.8-1 (Continued) g CONTAINMENT PSNETRATION CONDUCTOR OVERCURRENT PROTEClIVE DEVICES hw ,L DEVICE NUMBER SYSTEM

             'AND LOCATION                                                                                                                                                   POWERED

1. 6.9 KVAC from Switchgears (Continued)

a. Switchgear Bus 1A4 RCP #14

1) Primary Breaker IPCPX4 -

a) Relay 50H1-51 , b)~ Relay 26 , c) Relay 86M

2) Backup Breakers 1A4-1 or 1A4-2 -

a) Relay 51M2 b) Relay 51 for 1A4-1 c) Relay 51 for 1A4-2 ; d) Relay 86/1A4 '

2. 480 VAC from Switchgears 2.1 Device Location . Containment Recirc.

480V Switchgears 1EB1, IEB2, - Fans and CRDM Vent Fans 1EB3 and IEB4 , a. Primary Breakers - IFNAV1, 1FNAV2, IFNAV3, 1FNAV4, 1FNCB1 and IFNCB2

b. Backup Breakers - 1EB1-1, IEB2-1, 1EB3-1 and 1EB4-1 l

1) Long Time & Instantaneous Relays

I 50/51 50/51 g y (IEB1-1) 1FNAV2 (1EB2-1) I 50/51 50/51 ITRAV3 (1EB3-1) 17ggg4 (1EB4-1) i 50/51 50/51 g y (1EB3-1) M2 1EB4-1)
Associated circuit breaker shown in parentheses; e.g., IEB3-1, is backup to IFNAV3 and IFNCBl. ,

l COMANCHE PEAK - UNIT 1 3/4 8-23

TABLE 3.8-1 (Continued) CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE NUMBER SYSTEM AND LOCATION POWERED -

2. 480 V.AC from Switchgears (Continued)

2) Time Delay Relays 1 (1 01~1) 1 2 (1 B2-1) 2 1 k3(1B3-1) yfg 4 (1EB4-1) i 1 (1EB3-1) 1 CB2 (IE04'1) 2.2 Device Location - 480V Containment Polar Switchgear 1EB4 Crane

a. Primary Breaker - ISCCP1

b. Backup Breaker - 1EB4-1

1) 51 1SCCP1

2) 62 ISCCP1

3. 480 VAC from Motor Control Centers 3.1 Device Location -

MCC IEB1-2 Containment Numbers listed below. Primary and Backup - Both primary and backup Breakers breakers have identical trip ratings and are in the same These breakers MCC Compt. are General Electric type

                                                 .                                       THED or THFK with thermal-
                  .                                                                      magnetic trip elements.

MCC IEB1-2 G.E. COMPT. NO. BKR TYPE SYSTEM POWERED 4G THED Motor Operated Valve 1-TV-4691-4M THED Motor Operated Valve 1-TV-4693 COMANCHE PEAK - UNIT 1 3/4 8-24 1

w TABLE 3.8-1 (Continued') ' k CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE NUMBER

 ,       AND LOCATION

3. 480 VAC from Motor Control Centers (Continued)

MCC 1EB1-2 G.E. COMPT. NO. BKR. TY?E SYSTEM POWERED 3F THED Containment Drai, Tank Pump-03 SH THED Reactor Cavity Sump Pump-01 9M THED Reactor Cavity Sump Pump-02 , 7H THED Containment Sump #1 Pump-01 7M THED Containment Sump #1 Pump-02 6H THED RCP #11 Motor Space Heater-01

              .6M                   THED                                                               RCP '#1'3 Motor Space Heater-03 88 THED                                                               Incore Detector Drive "A"               .

8D ' THED Incore Detector Drive "B" 7B THED Incore Detector Drive "F" 50 THED Fuel Transfer System Reactor Side Cont. PNL FOR-01 3B THED Stud Tensioner Hoist Outlet-01 70 THED Hydraulic Deck Lift-01 4B THED Reactor Coolant Pump Motor Hoist Receptacle-42 8H THED RC Pipe Penetration Cooling Unit-01 . BM THED RC Pipe Penetration Cooling Unit-02 5H THED RCP #11 Oil Lif t Pump-01 SM THED RCP #13 Oil Lift Pump-03 10B THED Preaccess Filter Train Package Receptacle - 17 COMANCHE PEAK - UNIT 1 3/4 8-25

                                                                                                                                            ,                                     ~

t 3_" TABLE 3.8-1 (Continued) CONTAINMENT PENETRATION CONOUCTOR 5 OVERCURRENT PROTECTIVE DEVICES '- ' DEVICE NUMBER

         . AND LOCATION                                                          ,                                                                                        .

.- 3. 480 VAC from Motor Control Centers (Con'tinued)

MCC IEB1-2 G.E. ! . COMPT. NO. BKR. TYPE SYSTEM POWERED I 5B THED Containment L79 XFMR-14 (PML-C3) 10F THED S.G. Wet Layup Cire. Pump 01 ; (cpl-CFAPRP-01) 12M THED S.G. Wet Layup Circ. Pump 03 (CPI-CFAPRP-03) i . , 12H THFK Cont. Ltg. Transf. CPI-ELTRNT-28 i (NJLC-11AMC-12) ; 6D THED Refueling Machine (Manipulator i

        . .                                                                                            Crane-01)                                                                       t 2M                                       THED                                  RC Drain Tank Pump No. 1                                                        !

l' i 2F THED. Containment Ltg XFMR-16 ! (PNL C7 & C9) i , 1H THED Containment Ltg XFMR-12 (PNL C1 & C5). -

3M THED Preaccess Fan No. 11 i l

r 3.2_ Device Location - MCC IEB2-2 Containment !

Numbers listed below.

l Primary and Backup -

Both primary and backup {

Breakers breakers have identical trip ;

! ratings and are located in l the same MCC compt. These breakers are General Elec-l tric type THED or THFK with thermal-magnetic trip i . elements, '

MCC 1EB2-2 G . F. . i COMPT. NO. ER. TYPE . SYSTEM POWERED 4G THED Motor Operated Valve 1-TV-4692 l 4M THED Motor Operated Valve 1-TV-4694 f COMANCHE PEAK - UNIT 1 *'i 8-26 l

l

 -t         -       -      ,-- .-,.          , - - . , , , ,       , , .     ,-     . , , , - . .     - - , . -         , . . - , ,, - , ,      -.--       ,-.-e      , , , .

f. . .

TABLE 3.8-1 (Continued) CONTAINMENT PENETRATION CONDUCTOR f ,. ~

                                                            ] f;URRENT PROTECTIVE DEVICES                  p DEVICE NUMBER AND LOCATION                      .                -

3. 480 VAC from Motor Control Centers'(Continued)

MCC 1EB2-2 G.E. COMPT. N0; BKR. TYPE SYSTEM POWERED 3F THED Containment Drain Tank Pump-04 4 7H THED Containment Sump No.' 2 Pump-03 d 7M THED Containment Sump No. 2 Pump-04 6H THED RCP No. 12 Motor Space Heater-02 6M THED RCP No. 14 Motor Space Heater-04 5B THED Incore Detector Drive "C" 2B . THED Incore' Detector Drive "D" 78 ,- THED Incore Detector Drive "E" 50 THED Containment Fuel Storage Crane-01 3B THED Stud Tensioner Hvist Outlet-02 4B THED Containment Solid Rad Waste Compactor-01 10B THED RCC Change Fixture Holst Drive-01 10F THED Refueling Cavity Skimmer Pump-01 12B THED Power Receptacles (Cont. El. 841') 1M THED S.G. Wet Layup Cire. Pump 02 (CPI-CFAPRP-02) 12M THED S.G. Wet Layup Circ. Pump 04 (CPI-CFAPRP-04) - 8H THED RC Pipe Penetration Fan-03 8M THED RC Pipe Penetration Fan-04 SH THED RCP #12 Oil Lift Pump-02 COMANCHE PEAK - UNIT 1 3/4 6-27

I

BBLE3.8-1(Continued) p j CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES k .

DEVICE NUMBER - AND LOCATION * '

3. 480 VAC_from Motor

Control Centers (Continued) i MCC IEB2-2 G.E.

COMPT. NO. BKR. TYPE ESTEMPOWERED SM THED RCP #14 Oil Lift Pump-04 12H THED Preaccess Filter Train Package ; Receptables - 18 60 THED Containment Auxiliary Upper Crane-01 . 2F THED Containment Ltg. XFMR-13 (PNL C-2) 70 THED Containment Elevator-01 - 20 THED Containment Access' Rotating

                                .                                                  Platform-01              ,

2'M THED Reactor Coolant Drain Tank Pump-02 4 t 9F THED Containment Ltg. XFMER-17 (PNL C8 & C10) 9M THED Containment Ltg. XFMR-15 (PNL C4 & C6) 3M THED Preaccess Fan-12 i 1G THFK Containment Welding Machine Power Supply Unit

3. 3 Device Location -

MCC IEB3-2 Containment numbers i listed below. . I . Primary and Backup Breakers - Unless noted otherwise, both , primary and backup breakers have , identical trip ratings and are L located in the same MCC compt' ! j - These breakers are General i Electric type THE0 or THFK with t thermal-magnetic trip elements. l 1 > I t 'l .

COMANCHE PEAK - UNIT 1 3/4 8-28

TABLE 3.8-1 (Continued) CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECVIVE DEVICES DEVICE NUMBER AND LOCATION ,

3. 480 VAC from Motor Control Centers (Continued)

MCC 1EB3-2 G.E. - COMPT. NO. BKR. TYPE SYSTEM POWERED 8RF THED JB-15-1005 for Altern. Feed to Motor Operated Valve 1-8702A IG THED Motor Operated Valve 1-8112 9G -THE0 Motor Operated Valve 1-8701A I 9M THED ' Motor Operated Valve'l-8701B SM THED Motor Operated Valve 1-8000A SG THED. Motor Operated Valve 1-HV-6074 4G THED , HQ tor' Operated Valve 1-HV-6076

4M THE0* M9 tor Operated' Valve 1-HV-6078 2G THED Motor Operated Valve 1-HV-4696 2M THED Motor Operated Valve 1-HV-4701 3G THED Motor Operated Valve 1-HV-5541 3M THED Motor Operated Valve 1-HV-5543 1M ' THED' Motor Operated Valve 1-HV-6083 9RF THED Motor Operated Valve 1-HV-4782 9RM THED Motor Operated Valve 1-HV-8811A 6F THED Motor Operated Valve 1-HV-8308A 6M THED Motor Operated Valve 1-HV-8808C 7M THED Containment Ltg, XFMR-18

                                                           ,(PNL SC1 & SC3)

Primary protection is provided by Gould Tronic TR5 fusible switch with 3.2A fuse.

COMANCHE PEAK - UNIT 1 3/4 8-29

e TABLE 3.8-1 (Continued) [ F' ** CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT P?qTECTIVE DEVICES DEVICE NUMBER AND LOCATION

3. 480 VAC from Motor Control Centers (Continued)

MCC IEB3-2 G.E. COMPT. NO. BKR. TYPE SYSTEM POWERED . 8M THED Neutron Detector Well Fan-09 7F THFK Electric H Recombiner Power 2 Supply PNL-01 BRM THED Fire Protection Containment : Isolation M0V1-HV-4075C , 3.4 Device Location - MCC 1EB4-2 Containment numbers , listed below. l 4 ' Primary and Backup -

                                                    'Unless noted otherwise, both Breakers                            primary and backup breakers have identical trip ratings and are located in the same MCC compt.                                                                ;

These breakers are General , Electric type THED or.THFK with !

thermal-magnetic trip elements.

MCC 1EB4-2 G.E. COMPT. NO. BKR. TYPE SYSTEM POWERED IM THED JB-15-1230G, Altern. Power Supply Feed to Mov 1-8701B 8G THED Motor Operated Valve 1-8702A BM THED Motor Operated Valve 1-8702B 4M THED Motor Operated Valve 1-8000B 4G THED Motor Operated Valve 1-HV-6075 3G THED Motor Operated Valve 1-HV-6077 5 3M THED* Motor Operated Valve 1-HV-6079 , F

   *PrimaryprotectionisprovidedbyGouldTronicTR5fusibleswitchwith3.2A fuse.

i COMANCHE PEAK - UNIT 1 3/4 8-30

TABLE 3.8-1 (Continued) [ CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE NUMBER . Ak0 LOCATION -

3. 480 VAC from Motor Control Centers (Continued)

MCC 1EB4-2 G. E. COMPT. NO. BKR. TYPE SYSTEM POWERED 2G' THED Motor Opreated Valve 1-HV-5562 2M THED Motor Operated Valve 1-HV-5563 8RF THED Motor Operated Valve 1-HV-4783 BRM THED Sump to #2 RHR Pump MOV 1-8811B - 5F THED Accumulator' Iso. VLV, Mov-1-8808B J SM THED Accumulator Iso. YLV, Mov-1-88080 6M . THED Containment Ltg. XFMR-19 (PNL SC2 & SC4) 7M THED Neutron Detector Well Fan-10 6F THFK Elect. H 2 Recombiner Power Supply PNL-02 DEVICE NUMBER SYSTEM ' AND LOCATION POWERED

4. 480 VAC From Pane 1 boards For Pressurizer Pressurizer Heaters Heaters

a. Primary Breakers - General Electric Type TJJ Thermal Magnetic

, breaker. Breaker No. & Location - Ckt. Nos. 2 thru 4 of Pane 1 boards IEB1-1, : 1EB1-2, IEB2-2, 1EB3-2, IEB4-1, IEB4-2 and , Ckt. Nos. 2 thru 5 of Pane 1 boards IEB2-1 4 and IEB3-1. : i , b. Backup Breakers - General Electric Type THJS with longtime and insts solid state trip device with 400 Amp, sensor. Breaker No. & Location - Ckt. No. 1 of Pane 1 boards IEB1-1, IEB1-2, 1EB2-1, 1EB2-2, 1EB3-1, IEB3-2, 1EB4-1 and 1EB4-2. l COMANCHE PEAK - UNIT 1 3/4 8-31 i

TABLE 3.8-1 (Continued) : CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES hy

DEVICE NUMBER SYSTEM AND LOCATION .

POWERED

5. DC Power From Rod Control Power Cabinets Rod control Fuse Location - Rod control power Cabinets IAC, IBD, 2AC, 2BD and SCOE. ;

a. Primary Fuses FUSE LOCATION -

AND NUMBER SYSTEM POWERED , FU13 to FU20 Stationary Gripper Coils FU21 to FU24 Moving Gripper Coils FU25 to FU32 Stationary Gripper Coils FU33 to FU36 Moving Gripper Coils 1 , FU37 to FU44 Stationary Gripper Coils FU45 to FU52 Moving Gripper Coils , A51/FU1 & FU2 to Lift Coils A58/FV1 & FU2 :

b. Backup Fuses' '

FUSE LOCATION i AND NUMBER SYSTEM POWERED FU1 to FU9 Stationary Gripper Coils Movable Bus-Duct Moving Gripper Coils i Plug-in Unit A102-FU1 to FU3 Lift Bus-Duct Lift Coils Plug-in Unit A101-FU1 to FU3 l t g COMANCHE PEAK - UNIT 1 3/4 8-32 i

y &

TABLE 3.8-1 (Continued)

CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES I DEVICE NUMBER SYSTEM AND LOCATION POWERED

6. 120V Space' Heater Circuits Containment Recire. Fan from 480V Switchgears and CROM Vent. Fan Motor Space Heaters

a. Primary Breakers BKR. LOCATION WESTINGHOUSE

                                             & NUMBER                                       BKR. TYPE Swgr. 1EB1,                                    EB1010 Cubicle 3A CP1-VAFNAV-01 Space Heater Bkr.                                                                    ,

Swgr. IEB2, ' EB1010 Cubicle 3A - CP1-VAFNAV-02 Space Heater Bkr.

                                           'Swgr. 1EB3,                                     EB1010
                                            ' Cubicle 9A                                                                   -

CP1-VAFNAV-03 Space Heater Bkr. Swgr. 1EB4, EB1010 Cuoicle 9A

                                     .       CP1-VAFNAV-04 Space Heater Bkr.

Swgr. 1EB3, EB1010 Cubicle 8A, CP1-VAFNCB-01 Space Heater Bkr. Swgr. IEB4, EB1010 Cubicle 8A CP1-VAFNAV-02 Space Heater Bkr.

b. Backup Breakers BKR. LOCATION GENERAL ELECTRIC

                                             & NUMBER                                  BKR. TYPE Panel IEC3-2'                            TED .

Ckt. No. 3 Panel 1EC3-2 TED Ckt. No. 4 COMANCHE PEAK - UNIT 1 3/4 8-33

  ~ - - --- - - --

TABLE 3.8-1 (Continued)- Igj CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES w \ OEVICE' NUMBER SYSTEM

                     'AND LOCATION                                                 POWERED                  .

G. 120V Space Heater Circuits from 480V'Switchgears (Continued) t BKR. LOCATION GENERAL ELECTRIC

                                 & NUMBER                         BKR. TYPE
                   ~

Panel 1EC4-2 TED Ckt. No. 3 Panel IEC4-2 TED Ckt. No. 4

7. 120V Space Heater C.ircuit,s From 480V MCC's

a. Primary Fuses Location - Each MCC Starter Compartment MCC's IEB1-2, IEB2-2, IEB3-2 and 1EB4-2.

b. Backup Fuses- ,

FUSE LOCATION * . AND NUMBER SYSTEM POWERED , MCC 1EB1-2 Space Heater Circuits from . Compt. 12E, IFU MCCIEB1-2 MCC 1EB2-2 Space Heater Circuits from Compt. 12F, IFO MCC 1EB2-2 l MCC IEB3-2 Space Heater Circuits from l Compt. 7C, IFU MCC IEB3-2 j MCC IEB4-2 Space Heater Circuits from ! Compt. 6C, IFU MCC 1EB4-2 l 4 COMANCHE PEAK - UNIT 1 3/4 8-34 1

t , . TABLE 3.8-1(Continued g,p p { ' CONTAINMENT PENETRATION CONDUCTOR N b OVERCURRENT PROTECTIVE DEVICES'~ t DEVICE NUMBER SYSTEM i AND LOCATION . POWERED *

8. 125V DC Lighting Emergency DC Lighting f

a. Primary Breaker BREAKER LOCATION G. E. BKR. :

AND NUMBER TYPE DC Panelboard TFJ i 102-1, Ckt #6 - l b. Backup Fuse

                 .                                                                                FUSE LOCATION           .

i AND NUMBER DC Switchboard 102, Ckt. #1-2 .

9. 125V DC Control Power Various .

a. Primary Devices *- 3 Amp fuses in termination cabinets listed below with backup devices. *

b. Backup Breakers ;

GENERAL ELECTRIC CAB. NO. PANELBOARD NO. CKT. NO. BREAKER TYPE t 01 XE01-1 1,6,7,8,9,10 TED 02 XE02-1 1,3,6,7,8,9,10 TED 03 X02-3 8,9,12,14,17 TED 04 XE01-1 1,6,7,8,9,10 TED 05 1E02-1 7,10,12,15,16,17 TED

                     . 06                                                                                 X02-3                                     8,9,12,14,17                TED
                    ' 07                                                                                  IE01-1                                    7,10,14,17                  TED 68                                                                 XE02-1                                    1,3,6,7,8,9,10              TED

09 102-3 7,10,11,14,17 TED 10 1E01-1 7,10,14,17 TED 11 1E02-1 . 7,10,12,15,16,17 TED 13 1E01-1 7,10.14,17 TED 39A X02-1 11 TED COMANCHE PEAK - UNIT 1 3/4 8-35

it. rf . . i . TABLE 3.8-1 (Continued) CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE NUMBER . AND LOCATION -

10. 118V AC Instrument Distribution Panel Board 103-3

a. Primary Device -

b. Backup Breaker - GC Type TED

located in instrument Distribution Panel Board 103-CK #11 .

11. 120V AC Power for Personnel and Emergency Airlocks ,

a. Primary Devices ,

i

b. Backup Breakers GENERAL ELECTRIC PANELBOARD NO. CKT NO. BREAKER TYPE XEC2 34 TED XEC1-2 . 2 TED t

12. 118V AC Control Power L

a. Primary Devices

b. Backup Breakers GENERAL ELECTRIC l PANELBOARD NO. CKT. NO. BREAKER TYPE XEC12-1 3,5,7,9,10,12 TED XEC2-1 3,5,7,9,10,12 TED 102 12,22 . TED 1C3 12,14 TED IPC1

): 10,13 TED IPC2 10 TED

           ,   IPC4                       6,10                          TED 1EC1                       3,4,8,9                       TED 1EC2                       3,4,7,9                       TED                                                                         -

1ECS 3,8 TED IEC6 3,8 TED , I l .i COMANCHE PEAK - UNIT 1 3/4 8-36

   ,I . r                                                                                     .

TABLE 3.8-1 (Continued)

p CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PR0iECTIVE DEVICES

                                                                                                   . r . ri({'"
                                                                                                        .?

DEVICE NUMBER , AND LOCATION

13. Emergency Evacuation System Warning Lights Power-

Prinary Devices

_a.

b. Backup Bre'akers

SQUARE O SINGLC POLE PANELBOARD N3. UKT. NO. BREAKER TYPE "XEC3 3 FAL-12020 XEC4 3 FAL-12020

14. ORPI Data Cabinet. Dower Supplies

a. Primary Devices

t. Backup Breakers

                                           ..                                GENERAL ELECTRIC PANELBOARD NO.                CKT. NO.            BREAKER TYPE 1C14                          1,2              TED i

l l l ? l l l l .C0HANCHE PEAK - UNIT 1 3/4 8-37

ELECTRICAL POWER SYSTEMS

   -MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION AND BYPASS DEVICES
   -LIMITING CONDITION FOR OPERATION 3.S.A.3      The ticrmal overload protection and bypass devices, integral with.the motor starter of each valve listed in Table 3.8-2 shall be OPERABLE.

APPLICABILITY: Whenever the motor-operated valve is required to be OPERABLE. ACTION: With one or more of the thermal overload protection and/or bypass devices inoperable, declare the affected valve (s) inoperable and apply the appropriate ACTION Statement (s) for the affected valve (s). SURVEILLANCE REQUIREMENTS 4.8.4.3 The above required thermal overload protection and bypass devices shall be demonstrated OPERABLE:

a. At least'once per 18 months, by the performance of a TRIP ACTUATING DEVICE OPERATIONAL TEST of th'e bypass circuitry for those thermal overload devices which are either: ,

1. Continously bypassed and temporarily placed in force only when the valve motors are undergoing periodic or maintenance testing, or

2. No mally in force during plant operation and bypassed under I accident conditions.

b. At least once per 18 months by the performance of a CHANNEL i CALIBRATION of a representative sample of at least 25% of:

1. All thers.21 overload devices which are not bypassed, such that each non.-bypassed device.is calibrated at least once per 6 years.

2. All thermal overload d'evices which are continuously bypassed l

and temporarily placed in force only when the value motors are undergoing periodic or maintenance testing, and therme, overload . l devices normally in force and oypassed under accident conditions such that each thermal overload is calibrated and each valve is cycled through at least one complete cycle of full travel with the motor-operator when the thermal overload is OPERABLE and not bypassed, at least once per 6 years. , i COMANCHE PEAK - UNIT 1 3/4 8-38

TABLE 3.8-2 (t MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION , BYPASS DEVICE SYSTEM (S)' VALVE NUMBER (Continuous)(Accident Conditions)(No) AFFECTED~ 6 . W

                       ,                                t                                                                                                                                         .

4

 's I
                                                                                                                                                                                                                                    .h p                                                                                                                                                         .

l l l l. l I COMANCHE PEAK - UNIT 1 3/4 8-39

h nTT 3/4.9 REFUELING OPERATIONS [ I 3/4.9.1 BORON CONCENTRATION - LIMI ING CONDITION FOR OPERATION 3.9.1 The boron concentration of all filled portions o,f the Reactor Co,olant 5 stem and the refueling canal shall be maintained uniform and sufficient to

ensure that the more restrictive of the following reactivity conditions is met; either: ~

a. A K,ff.of 0.95 or less, or ~

b. A boron concentration of greater than or equal to 2000 ppm,*

Additionally, either valve 1C5-8455 or valves 105-8560, FCV-1118, ICS-8439, 105-8441 and 1CS-8453 shall be closed and secured in position. , APPLICABILITY: MODE 6. ACTION:

a. With the requirements a or b of the above not satisfied, immediately suspend all operations involving' CORE ALTERATIONS or positive reacti-vity changes and initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm boron or its equivalent until K gff is reduced .; less than or equal' to 0.95 'or .the bo'ron concentration is restored to great'er' than or equal to 2000 ppm, whichever is the more res.trictive,

b. If either valve 105-8455 or valves 1C5-8560~, FCV-111B, ICS-8439, 1C$-8441 and 105-8453 are not closed and secured in position, immediately suspend all operati'ons involving CORE ALTERATIONS or

positive reactivity changes and take action to isolate the dilution paths. Within 1 hour, verify the more restrictive of 3.9.1.a or 3.9.1.b or carry out Action a. above. SURVEILLANCE REQUIREMENTS

4. 9.1.1 The more restrictive of the above two reactivity conditions shall be determined prior to:

a. Removing or unbolting tae reactor vessel head, and

b. Withdrawal of any full-length control rod in excess of 3 feet from its fully inserted position within the reactor vessel.

4.9.1.0 The boron concentration of t.he Reactor Coolant System and the refueling ce:ial shall be determined by chemical analysis at least once per 72 hours.

4. 9.1. 3 Either valve 1C5-8455 or valves 105-8560, FCV-1118, 1C5-8439, M S-8441 and 105-8453 shall be verified closed'and secured in position by mechanical

stops or by removal of air or electrical power at least once per 31 days to verify that' dilution paths are isolated. ,

     *During initial fuel load, the boron concentration limitation for the refueling canal is not applicable provided the refueling canal level is verifiea to be below the reactor flange elevation at least once per 12 hours.

COMANCHE PEAK - UNIT 1 3/4 9-1

       'REFUE;ING OPERATIONS 3/4.9.2 INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.9.2 As a minimum, two source range neutron flux monitors shall be OPERABLE, each with continuous visual indication in the control room and one with. audible indication in the containment and- control room.

APPLICABILITY: MODE 6. ACTION:

a. With one of the abnve required monitors inoperable or not operating, immediately suspend'all operations involving CORE ALTERATIONS or positive reactivity changes.

b. With bdth of the above required monitors inoperable or not operating, determine the boron concentration of the Reactor Coolant System at least once per 12 hours.

l SURV'EILLANCE REQUIREMENTS 1 a 4.9.2 Each source range neutron flux monitor shall be demonstrated OPERABLE by performance of:

a. A CHANNEL CHECK at least once per 12 hours,

b. An AN'ALOG CHANNEL OPERATIONAL TEST within 8 hours prior to the initial start of CORE ALTERATIONS, and

c. An ANALOG CHANNEL OPERATIONAL TEST at least once per 7 days.

I l l l COMANCHE PEAK - UNIT 1 3/4 9-2

REFUELING OPERATIONS-3/4.9.3 OECAY TIME i

               - LIMITING CONDITION-FOR OPERATION 3.-9.3  The reactor shall be subcritical for at least 100 hours.
 ,              APPLICABILITY: During movement of irradiated fuel in the reactor vessel.                  .

ACTION: With'the reactor subcritical for less than 100 hours, suspend all operations involving movement of irradiated fuel _ in the reactor vessel. 9 SURVEILLANCE REQUIREMENTS __ , 4.9.3 The reactor shall be determined to have been subcritical for at least 100 hours by verification of the date and time of subcriticality prior to movement of irradiated fuel in the reactor vessel. S O 4 h 3/4 9-3

COMANCHE PEAK - UNIT 1

l . . . I REFUELING OPERATIONS A

 \                                                                                         '"A (tJ>
             '3/4.9.4- CONTAINMENT BUILDING PENETRATIONS                                \&36{

w LIMITING CONDITION FOR OPERATION f 3.9.4 The containment building penetrations shall be in the'following status:

a. The equipment hatch closed and held in place by a minimum of four bolts,

b. A minimum of one door in each airlock is closed, and

c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere shall be either:

1) Closed by an isolation valve, blind flange, or manual valve, or

2) Be capable of.be'ing closed by an OPERABLE automatic containment ventilation isolation valve.

APPLICABILITY: During CORE ALTERATIONS or movement of irradiated fuel.within the containmer.t. ACTION:

          ~.With'the req 0irements of the aboy? specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or movement of irradiated fuel in the containment building.

i SURVEILLANCE REQUIREMENTS , I 4.9.4 Each of the above required containment building penetrations shall be determined to be either in its closed / isolated condition or capable of being closed by an OPERABLE automatic containment isolation valve within 100 hours prior to the start of and at least once per 7 aays during CORE ALTERATIONS or movement of irradiated fuel in the containmeat building by: a. Verifying or the penetrations are in their closed / isolated cor.11 tion, b. Testing the containment purge and exhtust isolation valves per the applicable portions of Specification 4.6.4.2. I COMANCHE PEAK - UNIT 1 3/4 9-4

J REFUELING OPERATIONS 9 NA g 3/4.9.5 COMMUNICATIONS LIMITING CONDITION-FOR OPERATION.

3. 9. 5. Direct _ communications shall be maintained between the control room and personnel at the refueling station.

APPLICABILITY: During CORE ALTERATIONS. ACTICN: When direct communications between the control room and personnel at the refueling station cannot be maintained, suspend.all CORE ALTERATIONS.

   ,       SURVEILLANCE REQUIREMENTS-                                  -

' 4.9.5 Direct communications between the control room and personnel at the

       ,   refue' ling station shall be demonstrated within 1 hour prior to the start of and at least once per 12 hours during CORE ALTERATIONS.

A COMANCHE PEAK - UNIT 1 3/4 9-5 l t

0 REFUELING OPERATIONS k 3/4.9.6 REFUELING MACHINE LIMITING CONDITION FOR OPERATION 3.9.6 The refueling machine main hoist and auxiliary monorail hoist shall be used for movement of drive rods or fuel assemblies and shall be OPERABLE with:

a. The rtfueling machine main hoist used for movement of fuel assemblies havinr,:; 1) A minimum capacity of 2850 pounds, and; 2) An overload cutoff limit less than or equal to 2800 pounds.; b. The auxiliary monorail hoist used for latching, unlatching and
movement of control rod drive shafts having: .; 1) A minimum' capacity of 610 pounds, and; 2) A load indicator which shall be.used to prevent lifting loads in excess of 600 pounds.

APPLICABILITY: During movement of fuel assemblies and/or latching, unlatching or movement of control rod drive shafts within the reactor vessel. ACTION: With the requirements for refueling machine main hoist and/or auxiliary monorail hoist OPERABILITY not satisfied, suspend use of any inoperable refueling machine main hoist and/or auxiliary monorail hoist from operations involving the movement of fuel assemblies and/or latching, unlatching, and movement of control rod drive shafts within the reactor vessel. SURVEILLANCE REQUIREMENTS l 4.9.6.1 The refueling machine main hoist used for movement of fuel assemblies within the reactor vessel shall be demonstrated OPERABLE within 100 hours prior to the start of such operations ~by performing a load test of at least 2850 pounds and demenstrating an automatic load cutoff when the crane load exceeds 2800 pounds.

 '4.9.6.2 The auxiliary monorail hoist and associated load indicator used for latching, unlatching, movement of control rod drive shafts within the reactor vessel shall be demonstrated OPERABLE within 100 hours prior to the start of such operations by performing a load test of at least 610 pounds.                  .

e COMANCHE PEAK - UNIT 1 3/4 9-6

L

     .                                                                             - '- es REFUELING OPERATIONS, 3/>'.9.7'      CRANE TRAVEL - SPENT FUEL STORAGE AREAS

! LIMITING CONDITION FOR OPERATION 3.9.7 Loads inLexcess of 2150 pounds shall be prohibited from travel over fuel assemblies in a storage pool. APPLICABILITY: With fuel assemblies in a storage pool. ACTION:

a. With the requirements of the above specification not satisfied, place the crane load in a safe condition.

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

SURVEILLANCE REQUIREMENTS 4.9.7 Crane interlocks and physical stops which prevent crane travel with loads in excess of 2150 pounds over fuel assemblies shall be demonstrated OPERABLE within 7 days prior to crane use and at least once per 7 days thereafter during crane operation. 2 COMANCHE PE/.K - UNIT 1 3/4 9-7

9 REF'UELING OPERATIONS (. 3/4.9.8 ~ RESIOUAL HEAT REMOVAL AND COOLANT CIRCULATION

 .           HIGH WATER LEVEL
     . E LIMITING CONDITION FOR OPERATION 3.9.8.1 At least one residual heat removal (RHR) loop shall be OPERABLE and in operation.*

APPLICABILITY: MODE 6, when the water level above the top of the reactor vessel flange is greater than or equal to 23 feet.

           ' ACTION:

With no RHR loop OPERABLE and in operation, suspend all operations involving l an_ increase in the reactor decay heat load or a reduction in. boron concentration of-the Reactor Co,ola'nt System and immediately initiate corrective action to return the required RHR loop to OPERABLE and operating status as s6on as possible. Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours. SURVEILLANCE' REQUIREMENTS 4.9.8.1 At least one RHR locp shall be verified in operation and circulating reactor coolant at a flow rate of greater than or equal to 3800 gpm at least once per 12 hours. l l l l *The RHR . loop may be removed from operation for up.to 1 hour per 8-hour period i during the performance of CORE ALTERATIONS in the vicinity of the reactor vessel l hot legs. l COMANCHE PEAK - UNIT 1 3/4 9-8 l 1

REFUELING OPERATIONS LOW WATER LEVEL ki 5 *

     ~ LIMITING CONDITION FOR OPERATION 3.9'.8.2    Two independent residual heat removal (RHR) loops shall be dPERABLE, and at least one RHR loop shall be in operation.*

APPLICABILITY: MODE 6, when the water level above the top of the reactor vessel flange is less than 23 feet. ACTION:

a. With less than the required RHR loops OPERABLE, immediately initiate corrective action to return the required RHR loops to OPERABLE status, or to establish greater than or equal to 23 feet of water above the reactor vessel flange, as soon as possible.

b. With no RHR loop in operation, suspend all operations involving a

[ reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to operation. Close all containment penetrations providing direct access.from the containment atmosphere to the outside atmosphere within 4 hours. SURVEILLANCE REQUIREMENTS - 4.9.8.2 At least one RHR loop shall be verified in operation and circulating reactor coolant at a flow rate of greater than or equal to 3800 gpm at least once per 12 hours, i l

Prior to initial criticality, the RHR loop may be removed from operation for up to 1 hour per 8-hour period during the performance of CORE ALTERATIONS in the vicinity of the reactor vessel hot legs, i

COMANCHE PEAK - UNIT 1 3/4 9-9

i . REFUELING OPERATIONS 3/4.9.9 CONTAINMENT VENTILATION ISOLATION SYSTEM LIMITING CONDITION FOR OPERATION 3.9.9 .The Containment Ventilation Isolation System shall be OPERABLE. APPLICABILITY: During CORE ALTERATIONS or movement of irradiated fuel within the containment.

ACTION: - a.

 -                    With the Containment Ventilation Isolation System inoperable, close each of the affected ventilation penetration (s) providing direct access from the containment-atmosphere to the outside atmosphere.

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

SURVEILLANCE REQUIREMENTS . . 4.9.9 The Containment Ventilation Isolation System shall be demonstrated OPERABLE within 100 hours prior to the start of and at least once per 7 days during CORE ALTERATIONS by verifying that containment ventilstion isolation occurs on manual initiation and on a High Radiation test signal from each of the containment air radiation monitoring instrumentation channels. COMANCHE PF.AK - UNIT 1 3/4 9-10

REFUELING OPERATIONS 3/4.9.10 WATER LEVEL - REACTOR VESSEL

     . LIMITING CONDITION FOR OPERdTION 3.9.10 At least 23 feet of water shall be maintained over the top of the reactor vessel flange.

APPLICABILITY: During movement of fuel assemblies or control rods within the containment when either the fuel assemblies being moved or the fuel assemblies seated within the reactor vessel are irradiated while in MODE 6. - ACTION: . With the requirements of the above specification not satisfied, suspend all operations involving movement of fuel assemblies or control rods within the reactor vessel. SURVEILLANCE REQUIREMENTS . 4.9.10 The water level shall be determined to be at least its minimum required o depth within 2 hours prior to the start of and at least once per 24 hours thereafter during movement of fuel assemblies or control rods. COMANCHE PEAK - UNIT 1 3/4 9-11 1

REFUELING OPERATIONS , 3/4.9.11 WATER LEVEL - IRRADIATED FUEL STORAGE iI

          -LIMITING CONDITION FOR OPERATION 3.9.11L At least 23 feet of water shall be maintained over the top of irradiated fuel assemblies seated in the storage racks.

APPLICABILITY: Whenever irradiated fuel assemblies are in the storage rack. , ACTION:

a. With the requirements of the above specification not satisfied, suspend all movement of fuel assemblies and crane operations with loads in the fuel storage areas and restore the water level to within l its limit within 4 hours.

      .          b. The provisions of Specifications 3.0.3 and 3.0.4 are not applica.ble.

l SURVEILLANCE REQUIREMENTS 4.9.11 The water level above the storage racks shall be determined to be at least its minimum required depth at least once per 7 days when irradiated fuel assemblies are in the fuel storage racks. - L . 1 COMANCHE PEAK - UNIT 1 3/4 9-12

REFUELING OPERATIONS 3/4.9.12 FUEL STORAGE POOL AIR CLEANUP SYSTEM LIMITING CONDITION FOR OPERATION 3.9.12 Two independent Fuel Storage Pool Air Cieanup Systems shall be OPERABLE. APPLICABILITY: Whenever irradiated fuel is 'in the s' ' rage pool. ACTION:

a. With one Fuel Storage Pool Air Cleanup System inoperable, fuel movement within the storage pool or crane operation with loads over the storage pool may proceed provided the OPERABLE Fuel Storage Pool Air Cleanup System is capable of being powered from an OPERABLE .

emergency power source and is in operation and discharging through at' least one train of HEPA, filters .and charcoal adsorbers.

b. With no Fuel Storage Pool Air Cleanup System OPERABLE, suspend all operations involving moveinent of fuel within the storage pool or crane operation with loads over the storage pool until at least one Fuel Storage Pool Air Cleanup System is restored to OPERABLE status.

c. The provisions of Specificati'o'ns 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.12 The above required Fuel Storage Pool Air Cleanup Systems shall be demonstrated OPERABLE: i

a. At least once per 31 days on a STAGGERED TEST BASIS by ini',ating, l

from the control room, flow through the HEPA filters and c.3rcoal ' adsorbers and verifying that ti.' system operates for at least 10 continuous hours with the heaters operating;

b. At,least once per 18 months or (1) Ifter any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone

                 ' communicating with the system by:

8 i l l COMANCHE PEAK - UNIT 1 3/4 9-13 l i

REFUELING OPERATIONS ay g$k SURVEILLANCE REQUIREMENTS (Continued)

1) Verifying that the cleanup, system satisfies the in place

penetration and bypass leakage testing acceptance criteria of less than [*]% and uses the test procedure guidance in Regulatory Positions C.S.a, C.S.c, and C.S.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is cfm 2 10%;

2) Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accor-dance sith Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1l52, Revision 2, March 1978, for a methyl iodide penetration of less than [**]%;

and

3) Verifying a system flow rate of cfm i 10% during system operation when tested in accordance with ANSI N510-1975.

c. After every 720 hours of charcoal adsorber operation by verifying, within 31 days after removal, that'a laboratory analysis of a representative carbon sample obtained in accordance.with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2,' March 1978, meets the laboratory testing criteria ~ of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978,'for a methyl iodide penetration of less than [**]%.

d. At least once per 18 months by:

1) Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than [6] inches Water Gauge while operating the system at a flow rate of cfm i 10%,

2) Verifying that on a High Radiation test signal, the system automatically starts (unless already operating) and directs its.

exhaust flow through the HEPA filters and charcoal 'adsorber banks, f COMANCHE PEAK - UNIT 1 3/4 9-14

REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS (Continued) hi pa-

3) Verifying that the. system ma.intains the spent fuel storage pool -

area.at 'a negative pres'sure of greater than or equal to [1/4] inch Water Gauge relative to the outside atmosphere during system operation,

4) Verifying that the filter cooling bypass valves can be manually opened, and

5) Verifying that the heaters dissipate i kW when tested in accordance with ANSI N510-1975.

e. After each complete or partial replacement of a HEPA filter bank, by verifying that the cleanup system satisfies the in place penetration and bypass leakage testing acceptance criteria of less than [*]% in accor. dance with ANSI N510-1975 for a 00P test aerosol while operating the system at a flow rate of cfm i 10%.

f. After each complete or partial replacement of a charcoal adsorber bank, by verifying that the cleanup system satisfies the in place penetration and bypass leakage testing acceptance criteria of less than [*]% in accordance with ANSI N510-1975 for a halogenated

             ~ hydrocarbon refrigerant test gas while operating the system at a flow rate of         cfm i 10%.                         -
    *0.05% value applicable when a HEPA filter or charcoal adsorber efficiency of 99% is assumed, or 1% when a HEPA filter or charcoal adsorber efficiency of 95% of less is assumed in the NRC staff's safety evaluation. (Use the value assumed for the charcoal adsorber efficiency if the value for the HEPA filter is different from the charcoal adsorber efficiency in the NRC staff's safety evaluation).
   **Value applicable will be determined by the following equation:

P = 10 -E , when P equa'.s the value to be used in the test requirement (%), E is efficiency assumed in the SER for methyl iodide removal (%), and SF is. the safety factor to account for charcoal degradation'between tests (5 for systems with heaters and 7 for systems without heaters). COMANCHE PEAK - UNIT 1 3/4 9-15 1

=

   -3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN                                                                      .           ,

LIMITING CONDITION FOR OPERATION 3.10.1 The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 may be - suspended for measure' ment of control rod worth and SHUTDOWN MARGIN provided i reactivity equivalent to at least the highest estimated control rod worth is available for trip insertion from 0PERABLE control rod (s). APPLICABILITY., MODE 2. ACTION:

a. With any control rod not fully inserted and with less than the above reactivity equivalent available for trip insertion, immediately ini-tiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than.or equal to 7000 ppm boron or.its.

e'quivalent until the SHUTOOWN MARGIN required by Specification 3.1.i.1 is restored.

b. With all control rods fully inserted and the reactor subcritical by less than the above reactivity equivalent, immediately initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm boron or its. equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.

SURVEILLANCE REQUIREMENTS 4.10.1.1~ The position of each control rod either partially or fully withdrawn shall be determined at least once per 2 hours. 4.10.1.2 Each control rod not fully inserted shall be demonstrated capable of f-ull insertion when tripped from at least the 50% withdrawn position within 24 hours prior to reducing the SHUTDOWN MARGIN to less than the limits of Specification 3.1.1.1. e

COMANCHE PEAK - UNIT 1 3/4 10-1

SPECIAL TEST EXCEPTIONS g','{

                                                                                                >T w                        l 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS               i *i " ~

LIMITING CONDITION FOR OPERATION t-l . l 3.10.2 The group height, insertion, and power distribution limits of -

Specifications-3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.1, and 3.2.4 may be suspended

              'during the performance of PHYSICS TESTS provided:            ,

a. The THERMAL POWER is maintained less than or equal to 85% of RATED THERMAL POWER, and

b. The limits of Specifications 3.2.2 and 3.2.3 are maintain +;

and determined at the frequencies specified in Specification

        .                  4.10.2.2 below.

APPLICABILITY: H0DE 1. ACTION: With any of the limits of Specification 3.2.2 or 3.2.3 being exceeded while the requirements of Specifications 3.1.3.1, 3 1.3.5, 3.1.3.6, 3.2.1, and 3.2.4

            ' are suspended, either:

l. a. Reduce THERMAL POWER sufficient to satisfy the ACTION requirements of Specifications 3.2.2 and 3.2.3, or

b. B'e in HOT STANDBY within 6 hours.

i SURVEILLANCE REQUIREMENTS 4.10.2.1 The THERMAL POWER shall be determined to be less than or equal to 85% of RATED THERMAL POWER at least on e*per hour during PHYSICS TESTS. 4.10.2.2 The requirements of the below listed specif'ications shall be performed at least once per 12 hours during PHYSICS TESTS:

a. Specifications 4.2.2.2

b. Specification 4.2.2.3, and

c. Specification 4.2.3.2.

O / COMANCHE PEAK - UNIT 1 3/4 10-2

SPECIAL TEST EXCEPTIONS 3/4.10.3 PHYSICS TESTS ' LI!-ilTING CONDITION FOR OPERATION 3.10.3 The limitations of Specifications 3.1.1.3, 3.1.1.4, 3.1.3.1, 3.1.3.5, and 3.1.3.6 may be suspended during the performance of PHYSICS TESTS provided:

a. The THERMAL POWER does not exceed 5% of RATED THERMAL POWER,

b. The Reactor Trip Setpoints on the OPERABLE Intermediate and Power Range channels are set in accordance with Table 2.2-1 Functional Units 5 and 2b, and

c. The Reactor Coolant Syste'n lowest operating loop temperature (Tavg) is greater than or equal to 541 F.

APPLICABILITY: MODE 2. ACTION: a.' With the THERMAL POWER greater than 5% of RATED THERMAL POWER, immediately open the reactor trip breakers.

     .         b. With a' Reactor Coolant S'ystem operating loop temperature (Tavg) less than 541 F, restore T,yg to within'its limit within 15 minutes or be in at least HOT STANDBY within the next 15 minutes, j      SURVEILLANCE REQUIREMENTS                                 _

l 4.10.3.1 The THERMAL POWER shall be determined to be less than or equal to 5% of RATED THERMAL POWER at least once per hour during PHYSICS TESTS.

4.10.3.2 Each Intermediate and Power Range channel shall be subjected to an l ANALOG CHANNEL OPERATIONAL TEST within 12 hours prior to initiating PHYSICS l TESTS.

4.10.3.3 The Reactor Coolant System temperature (T,yg) shall be determined to be greater than or equal to 541*F at least once per 30 minutes during PHYSICS TESTS. COMANCHE PEAK UNIT 1 3/4 10-3

i SPECIAL TEST EXCEPTIONS C'C7 # sr 3/4.10.4 REACTOR COOLANT LOOPS {({(1 & LIMITING CONDITION FOR OPERATION 3.10.4 The limitations of Specification 3.4.1.2 may be suspended during the performance of hot rod drop time measurements in MODE 3 provided at least two reactor coolant loops as listed in Specification 3.4.1.2 are OPERABLE. APPLICABILITY: Ouring performance of hot rod drop time measurements. ACTION: With less than the above required reactor coolant loops OPERABLE during the performance of. hot rod drop time measurements, immediately open tne reactor trip breakers and comply with the provision of the action statements of Specification 3.4.1.2. SURVEILLANCE REQUIREMENTS 4.10.4 At least~the above required reartor coolant loops'shall be determined

OPERABLE within 4 hours prior to the initiation of' hot, rod drop time measure-ments by verifying current breaker alignments and indirated power availability and by verifying the indicated secondary side water level to be greater than or equal to 10% narrow range.

1 COMANCHE PEAK - UNIT 1 3/4 10-4 l

9 SPECIAL TEST EXCEPTIONS

     - 3/4.10.5 POSITION INDICATION SYSTEM --SHUTDOWN                         k   \ k{

LIMITING CONDITION FOR OPERATION 3.10.5 The limitations of Specification 3.1.3.3 may be suspended during't'he performance of individual shutdown and control rod drop time measurements provided;

a. Only one shutdown or control bank is withdrawn from the fully inserted position at a time, and

b. The digital rod position indicator is OPERABLE during the withdrawal of the rods.*

APPLICABILITY: M0CES 3, 4, and 5 during performance of rod drop time measurements. ACTION: With the required digital rod position indicator (s) inoperable or with more than one bank of rods withdrawn,-immediately open the Reactor trip breakers. SURVEILLANCE REQUIREMENTS __ 4.10.5 The above required digital rod position indicator (s) shall be determined to be OPERABLE within 24 hours prior to the start of and at least once per 24 hours thereafter during rod drop time measurements by verifying the Demand Position Indication System and the Digital Rod Position Indication System ! agree:

a. Within 12 steps when the rods are stationary, and

b. Within 24 steps during rod motion.

l l l *This requirement is not applicable during the initial calibration of the l Digital Rod Position Indication System provided: (1) K is maintained lessthanorequalto0.95,and(2)onlyoneshutdownoff[ontrolrodbank is withdrawn from the fully inserted position at one time. 1 l COMANCHE PEAK - UNIT 1 3.'4 10-5 1 r

3/4.11 RADIOACTIVE EFFLUENTS f\d 3/4.11.1 LIQUID EFFLUENTS ((,3c

CONCENTRATION . LIMITING CONDITION FOR OPERATION 3.11.1.1 ~The concentration of radicactive material released in liquid effluents to UNRESTRICTED AREAS (see Figure 5.1-3) shall oe limited to the concentrations specified in 10 CFR 20, Appendix B, Table II, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, , s the concentration shall be limited to 2 x 10 4 microcurie /ml total activity. APPLICABIt!TY: At all times. ACTION: With the concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS exceeding the above limits, immediately restore the concen- . tration to within the above limits. . SURVEILLANCE REQUIREMENTS 4.11.1.1.1 Radioactive liquid wastes.shall be sampled and analyzed accdrding to the sampling and analysis program of Table 4.11-1. 4.11.1.1.2 The results of the radioactivity analyses shall be used in ac.cordance with the methodology and parameters in the ODCM to assure that the concentrations at the point of release are maintained within the limits of Specification 3.1'1.1.1. l 4 l l l I l l l l - l l CCHANCHE PEAK - UNIT 1 3/4 11-1

TABLE 4.11-1 [O - RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM LOWER LIMIT MINIMUM OF DETECTION LIQUID RELEASE SAMPLING ANALYSIS TYPE OF ACTIVITY TYPE (LLO)(1) FREQUENCY FREQUENCY ANALYSIS (pCi/ml) ,- 1. Batch Waste P P Release Each Batch Each Batch Principal Gamma 5x10 7 Tanks (2) Emitters ( ) I-131 1x10 6

a. Waste P M Dissolved and 1x30 5 Monitor One Batch /M Entrained Gases Tanks (Gamma Emitters)

b. Laundry P M H-3 1x10 5 Ho.1 dup and Each Batch Comp'osite(4) r Gross Alpha 1x10 7

c. Waste Water P Q Sr-89, Sr-90 5x10 8 Holdup Each Batch Composite (4)

Tanks. Fe-55 1x10 6 l d. Condensate P P Principal Gamma 5x10 7 l Polisher Each Batch Each Batch l Backwash Emitters (3) Recovery I-131 1x10 6 Tanks (6)(7) l e. Component H-3 1x10 5 Cooling Water Drain Tank (7) l 2. Continuous 'W W Principal Gamma 5x10 7 Releases (5) Grab Sample Emitters (3) I-131 1x10 6 l a. Turbine H-3 1x10 5 l Bldg Sumps No. 1 & 2 Effluent (6)(7) , l COMANCHE PEAK - UNIT 1 3/4 11-2 - 4

                                                                      - * - - - - - - _ _ _                            L-_

TABLE 4.11-1 (Continued) h j i.. . TABLE NOTATIONS - (1)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 withonly5%p'real"signal.robabiTity represents a of falsely concluding that a blank observation For a particular measurement system, which may include radiochemical separation: 4' 8 b LLD = E V 2.22 x 106 Y exp (-Aot) l Where: LLD = the "a priori" lower limit of detection (microcurie per unit mass or volume), s b = the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (counts p'er minute), E = the counting' efficiency (counts per disintegration),- V = the sample size (units of mass or volume), 2.22 x 108 = the number of disintegrations per minute per microcurie, Y = the fractional radiochemical yield, when applicable, A = the radioactive decay constant for the particular radionuclide (sec 1), and at = the elapsed time between the midpcint of sample collection and the time of counting (sec). Typical values of E, V, Y, and at should be used in the calculation. It should be recognized that 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. O)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 by a method described in the ODCM to assure representative sampling. COMANCHE PEAK - UNIT 1 3/4 11-3

l TABLE 4.11-1-(Continued) g TABLE NOTATIONS (Continued) (3)The principal gamma emmiters for which the LLD specification ap' plies include the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, 2n-65, Mo-99, Cs-134, Cs-137., and Ce-141. Ce-144 shall also be measured,, but with an LLD of 5 x 10.e. This list does~not mean that only the'se nuclides are to be considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Semiannual Radioactive Effluent Release Report pursuant to Specification 6.9.1.4 in the format outlined in Regulatory Guide 1.21, Appendix B, Revision 1, June 1974. (4)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.

(5)A continuot.s 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. (6)To be representative of the quantities and concentrations of radioactive materials in liquid effluents, samples.shall be collected continuously in proportion to the rate of flow of the effluent stream. Prior to analyses, all samples taken for the composite shall be thoroughly' mixed in order for the composite sample to be representative of the efflu,ent release. e O COMANCHE PEAK - UNIT 1 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 unit, to UNRESTRICTt0 AREAS (see Figure 5.1 4) shall be limited:

a. During any calendar quarter to less than or equal to 1.5 mrems to the whole 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 whole body and to less than or equal to 10 mrem's to any organ.

APPLICABILITY: At all times. ACTION:

a. Wit'h the calculated dose from the release of radioactive materials in liquid effluents exce'eding any of the above limits, prepare and submit to the Cow,ission 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 lidits. This Special Report sh.all also include: (1) the results of radiological analyses of the drinking water source, and (2) the radiological impact on finished drinking water supplies wi.th regard to the requirements of 40 CFR 141, Safe Drinking Water Act.

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

SURVEILLANCE REQUIREMENTS

    . d.11.1.2 Cumulative dose contributions from liquid effluents for the current calendar quarter and the current calendar year shall be determined in' accordance with the methodo. logy and parameters in the ODCM at least once per 31 days.

4 COMANCHE PEAK - UNIT 1 3/4 11-5

7t ' RADI0 ACTIVE EFFLUENTS LIQUID RADWASTE TREATMENT SYSTEM

               ' LIMITING CONDITION FOR OPERATION 3.11.1.3 The liquid radwaste treatmJnt systen shall be OPERABLE and appropriite                    ,

portions of the system shall.be used to reduce releases of radioactivity when the projected doses due to the liquid effluent, from each unit, to UNRE5TRICTED AREAS (see Figure 5.1-4) would exceed 0.06 mrem t0 the whole body or 0.2 mrem to any organ in a 31-day period. , APPLICABILITY: At all times. ACTION: .

a. W4th radioactive liquid waste being discharged without treatment a..d ,

in excess of the above limits and any portion of the liquid radwaste treatment. system not in operation, prepare and submit to the Commis-sian 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 treateent, identification of any inoperable equipment or subsy;,tems, and the reason 4r the inoperability,

2. Action (s) taken to reatore the inoperabic eq'uipment to OPERAB,LE status, and

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

                        . The provisions of Specifications 3.0.3 and 3.0.4 are not applicable, SURVEILLANCE REQUIREMENTS 4.11.1.3.1 Doses due to liquid releases from each unit to UNRESTRICTED AREAS
    .c          shall be prejected at least once per 31 days in accordance with the methodology end parameters in the ODCM when liquid radwaste treatment systems are not being fully utilized.                            .

4.1111.3.2 The installed liquid radwaste treatment system shall be considered OPERABLE by caeting Specifications 3.11.1.1 and 3.11.1.2. i h.. A o'

l I' t l COMANCHE PEAK - UNIT 1 3/4 11-6 l

RADIOACTIVE EFFLUENTS g LIQUID HOLOUP TANKS * , . . LIMITING CONDITION FOR OPERATION 3.11.1.4 The quanti.ty of radioactive material contained in each unprotected outdoor tanks 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 unprotected outdoor tanks exceeding the above limit, immediately suspend all additions of radioactive material to the tank, within 48 hours reduce the tank contents'to within the limit, and desc. ribe the events leading to this condition in the next Semiannual Radioactive Effluent Release -

                    , Report,. pursuant to Specification 6.9.1.4.

b. The provisions of Specif.ications 3.0.3 and 3.0.4 are not applicable.

i 9 SURVEILLANCE REQUIREMENT 5 , 4.11.1.4 The quantity of radioactive material contained in each of the unprotected outdoor tanks shall be determined tc be within the above lirait 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. , 1 i i

Tanks included in this specification are those outdoor tanks that are not '

l sorrounded 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. 1 COMANCHE PEAK - UNIT 1 3/4 11-7

RADI0 ACTIVE 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 EXCLUSION AREA BOUNDARY (see Figure 5.1-1) shall be limited to the following:

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

b. For Iodine-131, for Iodine-133, for tritium, and for all radio-nuclides 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, immediately 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 to be within the above limits in accordance with the methodology and parameters in the ODCM. 4.11.2.1.2 The dose rate due to Iodine-131, Iodine-133, 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 methodology and parameters in the ODCM by obtaining representative s3mples and performing analyses in accordance with the sampling and analysis program specified in Table 4.11-2. l COMANCHE PEAK - UNIT 1 3/4 11-8

a , i-TABLE 4.11-2

g RADIOACTIVE GASEOUS WASTE SAMPLING ANI) ANALYSIS PROGRAM E g MINIMUM . LOWER LIMIT OF g - SAMPLING, ANALYSIS TYPE OF. DETECTI,0N (LLD)g)

   ,     GASEOUS RE!.[ASE TYPE FREQUENCY         FREQUENCY        ACTIVITY ANALYSIS                   (pCi/ml)             ,

k 1. Waste Gas Storage Tank P Each Tank P Each Tank Principal Gamma Emitters (2)' 1x10 4

    ,                                                                                                         ~

c Grab Sampic z 2, Containment Purge P P U or Vent Each Release fI Each Release ( Principal Gamma Emitters (2) 1x10 4

   -                           Grab Sample M            H-3 (oxide)           ,            1x10 6

3. a. Plant Vent M(3),(4),(S) g(3) Principal Gamma Emitters (2) 1x10 4-Grab Sample

                                                               ~H-3 (oxide)                         1x10 4             _

q 4. All Release Types Continuous (6) y(7) - g-131 1x10 12

as listed in'1., Radioiodine g 2., and 3. above Absorber Continuous (6) y(7) Principal Gamma Emitters (2) 1x10 81 Particulate 5 ample Continuous (6) M Gross Alpha 1x10 31 Composite Par- ;

ticulate Sample Continuous (6) Q Sr-89, Sr-90 1x10 81 Composite Par- - l ticulate Sample l e s~

a m .. . . . .. . . . . . TABLE 4.11-2 (Continued) TABLE NOTATIONS - (3)The LLD is defined,-for purposes of these specifications, as the smallest concentration of radioactive material in a s?.mple that will yield a net count, above system background, that will be detected with.95% probability . with only 5% probability of falsely concludirig that a blank observation ' represents a "real" signal. For a particular measurement system, which may include radiochemical separation: 4.66 s b LLD = exp (-Aat) - E V 2.22 x 106 'Y Where: LLO = the "a priori" lower limit of detection (microcurie per unit mass or volume),- s b = the standard deviation of the background counting rate or of the counting rate 'of a blank sample as appropriate. (counts per minute),' E = the counting efficiency (counts per disintegration), V = the sample siz'e (units of.' mass or volume), , 2.22 x 108 = th'e number of disintegrations per minute per microcurie, Y = the fractiona'i radiochemical yield, when applicable, A = the radioactive decay constant for the particular radionuclide (sec 1), and 7 at = the elapsed time between the midpoint of sample collection and the time of counting (sec). Typical values of E, V, Y, and at should be used in the calculation. It should be recognized that 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. 4 ^ COMANCHE PEAK - UNIT 1 3/4 11-10

TABLE 4.11-2 (Continued) d TABLE NOTATIONS (Continued) ,[ (2)The principal gamma emitters for which the LLD specification applies include the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 in noble gas releases and Mn-54, Fe-59, Co-58, Co-60,

      .Zn-65, Mo-99, I-131 Cs-134, Cs-137, Ce-l'41 and Ce-144 in Iodine and particu1&te releases. This list does not'mean that only these nuclides are to be considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Semiannual Radioactive Effluent Release Report pursuant to Specification 6.9.1.4 in the format outlined in Regulatory Guide 1.21, Appendix B, Revisi n 1, June 1974.

(3) Sampling and analysis shall also be performed following shutdown, startup, or a THERMAL POWER change exceeding 15% of RATED THERMAL POWER within a 1-hour period. (4) Tritium grab samples shall be taken at least once per 24 hours when the refueling canal is flooded. (5) 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'fue1 pool. (6)The ratio of the sample flow rate to the sampled stream flow rate shallebe

known for the time period covered by each doce or dose rate calculation made in accordance with Specifications 3.11.2.1, 3.11.2.2, and 3.11.2.3. ( ) Samples shall be changed at least once per 7 days and analyses shall be completed within 48 hours after changing, or after removal from sampler. Saapling shall 4.lso be performed at least once per 24 hours for at least - 7 days following each shutdown, startup, or THERMAL POWER change exceeding 15% of RATED THERMAL POWER within a 1-hour period and analyses shall be completed within 48 hours of changing. When samples cellected for 24 hours " are analyzed, the corresponding LL0s may be increased by a factor of 10. This requirement does not apply if: (1) analysis shows that the DOSE EQUIVALENT I-131 concentration in the reactor coolant has not increas6 ' 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. . o COMANCHE PEAK - UNIT 1 3/4 11-11

RADIOACTIVE EFFLUENTS ( DOSE - NOBLE GASES LIMITING CONDITION FOR OPERATION 3.11.2.2 The air dose due td noble ga.ses. released in gaseous effluents, from each unit, to areas at and beyond'the EXCLUSION AREA BOUNDARY (see Figure 5.1-1) shall be limited to the fe 'nwing: -

a. During any caleva. quarter: Less than or equal to 5 mrads for .

gamma radiation and less than cr equal.to 10 mrads for beta radiation, and

b. During any calendar year: Less chan or equal to 10 mrads for gamma radiation and less than or equai to 20 mrtds for beta radiation.

APPLICABILITY: At al.1 times. ACTION

a. With the calculated air dose from radioactive noble gases in' gaseous effluents exceeding any of the above limits, preoare and submit to the Qommission within 30 days, pu.suant to Specification 6.9.2, a Special Report that identtiice the cause(s) for exceeding the limit (s} i

              'and defines the corrective actions that have .sen taken to reduce the releases and the proposed ',orrective actions to be t: ken to assure that subsequent releases will be is compliance'with.the above limits.                                                                  '

b. The provisions of Specif. cations 3.0.3 and 3.0.4 ama not applicable.

SURVEILLANCE REQUIREMENTS ___ _ 4.11.2.2 Cumulative dose contributions for the current calendar quarter and current calendar year for noble gases shall be determined in accordance with the methodology and parameters in the 00CM at least once per 31 days. COMANCHE PEAK - UNIT 1 3/4 11-12 b

RADIOACTIVE EFFLUENTS DOSE - 10 DINE-131, IODINE-133, TRITIUM, AND RADIOACTIVE MATERIAL IN PARTICULATE FORM LIMITING CONDITION FOR OPERATION 3.11.2.3 The dose to a MEMBER OF THE PUBLI'C from Iodine-131, Iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released, from each unit, to areas-at and beyond the EXCLUSION AREA B0UNDARY (see Figure 5.1-1) 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 dose from the release of Iodine-131, Iodine-133, tritium, and radionuclides in particulate form with. half-lives greater than 8 days, in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant

                    .to Specificatioh 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit (s) and defines the corrective artions 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 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, Iodine-133, tritium and radionuclides in particulate form with half-lives greater than 8 days shall be determined in accordance with the methodology and parameters in the ODCM at least once per 31 days. CCMANCHE PEAK - UNIT 1 3/4 11-13

l: , RADIOACTIVE EFFLUENTS ( GASEOUS RADWASTE TREATMENT SYSTEM LIMITING CONDITION FOR OPERATION 3.-11.2.4 The PRIMARY PLANT VENTILATION SYSTEM and the GASEOUS WASTE PROCESSING SYS(EM thall be OPERABLE and appropriate portions of these systems shall be used to reduce-releases of radioactivity when the projected doses in 31 days d'Je to gaseous effluent releases, from each unit, to areas at and beyond the EXCLUSION AREA BOUNDARY (see Figure 5.1-1) would exceed:

a. 0.2 mrad to air from gamma radiation, or-

b. 0.4 mrad to air from beta radiation, or

c. 0.3 mrem to any organ of a MEMBER OF THE PUBLIC.

APPLICABILITY: At all times. ICTION:

a. With radioactive gaseous waste being discharged without treatment l- and in excess of the above limits, prepare and submit to the -

' Comission within 30 days, pursuant to Specification 6.9.2, a Special Report that includes the following information:

1. Ioentification of sny inoperable equipment or s'ubsystems, and

the reason for the inoperability,

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

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

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

SURVEILLANCE REQUIREMENTS _ _ 4.11.2.4.1 Doses due to gaseous releases from each unit to areas at and beyond the EXCLUSION AREA BOUNDARY shall be projected at least once per 31 days l in accordance with the methodology and parameters in the ODCM when Gasecus L Radwaste Treatment Systems are not being fully utilized. 4.11.2.4.2 The insta.11ed PRIMARY PLNG VENTILATION SYSTEM and GASEOUS WASTE PROCESSING SYSTEM shall be considered OPERABLE by mecting Specificati'ons 3.11.2.1 and 3.11.2.2 or'3.11.2.3. i l L COMANCHE PFAK - UNIT 1 3/4 11-14

RADIOACTIVE EFFLUENTS EXPLOSIVE GAS MIXTURE d'

                                                                                             .t,'

LIMITING CONDITION FOR OPERATION _ 3.11.2.5 The concentration of oxygen in the WASTE GAS HOLOUP 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 HOLOUP SYSTEM greater than 2% by volume but less than or equal to 4% by volume, reduce the oxygen concentration to the above limits within 48 hours.

b. With the concentration of oxygen in the WASTE GAS HOLOUP SYSTEM greater than 4% by volume and the hydrogen concentration. greater than 4% by volumei 'i.mmediately suspend all additions of waste gases to the system and reduce the concentration of oxygen to less than or equal to 4% by volume, then take ACTION a., above.

c. The provisions 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 HOLOUP SYSTEM shall be determined to be within the abcve limits by continuously monitoring the waste gases in the WASTE GAS HOLOUP SYSTEM with the hydrogen and oxygen monitors required OPERABLE by Table 3.3-13 of Specification 3.3.3.11. 9 9 9 C0HANCHE PEAK - UNIT 1 3/4 11-15 i

RADI0 ACTIVE CFFtVENTS GAS STORAGE TANKS LIMITING CONDITION FOR OPERATION

             .3.11.2/6 The quan'tity of radioactivity contained in sach gas storage tank        .

shall be limited 'to less than.or equal to 200,000 Curies of noble gases (considered as Xe-133 equivalent). 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, within 48. hours reduce the tank contents to within the limit, and describe the events leading to this condition in the next Semiannual Radioactive Effluent Release Report, pursuant to Specification 6.9.1.7.

b. The' provisions of Specifications 3.0.3 and 3.0.4 are nbt applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.6 The quantity of radioa'ctive material contained in each gas storage tank shal.1 be determined to be within the above limit at least once per 24 hours when radioactive materials are being added to the tank. I l 4 l l 1 COMANCHE PEAK - UNIT 1 3/4 11-16

RADIOACTIVE EFFLUENTS 3/4.11.3 SOLID RADIOACTIVE WASTES LIMITING CONDITION FOR OPERATION . 3.11.3 Radioactive wastes shall be solidified or dewatered in accordance with the PRQCESS CONTROL PR^ GRAM to meet shipping and transportat' ion requirements - during transit, and disposal site requirements when received at the disposal site. APPLICABILITY: At all times. ACTION:

a. With SOLIDIFICATION or dewatering not meeting disposal site and shipping and transportation requirements,. suspend shipment of the inadequately processed wastes and correct the PROCESS CONTROL PROGRAM, the procedures, and/or the Solid Waste System as necessary to prevent recurrence,

b. With SOLIDIFICATION or dewatering not performed in accordance with the PROCESS CONTROL PROGRAM, test the improperly'proces. sed waste in each' container to ensure that it meets burial ground and shipping requirements and take app'ropriate administrative action to prevent recurrence.

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

SURVEILLANCE REQUIREMENTS 4.11.3 SOLIDIFIC'ATION of at least one'r'epresentative test specimen from at least every tenth batch of each type of wet radioactive wastes (e.g., filter sludges, spent resins, evaporator bottoms, boric acid solutions, and sodium sulfate solutions) shall be verified in accordance with the PROCESS CONTROL PROGRAM:

a. If any test specimen fails to verify SOLIDIFICATION, the SOLIDIFICATION' of the batch under test shall be suspended until such time as ?dditional 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 SOLIDIFICATION parameters determined by the PROCESS _ CONTROL PROGRAM;

b. If the initial test specimen from a b' atch 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; and

c. With the installed equipment incapable of meeting Specification 3.11.3 or declared inoperable, restore the equipment to OPERABLE status or provide for contract capability to proces.s wastes as necessary to satisfy all applicable transportation and disposal requirements.

COMANCHE PEAK - UNIT 1 3/4 11-17 D-

RADI0 ACTIVE EFFLUENTS 3/4.11.4 TOTAL DOSE T - 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 a~nd to' radiation from uranium fuel c'cle y sources shall be limited to less than or equal to 25 mrems t6 the whole 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 calt:ulated doses from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Specifi-cation 3.11.1.2a., 3.11.1.2b., 3.11.2.2a., 3.11.2.2b., 3.11.2.3a., or 3.11.2.3b., calculations shall be made including direct radiation contributions from the units (including outside storage tanks etc.) to

        ,             determine whether the above limits of Specification 3.11.4 have been exceeded. If such is the case, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a 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 conformance with the above limits.
                     'This Special Report, as defined in 10 CFR 20.405(c), shall include an analysis that estinates 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 condition result-ing in violation of 40 CFR 190 has not already been corrected, the              ,

i Special Report shall include a request for a variance in accordance ' with the provisions of 40 CFR 190. Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete.

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

I 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, and 4.11.2.3, and in accordance with the methodology and parameters in the ODCM. t l 4.11.4.2 Cumulative dose contributions from direct radiation from the units (including outside storage tanks etc.) shall be determined in accordance with the methodology and parameters in the 00CM. This requirement is applicable only under conditions set forth in ACTION a. of Specification 3.11.4. '. COMANCHE PEAK - UNIT 1 3/4 11-18

i . 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/A.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, prepare and submit to

    .                                  the Commission, in the Annual Radiological Environmental Operating Report required by Specification 6.9.1.3, 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, 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 Specifications 3.11.1.2, 3.11.2.2, or 3.11.2.3. When more than '

one of the radionuclides in Table 3.12-2 are detected in the sampling medium, t.his report shall be submitted. if: concentration (1) con _ centration (2) + *> 1.0 reporting level (1) . reporting level (2) - When radionuclidas other than those in Table 3.12-2 are detected and are the result of plant effluents, this report shall be submitted if the potentisl annual dose

to a MEMBER OF THE PUBLIC from all radio-nuclides is 'wal to or greater than the calendar year limits of Specification 3.11.1.2, 3.11.2.2, or 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 conditior, shall be reported and described in the Annual Radiological Environmental Operating Report required by Specification 6.9.1.3.

                        ' *The methodology and parameters used to estimate the potential annual dose to a MEMBER OF THE PUBLIC shall be indicated in this report.

9 COMANCHE PEAK - UNI 1 3/4 12-1

1

RADIOLOGICAL ENVIRONMENTAL MONITORING,

        ' LIMITING CONDITION FOR OPERATION ACTION (Continued) c.. With milk or fresh leafy vegetation samples unavailable from one or more of the sample locations required by Table 3.12-1, identify specific locations for obtaining replacement samples and add them within 30 days to the Radiological Environmental Monitoring Program given in the 00CM. The specific locations fror.1 which samples were unavailable may then be deleted'from the monitoring program. Pursuant to Specification 6.14, submit in the next Semiannual Radioactive Effluent Release Report documentation for a change in the 00CM                             .

including a revised figure (s) and table for the 00CM reflecting 'he t new location (s) with supporting'information identifying the cause of , the unavailability of samples and justifying the selection of the new location (s) for obtaining samples.

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 specific locations given in the table and figure (s) in the 00CM,-and shall be analyzed pursuant to the requirements of Table 3.12-1 and the detection capabilities requ. ired by Table 4.12-1. 9 C0HANCHE PEAK - UNIT 1 3/4 12-2

TABLE 3.12-1 O

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 3

h m NUMBER OF . REPRESENTATIVE

 ;g EXPOSURE PAlif4AY          SAMPLES AND g                    SAMPLING AND          TYPE AND FREQUENCY g   AND/OR SAMPLE             SAMi'LE LOCATIONS                    COLLECTION FREQUENCY 0F ANALYSIS

[ 1. Direct Radiation I2} Forty routine monitoring Quarterly. Gamma dose quarterly. l $ stations either with two or

more dosimeters or with one . .

H instrument for measuring and ~ recording dose rate continu~ ously, placed as follows: An inner ring of stations, one in each meteorological sector in the ~

general area of the EXCLUSION AREA

 ,                             BOUNDARY; k                             An outer ring of stations, one in y

ry each meteorological sector in w the 6- to 8-km range from the site; and The balance of the stations to be placed in special interest areas such as population centers, nearby residences, schools, and in one or two areas to serve as control stations. 4

TABLE 3.17-1 (Continued) g RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM g NUMBER OF g REPRESENTATIVE EXPOSURE PATHWAY SAMPLES AND SAMPLING AND TYPE AND FREQUENCY

        ,                                           y) g  AND/OR SAMPLE        SAMPLE LOCATIONS                       COLLECTION FREQUENCY          OF ANALYSIS 7 2. Airborne g     Radioiodine and   Samples from five locations'           Continuous sampler oper-   Radioiodine Cannister:

p Particulates ation with sample collec- I-131 analysis weekly. g . tion weekly, or more Three samples (Al-A3) from frequently if required by close to the three EXCLUSION dust loading. Particulate Sampler: AREA BOUNDARY locations, in Gross beta radioactivity different sectors, of the analysis following highest calculated annual filter change- } and average ground-level D/Q; . gamma isotopic analysis of composite (by g One sample from the location) quarterly. y vicinity of a community y having the highest calcu-

1ated annual average ground-level D/Q; and One sample from a control location, as for example 15 to 30 km distant and in the least 3

prevalent wind direction.( )

3. Waterborne

a. Surface Squawk Creek Reservoir (6) Monthly Gamma isotopic analysis U)

monthly. Composite for Lake Granbury Monthly composite of tritium analysis quarterly. weekly grab samples when Lake Granbury is receiving letdown from SCR. Otherwise,

                                                             .          monthly grab sample.(8) g M

N

                                                                                                                                                      ~
                                                                                                             ,TABLE 3.12-1 (Continued) 8                                                                                                                                 .

g RADIOLOGICAL E.""!:;0NMENTAL MO,4ITORING PROGRAM N g NUMBER OF

                                         ,                                             REPRESENTATIVE g  EXPOSURE PATHWAY-                          SAMPLES AND                                     SAMPLING AND                 TYPE AND FREQUENCY x   AND/0R SAMPLE                             SAMPLE LOCATIONS (y)

COLLECTION FREQUENCY OF ANALYSIS [ 3. Waterborne (Continued) 5 w Control-Brazos River Monthly upstream of Lake Granbury

b. Ground Samples from one or two sources Quarterly. Gamma isotopic (5) and only if likely to be affected.I9) .

tritium analysis quarterly,

c. Orinking One sample of each of one to Grab sample at least I-131 analysis on each three of the nearest water once per 2-week period grab sample when the dose R

water supplies that could be when I-131 analysis is cal.culated for the con- ~ affected by its discharge. performed; monthly grab sumption of the water g otherwise. is greater than'I aren J, One sample from a control - locah on. r(10) . Composite for gross beta and gamma isotopic analyses (5) monthly. Composite for tritium analysis quarterly.

d. Sediment One sample from downstream. area Semiannually. Gamma isotopic analysis (5) from with existing or potential semiannually.

Shoreline recreational value. i

   ,.-                                           - - , . , - -                 m                           m          -y,..   - -                           ,

u. -

y TABLE 3'.12-1 (Continued) O RADIOLOGICAL ENVIRONMENTA,L MONITORING DROGRAM _ Y ~ h m NUMBER OF REPRESENTATIVE y EXPOSURE PATHW4Y SAMPLES AND SAMPLING AND TYPE AND FREQUENCY c g) g AND/OR SAMPLE SAMPLE LOCATIONS COLLECTION FREQUENCY OF ANALYSIS [ 4. Ingestion E Gamma' isotopic (5) and

a. Milk Samples from available milking. Semimonthly when ,
animals in three locations . animals t are on pasture; I-131 analysis semi . -

within 5 km distance having the monthly at other times. monthly when animals

highest dose potential. If there are oa pasture; monthly are none, then one sample from at other times.- available milking animals in each

of three areas between S to 8'km - distant. where dosesare calculated ~

 ,                                    to be greater than 1 arem per N

yr.(10) One sample from milking . .

animals at a control location 15 w 3 to 30 km distant and in the least I prevalent wind direction.(3)

b. Fish and One sample of each commercially Sample semiannually . Gamma isotopic analysis (S)

Inverte- and recreationally important on edible portions. brates species in vicinity of plant discharge area. One sample of same species in l areas not influenced by plant . discharge.

c. Food One sample of each principal At time of harvest ( . Gamma isotopic analyses Ib)

Products class of food products from on edible portion. any area that is irrigated. by water in which liquid plant wastes have beci discharged. M '

'I

TABLE 3.12-1 (Continued) , n o gz RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ~ g NUMBER OF

r" REPRESENTATIVE -

 ;g    EXPOSURE PATWAY                                 SAMPLES AND                                              SAMPLING AND y                                                                          gy)                                                                 TYPE AND FREQUENCY                         .

AND/0R SAMPLE SAMPLE LOCATIONS - COLLECTION FREQUENCY OF ANALYSIS h 4. Ingestion (Continued) E

c. Food Samples of three different Monthly during Gamma ~ isotopic (5) cad I-131

Products kinds of broad leaf vegeta-growing season. analysis. .

(Continued) tion grown nearest each of

two different offsite loca-tions of hig wst predicted annual average ground level D/Q if milk sampling it not "

performed.

 $                                                     One sample of each of the                             Monthly during                     Gamma isotopic (5) and I-131 y                                                     similar broad leaf vegeta-                            growing season.                    analysis.

ry tion grown 15 to 30 km dis-N tant in the least prevalent wind directionI ) if milk sampling is not performed. d Wp t

                                                                                                                                                                  . n..

(0

                                                                                                                                                        *                             ^

_ . , _ . _ , _ , , , _ . __ _ ._._.m - -. - ... , _ , _ _ . , , . . __ ,. - _ . . . . , ,

TABLE 3.12-1 (Continuad] TABLE NOTATIONS sN (1) , Specific parameters of distance and direction sector.from the centerline of one r.eactor, and additional description where, pertinent, shall be pro- , vided for each and every sample location in Table 3.12-1 in a table and figure (s) in the ODCM. Refer to NUREG-0133. "Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants," October 1978, and to Radiological Assessment Branch Technical Position, Revision 1, November 1979. Deviations are permitted from the required sampling schedule if specimens are unobtainable due to circumstances such as hazardous conditions, seasonal unavailability, and malfunction of auto-matic sampling equipment. If specimens are unobtainable due to sampling equipment malfunction, effort shall ba made to complete corrective; action , pr 3r to the end of the next sampling period. All deviations frorr the sampling schedule shall be documented in the Annual Radiological Environ-mental Operating Report pursuant to Specification 6.9.1.3. It is recog-nized that; at' times, it may not be possible or practicable to continue to obtain sampi c of the media of choice at the most desired location or time. In these instances suitable alternative media and locations may be chosen for the particular pathway in question and appropriate substitutions made within 30 days in the Radiological Environmental Monitoring Program given in the ODCM. Pursuant to Specification 6.14, submit in the next , Semiannual Radioactive Effluent Release Report documentation for a change in the ODCM including a revised .'igure(s)'and table for'the ODCM reflect-ing the new location (s) with supporting information identifying the cause

          'of the unavailability of samples for the pathway and justifying the selec-tion of the new location (s) for obtaining samples.

l l (2) One or more instruments, such as a pressurized ion chamber, for measuring 1 and recording dose rate continuously may be used in place of, or in addi-tion to, integrating dosimeters. For the purposes of this table, a thermoluminescent dosimeter (TLD) is considered to be one phosphor; two . i or more phosphors in a packet are considered as two or more dosimeters. ! l Film badges shall not be used as dosimeters for measuring direct radittion. , (The'40' stations is not an absolute number. The number of direct rat!ation i i monitoring stations 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. The frenuency of analysis or readout for TLD systems will depend upon the characteristics of the speci-fic system used and should be selected to obtain op'timum dose information within minimal fading.) l (3) The purpose of this sample is to obtain background information. If it is act practical to establish contrpl locations in accordance with the [ distance and wind direction criteria, other sites that provide valid ; background data may be substituted. , (4) Airborne particulate sample filters shall be analyzed for gross beta

radioactivity 24 hours or more after sampling to allow for radon and l thoron daughter decay. If gross beta activity in air particulate samples l is greater than 10 times the yearly mean of control samples, gamma ,

isotopic analysis shall be performed on the individual samples, i COMANCHE PEAK - UNIT 1 3/4 12-8 l

w . TABLE 3.12-1 (Cortinued) qs TABLE NOTATIONS (Continued) I (5) Gamma isotopic analysis means the identification and quantification of

  ,                 gamma-emitting radionuclides that may be      'ributable to the. effluents from the facility.                                             -      *

(6) Squaw Creek Reservoir is a closed cooling water basin which receives plant effluents at the circulating water discharge. The reservoir shall be sampled in an area at or beyond but not near the mixing zone. Als,o the reservoir shall be sampled at a distance beyond significant influence of the discharge. (7) Squaw Creek Reservoir is a closed cooling water basin which is composited naturally. (8) Lake Granbury may receive letdown from Squaw Creek Reservoir to control buildup of solids. This is the only pathway for plant effluents to Lake Granbury. The lake shall be sampled near the letdown dis' charge and at a distance beyond significant influence o'f the discharge. (9) Groundwater samples shall be taken when this source is tapped for drinking

                  'or irrigation purposes in areas whe~re the hydraulic gradient or recharge properties are suitable for contamination. '
      ,     (10) The~ dose shall be calculat'ed for'the maximum organ and age group, using          '

the methodology and parameters in the ODCM.

i (11) If harvest occurs more than once a year, sampling shall be performed during each discrete narvest. If harvest occurs continuously, sampling shall be monthly. Attention shall be paid to including samples of tuberous and root food products. f 1 COMANCHE PEAK - UNIT 1 3/4 12-9

TABLE 3.12-2 , O REPORTING LEVELS FOR RADI0 ACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES E 5 REPORTING LEVELS E a E WATER AIRBORNE PARTICULATE - FISH MILK F000 PRODUCTS '

ANALYSIS (pCi/1) OR GASES (pCi/m3 ) (pci/kg, wet) (pCi/1) (pci/kg, wet)

z . . U H-3 20,000* w Mn-54 1,000 30,000 Fe-59 400 10,000 Co-58 1,000 30,000 y s Co-60 300 10,000 Zn-65 300 20,000 - 5 Z r-Nb-95 400 1-131 2 0.9 - 3 100 Cs-134 30 10 1,000 60 1,000 Cs-137 50 20 2,000 70 2,000

 )                 Ba-La-140                        200                                                                                               300
                    *For drinking water samples. This is 40 CFR Part 141 value.                                            If no drinking water pathway exists, a value s                         of 30,000 pCi/1 may ht used.                                                                                                                                            .-

v

                                                                                                                                                                                  @=n
                                                                                                                                                                                    -1

__m -w-,,mne, ,, .- - - _ - - - - - - - - - - , - , y -- wn .-- - _, ,, ,

TABLE 4.12-1 O DETECTION CAPABILITIES FOR ENVIRO *! MENTAL SAMPLE ANALYSIS (I)~ (2) f LOWER LIMIT OF DETECTION (LLD)( } a g . 2

WATER AIRBORNE PARTICULATE FISH MILK F000 PRODUCTS SEDIMENT ANALYSIS (pCi/1) OR GASES (pCi/m3 ) (pCi/kg, wet) (pCi/1) (pCi/ke, wet) (pCi/kg, dry)

E ~ U Gross Beta 4 0.01 - e> H-3 2000* Mn-54 15 130 Fe-59 30 260 Co-58,60 15 130 R.

   .o g          Zn-65                   30                                                     260 Zr-Mb-95                15 I-131 1**                         0.07                                 1              60                             -

Cs-134 15 0.05 130 15 - 60 150 Cs-137 18 0.06 150 18 80 180 Ba-La-140 15 15 l

               *If no drinking water pathway exists, a value of 3000 pCi/1 may be used.

an , If no drinking water pathway exists, a value of 15 pCi/1 may be used. 1

l. -

1 C:: WA

      ,.--g -         e    v- --a       va- r----- -
                                                          -++r-e                  --         wv,-- y  w  so- -             -+.   -r,-y-            ,y

                                                                                               ., n "

TABLE 4.12-1 (Continued) ' TABLE NOTATIONS (1)This list does not mean that only these nuclides are to be considered. - Other peaks that are identifiable, together with those of the above

                      'nuclides, shall also be analyzed and reported in the Annual Radiological

Environmental Operating Report pursuant to Spedification 6.9.1.3.

(2) Required detection capabilities for thermoluminescent dosimeters used for environmental measurements shall be in accordance with the reconcenda-tions of Regulatory Guide 4.13. " (3)The LLO 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 "real" signal. For a particular measurement system, which may include radiochemical separation: 4.66 s 6 E - V - 2.22 - Y + exp(-AAt) Where: LLD = the "a priori" lower limit of detection (picoCuries per unit mass or volume), s b = the standard deviation of the background counting rate or of the j counting rate of a blank sample as appropriate (counts per. minute), . E = the counting efficiency (counts per disintegration), V = the sample size (units of mass or volume), i 2.22 = the number of disintegrations per minute per picocurie,

Y = the fractional radiochemical yield, when applicable, A = the radioactive decay constant for the particular radionuclide (sec 1), and At = the elapsed time between environmental collection, or end of the sample collection period, and time of counting (sec).

l Typical values of E, V, Y, and at should be used in the calculation. COMANCHE PEAK - UNIT 1 3/4 12-12

TABLE 4.12-1 (Continued) o TABLE NOTATIONS (Continued) It should be recognized that 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. AnaTyses shall be performed in such a manner that the stated LLDs will be achieved under routine conditions. Occasionally background fluctuations, unavoidable small sample sizes, the presence of interfering nuclides, or other uncontrollable circumstances may render these LLDS unachievable. In such cases, the contributing factors shall be identified and described in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.3. 1 0

1r - 4 i 4 . 6 C0KANCHE PEAK - UNIT 1 3/4 12-13 9 e

RADIOLOGICAL ENVIRONMENTAf. MONITORING 3/4.12.2 LAND USE CENSUS LIMITINGCON5ITIONFOROPERATION 3.12.2 A Land Use Ce'nsus'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 animai, the nearest residence, and the nearest garden

of greater than 50 m2 (500 ft 2) producing broad leaf vegetation.

APPLICABILITY: At all times. ACTION: With a Land Use Census identifying a location (s) that yields a a. calculated dose ca dose commitment greater than the values currently being calculated in Specification 4.11.2.3, pursuant to Specifica-tion 6.9.1.4, identify the new location (s) in the next Semiennual Radioactive Effluent Release Report,

b. With a Land Use Census identifying a location (s) that yields a calculated dose or dose commitment (via the same exposure pathway) 20% greater than at a location from which samples are cur.rently being obtained in.accordance with Specification 3.12.1, add the new location (s) within 30 days to the Radiological Environmental Moni- ,

toring Program given in the ODCM. The sampling location (s),.exclud-ing 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. Pursuant to Specification 6.14, submit in the next Semiannual Radioactive Effluent Release Report - dccumentation for a change in the ODCM including a revised figure (s) and table (s) for the ODCM reflecting the new location (s) with informa-tion supporting the change in sampling locations.

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 EXCLUSION AREA BOUNDARY in each of two different direction sectors with the highest predicted 0/Qs in lieu of the garden census.

Specifications for broad leaf vegetation sampling in Table 3.12-1, Part 4.c., shall be followed, including. analysis of control samples. G COMANCHE PEAK - UNIT 1 3/4 12-14

f.

RADIOLOGICAL ENVIRONMENTAL MONITORING =

SURVE}} ' ANCE REQUIREMENTS 4.12.2 The Land Use Census shall be conducted during the growing'sgason 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 consul. ting .                     '

local agriculture authorities. The results of the Land Use Census sh'll a be included in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.3. 4 4 e b e 6 9 i COMANCHE PEAK - UNIT 1 3/4 12-15

     '                                                                   ^               ' '

p s J .

'4

( AADI0 LOGICAL ENVIRONMENTAL MONITORIfG o h' . 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM M LIMITING CONDITION FOR @ ERATION 1. 3;12."3 Analyses shall be~ performed on all radioactive materials,'supp. lied as part' of an Interlaboratory Comparison Program that has been approved by the - Commi.ssion, that correspond to samples required by Table 3.12-1. APPI.ICABILITYi At all times. a e _

                      , t . .AtlId:

s , j i 'a . Witd analyses not being perforr..ed as required above, report the v .' corrective actions taken to prevent a recurrence to the Commission in the Annual Radiological Environmental Operating Report pursua e to Specification 6.9.1.3.

b. Toe prov.isions.of, Specifications 3.0.3 and 3.0.4 are not applicable..

SURVEILLANCE A? M REMENTS s '/

4. il. 3 The Inte.-laboratory Comparison Program shall be described .in the 00CM.

    .,                       A sut.;hary of the results obtained as part of the above required Interlaboratory
                           ,,Compa)isonProgramshall.beincludediintheAnnualRadiologicalEnvironmental

/. . . Oper0H f Report pursuant to, Specification 6.9.1.3.

l. , ; I- h , , i ,, t > l p., l , e' , p l \ l t .

                                                                                                                           /
                                                                                               )I                              '

s

                                                                                                                              \'                                                               l
                                                                                                                                   \

4 COMAl,CHE PEAK - UNIT 1 3/4 12-16

4 . . - . . - .

    -                                                                                                                                                         l 5

t < 1 P git

                                                           .                           BASES'FOR                              -

SECTIONS 3.0 AND 4.0

                                     .a LIMITING CONDITIONS-FOR OPERATION s

j' AND SURVEILLANCE REQUIREMENTS' t L-NOTE l The BASES. contained in succeeding pages summarize the reasons for the Specifications in Sections 3.0 < and 4.0, but-in accordance with 10 CFR 50.36 are not part of these Technical Specifications. . 0 *

     ,yi
 'b           ,
                                             -1 1
                                           +

i s t l l 7 i P l I COMANCHE PEAK . UNIT 1 B 3/4 0-0 i l

+, ;

A%K +

es 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.0 APPLICABILITY ) 1h 2 i I BASES Specification 3.0.1 through 3.0.4 establish.the general requirements applicable

            . to Limiting Conditions for Operation.        These requirements are based on the requirements for Limiting Conditions for Operation stat.ed in ,the Code of Federal Regulations, 10 CFR 50.36(c)(2):
                        "Limiting' conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility.

When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specification until the condition can be met." Specification 3.0.1 establishes the Applicability statement within each indi-vidual specification as the requirement for when (i.e., in which OPERATIONAL HCDES or other specified conditions) conformance to the Limiting Conditions for Operation is required for safe operation of the facility. The ACTION r%1uir.ements establish those remedial measures that must be taken within speci-fied time limi'ts when the requirements of a Limiting Condition for Operation are not met. There are two basic types of ACTION requirements. The first specifies the remedial measures that permit continued operation of the facility which is not further restricted by the time limits of the ACTION requirements.

In this case, conformance to the ACTION. requirements provide an acceptable level of safety for unlimited continued operation as long as the ACTION requirements continue to be met. The second type of ACTION requirement specifies a time limit in which conformance to the conditions of the Limiting Condition for Operation must be met. This time limit is the allowable outage time to restore an inoperable system or component to OPERABLE status or for restoring parameters within specified limits. If these actions are not completed witnin the allow-able outage time limits, a shutdown is required to place the facility in a MODE or condition in which the specification no longer applies. It is not intended that the shutdown ACTION requirements be used as an operational con-venience which permits (routine) voluntary removal of a system (s) or compo-nent(s) from service in lieu of other alternatives that would not result in redundant systems or components being inoperable. The specified time limits of the ACTION requirements are applicable from the . point in time it is identified that a Limiting Condition for Operation is not n.a t . The time limits of the ACTION requirements are also applicable when a system or component is removed from service for surveillance testing or investi-gation of operational problems. Individual specifications may include a speci-fied time limit for the completion of a Surveillance Requirement when equipment is removed from service. In this case, the allowable outage t*me limits of the ACTION requirenents are applicable when this limit expires if the surveillance , has not been completed. When a shutdown is required to comply with ACTION requirements, the plant may have entered a MODE in which a new specification becomes applicable. In this case, the time limits of the ACTION requirements would apply from the point in time that the new specification becomes applicable if the requirements of the Limiting Condition for Operation are not met. COMANCHE PEAK - UNIT 1 8 3/4 0-1 4

APPLICABILITY gog Ub BASES Specification 3.0.2 establishes that noncompliance with a specification exists when.the requirements of the Limiting Condition for Operation are not met and the associated ACTION requirements have not been implemented within the speci-fied time. interval. The purpose of this specification is to clarify that (1) implementation of the ACTION requirements within the specified time. interval constitutes compliance with a specification and (2) completion of the remedial measures of the ACTION requirements is not required when compliance with a Limiting Conditic.m of Operation is restored within the time interval ~specified in the associated ArTION requirements. Specification 3.0.3 establishes the shotdown ACTION requirements that must be implemented when a Limiting Condition for Operation is not met and the condi-tion is not specifically addressed by the associated ACTION requirements. The purpose of this specification is to delinaate the time limits for placing the l unit in a safe shutdown MODE when p.lant operation cannot be maintained within l the limits for safe operation defined by the Limiting Condi.tions for Operation and its ACTION requirements. It is not intended to be used as an operational convenience which permits (routine) voluntary removal of redundant systems or components from service in lieu of other alternatives that would not result in redundant systems or components being inoperable. One hour is allowed to pre-pare for a.n orderly shutdown before. initiating a change in plant operation. This time permits the operator to coordinate the reduction in electrical genera-tion with the' load dispatcher to ensure the stability and avail' ability of the electrical grid. The time limits specified to reach lower MODES of' operation permit the shutdown to proceed in a controlled and orderly manner that is well within the specified maximum cooldown rate and within the cooldown capabilities of the facility assuming only the minimum required equipment is OPERABLE. This reduces thermal stresses on components of the primary coolant system and the potential for a plant. upset that could challenge safety systems under con-ditions for which this specification applies. ! If remedial measures permitting limited continued operation of the facility under the provisions of the ACTION requirements are completed, the shutdown may be terminated. The time limits of the ACTION requirements are applicable from the point in time there was a failure to meet a Limiting Condition for Operation. Therefore, the shutdown may be terminated if the ACTION require-ments have been met or the time limits of the ACTION requirements have not expired, thus providing an allowance for the completion of the required actions. ine time limits of Specification 3.0.3 allow 37 hours for the plart to be in

the COLD SHUTDOWN MODE when a shutdown is required during the POWER MODE of

l operation. If the plant is in a lower MODE of operation when a shutdown is required, the time limit for reaching tne next lower MODE of operation applies.

However, if a lower MODE of operation is reached in less time than allowed, the total allowable time to reach COLD SHUTDOWN, or other applicable MODE, is not reduced. For example, if HOT STANDBY is reached in 2 hours, the time allowed to reach HOT SHUTDOWN is the next 11 hours be'cause the total time to reach' HOT SHUTDOWN is not reduced from the allowable limit of 13 hours. I i COMANCHE PEAY. - UNIT 1 8 3/4 0-2 l i -

APPLICABfLITY

       ' BASES Therefore, if remedial measures are completed that would permit a return to POWE,R operation, a penalty is not incurred by having to reach a lower MODE of operation in less than th'e total time a? lowed.

The same principle applies with regard to the allowable outage time limits of the ACTION requirements, if compliance with the ACT, ION requirements for one specification results in entry into a MODE or condition of operation for another specification in which the requirements of the Limiting Condition for Operation are not met. If the new specification becomes applicable in less time than specified, the difference may he added to the allowable outage time limits of the second specification. However~, the allowable outage time limits of ACTION requirements'for a higher MODE of operation may not be used to extend the allowable outage time that is applicable when a Limiting Condition for Operction is not met in a lower MODE of operation.

   .      The shutdown requirements of Specification 3.0.3 do not apply in MODES 5 and 6, because the ACTION requirements of individual specifications define the remedial measures to be taken.

Specification 3.0.4 establishes limitations on MODE changes when a Limiting Condition for-0peration is not met. It precludes placing the facility in a higher MODE of operation when the requirements for a Limiting Condition for Operation are not met and continued noncompliance ~to these conditions would result in a shutdown to comply with the ACTION requirements if a change in MODES were permitted. The purpose of this ipecification is to ensure that facility operation is not initiated or that higher MODES of operation are not entered when corrective action is being taken to obtain compliance with a speci-fication by restoring equipment to OPERABLE status or parameters to specified limits. Compliance with ACTloN requirements that. permit continued operation of the facility for an unlimited period of time provides an acceptable level of safety for continued operation without regard to the status of the plant I before or after a MODE change. Therefore in this case, entry into an OPERATIONAL MODE or other specified condition may be made in accordance with the. provisions of the ACT70N requirements. The provisions of this specification shoulc. not, however, be interpreted as endorsing the failure to exercise good practice in restoring systems or components to OPERABLE status before plant startup. When a shutdown is required to comply with ACTICN requirements, the provisions of Specification 3.0.4 do not apply because they would delay placing the facil-ity in a lower MODE of operation. Specifications 4.0.1 through 4.0.5 establish the general requirements applicable l to Surveiilance Requirements. These requirements are based on the Surveillance Requirements stated in the Code of Federal Regulations,10 CFR 50.36(c)(3): i i COMANCHE PEAK - UNIT 1 8 3/4 0-3 l l

                                                                                            .w APPLICABILI1f                                                                       ,

BASES "Surveillance requirements are requirements relating to tett, calibration, or inspection to ensure that the necessary quality,of systems ant' components is m'aintained, that facility operation wil.l'b'e within safety limits, snd that' the limiting conditions of operation will be met." - Specification 4.0.1 establishes the requirement that surveillances must be performed during the OPERATIONAL MODES or other conditions for which the re-quirements of the Limiting Conditions for Operation apply unless otherwise i stated in an individual Surveillance Requirement. The puroose of this specifi-cation is to ensure that surveillances are performed to verify the operational status of systems and components and that parameters are within specified limits to ensure safe operation of the facility when the plant is in a MODE or oth6r specified condition for which the associated Limiting Conditions for Operation are applicable. Surveillance Repirements do not have to be performed when the facility is in an OPERATIONAL MODE for which the requirements of the asso-ciated Limiting Condition for Oparation do not apply unless otherwise specified. The Surveillance Require ~ments associat'ed with a Special Test Exception are only applicable when the Special Test Exception is used as an allowable excep-tion to the requirements of a specification.

       . Specification 4.0.2' establishes.the conditions under which the speci.f-ied time interval for Surveillance Requirements may be extended. Item a. permits an allowable extension of the normal surveillance interval to facilitate surveil-lance scheduling and consideration of plant operating conditions that may not ba suitdble for conduct'ing the surveillance; e.g., transient conditions or other ongoing surveillance or maintehance activities.        Item b. limits the use l

of the provisions of item a. to ensure that it is not used repeatedly to extend the surveillance interval beyond that specified. The limits of Specification 4.0.2 are based on engineering judgment and the recognition that the most prob-able result of any particular surveillance being performed is the verificaion of conformance with the Surveillance Requirements. These provisions are suf-ficient to ensure that the reliability ensured through surveillance activities is not significantly degraded beyond that obtained from the specified surveil-lance interval. Specification 4.0.3 establishes the failure to perfo*m a Surveillance Require-ment within the allowed surveillance interval, defined by the provisions of Specification 4.0.2, as a condition that constitutes a failure to meet the OPERABILITY requirements for a Limiting Cor.dition for Operation. Under the - provisions of this specification, systems and components are assumed to be OPERABLE when Surveillance Requircments hwe been satisfactorily perform d ' within the specified time interval. However, nothing in this provision .s to be construed as implying that systems ca co.mponents are OPERABLE whu, they arc found or'known to be inoperable although still meeting the Surveillance Require-ments. .This specification also clarifies that the ACTION requirements are

applicable when Surveillance Requirements have not been colnpleted within the allowed

surveillance interval and that the time limits of the ACTION require-l ments apply from the. point in time it is identifirJ that a surveillance has not been performed and not at the time that the allowed surveillance. interval i

l COMANCHE PEAK - UNIT 1 B 3/4 0-4 i

APPLICABILITY ,np BASES hbb was exceeded. Completion of the Surveillance Requirement within the allowable outage time limits of the ACTION requirements restores compliance with the requirements of Specification 4.0.3., However, this does not negate the fact that the failure to have performeu the surveillance within the allowed surveil-lance interval, defined by the provicions of Specification 4.0.2, was a viola-tiori of the OPERABILITY requirements of a Liiniting Condition for Operation thai as subject to enforcement action. Further, the failure to perform a sur-veiliance within the provisions of Specification 4.0.2 is a violation of a Technical Specification requireaent and is, therefore, a reportable event under the requirements of 10 CFR 50.73(a)(2)(i)(B) because it is a condition pro-hibited by the plant's Technical Specifications. If the allowable outage time limits of the ACTION requirements are less than 24 hot.rs or a shutdown is required to comply with ACTION requirements, e.g. , , Specification 3.0.3, a 24-hcur allowance is provided to permit a delay in implementing the ACTION requirements., This provides an adequate time limit to complete' Surveillance RequireQents that have not been performed. The purpose of this allowarce is to permit the completion of a surveillance before a shutdown is required to comply with ACTION requirements or before other remedial measures would be required that may preclude completion of a surv'eil-lance. The basis for this allowance includes consideration for plant condi-tions, adequate planning, availability of personnel, the time required to perform the surveillance, arid the safety significance of the delay in cocplet-ing'the' required surveillance. This provision also provides a time 1imit for

 . the completion of Surveillance Requirements that become applicable as e consequence of MODE changes imposed by ACTION requirements and for completing Surveillance Requirements that are applicable when an exception to the requirements of Specification 4.0 A is al10wed. If a surveillance is not completed within the 24-hour allowance and de Surveillance Requirements are not met, the time limits of the ACTION requirementi are applicable at the time that the surveillance is terminated.

Surveillance Requirements do not base to be performed on inoperable equipment because the ACTION requirements define the remedial measures that apply. , However, the Surveillance Requirements have to be met to de w strate that ' inoperable equipment has been restored to OPERABLE status. 1 Specification 4.0.4 establishes the requirement that all applicable rueveil-lances must be met bofwe entry into an OPERATIONAL MODE or other condition of operation specified in the Applicability staternent. The purpose of this speci-l fication is to ensure that system and component OPERABILITY requirements or parameter limits are met before entry into a MODE or condition for which these systems and components ensure safe operation of the facility. This provicion applies to changes in OPERATIONAL MODES or other .%ecified conditions associated with plant shutdown as well as startup. l l Under the provisions of this specification, the applicable Surveillance l Requirements must be performed within the speciffed survM 11ance interval to ensure that the Limiting Conditions for Operation are met during initial plant startup or following a plant outGge. COMANCHE PEAK - UNIT 1 B 3/4 0-5

APPLICABILITY nn$hcaLl gKit BASES , When a shut is required to comply with ACTION requirements, tne p*ovisions of Specification 4.0.4 do not apply because this would delay placing '.he facility in a lower MODE of operation.' Specification 4.0.5 establishes the requirement that inservice inspection of A5ME Code Class 1, 2, and 3 components and inservice testing of ASME Code , Class 1, 2, and 3 pumps and valves shall be performed in accordance with a periodically' updated version of Section XI of the ASME Boiler and Pressure Vessel Code and Addenda as required by 10 CFR 50.55a. These requirements apply except when relief has been provided in writing by the Commission. This specification includes a clarification of the frequencies for performing the inservice inspection and testing activities required by Section XI for the ASME Boiler and Pressure Vessel Code and applicable Addenda. This clari-fication is provided to ensure consistency in surveillance intervals through-out the Technical Specifications and to remove any ambiguities relative to the

frequencies for performing the required inservice inspection and~ testing activities. Under the terms of this specification, the more restrictive requirements of the Technical Specifications take precedence over the ASME Boiler and Pressure Vessel Code and applicable Addenda. The requirements. of Specification 4.0.4 to. perform surveillance activities before entry into an OPERATIONAL MODE or other specified cond' tion takes precedence over the ASME Boiler and Pressure Vessel Co'de provision which al?qws pumps and valves to be tested up to one , week after return to normal operation. The Technical Specification definitien of OPERABLE does not allow a grace period before a component, that is not capable of performing its specified function, is declared inoperable and takes precedence over the ASME Boiler and Pressure Vessel Code provision which allows a valve to be incapable of performing its specified / unction for up to 24 hours before being declared inoperable, r i COMANCHE PEAK - UNIT 1 B 3/4 0-6

3/4.1 REACTIVITY CONTROL SYSTEMS e-. BASES 3/4.1.1 80 RATION CONTROL ' 3/4.1.1.1 arid 3/4.1.1. 2. SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that: (1) the reactor can be made

 -subcritical from all operating conditions, (2) the reactivity transients asso-ciated sith postulated accident conditions are controllable within~ accepio'oie limits, and (3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS T,yg. The most restrictive condition occurs at E0L, with T,yg at no load operating temperature, and is i associated with a postulated steam line break accident and resulting uncon trolled RCS cc6 down'. In the analysis of this accident, a minimum SHUT 00WN' MARGIN of 1.6% ak/k is required to control the reactivity transient. Accordingly, the SHUT 00WN MARGIN requirement is based upon this limiting condition and is. consistent with FSAR safcty analysis assumptions. With T,yg les's'than 200 F, the raactivity' transients resulting from a ps'tulated steam

line break cooldown are minimal and a 1% ak/k SHUTDOWN MARGIN piivides adequate protection.

! 374.1.1.3 MODERATOR TEMPERATURE COEFFICIENT Tne limitatiens on moderator temperature coefficient (MTC) are provided to ensure that the vaine of this coefficient remains within the limiting condition assumed in the FSAR accident and transient analyses. l The MTC values of this specification are applicable to a specific set of plant conditions; accurdingly, verification of MTC values at conditions other than those explicitly stated will rer;uire extrap'olation to those conditions in order to permit an accurate comparisor;. l The most negative MTC, value equivalent to the most positive moderator density coefficient (MDC), was obtained by incrementally correcting the MDC used in the FSAR analyses to nominal operating conditions. These corrections l l COMANCHE P'_AK - UNIT .1 B 3// 1-1

REACTIVITY CONTROL SYSTEMS BASES

    @DERATORTEMPERATURECOEFFICIENT(Continued)                                   ,

involved condition, subtracting the incremental of all rods inserted change in (most positive the MDC MDC) to anassociated with'a core-all rods withdrawn condition and, a corvarsion for the rate of change of moderator density with tempirature at RATED THERMAL POWER conditions. This value of the MDC was then transformed into the limiting MTC value -4.0 x 10 8 Ak/k/ F. The MTC value of -3.1 x 10 4 ak/k/ F represents a conservctive value (with corrections for burnup and soluble boron) at a core condition of 300 ppm equilibrium boron concentration and is obtained by making these corrections to the limiting MTC value of -4.0 x 10 4 ok/k/ F. The Surveillance Requirements for measurement of the MTC at the beginning and near the end of the fuel cycle are adequate to confirm that the MTC remains within its limits since this coefficient changes slowly due principally to the-mduction in RCS boron concentration asso:.icted with fuel burnup. 3/4.1.1.4 MINIMUM IEAPERATURti FOR CRITICALITY Thit specification ensures that the reactor will not be mMe critical with the Reactor Coolant System average. temperature less than 551 F. This l' imitation is required to ensure: (1).the moderator temperature co'afficient is within it analy7ed temperature range, (2) the trip instrumentation is within its normal operating range, (3) the P-12 interlock is above its setpoint, (4) the pressurizer is capable of baing in an OPERABLE status with a steam bubble, and (5) the reactor vessel is above its minimum RT temperature. NDT 3/4.1.2 B0 RATION SYSTEMS t The Boron Injection System ensures that negative reactivity control is available during each mode of facii Ry operation. The components required to perform this function include: (1) borated water sources, (2) charging pumps, (3) separate flow paths, (4) boric acid transfer pumps, (5) associated Heat Tracing Systems, and (6) an emergency power' supply from OPERABLE diesel generators. With the RCS average temperature above 200 F, a minimum of two boron injection fbw paths are required to ensure single functional capab'lity in the event an asu ced failure renders one of the flow paths inoperable. The boration capability of either flow path is sufficient to provide a SHUTOOWN MARCIN from expected operating conditions of 1.6% Ak/k after xenon decay and cuoldown to 200 F. The maximum expected boration capability requirement occurs at EOL from full power equilibrium xenon conditions and requires - 22,870 gallons of 70C0 ppm borated water from the boric acid stonge tanks or 479,900 gallons of 2000 ppm borated water from the refueling water storage tank (RWST). COMANCHE PEAK - UNIT 1 B 3/4 1-2

I REACTIVITY CONTROL SYSTEMS . BASES BORATION SYSTEMS (Continued)

        . .Witti the RCS temperature below 200'F, one Boron Injection ' System is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single Boron Injection System becomes inoperable.                -

The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps except the required OPERABLE pump to be inoperable below 275 F provides assurance

   .that a mass addition pressure transient can be relieved by the oparation of a single PORV.

The boron capability required below 200 F is sufficient to provide a

   -SHUTDOWN MARGIN of 1% Ak/k after xenon decay and cooldown from 200 F to 140 F. This condition requires either 6,385 gallons of 7000 ppm borated water from the boric acid storage tanks or 101,120 gallons of 2000 ppm borated water from the RWST.

The contained water volume limits include allowance for water not.available because of discharge lire location and other physical characteristics. lor limits on containe'd s ter voluma and. boron concentration of the RWST also ersure a pH value of between [8.5] an. [11.0] for the solution recirculated within containment e.fter a LOCA. This pH band minimizes the evolution of iodine and ainimizes the effect of chloride and caustic strets corrosion on mechanical systems and components. The OT'ERABILII/ of one Boron Injection System during REFUELING ensures , that this system is avsilable far reactivity control while in MODE 6. 3/4.1.3 HOVABLE CONTROL ASSEMBLIES The specifications Of this section eiisure that: (1) acceptable power distri~ bution limits are maintained, (2) the minimum SHUTOOWN MARGIN is maintained, and ( (3) the potential effects cf rod misalignment on associated accident aaalyses are ! limited. OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits. Verification that the Digital Rod Position Indicator agrees with the demanded position within i 12 steps at 24, 48, 120, and 223 steps withdrawn for the Control Banks and 18, 210, and 228 steps with-drawn for the Shutdown Banks provides assurances that the Digital Rod Position Indicator is operating correctly over the full range of indication. Since the. Digital Rod Position Indication System does r'ot indicate the actual shutdow1 rod position between 18 steps and 210 steps, only points in the indicated ranges - are picked for verification of agreement with demanded position. COMANCHE PEAK - UNIT 1 8 3/4 1-3

REACTIVITY CONTROL SYSTEMS BASES HOVABLE CONTROL ASSEMBLIES (Continued) The ACTION statements which permit limited variations from the basic - requireuents are accompani'ed by additional restrictions which ensure that the or.iginal design criteria are met. Misalignment of a rod requires measurement of peaking factors and a restriction in THERMAL POWER. These restrictions pro-viin assurance of fuel rod integrity during continued operation. 'In addition, those safety analyses affected by a misaligned rod are reevaluated to confirm that the results remain valid during future operation. The maximum rod drop time restriction is consistent with the assumed rod drop time used in the sefety analyses.. Measurement with T,yg greater than or l equal to 551*F and with all reactor coolant pumps operating. ensures that the measured drop times will be representative of insertio.n times experienced

                             'during a Reactor trip at operating conditions.

Control rod positions and OPERABILITY of the rod position indicators are required to be verified on a nominal basis of once per 12 hours with more fre-quent verifications required if an automatic monitoring channel is inoperable. These verification frequ'encies are adequate for assuring that the applicable LCOs.are satisfied. , l i l l l l l l l l l COMANCHE PEAK - UNIT 1 B 3/4 1-4

3/4.2 POWER DISTRIBUTION LIMITS mpW BASES The specifications of this section provide assurance of fuel integrity durincj Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by: (1) maintaining the minimum DNBR in the core greater than or equal to 1.30 during normal operation and in short-term transients, and (2) limiting the fisaion gas release, fuel pellet temperature, and cladding mechanical properties to within assumed design criteria. In addition, limiting the peak l linear power density during Condition I events provides assurance that the initial. conditions as'sumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200 F is not exceeded. The definitions of certain hot channel and peaking factors as used in thette specifications are as follows: Fq (Z) Heat Flux Hot Channel Factor, is defined as the. maximum local heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and. rods; F H Nuclear Enthalpy Rise Hot Channel Fact'or, is defined as the ratio of ' the int.egral of linear power along the rod with the highest integrated power to the average rod power; and

Radial ' Peaking Factor, is defined as the ratio of peak power ~ density F*Y(Z) to average power density in the horizontal plane at core elevation Z. 3/4.2.1 AXIAL FLUX DIFFERENCE The limits on AXIAL FLUX DIFFERENCE (AFD) assure that the Fq (Z) upper bound envelope of 2.32 times the normalized axial peaking factor is not exceeded during either normal operation or in the event of xenon redistribution following power changes. Target flux difference is determined at equilibrium xenon conditions. The full-length rods may be positioned within the core in accordance with their respective insertion limits and should be inserted near their normal position for steady-state operation at high power levels. The value of the target flux difference obtained under these conditions divided by the fraction of RATED THERMAL POWER is the target flux difference at RATED THERMAL POWER for the associated core burnup conditions. Target flux differences for other THERMAL POWER levels are obtained by multiplying the RATED THERMAL POWER value by the appropriate fractional THERMAL POWER level. The periodic updating of the target flux difference value is necessary to reflect core burnup considerations. 4 COMANCHE PEAK - UNIT 1 8 3/4 2-1

POWER DISTRIBUTION LIMITS dA BASES AXIAL FLUX DIFFERENCE '(Continued) Although it is intended that ,the plant will be operated with the AFD within the target band required by Specification 3.2.'1 about the target flux difference, during rapid plant THERMAL POWER reductions, control rod motion will cause the AFD to deviate outside of the target band at reduced THERMAL POWER levels. This deviation will not affect the xenon redistribution suffi-ciently to change the envelope of peaking factors which may be reached on a subsequent return to RATED THERMAL POWER (with the AFD within the target band) provided the time duration of the deviation is limited. Accordingly, a 1-hour penalty deviation limit cumulative during the previous 24 hours is provided for operation outside of the target band but within the limits of Figure 3.2-1 while at THERMAL POWER levels between 50% and 90% of RATED THERMAL POWER. For THERMAL POWER levels between 15% and 50% of RATED THERMAL POWER, deviations of the AFD outside of the target band are less significant. The penalty of 2 hours actual time reflects this reduced significance. Provisions f'or monitoring the AFD on an automatic basis are derived from

the plant process computer through the AFD Monitor Alarm. The computer deter-mines the 1-minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for two or more OPERABLE excore channels are outside the target band and the THERMAL POWER is greater than 90% of RATED THERMAL POWER._ During operation at THERMAL POWER levels between 50% and 90% and between 15% and 50% RATED THERMAL POWER, the computer outputs an alarm message when the penalty deviation accumulates beyond the' limits of 1 hour and 2-hours, respectively. Figure B 3/4 2-1 shows a typical monthly target band. 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR, and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR l The limits on heat flux hot channel factor, RCS flow rate, and nuclear l enthalpy rise hot channel factor ensure that: (1) th( design limits on peak local power density and minimum DNBR are not exceeded and (2) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200 F ECCS acceptance criteria limit. Each of these is measurable but will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3. This periodic surveillance is sufficient to ensure that the limits are maintained provided:

a. Control rods in a single group move together with no individual rod insertion differing by more than i 12 steps, indicated, from the group demand position;

b. Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.6; COMANCHE PEAK - UNIT 1 B 3/4 2-2

                                                                        %n r-A:      '

l i FIGURE B 3/4 2-1 TYPICAL INDICATED AXIAL FLUX DIFFERENCE VERSUS 1HERNAL POWER l COMANCHE PEAX ' UNIT 1 8 3/4 2-3 L

P POWER DISTRIBUTION LIMITS () BASES HEAT FLUX HOT CHANNEL FACTOR, and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued)

c. The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained; and d; The axial power die,cribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.

F H will be maintained within its limits provided Conditions a. through

d. above are maintained. As noted on Figure 3.2-3, RCS flow rate and F H may be "traded off" against one another (i.e., a low measured RCS flow rate is acceptable if the measured F H is also low) to ensure that the calculated DNBR will not be below the' design'DNBR value. The relaxation of F g as a function of THERMAL POWER allows changes in the radial power shape for all permissib1'e rod i.nsertion limits.

R es calculated in Specification 3.2.3 and used in Figure 3.7-3, accounts for.F H less than or equal to 1.49. This value i~s used in the various accident analyses where F H influences parameters other than DNBR, e.g., peak clad temperature, and thus is the maximum "as measured" value allowed. Fuel rod bowing reduces the value of DNB ratio. Credit is available to offset this reduction in the generic margin. The generic margins, totaling 9.1% DNBR completely offset any rod bow penalties. This margin includes the following:

a. Design limit ONBR of 1.30 vs 1.28,

b. Grid Spacing (K,) of 0.046 vs 0.059,

c. Thermal Diffusion Coefficient of 0.038 vs 0.051,

d. DNBR Multiplier of 0.86 vs 0.88, and

e. Pitch. reduction.

The applicable values of rod bow penalties are referenced in the FSAR. COMANCHE PEAK - UNIT 1 B 3/4 2-4

.g

POWER DISTRIBUTION LIMITS gQ{ ljnnth BASES HEAT FLUX HOT CHANNEL FACTOR, and RCS FLOW RATE AND NUCLEAR ENTHALPY' RISE HOT CHANNEL FACTOR (Continued) , When an qF measurement is taken, an allowance for both experimental error and manufacturing tolerance must be made. An allowance of 5% is appropriate for a full-core map taken with the Incare Detector Flux Mapping System, and a 3% allowance is appropriate for manufacturing tolerance. The Radial Peaking Factor, fxy(Z), is measured periodio ',y to provide assurance that the Hot Channel Factor, F (Z), remains within its limit. The 9 F limit for RATED THERMAL POWER (F RTP) as provided in the Radial Peaking xy Factor Limit Report per Specification 6.9.1.6 was determined from expected power control manuevers over the full range of burnup conditions in the core. WhenRCSflowrateandFharemeasured,noadditionalallowancesare necessary prior to comparison with the limits of Figures 3.2-3 and 3.2-4. Measurement errors of [2.1]% for RCS total flow rate and 4% for F have been H allowed for in determination of the design DNBR value. The measurement error for RCS total flow rate is based upon performing a

 ' precision heat balance and using'the result to calibrate the RCS flow rate indicators. Potential fouling of the feedwater venturi which might not be
 . detected could bias the result from the precision heat balance in a non-conservative manner.' Therefore, a penalty of [0.1]% for undetected fouling of the feedwater venturi is included in Figure 3.2-3.        Any fouling which might bias the RCS flow rate measurement greater than [0.1]% can be detected by monitoring and trending various plant performance parameters. If detected, action shall be taken before performing subsequent precision heat balance measurements,    i.e., either the effect of the fouling shall be quantified and compensated for in the RCS flow rate measurement or the venturi shall be cleaned to eliminate the fouling.

The 12-hour periodic surveillance of indicated RCS flow is sufficient to detect only flow degradation which could lead to operation outside the accept-able region of operation shown on Figure 3.2-3. 3/4.2.4 QUADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that the radial power distribu-tion satisfies the' design values used in the power capability analysis. Radial power distribution measurements are made during STARTUP testing and periodically during power operation. The limit of 1.02', at which corrective action is required, provides DNB

     .and linear heat generation rate protection with x y plane power tilts.. A limit of 1.02 was selected to provide an allowance for the uncertainty l      associated with the indicated power tilt.

COMANCHE PEAK - UNIT 1 B 3/4 2-5 I

POWER DISTRIBUTION LIMITS p -s r ' fk BASES _

  ,   QUADRANT POWER TILT RATIO (Continued)

The.2-hour time a'llowance for operation with a tilt condition' greater than 1.02 is prov'ided to allow ideatification and correction of a dropped or misaligned control rod. In the event such action action does not correct the tilt, the margin for uncertainty on F q is reinstated by reducing the maximum allowed power by 3% for each percent of tilt in excess of 1. For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the moveable incore detectors are used to confirm that

     .the normalized symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO. The incore detector monitoring is done with a full incore flux map or two sets of four symmetric thimbles.

3/4.2.5 DNB PARAMETERS The limits on the DNB-related parameters assure that each of the param-eters are maintained within the normal steady-state envelope of operation assumed in the transient and accident analyses. The limits are consistent with.the initial FSAR assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR of 1.30 throughout each analyzed transient. The indicated T,yg value of 592.7 F (conservatively rounded to 592 F) and the indicated pressurizer pressure value of 2207 psig correspond to analytical limits of 594.7 F and 2193 psig respectively, with' allowance for measurement uncertainty. The indicated uncertainties assume that the. reading from four channels will be averaged before comparing'with the required limit. The 12-hour periodic surveillance of these parameters through iristrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation. 1 i I i 1 I COMANCHE PEAK - UNIT 1 8 3/4 2-6 I

r t . p" J 3/4.3 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION The OPERABILITY of the Reactor' Trip System and the Engineered Safety Features Actuation System instrumentation and interlocks ensures thati (1) the . associated ACTION and/or Reactor trip will be initiated when the parameter monitored by each channel or combination thereof reaches its Setpoint (2) the specified coincidence logic is maintained, (3) sufficient redundancy is main-tained to permit a channel to be out-of-service for testing or maintenance, and (4) sufficient system functional capability is available from diverse parameters. The OPERABILITY of these systems is required to provide the overall reliability, redundancy, and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions. The integrated optiration of each of these systems is consistent with the assumptio.ns use_d in the. safety analyses. The Surveillance Requi.rements speci-fied'for these systems ensure that the overall system functional capability is maintained comparable to the original design standards. The periodic surveil-lance tests performed at the minimum frequencies are sufficient to demonstrate this capability. The Engineered Safety Features Actuation System Instrumentation Trip Setpoints specified in Table 3.3-4 are the nomina ~i values at which the bistables are set for.each functional unit. A Sotpoint is considered to be adjusted consistent with the nominal value when the "as measured" Setp.oint is within the band allowed for calibration accuracy. To accommodate the instrument drift assumed to occur between operational tests and the accuracy to which Setpoints can be measured and calibrated,

Allowable Values for the Setpoints have been specified in Table 3.3-4. Opera-l tion with Setpoints less conservative than the Trip Setpoint but within the l Allowable Value is acceptable since an allowance has been made in the safety l analysis to accommodate this error. An optional provision has been included for determining the OPERABILITY of a channel when its Trip Setpoint is found to exceed the Allowable Value. The methodology of this option utilizes the "as measured" deviation from the specified calibration point for rack and sensor components in conjunction with a statistical combination of the other l uncertainties of the instrumentation to measure the process variable and the l uncertainties in calibrating the instrumentation. In Equation 3.3-1, Z + R + 5 < TA, the interactive effects of the errors in the rack and the

sensor, and the "as measured" values of the errors are considered. Z, as specified in Table 3.3-4, in percent span, is the statistical summation . of errors assumed in the analysis excluding those associated with the sensor and rack drift and the accuracy of their measurement. TA or Total Allowance is the difference, in percent span, R or Rack Error is the "as measured" deviation, in the percent span, for the affected channel from the specified Trip Setpoint. S or Sensor Error is either the "as measured" deviation of COMANCHE PEAK - UNIT 1 B 3/4 3-1

~ _ .. . _ . ,.. .m . INSTRUMENTATION BASES

REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued)

           -the sensor from its calibration point or the value specified in Table 3.3-4, in percent span, from the analysis assumptions.                   Use of Equation 3.3-1 allows for a sensor draft factor, an increased rack drift factor, and provides a threshold value for REPORTABLE EVENTS.

The methodology to derive the Trip Setpoints is based upon combining all of the uncertainties in the channels. Inherent to the determination of the Trip Setpoints are the magnitudes of these channel uncertainties. Sensor and rack instrumentation utilized in these channels are expected to be capable of operating within the allowances of these uncertainty magnitudes. Rack drift in excess of the Allowable Value exhibits the behavior that the rack has not met its allowance. Being that there is a small statistical chance that this will happen, an infrequent excessive drift is expected. Rack or sensor drift, in excess.of the allowance that is more than occasional, may b.e indicative of more serious problems and should warrant further investigati'on.' The measurement of response time at the specified frequencies provides assurance that the Reactor trip and the Engineered Safety Features actuation associated with each channel.is completed within the time limit assumed in the safety analyses. No credit was taken.in the analyses for those channels with response times ir.dicated as not applicable. Response time may be demonstrated by any series of sequential, overlapping, or total channel test igeasurements provided that such tests demonstrate.the total channel response time as defined. Sensor response time verification'may be demonstrated by either: (1) in place, onsite, or offsite test measurements,'or (2) utilizing replacement sensors with certified response time. The Engineered Safety Features Actuation System senses selected plant parameters and determines whether or not predetermined limits are being exceeded. If they are, the signals are combined into logic matrices sensitive to combinations indicative of various accidents events, and transients. Once the required logic combination is completed, the system sends actuation signals to those Engineered Safety Features components whose aggregate function best serves the requirements of the condition. As an example, the following actions may be initiated by the Engineered Safety Features Actuation System to mitigate the consequences of a steam line break or loss-of-coolant accidenc: (1) ECCS pumps start and automatic valves position, (2) Reactor trip, (3) feed water isolation, (4) startup of the emergency diesel generators, (5) containment spray pumps start and automatic valves position (6) containment isolation, (7) steam line isolation, (8) turbine trip, (9) auxiliary feedwater pumps start and automatic valves position, (10) service water pumps start and automatic valves position, (11) Control Room Emergency Recirculation starts, and (12) essential ventilation systems (safety chilled water, electrical area fans, primary plant ventilation ESF exhaust fans, battery room exhaust fans, and UPS ventilation) start. - COMANCHE PEAK - UNIT 1 B 3/4 3-2

INSTRUMENTATION BASES REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued) To satisfy the recommendations set forth in Section 4.7 of IEEE 279-1971, in the e' vent that one of the three channels of high steam generator level protection is used for level control that channel shall be placed in the tripped condition until level control is returned to its normal channel.

                    .The Engineered Safety F.eatures Actuation System interlocks perform the following functions:

P-4 Reactor tripped - Actuates Turbine trip, closes main feedwater v'alves on T,yg below Setpoint, prevents the opening of the main feedwater valves which were closed by a Safety Injection or High Steam Generator Water Level signal, allows Safety Injection block so that components can be reset or tripped. Reactor not tripped prevents manual block of Safety Injection. P-11 On increasing pressurizer pressure,.P-11 automatically reinstates Safety Injection actuation on low pressurizer pressure and low steam line pressure. 'On decreasing pressure, P-11 allows the manual block. of Safety Injection actuation on low pressurizer pressure and low steam line pressure. P-12 On increasing reactor coolant loop temperature, P-12 automatically reinstates Safety Injection actuation on high steam flow coincident with either low-low T avg or low steam line pressure, and provides an arming signal to the Steam Dump System. On decreasing reactor - coolant loop temperature, P-12 allows the manual block of Safety Injection actuation on high steam flow coincident with either low-low T,yg or low steam line pressure and automatically removes the arming signal from the Steam Dump System. P-14 On increasing steam generator water level, P-14 automatically trips all feedwater isolation valves and inhibits feedwater control valve

modulation. ,

3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 RADIAT. ION MONITORING FOR PLA$T OPERATIONS The OPERABILITY of the radiation monitoring instrumentation for plant operations ensures that: (1) the associated action will be initiated when the radiation level monitored by each channel reaches its Setpoint, and (2) suffi-cient redundancy is maintained to permit a channel to be out-of-service for . testing or maintenance. The radiation monitors for plant operations senses COMANCHE PEAK - UNIT 1 8 3/4 3-3

                                                                                          ,n.        ,e ,

INSTRUMENTATION BASES 3/4.3.3.1 RADIATION MONITORING FOR PLANT OPERATIONS (Continued) radiation levels in selected plant. systems and locations and det. ermines whether or not predetermined limits are being exceeded. If they are,.the system sends

actuation signals to initiate alarms or actuate Control Room Emergency Recir-culation or actuate Containment Ventilation Isolation. 3/4.3.3.2 MOVABLE INCORE DETECTORS The OPERABILITY of the movable incore detectors with the specified minimum complement of equipment ensures that the measurements obtained from use of this system accurately represent the spatial neutron flux distribution of the core. The OPERABILITY of this system is demonstrated by irradiating each detector used and determining the acceptability of its voltage curve. For the purpose of measuring Fq (Z) or F H a full incore flux map is used. Quarter-core flux maps, as defined in WCAP-8648, June 1976', may be used in recalibration of the Excore Neutron Flux Detection System, and full incore flux maps or symmetric incore thimbles may be used for monitoring the QUADRANT POWER TILT RATIO when one Powar Range channel is inoperable. 3/4.3.3.3 SEISMIC INSTRUMENTATION The OPERABILITY of the seigmic instrumentation ensures that sufficient' caoability is available to promptly determine the magnitude of a seismic event and evaluate the response of those features important to safety. This capa-bility is required to permit comparison of the measured response to that used in the design basis for the facility to determine if plant shutdown is required pursuant to Appendix A of 10 CFR 100. The instrumentation is consistent with the recommendations of Regulatory Guide 1.12, "Instrumentation for Earthquakes," l April 1974. 3/4.3.3.4 METEOROLOGICAL INSTRUMENTATION l The OPERABILITY of the meteorological instrumentation ensures that suffi-l cient meteorological data are available for estimating potential radiation ( doses to the public as a result of routine or accidental release of radioactive l materials to the atmosphere. This capability is required to evaluate the need for initiating protective measures to protect the health and safety of the public and is consistent with che recommendations of proposed Revision 1 to Regulatory Guide 1.23, "Meteorological Programs in support of Nuclear Power l Plants," September 1980. - 3/4.3.3.5 REMOTE SHUTOOWN SYSTEM The OPERABILITY of the Remote Shutdown System ensures that sufficient capability is available to permit safe shutdown of the facility from locations outside of the control room. This capability is required in the event control room habitability is lost and is consistent with General Design Criterion 19 of 10 CFR 50. COMANCHE PEAK - UNIT 1 B 3/4 3-4

1m i 5 INSTRUMENTATION lNI a BASES

        , REMOTE SHUTDOWN SYSTEM (Continued)

The OPERABILITY of the Remote Shutdown Syst,em ensures that a fire will not"preclude achieving safe shutdown. The Remote Shutdown System instrumenta-

   . tion, control, and power circuits and transfer switches necessary to eliminate effects of the fire and allow operation of instrumentation, control and power circuits required to achieve and maintain a safe shutdown condition are inde-pendent of areas where a fire could damage systems normally used to shut down t.he reactor. This capability is consistent with General Design Criterion 3 and Appendix R to 10 CFR Part 50.

3/4.3.3.6 ACCIDENT MONITORING INSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters for which pie planned manually controlled operator actions are required to accomplish safety fonctions for recovery from Design Basis Accidents, as defined by the plant safety analysis. This capability is consistent with the recommendations of Regulatory Guide 1.97, Revision 2, "Instrumentation for Light-Water-Cooled Nuclear Poscer Plants to Assess Plant Conditions During and Following an Accident," December 1980 and NUREG-0737, "Clarification of TMI Action Plan Requirements," November 1980. - The provision that allows the. number of Steam Generator Water Level-Wide Range or Auxiliary Eeedwater Flow Rate Channels to. be reduced by combining . them into a Secondary Coolant Availability function is consistent with Action Plan requirement II.E.~1.2 of NUREG-0737 for Westinghouse Pressurized Water Reactors. The specific calibration provisions for the Containment Radiation (High Range) Monitor are in accordance with the provisions of NUREG-0737, Item II.F.1. 3/4.3.3.7 CHLORINE DETECTION SYSTEMS The OPERABILITY of the Chlorine Detection Systems ensures that sufficient capability is available to promptly detect and initiate protective action in

the event of an accidental chlorine release. This capability -is required to
protect control room personnel and is consistent with the recommendations of l

Regulatory Guide 1.95, Revision 1, "Protection of Nuclear Power Plant Control room Operators Against an Accidental Chlorine Release," January 1977. 3/4.3.3.8 LOOSE PART DETECTION SYSTEM The OPERA.BILITY of the Loose-Part Detection System ensures that sufficient i capability is available to detect loose metallic parts in the Reactor System and avoid or mitigate damage to Reactor System components. The allowable , out-of-service times and surveillance requirements are consistent with the ! recommendations of Regulatory Guide 1.133, "Loose-Part Detection Program for the Primary System of Light-Water-Cooled Reactors," May 1981. { i COMANCHE PEAK - UNIT 1 B 3/4 3-5

n r '- , INSTRUMENTATION Til BASES 3/4.3.3.9 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases 'of radioactive niaterials in' liquid effluents during actual or potent'ial releases of liquid effluents. The Alarm / , Trip Setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR 20. The OPERA-BILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of 10 CFR 50 Appendix A. 3/4.3.3.10 RADI0 ACTIVE GASE0US 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 msthodology and pa'rameters in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR 20. This instru-mentation also includes provisions for monitoring (and controlling) the con-centrations.of potentially explosive gas mixtures in the WASTE GAS HOLDUP SYSTEM. The OPERABILITY and use of this instrumentation is consistent with the requirements of Gen 6ral Design Criteria 60, 63, and 64 of 10 CFR 50 Appendix A. The sensitivity of any nobig gas activity monitors used to show compliance with

the gaseous effluent release requirements of Specification 3.11.2.2 shall be such that concentrations as low as 1 x 10 6 pCi/ml are measurable. 3/4.3.4 TURBINE OVERSPEED PROTECTION This specification is provided to ensure that the turbine overspeed protection instrumentation and the turMne speed control valves are OPERABLE and will protect the turbine from excessive overspeed. Protection from turbine excessive overspeed is required since excessive overspeed of the turt'ine could generate potentially damaging missiles which could impact and damage safety-related components, equipment or structures. 4 1 l COMANCHE PEAK - UNIT 1 B 3/4 3-6

                  .         +

6 3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate wi.th all reactor coolant loops in operation and maintain DNBR above 1.'30 during all normal operations and anticipated transients. In MODES 1 and 2 with one reactor coolant loop not in operation this specification requires that the plant be in at least HOT STANDBY within 6 hours. . In MODE 3, two reactor coolant loops provide sufficient heat removal capability for removing core decay heat even in the event of a bank withdrawal accident; nowever, a single reactor coolant loop provides sufficient heat removal capacity if a bank withdrawal accident can be prevented, i.e. , by opening the Reactor Trip System breakers. Single failure considerations require that two loops be OPERABLE at all times. , In MODES 3, 4, and 5, the operability of the required steam generators is based on maintaining a sufficient level to. guarantee tube coverage to assure heat transfer capability. In MODE 4, and in MODE 5 with reactor coolant loops filled, a si'ngle reactor coolant loop or RHR loop provides sufficient heat removal capability for remov.ing decay heat; but single failure considerations require that at' least two loops (either RHR or RCS) be OPERABLE. In MODE 5'with reactoi coolant loops not filled, a single RHR loop provides' sufficient heat removal capability for removing decay heat; but single failure considerations, and the unavailability of the steam generators as a. heat removing compor.ent, require that at least two RHR loops be OPERABLE. The operation of one reactor coolant pump (RCP) or one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System. The reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognitiois and control. The restrictions on starting an RCP with one or more RCS cold legs less than or equal to 350 F are provided to prevent RCS pressure transients, caused by energy additions from the Secondary Coolant System, which could exceed the limits of 10 CFR 50 Appendix G. The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 50 above each of the RCS cold leg temperatures. 3/4.4.2 SAFETY VALVES The pressurizer Code safety valves operate to' prevent the RCS from being pressurized above its Safety Limit of 2735 psig. (ach safety valve is designed to relieve 420,000 lbs per hour of saturated staam at the valve Setpoint. The relief capacity of a single safety valve is adequate to relieve any overpressure COMANCHE PEAK - UNIT 1 8 3/4 4-1

REACTOR COOLANT SYSTEM f E!bi,b jitrtin SASES _ REACTOR COOLANT LOOP 5 AND COOLANT CIRCULATION (Continued) condition which could occur during shutdown. In the event that no safety , valves are OPERABLE, an operating RHR loop, connected to the RCS, provides overpressure relief capability and will prevent RCS overpressurization. In } addition, the Overpressure Protection System provides a diverse means of protection against RCS overpressurization at low temperatures. During operation, all pressurizer Code safety valves must be OPERABLE to prevent the RCS from being pressurized above its Safety Limit of 2735 psig. The combined relief capacity of all of these valves is greater than the maximum surge rate resulting from a complete loss-of-load assuming no Reactor trip until the'first Reactor Trip System Trip Setpoint is reached (i.e., no cr:dit is taken for a direct Reactor trip on the loss-of-load) and also assuming no operation of the power, operated relief valves or steam dump valves. Demonstration of the safety valves' lift settings wili occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Code. 3/4.4.3 PRESSURIZER The limit on the maximum water volume in the pressurizer assures that the '

           . parameter is maintained within the normal steady-state envelope of operation assumed in the SAR. The' limit is consistent with the initial SAR assumptions.

The 12-hour periodic surveiliance is sufficient to ensure that t: . parameter is restored to within its limit following expected transient operation. The maximum water volume also ensures that a steam bubble is formed and thus the RCS is not a hydraulically solid system. The requirement that a minimum number of pressurizer heaters be OPERABLE enhances the capability of the plant to control Reactor Coolant System pressure and establish natural circulation. Pressurizer heater groups are powered from sources that meet the requirements of Item II.E.3.1 of NUREG-0737. e e COMANCHE PEAK - UNIT 1 8 3/4 4-2

REACTOR COOLANT SYSTEh ^"3D['T LTsiilf! ' 3 BASES q -- 3/4.4.4 RELIEF VALVES The power-operated relief valves (PORVs) and steam bubble function to relieve RCS pressure during all design transients up to and including the } design step load decrease with steam dump. Operation of the PORVs minimizes 1 the undesirable opening of the spring-loaded pressurizer Code safety valves. J Each PORV has a remotely operated block valve to provide a positive shutoff capability should a relief valve become inoperable. 3/4.4.5 STEAM GENE,RATORS The Surveillance Requirements for inspection of the steam generator tubes a ensure that the structural integrity of this portion of the RC$ will be main-J tained. The program for inservice inspection of steam generator tubes is based K on a modification of Regulatory Guide 1.83, Revisinn 1, inseryice inspection of steam generator tubing is' essential in order to ma. tain surveillance of the conditions of the tubes in the cvent that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions'that lead to corrosion. Inservice inspection of steam generator tubing also provides a means of characterizing'the nature and cause of any tube degradation so that corrective measures can be taken. Selected' tubes in the preheater section of each D4 and D5 steam generator have been modified to correct the tube vibration degradation phenomenon experi-enced by certain Westinghouse steam generators. The modification consisted of expanding these tubes in the vicinity of the support plates and is designed to limit the amplitude of vibrraion. These expanded tubes are subject to a special inspectior whenever the steam generators are opened for inservice eddy current testine, The plant is expected to be operated in a marner such that the secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam ger.arator tubes. If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking. The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the Reactor Coolant System and the Secondary Coolant System (reactor-to secondary leakage = 500 gallons per day per steam generator and a total leakage of 1 GPM to all steam generators). Cracks having a reactor-to-secondary leakage less than this limit'during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents. Operating plants have demonstrated that reactor-to-secondary leakage of 500 gallons per day per steam generator can readily be detected by radiation monitors of steam generator blowdown. Lcakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking' tubes will be located and plugged. C0KANCaE PEAK - UNIT 1 B 3/4 4-3 l

REACTOR COOLANT SYSTEM BASES STEAM GENERATORS (Continued) Wastage-type defe:ts are u'nli.kely with proper chemistry treatment of the secondary coolant. However, eve'n if a defect should devalop in service, it will be found during scheduled inservice steam generator tube examinations. Plugging will be required for all tubes with imperfections exceeding the plugging limit of 40% of t M tube nominal wall thickness. Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect deg'radation that has perietrated 20% of the original tube wall thickness. Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be reported to the Commission in a , Special Report pursuant to 10 CFR 50.72 within 4 hours from initial discove y and pursuant to Specification 6.9.2 within 30 days and prior to resumption c, plant operation:. .Such cases will be censide"ed by the Commission on a case-by-case' basis and may result in a requirement for analysis, laboratory examina-t,i ons , tests, additional eddy-current inspection, and revision of the Techn1:al Specifications, if necessary. , 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4. 4. 6.'1 LEAKAGE DETECTION SYSTEMS / . I The RCS Leakage Detection Systems required by this specification are provided to monitor and detect leakage from the reactor coolant pressure boundary. These Detection Systems are consistent with the recommendations of Regulatory Guide 1.45, "Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973. 3/4.4.6.2 OPERATIONAL LEAKAGE PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an irepending gross failure of the pressure boundary. Therefore, the presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to be promptly placed in COLD SHUTDOWN. Industry experience has shown that while a limited amount of leakage is exlected from the RCS, the untaendised ;:ertigo of this leakage can be reduced to a threshold value of less than 1 gpm. This thresh Id value is sufficiently low to ensure early detection of additional leakage. The tota'l steam generator' tube leakage limit of 1 gpm for all steam gen-erators not isolated from the RCS ensu,as that the dosage contribution from the

 ,     tube leakage will be limited to a small fraction of 10 CFR Part 100 dosq guide-line values in the event. of either a steam generator tube rupture or stea.n line break. The } gpm limit is consistent with the assumptions used in the analysis of these accidents. The 500 gpd leakage limit per steam generator ensures that steam generator tube integrity is maintained in the event of a main steam line rupture or under LOCA conditions.

COMANCHE PEAK - UNIT 1 B 3/4 4-4

  ~       ~

REACTOR COOLANT SYSTEM ' BASES OPERATIONAL LEAKAGE (Continued) The 10 gpm IDENTIFIED LEAKAGE limitation provides allowance'for a limited . amount of leakage from known sources whosc presence will not interfere with

,             the detection of UNIDENTIFIED LEAKAGE by the Leakago Detection Systems.

ine CONTROLLED LEAKAGE liraitation restricts operation when tne total flow supplied to the reactor coolant pump seals exceeds 40 gpm with the modulating valve in the supply line fully open at a nominal RCS pressure of 2235 psig. This limitation ensures that in the ovent of a LOCA, the safety injection flow will not be less than assumed in the safety analyses. The leakage from any RCS pressure isolation valve is sufficiently low to ~ ensure early detection of possible in-series valve failure. It is apparent that when pressure isolation is provided by two in-series valves and when failure of. one valve in the pair can go utidetected for' a substantial length

            'of time, verification of valve integrity is required. Since these valves are importa'nt in preventing overpressurization and rupture of the ECCS low pressure                 >

piping which could result in a LOCA, these valves should be tested periodically to ensure low probability of gross failure. The Surveillance Requirements for RCS pressure isolution valves provide added assurance of valve integri.ty thereby reducing the probability of gross valve failure and consequent intersys' tem'LOCA. . Leakage from the RCS pressure isolation valve is IDENTIFIED LEAKAGE ai.d will be considered as a portion of the allowed limit. 3/4.4.7 CHEMISTRY The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactnr Coolant System is minimized and reduces the potential for Reactor Coolant System leakage or failure due to stress corrosion. Mditaaining the chemistry within the Steady-State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant. The associated effects of exceeding the oxygen, chloride, and fluoride limits are time and temperature dependent. Corrosion studies show that operation may be continued with contaminant concentration levels i cess o' the Steady-State Limits, up to the Transient Limits, for the .pe,2 ed limi " time intervals without having a significant effect on the t . sal in: N . of the Reactor Coolant System. The time interval 9ers' c' % . aeation within the restrictions of the Transient Limits ptc4 h e corrective actions to restere the contamin6nt concen-tre a. >

                                ..         ;eady-State Limits.

i n w s:C hequirements provide adequate assurance that concentrations

             % excess of tv             's will be detected in suff icient time to take corrective

' " ;on. i COMANCHE PEAK - UNIT 1 B 3/4 4-5 L . -J

                                                                                            'l REACTOR COOLANT SYS7EM                                                      mn IN

BASES 3/4.4.8 SPECIFIC ACTIVITY ,

The limitations on the specific activity of the reactor coolant ensure that the resulting 2-hour doses at the EXCLUSION AREA BOUNDARY (EAB) will not exceed en appropriately small fraction of 10 CFR Part 100 dose guideline values following a steam generator tube rupture accident in conjunction with an assumed steady- state reactor-to-secondary steam generator leakage 76te of 1 gpm. The values for the 1.mits on specific activity represent limits based upon a parametric evaluation by the NRC of typical site locations. These values ere conservative in that specific site parameters of the CPSES site, such as EAB location and meteorological cor.ditions, were not considered in this evaluation. The ACTION statement permitting POWER OPERATION to continue for limited ti_me periods with the reactor coolant's specific activity greater than-1 microcurie / gram DOSE EQUIVALENT I-131, but within the allowable limit'shown on Figure 3.4-1, accommodates possit'le iodine sp'iking phenomenon which may occur following changes in THERMAL POWER. The sample analysis for determining the gross specific activity and E can exclude the radioiodines because of the low reactor coolant limit of 1 microcurie / gram DOSE EQUIVALENT'I-131, and because, if the limit is exceeded, the radio-iodine level i.< to be determined every 4 hours. If the gross specific activity level and radioiodine level in the reactor coolant were at their limits, the radioiodine contribution would be approximately 1E In a release of reactor coolant with a typical mixture of radioactivity, che actual radioiodine contri-bution would probably be about 20L The exclusiJn of radionuclides with half-lives less than 10 minutes from these determinations has been made for several reasons. The first consideration is th9 difficulty to identify short-lived radionuclides in a sample that requires a significant time to collect, transport, and analyze. The second consideration is the predictable delay time i between the postulated release of radioactivity from the reactor toolant to its release to the environment and transport to the EAB, which is relatable to at least 30 minutes decay time. The choice of 10 minutes for the half-life cutoff wac made because of thee nuclear characteristics of the typicci reautor cooleC radioactivity. The radionuclides in the typical reactor coolant have half-l lives of less than 4 minutes or half-lives of greater than 14 minutes, which allows a distinctio. between the radionuclides above and belce a ha h-life of 10 minutes. For these reasons the radionuclides that are excluded from consideration are expected to decay to very low levels before they could be transported from the reactor coolant to the EAB under any accident condition, i COMANCHE PEAK - UNIT 1 B 3/4 4'-6

I

REACTOR COOLANT SYSTEM . -3 s

b)

          " BASES                                                           _ _ _

SPECIFIC ACTIVI,TY (Continued) Bssed"upon the above considerations for excluding certain radionuclides trom the sample' analysis, the allowable time of 2 hours between sample taking and completing the initial analysis is based upon a typical _ time necessary to perform the sampling, transport the sample, and perform the analysis of about 90 minutes. .After 90 minutes, the gross count should be made in a reproducible geometry of sample and counter having rcproducible beta or gamma self-shielding

            . properties. The counter should be reset to a reproducible efficiency versus energy. It is not necessary to identify specific'nuclides.                          The radiochemical determination of nuclides sho'uld be based on multiple counting of the sample within typical counting basis following sampling of less than 1 hour, about 2 hours, aboJt 1 day, aboist I week, and about 1 month.

R. educing T to less than 600*F prevents'.t's release of activity should asteamgenerat8E9 tube rupture since the saturation.p'ressure of the reactor coolant is below the lift pressure of the atmospheric steam relief valves. The Surveillance Requi rements provide adequate assurance that excessive specific activity levels in the reactor coolant will be detected in sufficient time to

            .take corrective action. A reduction in frequency of isotopic analyses following power changes'may be permissible 'if justified by the data obtained.

3/4.4.9 PRESSURE / TEMPERATURE LIMITS The temperature and pressure changes during heatup and couldowr. 3re limited to be consistent with the requirements given in the ASME Boiler and Pressure Vessel Code, Section III, Apperdix G and 10 CFR 50 Appendix G.

1. The reactor coolant temperature and pressure and system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with Figures 3.4-2 and 3.4-3 for the service period specified thereon:

a. Allowable combinations of pressure and temperature for specific temperature change rates are below and to the right of the limit lines shown. Limit lines for cooldown rates between those presented may be obtained by inter'polation; and

b. Figures 3.4-2 and 3.4-3 define limits to assure prevention of r.on-ductile failure only. For normal operation, other inherent plant characteristics, e.g., pump heat addition end pressurizer heater capacity, may limit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges.

COMANCHE PEAK - UNIT 1 B 3/4 4-7

o29 , REACTOR COOLANT $YSTEM ' BASES PGESSURE/ TEMPERATURE LIMITS (Continued)

2. These lir.it linas shall bedalculatedperiodicalkyusingmethodsprovided below,

3. The secondary side of the steam generator muit not be pressurized above 200 psig if the tempera.ture of the steam generator is below 70 F,

4. The pressurizer heatup and cooldown rates shall not exceed 100*F/h and 200*F/h, respectively. The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater thin 625*F, and

5. System preservice hydrotests and inservice leak and hydrotests shall t>e performe'd at pressures in accordance with the requirements of ASME Boiler

               .and Pressure Vessel Code, Section XI.

The new 10 CFR 50, Appendix G rule addresses the metal temperature of t'he closure head flange and vessel f w,ge regions. This rule states that the mir.i-mum metal temperature of the closure flange region should be at least 120 F higher than the limiting RT NOT f r these regions when the pressure exceeds 20% of the preservice~ hydrostatic test pressure (621 psig for Westinghouse plants). For Comanche Peak Unit 1, the minimum temperature of the closure flange and the vessel flange regions is 160*F since the limiting RT NDT is 40*F (see Table B 3/4.4-1). The Comanche Peak Unit 1 cooldown curves'shown in Figure 3.4-3 are impacted by this new rule. The fracture toughness properties of the ferritic materials in the reactor vessel are determined in accordance with the NRC Standard Review Plan, ASTM , E185-73, and in a:cordance with additional reactor vessel requirer.ents. These propertie' are then evaluated in accordance with Appendix G of the 1976 Summer l Addenaa to Section III of the ASME Boiler and Pressure Versel Code and the calculation methods described in WCAP-7S24-A, "Basis for Heatur and Cooldow-Limit Curves," April 19'5. ' Heatup and cwldown limit curves are calculated using the most limiting value of the nil Actility reference temperature, RTNDT, at the end of [12] effective full power years (EFPY) of serv.cc life. The 16 EFPY service life period is chosen such that the limiting RT g at the 1/4T locatic.) in ! the core region is greater than the RT gp7 of the limiting :.mrradiated materi'.1. The selection of such a limiting RTg !ssures that all c.c.:.pner.;s in the Reactor Coolan.'. System will be operated conservatively in accordance with ' applicabk Code requirements. , l . ! C9MANCHE PEAK - UNIT 1 B 3/4 4-8

T_ABLE E 3/4.4-1 __ e o o REACTOR VESSEL TOUGHNESS k Q ASME 50 FT-LB/35 RT ' MIN. UPPER SHELF m COMP MATERIAL CU P NOTT MIL TEMP *F NOT FT-LB y COMP 0t8ENT CODE TYPE % % *F LONG TRANS *F LONG. TRANS 2 7 e C F.a O b hp

m G W 5 . e f e

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FIGURE f, 3/4.4-1 FAST NEUTRON FLUENCE (E>1MeV) AS A fi)NCTION OF FULL POWER SERVICE LIFE-4 Y 1 . R l ll * ; j -

                                                                                                                                                      .                                                                                                                      t 1                                                                                                 .
                                                                              . - , ,     , . , - .       ,.-.c.  , _    .r.,__,             . -. ,        ,..-.7 , ,     .,,, .             , . -,     .m     , , . , , . . . - . . . . . .            . , - ,            ,

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FIGURE B 3/4.4-2

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                                                                                                                                                                     ==4   ..                           .

EFFECT OF Flit)ENCE AND COPPER CONTENT ON SHIFT.0F RT - NOT FOR REACTOR VESSELS EXPOSED TO 550*F , ,

                                                                                                                                                                                            ~p 1
                               =                                                                              ,                                                                                          L k
                                                                                                  -m'~~: . -    ., , . . , . , , _ , , , , , _          _ _ __ _ _ _

e- .m , 7 . a . REACTOR C00LAhf SYSTEM PASES , _ PRESSURC/ TEMPERATURE LINITS (Continued) The reactor vessel materials have' been tested to deterdiine their initial RTNOT; the results cf these tests are shown in Table B 3/4.4-1. Reactor opera-tion and resultant fast neutron (C greater than 1 MeV) irradiation can cause an increase in the RT NDT. Therefore, an adjusted reference temperature, based upon the fluenca, copper content, and phosphorus content of the material in question, can be predicted using Figure B 3/4.4-1 aid the 13rgsst value of ARI N6T computed by either Regulatory Guide 1.99, Levision 1, "Effects of

         ' Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials,"

4r the Westinghouse Copper Trend Curves shown'in Figure B 3/4.4-2. The heatup and cooldown limit curves of Figures 3.4-2 and 3.4-3 include predicted adjust-ments for i.his shift in RT NOT at the end of 16 EFPY as well as adjustments for possibic errors in the pr6ssure and temperature sensing instruments. Values of ART NOT dctermined in this manner may be used until the results frca th9 material surveillance program, evaluated according to ASTM E185, are available. Capsules will be removed in accordance with the raquirements of ASTM E185-73 and 10 CFR Part 50, Appendix H. The surveillance specimen with-drawal sche'dule is shown in Table 4.4-5. The 16ad factor ' rep' resents the rela-tiohship between the fast neutron flux density at the location of the capsule and the inner wall of the reactor vessel. Therefore, the results obtained from the surveillarice specimens can be used to predict future radiation damage to the reactor vessel material by using the lead factor and the withdrawal time of.the capsule. Yhe heatup and cooldown curves must be racalculated when the 6RT determined from the surveillance capsule exceeds the calculated NDT t.RT NOT f r the equivalent capsule radiation exposure. Allowable pressure-temperature relationships for various heatup and cooldown rates are calculated using methods derived from Appendix G in Sec-tion III of the ASME Boiler and Pressure Vessel Code as required by Appradix G to 10 CFR Part 50, end these methods are discussed in detail in WCAP-7924-A. The general method for calculating heatup and cooldown limit curves is based upon the principles of the linear elastic fracture mechanics (LEFM) technology. In the calculation procedures a semielliptical surface defect with a depth of one quarter of the wall thickness, T, and a length of 3/2T is assumed to exist at the inside of the dessel wall as well as at the outside of'the vessel wall. The dimensions of this postulated crack, referred to in Appendix G of ASME Section III as the reference flaw, amply exceed tne current capabilities of inservice inspection techniques. Therefore, the reactor operation limit curves developed for this reference crack are cc'iservative and provide sufficient safety margins for protection against nonductile failure. To assure tilat the radiatior embrittlement COMANCHE PEAK - UNIT 1 8 3/4 4-12

      .                .'                                                              g.

REACTOR COOLANT SYSTEM BASES PRESSURE / TEM *ERATURE LIMITS (Continued) effects are accounted for in the calculation of the limit cur /es, the most limiting value of the nil-ductility reference temperature, RTN0T, is used and this includes the radiation-induced shift, ARTNOT, correspondin'g to - the end 6f the period for w' hic'h heatup and cooldown curves are generated. The ASME approach for calculating the alloN ble limit curves for various heatup and cooldown rates specifies that the total stress intensity f actor, Ky , for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, K IR' for the metal temperature at that time. K IR is obtained from the reference fracture toughness curve, defined in Appendix G to the ASME Code. The K IR curve is given by the equation: K7g = 26.78 + 1.223 exp [0.0145(T-RTNOT 160)] (1) Where: K 7g is the reference stress intea.sity factor as a function of the metal temperature T and the metal nil-ductility reference temperature RTNDT. Thus, the governing equation for the hea'up-cooldown' t analysis is defined in Appendix G of the ASME Code as fol. lows: , C K;g + kit <KIR' (2) Where: KIM = the stress intensity factor caused by membrane (pressure) stress, ! K It a the stress intensity factor caused by the thermal gradients, KIR = constant provided by the Code as a function of temperature relative to the RT NDT f the material, C e 2.0 for level A and B service limits, and l C = 1.5 for inservice hydrostatic and leak test operations. At any time during the heatup or cooldown transient, K IR is determined by the metal temperature at the tip of the postulated flaw, the appropriate value for RTNOT, and the reference fracture toughness curve. The thermal stresses resulting from temperature gradients through the vessel wall are calculeted and.then the corresponding thermal stress intensity factor, KIT, f r the. reference flaw is computed. From Equation (2) the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated. , COMANCHE PEAK - UNIT 1 B 3/4 4-13

REfCTOR COOLANT SYSTEM BASES _ _ _ PRESSUfE/TEMPERATURELTM11S(Continued) . C00 LOO 4f, . . Mr the calculation of the allowable pressure versus coolant temperature during cooldown, the Code reference flaw is' assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of.the wall because the thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates. Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations. From these relations, composite limit curves are constructed for each cooldown rate of interest. The use of the composite curve in the cooldown analysis is necessary

because control of the cooldown procedure is based on measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temp 3rature at the tip of the assumed flaw. During cooldown, the 1/4T vessel location is at a higher temperature than the fluid adjacent to the vessel 10. This condition, of course, is not true for the steady-state situa- -

tion. It.follows that at any given reactor coolant temperature, the AT developed during cooldown results.in a higher value of K IR at the 1/4T location for finite cooldown rate's than for steady-state operation. Furthermore, if conditions exist such that the increase in K IR exceeds Kyg, the calculated allowable pre',;re during cooldown will be greater th'an the steady-state value. The above procedures are needed because there is no direct control on temperature et the 1/4T ' location; therefore, allowable pressures may unkr.owingly be violated if the rate of enoling is decreased at various intervals along a < cooldown ramp. The use of the composite cut.e eliminates this problem and assures :onservative operation of the system for the entire cooldown period. HEaTUP Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rste conditions anuming the presence of a 1/4T defect at the inside of the wssel wall. The thermal gradients during heatup produce compressive stresses at the inside of the wall that alleviate the tensile stresses produced by-internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the K IR f r the 1/4T crack

during heatup is lower than the K IR f r the 1/4T crack during steady-state , conditions at the same coolant temperature. During heatup, especia'lly at the end of the transient, conditions may exist such that the effects of compressite thermal stresses ano different K 's f r steady-state and finite heatup rates IR I B 3/4 4-14 COMANCHE PEAK - UNIT 1

REACTOR COOLANT SYSTEM g 9qp NI BASES

                                                                . .uum PRESSURE / TEMPERATURE LIMITS (Continued) do not-offset each other and the pressure-temperature. curve based on steady-state conditions no longer represents a lower bo'und of all similar curves for finite heatup rates when the 1/4T flaw is considered. Therefore, both cases have to        -

be analyzed in order to assure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained. The second portion of the heatup analysis concerns the calculation of pressur.--temperature limitations for the case in which a 1/4T deep outside surface flaw is assumed. Unlike the situation at the vestoi inside surface,

          'the thermal gradients established at the outside surface during heatup produce stresses which are tensile in natu,re and thus tend to reinforce any pressure stresses present. These thermal stresses, of course, are dependent on both th.e rate of heatup and the time (or coolant temperature) along the heatup
   .        ramp. Furthermore, since the thermal stresses at,the outside art tensile and
           ' increase with increasing heatup' rate, a lower bound curve cannod be defined.

Rather, each heatup rate of interest must be analyzed on an in'avidual basis. Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are . produced as follows. A composite curve is constructed based on a point-by- . point compariso.. of the steady-state and finite heatup rate data. At any - given temperatt.re, the allowable pressure is taken to be the lesser of the three values taken frcm the curves under consideration. The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist such that over the course of the heatup ramp the controlng condition switches from the inside - to the outside and the pressure limit must at all times be based on analysis of the rost critical criterion. Finally, the composite curves for the heatup rate data and the cooldown l rate data are adjusted for possible errors in the pressure and temperature , sensing instruments by the values indicated on the respective curves. Although the pressurizer opera +!s in temperature ranges above those for which there i' reason for concern or nonductile failure, operating limits are provided to assure compatibility of operation with the fatigue analysis r performed in accord,nce with the ASME Code requirements. LOW TEMPERATURE qVERPRE3SURE PROTECTION The OPERABILITY of two PORVs or an RCS vent opening of at least 2.98 square j inches ensures that the RCS will be pretected from pressure transients which could exceed the limits of 10 CFR 50 Appendiy. G when one or more of the RCS cold legs are less than or equal to [275]'F. Either PORV has adequate relieving capa-bility to protect the RCS from overpressurization when the transient is limited tc either: (1) the start of'an idle RCP with the secondary water temperature of the steam generator less than or equal to 50*F above the RCS cold leg tempera-tures, or (2) the start of a HPSI pump and its injection into a water-solid RCS. COMANCHE PEAK - UNIT 1 B 3/4 4-15

e ~ . . i REACTOR COOLANT SYSTEM BASES The Maximum Allow d PORV Setpoint for the Low Temperature Overpressure Protection System (LTOPS) is derived by analysis which models the performance of the'LTOPS assuming various mass input and heat input transients. Operation with a PORV Setpoint less than or equal to the maximum Setpoint ensures that Appendix G criteria will not be violated with consideration for a maximum pressure overshoot beyond the PORV Setpoint which can occur as a result of time delays in signal processing and valve opening, instrument uncertainties, and single failure. To ensure that mass and heat input transients more severe than those assumed cannot occur, Technical Specifications require lockout of all but one safety injection pump and all but one centrifugal charging pump while in MODES 4, 5, and 6 with the reactor vessel head installed and disallow start of an RCP if secondary temp-erature is more than 50*F above primary temperature. , The Maximum Allowed PORV Setpoint for the LTOPS will be updated based on the results of examinations of reactor vessel. material irradiation surveillance i 4 specimens performed'as required by 10 CFR Part 50', Appendix H, and in accordance' . with the schedule in Table 4.4-5. 3/4.4.10 STRUCTURAL INTEGRITY . , The inservice inspection and testing programs for ASME Code Class 1,'2, and 3 components ensure that the structural integrity and operational readiness of these components will be maintained at an acceptable level throughout the life of the pl e t. These programs are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50.55a(g) except where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i). Components of the Reactor Coolant System were designed to provide access to permit inservice inspections in accordance with Section XI of the ASME . Boiler and Pressure Vessel Code, Edition and Addenda through . 3/4.4.11 REACTOR COOLANT SYSTEM VENTS Reactor Coolar.t System vents are provided to exhaust noncondensible gases and/or steam from the Reactor Coolant System that could inhibit natural circe'- ation core cooling. The OPERABILITY of least one Reactor Coolant System ve.4 path from the reactor vessel head, and the pressurizer steam space, ensures that the capability exists to perform this function, l The valve redundancy of the Reactor Coolant System vent paths serves to minimize the probability of inadvertent or irreversible actuation while ensuring I that a single failure of a vent valve, power supply, or control system does not ! ' prevent isolation of the vent path. The function, capabilities', and testing requirements of the Reactor Coolant , System vents are consistent with the requirements of Item II.B.1 of NUREG-0737, "Clarification of TMI Action Plant Requirements,'" November 1980, i COMANCHE PEAK - UNIT 1 B 3/4 4-16

3/4.5 EMERGENCY CORE COOLING SYSTEMS M. w1 - BASES _ 3/4.5.1 ACCUMULATORS . The 0."ERABILITY of each Reactor Coolant System (RCS) accumulator ensures that a sufficient volume of borated water will be immediately forced into the reactor core through each of the cold legs in the event the RCS pressure falls below the pressure of the accumulators. This initial surge of water into the core provides the initial cooling mechanism during large RCS pipe ruptures. The limits on accumulator volume, boron concentration and pressure ensure that the assumptions used for accumu.lator injection in the safety analysis are met. The accumulator power operated isolation valves are considered to be "operating bypasses" in the context of IEEE Std. 279-1971, which requires that bypasses of a protective function be removed automatically whenever permissive conditions are not met. In addition, as these accumulator isolation valves fail t:0 meet single failure criteria, removal of power to the valves is required by BTB ICSB 18. This is accomplished via key-lock control board cut-off switches. The limits for operation with an accumulator inoperable for any reason except an isolation. valve closed minimizes the time exposure of the. plant to a LOCA event occurring concurrent with failure of an additional accumulator' which may result in unacceptable peak cladding temperatures. If a closed isolation valve cannot be immediately opened, the full capability of one accumulator is not available and prompt action is required to place the reactor in a mode where this capability is not required. 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS The OPERABILITY of two independent ECCS subsystems ensures that sufficient emergancy core cooling capability will be available in the event of a LOCA assuming the loss of one subsystem through any single failure consideration. Either subsystem operating in conjunction with the accumulators is capable of supplying sufficient. core cooling to limit the peak cladding temperatures within acceptable limits for all postulated break sizes ranging from the double ended break of the largest RCS cold leg pipe downward. In addition, each ECCS subsystem provides long-term core cooling capability in the recirculation mode during the accident recovery period. With the RCS temperature below 350 F, one OP m E ECCS subsystem is acceptable without single failure cons.ideration , w basis of the stable reactivity condition of the reactor and the limittu core cooling requirements. The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the requirement to verify all charging pumps excep't the required COMANCHE PEAK - UNIT 1 B 3/4 5-1

EMERGENCY CORE COOLING SYSTEMS BASES ECCS SUBSYSTEMS (Continued) , OPERABLE charging pump to be inoperable below 350'F provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV. The requirement to remove power from certain valve operators is in accord-ance with Branch Technical Position ICSB-18 for valves that fail to meet single failure considerations. Power is removed via key-lock switches on the control board. . The Surveillance Requirements provided to ensure OPERABILITY of each component ensures that at a minimum, the assumptions used in the safety analyses are met and that subsystem OPERABILITY is maintained. Surveillance Requirements for throttle valve position stops and flow-balance testing provide' assurance that proper ECCS flows will be n.aintained in the event of a LOCA. Maintenance

of proper flow resis'tance and pressure drop in the piping system to each injection point is necessary to: (1) prevent total pump flow from exceeding runout conditions when the system is in its minimum resistance configuration, (2) provide the proper flow split between injection points in accordance with the assumptions used in the ICCS-LOCA analyses, and (3) provide an acceptable level of total ECCS flow to all injection points equal to.or above~that assumed in the ECCS-LOCA analyses. 3/4.5.4 REFUELING WATER STORAGE TANK 1 The OPERABILITY of the refueling water storage trnk (RWST) as part of the ECCS ensures that a sufficient supply of borated water is available for injec-tion by the ECCS in the event of a LOCA. The limits on RWST minimum volume and boron concentration ensure that: (1) sufficient water is available within containment to permit recirculation cooling flow to the core, (2) for small break LOCA and steam line breaks, the reactor will remain subcritical in the i cold condition following mixing of the RWST and the RCS water volumes with all control rods inserted except for the most reactive control assembly, and (3)

for large break LOCAs, the reactor will remain subcritical in the cold condition following mixing of the RWST and the RCS water volumes with all shutdown and control rods fully withdrawn, and (4) sufficient time is available for the operator to take manual action and complete switchover of ECCS and containment spray suction to the containment sump without emptying the RWST or losing suction.

The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics. The limits on contained water volume and boron concentration of the RWST

also ensure a pH value of between 8.5 and 10.5 for the solution recirculated

, within containment after a LOCA. This pH band minimizes the evolution of C l iodine and minimizes the effect of chloride and caustic stress coirosion on ! mechanical systems and components. 1 COMANCHE PEAK - UNIT 1 B 3/4 5-2

                       .. .~.

3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.' PRIMARY LONTAINMENT 3/4.6.1.4 CONTAINMENT INTEGRITY Primary CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the safety analyses. This restriction, in conjunction with the leakage rate limitation, will limit the EXCLUSION AREA BOUNDARY radiation doses to within the dose guidelino values of 10 CFR 100 during accident conditions. 3/4.6.1.2 CONTAINMENT LEAKAGE The limitations on c,ontainment leakage rates ensure that the total containment leakage volume will not exceed the value assumed in the safety ' analyses at the peak accident pressura, P,. As an added conservatism, the measured overall integrated leakage rate is further limited to less than or eciual' to 0.75 L, or 0.75 Lt . as applicable, during performance of the periodic test to account for possible degradation of the containment leakage barriers

  ,between leakoge tests.

The surveillance testing for measuring leakage rates is consistent with

 *the requirements of 10 CFR 50 Appendix J.

3/4.6.1.3 CONTAINMENT AIR LOCKS The limitations on closure and leak rate for the containment air locks l are required to meet the restrictions on CONTAINMENT INTEGRITY and containment leak rate. Surveillance testing of the air lock seals provides assurance that the overall air lock leakage will not become excessive due to seal damage during the intervals between air lock leakage tests. 3/4.6.1.4 INTERNAL PRESSURE The limitations on containment internal pressure ensure that: (1) the containment structure is prevented from exceeding its design negative pressure r differential of 5 psig with respect to the outside atmosphere, and (2) the containment peak pressure does not exceed the design pressure of 50 psig I during LOCA. The maximum peak pressure expected to be obtained from a LOCA event is 46.3 psig. The limit of 1.5 psig for initial positive containment pressure will limit the total pressure to 48.3 psig, which is less than design pressure - and is consistent with the safety analyses. l l l COMANCHE FEAK - UNIT 1 S 3/4 6-1 a

         ~ CONTAINMENT SYSTEMS                                                           g' , (

BASES 3/4.6.1.5 AIR TEMPERATURE The limitations on co6tainment average air temperature ensure that the overall containment average air temperature does not exceed the initial tem-perature condition assumed in the safety analysis for a LOCA or steam line break accident. Measurements shall be made at all listed locations, whether by fixed or portable instruments, prior to determining the average air temperature. 3/4.6.1.6 CONTAINMENT STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the contairment will be maintained comparable to the original design standards for the life of the facility. Structural integrity is required to ensure that the containment will withstand the maximum pressure of 48.3 psig in the event of a LOCA.' A visual inspection in conjunction with the Type A leakage tests is sufficient-to demonstrate this capability. ( 3/4.6.1.7 CONTAINMENT VENTILATION SYSTEM The 48-inch and 12-inch containment hydrogen purge supply and exhaust isolation valves are required to be locked closed during plant opefations since these' valves have.not been demonstrated capable of closing during a LOCA or steam line break accident. Maintaining these valves locked closed during plant operation ensures that excessive quantities of. radioactive materials will not be released via the Containment Ventilation System. To provide assurance that these containment valves cannot be inadvertently opened, the i valves are locked closed in accordance with Standard Review Plan 6.2.4 which includes mechanical devices to seal or lock the valve closed, or prevents j power from being supplied to the valve operator. i The use of the Containment Ventilation System during operations is

restricted to the 18-inch pressure relief discharge isolation valves since, t.nlike the 48-inch and 12-inch valves, the 18-inch valves are capable of closing during a LOCA or steam line break accident. Therefore, the SITE BOUNDARY dose guideline of 10 CFR 100 would not be exceeded.in the event of an l accident during containment PURGING operation. Operation with one pair of these valves open will be limited to 90 hours during a calendar year. The i

total time the containment purge (vent) system isolation valves may be open 1 during MODES 1, 2, 3, and 4 in a calendar year is a function of anticipated need and operating experience. Only safety-related reasons; e.g., containment pressure control or the reduction of airborne radioactivity to facilitate i parsennel access for surveillance snd maintenance activities, may be used to i suppirt the additional time requests. Only safety-related reasons should be used to justify the opening of these isolation valves during MODES 1, 2, 3, and 4 in any calendar year regardless of the allowable hours. . { l l COMANCHE PEAK - UNIT 1 B 3/4 6-2 L

CONTAINMENT SYSTEMS jik BASES CONTAINMENT VENTILATION SYSTEM (Continued) -

          ' Leakage integrity tests with a maximum allowable leakage rate for c'ontain -

ment ventilation valves kill provide early indication of resilient material seal degradation and will allow opportunity for repair before gross leakage failures could develes. The 0.60 L leakage limit of Specification 3.6.1.2b. shall not be excaeced when the leak $ge rates determined by the leakage integrity tests of these valves are added to the previously determined total for all valves and penetrations subject to Type B and C tests. 13/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS 3/4.6.2.1 CONTAINMENT SPRAY SYSTEM The OPERABILITY of~the Containment Spray System ensures that containment depressurization and cooling capability will be available in the event of a LOCA or steas line' break. The pressure reduction and resultant lower contain-ment leakage rate are consistent with the assumptions used in the safety analyses. The Containment Spray System which is composed of redundant trains, pro-vides post-accident cooling of the' containment atmosphere. However, the Con-tainment Spray System also provides a mechanism for removing iodine from the containment atmosphere and therefore.the time' requirements for restoring an-inoperable Spray System to OPERABLE status have been maintained consistent with'that assigned other inoperable ESF equipment. . 3/4.6.2.2 SPRAY ADDITIVE SYSTEM . The OPERABILITY of the Spray Additive System ensures that sufficient NaOH is added to the containment spray in the event of a LOCA. The limits on NaOH volume and concentration ensure a pH value of between 8.5 and 10.5 for the solution recirculated within containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components. The contained solution volume < l l'imit includes an allowance for solution not usable because of tank discharge line location or other physical characteristics. These assumptions are con-l sistent with tne iodine removal efficiency assumed in the safety analyses. l 3/4.6.3 CONTAINMENT ISOLATION VALVES l The OPERABILITY of the containment isolation valves ensures that the con-tainment atmosphere will be isolated from the outside environment in the event l of a release of radioactive material to the containment atmosphere o'r ' pres uri-l zation of the containment and is consistent with the requirements of General l Design Criteria 54 through 57 of 10 CFR 50 Appendix A. Containment iso-l 1ation within the tire limits specified for those isolation valves designed to i close automatically ensures that the release of radioactive material to the en-l vironmert will be consistent with the apsumptions used in the analyses for a LOCA. l t COMANCHE PEAK - UNIT 1 B 3/4 6-3 L

CONTAINMENT SYSTEMS %1 BASES 3/4.6.4 COMBUSTIBLE GAS CONTROL The OPERABILITY of the equipment and systems required for the detection and control o,f hydrogen gas ensures that this equipment will be available to maintain the hydrogen concentration within containment below its flammable limit during post ~LOCA conditions. Either recombiner unit is capable o) con-trolling the expocted hydrogen generation associated with: (1) zirconium-water reactions, (2) radiolytic decomposition of water, and (3) corrosion of metals within containment. These Hydrogen Control Systems are consistent with the recommendations of Regulatory Guide 1.7, "Control of Combustible Gas Concen-trations in Containment Following a'LOCA," March 1971. 9 t a l l l l i : COMANCHE PEAK - UNIT 1 B 3/4 6-4

i. .

4 4 4 g 3/4.7 PLANT SYSTEMS U at 4 BASES 3/4.7.I' TURBINE CYCLE-3/4.7.1.1 SAFETY VALVES The OPERABILITY of the main steam line Code safety valves ensures that the Secondary System pressure will be limited to within 110% (1305 psig) of its design pressure of 1185 psig during the most severe anticipated system opera-tional transient. The maximum relieving capacity is associated with a Turbine -

              - trip fr9m 100% RATED THERMAL POWER coincident with an assumed loss of condenser heat sink (i.e., no steam bypass to the condenser).

The specified valve lift settings and relieving capacities are in accor-dance with the requirements of Section III of the ASME Boiler and. Pressure Code, 1974 Edition. The total relieving capacity for all valves on all of the steam. lines is 18,190,884 lbs/h which is 120% of the to_tal. secondary' steam flow of 15,140,106 lbs/h.at 100% RATED THERMAL POWER. STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis of the reduction

        ,      in Secondary Coolant System steam flow and THERMAL POWER required by the.

reduced Reactor Trip settings of the Power Range Neutron Flux channels. The Reactor' Trip Setpoint' reductions are derived on the following bases: For four loop operation SP = (X) - (Y)(V) x (109) Where: SP = Reduced Reactor Trip Setpoint in percent of RATED THERMAL I POWER, V = Maximum number of inoperable safety valves per steam line, 109 = Power Range Neutron Flux-High Trip Setpoint for four loop operation, X = Total re!ieving capacity of all safety valves per steam , line in 1bs/ hour, and I i Y = Maximum relieving capacity of any one safety valve in 1bs/ hour l i COMANCHE PEAK - UNIT 1 B 3/4 7-1 f

PLANT SYSTEMS BASES 3/4.7.1.2 AUXILIARY FEE 0 WATER SYSTEM The OPERABILITY of the Auxiliary Feedwater System ensures that the ' Reactor Coolant System can be cooled down to less than 350*F from normal operating conditions in the event of a total loss-of-offsite power. Each electric motor-driven auxiliary feedwater pump is capable of deliver-ing a total feedwater flo.w of 430 gpm to.two steam generators at a pressure of 1221 psig to the entrance of the steam generators. The steam-driven auxiliary feedwater pump is capable of delivering a total feedwater flow of 860 gpm to four steam generators at a pressure of 1221 psig to the entrance of the steam-generators. This capacity is sufficient to ensure that adcquate feedwater flow is available to remove decay heat and reduce the Reactor Coolant System temp-erature to less than 350*F when the Residual Heat Removal System may be placed into operation. The Auxiliary Feeawater System is capable of deliverin~g a total feedwater . flow of 430 gpm at a pressu e of 1221 psig to the entrance of at least two steam generators while allowing for: (1) any possible spillage through the design worst case break of the main feedwater line; (2) the design worst case single. failure; and (3) r<teirculation flow. This capacity is sufficient to . ensure that adequate feedwater flow is available'to remove decay heat and reduce Reactor Coolant System temperature to less than 350 F at which point the Residual Heat Removal System may be placed in operation. 3/4.7.1.3 CONDENSATE STORAGE TANK The OPERABILITY of the condensate storage tank with the minimum water volume ensures that sufficient water is available to maintain the RCS at HOT STANDBY conditions for 18 hours with steam discharge to the atmosphere concur-rent with total loss-of-offsite power or 4 hours at HOT STANDBY followed by a cooldown to 350 F at a rate of 50 F/HR for 5 hours. The contained water volume limit includes an allowance fsr water not usable because of tank discharge line - location or other physical characteristics. NUREG-0737, Item II.E.1.1 requires a backup source to the CST which is the CPSES Station Service Water System, which can be manually aligned, if required. 3/4.7.1.4 SPECIFIC ACTIVITY The limitations on Secondary Coo.lant System specific activity ensure that the resultant offsite radiation dose will be limited to a small fraction of 10 CFR 100 c'ese guideline values in the event of a steam line rupture. This dose also includes the effects of a coincident 1 gpm primary-to-secondary tube leak in the steam generator of the affected steam line. These values are e.onsistent with the assumptions used in the safety analyses. COMANCHE PEAK - UNIT 1 B 3/4 7-2 L

PLANT SYSTEMS 4 k BASES

 .                                         , 3/4.7.1.5 MAIN STEAM LINE ISOLATION VALVES
                                .                          The OPERABILITY of the main steam line isolation valves ensures that no more than one steam generator will blow down in the event of a steam line rupture. This restriction is required to:                                                    (1) minimize the positive reac-tivity effects of the Reactor Coolant System cooldown associated with the blowdown, and (2) limit the pressure rise within containment in the event the steam line rupture occurs within containment. The OPERABILITY of the main steam isolation valves within the closure times of the Surveillance Require-ments are consistent with the assumptions used in the safety analyses.

3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION The limitation on steam generator pressure and temperature ensures that the pressure-induced stresses in the steam generators do not exceed the maximum allowable fracture toughness stress limits. The limitations of 70*F and 200 psig are based on a steam generator RTNDT f 60 F and are sufficient to prevent brittle fracture. 3/4.7.3 COMPONENT' COOLING WATER SYSTEM The OPERABILITY of the Component Cooling Water System ensures that suf-ficient cooling capacity is available for continu.ed operation of safety-related equipment during normal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the safety analyses. 3/4.7.4 STATION SERVICE WATER SYSTEM The OPERABILITY of the Stat.on Service Water System ensures that suffi-cient cooling capacity is availa;ie for continued operation of safety-related equipment during normal ano accident conditions. The redundant cooling capa-city of this system, assuming a single failure, is consistent with the assump-tions used in the safety analyses. 3/4.7.5 ULTIMATE HEAT SINK . The limitations on the ultimate heat sink level and temperatur'e ensure that sufficient cooling capacity is available either: (1) provide normal cool-down of the facility or (2) mitigate the effects of accident conditions within acceptable limits. COMANCHE PEAK - UNIT 1 B 3/4 7-3

PLANT SYSTEMS y K 'T g BASES L udi ULTIMATE HEAT SINK (Continued) . The limitations on minimum water level is t,ased on providing a 30-day cooling water supply to safety-related equipment without exceeding its design basis temperature and is consistent with the recommendations of Regulatory Guide 1.27, "Ultimate Heat Sink for. Nuclear Plants," rev. 2 (January 1976). The limitation on maximum temperature is based on the maximum allowable compo-nent temperatures in the Service Water and Component Cooling Water Systems, and the requirements for cooldown. The limitation on average sedimerit depth is based on the possible excessive sediment buildup in the service water intake channel. - 3/4.7.6 FLOOD PROTECTION The limitation of flood protection ensur.es that facility protective actions will be taken in the event of flood conditions. The only credible food condition that endangers safety related equipment is from water entry into the turbine building via the circulating water system from Squaw Creek Reservoir and then only if the level is above 778 feet Mean Sea Level. This corresponds to the elevation at which water could enter the electrical and control building endangering the safety chilled water system. The surveillance requirements are designed to implement level monitoring of Squaw Creek Reservo.ir should it reach an abnormally high level above 776 feet. The Limiting Condition - for Operation is designed to im'plement flood protection, by ensuring no open flow path via the Circulating Water System exists, prior to reaching the postulated flood level. 3/4.7.7 CONTROL ROOM HVAC SYSTEM The OPERABILITY of the Control Room HVAC System ensures that: (1) the control room ambient air temperature does not exceed the allowable tempera ^ure for continuous-duty rating for the equipment and instrumentation cooled by this system, and (2) the control room will remain habitable for operations personnel during and following all credible accident conditions. Operation of the system with the heaters operating for at least 10 continuous hours in a 31-day period is sufficient to reduce the buildup of moisture on the adsorbers and HCPA filters. The OPERABILITY of this system in conjunction with control room design provisions' is based on limiting the radiation exposure to personnel occupying the control room to 5 rems or less whole body, or its equivalent. This limitation is consistent with the requirements of General Design Criter-ion 19 of 10 CFR 50 Appendix A. ANSI N510-1975 will be used as a procedural guide for surveillance testing. , COMANCHE PEAK - UNIT 1 B 3/4 7-4

1 PLANT SYSTEMS U BASES 3/4.7.8 PRIMARY PLANT VENTILATION SYSTEM - ESF FILTRATION UNITS

The OPERABILITY of the ESF Filtra' tion Units' ensures that re.dioactive materials leaking from the ECCS equipmeilt within the cafeguarcs and auxiliary buildinJs following a LOCA are filtered prior to reaching the environment. These filtration units also ensure that radioactive materials leakage from within the fuel building are filtered prior to reaching the environment. Operation of the ESF filtration units with the heaters operating for at least 10 continuous hours in a 31-day period is sufficient to reduce the buildup cf moisture on the adsorbers and HEPA filters. The operation of the ESF filtra-tion units and.the resultant effect on offsite dosage calculations was assumed in the safety analyses. ANSI N510-1975 will be used as a proce' dural guide for surveillance testing. - 3/4.7.9 SNVBBERS All snubbers are required'0PERABLE to ensure that the structural integrity of the Reactor Coolant System and all othar safety-related cystems is main-tained during and following a seismic or other event iM tiating dynamic loads. Snubbers are classified and grcuped by design and manufacturer but not by size. For example, mechanical snubbers utilizing the sr.me design features of the 2-kip,10-kip and 100-kip capacity manufactured by Company "A" are of the same type. . The same design mechanic'al snubbers manufactured by Company "B" for the purposes of this Technical Specification would be of a different type, as would hydraulic snubbers from either manufacturer. A list of individual snubbers with detailed information of snubber loca-tion and size and of system affected shall be available at the plant in accor-

dance with 10 CFR 50 50.71(c). The accessibility of each snubber shall be determined and approved by the Static.; Operation Review Committee (SORC). The l determination shall be based upon the existing radiation levels and the l expected time to perform a visual inspection in each snubber location as well l as other factors associ:ded with accessibility during plant operations' (e.g. ,

temperature, atmosphere, location, etc.), and the recommendations of Regulatory Guides 8.8 and 8.10. The addition or deletion of any hydraulic or mechanical snubber shall be made in accordance with 10 CFR 50.59. The visual inspection frequency is based upon maintaining a constant level of snubber protaction to each safety-related system during an earthquake or severe transient. Therefore, the required inspection interval varies inversely with the observed sntuber failures on a given syster,and is determined by the number of iroperable snubbers found'during an inspection of each system. In order to establish the inspection frequency for each type of snubber on a safety-related system, it was assumed that the frequency of snubber failures and initiating events is cons *-'t with time ahd that the failure of any snubber on that system could cause the .ystem to be unprotected and to result in failure during an assumed initiating event. Inspections performed before that COMANCHE PEAK - UNIT 1 B 3/4 7-5 i .

i PLANT SYSTEMS BASES SNUCBERS (Continued) interval has elapsed may be used as a'new re'ference point to determine the next - -

       - inspection. However, the resul.ts of such early. inspections performed before the original required time interval has elapsed ~(nominal time less 25%) may not be used to lengthen the required inspection interval. 'Any inspection whose results require'a shorter inspection interval will override the previous schedule.

The acceptance criteria are to be used in the visual inspection to deter-mine OPERABILITY of the snubbers. For example, if a fluid port of a hydraulic snubber is found to be uncovered, the snubber shall be declared inoperable and shall not be. determined OPERABLE via functional testing. To provide assurance of snubber functional reliability, one of three functicnal testing methods is used with the stated acceptance criteria:. , e 1. Functionally test 10% of a type of snubber with an additional 10% tested fur each functional testing failure, or *

2. Functionally test a sample size and determine sample acceptance or rejection using Figure 4.7-1, or

3. Functionally' test a representative sample size and de.termine sample acceptance or rejection using the stated equation.

Figure 4.7-1 was developed using "Wald's Sequential Probability Ratio Plan" as described in "Quality Control and Industrial Statistics" by Acheson J. Duncan. l Permanent or other exemptions from the surveillance program for individual snubDers may be granted by the Commission if a justifiable basis for exemption i . is presented and, if applicable, snubber life destructive testing was performed to qualify the snubbers for the applicable design conditions at either the com-pletion of.their fabrication or at a subsequent date. Snubbers so exempted shall_be listed in the list of individual snubbers indicating the extent of the exemption,s. ! The service life of a snubber is established via manufacturer input and l information through consideration of the snubber service conditions and i associated installation and maintenance records (newly installed snubbers, seal o replaced, spring replaced, in high radiation area, in high tempersture area, etc.). The requirement to monitor the snubber service life is included to ensure that the snubbers periodically undergo a performance evaluattor in view' of their age and operating conditions. These records will provide statistical bases for future consideration of snubber service life. l l l COMANCHE PEAK - UNIT 1 B 3/4 7-6

PLANT SYSTEMS BASES 3/4.7.10 SEALE0 SOURCE CONTAMINATION The limit'ations on. removable ' contamination for sources requiring leak testing, including' alpha ' emitters, is based on 10 CFR 70.39(a)(3) limits for plutonium. This limitation will ensure that leakage from Byproduct, Source, ' and Special Nuclear Material sources will not exceed allowable intake values. Sealed sources are classified into three groups according to their use, with Surveillance Requirements commensurate with the probability of damage to a source in that group. Those sources which are frequently handled are required to be tested more often than those which are not. Sealed sources which are

       - continuously enclosed within a shielded mechanism (i e., sealed sources within radiation monitoring or boron measuring devices) are considered to be stored and need not be tested unless they are removed from the shielded mechanism.
              . 7. r. AREA TEMPERATURE MONITORING
                'The area temp.arature limitations ensure that safety-related equipment will not be subjected to temperatures in excess of their environmental qualification temperatures. Exposure to excessive temperatures may degrade equipment and can cause a loss of its OPERABILITY. The temperature limits include an allowance for
        'nstrument i               error of i      *F.                               .

O . e i . e 4 l l COMANCHE PEAK - UNIT 1 S 3/4 7-7 i

3/4.8 ELECTRICAL POWER SYSTEMS UbNA BASES 3/4.8_.1, 3/4.8.2, and 3/4.8.3 A.C. SOURCES, D.C. SOURCES, and ONSITE POWER DISTRIBUTION , . The OPERABILITY of the A.C. and D.C power sources and associated distribu-tion systems during operation ensures that sufficient power will be available to supply the safety-related equipment required for: (1) the safe shutdown of the facility, and (2) the mitigation and control of accident conditions within the facility. The minimum specified independent and redundant A.C. and D.C. power sources and distribution systems satisfy the requirements of General Design Criterion 17 of 10 CFR 50 Appendix A. The ACTION requirements specified for the levels of degradation of the power sources provide restriction upon continued facility operation corensurate with the level of degradation. The OPERABILITY of the power sources . consistent with. the initial' condition assumptions of the safety analyse and are based upon maintaining at least one redundant set of onsite A.C. and~DlC. power sources and associated distribution systems OPERABLE during accident conditions coincident with an assumed loss-of-offsite power and single failure of the other onsite A.C. Source. The A.C..and D.C. source allowable out-of-service times are bas e on Regulatory Guide 1.93, "Availability of Electrical Power Sources," December 1974 and Generic Letter 84-15, "Proposed Staff Position to Improve and Maintain Diesel Generator Reliability." When one diesel generator is inoperable, there is an additional ACTION r'equirement to verify that all required systems, subsystems, trains, components and devices, that depend on , the remaining OPERABLE diesel generator as a source of emergency power, are also OPERABLE, and that the steam-driven auxiliary feedwater pump is OPERABLE. This requirement is intended to provide assurance that a loss-of-offsite power event will not result in a complete loss of safety function of critical systems during the period one of the diesel generators is inoperable. The term, verify, as used in this context means to administratively check by examining logs or other information to determine if certain components are out-of-service for mai.m nance or other reasons It does not mean to perform the Surveillance Requirements needed to demonstrate the OPERABILITY of the component. The OPERABILITY of the minimum specified A.C. and D.C. power sources and associated distribution systems during shutdown and refueling ensures that: (1) the facility can be maintained in the shutdown or refueling condition for extended time periods, and (2) sufficient instrumentation and control capa-bility is available for monitoring and maintaining the unit status. The Surveillance Requirements for demonstrating the OPERABILITY of the diesel generators are in accordance with the recommendations of Regulatory Guides 1.9, "Selection of Diesel Generator Set Capacity for Standby Power Supplies," March 10, 1971; 1.108, "Periodic Testing of Diesel Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants," Revision 1, Augsst 1977; and 1.137, "Fuel-Oil Systems for Standby Diesel Generators," Revision 1, October 1979, Generic Letter 84-15, and Generic Letter 83-26, , "Clarification of Surveillance Requirements for Diesel Fuel Impurity Level l Tests." COMANCHE PEAK - UNIT 1 B 3/4 8-1

ELECTRICkLPOWERSYSTEMS .

BASES A.C. SOURCES, D.C. SOURCES, and ONSITE POWER DISTRIBUTION (Continued) The Diesel Generator Test schedule, Table 4~.8-1, is based on the recommenda-tions of Regulatory Guide 1.108, "Periodic Testing of D'esel Generator Units Used as Onsite Electric Power System's at Nuclear Power Plc 5," Revision 1, August 1977, and NRC Technical Report A-3230, "Evaluation of Diesel Unavailability and Risk Effective Surveillance Test Intervals," May 1986, and Generic Letter 84-15, "Proposed Staff Position to Improve and Maintain Diesel Generator R9 liability." The Surveillance Requirement for demonstrating the OPERABILITY of the station batteries are based on the recommendations of Regulatory Guide 1.129, "Maintenance Testing and Replacement of large Lead. Storage Batteries for Nuclear Power Plants," February 1978, Regulatory Guide 1.32 "Criteria for Safety Related Electric Power Systems for Nuclear Power Plants," Revision 2, and IEEE Std 450-1980, "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Large Lead Storage Batteries for Generating Stations and

                   . Substations."

Verifying average electrolyte temperature above th6 minimum for which the battery was sized, total battery terminal voltage'on float charge, connection resistance values, and the performance of battery service and discharge' tests ensures the effectiveness of the charging system, the ability to handle high discharge rates, and compa'res the battery capacity at that time with the rated capacity. , Table 4.8-2 specifies the normal. limits for each designated pilot cell and each connected cell for electrolyte level, float voltage, and specific gravity. The limits for the designated pilot cells float voltage and specific gravity, greater than 2.13 volts and 0.015 below the manufacturer's full charge specific gravity or a battery charger current that had stabilized at a low value, is characteristic of a charged cell with adequate capacity. The normal limitsforeachconnectedcellforfloatvoltageandspecificgravity, greater than 2.13 volts and not more than 0.020 below the manufacturer s full charge specific gravity with an average specific gravity of all the connected cells not more than 0.010 below the manufacturer's full charge specific gravity, ensures the OPERABILITY and capcoility of the battery, i Operation with a battery cell's parameter outside the normal limit but l within the allowable value specified in Table 4.8-2 is permitted for up to 7 days. During this 7-day period: (1) the allowable values for electrolyte level ensures no physical damage to the plates with an adequate electron transfer capability; (2) the allowable value for the average specific gravity of all the cells, not more .v 0.020 below the manufacturer's recommended full charge specific gravity, u r ibat the decrease in rating will be less e an

the safety' margin provided ; O ) the allowable value for an individual cell's specific gravity, rn:, , . .at an individual cell's specific gravity will not be more than 0.040 t4 a the manufacturer's full charge specific gravity and that the overall capability of the battery will be maintained I within an acceptable limit; and (4) the allowable value for an individual cell's float voltage, greater than 2.07 volts, ensures the battery's capability l

to perform its design function. COMANCHE PEAK - UNIT 1 B 3/4 8-2 l

ELECTRICAL POWER SYSTEMS d BASES 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES Containment electrical penetrations and penetration conductors are pr' tected by either deenergizing circuits not required during reactor operation or by demonstrating the OPERABILITY of primary and backup overcurrent protec-tion circuit breakers during periodic surveillance. This is based on the recommendations of regulatory guide 1.63 Revision 2 "Electric Penetration Assemblies in Containment S.ructures for Light-Water-Cooled Nuclear Power Plants." The Surveillance Requirement's applicable to lower voltage circuit breakers and fuses provide assurance of breaker and fuse reliability by testing at least ' 10% of each manufacturer's brand of circuit breaker and/or fuse. Each manu-facturer's molded case and metal case circuit breakers and/or fuses are grouped into representative samples which are then tested on a rotating-basis to ensure that al1 br'eakers and/or fuses are tested. If a wide variety exists within any manufacturer's brand of circuit, breakers and/or fuses, it is necessary to divide that manufacturer's breakers and/or fuses into groups and treat each group as a separate type of breaker or fuses for surveillance purposes. The OPERABILITY [or] [ bypassing) of the motor-operated valves thermal overload protection [ continuously) [or) [during accident conditions) [by integral bypass devices) ensures that the thermal overload protection [during accident conditions) will not prevent safety-related valves from performing ' their function. [The Surveillance Requirements for demonstrating the [0PERA-BILITY] [or] [ bypassing] of the thermal overload protection (continuously) [and) [or) [during accident conditions) are in accordance with Regulatory Guide 1.106, "Thermal Overload Protection for Electric Motors on Motor Oper-ated Valves," Revision 1, March 1977.] i f i l l l . COMANCHE PEAK - UNIT 1 B 3/4 8-3 t -

3/4.9 REFUELING OPERATIONS (( W 1 L1,6.f 454 BASES

                  .                             3/4.9.1 BORON CONCENTRATION-                                         ,

The limitations on reactivity conditions during REFUELING ensure that: (1) the reactor will remain suberitical during CORE ALTERATIONS, and (2) a uniform boron concentration is maintained for reactivity control in the water volume having direct access to the reactor vecsel. These limitations are consistent with the initial conditions assumed for the boron dilution incident in the safety analyses. The value of 0.95 or less for K,ff includes a 1% ok/k conservative allowance for uncertainties. Similarly, the boron

                                             . concentration value of 2000 ppm or greater includes a conservative uncertainty allowance of 50 ppm boron. The lock ag closed of the required valves during refueling operations precludes the possibility of uncontrolled boron dilution of the filled portion of the RCS.        This action prevents flow to the RCS of unborated water by closing flow paths from sources of unborated water.

3/4.9.2 INSTRUMENTATION

The OPERABILITY of the source range neutron flux monitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core.

3/4.9.3 DECAY TIME The minimum requirement for reactor suberiticality prior to movement of irradiated fuel assemblies in the reactor vessel ensures that sufficient - time has elapsed to allow the radioactive decay of the short-lived fission products. This decay time is consistent with the assumptions used in the safety analyses. 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS t The requirements on containment building penetration closure and OPERABILITY ensure that a release of radioactive matorial within containment will be restricted from leakage to the environment. The OPERABILITY and closure restrictions are sufficient to restrict radioactive material release from a fuel element rupture based upon the lack of containment pressurization potential while in the REFUELING MODE. l 3/4.9.5 COMMUNICATIONS Therequirementforcommunicationbcapabilitye.nsuresthatrefueling i station personnel can be promptly informed of significant changes in the facility status or co're reactivity conditions during CORE ALTEP,ATIONS. COMANCHE PEAK - UNIT 1 B 3/4 9-1

REFUELING OPERATIONS BASES . 3/4.9.6 REFUELING MACHINE The OPERABILITY requirements for the refueling machine ma'in hoist and auxiliary monorail hoist ensure that: (1) the main hoist will be used for l movement of fuel assemblies, (2) the auxiliary monorail hoist.will be used for latching, unlatching and movement of control rod drive shafts, (3) the main hoist has sufficient load capacity to lift a fuel assembly (with control rods), (4) the auxiliary monorail hoist has sufficient capacity to latch, unlatch and move the control rod drive shafts, and (5) the core internals and reactor vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations. 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE AREAS The restriction on movement of loads.in excess of the not:inal weight of a fuel and control rod assembly and associated handiing tool over other fuel assemblies in a storage pool ensures that in the event this load is dropped: (1) the activity release will be limited to that contained in a single fuel assembly, and (2) any possible. distortion of fuel in the storage racks will not result in a critical array. This assumption is consistent with the activity release assumed in the safety analyses, 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION , The requirement that at least one residual heat removal (RHR) loop be in operation ensures that: (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor vessel below 140*F as required during the REFUELING MODE, and (2) sufficient coolant circulation in maintained through the core to minimize the effect of a boron dilution incident and prevent boron stratification. The requirement to have two RHR loops OPERABLE when there is less than 23 feet of water above the reactor vessel flange ensures that a single failure of the operating RHR loop will not result in a complete loss of residual heat removal capability. With the reactor vessel head removed and at least 23 feet of water above the reactor pressure vessel flange, a large heat sink is avail-able for core cooling. Thus, in the event of a failure of the operating RHR loop, adequate time is provided to initiate emergency procedures to cool the core. 3/4.9.9 CONTAINMENT VENTILATION ISOLATION SYSTEM . The OPERABILITY of this system ensurcs that the containment ventilation ' penetrations will be automatically isolated upon detection of high radiation levels within the containment. The OPERABILITY of this system is required to ' restrict the release of radioactive material from the containment atmosphere . to the environment. COMANCHE PEAK - UNIT 1 B 3/4 9-2 1

t , REFUELING OPERATIONS I}. I, BASES 3/4.9.10 and 3/4.9.11 WATER LEVEL - REACTOR VESSEL and IRRADIATED FUEL STORAGE , The restrictions on minimum water level ensure that sufficient water depth is available to remove 99% of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly. The minimum water depth is consistent with the assumptions of the safety analysis, f 3/4.9.12 STORAGE POOL VENTILA1;3N SYSTEM The limitations on the Storage Pool Ventilation System ensure that all radioactive material released from an irradiated fuel assembly will be filtered through the HEPA filters and charcoal adsorber prior to discharge to the atmosphere. Operation of the system with the heaters operating for at least 10 continuous hours in a 31-day period-is sufficient.to' reduce the buildup of moisture on the adsorbers and HEPA filters. The OPERABILITY of this system and the resulting iodine removal capacity are consistent with the assumptions of the safety analyses. ANSI N510-1975 will be used as a procedural guide for surveillance testing. ' 4 l 9 4 COMANCHE PEAK - UNIT 1 B 3/4 9-3

3/4.10 SPECIAL TEST EXCEPTIONS BASES 3/4.10.1 SHUTDOWN MARGIN , This special test exception provides thht a minimum amount of control rod worth is immediately available for reactivity control when tests are performed for control rod worth measurement. This special test exception is required to permit the periodic verification of the actual versus predicted core reactivity condition occurring as a result of fuel burnup or bel cycling operations. 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUl' ION LIMITS This special test exception permits individual control rods to be posi-tioned outside of their normal group heights and insertion limits during the performance of such PHYSICS TESTS as those required to: (1) measure control rod worth, and (2) determine the reactor stability index and damping factor under xenon oscillation conditions. 3/4.10.3 PHYSICS TESTS This special test exception permits PHYSICS TESTS to be performed at less than or equal to 5% of RATED THERMAL POWER with.the RCS T,yg slightly lower than normally allowed so that the fundamental nuclear characteristics of the core and related instrumentation can be verified. In order for various' char-acteristics to be accurately measured, it is at times necessary to operate outside the normal restrictions of these Technical Specifications. For ~ instance, to measure the moderator temperature coefficient at BOL, it is necessary to position the various control rods at heights which may not normally be allowed by Specification 3.1.3.6 and the RCS T,yg may fall slightly below the minimum temperature of Specification 3.1.1.4. 3/4.10.4 REACTOR COOLANT LOOPS This special test exception is required to perform certain STARTUP and PHYSICS TESTS under no flow conditions. 3/4.10.5 POSITION INDICATION SYSTEM - SHUTDOWN This special test exception permits the Digital Rod Position Indicator (s) to be inoperable during rod drop time measurements. The exception is required since the data necessary to determine the rod drop time are derived from the induced voltage in the position indicator coils as the rod is dropped. This induced voltage is small compared to the normal voltage and, therefore, cannot be observed if the Digital Rod Position Indicator (s) remain OPERABLE. The exception to the requirement for the Digital Rod Position Indicator to be OPERABLE during the withdrawal of the rods for the initial calibration of the position indication system is required because the OPERABILITY of the Digital Rod Position Indication System can only be determined by withdraw,ing the , control rod. The limitation on Keff during this evolution provides the necessary assurance that inadvertent critically will be avoided. COMANCHE PEAK - UNIT 1 B 3/4 10-1

i

{.,Q 3/4.11 RADIOACTIVE EFFLUENTS ylLi BASES 3/4.11.1 LIQUID EFFLUENTS 3/4.11.1.1 CONCENTRATION This specification is provided to ensure that the concentration of radio-active materials released in liquid waste effluents to UNRESTRICTED AREAS will be less than the concentration levels specified in 10 CFR 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 10 CFR 50 Appendix I, to a MEMBER OF THE PUBLIC, and (2) the limits of 10 CFR Part 20.106(e) to the population. The concentratio.n limit for dissolved or entrained noble gases is based upon the assumption that Xe-135 is the con-trolling radioisotope rid 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. This specification app'l'iss 'to the re' lease of radioactive materials in liquid effluents from all units at the site. The required detection capabilities for radioactive materials in liquid

                      . waste samples are tabulated in terms of the lower limits of detection (LLDs).

Detailed discussion of the LLD, and other detection limits can be found in Currie, L. A., "Lower Limit of Detection: Definition and Elaboration of a - Proposed Position for Radiological. Effluent and Environmental Measurements," NUREG/CR-4077 (September 1984), and in the HASL Procedures Manual, HASL-300 . (revised annually). 3/4.11.1.2 OOSE This specification is provided to implement the requirements of Sections II.A, III.A and IV.A of 10 CFR 50 Appendix I. 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 ths releases of radioactive material in liquid effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." Also, for fresh water sites with drinking water supplies that can be potentially affected by plant operations, 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 the requirements of 40 CFR 141. The dose calculation methodology and parameters in the 00CM 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 expo-sure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the 00CM for calcu-lating the doses due to the actual release rates of radioactive materials in liqu'id effluents are consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of COMANCHE PEAK - UNIT 1 B 3/4 11-1 L

h k RADIOACTIVE EFFLUENTS BASES DOSE (Continued) Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR 50, Appendix I," Revision 1, October.1977 and Regulatory Guide 1.113, "Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Im~plementing Appendix I," April 1977. This specification applies to the release of radioactive materials in liquid effluents from each unit at the site. For units with snared Radwaste Systems, the liquid effluents from the shared system are to be proportional among the units sharing that system. - - 3.4.11.1.3 LIQUID RADIOWASTE TREATMENT SYSTEM The OPERABILITY of the Liquid Rarwaste Treatment System ensures that this . system will be available for use whersver liquid effluents require treatment prior to release to the environment. The requirement that the appropriate ; portions of this system be used when specified provides assurance that the i releases of radioactive materials in liquid effluents will be kept "as low as ' is reasonably achievable." This' specification implements the requirement $ of 10 CCR 50.36a, General Design Criterion 60 of 10 CFR 50 Appendix A and the design objective given in Section II.D of 10 CFR 50 Appendix I. The specified limits governing the use of appropriate portions of the Liquid Radwaste Treat-ment Sys*em were specified as.a suitable fraction of the dose design objectives set fort, in Section II.A of 10 CFR 50 Appendix I for liquid effluents.

            . This specification applies to the release'of radioactive materials in liquid effluents from each unit at the site.       For units with shared Radwaste     .

Systems, the liquid effluents from the shared system are to be proportioned among the units sharing that system. , 3/4.11.1.4 LIQUID HOLOUP TANKS The tanks listed i.n this specification include all those outdoor tanks both permanent and temporary that are not surrounded by liners, dikes, or walls capable of holding the tank contents and that do not have tank over-flows and surrounding area drains connected to the Liquid Radwaste Treatment System. Restricting the quantity of radioactive material contained in the speci-fied tanks provides assurance that in the event of an uncontrolled release of the tank's contents, the resulting concentrations would be less than the limits of 10 CFR 20, Appendix B, Table II, Column 2, at the nearest potable water supply and the nearest surface water supply in an UNRESTRICTED AREA. G 'l e COMANCHE PEAK - UNIT 1 B 3/4 11-2

_RADICACTIVE EFFLUENTS BASES 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 Exclusion Area B0UNDARY (EAB) from gaseous effluents from all units on the site will be within the annual dose limits of 10 CFR Part 20 to UNRESTRICTED AREAS. The annual dose limits are the doses associated with the concentrctions of 10 CFR 20, Appendix B, Table II, Column I. These limits provide reasonable assurance ~that radioactive material discharged in gaseous effluents will not result in the exposure of a MEMBER OF THE PUBLIC in an UNRESTRICTED AREA, -either within or outside the Eb, to annual average concen-trations exceeding the limits specified in Table I'. of 10 CFR Part 20 Appendix B

                                                                  '(10 CFR 20.106(b)). For MEMBERS OF THE PUBLIC who tray at times be within the EAB, the occupancy of that MEMDER OF THE PUBLIC will usually be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the EAB. The methodology of calculating doscs for such MEMBERS OF THE PUBLIC, .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 EAB to less than or equal to 500 mrems/ year to the whole body or to less than or equal to 3000.mrems/ year to the skin. These release rate limits also restrict, at all times, the cor-responding 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 radioactive materials in ' gaseous effluents from all units at the site. Activities unrelated to plant operation which may be permitted within the Exclusion Area it,clude the exercising of mineral rights and the maintenance of pipelines. The Applicants will have the necessary control to determine these ' activities and will require that all persons involved in them report to the CPSES Manager, Plant Operations or his designated representative prior to engaging in the activities. Publication recreational activities within the Exclusion Area are limited to Squaw Creek Reservoir and Squaw Creek Park. Appropriate and effective arrangements have been made (in coordination with the appropriate agen:ies) to control' access to, activities on, and the removal of persons and property from the reservoir in case of emergency. Arrangements for recreational use and emergency procedures governing such use have been completed. The Applicants have the authority to exclude or remove any person from this area at any time. , The required detection capabilities for radioactive material in gaseous waste samples are tabulated in terms of the lower limits of detection (LL0s). Detailed discussion of the LLD, and other detection limits can be found in Currie, L. A., "Lower Limit of Detection: Definition and Elaboration of a Proposed Position for Radiological Effluent and Environmental Measurements," NUREG/CR-4077 (September 1984), and in the HASL Procedures Manual, HASL-300 (revised annually). COMANCHE PEAK - UNIT 1 B 3/4 11-3

m v d 4 RADIOACTIVE EFFLUENTS BASES , DOSE-NOBLE GASES (Continued) 3/4.11.2.2 DOSE - NOBLE GASES This specification'is provided to implement the requirements ot Sections II.B. III.A and IV.A of 10 CFR 50 Appendix I. The Limiting Condition for Operation implements the guides set forth in Section I.B 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 to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." The Survuillance 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 under-estimated. The dose calculation methodology and parameters 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, "Revision I, October 1977 and Regulatory Guide 1.111 "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Efflu-ents in Routine Releases from Light-Water Cooled Reactors," levision 1, July 1977. The ODCM equations provided for determining the air deses at and beyond the EAB are based upon the' historical average atmospheric conoitions. ' This specification applies to the release of radioactive materials in gaseous effluents from each unit at the site. Since both units share the radwaste treatment systems, the gaseous effluents are proportioned among the units. 3/4.11.2.3 DOSE - 10 DINE-131, 10 DINE-133, TRITIUM, AND RADIOACTIVE MATERIAL IN PARTICULATE FORM This specification is provided to implement the requirements of Sections II.C, III.A and IV.A of 10 CFR Part 50 Appendix I. The Limiting Conditions 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 releases of radioactive materials in gaseous effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." The ODCM calculational Sethods specified in 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 dat,a such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The ODCM calculational methodology and parameters for calculating the doses due to the actual release rates of the subject materials are consistent with the methodology provided in Reguiatory 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 Guide 1.111. "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine r COMANCHE PEAK - UNIT 1 B 3/4 11-4

                                                                 'V RADI0 ACTIVE EFFLUENTS                                                        Ithlt a BASES DOSE - 10 DINE-131, IODINE-133, TRITIUM, AND RADIOACTIVE MATERIAL IN PARTICULATE FORM (Continued)

Releases from Lig'ht-Water-Cooled React' ors," Revision 1, July 1977. Theseeqpa-tions also provide-for determining the actual doses based upon the historical average atmospheric conditions. The release rate specifications for Iodine-131 Iodine-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days are dependent upon the existing radionuclide pathways to man in the areas at and beyond the EAB. The pathways that were examined in the development of the calculations were: (1) individual inhalation of airborne radionuclides, (2) deposition of radionuclides onto green leefy vegetation 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. This specification applies to the release of radioactive materials in

         ' gaseous, effluents from each unit at the site. For units with shared radweste treatment systems, the gesecus effluents from the shared system are proportioned among the units sharing that system.

3/4.11.2.4- GASEOUS RADWASTE TREATMENT SYSTEM The OPERABILITY of the GASEOUS WASTE PROCESSING SYSTEM and the PRIMARY

       . PLANT VENTILATION SYSTEM ensures that the systems will be available for use whenever' gaseous effluents require treatment prior to release to the environ-ment. The requirement that the appropriate portions of those systems be used, when specified, provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achiev-able." This specification implements the requirements of 10 CFR 50.36a, General Design Criterion 60 of 10 CFR Part 50 Appendix A and the design objec-tives given in Section II.D of 10 CFR Part 50 Appendix I. 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 Sections II.B and II.C of 10 CFR Part 50 Appendix I, for gaseous effluents.

This specification applies to the release of radioactive materials in gaseous effluents from each unit at the site. For units with shared radwaste treatment systems, the gaseous effluents from the shared rystem are propor-tioned among the units sharing that system. i C0HANCHE PEAK - UNIT 1 B 3/4 11-5 1

RADI0 ACTIVE EFFLUENTS BASES 3/4.11.2.5 EXPLOSIVE GAS MIXTURE This specification is provided to ensure that the concentration of poten-tially explosive' gas mixtures contained in the WASTE GAS HOLDUP' SYSTEM is - maintained below the flammability limits of hydrogen an'd 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. Maintaining the concentration of hydrogen and oxygen below their flammability limits provides assurance that the releases of radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of 10 CFR Part 50 Appendix A. 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 er.ctner Technical Specification. Restricting the quantity of radioacti'ity v cantaired in each gas storage tank provides assurance that in the event of an uncontrolled release of the tank's contents, the resulting whole body exposure to a MEMBER OF THE PUBLIC at the nearest EAB will not exceed 0.5 rem. This is consistent with Standard Review. Plan 1.1.3, Branch Technical Position ETSB 11-5, "Postulated Radioactive Releases Due to a Waste Gas System Leak or Failure," in NUREG-0800, July 1981. . 3/4.11.3 SOLID RADIOACTIVE WASTES This specification implements the requirements of 10 CFR 50.36a and General Design Criterion 60 of 10 CFR 50 Appendix A. The process parameters included in establishing the PROCESS CONTROL PROGRAM may include, but are not limiu d - to, waste type, waste pH, waste / liquid / SOLIDIFICATION agent /catalylt ratios, L vaste oil content, waste principal chemical constituents, and mixing and curing l times. 3/4.11.4 TOTAL OOSE i This specification is provided to meet the dose limitations of 10 CFR 190 that have been incorporated into 10 CFR'20 by 46 FR 1.8525. The specification requires the preparation and submittal of a Special Report whenever the calcu-lated doses due to releases of radioactivity and to radiation from uranium fuel cycle sources exceed 25 mrems to the whole body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrems. For sites containing up to four reactors, it is highly unlikely that the resultant dose to a MEMBER OF THE PUBLIC will exceed the dose. limits of 40 CFR 190 if the individual reactors remain within twice the dose design' objectives of Appendix I, and if direct radiation doses from the units (including outside storage tanks, ' etc.) are kept small. The Special Report will describe a course of action +1at should result in the limitation of the annual dose to a MEMBER OF THE PUBL45 to within the 40 CFR 190 limits. For the 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 contri- - butions from other nuclear fuel cycle facilities at the same site or within a COMANCHE PEAK - UNIT 1 B 3/4 11-6

E RADI0 ACTIVE EFFLUENTS BASES TOTAL DOSE (Continued) radius of 8 km must be considered. If.the dose to any MEMBER OF THE PUBLIC is estimated to exceed the requirements of 40 CFR 190, the Special leport with a request for a variance (-provided the release conditions.resultin) in violation of 40 CFR 190 have not already be in corrected), in accordance with the provi-sions of 40 CFR 190.11 and 10 CFR 20.405c, is considered to be a timely request' and fulfills the requirements of 40 CFR 190 until NRC staff action is completed. The variance only relates to the limits of 40 CFR 190, and does not apply in any way to the other requirements for dose limitation of 10 CFR 20, as addressed in Specifications 3.11.1.1 and 3.11.2.1. An individual 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. O o 1 COMANCHE PEAK - UNIT 1 B 3/4 11-7

3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING BASES 3/4.12.1 MONITORING PROGRAM The Radiological Environmental Monitoring Program required by this

           . specification provides representative measurements.of radiation and of radio-active materials in those exposure pathways and for those radionuclides that.

lead to the highest potential radiation exposure of MEMBERS OF THE PUBLIC resulting from the p' ant operation. This monitoring program implements Section 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. Guidance for this monitoring program is provided by the Radiological Assessment Branch Technical Position on Environ-mental Monitoring, Revision 1, November 1979. The initially specified monitoring program will be effective for at least the first 3 years of commercial operation. Following this period, program changes may be initiated based on operational

                                                                                                             ~
        ' experience.

The required detection capabilities for environmental sample analyses are tabulated in terms of the lower limits of detection (LL0s). The LLDs required by Table 4.12-1 are considered optimum for routine environmental measurements in industrial laboratorie's. It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measure-ment system and ho,t as an a pesteriori (after the fact) limit for a particular measurement. Detailed discussion of the LLD, and other' detection limits, can be found in Currie, L. A., "Lower Limit of Detection: Definition and Elaboration of a Proposed Position for Radiological Effluent and Environmental Measurements," NUREG/CR-4007 (September 1984), and in the HASL Procedures Manual, @SL-300 (revised annually). 3/4.12.2 LAND USE CENSUS This specification is provided to ensure that changes in the use of are'as at aad beyond the EXCLUSION AREA BOUNDARY are identified and that modifica- . tions to the Radiological Environmental Monitoring Program are made if required by the results of this census. The best information from the door-to-door survey, from aerial survey or from consulting with local agricultural authori-ties shall be used. This census satisfies the requirements of Section IV.B.3 of 10 CFR Part 50 Appendix I. Restricting the census to gardens of greater than 50~m2 provides assurance that significant exposure pathways via leafy vegetables will be identified and monitored since a garden of this size is t,he 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) a vegetation yield of 2 kg/m2, COMANCHE PEAK - UNIT 1 B 3/4 12-1 I

I RADIOLOGICAL ENVIRONMENTAL MONITORING BASES 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM t The requirement for participation in an approved Interlaboratory Comparison ! Program is provided to ensure that independent checks,on the precision and I accuracy of the measurements of radioactive materials in environmental sample . matrices are performed as part of the quality assurance program for environmental l monitoring in order to demonstrate that the results are valid for the purposes of Section IV.B.2 of 10 CFR Part 50 Appendix I. t 4 F t i il 4 t [ I COMANCHE PEAK - UNIT 1 B 3/4 12-2

I ' f SECTION 5.0 DESIGt4 FEATURES 4

5.0 DESIGN FEATURES N - 5.1 SITE EXCLUSION AREA . t 5.1.1 The Exclusion Area shall be as 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. MAP DEFINING UNRESTRICTED AREAS AND EXCLUSION AREA BOUNDARY FOR RADI0 ACTIVE . GA5EQU5 AND LIQUID EFFLUENTS 5.1.3 Information regarding radioactive gaseous and liquid effluents, which will allow identification of structures and release points as well as defini-tion of UNRESTRICTED AREAS within the EXCLUSION AREA BOUNDARY that are acces-sible to MEMBERS OF THE PUBLIC, shall be as shown in Figures 5.1-3 and 5.1-4. + The definition of UNRESTRICTED AREA used in implementing these Technical Specifications has been expanded over that in 10 CFR 20.'3(a)(17); The UNRESTRICTED AREA boundary may coincide with the EXCLUSION AREA BOUNDARY, as , defined in 10 CFR 100.3(a), but the UNRESTRICTED AREA does not include areas ' over water bodies. The concept of UNRESTRICTED AREAS, established at or beyond the. EXCLUSION AREA BOUNDARY, is utilized in the Limiting Conditions fer ! Operation to keep levels of radioactive materials in liquid and gaseous ;

   ,                                                     effluents as low as is reasonably achievable,. pursuant to 10 CFR 50.36a.                                                                       .

5.2 CONTAINMENT CONFIGURATION - 5.2.1 The 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 = 135 feet. l

b. Nominal inside height = 192.5 feet. (Done 67.5 feet; total =

260 feet)

c. Minimum thickness of concrete walls = 4.5 feet.

d. Minimum thickness of concrete roof.= 2.5 feet,

e. Minimum thickness of concrete floor pad = 12.0 feet,

f. Nominal thickness of steel liner wall = 3/8 inches. (Dome = 1/2 inch, Base Hat = 1/4 inch), and

g. Net free volume = 2,985,000 cubic feet.

! DESIGN PRESSURE AND TEMPERATURE a I . i 5.2.2 The containment building is designed and shall be maintained for a '

maximum internal pressure of E0 psig and a temperature of 280'F.

! COMANCHE PEAK - UNIT 1 5-1 l

O 4 I d 0 9 9 9 0 FIGURE 54 1-1 EXCLU510N AREA COMANCHE PEAK - UNIT 1 5-2 9

i , i FIGURE 5,1-2 LOW PGPULATION ZONE COMANCHE PEAK - UNIT 1 53

DI3kI"l 4 9 4 S FIGURE 5.1-3 UNRESTRICTED AREA AND EXCLUSION AREA BOUNDARY FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS COMANCHE PEAK - UNIT 1 5-4

DESIGN FEATURES 5.3 REACTOR CORE . FUEL ASSEMBLIES 5.3.1 The 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 144 inches. .The initial core loading shall have a maximum enrichment not to exceed 3.15 weight percent U-235. Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum enrichment not to exceed 3.5 weight percent U-235. CONTROL ROD ASSEMBLIES 5.3.2 The core shall contain 53 full-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 95.5% hafnium with the remainder zirconium. 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 accordanc'e with the Code requirements specified in Sec. tion [5.2) of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements,

b. For a pressure of 2,485 psig, and

c. For a temperature of 650*F, except for the pressurizer which is 680 F.

VOLUME 5.4.2 The total water and steam volume of the Reactor Coolant System is 12,500 + 100 cubic feet at a nominal T,yg of 589.5*F. 5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The primary meteorological tower shall be located as shown on Figure 5.1-1. O COMANCHE PEAK - UNIT 1 5-5

   ')\                                                                      '

l ~. DESIGN FEATURES dA i 5.6 FUEL STORAGE CRITICALITY 5.6.1.1 The spent fuel storage racks are' designed and shall be maintained .

                  . with:                                                                           ,                                         ,

a. A k,ff equivalent to less than or equal to 0.95 when flooded with .

unb' orated water, which includes a conservative allowance of 2.6% ok/k for uncertainties as described in Section 4.3 of the FSAR, and i b. A nominal 16 inch center-to-center distance between fuel assemblies ,, placed in the storage racks. 5.6.1.2 The k,ff for new fuel for the first core loading stored dry in the spent fuel storage racks shall not exceed 0.98 when aqueous foarr. moderation is assumed. 1 ORAINAGE 5.6.2 The spent fuel storage pool is designed and shall be maintained to > prevent inadvertent draining of the pool below elevation 854 feet. ', CAPACITY 5.6.3 The spent fue'l st'orage pool is designed and shall be maintained with a I storage capacity limit.ed to no more than 260 fuel assemblies, i ,

5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT i >

i 5.7.1 The components identified in Table 5.7-1 are designed and shall be , l maintained within the cyclic or transient limits of Table 5.7 1. t I t i I > l ; i I 1 COMANCHE PEAK - UNIT 1 5-6

l

TABLE 5.7-1 n h x COMPONENT CYCLIC OR TRANSIENT' LIMITS

1 CYCLIC OR DESIGN CYCLE y COMPONENT TRANSIENf LIMIT OR TRANSIENT 7c

Reactor Coolant System 200 heatup cycles at < 100*F/h Heatup cycle - T g and 200 cooldown cycles at to > 550*F. **9 from 1 200*F

                                            ~< 100*F/h.                                              CooTdown cycle - T      from

[ > 550*F to $ 200*F**9 - 200 pressurizer cooldown cycles Pressurizer cooldown cyc1'e at $ 200*F/h. temperatures from > 650*F to

                                                                                                    < 200*F.

80 loss of load cycles, without > 15% of RATED' THERMAL POWER to immediate Turbine or Reactor trip. D% of RATED THERMAL POWER.

{ 40 cycles of loss-of-offsite loss-of-offsite A.C. electrical A.C. electrical power. ESF Electrical System.- 80 cycles of loss of flow in one loss of only one reactor reactor coolant loop. - coolant pump. 400 Reactor trip cycles. 100% to 0% of RATED THERMAL POWER. 10 auxiliary spray Spray water temperature differential actuation cycles. > 320*F, but $ 625'F.

                                           ,200 leak tests.                                         Pressurized to > 2485 psig.

10 hydrostatic pressure tests. Pressurized to > 3107 psig.

Secondary Coolant System I steam line break. , Break in a > 6-inch steam line.

               *                                                                      ~

10 hydrostatic pressure tests. Pressurized to > 1481 psig. N

                                                                                                                           .             b

4 m h[Lild i 4 SECTION 6.0 ADMINISTRATIVE CONTROLS 4

ADMINISTRATIVE CONTROLS . Q 6.1 RESPONSIBILITY L.1.1 The Vice President, Nuclear Operations shall be responsible for overall operation of the site, while the Manager, Plant Operations shall be responsible for operation of the unit. The Vice President, Nuclear Operations and Manager, Plant Operations shall each delegate in writing the succession to this respon-sibility during his absence. - 6.1.2 The Shift Supervisor (or during his absence from the control room, a designated individual) shall be responsible for the control room command function. A management directive to this effect, signed by the (highest level of corporate management) shall be reissued to all station personnel on an annual basis. 6.2 ORGANIZATION OFFSITE 6.2.1 The corporate organization for unit management and technical support shall be as snown in Figure 6.2-1. UNIT STAFF 6.2.2 The unit organization shall be as shown in 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 shal'1 be in the control room when i

fuel is in the reactor. ~In addition, while the unit is in MODE 1, i 2, 3, or 4, at least one licensed Senior Operator shall be in the control room;

c. A Radiation Protection Technician

shall be on site when fuel is in the reactor;

d. All CORE ALTERATIONS shall be observed and directly supervised by either a licensed Senior Oparator or licensed Senior Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation;

e. A site Fire brigade of at least five members

shall be maintained on site at all tin.25. The Fire Brigade shall not include the Shift Supervisor and the [two] other members of the minimum shift crew necessary for safe shutdown of the unit and any personnel required for other essential functions during a fire emergency; and

     *The Radiation Protection Technician and Fire Brigade composition may be'less than the minimum requirements for a period of time not to exceed 2 hours, in order to accommodate unexpected absence, provided immediate action is taken to fill the required positions.

COMANCHE PEAK - UNIT 1 6-1

DRAFT ADMINISTRATIVE CONTROLS I q , I UNIT STAFF (Continued)

f. Administrative procedures shall.be developed and implemented to

.                          limit the workin i

functions (e.g.,g licensed' hours of Senior unit staff who perform Operators, licensed safety-related Operators, i Radiation Protection Technicians, auxiliary' operators, and key ; maintenance personnel).

The amount of overtime worked by unit staff members performing safety-related functions shall be limited in accordance with the NRC ;

Policy Statement on working hours (Generic Letter No. 82-12). ' i ! 4 1 i

j i ,

l i ! . i i ; ! I j COMANCHE PEAK - UNIT 1 6-2 l i j f , . F

9 I e 4 L 4 FIGURE 6.2-1 0FFSITE ORGANIZATION COMANCHE PEAK - UNIT 1 6-3

DR!!FI l FIGURE 6.2-2 UN;7 ORGANIZATION i l

COMANCHE PEAK - UNIT 1 6-4

TARLE 6.2-1 MINIMUM SHIFT CREW COMPOSITION . SINGLE UNIT FACILITY POSITION NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION MODE 1, 2, 3, or 4 MODE 5 or 6 SS 1 1 SRO 1 None R0 2 1 A0 2 1 STA 1* None SS - Shiit Supervisor vith e. Senior Operator license on Unit 1 SR0 - Individual with a Senior Operatur 4fcense. pn Unit 1 R0 - Individual with an Operator license on Unit 1 _ A0' - Auxiliary Operator STA - Shift Technical Advisor The shift crew composition may be orie less than the minimum requirements of Table 6.2-1 for a period of time not to exceed 2 hours in order to accommodate unexpected absence of on-duty shift 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 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 with a valid Senior Operator 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 Senior Operator license or Operator license shall be designated to assume the control room command function.

     *The STA position shall be manned in MODES 1, 2, 3, and 4 unless the Shift Supervisor or the individual with a Senior Operatoe license meets the qualifications for the STA as required by the NRC.

l COMANCHE PEAK - UNIT 1 6-5 i

ADMINISTRATIVE CONTROLS r3 g 6.2.3 UtmlL INDEPFNDENT SAFETY ENGINEERING GROUP (ISEG) FUNCTION 6.2.3.1 The ISEG shall function to examine unit operating characteristics, NRC issuances, industry advisories, Licensee Event Reports, and other sources - of unit design and operating experien~ce information, including units of similar design, which may indicate areas for improving unit safety. The ISEG shall make detailed recommendations for revised procedures, equipment modifi-cations, maintenance activities, operat. ions activities, or other means of improving unit safety to the Vice President, Nuclear Operations. COMPOSITION 6.2.3.2 The ISEG shall be composed of at least five, dedicated, full-time

 -engineers located on site. Each shall have a bachelor's degree in engineering or related science and at least 2 years professional level experience in his field, at least 1 yea, of which experience siall be in the nuclear field.

RESPONSIBILITIES 6.2.3.3 The ISEG shall be responsible for maintaining surveillance of unit activities to provide independent verification

that these activities are performed correctly and that human errors are r.duced as much as practical.

RECORDS 6.2.3.4 Records of acti'vities performed by the ISEG shall be prepared, main-tained, and forwarded each calendar month to Vice President, Nuclear Operations. 6.2.4 SHIFT TECHNICAL ADVISOR 6.2.4.1 The Shift Technical. Advisor (STA) shall provide advisory technical support to the Shift Supervisor in the areas of thermal hydraulics, reactor engineering, and plant a.ialysis'with regard to the safe operation of the unit. The Shift Technical Advisor shall have a bachelor's degree or equivalent in 7 scientific or engineering discipline and shall have received specific training in the response and analysis of the unit for transients and accidents, and in unit design and layout, including the capabilities of instrumentation and controls in the control room. 6.3 UNIT STAFF QUALIFICATIONS 6.3.1 Each member of the unit staff shall meet or exceed the minimum quali-fications of ANSI-N18.1-1971 for comparable positions, except for the Radia-tion Protection Manager ** who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975, for a Radiation Protection Manager. The licensed Operators and Senior Operators shall also meet or exceed the minimum qualifications of the supplemental requirements specified in Sections A and C of Enclosure 1 of the March 28, 1980 NRC letter to all licensees.

*Not responsible for sign-off function.

 **Until the Radiation Protection Manager meets all qualification per R.G.1.8, l     September 1975, an individual who meets all those qualifications shall

support the Radiaticn Protection Manager. ,

COMANCHE PEAK - UNIT 1 6-6

BER ADMINISTRATIVE CONTROLS UNIT STAFF QUALIFICATIONS (Cortinued) 6.4 TRAINING 6.4.1 A retraining and replacement training program for the unit staff shall be maintained under the direction of- the Vice President,' Nu' clear Operations , and shall meet or exceed the requirements and recommendations of ANSI-N18.1-1971 and Appendix A of 10 CFR 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 opera-tional experience. _ 6.5 REVIEW AND AUDIT 6.5.1 STATION OPERATIONS REVIEW COMMITTEE (SORC) FUNCTION 6.5.1.1 The SORC shall function to advise the Vice President, Nuclear Operations on all. matters rel.ated to nuclear safety. COMPOSITION 6.5.1.2 The 50RC shall be composed of the: Chairman: V'ce President, Nuclear Operations Member: [0perations Supervisor] Members to be . Member: [ Technical Supervisor] equivalent to .

               . Member:            [ Maintenance Supervisor].                  these posi-Member:           [ Plant Instrument and Control Engineer]    tions, but Member:           [ Plant Nuclear Engineer]                   of high        -

Member: [ Health Physicist] management position. 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 50RC activities at any one time.

         **Until the Radiation Protection Manager meets all qualification per R.G.1.8, September 1975, and individual who meets all those qualifications shall
          . support the Radiation Protection Manager.

l l COMANCHE PEAK - UNIT 1 6-7

ADMINISTRATIVE CONTROLS b I* i MEETING FREQUENCY ; 6.5.1.4 The 50RC shall meet at least once per calendar month and as convened by the 50RC Chairman or his designated alternate. QUORUM 6.5.1.5 The quorum of the SORC necessary for the performance of the 50RC responsibility and authority provisions of these Technical Specifications shall consist of the Chairman or his designated alternate and four members _ including altern.ites.

   -RESPONSIBILITIES                                                                    -

6.5. 1.6 The SORC shall be responsible for: .

a. Review of all Station Administrative Procedures;

b. Review of the safety evaluations for: (1) procedures, (2) change to procedures, equipment, systems or facilities, and (3) tests or experiments completed under the provision of 10 CFR 50.59 to verify that such actions did not constitute an unreviewed safety question;

c. Review of proposed procedures and' changes to procedures,. equipment, systems or. facilities which may involve an unreviewed safety ques- .

tion as defined in 10 CFR 50.59 or. involves a. change in Technical Specifications;

d. Review of proposed test or experiments which may involve an unreviewed safety question as defined in 10 CFR 50.59 or requires a change in Technical Specifications;

e. Review of proposed' changes to Technical Specifications or the Operating License;

f. Investigation of all violations of the Technical Specifications including the forwarding of reports covering evaluation and recom-mendations to prevent recurrence to.the Vice President, Nuclear Operations and to the ORC;

g. Review of reports of operating abnormalities, deviations from ex-pected performance of plant equipment and of unanticipated defici-encies in the design or operation of structures, systems or components that affect nuclear safety;

h. Review of all REPORTABLE EVENTS;

i. Review of the Security Plan and shall submit recommended changes to the ORC; 4

COMANCHE PEAK - UNIT 1 6-8

ADMINISTRATIVE CONTROLS RESPONSIBILITIES (Continued)

j. Review of the Emergency Plan and shall submit recommended changes to the ORC;

k. Review of changes,to the PROCESS CONTROL PROGRAM, the 0FFSITE. DOSE CALCULATION MANUAL, and Radwaste Treatment Systems;

1. Review of anya'ccidental, unplanned or uncontrolled radioactive release 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 Operations, and to the ORC;

m. Review of Unit operations to detect potential hazards to nuclear safety; and

n. Investigations or analysis of special subjects as requested by the Chairman of tho ORC or the Vice President, Nuclear Operations.

o Review of the Fire Protection Program and shall submit recommended changes to the ORC. 6.5.1.7 The SORC shall:

a. Recommend'in writing to the Vice. P. resident,. Nuclear Operations approval

       .                  or dirapproval of items considered under Specification 6.5.1.6a.

through e, i, j, k, and 1 above, prior to their implementation;

b. Render determinations in writing with regard to whether or not each item considered under Specification 6.5.1.6a. through e. and m.

constitutes an unreviewed safety question; and

c. Provide written notification within 24 hours to the Executive Vice President-Nuclear Engineering and Operations and the Operations Review Committee of disagreement between the SORC and the Vice President, Nuclear Operations; however, the Vice President, Nuclear Operations shall have responsibility for resolution of such dis-agreements pursuant to Specification 6.1.1.

RECORDS 6.5.1.8 The SORC shall maintain written minutes of each SORC meeting that, at a minimum, document the results of all SORC activities performed under the responsibility provisions of these Technical Specifications. Copies shall be provi'ded to the Vice President-Nuclear Operations and the Operations Review ! Committee. l l COMANCHE PEAK - UNIT 1 6-9

ADMINISTRATIVE CONTROLS 6.5.2 OPERATIONS REVIEW COMMITTEE (ORC) FUNCTION 6.5.2.1 The ORC shall function to provide independent _ 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 and electrical engineering, and

h. Quality ass.urance practices.

The ORC shall report to and advise the Executive Vice President, Nuclear Engineering and Operations on those areas of responsibility specified in Specifications 6.5.2.7 and 6.5.2.8. COMPOSITION 6.5.2.2 The ORC shall be composed of at least five* individuals of whom no more than minority are members having line responsibility for operations at CPSES. The Chairman and all members will be appointed by the Executive Vice President, Nuclear Engineering and Operations. The ORC members shall hold a Bachelor's degree in an engineering or physical science field or equivalent experience and a minimum of 5 years technical experience. It is the responsibility of the Chairman to ensure experience and competence is available to review problems in areas listed in Specification 6.5.2.la. through h. To a large measure, this experience end competence rests with the membership of the ORC. In specialized areas, this experience may be provided by personnel who act as consultants to the ORC, ALTERNATE _S I 6.5.2 J The Alterante for the Chairman and all alternate members shall be appv..'+2.d in writing by the Executive Vice President, Nuclear Engineering and Operations to serve on a temporary basis; however, no more than two alter- . nates shall participate as voting members in ORC activities at any one time. l t COMANCHE PEAK - UNIT 1 6-10

p' q & ADMINISTRATlVE CONTROLS U3

QUORUM CONSULTANTS 6.5.2.4. Consul'tants shall be utilized as determined by the Chairman, ORC to provid.e expert advice to the ORC'. MEETING FREQUENCY 6.5.2.5 The ORC shall meet at least once per calendar quarter during the initial year of unit operation following fuel loading and at least once per 6 months thereafter. QUORUM , 6.5.2.6 The quorum of the ORC necessary for the performance of the ORC review and audit functions of these Technical Specifications shall consist of not less than a majority of the appointed individuals (or their alternates) and

 . -the Chairman or.his designated alternate. No more than a minority of the quorum shall have line responsibility for operation of the unit.

REVIEW 6.5.2.7 The ORC shall be responsible for the review of:

a' . The safety evaluations for: (1) changes to ~ procedures, equipment, or systems; and (2) tests or experiments completed under the provision of 10 CFR 50.59, 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 10 CFR 50.59;

c. Proposed tests or experiments which involve an unreviewed safety question as defined in 10 CFR 50.59;

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 instructions having nuclear safety significance;

f. Significant operating abnormalities or deviations from normal and expected performance of unit equipment that affect nuclear safety;

g. All' REPORTABLE EVENTS;

h. All recognized indications of an unanticipated deficiency in some aspec't of design or operation of structures, systems, or components that could affect nuclear safety; and

1. Reports and meeting minutes of the SORC.

COMANCHE PEAK - UNIT 1 6-11

                                                                             .1 ADMINISTRATIVE CONTROLS                                                           3E AUDITS                                                          .

6. 5. 2. 8 Audits of unit activities shall be performed under the cognizance of the ORC. The audits shall be performed within the specified time interval with a maximum allowable extension not to exceed 25% of the specified interval pr'ovided the c'ombined time interval for any three consecutiv6 inter'vals shal'l not exceed 3.25 times the specified interval. These audits shall encompass:

a. The conformance of unit operation to provisions contained within the Technical Specifications and applicable license conditions at least once per 12 months;; b. The performance, training, and qualifications of the entire unit staff at least once per 12 months;; c. The results of actions taken to correct deficiencies occurring in unit equipment, structures, systems, or method of operation that affect nuclear safety, at least once per 6 months;; d. The performance of activities r'equired by the Operatiorial Quality Assurance Program to meet the criteria of Appendix B, 10 CFR 50, at least once per 24 months;; e. .The fire protection. programmatic controls including the implementing procedures at least once per'24 months by qualified licensee QA personnel; ,; f. The fire protection equipment and program implementation at least once per 12 months utilizing either a qualified offsite licensee fire protection engineer or an outside independent fire protection consultant. An outside independent fire protection consultant shall be used at least every third year;; g. The Radiological Environmental Monitoring Program and the results thereof at least once per 12 months;; h. The OFFSITE DOSE CALCULATION MANUAL and implementing procedures at least once per 24 months;; i. The PROCESS CONTROL PROGRAM and implementing procedures for processing i and packaging of radioactive wastes at least once per 24 months; I j. The performance of activities required by the Quality Assurance Program for effluent and environmental monitoring at least once per 12 months; and l
k. Any other area of unit operation considered appropriate by the

!- ORC or the Executive Vice President Nuclear Engineering and l Operations. - l COMANCHE PEAK - UNIT 1 6-12 I l i

W ADMINISTRATIVE CONTROLS MT RECORDS 6.5.2.9 Records of ORC activities shall be prepared, approved, and distribu-ted as indicated below:

 .  .       a. Minutes of each ORC meeting shall be prepared, approved, and for-warded to the Vice President, NLclear Operations and Lxecutive Vice President, Nuclear Engineering and Operations within 14 days following each meeting;

b. Reports of reviews encompassed by Specification 6.5.2.7 shall be prepared, approved, and forwarded to the Vice President Nuclear Operations and Executive Vice President, Nuclear Engineering and Operations within 14 days following completion of the review; and

c. Audit reports encompassed by Specification 6.5.2.8 shall be for-warded to the Vice President, Nuclear Operations and Executive Vice President, Nuclear Engineering and Operations and to the management positions responsible for the areas audited within 3.0 days after completion of the audit by the auditing organization.

6.5.3 TECHNICAL REVIEW AND CONTPOLS 6.5.3.1 Activities which affect nuclear safety shail be unducted as follows:

a. Procedures required by Specification ~6.8 and other procedures' which affect plant nuclear safety, and changes thereto, shall be prepared, reviewed and approved. Each such procedure or procedure change shall be reviewed by a qualified individual / group other than the individual / group which prepared the procedure or procedure change, but who may be from the same organization as the individual / group which prepared the procedure or procedure change. The Vice Presi-dent, Nuclear Operations, shall approve Station Administrative Procedures, Security Plan Implementing Procedures, and Emergency Plan Implementing Procedures. Other procedures shall be approved by the appropriate approval authority, as designated by the Vice Presi-dent, Nuclear Operations, in writing. Individuals responsible for procedure reviews shall.be members of the Nuclear Operations Staff previously designated by the Vice President, Nuclear Operations.
Changes to procedures which d,o not change the intent of approved t

procedures may be approved for implementation by two members of the l Nuclear Operations Staff, at least one of whom holds a Senior l Operator License, provided such approval is prior to implementation and is documented. Such changes shall be approved by the original approval authority within 14 days of implementation;

b. Proposed tests and experiments which affect plant nuclear safety and are not addressed in the Final Safety Analysis Report or-Technical Specifications shall be prepared, reviewed, and ,

COMANCHE PEAK - UNIT 1 6-13 l

ADMINISTRATIVE CONTROLS TECHNICAL REVIEW AND CONTROLS (Continued) approved. Each such test or experiment shall be reviewed by a qualified individual / group other than the individual / group which prepared the proposed test.or experiment. Proposed test and experi-ments shall be approved before implementation by the Manager, Plant Ope. rations. Individuals responsible for conducting such reviews. shall be members of the Nuclear Operations Staff previously designated by the Vice President, Nuclear Operations;

c. Proposed changes or modifications to plant nuclear safety-related structures, systems and components shall be reviewed as designated by the Vice President, Engineering and Crinstruction. Each such modification shall be reviewed by a qualified individual / group other than the individual / group which designed the modification, but who may be from the same organization as the individual / group which designed the modifications. Individuals / groups responsible for conducting such reviews shall be previously designated by the Vice President, Engineering and Construction. Proposed modifica--

tions to plant nuclear safety related structures, systems and components shall be approved by the Manager, Plant Operations prior to implementation;

d. Each review conducted in accordance with the requirements of Speci-fications 6.5.3.la, 6.5.3.lb, and 6.5.3.1c, shall include a deter-minati,on of whether or not additional cross-disciplinary review is necessary. If deemed necessary, such review shall be done in

               .accordance with the appropriate qualification requirements;

e. Each review shall include a determination of whether or not an unreviewed safety question is involved. Pursuant to NRC approval of items involving unreviewed safety questions shall be obtained prior to the Manager, Plant Operations, approval for implementation; and

f. The Security Plan and Emergency Plan, &nd implementing procedures, shall be reviewed at least once per 12 months. Recommended changes to the implementing procedures shall be approved by the Manager, .

Plant Operations. Recommended changes to the Plans shall be reviewed pursuant to the requirements of Specifications 6.5.1.6 and 6.5.2.8 and approved by the Manager, Plant Operations. NRC approval shall be obtained as appropriate. 6.6 REPORTABLE EVENT ACTION 6.6.1 The following actions shall be taken for REPORTABLE EVENTS:

a. The Commission shall be notified and a report submitted pursuant to the requirements of 10 CFR 50.73 and

, b. Each REPORTABLE EVENT shall be reviewed by the 50RC, and the results i of this review shall be submitted to the ORC and the Vice President Nuclear Operations. COMANCHE PEAK - UNIT 1 6-14

                                                                                 . . ..: a ADMINISTRATIVE CONTROLS 6.7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:

a. In accordance with 10 CFR 50.72,.the NRC Oper_ations Center, shall be notified by telephone as soon as practi. cal and in all cases within one hour after the violation has been determined.. The Vice President, Nuclear Operations and the Operations Review Committee (0RC) shall be notified within 24 hours.

b. A Licensee Event Report shall be prepared in accordance with 10 CFR 50.73.

c. The License Event Report shall be submitted to the Commission in accordance with 10 CFR 50.73, Vice President, Nuclear Operations and the Operations Review Committee (0RC) within 30 days after discovery of the event.

d. Critical operation of the unit shall not be resumed until authorized by the Nuclear Regulatory. Commission.

6.8 PROCEDURES AND PROGRAMS 6.8.1 Written procedures shall be established, implemented, and maintained covering the activities referenced below: -

a. The applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2,, February 1978; '.

b. The emergency operating procedures required to implement the require-ments of NUREG-0737 and Supplement 1 to NUREG-0737 as stated in Section 7.1 of Generic Letter No. 82-33;

c. Security Plan implementation;

d. Emergency Plan implementation;

e. PROCESS CONTROL PROGRAM implementation;

f. OFFSITE DOSE CALCULATION MANUAL implementation; and

g. Quality Assurance for effluent and environmental monitoring.

6.8.2 Each precedure of Specification 6.8.1, and changes thereto, shall be i reviewed by the 50RC and shall be approved by the Vice President, Nuclear ! Operations prior to implementation and reviewed periodically as set forth in administrative procedures. l 6.8.3 Temporary changes to procedures of Specification 6.8.1 may be made provided: t

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 Operator license on the unit affected; and l

COMANCHE PEAK - UNIT 1 6-15 1

l ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)

c. The change is documented, reviewed by the 50RC, and approved by the Vice President, Nuclear Operations within 14 days of implementation.

6.8.4 The following progiams shall be established; implemented, and 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 the Containment Spray System, Safety Injection System, Chemical and Volume Control System, RHR System, and RCS Sampling System. The program shall include the following:

1) Preventive maintenance and periodic visual inspection require-ments, and

2) 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 ai~rborne iodine concentration in vital areas under accident conditions. This program shall include the following: -

1) Trainit.g of personnel,

2) Procedures for monitoring, and

3) Provisions for maintenance of sampling and analysis equipment.

c. Secondary Water Chemistry A program for monitoring of secondary water chemistry to inhibit steam generator tube degradation and low pressure turbine disc stress corrosion cracking. This program shall include:

1) . Identification of a sampling schedule for the critical variables and control points for these variables,

2) Identification of the procedures used to measure the values of the critical variables,

3) Identification of process sampling points, which shall include monitoring the discharge of the condensate pumps for evidence of condenser in-leakage,

4) Procedures for the recording and management of data,

5) Procedures defining corrective actions for.all off-control point chemistry conditions, and COMANCHE PEAK - UNIT 1 .6-16

i ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)

6) A procedure identifying: (a) the authority responsible' for the t

interpretation of the data, and (b) the sequence and timing of administrative' events required to initiate co'rrective action.

d. Post-Accident Sampling A program which will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant gaseous effluents, and containment atmosphere samples under accident conditions. The' program shall include the following:

1) Training of personnel,

2) Procedures for sampling and analysis, and

3) Provisions for maintenance of sampling and analysis equipment.

6 4 e 4 9 . O COMANCHE PEAK - UNIT 1 6-17

w/ EJf. eJ a j ADMINISTRATIVE CONTROLS 6.9 REPORTING REQUIREMENTS ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code - of Federal Regulations, the following reports shall be submitted to the Regional Administrator of the Regional Office of the NRC unless otherwise noted. STARTUP REPORT

6. 9.1.1 be submitted A summary report of plant startup and power escalation testing shall 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 a different design or has been manufactured by a different fuel supplier,or thermal,                        and             (4) modifications hydraulic   performance thatofmay   have significantly altered the nuclear, the unit.

The initial Startup Report shall address each of the startup tests identi-fied in Ch. apter 14 of the Final Safety Analysis Report and shall inclu.de a description of the' measured values of the operating conditions or characteris . tics 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. Subsequent Startup Reports d

   .      shall    a' of changes.and/or   dress startup                        tests that are necessary to demonstrate the' acceptability modifications.

Startup Reports shall be submitted within: (1) 90 days following

completion of the Startup Test Program, (2) 90 days 'ollowing resumption or criticality, whichever is earliest. commencement of commercial power operation, or (3) 9 mon 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 operation), supplementary reports shall completed.be submitted at least every 3 months until all three events have been

     ,  ANNUAL REPORTS *

6. 9.1. 2 Annual Reports covering the activities of the unit as described below for year. the previous calendar year shall be submitted prior to March 1 of each following initial report The initial shall be submitted prior to March 1 of the year criticality.

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

       *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.

COMANCHE PEAX - UNIT 1 6-18

ADMINISTRATIVE CONTROLS ANNUAL REPORTS (Continued) greater than 100 mrem /yr and their associated man-rem exposure according to work and job functions

e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, an~d refueling.

The dose as"signments to various duty functions may be estimated based on pocket dosimeter, thermoluminescent dosimeter (TLD), or film badge measurements. Small exposures totalling less than 20% of the individual total dose need not be accounted for. In the aggragate, at least 80% of the total whole-body dose received from external sources should be assigned to specific major work functions;

b. The results of specific activity analys'es in which the primary coolant exceeded the limits of Specification 3.4.8. The following information shall be included: (1) Reactor power history starting 48 hours prior to the first sample in which the limit was exceeded (in graphic and tabular format); (2) Results of the last isotopic analysis for radio-iodine performed prior to exceeding the limit, results of analysis while limit was exceeded and results of one analysis after the radio-iodine activity was reduced to less than limit. Each result should include date and time of sampling and the radiciodine cencentrations; (3) Clean-up flow history starting 48 hours prior to the first sample in which the limit was exceeded; (4) Graph'of the I.-131 concentra-tion (pCi/gm) and one other radioidine isotope concentration (pCi/gm) as a function of time for the duration of the specific activity above the steady-state level; and (5) The time duration when the specific activity of the primary coolant exceeded the radioiodine limit.

ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT ** 6.9.1.3 Routine Annual Radiological Environmental Operating Reports covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year. The initial report shall be submitted prior to May 1 of the year following initial criticality and shall include copies of reports of the preoperational Radiological Environmental Monitoring Program of the unit for at least two years prior to initial criticality. 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 report period,

  *This tabulation supplements the requirements of $20.407 of 10 CFR Part 20.
 **A single submittal may be made for a multiple unit station.

COMANCHE PEAK - UNIT 1 6-19

ADMINfSTRATIVE CONTROLS ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT @ ntinued) including a comparison with preoperational studies, with operational controls, as appropriate, and with previous environmental surveillance reports, and an assessment of.the observed impacts of the plant operation on the environment.

   .The reports shall also include the results of the Land Use' Census required by Specification 3.12.2.

The Annual Radiological Envirormental Operating Reports shall include the results of' analysis of al'. radiologi 31 environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the Offsite Oose Calculation Manual, as well as summarized and tabulated results of 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 indivi-dual 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 a3 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

covering 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 and the corrective action taken if the specified program is not being performed as required by Specification 3.12.3;

    ' reasons for not conducting the Radiological Environmental Monitoring Program as

, required tcy specification 3.12.1, and discussion of all deviations from the

   . sampling sc.hedule of Table 3.12-1; discussion of environmental sample measure-ments that exceed the reporting levels of Table 3.12-2 but are no.t the result of' plant effluents, pursuant to ACTION b of Specification 3.12.1; and discussion of all analyses in which the LLD required by Table 4.12-1 was not achievable.

SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT ** 6.9.1.4 Routina Semiannual Radioactive Effluent Release Reports covering the operation of the unit during the previous 6 months of operation shall be submitted within 60 days after January 1 and July 1 of each year. The period of the first report shall begin with the date of initial criticality. The Semiannual Radioactive Effluent Release Reports shall include'a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit as outlined in Regulatory Guide 1.21, "Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants," Revision 1, June 1974, with data

        *0ne nap shall cover stations near the SITE BOUNDARY; a second shall include the more distant stations.
       **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.

COMANCHE PEAK - UNIT 1 6-20

ADMINISTRATIVE CONTROLS SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT (Continued) summarized on a quarterly basis following the format of Appendix B thereof. For solid wastes, the format for Table 3 in Appendix B shall be supplemented with thr,ee additional categories: class of solid wastes (as defined by 10 CFR Part 61), type of container (e.g., LSA, Type A, Type 8, Large. Quantity) and SOLIDIFICATION agent or absorbent (e.g., cement, urea formaldehyde). The Semiannual 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 on 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 liouid 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-ents, 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 methodology and parameters in the 0FFSITE DOSE CALCULATION MANUAL (ODCM). The Semiannual Radioactive Effluent Release Repo'rt to be submitted within 60 days after January 1 of each year shall also include an assessment of radia-tion doses to the likely most exposed MEMBER OF THE PUBLIC from reactor releases and other nearby uranium fuel cycle sources, including doses from primary efflu-ent pathways and direct radiation, for the previous calendar year to show con-formance with 40 CFR Part 190, "Environmental Radiation Protection Standards for Nuclear Power Operation." Acceptable methods for calculating the dose contribu-tion from liquid and gaseous effluents are given in Regulatory Guide 1.109, Rev. 1, October 1977. The Semiannual Radioactive Effluent Release Reports shall include a list and description of unplanned releases from the site to UNRESTRICTED AREAS of radioactive materials in gaseous and liquid effluents made during the reporting period. The Semiannual Radioactive Effluent Release Reports shall include any changes made during the reporting period to the PROCESS CONTROL PROGRAM (PCF) .

   *In lieu of submission with the Semiannual 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.

COMANCHE PEAK - UNIT 1 6-21 4

r ADMINISTRATIVE CONTROLS DMR SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT (Continued) and to the OFFSITE 00SE CALCULATION MANUAL (0DCM), pursuant to Specifica-ti.ons 6.13 and 6.14, respectively, as well as any major change to Liquid,

   ,   Gaseous,.or Solid Radwaste Treatment Systems pursuant to Specification 6.15.    -

It shall also include a listing of new locations for dose calculations and/or environmental monitoring identified by the Land Use Census pursuant to Specifi-cation 3.12.2. The Semiannual Radioactive Effluent Release Reports shall also include the following: an explanation as to why the inoperability of liquid or gaseous effluent monitoring instrumentation was not corrected within the time specified in Specification 3.3.3.10 or 3.3.3.11, respectively; and description of the events leading to liquid holdup tanks or gas storage, tanks exceeding the limits of Specification 3.11.1.4 or 3.11.2.6, respectively. MONTHLY OPERATING REPORTS

     - 6'. 9.1. 5 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 Resource Management, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, with a copy to the Regional Administrator of the Regional Office of the NRC, no later tMn the 15th of each' month following the calendar month covered by' the report.

RADIAL-PEAKING FACTOR LIMIT REPORT

6.9.1.6 The F xy limits for RATED THERMAL POWER (FRTP) shall be established x for at least each reload core and shall be maintained available in the Control Room. The limits shall be established and implemented on a time scale consis-tent with normal procedural changes. The analytical methods used to generate the F xy limits shall be tiicse previously reviewed and approved by the NRC.* If changes to these metnods are deemed necessary they will be evaluated in accordance with 10 CFR 50.59 and submitted to the NRC for review and approval prior to their use if the change is determined to involve an unreviewed safety question or if such a change would require amendment of previously submitted documentation. A report containing the F xy limits for all core planes containing Bank "0" control rods and all unrodded core planes along with the plot of predicted F Pp ,) axial core height (with the limit envelope for comparison) shall be provided to the NRC Document Control desk with copies to'the Regional Admin-istrator and the Resident Inspector within 30 days of their implementation.

      *WCAP 8385 "Power Distribution Control and Load Following Procedures" and WCAP 9272.A "Westinghouse Reload Safety Evaluation Methodology."

COMANCHE PEAK - UNIT 1 6-22

l ADMINISTRATIVE CONTROLS SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator of the Regional .0f fice of the NRC within the time period.specified for each report. 6.10 RECORD RETENTION 6.10.1 In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated. 6.10.2 The following records shall be retained for at least 5 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' rela' ted to '

nuclear safety; ,

c. All REPORTABLE EVENTS;

d. Records of surveillance activities, inspections, and calibrations required by these Technical Specifications;

e. Records of changes.made to the procedures required by Specification 6.8.1;

f. Records of radioactive shipments;

g. Records of sealed source and fission detector leak tests and results; and

h. Records of annual physical inventory of all sealed source material of record.

COMANCHE PEAK - UNIT 1 6-23 ,

a D1 AFT ADMINISTRATIVE CONTROLS RECORD RETENTION (Continued) 6.10.3 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 radioact;"e material released to the environs;

e. Records of transient or operational cycles for those unit components identified in Table 5.7-1;

f. Records of reactor. tests and expe.riments;

g. Records of training and qualification for current members of the unit staff;

h. Records of inservice inspections performed pursuant to these Technical Specifications;

i. Records' of quality assurance activities required by the Quality Assurance Man.ual;

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 50RC and the ORC;

1. Records of the service lives of all hydraulic and mechanical snubbers required by Specification 3.7.9 including the date at which the service life commences and associated installation and maintenance records;

m. Records of secondary water sampling and water quality; and

n. Records of analyses required by the Radiological Environmental Monitoring Program that would permit evaluation of the accuracy of the analysis at a latcr date. This should include procedures effective at specified times and QA records showing that these procedures were followed.

6.11 RADIATION PROTECTION PROGRAM ,

')

6.11.1 Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR 20 and shall be approved, maintained, and adhered to for all operations involving personnel radiation exposure. COMANCHE PEAK - UNIT 1 6-24

RAFf ADMINISTRATIVE CONTROLS 6.12 HIGH RADIATION AREA 6.12.1 Pursuant to paragraph 10 CFR Part 20.203(c)(5), in lieu of the "con-trol device" or "alarm signal" required by paragraph 10 CFR 20.203(c), each high radiation area, as defined in 10 CFR 20, in which the intensity of radia-tion is. equal to or less than 1000 mR/h at 45 cm (18 in.) from the radiation source or from any surface which the radiation penetrates shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP). Indi-viduals qualified in radiation protection procedures (e.g., Radiation Protec-tion Technician) or personnel continuously escorted by such individuals may be exempt from the RWP issuance requirement during the performance of their assigned duties in high radiation areas with exposure rates equal to or less than 1000 mf:/h, provided they are otherwise following plant radiation protec-

    . tion procedures for entry into such high radiation areas. Any individual or group of individuals permitted to enter such areas shall be provided wi.th or accompanied by one or more of the following:

a. A radiation monitoring device which continuously indicates the radiation dose rate in the area; or

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 levels in the area have been estab-lished and personnel have been made knowledgeable of them; or

c. An individual qualified in radiation protection procedure ~s with a -

radiation dose rate monitoring device, who is responsible for pro-- viding positive 'ontrol c over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the [ Radiation Protection Manager] in the RWP. 6.12.2 In addition to the requirements of Specification 6.12.1, areas acces-sible to personnel with radiation levels greater than 1000 mR/h at 45 cm (10 in.) from the radiation source or from any surface which the radiation penetrates shall be provided with locked doors to prevent unauthorized entry, and the keys shall be maintained under the administrative control of the Shift Supervisor on duty and/or radiation protection supervision. Doors shall remain locked except during periods of access by personnel under an approved RWP which shall specify the dose rate levels in the immediate work areas and the maximum allowable stay time for individuals in that area. In lieu of the stay time specification of the RWP, direct or remote (such as closed circuit TV cameras) continuous surveillance may be made by personnel qualified in radia-tion protection procedures to provide positive exposure control over the activities being performed within the area. For individual high radiation areas accessible to personnel with radia-tion levels of greater than 1000 mR/h that are located within large areas, such as PWR containment, where no enclosure exists for purposes of locking, and where no enclosure can be reasonably constructed around the individual area, that individual area shall be barric'aded, conspicuously posted, and a flashing light shall be activated as a warning device. COMANCHE PEAK - UNIT 1 6-25

ADMINISTRATIVE CONTROLS iib 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 change.s to the PCP: .

a. 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:

1) Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information;

2) A determination that the change did not red'uce the overall conformance of the solidified waste product to existing crite" a for solid wastes; and

3) Documentation of the fact that the change has been reviewed and found acceptable by the 50RC.

b. Snall become effective upon review and acceptance by the 50RC.

6:14 0FFSITE 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:

a. Shall be submitted to the Commission in the Semiannual Radioacti.ve Effluent Release Report for the period in which the change (s) was made effective. This submittal shall contain:

1) 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 ODCH to be chariged with each page numbered, dated and containing the revision number, together with appropriate analyses or evaluations justifying the change (s);

2) A determination that the change will not reduce the accuracy or reliability of dose calculations or Setpoint determinations; and

3) Documentation of the fai:t that the change has been reviewed and found acceptable by the SORC,

b. Shall become effective upon re' view and acceptance by the 50RC.

COMANCHE PEAK - UNIT 1 6-26

1 ADMINISTRATIVE CONTROLS 6.15 MAJOR CHANGES TO LIQUID, GASEOUS, AND SOLIO RADWASTE TREATMENT SYSTEMS

6.15.1 Licensee-initiated major changes to the Radwach Treatment Systems (liquid, gaseous, and solid):

a. Shall be reported to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the evaluation was reviewed by the 50RC. The discussion of each change shall contain:

1) A summary of the evaluation that led to the determination that the change could be made in accordance with 10 CFR 50.59;

2) Sufficient detailed information to totally support the reason for the change without benefit of additional or supplemental information;

3) A detailed description of the equipment, components, and processes involved and the interfaces with.other plant systems;

4) 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;

5) An evaluation of the change,.which shows the expected maximum exposures to a MEMBER OF THE PUBLIC in the UNRESTRICTED AREA and to the general population that differ from those previously estimated in the License application and amendments thereto;

6) 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 change is to be made;

7) An estimate of the exposure to plant operating personnel as a result of the change; and

8) . Documentation of the fact that the change was reviewed and found acceptable by the 50RC.

! b. Shall become effective upon review and acceptance by the 50RC. l

Licensees may choose to submit the information called for in this Specification as part of the annual FSAR update.

COMANCHE PEAK - UNIT 1 6-27

ENCLOSURE 2 SER ITEMS FOR WHICH TECHNICAL SPECIFICATIONS NEED TO BE PROPOSED

1. The Conianche Peak Steam Electric Station (CPSES) Safety Evaluation Report (SER).Sectionf4.4.1 states the following:

                 "The staff will require the applicant to present an acceptable method of accommodating the thermal margin ' reduction ... so that appropriato provisions may be incorporated in the Technical Specifications." The applicant should insert into the basis of the proposed Technical Spect-fications any plant-specific or generic margin that may be used to offset the reduction in departure from nucleate boiling ratio (CNBR) caused by rod bowing and should also reference the source and staff approval of each generic marsin."

The marked-up version of the Westinghouse Standard Technical Specifications (STS) submitted by TV Electric on October 30,1987(TXX-6905)doesnotseem to address this SER requirement. TU Electric needs to show that the require-ment has been incorporated into the October 30, 1987 marked-up STS or pro-pose appropr.iate technical specifications.to meet the requirement.

2. SER Section 4.4.1 requires that appropriate s'urveillance requirements be included in the Technical Specifications to recognize any rapid crud

   .      buildup in the reactor core. TV Electric has not proposed any surveil-lance requirements to meet this SER requirement and needs to provide the appropriate requirements.

3. Section 7.2.6 of Supplement 12 to the SER describes how the CPSES meets the generic implications of the anticipated transient without scram (ATWS) event at the Salem Nuclear Power Plant (Generic Letter 83-?8).

TV Electric modified the reactor trip system in order to meet Action Item 4.3 of the Generic Letter, Autcmatic Actuation of the Breaker Shunt Trip Attachment. That modification was found to be acceptible by the sta f f. However, Action Item 4.3 of Generic Letter 83-28 and Generic Letter 85-09 require that technical specificaticns be pr ansed for the automatic actuation of the breaker shunt trip attache s radification. This has not been included in the October 30, 1987 mar k . 'TS and TV Electric needs to provide the appropriate technice eped Hcations.

4. Section 7.3.1.5 of Supplement 3 to the SER stated tha m i .: x

. generator level channel is not operable, the high st. c m a. m level l protection channel selected for control of feedwater - NW l actuation will be placed in a high level trip condition. 9 S w: to be included in the Limiting Conditions for Operation (LCO, ' plant. TV Electric did not include this LCO in T.S. 3/4.3.2 of the cc.ooer 30, 1987 marked-up STS and needs to provide the appropriate LC0.

5. SER Section 8.2.4 requires specific Technical Specifications (LCOs),

surveillance requirements, and trip setpoints with minimum and maximum limits for the first and second level of undervoltage protection. Neither the October 1987 Final Draft Comanche Peak' TS nor the October 30, 1987 marked-up STS appear to include these requirements, particularly with regard to the minimum and maximum trip setpoints. TV Electric needs to provide the appropriate technical specifications.

t t _ >

6. SER Section 8.4.3 and Branch Technical Position ICSB 18 (PSB) requires all valves requiring a power lockout in order to meet the single failure criterion be listed in the Technical Specifications. Neither the October 1984 Final Draft Comanche Peak TS nor the October 30 marked-up STS includes such a list. TV Electric needs to provide the appropriate '

technical' specification.

7. SER Section 11.2.1.2 requires that because the filter /demineralizers for the condensate clean-up system are automatically backwashed, the Technical Specifications will require resin sampling and analysis for radioactivity.

Table 4.11-1, "Radioactive Liquid Waste Sampling and Analysis Program," does not appear to include this SER requirement. TV Electric needs to either show that the requirement is included in the Table or to include the requirement in the TS.

8. SER Section 15.4.5 states that when the emergency core cooling system (ECCS) pumps are operating under accident conditions, a negative pressure of at ,

    ! east 1/8 in water guage will be established by the emergency v'en'tilation.

This was a commitment made by TV Electric andn w'ich the staff required to be included in the Technical Specifications. The October 30, 1987 marked-up STS did not include this requirement. TV Electric needs to provide appropriate technical specifi. cations.

9. Section 22.2-II.F.2 of Supplement 6 to the SE.,. "Instrumentation for Detection of Inadequate Core Cooling (ICC), "requires,.' prior to fuel loading, that the technical specification for the ICC system be submitted by TV Electric and approved by the staff. Section 3/4.3, "Instrumentation,"

of the October 30, 1987 marked-up STS does not contain these technical specifications. TU Electric needs to provide the required technical specifications. O e

1

                    .                        ENCLOSURE 3 Comanche Peak Steam Electric Station, Unit 1 Technical Specification Review Milestones January 11-15, 1988       NRC/TV Electric meeting to resolve differences     ,

between TV Electric's marked-up STS and the first draft TS. February 8,1988 TV Electric responds to all action' items from 1 meeting held the week of January 11, 1988. March 21, 1988* NRC staff resolves all NRC and JTU Electric action items from site visit and any new differences or proposed changes. April 4, 1988 TV Electric begins review of Proof and Review copy of TS. . April 18, 1988 TV Electric provides comments on Proof and Review copy of TS. . May 9-13, 1988 .NRC Task Force Inspection. , May 16,1988* NRC. staff resolves NRC and applicant comments on Proof and Review copy of TS. (Sitevisitasnecessary.). May 30, 1988 TV Electric receives Final Draft version of TS. June 13, 1988 'NRC staff audit review of TS against FSAR and SER completed. June 20, 1988* NRC staff resolves differences between T5, FSAR, and SER identified during audit review. June 27, 1988 TV Electric certifies to NRC the accuracy of the TS in ccnformance with the FSAR, SER, as-built and safety analyses. 1 June 30, 1988 NRC completes preparation of Appendix A to the l operating license. l

Appropriate TU Electric staff should be readil'y available to discuss dif-ferences with NRC reviewer (s) in the time period prior to this milestone date.

l t . l l l \ '.i

1 i ENCLOSURE 2 SER ITEMS FOR WHICH TECHNICAL SPECIFICATICf.S NEED TU BE PROPOSED

1. The Comanche Peak Steam Electric Station (CPSES) Safety Evaluation Report (SER) Section 4.4.1 states the followiro:

                "The staff will require the applicant to present an acceptable method of accortnocating the thermal margin reducticn ... so that appropriate provisions may be incorporated in the Technicsl Specifications." The applicant should insert into the basis of the prog. 4 Technical Spec -

fications any plant-specific or generic margin that may be useo to off set the reduction in departure from rucleate boiling ratio (CNBR) caused by rod bowing and should also reference the source and staff approval of each generic margin." The marked-up version of the Westinghouse Standard Technical Specifications (STS) submitted by TV Electric on October 30, 1987 (TXX-6905) does not seem to address this SER requirement. TU Electric neeos to show that the require-ment has been incorporated into the October 30, 1987 marked-up STS or pro-pose appropriate technical trecifications to meet the requirerrent.

2. SER Section 4.4.1 requ1res that 6ppropriate surveillance requirerrents be included in the Technical Specifications to recognize any rapid crud buildup in t'e reactor core. TV Electric has not proposeo any surveil-lance requirements to niet this SEP requirement and needs to provide the ,

apprcpriate requirements.

3. Secticn 7.2.6 of Supplement 12 to the SES describes how the CPSES ceets the ceneric implications of the anticipated trarcient without scram (ATWS) event at the Salem Nuclear Power Plant ' aneric Letter 83-28).

TU Electric modif fed the reactor trip system ir :rder to meet Action Item 4.3 of the Generic Letter, Autenatic Actucion of the Breaker Shunt Trip Attachn ent. That ncdificaticn was found : be acceptable by the

          -taff. However, Action item 4.3 of Generic letter 83-28 and Generic Letter 85-09 reouire that technical specifications be proposed fer the autcratic actuation of the breaker shunt trip attachment modification.

This has not been included in the October 30, 1987 marked-up STS and TV Electric needs to provide the appropriate technical specifications.

4. Section 7.3.1.5 of Supplerrent 3 to the SER stateo that when a steam ganerator level channel is not cperable, the high steam generator level protection charnel selected for control of feedwater line isolation actuatien will be placed in a high level trip condition. This was tc be included in the Limiting Conditiens for Operation (LCO) of the plant.

TV Electric did not include this LCO in T.S. 3/4.3.2 of the October 30, 1967 r:arked-up STS and needs to provice the appropriate LCO.

c. SER Secticn 8.2.4 requires specific Technical Specifications (LCOs),

surveillance requirements, and trip setpoints with minimum anc n:aximum limits for the first and second level of undervoltage protection. Neither the Octcber 1987 Firal Draft Cemanche Peak TS nor the October 30, 1987 marked-up STS appear to incluce these reovirements, particularly with regard to the minin:cm and maximun trip setpoints. TV Electric needs to provice the appropriate technical specifications.

.g.

6. SER Section 8.4.3 and Branch Technical Position ICSB 18 (PSB) requires all valves requiring a power lockout in order to meet the single failure criterion be listed in the Technical Specifications. Neither the October 1984 Final Craft Comanche Peak TS nor the October 30 markeo-up STS includes such a list. TU Electric needs to provide the apprcpriate technical spec, ".ation.

7. SER Section 11.2.1.2 requires that because the tilter /cemineralizers for the condensote clean-up system are automatically backwashed, the Technical Specifications wil' reouire resin sampling ar.d analysis for radioactivity.

Table 4.11-1, "Radioactive liquid Waste Sampling and Analysis Program," does not appear to incluce this SER requirement. TU Electric needs to either shew that the requirement is incluoed in the Table or to include the requirement in the TS.

8. SER Section 15.4.5 states that when the emergency core cooling systet. (ECCS) ,

pumps are operating under accident conditiens, a negative pressure cf et least 1/8 in, water guage will be established by the emergency ventilatur. This was a commitment made by TV Electric and which the statf required to be includeo in the Technical Specifications. The October 30, 1987 marked-up STS did not include this requirement. TV Electric needs to provide appropriate technical specifications.

9. Section 22.2-II.F.? of Supplenent 6 to the SER, "Instrumentation fcr .

Detection of Inadequate Core Cooling (ICC), "requires, prior to fuel leading, that the technical specification for the ICC system be submitteo by TV Electric and approved by the staff. Sect n 3/4.3, "Instrumentation," of the October 30, '987 marked-up STS does not ontain these technical specifications. TV Electric needs to provide t"e required technical specificationr. ) 4 m 4 , - - ~ -

i ENCLOSURE 3 Comanche Feak Steam Electric Station, Unit 1 Technical Specificatior. Feview Milestones January 11-15, 1988 NRC/TV Electric neting to resolve differences between TV Elec; fe's marked-up STS and the first draft TS. February 8, 1988 TU Electric responds to all action items from neeting held the week of January 11, 1988.

March 21,1988* NRC steft resolves all NRC and Tu Electric action items from site visit and any rew differences or proposed changes. April 4. 1988 TV Electric begins review of Proof and Review copy of TS. April 18, 1988 TV Electric provides coments on Proof ard Peview copy of TS. May 9-13, 1988 NRC Task Force Inspection. May 16, 1988* NRC staff reselves NRC ano applicant cements en Proof and Review cepy of TS. (Site visit as necessa y.) May 30, 1988 TV Electric receives Final ' raft version of TS. Junc 13, 1988 'NPC staff audit review of " against FSAR and SER completed. June 20, 1988* NRC staff resolves differences between TS, FSAP, and SER identified durino audit review, June 27, 1988 TU Electric certifies to hRC the accuracy of the TS in ccnformance with the FSAR, SCR, as-built ard safety analyses. June 30, 1988 NRC completes preparation of Appendix A to the operatina license.

                             *Appropriete TL. Electric statt should be readily available to discuss dif-firencts with NRC reviewer (s) in the time period prior to this milestone date.}}

ML20147D900

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ML20147D900

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