ML20206H759
| ML20206H759 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 11/18/1988 |
| From: | Calvo J Office of Nuclear Reactor Regulation |
| To: | Goldberg J HOUSTON LIGHTING & POWER CO. |
| References | |
| NUDOCS 8811230303 | |
| Download: ML20206H759 (450) | |
Text
{{#Wiki_filter:-- - _ _ _ ___-_____ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - -- ___ _- -______ __ _____-______ - ________ -___ L() November 1@,1988 4 Docket Mos. 50-498
- and 50-499-l
.n Mr. J. H. Goldberg <
Group Vice-President, Nuclear - -
. Houston Lighting & Power Corpany P. O. Box 1700 llouston, Texas 77001 j
Dear fir. Goldberg:
SUBJECT:
FINAL DRAFT VERSION OF REVISED TECHNICAL SPECIFICATIONS - , 500Til. TEXAS PROJECT, UNITS 1 AND 2 i t
. Enclosed is the Final Draft version of the Technical Specifications (TS) for South Texas Project Units I and 2i The TS are being transmitted for review .
and Certification that the final diaft TS accurately reflect the as-built ; condition of Unit 2 and the FSAR. The TS to be issued with the operating ; license for Unit 2 are to be identical to the final draft TS unless changes < are formally requested, justified, and approved. Any cour.enic or requests for changes to the final draft TS nust be submitted on the docket along with justifications for any requested changes. ; The certification should be p;ovided by December 2, 1988.
; Sincerely,
/s/
Jose A. Calvo, Director ! ! Project Directorate - IV i Division of Reactor Projects - III, i ! IV, Y and Special Projects i ! Office of fiuclear Reactor Regulation ; i r i
Enclosure:
As stated t cc w/ enclosure: ; See next page f
.DISTRIEUTIAN
Bin:let (fFe7dN, 5 7 NRC PDR Local PDR PD4 Reading L. Rubenstein J. Calvo P. Noonan G. Dick ,
C. Abbate OGC-Rockville E. Jordan B. Grimes : ACRS (10) PD4 Plant File ! i PD4/LQ PD4/PE($ PD4/P ( PD4/D i; PNoonan CAbbate: Lr GDick b' JCalvo . 11//$88 ll/l5/88 11/if/88 II//f /88 { i i 8811230303 es123g n l gDR ADOCK 05 coo 499 'Q 3 i
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November 18, 1988 Docket flos. 50-498 and 50-499 Mr. J. H. Goldberg Group Vice-President,fiuclear Houston Lighting & Power Cornany P. O. Box 1700 Houston, Texas 77001
Dear Mr. Goldberg:
SUBJECT:
Fli!AL DRAFT VERSION OF REVISED TECHNICAL SPECIFICAT10iis - SOUTH TEXAS PROJECT, UNITS 1 At:0 2 Enclosed is the Final Draf t version of the Technical Specifications (TS) for South Texas Project, l' nits 1 and 2. The TS are being trar.sraitted for review and certification that the final draf t TS accurately reflect the as-built condition of Unit 2 and the FSAR. The TS to be issued with the operating license for Unit 2 are to be identical to the final draf t TS unless changes are forcally requested, justified, and approved. Any corrt.nts or requests for charges to the final dr6ft 1S must be submitted on the docket along with justifications for any requested changes. . The certification should be providet by Decen;ber 2, 1968. Sincerely,
/s/
Jose A. Calvo, Director Project Directorate - IV Division of Reactor Projects - Ill, IV, Y and Special Projects Office of liuclear Reactor Regulatior.
Enclosure:
As stated cc w/ enclosure: See next page DISTRIBUTION i;RC PDR Local PDR PD4 Reading D'oU et File L. Rubenstein J. Calvo P. iioonan G. Dict C. Abbate OGC-Rockville E. Jordan B. Grimes ACRS (10) PD4 Plant File PD4/PE(hh PD4/D I PD4/LC! Phoonan CAbbote:sr PD4/P( GDic N. *' e JCalvo 11/15 /E 8 11/if/88 11//J /E6 11/(()'F0
e
+* UNITED STATES
((g4E8Ipq%, g NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20666 t r a November 18, 1988
%,y ,/
es..* Docket Nos. 50-498 and 50-499 Mr. J. H. Goldberg Group Vice-President, Nuclear Houston Lighting & Power Company P. O. Box 1700 Houston, Texas 77001
Dear Mr. Goldberg:
SUBJECT:
Fit;AL DRAFT VERS 101: OF REVISED TECHNICAL SPECIFICATIONS -~ SOUTH TEXAS PROJECT, Uf;1TS 1 AND 2 Enclosed is the Final Draf t versien of the Technical Specificatior.s (TS) for South Texas Project, Units 1 and 2. The TS are being transmitted for review and certification that the firtal draf t TS 6ccurately reflect the as-built condition of Unit ? 6nd the FSAR. The TS to be issued with the operating license for Unit 2 are to be identical to the final draft TS unless changes are formally requested, justified, and approved. Any corrents or requests for chariges to the final draf t IS must be subr.itted on the docket alor.g with justifications for any requested changes. The certification should be provided by December 2, 1988. Sincerely, 7 % d , [et bW fJose A. Calvo, Director Project Directorate - IV Division of Reactor Projects - 111 IV, V and Special Projects Office of Nuclear Reactor Regulation
Enclosure:
As stated cc w/ enclosure: i See next page i
. Mr. J. H. Goldberg South Texas Project Houston Lighting and Power Company Cc:
Brian Berwick, Esq. Resident Inspector / South Texas Assistant Attctney General Project Environr, ental Protection Division c/o U.S. Nuclear Regulatory Commission P. O. Box 12548 P. O. Box 910 Capitol Station Bay City, Texas 77414 Austin, Texas 7P711 Mr. Jonathan Davis Mr. J. T. Westermeier Assistent City Attorney General Manager, South Texas Project City of Au~stin Houston Lighting and Power Company P. O. Box 1088
- P. O. Box 289 Austin, Texas 78767 Heusten, Texas 77483 Ms. Pat Coy Mr. R. J. Miner- Citizers Concerned About Nuclear-Chief Operating Officer . Power
- City of Austin Electric Utility 10 Singleton 721 Barton Springs Road Eureka Springs, Arkansas 72632 Austin, Texas 78704 Mr. M. A. McBurnett Mr. R. J. Costello Manager, Operations Support Licer sing Mr. H. T. Herdt Housten Lightir g 6r.d Power toepany City Public Service Board P. O. Eox 28g
< P. O. Box 1771 Wadsworth, Texas 77483 San Antoriio, Texas 7829f Mr. A. Zaccaria Jack R. Newman, Esq. Mr. K. G. Pess rewran & Poltziriger, P. C. Bechtel Corperatier. 1615 L Street, NV P. O. Box 216C Washingtor , D.C. 20036 Poustun, Texas 77001 Melbert Schwartz, Jr. , Esq. Mr. P. P. Verret E6ker & Ectts Mr. R. L. Range One Shell Plaza Cer tral Power and Light Company Houston, Texas 77002 P. O. Box 2121 Corpus Christi, Texas 78403 Mrs. Peggy Cuchern Executive Director Doub, Muntzing ar.d Glasgow Citizens for Equitable Utilities, Inc. Attorneys at Law Poute 1. Cox 1684 Suite 400 Brazoria, Texas 77422 808 Seventeenth Street, N.W. llashington, D.C. 20006 Mr. S. L. Posen General Manager, Operatiors Support Houston Lighting and Power Ccmpany P. O. Box 289 Wadsworth, Texas 77483
l'r. J. H. Goldberg South Texas Project Housten Lighting & Power cc: 5 Regional Administrator, Region IV U.S. Nuclear Regulatory Connission Office of Executive Director for Operations 611 Ryar. Plaza Drive Suite 1000 Arlington, Texas 76011 - Mr. Lanny Sinkin, Counsel for Intervenor Citizeris Concerned about f*uclear Power, Inc. Christic Institute 1324 North Capitol Street Washington, D.C. 20002 Licensing Representativt . Houston Lighting and Power Corpany Suite 610 Three l'etro Center Bethesda, Haryland 20814 Rufus S. Scott
/sscciate General Counsel Houston Lighting & Power Conpany P. O. Box 1700 Houston, Texas 77001 IliP0 Pecords Center 1100 Circle 75 Parkway Atlanta, Georgia 30339-3064 Juseph I'. Hencrie -
50 Ee11 port Lane Ee11 port, flew Yert 11713
- Gerald C. Vaughn, Vice President
. I:uclear Operations Houston Lighting & Power Company P. O. Box 289
- Kadsworth Texas 77483 R. W. Chewning, Chairrran l'uclear Safety Review Board Houston Lighting & Power Corrpany i
P. O. Box ?89 i Kadsworth, Texas 77483 l l t-
-.s - -- - a- A1 , , - -..
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t DRA F7~ FD) INDEX , l I i k I i ' I k I , i l I l , i l l ( 1 .
, t l ;
) - 1: 1953 l
! I -nn,-n-e - - . . - , _ - . , _ _ _ ~ . _
1.0 DEFINITIONS PAGE SECTION 1-1 1.1 ACTION........................................................ 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................................................. 1-2 1.7 CONTAINMENT INTEGRITY......................................... 1-2 1.8 CONTROLLED LEAKAGE............................................ 1-2 1.9 CORE ALTERATIONS.............................................. 1-2 1.10 DIGITAL CHANNEL OPERATIONAL TEST. ............................ 1-2 1.11 DOSE EQUIVALENT I-131......................................... 1- 3 1.12 E-AVERAGEDISINTEGR4TIONENERGY............................... TIME...................... 1-3 1.13 ENGINEERED SAFETY FEATURES RESPONSE 1-3 1.14 FREQUENCY N0TATION............................................ 1-3 1.15 GASEOUS WASTE PROCESSING SYSTEM............................... 1-3 1.15 IDENTIFIED LEAKAGE.......... ................................. 1-4 1.17 PASTER RELAY TEST............................................. 1-4 1.1B MEMBER (S) 0F THE PUBLIC....................................... 1-4 1.19 0FFSITE DOSE CALCULATION MANUAL............................... 1-4 1.20 OPERABLE - OPERABILITY........................................
................. 1-4 1.21 OPERATICNAL M0hr. - M0DE.....................
1-4 1.22 PHYSICS TESTS................................................. 1-4 1.23 PRESSURE BOUNDARY LEAKAGE...... .............................. 1-5 1.24 PROCESS CONTROL PR0 GRAM....................................... 1-5 1.25 PURGE - PURGIN3............................................... 1-5 1.26 QUADRANT POWER TILT RATI0..................................... 1-5 1.27 RATED THERMAL P0WER........................................... d i AMEN; MENT N05. AND SOUTH TEXAS - UNITS 1 & 2
'iu..
INDEX bh DEFINITIONS
,SECTION PAGE 1.28 REACTOR TRIP SYSTEM RESPONSE TIHE. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5 1.29 RE70RTABLE EVENT.............................................. 1-5 1.30 SHUTDOWN MARGIN............................................... 1-5 1.31 SITE B0VNDARY................................................. 1-6 1.32 SLAVE RELAY TEST.............................................. 1-6 1.33 SOLIDIFICATION................................................ 1-6 1.34 SOURCE CHECK.................................................. 1-6 1.35 STAGGERED TEST BASIS.......................................... 1-6 1.36 THERMAL P0WER................................................. 1-6 1.37 TRIP ACTUATING DEVICE OPERATIONAL TEST........................ 1-6 1.38 U N I D E N T I F I E D L E A KAG E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-6 1.39 UNRESTRICTED AREA............................................. 1-7 1.40 VENTING....................................................... 1-7 TABLE 1.1 FREQUENCY N0TATION....................................... 1-8 TABLE 1.2 OPERATIONAL M00ES........................................ 1-9 AMENDMENT NCS. AND SOUTH TEXAS - UNITS 1 & 2 ii I'0 V 17 n;,
FP 1 INDEX 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS PAGE SECTION 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 LIIL' - FOUR LOOPS IN OPERATION.... 2-2 2.2 LIMITING SAFETY _S,YSTEM SETTINGS 2-3 [ t 2.2.1 REACTOR TRIC r di J NSTRUMENTATION SETP0!NTS. . . . . . . . . . . . . . . . . !~ ' 2-4 , TABLE 2.2-1 REACTOR TRIf SYSTEM INSTRUMENTATION TRIP SETPOINTS...... t f BASES i PAGE SECTION 2.1 SAFETY LIMITS B 2-1 I 2.1.1 REACTOR C0RE................................................. PRESSURE.............................. B 2-2 2.1.2 REACTOR COOLANT SYSTEM 2.2 LIMITII'(iSAFETYSYSTEMSETTINGS B 2-3 2.2.1 REAC fJR TRIP SYSTEM INSTRUMENTATION SETPOINTS. . . . . . . . . . . . . . . . , } I f , l l ! i i ! I i i i I I ! i i f I ! i ! 'I ! south TEXAS - UNITS 1 & 2 iii 171s;g i
Fp INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILL.tNCE REQUIREMENTS PACE SECTION 3/4.0 APPLICABILITY................................................ 3/4 0-1 J 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL i Shutdown Margin - T,yg Greater Than 200'F............'..... 3/4 1-1 FIGURE 3.1-1 REQUIRED SHUTDOWN MARGIN YERSUS RCS CRITICAL BORON CONCENT RATION (MODES 1, 2, 3, AND 4). . . . . . . . . . . . . . . . . . . . . . 3/4 1-3 3/4 1-4 Shutdown Margin - T,yg Less Than or Equal to 200*F........ i FIGURE 3.1-2 REQUIRED SHUTDOWN MARGIN VERSUS RCS CRITICAL BORON CONCENTRATION (MODE S).................................... 3/4 1-5 ; Moderator Temperature Coefficient......................... 3/4 1-6 Minimum Temperature for Criticality....................... 3/4 1-8 3/4.1.2 BORATION SYSTEMS Flow Paths - Shutdown..................................... 3/4 1-9 g Flow Paths - Operating.................................... 3/4 1-10 [ ' Charging Pumps - Shutdown.... ............................ 3/4 1-11 2 Charging Pumps - Operating................................ 3/4 1-12 [ i Borated Water Sources - Shutdown.......................... 3/4 1-13 i Operating......................... 3/4 1-14 l i Borated Water Sources - t I 3/4.1.3 MOVABLE CONTROL ASSEMBLIES 1 Group Height.............................................. 3/4 1-16 : j ! TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE 3/4 1-18 [ EVENT OF AN INOPERABLE Full- LENGTH R00. . . . . . . . . . . . . . . . . . . . Position Indication Systems - Operating................... 3/4 1 19 i Position Indication Systems - Shutdown.................... 3/4 1-20 l j . Red Drop Time............................................. 3/4 1-21 3/4 1-22 l Shutdown Rod Insertion Limit.............................. 3/4 1-23 f i Control Rod Insertion Limits.............................. i FIGURE 3.1-3 ROD BANK INSERTION LIMITS VERSUS THERMAL POWER3/4 1-24 - i FOUR-LOOP 0PERATION....................................... i l AND SOUTH TEXAS - UNITS 1 & 2 iv AMENDMENT NOS. j
T 1;;;;g l i
FD I INDEX LIMITIN'i CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS PAGE 3ECT!0N 3/4.2 POWER DISTRIBUTION LIMITS 0!FFERENCE..................................... 3/4 2-1 3/4.2.1 AXIAL FLUX FIGURE 3.2-1 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF 3/4 2-4 RATED THERMAL P0WER...................................... 3/4 2-5 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - Fg (Z)..................... 3/4 2-6 FIGURE 3.2-2 K(Z) - NORMALIZED F n (Z) AS A FUNCTION OF CORE HEIGHT. 3/4 2-9 3/4.2.3 NUCLEAR ENTMALPY RISE HOT CHANNEL FACTOR................. 3/4 2-10 3/4.2.4 QUADRANT POWER TILT RATI0................................ 3/4 2-11 3/4.2.5 DNB PARAMETERS........................................... 3/4.3 INSTRUMENTATION REACTOR TRIP SYSTEM INSTRUMENTATION...................... 3/4 3-1 3/4.3.1 3/4 3-2 TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION................... 3/4 3-9 TABLE 3.3-2 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES.... TABLE 4.3-1 REACTOR YRIP SYSTEM INSTRUMENTATION SURVEILLANCE 3/4 3-11
, REQUIREMENTS.............................................
3/4.3.2 ENGINEEREr e SA5ETY FEATURES ACTUATION SYSTEM 3/4 3-16 INSTRUMENTATION............ ............................. TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM 3/4 3-18 INSTRUMENTATION.......................................... TABLE 3.3-4 ENCINEERED SAFETY FEATURES A;TUATION SYSTEM SETP0lNTS........................... 3/4 3-29 INSTRUMENTATION TRIP 3/4 3-37 TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES............. TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM 3/4 3-42 INSTRUMENTATION SURVEILLANCE REQUIREMENTS................ 3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitoring for Plant Operations................ 3/4 3 50 TABLE 3.3-6 RADIATION MONITORING INSTRUMENTATION 3/4 3-51 FOR PLANT OPERATIONS..................................... TABLE 4.3-3 RADIATION MONITORING IN!TRUMENTATION FOR PLANT 3/4 3-53 OPERATIONS SURVEILLAN;E REQUIREMENTS..................... 3/4 3 54 Novable Incore Detectors................................. 3/4 3-55 Seismic Instrumentation................ ................. 3/4 3-56 TABLE 3.3-7 SEISMIC MONITORING INST RUMENT ATION. . . . . . . . . . . . . . . . . . . . y AMENDMENT N05. AND SOUTH TEXAS - UNITS 1 & 2 7
FD INDEX l LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS j PAGE SECTION TABLE 4.3-4 SEISMIC MONITORING INSTRUMENTATION SURVEI: - REQUIREMENTS................................. . .. 3/4 3 57 ; Meteorological Instrumentation................ ......... 3/4 3-58 [ TABLE 3.3-8 METEOROLOGICAL MONITORING INSTRUMENTATION............. 3/4 3-59 j TABLE 4.3-5 METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS............................................. 3/4 3-60 Remote Shutdown System .............'..................... 3/4 3-61 l TABLE 3.3-9 REMOTE SHUTDOWN SYSTEM ............................... 3/4 3-62 ; TABLE 4.3-6 REMOTE SHUTDOWN MONITORING INSTRUMENTATION
- SURVEi LLANCE REQUli<EMENT S. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 3-66 Accident Monitoring Instrumentation...................... 3/4 3-67 4
TABLE 3.3-10 ACCIDENT MONITORING INSTRUMENTATION.................. 3/4 3-68 , TABLE 4.3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS............................................. 3/4 3-73 Chemical Detection Systems............................... 3/4 3-75 , TABLE 3.3-11 (This table number is not used.)..................... 3/4 3-77 Radioactive Liquid Effluent Monitoring Instrumentation... 3/4 3-79 , TABLE 3.3-12 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION 3/4 3-80 TABLE 4.3-8 RADI0 ACTIVE LIQUID EFFLUENT MONITORING
> INSTRUMENTATION SURVEILLANCE REQUIREMENTS................ 3/4 3-82 f Radioactive Gaseous Effluent Monitoring Instrumentation.. 3/4 3-84 ;
TABLE 3.3-13 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION....................... .................. 3/4 3-85 ; l , TABLE 4.3-9 RADICACTIVE GASEOUS EFFLUENT MONITORIN3 3/4 3-87 l INSTRUMENTATION SURVE!LLANCE REQUIREMENTS................ ' PROTECTION.............................. 3/4 3-89
; 3/4.3.4 TURBINE OVERSPEED i '
4 ' t 1 1 I i : d vi AMEh0 MENT NOS. AN? SOUTH TEXAS - UNITS 1 l. 2 [ f t' .9 ) * , i ! i
FD INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION Startup and Power Operation..............,............... 3/4 4-1 Hot Standby.............................................. 3/4 4-2 Hot Shutdown............................................. 3/4 4-3 Col d Shutdown - Loop s Fi11 ed. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-5 Cold Shutdown - Loops Not F111ed......................... 3/4 4-6 3/4.4.2 SAFETY VALVES Shutdown................................................. 3/4 4-7 0perating................................................ 3/4 4-8 3/4.4.3 PRESSURIZER.............................................. 3/4 4-9 3/4.4.4 RELIEF VALVES............................................ 3/4 4-10 3/4.4.5 STEAM GENERATORS......................................... 3/4 4-12 TABLE 4.4-1 HINIMUM NUMBER OF 5 TEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION............................. 3/4 4-17 TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION....................... 3/4 4-18 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems................................ 3/4 4-19
. Operational Leakage...................................... 3/4 4-20 TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES...... 3/4 4-22 3/4.4.7 CHEMISTRY................................................ 3/4 4-23 TABLE 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS............... 3/4 4-24 TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE
~
REQUIREMENTS.............................................. 3/4 4-25 3/4.4.8 SPECIFIC ACTIVITY........................................ 3/4 4-26 FIGURE 3.4-1 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIL ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY >l pCi/ gram DOSE EQUIVALENT I-131.................................... 3/4 4-28 TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM.................................................. 3/4 4-29 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System................................... 1/4 4-31 FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS - APPLICABLE UP TO 32 EFPY................................. 3/4 4 32 a AMENDMINT N35. AND 50VIH TEXAS UNITS 1 & C vii Mr ; , l i I I l
-g Y.V INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE FIGURE 3.4-3 REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS -
APPLICABLE UP TO 32 EFPY................................. 3/4 4-33 TABLE 4.4-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM - WITHDRAWAL SCHEDULE...................................... 3/4 4-31 Pressurizer.............................................. 3/4 4-35 Overpressure Protection Systems.......................... 3/4 4-36 FIGURE 3.4 4 NOMINAL MAXIMUM ALLOWABLE PORV SETPOINT FOR THE COLD OVERPRESSURE SYSTEM ........................ 3/4 4-38 3/4.4.10 STRUCTURAL INTEGRITY..................................... 3/4 4-39 3/4.4.11 REACTOR VESSEL HEAD VENTS......... ...................... 3/4 4-40 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS ............. .............................. 3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - T F.... 3/4 5-3 avg GREATER THAN OR EQUAL TO 350 3/4.5.3 ECCS SUBSYSTEMS - T avg LESS THAN 350 F................... 3/4 5-6 ECSS SUBSYSTEMS - T F....... 3/4 5-8 avg LESS THAN OR EQUAL TO 200 3/4.5.4 (This specification number is not used).................. 3/4 5-9 3/4.5.5 REFUELING WATER STORAGE TANK............................. 3/4 5-10 3/4.5.6 RESIDUAL HEAT REMOVAL (RHR) SYSTEM ...................... 3/4 5 11 1 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMA W CONTAINMENT Containrent Integrity...................... ............ 3/4 6-1 Containment Leakage.................. ................... 3/4 6 2 Containment Air Locks.................................... 3/4 6 5 Internal Pressure.................................... ... 3/4 6-7 Air Te perature........................................ . 3/4 6 8 Contain ent Structural Integrity......................... 3/4 6-9 Contain ent Ventilation Syster........................... 3/4 6-12 1/4.6.2 DEFR2SSURIZAT10N AND C00LIN3 SYSTEMS Containrent Spray System................................. 3/4 E 14 Spray Additive Syste...................................... 3/4 6-15 Containrent Cooling Syster. . . . . . . . . ......... .......... 3/4 6-17 AMEN 0 MENT N05. AN3 SOUTH TEXAS - UNITS 1 & 2 viii 0':V;y
o4 b INDEX LIMITING C0h0!TIONS FOR OPERAliM AND SURVEILLANCE REQUIREMENTS ' PAGE SECTION 3/4.6.3 CONTAINMENT ISOLATION VALVES............................. 3/4 6-18 3/4.6.4 COMBUSTIBLE GAS CONTROL Hydrogen Analyzers....................................... 3/4 6-19 Electric Hydrogen Recombiners............................ 3/4 6-20 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 4 LOOP 0PERATION......................................... 3/4 7-2 TABLE 3.7-2 STEA:1 LINE SAFETY VALVES PER L00P..................... 3/4 7-3 Auxiliary Feedwater System............................... 3/4 7-4 Auxiliary Feedwater Storage Tank......................... 3/4 7-6 Specific Activity........................................ 3/4 7-7 TABLE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM..................................... 3/4 7-8 Hain Steam Line Isolation Va1ves......................... 3/4 7-9 Atmospheric Steam Relief Valves ......................... 3/4 7-10 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION.......... 3/4 7-11 3/4 7-12 3/4.7.3 COMPONENT COOLING WAT ER SYSTEM. . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-13 3/4.7.4 ESSENTI AL COOLING WATER SYSTEM. . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-14 3/4.7.5 U LT I MAT E H E AT S I N K . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4.7.6 (This specification number is not used.) 3/4 7-16 3/4.7.7 CONTROL ROOM MAKEUP AND CLEANUP FILTRATION SYSTEM. . . . . . . . 3/4.7.8 FUEL HANDLING BUILDING (FHB) EXHAUST AIR SYSTEM.......... 3/4 7 19 3/4 7-21 3/4.7.9 SNUBBERS................................................. 3/4 7-26 FIGURE 4.7-1 S AMPLE PLAN 2) FOR SNUBEER FUNCTIONAL TEST. . . .. . . ... . . 3/4 7-27 3/4.7.10 SEALED SOURCE CONTAMINATION.............................. 3/4.7.11 (This specification number is not used.) 3/4.7.12 (This specification number is not used.) MONITORING.............................. 3/4 7-31 3/4.7.13 AREA TEMPERATURE 3/4 7-32 TABLE 3.7-3 ARE A T EMPERATURE MONITORING. . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-33 3/4.7.14 ESSENTIAL CHILLED WATER SYSTEM .......................... SOUTH T[XAS - UNITS 1 & 2 ix AMEN 0 MENT NDS. AO 17.' 1 ; 'c
l:'l? , INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS , PAGE SECTION' 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 SCHEDULE........................ 3/4 8 8 Shutdown................................................. 3/4 8-9 r 3/4.8.2 D.C. SOURCES , Operating................................................ 3/4 8-10 3/4 8-12 i TABLE 4.8-2 BATTERY SURVEILLANCE REQUIREMENTS..................... Shutdown................................................. 3/4 8-13 I 3/4.8.3 ONSITE POWER DISTRIBUTION 0perating................................................ 3/4 8-14 , Shutdown...............................:................. 3/4 8-16 , 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES Containment Penetration Conductor Overcurrent 3/4 8-17 ; Protective Devices ...................................... . 3/4.9 REFUELING OPERA 1;0NS ' CONCENTRATION...................................... 3/4 9-1 3/4.9.1 BORON 3/4 9-2 3/4.9.2 INST 9UMENTATION.......................................... 3/4 9-3 : 3/4.9.3 DECAY TIME............. ................................. 3/4 9-4 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS. . . . . . . . . . . . . . . . . . . . . . . . 3/4 9-5 , 3/4.9.5 COMMUNICATIONS........................................... 3/4 9-6 f 3/4.9.6 REFUELING MACHINE ....................................... 3/4 9-7 l 3/4.9.7 CRANE TRAVEL - FUEL HANDLING BUILDING. . . . . . . . . . . . . . . . . . . . 3/4.9.8 RESIOUAL HEAT REMOVAL AND COOLANT CIRCULATION 3/4 9-8 l High Water Leve1......................................... ' 3/4 9 9 Low Water Leve1.......................................... ' SYSTEM................. 3/4 9-10 , 3/4.9.9 CONTAINMENT VENTILATION ISOLATION ; 3/4 9-11 ; 3/4.9.10 WAT ER LEVEL - REFUELING CAVITY . . . . . . . . . . . . . . . . . . . . . . . . . . P b AND SOUTH TEXAS - UNITS 1 & 2 x AMEN 0 MENT N05. l 1 h .
FD INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS PAGE SECTION 3/4.9.11 WATER LEVEL - STORAGE POOLS Spent Fuel Pool ......................................... 3/4 9-12 In-Containment Storage Pool ............................. 3/4 9-13 3/4.9.12 FUEL HANDLING BUILDING EXHAUST AIR SYSTEM ............... 3/4 9-14 f 3/4.10 SPECIAL TEST EXCEPTIONS . t 3/4.10.1 SHUT 00WN MARGIN.......................................... 3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRICUTION LIMITS... 3/4 10-2 3/4.10.3 P HY S I C S T E ST S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 10-3 3/4 10 4 3/4.10.4 REACTOR COOLANT L00PS.................................... 3/4.10.5 POSITION INDICATION SYSTEM - SHUT 00WN.................... 3/4 10-5 3/4.11 RADIDACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS Concentration................ ........................... 3/4 11-1 00se..................................................... 3/4 11-2 Liquid Waste Processing System........................... 3/4 11-3 Liquid Holdup Tanks...................................... 3/4 11-4 3/4.11.2 GASEOUS EFFLUENTS ; Dose Rote................................................ 3/4 11-5 Dose - Noble Gases....................................... 3/4 11-6 Dose - Iodine-131, Iodine-133 Tritium, and Radioactive Material in Particulate Form............................. 3/4 11-7 Gaseous Waste Processing System.......................... 3/4 11-8 Explosive Gas Mixture.................................... 3/4 It-9 l Gas Storage Tanks........................................ 3/4 11-10 3/4 11 11 [ 3/4.11.3 SOLID RADICACTIVE WASTES................................. 3/4 11-13 l 3/4.11.4 TOTAL 00$E............................................... ( AND SOUTH TEXAS - UNITS 1 & 2 xi AMENDMENT N05. gQ.4 . \ 1 Gl%
y < fp INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PR0 GRAM....................................... 3/* 12-1 3/4.12.2 LAND USE CENSU5.......................................... s/4 12-3 3/4.12.3 INTERLABORATORY COMPARISON PR0 GRAM.................... .. 3/4 12-5 t
. t t
E i r i l t I l t a-SOUTH TEXA5 - UNITS 1 & 2 xit AMENDMENT N05. AND ! I's.'f,,,,) n. r
+ - . - . - ~ , - - - - - - ~ , - - _ - . . . - . , _ - - - , , - _ - - , , . _ . . ---_.-._-._.--,--_,.,.-,,.y -
,,y,. - - ~ , . - - , - - . - , . - -
FD : INDEX BASES SECTION PAGE r 3/4.0 APPLICABILITY............................................... B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS , 3/4.1.1 BORATION CONTR0L.......................................... B 3/4 1-1 I i 3/4.1.2 BORATION SYSTEMS........................................... B 3/4 1-2 l 3/4.1.3 MOVABLE CONTROL ASSEMBLIES................................ B 3/4 1-3 F 3/4.2 POWER DISTRIBUTION LIMITS................................... B 3/4 2-1 3/4.2.1 AXIAL FLUX DIFFERENCE..................................... B 3/4 2-1 . t ! 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR................. B 3/4 2-2 FIGURE B 3/4.2-1 TYPICAL INDICATED AXIAL FLVX DIFFERENCE VERSUS THERMAL P0WER............................................ !t 3/12-3 3/4.2.4 QUADRANT POWER TILT RAT 10..-.............................. B 3/4 2-5 3/4.2.5 DNB PARAMETERS............................................ B 3/4 2-5
- e 3/4.3 INSTRUMENTATION [
l 4 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION................ B 3/4 3-1 l 3/4.3.3 MONITORING INSTRUMENTATION................................ B 3/4 3-3 : 3/4.3.4 TURBINE OVERSPEED PROTECTION.............................. P. 3/4 3-6 l l 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION............. B 3/4 4-1 : 3/4.4.2 SAFETY VALVES............................................. B 3/4 4-1 j 3/4.4.3 PRtSSuRizER............................................... B 3/4 4-2 l 3/4.4.4 RELIEF VALVES............................................. B 3/4 4-2 j 3/4.4.5 STEAM GENERATORS.................................... 0 3/4 4 ? l 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE............................ B 3/4 4-3 l 3/4.4.7 CHEMISTRY................................................. B 3/4 4-4 t 3/4.4.B SPECIFIC ACTIVITY......................................... B 3/4 4-5 i i l f i xiii AMEN 0 MENT N35. AND SOUTH TEXAS - UNI 15 1 & 2 D'I: t;3 I i
,s FD ,
INDEX BASES SECTION PAGE
, 3/4.4.9 PRESSURE / TEMPERATURE LIMITS............................... B 3/4 4-6 TABLE B 3/4.4-la REACTOR VESSEL TOUGHNESS (UNIT 1)................ B 3/4 4-9 TABLE B 3/4.4-lb REACTOR VESSEL TOUGHNESS (UNIT 2)................ B 3/4 4-10 FIGURE B 3/4.4-1 FAST NEUTRON FLUENCE (E>1MeV) AS A FUNCTION OF J FULL POWER SERVICE LIFE.................................. B 3/4 4-11 3/4.4.16 STRUCTURAL INTEGRITY..................................... B 3/4 4-15 3/4.4.11 REACTOR VESSEL HEAD VENTS................................ B 3/4 4-15 1
3/4.5 EMERGENCY CORE COOLING SYSTEMS a. 3/4.5.1 ACCUMULATORS.............. ............................... 2 3/4 5-1 3/4.5.2 and 3/4.5.4 ECCS SUBSYSTEMS............................... B 3/4 5-1 3/4.5.5 REFUELING WATER STORAGE TANK.............................. B 3/4 5-2 3/4.5.6 RESIDUAL HEAT REMOVAL (RHR) SYSTEM ....................... B 3/4 5-3
- 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT....................................... B 3/4 6-1 l
3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS...................... B 3/4 6-3 3/4.6.3 CONTAINMENT ISOLATION VALVES.............................. B 3/4 6-4 l l 3/4.6.4 COMBUSTIBLE GAS CONTR0L................................... B 3/4 6-4 l 3/4.7 PLANT SYSTEMS I 3/4.7.1 TURBINE CYCLE......................................... ... 3 3/4 7-3 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION........... B 3/4 7-3 5 3/4.7.3 COMPONENT COOLING WATER SYSTEM............................ B 3/4 7-3 3/4.7.4 ESSENTIAL COOLING WATER SYSTEM............................ B 3/4 7-3 ! . 3/4.7.5 ULTIMATE HEAT SINK........................................ B 3/4 7-3 l 3/4.7.6 (Not used)' 3/4.7.7 CONTROL ROOM MAKEUP AND CLEANUP FILTRATION SYSTEM ........ B 3/4 7 4 j 3/4.7.8 FUEL HANDLING BUILDIN3 EXHAUST AIR SYSTEM................. B 3/4 7 4 3/4.7.9 SNUBBERE.................................................. B 3/4 7-4 l \ i xiv AMEN; MENT N05. AND i SOUTH TEXA5 UNITS 1 l. 2 W j ,.
s FD J_NDEX r BASES PAGE SECTION 3/4.7.10 SEALED SOURCE CONTAMINATION................ ............. B 3/4 7-6 3/4.7. H (Not used) 3/4.7.12(Notused) 3/4.7.13AREATEMPERATUREMONITORING............................... D 3/4 1-6 3/4.7.14 ESSENTIAL CHILLED WATER SYSTEM............................ B 3/4 7-6 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 DNSITE POWER DISTRIBUTION ............................... B 3/4 6-1 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES................... B 3# 1-3 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION....................................... B 3/4 9-1 3/4.9.2 INSTRUMENTATION........................................... B 3/4 9-1 3/4.9.3 DECAY TIME................................................ B"3/4 9-1 3/4.9.4 CONT AINMENT BUILDING PENETRATI0'45. . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 9-1 3/4.9.5 C0?F.UNICATIONS............................................ B 3/4 9-1 3/4.9.6 REFUELING MACHINE.,....................................... B 3/4 9-2 3/4.9.7 CRANE TRAVEL - FUEL HANDLING BUIL0!NG..................... B3/49-2 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCUL4? ION............. B 3/4 9-2 3/4.9.9 CONTAINMENT VENTILATION ISOLATION SYSTEM. . . . . . . . . . . . . . . . . . B 3/4 9-2 3/4.9.10 and 3/4.9.11 WATER LEVEL - REFUELING CAVIT) and B 3/4 9-3 STORAGE P00LS............................................. B 3/4 9-3 3/4.9.12 FUEL HANDLIN3 BUILDIN3 EXHAUST AIR SYSTEM ................
. 3/4.10 SPECIALTESTEXCEPT109 B 3/4 10-1 3/4.10.1 SHUTDOWN MARGIN...........................................
3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS.... B 3/4 10-1 B 3/4 10-1 3/4.10.3 PHYSICS tee.S............................................. B 3/4 10-1 3/4.10.4 REACTOR COCLANT L00PS..................................... 8 3/4 10 1 3/4.10.5 POSITION INDICATION SYSTEM - SHUT 00WN..................... xy AMEN 0 MENT N05. AND SOUTH TEXAS - UNITS 1 & 2 v' m. I i @
FD INDEX BASES PAGE SECTION 3/4.11 RADI0 ACTIVE EFFLUENTS B 3/4 11-1 3/4.11.1 LIQUID EFFLUENTS......................................... B 3/4 11-2 3/4.11.2 GASEOUS EFFLUENTS........................................ B 3/4 11-6 3/4.11.3 SOLID RADI0 ACTIVE WASTES................................. B 3/4 11-6 3/4.11.4 10T AL 00 S E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING
........................ B 3/4 12-1 3/4.12.1 MONITORING PROGRAM............
B 3/4 12-1 3/4.12.2 LAND USE CENSUS.......................................... B 3/4 12-2 3/4.12.3 INTERLABORATORY COMPARISON PR0 GRAM....................... i 1 l J i i I I . i i I AND ! xvi AMENDMENT N05. SOUTH TEXAS - UNITS 1 & 2 1 uo gy i
7
/9 INDEX DESIGN FEATURES PAGE SECTION 5.1 SITE 5-1 5.1.1 EXCLUSION AREA.................................................
5-1 5.1.2 LOW POPULATION Z0NE............................................ 5.1. 3 MAP DEFINING UNRESTRICTED AREAS AND SITE BOUNDARY FOR 5-1 RADI0 ACTIVE GASEOUS AND LIQUID EFFLUENTS.......................
- 5. 2 CONTAlNMENT 5-1 5.2.1 CONFIGURATION..................................................
5-1 5.2.2 DESIGN PRESSURE AND TEMPERATURE................................ 5-2 FIGURE 5.1-1 EXCLUSION AREA.......................................... 5-3 FIGURE 5.1-2 LOW POPULATION 20NE..................................... FIGURE 5.1-3 RESTRICTED AREA AND SITE BOUNDARY FOR RADI0 5-4 ACTIVE GASEOUS EFFLUENTS....................................... FIGURE 5.1-4 RESTRICTED Ak?A AND SITE BOUNDARY FOR RADI0 ACTIVE 5-5 LIQUID EFFLUEhTS ....................................... 5.3 REACTOR CORE 5-6 5.3.1 FUEL ASSEMBLIES........... .................................... 5-6 5.3.2 CONTROL ROD ASSEMBLIES......................................... 5.4 REACTOR COOLANT SYSTEM 5-6 5.4.1 DESIGN PRESSURE AND TEMPERATURE................................ 5-6 5.4.2 V0LUME......................................................... 5-6 5.5 METEOROLOGICAL TOWER LOCAT10N.................................... 5.6 FUEL ST0_Rg ; 5-6 i 5.G.1 CRITICAllTY.................................................... L*7 L. 2 CRAlNACE....................................................... 57 5.6.3 CAPACITY....................................................... 5-7 5.7 COMPONENT CYCLIC OR T R ANSI ENT LIMIT . . . . . . . . . . . . . . . . . . . . . . . . . . . FIGURE 5.6 1 SOUTH TEXAS PROJECT SPENT FUEL RACKS - REGION 5-8 2 REQUIRED BURNUP AS A FUNCTION OF INITIAL ENRICHMENT., LIMITS.................. 59 TABLE 5.7-1 COMPONENT CYCLIC 04 TRANSIENT AMENCMENT dS. ANC SOUTH TEXAS - UNITS 12, 2 wii
- g. . ,,
f ' s. . . . .
b _INDEX ADMINISTRATIVE CONTROLS _ PAGE _SEC710N 6-1 L1_ RESPONSIBILITY................................................ . 6.2 ORGANIZATION
............... 6-1 6.2.1 0FFSITE AND ONSITE ORGANIZATIONS...........
6-1 6.2.2 UNIT STAFF.................................................. TABLE 6.2-1 MINIMUM SHIFT CREW COMPOSITION-TWO UNITS WITH TWO 6-4 SEPARATE CONTROL R00MS................................ 6.2.3 INDEPENDENT SAFETY ENGINEERING GROUP (ISEG) 6-6 Function................................................... 6-6 Composition................................................ 6-6 Responsibilities........................................... 6-6 Records.............. ..................................... 6-6 6.2.4 SHIFT TECHNICAL ADVIS0R..................................... 6.3 (No(Used) 6-7 6.4 TRAINING......................................................
................................... 6-7 6.5 REVIEW AND AUDIT..........
6.5.1 PLANT OPERATIONS REVIEW COMMITTEE (PORC) 6-7 Function................................................... 6-7 Corposition................................................ 67 Alternates................................................. 6-7 Meeting frequency.......................................... 6-7 Quorum..................................................... 68
. Respansibilities...........................................
6-9 Records.................................................... xviii AMECMENT N05. AG SOUTH TEXAS UNITS 1 & 2 ii..;,....
I _k *, FD i INDEX ADMINISTRATIVE CONTROLS 4 PAGE ; SECTION 6.5.2 NUCLEAR SAFETY REVIEW BOARD'(NSRB) Function.............. .................................... 6-9 1 Composition................................................ 6-10 l Alternates................................................. 5-10 ~ C o n s u l t a r t s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... . . . 6-10 ; Meeting Frequency.......................................... 6-10 .[ 6-10 Quorum..................................................... ! Review..................................................... 6-10 : Audits..................................................... 6-11 l Records.................................................... 6-12 7 t 6.5.3 TECHNICAL REVIEW AND CONTROL f Activities ............... ................................ 6-12 . I t 6.6 REPORTABLE EVENT ACTI0N...................................... 6-13
- 6. 7. SAFETY LIMIT V10LATION....................................... 6-13 j t
r 6.A PROCEDURES AND PR0 GRAMS...................................... 6-14 ! 1 6.9 REPORTING REQUIREMENTS ! 6.9.1 ROUTINE REP 0RTS............................................ 6-16 ( Startup Report............................................. 6-16 Annual Reports............................................. 6-17 l Annual Radiological Environmental Operating ottert......... 6-17 l Semisnnual Radioactive Effluent Release Report............. 6 18 r r
- Monthly Operating Reports.................................. 6-20 ,
Radial Peaking Fa: tor Limit Report......................... 6-20 f l 6.9.2 SPECIAL REP 0RTS............................................ 6-20 6.10 RECORD RETENT10N........................................... 6-21 l i ( SOUTH TEXAS UNITS 1 & 2 xix AuEN0 MENT N35. AND : tiQ y ; 7 u. ,..
ED IA l .' i INDEX
~
a , f e ADMINISTRATIVE CONTROLS
'PAGE SECTION 6-22 6.11 RADIATION PROTECTION PR0 GRAM................................
6-22 C.12 H IGH R ADI AT I ON AR E A . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . (PCP).............................. 6-23
'6.13 PROCESS CONTROL PROGRAM 6-23 6.14 0FFSITE DOSE CALCULATION MANUAL (00CM).. ...................
6.15 MAJOR CHANGES TO LIQUID. GASEOUS. AND SOLIO SYSTEMS..............l................... 6-24 RADWASTE TREATMENT s. t b IF )i j AHENDMINT N05. AND $ xx SOUTH TEXAS - UNITS 1 & 2 45 . L' l'
FD SECTION 1.0 DEFINITIONS
- I0's 1 i 1; ;
F0 1.0 DEFINITIONS The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications. ACTION 1.1 ACTION shall be that part of a Technical Specification that prescribes remedial measures required under designated conditions. ACTUATION LOGIC TEST 1.2 An ACTUATION LOGIC TEST shall be the application of various simulated 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 OPERATIONAL TEST shall be the injection of a simulated signal into the channel as close tG the sensor as practicable to verify OPERABillTY of alarm, interlock and/or trip functions. The ANALOG CHANNEL ' OPERATIONAL TEST shall include adjustments, as necessary, of the alarm, inter-lock and/cr Yrip Setpoints so that the Setpoints are within the required range and accurary. AXIAL FLUX DIFFERENCE 1.4 AX1AL FLUX OlFFERENCE shal' be the difference in normalized flux signals between the top and bottom halves of a 2-section excore neutron detector. CHANNEL CALIBRATION 1 1.5 A CHANNEL CAllBRATION shall be the adjustment, as necessary, of th? channel so that it resDonds within tho required range and accurecy to kno.n vclues of input. The CHANNEL CAllBMTION shall encompass the entire channel including the sensors .snd alarm, interlock, and/or trip f unctions and may be perforred by any series of sequential, overlapping, or total channel steps so that the entire channel is calib-ated. CHANNEL CHECA ,
- 1. E. A CHANNEL CHECA shall be the Qualitathe assessment of channel behavior dui ing operation by observation. 'his determination shall include, where
) possQle, coed 3rison of the channel iridication and/or status with other indications and/or status cerived fror. indapendent instrutent channels measuri19 the sare parameter.
^ A"EN 3ENT N05. AN:
50VTm TE) A5 - UM' 5 1 - , uw
DEFINITIONS CONTAlhMENT INTEGRITY
- 1. 7 CONTAINMENT INTEGRITY shall exist when:
- a. All penetrations required to be closed during accident conditions are either:
- 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 Specification 3.6.3.
- b. All equipment hatches are closed and sealed,
- c. Each air lock is in compliance with the requirements of Specifica-tion 3.6.1.3,
- d. The containment leakage rates are within tt- limits of Specifica-tion 3.6.1.2, and
- e. The sealing mechanism associated with each penetration (e.g., welds, hellows, or 0-rings) is OPERAELE.
C0hTROLLED LEAKAGE
- 1. C CONTROLLED t.EAAAGE shall be that seal water ficw s'Jppl'ed to the raatter coLlant pump seals.
CORE ALTERATIONS 1.9 CORE ALTERATI0h5 ' hall be the Movement or manipulatior of any component within the reactor t.ressure vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATION shall not precluoe completion of movement of a component to a safe conservative position. DIGI TAL CHANNEL OPER ATIONAL TEST 1.10 A DIGITAL CHANNEL ')PERATIONAt. TEST shall consist of injecting simulated process data where available or c>ercising the digital computer hard,,are using data base manipulatian to serify OPERABILITY of alarm, interlock, and/or trip functions. 005E EqillyALENT !-131 1.11 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie / grad which alone would produce the saPe thyroid Core as the quantity and isotopic mixtcre of I-131, 1-132, 1-133, I-134, and 1-135 actually present. Tne thyroid dose conversion factors used for this calculation shall be those listed in Table E-7 of N;' R gJlatorf Guide 1.109, Resision 1, 0:tcoor 1977 50JTM TEAAS - UN!is 1 !. : 1-2 A"ENDMENT N 5. A',: b* 1;g
FP DFFINITIONS E - AVERAGE O!SINTEGRATION 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 gam:n energies per disintegration (MeV/d) for the isotopes, other than iodines, uitii iisi f lives greater than 15 minutes, m' king up at least 95% of the total non-iodine activity in the coolant. ENGINEERED SAFETY FEATURES RESPONSE TIME 1.13 The ENGINEERED SAFETY FEATURES (ESF) RESPONSE TIME shall be that time interval from when the monitored parameter esceeds its ESF Actuation Setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable. FREQUENCY NOTATION , 1.14 The FREQUENCY h3TATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1. CASEOUS WASTE PROCESS 1h3 SYSTEM 1.15 A GASEOUS WASTE PROCtSSIN3 SYSTEM shall be any systen designed and installed to redxe radioactise gaseous effluents by collecting Reactor Coolant System offgases fron the Reactor Coolant System Lnd providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environeent. IDENTIFIED LEAKAGE 1.16 IDENTIFIED LEAAAJE shall be:
- a. Leakage (except CC', TROLLED LEAKAGE) into closed systecs, such as pu o seal or vala pa:* ing leaks that are captured and concutte:: to a su ?
or collecting tank, or
- b. Leakage into the cuntainment atmosphere from sources that are both specifically located and knoan either not to interfere with the operation cf Leakage Detection Systems or not to be FRESSURE EDUN"AD LEAAAGE, or
- c. Rea: tor Coelant Syste- leakage through a stea generator to the Secondary Coolant System.
SOUTH T E ) : 5 - L",: T 5 1 & 2 13 AMEh: VENT N:3 A',:
~
1ifili l
FD DEINITIONS VASTER 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. MEMBER (S) 0F THE PUBLIC 1.18 HEMBER(S) 0F THE PUBLIC shall inciude 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 equipment 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 (00CM) sha'l 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 Rediological Monitoring Program. OPERAELE - OPERABILIT) 1.20 A system, subsyster, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s), and when all necessary attendant instrumentatio7, controls, electrical poner, cooling or seal water, lubrication or other auxiliary equipment that are required for the sy% tem, subsystem, train, component, or device to perform its function (s) are also capable of performing their related support function (s). OPERATIONAL MODE - M:0E 1.21 An OPERATIONAL MD E (i.e., M00E) shall Correspond to any one inclusive combination of core reactivity condit;on, power level, and average rea: tor coolant terperature specified in Table 1.2. PHYSICS TESTS 1.22 FHYSICS TEST $ shall be those tests performed to reasure the fundamental nuclear characteristics of the rea: tor core and related instrumentation: (1) described in Cnacter 14.0 of the FSAR, (2) authorized under the provisions of 10 CFR 50.59, or (3) other ise 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 Syster cocoonent booy, pipe wall, er vessel wi ll. SOUTW T E s 5 - U' ' T 5 1 E 2 ;-4 AMENDENI N05. A: t,:; y 7 g
I F.D i DEFINITIONS PROCESS CONTROL f'ROGRAM 1.24 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, tests, and determi'ations to be made to ensure that processing and pai!15lng of solid radioactive wastes based on demonstrated processing of ac' :M nr simulated wet solid westes will be accomplished in such a way as to e compliance with 10 CFR Parts 20, 61, and 71 and Federal and State r ; e lation,, t uri 1 ground requirements, and other require-
~
nents governing tb s P *adivactive waste. PURGE - PURGIN3 1.25 PURGE or PURGIN' W ; be any e.ontrolled process of discharging air or gas from a confinement te ' M 1.1 . "' ture, pressure, humidity, concentration or other vperating co. 9 . . ', n
- a manner that replacement air or gas is required to purify the intron +
QUADRANT POWER TILT RATIO 1.26 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore cetector calibrated output to the avera;e of the upper excore detector cali-brated outputs, or the ratio of the maximum lo er excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater. With one excore detector inoperable, the remaining three detectors shall be used for computing the average. RATED THERMAL POWER 1.27 RATED THEPMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3800 W t. REACTOR TRIP SYSTEM RESPONSE TIME 1.25 The REACTOR TRIP SNSTEM NESPONSE TIME shall be the time interval from when the monitored pcra eter exceeds its Trip ietpoint at 19e channel sensor until loss of stationary grippar coil voltage. REFORTAELE EVENT 1.29 A REPORIABLE EVENT shall be any of those conditions specified in Section 50.73 of 10 CFR Part 5's SHUTDOWN MARGIN 1.30 SHUTDCW's MAR 3IN shall be the instantaneous amount of reactivity by which the reactor is sLbcritical or would be subtritical from its present condition assum.ing all full-length red cluster asse3blies (shutcc.n and centrol) are fully inserted except for the single rod cluster asse-bly of highest reactivity worth which is assuted to be fully withdra n. SC L'T - T E 3 A s - L ',I T S 1 L : 1-5 AVE h;"E',T NM . A' ; C 1 i 15:;
fD DEFINITIONS SITE BOUNDARY 1.31 The SITE BOUNDARY shall be that line befand which the land is neither owned, nor leased, nor otherwise controlled by the licensee. SLAVE RELAY TEST 1.32 A SLAVE RELAY TEST shall be the energiz. d on 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
, 1.33 SOLIDIFICATION shall be the conversion of wet wastes into a form th.?t meets shipping and burial ground requirements.
SOURCE CPECK 1.34 A SOU0CE CHECK shall be the qualitative assestment of channel response when the channel sensor is exposed to a source of increased radioactivity. STAGGERED TEST BASIS 1.35 A S'A3GERED TEST BASIS shall consist of:
- a. A test schedule for n systems, subsystems, trains, or other des lg-nated components obtained by divioing the specified test interval into n equal subintervals, and
- b. The testing of one system, subsystem, train, or other designated component at the beginning of each subinterval.
THERMAL POWER 1.36 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant. TRIP AC'UATIN3 CEVICE OPERATIC',AL TEST 1.37 A TRIP ACTUATINS DEVICE OPERATIONAL TEST shall consist of operating the Trip Actuating Device and serifying OPERABILITY of alarr, interlock and cr trip functions. The TRIF A:TUATIN3 DEVICE OPERATIONA'. TEST shall incluce adjustrent, as necessar.i, of the Trip Acteating Device so that i' actuates at the required Setpcint within the required accuracy. UNIDENTIFIED LEAKAGE 1.3S UNIDENTIFIED LEANAGE shall be all leakage which is not IDENT:FIEC LEAKAGE or CONTROLLED LEAKAGE. SCUTH TEtAS - U',: TS1L2 1-E AMENDMEN' N;5. A'C .
,.<. l u. .
FD DEFINITIONS UNRESTRICTED AREA 1,39 An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials, or any area within the SITE BOUNDARY used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes. VENTING 1.40 VENTING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration, or other opercting 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. SOUTH TE ui - UNI's 1 & 2 1-7 AVEN:"ENT N:5. A'C HOV 1 ; h
7 Pb TABLE M FREQUENCY NOTATION NOTATION FREQUENCY S At least once per 12 hours. D At least once per 24 hours. 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. 50'J~ H 1 E x t.5 - UN I T S 1 !. 2 1-$ AMENDMENT N05. AN 1 ', G3.
m FD TABLE 1.2 OPERATIONAL MODES REACTIVITY % RATED AVERAGE COOLANT MODE CONDITION, K,.gf THERMAL POWER
- TEMPERATURE
- 1. POWER OPERATION > 0.99 > 5% > 350'F
- 2. STARTUP > 0.99
< 5% > 350*F
- 3. HOT STANDBY < 0.99 0 > 350'F
- 4. HOT SHUTENN < 0.99 O 350*F > T
> 200 g avg
- 5. COLD SHUTOOWN < 0.99 0 1 200'F
- 6. REFUELIN3** 1 0.95 0 1 140 F
- Excluding decay heat.
** Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.
SCUTH TEKA5 - UNIT 51 !. 2 1-9 ANEN:"ENT '4 5. AN:
.3
Fp , SECT 10t4 2.0 SAFETY LIMITS AtiD LIMITita SAFET) SNSTEti SETTit4GS 9
Fp 2.0 SAFETY LIMITS ANS LIMITING GAFETY SYSTEd_SEI_*lNGS , 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating Figure 2.1-1.loop coolant temperature (T,y9) she.Il not exceed the limits shown in APPLICABILITY: MODES 1 and 2. ACTION: Whenever the point defined by the combination of the highest operating ' cop average temperature and THERMAL POWER has exceeded the appropriate pre,surizer pressure line, be in HOT STANDEY within 1 hour, and comply with the require-ments of Specification 6.7.1. REACTOR COOLANT SYSTEM PRE 55URE 2.1.2 The Reactor Coolant System pressure shall not ex.ceed 2735 ps{g. APPLICABILITY: MDDE5 1, 2, 3, 4, and 5. ACTION: MODES 1 and 2: Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANDEY with the Reactor Coolant System pressure within its limit within I hour, and comply with the requirements of Specif' cation 6.7.1. MODES 3, 4 and 5: 1 Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes, ana corply with the requirements of Specification E.7.1. 5 : '. T - TE G5 - L',:15 . L : 2-1 A"ENC"ENT N25. A'O
' ' Gil -
. . . . . . . . ~ . _ _ . .
s . .. FC 4 ESO ma::rPira r 2a00 PSla = (0.662.I (0.70.642.4)
;0.652.7) ,
1 2250 PSta I % (0.90.632.5)
. i N. %
L (0.635.07\ . 2000 P514 10.90.62 % 1).
' 2,* ' ,3
- 8
- 3 3 '
b g
$ to.e:s.e)- I N m gi ,,) .\
A ,ee,,i \
.5 - iseu Psra
- N \\\_.
e goo \ N ie6.6DD X \J o i h C - 7 (i.ic. sac.i) ~ e-
/
. (l.20.532.t) ^ '
i ti.2o.seo.3) # ( **') 4:: tat 4At i i 560 I i . 540 t.20 0 .20 40 .60 ,80 I.CO FRACTION OF RATED THERMAL POWER FIGURE 2.1 1 i REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN CPERATION 1 AND ' 50'JTH TEXAS - UNITS 1 & 2 2-2 AMEh0 MENT NOS.
.. *
- I '
F.D SArETY LIMITS AND LIMITING SAFETY SYSTEM _ SETTINGS 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The Reactor Trip System Instrumentation and Interlock Setpoints shall be set consi, tent with the Trip Setpoint values shown in Table 2.2-1. APPLICABILITY: As shown for each channel in Table 3.3-1. ACTION: .
- a. With a Reactor Trip System Instrumentation or Interlock 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 Instrumentation or Interlock Setpoint less conservative than the value shown in the Allowable Value column of Table 2.2-1, either:
- 1. Adjust the Setpoint consistent with the Trip Setpoint value of Table 2.2-1 and determine within 12 hours that Equation 2.2-1 was satisfied for the affected channel, or
- 2. Declare the channel inoperable and apply the applic4ble ACTION stateTent requirerent of Specification 3.3.1 until the channel is restored to OPERABLE status with its Setpoint adjusted consistent with the Trip Setpoint value.
Equation 2.2-1 Z + R + 5 $ TA Where: 2 = The value frc Colutn 2 of Table 2.2-1 for the affected channel, R = The "as- easured" s alue (in percent span) of rack error for tre affecte: channel, S = Either the "as-ressured" value (in percent span)'of the sensor
~
error, or the value from Column 5 (Sensor Error) of Table 2.2 1 for the affected channel, and TA = The salue from Celu n TA (Total Allowance) of Table 2.2-1 fer the affected channel i
'o TEgAs - L',l's 1 & : 2-3 AMEN:"INT N:5. AN:
r
, , , TA8LE 2.2-1 o 'p R[ ACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS M TOTAL SENSOR T. ALLOWANCE ERROR '^
i t!NC T IONAL UNIT (TA) Z (5) TRIP SETPOINT All0WABLE VALUE
- 1. f tmual Reactor irip N.A. N.A. N.A. N.A. N.A.
*j ? Power Rango, Neutron Flus 4.56
- a. liigh setpoint 1. 5 0 <109% of RTP** <111.3% of RTP**
- b. Low Setpoint R. 3 4.56 0 125% of RTP** <27.3% of RTP**
- 3. Power Range, Neutron flux. I.6 0.5 0 <5% of RTP** with <6.3% of RTP** with liigh Positive Rate a time constant a time constant 12 seconds >2 seconds 4 Power Range, Neut ron I t u r , 1.6 0.5 0 <5% of RTP** with <6.3% of RTP** with liigh Nerptive Rate a time constant 3 time constant
>2 seconds ->2 seconds
- n. -
8- ',. lutermediate Ranqe, 17.0 8.41 0 <25% of RTP**
<31.1% of RTP**
~
Neutron flux
- 6. Source Range, Neutron flux 17.0 10.01 0 <105 cps $1.4 x 105 cps I. Overtemperature AT 6.8 4.66 1.5 + 0.9# See Note 1 See Note 2 H. Overpower AT 5. 5 1.74 1. 5 See Note 1 See Note 4
- 9. Pressurizer Pressure-Low 3.1 0.71 2. 0 >1870 psig >1862 psig r-A 10 Pressurizer Pre <.sure-High 3.1 0.71 2.0 <2380 psig 12388 psig th II. Pressurizer Water level-High 5.0 2.76 2.0 <92% of instrument 193.6% of instrument 7.' span span
[ !?. Reactor Coolant iIow-Low 4.0 3.19 0.6 >91.fC of loop >90.9% of loop y design flow
- ifesign flow *
~'i nop design f low = 95.400 opm
** RIP - RAllD THERMAL POWER
# 1. ';% span f or ai; 9.'C span f or Pressurirer Pressure l
- 9 g .
9
- v. TABLE 2.2-1 (Continued)
'd REACTOR TRIP SYST[*4 INSTRtrMENi ATION TRIP SETPOINTS i'
-4
'" total SENSOR ALLOWANCE ERROR
?A IlfNCTIONAL UNIT (TA) Z (S) TRIP SEfPOINT ALLOWABLE VALUE
[ 11 *tcam Generator W. iter 15.0 12.75 2.0 + 0.7## >33% of narrow l evel low-tow ->31.5% of narrow f; range ir.strument range instrument
;,1 span span
- 11. IInlervn1tage - Reactor lit. 5 0.3 0 >:0,014 volts >9408 volts c~~ _
Conlani. hn', l '. . tiniferfrequency - Reactor 3.4 0.01 h >$1.2 Hz >57.1 Hz Coolant Pumps I f. . Tur!>ine Trip e., a. Low Emergency Trip fluiel 237.1 100.8 131.3 >1245.8 psig >1114.5 psig J. Pressure ti. Turbine Stop Valve .N.A. N.A. N.A. < fully closed fully closed Closure I 7. 'afety Injettinn input N.A. N.A. N.A. N.A. N.A. from ESIA5 3 .5i0 <. pan for '. team Generator level; 0.7% span for Reference leg RIDS 6, _ O s p G
.n, b
mo e e n n H t i b re re E p
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- T u * *
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< $ > <I t N N I
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?
i I r - n P, n9 n1 l . P ps d- eP 0 3 a- a- a- e p r f n w 1 1 RP RP RP er i T eo 1 T ird o , - - r , r , nu r 1 T T o r r Ph P P r , is T c N l et c ex ex ex bs i D U rr t u wo wu wu wu re ur r t o a l oe ne ol ) ) 2 ol PI ol PF ol Pf TP t mc l A L t t I N 1B 1 A cn c oi R N aI a t q 0 e .
,. . e uo =
1 R a t c d. e f R AI P 1 I C N . . R u,R. 9 0
- I 1 1 7
, 1, f,
c~; gcu 7* 7; ' ' T h 4 g- 2 , i
TABLE 2.2-1 (Continued) E TAITLE f10TATIONS S fn f f 1. OVERiff4P[RAltJRE AT I I I *
" ST(I*'1[)(1+v3, 1 + 1 ;- , , ) < aio { K, - K 2 i
( 1
- vr.5 ) ( T (- !
- 1,;5) - T ' ) + K3 ( P - P' ) - f ( AI ) }
F. I; Where: SI = firasured ST by RCS Instrumentation; I* 1 h v2s
- tead-lag compensator on measured aT; r>
t,, = Tim constant utilized in lead-lag compensator for AT, 1 = 8 sec, v2 r 3 sec;
- 2 I
3, r I aq compensator on measured AT;
'> 13
= Time constant utilized in the lag compensator for AT, 13 = 0 sec; ai, = Indicated AT at RATED illERtML POWER; K, = 1.08; K2 = 0.0185/*f; I*t S 3,
= The function generated by the lead-lag compensator for T y;
dynamic compensation; T] i . , t r, = Time constants utilized in the Icad-lag compensator for T ,14 = 33 sec,
.; v,. = 4 sec; T = Averaqa temperature,I;
.i.
=
I+ r ,. .3 1aq compensator on measured Iavq;
=
,, v .. Time constant utilized in the measured T,y lag compensa*or, rs = 0 sec;
-7 l'
d h e
TABLE 2.2-1 (Continued)
} TABLE NOTATIONS (Continued) g NOII 1: (Continued)
T-
'e T' < 593.0T (flominal T avq at RATED THERMAL POWER);
3 K, = 0.000857/psig;
*J
'e P = Pressurizer pressure, psig;
& P' = ??35 psig (Nominal RCS operating pressure);
F) S = 1aplace t ransfonn operator, sec '; and f,(AI) is a function of the indicated difference between top and bottom detectors of the power- range neutron ion chambers; with gains to be selected based on measured instrument response during plant startirp tests such that: 3 (1) Fores t q b
' **"" ~ ' '
' '(a ) , w ere qt "" 9i > are Percent RARD THN POWfR in the top and bottom halves of the core respectively, and qt *U is total THERMAL b
POWIR in percent of RATID THERMAL POWER; (2) for each percent that the mary.itude of q a exceeds -39%, the AT Trip Setpoint shall b be automatically reduced by 1.55% of its value at RATED THERMAL POWER; and q (3) for each percent that the magnitude of q a exceeds +10%, the AT Trip Setpoint shall ls [c be automatically reduced by 1.52% of its value at RATED THERMAL POER. ~ [ fl01I ?- The channel's maximum Trip setenint shall not exceed its computed Trip Setpoint by more than o, 1.7% AI span. o E
,~
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- w. L % % ,
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>= C C C w #O &
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.- C # . 4 4 ea o g et ed C U - os rc et o w L L U M + 4' w m 7 0 4 Z ~4 ec : 4 M O ww L ed oe s == &
N >= > h
. aC mm og .O C.
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- ed e4 .-e w T w - + e ec .a aC J 4 4 0 4 0 0 ec L - &
p co m e ee ee ed es ea G G .= -.0a ** st ww C o o o o o L C +3 C C P- Z Z Z Z Z Z V 4 3 2 2
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w e e w o.
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.= 0 T T T *C % T N O $ T T
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) e ( u d u T p e n R m u i r E o n t o W c i n f O t c P s n C 0 t o ( l i C = A t
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. A n "0 n e 2 T a D . a 0 E 3 t E.
4 f " , T 9 , o 8 T 1 A 5 1 n B E R _ e e A L > 5 l T B t t t l A T o a , o a 1 l f n l f h r 9o s o n Vi . t f i at i l i a A n I d t d i
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/ n e e n l p 5 i t m i a t 3 f a u f e 1 e c rt e r S 0 d i d o .
0 d s f p s n n s i _
- 0. A I i A 0 T
r
= = = = = m u
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m-FD BASES FOR SECTION 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS { 9
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R NOTE The BASES contained in succeeding pages summarize the reasons for the Specifications in Section 2.0, but in accordance with 10 CFR 50.36 are not part of these Technical Specifications. h e
Fi 2.1 SAFETY LIMITS BASES
- 2.1.1 REACTOR CORE The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.
Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure fror, nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter uuring operation and therefore THERMAL POWER and reactor coolant temperature and pressure have been related to DNB through the W 3 R-Grid correlation. The W-3 R-Grid DNB correla-tion has been developed to predict the DNB flux and the location of DNB for axially uniform and nonuniform heat flux distributions. The local DNB heat flux ratio (DNBR) is defined as the ratio of the heat flux that would cause DNS at a particular core location to the local heat flux and is indicative of the
! margin to DNB.
The minimum value of the ONBR 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 ONB for all operating conditions. The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant Systen. pressure and ave age temperature for which the minimum DNBR is no less than 1.30, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid. Thesecurvesarebasedonanenthalpyhotchannelfactor,Fh,of1.52and a reference cosine with a peak of 1.61 for axial po er shape. An allowance is included for an increase in Fh at reduced po.er based on the empression: Fh=1.52(10.3(1-P)) Where P i.s tt.e fraction of RATED THERMAL POWER. These limiting heat flux conditiens are higher than those Calculated for the range of all control rods fully withdra n to the maxim o allodble control l rod inser* ion assuming the axial po.er imA*'ance is within the limits of the f t (41) function of the Overteeperature te ,, When the axial power imbalance is not within the tolerance, the axial power iebalance effect on the Over-temperature AT trips will reduce the Setpoints to provide protection consistent with core Safet) Lieits. i I SOUTH TEXAS - UNI's 1 & 2 E 2-1 AVEN; VENT N05, ANO ' ; 1;;- l tD
FL > SAFETY LIMITS RASES I I 2.1.2 REACTOR COOLANT SYSTEM PRESSURE l The restriction of this Safety Limit protects the integrity of the Reactor i Coolant System (RCS) from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere. i The reactor vessel, pressurizer, and the RCS piping, valves, and fittings ! are designed to Section III of the ASME Code for Nucleer Power Plants which t parmits 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 Iaquirements. { The entire RCS is hydrotested at 1254 (3110 psig) of det,ign pressure, to ! demonstrate integrity prior to initial operation. f l l i I 9 1 t f l I l 1 SOUTH TEXAS - U'dTS 11. 2 B 2-2 AYEN; MENT O S. A'O l
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- 2. 2 LIMITING SAFETY SYSTEM SETTINGS RASES !
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! 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS j The Reacter Trip Setpoint Limits specified in Table 2.21 are the nominal i ; values at which the Reactor trips are set for each functional unit. The Trip !
l Setpoints have been selected to ensure that the core and Reactor Ceolant i System are prevented from axceeding their safety limits during normal operation and design basis anticipated operational occurrences and to assist the Engi-f i neered Safety Features Actuation System in mitigating the consequences of 4 accidents. The Setpoint for a Reactor Trip System or interlock function is
. considered to be adjusted consistent with the nocinal value when the "as- 7 j
measured" Setpoint is within the band allowed for calibration accuracy. ; To accommodate the inst'ument drift assumed to occur between operational f l : tests and tte accuracy to wMch Setpoints can be measured and calibrated. l l Allowable Values for the Reactor Trip Setpoints have been specified in l ! Table 2.2 1. Operation with Setpoints less conservative than the Trip Set- . point but within the Allowable Value is acceptable since an allowance has been { l made in the safety analysis to accommodate this error. An optional provision j j 2 has been ircluded 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 reasured" deviation from the specified calibration ! point for rack and sensor components in conjunction with a statistical combi-
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j nation of the other uncertainties of the instrueentation to reasure the process 7
; variable and the uncertainties in calibrating the instrumentation. In Equa- l
? tion 2.2-1, 2 + R +.5 < TA, the interactive ef fects of the errors in the rack l 5 and the sensor, and the "as-measured" values of the errors are considered. 2, , j as specified in Table 2.2-1, in percent span, is th statistical summation of ( i errors assumed in the analysis excluding those associated with the sensor and ; rack drift and the accuracy of their reasurement. TA or Total Allowance is . the dif ference, in percent span, between the Trip Setpoint and the value used l l in the analysis for Reactor trip. R er Rack Error is the "as-reasured" devia- ; j tion, in percent span, for the affected channel from the specified Trip Fet- i point. 5 or Sensor Error is either the "as-ressured" deviation of the sensor I -l from its calibration point or the value specified in Table T. 2-1 in percent i span, from the analysis assucptions. Use of Equation 2.2-1 allows for a sensor drif t f actor and an increased rack drif t factor, and provides a threshold ' r valve for REPORTABLE EVENTS. ' f The rethodology to derive the itip 5etpoints is based upor, combining all ; of the uncertainties in the channels. Inherent to the determination of the ; Trip 5etpoints are the magnituces of these channel uncertainties. Sensors and
- other instrumentatien utilizec in these channels are expected te be capable of operating within the allowances of t.hese uncertainty ragnitudes. Rack crift in e* cess of the Allo =Able Value exhibits the behavior that the rack has not ret its allowance. Because there is a sea 11 statistical chance that this will happen, an inf requent excessive drif t is expected. Rack or sensor drift, i in excess of the alio ance that is core than occasional, may be indicative of {
more serious proble.-a and should warrant further investigation. l f i l l l SOUTH TEXA5 - UMTS 1 & 2 E 2-3 AviM MENT N05. A% ( i i l I l
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- LIMITING SAFETY SY M M SETi g
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REACTOR TRIP SYSTEM INST.1MENTAT10N $ETPotNTS (Continued) The various Reactor trip circuits automatically open he Reactor trip l breakers whenever a condition monitored by the Reactor Trip System reaches a : preset or calculated level. In addition to redundant ch,.nnels and trains, the ] design approach provides a Reactor Trip System which monitors numerous system i
; variables, therefore providing Trip System functional diversity. The functional l ! capability at the specified trip setting is required for those anticipatory or }
diverse Reactor trips for which no direct credit was assumed in the safety l analysis to enhance the overall reliability of the Reactor Trip System. The l J Reactor Trip System initiates a Turbine trip signal whenever Reactor trip is l initiated. This prevents the reactivity insertion that would otherwise result , from excessive Reactor Coolant System cooldown and thus avoids unnecessary i
- actuation of the Engineered Safety Features Actuation System. l 4 r i Manual Reactor Trip t r
I The Reactor Trip System includes manual Reactor trip capability. { i 4 Power Rangt, Neutron Flux [ In each of the Power Range Neutron Flux channels there are ',wo independent f 1 bistables, each with its own trip setting used for a High and Lt/w Range trip j i setting. The Low Setpoint trip provides protection during subcritical and low : power operations to mitigate the consequences of a power excurston beginning ! j from low po.er, and the High Setpoint trip provides protection during power i 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 pow e level ( j of approximately 10% of RATED THERMAL POWER) and is automatically reinstaict j 1 below the P-10 Setpoint, y pc.er Range, Neutron Flua, Hi;5 Rates The Power Range Positive Rate trip provides protection against rapid flus I increases which are characteristic of a rupture of a control rod drive housing. [ ( 5pecifically, this trip complements the Po.er Range Neutron Flux High and Low i j trips to ensure that the criteria are eet for rod e;ection from mid-po.er. ! I The Po.er Range hegative Rate trip provides protection for control rod c*op - accidents. At hign po.er a single or inultiple roc cro, 'accident could cause [ local flus cesting which could cause an unconservative local DNER to emist. The Po.er Range begatise Rate trip will prevent th)- from occurring by tripping the reactor. No credit is taken for operation of the Poner Range Negative Eate trip for those control rod crop accidents for which Ch6Rs will be greater than 1.30. ; [ SOUTH TE AAS - UNITS 1 & 2 B 2-4 AMENOMENT N05. AND 3 7 f;g j i
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F. f LIMITI*C, SAFETY SYSTEM SETTINGS
. 4 BASES _ _ . , _ __
Inte .tedie' sr mee Range, Neutron Flux e and Source Range, Neutron Flux trips p.evide core The - prottetio actor startup to mitigate the consequences of an uncon-trolled 5 control assembly bank withdrawal from a subcritical ,' conditfo. <, rips provide redundant protection to the Low Setpoint trip of the Pow- ..ge, Neutron Flux channels. The Source Range channels will initiate a Reector trip at about 105 counts per second unless manually blocked whe-) P-6 becomes active. The Source Range channels are automatically blocked above P-10. The Intermediate Range channels will initiate a Reactor trip at a current level equivalent to approximately 25% of RATED THERMAL POWER unless r.inually blocked when P-10 becomes activc. Overtemperature AT The Overtemperatu'2 AT trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power distribu-tion, provided that the transient is slow with respect to piping transit delays from the core to the temperature detectors, and pressure is within the range between the Pressurizer High and Low Pressure trips. The Setpoint is auto-matically 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 froer the core to the loop temperature detectors, (2) pressurizer pressure, and (3) { ial 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 pecks 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 AT Th.e Overpowee al trip provides assurance of fuel integrity (e.g., no fuel peliet melting and less than 1% cladding strain) under all possible overpower conditions, limits the required range for Overtemperature AT. trip, and provides a backup to the High Neutren Flux trip. The Setpoint it automatically varied with: (1) coolant temperature to correct for ter.perature-induced changes in density and heat capacity of water, and (2) rate of change of temperature for dynamic compensation for piping delays from the core to the loop temper.sture detectors, to ensure that the allowatile
" heat generation rate (kW/ft) is not exce Jed. The Overpower aT trip provides protection to mitigate the consequences of various size stere breaks as reported in WCAP-9226, "Reactor Core Respense to Excessive Secondary Steam Releases."
o SOUTH TEXAS - UNITS 1 & 2 B 2-b AMENDMEN! NDS. AND b 'I;fg , i _ _ _ _ . _ _ - _ _ ___ _ _ - , , _ _ - _ - _ _ - - - - - - - -
1 b LIMITING SAFETY SYSTEM SETTINGS . I BASES _ l Pressurizer Pressure i In each of the pressurizer pressure channels, there are two independent bistables, each with its own trip setting to provide for a High and Low Pressure trip thus limiting the pressure range 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 cvent of a loss of reactor coolant pressure. On decreasing power, the Low Setpoint trip is automatically blocked by P-7 (a power level of approximately 10% of RATED THERMAL POWER with turbine impulse chamber pressure at approximately 10% of full power equivalcat); and on increasing power, automatically reinstated by P-7. The High Setpoint trip functions in conjunction with the pressurizer relief and safety :sives to protect the Raactor Coclera System against system overprossure. Pressuri7er Water level The Pressurizer Hign Water Leve' trip is provided to prevent water relief through the pressurizer safety valves. On decreasing power, the Pressurizer High Water Level trip is automatically blocked by P-7 (a power level of approxi-mately 10% of RATED THERMAL POWER with a turbine impulse chamber pressure at approximately 10% of full power equivalent); and on increasi q power, auto-matica11y reinstated by P-7. Reactor Coolant Flow 1he Low Reactor Coolant Flow trips provide core protection 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 impulse chamber pressure at approximately 10% of full power equivalent), an automatic Reactor trip will occur if the flow in more than oo op drops below apnraalmately d2% of nominal full loop flow. Above P-8 (- '
- 1cyc1 of approximately 40% of RATED THERMAL POWER) an automatic Re u r trip will occur if the flow in any single loop drops below approximately 92% of nominal full loop flow. Conversely, on decreasing power bett.een P-8 and the P-7, an automatic Reactor trip will ot. cur on low reactor coolant flow in more than one loop, and below P-7 the trip function is auto-matic'11y blocked.
Steam Generator Water Level The Steam Generator Water Len 1 Low-Low trip pratscts the reactor from l loss of heat sink in the event of a sustained steam /feedwater flow mismatch resulting from loss et normal feedwater. The spec'fied Setpoint provides allowances for starting delays of the Auxiliary feedwater System. SOUTH TEXAS - UNITS 1 & 2 B 2-6 AMEN F ENT N05. AND
F LSM1TfNG SAFETY SYSTEM SETTINGS
/
BASES Undervoltage and Underfrequency - Reactor Coolant Pump Buses 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 the Low Flow Trip Setpeint is reached. Time delays are incorporated in the Underfrequency 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 si0nal 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 Underfrequency Reactor Coolant Pump Bus trips are automatically blocked by P-7 (a power level of approximately 10% of RATED THERfiAL P'NER with a turbino impulse chamber pressure at approximately 10% of full power equivalent); and on increasing power, reinstated automatically by P-7. Turbine Trip A Turbiae trip initiates a Reactor trip. On decreasing power, the Reactor trip from the Turbine trip is automatically blocked by P-9 (a power level of approximately 50% of RATED THERMAL POWER); and on increasing power, reinstated automatically by P-9. Safety injection Input from ESFAS If a Reactor trip has not already been generated by the Reactor Trip System instrumentation, the ESFAS automatic actuation logic channels will initiate a Reactor trip upon any signal which initiates a Safety injection. The ESFAS instrumentation channels which initiate a Safety Injection signal are shown in Table 3.3-3. Reactor Trip System Interlocks 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 tilock of Source Range trip and deenergizes the high voltage to the detectors). On decreasing power, Source Range Level trips are automatically reactivated and high voltage restore-P-7 On increasing power, P-7 attematica11y enables Reactor trips on low flow in more than one reactor coolant loop, reactor coolant pump bus undervoltage and underfrequency, pressurizer low pressure, and pressurizer high level. On decreasing power, the above listed trips are automatically blocked. SOUTH TEXAS - UNITS 1 & 2 B 2-7 AMENDMENT NDS. AND r ,
b LIMITING SAFETY SYSTEM SETTINGS BASES Reactor Trip System Interlocks (Continued) P-8 On increasing power, P-8 automatically enables Reactor trip on low flow in one or more reactor coolant loops. On decreasing power, the P-8 automatically blocks the above listed trip. P-9 On increasing power, P-9 automatically enables Reactor trip on Turbine trip. On decreasing power, P-9 automatically blocks Reacter trip on Turbine trip. 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 Provides input to P-7. 4 I I 6 4 1 SOUTH TEXAS - UNITS 1 & 2 B 2-8 AMENDMENT NOS. AND ,
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Eb SECTIONS 3.0 AND 4.0 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS l 4 I I 1 i
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A i l 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.0 APPLICABILITY LIMITING CONDITION FOR OPERATION 3.0.1 Compliance with the Limiting Conditions for Operation contained in the succeeding specifications is required during the OPERATIONAL MODES or other ccnditions specified therein; except that upon failure to meet the Limiting Conditions for Operation, the associated ACTION requirements shall be met. 3.0.2 Noncompliance with a specification shall exist when the requirements of the Limiting Condition for Operation and associated ACTION requirements are not met within the specified time intervals. If the Limiting Condition for Operation is restored prior to expiration of the specified time intervals, completion of the ACTION requirements is not required. 3.0.3 When a Limiting Condition for Operation is not met, except as provided in the associated ACTION requirements, within 1 hour action shall be initiated te 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 hours,
- b. At least HOT SHUT 00WN within the following 6 heurs, and !
- c. At least COLD SHUTOOWN withir. the subsequent 24 hours.
Where corrective measures are compl9ted that permit operation under the ACTION requirements, the action may be taken in accordance with the specified time limits as measured from the time of failure to meet the Limiting Condition for Operation. Exceptions to these requirements are stated in the individual specifications. This specification is not applicable in MODE 5 or 6. 3.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made when the conditions for the Limiting Condition for Operation are not met and the associated ACTION requires a shutdown if they are not met with E a specified time interval. Entry into an OPERATIONAL MODE or specified condition may be made in 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 or to OPERATIONAL MODES as required to comply with ACTION requirements. Exceptions to these requirements t are stated in the individual specifications. 3.0.5 Limiting Conditions for Operation including the associated ACTION requirements shall apply to each unit individually unless otherwise indicated as follows:
- a. Whenever the Limiting Conditions for Operation refers to systems or components which are shared by both units, the ACTION requirements will apply to both units simultaneously.
SOUTH TEXAS - UNITS 1 & 2 3/4 0-1 AMENDMENT NDS. AND ! i
F.D l l 3/4.0 APPLICABILITY l l LIMITING CONDITION FOR OPERATION (Continued)
- b. Whenever the Limiting Conditions for Operation applies to only one unit, this will be identified in the APPLICABILITY section of the specification; and
- c. Whenever certain portions of a specification contain operating parameters, Setpoints, etc., which are different for each unit,
- this will be identified in parentheses, footnotes or body of the requirement.
4 a j SOUTH TEXAS - UNITS 1 & 2 3/4 0-2 AMENDMENT NDS. AND d
FD l APPLICABILITY l SURVEILLANCE REQUIREMENTS 4.0.1 Surveillance Requirements shall be met during the OPERATIONAL H0 DES or other conditions specified for individual Limit *ng Conditions for Operation unless otherwise stated in an individur.i Surveillance Requirement. 1 4.0.2 Each Surveillance Requirement shall be pe>iormed within the specified time interval with: i
- a. A maximum allowable extension not to exceed 25% of 6he 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 Surveiiiance Requirement within the allowed sur-veillance interval, defined by SpecificatYon 4.0.2, shall constitute a f ailure to meet the OPERABILITY requirements for a Limiting Condition 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 hours to permit the completion of the surveillance when the allowed outage time limits of the ACTION requirements are less than 24 hours. Surveillance Requirements do not have to be performed on inoperable equipment. 4l0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made unless the Surveillance Requirement (s) associated with the Limiting Condi-tion 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 r rformed in accordance with Section XI of the ASME Boiler and Pre.v .re Vessel Code and applicable Addenda as required by 10 CFR vart 50, Section 50.55a(g), except where specific written
, relief has been granted by the Commission pursuant to 10 CFR Part 50, Section 50.55a(g)(6)(i);
- b. Surveillance intervals specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda for the inservice inspection and testing activitics required by the ASME Boiler and Pressure Vessel Code and applicable Addenda shall be applicable as follows in these Technical Specifications:
SOUTH TEXAS - UNITS 1 & 2 3/4 0-3 AMENCMENT N05. AND G 1715y
FC APPLICABILITY SURVEILLANCE REQUIREMENTS (Continued) ASME Boiler and Pressure Vessel Required frequencies for Code and applicable Addenda performing inservice terminology for inservice inspection and testing inspe: tion 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 ts. ting activities; ,
- d. Performance of the above inservice inspection and testing activities shall be in addition to other specified Surveillance Requirements; and I
- e. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any Technical Specification.
4.0.6 Surveillance Requirements shall apply to each unit individually unless otherwise indicated as stated in Specification 3.0.5 for individual specifica-tions or whenever certain portions of a specification contain surveillance parameters different for each ur.it, which will be identified in parentheses, footnotes or body of the requirement. SOUTH TEXAS - UNITS 1 & 2 3/4 0 4 AMENDMENT NDS. AND C." ; ,
FD 3/4.1 RFACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUTDOWN MARGIN - T, GREATER THAN 200'F LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to the limit as shown in Figure 3.1-1. APPLICABILITY: MODES 1, 2*, 3, and 4. l ACTION: With the SHUT 00WN MARGIN less than the limit as shown in Figure 3.1-1, imme-diately 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 equivalent until the required SHUTOOWN MARGIN is restored. j I SURVEILLANCE REOUIREMENTS 4.1.1.1.1 The SHUTOOWN MARGIN shall be determined to be greater than or equal ~ to the limit as shown in /igure 3.1-3: a. Within 1 hour af ter detection of an inoperable control rod (s) and at least once per 12 hours 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 2 with K,ff less than 1, within 4 hours prior to achieving reactor criticality by verifying that the predicted critical control rod position is within the limits of Specification 3.1.3.6;
- c. Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors of Specifica-tion 4.1.1.1.1d. below, with the control banks at the maximum inser-tion limit of Specification 3.1.3.6; and "See Special Test Exceptions Specification 3.10.1.
SOUTH TEXAS - UNITS 1 & 2 3/4 1-1 AMENDMENT NOS. AND
bI{} l REACTIVITY CONTROL SYSTEMS SURVEILLANCE REOUIREMENTS (Continued)
- d. When in MODE 3 or 4, at least once per 24 hours by consideration of the following factors:
- 1) Reactor Coolant System boron concentration,
- 2) Control rod position,
- 3) Reactor Coolant System average temperature, .
- 4) Fuel Durnup based on gross thermal energy ger.eretion,
- 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 4.1.1.1.1d., 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. 4 I 4 4 SOUTH TEXA5 - UNITS 1 & 2 3/4 1-2 AMENDMENT NDS. AND e i lin
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b REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN - T,yg LESS THAN OR EQUAL TO 200'F , LIMITING CONDITION FOR OPERATION 3.1.1.2 The SHUTOOWN MARGIN shall be greater than or equal to the limit as shown in Figure 3.1-2. APPLICABILITY: MODE 5. ACTION: With the SHUTDOWN MARGIN less than the limit as shown in Figure 3.1-2, imme-diately 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 equivalent until the required SHUTDOWN MARGIN is restored. SURVETILANCE REOUIREMENTS 4.1.1.2 The SHUTOOWN MARGIN shall be determined to be greater thar. or equal to the limit as shown in Figure 3.1-2:
- a. Within I hour after detection of an inoperable control rod (s) and at least once per 12 hours thereafter while the rod (s) is inoperable.
If the inoperable control rod is immovable or untrippable, the SH'TDOWN J MARGIN shall be verified acceptable with an increased ' allowance for the withdrawn worth of the immovable or untrippable control rod (s); -d
- b. At least once per 24 hours by consideration of the following factors:
- 1) Reactor Coolant System boron concentration,
, 2) Control rod position,
- 3) Reactor Coolant System average temperature, i 4) Fuel burnup based on gross thermal energy generation,
- 5) Xenon concentration, and
- 6) Samarium concentration.
SOUTH TEXAS - UNITS 1 & 2 3/4 1-4 AMEN 0 MENT NOS. t 40 W l 7 1;;;
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w H e-FJ R u 7 w 3 9 M .
'a w
5
;. FIGURE 3.1-2
- - o REQUIRED SHUTDOWN MARGIN VERSUS RCS CRITICAL BORON CONCENTRATION
.1 (MODE S) c h
- o
F6 REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION 3.1.1.3 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 Ak/k/'F for the all rods withdrawn, 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 Spt .fication 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 ak/k/*F within 24 hours or be in HOT STANDBY within the next 6 hours.
These withdrawal limits shall be in addition to the insertion 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 treasured MTC, the interim control rod withdrawal limits, and the predicted average core burnup necessary for restoring the positive MTC to within its limit for the all rods withdraan condition.
- b. With the MTC frore negative than the limit of Specification 3.1.1.3b.
above, be in HOT SHUTDOWN within 12 hours.
*With K,ff greater than or equal to 1.
**See Special Test Exceptions Specification 3.10.3.
SOUTH TEXAS - UNITS 1 & 2 3/4 1-6 AMENDMENT NDS. AND
- l. : . ; . . .
I l REACTIVITY CONTROL SYSTEMS SURVEILLANCE REOUIREMENTS 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 af Specifi- )
cation 3.1.1.3a., above, prior to initial operation above 5% of ' RATED THFRMAL POWER, after each fuel loading; and I
- b. The MTC shall be measured at any THERMAL POWER and compared to -3.1 x 10 4 Ak/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 Ak/k/'F, the MTC shall be remeasured, and compared to the EOL MTC limit of Specification 3.1.1.3b., at least once per 14 EFPD during the remainder of the fuel cycle.
SOUTH TEXA5 - UNITS 1 & 2 3/4 1-7 AMEN 0 MENT N05. AND
+ 1 , ,, ;
FT REACTIVITY CONTROL SYSTEMS MINIMUM TEMPERATURE FOR CRITICALITY LIMITING CONDITION FOR OPERATION 3.1.1.4 shall be greater than or equal to 561*F.The Reactor Coolant System lowest oper APPLICABILITY: MODES 1 and 2* **. ACTION: - With a Reactor Coolant System operating loop temperature (T,yg) less than 561*F, restore T,yg to within its limit within 15 minutes or be in HOT STANDBY within the next 15 minutes. SURVEILLANCE RE0dikEMENTS 4.1.1.4 be greaterThe Reactor than Coolant or equal System temperature (T'V9) shall be determined to to 561'f: .
- a. Within 15 minutes prior to achieving reactor criticality, and
- b. At least once per 30 minutes when the reactor is critical and the
' Reactor Coolant System T,yg is less than 571'F with the T,yg-T ref Oeviation Alarm not reset.
"With K,77 greater Gan or equal to 1.
**See Special Test Exceptions Specification 3.10.3.
SOUTH TEXAS - UNITS 1 & 2 3/4 1-8 AMENDMENT N05. AND lO k' 1 7 ,,;
FD REACTIVITY CONTROL SYSTEMS " 3/4.1.2 BORATION SYSTEMS FLOW PATHS - AHUTDOWN LIMITING CONDITION 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 emergency power source: 1
- a. A flow path from the Boric Acid Storage System via either a boric e acid transfer pump or a gravity feed connection, and a charging pump t to the Reactor Coolant System if the Boric Acid Storage System is OPERABLE as given in Specification 3.1.2.5a. for MODES 5 and 6 or es given in Specification 3.1.2.6a. for MODE 4; or
- b. The flow path from the refueling water storage tank via a charging pump to the Reactor Coolant System if the refueling water storage tank is OPERABLE as giver, in Specification 3.1.2.5b. for MODES 5 and 6 or as given in Specification 3.1.2.6b. for MODE 4.
l APPLICABILITY: MODES 4, 5, and 6. ACTION: With none of the above flow paths OPERABLE or capable of being powered from an l OPERABLE emergency power source, suspend all operations involving CORE ALTERATIONS or positive reactivity changes. SURVEILLANCE RE0VIREMENTS 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 heat traced portion of the flow path is greater than or equal to 65'F when a flow patn from the boric acid tanks is used, and
- b. At least once per 31 days by verifying that each valve (manual, !
, power-cperated, or automatic) in the flow path that is not locked,
. sealed, or otherwise secured in position, is in its correct position.
4 1 1
) SOUTP TEXAS - UNITS 1 & 2 3/4 1-9 AMENDMENT NOS. AND ' ** Jl gj; '
l t
PD REACTIVITY CONTROL SYSTEMS FLOW PATHS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.2 At least two of the fol'iowing three boron injection flow paths shall be OPERi3LE:
- a. The flow path from the beric Acid Storage System 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 charging pumps to the RCS.
APPLICABILITY: MODES 1, 2, and 3.* ACTION: With only one of the abov* required boron injection flow paths to the RCS OPERABLE, restore at least two boron injection flow paths to the RCS to OPERABLE status within 72 hours or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least the limit as shown in Figure 3.1-2 at 200*F within the next 6 hours; restore at least two flow paths to OPERABLE status within the next 7 days or be in COLD SHUT 00WN within the next 30 hours. SURVEILLANCE REOUIREMENTS 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 heat traced portion of the flow path from the boric acid tanks is greater than or equal to 65'F when it is a required water source;
- 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 otherwise secured in spo'ition, is in its correct position;
- 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 Safety Injection test signal; and
- d. 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.
*The provisions of Specifications 3.0.4 and 4.0.4 are not applicable for entry into MODE 3 for the charging pump declared inoperable pursuant to Specifica-tion 4.1.2.3.2 provided that the charging pump is restored to OPERABLE status within 4 hours or prior to the temperature of one or more of the RCS cold legs exceeding 375'F, whichever comes first.
SOUTH TEXAS - UNITS 1 & 2 3/4 1-10 AMENDMENT NDS. AND W ! ? 1:.,~3
l
$l l
REACTIVITY CONTROL SYSTEMS CHARGING PUMPS - SHUT 00WN 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 4**, 5, and 6. ACTION: With no charging pump OPERABLE or capable of being powered from an OPERABLE emergency power source, suspend all operations involving CORE ALTERATIONS or positive reactivity changes. SURVEILLANCE RE0VIREMENTS 4.1.2.3.1 The above required charging pump shall be demonstrated OPERABLE by verifying, on recirculation flow, that a differential pressure across the pump of greater than or equal to 2300 psid is developed when tested pursuant to Specification 4.0.5. 4.1.2.3.2 All charging pumps, excluding the above 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.
*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. **The provisions of Specification 3.0.4 and 4.0.4 are not applicable for entry into MODE 4 from MODE 3 for the charging pumps declared OPERABLE pursuant to Specification 4.1.2.4 provided that a maximum of one charging pump is OPERABLE within 4 hours af ter entry into MODE 4 from MODE 3 or prior to the terrperature of one or more of the RCS cold legs decreasing below 325'F, whichever comes first.
SOUTH TEXAS - UNITS 1 & 2 3/4 1-11 AMEN 0 MENT N05. AND tiOV t : .
,m
FP REACTIVITY CONTROL SYSTEMS CHARGING PUMPS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.4 At least two charging pumps shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3.* ACTION: With only one charging pump OPERABLE, restore at least two charging pumps to OPERABLE status within 72 hours or be in at least HOT STANDB\ and borated to a SHUTDOWN MARGIN equivalent to at least the limit as shown in Figure 3.1-2 at 200'F within the next 6 hours; restore at least two charging pJmps to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours. SURVEILLANCE REOUIREMENTS ,. 4.1.2.4 At least two charging pumps shall be demonstrated OPERABLE by verifying, on recirculation flov, that a differential pressure across each pump of greater than or equal to 2300 psid is developed when tested pursuant to Specification 4.0.5.
*The provisions of Specification 3.0.4 and 4.0.4 are not applicable for entry into MDDE 3 for the charging pumps declared inoperable pursuant to Specifica-tion 4.1.2.3.2 provided that the charging pump is restored to OPERABLE status withir 4 hours or prior to the temperature of one or more of the RCS cold legs exceedly 375'F, whichever comes first.
SOUTH TEXAS - UNITS 1 L 2 3/4 1-12 AMENDMENT NDS. AC
'-
- 1 7 !;g j
i
PD REACTIVITY CONTROL SYSTEMS BORATEDWATERSOURCES-SHUROM LIMITING CONDITION FOR OPERATION 3.1.2.5 As a minimum, one of the fc11owing borated water sources shall be OPERABLE:
- a. A Boric Acid Storage System with:
- 1) A minimum contained borated water volume of 2900 gallons,
- 2) A minimum boron concentration of 7000 ppm, and i +
- 3) A minimum solution temperature of 65'F.
- b. The refueling water storage tank (RWST) with:
- 1) A minimum contained borated water volume of 122,000 gallons for MODE 5 and 33,000 gallons for MODE 6, and f .
- 2) A boron con:entration between 2500 ppm and 2700 ppm.
APPLICABILITY: MODES 5 and 6. ACTION: With no borated water source OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes. SURVEILLANCE REOUIREMENTS 4.1.2.5 The above required borated wattr source shall be demonstrated OPERABLE at least once per 7 days by:
- a. Verifying the boron concentration of the water,
- b. Verifying the contained borated water volume, and
- c. Verifying the botic acid storage tank solution temperature when it is the source of bcrated water.
1 SOUTH TEXAS - UNITS 1 & 2 3/4 1-13 AMENDMENT N35. AND l'E ' ; ; p .
FD _ REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.6 As a minimum, the following borated water source (s) shall be OPERABLE as required by Specification 3.1.2.2 for MODES 1, 2, and 3 and one of the fol-lowing borated water sources shall be OPERABLE as required by Specifica-tion 3.1.2.1 for MODE 4:
- a. A Boric Acid Storage System with: *
- 1) A minimum contained borated water volume of 27,000 gallons,
- 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 458,000 gallons, and
- 2) A boron concentration between 2500 ppm and 2700 ppm.
APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: a. With the Boric Acid Storage System inoperable and being used as one of the above required borated water sources, restore the system to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and borated to a SHUT 00WN MARGIN equivalent to at least the limit as shown in Figure 3.1-2 at 200'F; restore the Boric Acid Storage System to OPERABLE status within the next 7 days or be in COLD SHUT 00WN within the next 30 hours.
- b. With the RWST inoperable, restore the tank to OPERABLE status
! within 1 hour or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTOOWN within the following 30 hours. I l 1 l l 4 SOUTH TEXA5 - UNITS 1 t. 2 3/4 1-14 AMENDMENT NDS. AND i ! ? lie 3 i
REACTIVITY CONTROL SYSTEMS SURVEILLANCE REOUIREMENTS , 4.1.2.6 Each borated water source shall be demonstrated OPERABLE at least once pe:' 7 days by: ,
- a. Verifying the boron concentration in the water, b Verifying the contained borated water volume of the water source, and
- c. Verifying the Boric Acid Storage System solution temperature when it l is the source of borated water.
4 4 1 7 4 l - i i l SOUTH TEXAS - UNITS 1 & 2 3/4 1-15 AMEN 0 MENT N05, AND
,,.1 ...<
J
FD REACTIVITY CONTROL SYSTEMS 3/4.1.3 MOVABLE CONTROL ASSEMBLIES GROUP HEIGHT LIMITING CONDITION FOR OPERATION 3.1.3.1 All full-length shutdown and control rods shall be OPERABLE and positioned within 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, determine that the SHUTOOWN MARGIN require-ment of Specification 3.1.1.1 is satisfied within 1 hour and be in HOT STANDBY within f hours.
- b. With one full-length rod trippable but inoperable due to causes ov.her than addressed by ACTION a,, above, or misaligne:1 from its group step counter demand height by more than i 12 steps 4
(indicated position), POWER OPERATION may continue provided that l within 1 hour: .i
- 1. The rod is restored to OPERABLE status within the above
, alignment requirements, or 4
- 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 3.1-3. The THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation, or
- 3. The rod is declared inoperable and the SHUT 00WN MARGIN requirement of Specification 3.1.1.1 is satisfied. POWER 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 analyzed results of these accidents 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 hours;
*See Special Test Exceptions Specifications 3.10.2 and 3.10.3.
SOUTH i.XAS - UNITS 1 f. 2 3/4 1-16 AMENDMENT N35. AND NOV 17 1.'.:
F.D , REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION ACTION (Continued) c) A power distribution map is obtained from the movable incoredetectorsandF(Z)andFhareverifiedtobe q within their limits within 72 hours; and d) The THERMAL POWER level is reduced to less than or equal to 75% of RATED THERMAL POWER within the next hour and within the following 4 hours the High Neutron Flux Trip Setpoint is reduced to less than or equal to 85% of RATED THERMAL POWER. i i
- c. With more than one rod trippable but inoperable due to causes other [
than addressed by ACTION a. above, POWER OPERATION may continue ' provided that: '
- 1. Within I hour, the remainder cf th" rods in the bank (s) with the inoperable rods are alignei r.o within i 12 steps of the t
inoperable rods while maintr'..iing the rod sequence and insertion limits of Figure 3.1-3. The THERMAL POWER level shall be l restricted pursuant to Specification 3.1.3.6 during subsequent operation, and ) t
- 2. The it: operable rods are restored to OPERABLE status within 72 hours.
t
- d. With more than one rod misaligned from its group step counter demand l
height by more than i 12 steps (indicated position), be in HOT STANDBY within 6 hours. 3 1 l i SURVIILLAM E RE M REMENTS ; 4.1.3.1.1 The position of each full-length rod shall t,e determined to be ^ within the group demand limit by verifying the individual rod positions at least once per 12 hours except during time intervals when the red position deviation monitor is inoperable, then verify the group positions at least once , per 4 hours. t 4.1.3.1.2 Each full-length rod not fully inserted in the core shall be 6 determined to be OPERABLE by moveeent of at least 10 steps in any one : direction at least once per 31 cays. ! t-i i l l SOUTH TEXA5 - UNITS 1 & 2 3/4 1-17 AMENCHENT N35. AND {
. . . : W.h I
TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT OF AN INOPERABLE FULL-LENGTH ROD Rod Cluster Control Assem'ly o Insertion Characteristics Rod Cluster Control Assembly Misalignment Loss of Reactor Coolant from Small Ruptured Pipes or from Cracks in Large Pipes Which Actuates the Emergency Core Cooling System . Single Rod Cluster Control Assembly Withdrawal at Full Power ; i Major Reactor Coolant System Pipe Ruptures (Loss-of-Coolant ' Accident) ( Major Secondary Coolant System Pipe Rupture ' Rupture of a Control Rod Drive Mechanism Housing (Rod Cluster Control Assembly Ejection) [ I i i l I $00TH TEXAS - UNITS 1 & 2 3/4 1-18 AMENDMEN) N35. AND
FD REACTIVITY CONTROL SYSTEMS POSITION INDICATION SYSTEMS - OPERATING LIMITING C0!!DITION FOR OPERATION 3.1.3.2 The Digital Rod Position Indication System and the Demand Position Indication System shall be OPERABLE and capable of determining the control rod positions within i 12 steps. APPLICABILITY: H0 DES 1 and 2. ACTION:
- a. With a maximum of one digital rod position indicator per bank inoperable either:
- 3. Determine the position of the nonindicating rod (s) indirectly by the movable incore detectors at laast once per 8 hours and immediately af ter any motion of the nonindicating rod which exceeds 24 steps in one direction since the last determination of the rod's position, or
- 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 8 hours,
- b. With a maximum of one demand position indicator per bank inoperable either:
- 1. Vsrify that all digital rod position indicators for the affected bank are OPERABLE and that the most withdrawn rod and the ledst withdrawn rod of the bank are within a maximum cf 12 steps of each other at least once per 8 hours, or
- 2. Reduce THERMAL POWER to less than 50% of RATED THEi MA'. POWER within 8 hours.
MflElkk1LEMEIMihD 4.1.3.2 Each digital rod position indicator shall be determined to be OPERABLE by verifying that the Demand Position Indication System and the Digital Rod Position :ndicatior System agree within 12 steps at least once per 12 hours en-cept during tiee intervals when the rod position deviation ronitor is inoperible, then r.Jpare the Deeand Position Indication System and the Digital Rod Position Indication System at least once per 4 hours. SOUTH TEXAS - UNITS 1 & 2 3/4 1-19 AMENCMENT h05. AND b" ' 1 'e :..
! REACTIVITYCONTRO15YSTEMS POSITION INDICA 110N SYSTEMS - SHUTDOWN ( LIMITING CONDITION FOR QQAIl071 c 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 i 12 stu ; for auch shutdown or control rod not fully inserted. APPLICABILITY: MODES _3a ** , 4* **, and 55 ** . ACTION: With less than the above required position indicator (s) OPERABLE, immediately open the Reactor Trip System breakers. SURVEILLANCE RE0VIREWENTS _ 4.1.3.3 Each of the above required digital rod position indicator (t , hall 'e. determined to be OPERABLE by verifying that the . digital rod positio ind* ' ors agree with the demand position indicators within 12 steps when exercisec' sa e the full-range of rod travel at least once per 18 months.
"With the Peactor Trip System breakers in the closed position.
**Sse Special Test Exceptions Specification 3.10.5.
SOUTH TEXAS - UNITS 1 & c 3/4 1-20 AMENDMINT NOS. AN3
*i1!
FO ! REACTIVITY CONTROL SYSTEMS ROD OROP TIME LIMITING CONDITION FOR OPER,ATION ____ 3.1.3.4 The individual full-leh0th (shutdown and control) rod drop time from the fully withdrawn position shall be less than or equal to 2.8 seconds from beginning of decay of stationsry gripper coil voltage to dashpot entry with;
- a. 6 T,yg greater than or equal to 561 F, and
- b. All reactor cooiant pumps operating.
, APPLICABILilX: MODES I and 2. i AETION: 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 prict to proceeding to i MODE 1 or 2. i , SURVEILLANCE REOUIREMENTS 4.1.3.4 The rod drop time of full-length rods shall be demonstrated through measurement prior to reactor critica'iity:
- a. For all rods following each removal of the reactor vessel head, I
- b. For specifically affected individual rods following any maintenance on or modification to the Control Rod Drive System which could
, sffect the drop time of those specific rods, and a
- c. At least onct per 18 months.
I j 4 l ' ) 1 1
- SOUTH TEXAS - UNITS 1 & 2 3/4 1-21 AMENDMENT NDS. AND N 0 i' 1
- 1; 3
FD REACTIVITY CONTROL SYSTEMS SHUTDOWN R00 INSERTION LIMIT LIMITING CONDITION FOR OPERATION 3.1.3.5 All shutdown rods shall be fully withdrawn. APPLICABILITY: MODES la and 2* **. ACTION: With a maximum of one shutdown rod not fully withdrawn, except for surveillance testing pursuant to Specification 4.1.3.1.2, within I hour either:
- a. Fully withdraw the rod, or
- b. Declare the rod to be inoperable and apply Specification 3.1.3.1.
SURVEILLANCE REOUIREMENTS 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 hours thereafter.
I
*See Special Test Exceptions Specifications 3.10.2 and 3.10.3.
**With K,ff greater than or equal to 1.
SOUTH TEXAS - UNITS 1 & 2 3/4 1-22 AMENDMENT N05. AND I I
FP REACTIVITY CONTROL SYSTEMS CONTROL ROD INSERTION LIMITS LIMITING CONDITION ft0R OPERATION ! i 3.1. 3, 6 The control banks shall be limited in physical insertion as shown in l Figure 3.1-3. APPLICABILITY: H0 DES 1* and 2* **. ! ACTION: I 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 hours, or
- b. Reduce THERMAL POWER within 2 hours,to less than or equal to that I fraction of RATED THERMAL POWER which is allowed by the bank posi- i tion using the above figure, or >
- c. Be in at least HOT STANDBY within 6 hours.
SURVEILLANCE RE0VIREMENTS 4.1.3.6 The position of each control bank shall be determined to be within the insertion limits at least once per 12 hours except during time intervals when the rod insertion limit monitor is inoperable, then verify the individual rod positions at least once per 4 hours. i i 1 i I l l l *See Special Test Exceptions Specifications 3.10.2 and 3.10.3.
**With K,f f greater than or equal to 1.
SOUTH TEXAS - UNITS 1 & 2 3/4 1-23 ANEN0hENT N05. AND fd h* j 7 M i
1 FD 300
~~
(O.23.259) (O.79,259) ~~ 2'wa' -
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1 ." / /_ s: i="-" / / O ): / SANK C / ew / /~ DC / / hd / / Em -
; *3
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O .2 4 .6 .8 f.O
- FRAOTION OF RATED THERMAL POWER FIGURE 3.1-3 ROD BANK INSERTION LIMITS VERSUS THERMAL POWER FOUR-LOOP OPERATION SOUTH TEXAS - UNITS 1 & 2 3/4 1-24 AW.ENDMENT N05. AND I'E 1 i ig;;
FD 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within the following target band (flux difference units) about the target flux difference:
- a. 1 5% for core average accumulated burnup of less than or equal to 3000 WD/MTU; and
- b. + 3%, -12% for core average accumulated burnup of greater tharJ3000 WD/MTV.
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- ' 4 cated AFD is within the Acceptable Operation Limits of Figure 3.2-1 and the cumu-i lative penalty deviation time does not exceed 1 hour during the previous 24 hours. ' 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 1 hour daring the previous 24 hours. ' i APPLICABILITY: MODE 1, above 15% of RATED THERMAL POWER.* 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:
- 1. Restore the indicated AFD to within the target band limits, or
- 2. Reduce THERMAL POWER to less than 90% of RATED THERMAL POWER. !
l b. With the indicated AFD outside of the above required target band for more than 1 hour of cumulative penalty deviation time during the previous 24 hours or outside the Acceptable Operation Limits of ) Figure 3.2-1 and with THERMAL POWER less than 90% but equal to or greater than 50% of RATED THERMAL POWER, reduce: -
- 1. THERMAL FMo u less than 50% of RATED THERPAL POWER within 30 minutes, and i 2. The Power Range Neutron Flux * ** - High Setpoint to less than or equal to 55% of RATED THERMAL P!WER within the next 4 hours.
p
*See Special Test Exceptions Specification 3.10.2.
j ** Surveillance testing of the Power Range Neutron Flux Channel may be performed pursuant to Specification 4.3.1.1 provided the indicated AFD is maintained within
- the Acceptable Operation Limits of Figure 3.2-1 A total of 16 hours operation i rnay be accumulated with the AFD outside of the above required target band during testing without penalty deviation, i
SOUTH TEXAS - L!NI'S 1 & 2 3/4 2-1 AMEN 0 MENT NOS. AN? I,; , y y 1 -, .. - _- - -
FD POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION ACTION (Continued)
- c. With the indicated AFD outside of the above required target band for more than 1 hour of cumulative penalty deviation time during the previous 24 hours 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 REOUIREMENTS 4.2.1.1 The indicated AFD shall be determined to be within its limits during POWER OPERATION above 15% of RATED THERMAL POWER by:
- a. Monitating 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 hours after restoring the AFD Honitor Alarm to OPERABLE status.
- b. Monitoring and logging the indicated AFD for each OPERABLE excore channel at least once per hour for the first 24 hours and at least once per 30 minutes thereafter, when the AFD Moniter Alarm is inoperable. The logged values of the indicated AFD shal' 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 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 OPERATION 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 Days.
The provisions of Specification 4.0 4 are not applicable.
- 4. 2.1. 4 The target flux difference shall be updated at least once per e
31 Effective Full Power Days by either determining the target flux difference SOUTH TEXAS - UNITS 1 & 2 3/4 2-2 AMENDMENT N05. AND
'a
- i. . w, ~
FD POWER DISTRIBUTION LlHITS SURVEILLANCE REOUIREMENTS (Continued) pursuant to Specification 4.2.1.3 above or by linear interpolation between the most recently measured value and the predicted value at the end of the cycle life. The provisions of Specification 4.0.4 are not applicable.
)
SOUTH TEXA5 - UNITS 1 & 2 3/4 2 3 AMENDMENT N05. AND
Pp 120 i I l l A . I I
- i
~
i I i l I ' ' ' ' '
% 100 IM '
i l g (-11.90) (11.90) i O ~ UNACCEPTABLE'/ I \iUNACCEPTABLE L CL t OPERATION ( i ODERATION J i l / l )k I i f I I /! I \I C I l !/ ACCEPTABLE i \i ! W i j
/ OPERATION [ _
\
h I : / l i i \ > I i 0 ; i /, i i X
. W I (-31,50) !
1 (31 50) ' H I I i e I l l l
<[ 40 '
C l I I l
. I l l i
o i . 20 , ; , , , i I . i ! ! ! I i i i i I l 1 0 40-30-20 -10 O 10 20 30 40 50
. FLUX DIFFERENCE (AI) %
FIGURE 3.2-1 AXIAL FLUX OlFFERENCE LIMITS A5 A FUNCTION OF RATED THERMAL POWER SOUTH TEXAS - UNITS 1 & 2 3/4 2 4 AMENDMENT NDS. AND W
FD POWER O!STRIBUTION LIMfTS 3/4.I 2 HEAT FLUX HOT CHANNEL FACTOR - Fg LIMITING CONDITION FOR OPERATION 3.2.2 F g(Z) shall be limited by the following relationships: FA (Z) ~< 2.50 [K(Z)) for P > 0.5 T Fg (Z) 1 5.0 (K(Z)) for P 5 0.5 Where: P = THERMAL POWER , and RATED THERMAL power K(Z) = the function obtained from Figure 3.2-2 for a given core height location. APPLICABILITY: M30E 1. - ACTION: With Fg (Z) exceeding its limit: a. Reduce THERMAL POWER at least 1% for each 1% gF (Z) exceeds the limit within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoint within the next 4 hours; POWER OPERATION may proceed for up to a total of 72 hours; subsequent POWER OPERATION may proceed provided the Overpower AT Trip Setpoint has been reduced at least 1% for each 1% F g (Z) exceeds the limit.
- 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 Fg (Z) is demonstrated through incore mapping to be within its limit.
SOUTH TEXAS - UNITS 1 & 2 3/42-5 AMEN: MENT N05. AND t. , ,
F.D l.2 (7.I40) i.O (13,0.936) N
- T w \
e O.e \ O 1 d C.6
--(l4.0.60)~
.)
I
@ O.4 Z
l -
- O.2 N ,
O.O O 2 4 6 e 10 12 14 16, CORE HEIGHT (PT) = FIGURE 3.2-2 K(Z) - NORMALIZED F g(Z) AS A FUNCTION OF CORE HEIGHT SOUTH TEXA5 - UNITS 1 & 2 3/4 2-6 AMENDMENT N05. AND
' 1 I 1553
FD POWER DISTRIBUTION LIMITS SURVEILLANCE RE0VIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable. 4.2.2.2 F xy shall be evaluated to determine ifgF (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,
- b. Increasing the measured F xy component of the power distribution map by 3% to account for manufacturing tclerances and further increasing the value by 51, to account for measurement uncertainties, C
- c. Comparing the F xy computed (Fx ) obtained in Specification 4.2.2.2b.,
above to: The F
- 1) xy limits for RATED THERMAL POWERx(FRTP) for the approprhte measured core planes given in Specification 4.2.2.2e. and f.,
below, and
- 2) The relationship:
Ff,=F x P (1+0.2(1-P)), - l Where F*Y is the limit for fractional THERMAL POWER operation expressed as a function of F RTP and P is the fraction of RATED xy THERMAL POWER at which F was measured, xy
- d. Remeasuring F according to the following schedule:
xy
- 1) When FC , is greater than the F RP limit for the appropriate x
measured core plane but less than the F relationship, additional power distribution maps shall be taken dF compared to F IP , and F,y either: a) Within 24 hours after exceeding by 20% of RATED THERMAL POWER or or determined, greater, the THERMAL POWER at which F*YC ,,, ),,g b) At least once per 31 Effective Full Power Days (EFPD), whichever occurs first. SOUTH TEXAS - UNITS 1 & 2 3/4 2-7 AMENDMENT N05. AC te. . .a ,' 2 - _ _ _ _ _ _ _ _ _ _ _ _ _ - - - Y
FD r POWER DISTRIBUTION LIMITS I SURVEILLANCE REQUIREMENTS (Continuedi C
- 2) When the F x is less than or equal to the FRTP limit for the x
appropriate measured core plane, additional power distribution l maps shall be taken and F,C compared to F,"andF,fatleast f once per 31 EFPD. j
- e. The F*Y limits used in the Constant Axial Offset Control analysis~ ,
for RATED THERMAL POWER y (F,RTP) shall be provided for all core planes containing Bank "D" control rods and all unrodded core planes in a Radial Peaking Factor Limit Report per Specification 6.9.1.6;
- f. The F,y limits of Specification 4.2.2.2e., above, are not applicable in the following core planes regions as measured in percent of core height from the bottom of the fuel: !
- 1) Lower core region from 0 to 15%, inclusive,
{
- 2) Upper core region from 85 to 100%, inclusive,
- 3) Grid plane regions at 22.4 1 2% 34 69.512% and 81.312%, inclusive,.212%, 46.01 2%, 57.812%, l and .
i 1
- 4) Core plane regions within i 2% of core height (i 3.36 inches) j 1
about the bank demand position of the Bank "D" control rods. l g. With F x exceedingF,h,theeffectsofF,y on F9 (Z) shall be
- evaluated to determine if9F (Z) is within its limits.
4.2.2.3 When qF (Z) is reasured for other than F,y determinations, an overall i measured qF (Z) shall be obtained from a power distribution map and increased ' by 3% to account for manufacturing tolerances and further increased by 5% to account for measurement uncertainty. ! i i-j i i ! I t l l i 2 : l ! SOUTH TEXAS - UNITS 1 & 2 3/4 ? c AMEhDMENT NOS, AND ( W,' t :. p. ., '
POWER DISTRIBUTION LIMITS 3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR LIMITING CONDITION FOR OPERATION 3.2.3 F g shall be less than 1,46 [1.0 + 0.3 (1-P)) THERMAL POWER Where: P = RATED THERMAL POWER ( APPLICABILITY: MODE 1. ACTION: With Fh exceeding its limit:
- a. Within 2 hours reduce the THERMAL POWER to the level where the LIMITING CONDITION FOR OPERATION is satisfied.
- b. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the limit required by ACTION a.,
above;THERMALPOWERmaythenbeincreased,providedFhisdemon-strated through incore mapping to be within its limit. SURVEILLANCE RE001REMENTS 4.2.3.1 The provisions cf Specification 4.0.4 are not applicable. 4.2.3.2 Ffg shall be demonstrated to be within its limit prior to operation above 75% RATED THERMAL POWER after each fuel loading and at least once per 31 EFPD thereafter by:
- a. Using the movable incore detectors to obtain a power distribution map at any THERMAL POWER greater than 5% RATED THERMAL POWER,
- b. Using the treasured value of F q which does not include an allowance for measurement uncertainty.
SOUTH TEXA5 - UNITS 1 & 2 3/4 2-9 AMENDv!NT N?S. AND ' ' -
D POWER DISTRIBUTION LIMITS i 3/4.2.4 QUADRANT POWER TILT RATIO i LIMITING CONDITION FOR OPERATION 3.2.4 The QUADRANT POWER TILT RATIO shall not exceed 1.02, f APPLICABILITY: MODE 1, above 50% of RATED THERMAL POWER". ! ACTION: With the QUADRANT POWER TILT RATIO determined to exceed 1.02: l
- a. Within 2 hours reduce THERMAL POWEit at least 3% from RATED THERMAL !
POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of I and similarly reduce the Power Range Neutron Flux-High Trip Setpoint within the next 4 hours. '
- b. Within 24 hours and every 7 days thereafter, verify that Fq (Z) (by i xy evaluation)andFfg are within their limits by performing Surveil-F lance Requirements 4.2.2.2 and 4.2.3.2. THERMAL POWER and setpoint ;
reductions shall then be in accordance with the ACTION statements of Specifications 3.2.2 and 3.2.3. r SURVEILLANCE RE001REW.ENTS ' 4 2.4.1 The QUADRANT POWER TILT RATIO shall be determined to be within the , limit above 50% of RATED THERMAL POWER by' t
- a. Calculating the ratio at least once per 7 days when the alarm is !
OPERABLE, and - l
- b. Calculating the ratio at least once per 12 hours during steady-state I
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 i inoperable by using the movable incare detectors to confirm indicated QUADRANT POWER TILT RATIO at least once per 12 hours by either:
- a. Using the four pairs of syr.Tetric thimble locations, or
- b. Using the movable incore detection system to monitor the QUADRANT l
POWER TILT RATIO subject to the requirements of Specification 3.3.3.2. t I l I
*See Special Test Exceptions Specification 3.10.2. '
t SOUTH TEXA5 - UNITS 1 & 2 3/4 2-10 AMENDMENT h?S. AN. NCIVj; { i
, _ _ . . _ .____________-.____.___.__.__..__.__._.c
POWER DISTRIBUTION LIMITS 3/4.2.5 DNB PARAMETERS LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB related parameters shall be maintained within the limits following: a. Reactor Coolant System T,yg, 5, 598'F
- b. Pressurizer Pressure, > 2201 psig*
- c. Reactor Coolant System Flow, > 395,000 gpm**
A PLICABILITY: MODE 1. O ACT!0N: With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours. SURVEILLANCE RE001REMENTS 4.2.5.1 Each of the parameters shown above shall be verified to be within its limits at least once per 12 hours. Additionally, RCS flow shall be demon-strated within its limit prior to operation above 75% RTP after each fuel load-ing. The provisions of Specification 4.0.4 are not applicable for ve.ification that RCS flow is within its limit. 4.2.5.2 The RCS flow rate indicators shall be subjected to a channel calibra-tion at least once per 18 months. 4.2.5.3 The RCS total flow rate shall be determined by precision heat balance seasurements at least once per 18 months. Within 7 days prior to performing the precision heat balance flow measurement, the instrumentation used for per-4 forming the precision heat balance shall be calibrated.
"Lim t Nt N .,11 cable during either a Thermal Power raep in excess of 5% of RTP per m...ute or a Thermal vower step in excess of 20% RTP, ** Includes a 3.5% flow eessurement uncertainty.
SOUTH TEXAS - UNITS 1 & 2 3/4 2-11 AMENDMENT N05. AND
'I liq
FD 3/4.7 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION LIM 111NG CONDIT109 FOR OPERATION 3.3.1 As a minimum, the Reactor Trip System instrumentation channels and interlocks of Table 3.3-1 shall be OPERABLE with 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. SURVEILLANCE RE0VIREMINTS 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 Reactor 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 to be within its 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 every N times 18 months where N is the total number of redundant channels in a specific Reactor trip function as shown in the "Total No. of Channels" column of Table 3.3-1. SCUTH TEXAS - UNITS 1 & 2 3/4 3-1 AMENCMENT N05. AND t.'i. 3- ,
, TABLE 3.3-1 g . .
g REACTOR TRIP SYSTEM INSTRUMENTATION h MINImm g TOTAL NO. CHANNELS CHANNELS APPLICABLE
, FUNCTIONAL UNIT Of CHANNELS TO TRIP OPERABLE IWDES ACTION
- 51. Manual Reactor Trip 2 1 2 1, 2 4*, 5*
1 g 2 1 2 3*, 10 v
,, 2. Power Range, Neutron Flux y a. High setpoint 4 2 3 1, 2 2
- b. Low Setpoint 4 2 3 18##, 2 2
- 3. Power Range, Neutron Flux 4 2 3 1, 2 2 liigh Positive Rate t 4. Power Range, Neutron Flux 4 2 3 1, 2 2 4- liigh Negative Rate
'> S. Intermediate Range, Neutron Flux 2 1 2 188#, 2 3
- 6. Source Range, Neutron Flux
- a. Startup 2 1 2 20# 4
- b. Shutdown 2 1 2 3*, 4*, 5* 10
- 7. Extended Range, Neutron Flux 2 0 2 3, 4, 5 4 C
- 8. Overtemperature AT 4 2 3 1, 2 6 o
- 9. Overpower AT 4 2 3 1, 2 - 6 g 10. Pressurizer Pressure--Low 4 2 3 1 6 r (Interlocked with P-7)
- 11. Pressurizer Pressure-High 4 2 3 1, 2 6 2-0 12. Pressurizer Water Level - High 4 2 3 1 6 (Interlocked with P-7)
*9 9
D _vn,m-----nn-n---n---, - ,- ---n,_.. . - - - - - - ----n- -- ,-,,--,-,,-,----n .,. - - - - , , - - - - - - - - - , . - . - - -
y, TABLE 3.3-1 (Continued) REACTOR TRID SYSTEM INSTRUMENTATION M MINIMUM 5 TOTAL NO. CllANNELS CHANNELS APPLICABLE 7 TUNCTIONAL UNIT Of CHANNELS TO TRIP OPERABLE MODES ACTION 3 13. Peactor Coolant flow--tow
] a. Single Loop (Above P-8) 3/ loop 2/ loop in 2/ loop in 1 6( }
any oper- each oper-p ating loop ating loop
- b. Two Loops ( Above P-7 and 3/ loop 2/ loop in 2/ loop 1 6 below P-8) two oper- each oper-ating loops ating loop
- 14. Steam Generator Water 4/stm. gen. 2/sta. gen. 3/stm. gen. 1, 2 6(I) w Level--Low-tow in any oper- each oper-D ating stm. ating sta.
w gen. gen. w
- 15. Undervoltaga -Reactor Coolant 4-1/ bus 2 3 1 6 Pumps (Interlocked with P-7) .
- 16. Underfrequency--Reactor Coolant 4-1/ bus 2 3 1 6 Pumps (Interloc6ed with P-7)
- 17. Turbine Trip (Interlocked with P-9) iE a. Low Emergency Trip Fluid T3 Pressure 3 2 2 1 6 i b. Turbine Stop Valve Closure 4 2 3 1 6 r,
r 3 o w
.-_ ._ _ - - - - - -.. . - . - . ._- ..._ .. =-- -- . - . . _ - - _ _ . . - - . . _.
5 2 I 1 a j y, TABLE 3.3-1 (Continued)
- 8 l
I 2 REACTOR 1 RIP SYSTEN INSTRUNENTATION
~
l O MININUN { $ TOTAL NO. CHANNELS CHANNELS APPLICABLE
- , TimC110NAL tmIT OF CHANNELS TO TRIP OPERABLE N00ES ACTION i
i c j $ IP, . Safety injection Input ! O from ESTA5 2 1 2 1, 2 9
- p. 19. Reactor Trip System Interlocks y a. Intere*diate Range Neutron Ilux, P-6 2 1 2 2M 8
- b. Low Power R-actor Trips Block, P-7 i F-10 Input 4 2 3 1 8 j t:* or j
P-13 Input 2 1 2 1 8 1 Y' j c. Power Range Neutron flux, P-8 4 2 3 -1 8
- d. Power Range Neutron 4 2 3 1 8 i Flux, P-9 i
k e. Power Rang
- Neutron j , Flux, P-10 4 2 3 1,2 8
,=
} sf f. Turbine Impulse Chamber j g Pressure, P-13 2 1 2 1 8 i
;5 20. Reactor Trip Breakers 2 1 2 1, 2 9, 12 r .
2 1 2 3*, 4*, 5* 10
?
e e
;. g
TABLE 3.3-1 (Continued) E ju REACTOR TRIP SYSTEM INSTRUMENTATION M HINIMUM E TOTAL NO. CHAN;iELS CHANNELS APPLICABLE
" ACTION iIINCTIONAL trNIT OF CHANNELS TO TRIP OPERABLE MODES 3 21. Automatic Trip and Interlock 2 1 2 1, 2 9 Logic 2 1 2 3*, 4*, 5* 10 O.
v H 9* ea s. v.
'l
's:
a - of% o 6 e (* 4 M e
fd TABLE 3.3-1 (Continued) TABLE NOTATIONS
*When the Reactor Trip System breakers are in the closed position and the Control Rod Drive System is capable of rod withdrawal.
##Below the P 6 (Intermediate Range Neutron Flux Interlock) Setpoint.
###Below the P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.
(1}The applicable MODES and ACTION statement for these channel's noted in Table 3.3-3 are more restrictive and, therefore, applicable. 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 hours or be in HOT STANDBY within the next 6 hours. ACTION 2 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied;
- a. The inoperable channel is placed in the tripped condition within 6 hours,
- b. The Minimtm Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours 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 THERPAL POWER and the Power Range Neutron Flux Trip Setpoint is reduced to less than or equal to 85% of RATED THERFAL POWER within 4 hours; or, the QUADRANT POWER TILT RATIO is monitered at least once per 12 hours per Specification 4.2.4.2.
SOUTH TEXA5 - UNITS 1 & 2 3/4 3 6 AWEN0*ENT N05. AND
' I .a
FL TABLE 3.3-1 (Continued) ACTION STATEMENTS (Continued) , i ACTION 3 - With the number of channels OPERABLE one less than the Minimum - Channels OPERABLE requirement and with the THERMAL POWER level:
- a. Below the P-6 (Intermediate Range Neutron Flux Interlock) .
Setpoint, restore the inoperable channel to OPERABLE I status prior to increasing THERMAL POWER above the P-6 : Setpoint, and '
- 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 I 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 - (Not Used) ACTION 6 - With the number of OPERABLE channels one less than the Total i Number of Channels, STARTUP and/or POWER OPERATION may proceed l provided the following conditions are satisfied i
- a. The inoperable channel is placed in the tripped condition l within 6 hours, and (
- b. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours l for surveillance testing of other channels per !
Specification 4.3.1.1. [; ACTION 7 - (Not Used) l ACTION 8-WithlessthantheMinimumNub.berofChannelsOPERABLE,within j 1 hour cetermine by observation of the associated permissive ' annunciator window (s) that the interlock is in its required state ! for the existing plant coadition, or apply Specification 3.0.3. j ACTION 9 - With the number of OPERABLC eh eneis one less than the Minimum ! Channels OPERABLE requirement, be in at least HOT STAN0BY t within 6 hours; however, one channel may be bypassed for up to : 2 hours for surveillance testing per Specification 4.3.1.1, ! provided the other channel is OPERABLE. i i r P e l l SOUTH TEAAS - UNITS 1 & 2 3/4 3-7 AMENDMENT N05. ANO
FC i Af f .,f, l : Continued) E'j.f N . ,AJEr' QTS (Continued) ACTION 10 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel
- to OPERABLE status within 48 hours or open the Reactor Trip System breakers within the next hour.
ACTION 11 - (Not Used) ACTION 12 - With one of the diverse trip tes;ures (undervo'ltage or shunt trip attachment) inoperable, restore it to OPERABLE status within 48 hours 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 0FERABLE status, J i f i SOUTH TEXAS - UNITS 1 & 2 3/4 3-8 AMEN; MENT N05. AND 143
TABLE 3.3-2 5 1 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES A h iUNCT10NAL UNIT RESPONSE TIE
- 1. Manual Reactor Trip N.A.
7 3 2. Powr Range, Neutron Flux
, $ 0.5 second*
[ 3. Power Range, Neutron Flux, y High Positive Rate M.A.
- 4. Power Range, Neutron Flux.
High Negative Rate 1 0.5 second*
- 5. Intermediate Range, Neutron Flux N.A.
- 6. Source Range, Neutron Flux $ 0.5 second*
E 7. Exte & d Range, Neutron Flux N.A.
- 8. Overtemperature AT 1 8.0 seconds *
- 9. Overpower AT $ 8.0 secor.ds"
- 10. Pressurizer Pressure--Low $ 2 seconds iju 11. Pressurizer Pressure--High 1 2 seconds
- 12. Pressurizer *.later Level-High ! 2 seconds 5
r
' Neutron detectors are ex M t from response time testing. Response time of the neutron flux signal portion f 3 of the channel shall be maasured from detector output or input of first electronic component in channel.
l :- .
*9 j
h.
.._-_- - - ._.___ - - ,- ..._ . . - - . . _ = . _ . - . - _ -. _.
>l t _ .
^
t
. TABLE 3.3-2 (Continued)
E g RfACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIE S M h FUNCTIONAL UNIT RESPONSE TIE
$ 13. Reactor Coolant flow--Low a
g a. Single Loop (Above P-8) $ 1 second y b. Two Loops (Above P-7 and below P-8) $ 1 second [ 14. Steam Generator h ter level--Low-Low $ 2 seconds
- 15. Undervoltage - Reactor Coolant Pumps 1 1.5 seconds
- 16. Underfrequency - Reactor Coolant Pumps 1 0.6 second ta 17. Turbine Trip s.
Y a. Low Emergency Trip Fluid Pressure N. A. g b. Turbine Stop Valve Closure N. A.
- 18. Safety injection Input from ESFAS N.A.
- 19. Reactor Trip System Inte locks M.A.
- 20. Reactor Trip Breakers N.A.
7-h 21. Automatic Trip and Interlock logic M.A. I$ Y 3, be
*O h
. . _ _ . _ _ = - -- _ . _ -
, TABLE 4.3-1 o
h REACTOR TRIP SYSTEN INSTRUMENTATION SURVEILLAfeCE REQUIREENTS m TRIP h ANALOG ACTUATING MODES FOR
, CHANNEL DEVICE idHICN CHANNE L CHANNEL OPERATIONAL OPERATION 4l. ACTUATION SURVEILLANCE
$~
TUNCTIONAL UNIT ClitCK CALIBRATION TEST TEST LOGIC TEST 15 REQUIRED w' 1. Manual Reactor Trip N.A. N.A. ,N.A. R(14) N.A. 1, 2, 3*, 4*, 5* [ 2. Power Range, Neutron . flux
- a. High Sctroint 5 D(2, 4), Q(17) N.A. M.A. 1, 2 P(3. 4),
Q(4,6), R(4, 5)
- b. Low Setpoint 5 R(4) 5/U(1) N.A. N.A. 1***, 2
{ 3. Power Range, Neutron M.A. R(4) Q(17) N.A. N.A. 1, 2 Flux, [ High Positive Rate
- 4. Powar Range, Neutron M.A. R(4) Q(17) N.A. N.A. 1, 2 Flux, High Negative Rate .
S. Intermediate Range, S R(4,5) 5/U(1) N.A. N.A. 1***, 2 Neutron Flux 3- 6. Source Range, Neutron S R(4, 5) S/U(1), E. Ilux Q(9)(17) N.A. N.A. 2**, 3, 4, 5 x M 7. Extended Range, S R(4) Q(12,17) N.A. N.A. 3, 4, 5
'5' Neutron Flux g 8. Overtemperature AT S R Q(17) N.A. N.A. 1, 2
- 9. Overpower AT S R Q(17) N.A. M.A. 1, 2
- 10. Pressurizer Pressure 2- --Low 5 R Q(17) P .. A. N. A.- 1 3
~ M1 w n.,,_, -- - - ,,,- -- - -- , ~ ~ - - - - - - - . - - - - r - . , - - - - - - . -- - ,-- .- . - - - , --
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . - _ _ _ _ _ _ _ - = _ _ - _ - _ _ . __ . - . __ __ - - - . - . .
, TABLE 4.3-1 (Continued)
E y REACTOR TRIP SYSTEN INSTRUNENTATION SURVEILLANCE REQUIREMENTS w 0 TRIP g ANA'_0G ACTUATING BRIDES FOR CHANNEL DEVJCE idHICH h CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE , 6 I(MCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST 15 REQUIRED l M y 11. Pressuriier Pressure
,, --High 5 R Q(17) N.A. N.A. 1, 2 1
n l 12. Pressurizer Water I level--High 5 R Q(17) N. A. N.A. 1 l 13. Reactor Coolant flow l
--Low $ R Q(17,18) N.A. M.A. I t'
** 14 Steam Generator Water y Level--Low-tow 5 R Q(17,18) N.A. M.A. 1, 2 l ,
j 15. Undervoltage - Reactor j roolant Pumps N.A. R N.A. Q(17) N.A. 1 l
- 16. Underfrequency -
Reactor Coolant Pumps N.A. R M.A. Q(17) N.A. 1
?
fj 17. Turbine Trip o .s
;g a. Low Emergency N.A. R M.A. S/U(1. 10) N.A. 1 5 Trip Fluid
, Pressure
- b. Turbine Stop N.A. R N.A. S/U(1,TO) N.A. I
! Valve Closure I y I P. , Safety Injection M.A. N.A. N.A. R N.A. 1, 2 j u Input from ESTAS t t y 4 .O.. -. (- _ , . . . . - - _ , _ _ . _ _ _ _ . _ _ , - _ , , _ , _ _ _ . _ . _ _ - _ - - . _ , _ . . _ . . . _ _ _ _ , . , _ _ - , . . _ _ . . _ .
7 tw _ ~' . .sy j
.r
, TABLE 4.3-1 (Continued) o 5
1 REACIOR TRIP SYSTEN INSTRUNENTATION SURVEILLANCE REQUIRENENTS M TRIP 1 5 ANALOG ACTUATING N00ES F0R CHANNEL DEVICE milch l i CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE E FUNCTIONAL 'JNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRE 0
- 19. Reactor Trip System Interlocks i
- a. Intermediate Range Neutron Flux, P-6 N.A. R(4) R N.A. N.A. -2 " *
- h. Low Power neactor Trips Block, P-7 N.A. R(4) R N.A. W.A. I
- c. Power Range Neutron w Flux, P-8 N.A. R(4) R N A. N.A. 1
- d. Power Range Neutron Y
w flux, P-9 N.A. R(4) R N.A. h.A. I
- e. Power Range Neutron Flux, P-10 N A. R(4) R N.A. N.A. 1, 2
- f. Turbine Impulse Chamber Pressure, P-13 N.A. R R N.A. N.A. 1
- 20. Reactor Trip Breaker N.A. N.A. N.A. N(7,11) N.A. 1, 2, 3*, 4*,_5*
l 'S 21 Automatic Trip and Interloc'< N A. N.A. N.A. N.A. M(7) 1, 2, 3*, 4*, 5* l
] Logic
=-
- 22. Reactor Trip Bypass Breaker N.A. N.A. N.A. N(15),R(16) N.A. 1, 2, 3*, 4*, 5*
5 r l E o l *N i f9 ? -. l .
.h D
b TABLE 4.3-1 (Continued) TABLE NOTATIONS
*When the Reactor Trip System breakers are closed and the Control Rod Drive System is capable of rod withdrawal.
**Below P-6 (Intermediate Range Weutron Flux Interlock) Setpoint.
***Below P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.
(1) If not performed in previous 31 days. (2) Comparison of calorimetric to excore power indication above 15% of RATED THERMAL POWER. Adjust excore channel gains con;istent with calorimetric power if absolute difference is greater than 2%. The provisions of Specification 4.0.4 are not applicable to entry into MODE 2 or 1. (3) Single point comparison uf incore to excore AXIAL FLUX DIFFERENCE' above 15% of RATED THERMAL POWER. Recalibrate if the absolu+ difference is greater than or equal to 3%. The provisions o Specification 4.0.4 are not applicable for entry into MODE 2 or 1. (4) Neutron detectors may be excluded from CHAN i CALIBRATION. (5) Detector plateau curves shall be obtained and evaluated. For the Inter-mediate Range and Power Range Neutron Flux channels the provisions of Specification 4.0.4 are not applicable for entry inte MODE 2 or 1. (6) Incore - Excore Calibration, above 75% of RATED THERMAL POWER. The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1. (7) Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS. (8) (Not Used) (9) Quarterly surveillance in MODES 3*, 4*, and 5* shall also include verification that permissives P-6 and P-10 are in their required state for existing plant conditions by observation of the permissive l annunciator window. l SOUTH TEXAS - UNITS 1 & 2 3/4 3-14 AMENT,..!NT N05. AND 1,; ; ,' N..g
FL TABLE 4.3-1 (Continued) TABLE NOTATIONS (Continued) (10) Setpoint verification is not applicable. (11) The TRIP ACTUATING DEVICE OPERATIONAL TEST shall independently verify the OPERABILITY of the undervoltage and shunt trip attachments of the Reactor Trip Breakers. (12) OPERABILITY shall be verified by a check of memory devices, input accuracies, Boron Dilution Alarm setpoints, output values, and software functions. (13) (Not used) (14) The TRIP ACTUATING DEVICE OPERATIONAL TEST shall independently verify the OPERABILITY of the undervoltage and shunt trip circuits for the Manual Reactor Trip Function. 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. (16) Automatic undervoltage trip. (17) Each channel shall be tested at least every 92 days on a STAGGERED TEST BASIS. (18) The surveillance frequency ar.d/or MODES specified for these channels in Table 4.3-2 are more restrictive and, therefore, applicable. . 1 i i i 1 SOUTH TEXAS - UNITS 1 & 2 3/4 3-15 AMENDMENT NDS. AO
' i ; ;g .
FL INSTRUMENTATION 3/4.3.2 ENGINEERED SACETY FEATURES ACTUATION SYSTEM INSTRUMENTATION , 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 values shown in the Trip Setpoint column of Table 3.3-4 and with RESPONSE TIMES as shown in Table 3.3-5. APPLICABILITY: 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 ESFAS Instrumentation or Interlock Trip Setpoint less conserva- l tive than the value shown in the Allowable Value column of Table ;
3.3-4, either:
- 1. Adjust the Sotpoint consistent with the Trip Setpoint value of Table 3.3-4, and determine within 12 hours 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: ] l 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 error, or the value from Column 5 (Sensor Error) of Table 3.3-4 for the affected channel, and l TA = The value from Column TA (Total Allowance) of Table 3.3-4 i for the affected channel. ; i
- c. With an ESFAS instrumentation channel or interlock inoperable, take the ACTION shown in Table 3.3-3. i i
SOUTH TEXAS - UNITS 1 & 2 3/4 3-16 AMENDMENT N05. AND I I
fi INSTRUMENTATION ! I I SURVEILLANCE RE001REMENTS 4.3.2.1 Each ESFAS instrumentation channel and interlock and the automatic actuation logic and relays shall be demonstrated OPERAB6E by performance of the ESFAS Instrumentation Surveillance Requirements specified in Table 4.3-2. 4.3.2.2 The ENGINEERED 5AFETY FEATURES RESPONSE TIME cf each ESFAS function shall be demonstrated to be within the limit at least ce per 18 months. Each test shall include at least one train so that:
- a. Each logic train is tested at least once per 36 months,
- b. Each actuation train is tested M least once per 54 months *, and
- c. One channel per function so 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 as shown in the "Total No.
of Channels" column of Table 3.3-3. i i
*1f an ESFAS instrumentation channel is inoperable due to response times exceeding the limits of Table 3.3-5, perform an engineering evaluation to determine if the test failure is a result of degradation of the actuation relays. If degradation of the actuation relays is determined to be the cause, increase the ENGINEERED SAFETY FEATURES RESPONSE TIME surveillance frequency such that all trains are tested at least once per 36 months.
3/4 3-17 AMENDMENT N05. AND SOUTH TEXAS - UNITS 1 & 2 z '.
- l ; ..
t..
o, TABLE 3.3-3 8
;! ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION A
x
? MINIMUM
, TOTAL NO. CHANNELS CHANNELS APPLICA8tE e FUNCTIONAL UNIT OF CitANNELS TO TRIP OPERABLE MDOES ACTION E.
Lt 1. Safety Injection (Reactor
,a Trip, Feedwater Isolation, ,
- p. Control Roo= Emergency n, Ventilation, Start Standby Diesel Generators, Reactor Containment fan Coolers, and Essential Cooling Water).
- a. Manual Initiation 2 1 2 1,2,3,4 19
. tf ' *- b. Automatic Actuation ) 97 Logic 2 1 2 1,2,3,4 14 I E$ j c Actuation Relays 3 2 3 1,2,3,4 14 i i
- d. Containment 3 2 2 1,2,3,4- 15 i Pressure--High-1
- e. Pressurizer 4 2 3 1, 2, 38- 20 3; Pressure--Low m
x gg f. Compensated Steam 3/ steam line 2/ steam line 2/ steam line 1, 2, 38 15 q gg Line Pressure-Low any steam line in each steam .
-d l line
=
0 1 ) i l 2) ta 9 i 1 1 11
m i TABLE 3.3-3 (Continued) 55 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION
, 1
! -4 m ) 5 MINIMUM MINIMUM f
- TOTAL NO. CHANNELS CHANNELS APPLICABLE I FUtiCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION i
E
] 2. Containment Spray J
- a. Manual Initiation 2 1 with 2 1,2,3,4 19 2 coincident switches
)
- b. Automatic Actuation Logic 2 1 2 1,2,3,4 14
, c. Actuation Relays 3 2 3 1,2,3,4 14 w 2: d. Containment Pressure-- 4 2 3 1,2,3 17 { g. High-3
- 3. Containment Isolation .
- a. Phase "A" Isolation 4
- 1) Manual Initiation 2 1 2 1,2,3,4 19 2 2) Automatic Actuation 35 Logic 2 1 2 1,2,3,4 14 55
[j 3) Actuation Relays 3 2 3 1,2,3,4 14 4 z 4) Safety Injection See Item 1. above for all Safety Injection initiating functions and y? requirements. 5 o N "O q l
TABLE 3.3-3 (Continued) ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION o S MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE OF CHANNELS TO TRIP OPERABLE MODES ACTION M FUNCTIONAL UNIT b *
- 3. Containment Isolation (Continued)
- b. Containment Ventilation E
Isolation d 1) Automatic Actuation 1, 2, 3, 4 18
- Logic 2 1 2 1,2,3,4 18
{ 2) 3) Actuation Relays *** SafetyInjection*** 3 2 3 See Item 1. above for all Safety Injection initia*ing functions and requirements.
- 4) RCB Purge y Radioactivity-High 2 1 2 1,2,3,4,5 ,,6,, 18
- 5) Containment Spray- See Item 2. above for Containment Spray manual initiating functions w Manual Initiation and requirements.
M 6) Phase "A" Isolation- See Item 3.a. above for Phase "A" Isolation manual initiating Manual Isolation functions and requirements.
- c. Phase "B" Isolation
- 1) Automatic Actuation 2 1 2 1,2,3,4 14 Logic
- 2) Actuation Relays 3 2 3 1,2,3,4 14
- 3) Containment Pressure-- 4 2 3 1,2,3 17 73 g High-3 5 4) Containment Spray- See Item 2. above for Conta?nment Spray manual initiating Manual Initiation functions and requirements.
G
- d. RCP Seal Injection Isolation w 1) Automatic Actuation 1 1 1 1,2,3,4 31 5 Logic and Actuation Relays 3 .-
4 e, s
TABLE 3.3-3 (Continued)
$ ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION M
y MINIMUM
;' TOTAL NO. CH.ANNELS CHANNELS APPLICABLE g; FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION
$ 2) Charging Header 1 1 1 1,2,3,4 31 5 Pressure - Low 5 Coincident with Phase See L.*m 3.a. above for Phase "A" Isolation initiating functions
~ "A" Isolation and requirements e
" 4. Steam Line Isolation
- a. Manual Initiation
- 1) Individual 2/ steam line 1/ steam line 2/ operating 1,2,3 24 t steam line u
y 2) System 2 1 2 1,2,3 23
- b. Automatic Actuation 2 1 2 1,2,3 22 Logic and Actuation Relays
- c. Steam Line Pressure -
Negative Rate--High 3/ steam line 2/ steam line 2/ steam line 3### 15 y any steam in each steam g line line 5 g d. Containment Pressure - 3 2 2 1,2,3 15 5 High-2 h
- e. Compensated Steam Line Pressure - Low 3/ steam line 2/ steam line any steam 2/ steam line in each steam 1, 2, 3# 15 line line N
0 l[ m e
d TABLE 3.3-3 (Continued) c3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION 5 1
- MINIMUM -
TOTAL NO. CHANNELS . CHANNELS APPLICABLE E? OF CHANNELS TO TRIP gRABLE MODES ACTION C FUNCTIONAL UNIT
- 5. Turbine Trip and Feedwater Isolation 55 a. Automatic Actuation 2- 1 2 1,2,3 25 E% logic and Actuation
- u. Relays .
$ b. Steam Gener.'ar 4/sta. gen. 2/sta. gen. 3/sta. gen. 1,2,3 20 Water Level-- in any oper . in each High-liigh (P-14) ating stm. operating gen. stm. gen.
u, c. (This functional unit number is not used.) li u, d. (This functional unit number is not used.) c'3
- e. Safety Injection See Item 1. for all Safety Injection initiating functions and requirements. .
- f. T,,g-Low coincident wi?.h Reactor Trip (P-4) 4 (1/ loop) 2 3 1,2,3 20 (Feedwater Isolation Only) 4 5'
a rve 5 5 r E o
- tj -
,~. ta
, TABLE 3.3-3 (Continued) o h ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION 5 MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE
[ FUNCTIONAL 'JNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION z h 6. Auxiliary feedwater [ a. Manual Initiation . 1/ pump 1/ pump 1/ pump 1,2,3 26 m b. Automatic Actuation Logic 2 1 2 1,2,3 22
- c. Actuation Rel.,ys 3 2 3 1,2,3 22
- d. Sim. Gen. Water Level--
Low-Low Start Motor- { Driven Pumps 4/sta. gen. 2/sta. gen. 3/sta. gen. 1, 2, 3 20
, and Turbine- in any stm. in each 4 Driven Pump gen. sta. gen.
- e. Safety Injection See Item 1. above for all Safety Injection initiating functions and requirements.
- f. Loss of Power (Motor See Item 8. below for all Loss of Power initiating functions and Driven Pumps Only) requirements.
- 7. Automatic Switchover to k= Containment Sump ****
M a. Aastomatic Actuation 3-1/ train 1/ train 1/ train 1, 2, 3. 4 19
'7] Logic and .*r'_~= tion, Relays G
- b. RWST Level--Low-Low 3-1/ train 1/ train 1/ train 1, 2, 3, 4 19 Coincident With: See Item 1. above for all Safety Injection iiltiating function:
$ Safety Injection and requirements.
-T.
6q .: s
Bs 4 1
! . TABLE 3.3-3 (Continued) -
O
; y ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION
] O 3; MINIMUM , , TOTAL NO. CHANNELS CHANNELS APPLICA8LE~ g FUNCTIONAL UNIT 01 CHANNELS TO TRIP OPERABLE MODES ACTION 1 = m 8. Loss of Power -
- a. 4.16 kV ESF Bus Under-
~ ; [ 4/ bus 2/ bus 3/ bus 1, 2, 3, 4 20 - , m voltage-Loss of Voltage
- b. 4.16 kV ESF Bus Under-
! voltage-Tolerable Degraded Voltage ! Coincident with SI 4/ bus 2/ bus 3/ bus 1, 2, 3, 4 20 ! Y q 3-
- c. 4.16 LV ESF Bus Under-
- y voltage - Sustained y Degraded Voltage 4/ bus 2/ bus 3/ bus 1, 2, 3, 4 20
- 9. Engineered Safety Features i
Actuation System Interlocks ) a. Pressurizer Pressure, 3 2 2 1,2,3 21 1 P-11 , h 73
- b. Low-Low Tavo, P-12 4 2 3 1,2,3 21
- c. Reactor Trip, P-4 2 1 2 1,2,3 23 .
4 7
- . t i &
v 9
- e, :.
w TABLE 3.3-3 (Continued) ENdINEEREDSAFETYFEATURESACTUATIONSYSTEMINSTRUMENTATION 5 MINIRJM TOTAL NO. CHANNELS CHANNELS APPLICABLE [ FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION
=- ] 10. Control Room Ventilation .- a. Manual Initiation 3(1/ train) 2(1/ train) 3(1/ train) All 2-
[ b. Safety Injection See Item 1. above for all Safety Injection initiating functions and requirements.
- c. Automatic Actuation Logic 3 2 3 All 27 and Actuation Relays
- d. Control Room Intake Air 2 1 2 All 28 w Radioactivity - High
[ e. Loss of Power See Item 8. above for all loss of Power initiating functions 4 and requirements. w II. FHB HVAC -
- a. Manual Initiation 3(1/ train) 2(1/ train) 3(1/ train) 1, 2, 3, 4 or 29, 30 with irradiated fuel in spent fuel pool E b. Automatic Actuation 3 2 3 1, 2, 3, 4 or 29, 30 9 Logic and Actuation with irradi-E Relays ated fuel in j spent fuel
_, Pool
% c. Safety Injection See Item 1. above for all Safety Injection initiating functions and requirements.
- d. Spent fuel Pool Exhaust 2 1 2 With irradi- 30
$ Radioactivity - High ated fuel in spent fuel pool
_[' ,, e
t FI> TABLE 3.3-3 (Continued) TABLE NOTATIONS
*** Function is actuated by either actuation train A or actuation train B.
Actuation train C is not used for this function.
**** Automatic switchover to containment sump is accomplished for each train i
using the corresponding RWST level transmitter. ;
# Trip function may be blocked in this H0DE below the P-11 (Pressurizer Pressure Interlock) Setpoint. .
##During CORE ALTERATIONS or movement of irradiated fuel within containment.
### Trip function automatically blocked above P-11 and may be blocked below P-11 when Low Compensated Steamline Pressure Protection is not blocked.
ACTION STATEMENTS ACTION 14 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours; however, one channel may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.2.1, provided the other channel is OPERABLE. ACTION 15 - With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed until performance of the next required ANALOG CHANNEL OPERATIONAL TEST provided the inoperable channel is placed in the tripped condition within 1 hour.
. ACTION 16 - (Not Used)
ACTION 17 - With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed previded the inoperable 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 hours for surveillance testing per Specification 4.3 2.1. ACTION 18 - With less than the Minimum Channels OPERABLE requirement, operation may continue provided the containment purge supply and exhaust valves are maintained closed. 3/4 3-26 AMENDMENT N05. AND SOUTH TEXAS - UNITS 1 & 2
Frc TABLE 3.3-3 (Continued) ACTION STAT _EMENTS (Continued) i ACTION 19 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restere the inoperable channel to OPERABLE status within 48 hours or be in at least HDT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. ACTION 20 - 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 lare satisfied:
- a. The inoperable channel is placed in the tripped condition within I hour, and
- b. The Minimum Channels OPERABLE requirement is met; however, one additional channel may be bypassed for up to 2 hours for surveillance testing of other channels per Specification 4.3.2.1.
ACTION 21 - With less than the Minimum Number of Channels OPERABLE, within 1 hour determine by observation of the associated permis:ive annunciator window (s) that the interlock is in its required state for the exist 3ng plant condition, or apply Specification 3.0.3. ACTION 22 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours and in at least HOT SHUTOOWN within the following 6 hours; however, one channel may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.2.1 provided the other channel is OPERABLE. ACTION 23 - With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 4B he'Jrs or be in at least HOT STANDBY within 6 hours and in at least HOT SHUT 00WN within the following 6 hours. , ACTION 24 - With the number of OPERABLE channels one less than the Total Nur.ber of Channels, restore the inoperable channel to OPERABLE status within 49 hours or declare the associated valve inoperable and take the ACTION required by Specification 3.7.1.5. ACTION 25 - With the t# umber of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours; however, one channel may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.2.1 provided the other channel is OPERABLE. SOUTH TEXAS - UNr.S 1 & 2 3/4 3-27 AMEN 0 MENT N05. AND ..,
~'I: ::;3
FL' M TABLE 3.3-3 (Continued) ACTION STATEMENTS (Continued) ACTION 26 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, declare the affected Auxiliary Feedwater Pump inoperable and take ACTION required by Specification 3.7.1.2. ACTION 27 - MODES 1, 2, 3, 4: With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTOOWN within the following 30 hours. MODES 5 and 6: With the number of OPERABLE channels less than the Minimum Channels OPERABLE requirement, restore the inoperable Channel to OPERABLE status within 48 hours or initiate and main-tain operation of the Control Room Makeup and Cleanup Filtration System (at 100% capacity) in the recirculation and makeup filtra-tion mode. ACTION 28 - MODES 1, 2, 3, 4: With the number of OPERABLE channels less than the Minimum Channels OPERABLE requirement, within I hour isolate the Control Room Envelope and maintain operation of the venti-lation system in the filtered recirculation mode. I MODES 5 and 6: With the number of OPERABLE channels less than the Minimum Channels OPERABLE requirement, within I hour initiate and maintain operation of the Control Room Makeup and Cleanup Filtration System (at 100% capacity) in the recirculation and makeup filtration mode. ACTION 29 - MODES 1, 2, 3, 4: With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours or either initiate and maintain operation of the FHB exhaust air filtration or be in at least HOT STANDBY within system (at 100% capacity)OLD SHUT 00WN within the following 30 the next 6 hours and in C hours. ACTION 30 - With irradiated fuel in the spent fuel pool: With the number of OPERABLE cnannels less than the Minimum Channels OPERABLE fuel movement within the spent fuel pool ~or crane ' requirement,th
-;1 ration wi loads over the spent fuel pool may prot.eed provided the FHB exhaust air filtration system is in operation and discharging through at least one train of HEPA filters and charcoal adsorbers.
ACTION 31 - With the Charging Header Pressure channel inoperable: a) Place the Charging Header Pressure channel in the tripped condition within one hour and b) Restore the Charging Header Pressure channel to operable status within 7 days or be in at least Hot Standby within the next 6 hours and in Cold Shutdown within the following 30 hours. AND SOUTH TEXAS - UNITS 1 & 2 3/4 3-28 AMENDMENT NDS. NOV I i 15;g
y
.i
~
TABLE 3.2-4 E h ENGINEERED SAFETY FEATURES ACTUATION SYSTEM _ INSTRUMENTATION TRIP SETPOINTS M TOTAL SENSOR ERROR h FUNCTI0dAL UNIT ALLOWANCE (TA) Z (S) TRIP SETPOINT- ALLOWABLE VALUE [ 1. Safety Injection (Reactor Trip, z Feedwater Isolation Control 3 Room Emergency Ventilation, Start Standby Diesel Generators, Reactor Containment Fan Coolers, and [ Essential Cooling water)
- 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.
- c. Actuation Relays N.A. N.A. N.A. N.A. M.A.
y d. Containment Pressure--High 1 3.6 0.71 2.0 $ 3.0 psig 5 4.0 psig
- e. Pressurizer Pressure--Low 13.1 10.71 2.0 1 1850 psigN 1 1842 psigN
- f. Compensated Steam Line 13.6 10.71 .
- 2. 0 1 735 psig 1 714.7 psig*
Pressure-Low
- 2. Containment Spray 2
A a. Manual Initiation N.A. N.A. N.A. N.A. N.A. 5
- b. Automatic Actuation Logic N.A. N.A. N.A. N.A. M.A.
g c. Actuation Relays N.A. N.A. N.A. N.A. M.A.
- d. Containment Pressure--Hign-3 3.6 0.71 2.0 $ 9.5 psig 5 10.5 psig
'A o
- a en
TABLE 3.3-4 (Continued) o 4 S ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS SENSOR ERROR
# TOTAL TRIP SETPOINT ALLOWABLE VALUE > FUNCTIONAL UNIT ALLOWANCE (TA) 2 (S)
- 3. Containment Isolation 5
- a. Phase "A" Isolation
- N.A. N.A. N.A. N.A. N.A.
- 1) Manual Initiation .
- 2) Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A.
o-
" N A. M.A. N.A. N.A.
- 3) Actuation Relays H.A.
- 4) Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and Allowable Values.
- b. Containment Ventilation Isolation N.A. N.A. N.A.
5 1) Automatic Actuation N.A. N.A. y logic N.A. N.A. N.A. e; 2) Actuation Relays N.A. N. A. .
- 3) Safety Injection See Ites 1. above for all Safety Injection Trip Setpoints and Allowable Values.
<5x10 4
# <6.4x10 4
- 4) RCB Purge 3.1x10- 1.8x10- 1.3x10-Radioactivity-liigh pCi/cc pCi/cc pCi/cc jiCi/cc jiCi/cc
- 5) Containment Spray - See Item 2. above for Containment Spray manual initiation Trip g Manual Initiation Setpoints and Allowable Values.
5 6) Phase "A" Isolation - See Item 3.a. above for Phase "A" Isolation manual initiation Manual Initiation Trip Setpoints and Allowable Values.
@O
- c. Phase "B" Isolation M.A. M.A. N.A.
b 1) Automatic Actuation N.A. N.A. Logic N.A. N.A. N.A. N.A. y 2) Actuation Relays N.A.
" 3.6 0.71 2.0 1 9.5 psig i 10.5 psig
_ 3) C:ntainment Pressure-- U High-3
- 4) Containment spray- See Item 2. above for Containment Spray manual initiation Trip 2
.. Manual Initiation Setpoints and Allowable Values. y e
i
~
TABLE 3.3-4 (Continued) 8 g ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS
;;; TOTAL SENSOR ERROR y FUNCTIONAL UNIT ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE
- d. RCP Seal Injection Isolation b 1) Automatic Actuation N.A. N.A. N.A. N.A. N.A.
- Logic and Actuation
" Relays -
e-m 2) Charging Header 4.6 1.01 2.0 1 561.0 1 494.5 Pressure - Low .. Coincident with See Item 3.a. above for Phase "A" Isolation Setpoints and Allowable Phase "A" Isolation Values ( 4 Steam Line Isolation R a. Manual Initiation N.A. N.A. N.A. N.A. N.A. ,
*= l Y b. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A. I
$ and Actuation Relays t
~
- c. Steam Line Pressure - 2. 6 0.5 0 -< 100 psi i 126.3 psi ** !
Negative Rate--liigh l
- d. Containment Pressure - 3.6 0.71 2.0 $ 3.0 psig < 4.0 psig ,
High-2 h e. Compensated Steam Line 13.6 10.71 2.0 1 735 psig 1 714.7 psig*
;5 Pressure - Low l x
- 35. Turbine Trip and Feedwater Isolation 2
o , P a. Automatic Actuation Logic N.A. N.A. N.A. M.A. N.A. I and Actuation Relays [ t i G b. Steam Generator Water 4.5 2.35 2.0+0.2# -< 87.5% of -< 88.9% of Level--liigh-liigh (P-14) narrow range narrow range instrument instrument
. span. span.
- c. (This functional unit number is not used.) y b
_ _ _ _ __ _ . . _ _ _ . _ . _ _. _ _ _ __ _ m y, TABLE 3.3-4 (Continued) 8 - - l ;! ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUNENTATION TRIP SETPOINTS t () TOTAL SENSOR ERROR :
$; FUNCTIONAL UNIT ALLOWANCE (TA) Z (S) TRIP SETPOIP.T ALLOWABLE VALUE ;
h 5. Turbine Trip and Feedwater i 35 Isolation (Continued) i C l
,, d. (This functional unit number is not used.)
$ e. Safety Injection See Item 1 above for all Safety Injection Trip Setpoints and Allowable Values.
f. T,,g-Low Coincident with 4.5 1.36 0.8 > 574*F > 571.1*F Reactor Trip (P-4) (Feedwater Isolation Only) T d-
- 6. Auxiliary feedwater Y
R* a. Manual Initiation N.A. N.A. M.A. N.A. N.A.
- b. Automatic Actuation Logic N.A. N.A. M.A. M.A. N.A.
- c. Actuation Relays N.A. N.A. N.A. N.A. M.A.
- d. Steam Generator Water 15.0 12.75 2.0+0.2f > 33.0% of > 31.5% of Level--Low-Low narrow range narrow range 3, instrument instrument gi span. span.
z E? e. Safety Injection See Item 1. above for all Safety Injection Trip '
$f Setpoints and Allowable Values.
w (_ a
?
E
.- es e "O
TABLE 3.3-4 (Continued) S ENGINEERED SAFETY FEATURES ACTUATION SYSTEN INSTRUMENTATION TRIP SETPOINTS M TOTAL SENSOR ERROR (S) TRIP SETPOINT ALLOWABLE VALUE h FUNCTIONAL UNIT ALLOWANCE (TA) Z_ [ 6. Auxiliary Feedwater (Continued) a w f. Loss of Power (Motor See Item 8. below for all Loss of Power Trip [ Driven Pumps Only) Setpoints and Allowable Values.
- 7. Automatic 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 5.0 1.21 2.0 -> 11% ~> 9.1%
Coincident Witn: Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and Allowable Values. w
- 8. Loss of Power
- a. 4.16 AV ESF Bus Undervoltage N.A. N.A. A.A. > 3107 volts > 2979 volts (Loss of Voltage) Gith a < 1.75 sith a < 1.93 second Ilse second time delay. delay.
4;>- b. 4.16 kV ESF Bus Undervoltage N.A. N.A. N.A. > 3835 volts > 3786 volts 5 (Tolerable Degraded Voltage Gith a < 35 sith a < 39 A Coincident with SI) second time second Ilme 5 delay. delay. b c. 4.15 kV ESF Bus Undervoltage N.A. N.A. N.A. > 3835 volts > 3786 volts (Sustained Degraded Voltage) sith a < 50 sith a < 55 second time second time 2 delay, delay.
-' 5 m
o 1
H
,, TABLE 3.3-4 (Continued) 8 3l ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS Eh TOTAL SENSOR ERROR g- FUNCTICNAL UNIT ALLOWANCE (TA)
Z (S) TRIP SETPOINT ALLOWABLE VALUE E 9. Engineered Safety Features 5 Actuation System Interlocks
~ <) ,, a. Pressuriier Pressure, P-11 N.A. N.A. N.A. $ 1985 psig i 1993 psig
(( b. Low-tow 1,, , P-12 N.A. N.A. N.A. > 563*F > 560.1*F
- c. Reactor Trip, P-4 N.A. N.A. M.A. N.A. N.A.
- 10. Control Room Ventilation g* a. Nanual Initiation M.A. N.A. N.A. N.A. N.A.
i' b. Safety Injection See Item 1. above for all Safety Injection Trip
;" Setpoints and Allowable Values.
- c. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A.
and Actuation Relays
- d. Control Room Intake Air 3.7x10 5 2.2x10 5 1.6x10 5 <6.1x10 5 <7.8x10 5 Radioactivity - High pCi/cc pCi/cc pCi/cc pCi/cc pCi/cc gg e. Loss of Power See Item 8. above for all Loss of Power Trip Setpoints and gg Allowable Values.
9
-* 11. FHB HVAC 5
y' a. Manual Initiation N.A. N.A. N.A. N.A. N.A. 3 o r: 13 b
TABLE 3.3-4 (Continued) {- ENGINEERED SAFELY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS M TOTAL SENSOR ERROR FUNCTIOMAL UNIT ALLOWANCE (TA) Z (S) TRIP SETPOINT . ALLOWABLE VALUE [2
- 11. FIS HVAC (Continued)
[ z O b. Automatic Actuation N.A. N.A. N.A. N.A. M.A. Logic and Actuation [ Relays .
- c. Safety injection See Item 1. abeve for all Safety Injection Trip Setpoints and A)?owable Values.
d, Spent fuel Pool Exhuast 3.1x10 4 1.8x10 4 1.3x10 4 <5.0x10 4 <6.4x10 4 Radioactivity
- liigh pCi/cc pCi/cc pCi/cc pCi/cc pCi/cc M.
Y 3: 9 W 9- - Z tn 7- o b 3 6 '
bb, TABLE 3.3-4 (Continued) TABLE NOTATIONS
- Time constants utilized in the lead-lag controller for Steam Line Pressure-Low are 12 2 50 seconds and T2 < 5 seconds. CHANNEL CALIBRATION shall ensure that these time constants are adjusted to these values.
**The time constant utilized in the rate-lag controller for Steam Line Pressure-Negative Rate-High is greater than or equal to 50 seconds. CHANNEL CALIBRATION shall ensure that this time constant is adjusted to this value. [
#2.0% span for Steam Generator Level; 0.2% span for Reference Leg RTDs
##Until resolution of the Veritrak transmitter uncertainty issue, the trip ,
setpoint will be set at 1 1869 psig, with the allowable value at 1 1861 psig.
###This setpoint value may be increased up to the equivalent limits of Specification 3.11.2.1 in accordance with the methodology and parameters of the ODCM during containment purge or vent for pressure control, ALARA and i respirable air quality considerations for personnel entry.
SOUTH TEXAS - UNITS 1 & 2 3/4 3-36 AMENDMENT N05. AND
1= 1 TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATION SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS
- 1. Manual Initiation
- a. Safety Injection (ECCS) N.A.
- b. Containment Spray N.A.
- c. Phase "A" Isolation N.A.
- d. Phase "B" '. solation N.A.
- e. Contal'eent Ventilation Isolation N.A. .
- f. Steam Line Isolation N.A.
- g. Feedwater Isolation N.A.
- h. Auxiliary Feedwater N.A.
- i. Essential Cooling Water N.A.
- j. Reactor Containment Fan Coolers N.A.
- k. Control Room Ventilation N.A.
- 1. Reactor Trip N.A.
- m. Start Diesel Generator N.A.
- 2. Containment Pressure--High-1
- a. Safety Injection (ECCS) s 27(1)/12(5)
- 1) Reactor Trip < 2(3)
- 2) Feedwater Isolation 12(3)
- 3) Phase "A" Isolation 33(1)/23(2)
- 4) Containment Ventilation Isolation ~
23(1)/13(2) (18-inch lines)
$) Auxiliary Feed.<ater < 60
- 6) Essential Cooling Water 62(1)/52(2)
- 7) Reactor Containment Fan Coolers 38(1)/28(2)
, 8) Control Room Ventilation [72(1)/62(2)
- 9) Start Standby Diesel Generators 1 12 e
SOUTH TEXAS - UNITS 1 & 2 3/4 3-37 AMEN 0 MENT NOS. ANO
f 7 gg
FC ' 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 III/12(5)
- 1) Reactor Trip 1 2(3)
- 2) Feedwater Isolation < 12(3)
I 3) Phase "A" Isolation 33(l')/23(2) 1
- 4) Containment Ventilation Isolation N.A.
- 5) Auxiliary Feedwater < 60
- 6) Essential Cooling Water k62(1)/52(2)
- 7) Reactor Containment Fan Coolers < 38(1)/28(2) l
- 8) Control Room Ventilation [72(1)/62(2)
- 9) Start Standby Diesel Generators 1 12
] l 4. (This item number is not used.)
- 5. Compensated Steam Line Pressure--Low
- a. Safety Injection (ECCS) 1 22(4)/12(5)
)
- 1) Reactor Trip 5 2(3)
- 2) Feedwater Isolation < 12(3)
]
- 3) Phase "A" Isolation 33(1)/23(2)
! 4) Containment Ventilation Isolation N.A.
- . 5) Auxiliary Feedwater 1 60
! 6) Essential Cooling Water < 62(1)/52(2) ! 7) Reactor Containment Fan Coolers [38(1)/28(2) i i i l 1 l l ! SOUTH TEXA5 - UNITS 1 & 2 3/4 3-36 AMENDMENT N05. AND l I :
- UR
Ft ' TABLE 3.3-5 (Continued) ENGINEERED SAFETY FEATURES RESPONSE TIMES . INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS
- 5. Compensated Steam Line Pressure--Low (Continued)
- 8) Control Room Ventilation < 72I1)/62(2) .
- 9) Start Diesel Generators 7 12 i I
- b. Steam Line Isolation [8I3)
- 6. Containment Pressure--High-3
}
- a. Containment Spray 1 30I1)/20(2)
- b. Phase "B" Isolation i 28I1)/18I2) t
. 7. Containment Pressure--High . 1
- Steam Line Isolation <7 I3}
s
- 8. Steam Line Pressure - Negative Rate--High l.
Steam Line Isolation N.A.
- 9. Steam Generator Water Level--High-High l a. Turbine Trip 1 3(3)
! b. Feedwater Isolation 1 12(3) (
- 10. Steam Generator Water Level--Low-Low
- a. Motor-Driven Auxiliary j Feedwater Pumps 1 60
- b. Turbine Oriven Auxiliary Feedwater Pump i 60 i.
i i i
- 11. RWST Level--Low-Lnw Coincident with Safets Injection }
Automatic Switchover to Containment Sump I2) } 1 32
- 12. Loss of Power i
j a. 4.16 kV E.SF Bus Undervoltage 1 12 (Loss of Voltage) i b. 4.16 LV ESF Bus Undervoltage i 49 l (Tolerable Degraded Voltage Coincident with Safety Injection) l I 1 SOUTH TEXAS - UNITS 1 l. 2 3/4 3-39 AMINDMENT NOS. AND
' ' '::5 ~.
Fi TABLE 3.3-5 (Contint,ed) ENGINEERED SAFETY FEATURES RESPONSE TIMES 2 INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONOS
- 12. Loss of Power (Continued)
- c. 4.16 kV ESF Bus Undervoltage < 65
~
(Sustained Degraded Voltage)
- 13. RCB Purge Radioactivity-High j a. Containment Ventilation Isolation (48-inchlines) , 1 73(2'1
- b. Containment Ventilation Isolation (18-inchlines) $ 23(2I
- 14. (This item number is not used.)
- 15. (This item number is not used.)
- 16. Low Coincident with Reactor Trip T,yg Feedwater Isolation N.A.
i
- 17. Control Room Intake Air Radioactivity - High Control Room Ventilation 5 78(2)
! 18. Spent Fuel Pool Exhaust Radioactivity - High FHB HVAC Emergency Startup 1 42(2)
- 19. Charging Header Pressure - Low N.A.
4 i i i I i l l AND SOUTH TEXAS - UNITS 1 & 2 3/4 3-40 AMENDMEN1 N05. l
- ' ; la;;
i _ - - _ _ _ . . . ,__ _ _ ._, ,,_ _ ._---- -- m ,, _ - _ . ._
F
.y TABLE 3.3-5(Continued)
TABLE NOTATIONS f (1) Diesel generator starting and sequence loading delays int sded. (2) Diesel generator starting delay not included, sequence loading delay is included. Offsite power avaiTaSle. (3) Not depennt upon diesel generator starting or sequence loading delays. (4 ', Diesel ge... ator starting and sequence loading delay included.
. Low Head Safet) Injection pumps not included.
(5) Diesel generator starting delays not included, sequence loading delay is included. Low Head Safety InfecIion pumps not included. 4" . t SOUTH TEXAS - UNITS 1 & 2 3/4 3-41 AMENDMENT N05. AN3 l
4
~
y, TABLE 4.3-2 2
;p ENGINEERED SAFETY IEATURES ACTUATION SYSTEM INSTRUNENTATION
_4 St RVEILLANCE REQUIREMENIS R
$; DIGITAL OR TRIP
, ANALOG ACTUATING MODES c- CHANNEL DL'! ICE MASTER SLAVE FO*4 WPICH 5
CHANNEL CHANNEL CHANNI L OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE FUNCTIONAL UNI', CHECK CALIBIATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED I 1. Safety Injec tion (Reactor
,, Trip, feedwater Isolation, Control Room Emergency Ventilation, Start Standby Diesel Genera. ors, Reactor Containment fan Coolers, and Essential Joolir; Water)
T#-
- a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1, 2, 3, 4 Y
*; b. Automa'.ic Actuation N.A. N.A. N.A. N.A. M(1) N.A. N.A. 1, 2, 3, 4 Logic
- c. Actuation Relays N.A. N.A. N.A. M.A. N.A. M(5) Q(4,5) 1, 2, 3, 4
- d. Containment Pressure- S R M N.A. N.A. N.A. M.A. 1, 2, 3, 4 High-1 4
~~
- e. Pressurizer Pressure- S R M N.A. M.A. M.A. N.A. 1, 2, 3
'gg Low U e..,. pens *-d Steam Line S R M N.A. N.A. N.A. N.A. 1, 2, 3 pg Pressure ..e w
o 7
*O
- 't}
t, -
TABLE 4.3-2 (Continued) S ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION
] $URVEILLANCE REQUIRL%NIS m
7 DIGITAL OR TRIP ANALOG ACTUATING MODES CHANNEL DEVICE MASTER SLAVE FOR WHICH E CHANNEL CHANNEL CHAWNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE O FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED [ 2. Containment Spray
- 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) N.A. M.A. 1, 2, 3, 4 Logic
- c. Actuation Relays N.A. N.A. N.A. N.A. M.A. M(6) Q 1,2,3,4 w d. Containment Pressure- S R M N.A. N.A. M.A. M.A. 1, 2, 3 E High-3 w
- 3. Containment Isolation
- a. Phase "A" Isolation
- 1) Manual Initiation N.A. N.A. N.A. R N.A. N.A. M.A 1, 2, 3, 4
#, 2) Automatic Actuation M.A. N.A. N.A. N.A. M(1) M.A. M.A. 1,2,3,4 y logic b5 3) Actuation Relays M.A. N.A. N.A. N.A. N.A. M(6) Q(4) 1, 2, 3, 4 h 4) Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.
- b. Containment Ventilation Isolation b 1) Automatic Actuation N.A. N.A. N.A. N.A. M(1) N.A. N.A. 1, 2, 3, 4
_ Logic
- 2) Actuation Relays N.A. N.A. N.A. N.A. N.A. M(6) Q 1, 2, 3, 4
TABLE 4.3-2 (Continued) v.
- r ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRthENTATION 5bRVEILLANCE RtQUIRLMLMIS M DIGITAL OR TRIP 5
ANALOG ACTUATING MODES MASTER SLAVE FOR ietICH-CHANNEL DEVICE CHANNEL DIANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE E FUNCTIONAL UNIT CilECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED
- 3. Containment Isolation (Continued)
- 3) Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.
[ eo 4) RC8 Purge Radioactivity-High 5 R M N.A. M.A. N A. M.A. 1,2,3,4,5*,6* ,
- 5) Containment Spray - See Item 2. above for Containment SP ay manual initiation Surveillance Manual Initiation Requirements-w 6) Phase "A" Isolation- See Item 3.a.. above for Phase 5A" Isolation manual initiation D Manual Initiation Surveillance Requirements. ,
- c. Phase "B" Isolation N.A. N.A. N.A_ N.A. M(1) N.A. N.A. 1,2,3,4
- 1) Automatic Actuation Logic .
- 2) Actuation Relays N.A. M.A. N.A. N.A. M.A. M(6) Q 1,2,3,4 S R M N.A. M.A. N.A. N.A. 1,2,3
- 3) Containment Pressure--High-3 M 4) Cor.tainment Spray- See Item 2. above for Containment Spray manual initiation Surveillance Requirements.
'J' a
Manual Initiation 6, d. RCP Seal Injection . 5 Isolation N.A. 1,2,2,4 b 1) Automatic Actuation N.A N.A. N.A. M.A. Q Q Logic and Actuation Relays
- 2) Charging Header 5 R N.A. M.A. N.A. N.A.- M.A. 1,2,3,4 Pressure - Low Coincident with See Item 3.a. above for Phase "A" surveillance requirements.
Phase "A" Isolation M.
j m TABLE 4.3-2 (Continued)
! 8 g ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION y SURVEILLANCE REQUIRtrftNTS m
h DIGITAL OR TRIP ANALOG ACTUATING MODES i CHANNEL DEVICE MASTER SLAVE FOR WilCH c 5 CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE g FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED j [ 4. Steam Line Isolation
- a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1, 2, 3 j
- b. Automatic Actuation N.A. N.A N.A N.A. M(1) M(6) Q 1,2,3 Logic and Actuation Relays N.A, M.A. N.A. 3
$ c. Steam Line Pressure-Negative Rate-High S R M N.A.
- d. Containment Pressure - N.A. N.A. N.A. M.A. 1, 2, 3
- $ High-2 S R M
- e. Compensated Steam Line S R M N. A. - N.A. M.A. N.A. 1, 2, 3 Pressure-Law
- 5. Turbine ! rip and Feedwater Isolat'on
- a. Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(6) Q(4) 1, 2, 3 5 Logic and Actuation
, ;c,
$ Relays 5 b. Steam Gem rator Water S R M N.A. M.A. N.A. M.A. 1, 2, 3 P Level-High-High (P-I4)
>- c. (This functional unit number is not used.)
- d. (This functional unit number is not used.)
- e. Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.
1
; y s
TABLE 4.3-2 (Continued) E ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUNENTATION S x SURVEILLANCE REQUIREMENTS
-4 DIGITAL OR TRIP R ANALOG ACTUATING MODES E CilANNEL DEVICE MASTER SLAVE FOR WHICH
. CilANNEL CilANNEL CilANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE ruNCTIONAL UNIT CllECK CALIBRATION TEST _ TEST LOGIC TEST TEST TEST IS REQUIRED g] f. T avg
-Low Coincident 5 R M N.A. N.A. N.A. N.A. 1, 2, 3
~ with Reactor Trip (P-4) 9 (feedwater Isolation m Only)
- 6. Auxiliary feedwater
- a. Manual Initiation N.A. N.A. N.A. R N.A. M.A. N.A. 1,2,3
- b. Automatic Actuation N.A. N.A. N.A. N.A. M(1) N.A. N. 4. 1,2,3 Logic
{ 1,2,3
- c. Actuation Relays N.A. N.A. N.A. N.A. N.A. M(6) Q y
$ d. Steam Generator Water 5 R M N.A. M.A. N.A N.A 1,2,3 Level--Low-Low
- e. Safety Injection See Ites 1. above for all Safety Injection Surveillance Requirements.
- f. Loss of Power See Item 8. below for all Loss of Power Surveillance Requirements.
- 7. Automatic Switchover to y Containment Sump H;
5 a. Automatic Actuation N.A. N.A. N.A. N.A. M(6) M(6) Q 1, 2, 3, 4 i?, Logic and Actuation 5 Relays 5 b. RWST Level--Low-Low 5 R M N.A. N.A. N.A. N.A 1, 2, 3, 4 Coincident With: g Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements. t2 g -
- 8. Loss of Power
- a. 4.16 kV ESF Bus N.A. R N.A M N.A. N.A. N.A. 1,2,3,4 Undervoltage (Loss i of Voltage) M
, TABLE 4.3-2 (Continued) o 5 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION
[ SURVEILLANCE RLQUIREMfNIS m 5 DIGITAL Oo TDTD ANALOG ACTUATI.1G MODES CHANNEL DEVICE MASTER SLAVE FOR %dHICH E CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATICNAL ACTUATION RELAY RELAY SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST _ LOGIC TEST TEST TEST IS REQUIRED
]
[ 8. Loss of Power (Continued)-
- b. 4.16 kV EST Bus N.A. R N.A. M N.A. M.A. N.A. 1, 2, 3, 4 Undervoltage (Tolerable Degraded Voltage Coincident with SI) w c. 4.16 kV EST Bus N.A. R N.A. M M.A. N.A. N.A. 1, 2, 3, 4 1 Undervoltage (Sustained w Degraded Voltage) s.
! 9. Engineered Safety features Actuation System Interlocks .
- a. Pressurizer N.A. R M N.A. N.A. M.A. N.A. 1, 2, 3 Pressure, P-11 4
i fj b. Low-Low T,, , P-12 N.A. R M M.A. N.A. M.A. N.A. 1, 2, 3 O j i[i c. Reactor Trip, P-4 N.A. N.A N.A. R h.A. M.A. M.A. 1,2,3 a 7 10. Control Room Ventilation
- a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. All R
o Os U
-. -__ ~ - _ - . = _- - ..
,,, TABLE 4.3-2 (Continued)
E
;! ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION
_4 SURVEILLANCE REQUIREMENIS O
?! DIGITAL TRIP
, ANALOG ACTUATING MODES c CHANNEL DEVICE MASTER SLAVE FOR WHICH
$ CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEI*. LANCE gl IUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LDGIC TEST TEST TEST IS REQUIRED
[ 10. Control Room Ventilation (Continued) m
- h. Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.
- c. Automatic Actuation N.A. N.A. N.A. N. A. M(6) N.A. N.A. All Logic and Actuation Relays R
3-
- d. Control Room Intake Air S R M N.A. N.A. N.A. M.A. All 4' Radioactivity-High
- e. Loss of Power See Items 8. above for all Loss of Power Surveillance Requirements.
- 11. THB HVAC
- a. Manual Initiation N.A. N.A. N.A. R N.A. M.A. N.A. 1,2,3,4 or with y irradiated gg fuel in the
?g spent fuel g; pool 2; b. Automatic Actuation N.A. N.A. N.A. N.A. M(6) N.A. M.A. 1, 2, 3, 4, y' Logic and Actuation or with i
Relays irradiated fuel in the y spent fuel o pooj
?
ce
! *J
.b.I.
, TABLE 4.3-2 (Continued) 8 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION M SURVEILLANCE REQUIREMEN15 M DIGITAL OR TRIP 5
ANALOG ACTUATING N00ES CHANNEL DEVICE MASTER SLAVE FOR WHICH CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE y filNCTIONAL UNIT CHECK CAilBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED d II. FHB HVAC (Continued) [ c. Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.
- d. Spent Fuel Pool S R M N.A. N.A. N.A. N.A. With Exhaust Radio- irradiated activity-High fuel in spent fuel pool.
w D TABLE NOTATION (1) Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS. (7) (This notation number is not'used.) (3) (This notation number is not used.) (4) Except relays K807. K814. K829 (Train 8 only), K831, K845, K852 and K854 (Trains 8 and C only) 2, which shall be tested at least once per 18 months during refueling and during each COLD SHUTDOWN ij, exceeding 24 hours unless they have been tested within the previous 92 days. o J. (S) Except relay K815 which shall be tested at indicated interval only when reactor coolant pressure 5 is above 700 psig. b
^
(6) Each actuation train shall be tested at least every 92 days on a ST1GGERED TEST BASIS. Testing of each actuation train shall include master relay testing of both logfc trains. If an ESFAS instru-mentation channel is inoperable due to failure of the Actuation Logsc Test and/or Master Relay Test. 2 increase the surveillance frequency such that each train is tested at least every 62 days on a 5 STAGGERED TEST BASIS unless the failure can be determined by perfornance of aa engineering evaluation g to be a single random failure. . *During CORE ALTERATIONS or mcvement of irradiated fuel within containment.
F. INSTRUMENTATION 3/4.3.3 MONITORING INSTRUMENTATION RADIATION MONITORING FOR PLANT OPERATIONS LIMITING CONDITION FOR OPERATION 3.3.3.1 The radiation monitoring instru. mentation 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 hours or declare the channel inoperable.
- 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 Specification 3.0.3 are not applicable.
SUovEILL MOE REOUIREMENTS 4.3.3.1 Each radiation monitoring instrumentation channel for plant operations shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CAllBRATION and DIGITAL CHANNEL OPERATIONAL TEST for the MODES and at the frequencies shown in Table 4.3-3. SOUTH TEXAS - UNITS 1 & 2 3/4 3-50 AMENDMENT N05. AN? IG ; : ,
,, TABLE 3.3-6 a
{f RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS
=
5 MINIMUM CHANNELS CHANNELS APPLICABLE ALARM / TRIP IUNCTIONAL UNIT 10 TRIP / ALARM OPERABLE MODES SETPOINT ACTION [j 1. Containment , (( a. Containment Atmosphere N.A 3 All N.A. 31
,, Radioactivity-liigh
- b. RCS Leakage Detection
- 1) Particulate Radioactivity N.A. I 1, 2, 3, 4 N.A. 34
- 2) Gascous Radioactivity N.A. I 1, 2, 3, 4 N.A. 34 T.
4 S i6 5 i5 (3 r E o M N m
[Il2 ', TABLE 3.3-6 (Continued) ACTION STATEMENTS ACTION 31 - With less than the Minimum Channels OPERABLE requirement, l opt ration may continue for up to 30 days provided grab samples ; of the containment atmosphere are obtained and analyzed at least once per 24 hours. . ACTION 32 - (Not Used) ; i ACTION 33 - (Not Used) { ACTION 34 - Must satisfy the ACTION requirement for Specification 3.4.6.1. i e i 1 l i 1 I a ) i l i
- I i,
SOUTH TEXAS - UNITS 1 & 2 3/4 3-52 AMENDMENT NDS. AND
!.~i 1 1 li::
y, TABLE 4.3-3 o
$h RADIATION MONITORING INSTRUMENTATION FOR PLANT
_, OPERAll0NS SURVElltANCL REQUIREMINIS 7 3; D.GITAL
, C'lANNEL MODES FOR WHICH c CHANNEL CHANNEL OPERATIONAL SURVEILLANCE y; FUNCTIONAL UNIT CHECK CALIBRATION , TEST IS REQUIRED C
,, 1. Containment e-s, a. Containment Atmosphere Radioactivity-High 5 R H All
- b. RCS Leakage Detection
- 1) Particulate Radio- S R M 1, 2, 3, 4 q>, activity 4
g> 2) Gaseous Radioactivity S R M 1, 2, 3, 4 E _4 . t) v.
\
't) k7
; FL-INSTRUMfNtAt10N MOVABLE INCORE DETECTORS ,
i LIMITING CONDITION FOR OPERATION ! 3.3.3.2 The Movable Incore Detection System shall be OPERABLE with: I
- a. At least 75% of the detector thimbles, i
- b. A minimum of two detector thimbles per core quadrant, and,
- c. Sufficient movable detectors, drive, and readout equipunt to map these thimbles.
2 APPLICABILITY: When the Movable Incore Detection System is used for:
- a. Recalibration of the Excore Neutron Flux Detection System, or
- b. Monttsring the QUADRANT POWER TILT RATIO, or N
- c. Measurement of F 3q, Fq(2) and F,y, q
ACTION: ! With the Movable Incore Detection System inoperable, do not use the system for the above applicable monitoring or calibration functions. The provisions of Specification 3.0.3 sre not applicable. SURVEILLANCE RE0VIREMENis
- 4.3.3.2 The Movable Incore Detection System shall be demonstrated OPERABLE at least once per 24 hours 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 F 3g, F (Z) and F,y, 9
I I l 1 1 t ) l SOUTH TEAAS - UNITS 1 & 2 3/4 3-54 AMENDMENT h:5. AC l'DJ7rg:3 i l- _ -
t F1 i , t INSTRUMENTATION ! i SEISMIC INSTRUMENTATION , LIMITING CONDITION FOR OPERATION 3.3.3.3 The seismic monitoring instrumentation shown in Table 3.3-7* shall be ! OPERABLE. ' } i APPLICABILITY: At all times. t ACTION: [
- a. With one or more of the above required seismic monitoring instruments inoperable for more than 30 days, prepare and submit a Special >
Report to the Comnission pursuant to Specification 6.9.2 within the t next 10 days outlining the cause of the malfunction and the plans r 3 for restoring the instrument (s) to OPERABLE status. This ACTION is l applicable to both units simultaneously. , r
- b. The provisions of Specification 3.0.3 are not applicable, j SURVEILLANCE REOUIRNENTS i 3
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 OP: RATIONAL TEST at the frequencies shown in , Table 4.3-4. l 4.3.3.3.2 Each of the above required seismic monitoring instruments actuated f during a seismie event shall be restored to OPERABLE status within 24 hours and l a CHANNEL CAllBRATION performed within 10 days following the seismic event. [ Data shall be retrieved from actuated instruments and analyzed to determine the , magnitude of the vibratory ground motion. A Special Report shall be prepared t and submitted to the Comission pursuant to Specifiestion 6.9.2 within 14 days describing the magnitude, f equency spectrum, and resultant effect upon facility , features important to safety. , j
. I r
[ I "The instrumentation may be shared with additional units at a common site provided seismic instrueentation and corresponding Technical Specificatient meet the recorrencations of Regulatory Guide 1.12, Rev hion 1, April 1974. i SOUTH TEAAS - UNITS 1 & 2 3/4 3-55 AMENDMENT N05. AND E I N 1 7 *,"f3 l t i l f i
FCl 4 TABLE 3.3-7 SEISMIC MONITORING INSTRUMENTATION l ? MINIMUM 3 INSTRUMENTS AND SENSOR LOCATIONS MEASUREMENT INSTRUMENTS (Unit 1 only) ! _ RANGE OPERABLE !
- 1. Triaxial Time History Accelerometers *** I
- a. Free Field 13g 1 I
- b. Contaiteent Bldg. Foundation 13g 1 {
(Tendon Gallery E1. -36'9") l
- c. Outside Face Containment Shell 23g 1 (Reactor Containment Building E1. 68'0") {
- d. Steam Generator Upper Lateral Support 13g 1 :
, (ReactorContainmentBuildingE1.66'7\") !
- e. Fuel Handling Building Foundation 13g 1 !
(Fuel Handling Building E1 -29'0") l
- f. Mechanical Electrical Auxiliary Building 13g 1 !
2 (Mechanical Electrical Auxiliary ! Building E1. 35'0") I
' l
- 2. Triaxial Peak Accelerographs
- a. Spent Fuel Pool Heat Exchanger t3g 1
{ (Inlet Line Fuel Handling Building ( } H E1. 64'55") [ b, Reactor Vessel 23g 1 (Reactor Containeent Building E1. 6S'0") l l r i c. Cold Leg of RC Piping 13g 1 l ) (ReactorContainmentBuilding i i E1. 34'3") i r
- 3. Self-Contained Triaxial Accelerograph 13g 1
- (At Reactor Containment Building I l
Foundation Tendon Gallery E1. -36'9") i ) 4. Triamial Seismic S. itch' **
- 0.C3 to 3g 1 3
- 5. Triaxial Seismic Trigger * ** H 0.003 to 0.3g 1 l 6, Response Spectrum Analyzer * ** 1 to 32 Hz 1 f
- 7. Magnetic Tape Recorders **
0.1 to 33 Hz 6 [ f
- 8. Playback Syste ** N.A. 1
- *With reactor control room indication and alarm in Unit 1 (Alarm only in i
Unit 2) { 4
**At seismic monitoring panel in Control Room, Unit 1
*** Accelerometer data is gathered and analyzed by the Response Spectrum Analyzer (Item 6).
1 #Triasial seistic switch is set at the OBE acceleration level of 0.05g i horizontal and 0.033g vertical. i j H Triaxial seismic trigger is set at 0.02g all ases. ( i SOUTH TEXAS - UNITS 1 & 2 3/4 3-56 AMENDu.ENT N05. AND '
. ,g , , (
4 U.! ! l f
f9 TABLE 4.3-4
' SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS ANALOG CHANNEL INSTRUMENTS AND SENSOR LOCATIONS CHANNEL CHANNEL OPERATIONAL (Unit 1 only) CHECK CALIBRATION TEST
- 1. Triaxial Time-History Accelerometers ***
SA
- a. Free Field M R
- b. Containment Bldg. Foundation M R SA l (Tendon Gallery El. -36'9")
M R SA l
- c. Outside Face Containment Shell (Reactor Containment Building El. 68'0") i
- d. Steam Generator Upper Lateral Support M R SA l (Reactor Containment Building El. 66'71") s j
1 e. Fuel Handling Building Foundation M R SA I (Fuel Handling Building El. -29'0") -
- f. Mechanical Electrical Auxiliary M R SA l i Building (Mechanical Electrica'l l
Auxiliary Building El 35'0") [
- 2. Triaxial Peak Accelerographs
- a. Spent Fuel Pool Heat Exchanger N.A. R N.A. t i
(Inlet Line Fuel Handling Building El. 64'5\") N.A. l
- b. Reactor Vessel N.A. R (Reactor Containment Building i El. 68'0")
- c. Cold Leg of RC Piping N.A. R N.A. !
(ReactorContainmentBuilding f El. 34'3")
- 3. Self-Contained Triaxial A;celerograph M R SA (At ReactorTendon Foundation Containment Gallery BuH El. ding 36 '9")
M R SA
- 4. Triaxial Seismic Switch * **
- 5. Triaxial Seismic Trigger * ** M R SA
- 6. Response Spectru.n Analyzer * ** M R SA M R SA
- 7. Magnetic Tape Recorders **
M R N.A.
- 8. Playback Syster**
*With reactor control room indication and alarm in Unit 1 (Alarm only in Unit 2)
**At seisric monitoring panel in Control Room, Unit 1
*** Accelerometer data is gathered and analyzed by the Response Spectru-Analyzer (Item 6).
3/4 3-57 AMENFENT N]S. AND SOUTH TEXAS - UNITS 1 & 2 t;5' 1 1W
FD INSTRUMENTATION METEOROLO31 CAL INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.4 ine eeteoroiogical monitoring instrumentation channels shown in Table 3.3 8 shall be OPERABLE. APPLICABILITY: At all times. ACTION:
- a. With 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. This ACTION is applicable to both units simultaneously.
- b. The provisions of Specification 3.0.3 are not applicable.
SUAVEILLANCE RE0VIDEME!il! 4.3.3.4 Each of the above meteorological monitoring instrumentation channels shall be det.onstrated OPERABLE by the performance of the CHANNEL CHECK and CHANNEL CALIBRATION at the frequencies shown in Table 4.3 5. t I l I i l SOUTH TEXAS - UNITS 1 & 2 3/4 3-59 AMENDMENT h05. AND i-
;; g 1
FD TABLE 3.3 8 METEOROLOGICAL MONITORING INSTRUMENTATION MINIMUM INSTRUMENT LOCATION OPERABLE
- 1. Wind Speed
- a. Primary meteorological Nominal Elev. 10m 1 tower and backup mete-orological toiser
- b. Primary mettarological Nominal Elev. 60m 1 tower
- 2. Wind Direction
- a. Primary meteorological Nominal Elev 10m 1 tower and backup mete-orological tower
- b. Primary meteorological Nominal Elev '60m 1 tower
- 3. Air Temperature - AT
- a. Primary meteorological Nominal Elev. 10m 1 tower f
- b. Primary meteorological Nominal Elev. 60m 1 tower SOUTH TEXA5 - UNIT 5 1 & 2 3/4 3-59 AMEN 0 MENT N05. AN: I. . . ; ; ; ,,
i
U TA4LE 4.3-5 ! METEOROLOGICAL MONITORING INSTRUMENTATION ! SURVEILLANCE REQUIREMENTS [ t CHANNEL CHANNEL l
!NSTRUMENT CHECK CAL!BRATION
- 1. Wind Speed i
- a. Nominal Elev. IDS 0 , 5A !
- b. Nominal Elev. 60m D SA I
i
- 2. Wind Direction
[
- a. Nominal Elev. 10m D SA f
- b. Nominal Elev 60m D SA
- 3. Air Temperature - AT ,
i
- a. Nominal Elev. 10m D SA l t
- b. Nominal Elev. 60m D SA l t
i I r i i I t i I 4 SOUTH TEXA5 - UNITS 1 & 2 3/4 3 60 AMENDMENT N35. AND .
"*,,1 : %;
FF INSTRUMENTATION REMOTE SHUT 00WN SYSTEM 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, transfer switches, power or control circuits 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 SHUTOOWN within the next 12 hours,
- b. With the number of OPERABLE remote shutdown monitoring channels, transfer switches, power or control circuits less than the Total Number of Channels as required by Table 3.3 il, within 60 days restore the inoperable channel (s) to OPERABLE status or, pursuant to Specification 6.9.2, submit a Special Report that defines the corrective action to be taken.
- c. The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REOUIREu!NTS 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, 4.3.3.5.2 Each Remote Shutdown System transfer switch, power and control circuit including the actuated components, shall be demonstrated OPERABLE at least once per 16 months. SOUTH TEXAS - UNITS 1 & 2 3/4 3-61 AMENDMENT N05. ANO Xig
m TABLE 3.3-9 8 - . y REMOTE SHUTD0hM SYSTEM h TOTAL NO. MINIftfM f/; READOUT OF CHAMIELS
. IN'.1RIMENT LOCATION CHANNELS OPERABLE h 1. Neutron Flux - Extended Range U
y a. Startup Rate - ASP *-QDPS** 2 2 o~ y b. Flux level ASP-QDPS 2 2
- 2. Reactor Trip Breaker ASP-QDPS 1/ trip breaker 1/ trip breaker Indication Reactor Trip Switchgear y 3. Reactor Coolant Temperature- -
*- Wide Range
, a. Hot leg ASP-QOPS 4-1/ loop 1/ loop - 3 loops #
- b. Cold leg ASP-QOPS 4-1/ loop 1/ loop - 3 loops #
- 4. Reactor Coolant Pressure- ASP-QOPS 3 2 l Wide Range / Extended Range 1
- 5. Pressurizer Water Level ASP-QOPS 4 2
]
Ig 6. Steae Line Pressure ASP-QOPS 4-1/ steam line 1/ steam line - iy 3 steam lines # g 7. Steam Generator Water Levei- ASP-QDPS 4-1/ steam 1/ steam generator - y Wide Range generator 3 steam generators # 1 3 o b" i
# M 1
1
% %1 . . - - . . - - . - . . - - . - . - , - - - - - - , , . - , - - - - - - - - - - - . - - - - - - . - - - - - - - - . - - . - - , . . - - , - . - - . ~ . . , - . . . _ - , -,- . - - -
, TABLE 3.3-9 (Continued)
REMOTE SHUTDOWN SYSTEM M TOTAL NO. MINIORM 5 READOUT OF CHANNELS 7 INSTRtiMfMT LOCATION CHANNELS OPERABLE 2 8. Auxiliary feedwater Flow Rate ASP-QDPS 4-1/ steam 1/ steam generator - g , generator 3 steam generators # [ 9. Auxiliary feedwater Storage ASP-0DPS 3 2 Tank Water Level
- 10. Core Exit Thermocouples ASP-QDPS ## 4 therm 0 Couples /
core quadrant TOTAL NO. MINIMUM w TRANSFER SWITCHES AND TRAr.SFER SWITCH CONTROLS OF CHANNELS A A550CIATfD CONTROLS LOCATIONS LOCATION _ CHANNELS OPERA 8LE
- 1. Steam Generator PORVs ZLP-653 (Train A) ASP 4 2#
ZLP-654 (Train B) ZLP-655 (Train C) . ASP (Train 0)
- 2. Reactor Head Vent Throttle ZLP-700 (Train A) ASP 2 1 Valves ZLP-701 (Train B) 7-J, 3. Reactor Head Vent Isolation ZLP-700 (Train A) ASP 2 pair 1 pair.
Q Valves ZLP-701 (Train B) 5 4 ATW Pumps and Valves ZLP-653 (Train A-AFW Pump) ASP 4 2# z ZLP-700 (Train A-AFW Valves)
*3
~
ILP-654 (Train B-AFW Pump) ZLP-701 (Train B-AFW Valves) ZLP-655 (Train C-AIW Pump) 2 ZLP-709 (Train C-AFW Valves) 5 ASP (Train D)
l
~
a w TABLE 3.3-9 (Continued) 8 y REMOTE SHUTDOWN SYSTEM h TOTAL NO MINIRM g TRANSFER SWITCHES AND TRANSFER SWITCH CONTROLS OF CHANNELS
, ASSOCIATED CONTROLS LOCATIONS LOCATION CHANNELS OPERA 8tE 5 5. Centrifugal Charging Pumps ZLP-653 (Train A) ASP 2 1 d ZLP-655 (Train C) -
[ 6. Boric Acid Transfer Pumps ZLP-653 (Train A) kSP 2 1 m ZLP-655 (Train C)
- 7. Pressurlier PORVs and Block ZLP-700 (Train A) ASP 2 1 Valves ZLP-701 (Train 8)
- 8. Accumulator Discharge ZLP-653 (Train A) ASP 3 3 R Isolation Valves and Power ZLP-654 (Train 8)
#- Lockouts ZLP-655 (Train C)
(48 h 9. Letdown Stop Valves ZLP-700 (Train A) ASP 2 1 ZLP-709 (Train C) l
- 10. CCW Pumps and lleat Exchanger ZLP-653 (Train A) ZLP-653 (Train A) 3 2 Outlet Valves ZLP-654 (Train 8) ZLP-654 (Train 8)
ZLP-655 (Train C) ZLP-655 (Train C) y '.1. "CW Pumps ZLP-653 (Train A) ZLP-653 (Train A) 3 2 5 ZLP-654 (Train 8) ZLP-654 (Train 8) I y ZLP-655 (Train C) ZLP-655 (Train C) E
- 12. EAB HVAC Fans ZLP-700 (Train A) ZLP-700 (Train A) 3 2
,"5 ZLP-653 (Train A- ZLP-653 (Train A-P Battery Room and Battery Room Electrical Penetration and Electrical Spac-Fans) Penetration 3: Space Fans) o 7
ZLP-701 (Train 8) ZLP-701 (Train 8) ,
; ZLP-654 (Train 8- ZLP-654 (Train 8-
_ Battery Poom and Battery Room Electrical Penetration and Electrical
- . . Space Fans) Penetration
*: Space Fans)
TABLE 3.3-9 (Continued) y REMOTE SHUTDOWN SYSTEM M TOTAL E30. MINIPRM 5 TRANSFER SWITCHES AND TRANSFER SWITCH CONTROLS OF CHAISIELS ASSOCIATED CONTROLS LOCATIONS LOCATION CHANNELS OPERA 8LE E O 12. [AB HVAC fans ZLP-709 (Train C) ZLP-709 (Train C) (Continued) ZLP-655 (Train C- ZLP-655 (Train C-
. Battery Room and Battery Roon
" Electrical Penetration and Electrical Space Fans) Penetration Space Fans)
- 13. Reactor Containment fan ZLP-700 (Train A) ZLP-700 (Train A) 6 3 i w Coolers ZLP-701 (Train B) ZLP-701 (Train 8) l D ZLP-709 (Train C) ZLP-709 (Train C) i w h
- ASP - Auxiliary shutdown Panel
**00P5 - Qualified Display Processing System
! #Must be in the same OPERABLE RCS loop / secondary loop. - i ##A total of 50 thermocouples are provided with 25 thermocouples on each of two trains. Quadrants 8 and ! D have 6 themocouples per train each. Quadrants A and C each have 6 thermocouples on one train and ! 7 thermocouples on the other train. Tha provisions of ACTION b. are not applicable as long as each j j quadrant has 4 thermocouples per train CPERABLE. ! 5 i 5, i 5 . 5 !l , :n-e 1 - 1 u__ _ _ _ _ _ _ _ ___ _. __ -. -. __ _ . - - .. _ -_ . ._ .
, TA8tE 4.3-6 8 REMOTE SWJTtems MONIIORING I;45TRUMENTATION 2 50RVEILLANCE REQUIRfMtNIS h CHANNEL Ct;ECK CHAleIEL 3 INSTRUMENT CALIBRATION h 1. Neutron Flux - Extended Range 5
g a. Startup Rate M R
~,. b. Flux Level M R
~
- 2. Reactor Trip Breaker Indication % N.A.
- 3. Reactor Coolant Temperature-Wide Rany
- a. Hot Leg M R R.
s Y b. Cold leg M R E
- 4. Reactor Coolant Pressure-Wide Range / Extended Range M R
- 5. Pressurizer Water Level M R
- 6. Steam Line Pressure M R M
f'V 7. Steam Generator Water Level-Wide Range M R
@ 8. Auxiliary feedwater T10w Rate M R .
,$ 9. Auxiliary feedwater Storage
= Tank Water Level M R
- 10. Core Exit Thermocouples M R D
u l 1 a a _ . . ... _ ., _ __. _ _ - ..~. . _ _ . . - . _ , _ - _ . _ . - _ -- _--... _ ___ _ __ __ _.._. . -- ._-
FC INSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATION Lip 1 TING _CONDITIONFOROPERATION _ , , , 3.3.3.6 The accident monitoring instrumentation channels shown in Table 3.3-10 shall be OPERABLE. APPLICABILITY: H0 DES 1, 2, and 3. ACTION:
- a. As shown in Table 3.3-10.
- b. The provisions of Specification 3.0.4 are not applicable.
SLTYEILLMCE REQ'llREMENTS 4.3.3.6 Each accident monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION at the frequencies shown in Table 4.3-7. l l SCUTH TE AAS - UNITS 1 & 2 3/4 3-67 AuEN;v.ENT NOS. AN?
l l l 1 . l
, TABLE 3.3-10 g . .
p ACCIDENT MONITORING INSTRUNENTATION ! h TOTAL MINIMUN 3; NO. OF CHANNELS
, INSTRUM[NT CHANNELS OPERABLE ACTION 5 1. Containment Pressure 4 1 38 0
y 2. Reactor Coolant Outlet Temperature -
, T,gy (Wide Range) 1/ loop 1/ loop 35 u
- 3. Reactor Coolant In;et Temperature -
Tpt n (Wide Range) 1/ loop 1/ loop 35
- 4. -
r* Coolant Pressure - Wide Range 3 1 37 Lended Rame
- 5. Pressurizer W.ter Level 4 1 38
- 6. Steam tir.e Pressure 4/ steam generator 1/ steam generator 38
- 7. Steam Generator Water Level -
Narrow Range 4/ steam generator 1/ steam generator 38 l l 8. Steam Generator Water Level - l Wide Range 1/ steam generator 1/ steam generator 35 ij 9. Refueling Water Storage lank Water
<g Level 3 1 37 m -
5 10. Auxiliary Feedwater Storaga Tank y Water tevel 3 1 37 m
- 11. Auxilia.y Feedwater Flow 1/ steam generator 1/ steam generator 35
> 12. Reactor Coolant System Subcooling 6 Margin Monitor 2 1 36 F
E
, w- .
9 TABLE 3.3-10 (Continued) o ' ACCIDENT MGMITORING INSTRUMENTATION , {5 M TOTAL MINIKIM 5 NO. OF CHANN'.LS
'[ INSTRUMENT CHANNELS LPERABLE ACTION EE 13. Containment Water Level (Narrow Range) 2 1 36 .
(( ,
- 14. Containment Water Level (Wide Range) 3 1 37 l[
Core Exit Thermocouples ** 4 thermocouples/ core 42 15. quadrant
- 16. Steam Line Radiation Monitor 1/ steam line 1/ steam line 40 w 17. Containment - High Range Radiation 3: Monitor 2 1 33
- 18. Reactor Vessel Water Level (RVWL) 2* 1* 41
- 19. Neutron Flux (Extended Range) 2 1 35
- 20. Containment Hydrogen Concentration 2 1 3:
- 21. Containment Presscre (Extended Range) 2 1 35 E
m 22. Steam Generator Blowdown R-d'ation is hanitor 1/ blowdown line 1/ blowdown line 40 M 25 23. Neutron Flux - Startup Rate z (Extended Range) 2 1 36 0
*A channel is eight sensors in a probe. A channel is OPERABLE if four or more sensors, one or more in the upper section and three or more in the lower section, are OPERABLE.
3? **A total of 50 thermocouples are provided with 25 thermocouples on each of two trains. Quadrants B and D have 6 thermocouples per train each. Quadrants A and C each have 6 thermocouples on one train and 7 thermocouples on the other train. No ACTION is required as long as each quadrant has 4 thermocoupies per train OPERABLE. k.
F. TABLE 3.3-10 (Continued) ACTION STATEMENTS ACTION 35 - With the number of OPERABLE channels less than the Minimum Chann21s Operable requirement, restore at least one Inoperable channel to OPERABLE status witnin 4s nours, or be in at least HOT SHUTDOWN within the next 1.1 hours. ACTION 36 - a. With the number of OPERABLE channels one less than the Total Number of Channels requirements, restore one inoperable channel to GPERABLE status within 7 days, or be in at least HOT SHUTDOWN within the next 12 hours.
- b. With the number of OPERABLE channels less than the Minimum
. Channels Operable requirements, restore at least one inoperable channel to OPERABLE status within 48 hours, or be in at least HOT SHUTOOWN within the next 12 hours.
ACTION 37 - a. With the number of OPERABLE channels one less than the Total o ' Number of Channels requirements, restore the inoperable channel < to OPERABLE status within 31 days, or be in at least HOT h.N SHUTDOWN within the next 12 hours. Se ' j
- b. With the number vf OPERABLE channels two lets than the Total Number of Channels requirement, restore at least one inoperable channel to OPERABLE status within 7 days, or be in at least HOT SHUTDOWN within the next 12 hours.
- c. With the number of OPERABLE channels less than the Minimum Channels Operable requirement, restore at least one inoperabir-channel to OPERABLE status within 48 hours or be in et least HOT SHUTDOWN within the next 12 hours.
ACTION 38 - a. With the number of OPERABLE channels one less than the Total Number of Channels requirements, restore the inoperable channel to OPERABLE status within 90 days, or be in at least HOT SHUTDOWN vithin the next 12 hours.
- b. With the number of OPERABLE channels two less than the Total Number of Channels requirements, estnre the inoperable channel to OPERABLE status withir. 31 days, or be in at least HOT SHUT 00WN within the next 12 hours,
- c. With the number of OPERABLE channels three less 9'an the Total Number of Channels requirement, restore at lea ' cne inoperable .
channel to OPERABLE status within 7 days, or be n at least HOT l SHUTDOWN within the next 12 hours. t With the number of OPERABLE channels less than ihr Folicum Channels Operable requirement, restore at leas' on.- inoperabia channel to OPERABLE status within 48 hourt or sa ' ot lear.' HOT SHUTD0nw within the next 12 hours. SOUTH TEXAS - UNITS 1 & 2 3/4 3-70 AMENCMENT NDS. AND l .. .5
F TABLE 3.3-10 (Continued) ACTION STATEMENTS (Contince(L ACTION 39 - a. With the number of OPERABLE channels one less than the Total Number of Channels requirements, restore one inoperable channel to OPERABLE status within 7 days, or be in at least ii0i SHUTDOWN within the next 12 hours.
- b. With the number of OPERABLE channels less than the Minimum Channels Operable requirements, restore at least one inoperable channel to OPERABLE status within 72 hours, or be in at least HOT SHUTDOWN within the next 12 hours.
ACTION 40 - With the number of OPERABLc channels less than the Minimum Channels Operable requirements, restore at least one inoperable channel to OPERABLE status within 72 hours, or be in at least HOT SHUTDOWN within the next 12 hours. ACTION 41 - a. With the number of OPERABLE channels one less than the Required Number of Channels, either restore the system to OPERABLE status within 7 days if repairs a.re feasible without shutcing down or prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status,
- b. With the number of 0F' G E Channels one less than the Minimum Channels GPERABLE in e 3.3-10, either restore the inoperable channel (s) to OPERABLE scatus within 48 hours if repairs are feasible without shutting down or:
- 1. Initiate an alternate method of monitoring the reactor vessel inventory;
- 2. Prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule fo,r restoring the system to OPERABLE status; and
- 3. Restore the system to OPERABLE status at the next scheduled refueling.
ACTION 42 - a. With the number of OPERABLE channels less than 4 thermocouples per quadrant per train, restore these thermocouples to OPERABLE status within 31 days, or be in at least HOT SHUTDOWN within the next 12 hours,
- b. With the number of OPERABLE channels less than 6 thermocouples per quadrant, restore these thermocouples to OPERABLE status within 7 days, or be in at least HOT SHUT 00WN within the next 12 hours.
!OUTH TEXAS - UNITS 1 & 2 3/4 3-71 AMENDMENT N35. AND 9
, . pi TABLE 3.3-10 (Continued)
ACTION STATEMENT 3 (Continued)
- c. With the number of OPERABLE channels less than 4 thermocouples per quadrant, restore these tu rmocouples'to OPERABLE status within 48 hours, or be in at least _ HOT SHUTCOWN.within the next 12 hours.
] I l i l t l l l 3/4 3-72 AMEh0 MENT N05. AND SOUTH TEXA5 - UNITS 1 & 2 II.' . ,. . t -*
-l TABLE 4.3-7 E .'
5 z ACCI' DENT MONITO..NG INSTRUMENTATION SURVEILLANCE REQUIREMENTS M CHANPEL CHANNEL I- IGTRUMENT CHECK CALIBRATION E 1. Containment Pressure M R
-e .
[ 2. Reactor Coolant Outlet Temperature - THOT (Wide Range) M R [ 3. Reactor Coolant Inlet Temperature - TCOLD (Wide Range) M R
- 4. Reactor Coolant Pressure - Wide Range and Extended Range M R S. Pressurizer Water Level M R
- 6. Steam Line Pressure M R 'I y 7. Steam Generator Water level - Narrow Range M R
- 8. Steam Generator Water Level - Wide Range M R
- 9. Refueling Water Storage Tank Water Level M R
- 10. Auxiliary feedwater Storage Tank Water Level M R
- 11. Auxiliary Feedwater Flow M R 5 12. Reactor Coolant System Subcooling Margin Monitor M R A
5 13. Containment Water Level (Narrow Range) M R f 14 Containment Water Level (Wide Range) M R
- 15. Core Exit Thermocouples M R b 16. Steam Line Radiation Monitor M R o
l 17. Containment - High Range Radiation Monitor M R s
-, e--
. ~
, TABLE 4.3-7 (Continued) g . .
y ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS h CHANNEL CHANNEL
~
3 INSTRUMENT CHECK CALIBRATION .- [ 18. Reactor Vessel Water Level (RWL) M R 5 g 19. Neutron Flux (Extended Range) M- R - [ 20. Containment Hydrogen Concentration M R
~
- 21. Containment Pressure (Extended Range) M R
- 22. Steam Generator Blowdown Radiation Monitor- M R
- 23. Neutron Flux - Startup Rate (Extended Range) M R R.
2 E M ' l 9
~
1 5 Y N, c
*9 k
n
FL INSTRUMENTATION CHEMICAL DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.3.3.7 Two independent Chemical Detection Systems shall be OPERABLE with their Alarm / Trip betpoints adjusted to actuate at the following concentrations:
- a. Vinyl Acetate 5 10 ppm
- b. Anhydrous Ammonia / 1 5 ppm Ammonium Hydroxide /
Hydrazine APPLICABILITY: All MODES. ACTION:
- a. With one Chemical Detection Syster. inoperable, restore the inoperable system to OPERABLE statut within 7 days or within the next 6 hours initiate and maintain operation of the Control Room Emergency Venti-lation System in the recirculation mode of operation,
- b. With both Chemical Detection Systems inoperable, within 1 hour initi-ate and maintain operation of the Control Room Emergency Ventilation System in the recirculation mode of operation.
SURVEILLANCE REOUIREMENIS 4.3.3.7 Each Chemical Detection System shall be demonstrated OPERABLE by per-formance of a CHANNEL CHECK at least once per 12 hours, an ANALOG CHANNEL OPERATIONAL TEST at least once per 31 days and a CHANNEL CALIBRATION at least once per 18 months. SOUTH TEXAS - UNITS 1 & 2 3/4 3-75 AMENDMENT N05. AND h3. j ; t.a
FC INSTRUMENTATION 3.3.3.8 (This specification number is not used.) I i i l a i I ( SOUTH TEXA5 - UNITS 1 & 2 3/4 3-76 AMENDMENT NDS. AND l I'*h' 1 ~ 12. L
FL 1 l I TABLE 3.3-11 l (This table number is not used.) SOUTH TEXAS - UNITS 1 & 2 3/4 3-77 AMENDMENT N05. AND lid'. i 7 '..i.
FC ( INSTRUMENTATION 3.3.3.9 (This specification number is not used.) SOUTH TEXAS - UNITS 1 & 2 3/4 3-78 AMENDMENT NDS. AND I' 6 ' i .,,,
f{) INSTRUMENTATION RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.10 The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.3-12 shall be OPERABLE with their Alarm / Trip Setpoints set to ensure that the limits of Specification 3.11.1.1 are not exceeded. The Alarm / Trip Setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (ODCM). APPLICABILITY: At all times. ACTION:
- a. With . radioactive liquid effluent monitoring instrumentation channel Alarm / Trip Setpoint less conservative than required by the above 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 3.3-12. Restore the inoperable instrumentation to OPERABLE status within the time specified in the ACTION, or explain in the next Semiannual Radioactive Effluent Release Report pursuant to Specification 6.9.1.4 why this inoperability was not corrected within the time specified.
- c. The provisions of Specification 3.0.3 are not applicable.
SUMEILLANCE RE0_UIREMEtiTS 4.3.3.10 Each radioactive liquid effluent monitoring instrumentation ch?.nnel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE < CHECK, CHANNEL CALIBRATION, and DIGITAL CHANNEL OPERATIONAL TEST at the frequencies shown in Table 4.3-8. I SOUTH TEiAS - UNITS 1 & 2 3/4 3-79 AMENDMENT N05. AND I,. .. .
?
4
. TABLE 3.3-12 g . .
; RADIOACTIVE LIQUID EFFLUENT M0r_.*0 RING INSTRUMENTATION A
x
? MINIPRJM
, CHANNELS e INSTRUMENT OPERA 8LE ACTION 5
g 1. Radioactivity Monitors Providing Alarm and - y Automatic Termination of Release - e. m Liquid Waste Processing Discharge Monitor 1 43
- 2. Flow Rate %asur ment Devices Liquid Waste Processing Discharge Line 1 46 Ra 8
?
9 M . 9-5 k N o . ,Y.
~.
. = i i f: .e _ _ _ _ _ . - . - . ~ _ . . - . -. . - - _ - _ . . . . . . - _ . _ - - _ __ -. . ._. -
F& TABLE 3.3-12 (Continued) ACTION STATEMENTS ACTION 43 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effiuent releases via this pathway may continue for up to 14 days provided that prior to initiating a release:
- a. At least two independent samples are analyzed in accordance with Specification 4.11.1.1.1, and
- b. At least two technically qualified members of the facility staff independently verify the release rate calculations and discharge line valving.
Otherwise, suspend release of radioactive effluents via this pathway. , ACTION 44 - (Not Used) ACTION 45 - (Not Used) ACTION 46 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided the flow rate is estimated at least once per 4 hours during actual releases. Pump performance curves generated in place may be used to estimate flow. SOUTH TEXAS - UNITS 1 & 2 3/4 3 81 AMENDMENT NDS. AND ,
1.b 5E
,, TABLE 4.3-8 8
y RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS M x 2-v. DIGITAL CHANNEL-c. CHANNEL SOURCE CHANNEL OPERATIONA. 25 , INSTRUMENT CHECK CHECK CALIBRATION TEST , d
,, I. Radioactivity Monitors Providing
- p. Alarm and Automatic Termination
,, of Release l Liquid Waste Processing Discharge Monitor D P R(3) Q(1)
L
- 2. Flow Rate Measurement Devices I
l Sa Liquid Waste Processing Discharge Line D(4) N.A. R N.A. z. Y
?? .
M
'a.
5 r l 3? i :- d' 1 , h
'i1 '
* - [
bl TABLE 4.3-9 (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. Monitor failure.
(2) (Not Used) (3) The initial CHANNEL CALIBRATION shall be performed using one or more of the referer.ce 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 shall be used. (4) CHANNEL CHECK shall consist of verifying in'dication of flow during periods of release. CHANNEL CHECK shall be made at least once per 24 hours on days on which continuous, periodic, or batch releases are made. 1 4 4 I SOUTH TEXAS - UNITS 1 & 2 3/4 3 83 AMEN 0 MENT N35. AND i i 4
,- , --- ..,n, - - - , , ,-_
?I INSTRUMENTATION RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.11 The radioactive gaseous effluent monitoring instrumentation channels shown in Table 3.3-13 shall be OPERABLE with their Alarm / Trip Setpoints set to ensure that the limits of 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 param-sters in the ODCH. - APPLICABILITY: As shown in Table 3.3-13 ACTION:
- a. With a radioactive gaseous effluent monitoring instrumentation channel Alarm / Trip Setpoint less conservative than required by the above specification, immediately suspend the release of radioactive gaseous effluents monitored by the affected channel, or declare the channel inoperable,
- b. With the number of OPERABLE radioactive gaseous effluent monitoring instrumentation channels less than the Minimum Channels OPERABLE, take the ACTION shown in Table 3.3-13. Restnre the inoperable instrumentation to OPERABLE status within the time specified in the ACTION, or explain in the next Semiannual Radioactive Effluent Release Report pursuant to Specification 6.9.1.4 why this inoperability was not co.rected within the time specified,
- c. The provisions of Specification 3.0.3 are not applicable.
iL8Y11L], M E E @lREMENTS 4.3.3.11 Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST or DIGITAL l CHANNEL OPERATIONAL TEST, as applicable, at the frequencies shown in Table 4.3-9. 1 1 1 1 i SOUTH TEXAS - UNITS 1 & 2 3/4 3-84 AMENDMENT NDS. AND
j: '
.i .
TABLE 3.3-13 h R'ADI0 ACTIVE CASE 005 EFFLUENT MONITORING INSTRUMENTATION 5 MINIMUM CHANNELS 7 INSTRUMENT OPERABLE APPLICABILITY ACTION E 1. GASEOUS WASTE PROCESSING SYSTEM Explosive
] Gas Monitoring System
[ 0xygen Monitor (Process) 1 ** 51 m
- 2. Condenser Evacuation System
- a. Condenser Air Removal System Discharge Header Noble Gas Activity Monitor 1 .
- 49 w
1 b. Flow Rate Monitor 1
- 48 w
J> u.
- c. Sampler Flow Rate Monitor 1
- 48
- 3. Unit Vent
- a. Noble Gas Activity Monitor 1
- 49
- b. Iodine Monitor or Iodine Sampler 1
- 53 5 c. Particulate Monitor or Particulate Sampler 1
- 5:t 4
5 d. Flow Rate Monitor 1
- 441 f e. Sampler Flow Rate Monitor 1
- 48 l
c o se b D _ - _ . . , - _ _ . _ , . _ _ . . - _ _ _ _ _ . _ . _ _ _ . , _ - - _ . _ - _ _ . . _ _ _ - . _ _ _ .-~__ _ ____ . _ . .
F& TABLE 3.3-13 (Continued) TABLE NOTATIONS
- At all times.
** During GASEOUS WASTE PROCESSING SYSTEM operation.
ACTION STATEMENTS
~
ACTION 47 - (Not used) ACTION 48 - With the number of channels OPERABLE less than require; by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided the flow rate is estimated at least once per 4 hours. ACTION 49 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided grab sanples are taken at least once per 12 hours and these samples are analyzed for radioactivity within 24 hours. ACTION 50 - (Not used) ACTION 51 - With the nua.ber of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, operation of this GASEQUS WASTE F't0 CESSING SYSTEM aiay continue provided grab samples are collected at least once per 4 hours and analyzed within the following 4 hours. ACTION 52 - (Not used) AC' TION 53 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via the affected pathway m6y continue for up to 30 di.ys provided samples are continuously collected with auxiliary sampliag equipment as required in Part A of the ODCM. SOUTH TEXAS - UNITS 1 & 2 3/4 3-86 AMENDMENT NOS. 60 l.i. . ; ; .
, TABLE 4.3-9 c) h RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSlRUMENTATION SURVEILLANCE REQUIREMENTS M , ANALOG OR 5
" DIGITAL CHANNEL MODES FOR Wi!CH CHANNEL SOURCE CHANNEL OPERATIONAL SURVEILLANCE E IMSTRUMENT CHECK CHECK CALIBRATION TEST IS REQUIRED
- 1. CASEOUS WASTE PROCESSING SYSTEM
[ Explosive Gas Monitoring System 0xygen Monitor (Process) D N.A. Q(S) M
- 2. Condenser Evacuation System
- a. Condenser Air Reeval System w Discharge Header hoble Gas 1 Activity Monitor D M R(3) Q(2) *
- b. Flow Rate Monitor D N.A. R N.A. *
- c. Sampler Flow Rate Monitor D N.A. R , Q
- 3. Unit Vent
- a. Noble Gas Activity Monitor D M R(3) Q(2)
A b. Iodine Monitor or D M R(3) Q(2)
- Q Iodine Sampler W N.A. N.A. M.A.
5 4 c. Particulate Monitor or D M R(3) Q(2) z Particulate Sampler W N.A. M.A. N.A.
- d. Flow Rate Monitor D N.A. R N.A. *
- e. Sample Flow Rate Monitor D N.A. R
- Q T1 M
*W s
Fs' TABLE 4.3-9 (Continued) TABLE NOTATIONS
- At all times.
During GASEOUS WASTE PROCESSING SYSTEM operation. (1) (Not Used) (2) The DIGITAL CHANNEL OPERATIONAL TEST shall also demonstrate that control room alarm annuncietion occurs if any of the fo' lowing conditions exists:
- a. Instrument indicates measured levels above the Alarm Setpoint, or
- b. Monitor failure.
(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 shall be used. (4) (Not Used) (5) The CHANNEL CALIBRATION shall include the use of a standard gas sample containing a nominal two volume percent oxygen, balance nitrogen. A SOUTH TEXAS - UNITS 1 & 2 3/4 3 8S AMENDMENT SOS. AND I- 1 7 g;;.
b INSTRUMENTATION 3/4.3.4 TURBINE OVERSPEED PROTECTION LIMITING CONDITION FOR OPERATION , 3.3.4 At least one Turbine Overspeed Protection System shall be OPERABLE. 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 hours, or close at least one valve in the affected steam line(s) or isolate the turbine from the steam supply within the next 6 hours,
- b. With the above required Turbine Overspeed Protection System otherwise inoperable, within 6 hours isolate the turbine from the steam supply.
SURVEILLANCE REOUIREMENTS 4.3.4.1 The provisions of Specification 4.0.4 are not applicable. 4.3.4.2 The above required Turbine Overspeed Protection System shall be demonstrated OPERABLE:
- a. At least once per 31 days in MODES 1 and 2 when the main turbine is operating by cycling each of the following valves through at least one complete cycle from the running position:
- 1) Four high pressure turbine stop valves,
- 2) Four high pressure turbine governor valves,
- 3) Six low pressure turbine reheat stop valves, and
- 4) Six low pressure turbine reheat intercept valves.
- b. At least once per 31 days in MODES 1 and 2 when the main turbine is operating by direct observation of the movement of each of the above valves through one co.?plete c,vele from the running position,
- c. At least once per 18 months by performance of a CHANNEL CAllBRATION on the Turbine Overspeed Protection Systems, and
- 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, a3d Stems and verifying no unacceptable flaws or excessive corrosion. If unacceptable flaws or excessive corrosion are found, all other valves of that type shall be inspected."
Disassembly and inspection of the low pressure turbine reheat intercept valves are not required prior to the end of the first 40 month interval. SOUTH TEXAS - UNITS 1 & 2 3/4 3-89 AMENDMENT N05. AND _. _ _mi i-ii
FO 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION STARTVP AND POWER OPERATION LIM 1I1HQ_DQHD111DH_LQF._0PERATION , 3.4.1.1 All reactor coolant loops shall be in operation. , 1 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 hours. 1 SURVEILLANCE REOUIREMENTS 4.4.1.1 The above required reactor coolant loops shall be verified in ! operation and circulating reactor coolant at least once per 12 hours. l I l t i e
*See Special Test Exceptions Specification 3.10.4 SOUTH TEXAS - UNITS 'i & 2 3/4 4-1 AMEN 0 MENT N05. AND MOV 1 * '."'.'-
, - s. . .
b REACTOR Cfs0LANT SYSTEM HOT STANDBY QMITINGCONDITIONFOROPERATION 3.4.1.2 At least two of the reactor coolant loops listed below shall be 3 OPERABLE and with two reactor coolant. loops in operation when the Reactor Trip System breakers are closed and one reactor coolant loop in operation when the Reactor Trip 5"item breakers are open:*
- a. Reactor Coolant Loop A and its associated steam generator and reactar coolant pu'sp.
- b. Rea!: tor Coolant 1.oop 8 and its associated steam generator and reactor coolant pump,
- c. Reactor Ceclant Loop C and its associated steam generator and reactor coolant pump, and
- d. Reactor Coolant Loop D and its associated steam generator and reactor coolant pump.
APPLICABILITY: H0DE 3. ACTION:
- a. With less than the above required reactor coolant loops OPERABLE, 1
restore the required loops to OPERABLE status within 72 hours or be in HOT SHUTDOWN within the next 12 hours. l b. With only one reactor coolant loop in operation and the Reactor Trip System breakers in the closed position, within 1 hour open the Reactor i Trip System breakers.
- c. With no reactor coolant loop in operation, suspend all operations I 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.
! 193%ULLEWQ41REMENTS , l 4.4.1.2.1 At least the above required reactor coolant purrps, if not in opera-j tion, shall be determined OPERABLE once per 7 days by verifying correct breaker alignments and indicated poner availability. 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 10% narrow range at least once per 12 hours. 4.4.1.2.3 The required reactor coolant loops shall be verified in operation and circulating reactor coolant at least once per 12 hours. l *All reactor coolant pumps may be deenergized for up tc, I hour provided: (1) no operations are permitted that would cause dilution of the Reactor Coolant Syste.: boron concentration, and (2) core outlet temperature is maintained at least 10'F below saturation temperature. SOUTH TEXA5 - UNITS 1 & 2 3/4 4-2 AMENDMENT NDS. AND s ..
9 , 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 i one of these loops shall be in operation:* j
- a. Reactor Coolant Loop A and its associated steam generator and ;
reactor coolant pump,** , j
- b. Reactor Coola'nt Loop B and its associated steam generator anc reactor coolant pump,** ,
- c. Reactor Coolant Loop C and its associated steam generator and reactor coolant pump,** ;
i
- d. Reactor Coolant Loop D and its associated steam generator and l reactor coolant pump,**
- e. RHR Loop A with valve CV0198 locked or pinned in position to limit flow to 125 gpm, ;
, f. RHR Loop B with valve CV0198 locked or pinned in position to limit flow to 125 gom, and
. g. RHR Loop C with valve CV0198 locked or pinned in position to limit flow to 125 gpm.
APPLICABILITY: MODE 4. ACTION:
- a. With less than the ab3ve required loeps OPERABLE, immediately "
initiate corrective action to return the required loops to OPERABLE ' status as soon as possible; if the remaining ODERABLE loop is an RHR loop, be in COLD SHUTDOWN within 24 hours. b' . With no loop in operation, suspend all operations involving a reduc-l tion in boron concentration of the Reactor Coolant System and i immediately initiate corrective action to return the required loop '
- to operation.
'All reactor coolant pu?ps 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 with one or more of the Reactor Coolant System cold leg teeperatures less than or equal to 350'F unless the secondary water temperature of each steam generator is less than 50'F above each of the Reactor Coolant System cold leg temperatures.
5 LOUTH TEXAS - UNITS 1 & 2 3/4 4-3 AMENDMENT N05. AND 4 RV i i i,.a
FD REACTOR COOLANT SY' STEM HOT SHUTDOWN SURVEILLANCE REOUIREMENTS 4.4.1.3.1 The required reactor coolant pump (s) ar.d/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 generator (s) shall be determined OPERABLE by l Verifying secondary side water level to be gre6ter than or equal to 10% narrow t range at least once per 12 hours. 4.4.1.3.3 At least one reactor coolant loop, or one RHR loop with valve CV0198 locked or pinned in position to limit flow to 125 gpm shall be verified in operation and circulating reactor coolant at least once per 12 hours, i
- t 4
i I 1 l l SOUTH TEXAS - liNITS 1 & 2 3/4 4-4 AMENDMENT N05. AND NOV :;tg;
FD 1 l REACTOR COOLANT SYSTEM 1 COLD SHUT 00WN - LOOPS FILLED I LIMITING CONDITION FOR ODERATION 3.4.1.4.1 At least one residual heat removal (RHR) loop with valve CV0198 locked or pinned in position to limit flow to 125 gpm shall be OPERABLE and in opera-tion *, and either: i
- a. One additional RHR loop shall be OPERABLE **, or i
- b. The secondary side water level of at least two steam generators shall be greater than 10% narrow range.
l APPLICABILITY: MODE 5 with reactor coolant loops filled ***. { ACTION: l
- a. With twe of the RHR loops inoperable and with less than the required .
l steam generator water level, immediately initiate corrective action < to return one of the inoperable RHR loops to OPERABLE status or restore j the required steam generator 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 i loop to operation.
SURVEILLANCE RE00lPEMEN15 4.4.1.4.1.1 The secondary side water level of at leart two steam generators when required shall be determined t' he within limits at least once per 12 hours. 4.4.1.4.1.2 At least one RHR loop with valve CV0198 locked or pinned in posi-tion to limit flow to 125 gpm shall be determined to be in operation and cir-culating reactor coolant at least once per 12 hours. i
*The RHR pump may be deenergized for up to I ho'Jr provided: (1) no operations are permitted that wouid cause dilution of the Reactor Coolant System boren concentration, and (2) core outlet temperature is maintained at least 10*F below saturation temperature.
**Two RHR loops may be inoperable for up to 2 hours for surveillance testing i provided the other RHR loop is OPERABLE and in operation, j ***A reactor coolant pump shall not be started with one or more of the Reactor j Coolant System cold leg temperatures less than or equal to 350*F unless the secondary water temperature of each steam generator is less than 50*F above eacn of the Reactor Coolant System cold leg temperatures.
A 3/4 4-5 AMENDMENT N05. AND SOUTH TEXA5 - UNITS 1 & 2 j h. ; , ". J. ;
i FL l REACTOR COOLANT SYSTEM COLD SHUTDOWN - LOOPS NOT FILLED 2 LIMITING CONDITION FOR OPERATION 3.4.1.4.2 At least two residual heat removal (RHR) loops shall be OPERABLE
- and at least one RhR loop shall be in operation.**
APPLICABILITY: MODE 5 with reactor coolant loops not filled. ACTION:
- a. With less than the above required RHR loops OPERABLE, imediately initiate corrective action to return the required RHR loops to OPERABLE status as soon as possible,
- b. With no RHR loop in operation susperJ all operations involving a reductioninboronconcentratIonoftheReactorCoolantSystemand immediately initiate corrective action to return the required RHR loop to operation.
SURVEILLANCE RE0VIREMENTS _ 4.4.1.4.2.1 At least one RHR loop shall be determined to be in operation and circulating reactor coolant at least once per 12 hours. 4.4.1.4.2.2 Valves FCV-110B, FCV-111B, CV0201A, and CV0221 shall be verified closed and secured in position by mechanical stops or removal of air or elec-trical power at least once per 31 days. 4.4.1.4.2.3 Valve CV0215 shall be verified closed and secured in position by a permanent restraint with the handwheel remaved prict to entering MODE 5 with the reactor coolant loops not filled.
"Two RHR loops may be inuperable for up to 2 hours for surveillance testing provided the other RHR loop is OPERABLE and in operation.
**The RHR pu?p eay be deenergized for up to I hour provided: (1) no opera-tions are perr.itted that would cause dilution of the Reatter Coolant Syster boron concentration, and (2) core outlet temperature is raintained at least 10'F below saturation terperature.
SOUTH TEXAS - UNITS 1 & 2 3/4 4-6 AND AMENDMENT NOS. 4'. ): ,, q:
// F1 l l REACTOR ;COLANT SYSTEM l
1 3/4.4.2 SAFETY VALVES i SHUTDOWN .i ., LIMITING CONDITION FOR OPERATION 3.4.2.1 A minicum of one pressurizer Code safety valve shall be OPERA 8LE with a lift setting of 2485 psig i 1%.* ~ APPLICABILITi. MODES 4 and 5. ACTION: With no pressurizer Code safety valve OPERABLE, immediately suspend all operations involving positiva reactivity changes and place an OPERABLE RHR loop into operation in the shutdown cooling mode. SURVEILLANCE REOUIREMENTS 4.4.2.1 No additional requirements other than those required by Specification 4.0.5. i i
, *The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressuee. ,
500TH TEXAS - UNIT.S 1 & 2 3/4 4 7 AMENDMENT NDS. ?.ND K I' 1 7 D',',
FD 1ACTORCOOLANTSYSTEM OPERATING
- 11MITING CONDITION FOR_0PERATION 3.4.2.2 All oressuri:er Code safety valvet shall be OPERABLE with a lift set-ting of 2485 psig i 1%.*
APPLICABILITY: H0 DES 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 hours and in at least HOT SHUTDOWN within the following 6 hours. SURVEILLANCE REQUIREMENTS 4.4.2.2 No additional requirements other than those reqdred by
- Specification 4.0.5.
i i i I i l 3 "The lift setting pres:ure shall correspond '.o ambient conditions of the valve
- at nominal operating temperature and oressure.
I i SOUTH TEXA5 - UNITS 1 & 2 3/4 4-8 AMENDMENT N05. AN?
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5 REACTOR COOLANT SYSTEH 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 1816 cubic feet, and at least two groups of pressurizer heaters supplied by ESF power each having a capacity of at least 175 kW. APPLICABILITY: MODES 1, 2, and 3. - ACTION:
- a. With only one group of pressurizer heaters supplied by ESF power OPERABLE, restore at least two groups to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
- b. With the pressurizer otherwise inoperable, be in St least HOT STANDBY with the Reactor Trip System breakers open within 6 hours and in HOT SHUTDOWN within the following 6 hours.
SURVEILLANCEREOUIPlgNTS 4.4.3.1 The pressurizer water volume shall be determined to be within its limit at least once per 12 hours. 4.4.3.2 The capacity of each of the above required groups of pressurizer heaters supplied by ESF power shall be verified by energizing the heaters and measuring circuit current ct least once per 92 days. SOUTH TEXA5 - UNITS 1 & 2 3/4 4-9 AMENDMENT N05. AND
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O REACTOR COOLANT SYSTEM 3/4.4.4 RELIEF VALVES LIMITING CONSITION FOR OPERATION 3.4.4 All power-operated relief valves (PORVs) and their associated block valves shall be OPERABLE. APPLICABILITY: H] DES 1, 2, and 3. ACTION:
- a. With one or more PORV(s) inoperable, because of excessive seat leakage, within 1 hour either restore the PORV(s) to ODERABLE status or close the associated block valve (s); otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTOOWN within the following 30 hours.
- b. With one PORV inoperable due to causes other than excessive ciat leakage, within 1 hour either restore the FORV to OPERABLE status er close the associated block valve and remove power from the block valve; restore the PORV to OPERABLE status within the following 72 hours or be in HOT STANDBY within the next 6 hours and in COLD SHUT 00WN within the following 30 hours,
- 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 power from the block valve (s) and be in HOT STANDBY within the next 6 hours and COLD SHUTOOWN within the following 30 hours,
- d. With one or more block valve (s) inoperable, within 1 hour:
(1) restore the block valve (s) to OPERABLE ststus, er close the t' lock valve (s) and remove power from the block valve (:), or close the PORV and remove power from the PORV; and (2) apply the ACTION b. or c. above, as appropriate, for the isolated P0th('s).
- e. The provisions of Specification 3.0.4 are not applicabit.
I i l SOUTH TEXA5 - UNITS 1 & 2 3/4 4-10 AMENDMENT NDS. ANO ,
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fh I REACTOR COOLANT SYSTEM RELIEF VALVES i g EILLANCE REOUIREMENTS 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. Performing a CHANNEL CALIBRATION, and
- b. Operating the valve through one complete cycle of full travel.
4.4.4.2 Each block valve shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel unless the block valve is closed with power removed in order to meet the requirements of ACTION b, or c. in Specification 3.4.4. l i ! i ) . i I h l l l l i SOUTil TEXAS - UNITS 1 & 2 3/4 4-11 AMENDMENT NDS. AND
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M REACTOR COOLANT SYSTEM
?/4.4.5 STEAM GENERATORS LIMITING CONDITION FOR OPERATION 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 increc. sing T,yg above 200'F. SURVEILLANCE REOUIREMENTS 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 Sample Selection and In'spection - 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 Stean 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 the total 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 be from these critical arees;
- b. The first sample of tubes selected for each inservice' inspection (subsequent to the preservice inspection) of each steam generator
. shall include:
- 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 SOUTH TEXAS - UNITS 1 & 2 3/4 4-12 AMENDMENT NOS. AND h !* ,' 1 ~ iL'y
Ptll REACTOR COOLANT SYSTEM l i STEAM GENERATORS SURVEILLANCE REOUIREMENTS (Continued)
- 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 inservice inspection may be subjected to a partial tube inspection provided:
- 1) The tubes selected for these samples include the tubes from those areas of the tube sheet array where tubes with 4 imperfections were previously found, and i
- 2) The inspections include those portions of the tubes where irtperfections were previously found.
The results of each sample inspection shall be classified into one of the following three categories: Category Inspection Results 4 C-1 Less than % of the total tubes inspected are degraded j 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% j 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. i ! Note: In all inspections, previously degraded tubes must exhibit
, significant (greater than 10t.) furtner wall penetrations to
) be included in the above percentage calculations. 1 a SOUTH TEXA5 - UNITS 1 & 2 3/4 4-13 AMEfCMENT N05. AND L. .
FP REACTOR COO! AESYSTEM STEAM GENERATORS SURVEILLANCE RE0UIREMENTS (Contig_ued) 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 but within 24 calendar months of initial criticality.
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 l observed degradation has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of I once per 40 months; i b. If the results of the inservice inspection of a steam generator
- conducted in a cordance with Table 4.4-2 at 40-month intervals fall in Category C-3, the inspection frequency shall be increased to at I least once per 20 months. The increase in inspection frequency
- shall apply until the subsequent inspections satisfy the criteria of i Specification 4.4.5.3a.; the interval may then be extended to a maximum of once per 40 months; and
- c. Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection i specified in Table 4.4-2 during the shutdown subsequent to any of j the following conditions:
- 1) Primary-to-secondary tube leaks (not including leaks originating I
from tube-to-tube sheet welds) in excess of the limits of j Specification 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.
SOUTri TEXAS - UNITS 1 & 2 3/4 4-14 AMECMENT CS, AC
Fo l l i REACTOR COOLANT SYSTEM l l STEAM GENERATORS l SURVEILLANCE RE0UIREMENTS (Continued) 4.4.5.4 Acceptance Criteria
- a. As used in this specification:
- 1) Imperfection means an exception to the dimensions, finish, or contour of e tube from that required by fabric,ation drawings or specificationt. Eddy-current testing indications below 20% of the nominal tuse wall thickness, if detectable, may be considered as imperfections;
- 2)
Dearadation means a service-induced cracking,
wastage, wear, or general corrosion occurring on either inside or outside of a tube;
- 3) Decraded Tube means a tube containing imperfections greater than or equal to 20% of the nominal wall thickness caused by degradation;
- 4) % Degradation means the percentage of the tube wall thickness affectso or removed by degradation;
- 5) Defect means an imps. Sction of such severity that it exceeds the plugging limit. A tube containing a defect is defective;
- 6) Pluacina Limit means the imperfection depth at or beyond which the tube sna'iT be removed from service and is equal to 40% of the nominal tube wall thickness;
, 7) Unserviceable describes the condition of a tuM 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 feed ater line break as specifieo in Specification 4.4.5.3c., above;
- 8) Tube Inspection neans an inspection of the steam generator tube f rom the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg; and
, 9) Preservice Inspection means an inspection of the full length of each tube in each steam generator performed by eddy current techniques prior to servict to establish a baseline condition of the tubing. This inspection shall be performed prior to initial PodER OPERATION using the equipment and techniques expected to be used during subsequent inservice inspections, i
i. t SOUTH TEXA5 - UNITS 1 & 2 3/4 4-15 AMENDMENT N05. AND l . 7 l 1 *. 'i . . ', l i-
REACTOR COOLANT SYSTEM i STEAM GENERATORS ! SURVEILLANCE REQUIREMENTS (Continuedi ;
- b. The steam generator shall be determined OPERABLE after completing 1 the corresponding actions (plug all tubes exceeding the plugging j limit and all tubes containing through-wall cracks) required by Table 4.4-2.
i 4.4.5.5 Reports
- a. Within 15 days 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 inspection !
. shall be tubmitted to the Commission in a Special Report pursuant to r l Specification 6.9.2 within 12 months following the completion of the ,
inspection. This Special Report shall include: I
- 1) Number and extent of tubes inspected,
- 2) Location and percent of wall-thickness penetration for each
. indication of an imperfection, and
- 3) Identifica lon of tubes plugged.
) , c. Results of steam generator tube inspections which fall into Cate- t gory C-3 shall be reported in a Special Report to the Commission ! pursuant to Specification 6.9.2 within 30 days and prior to resump- !
- tion of plant operation. This report shall provide a description of j i.vestigations conducted to determine cause of the tube degradation
- i and corrective measures taken to prevent recurrence.
. I
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l l 4 1 1 1 ! f t , SOUTH TEXAS - UNITS 1 & 2 3/4 4-16 AMEN 0 MENT N05. AND ; I l i .. ....
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SOUTH TEXA$ - UNITS 1 & 2 3/4 4-17 AMENDMENT NOS. AND
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TARLE 4.4-2 E c g STEAas CENER ATOR TtMtt INSPECTION A I-7 w 1ST SAesPLE INSPECT 904 2ND SAasPLE INSPECTION 3RO SAasPLE INSPECTION
. eg Sire Rewet Action Required Remet Actiori Required Rewit Acteen Ikpuered .
5 0 A --'2:.._ = of C-1 None M. A. N. A. M. A. N. A. d S Tubestw
- S. G.
- p. C Ptmo defective tutwo C-1 Mone M. A. N. A.
" *"d W "dd* Plug defect.ve outws C-1 None 25 totes in tSis S. G. C-2 and inerect addnionet C-2 Plug defective tubes 45 tutws in etws S. G.
Perforsse action for C-3 C-3 resuet of firse : esnapse w 2 rerforwe action for i C-3 C ~l result of first N. A. N. A. 9 sasnple
% segwct oss subn 6n AN &
C-3 i this S. G., pews e S. G.s are Mea
- M. A. N. A. '
e,cew v.ews and C-1 i Ineswet 25 tubes in sesne S. G.s pg,gense action ser rech estwe S. G. M. A. N. A-
;E C-2 but no C--2 rems 8t of Second 1 5 additional g j poetificeeiew to 90MC S. G. are
.j, h purment se 150.72 C-3 Ca r thil2)of to CFR Additional 86 aft tubes in Part 50 M
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',. G. is C-3 each S. G. and plug (y defectose outes.
psovitication se MMC N. A. M A- I 3 pursusne se 550.72 6 lbH2) of 10 CF R
)
Part50 S-3 % N h h
- d m N h h unit, and n h h W W ineene gnweseen ;-_W n during on ingwetion Y
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FC' REACTOR COOLANT SYSTEM i 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITINO CONDITION FOR OPERATION , 3.4.6.1 The following Reactor Coolant System Leakage Detection Systems shall be OPERABLE: I
- a. The Containment Atmosphere Gaseous Radioactivity Monitoring System,
- b. The Containment Normal Sump Level and Flow Monitoring System, and
- c. The Containment Atmosphere Particulate Radioactivity lionitoring l' System.
APPLICABILITY: MODES 1, 2, 3, and 4. i ACTION:
- a. With a. or c. of the above required Leakage Detection Systems in-operable, operation may continue for up to 30 days provided grab samples of the containment atmosphere are obtained and analyzed 'or ,
gaseous and particulate radioactivity at least once per 24 hours when ! the required Gaseous or Particulate Radioactive Monitoring System is ! inoperable; otherwise, be in at least ll0T STANDBY witnin the next 6 hours and in COLD SHUTOOWN within the following 30 hours. ;
- b. With b. of the above required Leakage Detection Systems inoperable, f be in at least HOT STANDBY within the next 6 hours and in COLD SHUT- l 00WN within the following 30 hours. [
- c. With a. and c. of the above required Leakage Detection Systems inoperable: {
- 1) Restore either Monitoring System (a. or c.) to OPERABLE status l within 72 hoe:rs and >
- 2) Obtain and analyze a grab sample of the containment atmosphere for !
gaseous and particulate radioactivity at least once per 24 hours, and
- 3) Perfore a Reactor Coolant System water inventory balance at l 1 east once per 8 hours, j Otherwise, be in at least HOT STANDCY within the next 6 hours and in COLD SHUTOOWN within the following 30 hours.
SVEYEILLAME REOUIREk'ENTS 4.4.6.1 The Leakage Detection Systems shall be demonstratel OPERABLE by*
- a. Containtrent Atmosphere Gaseous and Particulate Mo titoring Systems !
, performance of CHANNEL CHECK, CHANNEL CALIBRATION. and DIGITAL CHANNEL OPERATIONAL TEST at the frequencies specified in (able 4.3-3, and g
- b. Containment Nores1 Surp Level and Flow Monitorin) System performance ',
of CHANNEL CALIERATION at least once per 18 men'.hs. l l SOUTH TEXAS - UNITS 1 & 2 3/4 4-19 Ak'Eh0 MENT N05. AND [ s '. . l
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1: REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITINS CONDITION FOR OPERATION 3.4.6.2 Reacto.' Coolant System leakage shall be limited to:
- a. No PRESSURE BO'tNDARY LEAKAGE,
- b. 1 gpm UNIDENTIFIED LEAKAGE, 1 c. 1 gpm total reactor-to-secondary leakage through all steam J generators and 500 gallons per day through any one steam generator,
- d. 10 gpm IDENTIFIED LEAKAGE from the Reactor Coolant System, and
- e. 0.5 gom leakage per nominal inch of valve size up to a maxieum of
, 5 gpm at a Reactor Coolant System pressure of 2235 1 20 psig from any Reactor Coolant System Pressure Isolation Valte specified in Table 3.4-1.*
APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: l ) a. With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY
- within 6 hours and in COLD SHUTDOWN within the following 30 hours.
l b. With any Reactor Coolant System leakage greater than any one of the , above limits, excluding PRESSURE BOUNDARY LEAKAGE and leakage from i Reactor Coolant System Pressure Isolation Valves, reduce the leakage 3 rate to within limits within 4 hours or be in at least HOT STANDBY
- within the next 6 hours and in COLD SHUTDOWN within the following
] 30 hours. ]
- c. With any Reactor Coolant System Pressure Isolation Valve leakage
! greater than the above limit, isolate the high pressure por ion of I the affected system from the low pressure portion within 4 hours by
- use of at least two clo m*d ranual or deactivated auton.atic valves.
I or be in at least HDT S U NOBY within the next 6 hours and in COLD
. SHUTOOWN within the foll. wing 30 hours.
i j
- Test pressures less than 2235 psig but grei.ter than 150 psig are allowed.
Observedleakageshallbeadjustedfortheactualtestpressureupto2235psig assuming the leakage to be directly proportional to pressure differential to the one-half power. I i SOUTH TEXAS - UNITS : &2 3/4 4-20 AMECMENT N05. AND i;U l t i 1
FD . REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE SURVEILLANCE REQUIREMENTS (Continued) 4.4.6.2.1 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by:
- a. Monitoring the containment atmosphere gaseous radioactivity and par-ticulate radioactivity channels at least once per 12 hours;
- b. Monitoring the containment normel sump inventory and discharge at least once per 12 hours;
- c. Performance of a Reactor Coolant System water inventory balance at least once per 72 hours; and
- d. Monitoring tne Reactor Head Flange Leakoff System at least once per 24 hours.
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 SHUTDOWN for 72 hours or more and if leakage testing has not been performed in the previous 9 months,
- c. Prior to returning the valve to service following maintenance, repair or replacement work on the valve, and
, d. Within 24 hours following valve acsuation due to automatic or manual action or flow through the valve except for valves XRH0060 A,B.C and XRH0061 A,B,C.
- e. As outlined in the ASME Code, Section XI, paragraph IWV 3427(b).
The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4. SOUTH TEXAS - UNITS 1 & 2 3/4 4-21 AMEN 0 MENT N05. ANO
ll TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES VALVE NUMBER FUNCTION XSIOOO7 A, B, C HHS! Cold Leg Injection Check Valves (RCS Loops 1, 2, 3) X$10009 A, B, C HHSI Hot Leg Recirculation Check Valves (RCS Loops 1, 2, 3) XS10010 A, B, C LHSI/HHSI Hot Leg Recirculation Check Valves (RCS Loops 1, 2, 3) XRH0020 A, B, C LHSI Ho', Leg Recirculation Check Valves (RCS Loops 1, 2, 3) XRH0032 A, B, C LHSI/RHR Cold Leg Injection Check Valves (RCS Loops 1, 2, 3) XS!0038 A, B, C LHSI/HHSI/RHR/ Accumulator Cold Leg Injection Check Valves (RCS Loops 1, 2, 3) XS!0046 A, B, C Accumulator Cold Leg Injection Check Valves (RCS Loops 1, 2, 3) XRH0060 A, B, C RHR Suction Isolation Valves (RCS Loops 1, 2, 3) XRH0061 A, B, C RHR Suction Isolation Valves (RCS Loops 1, 2, 3) SCUTH TEXAS - UNITS 1 & 2 3/4 4-22 AMEN MENT h05. ANO
f REACTOR COOLANT SYSTEM 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 hours or ce in at least HOT STANDSY within the next 6 hours and in COLD SHUTOOWN within the following 30 hours; and
- b. With any one or more chemistry parameter in excess of its Transient Limit, be in at least HOT STANDBY within 6 hours and in COLD SHUT 00WN within the following 30 hours.
At All Other Tires: With the concentration of either chloride or fluoride in the Reactor Coolant System in excess of its Steady-State Limit for more than 24 hours or in excess of 't: Transier.t Limit, reduce the pressurizer pressure to less than or eaus! to 500 psig, if applicable, and perform an engineering evaluation to ceterr.ine 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 pr!or to proceeding to MODE 4. S E Y @ Laji; L g g g E3'p; 3 4.4.7 The Reactor Coolant System chemistry shall be determined to be within the limits by analysis of those parateters at the frequencies specified in Table 4.4-3. SOUTn TE)AS - UNITS 1 & 2 3/4 4-23 AMENCuENT N05. AND L, , .,,,
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FC TABLE 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS STEADY-STATE TRANSIENT PARAMETER LIMIT LIMIT Dissolved Oxygen * < 0.10 ppm $ 1.00 ppm Chloride < 0.15 ppm 5 1.50 ppm Fluoride 3 0.15 ppm 5 1.50 ppm
- Lieit not applicable with T, less than er equal to 250*F.
SOUTH TEXA5 - UNITS 1 & 2 3/4 4-24 AMEN:HENT h05. ANC
TA8LE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVE!LLANCE REQUIREMENTS SAMPLE AND PARAMETER ANALYLIS FREQUENCY Dissolved Oxygen
- At least once per 72 hours Chloride At least once per 72 hours Fluoride At least once per 72 hours
- Net required with T,yg less than or equal to 250'F.
SOUTil TEXAS - UNITS 1 & 2 3/4 4-25 AMENDMENT N05. AND
- 1. . : ; 1933
F1 REACTOR COOLANT SYSTEM 3/4.4.8 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.8 The specific activity of the reactor coolant shall be limited to:
- a. Less than or equal to 1 microcurie per gram DOSE EQUIVALENT l-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 EQUIVALENT I-131 for more than 48 hours during one continuous time interval, or exceeding the limit line shown on Figure 3.4-1, be in at least HOT STANDBY with T avg less than 500'F within 6 hours; and
- b. With the gross specific activity of the reactor coolant greater than 100/l microCuries per gram, be in at least HOT STANDBY with T less avg than 500'F within 6 hours.
MODES 1, 2, 3, 4, and 5: With the specific activity of the reactor coolant greater than 1 microcerie per gran DOSE EQUIVALENT l-131 or greater than 100/E micro-Curies per gram, perform the sampling and analysis requirer.ents of item 4.a) of Table 4.4-4 until the specific activity of the reactor coolant is restored to within its limits.
*Witn T greater than or equal to 530*F.
.l SOUTH TEXAS UNITS 1 & 2 3/4 4-26 AMEN: MENT N05. AND
FF l l l REACTOR COOLANT SYSTCH SPECIFIC ACTIVITY l SURVEILLA* ICE RE0VIREMENTS _ 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, 4
$0JTH TEXAS - UNITS 1 & 3/4 4-27 AEN; MENT N05. A*C
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a o O - 20 30 40 50 60 70 80 90 100
' PERCENT OF RATED THERMAL POWER FIGURE 3.4-1 MSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACT!v!TY LIMIT VERSUS PERENT OF RATED THERMAL POWER WITH THE REA; TOR COOLANT SFECIFIC ACTIVITY >l p;i/gra- 005E EQUIVALENT I-131 3/4 e-2E AMEN 0 MENT N05. AND SOUTH TEXA5 - UNITS 1 & 2
.: 9 t,~
TABLE 4.4-4 o S RfACTOR COOLANT SPECIFIC ACTIVITY 'AIPLE
- AND ANALYSIS PROGRAM A
5* MODES IM inflCH SAMPLE TYPE OF MEASURDIENT SAMPLE AND ANALYSIS FREQUENCY AIS ANALYSIS REQUIRED AND ANALYSIS g -- 0 1. Cross Radioactivity At least once per 72 hours. 1, 2, 3, 4 w Octermination .
- 2. Isotopic Analysis for DOSE EQUIVA- 1 per 14 days. 1 LENT I-131 Concentration
~
- 3. Radiochemical for E Determination
- 1 per 6 months ** I
- 4. Isotopic Analysis for Iodine a) Once per 4 hours, 1#, 2#, 38, 48, 5#
w D Including I-131, 1-133, and I-135 whenever the specific 4 activity exceeds 1
- pCi/ gram DOSE
"' EQUIVALENT I-131 or 100/E pCi/ gram of gross radioactivity, and b) One sample between 2 1,2,3 f and 6 hours following j a THERMAL POWER change j g exceeding 15%
- M
$ of the RATED THEllMAL 4 POWER within a 1-hour 5 period.
5 y-i k _. o I t,
F0 TABLE 4.4-4 (Continued) TABLE NOTATIONS ] i j- "Aradiochemical.analysisforEshallconsistofthequantitativemeasurement of the specific activity for each radionuclide, except for radionuclides with i half-lives less than 15 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 bssed upon those energy peaks identifiable with a 95% confidence level.
- j. ** 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 hours or longer.
! #Until the specific activity of the Reactor Coolant System is restored within its limits, 4 f I i j l ll k ) i 2 i l i SOUTH TI AAS - UNITS 1 & 2 3/4 4-30 AMENDMENT N05. Ah?
**
- 4 as yQ
4 REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE /TEMPERATURELIMig REACTOR COOLANT SYSTEM LIMITING CONDITION Foo OPERATION 3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the lim:t lines shown on Figures 3.4-2 and 3.4-3 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with;
- a. A maximum heatup of 100'F in any 1-hour period,
- b. A maximum cooldoon of 100*F in any 1-hour period, and
- c. A caximum temperature change of less than or equal to 10'F in any 1-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.
APPLICABILITY: At all times. ACTION: With any of the above limits exseeded, restore the terperature and/or pressure to within the limit within 30 minutes perform an engineering evaluation to determine the effe: cts of the out-of-limit condition on the structural integrity of the Rea: tor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation or be in at least HOT STANDBY within the next 6 hours and reduce the RCS T,y and pressure to less than 200'F and ] S00 psig, respectively, within the ellowing 30 hours. 1 i 511EVEILLANCE AEQUIEEWNTS j ^ 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within tre limits at leest once per 30 minutes during system i heatup, cooldo n, and inservice leak and hydrostatic testing operations. I i 4.4.9.1.2 The reactor vessel material irradiation surveillance specimens
- shall be removed and examined, to determine changes in material properties, as required by 10 CFR Part 50, Appendix H, in accordance with the schedule in Table 4.44. The results of these examinations shall be used to update l Figures 3.4 2 and 3.4-3.
J l l 1 j SOUTH TEXA5 - UNITS 1 & 2 3/4 4-31 Au!NDMENT N05. AND /. Y 4 5 a
, , _ . . _ _ . _ . _ . _ , . . . . . -----me~~,------n-
yo MATERIAL PROPERTY BASIS CONTROLLIN3 MATERIAL - RV RT tCT INITIAL: lO'F INTERMEO! ATE SHELL R-1606-3 RTNDT AFTER 32 EFPY COPPER CONTENT: CONSERVATIVELY t/4, gl'F ASSUMED AS O.10 VTX 3/4T, 64'F CURVE APPLICABLE FOR HEATUP RA"ES UP TO LOO'/HR FOA THE SERVICE PERIOD UP TO 32 EFPY 40 CONTAINS MAR 3!NS OF lo'F AND 60 PSIG FOR POSSIBLE INSTRUMENT ERRORS 3000 l l E # - g t eta TEST us:7 y / f,: 2000
\ l\ { l
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B ASED ON INSERv!CC WDR3ST AT!C TEST MATLP RATES y TE W AAT W E (242') LP 10 100' F/m N E
! WT
; 32 UPY l ' ' ' '
O O 100 200 3.0 400 INDICATED TEMPERATURE ('F) FIGURE 3.4-2 REACTOR COOLANT SiSTEM HEATUP LIMITATIONS - APPLICAoLE UP TO 32 EFPY SOUTH TEXA5 - UNITS 1 & 2 3/4 4-32 A.u,Eh?v.ENT N05. AND
FP HATERIAL PROPERTY BASIS CONTROLLING MATERIAL - RV RTNDT INITIAL: lO'T INTERMEDIATE SHELL R-1606-3 RTNOT AFTER 32 EFPY COPPER CONTENTS: CONSERVATIVELY 1/4, gl'T ASSUMEO AS 0.10 VTX 3/4T. 64'F SINGLE CURVE APPLICABLE FOR COOLOOWN RATES UP TO LOO */HR FOR , THE SERVICE PERIOD UP TO 32 EFPY, AND CONTAINS MAR 3!NS OF l IO'r AND 60 PS!G FOR POSS!5LE INSTRUMENT ERRORS 3000 ,
/
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'2000 i I , /l w l l l l l l [
B m l l /- E I f) B Q 1000 , lii
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I h ,_COMO.*4 R ATE 'F/HR (62i PSIG) I g e _ _l , _ _ g_ _ l ! lDI l l l l(120'F) l l O
.l 4 l I ! l l l l O 100 200 300 INDICATED TEFFERATURE (*F) i FIGURE 3.4-3 REA010R COOLANT SYSTEM C00LC n'i LIMITATIONS - AFFL! CABLE UP TO 32 EFh 3/4 4-33 A.MEh0 MENT hD5. ANO 4
SOUTH TEXA5 - UNITS 1 & c t.. . : : ...:
FC TABLE 4.4-5 REACTOR VESSE' MATERIAL SURVEILLANCE PROGRAM - WITHDRAWAL SCHEDULE CAPSULE VESSEL LEAD NUMBER LOCATION FACTOR WITH0RAWAL TIME (EFPY) U 58.5* 4.00 First Refueling Y 241' 3.69 5 V 61* 3.69 9 X 238.5' 4.00 15 W 121.5* 4.00 Standby Z '01.5' 4.00 Standby (' SD'JTH T EX A5 - UNIT 5 1 & 2 3/4 4 34 AuEN:YENT A05. A'O l.: . ; -_.
REACTOR COOLANT SYSTEM PRE 550312ER LIMITING CONDITION FOR OPERATION 3.4.9.2 The pressurizer temperature shall be limited to:
- a. A maximum neatup 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 621*F.
APPLICABILITY: At all 'imes. ACTIOP(: With the pressurizer terperature limits in excess of any of the above limits, restore the terrperature to wit?,in the limits within 30 minutes; perform an engineering evaluation t: <ctermine the effects of the out-of-limit condition on the f,tructural inty y of the pressurizer; determine that the pressurizer remains acce; table it entinued operation or be in at least HOT STANDBY within the next 6 hours and reduce the pressurizer pressure to less than 500 psig ! within the following 30 hours. JEYilU2CE_REMEEMENTC 4.4.9.2 The pressurizer temperatures shall be determined to be within the limits, at least on:e per 30 rninutes during system heatup or cooldown. The spray water tertperature differential shall be determined to be within the limit at least once per 12 hours during auxiliary spray operation, i i l J AN".
$0UTH TExA5 - UNIT 5 1 & 2 3/4 4-35 Av.EN:U.Ehi N:5
/. . . : ; , , -
FD REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS LIMITING C0i4DITION FOR OPERATION 3.4.9.3 At least one of the following Overpressure Protection Systems shall be OPERABLE:
- a. Two power-operated relief valves (PORVs) with lift settings which do not exceed the limit established in Figure 3.4-4, or
- b. The Reactor Coolant System (RCS) d1 pressurized with an RCS vent of greater than or equal to 2.0 square inches.
APPLICABILITY: MODES 4 and 5, and H0DE 6 with the reactor vessel head on. ACT10fi:
- a. With one PORV inoperable, restore the inoperable PORV to OPERABLE status within 7 days or depressurize and vent the RCS through at least a 2.0 square inch vent within the next 8 nours.
b With both PORVs inoperable, depressurize and vent the RCS through at least a 2.0 square inch vent within 8 hours,
- c. In the event either the PORVs or the RCS vent (s) are used to miti-gate 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 initi-ating the transient, the effect of the PORVs or RCS vent (s) on the-transient, and any corrective action necessary to prevent recurrence. ,
L 4 3/4 4-36 AMENDMENT N05. AND SOUTH TEXAS - UNITS- 1 & 2 l l
d 800 c ( ; - i B , f ; 4J e. l$700 I Wu, I IO db v I j
. @g 600 g zo )
x I
- _s
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[ I ! I I 500 50 100 200 300 400 TRTD - AUCTIONEERED LOW MEASURED RCS TEMPERATURE (*F) FIGURE 3.4-4 NOMINAL MAXIM'JM ALLOWABLE PORV SETPOINT FOR THE COLD OVERPRESSURE SYSTEM SOUTH TEXA5 - U$stIS 1 & 2 3/4 4-37 AP.ENDMENT N05. AND 1,'; . , . f
FL REACTOR COOLANT SYSTEM OVERPRESSURE PROTFCTION SYSTEMS . SURVEILLANLE REOUIREMENTS' 4.4.9.3.1 Each PORV shall be demonstrated OPERABLE by:
- a. Performance of an ANALOG CHANNEL OPERATIONAL TEST on the PORV ;
actuation channel, but excluding valve operation within 31 days ' prior to entering a condition in which the PORV Is required OPERMLE ' and at least once per 31 days thereafter when the PORV is require OPERABLE;
- b. Performance of a CHANNEL CALIBRATION on the PORV actuation channel at least once per 18 months; and
- c. Verifying the PORV block valve is open at least once per 72 hours 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 hours
- when tre vent (s) is being used for overpressure protection.
4.4.9.3.3 The positive displacement pump shall be demonstrated inoperable ** at least once per 31 days, except when the reactor vessel head is removed i or when both centrifugal charging pumps are U.soerable and secured, by verifying that the motor circuit breakers a e secured in the open position.*** l
*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.
- **The provisions of 3.0.4 and 4.0.4 are not applicable for entry into MODE 4 from MODE 3 for the positive displacertent pump declared inoperable pursuant to Specification 4.4.9.3.3 provided that the positive displacement pump is declared INOPERABLE within 4 hours after entry into MODE 4 from MODE 3 or prior to the temperature of one or more of the RCS cold legs decreasing below 325 F, whichever comes first.
*The positive displacement pump may be energized for testing provided the discharge of the pump has been isolated from the RCS by a closed isolation val'.1 with power removed from the valve operator, or by a manual isolation valve secured in the closed position.
SOUTH TEXAS - IINITS 1 & 2 3/4 4-38 AMENDMENT N05. AC
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1 FL ' l 1 REACTOR COOLANT SYSTEM 3/4.4.10 STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.4.10 The structural integrity of ASME Code Class 1, 2, and 3 components shall be maintained in accordance with Specification 4.4.10. APPLICABILITY: All MODES. ACTION:
- a. With the structur:1 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 conformir.g to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected corponent(s) prior to increasing the Reactor Coolant System temperature above 200 F.
- c. With the structural integrity of any ASME Code Class 3 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the af fected corrponent(s) from service.
MilLLAEE REQUIREMESU 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 Regulatory Position C.4.b of Regulatory Guide 1.14, Revision 1, August 1975. i l l SOUTH TE/AS - UNITS 1 & 2 3/4 4-39 AMENDMENT NDS. AND IS V j , g .,..
f.l REACTOR COOLANT SYSTEM 3/4.4.11 REACTOR VESSEL HEAD VENTS LIMITING CONDITION FOR OPERATION ,,,,, 3.4.11 Two reactor vessel head vent paths each consisting of two vent valves and a control valve powered from emergency busses shall be OPERABLE and closed. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:
- a. With one of the above reattor vessel head vent paths inoperable, i STARTUP and/or POWER OPERATION may continue provided the inoperable vent path is maintaired closea with power removed from the valve actuators of all the vent valves in the inoperable vent path; restore the inoperable vent path to OPERABLE status within 30 days, or, be in HOT STANDBY within 6 hours and in COLD SHUTOOWN within the following 30 hours,
- b. With two reactor vessel head vent paths inoperable, maintain the inoperable vent paths closed with power removed from the valve actua-tors of all the vent valves in the inoperable vent paths, and restore at least one of the vent paths to OPERABLE status within 72 hours or be in HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE RE0VIREMENTS 4.4.11 Each reactor vessel head vent path shall be demonstrated OPERABLE at least once per 18 months by:
- a. Verifying all anual isolation valves in each vent path are locked in the open position,
- b. Cycling each vent valve through at least one complete cycle of i full travel from the control room, and ;
- c. Verifying flow through the reactor vessel head vent pcths during venting.
I 1 e i 1 SOUTH TEXAS - UNITS 1 & 2 3/4 4-40 AMENDMENT N05. AND \ h1 . : ; ....
...a l
l FP ' 3/4.5 EMERGENCY CORE COOLING SYSTEMS i 3/4.5.1 ACCUMULATORS LIMITING CONDITION FOR OPERATION 3.5.1 Each Safety injection System accumulator shall be OPERABLE with: t
- a. The isolation valve open and pcuer removed,
- b. A contained borated water volume of between 8800 and 9100 gallons, ,
- c. A boron concentration of between 2400 and 2700 ppm, and
- d. A nitrogen cover pressure of between 590 and 670 psig.
APPLICABILITY: MODES 1, 2, and 3*. ACTION:
- a. With one accumulator inoperable, except as a result of a closed !
isolation vaive or the boron concentrction outside the required limits, restore the inoperable accumulator to OPERABLE status within I hour or be in at least HOT STANDBY within the next 6 hours and reduce pressurizer pressure to less than 1000 psig within the follow-irq 6 hours. i
, b. With one accumulator inoperable due to the isolation valve being i closed, either open the isolation valve within 1 hout or be in at least HOT STANDBY within the next 6 hours and reduce pressurizer pressure to less than 1000 psig within the following 6 hours.
, c. With the boron concentration of one accumulator outside the required
- 1 limit, restore the boron concentration to within the required limits within 72 hours or be in at least HOT STANDBY within the next 6 hours and reduce pressurizer pressure to less than 1000 psig within the j following 6 hours.
SEVil11ANCE REQUllE TS 4.5.1.1 Each accumulator shall be demonstrated OPEP.ABLE: .
- a. At least once per 24 hours by: l
- 1) Verifying, by the absence of alarms, the contained borated j water volume and nitrogen cover pressure in the tanks, and
- 2) Verifying that each accumulator isolation valve is open.
- b. At least once per 31 days and within 6 hours after each solution ,
volume increase of greater than or equal to 1% of tank volume by verifying the boron concentration of the accumulator solution; and
- Pressurizer pressure above 1000 psig. :
SOUTH TEXAS - UNITS 1 & 2 3/4 5-1 AMfNDMENT N05. AND l f l 1
- D3 ;
I l
Fo EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE RE0UIREMENTS (Continued)
.'. c. At least once per 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, 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 by the performance of an ANALOG CHANNEL OPERATIONAL T2ST, and
- b. At least once per 18 months by the performance of a CHANNEL CAllBRATION.
l l 1 l SOUTH TEXAS - UNITS 1 & 2 3/4 5-2 AMENDMENT N05. AND
. .C / r . .. .
A
- I t-/2 EMERGENCY CORE C01 LING SYSTEMS 3/4.5.2 ECCS SUBSYSTEMS - T,yg GREATER THAN OR EQUAL TO 350'F
_ LIMITING CONDITION FOR OPERATION 3.5.2 Three independent Emergency Core Cooling System (ECCS) subsystems shall be OPERABLE with each subsystem comprised of:
- a. One OPERABLE high Head Safety Injection pump,
- b. One OPE \BLE Low Head Safety Injection pump
- c. One OPERABLE RHR heat exchanger, and
- d. An OPERABLE flow path capable of taking suction from the refueling water storage tank on a Safety injection signal and automatically transferring suction to the containment sump during the recirculation phase of operation through a High Head Safety Injection pump and into the Reactor Coolant System and through a low Head Safety injec-tion pump and its respective RHR heat exchanger into the Reactor Coolant System.
APPLICABILITY: MODES 1, 2, and 3.* ACTION:
- a. With less than the above subsystems OPERABLE, but with at least two High Head Safety inject:en pumps in an OPERABLE status, two Low Head Safety injection pumps and associated RHR heat exchangers in an OPERABLE status, and sufficiert flow paths to accommodate these OPERABLE Safety Injection pumps and RHR heat exchangers, restore the inoperable subsystem (s) to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in HOT SHUT 00W'd within the following 6 hours.
- 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 Co~ mission 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 talue exceeds 0.70.
'The provisions of Specifications 3.0.4 and 4.0.4 are not applicable for entry into MODE 3 for the Safety injection pumps declared inoperable pursuant to Specification 4.5.3.1.2 provided that the Safety injection pumps are restored to OPERABLE status within 4 hours or prior to the temperature of one or more of the RCS cold legs exceeding 375'F, whichever comes first. SOUTH TEXAS - UNITS 1 & 2 3/4 5-3 AMEhCMENT NDS. AND h.. I 7 1553
FO EMERGEF0Y CORE COOLING SYSTEMS l SURVEILLANCE REOUIREMENTS 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE:
- a. At least once per 24 hours by verifying that the following valves are in the indicated positions witn power to the valve operators removed:
Valve Number Valve Function Valve Position XSIOOO8 A,B,C High Head Hot Leg Closed Recirculation Isolation XRH0019 A,B,C Low Head Hot Leg Closed Recirculation Isolation
- 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 locked, sealed, or otherwise secured in position, is in its correct position.
~
- c. By a visual inspection which verifies that no loose debris (rags, trash, clothing, etc.) is present in the containment which could be transported to the containment sump and cause restriction of the pump suctions during LOCA conditions. This visual inspection shall be performed:
- 1) For all accessible areas of the containment prior to establish-ing CONTAlhMENT 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 a visual inspection of the contain-rnent 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.
SOUTH TEXAS - UNITS 1 & 2 3/4 5-4 AMLNDMENT N35. AND I:: ; - ; *, ,
~
\
Pt 1 EMERGENCY CORE COOLING SYSTEMS 3/4.5.3 ECCS SUBSYSTEMS - T,yg LESS THAN 350*F LIMITING CONDITION FOR OPERATION 3.5.3.1 As a minimum, the following ECCS components shall be OPERABLE:
- a. Two OPERABLE High Head Safety Injection pumps,*
- b. TwoOPERABLELowHeadSafetyInjectionpumpsandtheirassociated RHR heat exchangers, and
, c. Two OPERABLE flow paths capable of taking auction from the refueling water storage tank upon being manually realigned and transferring suction to the containment sump during the recirculation phase of operation through a High Head Safety Injection pump and into the ReactorCoolantSystemandthroughaLowHeadSafetyInjectionpump t and its respective RHR heat exchanger into the Reactor Coolant Systi:m.
APPLICABILITY: MODE 4. ACTION:
- a. With less than the above-required ECCS components OPERABLE because of the inoperability of either the High Head Safety Injection pumps or the flow paths from the refueling water storage tank, restore at least the required ECCS components to OPERABLE status within 1 hour or be in COLD SHUTDOWN within the next 20 hours,
- b. With less than the above-required ECCS components OPERABLE because of the inoperability of either the residual heat removal heat exchangers ortheLowHeadSafetyinjectionpumps,restoreatleasttherequired ECCS components to OPERABLE status or maintain the Reactor Coolant System T,yg less than 350 F by use of alternate heat removal methods.
- c. IntheeventtheECCSisactuatedandinjectswaterintotheReactor Coolant System, a Special Report shall be prepared and submitted tc the Com-ission pursuant to Specification 6.9.2 within 50 days describ-3 ing the circumstances of the actuation and the total accumulated actuation cycles to date. The current value of the usage factor for eachaffectedSafetyinjectionnozzleshallbeprovidedinthis
- Special Report whenever its value exceeds 0.70.
l i
*A maximum of one High Head Safety injection pump shall be OPERABLE and & second i High Head Safety injection pump shall be OPERABLE except that its breaker shall be racked out (the third HHSI pump shall have its breaker racked out) within:
! (1) 4 hours after entering MODE 4 from MODE 3 or prior to the temperature of j one or morr of the RCS cold legs decreasing below 325'F, whichever comes first; i or (2) 4 hours after entciing MDDE 4 from MODE 5 or prior to the temperature
- of one or more of the RCS cold legs exceeding 225'F, whichever comes first.
SOUTH TEXAS - UNITS 1 & 2 3/4 5-6 AMENDMENT NDS. AND l I'.'.' 1 ; 3 ,3
Fl EMERGENCY CORE C00LfNG SYSTEMS ECCS SUBS,15,TEMS_- T,y LESS THAN 350'F SURVEILLANCE REOUIREMENTS 2 4.5.3.1.1 The ECCS components shall be demonstrated OPERABLE per the applicable requirements of Specification 4.5.2. 4.5.3.1.2 All High Head Safety Injection pumps, except the above allowed OPERABLE pumps, shall be demonstrated inoperable
- by verifying that the motor circuit breakers are secured in the open position within 4 hours after entering MODE 4 from H0DE 3 or prior to the te.nperature of one or more of the RCS cold .
legs decreasing below 325*F, whichever comes first, and at least once per 31 days thereafter. t i 1 . I I
; l
! I ! l I i. r
- I s f
*An inoperable pump may be energized for testing or for filling accumulators t l
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. l T
i 3/4 5 7 AMENDMENT N05. AND ! l SOUTH TEXAS - UNITS 1 & 2 ' t s . , a 155b l
f~ l ' i I l EMERGENCY CORE COOLING SYSTCMS ECCS SUBSYSTEMS - T,y LESS THAN OR EQUAL TO 200*F LIMITING CONDITION FOR OPERATION l 3.5.~.2 All High Head Safety Injection pumps shall be inoperable. APPLICABILITY: MODE 5 and MODE 6 with the reactor vessel head on. ACTION: With a Safety Injection pump OPERABLE, restore all High Head Safety Injection pumps to an inoperable status within 4 hours. SURVEILLANCE REOUIREMENTS 4.5.3.2 All High Head Safety Injection pumps shall be demonstrated inoperable
- by verifying that the .sator circuit breakers are secured in the open position at least once per 31 days.
i
*An inoperable pomp may be energized for testing or for filling accumulators 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 6 manual i isolation valve secured in the closed position.
3/4 5-8 AMEN 0 MENT N05. AND SOUTH TEXAS - UNITS 1 & 2
'- : i .;,;
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Fi EMERGENCY CORE COOLING SYSTEMS 3/4.5.4 (This specification number is not used.) SOUTH TEXAS - UNITS I & 2 3/4 5-9 AMEt0 MENT ?CS. AND b
Pi EMERGENCY CORE COOLING SYSTEMS 3/4.5.5 REFUELING WATER STORAGE TANK i LIMITING CONDITION FOR OPERATION 3.5.5 The refueling water storage tank (RWST) shall be OPERAP8.E with:
- a. A minimum contained borated water volume of 458,000 gallons, and ,
- b. A boron concentration between 2500 ppm and 2700 ppm.
APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With the RWST inoperable, restore the tank to OPERABLE status within I hour or be in at least HOT STANDBY within 6 hours and in COLD SHUTOOWN within the following 30 hours. SEPVEILLANCE RE0VIREMENTS 4.5.5 The RWST shall be demonstrated OPERABLE at least once per 7 days by:
- a. Verifying the contained borated water volume in the tank, and
- b. Verifying the boron concentration of the water.
A SOUTH TEXA5 - UNITS 1 & 2 3/4 5-10 AMENOMENT N05. ANO
- l. (
1 Ff f EMERGENCY CORE COOLING SYSTEMS 3/4.5.6 RESIDUAL HEAT REMOVAL (RHR) SYSTEM i1MITING CONDITION FOR OPERATION 3.5.6 Three independent Residual Heat Removal (RHR) loops shall be OPERABLE with each loop comprised of:
- a. One OPERABLE RHR pump,
- b. One OPERABLE RHR heat exchanger, and
- c. One OPERABLE flowptth capable of taking suction from its associated RCS hot leg and discharging to its. associated RCS cold leg
- APPLICABILITY: MODES 1, 2 and 3.
ACTION: ,
- a. With one RHR loop inoperable, restore the required loop to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours ar in HDT SHUTDOWN within the following 6 hours,
- b. With two RHR loops inoperable, restore at least two RHR loops to OPERABLE status within 24 hours or be in at least HOT STANDBY within 6 hours and in HOT SHUTDOWN within the following 6 hours.
- c. With three RHR loops inoperable, immediately initiate corrective action to restore at least one RHR loop to OPERABLE status as soon as possible.
SURVEILLANCE RE0VIDEMENTS _ _ _ . . _ . . _ _ _. . _ _ . _ _ . _ . _ __ . _ _ , . 4.5.6.1 Each RHD loop shall be demonstrated OPERABLE pursuant to the require-ments 'f Specification 4.0.5. 4.5.6.2 At least once per 18 months by verifying automatic isolation and inter-lock action of the RHR system from the Reactor Coolant System to ensure that:
- a. With a simulated or actual Reactor Coolant System pressure signal greater than or equal to 350 psig, the Interlocks prevent the valves from being opened, and
- b. With a simulated or actual R:*. tor Coolant System pressure signal less than or equal to 700 psig, the interlocks will cause the valves to automatically close.
"Valves May-0060 A, B, and C and MOV-0061 A, B, and C may have power removed to support the FHAR (Fire Huarri Atialysis Report) assumptiens.
SOUTH TEXAS - UN!($ 1 & 2 3/4 5-11 AkENDMENT N05. AND g , 1 , V.i _- - ________ _ ____ _ D
FD 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT CONTAINMENT INTEGRITY , LIM 1IlELIQEDlIl0h'. FOR OPERATION 3.6'.1.1 Primary CONTAINMENT INTEGRITY shall be maintained. APPLICABILITY: MODES 1, 2, 3, and 4. , ACTION: , Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY within I 1 t. cur or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SUR/EILLANCE IEOUIREMENTS , t 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 automttic isolation valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic valves secured ir their positions, except as provided in Specification 3.6.3;
- b. By verifying that each containment air lock is in compliance with the requirements of Specification 3.6.1.3; and
- c. After each closing of each penetration subject to Type B testing, except the containment air locks, if opened following a Type A or B test, by leak rate testing the s;al with gas at a pressure not less than P,, 37.5 r.:'g, and verifying that when the measured leakage rate ,
for these seals is added to the leakage rates determined pursuant to Specification 4.6.1.2d. for all other Type B and C penetrations,
- the combined leakage rate is less than 0.60 L,. <
i I i
*Except valves, blind flanges, and deactivated automatic valves which are ,
located inside the containment and are locked, sected or otherwise secured in the closed position. These penetrations shall be verified closed during each COLD SHUT 00WN ex ept that such verification need not be performed more often than once per 92 days. SOUTH TEXAS - UNITS 1 & 2 3/4 6-1 AMENDMENT NDS. AND I
- - - _ . - _ . . - . . . ., , .. . . - . _ _ . . , . _ _ . _ , _ _ . ~
$Q CONTAINMENT SYSTEMS CONTAINMENT LEAKAGE LIMITING CONDITION FOR QPERATION
- 3. 6.1. 2 Containment leakage rates shall be limited to:
- a. An overall integrated leakage rate of less than or equal to L,, 0.30%
by weight of the containment cir per 24 hours at P,, 37.5 psig.
- b. A combined leakage rate of less than 0.60 L, for all penetrations and valves subject to Type B and C tests, when pressurized to P,.
APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With either the measured overall integrated containment leakage rate exceeding 0.75 L, or the measured cctbined leakage rite for all penetrations and valves subject to Types B and C tests exceeding 0.60 L , restore the overall inte-grated leakage rate to lecs than 0.75 L, and the combined leakage r&te for all penetrations subject to Type B and C tests to less than 0.60a L prior to in-creasing the Reactor Coolant Systen temperature above 200 F. SURVE1LLEEJEQVIFEML___ _ _ 4.6.1.2 The containmeat leakage rates shall be demonstrated at the following test schedule and shall be determined in conformance with the criteria speci-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 Irtegrated Containment Leakage Rate) shall be conducted at 40 1 10 month intervals during shutdown at a pressure not less than P,, 37.5 psig, during each 10 year service period. The third test of each set shall be conducted during the shutuce for the 10 year plant inservice inspection; ,
SOUTH TEXA5 - UNITS 1 & 2 3/4 6-2 AMENDMENT N?5. AND
FD CONTAINMENT SYSTEMS SURVEILLANCE REOUIREMENTS (Continued)
- b. If any periodic Type A test fails to meet 0.75 L,, the test schedule for subsequent Type A tests shall be reviewed and approved by the Commission. If two consecutive Type A tests fail to meet 0.75 L,, I a Type A test shall be performed at least every 18 months until two consecutive Type A tests meet 0.75 L, at which time the above test l schedule may be resumed; ,
I
- c. The accuracy of each Type A test shall 3e verified by a supplemental test which:
- 1) Confirms the accure.cy of the test by verifying that the supple- j mental test result, L c, is in accordance with the following -
equation: lLc '(l am *l o)l 10.25 L, , where L am is the measured Type A test leakage and gL is the superimposed leak;
- 2) Has a duration sufficient to establish accurately the change in leakage rate between the Type A test and the supplemental test; and
- 3) Requires that the rate at which gas is injected into the contain-ment or bled from the containment during the supplemental test is between 0.75 L, and 1.25 L,.
- d. Type B and C tests shall be conducted with gas at a pressure not less than P,, 37.5 psig, at intervals no greater than 24 months except for tesis involving:
- 1) Air locks, ,
- 2) Purge supply and exhaust isolation valves with resilient material seals, and
- 3) Penetrations using continuous Leakage Monitoring Systems. (
1
- e. Air locks shall be tested and demonstrated OPERABLE by the require- !
l ments of Specification 4.6.1.3; - i
- f. Purge supply and axhaust isolation valves with resilient material ,
seals shall be tested and demonstrated OPERABLE by the requirements ! of Specification 4.6.1.7.2 or 4.6.1.7.3, as applicable; ; 4 SOUTH TEXAi - UNITS 1 l. 2 3/4 6*3 AMEN 0 MENT N05. AND
..:'; .g
---e.
J
, 1 CONTAINMENT SYSTEMS SURVEILLANCE REOUIREMENTS (Continued) __
- g. Leakage from isolation valves that are sealed with fluid from a Seal System may be excluded, subject to the provisions of Appendix J, Section III.C.3, when determining the combined leakage rate provided the Seal System and valves are pressurized to at least 1.10 P,,
41.25 psig, and the seal system capacity is adequate to maintain system pressure for at least 30 days;
- h. Type B tests for penetrations employing a continuous Leakage Monitor-ing System shall be conducted at P greater than once per 3 years; and,, 37.5 psig, at intervals no
- i. The provisions of Specification 4.0.2 are not applicable.
1 i i J t 1 SOUTH TEXAS - UNITS 1 & 2 3/4 6-4 AMENDMENT N05. AND
. . . !
- UE3 i
FD ; i CONTAINMENT SYSTEMS CONTAINMENT AIR LOCKS gM8 FOR OPERATION
- 3. 6.1. .! Each containment air lock shall be OPERABLE with:
- a. Both doors closed except wnen the air lock is being used for normal transit entry and exit through the containment, then at least ene air lock door shall be closed, and
- b. An overall air lock leakage rate of,less than or equal to 0.05 L, at P,, 37.5 psig.
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 hours or lock the OPERABLE air lock door closed;
- 2. Operation nay then continue until performance of the next required overall air lock leakage test provided that the OPERABLE 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 hours and in COLD SHUTDOWN within the following 30 hours; 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 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTOOWN within the following 30 hours.
SOUTH TEXAS - UNITS 1 & 2 3/4 6-5 AMENDMENT NOS. ANS
CONTAINMENT SYSTEMS SURVEILLANCE REOUIREMENTS 4.6.1.3 Each containment air lock sheli be demonstrated OPERABLE: i
- a. Within 72 hours fo'. lowing each closing, except when the air lock is i being used for multiple entries, then at least once per 72 hours, by verifying seal leakage is less than 0.01 L, as determined by precision flow measurements when measured for at least 30 seconds with the volume between the seals at a constant pressure of 37.5 psig;
- b. By conducting overall air lock leakage tests at not less than P,, ,
37.5 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 capabi'.ity **
- c. At least once per 6 months by verifying that only one door in each air lock can be opened at a time.
- d. By verifying at least once per 7 days that the instrument air pres-sure in the header to the perrionnel airlock seals is > 90 psig,
- e. By verifying the door seal pneumatic system OPERABLE at least once per 18 months by conducting a seal pneumatic system leak test and verifying that system pressure does not decay more than 1.5 psi from 90 psig minimum within 24 hours.
*The provisions of Specification 4.0.2 are not applicarle.
l "This represents an exemption to Appendix J, paragraph 111.D.2 of 10 CFR Part 50. SOUTH TEXA5 UNIis 1 & 2 3/4 6-6 AMEN 0 MENT N05. AND l.. 1 i h.3
FD CONTAINMENT SYSTEMS INTERNAL PRES 5URE 4 LIMITING CONDITION FOR OPERATION 3.6.1.4 Primary containment internal ,eressure shall be maintained between
-0.1 and +0.3 psig. I APPLICABILITY: ;40 DES 1, 2, 3, and 4.
ACTION: With the containment iriternal pressure outside of the limits above, restore the internal pressure to within thy 'simits within I hour or be in at least HOT STANDBY within the next 6 hours asc in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REOUIREMENTS 4.6.1.4 The primary containment internal pressure shall te determined to be within the limits at least once per 12 hours. 4 i l i i i ? n ' u a
. l
> l' i i I I i SOUTH TEXAS - UNITS 1 & 2 3/4 6 7 AMENDMENT NDS. AND l Li I i li!: I i ,
CONTAINMENT SYSTEMS AIR TEMPERATURE , LIMITING CSNDITION 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 t.he average air temperature to within the limit within 8 hours, or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTOOWN within the following 30 hours. SURVEILLANCE RE0VIREMENTS
- 4. 6.1. 5 The primary containment average air temperature shall be the arith-metical average of a minimum of four RCCC inlet temperatures and shall be determined at least once per 24 hours. .
i 4 a SOUTH TEXAS - UNITS 1 & 2 3/4 6-8 AMENDMENT NOS. AND
. J i 7 12:3
t ~FP CONTAINMENT SYSTEMS CONTAINMENT STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.6 The structural integrity of the containment shall be maintained at a 7 level consistent with the acceptance criteria in Specification 4.6.1.G. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:
- a. With more than one tendon with an observed lift-off force between the predicted lower limit and 90% of the predicted lower limit or with one tendon below 90% of the predicted lower limit, restore the tendon (s) to the required level of integrity within 15 days and pttform an engineering evaluation of the containment and provide a Special Report to the Commission within 30 days in accordance with Specification 6.9.2 or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- b. With any abnormal degradatien of the structural integrity other than ACTION a. at a level below the acceptance criteria of Specification 4
4.6.1.6 restore the containment to the required level of integrity withia 72 h]urs and perform an engineering evaluation of the contain-ment and provide a Special Report to the Commission within 15 days in accordance with Specification 6.9.2 or be in at least HOT STANDBY ! within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. i SURVEILLANCE REOUIREMENTS 4.6.1.6.1 Containment Tendons. The containment tendons' structural j integrity shall be demonstrated at the end of 1, 3, and 5 years following the initial containment structural integrity test and at 5 year intervals thereafter. The tendons' structural integrity shall be demonstrated by:
- a. Determining that a random but representative sample of at least 13 tendons (4 inverted U and 9 hoop) each have an observed lift-off force within predicted limits for each. For each subsequent inspec-
. tion, one tendon from each group may be kept unchanged to develop a ristory anc' to correlate the observed data If the observed lift-off force of any one tendon in the original sample population lies between the predicted lower limit and 90% of the predicted lower limit, two tendons, one on each side of this tendon, should be checked for their lif t-of f forces. If both of these adjacent tendors
- are found to be within their predicted limits, all three tendons should be restored to the required level of integrity. This single deficiency may be considered unique and acceeptable. Unless there is i abnortral degradation of the containment during the first three inspections, the sample population for subsequent inspections shall include at least 5 tendons (2 inverted U and 3 boop);
l SOUTH TEXAS - UNITS 1 & 2 3/4 6-9 AMENDMENT N35. AND r l
R CONTAINMENT SYSTEMS SURVEILLANCE REOUIREMENTS (Continued)
- b. Perfortning tendon detensioning, inspections, and material tests on a previously stressed tendon from each group (invertea U and hoop).
A randomly selected tendon from each group shall be completely detensioned in order to identify broken or damaged ' fires and deter-mining that over the entire length of the removed wire that:
- 1) The tendon wires are free of corrosion, cracks, and damage,
- 2) There are not changes in the presence or physical appearance of the sheathing filler greast,, and
- 3) A minimum tensile strength of 240,000 psi (guaranteed ultimate strength of the tendon material) for at least three wire samples (one from each end and one at mid-length) cut from each removed wire. Failure of any one of the wire samples to meet the mini-mum tensile strength test is evidence of abnormal degradation of the containment structure.
- c. Performing tendon retensioning of thost ter. dons detensioned for inspection to their observed lift-off force with a tolerance limit of +6%. During retensioning of these t waons, the changes in load and elongation should be measured simultaneously at 20%, 60%, and 100%
of the maximum jacking force. If the elongation corresponding to a specific load differs by more than 5% from that recorded during installation, an investigation should be made to ensure that the difference is not related to wire failures or slip of wires in anchorages;
- d. Assuring the observed lift-off stresses exceed the average minimum l design value given below, which are adjusted to account for elastic and time dependent losses; and i inverted U 126 ksi Hoop: Cylinder 128 ksi Dome 123 ksi
- e. Verifying the OPERABILITY of the sheathing filler grease by:
i
- 1) No voids in excess of 5% of the net duct volume, I 2) Minimum grease coverage exists for the different parts of the anchorage system, and
- 3) The chemical properties of the filler material are within the tolerance limits as specified by the manufacturer.
i I SOUTH TEXAS - UNITS 1 & 2 3/4 6-10 AMEN 0 MENT N05. AND
! ?.' . .
fDl CONTAINMENT SYSTEMS SURVEILLANCE RE001REMENTS (Continued) 4.6.1.6.2 End Anchorages and Adjacent Concrete Surfaces. The structural integrity of the end ancnorages of all tendons inspected pursuant to Specifi-cation 4.6.1.6.1 and the adjacent concrete surfaces shall be demonstrated by determining through inspection that no apparent changes have occurred in the visual appearance of the end anchorage or the concrete crack patterns adjacent to the end anchorages. Inspections of the concrete shall be performed during the Type A containment leakage rate tests (reference Specification 4.6.1.2) while the containment is at its maximum test pressure.
- 4.6.1.6.3 containment Surfaces. The structural integrity of the exposed accessible interior and exterior surfaces of the containment, including the liner plate, shall be determined during the shutdown for ch Type A contain-ment laakage rate test (reference Specification 4.6.1... sy 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 abncrmal degradation, t
1 i I l ! l ! t i t l i e I j SOUTH TEXA5 - UNITS 1 & * '/4 6-11 AMEN 0 MENT NC5. AND ,. ,
*. .. ,,e
FD CONTAINMENT SYSTEMS CONTAINMENT VENTILATION SYSTEM i LIMITING CONDITION FOR OPERATION __ 3.6.1.7 Each containment purge supply and exhaust isolation valve shall be OPERABLE and:
- a. Each 48-inch containment shutdown purge supply and exhaust isola-tion valve shall be closed and sealed closed, and "
- b. The 18-inch supplementary containment purge supply and exhaust ,
isolation valves shall be closed to the maximum extent practicable but may be open for supplementary purge system operation for pres-sure control, for ALARA and respirable air quality considerations for personnel entry and for surveillance tests that require the . valves to be open. ! APPLICABILITY: MODES 1, 2, 3, and 4. { j ACTION:
- a. With a 48-inch .ontainment purge supply and/or exhaust isolation valve open or not sealed closed, close and/or seal close that valve l or isolate the penetration (s) within 4 hours, otherwise be in at i
least HOT STANDBY within the next 6 hours and in COLD SHUTOOWN , wittiin the following 30 hours. !
- b. With the 18-inch supplementary containment purge supply and/or exhaust isolation valve (s) open for reasons other than given in ;
Specification 3.6.1.7.b. above, close the open 18-inch valve (s) or isolate the penetration (s) within 4 hours, otherwise be in at least . HOT STANDBY within the next C hours, and in COLD SHUTOOWN within the following 30 hours, i
- c. With a containment purge supply and/or exhaust isolation valve (s) l having a measured leakage rate in excess of the limits of Specife I cations 4.6.1.7.2 and/or 4.6.17.3, restore the irioperable valve (s)
] to OPERABLE status or isolate the penetrations so that the measured leakage rate dogs not exceed the limits of Specifications 4.6.1.7.2 l and/or 4.6.1.7.3 within 24 hours, otherwisk be in at least HOT ! STANDBY within the next 6 hours, and in COLD SHUTOOWN within the t i following.30 hours. l > i
- I 1
e ! SOUTH TEXAS - UNITS 1 & 2 3/4 6-12 AMEN 0 MENT N05. AND
- a
F T-CONTAINMENT SYSTEMS SURVEILLANCE REOUIREMENTS _ 4.6.1.7.1 Each 48-inch containment ctrge supply and exhaust isolation valve shall be verified to be sealed closed at least once per 31 days. 4.6.1.7.2 At least once per 6 months on a STAGGERED TEST BASIS, the inboard and outboard isolation valves with resilient snaterial seals in each sealed closed 48-inch containment purge supply and exhaust penetration shall be demonstrated OPERABLE by verifying that the measured leakage rate is less than 1 0.05 L, when pressurized to P,. 4.6.1.7.3 At least once per 3 months each 18-inch supplementary containment purge supply and exhaust isolation valve with resilien.t material seals shall be demonstrated OPERABLE by verifying that the measured leakage rate is less than 0.01 L, when pressurized to P,. 4.6.1.7.4 At least once per 31 days each 18-inch supplementary contcinment purge supply and exhaust isolation valve shati Lc verified to be closed or open in accordance with Specification 3.6.1.7.b. i i a l SOUTH TEXA5 - UNITS 1 & 2 3/4 6-11 AMEN 0 MENT N05. AN *, I- : ;,,;
x FC
, CONTAINMENT SYSTEMS 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT SPRAY SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.1 Three independent Containment Spray Systems shall be OPERABLE with each
- Spray System capable of taking suction from the RWST and transferring suction to the containment sump.
APPLICABILITY: MODES 1, 2, 3, And 4. ACTION: With one Containment Spray Syst o inoperaDie, restore the inoperable Spray System to OPERABLE status within 72 hours or be in at least HOT STANDBY within the ne. 6 hoars; restore the inoperable Spray System to OPERABLE status within ..e ned. 48 hours or be in COLD SHUTDSWN within the following 30 hours. SURVEILLANCE REOUIREMENTS - 4.6.P.1 Ee.h Containment Spray Systea shall be demonstrated OPERABLE:
- 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. By verify'ng, that on recirculation flow, each pump develops a i
differential pressure of greater than or equal to 283 psid when tested pursuant to Specification 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 pocition on a Contr.inment Pressure High 3 test i
signal, and
- 2) Verifying that each spray pump : tarts automatically on a Contain-rent Pressure High 3 test signal coincident with a sequencer start signa'.
- d. At Itast once per 5 y2ars by performing an air or smoke flow test through each spray header and verifying each spray no:zle is
, unobstructed.
i 50t'N TEXAS o UN.th 3/4 6-14 AMEN 0 MENT N05. AND iD 1i E:
- . r, - - -,,, . . - - . - - - - , ,
F-CONTAI'iMENT SYSTEMS _ SPRAY ADDITIVE SYSTEM l l LIMITI4G CONDITION FOR OPERATION 3.6.2.2 The Spray Additive System shall be OPERABLE with:
- a. Three spray additive tanks each containing a volume of between 1061 and 1342 gallons of between 30 and 32% by weight NaOH solution, and
- b. Three spray additive eductors each capable of adding NaOH solution from its associated spray additive tank to its Containment Spray System pump flow.
APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With the Spray Additive System inoperable, restore the system to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours; restore the Spray Additive System to OPERABLE status within the next 48 hours or be in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REOUIREMENTS 4.6.2.2 The Spray Additive System shall be demonstrated OPERABLE: a. At least once per power-operated, 31 days by) verifying that each valve (manual,in the flow path that i or automatic 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 each spray additive 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 Pressure High 3 test signal; and SOUTH TEXAS - UNITS 1 & 2 3/4 6-15 AVENOMENT N05. AND i
fC CONTAINMENT SYSTEMS _SURYIILLANCE RE0VIREMENTS (Co tinued)
- d. At least once per 5 years by ve ifying:
- 1) Each eductor suction flow rate is greater than or equal to 30 gpm using the RWST as the test source to the eductor inlet, 'ad under the following conditions:
a) CS pump suction pressure is > 15 psig, b) Valve C50019A, B, or C, as applicable, is in the full open position, and c) CS pump recirculation flow rate to the RWST is 800 gpm i 100 gpm.
- 2) The lines t'etween the spray aoditive tank and the eductors are not blocked by verifying flow.
SOUTri TEXAS - UNITS 1 & 2 3/4 6-16 AMENDMENT NDS. AND
fi i CONTAINMENT SYSTEMS CONTAINMENT COOLING SYSTEM M M OR OPERATION 3.6.2.3 hree independent groups of Reactor Containment Fan Coolers (RCFC) shall be OPERABLE with a minimum of two units in two groups and one unit in the , third group. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With one group of the above required Reactor Containment Fan Coolers inoperable, restore the inoperable group of RCFC to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hoers. ! SURVEILLANCE REOUIREMENTS j 4.6.2.3 Each group of Reactor Containment Fan Coolcrs shall be demonstrated OPERAFLE:
- a. At least once per 31 days by:
, 1) Starting each non-operating fan group from the control room, and i verifying that each fan group operates for at least 15 minutes, i and
- 2) Verifying a cooling water flow rate of greater than or equal to 550 gpm to each cooler.
- b. At least once per 18 months by verifying that each fan group starts i automatically on a Safety Injection test signal.
l l i l l a SOUTH TEXAS - UNITS 1 & 2 3/4 6-17 AMEN 0 MENT N05. AND . .i
\..g 1
R i CONTAINMENT SYSTEMS 3/4.6.3 CONTAINMENT IS0lATION VALVES LIMITING CONDITION FOR OPERATION 1 3.6.3 The containment isolation valves shall be OPERABLE with isolation times , less than or equal to the required isolation times. l APPLICABILITY: MODES 1, 2, 3, and 4. l ACTION: , With one or more of the isolation valve (s) inoperable, maintain at least one ; isolation valve OPERABLE in each affected penetration that is open and; j i
- a. Restore the inoperable valve (s) to OPERABLE status within 4 hours, ;
or
- b. Isolate each affected penetration within 4 hou*s by use of at least one deactivated actomatic valve secured in the isolation position, or
- c. Isolate each affected penetration within 4 hours by use of at least ,
one closed manual valve or blind flange, or
- d. Be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. ;
l SURVEILLANCE REOUIREMENTS 4.6.3.1 The isolation valves shall be demonstrated OPERABLE prior to returning the v31ve to service after maintenance, repair or replacement work is performed on the valve or its associated actuator, control or power circuit by perform-4 ance of a cycling test, and verification of isolation time. . 4.6.3.2 Each isolation valve shall be demonstrated OPERABLE during the COLD i i SHUTDOWN or REFUEL!N3 MODE at least once per 18 Lonths by:
- a. Vn tiving that on a Phase "A" Isolation test si0nal, each Phase "A" i isvietion valve actuates to its isciacion position; ,
l b. Verifying that on a Containment Ventilation Isolation test signal, ! each purge and exh'ust valve actuates to its isolation position; and f t
- c. 'lerifying that on a Phase "B" Isolation test signal, err.h Phase "B" isolation valve actuates to its isolation position.
- 4.G.3.3 The isolation ti,ne of each power-operateri or automatic vai<e shall be
' determined to be within its limit when tested pursuant to Specificotion 4.0.5. i i 3/4 6-10 AMENDMENT N05. AND
'30tTH TEXAS - UNITS 1 & 2 t, n u l
1 l w .
b CONTAINMENT SYSTEMS 3/4.6.4 COMBUSTIBLE GAS CONTROL HYDROGEN ANALYZERS LIMITING CONDITION FOR OPERATION 3.6.4.1 Two independent containment hydrogen analyzers shall be OPERABLE. APPLICABILITY: H0 DES 1 ar.d 2. ' ACTION:
- a. With one hydrogen analyzer inoperable, restore the inoperable analyzer to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 6 hours, i l
- b. With both hydrogen analyzers inoperable, restore at least one analyzer to i
OlCRABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours.
- SURVEILLANCE REOUIREWINTS i *a "==
4.6.4.1 Each hydrogen analyzer shall be demonstrated OPERABLE by the pe-form- , ance of a CHANNEL CHECK at least once per 12 hourt, an ANALOG CHANNEL ' OPERATIONAL TEST at least once per 31 days, a channel OPERABILITY verification , at least once per 92 days on a STAGGERED TEST BASIS using sample gas contain- - ing one volume percent hydrogen, balance nitrogen, and by performing a CHANNEL CALIBRATION at least once per 18 months using sample gas containing ten volume t percent hydrogen, balance nitrogen. l T I 0 l j SOUTH TEXAS - UNITS 1 & 2 3/4 6-19 AMENCMiNT N05. AND l l -, : :;3
n FP CONTAINMENT SYSTEMS ELECTRIC HYDROGEN RECOMBINERS
- LIMITING CONDITION FOR OPERATION 3.6.4.2 Two independent Hydrogen Recombinor Systems shall be OPERABLE.
i APPLICABILITY: MODES 1 and 2. i ACTION:
; With one Hydrogen Recombiner System inoperable, restore the inoperable system i to OPERABLE status within 30 days or be in at least HOT STANDBY within the l next 6 hours.
Ely11tLANCE RE0SIREWINTS , l 4.6.4.2 Each Hydrogen Recombiner System shall be demonstrated OPERABLE: *
- a. At least once per 6 months by verifying, during a Hydrogen Recombiner ,
System functional test, that the minimum heater sheath temperature ' increases to greater than or equal to 1000'F within 90 minutes at ! 52 kW. Upon reaching 1000*F, increase the power svtting to maximum
- power for 2 minutes and verify that the power meter reads greater
- than or equal to 65 kW, and l l
- b. At least once per 18 months by. .
l
- 1) Performing a CHANNEL CALIBRATION of all recombiner instrun.enta- !
l tion and control circuits, j
- 2) Verifying through a visual examination that tnere is no !
evider.ce of abnormal conditions within the recombiner enclosure ! (i.e., loose wiring or structural connections, deposits of j foreign materials, etc.), and , i Verifying the integrity of all heater electrical circuits by 3) performing a resistancc to ground test following the above ; required functional test. The resistance to gropnd for any [ heater' phase shall be greater than or equal to 10,000 ohms. j l l i l e t SOUTH TEXAS - UNITS 1 & 2 3/4 6-20 AMENMENT N05. AND 1.: . ; g f 1
f:$> 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE SAFETY VALVES LIMITING CONDITION FOR OPERATION 3.7.1.1 All main steam line Code safety valves associated with each steam generator shall be OPERAELE with lift settings as specified in Table 3.7-2. APPLICABILITY: MODES ), 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 I within 4 hours, either the inoperable valve is restored to OPERABLE !
status or the Power Range Neutron Flux High Trip Setpoint is reduced per Table 3.7-1; otherwise, be in at least HOT STANDBY within the l next S hours and in COLD SHUTDOWN within the following 30 hours. i
- b. The provisions of Specification 3.0.4 are not applicable. ,
EURVEILLANCE REOUIREMENTS ___ l 4.7.1.1 There are no additional requirements other than those required by Specification 4.0.5. 1 i l ) ! i l 1 I I t SOUTH TEXAS - UNITS 1 & 2 3/4 7-1 AMENDMENT N05. AND i I- I 7 ?E; ! t
- i. .--_ -
__________.I
FD: TABLE 3.7-1 MAX 1 MUM ALLOWABLE POWER RAnE NEUTRON FLUX HIGH SETPOINT WITH l INOPERABLE STEAM LINE 5AFETY VALVE 5 DURING 4 LOOP OPER M i j i 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 37 ! 2 65 i i 3 43 i i I [ i I l' l r t c 1 F i i ! i l J
- l 1
J i J l l i i SOUTH TEXA5 - UNITS 1 & 2 3/4 7-2 AMENDMENT N05. AND t;r, t W..
FP TABLE 3.7-2 STEAM LINE SAFETY VALVES PER LOOP VALVE NUMBER LIFT SETTING (t 1%)* OkIFICE SIZE LOOP A LOOP B LOOP C LOOP O
- 1. 7T67410 F5FT420 PTF7430 PTF7440 1285 psig 16 in.2
- 2. PSV-7410A PSV-7420A PSV-7430A PSV-7440A 1235 psig 16 in.2 1 3. PSV-74108 PSV-7420B PSV 7430B PSV-7440B 1505 psig -
16 in.2
- 4. PSV-7410C PSV-7420C PSV-7430C PSV-7440C 1315 psig 16 in.2
- 5. PSV-74100 PSV-74200 PSV-7430D PSV-74400 1325 psig 16 in.2 i
.1 I i l f i a f 1 - ) l l 1 *The lift setting pressure shall correspond to ar.bient conditions of the valve at reminal operating temperature and pressure. 1 3/4 7-3 AMEN 0 MENT N05. AND i SOUTH TEXA5 - Uhl151 & 2 i
~ FL PLANT !YSTEMS AUXILIARY FEE 0 WATER SYSTEM ypl*JG CONDITION FOR OPERA 110N 3.7.1.2 At least four independent steam generator auxiliary feedwater pumps and associated flow paths shall be OPERABLE with:
- a. Three motor-driven auxiliary feedwater pumps, each capable of being powered from separate emergency busses, and
- b. One stean turbine-driven auxiliary feedwater pump capable of being powered from an OPERABLE steam supply system.
APPLICABILITY: MODES 1, 7, and 3.
- ACTION
- a. With the Train A motor-driven auxiliary feedwater pump inoperable, initiate corrective actions to restore the puep to OPERABLE status as soon as possible. The provisions of Specification 4.0.4 are not applicable,
- b. With any of the following combinat:ons of auxiliary feedwater pumps inoperable:
- 1) Train B or Train C motor-driven pump,
- 2) Train D turbine driven pump and any one motor-driven pump,
- 3) Train A and either Train 6 or Train C motor-driven pump, or
- 4) Train D tuebine-driven purp Restore the affected auxiliary feedwater pump (s) to OPERABLE status within 72 hours or be in at least HOT STANOCY within the next 6 hours and in HOT SHUT 00WN within the following 6 hours,
- c. With Train B and Train C motor driven pumps, or any three auxiliary feedwater pumps inoperable, be in at least H01 STANDBY within 6 hours and in HOT SHUT 00WN within the following 6 hours, j d. With four auxiliary feed ater pumps inoperable, immediately initiate
- corrective action to restore at least one auxiliary feedwater purp
! to OPERABLE status as soon as possible. ll EIILLeicIJEg1REMErTs , 4.7.1.2.1 Each auxiliary feedwater pump shall be demonstrated OPERABLE: j a. At least once per 31 days on a STAGGERED TEST BASIS by: i 1) Verifying that each motor-driven puep develops a discha ,e
- pressure of greater than or equal to 1454 psig at a flow of i greater than or equal to 540 gpm;
- 2) Verifying that the steam turbine-driven pump develops a
- discharge pressure of greater than er equal to 1454 psig at t
a flow of greater than or equal to 540 gem when the secondary l steam supply pressure is greater than 1000 psig. The provisions i of Specification 4.0.4 are not applicable for entry into MODE 3; 3/4 7-4 AMENDMENT N05. AND SOUTH TEXAS - UNITS 1 & 2
. 1 7 1933
' fI l
l PLANT SYSTEMS SURVEILLANCE REOU!REMENTS (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 correct position whenever the Auxiliary Feedwater System is placed in automatic control 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 auxilitry feedwater pump starts as designed automatically upon receipt of an Auxiliary Feedwater Actuation test signal.
- 3) Verifying that each auxiliary feedwater flow regulating valve limits the flow to each steam generator between 550 gpm end 675 gpe.
4.7.1.2.2 An auxiliary feedwater flow path to each steam generator shall be demonstrated OPERABLE following each COLO SHUTDOWN of greater than 30 days prior to entering MODE 2 by verifying rarmal flow to e e.h steam generator. SOUTH TEXA5 - UNITS 1 & 2 3/4 7-5 AMENDMENT N05. AND 1 ~. G C
fT'/ i PLANT SYSTEMS AUXILIARY FEEDWATER STORAGE TANK LIMITING CONDITION FOR OPERATION 3.7.1.3 The auxiliary feedwater storage tank (AFST) shall be OPERABLE with a 4 contained water volume of at least 518,000 gallons of water. APPLICABILITY: MODES 1, 2, and 3. ACTION: 1
- With the AFST inoperable, within 4 hours restore the AFST to OPERABLE status or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
SURVEILLANCE RFOUIREWENTS i 4.1.1.3 The AFST shall be demonstrated OPERABLE at least once: per 12 hours by verifying the contained water vo'iume is within its limits. 4 i l I l . 4 \ i i l l I SOUTH TEXA5 - UNITS 1 & 2 3/4 7-6 AMENDMENT NDS. AND i
~^
w w , w,
hl PLANT SYSTEMS SPECIFIC ACTIVITY Lidll]NG CONDITION FOR OPERATION 3.7.1.4 The specific ectivity of the Secondary Coolant System shall be less than or equal in 0.1 microcurie / gram DOSE EQUIVALENT I-131. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With the specific activity of the Secondary Coolant System greater than 0.1 microcurie /gra' DOSE EQUIVALENT I-131, be in at least HOT STU 0BY within 6 hours ard in COLD SHUT 00WN within the following 30 hours. 123YLlLLA!iCI_ REQ'J1REMENT S 4 4.7.1.4 The sp?cific activity of the Secondary Coolant System shall be determined to be within the limit by performance of the sampling and analysis program of Table 4.7-1. I SCUTH TEXAS - UNITS 1 & 2 3 /3 7-7 AMENDMENT NOS. AN: L, . 4 . .. l l
- _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ . _A
FD l l 'ABLE 4.7-1 l SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY l SAMP'E AND ANALYSIS PROGRAM TYPE OF HEASUREMENT SAMPLE AND ANALYSIS AND ANALYSIS FREQUENCY
- 1. Gross Radioactivity At least once per 72 hours.
Determination
- 2. Isotopic Analysis for DOSE e) Once per 31 days, wnen-EQUIVALENT 1-131 Concentration ever the gross radio-activity determination indicates concentrations greater than 10% of the allowable limit for radiciodines.
b) Once per 6 months, when-ever the gross radio-activity de'.ermina*. ion indicates concentrations less than or equal to 10% of the allowable limit for radiciodines. 9 4 SOUTH TEXAS - LNITS 1 & 2 3/4 7-8 AMENDMINT NOS. AND
"-1.i..
Dj i PLANT SYSTEMS l MAIN STEAM LINE ISOLATION VALVES l LIMITINO CONDITION FOR OPERATION 3.7.1.5 Each main steam line isolation valve (MSIV) shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3. ACTION: ' l t MODE 1: t With one M51V inoperable but open, POWER OPERATION may continue ! I provided the inoperable valve is restored to OPERABLE status within j
- 4 hours; otherwise be in HOT STANDBY within the next 6 hours and t 1 in HOT SHUTD0hH within the following 6 hours. ;
$ MODES 2 and 3: ! With one M$1V inoperable, subsequent operation in MODE 2 or 3 may proceed - i J provided f.f)e isolation valve is maintained closed. Otherwise, be in HOT
- STANDBY within the next 6 h6urs and in HOT SHUTDOWN within the following ;
j 6 houn. i
; gEVElltaNCEREQUIREMENTS . . . .
l 4.7.1.5 Each M5IV shall be demonstra*.ed OPERABLE by verifying full closure : within 5 seconds when tested pursuant to Specification 4.0.b. The provisions t of Specifiu tion 4.0.4 are not applicable for entry into MODE 3. ; i ! l l l 4
\
l I i
- i
?
j ! e , I I I I I i
- l l [
l l \ I 3/4 7-9 AVENDMEM h35. AND [ i SOUTH TEXA5 - UNITS 1 & 2 i t : . n a, ' f
.---r,y -ny, ,,-r--,.-,-,-v-,,,,--.-,,v,. , , , , - . - - - - - , _ , , - -
FC PLANT SYSTEMS ATHOSPHERIC STEAM RELIEF VALVES LIMITING CONCITION FOR OPERATION 3.7.1.6 At least four atmospheric steam relief valves and associated manual controls shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3, and 4.* ACTION:
- a. With one less than the required atmospheric steam reliet valves OPERABLE,
- ' restore the required atmospheric steam relief valves to OPERABLE status within 7 days; or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTOOWN within the following 6 hours and place the required RCS/
RHR loops in operation for decay heat removal. I
- b. With two less than the required atmospheric relief valves OPERABLE, re-store at least three atmospheric relief valves to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours and platt the required RCS/RHR loops in operation fer decay heat removal, j
2 3 SUDVEILLANCE REOL'IREMENTS ,,, ' 4.7.1.6 Each atmospheric relief valve and associated manual controls shall be l demonstrated OPERABLE prior to startup following any refueling shutdown or ! COLD SHUTDOWN of 30 days or longer, by verifying that all valves will open and
- close fully by operations of manual cont ols.
i i I . j i I l "When steam generators are teing u.ed for ce:a) heat removal. 4 I SOUTH TEXAS - UNITS 1 & 2 3/4 7-10 AMENDPENT NDS. AND 4
-__.,_,__-.--._,y__ _.m - . - , , . , - - - , _
F1 PLANT SYSTEMS 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION LIMITING CONDITION FOR OPERATION i 3.7.2 The temperatures of both the reactor and secondary coolants in +'ie steam genuators 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: 1 i
- 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 engineering evaluation to determine the effect of i the overpressurization on the structural integrity of the steam generator. Determine that the steam generator remains acceptable for continued operation prior to increasing its temperatures above 200'F. I 4 ! grfLJANCE REQUIDiplNTS i 4.7.2 The pressure in each side of the steam generator shall be determined to ^ i be less than 200 psig at least once per hour when the temperature of either
- the reactor or secondary coolant is less than 70'F.
l l i l i I j SOUTH TEXAS - UNIT 51 & 2 3/4 7-11 AMENDMENT NOS. AND L" I ? ::.;; i
I FL PLANT SYSTEMS 3/4.7.3 COMPONENT COOLING WATER SYSTEM WITINGCONDITIONFOROPERATION 3.7.3 At least three independent component cooling water loops shall be OPERABLE. RPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With only two component cooling water loops OPERABLE, restore at lea:t three loops to ')PERABLE status within 72 hours or be in at. least HOT STANDBY within the next C hours and in COLD SHUTDOWN within the following 30 hours. H M ILLAM E REQUIDE"ENIS 4.7.3 At least three co*p0nent cooling water loops shall be demonstrated OPERABLE. .
- a. At least once per 31 days by verifying that each valve outside con-tainment (manual, pr.er operated, or automatic) servicing safety-related e;uipment that is not locked, sealed, or otherwise secured in position is in its correct position; and
- b. At least once per 18 months during shutdown, by verifying that:
- 1) Each automatic valve servicing safety-related equipment or isolating the non-nuclear safety portion of the system actuates to its correct position on a Safety injection, Loss cf Offsite Power. Containment Phase "B' Isolation, or Low Surge Tank test signal, as applicable.
- 2) Each Co penent Cooling Water System pu?p starts automatically on a Safety Injection or Lnss of Offsite Po er test signal, and
- 3) The surge tank level instrucentation which provides automatic isolation of P0rtions of the systen is demonstrated OPERABLE by perfornance of a CHANNEL CAllBRATION test,
- c. By verifying that each valve inside containment (manual, power-operated, or auto atic) servicing safety-related equipment that is not locke:, stalec, or C.hnr.ise secured in position is in its cer-rett positien prior te entering MODE 4 tollo ing each COLD SHUTD0iN of greater than 72 hours if not perforced within the previous 31 days.
SOUTH TEAAS - UNITS 1 & 2 3/4 7-12 AMEN 0 MENT N05. AND 5 I 7 i;f,;
L Fi PLANT SYSTEMS 3/4.7.4 ESSENTIAL COOLING WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.4 At least three independent essential cooling water loops shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3, and 4.
; ACTION:
With only two essential cooling water loops OPERABLE, resto e at least three loops tc OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. , SURVEILLANCE RE0VIREMENTS 1 ) 4.7.4 At least three essential cooling water loops shall be demonstrated j OPERABLE:
- a. At least once per 31 days by verifying thet each valve (manual, power-i operateo, Jr automatic) servicing safe;y-related equipment that is
! not locked, sealed, or otherwise secured in position is in its correct position;
- b. At least once per 18 months during shutdown, by verifying that:
- 1) Each automatic valve servicing safety-related equipment actuates to its correct position on a Safety Injection, ECW pump start, i screen wash booster pump start and essential chiller start test signals, as applicable,
- 2) Each Essential Cooling Water pump starts automatically on a Safety Injection or a Loss of Offsite Power test signal, and
- 3) Each screen wash booster pump and the traveling screen start automatically on a Safety Injection test signal.
SOUTH TEXA5 - UNITS 1 & 2 3/4 7-13 AMENDMENT NDS. AND I - . : '. n.
FI PLANT SYSTEMS 3/4.7.5 ULTIMATE HEAT SINK _ LIMITING CONDITION FOR ODERATION 3.7.5 The ultimate heat sink shall be OPERABLE with:
- a. A minimum wh6er level at or above elevation 25.5 feet Mean Sea Level, USGS datum, and
- b. An Essential Cooling Water intake temperature of less than or equal to 99'F.
APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With the requirements of the above specification not satisfied, be in at least HOT STANDBY within 6 nours and in COLO SHUTOOWN within the follo.ing 30 hours. This ACTION is applicable to both units simultaneously. SVEi'IILLM:E FEQUIEE T J( 4.7.5 The ultimate heat sink shall be determ ned OPERACLE at least once per i 24 hours by verifying the intake water temperature and water level to be within their limits. I i I SOUTH TEtAS - UNITS 1 & 2 3/4 7-14 AMEN? MENT N05. AND
m FO PLANT SYSTEMS 3/4.7.6 (This specification number is not used.) AMENDMENT hDS. AND SOUTH TEXA5 - UNITS 1 & 2 3/4 7-15 '
FF PL. ANT SYSTEMS 3/4.7.7 CONTROL ROOM MAKEUP AND CLEANUP FILTRATION SYSTEM LIMITING CONDITION FOR OPERATION 3.7.7 Three independent Control Room Makeup and Cleanup Filtration Systems shall b OPERABLE. APPL.ICABILITY: All MODES. ACTION: MODES 1, 2, 3 and 4:
- a. With one Control Room Makeup and Cleanup Filtration System inoper-able, restore the inoperable system to OPERABLE status within 7 days or be in at least HOT STANDBY ttnin the next 6 hours and in COLD SHUTDOWN within the following 30 hours,
- b. With two Control Room Makeup and Cleanup Filtration Systems inoper-
;ble, restore at least two systems to OPERABLE status ,sithin 24 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
M] DES b and 6:
- a. With one Control Room Makeup and Cleanup Filtration System inoper-able, restore the inoperable system to OPERABLE status within 7 days or initiate and maintain operation of the remaini.7g OPERABLE Control Room Makeup and Cleanup Filtration Systems in the recirculation and makeup air filtration mode.
- b. With two Control Room Maxeup and Cleanup Filtration Systems inoper-able, or with tLe OPERABLE Control Room Makeup and Cleanup Filtration System, required to be in the recirculation and makeup air filtration mcde by ACTION a., not capable of being powered by an OPERABLE emergency po.,er source, suspend all operations involving CORE ALTERA-TIONS or positive reactivity changes.
SVh'OLLAli!LEEMIEEEMS 4.7.7 Each Control' Room Makeup and C l anup Filtration System shall be deron-strated OPERABLE:
- a. At least once per 12 hours by verifying that the control room air temperature is less than or equal to 78'F;
- b. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control roor, flow through the HEPA filters and charcoal adsorbers of the makeup and cleanup air filter units and verifying that the syster operates for at least 10 continuous hours with the makeup filter unit heaters eperating; SOUTH TEXAS - UNITS 1 & 2 3/4 7-16 AMENDMENT N05. AND N;;
^
FD l l PLANT SYSTEMS SURVEILLANCE RE0VIRU*ENTS (Continuedi __
- c. At least once per 18 months or (1) af ter any structural mainter,ance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone communi-cating with the system by:
- 1) Verifying that the makeup and cleanup systems satisfy the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05% for HEPA filter banks and 0.10% for charcoal adsorber banks and uses the test procedure guidance in Regulatory Positions C.S.a. C.5.c. and C.5.d of Regulatory
- Guide 1.52 Revision 2, March 1978, and the system flow rate is 6000 cfm i 10% for the cleanup units and 1000 cfm i 10% for the ,,
makeup units;
- 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% when tested at a temperature of 30'C and a relative humidity of 70%; and
- 3) Verifying a system flow rate of 6000 cfm i 10% for the cleanup units and 1000 cfm i 10% for the makeup units during system operation when tested in accordance with ANSI H510-1980.
- d. After every 720 hours 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 Regulatory Guide 1.52, Revision 2, March 15/6, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2 March 1978, for a eethyl iodide penetratica of less than 1.0% when tested at a temperature of 30'C and a relative humidity of 70%;
- 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 6.1 inches Water Gauge for the makeup units aid 6.0 inches Water Gauge for the cleanup units while operating t se system at a flow rate of 6000 cfm 2 10% for the cleanup units and 1000 cfm 2 10% for the makeup units;
- 2) Verifying that en a control room emergency ventilation test signal (High Radiation and/or Safety Injection test signal), the system automatically switches into a recirculation and makeup air filttation rode of operation with flow through the HEFA tilters and charcoal aasorber banks o' the cleanup and e neup units; SOUTH TEXAS - UNITS 1 & 2 3/4 7-17 A"ENDMENT N05. ANO
' I i m,
FI PLANT SYSTEMS SURVEILLANCE RE0VIBEMENTS 3Centinued)
- 3) Verifying that the system maintains the control room envelope at a positive pressure of greater than or equal to 1/8 inch Water Gauge at less than or equal to a pressurization flow of 2000 cfm rt'ative to adjacent areas during system operation;
- 4) Vt;rifying that the makeup filter unit heaters dissipate 4.5 1 0.45 kW when tested in accordance with ANSI N510 1980; and Verifying that on a High Toxic Gas test signal, the system autoa3tically switches into a recirculation mode of operation by isolating the normal supply and exhtust flow within 25 teconds,
- f. After each co plete or partial replacement of a HEPA filter bank, by verifying that the HEPA filter bank satisfies the in place penetration and bypass leakage testing acceptance criteria of less than 0.05% in accorcance with ANSI N510-1980 for a 00P test aerosol while cperating the system at a flow rate of 6000 cfm i 10% for < cleanup units and 1000 cfm i 10% for tne makeup units; and
- g. After each cc plete or partial replaceeent of a charcoal adsorber bank, by verifying that the charcoal adsorber oank satisfies the in place penetration and bypass leakage testing acceptance criteria of less than 0.10', in accordance with ANSI F510-1980 for a balogenated 5ydrocarbon ref rigeret test gas while operatir.g the system at a flow rate of 6000 cfm 210% for the cleanup units and 1000 efm 20% for the rakeup units.
SOUTH TEXAS - U':T5 1 & 2 3/4 7-19 AMENOMENT N05. AND
** ' 1 i nij l
L
FD PLANT SYSTEMS 3/4.7.8 FUEL HANDLING BUILDING (FHB) EXHAUST AIR SYSTEM ggl@_CONSITION FOR OPERATION 3.7.8 The FHB Exhaust Air System comprised of the following components shall be OPERABLE:
- a. Two independent exhaust air filter trains,
- b. Three independent exhaust booster fans,
- c. Three independent main exhaust fans, and
- d. Associated dampers.
APPLIC'.BILITY: MODES 1, 2, 3, and 4. ACTION: With iess than the above FHB Exhaust Air System components OPERABLE but with at least one FHB exhaust air filter train, two FHB exhaust booster fans, two FHB main exharst fans and associated dampers OPERABLE, restore the inoperable system to OPEhA3tE status within 7 days or be in at least H01 STANO3Y within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. EVEILLAME REWIDEMENTS 4.7.8 The Fuel Handling Builciing Exhaust Air System shall be demonstrated OPf' TABLE:
- a. At least once per : 4 5 on a STAGGERED TEST BASIS by initiating, from the control rv flow through the HEPA filters and charcoal adsorbers and veri 1/ing that the system operates for at least 10 continuous hcurs with the hcaters operating with two of the three exhaust booster f ans ar.d two of the three main exhaust fans operat-ing to maintain adequate air flow rate;
- b. At least once per 18 months and (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following paintin's, fire, or chemical release in any ventilation zone cor,muni-cat 'ng with the system by:
- 1) Verifying that the cleanup system satisfies the in-place penettation and bypass leakage testing acceptance criteria of less than 0.25% for HEPA filter banks and 0.10% for charceal adsorocr banks and uses the test procedure guidance in Regula-tory Positions C.!..a. C.S.c and C.5.d of Regulatory Guide 1.52, Revision 2, March 1.978, and the system flow rate is 29,000 cfm
- 10%;
- 2) Verifying, witt.in 31 days after removal, that a laboratory analysis of a representative caroon sample cttained in accer-dance with Regulatery Position C.6.t of Regulatory Guidc 1.f 2, AND SOUTH TEXAS - UN1151 & 2 3/4 7-19 AMENDMENT N05.
a o i;
PLANT SYSTEMS l t' SURVEILLANCE REOUIREMENTS (Centinued) j Rev'ision 2, March 1978, meets the laboratory testing criteria 1 of Regulatorv Position C.6.a of Regulatory Guide 1.$2. Revi- t sion 2, Maren 1978, for a methyl iodide penetration of less than 1.0% when tested at a temperature of 30'C and a relative ; humidity of 76%; and t
- 3) Verifying a system flow rate of 29,000 cfm i 10% during system operation with two of the three exhaust booster fans and two of the three main exhaust fans operating when tested in accordance with ANSI N510-1980. All combinations of two exhaust booster fans and two main exhaust fans shall be tested.
- c. After every 720 hours of charcoal adsorber operation, b/ verifying, [
within 31 days after removal, that a laboratory analysis of a i representative carbon sample obtained in accordance with Regulatory l Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, l mee:s the laboratory testing criteria of Regulatory Position C.6.a i of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl iodide penetration of less than 1.0% when tested at a temperature of t 30'C and a relative humidity of 70%, , f i i d. At least once per 18 months by: y f
- 1) Verifying that the pressure drop across the combined HEPA f
! filters and charcoal adsorber banks is less than 6 inches t Water Gauge while operating the system at a flow rate of 29,000 l cfm i 10%,
- 2) Verifying that the system starts on High Radiation and Safety ;
1 Injection test signals and directs flow throug~ the HEPA ! I filters and charcoal adsorbers, I
- 3) Verifying that the system maintains the FHB at a negative f pressure of greater than or equal to 1/8 inch Water Gauge relative to the outside atmosphere, and [
I j 4) Verifying that the heaters dissipate 50 1 5 kW when tested in [ accordance with ANSI N510-1980.
- e. After each complete or partial replacement of a HEPA filter bank, by verifying that the HEPA filter bank satisfies the in-place pene- [
tration and bypass leakage testing acceptance criteria of less than 0.05% in accordance with ANSI H510-1980 for a DOP test aerosol while opetating the system at a flow rate of 29,000 cfm i 10%; and I
- f. After each complete or partial replacement of a charcoal adsorber I bank, by verifying that the charcoal adsorber bank satisfies the in-place penetration and bypass leakage testing acceptance criteria of
' less than C.10% in accordance with ANSI N510-1980 for a halogenated
- hydrocarbon refrigerant test gas while operating the system $t a i flow rate e' 29,000 cfm 2 10%. ,
f f f AND I SOUTH TEXA5 - LWITS 1 & 2 3/4 7-20 AMENDMENT N05. 1, ,. 4 . - I l !
- - , - + - - -,--,~e-,,,m.e, _ - - - + . - w m o mm m, amm--m ,-4-e, --n--,r-~,w- ,-,---,--,--,e r, e m+-- -
0 1 PLANT SYSTEMS 3/4.7.9 SNUBBERS LIMITING CONDITION FOR OPERATION 4 4 3.7.9 All snubbers shall be OPERABLE. The only snubbers excluded from the
- requirements are those installed on nonsafety-related systems and then only if their f ailure or f ailure 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 snubb'ers located on
; systems required OPERABLE in those MODES.
a i ACTION: I With one or more snubbers inoperable on any system, within 72 hours replace or re-store the inoperable snubber (s) to OPERABLE status and perform an engineering eval-uation per Specification 4.7.9g. on the attached component or declare the attached
- system inoperable and follow the appropriate ACTION statement for that system.
i SURVEILLANCE REOUIREWENT< 4.7.9 Each snubber shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program in lieu of the requirements of Specification 4.0.5. }'
- a. Inspection Types As used in this specification, type of snubber shall mean snubbers l of the same design and manufacturer, irrespective of capacity.
- b. Visual Inspections l ,
i i Snubber, are categorized as inaccessible or accessible durino reactor operation. Each of these groups (inaccessible and accessible)The may l be inspected independently according to the schedule below. first inservice visual inspection of each type of snubber shall be performed af ter 4 months but within 10 conths of comencing POWER l I OPERATION and shall include all snubbers. If all shubbers of each type are found OPERABLE during the first inservice visual inspection, l ' the second inservice visual inspection shall be performed at the first i refueling outage. Otherwise, subsequent visual inspections shall be l performed in accordance with the following schedule: l l i 1 l i AMENDMENT NDS. AND SOUTH TEXAS - UNITS 1 & 2 3/4 7-21 ' l 1 L . .. - -
..g. ._
FD i PLANT SYSTEMS SURVEILLANCE REOUIREMENTS (Continued 1
- b. Visual Inspections (Continued) ;
J No. of Inoperable Snubbers of Each Type Subsequent Visual i per Inspection Period Inspection Period * ** j 0 18 months z T ~ { i 1 12 smnths t *5% ;
, 2 6 months 1 25% !
3,4 124 days i 25% ! 5,6,7 62 days i 25% l 1 8 or more 31 days t 25% l } c. Visual Inspection Acceptance Criteria ' , Visual inspections shall verify that: (1) there are no visible } I indications of damage or impaired OPERA 31LITY, (2) attachments to : l the foundation or supporting structure are functional, and (3) fasten- } i ers for attachment of the snubber to the component and to the snubb.r ' j anchorage are functional. Snubbe.'s which appear inoperable as a !
; result of visual inspect ons may be determined OPERABLE for the i
i i purpose of establishing the next visual inspection interval, provided ! l' that: (1) the cause of the rejection is clearly established and l remedied for that particular snubber and for other snubbers irrespec- l tive of type that may be generically susceptible; and (2) the affected i i snubber it functionally tested in the as-found condition and determined l 1 OPERABLE per Specification 4.7.9f. All snubbers connected to an i inoperable common hydraulic fluid reservoir shall be counted as ;
; inoperable snubbers. j J '
j d. Transient Event Inspection - 1 l
! An inspection shall be performed of a'l snubbers attached to sections l
) of systems that have experienced unervected, potentially damaging { ! transients as determined from . re/iew of optrational data and a j visual inspection of the systets w thin 6 mot.ths following such an , i event. In addition to satisfying the visual inspection acceptance ; i criteria, freedom-of-motion of mechanical snubbers shall be verified ] using at least one of the following: (1) manually induced snubber
; movement; or (2) evaluation of in place snubber piston setting; or i . (3) stroking the trechanical snubber through its full rarge of travel.
1 . t I
- "The inspection interval for % cn type of snobber shall not be lengthened ['
1 more than one step at a ttre unless a generic prwM em has been identified { and corrected; in that event the inspection interval may be lengthened one f I step the firrt time and No steps thereaf ter if no inoperaba snubbers of j that type are found. ! **The provisions of Specificat3cn 4.0.2 are net applicable. 5 1 i SOUTH TEXAS - UNITS 1 & 2 3/4 7-22 AMEN? MENT N05 AND ,
.. , ,i:
)
.Q PLANT SYSTEMS SURVEILLANCE RE00fREMENTS (Continued)
I
- e. Function'al Tests i During the first refueling shutdown and at least once per 18 months thereafter durinC shutdown, a representative sample of snubbers 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 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 101,of the total of each type of snubber shall be functionally tested either in place or in a bench test. For each snubber of a type that does not meet the functional test acceptance criteria of Specification 4.7.9f., an additional lot, s' 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
! 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 J
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-l 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 snubberc of that type shall
{ be functionally tested. If at any time the point plotted falls j 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 1 tested until the point falls in the "Accept" region or the j "Reject" region, or all the snubbers of that type have been tested; or l 3) An initial representative sample cf 55 snubbers shall be func-tionally tested. For each snubber type which dwes 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 i multiplied by the f actor,1 + C/2, where "C" is the nueber of snubbers found which do not meet the functional test acceptance l criteria. The results from this saeple plan shall be plotted I using an "Accept" line which follows the equation N = 55(1 ; + C/2). Each snut.ber point should be plotted as soon as the l snubber is tested. If the point plotted falls Sn or belcw the j "Accept" line, testing of that type of snubber may be terminated.
1 l l i l SOUTH TE AAS - UNITS 1 & 2 3/4 7-23 AMEN >'.ENT N0 5. AND i l l L _
F2 PLANT SYSTEMS SURVEILLANCE REOUIREMENTS (Continued)
- e. Functional Tests (Continued)
If the point plotted falls above the "Accept" line, testing must continue until the point falls in the "Accept" region cr 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 the day of equipment failure are retested. The representative satple selected for the functional test sample plans shall be randomly selected fror the snubbers of each type and reviewed before beginning the testing. The review shall ensure, as far as practicable, that they are representative 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 snubber, the functional test results shall be reviewed at that time to determine if additional samples should be limitet to the type of snubber which has failed the functional testing. l
- f. Functional Test Acetotance Criteria The snubber functional test shall verify that:
- 1) Activation (restraining action) is achieved within the specified ,
range in both tension and compression; '
- 2) Snubter bleed, or release rate where required, is present in both trnsion anc co.pression, within the specified range; l
- 3) Wntrt required, the force required to initiate or maintain motion i of the,5nutber is within the specified range in both directions (
of travel; and
- 4) For snutbers specifically required not to displace under contin-uous load, the ability of the Snubber to withstand load without cisplacerent.
Testing metheos may be used to measure parametars indirectly or para *eters etner than those specified if those results can be correlated to the specified parameters through established methocs. i SOUTH TE uS - UNITS 1 & 2 3/4 7-24 AMEN MENT NDS. AN: t, , , ; . )
F9 PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued 1
- g. Function ~al Test Failure Analysis An engineering evaluation shall be made of each failure to meet the functional test acceptance 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 of 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 f ails 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-tio "I!y tested. This testing requirement shall be independent of the requirerents stated in Specification 4.7.9e. for snubbers not meeting the functional test acceptance criteria,
- h. Functional Testino 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 ret the acceptance criteria subsequent to their most recent service, and the freedot-of-motion test must have been performed within 12 months before being installed in the unit.
- i. Snubber Service Life Procram The service life of hydraulic and rechanical snubbers shall be monitorea 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 establishec based on engineering inferration and shall be extended or shortened based on monitored test res91ts and failure history. Critical parts shall be replaced so that the maximum service life will not be exceeded during a period when the snubber is required to be OPERABLE. The parts replacements shall be cocu-rented and the docu~entation shall be retained in accordance with Specification 6.10.3.
SOUTH TEXAS - UNITS 1 & 2 3/4 7-25 AMENDMENT N05. ANO
. 7 b ,<
Fi 10 S 8 jf REJECT I w 6 e# CONTINUE
, TESTING ,
i 2
, 7 JV i
ACCEPT ! 1 e )
/
O 10 20 30 40 50 60 70 80 90 100 i
- N ,
l I 1 i I FIGURE 4.7-1 I SAM LE PLAN 2) FOR SNUEEER FUN;TIONAL TEST l SOUTH TEXA5 - UNITS 1 & 2 3/4 7-26 AMEN MENT h05. AN? 1 I E , ; i ., l \ _ - _ _ _ _ _ - - _ _ . - . _ - _ .
FP PLANT SYSTEMS 3/4.7.10 SEALE0 SOURCE CONTAMINATION LIMITING CONDITION FOR OPERATION 3.7.10 Each sealed source containing radioactive naterial either in excess of 100 microcuries of beta and/or gamma emitting material or 5 microcuries of alpha emitting material thall be free of greater than or equal to 0.005 microcurie of removsble contamination. APPLWABILITY: At all times. ACTION:
- a. With a sealed source having removable contamination in excess of the above limits, immeciately 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 Ccmmission Regulations,
- b. The provisions of Specification 3.0.3 are not applicable.
l'RVEILL ANCE REQUDEPENTS 4.7.10.1 Test Requirements - Each sealed source shall be tested for leakage and/or contamination by:
- a. The licensee, or
- b. Other perso.is specifically authorized by the Com-ission or an Agreement State.
The test reth0d shall have a detection sensitivity of at least 0.005 microcurie p.:r test sa?ple. 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 et the frequency described below,
- a. Sources in use - At least once per 6 months for all sealed sources containing radioactive ruterials:
- 1) Vith a half-life reater w than 30 days (excluding Hydrogen 3),
and
- 2) In any form other than gas.
SOUTH TEKAS - UN:TS 1 & 2 3/4 7-27 AuENCu!NT N;S. AND
.l
FI PLANT SYSTEMS SURVEILLANCE RE0VIRf"ENTS (Continued)
- b. Stored s'ources 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 sources 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 source or fission detector leakage tests reveal the presence of greater than or equal to 0.005 microcurie of removable contamination. SOUTH TEXA5 - UNITS 1 L 2 3/4 7-26 AMU 4Mr.NT N05. AN.
) i::?
FP PLANT SYSTEMS 3.7.11 (Th', specification number is not used.) 3/4 7-29 AMEhDMENT N05. AND SOUTH TEXA5 - UNITS 1 & 2
FP PLANT SYSTEMS 3.7.12 (This specification number is not used.) SOUTH TEXA5 - UNITS 1 & 2 3/4 7-30 AuEh0"ENT N05. AND ih3
'n FP I i PLANT SYSTEMS 3/4.7.13 AREA TEMPERATURE MONITORING a LIMITING CONDITION FOR OPERATION 3.7.13 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 the temperature inside any QDPS auxiliary processing cabinet ! exceeding 110*F for more than 12 hours, prepare an engineerirg evalu- , ation within the next 24 hours to determine the temperature effects t on QDPS OPERABILITY and service life. The provisions of Specifica- ; l tion 3.0.3 are not applicable. , J
- b. With one or more areas exceeding the temperature limit (s) sho.n I in Table 3.7-3 for more than 8 hours, prepara and submit to the !
Commissirr within 30 days, pursuant to Specification 6.9.2, a ) Special Peport that provides a record of the cumulathe time and ; the amount by which the temperature in the affected area (s) i exceeded the limit (s) and an analysis to demonstrate the continued , OPERABILITY of the affected equipment. The provisions of Specifi-i cation 3.0.3 are not applicable.
- c. With one or more areas exceeding the temperature limit (s) shown in r j Table 3.7 1 by more than 30*F prepare and submit a Special Report as required by ACTION b. above and within 4 hours either restore j the area (s) to within the temperature limit (s) or declare the equip- i ment in the affected area (s) inoperable, j
- I S E EILLaM i RE001REu!NTS
; 4.7.13 The terperature in each of the areas shown in Table 3.7-3 shall be !
l determined +.o be within its limit at least once per 12 hours. l I
- i t
I i ? l i ! Sou - s . u~nS u 2 3y 7.n m m N1 ~as. A~a l
-tim:
l
FD TABLE 3.7-3 AREA TEMPERATURE MONITORING AREA TEMPERATURE LIMIT ('F)
- 1. Relay Room < 78 (Electrical Auxiliary Building E1. 35'0")
- 2. Switchgear Rooms < 85 (Electrical Auxiliary Building E1. 10'0",
15'0",60'0")
- 3. Electrical Penetration Spaces < 103 (Electrical Auxiliary Building F1. 10'0",
35'0", 60'0")
- 4. Safety Injection and Contalnmant Spray ~
< 101 Pump Cubicles (Fuel Handling 6. 'di'g El. -29'0")
- 5. Corponent Cuoling Water Pump Cubicles -
< 112 (Mechanical Auxiliary Building El. 10'0")
- 6. Centrifugal Charging Pump Cubicles 5 132 (Mechanical Auxiliary Building E1. 10'0")
- 7. Hydrogen Analyzer Roor. $ 102 (Mechanical Auxiliary Building E1. 60'0")
- 8. Boric Acid Transfer Puep Cubicles < 101
~
(Mechanical Auxiliary Building El.10'0")
- 9. Standby Diesel Generator Rooms < 101*
~
(Diesel Generator Euilding El 25'0")
- 10. Essential Cooling Water Pu?p Roers 1 101 (Essential Cooling Water Intake Structure El. 34'0")
- 11. Isolation vahe Cubicles -
< 101
. (Isolation Valve Cubicle E1. 10' 0")
- 12. Qualified Display Frocessing System Roora 1 94**
(Electrical Auxiliary Building E1. 10'0")
*
- Temperature limit is 1 120'F when testing the standby diesel generator pursuant to Surseillance Re:;uirement 4.8.1.1.2,e.7).
**Measurerer.t insice CDPS a ailiary processing cabinets UNITS I & 0 3/4 7-J2 AMENy'ENT N05. AND SOUTH TEXAS
- 1:G m
Ft PLANT SYSTEMS 3/4.7.14 ESSENTIAL CHILLEO WATER SYSTEM LIMITING CONDITION F00 OPERATION 3.7.14 At least three independent Essenti , Chilled Water System loops shall be OPERABLE. APPLICABILI'if: H0 DES 1, 2, 3, and 4. ACTION: With enly two Essential Chilled Water System loops OPERABLE, restore three loops to OPERABLE status within 72 hours or be in at least HOT STANOBY within the next 6 hours and in COLD SHUTOOWN within the following 30 hours. SUcVEILLANCE REOUIDEMENTS 4.7.14 The Essential Chilled Water System shall be demonstrated OPERABLE by;
- a. Performance of surveillances as required by Specification 4.0.5, and
- b. At least once per 18 renths by demonstrating that the system starts automatically on a Safety Injection test signal, l
3/4 7-33 AMENDYFNi h25. AND SOUTH TEXA5 - UNITS 1 & 2 l -' 1 ? Ig,
Fp
,3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES OPERATING LIMITING CONDITION FOR OPERATION
- 3. 8.1.1 As a minimum, the following A.C. electrical power sources shall be OPERABLE:
- a. Two physically independent circuits between the offsite transmission network and the onsite Class 1E Distribution System", and
- b. Three separate and independent standby diesel generators, each with a separate fuel tank containing a minimum volume of 60,500 gallons '
of fuel. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:
- a. With one offsite circuit of the above-required A.C. electrical power sources inopera51e, demonstrate the OPERABILITY of the remaining A.C.
sources by performing Surveillance Requirerent 4.8.1.1.1.a within 1 hour and at least once per 8 hours thnreafter. Demonstrate the OPERAB161TY cf each standby diesel generator that has not been suc-cessfully tested within the past 24 hours by performing Surveillance R,guirement 4.8.1.1.2.a.2) for each such standby diesel generator, separately, within 24 hours. Restore the offsite circuit to OPERABLE status within 72 hours or be in at least HOT SHUT 00hh within the n o t 12 hours and in COLD SHUT 00hN within the following 24 hours,
- b. With a standby diesel generator inoperable, demonstrate the OPERABILITY of the above-required A.C. offsite sources by performing Surveillance Requirement 4.8.1.1.1.4 within 1 hour and at let.st once per 8 hours thereafter. If the standby diesel generator became in- ,
operable due to any cause other than preplannred preventive main- I tenance or testing, der.onstrate the OPERABILITY of the remaining 0FERAB;E 4tandby diesel generators by performing Survet11ance Require-rent 4.E.1.1.2.a.2) and for each such standby diesel generator, , seca .tely, within 24 hours." Restore the ineperaole standby diesel l;
.t o r to OPERAELE status within 72 hours or be in at least HJT 56.ufD?^ within the next 12 hours and in COLD SHJT00kN within the I folio.ing 24 hours. . c. Vith one offsite circuit and one standby diesel generator of the above required A.C. electrical pc.er sources inoperable, demonstrate the OFERAEILIT) of the rem ining A.C. sources by perfore.ing Specifica-tien 4.6.1.1.la. witnin 1 hour and at least once per 8 hours there-after; ano if the standby diesel generator became inoperable due to "This test is required to be completed regardless of when the inoperable standby i diesel generator is restored to OPERABILITY.
I "Less cf ore 13.5 kV Standty bus to 4.16 kV Esr tus line cc .stitutes icss ef s se offsite so.;rce. Less ef t.o 13.6 LV Standh busses to 4.16 kV E5r tus lines constitutes less of t o effsite scurces. SOUTH TEAA5 - LNITS 1 & 2 3/4 E 1 AENMNT N05. AND
... 1 '. . . . ' .
FD ELECTRICAL POWER SYSTEMS { I i 4 LIMITING CONDITION FOR ODERATION , j ACTION (Continued) ! any cause other than preplanned preventive maintenance or testing, l i demonstrate the OPERABILITY uf the remaining OPERABLE standby diesel ' l generators by performing Surveillance Requirement 4.8.1.1.2a.2) within 8 hours"; restoro at least one of the inoperable sources to OPERABLE 2 status within 12 hours e* be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN witnin the following 30 hours. Restore
- at least two offsite circuits and three standby diesel generators to 4
OPERABLE status within 72 hours from the time of initial loss or be . in at least HOT STAN9BY within the next 6 hours and in COLD SHUTDOWN 4 within the following 30 hours.
- d. With one standby diesel generator inoperable in addition to ACTION b.
l or c. above, verify that: l 1. All required systems, subsystems, trains, components, and de-i vices that depend on the remaining OPERABLE diesel gentrator j as a source of emergency po.er are also OPERABLE, and ! 2. When in MODE 1, 2, or 3, the steam driven auxiliary feed.ater puep is OPERABLE. l If these conditions are not satisfied within 2 hours be in at least HOT STANDBY within the next 6 hours and in COLD SHUTOOWN within ths following 30 hours. f l e. With two of the above required offsite A.C. circuits inoperable, j demonstrate the OPERABILITY of three stancy diesel generators by . performing the requirements of Specification 4.8.1.1.2a.2) within 8 hours unless the standby diesel generators are already coerating; restore at least one of the inoperable offsite sources to OPERABLE status within 24 hours or be in at least HDT STANDBY within the next l 6 hours. With only one offsite source restored, restore at least 4 two off site circuits to OPERABLE status within 72 how-t from time of initial less or be in at 1 cast HOT STANDBY within the next 6 hours j and in COLD SHUT 00.N within the following 30 heurs. f f. With two or three of the above required standby diesel gentrators inoperable, dewenstrate the OPERABILITY of two offsite A.C. circuits by perforcing the requirements of Specification 4.8.1.1.la, within I hour and at least on:e per B hours thereafter; restore at least two standty diesel generators to OPERAELE status within 2 nours or te in at least H;1 STA'CSY within tne next 6 hours and in COLD SHUT 00*H within the fo11 ewing 30 hours. Restore at least three standby diesel generators to OPERAELE status within 72 hours from titre of initial loss or be in least HOT STANDBY vithin the next 6 hours and in COLO
. SHUT 00.N within the following 30 hours.
"This test is reagire: te be corpleted regardless of when the inoperable stand;y ciesel generator is restered to CFERASILITi.
SOUTH TE)AS - UNITS 1 & 2 3/4 6-2 AMIN MENT N05. ANO e'. 4 . ; *,E.
Pt ELECTRICAL POWER SYSTEMS SURVEILtBNCE REQUIREMENTS 4.8.1.1.1 Each of the above required independent circuits between the offsite transmission network and the Onsite Class 1E Distribution System shall be:
- a. Determined OPERABLE at least once per 7 days by verifying correct breaker alignments, indicated power availability, and
- b. Orimonstrated OPERABLE at least once per 18 months during shutdown by transferring the unit power supply from the norttal circuit to each of the alternate circuits.
4.8.1.1.2 Each standby diesel generatar shall be demonstrated OPERABLE:
- a. In accordance with the frequency specified in Table 4.8-1 on a l
STAGGERED TEST EASIS by:
- 1) Verifying the fuel level in its assnciated fuel tank,
- 2) Verifying the diesel starts from arbient condition and accele-rates te 600 rp- (nominal) in less thati or equal to 10 seconds.*
The genera?.or vcitage and frequency shall be 4160 416 volts and 60 1 1.2 n: within 10 seconds
- after the start signal.
The diesel generator shall be started for this test by using one of the folle.ing signals: a) Manual, or b) Sirulated less-of-offsite po er by itself, or I c) Simulated loss-of-of f site pc.er in conjunction with a 1 Safety injection test signal, or d) A Safety Injection test signal by itself.
- 3) Verifyirg the generator is synchronized, loaded to greater than or ec.si to 5500 kW in less than or equal to 10 M nutes', and operates with a lead greater than or equal to 5500 kW for at least (0 rirutes, an.
- 4) Verifying the standby diesel generator is aligned to provide starc;y pe=tr to the associated erergency busses.
l "These diesel generater starts f rcr. art ent conditions shall te perforred only once per 184 cays in these surveillance tests and all other engine starts for i the purpose of this surveillance testing shall be preceded by an engine prelube period and/or otrer warr.p procedures such as gradual leading (>150 sec) recom-rended ty the raVacture- so tnat the rechanical stress anc wear on the i ciesel engine is riniri:e-l SCJTH TEMS - LN Ti 1 & 2 3/4 6 3 AMEC MENT N25. AN: h'17 Z', l l
FC : ELECTRICAL POWER SYSTEMS ( SURVEILLANCE REOUIREMENTS (Continuadi i
- b. At least once per 31 cays 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 its associated ; fuel tank; r
- c. By saepling new fuel oil in accordance with ASTM-D4057 prior to f addition to storage tanks and:
- 1) By verifying in accordance with the tests specified in i ASTM 0975-81 prior to additior,tc the storage tanks that the !
I sarple has: a) An API Gravity of within 0.3 degrees at 60'F or a specific gravity of within 0.0016 at C')/60*F, when compared to the l supplier's certificate, or an absolute specific gravity at l 60/60*F of greater than or equal to 0.83 but less than or t equal to 0.89, or an API gravity of greater than or equal to , 27 cegrees bat less than or equal to 39 degrses; b) A kinematic viscosity at 40*C of greater than or equal to 1.9 centistokes, but less than or equal to 4.1 centistokes I if gravity was not detertined by comparison with the ! supplier's certification; i c) A flash point equal to or greater than 125'F; and { l d) A clear and bright appearance with proper color when tested i I in accordance with ASTM 04176 82. I
- 2) By verifying within 30 days of obtaining the sample that the t other properties specified in Table 1 of ASTM 0975-81 are ret when tested in accordance with ASTM 0975-81 except that the [
enalysis for sulfur may be performed in accordance with l ASTM 01552 Ti, ASTM-02622 82, or ASTH-04294 83.
- d. At least once every 31 days by obtaining a sample of fuel oil in accordance with ASTM-02276 78, and verifying that total particulate centa-ination is less than 10 mg/ liter when checked in accordance with ASTH-02276-76, Msthod A;
- e. At least once per 18 months, during shutdown, by:
- 1) Subjecting the diesel to an inspection in accordance with procecures preparac in conjunction with its stanufacturer's recomendations for this class of sundby service;
- 2) Verifying the generator capability tc reject a load of greater [
than or equal to 765.3 kW while maintaining volt 3ge at l f 4160 416 volts and frequency at 60 4.5 N:: AV 50'JTh TEXAS - OMTS 1 & 2 3/4 6-4 AMECMENT M S. ( I l.. . , , .. - l
FL ELECTRICAL POWER SYSTEMS SURVEILLANCE REOUIREMENTS (Centinuedi
- 3) Verifying the generator capability to reject a load of 5500 kw witt out tripping. The generator voltage shall not exceed 5262 volts during and following the load rejection;
- 4) Simulating L loss-of-offsite power by itself, and:
a) Verify ng deenergization cf the ESF busses and load sheddiig from the ESF busses, and b) Verifying the diesel starts on the auto-start signal within 10 seconds, energizes the suto-coanected shutdown loads through the load sequencer and operates for greater than or equal to 5 minutes while its generetor is leaded with the shutdown loads. After energization, the steady-state voltage and frequency of the ESF busses shall be maintained at 4160 2 416 volts and 601 L 2 Hz during this test.
- 5) Verifying that on a Safity Injection test signal, without loss-of-of f site pomer, 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 4160 1 416 volts and E0 + 1.2 Hz 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) Simulating a loss-of-of fsite pcwer in conjunction with a Safety injection test signal, and- ,
i a) Verifying deenergizattan of the ESF busses and load shedding from the ESF busses; b) Verifying the diesel sta7ts on the auto-start signal with-in 10 se:cnds, energizes the auto-connected ESF (a:cicent) , loads through the load sequencer and operates fer greater than or ecual to 5 minutes while its generater is leaded with the E5F loads. After energization, the steady-state voltage and frequency of the ESF busses shall be raintainec , at 4160 + 416 volts and 60 2 1.2 Hz during this test; and c) Verifying that all auto atic diesel generate trips, except engine oserspeed, generator cif ferential, and loi. lube oil pressure are automatically bypassed upon loss of voltage on the ESF bus concurrent with a Safety injection : Actuation signal.
- 7) Verifying the standby diesel generator crerates for at least 24 h0ves. D. ring the first 2 h:urs of this test, the ciesel 50JTH TEXAS - UNITS 1 & 2 3/4 S-5 AMEN 0 MENT N05. AN:
F. ELECTRICAL POWER SYSTEMS SURVEILLANCE REOUIREMENTS (continued) I generator shall be loaded to greater than or equal to 5935 kW* and during the remaining 22 hours of this test, the diesel ,i generator shall be loaded to greater than or equal to 3500 kW. The genera.or voltage and frequency shall be 4160 + 416 volts , and 60 + 1.2 Hz witnin 10 seconds af ter the start signal the , steady state generator voltage and frequency shall be maintained within these limits during this test. Within 5 minutes after com- i pleting this 24 hour test, perform Spe:ificatio'n 4.8.1.1.2e.6)b);**
- 8) Verifying that the auto-connected loads to each standby diesel ;
e generator do not exceed the 2000-hour rating of 5935 kW; i
- 9) Verifying the standby diesel generator's capability to:
' a) Synchronize with the offsite power source while the [ generator is leaded with its ESF loads upon a t il simuiated restoration of offsite power, l a b) Transfer its loads to the offsite power source, and l 1 I c) Be restored to its standby status. i i
- 10) Verifying that with the standby diesel generator operating in a l .
test. mode, connected to its bus, a simulated Safety Injection i 8 signal overrides the test mode by: (1) returning the diesel gen-erator to standby operation, and (2) automatically energizing [ ) the E5F loads with offsite power; i*
- 11) verifying that the automatic lead sequence tirer is OPERABLE i with the first sequenced load verified to be loaded between '
j 1.0 second and 1.6 secondt, and all other load blocks within 1 10% of its oesign interval;
- 12) Veri */'ng that the stancey diesel generater etergency stop lock- j out feature pre"ents diesel generator starting; and !
l l t
*1f future load canditinns enceed the 2000-hour ratics (5935 kW) of the diesel generator, the diesel generator will be tested at the 2-hour rating I
{ (6050 kW) thereafter. ( 4
**lf Specification 4.8.1.1.2e.6)t) is not satisf actorily Instead, the standby diesel corpleted, it is net !
necessary to repeat the preceding 24-hour test. j i generator may be operated at $500 kV for 1 bour or until operating terperatu,e > has stabilized. l i 1 3/4 6 6 AVEh0MiliT N05. A'O . SOUTri TEX A5 - L'dTS 1 !. 2 i b*Y j 6 e..', f t
FL ELECTRICA' POWER SYSTEMS SURYEILLEE REtmEMENTS _ff entinuedi
- 13) Demon,'. rating the OPERABILITY of the automatic load shed bypass and the manual load shed reinstatement features of the load sequencer,
- f. At least once per 10 years or after any rnodifications which could affect standby diesel generator interdependence by starting all standty diesel generators simultaneously, during shutdown, and l verifying that all standby diesel generators accelerate to at least i
600 rpm in less than or equal to 10 seconds; and
- g. At least once ptr 10 years by:
- 1) Draining each fuel tank, removing the accueulated sedia.cnt and cleaning the tank using a sodiui hypochlorite solution, or I equivalent, and
- 2) Perforring a pressure test of those portions of the diesel fuel oil system designed to Section Ill, subsection ND of the ASME Code at a test pressure equal to 110% of tha systeu design pressure.
4.E.1.1.3 Reports - All standby diesel generator failures, valid or nonvalid, shall be reported to the Cc-missien in a Special Report pursuant to Specifica-tion 6.9.2 within 33 days. Reports of standby diesel generator failures shall include the inforestion recon ended in Regulatory Position C.3.b of Regulatory Guide 1.108, Resision ., August 1977. If the nu-ber of failures in the last 100 valid tests (en a per nuclear unit basis) is greater than or equsi to 7 the report shall be supplerented to include the additional information reco-- nended in Regulatory Position C.3 b of Acgulatory Guide 1.105, Revision 1, August 1977. 5:,Tr TE>AS - L',;is ; & : 3/4 E 7 A*'EN:ENT h:5. AN: I e
F; lable4.8-1 O!ESEL GENERATOR TEST SCHEDULE NUMBER OF F/ LURES IN NUMBER OF FAILURES IN LAST 20 VAllo TESTS
- _ LAST 100 VAL 10 TESTS
- TESTFREQUENg 14 Once per 31 days 51 1 2** 15 Onct per 7 days
- Criteria for determining nu-ber of failures and nu-ber of valid tests shall be in a:cordan:e with Fegulatory Position C.2.e of Regulatory Guide 1.105, but determined on a per diesel generator basis.
For the purpcse cf detereining the required test frecuency, the previous test failu e count ray te redu:ed to zero if a co plete diesel overhaul to ii6e-ne. l condition is corpleted, presided that the overhaul, in:1uding apprcpriate post-raintenance cperation and testing, is specifically appresedThe by reliability the ranu-fa:turer d if acceptable reliability has been demonstrated. criterion shall be the suce.essful completion of 14 conse:utive tests in a single series. Ten cf snese tests .hsll te in accordance with the -outine Surveillance Require e .ts 4.5.1.1,2.a.2 and 4.E.1.1.2.a.3 an-J four tests in ' a:cercan:e with tne 154 day te ting requirerent of Surseillance Require ents 4.5.1.1.2.a.2 and 4.8.1.1.2.a.3. If this criterien is not satisfies curing the first series of tests, any alternate criterien to be used to transvalue the failure count to zero requires SRC a;prosal,
**The assceiated test frequency shall be maintali.ed until seven consecutise failure free de-an05 have been perforced and the nu-ber of failures in the ,
last 20 salid crards has teen reda:ec te7e. AEN:ENT N: 5. A%: SOUTH TExA5 - llN:71 & 2 3M S-E 1
FI ELECTRICAL POWER SYSTEH5 A.C. SOURCES SHUTDOWN LIMITIN3 CONDillCN F0D OPERATION 3.8.1.2 As a miniruft, 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. Two standby diesel generators each with a separate fuel tank contain-ing a minimum volume of 60,500 gallons of fuel.
APPLICABILITY: MODES 5 and 6. A1 TION: With less that the above minima required A.C. electrical 7,0 er sources l OPERABLE, im ediately suspend all operations i.wolving CORE ALTERATIONS, posi-l tive reactivity changes, rosement of irradiated fuel, or crane operation with l loads over the spent fuel pool, and within 8 hours, depressurize and vent the Reactor Coolant Syster through a greater than or equal to 2.0 square inch vent. l
!a addition, whe, in M :E 5 with the reactor coolant loops not filled, or ir, l MODE 6 with the water lesel less than 23 feet above the reactor vessel flange. l imediately initiate corrective action to restore the required sources to l
OPERABLE status as soon as possible.
)LDXEILLh(W'd'!fCiB 4.8.1.2 The above recaired A.C. electrical oever sources shall be de onstrated OPERABLE by the perferran:e cf each of the require ents of Specifications 4.8.1.1.1, 4.8.1.1.2 (except fcr Specification 4.6.1.1.Pa.3), and 4.E.1.1.3.
50'JT H T E
- A5 - L" T5 1 & 2 3/4 E-9 AMEN 0 MENT N05. AN:
- l. %
FD ELECTRICALr?ERSYSTEMS 3/4.8.2 D.C. SOURCES OPERATIS LIMITIN3 CONDITION FOR OPERATION 3.8.2.1 As a minimum, the following D.C. electrical sources shall be OPERABLE:
- a. Channel 1 125-volt Battery Bank EIA11 (Unit 1), E2A11 (Unit 2) and its two associated chargers,
- b. Channe) 11125 volt Battery Bank E1011 (Unit 1), E2011 (Unit 2) and its associated full capacity charger,
- c. Channel !!! 125-volt Battery Bank E1811 (Unit 1), E2011 (Unit 2) and its associated full capacity charger, and
- d. Channel IV 125-volt Battery Bank E1C11 (Unit 1), E2C11 (Unit 2) and its two associated chargers.
APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:
- a. With one of the required battery banks, and/or one of the required chargers for the Channels II or !!! inoperAle, restore +.he inoperable battery bank and/or charger to OPERABLE status w!'hin 2 hours or te in at least NOT STANOBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- b. With only one charger on Channel I or l'/ OPERABLE, demonstrate the OPERABILITt of the associated battery bank by performing Survel11ance Requireeent 4.6.2.1.a.1) within 1 hour and at least once per 8 hours thereafter. If any Category A lir.it in Table 4.8 2 is not ret, declare the battery inoperable. Rettore the ineperable charger to OPERABLE status within 24 hours or be in at least HOT STAN0Ev within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVilltaT_E PE091tEPENTS 4.8.2.1 Each 125 volt battery bank and charger shall be demonstrated OPERABLE:
- a. At least once per 7 days by verifying that:
- 1) The parameters in Table 4.6 2 rett the Category A limits, and
- 2) The total battery terminal voltage is greater than or equal to 129 volts on float charge.
SOUTH TEXA5 - UNITS 1 & 2 3/4 6 10 AYENMENT N05. ANO h 1 i ;In
i ELECTRICAL POWER SYSTEM 5 M311LLAELHOI.tEINILLCen11%ed)
- b. At least once per 92 days and within 7 days after a batteef ctischarge with battery terrinal voltage belo 110 volts, or battery overcharge with battery terminal voltage above 135 volts, by verifying that:
- 1) The parareters in Table 4.8 2 rett the Categcry B limits,
- 2) There is no visible corrosion at either cell-to cell or terminal connections, or the connection resistance of these it::cs is less than or equal to 150 x 10 $ che, and
- 3) The averact electrolyte temperature of six connected cells is above 65' f.
- c. At least er:e per 15 ranths by verifying that:
- 1) The cells, cell plates, and battery ra:ks shew no visual indicatien of physical drage or abnorcal deterioration,
- 2) Tre cell .c-: ell a d terrinal connections are clean, tight, a-<
coatee with antit:rresion raterial,
- 3) Tte res:stan:e cf ea:h cell to-cell and terrinal conne:tien is less than cr e;.al te 150 x 10 ' ch , and
- 4) The battery charger .ill supply at least 300 a peres at 125 solts fer at least S hou 5.
- d. At least er:e cer 15 r:nths, durin; shutd:.n, ty set ifyino that the l batterycara:it.,isa:ecuatetosu;plyandrnintaininOi!RAELE status all cf the a:tual or siralate: E5F icads fc the design d.ty c):',e .'e t* e t attery is sLt.ie:ted te a tatter) ser,i:e test;
- e. At least en:e re 6: r:nths, d;rin shut:: , ty serifyin; that tre battery ca;3:it., is at least EM of the craf a:turer's ratir; w*e .
sutje:te: te a perfer v:e discharge test. On:e per 60-r: nth inter $al this ptrfer r:e discharge test ray be perferre: in lies cf the tatte y sersite test required by Sre:ification 4.E.2.lc.; an:
- f. At least cr:e per li r:nths, durir; shutd .9, by gisin; perforr y:e dis:targe tests c' tattery capa:ity to arj tatterv tha'. st;.; sig .s of ctgra:aticn er tas rea:he: E5% of t.he service life eare:te: fer the a;;1icatic . Degradaticn is indi:atta when the tatt.ery rca:ity decps r:re than IC\ of rated ctpacity f ree its avera;e on ;,revious performan:e tests, er is telo. 93 of the ranufa:turer's ratin;.
5: Jir TE di a :is 1 !. 2 3 4 i-11 AMEN:"INT N:: AN: bit
FC TABLE 4.8-2 BATTiRY SVRVEILLANCE REQUIREMENTS CATEGORY AW CATEGORY BW PARAMETER LIMITS FOR EACH LIMITS FOR EACH A'.LOWABLE W DESIGNATED PILOT CONN *CTED CELL VALUE FOR EACH CELL CONNECTiD CELL Electrolyte > Minimum level > Minimum level - Above 'op of Level indication m:-k, indication mark, pistes, and < 1/4" above and < 1/4 above and not maxieum level maximum leve' overflowing indication mark indication mark Float Voltage 1 2,13 volts U) > 2.07 volts 1 2.13 volts Not more than O.020 below the
***'*9**'*'
Specific > 1.200(5) > 1.195 connected cell: Gravity (4) Averaga of all Average of all connected cells conne:ted rolls
> 1,205 1 1.195(5) taste NOTATIONS (1) For any Category A para eter(s) outside the limit (s) shp.n the battery ray be censicered CPEAABLE provided that within 24 hours all Category B reasure ents are taken 6md found to be within their allo =4ble salves, and provided all Category A and E paraTeter(s) are r: stored to within lieits within the nest ( caps.
(2) For any Cete; ry E para eter(s) outsid) the limit (s) sho.n, the tattery eay te ceaidered OiERAELE provided that the Category B para eters are within their allo.atie values and provided that Category B para eters(s) are restorec to within lirits within 7 days. (3) Any Category 5 para eter n:t ithin its allo.able salue indicates an ino;erable battery. (4) Cerre:ted for electrolyte tercerature and level. (5) Or battery charging rurrent is less than 2 amps when on charge. (6) Cctrected for average ele:trolyte terperature. ! l r i SCUTn TE u i - UN!T5 1 & 0 3/4 6-10 ave'O E'J h:5. AN:
...;... (
I
ELECTRICAL POWER SYSTEMS ' D.C. SOURCES
. SHUT 00WN' LIMITING CONDITION FOR OPERATION _ _ _ _ _
3.8.2.2 As a minimum, Channel I 125-volt Battery Bank EIA11 (Unit 1), E2A11 (Unit 2), and Channel IV 125-volt battery bar.k E1C11 (Unit 1), E2C11 (Unit 2), and their two associated chargers shall be OPERABLE. APPLICABILITY: MODES 5 and 6. ACTION: With the required battery banks and/or charger (s) inoperable, immediately sus-pend all cperations involving CORE ALTERATIONS, positive reactivity changes or movement of irradiated fuel; initiate ccrrective action to restore the required battery banks and/or chargers to OPERALLE status as soon as possible, and within 8 hours, depressurize and vent the Rractor Coolant System through a 2.0 square inch vent. SURVEILLANCE g ?J E nil ___ _ 4.8.2.2 The above required 125-volt battery banks and chargers shall be demonstrated OPERABLE in accordance with Specification 4.8.2.1. , i i {,
- SOUTH TEXAS - UNITS 1 & 2 3/4 8 13 AMENDMENT N05. AND l.0V 17 i.e,
FL ELECTRICAL POWER SYSTEMS 3/4.8.3 ONSITE POWER DISTRIBUTION OPERATING LIMITING CONDITION FOR OPERATION 3.8.3.1 The following electrical busses shall be energir.d in the specified manner:
- a. Train A A.C. ESF Busses consisting of:
- 1) 4160-Volt ESF Bus # E1A (Unit 1), E2A (Unit 2), and
- 2) 480-Volt ESF Busses # E1A1 and E1A2 (Unit 1), E2A1 and E2A2 '
(Unit 2) from respective load center transformers. Train B A.C. ESF Busses consisting of: b.
- 1) 4160-Volt ESF Bus # ElB (Unit 1), E2B (Unit 2), and
- 2) 480-Volt ESF Busses # ElB1 and E182 (Unit 1), E2B1 and E2B2 (Unit 2) from respectiva load center transformers,
- c. Train C A.C. ESF Busses consisting of:
- 1) 4160-Volt ESF Lus # E1C (Unit 1), E2C (Unit 2), and
- 2) 480-Voit ESF Busses # E1C1 and E1C2 (Unit 1), E2C1 and E202 (Unit 2) from respective load center transformers.
- d. 120-Volta.C.VitalDistributionPanelsDP1201andDP001 energized from their associated inverters connected to D.C. Bus # E1A11 (Unit 1), E2A11* (Unit 2),
- e. 120-Volt A.C. Vital Distribution Panel DP1202 ener ized from its associated inverter connected to D.C. Bus # E1011*g(Unit 1), E2011*
(Unit 2), l f. 120-Volt A.C. Vital Distribution Panel DP1203 ener ized from its
- associated inverter connected to D.C. Bus # ElB11*g(Unit 1), E2B11*
(Unit 2), ,
- g. 120-Volta.C.VitalDistributenPanelsDP1204andDP002 energized from their associated inverters connected to D.C. Bus # E1C11 (Unit 1), E2C11* (Unit 2),
- b. 125-Volt D.C. Bus E1A11 energized from Battery Bank E1A11 (Unit 1),
! E2A11 (Unit 2),
- i. 125-Volt D.C. Bus E1011 energized from Battery Bank E1011 (Unit 1),
E2011 (Unit 2), J. 125-Volt D.C. Bus ElB11 energized f rom Battery Bank E1811 (Unit 1), , E2B11 (Unit 2), and
- k. 125-Volt D.C. Bus E1C11 energized f rom Battery Bank E1C11 (Unit 1), ,
E2C11 (Unit 2).
*The inverter (s) asso:iated with one channel may be disconnected from its D.C.
bus for up to 24 hours as necessary, for the purpose of performing an equalix-ing charge on its associated battery bank provided: (1) its vital distribu-tion panels are energized, and (2) the vital distribution panels associated with the other battery banks are energized from their associated inverters ' and connected to their associated D.C. busses. SOUTH TEXAS - UNITS 1 & 2 3/4 8-14 AMENDMENT N05. AND ,
~ - - fr[ ELECTRICAL POWER SYSTEMS l l LIMITING CONDITION FOR OPERATION (Continued) ' APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:
- a. With one of the required trains of A.C. ESF busses not fully energized, reenergize the train within 8 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- b. With one A.C. vital distribution panel either not energized from its associated inverter, or with the inverter not connected to its asso-ciated 0.C. bus: (1) reenergize the A.C. distribution panel within 2 hours or be in at least HOT STANDBY within the next 6 hours anu in COLD SHUTOOWN within the following 30 hours; and (2) reenergize the A.C. Vital distribution panel 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 SHUTDOWN within the following 30 hours,
- c. With one D.C. bus not energized trom its associated battery bank, reenergize the D.C. bus from its associated battery bank within 2 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUT 00VN within the following 30 hours.
SURVEILLANCE REOUIREVENTS 4.8.3.1 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.
?OUTH TEXAS - UNITS 1 & 2 3/4 8-15 AMENDMENT NDS. AND
- 1. :. .
- 1;J;
?l ELECTRICAL POWER SYSTEMS ONSI1E POWER DISTRIBUTION SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.3.2 As a minimum, the following electrical busses shall be energized in the specified manner:
- a. Train A and Train C of A.C. ESF busses EIA and E1C (Unit 1), E2A and E2C (Unit 2), each consisting of one 4160-volt ESF bus and two 480-volt A.C. ESF load centers,
- b. Four 120-volt A.C. vital distribution panels consisting of OP001, DP1201, DP002 and DP1204 energized from their associated inverter
- connectedtoItsrespectiveD.C.busE1A11andE1C11(Unit 1),E2A11 and E2C11 (Unit 2), and ,
- c. Channel I and Channel IV 125-volt D.C. busses energized from their associated battery banks E1A11 and E1C11 (Unit 1), E2A11 and E2C11 (Unit 2).
APPLICABILITY MODES 5 and 6. ACTION: With any of the abovo required electrical busses not energized in the required manner, immediately suspend all operations involving CCRE ALTERATIONS, positive reactivity changes, or movement of irradiated fuel, initiate corrective action to energize the required electrical busses in the specified manner as soon as p3ssible, and within 8 hours, depressurize and vent the RCS thre;gh at least a 2.0 square inch vent. SURVEILLANCE RE021REMENTS 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. AND SOUTH TEXA5 - UNITS 1 & 2 3/4 8-16 AMEN 0 MENT N05. C. I 71:3
N ELECTRICALPOWERSYSTFF, 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES LIMITING CONDITION FOR OPERATION __ 3.8.4.1 For each containment penetration provided with a penetration conductor overcurrent protective device (s), each device shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With one or more of the containment penetration conductor overcurrent protective device (s) inoperable: . a. Restore the protective device (s) to OPERABLE status or deenergize the circuit (s) by tripping the associated backup circuit breaker or racking out 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 circuit breaker racked out or removed at least once per 7 days thereafter; or
- b. Be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE RE0VIREMENTS l 4.8.4.1 Protective devices required to be OPERABLE as containment penetration conductor overcurrent protective devices shall be demonstrated OPERABLE: i
- a. At least once per 18 months:
- 1) By verifying that the medium' voltage 13.3 kV circuit breakers are OPERABLE by selecting, on a rotating basis, at least 101, of
' the circuit breakers, and performing the following: a) A CHANNEL CALIBRATION of the associated protective relays, b) An integrated system functional test which includes simulated automatic actuation of the system and verifying that each relay and associated circuit breakers and control circuits function as designed, and 3/4 8 17 AMENDMENT N05. AND SOUTH TEXAS - UNITS 1 & 2 k + J 71;g l i
- e hi
!I/
ELECTRICAL POWER SYSTEMS
\'
SURVEILLANCE RE00fREMENTS (Centinued) c) 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 e type shall also be functionally tested until no more failures are found or all circuit breakers of that type have been functionally tested.
- 2) By selecting and functionally testing a rep'resentative sample of at least 10% of each type of lower voltage circuit breakers.
Circuit breakers selected for functional testing shall be selected on a rotating basis. Testing of thu e circuit breakers shall consist of injecting a current with a value equal to 300% of the pickup of the long-time delay trip element and 150% of the pickup cf 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 monufacturer. The instantaneces element shall be testedbyinjectingacurrentequelto120%ofthepickup 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. The instantaneous . element for molded case circuit breakers shall be tested by injecting a current for a frame size of 250 amps or less with , tolerances of +40% -25% and a frame size of 400 amps or greater of225%andverifyIngthatthecircuitbreakertripsinstanta-neously with no apparent time delay. 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 function-i ally tested until no more failures are found or all circuit ,
breakers of that type have been functionally tested; and
- b. At least once per 60 months by subjecting each circuit breaker to an inspection and preventive maintenance in accordance with procedures i ,
prepared in conjunction witn its manufacturer's recommendations. I i SOUTH TEXAS - UNITS 1 & 2 3/4 8-18 AMENDMENT NDS. AND l
FL i 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION LIMITING CONDITION FOR OPERATION 3.9.1 The boron concentration of all filled portions of the Reactor Coolant System 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, v
- b. A boron concentration of greater than or equal to 2500 ppm.
APPLICABILITY: MODE 6." ACTION: , With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes and initiate and continte 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,ff is reduced to less than or equal to 0.95 or the boron concentration is restored to greater than or equal to 2500 ppm, whichever is the more restrictive. SURVEILLANCE REOUIREMENTS 4.9.1.1 The more restrictive of the above two reactivity conditions shall be determined prior to:
- a. Removing or unbolting the 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.
- 4.9.1.2 The boron concentration of the Reactor Coolant System and the refueling i canal shall be determined by chemical analysis at least once per 72 hours.
- 4.9.1.3 Valves FCV-1108, FCV-111B, CV0201A and CV0221 shall be verified i closedandsecured'inpositionbyrechanicalstopsorbyremovalofairor l
electrical power at least once per 31 days. 4.9.1.4 Valve CV0215 shall be verified closed and secured in position by a permanent restraint with the handwheel removed prior to entering MDDE 6. 4 i "The reactor shall be maintained in MODE 6 whenever fuel is in the reactor ! vessel with the vessel head closure bolts less than fully tensioned or with the head removed. SOUTH TEXAS - UNITS 1 & 2 3/4 9 1 AMENDMINT N05. AND kOP j,%, i 1
F0 REFUELING 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 above required monitors inoperable or not operating,
- immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes.
- b. With both of the above required monito'rs inoperable or not operating, determine the boron concentration of the Reactor Coolant System at 1 cast once per 12 hours. .
SUCVEILLANCE REOUIREMENTS 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 ANALOG 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.
l 4 I l 3/4 9-2 AMENOMENT N05. AND SOUTH TEXAS - UNITS 1 & 2 i r ya
I fE i REFUELING OPERATIONS 3/4.9.3 DECAY TIME LIMITING CONSITION FOR ODEPATION 3.9.3 The reactor shall be suberitical for at least 42 hours. APPLICABILITY: During movement of irradiated fuel in the reactor vessel. ACTION-i With the reactor suberitical for less than 42 hours, suspend all operations involving movement of irradiated fuel in the reactor vessel.
~
LIMITING CONDITION F09 OPERATION
- 11. 9 . 3 The reactor shall be determined to havee b'en suberitical for at least 42 hours by verification of the date and time of subcriticality pricr to movement of irradiated fuel in the reactor vessel.
I b I 1 e I SOUTH TEXA5 - UNITS 1 & 2 3/4 9-3 AMENDMENT NDS. AND
!: *1i19?$
l
.7D REFUELING OPERATIONS 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS M'f4 FOR OPERATION 3.9.4 The containment building penetrations shall be in the following status:
- a. The equipment door clcsed 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 being closed by an OPERABLE automatic containment purge and exhaust isolation valve.
APPLICABILITY: During CORE ALTERATIONS or movement of irradiated fuel within the containment. ACTION: With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or movement of irradiated fuel in the containment building. SEYll!.LA3cE REOUIREMENTs 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 purge and exhaust isolation valve within 100 hours prior to the start of and at least once per 7 days during , CORE ALTERATIONS or movement of irradiated fuel in the containment building '. by:
- a. Verifying the penetrations are in their closed / isolated condition, or
- b. Testing the containment purge and exhaust isolation valves per the applicable portions of Specification 4.6.3.2.
4 SOUTH TEXAS - UNITS 1 & 2 3/4 9 4 AMEN 0 MENT N05. AND 4 - l L
N REFUELING OPERATIONS 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. ACTION: When direr.t communications between th( control room and personnel at the refueling station cannot be maintained, suspend all CORE ALTERATIONS. SURVEILLANCE REOUIREMENTS 4.9.5 Direct communications between the control room and personnel at the refueling station shall be demonstrated within 1 hour prior to the start of and at least once per 12 hours during CORE ALTERATIONS. O l l l AND SOUTH TEXAS - UNITS 1 & 2 3/4 9-5 AMENDHENT NDS. bW 1 7 }l,,:; i { t
R REFUELING OPERATIONS 3/4.9.6 REFUELING MACHINE LIMITING CONDITION FOR OPERATION 3.9.6 The refueling machine and auxiliary hoist shall be used for movement of thimble plugs, drive rods or fuel assemblies and shall be OPERABLE with:
- a. The refueling machine used for movement of fuel assemblies having:
- 1) A minimum capacity of 3300 pounds, and ,,
- 2) An automatic overload cutoff less than or equal to 3250 pounds,
- b. The auxiliary hoist used for latching and unlatching drive rods and for thimble plug handling operations having:
- 1) A minimum capacity of 760 pounds, and
- 2) A 1,000 pound load indicator which shall be used to monitor lifting loads for these operations.
~ APPLICABILITY: During movement of thimble plugs, drive rods or fuel assemblies within the reactor vessel. ACTION: With the requirements for the refueling machine and/or hoist OPERABILITY not satisfied, suspend use of any inoperable refueliig machine and/or auxiliary hoist from operations involving the movement of thimble plugs, drive rods and fuel assemblies w?Lhin the reactor vessel. LUcVEILLANCE_EEQUIREMENTS 4.'9.6.1 Each refueling machine used for mov v ent 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 3300 pounds and demonstrating an automatic load cutoff when the refueling machine load exceeds the setooints of Specification 3.9.6a.2). 4.9.6.2 Each auxiliary hoist and associated load indicator used for movement of drive rods 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 760 pounds. SOUTH TEXAS - UNITS 1 & 2 3/4 9 6 AMENDMENT NDS. AND kee
FL REFUELING OPERATIONS 3/4.9.7 CRANE TRAVEL - FUEL HANDLING BUILDING {IMITINGCONDITIONFOROPERATION , 3.9.7 Loads in excess of 2,500 pounds shall be prohibited from travel over fuel assemblies in the spent fuel pool except when carried by the single-failure proof 15-ton hoist of the FHB crane. APPLICABILITY: With fuel assemblies in the spent fuel pool. ACTION:
- a. With the requirements of the above specification not satisfied, place the crane load in a safe condition.
- b. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REOUIREMENTS l i 4.9.7 Loads shall be verified less than or equal to 2,500 pounds prior to movement over fuel assembliss in the spent fuel pool unless they are carried by the single-failure proof 15-ton hoist of the FHB crane. 4 I 4 i d SOUTH TEXA5 - UNITS 1 & 2 3/4 9-7 AMEN 0 MENT N05. ANO liI/171M9
FL REFUELING OPERATIONS 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIF;ULATICN HIGH WATER LEVEL jIMITINGCONDITIONFOROPERATION 3.9.8.1 At least one residual heat removal (RHR) loop shall be OPERABLE and in operation.* APPLICABILITY: MODE 6, when tne water 1: vel above the top of the reactor vessel flange is greater ;1n or equal to 23 feet. ACTION: With no RHR loop OPERABLE and in operation, suspend all operations involving an increase in the reactor decay heat load or a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to OPERABLE and operating status as soon as possible. Close all containment penetrations providing direct access from the containment atmosphere to the outside atm'osphere within 4 hours. SURVEILLANCE REOUIREMENTS 4.9.8.1 At least one RHR loop shall be verified in operation and circulating reactor coolant at a flow rate of greater thar, or equal to 3000 gpm at least once per 12 hours. i "The RHR loop may be removed from operation for up to I bour per 8-hour period during the performance of CORE ALTERATIONS in the vicinity of the reactor vessel hot legs. SOUTH TEXAS - UNITS 1 & 2 3/4 9 8 AMENDMENT N05. AND NOV 17 IL; l
-- - - _ _ - _ = _ _ _ _ _ - _ _
f6 l REFUELING OPERATIONS I LOW WATER LEVEL LIMITING CONDITION FOR OPERATION , 3.9.8.2 Two independent residual heat removal (RHR) loops shall be OPERABLE, and at least one RHR loop shall be in operation." i APPLICABILITY: MODE 'a, when the water level above the top of the reactor vessel flange is less than 23 feet. ACTION:
- a. With less than the reqJired 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 2 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 i
atmosphere within 4 hours. SURVEILLANCE REOUIcEVENTS 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 3000 gpm at least once per 12 hout4. 1 l
"Prior to initial criticality, the RHR loop may be removed from opt.'ation for up to I hour per 8 hour period during the performance of CORE ALTERATIONS in the vicinity of the reactor vessel hot legs.
SOUTH TEXAS - UNITS 1 & 2 3/4 9 9 AMENDMENT N05. AND 1111*.3
f:L REFUELING OPERATIONS 3/4.9.9 CONTAINMENT _ VENTILATION ISOLATION SYSTEM LIMITING CONDITION FOR OPERATION 3.9.9 The Containment Ventilatien Isolation System shall be OPERABLE. APPLICABILITY: During CORE ALTERATIONS or movement of irradiated fuel within the containment. ACTION: t
- a. With the Containment Ventilation Isolation System inoperable, close each of the purge and exhaust penetrations providing direct access from the containment atmosphere to the outside atmosphere,
- b. The provisions of Specification 3.0.3 are not applicable.
SUDVEILLANCE RE0VICEMENTS 4.9.9 The Containment Ventilation Isolation $ystem shall be demonstrated OPERABLE within 100 hours prior to the start of and at least once per 7 days l during CORE ALTERATIONS by varifying that containment ventilation isolation occurs on manual iriitiation and on a High Radiation test signal from each of the RCB purge radiation monitoring instrumentation channels. . t r
. 1 i
f l 4 3/4 9-10 AMENDMENT N05. AND i SOUTH TEXAS - UNITS 1 & 2 ucy i 1 iSTS ! l
b REFUELING OPERATIONS 3/4.9.10 WATER LEVEL - REFUELING CAVITY LIMITING CONDITION FOR ODERATION 3.9.10 At least 23 feet of water shall be maintained over the top of the reactor vessel flar.ge. APPLICABILITY: During movement of fuel assemblies or control rods
- within the refueling cavity when either the fuel assemblies being moved or the fuel assem-bli25 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. SURVElLLANCE REOUIREMENTS 4.9.10 The water level shall be determined to be at least its minimum required 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. l t 1 i !
- Water level require.ments are not applicable when control rods are moved in
! conjunction with the head package during a rapid refueling. l SOUTH TEXAS - UNITS 1 & 2 3/4 9 11 AMENDMENT NDS. AND
} '. " , ' .
FD I REFUELING OPERATIONS 3/4.9.11 WATER LEVEL - STORAGE POOLS SPENT FUEL POOL LIMITING CONDITION FOR OPERATION 3.9.11.1 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 spent' fuel pool. ACTION:
- a. With the requirem'ents of the above specification rot satisfied, ,
suspend all movement of fuel assemblies and crane operations with loads in the fuel storage areas and restore the water level to within its limit within 4 ho. s.
- b. The provisions of Specification 3.0.3 are not applicable, f
SUDVEILLANCE REOUIREMENTS 4.9.11.1 The water level in the spent fuel pon'. shall be determined '.o be at least its minimum required depth at least once per 7 days when irradiated fuel assemblies are in the spent fuel pool. I 1 . 1 I
. l l
i l [ SOUTH TEXAS - UNITS 1 & 2 /4 9 12 AMENDMENT N05. AND i h'T, * ,iij c . - -
FT-REFUELING OPERATIONS IN-CONTAINMENT STORAGE POOL LIMITING CONDITION FOR ODERATION 3.9.11.2 At least 23 feet of water shall be maintained over the top of irradiated fuel assemblics seated in the storage racks. APPLICABILITY: Whenever irradiated fuel assemblies are in the in-containment storage pool. ACTION:
- a. With the requirements of the above specification not satisfied, sus-pend all movement of fuel assemblies and crane operations with loads in the fuel storage areas and estore the water level to withi.) its limit within 4 hours,
- b. The provisions of Specification 3.0.'3 are not applicable.
SURVEILLANCE REOUIREMENTS , 4.9.11.2 The water level in the in-containment storage pool shall be deter-
, mined to be at least its minimum required depth at least once per 7 days when irradiated fuel assemblies are in the in-containment storage pool.
l i e l P i . I i . i 1 SOUTH TEXA5 - UNITS 1 & 2 3/4 9 13 AMENOMENT N35. AND i t
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F( REFUELING OPERATIONS 3/4.9.12 FUEL HANDLING BUILDING EXHAUST AIR SYSTEM ; l I LIMITING CONDITION FOR OPERATION l 3.9.12 The FHB Exhaust Air System comprised of the following components shall be OPERABLE:
- a. Two independent exhaust air filter trains,
- b. Three independent exhaust booster fans,
- c. Three independent main exhaust fans, and
- d. Associated dampers. -
. APPLICABILITY: Whenever irradiated fuel it in the spent fuel pool, l
l ACTION: , i
- a. With less than the above FHB Exhaust Air System components OPERABLE t but with at least one FHB exhaust air filter train, two FHB exhaust '
booster fans, two FHB main exhaust fans, and associated dampers OPERABLE, fuel movement within the spent fuel pool or crane operation I with loads over the spent fuel pool may proceed provided the OPERABLE FHB Exhaust Air System components are capable of being powered from an OPERA 3LE emergency power source and are in operation and discharg-ing through at least one train of HEPA filters and charcoal adsorbers,
- b. With no FHB exhaust air filter train, or less than two FHB exhaust booster fans, or less than two FHB taain exhaust fans and associated l dampers OPERABLE, suspend all operations involving movement of fuel l within the spent fuel pool or crane operation with loads over the '
spent fuel pool until at least one FHB exhaust air filter train, two ! FHB exhaust booster fans, two main exhaust fans, and associated I dampers are restored to OPERABLE status,
- c. The provisions of Specification 3.0.3 are not applicable. i i
M y n L M E WIREMENTS . 4.9.12 The above required FHB Exhaust Air Systems shall be demonstrated OPERABLE: I I
- a. At least once per 31 days on a STAGGERED TEST BAS!$ by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the ,ystem operates for at least 10 continuous hours with the heaters operating with two of the three i exhaust booster fans and two of the three main exhaust fans operating to maintain adequate air flow rate; I
SOUTH TEXAS - UNITS 1 & 2 3/4 9-14 AMENDMENT NDS. AND f
F. REFUELING OPERATIONS JURVEILLANCE REQUIREMENTS (Continued)
- b. At least once per 18 months and (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone communicating with the system by:
- 1) Verifying that the cleanup system satisfies the in place penetratinn and bypass leakage testing acceptance criteria of less than 0.05% for HEPA filter banks and 0.10% for char.- ,
coal adsorber banks and uses the test procedure guidance in i Regulatory Positions C.5.a, C.5.c. and C.5.d of Regulatory ; Guide 1.52, Revision 2, March 1978, and the system flow rate
, is 29,000 cfm i 10%;
- 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 Regulato y Position C.6.a of Regulatory Guide 1.52. Revi-sion 2, March 1978, for a methyl iodide penetration of less than 1.0% when tested at a temperature of 30'C and a relative humidity of 70%; and
- 3) Verifying a system flow rate of 29,000 cfm i 10% during system operation with two of the three exhaust booster fans and two of the three main exhaust fans operating when tested in accordance with ANSI N510-1980. All combinations of two exhaust booster fans and two main exhaust fans shall be tested.
- c. After every 720 hourc 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 1.0% when tested at a temperature of 30*C and a relative humidity of 70%.
- d. At least onde per 18 m.onths 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 flos rate of 29,000 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, SOUTH TEXA5 - UNITS 1 & 2 3/4 9 15 AMENDMENT N05. AND C J 7 n*J
f:s REFUELING OPERATIONS SURVEILLANCE REOUIREMENTS (Continued)
- 3) Verifying that the system maintains the spent fuel storage pool area at a negative pressure of grep.?.er than or equal to 1/8 inch Water GaugJ relative to the 'sutside atmosphere during system oparation, and
- 4) Verifying that the heaters dissipate 50 t 5 kW when tested in accordance with ANSI N510-1980,
- e. After each complete or partial replacement of a HEFA filter bank, by i verifying that the HEPA filter bank satisfies the in place penetration and bypass leakage testing acceptance criteria of less than 0.05% in accordance with ANSI H510-1980 for a DOP test aerosol while operating the system at a flow rate of 29,000 cfm's 10%.
. f. Af ter each complete or partial replace' ment of a charcoal adsorber bank, by verifying that the charcoal adsorber bank satisfies the in place penetration and bypass leakage testing acceptance criteria of less than 0.10% in accordance with ANSI H510-1990 for a halogenated hydrocarbon refrigerant test gas while operating the system at a flow rate of 29,000 cfm 2 10%. ,
i 4 h t SOUTH TEXA5 - UNITS 1 & 2 3/4 9-16 AMENDMENT N05. AND NOV 17 li i i I
FTl 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUT 00WN MARGIN LIMITING CONDITION FOR OPERATION 3.10.1 The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 may be suspended for measurement of control rod worth and SHUTDOWN MARGIN provided reactivity equivalent to at least the highest estimated control rod worth is available for trip insertion from OPERABLE control rod (s). APPLICABILE: MODE 2. ACTION:
- a. With any full-length control rod not fully inserted and with less than the above reactivity equivalent available for trip insertion, immediately initiate and cor.tinue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm boren or its equivalent until the SH'JTDOWN MARGIN required by Specification 3.1.1.1 is restored.
- b. With all full-length control rods fully inserted and the reactor suberitical by less than the above reactivity equivalent, immedi-ately initiate and continue boration at greater than or equal to 30 gom of a solution containing greater than or equal to 7000 ppm boron or its equivalent until the SHUTOOWN MARGIN required by Specification 3.1.1.1 is restored.
SURVEILLANCE REOUIREMENTS 4.10.1.1 The position of each full-length control rod either partially or fully withdrawn shall be determined at least once per 2 hours. 4.10.1.2 Each full-length control rod not fully inserted shall be demonstrated capable of full insertion when tripped from at least the 50% withdrawn position
- within 24 hours prior to reducing the SHUTOOWN MARGIN tu less than the limits of Specification 3.1.1.1.
SOUTH TEXAS - UNITS 1 & 2 3/4 10-1 AMENDMENT N05. AND l';'>.,, 7
FL SPECIAL TEST EXCEPTIONS 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION 3.10.2 The group height, insertion, and power distribution limits of Specifications 3.1.3.1, 3.1.3.5, 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 i: 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 maintained and determined at the frequencies specified in Specification 4.10.2.2 below.
APPLICABILITY: MODE 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:
- a. Reduce THERMAL POWER sufficient to satisfy the ACTION requirements of Specifications 3.2.2 and 3.2.3, or
. b. Be in HOT STANDBY within 6 hours.
SUDVEILLANCLMQVIREMENTS 4.10.2.1 The THERMAL POWER shall be determin j to De 16:e than or equal to
- 85% of RATED THERMAL POWER at least once per hour during PHYSICS TESTS.
4.10.2.2 The requirements of the below listed specificatisns shall be performed at least once per 12 hours during PHYSICS TESTS:
- a. Specifications 4.2.2.2 and 4.2.2.3, and
- b. Specificatic.1 4.2.3.2.
a SOUTH TEXA$ - UNITS 1 & 2 3/4 10-2 AMEN 0 MENT N05. AND
"~* 2 i 1;;;
y FC l SPECIAL TEST EXCEPTIONS 3/4.10.3 PHYSICS TESTS i LIMITING 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: 1
- a. The THERMAL POWER does not exceed 5% of RATED THERMAL POWER,
- b. The Reactor Trip Setpoints on the OPERA 8LE Intermediate and' Power
- Range chennels are set at less than or equal to 25% of RATED THERMAL POWER, and
- c. The Reactor Coolant System lowest operating loop temperature (T,y9) is greater than or equal to 551'F.
APPLICABILITY: A0DE 2. , j ACTION: -
- a. With the THERMAL POWER greater than 5% of RATED ', DERMAL POWER, t immediately open the Reactor trip breakers.
- b. With a Reactor Coolant System operating loop temperature (T**9) less than 551'F, restore T,yg to within its limit within 15 minutes or be in at least HOT STANDBY within the next 15 minutes, i SURVEILLANCE RE0VIDEMENTS ,
4.10.3.1 The THERMAL POWER shall be cetermined 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 sabjected to an
- ANALOG CHANNEL OPERATIONAL TEST within 12 hours prior to initiating PHYSICS TESTS.
4.10.3.3 The Reactor Coolant System temperature (T,yg) shall be determined to be greater than or equel to $51'F at least once per 30 minutes during PHYSICS TESTS. . SOUTH TEXA$ - UNITS 1 & 2 3/4 10-3 AMENOMENT N05. AND
' I lig
Ft SPECIAL TEST EXCEPTIONS 3/4.10.4 REACTOR COOLANT LOOPS ; l LIMITING CONDITION FOR OPERATION
, 3.10.4 The limitations of Specification 3.4.1.1 may be suspended during the f performance of STARTUP and PHYSICS TESTS provided ,
- a. The THERMAL POWER does not exceed the P-7 Interlock Setpoint, t and }
4 b. The Reactor Trip Setpoints on the OPERA 3LE Intermed' ate and Power ! 2 Range channels are set less than or equal to 25% of .a.ATED THERMAL , POWER. ;
. APPLICABILITY: During operation below the P-7 Interlock Setpoint.
ACTION: , With the THERMAL POWER greater than the P-7 Interlock Setpoint, immediately i open the Reactor trip breakers. t SURVEILLANCE REOUIREMENTS ___ . 4.10.4.1 The THERMAL POWER shall be detertrined to be less than P-7 Interlock Setpoint at least once per hour during STARTUP and PHYSICS TESTS. l l 4.10.4.2 Each Intermediate and Power Range channel, and P-7 Interlock shall ' be subjected to an ANALOG CHANNEL OPERATIONAL TEST within 12 hours prior to . initiating STARTUP and PHYSICS TESTS. t 1 i 4 I
- l l -
c i ! a i l SOUTH TEXAS - UNITS 1 & 2 3/4 10-4 AMENDMENT N35. AND 1.c.. ; J
FL SPE0!AL TEST EXCEPTIONS 3/4.10.% POSITION INDICATION SYSTEM - SHUTOOWN LIMITING CONDITION FOR OPERATION 3.10.5 The limitations of Specification 3.1.3.3 may be suspended during the performance of individual full-length shutdown and control rod drop time l measurements provided;
- a. Only one shutdown or control bank N withdrawn from tPs fully inserted position at a time, and
- b. The rod position indicator is OPERABLE during the withdrawal of the rods.*
APPLICABILITY: MODES 3, 4, and 5 during performance of rod drop time measurements. ACTION: With the Position Indication Systems inoperable or with more than one bank of rods wiihdrawn, imr.iediately open the Reactor trip breakers. l SURVEILLANCE RE001pevry7s 1 i 4.10.5 The above required Position Indication Systems shall be deterc;'od to be OPERABLE within 24 hours prior to the start of and at least oncv per ! 24 hours thereaf ter during rod tirop time measurements by verifying the Demand Position Indication System and the Digital Rod Position Indication System agree: 'l
- a. Within 12 steps when the rods are stationary, and
- b. Within 24 steps during rod motion.
2 i . 1 1 I i 1 "This requirement is not applicable during the initial calibration of the Digital Rod Position Indication Sistem provided: (1) K,ff is maintained less than or equal to 0.95, and (2) only one shutdown or control rod bank is withdra.n from the fully inserted position at one time. 1 3/4 10-5 AMENDMENT NDS. AND I SOUTH TEXAS - UNITS 1 & 2 e f.:
i FP; 3/4.11 RA'J0 ACTIVE EFFLUENTS ; 3/4.11.1 LIQUID EFFLUENTS CONCENTRATION LIMITING CONDITION FOR OPERATION. 3.11.1.1 The concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS (see Figure 5.1-4) shall be limited tc the concentrations , specified in 10 CFR Part 20, Appendix B, Table II, Column 2 for radionuclides other than dissolved or .v trained noble gases. For dissolved or entrained ' noble gases, the concentistiun shall be limited to 2 x 10 4 microcurie /mi
.otal activity. l APPLICABILITY: At o11 times.
ACTION: With the concentration of radioactive material released in liquid effluentf to UNRESTRICTED AREAS exceeding the sbove limits, immediately restore the corcen-tration to withiri the above limits. SU~.VEILLANCE REOUIREMENTS l 4.11.1.1.1 Radioactive liquid wastes shall be sampled and analyzed according to the sampling and an lysis program specified in Part A of the ODCH. 4.11.1.1.2 The results of the radioactivity analyses shall be used in accordance - 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 3.11.1.1. t t
. l i
\
l i SOUTH TEXAS - UNITS 1 & 2 3/4 11-1 AMENDMENT N05. AC l l
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r I
FL MA010 ACTIVE EFFLUENTS l DOSE i LIMITING CONDITION FOR OPERATION y 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 UNRESTRICTED AREAS (see Figure 5.1-4) shall be limited:
- a. During any calendar quarter to less than or at;al to 1.5 mress to the whole body and to less than or equal to 5 mrems to any organ, 4 and
- b. During any calendar year to less than or equal to 3 mrams to the ;
whole body and to less than or equal to 10 mrems to any organ. ; APPLICABILITY: At all tiees. I ACTION: i i
- a. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, prepare and submit to the Commissinn within 30 days, pursuant to Specifi u tion 6.9.2, a Special Report that identifies the cause(s) for exceeding 1
the limit (s) and defines the corrective actions that have been taken } to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with
- the above limits. This Special Report shall also include
- (1) the
- results of radiolo ital anelyses of the drinking water source, and ;
- (2) the radiologic 1 impact on finished drinking water supplies with f
! regard to the requirements of 40 CFR Part 141, Oafe Dri-king Water Act.* , b. The provisions of Specification 3.0.3 are not applicable. i j L1L'lLLAN 0VIC'MENTS _ 4.11.1.2 Cumulative dose % '.ributicas from liquid ef fluents for the c.urren calendar quarter and the current calendar year shall be determined in accordance l with the methodology and parameters in the ODCM at least once per 31 days, l i I . ) "The requirements of ACTION a.(1) and (2) are applicable only if drinking water I supply is taken froft the receiving water body within 3 miles of the plant ! discharge. In the case of river-sited plants, this is 3 miles downstream only, i l SOUTH TEXA5 - UNITS 1 & 2 3/4 11-2 AMENDMENT C5. AO ,1 1 M l ,f h r
- , _ _r~~,, , - <. _ . ._-._-_.--m=,-_ =~
Fo RADI0 ACTIVE EFFLUENTS J LIQUID WASTE PROCESSING SYSTE4 _ LIMITING CONDITION FOR OPERATION l 3.11.1.3 The Liquid Waste Processing System shall be OPERABLE and appropriate portions of the system shall te used to reduce releases of radioactivity when ' the projected doses due to the liquid effluent, from each unit, to UNRESTRICTED AP.EAS (see Figure 5.1-4) would exceed 0.06 arem to the whole body or 0.2 arem to any organ in a 31-day period. APPLICABIt!TY: At all times, t ACTION: ,
- a. With radioactive liquid waste being discharged without treatment and in excess of the above limits and any portion of the Liquid Waste Processing System not in coeration, prepare and submit to the Commis-sion within 30 days, pursuant to Specification 6.9.2, a Special Report ,
that includes the following information:
- 1. Explanation of why liquid radmaste was being discharged without treatment, identification of any inoperable equipment or subsystems, and the reason for the inoperability, i 2. Action (s) taken to restore the inoperable equipment to OPERASLE t status, and
- 3. Sum.ary description of action (s) taken to prevent a recurrence, i
! b. The provisions of $pecifications 3.0.3 are not applicable. ; i . r it,iDJ11LLb E REOUIDEu!NTS }4 4.11.1.3.1 Doses due to liquid releases from each unit to UNRESTRICTED AREAS [ shall be projected at least nnee per 31 days in accordance with the methodology 6 and parameter > in the ODCM when Liquid Waste Processing Systems are not being l 4 fully utilized, i 4.11.1.3.2 The '.nstalled Liquid Waste Processing System shall be considered j
- OPERABLE by meeting Specifications 3.11.1.1 and 3.11.1.2.
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l i l ! 3/4 11-3 AMEN 0 MENT N35. AND ! SOUTH TEXAS - UNITS 1 & 2 # -
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i RADIOACTIVE EFFLUENTS LIQUID HUI.0VP TANKS
- l r
l LIMITING CONDITION FOR ODERATION _ 3.11.1.4 The quantity of radioactive material contained in each unprotected outdoor tank shall be limited to less than or equal to 10 Curies, excluding i tritium and dissolved or entrained noble gases: j APPLICABILITY: la all times, t ACTION: : i
- a. With the quantity of radioactive material in any unprotected outdoor tank exceeding the above limit, immediately suspend all additions i of radioactive material to the tank, within 48 hours reduce the *=nk '
contents to within the limit, and describe the events leading to this condition in the next Semiannual Radioactive Effluent Release Report, pursuant to Specification 6.9.1.4. I b. The provisions of Specification 3.0.3 are not applicable. 1 SURVEILLANCE REOUIREMENTS 4.11.1.4 The quantity of radioactive material contained in each unprotected 1 outdoce tank shall be determined to be within the above limit by analyzing a representative sample of the tank's contents at least once per 7 days when i radioactive materials are being added to the tank. 1 l 4 4
\
- Tanks included in this specification are those outdoor tanks that are not surrounded by liners, dikes, or walls capable of holding the tank contents and that do not have tank overflows and surrounding arec drains connected to the Liquid Waste Processing System.
3/4 11-4 AMENDMENT N05. AN? SOUTH TEXAS - UNITS 1 & 2 noi i 1 W
F. RADICACT!VE EFFLUENTS 3/4.11.2 GASE0US EFFLUENTS d, DOSE RATE LIMITING CONDITION FOR OPERATION 3.11.2.1 The dose rate due to radicactive materials released in gaseous , effluents from the site to areas at and beyond the SITE 80VNJARY (see Figure 5.1-3) shall be limited to the following:
- a. For noble gases: Less than or equal to 500 mress/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-
. nuclidesinparticciateformwIthhalf-livesgreaterthan8 days:
Less than or equal to 1500 arems/yr to any organ. APPLICABILITY: At all times. j J *
. ACTION:
With the dose rate (s) exceeding the above limits, immediately restore the release rate to within the above limit (s), j SURVEILLANCE RE0'JIREMENTS
! 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 00CM.
- 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 r.ethodology and parameters in the 00CM by obtaining
, representative samples and performing analyses in accordance with the sampling and analysis program. specified in Part A of the 00CM. i I i ' i l, i SOUTH TEXA5 - UNITS 1 & 2 3/4 11-5 AMENOMENT N05. AND
Ft RADI0 ACTIVE EFFLUENTS DOSE - NOBLE GASES LIMITING CONDITION FOR OPERATION 3.11.2.2 The air dose due to noble gases released in gaseous effluents, from each unit, to areas at and beyond the SITE BOUNDARY (see Figure 5.1-3) shall be limited to the following:
- a. During any calendar quarter: Less than or equal to 5 mrads for gamma radiation and less than or equal to 10 mrads for beta. radiation, and
- b. During any calendar year: Less than or equal to 10 mrads for gamma radiation and less than or equal to 20 mrads for beta radiation.
APPLICABILITY: At all timu. ACTION
- a. With the calculated air dose from radioactive noble gases in gaseous i effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit (s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to a
assure that subsequent releases will be in compliance with the above limits.
- b. The provisions of Secif; cation 3.0.3 are not applicable.
i SURVEILLANCE RE0VIDEuENTS ! 4.11.2.2 Cu ulative 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. i SOUTH TEXA5 - UNITS 1 & 2 3/4 11-6 AMENDMENT N05. AND h0V 171953
F. l RADI0 ACTIVE EFFLUENTS l DOSE - 10 DINE-131, 10 DINE-133, TRITIUM, AND RADI0 ACTIVE MATERIAL IN PARTICULATE FORM LIMITING foNDIT104 FOR OPERATION . ! 3.11.2.3 The dose to a MEMBER OF THE PUBLIC from Iodine-131, Iodine-133, I 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 . i beyond the SITE BOUNDARY (see Figure 5.1-3) shall be limited to the following- '
- a. During any calendar quarter: Less than or equal to 7.5 arems to any '
organ and, j
- b. During any calendar year: Less than or equal to 15 areas to any !
organ. APPLICABILITY: At all times. ; ACTION: l
- a. With the calculated dose from the release of Iodine-131, lodine-133, tritiurr, and radionuclides in particulate form with half-lives l
greater than 8 days, in gaseous effluents exceeding any of the above limits, prepare and submit the the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that identifie the caust(s) e for exceeding the limit (s) and defines the corrective actions that have i been taken to reduce the releases and the proposed corrective actions i to be taken to assure that subsequent releases will be in compliance [' with the above limits. j b. The provisions of Specification 3.0.3 are not applicable.
}
i j SUDVEILLANCE RE00! CEMENTS j 4.11.2.3 Cumulative dose contributions for the current calendar quarter and I current calendar year for lodine-131. Iodine 133, tritium and radionuclides > in particulate form with half-lives greater than 8 days shall be determined ; in accordance with the rethodology and parameters in the 00CM at least once ; per 31 days. : I ( l l I SOUTH TEXAS - UNITS 1 & 2 3/4 11-7 AMENDMENT NOS, AND [ 1 i li.,
' F,t.
RAD 20 ACTIVE EFFLUENTS 1 t GASE0US WASTE PROCESSING SYSTEM { t LIMITING CONDITION FOR OPERATION : 3.11.2.4 The GASEOUS WASTE PROCESSING SYSTEM shall be OPERABLE and appropriate i portions of this system shall be used to reduce releases of radioactivity when the projected doses in 31 days due to gaseous effluent releases, from each unit, ;
- to areas at and beyond the SITE BOUNDARY (see Figure 5.1-3) would exceed
- i i a. 0.2 erad to air from gamma radiation, or i j b. 0.4 mra to air from beta radiation, or !
- c. 0.3 mrem to any organ of a MEMBER OF THE PUBLIC. f 1 .
APPLICABILITY: At all times. l t i ACTION: I l t l a. With radioactive gaseous waste being discharged without treatment i and in excess of the above limits, prepare and submit to the j Commission within 30 days, pursuant to Specification 6.9.2, a 1 Special Report that includes the following information: Identification of any inoperable ageipment or subsystems, and
- 1. ,
the reason for the inoperability. [ l 2. Action (s) taken to restore the inoperable equipment to OPERABLE l 1 status, and r [
- 3. Summary description of action (s) taken to prevent a recurrence. ,
l
- b. The provisions of Specification 3.0.3 are not applicable. ;
i SUCNEILL ANCE REMIRE6'ENTS _ _ I . 4.11.2.4.1 Doses due to gaseous releases from each unit to areas at and I beyond the SITE BOUNDARY sna11 be projected at least once per 31 days in ! , accordance with the methodology and parateters in the 00CM when the GASEOUS ! 1 . WASTE PROCESSIN3 SYSTEM is not being fully utilized. ! ! 4.11.2.4.2 The installed GASEOUS WASTE PROCESSIN3 SYSTEM shall be considered I l OPERABLE by meeting Specifications 3.11.2.1, and either 3.11.2.2 or 3.11.2.3. : i l I t ! f \ I l I l t SOUTH TEXAS - UNITS 1 & 2 3/4 11-8 AMEN? MENT NOS. AND '
- i1503
F RADI0 ACTIVE EFFLUENTS EXPLOSIVE GAS MIXTURE . LIMITING CONDITION FOR OPERATION 3.11.2.5 The concentration of oxygen in the GASE0US WASTE PROCESSING ! SYSTEM inlet shall be limited to less than or equal to 35 by volume. !
] APPLICABIL_ITY: At all tiles.
1 ACTION: ,
- a. With the concentration of oxygen in the GASEOUS WASTE PROCEF51NG SYSTEM inlet exceeding the limit, restore the concentratier to I within the limit within 48 hours. )
- b. The provisions of Specification 3.0.3.are not applicable.
r SURVEILLANCE RE001REWENTS 4.11.2.5 The concentration of oxygen in the GASEOUS WASTE PROCESSING SYSTEM , i shall be determined to be within the above limits by continuously monitoring the waste gases entering the GASEOUS WASTE PROCESSING SYSTEM with the oxygen monitor required OPERABLE by Table 3.3-13 of Snacification 3.3.3.11. l t, n t 1 l 1 1 l i
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. , i 1 ; I l i I l J SOUTH TEXAS - UNITS 1 & 2 3/4 11-9 AMENOMENT N05. AND i t Pf , ,' ] I * ** b
Ft RAD 10ACTI' e EFFLUENTS GAS STO'. AGE TANKS l 1 LIMITI.G CONDITION FOR OPERATION , 3.11.2.6 The quantity of radioactivity o ntained in each gas storage tank shall be limited to less than or equal U 1.0 x 105 Curies of noble gases (considered as Xe 133 equivalent). 2 APPLICABILITY: At all times. ACTION:
- a. With the quantity of radioactive material ir. 2ny gas storage tank
- exceeding the above ?imit, immediately suspend all additions of radioactive material to the tank, within 48 hours reduce the tank 3
contents to within the limit, and describe the events lear!!ng to this condition in the next Semiannual Radioactive Effluent Release Report, pursuant to Specification 6.9.1.4. ,
- b. The provisions of Specification 3.0.3 are not applicable.
- 5UDVEILLANCE REOUIDEMENTS 4.11.2.6 The quantity of radioactive material contained in each gas s' arage tank shall be determined to be within the above limit at least once per 24 hours when radioactive materials are being added to the tank.
1 i { SOUTH TEXA5 - UNITS 1 & 2 3/4 11-10 AMEN 0 MENT N05. AND 1 7 13E:
F, RADI0 ACTIVE EFFLUENTS t 3/4.11.5 SOLIO RA010 ACTIVE WASTES LIMITING CONDITION F0D OPERATION ; 3.11.3 Radioactive wastes shall be solidified or dewatered in accordance with the PROCESS CONTROL PROGRAM to meet shipping and transportation requirements during transit, and disposal site requirements when received at the disposal l site. t APPLICABILITY: At all times. $ ACTION:
- a. With SOLIDIFICATION or dewatering'not meeting dispoJa1 site and '
- shipping and transportstion 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 l recurrence.
- b. With SOLIDIFICATION or dewatering not performed in accordance with l 4 the PROCESS CONTROL PROGRAM, test the improperly processed waste in 4
each container to ensure that it meets burial ground and shipping ! requirerents and take appropriate administrative action to prevtat [
! recurrence.
- c. The provisions of Specification 3.0.3 are not applicable. >
SURVEfttaNCE REOUIREMENTS . 4.11.3 SOLIDIFICATION of at least one re,aresentative test spedmen 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 i 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 additional test specimens can be obtained, alternative SOLIDIFICATION parateters r can be determined in accordance with the PROCESS CONTROL PROGRAM, i and a subsequent test ver fies SOLIDIFICATION. SOLIDIFICATION of the batch may then be resumed using the alternative SOLIOIFICATION parameters determined by the PROCESS CONTROL PROGRAM;
- b. If the initial test specimen from a batch of waste fails to verify SOLIO!FICATION, the PROCESS CONTROL PROGRAM shall provide for the l collection and testing of representative test specimens from each j 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 i in Specification 6.13, to assure SOLIDIFICATION of subsequent batches j of waste; and j t SOUTH TEXAS - UNITS 1 & 2 3/4 11-11 AMENDMENT N05. AND
= _ _ _ _ _ - _ - _ - _ _ _ _ _ _ _ _ _ _ _ _
f *p . I j RADI0 ACTIVE EFFLUENTS ; SURVEILLANCE REQUIREMENTS (continued) -
- c. With the installed equipment inceoable of meeting Specification
; .3.11.3 or declared inoperable, restore the equipment to OPERABLE status or provide for contract capability to process wastes as necessary to satisfy all applicable transport.ation and disposal requirements.
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[ < 1 i i4 1 . 1 \ 1 I i l 1 1 SOUTH TEXAS - UNITS 1 & 2 3/4 11-12 AMENOMENT N05. AND i
Fu RADI0 ACTIVE EFFLUENTS 3/4.11.4 TOTAL DOSE LIMITING CONDITION FOR OPERATION 3.11.4 The annual (calendar year) dose or dose commitment to any MEMBER OF l THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel j cycle sources shall be limited to le.s than or equal to 25 mrems to the whole i body or any organ, except the thyroid, which shall be limited to less than or ) equal to 75 mrems. Appl!CABILITJ: At all times. ACTION:
- a. With the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of 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 incluoing direct radiation contributions from the units snd from outside storage tanks to deter-nine tihether 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 achievirg conformance with the above limits.
This Special Report, as defined in 10 CFR 20.405(c), shall include an analysis that estimates the radiation exFosure (dose) to a MEMBER OF THE PUBLIC from uranium fuel cycle sources, including all effluer.t 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 dese(s) exceeds the above limits, and if the release condition result-ing in violation of 40 CFR Part 190 has not already been corrected, the Special Report shall include a request for a variance in accor-dance with the provisions of 40 CFR Part 190. Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete,
- b. The provisions of Specification 3.0.3 are not applicable.
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SURVEILLANCE RE001cE6'ENTS 4.11.4.1 Cunulative 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 ODCH. 4.11.4.2 Cumulative dose contributions from direct radiation from the units and from radwaste storage tanks shall be determined in accordance with the methodology and parameters in the ODCH. This requirement is applicable only under conditions set forth in ACTION a of Specification 3.11.4. SOUTH TEXAS - UNITS 1 & 2 3/4 11-13 AMLNDMENT NDS. AND I 195 5
Fi 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM LIMITING CONSITION FOR OPERATION 3.12.1 The Radiological Environmental Monitoring Program (REMP) shall be conducted es specified in the ODCH. APPLICABILITY: At all times. AP. TION:
- a. With the Radiological Environmental Monitoring Program not being conducted as specified, prepare and submit to the Commission, in the Annual Radiological Environmental Operating Report required by Specification 6.9.1.1, a description of .the reasons for not conduct-ing the program as required and the plans for pre'!enting a recurrence,
- b. With the level of r&dioactivity as the result of plant effluents in an environmental sampling medium at a specified location exceeding the reporting levels of the REMP when averaged over any calendar quarter, prepare and submit to the Co =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 to be taken to reduce radioactive effluents so that the potential annual dose
- to a MEMBER OF THE PUBLI; is less than tb4 calendar year limits of Specificatiens 3.11.1.2, 3.11.2.2, or ?.11.z.3. When more than one of the radionuclides in the REMP are detected in the sampling redium, this repcrt shall be submitted if:
concentration (1) concentratinn (2)- + **'>- 1.0 reporting level (1) + reporting level (2) When radionuclides other than those listed in the REMP are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose
- to a MEMBER OF THE PUBLIC from all radionuclides is equal to or greater than the cM1endar year limits of Specification 3.11.1.2, 3.11.2.2, or 3.11.2.3. This report is not required if the reasured level of radioactivity was not the result of plant ef fluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report required by Specification 6.9.1.3.
'The methodology and para-eters used to estimate the potential annual dose to a MEMEER OF THE FUBLIC shall be indicated in this report.
3/4 I?-1 AMEN 0 MENT N05. AN: SOUTH TEXA5 - UNITS 1 & 2
1;;;
RAD 10 LOGICAL ENVIRONMENTAL MONITORING LIMITING CONDITION FOR OPERATION ACTION (Continued)
- c. With milk or fresh leafy vegetable samples unavailable from one et more of the sample locations required by the REMP, identify specific locations for obtaining replacement samples and add them within 30 days to the Radiological Environmental Monitoring Program given in the ODCM. The specific locations from which samples were unavafiable may then be deleted from the monitoring program. Pursuant to Spe-cification 6.14, submit in the next Semiannual Radioactive Effluent Release Report documentation for a change in the 00CM including a re-vised figure (s) and table for the 00CM reflecting the new location (s) with supporting information identifying the cause of the unavailabil-ity of samples and justifying the selection of the new location (s) for obtaining samnles,
- d. The provisions of Specification 3.0.3 are not applicable.
SUDNEILLANCE REOUIDEMENTS 4.12.1 The radiological environmental monitoring samples shall be c911ected pursuant to the REMP from. the specific locations given in the table and figure (s) in the ODCM, and shall be analyzed pursuant to the requirements of and the detection capabilities required by the REMP. 3/4 12-2 AMEN 0 MENT N05. AND . SOUTH TEXA5 - UNITS 1 & 2 ,
f RADIOLOCICA!, ENVIRONMENTAL MONfTORING 3/4.12.2 LAND USE CENSUS LIMITIN3 CONDITION FOR ODEDATION l 3.12.2 A Land Use Cenws shall be conducted and shall identify within a distance of 8 k.m (5 miles) the location in each of the 16 meteorological sectors of the nearest milk animal, the nearest residence, and the nearest garden
- of greater than b0 m 2 (500 ft 8) producing broad leaf vegetation.
APPLICABILITY: At all times. ACTION:
- a. With a Land Use Consus identifying a location (s) t5at yields a calculated dose or dose coreitment greater than the values currently being calculated in Specification 4.11.2.3, pursuant to Specifica-tien 6.9.1.4, identify the new location (s) in the next Semiannual Radioacthe Effluent kelease Report.
- b. With a Land Use Census identifying.a location (s) that yields a calculatcc dose or dose cor.mitment (via the same exposure path.ay) 20% greater than at a location from which samples are currently being ebtained in accorcance witn 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 com-iteent(s), via the sate exposure path ay, may be deleted fron this ronitoring 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 docu?entation for a change in the ODCM including a revised figure (s) and table (s) for the CDCM reflecting the new location (s) with inforea-tion ,upporting the change in sampling locations,
- c. The provisions of Specification 3.0.3 are not applicable.
- Broad leaf vegetation sa-pling of at least three dif ferent kinds of vegetation may be perforred at the SITE EDUNDARY in each of two different direction sectors with the highest predicted 0/Qs in lieu of the garden census. Speci-fications for brca: leaf segetatien sa pling in the REW shall be follo ed, inclu::ing analysis of control sa ples.
SOUTH TE AS - UNITS 1 & 2 3/4 12 3 AMEN; MENT h05. AS?
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f RADIOLOGICAL ENVIRONMENTAL MONfTORING SURVEILLANCE RE001REMENTS 4.12.2 The Land Use Census shall be conducted at least once per 12 months using that information that will provide the best results, such as by a door-to-door survey, aerial survey, or by consulting local agriculture authorities, l as described in the 00CM. The results of the Land Use Census shall be included in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.3. I t i 1 SOUTH TEhAS - UNITS 1 & 2 3/4 12-4 AMEN; MENT N05. AN? { .- - . . .
f7 RADIOLOGICAL ENv!RONMENTAL MONITORING 3/4.12.3 INTERLABOR.* TORY COMPARISON PROGRAM LIMITINGCON21 TION {0RODED.ATION 3.12.3 Analyses shall be performed on all radioactive materials, supplied as part of an Interlaboratory Comparison Program that has been approved by the Commission, that corrispond to samples required by the REMP. APPLICABILITY: At all tires. ACTION:
- a. With analyses not b0ing performed as required above, report the corrective actions taken to prevent a recurrence to the Commission '
in the Annual R6diological Envirtnmental Operating Report purs. ant to Specification 6.9.1..).
- b. The provisions of Specification 3.0.3 are not applicable.
it'DVEILLANCE REQ 2]E.EMENTS 4.12.3 The Interlaboratory Comparison Program shall be described in the 00;M. A sum. mary of the results obtained as part of the above required Interlaboratory Comparison Program shall be included in the Annual Radiological Environmentrl Operating Report pursuant to Specification 6.9.1.3. > SOUTH TEXA5 - UNITS 1 & 2 3/4 12-5 AMEh0 MENT h05. ANO l'T V 1 7 1;j!
FD ' i EMERGENCY CORE COOLING SYSTEMS j M @ M g TS (continued).
- e. At least,once per 18 P.onths, during shutdowa, by: )
f
- 1) Verifying tnat each automatic valve in the flow path actuates to l its cocrect position an an Automatic Switchover to Containment i Sump test signAi, ano ;
I ., 2) Varifying that each of the following pumps start automatically 4 u;.on receipt of a Safety. Injection test signal: , j a) High Head Safety Injection pump, and [ b) Low Head Safety Injection pump.
- f. By verifying that each of the following pumps develops the indicated [
differential pressure on recirculation flow when tested pursuant to Specification 4.0.5: . l
- 1) High Head Safety Injection pump > 1480 psid, and
- 2) Low Head Safety In, action pump > 286 psic.
, j
- g. By performing a flow test, during shutdown, following completion of l modifications to the ECC5 subsystems that alter the subsystem flow (
characteristics and verifying that. , t
- 1) ForHighHeadSafetyInjectionpumplines,withtheHighHead Safety Injection pump running, the pump flow rate is greater ,
than 1470 gpm and less than 1620 gpm. lt
- 2) For low Head Safety Injection pump lines, with the Low Head l Safety Injection pump running, the pump flow rate is greater l than 2550 gpm and less than 2800 gpm. !
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I [ i I SOUTH TEXA5 - L.,ITS 1 & 2 3/4 5-5 WENOMENT NDS. AO i
*' } I *${3
F. BASES FOR SECTIONS 3.0 AND 4.0 LIMITIN3 CONDITIONS FOR OPERAT.'ON AND SURVEILLANCE REQUIRtvENTS O e
FJ NOTE The B',5ES contained 'i succeedinc pages sumarize the reasons for the ,,ecifications in Sections 3.0 and 4.0, but in accordance with 10 CFR 50.36 are not part of these Technical Specifications. l l 1 f a l l l l
FL i 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVE!LLANCE REQUIREMENTS I 3.4.0 APPLICABILITY 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 stated in the Code of Federal Regulations, 10 CFR 50.36(c)(2)- ,
"Limiting conditions for operation are the lowest functional capability i or performance levels of equipment required for safe operation of the l facility, When a limiting condition for operation of a nuclear reactor is not t met, the licensee shall shut down the reactor or follow any remedial action. L permitted by the technical specification until the condition can be met." l Specification 3.0.1 establishes the Applicability' statement within each [
individual specification as the requirement for whwn (i.e., in which ' OPERATIONAL MODES or other specified conditions) conformance to the Limiting > Conditions for Operation it required for safe operation of the facility. The : ACTION requirements establish those remedial measures that must be taken within specified tire limits when the requireteents of a Limiting Condition for Operation are not ret. , There are two basic types of ACTION requirements. The first specifies the reeedial measures that permit continued operation of the facil uy which is not further restricted by the tire limits of the ACTION requirements. In this : t case, con.'ormance to the ACTION requirements provides an acceptable level of safety for unlimited continued operation as long as the ACTION requirements ! continue to be met. The second type o' ACTION requirement specifies a time limit in which conformance to the conditions of the Limiting Condition for I Operation rust be met. This time limit is the allowable outage tire to t restore an inoperable system or component to OPERABLE status or for restoring
- i parameters within specified limits. If these actions are not completed within the allo.able outage tire limits, a shutdown is required to place the facility ;
in a MODE or condition in which thc specification no longer applies, it is l not intended that the shutdown ACTION requirements te used as an operational ; convenience which pereits (routine) voluntary reroval of a system (s) or component (s) from service in lieu of other alternatives that would not result l in redundant systems or components being inoperable. [ The specified tire lirits of the ACTION requirements are applicable from the ! point in tine it is identified that a Limiting Conditicn for Operation is not ret. The tire limits of the ACTION requirements are also applicable when a ; system or component is removed from service for surveillance testing or investigation of cperational problems. Individual specifications may include j a specified tire limit for the corpletion of a Surveillance Requir Tent when equipment is removed from service. In this case, the allowable outage tire r I 5 3/4 0-1 AMEN; MENT N05. ANO l SOUTH TEXA5 - UNITS 1 & 2 L': , 2 : ,,
F. l i i l 3.4.0 APPLICABILITY _ ( l BASES (Continued) ~ I i i ! limits of the ACTION requirements a*e applicable when this limit expires if t the survelliance has not been comp 1sted. When a shutdawn is required to comply with ACTION requirements, the plant say have entered a MODE in which a [ ! new specification becomes applicable. In this case, tae time limits of the i d ACTION requirements would apply from the point in thee that the new r
- specification becomes applicaole if the t@r**c;its of the Limiting Condition !
] for Operation are not ret, j 9 - Specification 3.0.2 establishes thut noncompliance with a specification exists [ ! when the requirements of the Limiting Condition for Operation are not met and
< the associated ACTION requirenents have not been implemented within the }
i specified time interval. The purpose of this specification is to clarify that f (1) implerentation of the ACTION requirerents within the specified time i
- interval constitutes compliance with a specification and (2) completion of the t a remedial measures of the ACTION requirements is not required when compliance I with a Limiting Condition for Operation is restored within the tire interval l specified in the associated ACTION requirements, t
{ Specificatien 3.0.3 establishes the shutdown ACTION requirements that must be irplementee when a Limiting Condition for Operation is not ret and the { { condition is not spfeifically addressed by the associated ACTION requirements, j l The purpose of this specification is to delineate the time limits for placing + 1 the unit in a safe shutdo n MODE when plant operation cannot t'e maintained I j within the limits for safe operation defined by the Limiting Conditions for ! Operation and its ACTION requireeents. It is act intended to be used as an F 4 operational convenience which permits (ruutine) voluntary removal of redundant { ! syster.s or coeponents from stevice in lieu oT other alternatives that would not r j result in redundant systems or components be.'ng inoperable. One hour is I , allowed to prepare for an orderly shutdown before initiating a change in plant i i operation. This tire permits the operator to coordinate the reduction in i i electrical generation with the load dispatcher to ensure the stability and i
- availability of the electrical grid. The tire limits specified to rea
- h lower i MODES of operation permit the shutdow.n to proceed in a controlled and orderly f ranner that is well within the specified maximut. cooldown rate and within the !
cooldown capabilities of the facility assuming only the n'nimum riquired I equiprent is CFERAELE. This reduces thermal stresses on coeponeras of the primary coolant syste and the potential for a plant upset that ,:ould challenge safety systets uncer conditions for which this specification applies. I, remedial ressures permitting limited continued operation of *,he f acility under the provisiens of the ACTION requirements are completed, the shutdown ; ! may be terminated. The tire limits of the ACTION requirements are applicable ; ! f rom the point in tire there sin a f ailure to Peet a Limiting Condition for i Operation. Therefore, the 6hutco*n eay be t,erminated if t,he ACTION requirerents have been net or the tire limits of the ACTION require ents have not empired, thus providing an allo.ance for the completion of the required actions. SOUTH TEXA5 - UNITS 1 & 2 E 3/4 0-2 AMEN: MENT NDS. ANO
- m3
F 3.4.0 APPLICABfLITY BASES (Continued) The time limits of Specification 3.0.3 allow 37 hours for the plant to be in the COLD SHUTDOWN NDE when a shutdown is required during the POWER NDE of operation. If the plant is in a lower MODE of operation when a shutdown is required, the time limit for reaching the next lower N DE of operation ap-plies. However, if m Swer MODE of operation is reached in less time than allowed, the total als sable tire to reach COLD SHUTDOWH, or other appli',able MODE, is not reduced. ior example, if HOT STANDBY is reached in 2 hours, the time allowed to reach HOT SHUTDOWN is the next 11 hours because the total tire to reach HOT SHUTDOWN is not reduced from the allowable limit of 33 hours. Therefore, if remedial measures are completed that would permit a return to POWER operation, a penalty is not incurred by having to reach a lower MODE of cperation in less than the total time allowed. The same principle applies with regard to the allowable outage time limits of the ACTION requirements, if compliance with the ACTION requir ments 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 ret, if the new specification becomes applicable in less tire than specified, the difference may be added to the allowable outage tire limits of the second specification. Homever, the allowable outage tire limits of ACTION requirements for a higher MODE of operation ray not be used to extend the allowable outage tire that is applicable when a Liriting Condition for Operation is not ret in a lower MODE of operation. The shutdown requiretents of Specification 3.0.3 do not apply is MODES 5 and 6, because the ACTION requirements of individual specifications define the remedial reasures to be taken. Specification 3.0.4 establishes limitations on MODE changes when a Limiting Condition for Operation is not ret. It precludes placing the facility in a higher MODE of operatier when the requirements for a Limiting Condition for Operation are not ret and cortinued noncompliance to these conditions would result in a shutdown to comply with the ACTION requirements if a change in NDES were perritted. The purpose of this specification is to ensure that facility operation is not initiated or that higher MODES of operation are net entered when ccrrectist action is being taken tc obtain co plianct with a specification by restoring equip"ent to OPERABLE status or parameters to specified limits. Co p11ance with ACTION requirements that pereit continued operation of the facility for an unlieited period of tire provides an accept-able level of safety for continueJ cperation without regard to the status of the plant before or after a MODE change. Therefore, in this case, entry into an OPERATIONAL MDDE or other specified condition ray be race in accordance with the provisions of the ACTION requireeents. The provisions of this specification should net, henever, be interpreted as endorsing the failure to enercise good practice in restoring systets or components to OPERAELE status before plant startup. 50'JTH TE AAS - UN!T51 & 2 E 3/4 0-3 Au!N MENT N05. AN: D' 17f;g
! i i fp > 2 I 3.4.0 APPLICABILITY 5ASES (Continued) i When a shutdown is required to comply with ACTION requirements, the provisions f i of Specification 3.0.4 do not apply oecause they would delay placing the l a facility in a loser MODE of operation. j j$ ecification 3.0.5 delineates the applicability of each specification to 1 unWTand Unit 2 Operation. f Specifications 4.0.1 throuch 4.0.5 establish the general requirements I applicable to Surveillance Requirements. These requirements are based on the r Surveillance Requirements stated in the Code of Federal Regulations, ! 10 CFR 50.36(c)(3): ?
" rvaillance requirerents are requirements relating '.o test, calibra- I
! tier . inspection to ensure that the necessary quality of systems and ' components is esintained, that facility operation will de witLin safety I j limits, and that the limiting conditiers of operation will be set." l j Specification 4.0.1 establishet the requirement.that surveillances must be ! performed during trie OFIRATIONAL MODES or other conditions for which the requirements of the Limiting Conditions for Operation apply unless otherwise stated in an individual Surveillance Requirerent. The purpose of this speci-1 fication is to ensure that surveillances are perforced to verify the cpera-l tional status of systems and components and that parameters are within spect- { fled limits to ensure safe operation of the facility when the plant is in a 3 MODE or other specified condition for which the allociated Limiting Conditions ! for Operation are applicable. Surveillance Requirements do not have to be 1 performed when the facility is in an OPERATIONAL MODE for which the requirerents i of the associated Limiting Condition for Operation do not apply nless otherwise { specified. The Surveillance Requirements associated with a Special Test i Exception are only applicable when the Special Test Exceotion is used as an j alle.able exception to the requirerents of a specification. 4 Specification 4.0.2 establishes the condittens under which the specified tire 2 intersal for 5urseillance Requirerents ray be extended. Item a. permits an ( allowable extension of the nornal surveillance interval to facilitate e surveillance scheduling and consideration of plant cocrating conditions that ! ray not be suitable for condutting the surveillance; e.g., transient conditions
- or other ongoing surveillance or maintenance activities. Item b. limits the use of the provisions of itet a to ensurt 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 judstent and the recognition that the rest probable result of any particular surveillant.e being performed is the verification of conformance with the Surveillance Requireeents. These provisions are sufficient to ensure that the reliability ensured through surveillance activit,ies is not significantly degraded beyond that obtained from the specified surveillance interval.
Specification 4.0.3 establishes the failure to perform a Surveillance Requirerent within the allo =ed surveillance interval, defined b) the provisions of Specification 4.0.2, as a condition that constitutes a failure to rett the SOUTH TEXA5 - UNITS 1 & 2 E 3/4 0 4 AMEN 0YENT N05. ANO
FL i, ! 3.4.0 APPLICABILITY l t 4 BASEA (Continued) j L 3 OPERABILITY requirements for a Limiting Condition for Operation. Under the ; i provisions of this specificap on, systems ard components are assumed to be i OPERABLE when Surveillance Requirements have been satisfactorily performed , 4 within the specified time interval. However, nothing in this provision is to
- be construed as implying that systems or components are OPERABLE when they are ,
found or known to be inoperable although still meeting the Surveillance l Requirements. This specification also clarifies that the ACTION requirements ! are applicable when Surveillance Requirements have not been completed within l the allowed surveillance interval Ond that the time limits of the ACTION 6 requiraments apply from the point b' time it is identified that a surveillance I has not been performed and not at tht time that the allowed surveillanen inter- l val was exceeded. Completion of the Surveillance Requirement within the allow- l able outage time limits of the ACTION -)quirements restores compliance with i the requirements of Specification 4.0.3. However this does not negate the ! factthatthefailuretohaveperformedthesurvel11ancewithintheallowed : surveillance interval, defined by the provisions of Specification 4.0.2, was a l violation of the OPERABILITY requirements of a Limiting Condition for Opera- l tion that is subject to enforcement action. .Further, the failure to perform i a surveillancs within the provisions of Specification 4.0.2 is a v.alation cf ; a Technical Specification requirerent and is, therefore, a reportable event ! under the requirements of 10 CFR $0.73(aM2)(1)(B) because it is a condition ; prohibi ed by the plant's Technical Specifications. [ r If the allowable outage time limits of the ACTION requirerents are less than : 24 hours or a shutdo n is required to cotply with ACTION requireeents, e.g., I Specification 3.0.3, a 24 hour allowance is provided to permit a delay in i implementing the ACTION requirements. This provides an adequate time limit to ! complete Surveillance Requirements that aave not been performed. The purpose [ of this allowance is to permit the completion of a surveillance before a shut- [ down is required to comply with ACTION requirements or before other remedial : measures would be required that may preclude completion of a surveillance, j The basis for this allowance includes consideration for plant conditions, ade- i quate planning, availability of personnel, the time required to perform the i surveillance, and the safety significance of the delay in completing the { required surveillance. This provision also provides a tire limit for the cor- ! pletion of Surveillance Requirerents that becore applicable as a consequence of MODE channes irpesed by ACTION requirements and for completing Surveillance Requirerent' t. hat are applicable when an exception to the requirements of i Specificati a 4.0.4 is allo =ed. If a surveillance is not coepleted within the + 24-hour allowance, the tire limits of the ACTION requirerents are applicable , at that t he. When a surveillance is performed within the 24-hour allowance and the Surveillance Requirements are not ret, the time lieits of the ACTION ! reovirements are applicable at the tire that the surveillance is terminated. Surveillance Requirerents do not have to be perforced en inoperable equiprent ; because t,he ACTION requirements define the remedial measures that apply. Ho=- ' ever, the Surveillance Requirements have to be ret to demonstrate that inoper-able equipeent has been restored to OPE' TABLE status. , I i SOUTH TEXA5 - UNITS 1 & 2 E 3/4 0-5 AMihCMENT h:5. AND E W ] ; t.
FL 3.4.0 APPLICABILITY BASES (Continued) i
. Specification 4.0.4 establishes the requirement that all applicable surveil- '
I lances must be met before entry into an OPERATIONAL MODE or other condition of operation specified in the Applicability statement. The purpose of this i specification is to ensure that system and component OPERASILITY requirements or parameter limits are met before entry into a M30E or condition for which ' I these systems and components ensure safe operation of the facility. This i provision applies to changes in OPEkATIONAL MODES or other specified conditions < associated with plant shutdown as well as startup. l Under the provisions of this specification, the applicable Surveillance Requirements must be performed within the specified surveillance interval to f } ensure that the Limiting Conditions for Operation are met during initial plant ! ! startup or follo.ing a plant cutage. L 1 t When a shutdown is required to comply with ACTION requirements, the provisions ! of Specification 4.0.4 do not apply because this would delay placing the facil-j ity in a lower MODE of operation. , t j 5pecification 4.0.5 establishes the requirement that inservice inspection of 1 A5ME Coce Class 1, 2, and 3 coeponents and inse vice testing of ASME Code i Class 1, 2, and 3 pumps and valves shall be performed in accordance with a i . periodically updated version of Section XI of the ASME Boiler and Pressure j ! Vessel Code and Addenda as required by 10 CFR 50.55i. These requirements apply : l except when relief has been provided in writing by the Comission. ! l 1 l This specification includes a clarification of the frequencies for performing the inservice inspection and testing activities requirec by Section XI of the ASME Boiler and Pressure Vessel Code and applicable AdJenda. This :larifica- i tion is provided to ensure consistency in surveillance intervals throughout ! J the Technical Specifications and to remove any ambiguities relative to the J frequencies for perforcing the required inservice inspection and testing activities. i I 1 Under the teres cf this specification, the rore restrictive requirements of [
- the Technical Specificatiens take precedence over the A5ME Boiler and Pressure !
! Vessel Code and applicable Accenda. The requirements of Specification 4.0.4 l 1 to perform surveillance activities before entry into an OPERATIONAL MODE or l 1 other specified condition takes precedence over the ASME Boiler and Pressure ! ! Vessel Code provision which allo s peps and valves to be tested up to one l J week after return to normal operation. The Technical Specification definition > ! of OPERABLE does not alio a grace period before a component, that is not cap- l } able of performing its specified function, is declared ineptrable and takes l 1 precedence over the ASME Boiler and Pressure Vessel Code provision which j l allons a valve to be incapable of performing its specified function for up to
- 24 hours before being declared inoperable. ,
1 l i Specification 4.0.6 delineates the applicability of the surveillance activities ! ] to Unit 1 anc Unit 2 operations. l l ) I i i SOUTH TEXAS - UNITS 1 & 2 8 3/4 0 6 Au!NDMENT h05. AND l 4 r 1 ; a ! i l
Ft : 3/4.1 REACTIVITY CONTROL SYSTEMS ; RASES 3/4.1.1 BOR G10N CONTROL : 3 /4.1. ?. . I and 3/4'.1.1.? SHUTDOWN PARGIN A suf ficient SHUTD0=H MARGIN ensures that: (1) the reactor can be modo sub:ritical from all operating conditions, (2) the reactivity transients asso- t ciated with postulated accident conditions are controllable within acceptable , limits, and (3) the reactor will be maintained sufficiently suberitical to l preclude inadvertent criticality in the shutdown condition. .. - i SHUT 40WN MARGIN requirements vary throughout core life es a function of : fuel depletion, RCS boron concentration, and RCS T ,g. In MODES 1 and 2, the ! most restrictive condition occur > at E0L, with T,yg at no load operating ; I temperature, and is associated with a postulated steam line brtak accident and ! resulting uncontrolled RCS cooluown. In the anaiysis of this accident, a l j minimum $HUTOOwH MARGIN of 1.75% ok/k is required to control the reactivit) transient. The 1.75% at/k SkUTDOWH MARGIN is the Gesign basis minimum for the ; 14 foot fuel using Hafnium control rods (Ref. FSAR Table 4.3-3). Accordingly, {' the SHUTD0hh MARGIN requirement for MODES 1 and 2 is based upon this limiting condition anti is consistent with FSAR safety analysis assumptions. In MODES ! s 3, 4, and 5, the most restrictive condition occurs at BOL, when the boron concen- [
- tration is the greatest, in these modes, the required SHUTOOWN MARGIN is cor.- i
- posed of a constant requirement and a variable requirement, which is a function !
} of the RCS boron concentration. The constant $HUTD0hH MARGIN requirerent of l 1.75% ok/A is based on an uncontrolled RCS cooldown from a steamline bresk ac,:ident. The variable SHUT 00hh MARGIN requirement is based on the results ! of a boron dilution accident analysis, where the SHUTOOWN MARGIN is varied as a { l function of RCS boron concentration to guarantee a minimum of 15 minutes for i ~ operator action after a boren dilution alarm, prior to a loss of all 5HU100WN
- l i
- MARGIN.
1 ! The boron dilution analysis assured a coreen RCS velume, and maxinu?. f a dilutten flew rate for MODES 3 and 4, and a different voluwe and flow rate for ! MODE 5. The MWE 5 conditions ansced limited mining in the RCS and ecoling i 1 wit'. the RHR syste- only. In MODES 3 and 4 it was assured that, at least one i reactor coolant pu p was operating. If at least one reactor coolant pu?p is l not operating in MODE 3 or 4, then the 5HUTDOWH MAR 5!N requirements for MODE 5 [ shall apply. l ! 3/4.1.1.3 MODERATOR TE@ ERAfuRE COEFFICIENT d i i The limitations on mor,erater temperature coefficient (MTC) are provided :
- to ensure that the value el this coef ficient remains within the liutt.ing condition assumed in the FSAR accident and transient analyses. [
The MTC values of ihis specification are applicable to a specific set of ! i plant conditions; accordingly, verification of MTC values at conditions other ; than those explicitly stated will require extrapolation tc those conditions in ; j order to permit an accurate cerparison, 1 J
! SOUTH TEXA5 - UNITS 1 & 2 B 3/4 1-1 AMEN 2 MINT N05. ANO [
4
~
Fi i I REACTIVITY CONTROL SYSTEMS 1 EA5t5
$0ERATORTEMPERATURECOEFFICIENT(Continued)
The most negative MTC, value equiveient to the most positive moderator density coefficient (MDC), was obtained by incrementally correcting the MOC used in the FSAR analyses to nominal operating conditions. These corrections involved sui,tearting the incremental change in the MOC associated with a core , condition of all G d: inserted (most positive MDC) to an all rods withdrawn condition and, a conversion for the rate of change of moderator density with temoerature at RATED THERML POWER conditions. This value of the MDC was then transformed into the limiting MTC value -4.0 x 10 4 Ak/k/* F. The MTC value of -3.1 x 10 4 Ak/k/'F represents a conservative value (with corrections for burnup and soluble baron) at a core condition of 300 ppm equilibrium boron concentration and is cbtained by making these corrections to the limiting MTC value of 4.0 x 10 4 Ak/k/'F. 4 The Surveillance Requirements for reasurement of the KTC at the beginning t and near the end of the fuel cycle are adequate to confirm that the MTC remains j within its limits since this coefficient changes slowly due principally to the i reduction in RCS boron concentration associated with fuel bur mp. 3/4.1.1.4 MINIPJM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical ! with the Reactor Coolant System everage terperature less than 561'F. This j limitation is required to ensure: (1) the moderator terperat:*re coef ficient is within its analyzed teeperature range, (2) the trip instrueentatiun is within 1 its normal operating range, (3) the pressurizer is capable of being in an OPERABLE status with a steam bubble, and (4) the reactor vessel is'.nove its minimum Rig )7 temperature.
- 3/4.1.2 BORAT!0N SYSTEMS
) The Boron Injection System ensures that negative reactivity control is available curing each rode of facility operation. The components required to ] l perform this functicn include: (1) berated water sou ces, (2) charging pu-ps, j (3) separate ficw paths. (4) boric acid transfer put;s, end (5) an emergency j power supply fro?. OPL 6 ELE diesel generators. 1 ) With the RCS average tee;e ature above 350'F, a minimam of two boron injection flo. paths are required to ensure single functional capabllity in
- the event an assured f ailure renders one of the flow paths twp* 4ble. The
; boration capability of either flo. path is sufficient to per 9 .* $ HUT 00.'N MARGIN f rom expected operatik; conditions of 1.75% AL/A af t m < decay and coolde.n to 200'F. The raaien expected boration capalli W rerent occurs at ICL from full po.er easilierium menon conditions an, 'dres 27,000 gallons of 7000 ppm borated water from the boric acid sh rage systet 1 or 458,000 gallons of 2500 ppm borated water from the refueling water sto sge j tank (NST). The RWST solu e is an ECCS requirement and is core than adequate J for tf
- required boration capability.
SOUTH TEXAS - UNITS 1 & 2 E 3/4 1-2 AMENNENT h'5. AND
*I: q
E l l REACTIVITY CONTROL SYSTEMS _ BASES l BORATION SYSTEMS (Continued) With the RCS temperature f.e'.ow 350'F, one boron injection flow path / source 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 flow path / source becomes inoperable. The limitation for a maximum of one charging pump to be OPERABLE and the Surveillance Requirement to verify all charging preps except the requireo OPERABLE pump to be inoperable below 350'F provides assurance that a mass addi-tion pressure transient can be relieved by the operation of a single PORV. The boration capability required below 200'F.is sufficient to provide a variable SHUTDOWN MARGIN based on the results of a boron dilution accident analysis where the SHUT 00WN MARGIN is varied as a function of RCS horon concen- . tration after xenon decay and coo?down from 200'F to 140*F. 91s condition requires either 2900 gallons of 7000 ppm borated water from . s boric acid storage system or 122,000 gallons of 2500 ppm borated water from the RWST for MODE 5 and 33,000 gallons of 2500 ppm borated water from the RWST for H00E 6. The contained water volume limits include allowance for water not available because of discharge line location and other physical characteristics. l The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between 7.5 and 10.0 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 cerrosion on ! mechanical systems and components. l The OPERABILITY of one Boron Injection System during REFUELING ensures l that thi. system is available for reactivity control while in MODE 6. 3/4.1.3 H3VABLE CONTROL ASSEMBLIES , lhe specifications of this section ensure that: (1) acceptable powe-distribution limits are maintained, (2) the minimum SHUT 00wN MARGIN is main-tained, and (3) the potential effects of rod misalignment on associated acci-dent analyses arr limited. OPERASILITY of the control rod position indicators
. is required to determine control rod positions and thereby ensure compliance - with the control rod aligncent and insertion limits. Verificadon that the Digital Rod Position Indicator agrees with the demanded position within i 12 .t steps at 24, 48, 120, and 259 steps withdrawn for the Control Banks and 18, 234, and 259 steps withdrawr. for the Shutdown Banks provides assurances that the Digital Rod Position Indicator is operating correctly over tho full range of indication. Sirce the Digital Rod Position Indication System does not indicate the actual shutdown rod position between 18 steps and 234 steps, only points in the indicated ranges are picked for verification of agreement with demanded position.
50U1H TEXAS - UNITS 1 & 2 B 3/4 1-3 AMENDMENT N05. AND
' -~
1 sm I
FE REACTIVITY CONTROL SYSTEMS BASES ,, MOVABLE CONTDOL ASSEMBLIES (Continued) The ACTION statements which permit limited variations from the basic requirements are accompanied by additional restrictions which ensure that the original design criteria are met. Misalignment of a t-od requires measurement of peaking factors and a restriction in THERHAL POWER. These restrictions pro-vide 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 safety analyses. Me3surement with T,yg greater than or equal to 561*F and with all reactor coolant pumps' operating ensures that the measured drop times will be repre$eatative of insertion times experienced during a Reactnr trip at operating conditions. 4 Control rod positions and OPLRABILITY of the rod position indicators are required to be verified on a nominal basis of onea per 12 hours with more fre-quent verifications required if an automatic monitoring channel is inoperable. These verification frequencies art adequate for assuring that the applicable LCOs are satisfied. l a d l a L SOUTH TEXAS - UNITS 1 & 2 B 3/4 1-4 AMEN 0 MENT N05. AND , J 7 lid 2
- F. 3/4.2 POWER DISTRIBUTION LIMITS l l l BASES The specifications of this section provide assurance of fuel integrity " during 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 fission gas release, fuel pellet temperature, and cladding mechanical pro- " perties to within assumed design criteria. In addition, limiting the peak ! linear power density during Condition I events provides assurance that the ' initial conditions assumed for the LOCA analyses are met and the ECCS. ace ance criteria limit of 2200'F is not exceeded. ; The definitions of certain hot channel and peaking factors as used in 7 these specifications are as follows: ! Fg (Z) Heat Flux Hot Chancel 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 i fuel pellets and rods; , F g Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of l the integral of linear power along the rod wi6h the highest integrated l power to the average rod power; and l F,j(Z) Radial Peaking Factor, is defined as the ratio of peak power density ! to average power density in the horizontal plane at core elevation Z. ! 3/4,2.1 AXIAL FLUX DIFFERENCE The limits on AXIA'. FLUX DIFFERENCE (AFD) assure that the Fg (Z) upper t bound envelope of 2.50 times the normalized axial peaking factor is not exceeded i during either normal operation or in the event of xenon redistribution following ; power changes. Target flux difference is determined at 2quilibrium xenon conditions. The full-length rods may be positioned within the core in accordance with their l - respective insertion limits and should be inserted near their normal position I for steady-state o;.eration 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 ' l , associated core burnup conditions. Target flux differences for other THERMAL : POWER levels are obtained by multiplying the RATED THERMAL POWER value by the I acpropriate fractional THERMAL POWER level. The periodic updatir, of the target t flux difference value is necessary to reflect core burnup consider sions. l ! l l i i l ( SOUTH TEXAS - UNITS 1 & 2 B 3/4 2-1 AMENDMENT NDS. AND l j i I
E POWER DISTRIBUTION LIMITS 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 for 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 i.i alarm message when the penalty deviation accumulates beyond the limits or 1 hour and 2 hours, respectively. Figure B 3/4 2-1 shows a typical mor.thly target Dand. 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and NUCLEAR ENTHALPY RISE HDT CHANNEL FACTOR The limits on heat flux hot channel factor and nuclear enthalpy rise hot channel factor ensure that: (1) the design limits on peak local power density and minimum DNER are not exceeded and (2) in the event of a LOCA the pesk fuel clad te perature will not exceed the 2200*F ECCS acceptance criteria lirr.i+. Each of these is measurable but will normally only be determined pe'riodically 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 inove together with no individual rod i
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; i
SOUTH TEXAS - UNITS 1 & 2 B 3/4 2-2 AMENDMENT N05. AND
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, FIGURE B 3/4.2-1 TYPICAL INDICATED AXIAL FLUX DIFFERENCE VERSUS THERMAL POWER i
l l B 3/4 2-3
- SOUTH TEXAS - UNITS 1 & 2 AMENDMENT N05. AND
, I. , , , , I i
F POWER DISTRIBUTION LIMITS BASES HEAT FLUX HOT CHANNEL FACTOR and NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued)
- c. Thri control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained; and
- d. The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.
Fh will be maintained within its limits provided Conditions a. through
- d. above are maintained. The combination of the RCS flow requirement (395,000 gptr)andtherequirementonFfg guarantees that the DNBR used in the safety analysis will be met. TherelaxationofFfg as a function of THERMAL POWER allows changes in the radial power shape- for all permissible rod inser-tion limits. ,
. When F q is measured, no additional allowances are necessary prior to comparison with the limit. Ameasurementerrorof4%forFhhasbeenallowed for in the determination of the design DNBR value.
Fuel rod bowing reduces the value of DNB ratio. Credit is available to offset this reduction in the generic margin. The generic margins, totaling 3.3% DNBR completely offset any rod bow penalties. This margin includes the following:
- a. Design limit DNBR of 1.30 vs 1.28,
- b. Grid Spacing (K,) of 0.059 vs 0.066, and
- c. Thermal Diffusion Coefficient (for use in modified spacer factor) of 0.059 vs 0.061.
The applicable values of rod bow penalties are explained in FSAR Sec-tion 4.4.2.2.5. l I SOUTH TEXAS - UNITS 1 & 2 B 3/4 2-4 AMENDMENT NOS. AND ,, I I 1969
f$ ' POWER DISTRIBUTION LIMITS BASES HEAT FLUX HOT CHANNEL FACTOR 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 Incore Detector Flux Mapping System, and a 3% allowance is appropriate for manufacturing tolerance. The Radial Peaking Factor, F,y(Z), is seasured periodically to provide assurance that the Hot Channel Factor, F (Z), remains within its limit. The RTPn
. F xy limit for RATED THERMAL POWER (F
) as provided in the Radial Peaking 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.
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 ONB and linear heat generation rate protection with x y plane power tilts. A limit l of 1.02 was selected to provide an allowance for the uncertainty associated With the indicated power tilt. The 2-hour time allowance for operation with a tilt condition greater than 1.02 is provided to allow identificai. ion and correction of a dropped or misaligned control rod. In the event such action does not correct the tilt, the margin for uncertairty on F is g reinstated by reducing the maximum allowed power by 3% for each percent of tilt in excess of 1. 1 For purposes of monitoring QUADRANT POWER TILT RATIO when one excore . detector is inoperable, the moveable incore detectors are used to confirm that l the normalized symmetric power distribution is consistent with the QUADRANT : POWER TILT RATIO. The intore detector monitoring is done with a full incore flux map or two sets of four symmetric thimbles. The two sets of four sym.tetric thimbles is a unique set of eight detector locations. These locations are C-8, E-5, E-11, H-3, H-13, L-5, L-11, N-8. i 3/4.2.5 DNB PARAMETERS f The limits on the DNB-related parameters assure that each of the parameters are maintained within the normal steady-state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the SOUTH TEXAS - UNITS 1 & 2 B 3/4 2-5 AMENDMENT N05. AND 6 .* .,,,
(I n POWER DISTRIBUTION LIMITS ! l l BASES 3/4.2.5 DNB PARAMETERS (Continued) 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 598'F and the indicated pressurizer pressure value of 2201 psig are provided assuming that the readings from four channels will be averaged before comparing with the required limit. The flow requirement - (395,000 gpm) includes a measurement uncertainty of 3.5%. The 12-hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored witnin their limits following load changes and other expected transient operation. I d SOUTH TEXA5 - UN115 1 & 2 B 3/4 2-6 AMENDMENT N05. AND
FL 3/4.3 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTAL 10N The OPERABILITY of t5e Reactor Trip System and the Engineered Safety Features Actuation System instrumentation and interlocks ensures that: (1) the a r -iated ACTION and/or Reactor trip will be initiated when the parameter monitt. ed by each channel or combination thereof reacher, its Setpoint, (2) the specified coincidence logic is maintained, (3) sufficient redundancy is main-tair.d to permit a channel to be out-of-service for testing or maintenance, ' and (4) sufficient system functional capability is available from diverse I 'nmeters. . 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 acci' dent and transient conditions. The integrated operation of each of these systems is consistent with the assumptions used in the safety analyses. The Surveillance Requirements 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 sufficitnt to dem)nstrate this capability. Specified surveillance intervals and surveillance c.nd maintenance outage times have been determined in accordance with WCAP-10271, "Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrumentation System," and supplements to that report.
Surveillance intervals and out of service times were determined based on maintaining an appropriate level of reliability of the Reactor Protection System instrumentation. The Engineered Safety Features Actuation System Instrumentation Trip Setpoints specified in Table 3.3-4 are the nominal values at which the bistables are set for each functional unit. A Setpoint is considered to be adjusted consistent with the nominal value when the "as measured" Setpoint is within the band allowed for calibration accuracy. To accommodate the instrument drif t 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-tion with Setpoints less conservative then the Trip Setpoint but within the Allowable Value is acceptable since an allowance has been made in the safety analysis to accomodate this error. An optional provision has been included for determining the OPERABILITY of a channel when its Trip Setpoint is found to exceed the Allewable Value. The methodology of this option utilizes the "as measurtd" deviation from the specified calibration point for rack and sensor components in conjunction with a Statistical combination of the other uncertainties of the instrumentation to measure the process variable and the uncertainties in calibrating the instrumentation. In Equation 2.2-1, Z + R + S < TA, the interactive effects of the errors it, the rack and the sensor, an3 the "as measured" values of the errors are considered. 2, as specified in Table 3.3-4, in percent span, is the statistical summation of errors assu9 d in the analysis excluding those associated with the sensor and rack drift and the accuracy of their measurement. TA or Total Allowance SOUTH TEXAS - UNITS 1 & 2 B 3/4 3-1 AMENDMENT NDS. AND 1 i f,9; l
F INSTRUMENTATION BASES REACTOR TRIP SfSTEM .ad ENGINEERED SAFETY FEATURES ACTUA. TION SYSTEM INSTRUMENTAT10N (Continued) is the difference, in percent span, between the trip setpoint and the value used in the analysis for the actuation. 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 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 2.2-1 allows for a sensor drift factor, an increased rack drift factor, and provides a threshold volue 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 unce,rtainty 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 cccasional, may be indicative of more serious problems and should warrant further investigation. 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 +' ken in the analyses for those channels with response times indicated as not t,J11 cable. Response time may be demonstrated by any series of sequential, overlapping, or total channel test measurements 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 times. The Engineered Safety Features Actuation System senses selected plant para-meters and determines whether or not predetermined limits are being exceeded. If they are, the signals are combined into logic matrices sensitive to combina-tions indicative of various accidents, events, and transients. Once the re-quired 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 accident: (1) Safety In-jection pumps start, (2) Reactor trip. (3) feedwater isolation, (4) startup of the standby diesel generators, (5) containment spray pumps start and automatic valves position, (6) containment isolation, (7) steam line isolation, (8) Tur-bine trip, (9) auxiliary feedwater pumps start and automatic valves position, (10) rea. tor containment fan coolers start, (11) essential cooling water pump, start and automatic valves position, (12) Control Room Ventilation Systems start, and (13) component cooling water pumps start and automatic valves position. SOUTH TEXAS - UNITS 1 & 2 B 3/4 3-2 AMENDMENT NDS. AND
'l G
FL INSTRUMENTATION BASES REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued) Radiation Monitoring Bases are discussed in Section 3/4.3.3.1 below. The Engineered Safety Features Actuation System interlocks perform the following functions: P-4 Reactor tripped - Actuates Turbine trip via P-16, closes main feed-water valves 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 and 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 or low compen-sated s'.eamline pressure signals, reinstates steamline isolation on low compensated steamline pressure signals, and opens the accumulator discharge isolation valves. On decreasing pressure, P-11 allows the manual block of Safety Injection actuation on low pressurizer pres-sure or low compensated steamline pressure signals, allows the manual block of steamline isolation on low compensated steamline pressure - signals, and enables steam line isolation on high negative steam line pressure rate (when steamline pressure is manually blocked). P-12 On increasing reactor coolant loop temperature, P-12 automatically provides an arming signal to the Steam Dump System. On decreastig reactor coolant loop temperature, P-12 automatically removes the arming signal from the Steam Dump System. P-14 On increasing steam generator water level, P-14 automatically trips ' the turbine and the main feedsater pumps, and closes all feedwater isolation valves and feedwater control valves. 3/4.3.3 MONITORIN3 INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING FOR PLANT 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 or combination thereof reaches its Setpoint, (2) the specified coincidence logic is maintained, and (3) suffi-cient redundancy is maintained to permit a channel to be out of service for testing or maintenance. The radiation monitors for plant operations sense radiation levels in sel.ected plant systems and locations and determine whether or not predetermined limits are being exceed ~. If they are, the signals are combined into logic matrices sensitive to e , ations indicative of various accidents and abnormal conditions. Once thi - iired logic combination is completed, the system sends actuation signals ?- initiate alarms or automatic ' isolation action and actuation of Eeergency Exhaust or Ventilation Systems. SOUTH TEXAS - UNITS 1 & 2 B 3/4 3-3 AMENDMENT N05. AND l 1559
F1-INSTRUMENTATION BASES 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 i 'emonstrated by irradiating each detector used and determining the acceps.oility of its voltage curve. For the purpose of measuring Fq (Z) or F q a fuM .incore flux map is used. Quarter-core flux map . .: e fined in WCAP-8648, June 1976, may be used in recalibration of the Excore i.eutron Flux Detection System, and full incore flux maps or symmetric iner v thimbles may be' used for monitoring the QUADRANT POWER TILT RATIO when one et' Range channel is.ineperable. 3/4.3.3.3 SEISMIC INSTRUMENTATION The OPERABIlf.Y of the s.tismic instrumentation ensures that sufficient capability is avui?able to prochtly determine the magnitude of a seismic event and evaluate the response of those features important to safety. This capa-bility is required to permit com'arison p 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 Part 100. The instrumentation is consistent with the recomu ndations of Regulatory Guide 1.12. "Instrumentation for Earth-quakes," April 1974. 3/4.3.3.4 METj.R0LOGICALINSTRUMENTATION The OPERABILITY of the meteorological instrumentation ensures that sufficient meteorological data are available for estimating potential radiation doses to the public as a result of routine or accidental release of radioactive materials to the atmosphere. This capability is required to evpluate the need for initiating protective measures to protect the health and safety of the public and is consistent with the reconmendations of Regulatory Guide 1.23, "Onsite Meteorological Programs," February 1972. 3/4.3.3.5 REM 3TE SH'JTDOWN 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 Part 50. SOUTH TEXAS - UNITS 1 / 2 B 3/4 3 4 AMEN 0 MENT NDS. AND
"~ ' I : 1p-
f'l INSTRUMENTATION , BASES , REMOTE SHUTDOWN SYSTEM (Continued) The OPERABILITY of the Remote Shutdown System ensures that a fire will not preclude achieving safe shutdown. The Remote Shutdown System instrumentation, 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 independent of areas where a fire could damage systems normally used to shut down the 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,1NSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables following an accident. This capability is consis- ' tent with the recommendations of Regulatory Guide 1.97, Revision 2 "Instrumen-tation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident," December 3980 and NUREG-0737, "Clarification of TM1 Action Plan Requirements," Nevember 1980. The instrumentation listed in Table 3.3-10 corresponds to the Category 1 instrumentation for which selection, ; design, qualification and display criteria are described in Regulatory Guide 1.97, Revision 2. 3/4.3.3.7 CHEMICAL DETECTION SYSTEMS The OPERABILITY of the Chemical Detection Systems ensures that sufficient capability is available to promptly detect and initiate protective action in the event of an accidental chemical release. This capaDility is required to protect control room personnel and is consistent with the recommendations of Regulatory Guide 1.78, "Assu ptions for Evaluating the Habitability of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical Release," June 1974. 3 /4. 3. 3. 8 (Not Used) I 4 1 l 8 3/4 3-5 AMEN 0 MENT NDS. AND SOUTH TEXA5 - UNITS 1 & 2 i d I id.,
FG INSTRUMENTATION BASES 3/4.3.3.9 (Not Used) 3/4.3.3.10 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The Alarm / Trip Setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the 00CM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50. 3/4.3.3.11 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION The radioactive gaseous s ffluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous wifluents i during actual or potential releases of gaseous effluents. The Alarm / Trip Setpoints for these instrueents 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 Part 20. This instrumentation also includes provisions for monitoring (and controlling) the concentrations of potentially explosive gas mixtures in the GASEOUS WASTE PROCESSING SYSTEM. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Crite*ia 60, 63, and 64 of Appendix A to 10 CFR Part 50. The sensitivity of any n0ble gas activity monitors used to show compliance with the gaseous effluent release requirements of Specification 3.21.2.2 shall be such i that concentrations as los as 1 x 10 8 pCi/cc are measurable. 3/4.3.4 TURBINE OVERSPEED PROTECTION i This specification is provided to ensure that the turbine overspeed protection instrumentation and the turbine speed control valves are OPERABLE l and will protect the turbine from excessive overspeed. Protection from turbine excessive overspeed is required since excessive overspeed of the turbine could generate potentially damaging missiles which could impact and damage safety-related components, equipment, or structures. , l i SOUTH TEXA5 - UNITS 1 & 2 B 3/4 3-6 AMENDMENT N05. AND i I I )
b 1 3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to Operate with all reactor coe' ' 35 in operation and maintain DNBR above 1.30 during all normai as and ! anticipated transients. In MODES 1 and 2 with one reacti ;1t loop not in . operation this specification requires that the plant be u 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; however, a single reactor coolant loop provides sufficient heat : removal capacity if a bank withdrawal accident can be prevented, i.e., by I opening the Reactor Trip System breakers. Single failure considerations f require that two loops be OPERABLE at all times. ; In MODE 4, and in MODE 5 with reactor coolant loops filled, a single l
*eactor coolant loop or RHR lovp provides sufficient heat removal capability I . for removing decay heat; but single failure considerations require that at !
least two loops (either RHR or RCS) be OPERABLE. '
^
In MODE 5 with reactor coolant loops not filled, a single RHR loop provides sufficient heat removal capability for removing oecay hest; but single failure considerations, and tha unavailability of the steam generators as a heat i removing component, require that at least two RHR loops be OPERABLE. ! The boron dilution analysis assumed a common RCS volume, and maximum di-lution flee rate for MODES 3 and 4, and a different volume and flow rate for ! MODE 5. The MODE 5 concitions assumed limited mixing in the RCS and cooling with the RHR system only. In MODES 3 and 4, it was assumed that at least one ' reactor coolant pump was operating. If at least one reactor coolant pump is ' not opera'.ing in MODE 3 or 4, then the maximum possible dilution flow rate must be limited to the value assumed for MODE 5. The operation of one reactor coolant pump (RCP) or one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual l reactivity changes during boren concentration reductions in the Reactor Coolant System. The reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognition and control. - 1 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, cause by energy additions from the Secondary Coolant System, which could exceed the
. limits of Appendix G to 10 CFR Part 50. 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 terperature of each l steam generator is less than 50'F above each of the RCS cold leg temperatures. '
3/4.4.2 SAFETY VALVES The pressuriier Code safety valves operate to prevent the RCS from being ! pressurized above its Safety Limit of 2735 psig. Each safety valve is designed ! to relieve 504,950 lbs per hour of saturated steam at the valve setpoint of f 2500 psia. The relief capacity of a single safety valve is adequate to relieve l any overpressure condition which could occur during shutdown, in the event that no safety valves are OPERABLE, an operating RHR loop, connected to the , i i
+
SOUTH TEXAS - UNITS 1 & 2 B J/4 4-1 AMENDMENT N05. AC
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F. r REACTOR COOLANT SYSTEM i BASES
- r I
3AFETY VALVES (Continued) RCS, provides overpressure relief capability and will prevent RCS overpressuri-zation. In addition, the Overpressure Protection System provides a diverse means of protection against RCS overpressurization at low temperatures, j 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 credit is taken for a direct Reactor trip on the turbine trip resulting from l loss-of-load) and also assuming no operation of the power-operated relief valves f or steam dump valves. ! Demonstration of the safety valves' lift settings will occur only during f 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 12-hour periodic surveillance is sufficient to insure that the parame-ter 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 hydriulically 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 circulaticn. 3/4.4.4 RELIEF VALVES The power-operated relief valves (PORVs) and steam bubble function to relieve RCS pressure during all design '.ransients up to and including the design step load decrease with steam d.wp. Operation of the PORVs minimizes the undesirable opening of the spring-loaded pressurizer Code safety valves. Each PORV has a remotely operated block valve to provice a positive shutoff capability should a relief valve become inoperable. 3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be main-ta'ined. The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1. Inservice i inspection of steam generator tubing is essential in order to maintain surveil- l lance of the conditions of the tubes in the event that there is evidence of l mechanical da-age or progressive degradation due to design, manuf acturing 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. a i i SOUTH TEXAS - UNITS 1 & 2 B 3/4 4-2 AML,,.<ENT NDS. AV j b 1 < 1;;
F REACTOR COOLANT SYSTEM BASES STEAM GENERATORS (Continued) The plant is expected to be cperated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes. If the secondary coolant chemistry is not maintained within these limitt, 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 (primary-to-secondary leakage = 500 gallons per day per steam generator). Cracks having a primary-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 primary-to-secondary leakage o_f 500 gallons per day per steam generator can readily be detected by radiation r.onitors of steam generator blowdown. Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged. Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant. However, even if a defect should develop in service, it will be found during scheduled inservice steam generator tr*;e examinations.
~
Plugging will be required for all tubes with imperfections exceeding the i plugging limit of 40% cf the tube nominal wall thickness. Steam generator . 4 tute inspections of operating plants have demo,strated the capability to ! reliably detect degradt. tion that has penetrate 9 20% of the original tube wall thickness. Whenever the results of any steam, generator tubing insersice inspection fall intc. Category C-3, these results will be promptly reported to the Commis(ion in a Special Report pursuant tc Specification 6.9.2 withic 30 days and prior to resumption of plant operation. Such cases will be contidered by the Ccmmif3 ion on a case-by-case basis and may result in a requirement ft r analysis, lacoratory examinations, tests, t.dditional eddy-current inspection, and revision of the Technical Specifications, if necessary. 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS f The RCS Leakage Detection Systems required by this specification are i provided to monitor and detect leakage from the reactor coolant pressure boundary. These Detection Syster.5 are consistent with the recomendations of Regulatory Guide 1.45, "Reacter Coolant Pressure Boundary Leakage Detection . Systems," May 1973. : 3/4.4.6. OPERATIONAL LEAKAGE PRESSURE BOUNDAf.Y LEAKAGE of any magnitude is unacceptable sin;e it may be indicative of an iepending gross failure of the pressure boundary. Therefore, i i SOUTH TEXAS - UNITS 1 & 2 B 3/4 4-3 AMENDMENT N05. AND 1 ; ;,
# %]
REACTOR COOLANT SYSTEM BASES _ OPERATIONAL LEAKAGE (Continued) the ptasence of any PRESSURE BOUNDARY LEAKAGE requires the unit to be promptly placed in COLD SHUTOOWN. Industry experience has shown that while a limited amount of leakage is i expected from the RCS, the unidentified portion of this leakage can be reduced to a threshold value of less tht.n 1 gpm. This threshold value is sufficiently low to ensure early detection of additional leakage. l The total steam generator tube leakage limit of I gpm for all steam generators not isolated from the RCS onsures that the dosage contribution from the tube leakage will be limited to a small fraction of 10 CFR Part 100 dose guideline values in the event of either a steam generator tube ruptare or steam line break. The 1 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 i event of a main steam line rupture or under LOCA conditions. , The 10 gpm IDENTIFIED LEAKAGE limitation provides allowance for a limited amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the Leakage Detection Systems. The specified allowed leakage from any RCS pressure isolation valve is sufficiently low to ensure early detection of possible in-series check valve frilure. It is apparent that when pressure isolation is provided by two in-series check valves and when failure of one valve in the pair can go undetected for a substantial length of time, verification of valve integrity is required. Since these valves are important in preventing overpressurization and rupture of the ECCS low pressure piping which could result in a LOCA that bypasses containment, these valves should be tested periodically to ensure low probability of gross failure. The Surveillance Requirements for RCS pressure isolation valves provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA. Leakage from the RCS pressure isolation valve is IDENTIFIED LEAKAGE and 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 Reactor Coolant System is einimized and reduces the potential for Reactor Coolant Systee leakage or failure due to stress corrosion. Maintaining SOUTH TEXAS - UNITS 1 & 2 8 3/4 4 4 AMENDMENT N05. AND
'."J
N REACTOR COOLANT SYSTEM BASES CHEMISTRY (Continued) 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. Corrosinn studies show that operation may be continued with contaminant concentration levels in excess of the Steidy-State Limits, up to the Transient Limits, for 4 the specified limited time intervals without having a significant 6ffect on the structural integrity of the Reactor Cool. ant System. The time interval permitting continued operation within the restrictions of the Transient Limits
, provides time for taking corrective actions to restore the contaminant concen-trations to within the Steady-State Limits.
The Surveillance Requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action. . 3/4.4.8 SPECIFIC ACTIVITY The limitations on the specific activity of the reactor coolant ensure that the resulting 2 hour doscs at the SITE BOUNDARY will not exceed an { appropriately small fraction of 10 CFR Part 100 dose guideline values following i a steam generator tube rupture accident in conjunction with an assumed steady-state reactor-to-secondary steam generator leakage rate of 1 gpm. The values for the limits on specific activity represent limits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative in that specific site parameters of the STPEGS site, such as SITE BOUNDARY location and metearological conditions, were not considered in this evaluation. The ACTION statement permitting POWER JPERATION to continue for limited time periods with the reactor coolant's sptcific activity greater than 1 microcurie / gram 00SE EQUIVALENT I-131, but within the allowable limit shown on Figure 3.4-1, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER. The sample analysis for determining the gross specific activity and I can
. exclude the radiciodines because of the low reactor coolant limit of 1 microcurie /
gram DOSE EQUIVALENT l-131, and because, if the limit is exceeded, the radiciodine level is to be determined every 4 hours. If the gross specific activity level and radiciodine level in the reactor coolant were at their limits, the radioiodine contribution would be approximately 1%. In a release of reactor coolant with a typical mixture of radioactivity, the actual radio-iodine contribution would probably be about 20L The exclusion of radio-nuclides with half-lives less than 15 minutes from these determinations has B 3/4 4-5 AMEN 0 MENT N05. AND SOUTH TEXAS - UNITS 1 & 2 f71;is
- Fi - REACTOR COOLANT SYSTEM RASES SPECIFIC ACTIVITY.(Continued) been made for several reasons. The first consideration is the difficulty to ! I identify short-lived radionuclides in a sample that requires a significant
- time to collect, transport, and analyze. The second consideration is the l predictable delay time between the postulated release of radioactivity from )
the reactor coolant to its release to the environment and transport to the ! SITE BOUNDARY, which is relatable to at least 30 minutes decay time. The ! choice of 15 minutes for the half-life cutoff was made because of the nuclear , characteristics of the typical reactor coolant radioactivity. The radionuclides - in the typical reactor coolant have half-lives of less than 4 minutes or - i half-lives of greater than 14 minutes, which allows a distinction between the i radionuclides above and below a half-life of 15 minutes. For these reasons i the radionuclides that are excluded from consideration are expected to decay t j to very low levels before they could be transported from the reactor coolant [ l to the SITE BOUNDARY under any accident condition. 1 Based upon the above considerations for excluding certain radionuclides I
. from 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 l 2 perform the sampling, transport the satple, and perform tha analysis of about l 90 minutes. After 90 minutes, tha gross count should be made in a reproducible ! geometry of saeple and counter having reproducible beta or gam s self-shieldir.g t properties. The counter should be reset to a reproducible efficiency versus It is not necessary to identify specific nuclides. The radiochemical j enargy. : determination of nuclides she, tid be based on multiple counting uf the satple l within typical counting basis following sampling of less than 1 hour, about t 2 hours, about 1 day, about 1 week, and about 1 month. [ Reducing T to less than 500*F prevents the release of activity should [ a steam generator tube rupture since the saturation pressure of the reactor ; coolant is below the lift pressure of the atmospheric steam rtlief valves.
- The Surveillance Requirements provide adeouate assurance that excessive specific activity levels in the reactor coolant will be detected in suf ficier.t tire to l take corrective action. A reduction in frequency of isotopic analyses following ;
power changes may be pcrmissible if justified by the data obtained, j 3/4.4.9 PRES $URE/TE N RATUTE E LIMITS [ The tercerature and pressure changes during heatup and cooldown are I limited to be consistent with the requirements given in the ASME Boiler and i Pressure Vessel Coce, Section 111, Appendix G: !
- 1. The reactor coolant temperature and pressure and system heatup and cooloown !
rates (with the exception of the pressurizer) shall be limited in accordance l with Figures 3.4 2 and 3.4-3 for the service period specified thereon: 1 SOUTH TEXA5 - UNITS 1 & 2 B 3/4 4 6 AMENDMENT h05. A'O I _ _ _ _ _ _ _ - ~ _ _ _ _ . _ . _ . _ _
Ft i REACTOR COOLANT SYSTEM L i l BASES PRESSURE TEMPERATURE LIMITS (Continued) !
- a. Allowable combinations cf pressure and temperature for s)ecific temperature change rates are below and to the right of tie limit lines shown. Limit lines for cooldown rates between those presentad ;
may be obtained by interpolation; and l
- b. Figures 3.4-2 and 3.4-3 define limits to assure prevention of i non-ductile failure only. For normal operation, other inherent plant characteristics e.g., pump heat addition and pressuri % e heater ,
capacity, may limit the heatup and cooldown rates that can be , achieved over certain pressure-temperature rangen. i
- 2. These limit lines shall be calculated periodically using methods provided l below, :
- 3. The secondary side of the steam generator must not be pressurized above .
200 psig if the temperature 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 than 621*F, and
- 5. System preservice hydrotests and inservice leak and hydrotests shall be performed at pressures in accoroance with the requirements of ASME Boiler and Pressure Vessel Code, Section XI.
The fracture toughness properties of the ferritic materials in the reactor vessel are determined in acectdance with the NRC Standard Review Plan, ASTM EISS-73 and in accordance with additional reactor vessel requirements. These
- properties are then evaluated in accordance with Appendix G of the 1976 Sumer i Addenda to Section III of the ASME Boiltr and Pressure Vessel Code and the
! calculation methorts described in WCAP-7924-A, "Basis for Heatup and Coaldewn limit Curves," April 1975, ) Heatup and cooldown limit curves are calculated using the most limiting j value of the nil-ductility reference temperature, RTOT, at the end of 32 effec- ! tive full power years (EFPY) of service life. The 32 EFPY service life period i is chosen such that the limiting RTMT at the 1/4T location in the core region j is greater than t.e RTW T of the limiting unirradiated material. The selection l of such a limiting RTW T assures that all components in the Reactor Coolant System will be operated conservatively in accordance with applicable Code
- requirements.
l The reactor vessel materials have been tested to determine their initial
- RTWT; the results of these tests are shown in Tables B 3/4.4-14 and B 3/4.4 lb.
) Reactor operation and resultant fast neutron (E greater than 1 MeV) irradiation l l l SOUTH TEXAS - UNITS 1 & 2 B 3/4 4 7 AMECMENT C 5. AC i.
Fl , REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) I can cause an increase in the RT HDT. Therefore,anadjustedreferencetempera-ture, based upon the fluence, copper content, and phosphorus content of the material in question, can be predicted using Figure B 3/4.4-1 and the value , of ART computed by Regulatory Guide 1.99, Revision 1, "Effects of Residual l NDT Elements on Predicted Radiation Damage to Reactor Vessel Materials." The heat- : up and cooldown limit curves of Figures 3.4-2 and 3.4-3 include predicted - adjustments for this shift in RTNDT at the end of 32 EFPY as well as adjustmentsforpossibleerrorsinthepressureandtemperaturesensing i instruments. Values of ART NDT determined in this manner may be used until the results t from the material surveillance program, evaluated according ta ASTM E185, are available. Capsules will be removed in accordance with the reauirements of ASTM E185-73 and 10 CFR Part 50, Appendix H. The surveillance coecimen with- i drawal schedule is shown in Table 4.4 5. The lead factor represets the rela- , tionship between the fast neutron flux density at the location of e e capsule ! and the inner wall of the reactor vessel. Therefore, the results obtained from t the surveillance specimens can be used to predict future radiation damap to the reactor vessel material by using the lead factor and the withdrawal time of : the capsule. The heatup and cocidown curves must be recalculated when the ART OT determined from the surveillance capsule exceeds the calculated aRTg7 ; for the equivalent car,ule radiation exposure. ! Allowabie pressure-tee.perature relationships for various heatup and ccol- [ down rates are calculated using methods derived from Appendix G in Section !!! ; of the ASME Boiler and Pressure Vessel Code as required by Appendix G to 10 CFR : Part 50, and these metheds are discussed in detail in WCAP-7924-4. 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 lengtn of 3/2T is ass eed to exist l at the inside of the vessel wall as well as at the outside of the vessel wall. i The dicensions of this postulated crack, referred to in Appendix G of ASME Sec- l tion III as the reference flaw, a? ply exceed the current capabilities of inser- l vice inspection techniques. Therefore, the reactor operation limit curves de- : veloped for this reference crack are conservrtive i and provide sufficient safety l
- margins for protection against nonductile failure. To assure that the radiation [
erbrittlement effects are accounted for in the calculation of the licit curves, ! the m0st liriting value of the nil ductility reference terperr u ,, r Q. , i s [ used and this includes the radiation-induced shift, aRTWT, corresponding to the ; end of the period for which heatup and cooldown curves are generated. ; The ASME approach for calculating the allowable limit curves for various . heatup and cooldown rates specifies that the tJtal stress intensity factor, I K , for the combined therfral and pressure stresses at any time during heatup , 3 , or cooldo.in cannot be greater than the reference stress intensity facter, Kg, , for the fretal terperature at that tire. K g is obtained f ro : the referen:e j SOUTH TEMS - UMTS 1 & 2 B 3/4 4 8 AMEC MENT M S. AC ! L h
- . - _ - . - . - - - - . - - - _ - . . - - _ _ - - - - - . - - - . . . . . . ~ . . - . _ . - - . - . . .-. _ . . ,. n TABLE B 3/4.4-13 o, REACTJR VESSEL TOUG::MiSS (UNIT 1) E! - - M Average Upper s Shelf Energy
- Normal to 50 ft-Ib Principal Principal E y 35 mil RT Working Working
- Ce F NDT Temp. NOT Direction . Direction Cneponent Code Grade (1) (%) (*F) (*F) (*F) (ft-lb) (ft-lb)
{}
Closure head dome RI616-1 A5338 CL 1 0.07 0.018 -30 80 20 116 -
k* Clo.ure head torus R1615-1 A5338 CL 1 0.04 0.010 -!D <30 -30 152 - + Closure head torus R1615-2 A5338 CL 1 0.11 0.012 -30 <30 -30 196 - Closure head torus R1615-3 A5338 CL 1 0.07 0.011 -40 <20 -40 132 - Closure head torus R1615-4 A5338 CL 1 0.13 0.018 -30 <50 -10 133 - Closure head flange R1692-1 A508 CL 2 0.05 0.007 0 <60 0 109 - to vessel flange RI601-1 A508 CL 2 0.02 0.017 -10 <50 -10 160 5 - t: Inlet nozzle R1613-1 A508 CL 2 - 0.009 -10 <50 -10 140 -
- Inlet nozzle R1613-2 A508 CL 2 -
0.013 0 <60 0 130.5 - t Inlet nozzle' R1613-3 A508 CL 2 0.09 0.G09 -20 <40 -20 175 - i u) Inlet nozzle R1613-4 A508 CL 2 - 0.006 20 <80 20 128- - Outlet nozzle R1614-1 A508 CL ? - 0.006 10 <70 10 106 - Outlet nozzle R1614-2 A508 CL 2 - 0.004 0 <60 0 114 - Outlet nozzle R1614-3 A508 CL 2 - C.009 -30 <30 -30 129 - Outlet nozzle rig 14-4 A508 CL 2 - 0.006 10 <70 10 118 - Nozz!c shell P1607-1 A5338 CL 1 0.08 0.032 0 110 50 89 -
,, Nozzle shell R1607-2 A5338 CL 1 0.08 0.012 -20 110 50 85 -
di fuzzle shell R1607-3 A5338 CL 1 0.07 0.010 -50 90 30 82 - f; Inter. shell R1606-1 A5338 CL i 0.04 0.609 -40 70 10 109.5 130 j5 Inter. shell R1606-2 A5338 CL 1 0.04 0.008 -20 60 0 . 94 119 5 Inter. shell R1606-3 A5338 CL 1 0.05 0.007 -20 .70 10 105.5 132
-. Lower shell R1622-1 A5338 CL 1 0.05 0.006 -30 30 -30 111 143 6{ lower shell R1622-2 A5338 CL 1 0.07 0.006 -30 30 -30 122 149 Lower shell R1622-3 A5338 CL 1 0.05 0.007 -30 30 -30 127 148 Ecttom head torus R1617-1 A5338 CL 1 0.14 0.012 -50 <10 -50 143 -
,, Bottom head torus R1618-1 A5330 CL 1 0.08 0.015 -50 <1P -50 128 -
g; Inter. and lower Gl.70 SAW 0,03 0.004 -50 (10 -50 *158 - shell vert. welds Inter. and lower [3.13 SAW 0.03 0.007 -70 <10 -70 *100 - _ shell girth weld
- Normal to principal weldinq direction .
TABLE B 3/4.4-Ib REACTOR VESSTL TOL'GHNESS (UNIT 2) E y Average Upper Shelf Energy m Normal to 5 50 ft-lb Principal Principal 3S "iI W rking Working T RT Cu P NDT Temp. NDT Direction Direction [ Co-c.onent Code Grade (1) (t) (*f) (*T ) (*F) {ft-lb) (ft-Ib) Closure head dow R3012-1 A5338 CL 1 0.94 0.00R -40 <20 -40 144 - Cin<-ure head tc as R3013-1 A5338 CL 1 0.13 0.009 -30 40 -20 128 - Closure head torus R3013-2 A5338 CL 1 0.13 0.009 -30 70 19 127 - Clo<.ure head torus R3013-3 A533B CL 1 0.15 0.012 -30 60 0 134 - Closure head torus R3013-4 A5338 CL 1 0.15 0.012 -30 60 0 138 - Closure head tlanqo R3002-1 A508 CL 2 0.06 0.008 -50 <10 -50 142 - V.ssel flange R3001-1 A503 CL 2 0.04 0.C08 -10 <$0 -10 146 - Inlet nozzle R20ll-1 A508 CL 2 0.10 0.011 -40 <20 -40 165 - Inlet nozzla R20ll-2 A508 CL 2 0.10 0.011 -20 <40 -20 136 - T Inlet nozzle R20ll-3 A508 CL 2 0.12 0.009 -20 <40 -20 128 - Inlet nozzle R2011-4 A508 CL 2 0.11 0.009 -20 <40 -20 132 - 7 Dottet nozzle R2012-1 A508 CL 2 - 4.006 10 <70 10 132 - Outlet nozzle R2012-2 A508 CL 2 0.007 10 <70 10 132 - Outlet nozzle R2012-3 A508 CL 2 - 0.004 0 <60 0 121 - Outlet nozzle R2012-4 A508 CL 2 - 0.007 10 <70 10 126 - N rzte shell R2505-1 A5338 CL 1 0.05 0.009 -40 60 0 114 -
% :ile shell R2505-2 A533B CL 1 0.07 0.009 -30 60 0 124 -
Nozzle shcIl R2505-3 A533B CL 1 0.05 0.003 -50 50 -10 127 -
> Inter. shell R2507-1 A5338 CL 1 0.04 0.006 -10 <50 -10 109 137 iE Inter. shell R2507-2 A5338 CL 1 0.05 0.006 -10 <50 -10 129 145 5 Inter. shell R2507-3 A5338 Ct. 1 0.05 0.005 -40 20 -40 122 149 iE Lowar shell R3022-1 A5338 CL 1 0.03 0.002 -30 30 -30 124 141 5 tower shell R3022-2 A5338 CL 1 0.04 0.003 -40 20 -40 118 141
- - !ower shell R3022-3 A5338 CL 1 0.04 0.004 -40 20 -40 123 126 t Bottoa Head Torus R3020-1 A5338 CL 1 0.11 0.009 -30 100 40 86 -
Lottom Head Torus R3021-1 A5338 CL 1 0.09 0.008 -60 0 -60 132 - Inter. Shell teams G3.02 Sub Arc 0.05 0.001 -70 <10 -70 146* -
> Lower Shell Seams E3.12 Sub Arc 0.05 0.008 -70 <10 -70 101* -
6 Inter. to tower E 3.12 A am Girth ' [3.12
- Normal to principal welding direction T
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_I !l lll..; 1 i i i i ! 2 - t. l I{ ' ' IOI7 O 10 20 30 40 50 60 EFFECTIVE FULL POWEh (YEARS) FIGURE B 3/4.4-1 FAST NEUTRON FLUEN:E (E>1MeV) A5 A FUN; TION OF FULL 90n'IR SERVICE LIFE E 3/4 4 11 Av.IN;u.ENT N05. AND ,,y 50'JTH TEXA5 - UNITS 1 & 2 C '.
FP REACTOR COOLANT SYSTEM BASES iRESSURE/ TEMPERATURE LIMITS (Continued) fracture tought.ess curve, defined in Appendix G to the ASME Code. The K IR curve is given by the equation: KIR = 26.78 + 1.223 exp [0.0145(T-RTNDT + 160)] (1) Where: K is the reference stress intensity factor as a function of the metal IR temperature T and the metal nil-ductility reference temperature RTNDT. Thus, the governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows: CKIM + kit IKIR (2) Where: K;g = the stress intensity factor caused by membrane (pressure) stress, Kyg = the stress intensity factor caused by the therrral gradients, KIR = constant providd by the Code as a' functio.1 of temperature relative to the 9.T MT of the material, C = 2.0 for level A and 6 service limits, and C = 1.5 for. inservice hydrostatic and leak test operations. At any time during the heatup or cooldown transient, K;g is determinec bs the metal temperature at the tip of the postulated flaw, the apprcpriete value for RTNDT, and the reference fracture tcughness curve. The thermal stresses resultilg from terperature gradients through the vessel wali 3re calculated and then the corresponding thermal stress irtensity factor, KIT, for the reference flaw is corputed. Frcr Equation (2) the pressure stress intensiiy factors are obtained and, fro. these, the allowable pressures are calculated. C00LDOW'1 For the calculatlen of the allowatie pressure versus coola'nt te perature during cooldown, t' Ccce reference flaw is assved to exist at the inside of int vessel wall. .Jring cooldown, the controlling location of the flaw is a hays at the inside of the wall because the therr.31 gradients produce tensile stre .5 at the inside, which increase with increasing cooldoon rates. Allo.atle pressure-terperature relations are generated for both steady-state and finite cooldown rate situstions. From these relations, corposite lir.it curves are constructed for each coold0wn rate of interest. The use of the corposite curve in the cooldown analysis is necessary because control of the cooldoon procedure is based on ressurerent of reacter coolant terperature, whereas the liriting pressure is actually dependent en the raterial teiperature at the tip of the asse ed fla.. During cocido.r., the SOUTH TEXA5 - UNlTS 1 & 2 B 3/4 4-12 AYEOYENT NDS. A'O
Pl i REACTOR COOLANT SYSTEM BASES _ PRESSURE / TEMPERATURE LIMITS (Continued) 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 toolant temperature, the AT ! developed during cooldown results in a higher value of Kgg at the 1/4T location l for finite cooldown rates than for steady-state operation. Furthermore, if I conditions exist such that the increase in K IR exceeds Kgg, the ca.lculated [ allowable pressure during cooldown will be greater than the steady-state value. j I The above procedures are needed because there is no direct control on i temperature at the 1/4T location; therefore, allowable pressures may unknowingly ! be violated if the rate of cooling is decreased at various interv315 along a l cooldown ramp. The use of the composite curve eliminates this problem and : assures conservative operation el the system for the entire cooldown period. l HEATUP Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable i pressure-temperature relationships are developed for steady-state conditions i as well at firite heatup rate conditions assuming the presence of a 1/4T , defect at the ir41de of the vessel wall. The thermal gradients during heatup i
- produce compressive stresses at the inside of the wall that alleviate the
.' tensile stresses produced tsy internal pressure. The metal temperature at the i j crack tip lags the coolant temperature; therefore, the Kyp, for the 1/4T track { j during heatup is lower than the Kg f r the 1/4f crack during steady-state l conditions at the same coolant ten erature. During teatup, especially at the I
! end of the transient, conditions msy exist stch that the effects of compressive j j thermal stresses and different Kgg's fer ste6dy-state and finite heatup rates j do not of f set each other and the pressure-temperature et,rve based on stead /-state
- conditions no longer represents a lower bound of all similar curves for finite ,
heatup rates when the 1/4T flaw is considered. Therefore, both cases have to i be analyzed in order to assure that at any coolant terperature the lower value { 4 of the allowable pressure calculated for steady state and 'linite heatup rates ; I > is obtained. The second portion of the heatup analysis concerns the esiculation of r pressura teeperature lim'.tations for the case in which a 1/4T deep outside ( surface flaw is assumed. Unlike the situation at the vessel inside surface, L the thereal gradients established at the outside surface daring heatup product i
' stresses which are tensile in nature and thus tend to reinforce any pressure [
I stresses present. These thermal stresses, of course, are dependent on both ; I the rate of heatup and the time (or coolant temperature) along the heatup , l ramp. Furthermore, since the thermal stresses at the outside are tensile and f i l i l I SOUTH TEXA5 - UNIls 1 & 2 B 3/4 4 13 AME@ MENT h05. AK ( l ! ! l } l
Fl REACTOR COOLANT SYSTEM r BASES , PRESSURE / TEMPERATURE LIMITS (Continued) . increase with increasing heatup rate, a lower bound curve cannot be defined. Rather, oach heatup rate of interest must be analyzed on an individual 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 coaposite curve is constructed based on a point-b>> point comparison of the steady-state and finite heatup rate data. At any !' ben temperature, t5e allowable pressure is taken to be the le',ser of the . three values taken frce 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 controlling conditiun switches from the inside ;
- to the outside and the pressure limit must at all times be based on analysis i of the most critical criterion. r Finally, the composite curves for the heatup rate data and the cooedown ,
t rata data are adjusted for possible e"rors in the pressure and temperature i sensing instrumnLs by the values indicated on the respective curves. ; Although the pressurizer operates in ter.perature ranges above those for r
- which thwe is reason for concern of nonductile failure, operating limits are provided to assure compatibility of operation with the fatigue analysis 1 performed in accoreance with the ASME Code requirements.
1 l LOW TEMEERATURE OVERPRESSURE PROTECTION The OPERABILITY of two PORVs or an RCS vent openirg of at leu t 2.0
; square inches ensures that the RCS will be protected from pressure transients
! which cuuld exceed the limits of Appendix G to 10 CFR Ptrt 50 when one or more
- of the RCS cold legs are less than or equil to 350'F. Either PORV has
> adequate relieving capability to protect the RCS from overpressurization when ! the transient is limited to either: (1) the start of an idle RCP with the 4 secondary water temperatura of the steam gen'erator less than or equal to 50*F l above the RCS cold leg terperatures, or (2) the maximum credible mass injection flow rate due to the startup of a single HHS1 pump plus *.M gpm net charging i flow, while the ROS is in a water solid condition and ths RCS temperature is betweer 350'F and 200'F. For RCS teeperatures less than 200*F, the maximu overpressure event con-l sists of operating a centrifugal charging purp with complete tereinatien of letdown and a failure of the charging flow control valve to the full flow ) ondition. 4 i The Maximum Allowed PORY Setpoint for the Cold Overpressure Hitigation i System (COMS) is derived by analysis which models the performanct 'f the COMS 4 assuming various mass input and heat input transients. Operation with a PORV 4 Setpoint less than or equal to th9 maxinum Setpoint ansures that Appendix G criteria will not t.e violated witn consideration for a eariru . pressure i 50llTH TEXAS - UNITS 1 & 2 B 3/4 4-14 AME C 'ENT NOS. AC
FC l [ REACTOR COOLANT SYSTEM L i BASES LOW TEMPERATURE OVERPRESSURE PP.0TECTION (Continued) ( overshoot beyond the PORV Setpoint which can occur as a result of time delays instrument uncertainties, and single i
) in signalToprocessing and valve opening, heat input transients more severe than those failure. ensure that mass and '
assumed cannot occur, Technical Specifications require lockout of all high head safety injection pumps while in MODE 5 and MODE 6 with the reactor vessel head on. All but one high head safety ;njection pump are required to be locked out j in MODE 4. Technical Specifications also require lockout of the positiva displacement pump and all but one charg n pump while in MODES 4, 5, and o with ; the reactor vessel head installed and d s llow start of an RCP if secondary ; temperature is more than 50'F above primary temperature. b : The Maximum Allowed PORV Setpoint for the COMS will be updated based on the i t results of examinations of reactor vessel material irradiation surveillance I specimens perforced as required by 10 CFR Part 50, Appendix H, and in accordance i with the schedule in Table 4.4-5. - 3 ! 3/4.4.10 STPUCTURAL IN7EGRITY l t i 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 co:rponents will be maintained at an acceptable level throughout the l life of the r Wt. These programs are in accordance with Section XI of the E j ASME Boilr" v.a Pressure Vessel Code and applicable Addenda as required by , I 10'CFR 50.Ma(g) except where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i). ; r l Components of the Reactor Coolant System were designed to provide access ( to permit inservice inspections in accordance with Section XI of the ASM2 l Boiler and Pressure Vescel Code, 1974 Edition and Addenda through Winter 1975. ! 3/4.4.11 REACTOR VESSEL HEAD VENTS f Reactor vessel head vents are provided to exhaust noncondensible gases f and/or steam fro- the Reactor Ccolant System that could inhibit natural l ' circulation core coaling. Tne OPERABILITY of at least two reactor vessel head vent paths ensures that the capability exists to perform this function. I The valwe redundanc3 of the reactor vessel head ver' paths serves to eini-4 i mize the probability of inadvertent or irreversible actuation while ensuring i that a sin le failure of a sent valve, power supply, or control syster. does not prevent is lation of the vent path. The function, capabilities, and testing requirements of the reactor vessel head vents are consistent with the requirements of Item II.B.1 of NUREG-0737, j
"Citrification of TM1 Action Plan Requiremencs," November 1960.
I B 3/4 4-15 AMEN E . N05. AC s SOUTH TEXA5 - MITS 1 & 2 i .
e D' 3/4.5 EMERGENCY CORE COOLING SYSTEMS ! 2 l 4 BASES y 3/4.5.1 ACCUMULATORS l The OPERASILITY of each Reactor Coolant System (RCS) accumulator ensures that a sufficient volume of borated water will be immediately forced into the reactor core through three 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.
1 J The limits on accumulator volume represent a spread about an average value used in the safety analysis and have been demonstrated by sensitivity I studies to vary the peak clad temperature by less than 20'F. The limit on accumulatorpressureensuresthattheassumptionsusedforaccumulatorinjec-tion in the safety analysis are met. i The accumulator power operated isolation valves are considered to be "operating bypasses" in the context of IEEi 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 to meet single failure criteria, renoval of power to the valves is required. ] The limits for operation with an accumulator inoperable for any reason i except an isolation valve closed ninimizes the time exposure of the plant to a Lfi'. event occurring concurrent with failure of an additional accumulator which may result in unacceptable peak ciadding temperatures. If a closed isolation I i 1 valve cannot be opened within one hour, the full capability of one accumulator is not available and prompt action it required to place the reactor in a mode wl.ere this capability is not required. 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS j The OPERABILITY of three independent ECCS subsystems ensures that suffi-cient emergency core coo:ing capability will be available in the event of a l LOCA assuming the loss of one subsystem through any single failure considera-i tion. Each subsystem operating in conjunction with the accumulators is cap- ! able of supplying sufficient core cooling to limit the pea. cladding tempera-l tures within acceptable limits for all postulated break sizes ranging frot the douole ended bre=L of the le* gest RCS cold leg pipe downward. One ECCS is assumed to discharge corpletely throu;;h the postulated break in the RCS loop. )
- Thus, three trains at: required to satisfy the single f ailure criterion. Note that the centrifugal charging p eps are not part of ECCS and that the RHR pumps are not used in tb6 injection phase of the ECCS. Each ECCS subsyste'n and the
' RHR purps and heat exchanges provide long-term core cooling capability in
- recirculation mode during the accidert recovery period.
Vnen the RCS temperature is below 350*f, the ECCS reavirements are balanced between the limitations imposed by the low temperature overpressure protection i and the requirements necessary to mitigate the consequences of a LOCA below j 350'F. At these temperatures, single failure considerations are not required bec ase of the stable reactivity conoition of the reactor ana the lir.ited core l' cooling requirerents. Only a single Lew Head Safety injectionHo.ever, puPp ist.o required l are I to mitigste the effects of a large-break LOCA in this rode. i AND 50'J1H TEXAS - UNITS 1 & 2 E 3/4 5-1 AMENDMENT N35. l
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F0 t i EMERGENCY CORE COOLING SYSTEMS l BASES ECCSSUBSYSTEMS(Continuai) provided to accommodate the possibility that the break occurs in a loop con-taining one of the Low Head pumps. Low Head Safety In ection pumps are not required inoperable below 350 F because their shutoff ead is too low to impact the low temperature overpressure protection limits. Below 200'F (MODE 5) no ECCS pumps are required, so the High Head Safety l Injectionpumpsarelockedouttopreventcoldoverpressure. The Surveillance Requirements provided to ensure OPERABILITY of each l component ensure that, at a minimum, the assumptions used in the safety anal-i 2 . yses are met and that subsystem OPERABILITY is maintained. Surveillance Requirements for flow testing provide assurance that proper ECCS flows will be maintained in the event of a LOCA. l 3/4.5.4 (not used) j 3/4.5.5 REFUELIN3 WATER STORAGE TANK l The OPERABILITY of the refueling water storage tank (RWST) as part of the ECCSensuresthatasufficientsupplyofboratedwaterisavailableforinjec-l i tion Sy the ECCS in the event of a LOCA or a steamline break. The limits on l RWST minimum volure and boron concentration ensure that: (1)sufficientwater i is avi.ilable within containeent to permit recirculation cooling flow to the l the reactor will remain subtri.ical in the cold condition (68'F to core, 212'F)(2)llowing fo a small break LOCA assuming complete mixing of the RWST, RCS, j Spray Additive Tank, Containment Spray System and ECCS water volumes with all control rods inserted except the most reactive control rod assembly (ARI-1), l (3) the reactor will remain subcritical in cold condition following a large J break LOCA (break flow area > 3.0 ft )2 assuming complete mixing of the RWST, i RCS, Spray Additive Tank, Containment Spray System and ECCS water volumes and j other sources of water that ma" eventually reside in the sump post-LOCA with all cor. trol rods assu ed to be cut (ARO), and (4) long term suberiticality j folicwing a steamline break assuming ARI-1 and preclude fuel failure, The maximum allowable value for the RWST boren concentration forcs the basis for determinina the tire (post-LOCA) at which operator action is required to switch over the EECS to het leg recirculation in crder to avoid precipita-l . tion of De soluble boron, f The cent.ined water volume limit includes an allowance for water not usable because of tani discharge line location or other physical characteristics. l The lir.%s en contained water volu e and boron concentration of the RWST also ensurc a pH value of between 7.5 and 10.0 for the solution recirculated within %ntainment after a LOCA. This pH band ainimizes the evolution of iodine ' and minimizes the ef fect of chloride and caustic stress corrosion on mechanical systems and components. AMECMEM M5. A'C SCUTH TEAA5 - UNITS 1 & 2 B 3/4 5-2 W. .s
F7i EMERGENCY CORE COOLING SYSTEMS BASES 3/4.5.6 RESIDUAL HEAT REMOVAL (RHR) SYSTEM The OPERABILITY of the RHR system ensures adequate heat removal capabili-ties for Long-Term Core Cooling in the event of a small-break loss-of-coolant accident (LOCAy, an isolatable LOCA, or a secondary break in MODES 1, 2, and 3. The limits on the OPERABILITY of the RHR system ensure that at least one RHR loop is available for cooling including single active failure criteria. The surveillances ensure that RHR system isolation valves close upon an overpressure protection system signal. 50'JTH TEXAS - UNIi51 & 2 E 3/4 5-3 AMEN; MENT N25. AN:
70 3/4.6 CONTAINMENT SYSTEMS BASES S/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 CONTAINMENT INTEGRITY Primary CONTAINMENT INTEGRITY ensures that the release of radioactive materi Als from the containment atmosphere will It restricted to those leakage paths and asst.,ciated leak rates assumed in the safety analyses. This restriction, in conjunction with the leakage rate limitation, will limit the SITE BOUNDARY radiation doses to within the dose guideline values of 10 CFR Part 100 during accident conditions. 3/4.6.1.2 CONTAINMENT LEAKAGE The limitations on containment leakage rates ensure that the total containment leakage volume will not exceed the value assumed in the safety analyses at the peak accident pressure, P,. As an added conservatism, the mercured overall integrated leakage rate is further limited to less than or equal to 0.75 L, during performance of the periodic test to account for possible degradation of the containment leakage barriers between leakage tests. Tre surveillance testing for measuring leakage rates is consistent with the requirerents of Appendix 1 of 10 CFR Part 50. 3/4.6.1.3 CONTAINMENT AIR LOCKS The limitations on closure and leak rate for the containment air locks 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 in :ts. 3/4.6.1.4 INTERNAL PRESSURE The lireitatiens on containment internal pressure ensure that: (1) the containment structure is prevented from exceedin0 its design negative pressure diffe,*ential with respect to the outside atmosphere c 3.5 psig, and (2) the containment peak pressure does not exceed the design pressure of 56.5 psig during LOCA or steam line break conditions. The maximum peak pressure expected to be obtained from a LOCA or steae line break event is 37.5 psig. The limit of 0.3 psig for initial positive containeent pressure will limit tne total pn ssure to 37.5 psig, which is less than design pressure and is consistent with tha safety analyses. SCUTH TEXAS UNITS 1 & 2 B 3/4 6-1 AMENDMENT NDS. AC I I _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ I
! k i ? CONTAINMENT SYSTEMS ; BASES 3/4 6.1.5 AIR TEMPERATURE The limitations on containment average air temperature ensure that the over-all containment average air temper 4ture does not exceed the initial temperature , condition assumed in the safety analysis for a LOCA or steam line break ; accident. Measurements shall be made by fixed instruments, prior to determin- l ing the average air temperature. 3/4.6.1.6 CONTAINMENT STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the containment will ; be maintained comparable to the original design standards for the life of the ; facility. Structural integrity is required to ensure that the co.dainment will with- i stand the maximum pressure of 37.5 psig in the event of a LOCA or steam line i break accident. The measurement of containment tendon lif t-of f force, the tensile tests of the tendon wires, the visual examination of tendons, anchorages i and exposed interior and exterior surfaces of the containment, and the Type A ! leakage test are sufficient to demonstrate this capability. i The Surveillance Requirements for demanstrating the containment's structural : integrity are in compliance with the recommendations of Regulatory Guide 1.35, ,
"Inservice Inspection cf Ungreated Tendons in Prestresseo Concrete Containment Structures," and proposed Regulatory Guide 1.35.1, "Determining Prestressing ;
Forces for Inspection of Prestressed Concrete Containments," April 1979, t The required Special Reports from any engineering evaluation of containment t abnormalities shall include a description of the tendon condition, the condition ; of the concrete (especially at tendon anchorages), the inspection procedures, the , tolerances on cracking, the results of the engineering evaluation, and the correc-tive actions taken. 3/4.6.1.7 CONTAINMENT VENTILATION SYSTEM The 46-inch containment purge supply and exhaust isolation valves are required to be sealed closed during plant op'erations since these valves have not ; been demonstrated capable of closing during a LOCA or steam line break accident. I Maintaining these valves sealed closed during plant operation ensures that exces-si,ve quantities of radioactive materials will not be released via the Containment l Purge Systera. To provide assurance that these containment valves cannot be inac- [ vertently opened, the valves are sealed closed in accordance with Standard Revie. j Plan 6.2.4 which includes rechanical devices to seal or lock the valve closed, or ; prevents poner from being supplied to the valve operator. The use of the containment purge lines is restricted to the 18-inch purge supply and exhaust isolation valves since, unlike the 48-inch valves, the 18 inch valves are capable of closing daring a LOCA or steam line break accident. There-E 3/4 6-2 AMENDMENT NM. AND SOUTH TEAAS - UNITS 1 & 2 1 ' M3
FD CONTAINMENT SYSTEMS i BASES [0NTAINMENTVENTILATIONSYSTEM(Continued) 4 fore, the SITE BOUNDARY M a guideline of 10 CFR Part 100 would not be exceeded in the event of an accidt'r +. ring containment PURGING operation. Leakage integrity 'wii vith a maxion allowable leakage rate for containment purge supply and exhaus; st. isn o will provide early indication of resilient material seal degradatite . M' 4116 t '.portunity for repair before gross leak-age failures could develop .. akage limit of Specification 3.6.1.2b. shall not be exceeded wher
- N w. P t. fate determined by the leakage integrity
. lu.ksge tests of these valves are - .f to te perviously determined total for all valves and penetrations subject t ! p. k e., s U lts.
3/4 6.2 DEPRESSURIZATION Ah @%y. g , 3/4.6.2.1 CONTAINMENT SPRAY SYL i _ The OPERABILITY of the Containment Spray System ensures that containment depressurization and cooling capability will be available in the avent of a LOCA or steam line break. Tne pressure reduction and resultant lower containment leakage rate are consistent with the atsueptions used in the safety analyses. The Containmert Spray System and the Containment Cooling System both pro-vide post-accioent cooling of the containment atmosphere. However, the Con-tainment Spray System also provides a nechanism fcr removing iodine from the containment atmosphere and therefore the time requirements fcr 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 and containment suep in tne event of a LOCA. The limits on NaOH volu-e and concentration ensure a pH value of between 7.5 and 10.0 Ier 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 systres and corponents. The con-tained solution volure limit includes an allowan'e for solution not usable
, because of tank discharge line location or other .3hysical characteristics.
These assumptions are consistent with the iodine removal efficiency assumed in the safety analy ses. SOUTH TEXAS - UNITS 1 & 2 B 3/4 6-3 AMENDMENT NOS. AND
.: 4?
FP CONTAINMENT SYSTEMS I i RASES , , 3/4.6.2.3 CONTAINMENT COOLING SYSTEM The OPERABILITY of the Containment Cooling System ensures that: (1) the l 1 containment air temperature will be maintained within limits during normal , operation, and (2) adequa+.e heat removal capacity is available when operated ,
- in conjunction with the Containment Spray Systems during post-LOCA conditions.
STPEG5 has three groups of Reactor Containment Fan Coolers with two fans in each group (total of six fans). Five fans are adequate to satisfy the safety requirements including r. ingle failure, , 3/4.6.3 CONTAINMENT ISOLATION VALVES I The OPERABILITY of the containment isolation valves ensures that the containment &tmosphere will be isolated from the outside environment in the t event of a release of radioactive raterial to the containment atmosphere or pressurization of the containment and is consistent with the requirements of General Design Criteria 54 through 57 of Appendix A to 10 CFR Part 00. Contain-l' ment isolation within the time limits specified for those isolation valves i designed to close automatically ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses { i for a LOCA. i 3/4.6.4 COMBUSTIBLE GAS CONTO 0L I a The OPERABILITY of the equipment and systems required for the detection l l and control of 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 recorbiner unit is capable cf con-i [ j trelling the expected hydrogen generation associated with: (1) zirconium-water [ j reactions, (2) radiolytic cecomposition of water, and (3) corrosion of metals j within containeent. These Hydrogen Control Systems are consistent with the re- 1 ' commendations of Regulatory Guide 1.7, "Control of Combustible Gas Concentra- ! i tions in Containment Following a LOCA," Revision 2 November 1978. j i i ) ! i >
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I E D j , I E 3/4 6-4 AMENDP.ENT N?S. A'O ; j SO'JTH TEXAS - UNITS 1 & 2 I i I
Fl - 3/4.7 PLANT SYSTEMS l BASES __ l 3/4.7.1 TURBINE CYCLE i 3/4.7.1.1 SAFETY VALVES The OPERABILITY of the main steam line Code safety valves ensures that c tho Secondary System pressure will be limited to within 110% (1413.5 psig) of ; 1 its design pressure of 1285 psig during the most severe anticipated system i operational transient. The maximum relieving capacity is associated with & Turbine trip from 100% RATED THERMAL POWER coincident with an assumed Ices sf condenter heat sink (i.e., no steam bypass to the condenser). The specified valve lift settings and relieving capacities ate in l
- accordance with the requirements of Section III of the ASME Boiler and Pressure Code, 1971 Edition. The total relieving capacity for all valves on ,
all of the steam lines is 20.65 x 105 lbs/h which is 122% of the total second-ary steam flow of 16.94 x 106 lbs/h at 100% RATED THERMAL POWER. A minimum of [ two OPERABLE safety valves per steam generator ensures that sufficient reliev- i ing capacity is available for the allowable THERMAL POWER restriction in i Table 3.7-1. L r STARTUP and/or POWER OPERATION is allowable with safety valves inoperable l
' within the limitations of the ACTION requirements on the basis of the reduc- !
1 tion in Secondary Coolant System Ateam flow and THERMAL POWfR required by the l l reduced Reactor trip settings of the Power Range Neutron Flux channels. The t Reactor Trip Setpoint reductions are derived on the following bases: Sp , (X) (Y)(V) x (109) Where: 1 SP = Red w.a Reacter Trip Setpoint in percent of RATED THERMAL POWER, V = Maximu'. neber cf inoperable safety valves per steam line, i 109 = Power Range Neutren Flux-High Trip Setpoint for four loop cperation,
- X = Total relieving capacity of all safety valves per steam l line in 1bs/ hour, and Y = Pasiru- relieving capacity of any one safety valve in 1bs/nour i
i 5 l SOUTH TEXAS - UNITS 1 & 2 B 3/4 7-1 AMEND".ENT N05. AV l
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i 1
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, PLANT SYSTEMS j BASES i 4 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. !
I Each auxiliary feedwater pump is capable of delivering a total feedwater flow of S40 gpm at a pressure of 1324 psig to the entrance of the steam genera- ; tors. This capacity is sufficient to ensure that adequate feedwater flow is i available to remove decay heat and reduce the Reactor Coolant System tempera- i ture to less than 350'F when the Residual Heat Removal System may be placed [ into operation. The AFW pumps are tested using the test line back to the AFST ;
~
and the AFW isolation valves closed to prevent injection of cold water into the steam generators. The STPEGS isolation valves are active valves required - to open on an AFW actuation signal. Specification 4.7.1.2.1 requires these I valves to be verified in the correct position. ! 3/4.7.1.3 AUXILIARY FEEDWATER STORAGE TANK (AFST) ! The OPERABILITY of the auxiliary feedwater ' storage tank with the minimum water volume ensures that sufficient water is available to maintain the RCS at i HOT STANDBY conditions for 4 hours with steam discharge to the atmosphe;e concurrent with total loss-of-offsite power followed by a cooldown to 350'F at 25'F per hour. The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical 3 characteristics. l l 3/4.7.1.4 SPECIFIC ACTIVITY ! The limitations on Secondary Coolant System specific activity ensure that
- the resultant offsite radiation dose will be limited to a small fraction of 10 CFR Part 100 dose guideline values in the event of a steam line rupture.
], This dose also includes the ef fects of a coincident 1 gpm primary-to-secondary , tube lean in the steam generator of the affected steam line. These values are l consistent with the assumptions used in the safety analyses. ) 3/4.7.1.5 MAIN STEAM LINE ISOLATION VALVES , I The OPERABILITY of the main steam line isolation valves ensures that no l
- more than one stear generator will blos down in the event of a steam line rupture. This restriction is required to: (1) minimize the positive reac-1 tivity effects of the Reacter Coolant System cooldown associated with the i 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-i ments are consistent with the assumptions used in the safety analyses.
l I SCUTh TEXA5 - UM T5 1 & 2 B 3/4 7-2 AMENDMENT N05. AC G3
b PLANT SYSTEMS BASES 3/4.7.1.6 ATMOSPHERIC STEAM RELIEF VALVES The atmospheric steam relief valves are required for decay heat removal and safe cooldown in accordance with Branch Technical Position RSB 5-1. In the safety analyses, operation of the atmospheric steam relief valves is assumed in accident analyses for mitigation of small break LOCA, feedwater line break, loss of normal feed ater and less-of-offsite power. 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 .taximum allowable fracture toughness stress limits. The limitations of 70 F and 200 psig are based on a stea'n generator RTNDT of 10*F and are sufficient to prevent brittle facture. 3/4.7.3 COMPONENT C00LIN3 WATER SYSTEM The OPERABILITY of the Coeponent Cooling Water System ensures that suf-ficient cooling capacity is available for continued operation of safety-related equipment during normal and accident conditions. The redundant cooling capacity of this system, assu.ing a single failure, is consistent with the assumptions used in the safety analyses. 3/4.7.4 ESSENTIAL C00LIN3 WATER SYSTEM The OPERABILITY of the Essential Cooling Water System ensures that suffi-cient cooling capacity is available for continued operation cf safety-related equipment during norral and accident condit. ions. The redundant cooling capac-ity of this system, assuming a single failure, is c -3istent with the assump-tion. used in the safety analyses. 3/4.7.5 ULTIPATE HEAT SINE The limitations en the ultieste heat sink lesel and temperature ensure j that sufficient cooling capacity is available either: (1) provide normal ! cooldown of the facility or (2) mitigate the effects of accident conditions within acceptable limits. The lir.itations on rinier water level and muimum temperature are based on providing a 30-day cooling water supply to safety-related equip ent without exceeding its design basis ten erature and is consistent with the recor enc-ations of Regu'.atory Guice 1.27, "Ultirate Heat Sink for Nuclear Plants," March 1974. 3/4.7.6 (Not used) t 6 3/4 7-3 AMEN 0"ENT A:5. AN: SOUTH TEXA5 - UNITS 1 & 2
R PLANT SYSTEMS BASES 3/4.7.7 CONTROL ROOM MAKEUP AND CLEANUP FILTRATION SYSTEM The OPERABILITY of the Control Room Makeup and Cleanup Filtration System ensures that: (1) the ambient air temperature does not exceed the allowable temperature for continuous-duty rating for the equipment and instrumentation cooled by this system, and (2) the control room will remain habitatle 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 HEPA 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 Criterion 19 of Appendix A, 10 CFR Part 50. ANSI N510-1980 will be used as a pro:edural guide for surveillance testing. 3/4.7.8 FUEL HANDLING EUILDIN3 EXHAUST AIR SYSTEM TM OPERABILITY of the Fuel Handling Building Exhaust Air System ensures that redioactive materials leaking from the ECCS equipment within the FHB fol-lowing a LOCA are filtered prior to reaching the environment. Operation of the system with the heaters operating for at least 10 continuous hours in a 31-day period is sufficient to reduta the buildup of moisture on the adsorbers and HEPA filters. The operation of this system and the resultant effect on offsite dosage calculations was assumed in the safety analyses. ANSI h510-1980 will be used as a procedural guide for surveillance testing. 3/4.7.9 SNUBBERS All snubbers are required OPERABLE to ensure that the structural integrity of the Reactor Coolant Syste and all other safety-related systems is rain-tained during and following a seiseic or other event initiating dynamic loads. Snubbers are classified and grouped by design and ennufacturer but not by size. For exarple, rechanical snutbers utilizing the sa e design features of the 2-kip, 10-kip and 100-kip capacity manufactured by Company "A" are of the sare type. The sa~e design rechanical snubbers manufactured by Co pany "B" f cr the purposes of this Technical Spe:ification would be of a dif f erent type, as would hydraulic snubbers frum either manufacturer. A list of individual snubber 5 with detailed inferration cf snubber locatic- , and size and of syste affe:ted shall be available at the plant in a:cordan:e ' with Section 50.71(c) of 10 CFR Part 50. The accessibility of each snubber shall be deterrined and a;; proved by the Plant Operations Review Committee. The determination shall be based upon the existing radiation lesels and the especte0 time to perforn a visual inspection in each snubber location as well as other fa: tors associated with accessibility during plant operations (e.g., temperature, , atmosphere, location, etc.), aad the recem-endations of Regalatory Guides 8.E I an: E.10. The accition or dele *.ien of any hydraulic or rechanical snLtter shall be race in accordance with Se:tien 50.59 of 10 CFR fart 50 SOUTH TEXAS - UMTS 1 & 2 E 3/4 7-4 AviC"EM h:5. AN:
1 FJ PLANT SYSTEMS BASES i i SNUBBERS (Continued) ~ The visual inspection frequency is based upon maintaining a constant
- 1 level of snubber protection to each safety-related system during an earthquake or severe transient. Thereferr., the required inspection interval varies inversely with the observed snubber failures on a given system and is determined j by the number of inoperable snubbers found during an inspection of each system.
1 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 constant with time and that the failure of any Snubber on that system could cause the system to be unprotected and to result in faGure during an assumed initiating event. Inspections performed before that int %rval has elapsed may be used as a new reference point to determine the next inspect'.ot.. However, the results of such early inspections performed before the original
.' required tire interval has elapsed (nominal time less 25%) may not be used *o i lengthen the required inspection interval. Any inspection whose results .equire
) a shorter inspection interval will override the previous schedule. ! The acceptance criteria are to be used in the visual inspection to determine i 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. l To provide assurance of snubber functional reliability, one of three 1 functional testing e,ethods is used with the stated acceptance criteria: i ! 1. Functionally test 10% of a type of snubber with an additional l 10% tested for each functional testing failure, or
- 2. Functionally test a satple size and determine sample acceptance or
. rejection using Figure 4.7-1, or l 3. Functionally test a representative sample size and determine sample acceptance or rejection using the stated equation.
l ] Figure 4.7-1 was developed using "Wald's Sequential Probability Ratio j Plan" as described in "Quality Control and Industrial Statistics" by Acheson J. Duncan, f Permanent or other exerptions fro'n the surveillance program for individual j snubbers ray be granted by the Co.raission if a justifiable basis for eserption is presented and, if applicable, snubber life destructive testing was perforced ! to qualify the snubbers for the applicable design conditions at either the cor-
- pletion of their fabrication er at a subsequent date. Snubbers so exempteu j shall be listed in the list of individual snubbers indicating the extent i of the exemptions.
The service life of a snubber is established via ranufactuter input and inferration throu;5 consideration of tne snutber service corditions anc
$0'JTH TEXA5 - U'GTS a &2 B 3/4 7-5 AMECMENT h25. AC l
l
FFA _P_LANT SYSTEMS EASES SNUBBERS (Continued) associated installation and maintenance records (newly installed snubbers, seal replaced, spring replaced, in high radiation area, in high temperature area, etc.). The requirerent to monitor the snubber service life is included to ensure that the snubbe*s periodically undergo a performance evaluation in view of their age and operating conditions. These records will provide statistical bases for future consideration of snubber service life. 3/4.7.10 SEALED SOURCE CONTAMINATION The limitations 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 frcm Byproduct, Source, and Special Nuclear Material sources will not exceed allowable intake values. i Sealed sources are classified into three groups according to their use, with Surveillance Requirerents 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 rechanism (i.e., sealed sources within radiation monitoring or boron reasuring devices) are considered to be stored and seed not be tested unless they are re. moved from the shielded mechanist. 3/4.7.11 (Net Usec) 3/4.7.12 (Net Used) 3/4.7.13 AREA TEMrERATURE MONITORIN3 The area temperature limitations ensure that safety-related equiprent will i not be subjected to te peratures in encess of their environrentt i valification terperatures. Exposure to encessive terperatures eay degrace equipment and can cause a loss cf its 0FERAEIL!TY. The terperature licits include an allo.ance for instrurent error of 3*F rasies . 3/4.7.14 ESSENTIAL CHILLED WATER SNSTEP The OPERABILIT) cf the Essential Chilled Water Systen ensures that suffi-cient ccoling capacity is available for continued operatic, of safety-relateo equiprent curing norral anc a:cicent corditions. The recundant cooling capacity 4 of this systeT, assu in; a single failure, is consistent with the assur;tions used in the saf ety analy ses. t TEXA5 - LNIT5 1 & 2 E 3/4 7-E AviN:"ist N;5. AN: 50 J'-
F0 3/4.8 ELECTRICAL POWER SYSTEMS BASES - 3 /4. 8.1, 3/4. 8. 2, a nd 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 durirg 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 minimu specified independent and redundant A.C. and D.C.
- power sources and distribution systems satisfy the requirements of General Design Criterion 17 of Appendix A to 10 CFR Part 50.
3 The ACTION tequirerents specified for the levels of degradation of the t power sources provide restriction upon continued f acility operation comensurate with the level of degradation. The OPERASILITY of the power sources are consistent with the initial condition assu ptions of the safety analyses and are based upon raintaining at least tmo redundant gets of onsite A.C. and 0.C. power sources and asscciated distribution systems OPERABLE during accident conditiens coincident with an assured less-of-offsite po.er and single failure of the other onsite A.C. source. The A.C. and 0.C. source allowable out-of-service times are based en Regulatory Guide 1.93, "Availability of Electrical Power Sources,' Cecember 1974. When one standoy diesel generator is inoperable, there is an additioral A; TION require ent to verify that all required systees, subsystems, trains, ccrponents and devices, that depend on the remaining OPERABLE l standoy diesel generators as a source of emergency power, are also OPERABLE, and that the stea -driven auxiliary feed ater pump is OPERABLE. This require-cent is intended to provide assurance that a loss-of-of fsite power event will not result in a coeplete loss of safety function of critical systems during the period one of tne standoy diesel generators is inoperable. The term, verify, 2 as used in this contest resns to administratively check by examining logs or other informaticn to cetermine if certain co ponents are out-of-service for maintenance er othe- reas:ns. It d:es not mean to perform the Surveillance Require ents neeced to ceronstrate tne OPERABILITY of the component. 2 The OPERAEILIT) c' the rinire- specified A.C. and 0.C. pc.er seurces and associated cistritatica syste ! caring shutco n and refueling ensures that: (1) the f acility can te r.aintained in the shutdc.n or refueling conditien for extended tire periods, a M (2) sufficient instrurentation and control capa-4 l tility is available fcr remiterir; and raintaining the unit status. The Surveilla*:e Ke;;uire ents for deT:nstrating the OPERAEILITY ef the diesel generators are in a:cercance with the recorsendations of Fegulatory Guide. 1.9, "Selt:tien of Ciesel Generator Set Capacity for Stanooy Power Supplies," Revisien , De:e eer 1979; 1.10E, "Periodic Testing of Diesel Generator Units Use: as Onsite Electric Po.ee Systers at huclear Po er Plants," 4 Revision 1, August 1977; and ASTM 0975-E1, ASTM 01552-79, ASTM 02622 82, ASTM 04294-83, and ASTM 02276-76. The standoy diesel generators auxiliary systems J are designed to circulate ware oil and mater through the diesel while the diesel is n:t ru ning, to preclude cold a-bient starts. For the purposes of a surveillan:e testing, a-tie-t conditic s are consicered to be the het pre-
)
luce cen:iticr. AN Sr -
- LAIT 5 1 & 2 E 3/4 E-1 AMEN: MENT h:5.
t.;
Ff l ILECTR1tAL 90WER SYSTEMS BASES A.C. SOURLES, L.C. SOURCES, and ONSITE POWER DISTRIBUTION (Continued) The Surveillance Requirements 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 huclear Fo.er Plants," February 1978, and IEEE Std 450-1980, "1EEE Recommended Practice for Maintenance, Testing, and Replacement of Large Lead Storage Batteries for Generating Stations and Substations." Verifying average electrolyte temperature above the minimum for which the battery was sized, total battery terminal voltage on float charge, connection resistance values, and the perforeance of battery service and discharge tests ensures the effectiveness of the charging system, the ability to handle high discharge rates, and compares the battery capacity at that time with the rated capacity, i 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 fer 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, are characteristic of a charged cell with ariequate capacity. The norral limits for each connected cell for float voltage and specific gravity, greater than 2.13 volts and not more than 0.020 below the ranufacturer's full charge specificgravitywithanaveragespecificgravityofalltheconnectedcells not more than 0.010 belon the manufacturer s full charge specific gravity, ensures the OPERASILIT) and capability of the battery. l Operation with a battery cell's parameter outside the normal limit but within the allowable value specified in Table 4.8-2 is permitted for up to . 7 days. During this 7-cay period: (1) the allowable values for electrolyte level ensures no physical da age to the plates with an adequate electron transfer capability; (2) the allo.able value for the average specific gravity of all the cells, not rore than 0.020 below the ranufacturer's recom ended full charge specific gravity, ensures that the decrease in rating will be less than the safety rargin presided in sizing; (3) the allo.able value for an incividual cell's specific gra,ity, ensures that an individual cell's specific grasity will not be more than 0.040 t:elo. the ranuf acturer's full charge specific gravity and that the overall capability of the battery will be raintained l
! within an acceptable limit; and (4) the allowable value for an individJal cell's ficat voltage, greater than 2.07 volts, ensures the battery's capability to perfore its cesign fun:ticn.
I i E 5/4 E-2 Avi'C#ENT h05. A'C i 50J1H TE MS - U'i:T5 1 & 2 i
FD ELECTRICAL POWER SYSTEMS BASES , 3/4.8.4 ELECTRICAL EQUIPMENT PPOTECTIVE DEVICES Containment electrical penetrations and penetration conductors are pro-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. The Surveillance Requirements applicable to lower voltage circuit breakers provide assurance of breaker reliability by testing a representativ: sa,ple of at least 10% of each manuf acturar's brand of circuit breaker. Each manufacturer's molded case and metal case circuit breakers are grouped into representative samples which are then tested on a rotating basis to ensure that all breakers are tested. If a wide variety exists within any manufacturer's brand of circuit breakers it is necessary to divide that manufacturer's breakers into groups and treat ea:h group as a separate type of breaker for surveillance purposes. The molded case circuit breakers will be tested in accordance with NEMA Standard Publication No, AB-2-19EO, For a frame size of 250 amperes or less, the field toltrance of the high and low setting of the injected current will be within + 40%, -25't of the setpoint (pickup) value. For a frame size of 400 a peres or greater, the field tolerance will be t 25% of the setpoint (pickup) value. The circuit treakers chould not be affected when tested with-in their tolerance. E 3/4 E 3 A*Eh TENT N25. AC SCUle TE)A5 - UNIT 5 1 &
b 3/4.9 AEFUELING OPERATIONS BASES __ 3/4.9.1 BORON CONCENTRATION The limitations on reactivity conditions during REFUELING ensure that: (1) the reactor will remain subcritical during CORE ALTERATIONS, and (2) a uniform boron concentration is maintained for reactivity control in the water volume having direct access to the reactor vessel. These limitations are censistent 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% ak/k conservative allowance for uncertainties. Similarly, the boron concentration value of 2500 ppm or greater includes a conservative uncertainty allowance of 50 ppm boron. The locking 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 ': '3
,. OECAY TIME The minimum requirement for reactor subcriticality 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 tire is consistent with the assumptions used in the safety analyses for the rapid refueling design.
2/4.9.4 CONTAINMENT BUILDIN3 PENETRATIONS The requirements on containrent building penetration closure and OPERABILITY ensure that a release of radioactive raterial within contain?ent will be rest *icted f rom leauge to the environment. The OPERABILITY and closure restrictions are sufficiert to restrict radioactive raterial release fro? a fuel eierent rupture based upon the lack of containeent pressurization potential while in the REFUELIN3 MODE. 3/4.9.5 _CO'WNI C AT !ON S The recuirerent fer co- unications capability ensures that refueling station persevel can be pre ptly inferred of significant changes in the f acility statu. or core rea:tivity ccnditions during CORE ALTERATIONS. 50JiH TE) AS - UN:T51 & 2 E 3/4 9-1 AuEN; MENT N05. AN:
?g i
FD l l I REFUELING OPERATIONS RASES l 3/4.9.6 REFUELING MACHINE , The OPERABILITY requirements for the refueling machine and auxiliary i hoist ensure that: (1) the refueling machine and auxiliary hoist will be used I I for movement of drive rods and fuel assemblies. (2) the refueling machine has
. sufficient load capacity to lift a drive rod or fuel assembly, and (3) the core i internals and reactor vessel are protected from excessive lifting force in the '
event they are inadvertently engaged during lifting operations. , t
- 3/4.9.7 CRANE TRAVEL - FUEL HANDLING BUILDING
- The restriction on movement of loads in excess of the nominal weight of a 'I fuel and control rod assembly and associated handling tool over other fuel [
l assemblies in the storage pool, unless handled by the single-failure proof main , i hoist of the FHS 15-ton crane, ensures that in the event this load is dropped: ! j (1) the activity release will be limited to that contained in a single fuel l assembly, and (2) any possible distortion of fuel in the storage racks will not I
- result in a critical array. This assuatption is consistent with the activity l' release assumed in the safety analyses.
} 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION s l The requirement that at least one residual heat removal (RHR) loop be in [ t op,eration ensures that: (1) sufficient cooling capacity is available to remove l decay heat and maintain the water in the reactor vessel below 140'F as required ! l during the REFUELING MODE, and (2) sufficient coolant circulation is maintained , } j through the core to minimize the effect of a boron dilution incident and prevent i ? boron stratification. ! The requirement to have two RHR loops OPERABLE when there is less than i 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 i removal capability. With the reactor vessel head removed and at least 23 feet l of water above the reactor pressure vessel flange, a large heat sink is avail
- l able for core cooling. Thus, in the event of a failure of the operating i RHR locp, adequate time is provided to initiate emergency procedures to cool the core.
i 1 l j . 3/4.9.9 CONTAINMENT VENTILATION ISOLATION SYSTEM [ I t The OPERABILITY of this system ensures that tne containment purge an- ) exhaust penetrations will be automatically isolated upon detection of high ! radiation levels in the purge exhaust. Tne OPERABILITY of this system is ! l required to restrict the release of racioactive material from the containment I atmosphere to the environment. I , t ! I t SOUTH TEXA5 - UNITS 1 1 2 B 3/4 9-2 AuENDMENT N05. N r t
F1 REFUELING OPERATIONS BASES 3/4.9.10 and 3/4.9.11 WATER LEVEL - REFUELING CAVITY and STORAGE POOLS The restrictions on minimum water level ensure that sufficient water i depth is available to remove 99% of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel asserrbly. The minimum water depth is consistent with the assurptions of the safety analysis. 3/4.9.12 FUEL HANDLING BUILDING EXHAUST AIR SYSTEM The limitations on the Fuel Handling Building Exhaust Air 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 build-up of rnoisture on the adsorbers and HEPA filters. The OPERABILITY of this system and the resulting iodine removal capacity are consistent with the assurp-tions of the safety analyses. ANSI N510-1980 will be used as a procedural guide for surveillance testing. t I sm Ttus - usns 1 t e an 9-3 AeEs:'ust s:s. m l
b l 3/4.10 SPECIAL TEST EXCEPTIONS i EMis ; 3/4.10.1 SHUTDOWN PARGIN This special test exception provides that a minimum amount of control rod worth is immediately av&ilable for reactivity control when tests are performed for control rod worth neasureeent. 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 fuel cycling operations. 3/4.10.2 GROUP HEIGHT. INSERTION, AND POWER DISTRIBUTION LIMITS This special test exception permits individual control rods to be positioned outside of their norral group heights and insertion limits during the performance of such PHYSICS TESTS as those required to: (1) r.easure control rod worth, and (2) determine the reactor stability index and damping factor under xenon oscillation conditions. 3/4.10.3 PHYS!CS TEST _5 This special test exception permits PHYSICS TESTS to be performed at less than or equal to 5% of RATED TnERMAL P0a'ER with the RCS i slightly lo.er than ner? ally allemed 50 that the fundstental nuclear characteristics of the core ar.d related instru entation can be serified. In order for various charac-teristics to be accurately ressured, it is at tires necessary tc operate outside the norral restricticns of these Technical Specifications. For instance, to reasure the roderater te perature 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 which in turn may cause the RCS T,y9 to fall slightly below the minima?. temperature of Specification 3.1.1.4. 3/4.10.4 REACTOR COOLANT LO:05 This special test exception permits reactor criticality under no fic. conditiens and is recuired to perfor. certain STARTUP and PHYSICS TESTS while at 10. THERMAL F0aER le.cls. 3/4.10.5 FOSITION IND::A'!0% $YSTEM - SHUTD0aN This special test enception oerrits the Position Indication Syste s to be incperable during rod crcp tire reasure?ents. The esception is required since the data necessary te ceterrine the rcd drcp tire are derised fro 9 the induced voltage in the pcsitic" insicator coils as the rod is orepped. This induced voltage is seali cc ;sud te tre norral soltage and, theref ore, cannet be observed if the F sitien Incication Systers remain C5ERASLE. SOUTH TERA 5 - LN T5 1 & 2 E 3/4 1C 1 AYIN: MENT N:5 AN: A. j,
Fl 3/4.11 RADIOACTIVE EFFLUENTS
$5ES 3/4.11.1 LIQUID FFFLUENTS 3/4.11.1.1 CON 0ENTRATION This specification is provided to ensure that the concentration of radio-active eaterials released in liquid waste effluents to UNRESTRICTED AREA $ will be less than the concentration levels specified in 10 CFR Part 20, Appendix E.
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 !!.A design objectives of Appendix I, 10 CFR Part 50, to a MEMBER OF THE PUBLIC, and (2) the limits of 10 CFR Part 20.106(e) to the population. The concentration limit for dissolved or entrained noble gases is based upon the assueption that Xe-13T is the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent concentratien in water using the eethods described in Inter-national Cor. mission on Radiological Protection (ICRP) Publication 2. This specification applies to the release of radioactive eaterials in liquid effluents from all units at the site. . The required dete tion capabilities for radioactive eaterials in liquid waste sa?ples are tabulated in teres of the lower limits of detection (LLDs). Detailed discussion of the LLD, and other detection limits can be found in HA5L Procedures Manual, HA5L-300 (redsed annually), Currie, L. A., "Limits for Qualitative Detection and Quantitative Determination - Application to Radio-chemistry," Anal. Che*. 40, 566-93 (1958), anc Hartwell, J. K., "Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-5A-215 (June 1975), 3/4.11.1.2 005E This specification is provided to implement the requirements of Sec-tions II. A. Il!. A and IV. A of Appendim I, 10 CFR Part 50. The Limiting Qndi-tien for 0;eraticn irplements the guides set forth in Section II. A cf Acpendix 1. The ACTION state ents p* ovide the required eperating flexitility anc at the sa e tire irpierent the guides set fcrth in Section IV.A of Appendia I to assure that the releases of radioactive raterial in liquid effluento to UNRESTRICTED AREAS will be kept "as 10 as is reasonably achievable." The dose calculation retno-dology and paraaete s in the ODCM irplement the requirements in Section Ill.A of AppendiA 1 that conferrance with the guides of Appendix ! be sho.n by calcula-tional procedures based on recels and dats, such that the actual exposure of a MEMEER OF THE NELIC through a;propriate pathways is unlikely to be substa9-tially underestimated. Tne equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid efflu-ents are consistent with the rethedology provided in Regulatory Guide 1.109, , "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Corp 11ance with 10 CFR Part 50, Appendix I" Revi-sicq 1, Octeter 1977 and Regulatory Guide 1.113, "Esticating Aquatic Dispersion of Ef fit.ents f rc- Accicental and Routine Rea:ter Releases for the Furpcse of leplerenting Appencis 1,' April 197L SOUTH TE) A5 - UNHS 1 & 2 6 3/4 11-1 AMEC MENT h:5. AV
F RADI0 ACTIVE EFFLUENTS L'EU= OOSE (Continued) This specification applies to the release of radioactive materials in liquid effluents from each unit at the site. 3/4.11.1.3 LIQUID WASTE PROCESSIN3 SYSTEM The OPERABILITY of the Liquid Waste Processing System ensures that 'his system will be available for use whenever 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 releases of radioactive materials in liquid effluents will be kept "as low as is reasonably achievable". This specification irplements the requirements of 10 CFR 50.36a, Genera', Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design cbjective given in Section II.D of Appendix ! to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the Liquid Waste Processing Systen were specified as a suitable fraction of the dose design objectives set forth in Section II.A of Appendix 1, 10 CFR Part 50, for liquid effluents. This specification applies to the release of radioactive r.aterials in liquid effluents from each unit at the site. 3/4.11.1.4 LIQUID POLOUP TANK 5 The tanks listed in this specification include all those outdoor rad aste tanks that are not surrounded by liners, dikes, or walls capable of holding the tank contents and that do not have tank overflows and surrounding area drains connected to the Liquid Waste Processing Syst e. Restricting the quantity of radioactive rnaterial contained in the specified tanks provides assurance that in the event of an uncontrolled release of the tanks' contents, the resultirg concentrations would be less than the litits of 10 CFR Fart 20, A;pendia E, Table !!, Colu-n 2, at the nearest potable water suppij and tne nearest surf ace water supply in an UNRESTRICTED ARF'.. 3/4.11.2 GASEOUS EFFLUENTS 3/4.11.2.1 005E dATE This specification is provided to ensure that the dose at any tire at and beyond the SITE EDUND'R) frem gasecus effluents fror all units on the site will be within the annual dese lirits of 10 CFR Part 20 to UNRESTRICTED AREA 5. Tte ann,,31 dase licits are the deses associated with the concentrations of 10 CFR Part 20, Appendix E. Table 11, Colv n 1. These limits provide reasonatie assurance that radioactive raterial discharged in gaseous ef fluents will not result in the exposure of a MEMBER OF THE PUBLIC in an UNRESTRICTED AREA,
- either within or outsice the SITE BOUNDARY, to annual average conctntratiers exceeding the litits specified in Appendix B, Table Il of 10 CFR Part 20 (10 CFR Fart 20.1C6(t)). For PEMEERS OF TriE PUBLIC who ray at tires be within SOUTH TD AS - LNTS 1 & 2 E 3/4 11-2 AMEN; MENT N05. AND
h$ RA010 ACTIVE EFFLUENTS BASES DOSE RATE (Continued) the SITE BOUNDARY, the occupancy of that MEMBER OF THE PUBLIC will usually be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the SITE BOUNDARY. Examples of calculations for such MEMBERS OF THE PUBLIC, with the appropriate occupancy factors, shall be given in the 00CM. The specified release rate limits restrict, at all tires, the corresponding gama and beta dose rates above background to a HEMBER OF THE PUBLIC at or beyond the SITE BOUNDARY to less than or equal to 500 mrees/ year to the whole body or to less than or equal to 3000 mrems/ year to the skin, r These release rate limits also restrict, at all times the corresponding thyroid dose rate above background to a child via the inhalation pathm'ay to less thkn or equal to 1500 mrems/ year. This specification applies to the release of radioactive materials in gaseous effluents from all units at the site. The required detection capabilities for radioactive materials in gasecus waste samples are tabulated in terms of the lower limits of detection (LL0s). Detailed discussion of the LLO, and other detection limits can be found in NASL Procedures Manual HASL-300 (revised annually), Currie, L. A., "Limits for Qualitative Detection and Quantitative Deterrination - Application to Radio-chemistry," Anal. Cher. 40, 586-93 (196B), and Hartwell, J.K., "Detection Linits for Radicanalytical Countine; Techniques," Atlantic Richfield Hanford Corpany Report ARH-5A-215 (Juts . 0 5)- 3/4.11.2.2 00$E - NOBLE GASES This specification is provided to irplerent the requirements of Sections
!!.B III.A and IV.A of Appendix 1, 10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section !!.B of Appendix 1.
The ACTION staterents provide the required operating flexibility and at the same tire implerent the guides set forth in Secticn IV.A of Appendix I to assure that the releases cf radioactive eaterial in gaseous effluents to UN'ESTRICTED AREAS will be 6ept "as le as is reasonably achievable." Tre Surveillance Requirerents irple ent the require ents in vection Ill.A of Appendix ! tnat c9nferrance with the guides of Appendix ! be shewn by calcula* f ticnal procedures based on rodels and data such that th. actual exposure of a MEMBER OF THE PUELIC through a:prcpriate path ays is un'likely to te substantially uncerestirated. The dose calculation rethocology and para et,ers established 1 in the 00:M for calculating the coses due to the actual release rates cf radicactive raterials in liquid ef fluents are consistent with the rethodeleg., provided in Regulatory Guide 1.109, "Calculation of Anrual Doses to Man fr:- Routine Releases of Reactor Ef fluents for the Purpcse of Evaluating Co p11ance with 10 CFR Par' 50, Appsndix I" Revision 1, Octeter 1977 and Regulatory l Guide 1.111, Mothocs f or Esticating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors," Revision 1," July 1977. Tne 00Cv equations provided for determining the air l d0ses at and te>cra tre $1TE BOLVRf are basec upOn the histcrical average . ate:sp"eric cenditiens. I 50 Viri TExA5 - UNITS 1 & 2 E 3/4 11-3 A"iN:"ENT N:L A'O J
FL RADI0 ACTIVE EFFLUENTS $$ES This specification applies to the release of radioactive eaterials in gasecas effluents from each unit at the site. 3/4.11.2.3 DOSE - 100!NE-131, 100!NEy133, TRITIUM, AND RADI0 ACTIVE MATERIAL E PARTICULATE FORK This specification is provided to implement the requirements of Sections II.C. III.A and IV.A of Appendix I, 10 CFR Part 50. 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 ! to assure that tr.e releases of radicar.tive materials in gaseous effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." The 00CM calculational ; rethods specified in the Surveillance Requirements irplement the requirements in Section III. A of Appendix ! that conforrance with the guides of Appendix ! be shown by calculational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate path.ays is unlikely to be substantially underestimated. The ODCM calculational methodology and parateters for calculating the doses due to the actual release rates of the subject materials are consistent with the rethodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Ef fluents for the Purpose of Evaluating Corpliance with 10 CFR Part 50 Appendix I," Revision 1, Octobtr 1977 and Regulatory Cuide 1.111. "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Ef fluents in Routine Releases f rom Light-Water-Cooled Reactors," Revision 1. July 1977. These ecua-tions also provide for determining the actual doses based upon the historical average atmospheric conditions. The release rate specifications for Iodine 131, lod:ne-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days are (Jependent upon the existing radionuclide pathways to r.an, in the areas at and beyond the SITE BOUNDARY. The pathnays that were examined [ in the development of these calculations were: (1) individual inhalation of air- i torne radionuclides, (2) der:sition of radionuclides onto green leafy segetation ' with subsequent consu ption by ran, (3) deposition onto grassy areas where r, ilk anirals anc rest produ:ing anicals graze with consurption of the rill and Feat by can, and (4) deposition on the grcund with subsequent enposure to man. This specificatien applies to the release of rrdioactise materials in gaseous effluents fro? ta:h unit at the site. 3/4.11.2.4 GASEOUS V'5TF FROCE551N3 SYSTEM The OPERASILITY ef ths GA5EOUS WASTE PROCESSIN5 SYSTEM ensures that tre systers will te available fcr use wheneser gaseces effluents require treatnert prior to rel(ase to the environrent. Tne requirerent that the appropriate po" tions of these syste s be usec, when specifi. provices reascnable assurance that the releases of radioactive caterials in gaseous effluents will be kept "as low as is reasonably achievable". This specification irple ents the re-quirerents of 10 CFR 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 5: 390 tte cisign coje:tists given in Se:tien 11.0 cf Appeadix ! tc 10 CfR Part 50. Tre :mcifie: lirits go.erning the t.se cf re;repriate pertiens , SOU% TEXAS - UNIT 51 !. 2 E 3/4 11-4 A"EN:"ENT N'5. A'C
FA010 ACTIVE EFFLUENTS fLMES CASE 0VS WASTE PROCESSINO SYSTEM (Continued) of the system were specified as a suitable fraction of the dose design objec-tives set forth in Section !!.B and II.C of Appendix I, 10 CFR Part 50, for gaseous effluents. This specification applies to the release of radioactive materials in gaseous effluents from each unit at the site. 3/4.11.2.5 EXPLOSIVE GAS MIXTURE This specification is provided to ensure that the concentration of poten-tially explosive gas mixtures contained ai the GASEOUS WASTE PROCE551N3 SYSTEM is maintained below the flar. ability limit of oxygen. The concentration of oxygen in the inlet header to the GASE0V5 WASTI PROCESSING SYSTEM is contin-uously monitored and a high level alarm isolates the GASEOUS WASTE PROCE55th3 SYSTEM. Provision is made to ranua11y purge the system with nitrogen and/or isolate the source of oxygen. Maintaining the concentration of oxygen below its flam ability limit (4% by volu-e) provides assarance that the releases of radioactive raterials will be controlled in conformance with the requirerents of General Design Criterion 60 of Appendix A to 10 CFR Part 50. Because of analyzer variabilities, a safety rargin of 1% by volure is apy *ed. Therefore, the liritin0 condition for operatien is raintaining oxygen cen:entration belo. 34 by volu e. 3/4 11.2.6 GAS STORA3! TAMS The tanks include:1 in this specification are those tanks for which the quantity of radioactivity contained is not limited directly or indirectly by another Technical Specificatien. Restricting the quantity of radioactivity contained in ea:h gas storage tank presides assurance that in the event of an uncontrolled release of the ta%'s contents, the resulting whole body esposure te a MEMEER OF InE FLELIC at the resrest SITE EOU'CARY will not exceed 0.5 re . This is consistent with Staadard Resie. Plan 11.3, Branch Technical Fcsitien ETSB 115, "Fostulatec F acica:tise Releases Due to a waste Gas Syste- Lea. er Failure," in huREG-05:0, July 1951. Since only the ga- a bocy dose fa: tor (DFB3 ) is us d in the analysis, tre >e 133 equisalent is determined frc- the DFEg s al t.e for ke 133 as corparea to the cc posite CFE, for the a:tual risture in the ta s n .
- 5 e e
h RADI0 ACTIVE EFFLUENTS
$MLS _
3/4.11.3 SOLIO RADICACTIVE WASTES This specification irplements the requirements of 10 CFR 50.36a and General Design Criterion 60 of Appendix A to 10 CFR Part 50. The process parameters included in establishing the PROCESS CONTROL PRO *.iRAM may include, but aro not limited to, waste type, waste ph, waste / liquid /50L101FICATION agent / catalyst ratios, waste oil content, waste principal chemical constituents, and mixing and curing times. 3/4.11.4 TOTAL DOSE
- this specificatien is provided to reet the dose limitations of 40 CFR Part 190 that have been incorporaied into 10 CFR Part 20 by 46 FR 18525. The specification requires the preparation and submittal of a Special Report when-ever the calculatec c:ses due to releases of radioactivity and to radiation fior uranium fuel cycle sources es:eed 25 mrems to the whole body or any organ, except the thyroid, which shall te limited to less than or equal to 75 mrees. For sites containing up te fcur reactors, it is highly unlikely that the resultant dose to 3 MEMEER CF THE PUELIC will exceed the d:se limits of 40 CFR Part 190 if the individual rea:tcrs re sin within t ice the dose design obje:tives of Appendix 1, and if dire:t radiation d0ses fro
- the rea: tor units and outside storage tanks are ke;t s all. The Special Report will describe a course of I action that should result in the limitation of the annual dose to a MEMBER OF THE PUBLIC to .ithin tre 40 CFR Part 190 limits. For the purposes of the Special Rep:rt, it may be assuae: that the dose commiteent to the MEMBER of the PUBLIC from other uraniu- fuel cycle sources is negligible, with the enception that dose contributions fron other nuclear fue" cycle facilities at the sa*e site or within a racius of 8 km must be considered. If the dose to any MEMBER OF THE PLELIC is estimated to exceed the requirements of 40 CFR 4
Part 190, the Special Fe; ort with a request for a variance (provided the release conditions resulting in violation of 40 CFR Part 190 have not already been corre:tec), in accordan:e with the provisiens of 40 CFR 190.11 and 10 CFR 20.4 M : is censicered te t.e a tirely request and fulfills the recuire ants cf 40 CFR Part 19D until ht staff a:tica is co pleted. The variance enly relates to the lirits of C CFF fa n IE , and d:es not ap:1y in any way to the other requirements for c:se liritatien cf 10 CFR Fart <*0, as addressed in Spe:ifi-cations 3.11.1.1 and 3.11.2.1. An individsal is rot considered a MEMEEE Of THE PUELIC durir; any peric: in wnich he/she is enga;ed in carrying out any operation that is part of the nu: lea
- fuel cy:le.
50;in TEiAS - L'dTS 1 & 2 E 3/4 11-6 AMEC "ENT M 5. AV
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FC 3/4.12 RADIOLOGICAL ENVID.0 MENT.'.L MONITORING BASES 3/4.12.1 MONITORING PROGRAM The Radiological Environmental Monitoring Program requirec, oy tais specification provides representative measurements of radiation and of radio-active materials in thoss exposure pathways and for those rad'onuclides that ' lead to the highest potential radiation exposure of MEMBERS OF THE PUBLIC resulting from the plant operation. This monitoring program implements Sec-tion IV.B.2 of Appendix i to 10 CFR Part 50 and thereby supplements the Radio-logical Effluent Monitoring Program by verifying that the m asurable concentra-tions of radioactive raterials 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 pro-vided by the Radiological Assessment Branch Technical Position on Environmental Monitoring. The initially soecified monitoring program will be effective for at least the first 3 years of com*ercial operation. Following this period, progra't changes mey be initiated based on operational experience. The required detection capabilities for environmental sample analyses are tabulated in teres of the lo.er limits of detect. ion (LLDs). The LL0s required by the ODCM are consicered optimas for routine environmental reasurements dr. industrial laboratories. It shovid be recognized that the LLD is defired as an a priori (before the fact) limit representing the capability of a heasure-ment system and net as an a tolterieri (after the fact) limit for a Tarticular reasurerent. Detailed discussion of the LLO, and other detection limits, can te found in HASL Procedures Panual HA5L-300 (revised annually), Currie, L. A. , "Limits for Qualitative Detection anc Quantitative Determination - Application to Radiocheristry," Anal. Cne- 40, 556-93 (1968), and Hartwell, J. A., "Detection Limits for Radicanalytical Counting Terhniques" Atlantic Richfield Hcnford Company Report AN-Sa-215 (June 1975). 3/4.12.2 LAND USE CEN WS This spe:iffratica is rrtvided to ensure that changes in the use of areas at and beyend the SITE EX' Cur are icentified and that modifications to the Radiological Environmental P:nitoring Program given in the 00CM are made if icquired by the rescits cf this census. The best inferration f'rce the coer-to-door survey, frc serial surwey or froa consultirig with local agricultural authorit'en shall t'e used. This census satisfies the require ents of Sec-tion iV.B.3 of AoDerd a 1 to 10 CFR Fart 50. Restricting the census to garcens of greater than E; r; prosices assuran:e that significant espesure path.sys via r leafy vegeta!1es will Le iceqified and ronitored sinc 6 a garden of this size is the r.inire re;uired to precuce the cuantity (26 'g/> ear) c' leafy vege-tables assc ed in Eegulaterv Guice 1.109 for consumpt.on by a child. To ceter- ' eine this rinimum garden size, the f ollo.ing assuwptio is were made: (1) 2Ct
- of the garden =,ts used for gro.ing broad leaf vegetation (i.e., similar to lettuce and catonge), and (2) a segetation yield of 2 kg/r:.
i l S NT H T E A AS - LN T 5 1 & 2 E 3': 12-1 AuEN;"ENT N;5. AC f
f;l RADIOLOGICAL.INVIRONMENTAL HONITORING BASES __ 3/4.12.3 INTERL' .;< '9MPARISON PROGRAM The require w - trticipation in an approved Interlaborator/ Comparison Program is prov'.c .re that independent checks on the precision and accuracy of the , ts of radioactive materials in environmenta' sample matrices are p. .- 3 part of the quality assurance program for environ-mental ronitoring . order to demonstrate that the results are valid for the purposes of Section IV.B.2 of Appendix I to 10 CFR Part 50.
.c 1
i B 3/4 12-2 AMENOMENT NOS. AN3 I SOUTH TEXAS - UNITS 1 & 2 l
g SECTION 5.0 DESIGN FEATURES i l i l lt $ ' , , i l
h 5.0 DESIGN FEATURES 5.1 SITE-EXCLUSION AREA 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 SITE BOUNDARY FOR RADI0 ACTIVE GASEOUS AND ETUUTUlFFLUENTS 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 SITE BOUNDARY that are accessible to MEMBERS OF THE PUBLIC, shall be as shown in Figures 5.1-3 and 5.1-4. The definition or UNRE51RiCIED AREA usea in implementing these Technical Speci-fications has been expanded over that in 10 CFR 20.3(a)(17). The UNRESTRICTED AREA boundary may coincide with the Exclusion (fenced) 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 SITE BOUNDARY, is utili7ed in the Limiting Conditions for Operation to keep levels of radioactive materials in liquid and gaseous effluents as low as is reason-ably achiesable, pursuant to 10 CFR 50.36a. 5.1 CONTAINMENT CONFIGURATION 5.2.1 The reactor containment building is a steel-lined, reinforced concrete building of cylindrical shape, with a dome roof and having the following design features:
- a. Nominal inside diameter = 150 feet.
- b. Nominal inside height = 241.25 feet.
- c. Minimum thickness of concrete walls = a feet,
- d. Minimum thickness of concrete roof = 3 feet.
- e. Minimum thickness of concrete floor eat = 18 feet,
- f. Nominal thickness of steel liner = 3/8 inches.
- g. Net free volume = 3.56 x 108 cubir. feet.
DESIGN PRESSURE AND TEMPERATURE 5.2.2 The reactor containment building is otsigned and shall be maintaineo for a maximum internal pressure of 56.5 psig and a temperature of 286'F. , 5-1 AMENDMENT NCS. AND SOUTH TEXAS - UNITS 1 & 2 l~,.* i . .
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RESTRICTED AREA AND SITE BOUNDARY FOR RADI0 ACTIVE LIQUID EFFLUENTS M.
FP DESIGN FEATURES , __ 5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The core shall contain 193 fuel assemblies with each fuel assembly con-taining 264 fuel rods clad with Zircaloy-4. Each fuel rod shall have a nominal active fuel length of 168 inches. The initial core injing shall have a maxi-mum enrichment of 3.5 weight percent U-235. Reload fvkl shall be similar in physical design to the initial core loading and sha!' have a maximum enrichment of 3.5 weight percent U-235. ; CONTROL R00 ASSEMBLIES 5.3.2 The core shall contain 57 full-length control rod assemblies. The full-length control rod assemblies shall contain a nominal 158.9 inches of absorber material. The absorber material shall be hafnium. All control rods shall be a clad with stainless steel tubing. 5.4 REACTOR COOLANT SYSTEH. DESIGN PRESSURE AND TEMPERATURE 5.4.1 The Reactor Coolant System is designed and shall be maintained:
- a. In accordance with the Code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements,
- b. For a pressure of 2485 psig, and
- c. For a temperature of 650 F, except for the pressurizer which is t 680 F.
VOLUME , 5.4.2 The total water and steam volume of the Reactor Coolant System is 13,814 ,
+ 100 cubic feet at a nominal T,yg of 561'F.
5.5 METEOROLOGICAL TOW s LOCATION ( 5.5.1 The meteorological towers shall be located as shown on Figure 5.1-1. 5.6 FUEL STORAGE 7 CRITICALITY j 5.6.1 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 i unborated water, which includes a conservative allowance of I
5-6 AMENDMENT N05. AND SOUTH TEXAS - UNITS 1 & 2 li . . , ,
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0 DESIGN FEATURES 0.0185 ok for Re ion 1 uncertaiaties and tolerances and 0.0259 Ak for Re ion 2 uncertainties and tolerances.
- b. A nominal 10.95 ir.ches center to center distance between fuel assemblies in Region 1 of the storage racks and a nominal 9.15 inches center to center distance between fuel assemblies in Region 2 of the storage racks,
- c. Neutron absorber (Boraflex) installed between spent fuel assemblies in the storage racks in Region 1 and Region 2.
- d. Region 1 of the spent fuel storage racks can be use'd to store fuel which has a U-235 enrichment less than or equal to a nominal 4.5 weight percent. Region 2 can be used to store fuel which has achieved sufficient burnup such that storage in Regior. 1 is not required. The initial enrichment vs. burnup requirements of Fi ure 5.6.1 shall be met prior to storage of fuel assemblies in Re ion 2.
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 foam moderation is assumed. DRAINAGE 5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 62 feet-6 inches. CAPACITY 5.6.3 The spent fuel storage pool ia designed and shall be maintained with a storage capacity limited to no more *;han 1969 fuel assemblies. 5.7 COMPONENT C'CLIC OR TRANSIENT LIMIT 5.7.1 The compoaents identified in Table 5.7-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.7-1. SOUTH TEXAS - UNITS 1 & 2 57 AMENDMENT N05. AND Y
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TABLE 5.7-1 8 COMPONENT CYCLIC OR TRANSIENT LIMITS f 5 CYCLIC OR DESIGN CYCLE TRANSIENT LIMIT 01 TRANSIENT 7 COMPONENT E Reactor Coolant System 200 heatup cycles at < 100'F/h Heatupcycle-T avg from < 200 F U and 200 cooidown cycl ~s at to > 550 T.
< 100 F/h. Cooldown cycle - T avg from
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5 200'F. 80 loss of load cycles, without > 15% of RATED TiMRMAL POWER to immediate Turbine or Reactor trip. 0% of RATED TPcRMAL POWER. 40 cycles of loss-of-of fsite loss-of-offsita A.C. electrical A.C. electrical power. ESF Electric?1 System. 80 cycles of loss of flow in one Lossoftaiyonereactor reactor coolant loop. coolant pump. 400 Reactor trip cycles. 100% to 0% of RATED THERMAL POWER. 10 auxiliary spray Spray water temperature differential 4 actuation cycles. > 621'F. r; 200 leak tests. Pressurized to > 2485 psig. 5 10 hydrostatic pressure tests. Pressurized to > 3110 psig. g Y' Secondary Coolant System 1 steam line break. Break in a > 6-inch steam line. 10 hydrostatic pressure tests. Pressurized to > 1600 psig. D. b i
FP SECTION 6.0 ADMINISTRATIVE CONTROLS E h
r FP ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The Plant Manager shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility 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 Group Vice President, Nuclear shall be reissued to all station personnel on an annual basis. 6.2 ORGANIZATION OFFSITE AND ONSITE ORGANIZATIONS 6.2.1 Onsite and offsite organizations shall be established for unit respectively. Th1 onsite and offsite operation and corporate management,itions for activities affecting the safety organizations shall include the pos of the nuclear power plant,
- a. responsibility, and communication shall be Lines of authority,ined established and def for the highest management levels through intermediate levels to and including all operating organization positions. These relationships shall be documented updated, as appropriate, in the form of organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or equivalent forms of documentation. These requirements shall be documented in the FSAR.
- b. The Plant Manager shal 'a responsible for overall unit safe operation and shall have v itrol over those usite activities necessary for safe operation and maintenance of the plant.
- c. The Vice President, Nuclear Plant Operations, shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety.
- d. The individuals who train the operating staff ard those who carry out health physics and quality assurance functions may report to the appropriate onsite canager; however, they shall have sufficient organizational freedom to ensure their independence for operating pressures.
UNIT STAFF G.2.2 The unit staff shall be as follows:
- a. Each on-duty shift shall be coeposed of at least the minimum shift crew composition shown in Table 6.2-1; AMENDMENT N05. At )
SOUTH TEXA5 - UNITS 1 l. 2 6-1
- . V :
FD l I ADMINISTRATIVE CONTROLS l UNIT STAFF (Continued)
- b. At least one licensed Operator shall be in the control room when I fuel is in the reactor. In addition, while the unit is in MODE 1, 2, 3, or 4, at least one licensed Senior Operator shall be in the control room;
- c. A Health Physics Technician
- shall be on site when fuel is in the reactor;
- d. All CORE ALTERATIONS shall be observed and directly supervised by either a licensed Senior Operator 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 times. The Fire Brigade shall not include the Shift Suoervisor 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
- f. Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safety-related functions (e.g. , licensed Senior Operators, licensed Operators, health physicists, auxiliary operators, and key maintenance personnel).
- g. Senior reactor operator licenses shall be held by:
Plant Operations Manager Operations Manager Shift Supervisors Unit Supervisors Reactor operator licenses shall be held by: Reactor Operators Adequate shift coverage shall be maintained without routine heavy use of overtime. Theobjectiveshallbetohaveoperatingpersonnel work a nominal 40-hour week while the unit is operating. Hoaever, in the event that unforeseen problems require substantial amounts of overtire to be used, or during extended periods of shutdo n for re-fueling,majorraintenance,ormajorplantmodification,onatempo-rary basis the following guidelines shall be followed (excepc for shift technical advisor personnel):
- 1. An individual should not be permitted to work more than 16 hours straight, excluding shift turnover time.
'The Health Physics Technician and Fire Brigade composition may be less than the minimu requirements for a period of time not to exceed 2 hours, in order to accormodate unexpected absence, provided immediate action is taken to fill the required positiens.
6-2 AMENNENT N05. A'C SOUTH TEXAS - UNITS 1 & 2 a.
y FP ADMINTSTRATIVE CONTROLS UNITSTAFF(Continued]
- 2. An individual should not be permitted to work more than 16 hours in any 24-hour period, nor more than 24 hours in any 48-hour period, nor more than 72 hours in any 7-day period, all excluding shift turnover time.
- 3. A break of at least 8 hours should be allowed between work periods, including shift turnover time.
- 4. Except during extended shutdown periods, the use of overtime should be considered on an individual basis and not for the entire staff on a shift.
Any deviation from the above guidelines shall be authorized by the f Plant Manager or his deputy, or higher levels of management, in accordance with established procedures and with documentation of the basis for granting the deviation. Controls shall be included in the l pr;:edures such that individual overtime shall be reviewed monthly by ! the Plant Manager er his designee to assure that exce".tve hours have not been assigned. Routine deviation from the above guidelines is not ! authorized. l I r c
~,
6-3 AMENDMENT N05. AND SOUTH TEXAS - UNITS 1 & 2
. tE?
F TABLE 6.2-1 J SHIFT CREW COMPOSITION MINIM'M TWO UNITS WITH TWO SEPARATE CONTROL ROOMS WITH THE OPPOSITE UNIT IN MODE 5 OR 6 OR DEFUELED POSITION NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION MODE 1, 2, 3, or 4 HODE 5 or 6 SS l' 1" SRO 1 None R0 2 1 A0 2 2** STA 1*** None WITH THE OPPOSITE UNIT IN MODE 1, 2, 3, OR 4 POSITION NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION MODE 1, 2, 3, or 4 MODE 5 or 6 SS l' l' SRO 1 None R0 2 1 A0 2 1 STA 1* *** None SS - Shif t Supervisor with a Senior Operator license SRO - Individual with a Senior Operator license RO - Individual with an Operator license AO - Auxiliary Operator STA - Shift Technical Advisor The shift crew corposition ray be one less than the ninimur, requirements of Table 6.2-1 for a period of tire not to exceed 2 hours in order to acc r. odate unexpected absence of on-duty shift cre. merbers provided ir ediate action is taken to restore the shift cre ccG osition to within the minimu, requirerents of Table 6.2-1. This provision does not permit any shift cre. pcsition to be unmanned upon shift chang? due to an oncer.ing shift cre nn being late or absent.
- Individual may fill the saPe position on the opposite Unit.
**0ne of the two required inriividuals may fill the same position on the opposite Unit. *"The STA position shall be caved in MODES 1, 2, 3, and 4 unless the Shif t Supervisor or the indivicual .ith a Senior Operator license reets the qualifications fcr thc STA as required by the NR".
SOUTH TEAAS - UNITS 1 & 2 64 AMEN;MEhi h?5. AN: h,,
FD _ ADMIN 7STRATIVE CONTROLS During any absence of the Shift Supervisor from the control room while the unit is in MODE 1, 2, 3, or 4, an individual (other than the Shif t Technical Advisor) with a valid Senior Operator license shall be designated to assume the control room com.and 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 licerse shall be designated to assume the control room conmand function. l te'.Eh0 MENT tCS- A'O SOUTH TE u5 - U',IT5 1 L 65 l i (
FP ADMIN 1STRATIVE CONTROLS 6.2.3 INDEPENDENT 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 ofunitdesignandoperatingexperIenceinformation,includingunitsofsimilar design, which may indicate areas for improving unit safety. The ISEG shall make detailed recommendations for revised procedures, equipment modifications, maintenance activities, operations activities, or other means of improving unit safety to the Manager, Nuclear Safety Review Board. 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 year of which experience shall be in the nuclear fleid. RESPONSIBILITIES 6.2.3.3 The ISEG shall be responsible for maintaining surveillance of unit activities to provide indepenaent verification
- that these activities ar.
performed correctly and that human errors are reduced as much as practical. RECORDS 6.2.3.4 Records of activities performed by the ISEG shall be prepared, main-tained, and forwarded each calendar month to the Manager, Nuclear Safety Review Board. 6.2.4 SHIFT TECHNICAL ADVISOR 6.2.4.1 The Shif t Technical Advisor shall provide advisory technical support to the Shift Supervisor in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit. The Shift Technical Advisor shall have a bachelor's degree or equivalent in a 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 (Not Used) I "Not responsible for sign-off function. 1 I AMENDMENT NDS. AND SOUTH TEXAS UNITS 1 & 2 6-6
.ru
Fi ADMINISTRATIVE CONTROLS 6.4 TRAINING 6.4.1 A retraining and replacement training program for the unit staff shall be maintained under the direction of the Training Manager and shall meet or I' exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and Appendix A of 10 CFR Part 55 and the supplemental requirements specified in Sections A and C of Enclosure 1 of the March 28, 1980 NRC letter to all licensees, and shall include fam:liarization with relevant industry operational expeilence. l 6.5 REVIEW AND AUDIT l 6.5.1 PLANT OPERATIONS REVIEW COMMITTEE (PORC) FUNCTION 6.5.1.1 The PORC shall function to advise the Plant Manager on all matters related to nuclear safety. COMPOSITION
- 6. 5.1. 2 The PORC shall be composed of the:
Member: Plant Superintendent Member: Technical Services Manager Member: Plant Operations Manager Member: Plant Eng'neering Manager Member: Maintenance Manager Member: Quality Engineering Manager The PORC Chairman shall be appointed in writing from aniong these rembers by the Plant Manager, except for the Quality Engineering Manager. If the Technical Services Manager does not reet the qualifications of a Radiation ProtectionManagerasdefinedinRegulatoryGuide1.8(PersonnelSelectionand Training-Revision 1-R), then the P0n; composition will include the Health Physics Manager. ALTERNATE 5 6.5.1.3 All alternate rerbers shall be appointed in writing by the Plant
- Manager to serve on a terporary basis; he ever, no more than two alternates I shall participate as voting retters in PORC activities at any one tire.
l MEETIN3 FREQUENCi 1 i 6.5.1.4 The PORC shall reet at least once per calendar month and as convened I by the PORC Chairtan or his designated alternate. QUORUM 6.5.1,5 The quorum of tne PORC necessary for the performance of the PORC responsibility and authority provisions of these Technical Specifications shall corsist of tre Chairrar or his cesignated alternate ard three other re-ters including alternates. SOUTH TExA5 - UNITS 1 & 2 E-7 AMEN:M N N:i ANC g.y! 1, V ?
FL ADMINISTRATIVE CONTROLS RESPONSIBILITIES 6.5.1.6 The PORC shall be responsible for:
- a. Review of all safety-related station _ administrative procedures and changes thereto. ,
- b. Review of safety evaluations for (1) procedures, (2) changes to l procedures, structures, components, or systems, and (3) tests or L experiments completed under the provision of 10 CFR 50.59 to verify I I
that such actions did not constitute an unreviewed safety question.
- c. Review of proposed (1) procedures, (2) changes to pr6cedures, struc-tures, components, or systems, and (3) tests or experiments which may ;
involve an unreviewed safety question as defined in 10 CFR 50.59.
- d. Review of all programs required by Specification 6.8 and changes thereto. j
- e. Review of proposed changes to the Technical Specifications or the Operating License,
- f. Review of all REPORTABLE EVENTS.
- g. Review of reports of significant operating abnormalities or devia-tions from normal and expected performance of plant equipment or systems that affect nuclear safety, "
- h. Review of reports of unanticipated deficiencies in the design or operation of structures, systems, or components that affect nuclear '
safety. Review of the Security Plan and implementing procedures and changes ( i. thereto. j. Review of the Emergency Plan and implementing procedures and changes thereto.
- k. Review of the PROCESS CONTROL PROGRAM and implementing procedures and f changes thereto.
I
- 1. Review of the OFFSITE DOSE CALCULATION MANUAL and implementing proce-dures and changes thereto,
- m. Performance of special reviews, investigations, or analyses and reports thereon as requested by the Plant Manager or the Nuclear Safety Review Board (NSRB).
^
- n. Review of any accidental, unplanned, or uncontrolled radioactive re- [
- lease including the preparation of reports covering evaluation, rec- >
om endations, and disposition of the corrective action to prevent ; recurrence and the forwarding of these reports to the Plant Man 1ger [ i and to the NSRE.
- o. Reports of violations of codes, regulations, orders, Technical Speci- l fications, or Operating License requirements having nuclear safety :
L significance or reports of abnormal degradation of systems designed to contain radioactive material.
- p. Review of the Fire Protection Program, quality-related implementing procedures and changes thereto.
6-8 AMEhDMENT N05. AND SOUTH TEXA5 - UNITS 1 & 2
?f&
L
- - , - - - ~ _ _ _ , , , , _ _ - . _ - _ _ . , _ _ , _
FT ADMINISTRATIVE CONTROLS RESPONSIBILITIES (Continued) 6.5.1.7 The PORC shall: ,
- a. Recommend in writing to the Plant Manager approval or disapproval of items considered under Specification 6.5.1.6a. through e, prior to their implementation, and items considered under Specifica-tion 6.5.1.61. through 1.
- b. Render determinations in writing with regard to Ehether or not each item considered under Specification 6.5.1.6a. through e. and o. con-stitutes an unreviewed safety question; and .
- c. Provide written notification within 24 hours to the Group Vice President-Nuclear and the Nuclear Safety Review Board of disagreement between the PORC and the Plant Manager; however, the Plant Manager shall have responsibility for resolution of such disagreements pur-suant to Specification 6.1.1.
RECORDS 6.5.1.8 The PORC shall maintain written minutes of each PORC meeting that,
- at a minimum, document the results of all PORC activities performed under the responsibility provisions of these Technical Specifications. Copies shall be
! provided to the Group Vice President-Nuclear and the Nuclear Safety Review i Board. 5.E.2 NUCLEAR SAFETY REVIEW BOARD (NSRB) FUNCTION 6.5.2.1 The NSRB shall function to provide independent review and audit of designated activities in the areas of:
. a. Nuclear power plant operations,
- b. Nuclear engineering, i c. Chedstry and radiochemistry,
- d. Metallurgy,
- e. Instrumentation and control,
- f. Radiological safety, g, Mechanical and electrical engineering,
- h. Civil engineering,
- i. Training,
- j. Nuclear ass"rance,
- k. Nuclear licen'ing,
- 1. Plant security, and
- m. Environmental iepact.
The NSRB shall report to and advise the Group Vice President-Nuclear on those areas of responsibility specified in Specifications 6.5.2.7 and 6.5.2.8. 69 A"ENDMENT NOS. AND SOUTH TEXAS - O'41TS 1 & 2
'~1 1;;-
- FD
' ADMINISTRATIVE CONTROLS COMPOSITION 6.5.2.2 The N5RB shall be composed of the following, and other members shall' be appointed in writing by the Group Vice President, Nuclear Chairman General Manager, NSRP Member: General Manager, South Texas Project Management Member: Vice President, Nuclear Plant Operations Member: General Manager, Nuclear Assurance .
Member: General Manager, South Texas Project Operations Support ALTERNATES 6.5.2.3 All alternate members shall be appointed in writing by the Group Vice President-Nuclear to serve on a temporary basis; however, no more than two alternates rhall participate as voting members in NSRB activities at any 4 one time.
- CONSULTANTS 6.5.2.4 Consultants shall be utilized as determined by the N5RB Chairman to providt 'xpert advice to the NSRB.
MEETING FREQUENCY 6.5.2.5 The NSRB 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. l . QUORUM 6.5.2.6 The quorum of the NSRB necessary for the performance of the NSRB review and audit functions of these Technical Specifications shall consist of the Chairman or his designated alternate and at least a majority of NSRB members including alternates. No more than a minority of the quorum shall have line responsibi,'ity for operation of the unit. l l REVIEW l 6.5.2.7 The NSRB shall t,e responsible for the review of:
- a. The safety evaluations for: (1) changes to procedures, equipment, l
or syste.a; and (2) tests or experiments completed under the provision of 10 CFE'50.59, to verify that such actions did not constitute an unreviewed safety qucstion;
- b. Proposed change 1 to procedures, equipment, or systems which involve an unreviewed safety question as defined in 10 CFR 50.59; j
i 6-10 AMEOMENT N05. A'O SOUTri 1EXA5 - UNITS 1 & 2 n, - - --,,.- , . - ,---wr-g-y -,_g----.n. , , , , , - - - - _ , , , _ , - , . . , - - - - - . , _ ,,y,- ,.-, ,,,e----,----,
FC ADMINISTRATIVE CONTROLS REVIEW (Continued)
'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 aspect of design or operation of structures, systems, or components that could affect nuclear safety; and .
- i. Reports and meeting minutes of the PORC.
AUDITS 6.5.2.8 Audits of unit activities shall be perforraed under the cognizance of the NSRB. 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, syst2ms, or method of operation that affect nuclear safety, at least once per 6 months;
- d. The performance of activities required by the Operational Quality Assurance Program to meet the criteria of Appendix B, 10 CFR Part 50, at least once per 24 months;
- e. The fire protection prograuatic controls including the impleeenting procedures at least once per 24 months by qualified licensee QA personnel;
- f. The fire protec'. ion equipment and program impleme .tation at least once per 12 meiths utilizing either a qualified offsite licensee fire protect'on engineer or an outside independent fire protection
- consultant. An outside independent fire protection consultant shall be used at least every third year; SOUTH TEXAS - UNITS 1 & 2 6 11 Av.EtCMENT N05. A'O I'717;;g
f' MMINISTRATIVE CONTROLS AUDITS (Continued)
- 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 and packaging of radioactive wastes at least once per 24 months; [
- j. The performance of activities required by the Quality Assurance '
Program for effluent and environmental monitoring at least once per > 12 months; and
- k. Any other area of unit operatinn considered appropriate by the NSRB or the Group Vice President-Nuclear.
I R2CORDE
- 6. 5. 2. 9 Records of NSRB activities shall be prepared, approved, and dis-tributed as indicated below; ,
i
- a. Minutes of each NSRB meeting shall be prepared, approved, and forwarded to the Group Vice President-Nuclear within 14 days following each reeting;
- b. Reports of reviews encompassed by Specification 6.5.2.7 shall be i
prepared, approved, and forwarded to the Group Vice President-Nuclear within 14 days following completion of the review; and
- c. Audit reports encompassed by Specification 6.5.2.8 shall be forwarded i
to the Group Vice President-Nuclear and to the management positions l responsible for the veas audited within 30 days after completion of i the audit by the auGiting organization. J G. 5. 3 TECHNICAL REVIEW AND CONTROL 4 ACTIVITIES i
- 6.5.3.1 Activities that affect nuc1 tar safety shall be conducted as follows
4 l a. Procedures requirec by Sp.cification 6.8, and other procedures that
- affect nuclehr safety, anJ changes thereto, shall be prepared, re-l viewed, and approved. F sch such procedure, or change thereto, shall be revie ed by an individual / group other than the individual / group i who prepared the procedure, or change thereto, but who may be from the same organization as the individual / group who prepared the pro-cedure, or change thereto. Procedures other than station adminis-l trative procedures shall be approved as delineated in writing by the
! Plant Manager, Plant Superintendent, or the head of the responsible
- department prior to implementation. The Plant Manager shall approve
! station administrative procedures, security plan implementing proce-l dures, and emergency plan implementing procedures. Temporary changes , to procedares, which clearly do not change the intent of the approved I procedures, shall be approved prior to implementation by two metbers i l SOUTH TEXA5 - UNiiS 1 & 2 6-12 AMENDMENT NDS. AND i, f, . ; }h ' l i i
FD ADMINISTRATIVE CONTROLS ACTIVITIES (Continued) f of the plant staff, at least one of whom holds a Senior Reactor Operator's License. Changes to procedures that may involve a change ; to the intent Of the original procedure shall be approved by the individual authorized to approve the procedure prior to implementa-tion of the change. '
- b. Proposed changes or modifications to safety-related structures, Plant '
systems, and components shall be reviewed as designated b Manager. ! group other than the individual / group who designed the modification, but who may be from the same organization as the individual / group who designed the modification. Proposed modifications to safety-related structures, systems, and components sh.11 be approved by the ; Plant Manager prior to implementation.
- c. Proposed tests and experiments that affect nuclear safet/ and that j are not addressed in the Final Safety Ana ysis Report shall be pre-Each such pared, reviewed, and approved prior to implementation.
test or experiment shall be reviewed by an individual / group other l than the individual / group who prepared the test or experiment but who may be from the sete organizdtion as the inuividual/ group who prepared the test or experiment. Proposed tests and experiments ; shall be approved by the Plant Manager.
- d. Individuals responsible for reviews p rformed in accordance with Specification 6.5.3.1 (a) through (c) shall be members of the Each plant i
management staff previously designated by the Plant Mana If deemed necessary, such i cross-disciplinary review is necessary. : review shall be performed by qualified personnel of the appropriate di::ipline. I ) e. Each review will include a determination of whether or not an un-reviewed safety question is involved. Pursuant to 10 CFR 50.59, ! ' NRC approval of items involving an unreviewed safety question will , be obtained prior to Plant Manager approval for implen ntation. 6.5.3.2 Records of the above activities shall be provided to the Plant Manager, i PORC, and/or the N5RB as necessary for required reviews. t 6.6 REPORTABLE EVENT ACTION 6.6.1 The following actions shall be taken for REPORTABLE EVENTS: t I l ' a. The Con.ission shall be notified and a report submitted pursuant to ; ' the requirements of Section 50.73 to 10 CFR Part 50, and l
- b. Each REPORTABLE EVENT shall be reviewed by the PORC, and the rtsults of this review shall be submitted to the NRSB and the [
l Group Vice President-Nuclear. i i i t
! 6.7 SAFETY LIMIT VIOLATION i
6.7.1 The f ollowing actions shall be taken in the event a Safety Limit is : l viol ted: f AMENDMENT N05 A'C 6-13
$0VTH TEXA5 - UNITS 1 & 2
' ' ' 1 7 S3; j
FP ADMfNTSTRATIVE CONTROLS SAFETY LIMIT VIOLATION (Continued)
- a. The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within 1 hour. The Group Vice President-Nuclear and the NSRB shall be nstified within 24 hours;
- b. A Safety Limit Violetion Report shall be prepared. The report shall be reviewed by the PORC. This report shall describe: (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems, or structures, and (3) corrective action taken to prevent recurrence;
- c. The Safety Limit Violation Report shall be submitted to the Commission, the NSRB, and the Group Vice President-Nuclear within 14 days of the violation; and
- d. Operation of the unit shall not be resumed until authorized by the Commission.
6.8 PROCEDURES AND PROGRAMS 6.8.1 Written procedures shall be established, implemented, and maintained covering the &ctivities referenced below:
- a. The applicable procedures recom. ended ir Appe . dix A of Regulatory Guide 1.33, Revision 2, February 197fo,
- b. The emergency operating procedures required to implement the require-ments of NUREG-0737 and Supplement 1 to NUREG-0737 as stated in Generic Letter No. 82-33;
- c. Security Plan implementation;
- d. Emergency Plan irple entation;
- e. PROCESS CONTROL PkOGRAM implementation;
- f. OFFSITE DOSE CALCULATION MANUAL implementation;
- g. Quality Assurance Progran, for effluent and environmental monitoring; and
- h. Fire Protection Program implementation.
6.8.2 Each procedure of Specification 6.8.1,' and char.ges theret:, shall be reviewed and approved prior to implementation and reviewed periodicali.s :is set forth in Specification 6.5.3 and administrative procedures. 6.6.3 The following prograts shall be established, irplemented, a90 maintained:
- a. Prirary Coolant Sources Outside Contain ent A progra- to reduce leakage frem those portions of systems outsice containment that could contain highly radioactive fluids during a serious transient or acciuent to as low as prutical levels. The systems include the containment spray, Safety Injection, containment hydrogen ronitoring, post-accident sampling and primary saepling.
The program shall include the following: SOUTH TEXAS - UNITS 1 & 2 (-14 A"ENP'ENT NOS. A'C
..i
, tw --
FD ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)
- 1) Preventive maintenance and periodic visual inspection requirements,- l
.and
- 2) Integrated leak test requirements for each system at refueling l cycle intervals or less. ,
- b. In-Plant Radiation Monitoring A program which will ensure the capability to accurately determine the cirborne iodine concentration in vital areas under accident ,
conditions. This program shall include the following: l
- 1) Training of personnel, i
- 2) Procedures for monitoring, and ;
- 3) Provisions for maintenance of sampling and analysis equipment, j
- c. Secondary Water Chemistry A program for monitoring t,r secondary water chemistry to inhibit steam generator tube degradation. This program shall include:
- 1) Identification of a sampling schedule for the critical variables and control points for these variables, l
- 2) Identification of the procedures used to measure the values of 7 the critical variables,
- 3) Identification of process sampling points, which shall include i monitoring the discharge of the condensate pumps for evidence i of condenser in-leakage, J
- 4) Procedures for the recording and management of data, t
- 5) Procedures defining corrective actions for all off control poie :
chemistry conditions, and l
- 6) A procedure identifying: (a) the authority responsible for the !
interpretation of the data, and (b) the sequence and timing of l administrative events required to initiate corrective action. !
- d. Post-Accident 53?plino A program which will ensure the capability to obtain and analyze ,
reactor coolant, radioactive iodines and particulates in plant l gaseous effluents, and containcent atmosphere samples under accident ! conditions. The program shall include the following: ! i
- 1) Training of personnel, (
- L j 2) Procedures for saepling and analysis, and ;
- 3) Provisions for raintenance of sampling aid analysis equiprent. l A'O i i SOUTH TEXAS - UNITS I & 2 6-15 AMENDMiNT N05.
l' #'7' l i ;; ; f i i
ii: f f, ADM2NISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)
- e. Accident Monitoring Instrumentation A program which wil' ensure the capability to monitor plant variables and systems operating status during and following an accident. This program shall include those instruments provided to indicate system operating status and futnish information regarding the release of radioactive materials (Category 2 and 3 instrumentation as defined in Regulatory Guide 1.97, Revision 2) and provide the following:
- 1) Preventive maintenance and periodic surveillance of instrumen-tation,
- 2) Pre planned operating procedures and backup instrumentation to be used if one or more monitoring instruments become inoperable, and
, 3) Administrative procedures for returning inoperable instruments
- to OPERAELE status as soon as pra:ticabic.
6.9 PEPORTING REOUIREMENTS 2 ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code i of Federal Pegulations, the following reports shall be submitted to the Regional l Administrator of the Regional Office of the NRC unless otherwise noted. STARTUp REPORT 4 6.9.1.1 A sumary report of plant startup and power escalation test ng shall be submitted following: (1) receipt of an Operating 1.icense, (2) amer. tent to t.e license involving a planned increase in power level, (3) installation of
; fuel that has a dif ferent design or has been manufactured by a dif ferer t fuel supplier, and (4) modifications that may hav- significantly altered the nuclear, thermal, or hydraulic oerformance of the unit.
The Startup Report shall address each of the tests identified in the Final l Safety Analysis Report and shall include a description of the reasured values i of the operating conditions or characteristics obtained during the test progra*. ! and a comparison et these values wi.h design predictions and specifications, n Any corrective actions tnat were required to obtain satisfactory operation shall ! also be describec. An, adcitional specific details required in license condi-tions basca on other co-*itrents shall be included in this report. l l Startup lieports shall be sub-itted within: (1) 90 days following coepletion of the Startup Test Program, (2) 90 days following resumption or comencement of comercial power operation, or (3) 9 months following initial criticality, whichever is earliest. If the Startup Report does not cover all three events (i.e., initial criticality, completion of Startup Tr;t Program, and resuvption or conencetent of co- ercial operation), suppleent ry reports shall be sub-mitted at least every 3 ronths until all three events have been coPpleted. SOTH TEXAS - UNITS 3 !, 2 6-16 AMEO MENT N05. AO n-
Ft , ADMfNISTRATfVE CONTROLS , ANNUAL REPORTS
- 6.9.1.2 Annual Reports covering the activities of tae uait as described below for the previous calendar year shall be submitted prN to March 1 of each year. The initial report shall be submitted prior to March 1 of the year following initial criticality.
Reports required on an ann c. basis shall include:
- a. A tabulation on an annual basis of the number of station, utility, and other personnel (including contractors) re<11ving exposares greater than 100 mren/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, and refueling).
. The dose assign. tents to various duty functions may be estimated based on pocket dosimeter, thermolumir.escent dosimeter (TLD), or film badge measurements. Small exposures totalling less than 20% of the individual total dose need not be accounted for. In the aggregate, at least 80% of the total whole-Dody dose received from external sources should be assigned to specific major work furetiont; and
- b. The results of specific activity analyses in which the primary coolant exceeded the limits of Specification 3.4.8. The following information shall be included: (1) Redctor power history starting 48 hours prior to the first sarple in which the limit was erceeded (in graphic and tabular format); (2) Results of the last isotopic analysis for radioiodine performed prior to exceeding the limit, <
results of analysis while limit was exceeded and results of one analysis after the radiciodine activity was reduced to less than limit. Each result should include date and tire of s.mpling and the radioiodine concentrations; (3) Clean-up flow history starting 48 hours prior to the first sample in which the limit was exceeded; (4) Graph of the 1-131 concentration (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 tire duration when the specific activity of the primary coolant exceeded the radiciodine limit. ANNUt.L PAD 10 LOGICAL ENVIR0w!NTAL OPERATIN3 REPORT
- 6.9.1.3 Routine Annual Radiological Environmental Operating Reports cov, ring 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 follo ing initial criticality. T e Annual Radtological Environ ental Operating Reports shall include lunar ., interpretations, and an analysis of trends of the results of the radiological environeental surveillance activities for the report period,
*A single submittai may be made for a multiple unit station. The submittal should co?bine those sections that are common to all units at the station.
**This tabulation st;ple ents the requirements of 520.407 ef 10 CFR Part 20.
50llTH TEXA5 - UNIT 5 1 & 2 6-17 AMEN?ENT N25. AC h 2.' : .
M i ADMINTSTRATIVE CONTROLS : ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT (Continued) l including a comparison with preoperational studies, with operational controls, I as appropriate, and with previous environmental surveillance reports, and an ! i 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. l i l The Annual Radiological Environmental Operating Reports shall include the i results of analysis of all radiological environmental samples and of all i environmental radiation measurements taken during the period pursuant to the locations specified in the table ar.d figures in the Offsite Dose Calculation { Manual, as well as summarized and tabulated results of these analyses and- ! i measurements in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979. In the event that some indivi- ! 3 dual results are not available for inclusion with the report, the report shall i a be submitted noting and explaining the reasons for the missin results. The [ > missing data shall be submitted as soon as possible in a supp ementary report. j l 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
- l the Interlaboratoiy Comparison Program and the corrective action taken if the ,
i specified program is not being performed as required by Specification 3.12.3; I reason for not conducting the Radiological Encironmental Monitoring Program as ! } required by specification 3.12.1, and discussion of all deviations from the ; j saepling schedule; discussion of environmental sample measurements that exceed a the reporting levels but are not the result of plant effluents, pursuant to , ACTION b. of Specification 3.12.1; and discussion of all analyses in which the i j LLD required was not achievable, f SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT ** f t
- 6.9.1.4 Routine Semiannual Radicar'ive Effluent Release Reports covering the i i
j operation of the unit during the previous 6 months of operation shall be I ) submitted within 60 days af ter January 1 and July 1 of each year. The period { of the first report shall begin with the date of initial criticality. [ 1 i The Semiannual Radioactive Effluent Release Reports shall include a sumary t I of the quantities of radioactive liquid and gaseous effluents and solid waste l
; released frox the unit as outlined in Regulatory Guide 1.21, "Measuring, Evalua-1 ting, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive l
Materials in Liquia and Gaseous Effluents from Light-Water-Cooled Nuclear Po.er Plants," Revision 1, June 1974, with data summarized on a quarterly basis folloc ing the format of Appench B thereof. For solid wastes, the format for Table 3 i ) l l
! *This tabulation supplements the requirements of $20.407 of 10 CFR Part 20. f
. **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; I l' however, for units with separate radwaste systems, the subrittal shall specify the releases of radioactive raterial froe each unit. 6-IS AMECMENT N05. AC SOUTH TEXA5 - UNITS 1 & 7
'1: 1;;;
I f I
FJ
,ADMINfSTRATfVE CONTROLS SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (Continued) in Appendix B shall be supplemented with three additional categories: class of solid wastes (as defined by 10 CFR Part 61), type of container (e.g., LSA, Type A, Type B Large Quant ity) 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 annuel 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 nf wind speed, wind direction, and atmospheric stability.* This same report shall include an assessment of ' the radiation doses due to the radioactive liquid and gaseous effluents released from the unit or station during the previous calendar year. This same report shall also include an assessment of the radiation doses from radioactive liquid and gaaeous effluents to MEMBERS OF THE PUBLIC due to their activities inside the SITE BOUNDAR) (Figures 5.1-3 and 5.1-4) during the report period. All assueptions used in making these assessments, i.e., specific activity, exposure time, and location, shall be included in these reports. The reteorological conditions concurrent with the tire of release of radioactive materials in gaseous effluents, as determined by sampling frequency and reasurement, shall be used for determining the gaseous pathway doses. The assessment of radiation doses shall be perforced in accordance with the rethodology and parameters in the OFFSITE DOSE CALCULATION MMUAL (ODCM). The Semiannual Radica:tive Effluent Release Report to be submitted within 60 days after January 1 of each year shall also include an assessment of radiation doses to the likely most exposed MEMBER OF THE PUBLIC from reactor releases and other nearby uranium fuel cycle sources, including doses from primary effluent pathways and direct radiation for the previous calendar year toshcwconforman:ewith40CFRPart190,"EnvIronmentalRadiationProtection Standards for Nuclear Power Operation." Acceptable rethods fer calculating the dose contribution from licuid and gaseous effluents are given in Regulatcry Guide 1.109, Rev. 1, October 1977. The Semiannual Radies:tive E,"fluent Release Reports shall include a list and description of unplanned releases from the site to UNRESTRICTED AREA 5 of radica:tive materials in gaseous and liquid effluents made during the reporting period. - The Semiannual Radioactive Effluent Release Reports shall include any chances made during the re;orting period to the PROCESS CONTROL PROGRAM and the 60;M, pursuant to Specifications 6.13 and 6.14, respectively, as well as any enjo? change to Liquid, Gaseous, or Solid Radwaste Treatrent Systers p'. suant to Specification 6.15. It shall also include a listing of new loca-tions for dose calculatinns and/or ensironmental tonitoring identified by the Land Use Census pursuant to Specification 3.12.2.
'In lieu of submission with the Semiannual Radioactive Effluent Release Report, the licensee has the option of retaininc this sum a y of recuired rete:rc-Icgical cata on site in . file that shall be provide 0 to the NR: upon request.
6-19 A"EN NENT N?5. A'O SCUTH TE AA5 - LNT51 & 2
FC ADMINISTRATIVE CONTROLS SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT (Continued) The Semiannual Rsdioactive 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 than 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 limitsforRATEDTHERMALPOWER(FRTP)shallbeestablished x for at least each reload core and shall be maintain d available in the Control Rootn. The limits shall be established and implemented on a time scale consist-ent with normal procedutal changes. The analytical methods used to generate the xF , limits shall be those previously approved by the NRC.* If changes to these met ods 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 changes are determined to involve an unreviewed safety question or if such a change would require amendment of previously submitted documentation. A report containing the F,y limits for all core planes containing BankT "0" con-trol rods and all unrodded core planes and the plot of predicted (Fq ,ppe]) y, Axial Core Heicht with the limit envelope shall be provided to the NRC Document Control desk with ecoies to the Regional Administrator and the Resident Inspec-ter within 30 days of their implementation. SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator of the Regional Office of the NRC within the time period specified for each report.
*WCAP-83S5, "Po er Distribution Control and Load Follow Procedures"; WCAP-9272.A, "n'estinghouse Reload Saf ety Evaluation Methodology."
6-20 AMENDMENT N05. AND SOUTH TEXA5 - UNITS 1 & 2 U ; ! 7 1;.::
F1 ADMINISIMI1VE CONTROLS 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 a d legs of unit operation covering time interval at each power level;
- b. Records and legs of principal maintenance activities, inspections, repair, and replacement of principal items of equipment related to nuclear safety;
- c. All REPORTABLE EVENTS;
- d. Records of surveillance activities inspections, and calibrations requiredbytheseTechnicalSpecifIcations;
- 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.
6<10.3 The following records shall be retained for the duration of the unit Operating License:
- a. Records and drawing cha.tges 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 radiatien exposure for all individuals entering radiation control areas;
- d. Records of gaseous and liquid radioactive material released to the environs;
- e. Records of transient or operational cycles for those unit components identified in Table 5. 7-1;
- f. Records of reactor tests and experirrents;
- g. Records of training and qualification for current cerbers of the unit staff;
- h. Records of inservice inspections performed pursuant to these Technical Specifications;
- i. Records of cuality assurance activities required by the Operational Q.ality Assuran:e Plan; 6-21 AMEN:"ENT N:5. AN:
SCGTH TEM 5 - UN:T51 ! 2
P ADMINISTRATIVE COSLRDLS RECORD RETENTION (Continued)
- 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 PORC and the NSRB;
- 1. Records of the service lives of all hydraulic and rechanical 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 samplirg 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 later date. This should include procedures effective at specified times and QA records showing that these procedures were folloi,ed.
6.11 RADIATION PROTECTION PR03 RAM 6.11.1 Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained, and adhered to for all operations involving personnel radiation exposure. 6.12 HIGH RADIATION AREA 6.12.1 Pursuanttoparagraph20.203(c)(5)of10CFRPart20,inlieuofthe "control device" or alare signal" required by paragraph 20.203(c),each high radiation area, as defined in 10 CFR Part 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). Individuals qualified in radiation protection procedures (e.g., Health Physics Technician) or personnel continuously escorted by such individuals may be exempt from the RwP issuance requirement during the perforrance of their assigned duties in high radia-tion areas with exposure rates equal to or less than 1000 mR/h, provided they are other.ise following plant raciation protection procedures for entry into such high radiation areas. Any indivicual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following;
- a. A radiatio, conitoring device which continuously indicates the
- radiation dose .2te in the area; or
- b. A radiation r.cnitoring device which continuously integrates the radiation cose rate in the area and alarms when a preset integrated dose is receised. Entry int, such areas with this monitoring device may be rade after the cose rate le els in the area have been established and personnel have been made knowledgeable of them; or
- c. An individual qualified in radiation protection procedures with a radiation dose rate monitoring device, who is responsible for providing positive control over the activities within the area and shall perforr periedic radiation sen eillance at the frecurcy specified by the he31th and Safety Scrsites Panager in the Re.
SOUTH TExA5 - UN:TS 1 & 2 6-22 AMEN 0 MENT N05. A'C
! ~ 19
P3
@ilNISTRATIVE CONTROLS HIGH RADIATION AREA (Continued) 6.12.2 In addition to the requirements of S>ecification 6.12.1, areas accessible l to personnel with radiation levels greater tian 1000 mR/h at 45 cm (18 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 health physics 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 radiation protection
! procedures to provide positive exposure control over the activities being performed within the area.
For individual high radiation areas accessible to personnel with radiation i 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 l that j no enclosure individual areacan bebereasonably shall barricaded,constructed conspicuously around theand p6sted, individual a flash area,ing light shall be activated as a warning device. 6.13 PROCESS CONTROL PROGRAM (PCP) 6.13.1 The PCP shall be approved by the Commission prior to implementation. 6.13.2 Licensee-initiated changes to the PCP:
- 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:
l; 1) Sufficiently detailed information to totally support the rationale for the chcnge without benefit of additional or ! supplemental information; i 2) A determination that the change did not reduce the overall i conformance of the solidified waste product to existing critoria for solid wastes; and R
- 3) Docu entation of the fact that the change has been revie.ed and
- found acceptable by the PORC.
- b. Shall become effective upon review and acceptance by the POR:.
6.14 0FFSITE DOSE CALCULATION MANUAL (ODCM) 1 6.14.1 The ODCM shall be approved by the Commission prior to implementation. j ! 6.14.2 Licensee-initiated changes to the ODCM: l l a. Changes to Part A shall be subtitted to and apptrved by the NRC staft ' prior to irplementation, l h. Changes to Part B shall be submitted to the Commission in the Semiannual Radioactive Effluent Release Report for the period in l which the change (s) was made effective. This su5mittal shall j contain: SOUTH TEKAS - UNITS 1 & 2 6-23 AMEh: MENT N;5. AV
FJ . ADMINISTRATIVE CONTROLS l OFFSITE DOSE CALCULATION MANUAL (ODCM) (Continued)
- 1) Sufficiently detailed information to totally support the ,
rationale for the change without benefit of additional or supplemental information. Information submitted should consist ; l of a package of those pages of the 00CM to be changed with each page numbered, dated and containing the revision number, together " withappropriateanalysesorevaluationsjustifyingthe change (s);
- 2) A determination that the change will not reduce the accuracy or reliability of dose calculations or Setpoint determinations; i
and
- 3) Documentation of the fact that the change has been reviewed and ;
found acceptable by the PORC.
- c. Changes to Part B shall become effective upon review and acceptance ;
by the PORC. ; 6.15 MAJOR CHAN3ES TO LIQUID, GASEOUS, AND SOLIO RADWASTE TREATMENT SYSTEMS * , 6.15.1 Licensce-initiated major changes to the Radwaste Treatment Systems (liquid, gaseous,andsolid). . p t
- a. Shall be reported to the Commission in'the Semiannual Radioactive f Ef fluent Release Report for the period in which the evaluation was ;
reviewed by the PORC. 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; ; i
- 2) Sufficient detailed information to totally support the reason [
for the change without benefit of additional or supplemental l information;
- 3) A detailed description of the equipment, components, and processes j involved and the interfaces with other plant systems; r
- 4) An evaluation of the chan e, which shows the predicted releases '
of radioactive materials n liquid and gaseous effluents and/or quantity of solid waste that differ from those previously pre- I dicted in the License application and amenotents'thereto; l
- 5) e, which shows the expected maximum l
- An evaluation exposures of the to a MEMSER OF chanfHE PUM.lc in the UNRESTRICTED AR [
and to the general population that differ from those previously [ l estimated in the License application and amend ents thereto; 7
- 6) A corparison 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, , L
- Licensees may chcose to submit the information called for in this Specification !
! as part of the annual FSAR update. 6-24 AMENMENT M S. AC SOUTH TEXA5 - UNITS 1 & 2 t
F4 ADMINISTRATIVE CONTRQLS MAJOR CHANGES TO LIQUID, GASEOUS, AND SOLID RADWASTE TREATHENT SYSTEM 5 (Continued)
- 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 PORC.
- b. Shall become effective upon review and acceptance by the PORC.
SOUTH TE uS - UN!is 1 & 2 6-25 AMEN MENT N 5. A: I. , .
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