ML20235K552
| ML20235K552 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 07/07/1987 |
| From: | Perch R Office of Nuclear Reactor Regulation |
| To: | Goldberg J HOUSTON LIGHTING & POWER CO. |
| References | |
| NUDOCS 8707160353 | |
| Download: ML20235K552 (133) | |
Text
.___ ____ - _ _______ _
July 7,1987 l
Docket No. 50-498 Mr. J. H. Goldberg Group Vice President, Nuclear Houston Lighting and Power Company l P. O. Box 1700 '
Houston, Texas 77001
Dear Mr. Goldberg:
(
SUBJECT:
CERTIFICATION OF REVISED FINAL DRAFT TECHNICAL l SPECIFICATIONS FOR SOUTH TEXAS PROJECT, UNIT 1 Final Draft Technical Specifications (TS) for South Texas Project, Unit I were issued by the staff on May 18, 1987 for your certification. Your comments and proposed revisions to the Final Draft TS were provided to the staff in letters i dated June 5, 1987 and June 30, 1987. l The staff has reviewed your submittals and revised the Final Draft TS issued May 18, 1987 with the enclosed draft page changes. The Final Draft TS issued May 18, 1987 as revised by the draft page changes will be incorporated into the Operating License as Appendix A.
The staff requests that you certify under oath and affirmation that the Final l Draft TS issued May 18, 1987 as revised by the enclosed draft page changes are l consistent with the Final Safety Analysis Report (FSAR), Safety Evaluation 1
Report (SER) and the as-built facility.
If you have any questions concerning this matter, please contact Robert Perch l at (301) 492-9401.
Sincerely, 10l Robert L. Perch, Project Manager Project Directorate - IV Division of Reactor Projects - III, IV, V end Special Projects
Enclosure:
j As stated "
cc w/ enclosure: J See n xt page i QI TRIBUTION !
pocketFile NRC PDR Local PDR j PD4 Reading D. Crutchfield F. Schroeder i P. Noonan R. Perch P. Kadambi i OGC-Bethesda E. Jordan J. Partlow i ACRS (10) PD4 Plant File -
'[ i PD4/L PD4/PM D PNoon RPerch: p; JCalvo 7/y /87 7/ 7 /87 ,
7/}/87 8707.160353 870707 j PDR ADOCK 05000498 !
P PDR I
s %
1 l
July 7,1987 Docket No. 50-498 Mr. J. H. Goldberg Group Vice President, Nuclear Houston Lighting and Power Company P. O. Box 1700 Houston, Texas 77001
Dear Mr. Goldberg:
SUBJECT:
CERTIFICATION OF REVISED FINAL DRAFT TECHNICAL SPECIFICATIONS FOR SOUTH TEXAS PROJECT, UNIT 1 Final Draft Technical Specifications (TS) for South Texas Project, Unit 1 were issued by the staff on May 18, 1987 for your certification. Your comments and proposed revisions to the Final Draft TS were provided to the staff in letters dated June 5, 1987 and June 30, 1987.
l The staff has reviewed your submittals and revised the Final Draft TS issued May 18, 1987 with the enclosed draft page changes. The Final Draft TS issued ,
I May 18, 1987 as revised by the draft page changes will be incorporated into the Operating License as Appendix A. f The staff requests that you certify under oath and affirmation that the Final Draft TS issued May 18, 1987 as revised by the enclosed draft page changes are {
j consistent with the Final Safety Analysis Rieport (FSAR), Safety Evaluation l Report (SER) and the as-built facility. J
- If you have any questions concerning this matter, please contact Robert Perch at (301) 492-9401.
Sincerely, l
Mi Robert L. Perch, Project Manager Project Directorate - IV i
' ' Division of Reactor Projects - III, IV, V and Special Projects
Enclosure:
As stated cc w/ enclosure:
See next page DISTRIBUTION Docket File NRC PDR Local PDR PD4 Reading D. Crutchfield F. Schroeder P. Noonan R. Perch P. Kadambi OGC-Bethesda E. Jordan J. Partlow l ACRS (10) PD4 Plant File d PD4/ PD4/PM D RPerch: y JCalvo j PNoo 7/7/87 7/ 7 /87 7/}/87 l
m M
Mr. J. H. Goldberg Houston Lighting and Power Company South Texas Project cc: Resident Inspector / South Texas Brian Berwick, Esq.
Assistant Attorney General Project c/o U.S. Nuclear Regulatory, Commission Environmental Protection Division P. O. Box 910 P. O. Box 12548 Capitol Station Bay City, Texas 77414 Austin, Texas 78711 Mr. Jonathan Davis l
Mr. J. T. Westermeir Assistant City Attorney Manager, South Texas Project City of Austin Houston Lighting and Power Company P. O. Box 1088 P. O. Box 1700 Austin, Texas 78767 Ms. Pat Coy Mr. M. B. Lee Mr. J. E. Malaski Citizens Concerned About Nuclear City of Austin Power P. O. Box 1088 5106 Casa Oro Austin, Texas 78767-8814 San Antonio, Texas 78233 l
Mr. Mark R. Wisenberg Mr. A. von Rosenberg Manager, Nuclear Licensing Mr. M. T. Hardt Houston Lighting and Power Company City Public Service Board P. O. Box 1700 (
Houston, Texas 77001 )
P. O. Box 1771 San Antonio, Texas 78296 Mr. A. Zaccaria Jack R. Newman, Esq. Mr. K. G. Hess Newman & Holtzinger, P.C. Bechtel Corporation 1615 L Street, NW P. O. Box 2166 Washington, D.C. 20036 Houston, Texas 77001 Melbert Schwartz, Jr.', Esq. Mr. T. V. Shockley Baker & Botts Mr. R. L. Range One Shell Plaza Central Power and Light Company l
Houston, Texas 77002 P. O. Box 2121 Corpus Christi, Texas 78403 Mrs. Peggy Buchorn Executive Director Citizens for Equitable Utilities, Inc.
Route 1, Box 1684 Brazoria, Texas 77422 l
l
' ' ~
\ '.
Houston Lighting & Power Company South Texas Project cc:
Regional Administrator, Region IV U.S. Nuclear Regulatory Commission Office of Executive Director for Operations 611 Ryan Plaza Drive, Suite 1000 Arlington, Texas 76011 Mr. Lanny Sinkin, Counsel for Intervenor Citizens Concerned about Nuclear Power, Inc.
Christic Institute 1324 North Capitol Street Washington, D.C. 20002 Licensing Representative Houston Lighting and Power Company Suite 610 Three Metro Center Bethesda, Maryland 20814
+ ._ ,
6 .n DRAFT INDEX jut. 7 1987 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS PAGE SECTION 3/4.2 POWER DISTRIBUTION LIMITS AXIAL FLUX DIFFERENCE..................................... 3/4 2-1 3/4.2.1 FIGURE 3.2-1 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL P0WER......................................
3/4 2-4 3/4 2-5 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - 9F (Z) . . . . . . . . . . . . . . . . . . . . .
FIGURE 3.2-2 K(Z) - NORMALIZED qF (Z) AS A FUNCTION OF CORE HEIGHT.
3/4 2-6 3/4 2-9 3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR.................
3/4 2-10 3/4.2.4 QUADRANT POWER TILT RATI0................................
3/4.2.5 DNB PARAMETERS........................................... 3/4 2-11 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION. . . . . . . . . . . . . . . . . . . . . . 3/4 3-1 TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION. . . . . . . . . . . . . . . . . . .
3/4 3-2 TABLE 3.3-2 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES.... 3/4 3-9 TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE R EQU I R EM E NT S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
3/4 3-11 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM 3/4 3-16 INST RUMENT AT ION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION.......................................... 3/4 3-18 TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETP0lNTS. . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 3-29 TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES............. 3/4 3-37 TABLE 4.3-2 ENGINEERE6 SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS. . . . . . . . . . . . . . . . 3/4 3-42 3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitoring for Plant Operations. . . . . . . . . . . . . . .. 3/4 3-50 TABLE 3.3-6 RADIATION MONITORING INSTRUMENTATION FOR PLANT 0PERATIONS..................................... 3/4 3-51 TABLE 4.3-3 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS SURVEILLANCE REQUIREMENTS... ................. 3/4 3-53 Movable Incore Detectors. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 3-54 l Seismic Instrumentation.................................. 3/4 3-55 TABLE 3.3-7 SEISMIC MONITORING INSTRUMENTATION.................... 3/4 3-56 y
SOUTH TEXAS - UNIT 1
. 3 DRAFT INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS l PAGE SECTION TABLE 4.3-4 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE 3/4 3-57 REQUIREMENTS.............................................
Meteorological Instrumentation........................... 3/4 3-58 3/4 3-59 TABLE 3.3-8 METEOROLOGICAL MONITORING INSTRUMENTATION. . . . . . . . . . . . .
TABLE 4.3-5 METEOROLOGICAL MONITORING INSTRUMENTATION 3/4 3-60 SURVEILL REQUIREMENTS.............................................
Remote Shutdown System ..................................
3/4 3-61 ,
3/4 3-62 I TABLE 3.3-9 REMOTE SHUTOOWN SYSTEM ...............................
TABLE 4.3-6 REMOTE SHUTDOWN MONITORING INSTRUMENTATION 3/4 3-66 SURVEILLANCE REQUIREMENTS................................
Accident Monitoring Instrumentation. . . . . . . . . . . . . . . . . . . . . . 3/4 3-67 3/4 3-68 TABLE 3.3-10 ACCIDENT MONITORING INSTRUMENTATION..................
TABLE 4.3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE 3/4 3-73 REQUIREMENTS.............................................
Chemical Detection Systems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 3-75 3/4 3-77 TABLE 3.3-11 (This table number is not used. ). . . . . . . . . . ... . . . . . . . .
Radioactive Liquid Effluent Monitoring Instrumentation... 3/4 3-79 {
3/4 3-80 TABLE 3.3-12 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATIO TABLE 4.3-8 RADI0 ACTIVE LIQUID EFFLUENT MONITORING 3/4 3-82 INSTRUMENTATION SURVEILLANCE REQUIREMENTS. . . . . . . . . . . . . . . .
Radioactive Gaseous Effluent Monitoring Instrumentation.. 3/4 3-84 TABLE 3.3-13 RADI0 ACTIVE GASE0US EFFLUENT MONITORING 3/4 3-85 I NST RUMENT AT ION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
TABLE 4.3-9 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING 3/4 3-87 INSTRUMENTATION SURVEILLANCE REQUIREMENTS. . . . . . . . . . . . . . . .
3/4 3-89 3/4.3.4 TURBINE OVERSPEED PROT ECTION. . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
i SOUTH TEXAS - UNIT 1 vi u , ,.,.. . . - .. . - . _ . . . . . - . . , - . .
i 3 DRAFT JUL 7 1987 INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS PAGE SECTION 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION Startup and Power 0peration.............................. 3/4 4-1 3/4 4-2 Hot Standby..............................................
3/4 4-3 Hot Shutdown.............................................
Cold Shutdown - Loops Filled............................. 3/4 4-5 Cold Shutdown - Loops Not Filled. . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-6 3/4.4.2 SAFETY VALVES 3/4 4-7 5hutdown.................................................
3/4 4-8 0perating................................................
3/4 4-9 3/4.4.3 PRESSURIZER..............................................
3/4 4-10 3/4.4.4 RELIEF VALVES............................................
3/4 4-12 3/4.4.5 STEAM GENERATORS.........................................
TABLE 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE 3/4INSPECTED 4-17 DURING INSERVICE INSPECTION.............................
3/4 4-18 TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION. . . . . . . . . . . . . . . . . . . . . . .
3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems................................ 3/4 4-19 Operational Leakage...................................... 3/4 4-20 3/4 4-22 TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES......
3/4 4-23 3/4.4.7 CHEMISTRY................................................
3/4 4-24 TABLE 3.4-2 REACTOR 000LANT SYSTEM CHEMISTRY LIMITS. . . . . . . . . e . . . . .
TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE. 3/4 4-25 REQUIREMENTS.............................................
SPECIFIC ACTIVITY........................................
3/4 4-26 3/4.4.8 FIGURE 3.4-1 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY >l pCi/ gram 3/4 4-28 DOSE EQUIVALENT I-131....................................
l TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND 3/4ANALYSIS 4-29 PR0 GRAM.................................................. l 3/4.4.9 PRESSURE / TEMPERATURE LIMITS 3/4 4-31 Reactor Coolant System...................................
FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS - 3/4 4-32 APPLICABLE UP TO 32 EFPY.................................
SOUTH TEXAS - UNIT 1 vii
. 6 DRAFT INDEX g 7 jgg7 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREPINTS PAGE SECTION FIGURE 3.4-3 REACTOR COOLANT SYSTEM C00LOOWN LIMITATIONS - 3/4 4-33 APPLICABLE UP TO 32 EFPY.............. ..................
TABLE 4.4-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM - 3/4 4-34 WITHDRAWAL SCHEDULE......................................
Pressurizer..............................................
3/4 4-35 Overpressure Protection Systems.......................... 3/4 4-36 FIGURE 3.4-4 NOMINAL MAXIMUM ALLOWABLE PORV SE190 INT 3/4 4-37 FOR THE COLD OVERPRESSURE SYSTEM . . . . . . . . . . . . . . . . . . . . . . . .
3/4 4-39 3/4.4.10 STRUCTURAL INTEGRITY.....................................
3/4 4-40 3/4.4.11 REACTOR VESSEL HEAD VENTS. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS ............................................ 3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - T avg GREATER THAN OR EQUAL TO 350*F. .. . 3/4 5-3 3/4.5.3 ECCS SUBSYSTEMS - T LESS THAN 350*F................... 3/45-6 avg ECSS Subsystems - T avg Less Than or Equal to 200*F. . . . . . . 3/4 5-8 3/4.5.4 (This specification number is not used.)
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 3/4.6 CON _TAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Containment' Integrity.................................... 3/4 6-1 Containment Leakage...................................... 3/4 6-2 Contai nme nt Ai r Loc ks . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6-5 Internal Pressure........................................ 3/4 6-7 Air Temperature........................................... 3/4 6-8 Containment Structural Integrity......................... 3/4 6-9 Contai nment Ventil ati on Sys tem. . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6-12 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS Containment Spray System................................. 3/4 6-14 Spray Additive System.................................... 3/4 6-15 Containment Cooling System............................... 3/4 6-17 SOUTH TEXAS - UNIT 1 viii
. 6 DRAFT INDE'( JUL 7 1987 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS PAGE SECTION CONTAINMENT ISOLATION VALVES.............................
3/4 6-18 3/4.6.3 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 1 3/4.7.1 TURBINE CYCLE Safety Valves............................................ 3/4 7-1 TABLE 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH IN0PERABLE STEAM LINE SAFETY VALVES DURING 3/4 7-2 4 LOOP 0PERATION.........................................
3/4 7-3 TABLE 3.7-2 STEAM LINE SAFETY VALVES PER L00P. . . . . . . . . . . . . . . . . . . . .
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 SAMPLE3/4 7-8 AND ANALYSI S P R0 GRAM. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
Main Steam Line Isolation Va1ves. . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-9 Atmospheric Steam Relief Valves ......................... 3/4 7-10 3/4 7-11 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION. . . . . . . . . .
COMPONENT COOLING WATER SYSTEM...........................
3/4 7-12 3/4.7.3 3/4 7-13 3/4.7.4 ESSENTI AL COOLING WATER SYSTEM. . . . . . . . . . . . . . . . . . . . . . . . . . .
3/4 7-14 3/4.7.5 ULTIMATE HEAT SINK.......................................
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 7-19 3/4.7.8 FUEL HANDLING BUILDING (FHB) EXHAUST AIR SYSTEM. . . . . . . . . .
3/4 7-21 3/4.7.9 SHUBBERS.................................................
3/4 7-26 FIGURE 4.7-1 SAMPLE PLAN 2) FOR SNUBBER 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.)
3/4 7-31 3/4.7.13 AREA TEMPERATURE M0NITORING. . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
M0NITORING........................... 3/4 7-32 TABLE 3.7-3 AREA TEMPERATURE 3/4 7-33 3/4.7.14 ESSENTI AL CHILLED WATER SYSTEM . . . . . . . . . . . . . . . . . . . . . . . . . .
SOUTH TEXAS - UNIT 1 ix
. n DRAFT INDEX JUL 7 1987 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS PAGE SECTION 3/4.9.11 WATER LEVEL - STORAGE P0OLS 3/4 9-12 Sp ent F uel P ool . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
In-Contai nment Storage Pool . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 9-13 3/4 9-14 3/4.9.12 FUEL HANDLING BUILDING EXHAUST AIR SYSTEM ...............
3/4.10 SPECIAL TEST EXCEPTIONS _
3/4 10-1 3/4.10.1 SHUTDOWN MARGIN..........................................
3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS... 3/4 10-2 3/4 10-3 3/4.10.3 P HY SI C S T E ST S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
3/4 10-4 3/4.10.4 REACTOR COOLANT L00PS....................................
3/4 10-5 3/4.10.5 POSITION INDICATION SYSTEM - SHUTD0WN. . . . . . . . . . . . . . . . . . . .
3/4.11 RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS 3/4 11-1 Concentration............................................
00se.....................................................
3/4 11-2 Liquid Waste Processing System........................... 3/4 11-3 Liqui d Hol dup Tanks . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 11-4 3/4.11.2 GASEOUS EFFLUENTS Dose Rate................................................
3/4 11-5 D o s e - N ob l e G a s e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
3/4 11-6 Dose - Iodine'131, Iodine-133, Tritium, and Radioactive Material i n Particul ate Form. . . . . . . . . . . . . . . . . . . . . . . . . . . . .
3/4 11-7 Gaseous Waste Processi ng System. . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 11-8 Explosive Gas Mixture.................................... 3/4 11-9 Gas Stcrage Tanks........................................ 3/4 11-10 3/4 11-11 3/4.11.3 SOLID RADI0 ACTIVE WASTES.................................
3/4 11-13 3/4.11.4 TOTAL 00SE...............................................
SOUTH TEXAS - UNIT 1 xi
.___._.m.__ __ . . - - . . . . . . . .
DRAFT INDEX JUL 7 1987 BASES PAGE SECTION B 3/4 0-1 3/4.0 APPLICABILITY...............................................
3/4.1 REACTIVITY CONTROL SYSTEMS B 3/4 1-1 3/4.1.1 B0 RATION CONTR0L..........................................
B 3/4 1-2 3/4.1.2 B0 RATION SYSTEMS..........................................
B 3/4 1-3 3/4.1.3 MOV AB LE CO NT RO L AS S EMB LI E S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
B 3/4 2-1 3/4.2 POWER DIST RIBUTION LIMITS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
B 3/4 2-1 3/4.2.1 AXIAL FLUX DIFFERENCE.....................................
3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and B 3/4 2-2 l
HUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR.................
) FIGURE B 3/4.2-1 TYPICAL INDICATED AXIAL FLUX DIFFERENCE VERSUS B 3/4 2-3 l THERMAL P0WER............................................
B 3/4 2-5 3/4.2.4 QUADRANT POWER TILT RATI0.................................
B 3/4 2-5 3/4.2.5 DNB PARAMETERS............................................
l l
3/4.3 INSTRUMENTATION l
3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM and ENGINEERED SAFETY B 3/4 3-1 I FEATURES ACTUATION SYSTEM INSTRUMENTATION................ 1 B 3/4 3-3 3/4.3.3 MONITORING INSTRUMENTATION................................
PROTECTION.............................. B 3/4 3-6 3/4.3.4 TURBINE OVERSPEED 3/4.4 REACTOR COOLANT SYSTEM B 3/4 4-1 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION. . . . . . . . . . . . .
B 3/4 4-1 3/4.4.2 SAFETY VALVES.............................................
B 3/4 4-2 1
3/4.4.3 PRESSURIZER...............................................
B 3/4 4-2 3/4.4.4 RELIEF VALVES.............................................
B 3/4 4-2 3/4.4.5 ST E AM GENERATO RS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
B 3/4 4-3 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE............................
B 3/4 4-4 3/4.4.7 CHEMISTRY.................................................
ACTIVITY.........................................
B 3/4 4-5 3/4.4.8 SPECIFIC xiii SOUTH TEXAS - UNIT 1
. 8 DRAFT INDEX JUL 7 1987 BASES PAGE SECTION 3/4.4.9 PRESSURE / TEMPERATURE LIMITS...............................
B 3/4 4-6 l TABLE B 3/4.4-1 REACTOR VESSEL TOUGHNESS.......................... B 3/4 4-9 FIGURE B 3/4.4-1 FAST NEUTRON FLUENCE (E>1MeV) AS A FUNCTION OF FULL POWER SERVICE LIFE.................................. B 3/4 4-10 3/4.4.10 ST RU CT U RA L I NT EG R I TY. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
B 3/4 4-14 3/4.4.11 REACTOR VESSEL HEAD VENTS. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
B 3/4 4-14 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS.............................................. B 3/4 5-1 l 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS............................... B 3/4 S-1 3/4.5.4 (Not used) 3/4.5.5 REFUELING VATER 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 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS. . . . . . . . . . . . . . . . . . . . . .
B 3/4 6-3 ;
I 3/4. 6. 3 ' CONTAINMENT ISOLATION VALVES. . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
B 3/4 6-4 3/4.6.4 COMBUSTIBLE GAS CONTR0L................................... B 3/4 6-4 1
l l 3/4.7 PLANT SYSTEMS 1 3/4.7.1 TURBINE CYCL 5............................................. B 3/4 7-1 j 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION........... B 3/4 7-3 ,
3/4.7.3 COMPONENT COOLING WATER SYSTEM............................
B 3/4 7-3 i 3/4.7.4 ESSENTIAL COOLING WATER SYSTEM............................
B 3/4 7-3 3/4.7.5 U LT IMAT E HE AT S I N K . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
B 3/4 7-3 1
3/4.7.6 (Not used) 3/4.7.7 CONTROL ROOM MAKEUP AND CLEANUP FILTRATION SYSTEM. . . . . . . . .
B 3/4 7-4 3/4.7.8 FUEL HANDLING BUILDING EXHAUST AIR SYSTEM................. B 3/4 7-4 >
3/4.7.9 SNUBBERS.................................................. B 3/4 7-4 SOUTH TEXAS - UNIT 1 xiv l
~,- - . _ . _ __ . , _ _ _ . _ . . _ ,
,.. . ,. . . . . . - - -, l DRAFT INDEX JUL 7 1987 BASES PAGE SECTION B 3/4 7-6 3/4. 7.10 SEALED SOURCE CONTAMINATION. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
3/4.7.11 (Not used) 3/4.7.12 (Not used)
B 3/4 7-6 3/4. 7.13 AREA TEMPERATURE MONITORING. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
B 3/4 7-6 3/4.7.14 ESSENTI AL CHILLED WATER SYSTEM. . . . . . . . . . . . . . . . . . . . . . . . . . . .
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 B 3/4 8-1 ONSITE POWER DISTRIBUTION.................................
DEVICES................... B 3/4 8-3 3/4. 8.4 ELECTRICAL EQUIPMENT PROTECTIVE l
3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION....................................... B 3/4 9-1 3/4.9.2 INSTRUMENTATION........................................... B 3/4 9-1 3/4.9.3 DECAY TIME................................................ B 3/4 9-1 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS. . . . . . . . . . . . . . . . . . . . . . . . .
B 3/4 9-1 3/4.9.5 COMMUNICATIONS............................................ B 3/4 9-1 l B 3/4 9-2 i 3/4.9.6 REFUELING MACHINE......................................... l 3/4.9.7 CRANE TRAVEL - FUEL HANDLING BUILDING. . . . . . . . . . . . . . . . . . . . . B 3/4 9-2 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION. . . . . . . . . . . . . B 3/4 9-2 3/4.9.9 CONTAINMENT VENTILATION ISOLATION SYSTEM.................. 8 3/4 9-2 3/4.9.10 and 3/4.9.11. WATER LEVEL - REFUELING CAVITY and STORAGE P00LS............................................. B 3/4 9-3 3/4.9.12 FUEL HANDLING BUILDING EXHAUST AIR SYSTEM . . . . . . . . . . . . . . . . B 3/4 9-3 3/4.10 SPECIAL TEST EXCEPTIONS.
3/4.10.1 SHUTDOWN MARGIN . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
B 3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS.... B 3/4 10-1 3 /4 .10. 3 P HY S IC S T E ST S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
B 3/4 10-1 3/4.10. 4 REACTOR COOLANT L00 P S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
B 3/4 10-1 3/4.10. 5 POSITION INDICATION SYSTEM - SHUTD0WN. . . . . . . . . . . . . . . . . . . . .
B 3/4 10-1 SOUTH TEXAS - UNIT 1 xv
~ - - - - -- -. ~. - -.
2TL_-_:T r_:TT
DRAFT II'DEX 7 1987 JUl.
DESIGN FEATURES SECTION PAGE 5.1 SITE 5.1.1 EXCLUSION AREA................................................. 5-1 5.1.2 LOW POPULATION Z0NE............................................ 5-1 5.1. 3 MAP DEFINING UNRESTRICTED AREAS AND SITE. BOUNDARY FOR RADI0 ACTIVE GASE0US AND LIQUID EFFLUENTS....................... 5-1 5.2 CONTAINMENT 5.2.1 CONFIGURATION.................................................. 5-1 5.2.2 DESIGN PRESSURE AND TEMPERATURE................................ 5-1 FIGURE 5.1-1 EXCLUSION AREA.......................................... 5-2 FIGURE 5.1-2 LOW PO P U LAT I ON Z0N E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-3 FIGURE 5.1 3 RESTRICTED AREA AND SITE B0UNDARY FOR RADI0 ACTIVE GASEOUS EFFLUENTS....................................... 5-4 FIGURE 5.1-4 RESTRICTED AREA AND SITE B0VHDARY FOR RADI0 ACTIVE LIQUID EFFLUENTS ....................................... 5-5
- 5. 3 REACTOR CORE 5.3.1 FUEL ASSEMBLIES................................................ 5-6 5.3.2 CONTROL R0D ASSEMBLIES......................................... 5-6 5.4 REACTOR COOLANT SYSTEM 5.4.1 DESIGN PRESSURE AND TEMPERATURE................................ 5-6 5.4.2 V0LUME..........'............................................... 5-6 5.5 METEOROLOGIC AL TOWER LOCATION. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-6
- 5. 6 FUEL STORAGE 5.6.1 CRITICALITY.................................................... 5-6 5.6.2 DRAINAGE....................................................... 5-7 5.6.3 CAPACITY....................................................... 5-7 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT. . . . . . . . . . . . . . . . . . . . . . . . . . . 5-7 TABLE 5.7-1 COMPONENT CYCLIC OR TRANSIENT LIMITS. . . . . . . . . . . . . . . . . . 5-8 SOUTH TEXAS - UNIT 1 xvii
DRAFT jut. 7 t367 INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.5.2 NUCLEAR SAFETY REVIEW BOARD (NSRB)
Function................................................... 6-9 Composition................................................ 6-10 Alternates................................................. 6-10 Consultants................................................ 6-10 Meeting Frequency.......................................... 6-10 Quorum.....................................................
6-10 Review..................................................... 6-10 Audits..................................................... 6-11 Records.................................................... 6-12 6.5.3 TECHNICAL REVIEW AND CONTROL Activities ................................................ 6-12 6.6 REPORTABLE EVENT ACTI0N...................................... 6-13 6.7 SAFETY LIMIT VIOLATION....................................... 6-13 6.8 PROCEDURES AND PR0 GRAMS...................................... 6-14 6.9 REPORTING REQUIREMENTS 6.9.1 ROUTINE REP 0RTS.,........................................... 6-16 Startup Report............................................. 6-16 Annual Reports............................................. 6-17 Annual Radiological Environmental Operating Report......... 6-17 Semiannual Radioactive Ef fluent Release Report. . . . . . . . . . . . . 6-18 Monthly Operating Reports.................................. 6-20 Radial Peaking Factor Limi t Report. . . . . . . . . . . . . . . . . . . . . . . . . 6-20 6.9.2 SPECIAL REP 0RTS............................................ 6-20 l
6.10 RECORD RETENTION........................................... 6-21 SOUTH TEXAS - UNIT 1 xix
1 DRAFT l JUL 7 1987 l 680 j i UNACCEPTABLE ;
1 I
660 , sed %s 2400 PSIA _
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(O 640 N)
(o.ss2.7 '%I, 2250 PSIA ^%
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(o,ess.eTN % 2000 PSIA 'N % h m y % (o.90,625.,1) ,
S k1.20,599.3)
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(O.625.6) % % (l.OO,618.9) 186 PSIA N
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UNACCEPTABLE l 560 l
l 1 \
540 1.20 O .20 .40 .60 .80 1.00 !
I FRACTION OF. RATED THERMAL POWER i i
!l FIGURE 2.1-1 l REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION l l
SOUTH TEXAS - UNIT 1 2-2 I
DRT JUL 1 191:7 LIMITING SAFETY SYSTEM SETTINGS BASES Pressurizer Pressure 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 event 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 equivalent); and on increasing power, automatically reinstated by P-7.
The High Setpoint trip functions in conjunction with the pressurizer relief and safety valves to protect the Reactor Coolant System against system overpressure.
Pressurizer Water Level The Pressurizer High Water Level 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 increasing power, auto-matica11y reinstated by P-7.
Reactor Coolant Flow The 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 one loop drops below approximately 92% of nominal full loop flow.
Above ?-8 (a power level of approximately approximately 40% of RATED THERMAL POWER) an automatic Reactor trip will occur if the flow in any single loop drops below approximately 92% of nominal full loop flow. Conversely, on decreasing power between P-8 and the P-7, an automatic Reactor trip will occur on low reactor coolant flow in more than one loop, and below P-7 the trip function is automatically blocked.
Steam Generator Water Level The Steam Generator Water Level Low-Low trip protects the reactor from loss of heat sink in the event of a sustained steam /feedwater flow mismatch l resulting from loss of normal feedwater. The specified Setpoint provides allowances for-starting dolcys cf -ths-Auxi.11ary Feedwater System.
SOUTH TEXAS - UNIT 1 B 2-6
- _ ~_. _ . _ _ _ _
DRAFI JUL 7 1987 LIMITING SAFETY SYSTEM SETTINGS I
BASES Undervoltage and Underfrequency - Reactor Coolant Pump Buses l
The Undervoltage and Underfrequency Reactor' Coolant Pump Bus trips pro- l vide core protection against DNB as a result of complete loss of forced coolant I flow. The specified Setpoints assure a Reactor trip signal is generated before the Low Flow Trip Setpoint is reached. Time delays are incorporated in the Underfrequency and Undervoltage trips to prevent spurious Reactor trips 3 from momentary electrical power transients.- For undervoltage, the delay is j set so that the time required for a signal to reach the Reactor trip breakers j 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 ,
I 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 THERMAL POWER with a turbine impulse chamber pressure at approximately !
10% of full power equivalent); and on increasing power, reinstated automatically ;
by P-7. ,
A Turbine 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 f 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 i 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: j P-6 On increasing power, P-6 allows the manual block of the Source Range trip (i.e., prevents premature block of Source Range trip and j
deenergizes the high voltage to the detectors). On decreasing power, Source Range Level trips are automatically reactivated and high i voltage restored.
P-7 On increasing power, P-7 automatically enables Reactor trips on low flow in more than one reactor coolant loop, reactor coolant pump bus undervoltage and underfrequency, pressurizer low pressure, and i pressurizer high level. On decreasing power, the above listed trips are automatically blocked.
SOUTH TEXAS - UNIT 1 B 2-7
- ^ *N#W M + t a..mem.- .e
- ---_____E _ _ _ _ _ . _ _ _ _ _ _ __
DRAFT REACTIVITY CONTROL SYSTEMS JUl. 7 1987 R00 OROP TIME LIMITIN(LCONDITION FOR OPERATION 3.1.3.4 The individual full-length (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 stationary gripper coil voltage to dashpot entry with:
- a. T,yg greater than or equal to 561*F, and
- b. All reactor coolant pumps operating.
APPLICABILITY: MODES 1 and 2.
ACTION:
With the drop time of any full-length rod determincd to exceed the above limit, restore the rod drop time to within the above limit prior to proceeding to MODE 1 or 2.
SURVEILLANCE RE0VIREMENTS 4.1. 3. 4 The rod drop time of full-length rods shall be demonstrated through
. measurement prior to reactor criticality:
- a. For all rods following each removal of the reactor vessel head,
- b. For specifically affected individual rods following any maintenance on or modification to the Control Rod Drive System which could affect the drop time of those specific rods, and
- c. At least once per 18 months.
SOUTH TEXAS - UNIT 1 3/4 1-21
- ______:_:_ :L -- -- , -
]
l DRAFT l 3/4.2 POWER DISTRIBUTION LIMITS JUL 7 1987 )
l 3/4.2.1 AXIAL FLUX DIFFERENCE 11HITING 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. i 5% for core average accumulated burnup of less than or equal to 3000 MWD /MTU; and l
- b. + 3%, -19% for core average accumulated burnup of greater than 3000 MWD /HTU.
The indicated AFD may deviate outside the above required target band at greater than or equal to 50% but less than 90% of RATED THERMAL POWER provided the indi-cated AFD is within the Acceptable Operation Limits of Figure 3.2-1 and the cumu-lative penalty deviation time does not exceed 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> during the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> during the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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.
- b. With the indicated AFD outside of the above required target band for i more than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of cumulative penalty deviation time during the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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: l
- 1. THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes, and
- 2. The Power Range Neutron Flux * ** - High Setpoint to less than or '
equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
- See Special Test Exceptions Specification 3.10.2.
- Surveillance testing of the Pcwer 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 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> operation may be accumulated with the AFD outside of the above required target band during testing without penalty deviation.
SOUTH TEXAS - UNIT 1 3/4 2-1
- %-e gen e p w g ,w.e g- --e. a. g me .e>e- n m. . - -
- L POVER DISTRIBUTION LIMITS 7g 3/4.2.2 HEATFLUXHOTCHANNELFACTOR-Fg LIMITIRGl0NDITION FOR_QPERATION 3.2.2 Fq (Z) shall be limited by the following relationships:
F0 (Z) 5 2.50 [K(Z)] for P > 0.5 P
F9 (Z) $ 5.0 [K(Z)] for P $ 0.5
, and Where: P _ RATED THERMAL POWER THERMAL POWER K(Z) = the function obtained from Figure 3.2-2 for a given core height location.
APPLICABILITY: MODE 1.
ACTION:
With Fq (Z) exceeding its limit:
- a. Reduce THERMAL POWER at least 1% for each 1% Fq (Z) exceeds the limit within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoint within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; subsequent POWER OPERATION may proceed provided the Overpower AT Trip Setpoint has been reduced at least 1%
for each 1% F 9 (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.F (Z) is demonstrated through incore mapping to be 9
within its limit.
o SOUTH TEXAS - UNIT 1 3/4 2-5
DRAFT ggt 7 tgg7 PCVER DISTRIBUTION LIMITS 3/4. 2. 4 QUADRANT POWER TILT RATIO LIMITING CONDITION FOR OPERATION 3.2.4 The QUADRANT POWER TILT RATIO shall not exceed 1.02. 1 l
APPLICABILITY,: MODE 1, above 50% of RATED THERMAL POWER *.
ACTION:
With the QUADRANT POWER TILT RATIO determined to exceed 1.02:
- a. Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> reduce THERMAL POWER at least 3% fron RATED THERMAL i POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess l of 1 and similarly reduce the Power Range Neutron Flux-High Trip l Setpoint within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and e ery 7 days thereafter, verify that F9 (Z) (by l b.
F xy evaluation) and Fg are within their limits by performing Surveil-l lance Requirements 4.2.2.2 and 4.2.3.2. THERMAL POWER and setpoint l reductions shall then be in accordance with the ACTION statements of Specifications 3.2.2 and 3.2.3. j S' SURVEILLANCE REQUIREMENTS _ __
4.2.4.1 The QUADRANT POWER TILT RATIO shall be determined to be within the limit above 50% of RATED THERMAL POWER by:
- a. Calculating the ratio at least once per 7 days when the alarm is OPERABLE, and
- b. Calculating the ratio at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during steady-state operation when the alarm is inoperable.
- 4. 2. 4. 2 The QUADRANT ' POWER TILT RATIO shall be determined to be within the limit when above 75% of RATED THERMAL POWER with one Power Range channel inoperable by using the movable incere detectors to confire indicated QUADRANT POWER TILT RATIO at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by either:
- a. Using the four pairs of symmetric thimble locations, or
- b. Using the movable incore detection system to monitor the QUADRANT POWER TILT RATIO subject to the requirements of Specification 3.3.3.2.
- See Special Test Exceptions Specification 3.10.2.
SOUTH TEXAS - UNIT 1 3/4 2-10
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TABLE 3.3-S ENGINEERED SAFETY FEATURES RESPONSE TIMES RESPONSE TIME IN SECONOS INITIATION SIGNAL AND FUNCTION
- 1. Manual Initiation Safety Injection (ECCS) N.A.
a.
Containment Spray N.A.
b.
- c. Phase "A" Isolation N.A.
- d. Phase "B" Isolation N.A.
Containment Ventilation Isolation N. A.
e.
Steam Line Isolation N.A.
f.
- 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) < 27 III/12(5)
- 1) Reactor Trip 2 I3) 12 I3)
- 2) Feedwater Isolation
- 3) Phase "A" Isolation 33II)/23(2)
- 4) Containment Ventilation Isolation k 23(1)/13(2)
- 5) Auxiliary Feedwater < 60
- 6) Essentia1 Cooling Water 62II)/52(2)
- 7) Reactor Containment Fan Coolers [38(I}/28(2)
- 8) Control Room Ventilation. .
< 72II)/62(2)
- 9) Start Standby Diesel Generators < 12 SOUTH TEXAS - UNIT 1 3/4 3-37
DRAFT TABLE 3.3-5 (Continued) JUL 7 1937 ENGINEERED SAFETY FEATURES RESPONSE TIMES RESPONSE TIME IN SECONDS INITIATING SIGNAL AND FUNCTION
- 3. Pressurizer Pressure--Low
- a. Safety Injection (ECCS) $ 27(1)/12(5)
- 1) Reactor Trip < 2(3)
- 2) Feedwater Isolation 12(3)
- 3) Phase "A" Isolation [33(1)/23(2)
Containment Ventilation Isolation N.A.
4)
Auxiliary Feedwater < 60 5)
- 6) Essential Cooling Water 62(1)/52(2)
- 7) Reactor Containment Fan Coolers 38(1)/28(2)
- 8) Control Room Ventilation h72(1)/62(2)
- 9) Start Standby Diesel Generators 1 12
- 4. Compensated TCOLD 40 U" l
- a. Safety Injection (ECCS) N.A.
Reactor Trip N.A.
1)
- 2) Feedwater Isolation N.A.
- 3) Phase "A" Isolation N.A.
- 4) Containment Ventilation Isolation N.A.
- 5) Auxiliary Feedwater N.A.
Essential Cooling Water N.A.
6)
- 7) Reactor Containment Fan Coolers N.A.
- 8) Control, Room Ventilation N.A.
- 9) Start Diesel Generators N.A.
- b. Steam Line Isolation N.A.
- 5. Compensated Steam Line Pressure--Low
- a. Safety Injection (ECCS) 1 22(4)/12(5)
- 1) Reactor Trip < 2(3)
- 2) Feedwater Isolation k12(3)
- 3) Phase "A" Isolation 1 33(1)/23(2)
Containment Ventilation Isolation N.A.
4)
Auxiliary Feedwater < 60 5)
- 6) Essential Cooling Water 62(1)/52(2)
- 7) Reactor Containment Fan Coolers [38(1)/28(2)
SOUTH TEXAS - UNIT 1 3/4 3-38
, y
? - *'
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- P w= > ~
DRAFT TABLE 3.3-5 (Continued) ggt 7 1937 ENGINEERED SAFETY FEATURES RESPONSE TIMES RESPONSE TIME IN SECONDS, INITIATING SIGNAL AND FUNCTION
- 5. Compensated Steam Line Pressure--Low (Continued)
Control Room Ventilation < 72(1)/62(2)
- 8) 7 12
- 9) Start Diesel Generators
- b. Steam Line Isolation k8(3)
- 6. Containment Pressure--High-3 Containment Spray < 30(1)/20(2) a.
- b. Phase "B" Isolation [28II)/18(2)
- 7. Containment Pressure--High-2 Steam Line Isolation 17(3)
- 8. Steam Line Pressure - Negative Rate--High N.A.
Steam Line Isolation
- 9. Steam Generator Water Level--High-High
- a. Turbine Trip < 3(3)
- b. Feedwater Isolation k12(3)
- 10. Steam Generator Water Level--Low-Low
- a. Motor-Driven Auxiliary Feedwater Pumps < 60
- b. Turbine-Driven Auxiliary Feedwater Pump < 60
- 11. RWST Level--Low-Low Coincident with Safety Injection Automatic Switchover to Containment Sump i32(2)
- 12. Loss of Power
SOUTH TEXAS - UNIT 1 3/4 3-39
_=_________ _ - _ - _ _ - _
- . - - - - - - ~ - -- - - -
1 d &
DRAFT TABLE 3.3-5 (Continued) 7g ENGINEERED SAFETY FEATURES RESPONSE TIMES RESPONSE TIME IN SECONDS INITIATING SIGNAL AND FUNCTION
- c. -
(Sustained Degraded Voltage)
- 13. RCB Purge Radioactivity-High
- a. Containment Ventilation Isolation (48-inch lines) 1 73(2)
- b. Containment Ventilation Isolation < 23(2) l (18-inch lines) _
- 14. Compensated Tcold~~l
- N.A. l
- a. Turbine Trip N.A.
- b. Feedwater Isolation l l
- 15. Feedwater Flow - High coincident with 2 of 4 Loops Having Either Reactor Coolant Flow - Low or T,yg - Low Turbirae Trip - Reactor Trip N.A.
.- a.
N.A.
- b. Feedwater Isolation
- 16. T,yg - Low Coincident with Reactor Trip N.A.
Feedwater Isolation
- 17. Control Room Intake Air Radioactivity - High Control Room Ventilation 178(2)
- 18. Spent Fuel Pool Ekhaust Radioactivity - High FHB HVAC Emergency Startup 5 42(2)
SOUTH TEXAS - UNIT 1 3/4 3-40
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DRAFT TABLE 3.3-7 g 7 ggg7 SEISMIC HONITORING INSTRUMENTATION MINIMUM INSTRUMENTS i MEASUREMENT RANGE OPERABLE INSTRUMENTS AND SENSOR LOCATIONS
- 1. Triaxial Time-History Accelerometers ***
13g 1
- a. Free Field ;
- b. Containment Bldg. Foundation i3g 1 (Tendon Gallery El. -36'9") l i3g 1 i
- c. Outside Face Containment Shell I (Reactor Containment Building E1. 68'0")
- d. Steam Generator Upper Lateral Support f3g 1 (Reactor Containment Building E1. 66'7\")
- e. Fuel Handling Building Foundation f3g 1 (Fuel Handling Building El. -29'0")
- f. Mechanical Electrical Auxiliary Building i3g 1 (Mechanical Electrical Auxiliary Building El. 35'0")
- 2. Triaxial Peak Accelerographs
- a. Spent Fuel Pool Heat Exchanger i3g 1 (Inlet Line Fuel Handling Building El. 64'5%")
- b. Reactor Vessel i3g 1 (Reactor Containment Building E1. 68'0")
3g 1
- c. Cold leg of RC Piping (Reactor Containment Building E1. 34'3")
- 3. Self-Contained Triaxial Accelerograph 13g 1 (At Reactor Containtnent Building Foundation Tendon Gallery E1. -36'9") ,
0.03 to 3g 1*
- 4. Triaxial Seismic Switch ** #
- 5. Triaxial Seismic Trigger ** ## 0.003 to 0.3g 1*
1 to 32 Hz 1*
- 6. Response Spectrum Analyzer **
- 7. Magnetic Tape Recorders ** 0.1 to 33 Hz 6 N.A. 1
- 8. Playback System **
- With reactor control room indication and alarm
- At seismic monitoring panel in Control Room, Unit 1
- Accelerometer data is gathered and analyzed by the Response Spectrum Analyzer (Item 6).
- Triaxial seismic switch is set at the OBE acceleration level of 0.05g horizontal and 0.033g vertical.
- Triaxial seismic trigger is set at 0.02g all axes.
SOUTH TEXAS - UNIT 1 3/4 3-56
DRAFT TABLE 4.3-4 jut, 7 1987 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS ANALOG CHANNEL CHANNEL CHANNEL OPERATIONAL CHECK CALIBRATION TEST INSTRUMENTS AND SENSOR LOCATIONS
- 1. Triaxial Time-History Accelerometers ***
M R SA
- a. Free Field M R SA
- b. Containment Bldg. Foundation (Tendon Gallery E1. -36'9")
M R SA
- c. Outside Face Containment Shell (Reactor Containment Building El. 68'0")
M R SA
- d. Steam Generator Upper Lateral Support (Reactor Containment Building E1. 66'7\")
R SA
- e. Fuel Handling Building Foundation M (Fuel Handling Building E1. -29'0")
- f. Mechanical Electrical Auxiliary M R Building (Mechanical Electrical Auxiliary Building El. 35'0")
- 2. Triaxial Peak Accelerographs N.A. R N.A.
- a. Spent Fuel Pool Heat Exchanger (Inlet Line Fuel Handling Building E1. 64'5\")
N.A. R N.A.
- b. Reactor Vessel (Reactor Containment Building E1. 68'0")
H.A. R N.A.
- c. Cold Leg of RC Piping (Reactor Containment Building E1. 34'3") ,
- 3. Self-Contained Triaxial Accelerograph M R (At Reactor Containment Building Foundaticn Tendon Gallery El. -36'9")
M R SA
- 4. Triaxial Seismic Switch * **
- 5. Triaxial Seismic Trigger * ** M R M R SA
- 6. Response Spectrum Analyzer * **
M R SA
- 7. Magnetic Tape Recorders **
M R N.A.
- 8. Playback System **
- With reactor control room indication and alarm
- At seismic monitoring panel in Control Room, Unit 1
- Accelerometer data is gathered and analyzed by the Response Spectrum Analyzer (Item 6).
SOUTH TEXAS - UNIT 1 3/4 3-57
~~ ~
l DRAFT INSTRUMENTATION g 7 jgg7 REMOTE SHUT 00WN SYSTEM i
LIMITING COMITION FOR OPERAll0N I 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. l APPLICABILITY: MODES 1, 2, and 3.
I ACTION:
- a. With the number of OPERABLE remote shutdown monitoring channels, j 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 SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- b. With the number of OPERABLE remote shutdown monitoring channels, I transfer switches, power or control circuits less than the Total l Number of Channels as required by Table 3.3-9, within 60 days l restore the inoperable channel (s) to GPERABLE 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.
4 SilEVEltLANCE RE0VIREMENTS
- 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 18 months.
SOUTH TEXAS - UNIT 1 3/4 3-61 v~ _, _ .._ . _ _ _
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i DRAFT 7g INSTRUMENTATION RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION LIMITINGENDlIl0E 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 a 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 l
next Semiannual Radioactive Effluent Release Report pursuant to l Specification 6.9.1.4 why this inoperability was not corrected within the time specified.
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. i SURVEILLANCE REQUIREMENTS 4.3.3.10 Each radioactive liquid effluent monitoring instrumentation channel 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.
1 SOUTH TEXAS - UNIT 1 3/4 3-79
DRAFT TABLE 3.3-13 (Continued) 7 1987 JUL TABLE NOTATIONS
- At all times.
- During GASE0US WASTE PROCESSING SYSTEM operation.
ACTION STATEMENTS ACTION 47 - (Not used)
ACTION 48 -
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 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
ACTION 49 - With the number of channels OPEP.ABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this' pathway may continue for up to 30 days provided grab samples are taken at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and these samples are analyzed for radioactivity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
ACTION 50 - (Not used)
. . ACTION 51 - With the number of channels OPERABLE less than required by the
- Minimum Channels OPERABLE requirement, operation of this GASEOUS WASTE PROCESSING SYSTEM may continue provided grab samples are collected at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and analyzed within the following 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
ACTION 52 - (Not used)
ACTION 53 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via the -TGcted pathway may continue for up to 30 days provided samples are continuously collected with auxiliary sampling equipment as required in Part A of the ODCM.
SOUTH TEXAS - UNIT 1 3/4 3-86
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DRAFT JUL 7 1987 INSTRUMENTATION 3/4. 3. 4 TURBINE OVERSPEED PROTECTION 4
LIMITING CONDIIION FOR OPERATION l 3.3.4 At least one Turbine Overspeed Protection System shall be OPERABLE.
APPLICABILITY: MODES 1, 2, and 3. l l
ACTION: i I
- a. With one stop valve or one governor valve per high pressure turbine steam line inoperable and/or with one reheat stop valve or one reheat intercept valve per low pressure turbine steam line inoperable, restore the inoperable valve (s) to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, j or close at least one valve in the affected steam line(s) or isolate the turbine from the steam supply within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
\
l
- b. With the above required Turbine Overspeed Protection System otherwise j inoperable, within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> isolate the turbine from the steam supply.
l S
l
_ SURVEILLANCE RE0VIREMENTS l
l 4.3.4.1 The provisions of Specification 4.0.4 are not applicable.
l 4.3.4.2 The above required Turbine Overspeed Protection System shall be demonstrated OPERABLE:
- a. At least once per 7 days 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 by direct observation of the sovement of each of the above valves through one complete cycle from the running position,
- c. At least once per 18 months by performance of a CHANNEL CALIBRATION 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, and 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.
f SOUTH TEXAS - UNIT 1 3/4 3-89
j l
DRAFT l t, REACTOR COOLANT SYSTEM JUL 7 587 I i
i HOT STANDBY
\
l 1
LlHITINGl0EDlIIDN FOR OPERATION 3.4.1.2 At least two of the reactor coolant loops listed below shall be 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 System breakers are open:^
- a. Reactor Coolant Loop A and its associated steam generator and reactor coolant pump,
- b. Reactor Coolant Loop B and its associated steam generator and reactor coolant pump,
- c. Reactor Coolant Loop C and its associated steam generator and reactor coolant pump, and j
- d. Reactor Coolant Loop D and its associated steam generator and reactor coolant pump. l APPLICABILITY: MODE 3.
l ACTION: l
- a. With less than the above required reactor coolant loops OPERABLE, restore the required lecps to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be !
in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- b. With only one reactor coolant loop in operation and the Reactor Trip System breakers in the closed position, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> open the Reactor Trip System breakers.
- c. With no reactor coolant loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to ret. urn the required reactor coolant loop to operation.
MRVEILLANCE RE0_UIREMENTS 4.4.1.2.1 At least the above required reactor coolant pumps, if not in opera-tion, shall be determined OPERABLE once per 7 days by verifying correct breaker alignments and indicated power 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 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
j 4.4.1.2.3 The required reactor coolant loops shall be verified in operation and i circulating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- All reactor coolant pumps may be deenergized for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> provided: ,
l (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.
SOUTH TEXAS - UNIT 1 3/4 4-2 w,wmeep m.- q pw o r *w+ e**petgyes-psqyom - t y sy = e mp g v parwee -ymp _ _'_7_*_ _ _ _ _ _ _ _ __,e, --_
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DRAFT REACTOR COOLANT SYSTEM 7 1987 JUL COLD SHUTDOWN - LOOPS NOT FILLED LIMITiliG 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.** l l
APPLICABILITY: MODE 5 with reactor coolant loops not filled.
ACTION:
j
initiate corrective action to return the required RHR loups to OPERABLE status as soon as possible. l I
i 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 l loop to operation.
SURVEILLANCE REQUIREMENTS 4.4.1.4.2.1 At least one RHR loop shall be determined to be in operation and l circulating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. !
4.4.1.4.2.2 Valves FCV-1108, FCV-111B, CV0201A, and CV0221 shall be verified l closed and secured in position by mechanical stops or rencval of air or elec- ;
trical power at least once per 31 days.
l
- Two RHR loops may be inoperable for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing provided the other RHR loop is OPERABLE and in operation.
- The RHR pump may be deenergized for up to I hour provided: (1) no opera-tions 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.
SOUTH TEXAS - UNIT 1 3/4 4-6
, - _ _ - ~ _ _ _ _ _ _ . - _ __. - - - _. -: - _ . - -
DRAFT REACTOR COOLANT SYSTEM 7 1987 JUL 3/4.4.4 RELIEF VALVES LIMIUNG CONDITION F0.R OPERATION 3.4.4 All power-operated relief valves (PORVs) and their associated block J valves shall be OPERABLE.
APPLICABILITY: MODES 1, 2, and 3.
ACTION:
- a. With one or more PORV(s) inoperable, because of excessive raat leakage, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the PORV(s) to OPER*ME status or close the associated block valve (s); otherwise, be in ai. least HOT STAUDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTOOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. l
- b. With one PORV inoperable due to causes other than excessive seat leakage, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the PORV to OPERABLE status or close the associated block valve and remove power from the block l valve; restore the PORV to OPERABLE status within the following 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTOOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
. c. With both PORV(s) inoperable due to causes other than excessive seat l' 1eakage, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 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 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. ,
- d. With one or more block valve (s) inoperable, within 1 hour: ,
(1) restore the block valve (s) to OPERABLE status, or close the !
block valve (s) and remove power from the block valve (s), or close j the PORV and remove power from the PORV; and (2) apply the ACTION b. i or c. above, as appropriate, for the isolated PORV(s).
i
- e. The provisions of Specification 3.0.4 are not applicable. !
i l
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i l
SOUTH TEXAS - UNIT 1 3/4 4-10 i i
r ~~~_ r _ m _r_ .___L~~_~___ _:. _ . ~*c-* *:r ~ - ~ cn ~ ~ ~ ~ :~~ ~ ~
DRAFT REACTOR COOLANT SYSTEM 7 1987 JUL 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPIRATION 3.4.6.1 The following Reactor Coolant System Leakage Detection Systems shall be OPERABLE:
- 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 Monitoring System.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
With a. or c. of the above required Leakage Detection Systems inoperable, operation may continue for up to 30 days provided grab samples of the contain-sent atmosphere are obtained and analyzed for gaseous and particulate radioac-tivity at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the required Gaseous or Particulate Radioactive Monitoring System is inoperable; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. 1
- a. With b. of the above required Leakage Detection Systems inoperable, i be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUT-DOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. l
- b. With a. and c. of the above required Leakage Detection Systems inoperable:
- 1) Restore either Monitoring System (a. or c.) to OPERABLE status within 7,2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and
- 2) Obtain and analyze a grab sample of the containment atmosphere for gaseous and particulate radioactivity at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and
- 3) Perform a Reactor Coolant System water inventory balance at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
Otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE RE0VIREMENTS
- 4. 4. 6.1 The Leakage Detection Systems shall be demonstrated OPERABLE by:
- a. Containment Atmosphere Gaseous and Particulate Monitoring Systems performance of CHANNEL CHECK, CHANNEL CALIBRATION, and DIGITAL CHANNEL OPERATIONAL TEST at the frequencies specified in Table 4.3-3, and
- b. Containment Normal Sump Level and Flow Monitoring System performance of CHANNEL CALIBRATION at least once per 18 months.
SOUTH TEXAS - UNIT 1 3/4 4-19
DRAFT TABLE 3.4-1 JUL 7 1987 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES VALVE NUMBER FUNCTION XS10007 A, B, C HHSI Cold Leg Injection Check Valves (RCS Loops 1, 2, 3)
XSI0009 A, B, C HHSI Hot Leg Recirculation Check Valves (RCS Loops 1, 2, 3)
XSIOO10 A, B, C LHSI/HHSI Hot Leg Recirculation Check Valves (RCS Loops 1, 2, 3)
XRH0020 A, B, C LHSI Hot Leg Recirculation Check Valves (RCS Loops 1, 2, 3) e.
XRH0032 A, B, C LHSI/RHR Cold Leg Injection Check Valves 4 (RCS Loops 1, 2, 3)
XSI0038 A, B, C LHSI/HHSI/RHR/ Accumulator Cold Leg Injection Check Valves (RCS Loops 1, 2, 3)
XSI0046 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 t I
(RCS Loops 1, 2, 3) 4 3/4 4-22 )
SOUTH TEXAS - UNIT 1 J__ , _ __. - ___ . _
l
4 4 .
DRAFT :
REACTOR COOLANT SYSTEM JUL 7 1987 3/4.4.8 SPECIFIC ACTIVITY LIMITlBG 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 I-131, and
- b. Less than or equal to 1006 microCuries per gram of gross radioactivity.
l APPLICABILITY: MODES 1, 2, 3, 4, and 5.
ACTION:
H0 DES 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 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> 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 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; and i
- b. With.the gross specific activity of the reactor coolant greater than 100/E microcuries per gram, be in at least HOT STANDBY with T avg less j than 500*F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
MODES 1, 2, 3, 4, and 5:
With the specific activity of the reactor coolant greater than 1 microcurie per gram DOSE EQUIVALENT I-131 or greater than 1006 micro-Curies per gram, perform the sampling and analysis requirements of Item 4.a) of Table 4.4-4 until the specific activity of the reactor coolant is restored to within its limits.
- With T,yg greater than or equal to 500 F.
SOUTH TEXAS - UNIT 1 3/4 4-26
, 4 i
DRhfT '
JUL 7 1987 il I e I\ 'I
( 250
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I F \
E ~ 200 \
15d \
\ i UNACCEPTABLE j OPERATION
- \3 l
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OO OPERATION %
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20 30 40 50 60 70 80 90 100 PERCENT OF RATED THERMAL POWER FIGURE 3.4-1 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY >1 pCi/ gram DOSE EQUIVALENT I-131 SOUTH TEXAS - UNIT 1 3/4 4-28
4 l
DRAFT Jul. 7 1987 MATERIAL PRCPERTY BASIS CONTROLLING MATERIAL - RV RTNOT INITIAL: lO'F INTERMEDIATE SHELL R-1606-3 RT NOT AFTER 32 EFPY COPPER CONTENT: CONSERVATIVELY I/4, 91*F ASSUMED AS 0.10 WT% 3/4T, 64*F l i
CURVE APPLICABLE FOR HEATUP RATES UP TO IOO'/HR FOR THE SERVICE PERIOD UP TO 32 EFPY AND CONTAINS MARGINS OF lO'F AND 60 PSIG FOR POSSIBLE INSTRUMENT ERRORS 3000 I
"e .
@ LEAK TEST LIMIT- >
J 7
f f Q.
2000 l [ l l
$ Y' Y
/ i D
W [ !
W Y W I
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[
B' j- / !
O
/ e CRITICALITY Z -
LIMIT H SASED ON INSERVICE f- HYDROSTATIC TEST T N N M E (242*)
HEATUP RATES J IDR ICE UP TO IOO'F/m g s2 Ever O
O 100 200 300 400 INDICATED TEMPERATURE (*F)
FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS - APPLICABLE UP TO 3 SOUTH TEXAS - UNIT 1 3/4 4-32
DRAFT MATERIAL PROPERTY BASIS JUL 7 1987 CONTROLLING MATERIAL - RV RTNDT *NITIAL: lO'F INTERMEDIATE SHELL R-1606-3 RTNDT AFTER 32 EFP(
COPPER CONTENTS: CONSERVATIVELY if4, gg.y ASSUMED AS 0.IO WT% 3/4T, 64'F i SINGLE CURVE APPLICABLE FOR COOLDOWN RATES UP TO IOO*/HR FOR THE SERVICE PERIOD UP TO 32 EFPY. AND CONTAINS MARGINS OF IO'F AND 60 PSIG FOR POSSIBLE INSTRUMENT ERRORS 3000 1
2 I
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- 2000
/
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m )
m W
Z
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H l000 -
COOLDOWN RATE 'F/HR J cezi esio) o z _ . _ . __ _
_9 .7 25p 50 l
'0# l ( 120'r)
I O
O 100 200 300 INDICATED TEMPERATURE ( *F)
FIGURE 3.4-3 l REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS - APPLICABLE UP TO 32 EFPY
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l SOUTH TEXAS - UNIT 1 3/4 4-33
. . _ . , . ~ ,._- . . _ . _ _ _ . . _ _ -- .
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- s. .
DRAFT JUL 7 E7 800 Ltl J
to o '
Jm O
E$ 700 s-r Xz 4m Io 0.
YN Y
- zw j
@ g 600 j g
zo y
i'
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Z l m Q.
500
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50 I00 200 300 400 j I
TRTD - AUCTIONEERED LOW MEASURED RCS TEMPERATURE (*F) i FIGURE 3.4-4 NOMINAL MAXIMilM ALLOWABLE PORV t SETPOINT FOR THE COLD OVERPRESSURE SYSTEM SOUTH TEXAS - UNIT 1 3/4 4-37
DRAFT REACTOR COOLANT SYSTEM g 7 g7 OVERPRESSURE PROTECTION SYSTEMS SL'EVEILLANCE RE0llREMENIS
- a. Performance of an ANALOG CHANNEL OPERATICHAL TEST on the PORV actuation channel, but excluding valve operation, within 31 days prior to entering a condition in which the PORY is required OPERABLE and at least once per 31 days thereafter when the PORV is required OPERABLE; l
- b. Performance of a CHANNEL CALIBRATION on the PORY actuation channel at least once per 18 months; and
- c. Verifying the PORV block valve is open at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> when the PORV is being used for overpressure protection.
4.4.9.3.2 The RCS vent (s) shall be verified to' be open at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
- when the vent (s) is being used for overpressure protection. .
I i
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1
- 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.
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SOUTH TEXAS . UNIT 1 3/4 4-38 l
j
4 DRH REACTOR COOLANT SYSTEM g 7 79g7 3/4.4.11 REACTOR VESSEL HEAD VENTS LIMITING C0HDITION 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:
l
- a. With one of the above reactor vessel head vent paths inoperable, '
STARTUP and/or POWER OPERATION may continue provided the inoperable vent path is maintained closed 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 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following J
30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. t
- 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 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the
' following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
S
_ SURVEILLANCE REQUIREMENTS 4.4.11 Each reactor vessel head vent path shall be demonstrated OPERABLE at least once per 18 months by:
- a. Verifying all manual isolation valves in each vent path are locked in the open position,
- b. Cycling each vent valve through at least one complete cycle of full travel from the control room, and
- c. Verifying flow through the reactor vessel head vent paths during venting.
I SOUTH TEXAS - UNIT 1 3/4 4-40 1 - - -. _ _.
1 DRAFI 3/4.5 EMERGENCY CORE COOLING SYSTEMS _ JUL 7 1987 j
3/4. 5.1 ACCUMULATORS l
Llw1TINGJONDITION FOR OPERATION l
3.5.1 Each Safety Injection System accumulator shall be OPERABLE with: l j
a, The isolation valve open and power removed,
- b. A contained borated water volume of between 8800 and 9100 gallons,
- c. A boron concentration of between 2400 and 2600 ppe, and
- d. A nitrogen cover pressure of between 590 and 670 psig.
APPLICABILITY: MODES 1, 2, and 3*. !
l ACTION: l
- a. With one accumulator inoperable, except as a result of a closed isolation valve or the boron concentration outside the required l limits, restore the inoperable accumulator to OPERABLE status with 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce pressurizer pressure to less than 1000 psig within the follow-ing 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
- b. With one accumulator inoperable due to the isolation valve being closed, either open the isolation valve within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce pressurizer pressure to less than 1000 psig within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
With the boron concentration of one accumulator outside the required
- c. !
limit, restore the boron concentration to within the required limits within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> l l
and reduce pressurizer pressure to less than 1000 psig within the l following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
SURVEILLANCE RE0VIREMENTS 1
- 4. 5.1.1 Each accumulator shall be demonstrated OPERABLE:
- a. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by:
- 1) Verifying, by the absence of alarms, the contained borated 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 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 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 i
- Pressurizer pressure above 1000 psig.
SOUTH TEXAS - UNIT 1 3/4 5-1
DRAFT EMERGENCY CORE COOLING SYSTEMS g 7g f
SU1YEILLANCE REQUIREMENTS (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:
i
- a. At least once per 31 days by the performance of an ANALOG CHANNEL J
OPERATIONAL TEST, and
- b. At least once per 18 months by the performance of a CHANNEL CALIBRATION.
i SOUTH TEXAS - UNIT 1 3/4 5-2 l
, , 1 l
DRAFT EMERGENCY CORE COOLING SYSTEMS 7 )gg SRVE11 LANCE _REW1REMENIL 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE:
- a. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying that the following valves are in the indicated positions with power to the valve operators removed:
Valve Number Valve Function Valve Position XSI0008 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. I
- 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 CONTAINMENT INTEGRITY, and
- 2) Of the areas affected within containment at the completion of each containment entry when CONTAINMENT INTEGRITY is established.
- d. At least once per 18 months by a visual inspection of the contain-ment 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 - UNIT 1 3/4 5-4
DRAFT l EWERGENCY CORE COOLING SYSTEMS g 7 gg SURVEILLANCE RE001REMENTS (Continued)
- e. At least once per 18 months, during shutdown, by:
- 1) Verifying that each automatic valve in the flow path actuates to its correct position on an Automatic Switchover to Containment Sump test signal, and
- 2) Verifying that each of the following pumps start automatically upon receipt of a Safety Injection test signal:
a) High Head Safety Injection pump, and l
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:
- 1) High Head Safety Injection pump t 1480 psid, and
- 2) Low Head Safety Injection pump t 286 psid.
- g. By performing a flow test, during shutdown, following completion of modifications to the ECCS subsystems that alter the subsystem flow characteristics and verifying that:
- 1) For High Head Safety Injection pump lines, with the High Head !
l Safety Injection pump running, the pump flow rate is greater than 1500 gpm and less than 1600 gpm.
- 2) For low Head Safety Injection pump lines, with the Low Head Safety Injection pump running, the pump flow rate is greater '
than 2700 gpm and less than 2800 gpm.
l SOUTH TEXAS - UNIT 1 3/4 5-5
(
EMERGENCY CORE COOLING SYSTEMS JUL 7 1987 3/ 4. 5. 4 (This specification number is not used.)
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SOUTH TEXAS - UNIT 1 3/4 5-9 l
1
DRAFT EMERGENCY CORE COOLING SYSTEMS 7g 3/4. 5. 5 REFUELING WATER STORAGE TANK LIWITIN.LC0f4flJ10H F0R OPERATION 3.5.5 The refueling water storage tank (RWST) shall be OPERABLE 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 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at least H0T STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
l l
SURVEILLANCE REQUIREMENTS 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.
SOUTH TEXAS - UNIT 1 3/4 5-10
DR,u 5 EwERGENCY CORE COOLING SYSTEMS Jul. 7 1987 3/4.5.6 RESIDUAL HEAT REMOVAL (RHR) SYSTEM tIMITING CDHDITION FOR OPERATION 3.5.6 Three independent Residual Heat Removal (RHR) loops shall be OPERABLE vith each loop comprised of:
- c. One OPERABLE flowpath capable of taking suction from its associated RCS hot leg and discharging to its associated RCS cold leg.*
APPLICABILITY: H0 DES 1, 2 and 3.
ACTION:
- a. With one RHR loop inoperable, restore the required loop to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANOBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
- b. With two RHR loops inoperable, restore at least two RHR loops to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
- 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 REOUIREMEgTS 4.5.6.1 Each RHR loop shall be demonstrated OPERABLE pursuant to the require-ments of 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 Reactor Coolant System pressure signal less than or equal to 700 psig, the interlocks will cause the valves to automatically close.
rValves MOV-60 A, B, and C and MOV-61 A, B, and C may have power removed to support the FHAR assumptions.
SOUTH TEXAS - UNIT 1 3/4 5-11
= _ _ _ . _:- -- - - . . . .-. -_ _ _
l DRAFT JUL 7 1987 CONTAINMENT SYSTEMS l SURVEILLANCE REQUIREMENTS (Continued 1
- 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,,
a Type A test shall be performed at least every 18 months until two consecutive Type A tests meet 0.75 L a at which time the above test schedule may be resumed;
- c. The accuracy of each Type A test shall be verified by a supplemental test which:
- 1) Confirms the accuracy of the test by verifying that the supple-mental test result, L c, is in accordance with the appropriate following equation:
ll e-(L am u o)l 10.25 L, is the j where L,, is the measured Type A test leakage and L a superimposed leak; i
- 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,. 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 tests involving:
- 1) Air locks,
- 2) Purge supply and exhaust isolation valves with resilient material seals, and
- 3) Penetrations using continuous Leakage Monitoring Systems.
- e. Air locks shall be tested and demonstrated OPERABLE by the require-ments of Specification 4.6.1.3;
- f. Purge supply and exhaust isolation valves with resilient material seals shall be tested and demonstrated OPERABLE by the requirements of Specification 4.6.1.7.3 or 4.6.1.7.4, as applicable; 8
SOUTH TEXAS - UNIT 1 3/4 6-3
DRAFT JUL 7 1987 CONTAINMENT SYSTEMS S FVEILLANCE RE0VIREMENTS 4.6.1.3 Each containment air lock shall be demonstrated OPERABLE:
- a. Within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following each closing, except when the air lock is being used for multiple entries, then at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, by verifying seal leakage is less than 0.01 L, as determined by precision i 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 l
- 2) Prior to establishing CONTAINMENT INTEGRITY when maintenance l
has been performed on the air lock that could affect the air lock sealing capability.**
l
- c. At least once per 6 months by verifying that only one door in each air lock can be opened at a time.
By verifying at least once per 7 days that the instrument air pres- l d.
sure in the header to the personnel 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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. j l
l l
l l
1
- The provisions of Specification 4.0.2 are not applicable.
- This represents an exemption to Appendix J, paragraph III.D.2 of 10 CFR Part 50.
4 SOUTH TEXAS - UNIT 1 3/4 6-6
-i-.
DRAFT CONTAINMENT SYSTEMS 7g E!RVEILLANCE REOUIREMERTS (Continued)
- b. Performing tendon detensioning, inspections, and material tests on a previously stressed tendon from each group (inverted U and hoop).
A randomly selected tendon from each group shall be completely detensioned in order to identify broken or damaged wires and deter-mining that over the entire length of the removed wire that:
- 1) The tendon wires have not undergone corrosion, cracks, or physical damage in excess of that allowed by ASTM A421-77.
- 2) There are not changes in the presence or physical appearance of the sheathing filler grease, and
- 3) A minimum tensile strength of 240,000 psi (guaranteed ultimate strength of the tendon material) for at least three wire sampics (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-l mum tensile strength test is evidence of abnormal degradation of the containment structure.
- c. Performing tendon retensioning of those tendons detensioned for inspection to their observed lif t-off force with a tolerance limit of +6%. During retensioning of these tendons, the chan and elongation should be measured simultaneously , 60%, andat 20%ges in l 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 1 design value given below, rhich are adjusted to account for elastic !
and time dependent losses; and Inverted U 126 ksi Hoop: Cylinder 128 ksi Dome 123 ksi
- e. Verifying the OPERABILITY of the sheathing filler grease by:
- 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.
SOUTH TEXAS - UNIT 1 3/4 6-10
~ . _ . _ . _ . .~ . _ _ . ._ ._ _ ._. ,,
7- - h
i DRAFT CONTAINMENT SYSTEMS JUL 7 1987 l
CONTAINMENT VENTILATION SYSTEM 1 LIw1 TING CONDIllDN 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.
ACTION:
- a. With a 48-inch containment purge supply and/or exhaust isolation valve open or not sealed closed, close and/or seal close that valve or isolate the penetration (s) within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, otherwise be in at
. least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN
- within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
- 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 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, otherwise be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
- c. With a contai.nment purge supply and/or exhaust isolation valve (s) having a measured leakage rate in excess of the limits of Specifi-cations 4.6.1.7.2 and/or 4.6.1.7.3, restore the inoperable valve (s) to OPERABLE status or isolate the penetrations so that the measured leakage rate does not exceed the limits of Specifications 4.6.1.7.2 and/or 4.6.1.7.3 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, otherwise be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in COLD SHUIDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SOUTH TEXAS - UNIT 1 3/4 6-12
DRAFT CCWAINMENT SYSTEMS 7g SURVEILLANCE REQUIREMENTS
(
4.6.1.7.1 Each 48-inch containment purge supply and exhaust isolation valve I
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 i and outboard isolation valves with resilient material 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 0.05 L, when pressuri:ed to P3.
4.6.1.7.3 At least once per 3 months each 18-inch supplementary containment f purge supply and exhaust isolation valve with resilient material seals shall be i i
demonstrated OPERABLE by verifying that the measured leakage rate is less than 0.01 L, when pressurized to P3 .
4.6.1.7.4 At least once per 31 days each 18-inch supplementary containment purge supply and exhaust isolation valve shall be verified to be closed or open in accordance with Specification 3.6.1.7.b.
1 I
I i
SOUTH TEXAS - UNIT 1 3/4 6-13
_ . . _ _ . - ____ _ _ _ . . ~ . _ ._ . . . _ _ . _ .-
DRAFT CONTAINMENT SYSTEMS g 7 g7 SPRAY ADDITIVE SYSTEM L MITING 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 Na0H soletion, and
- b. Three spray additive eductors each capable of adding Na0H 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 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore the Spray Additive System to OPERABLE status within the next 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in COLD SHU7DOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS 4.6.2.2 The Spray Additive System 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. 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 varifying that each automatic valve in the flow path actuates to its correct position on a Containment Pressure High 3 test signal; and
. SOUTH TEXAS - UNIT 1 3/4 6-15
CONTAINMENT SYSTEMS JUL 7 1987 SURVEILLANCE REQUIREMENTS (Continued)
- d. At least once per 5 years by verifying:
- 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, and under the following conditions: 1 a) CS pump suction pressure is > 15 psig, b) Valve CS0019A, B, or C, as applicable, is in the full open position, and c) CS pump recirculation flow rate to the RWST is 800 gpa i 100 gpm.
- 2) The lines between the spray additive tank and the eductors are not blocked by verifying flow, i
9 SOUTH TEXAS - UNIT 1 3/4 6-16
+genW = a i.w=wp ;q ++
- evapgeper-e r * -w w ; *yMfs %r* -w,3 -+ eyq *,w g -y,
DRJT CONTAINMENT SYSTEMS 3/4.6.3 CONTAINMENT ISOLATION VALVES l LIMITlHG CONDITIOM FOR OPERAT10B 3.6.3 The containment isolation valves shall be OPERABLE with isolation times less than or equal to the required isolation times.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
l Vith one or more of the isolation valve (s) inoperable, maintain at least one I isolation valve OPERABLE in each affected penetration that is open and:
- a. Restore the inoperable valve (s) to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, ;
or j
- b. Isolate each affected penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least l one deactivated automatic valve secured in the isolation position, j or
- c. Isolate each'affected penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least one closed manual valve or blind flange, or l l
- d. Be in at least HOT STAN0BY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
i SURVEILLANCE REOUIREMENIS ;
I 4.6.3.1 The isolation valves shall be demonstrated OPERABLE prior to returning i
the valve to service af ter maintenance, repair or replacement work is performed i on the valve or its associated actuator, control or power circuit by perform-ance of a cycling test, and verification of isolation time.
4.6.3.2 Each isolatio'n valve shall be demonstrated OPERABLE during the COLD SHUTDOWN or REFUELING MODE at least once per 18 months by:
- a. Verifying that on a Phase "A" Is'olation test signal, each Phase "A" isolation valve actuates to its isolation position;
- b. Verifying that on a Containment Ventilation Isolation test signal, each purge and exhaust valve actuates to its isolation position; and
- c. Verifying that on a Phase "B" Isolation test signal, each Phase "B" isolation valve actuates to its isolation position.
4.6.3.3 The isolation time of each power-operated or automatic valve shall be determined to be within its limit when tested pursuant to Specification 4.0.5.
SOUTH TEXAS - UNIT 1 3/4 6-18
_ _________._--a --- - - -- -- - - - . - - . - ~,
CONTAINMENT SYSTEMS JUL 7 567 3/4.6.4 COMBUSTIBLE GAS CONTROL HYOR0 GEN ANALYZERS LIWIIJBG CONDIIl0N FOR OPERATION 3.6.4.1 Two independent containment hydrogen analyzers shall be OPERABLE.
APPLICABILITY: MODES 1 and 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 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
- b. With both hydrogen analyzers inoperable, restore at least one analyzer to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
SURVEILLANCE RE0VIREMENTS 4.6.4.1 Each hydrogen analyzer shall be demonstrated OPERABLE by the performance l of a CHANNEL CHECK at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, 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 containing one volume percent hydrogen, balance nitrogen, and by performing a CHANNEL CALIBRATION at least once per 18 months using sample gas containing ten volume percent hydrogen, balance nitrogen.
SOUTH TEXAS - UNIT 1 3/4 6-19
- e 4, , r mesqPp e s***We*=1M * *N48*.
- 0"* 4 N
- P4eam-esse % ' t et e.WDernGF DP*
- p Wuurg vP tr ge, GL8Pf3
DRTI CONTAINMENT SYSTEMS 7 1987 JUL ELECTRIC HYOR0 GEN RECOMBINERS LIWITING CONDITION FOR OPERATION
APPLICABILITY: MODES 1 and 2.
ACTION:
With one Hydrogen Recombiner System inoperable, restore the inoperable system to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 6 hcurs.
SURVElfLANCE REQUIREMENTS 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 i 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 setting to maximum power for 2 minutes and verify that the power meter reads greater than or equal to 65 kW, and
- b. At least once per 18 months by:
- 1) Performing a CHANNEL CALIBRATION of all recombiner instruments-tion and control circuits,
- 2) Verifying through a visual examination that there is no evidence of abnormal conditions within the recombiner enclosure (i.e. , loose wiring or structural connections, deposits of foreign. materials,etc.),and
- 3) Verifying the integrity of all heater electrical circuits by i performing a resistance to ground test following the above l required functional test. The resistance to ground for any l heater phase shall be greater than or equal to 10,000 ohms.
SOUTH TEXAS - UNIT 1 3/4 6-20
_ . . - . _ . _ . _ _ _ . . _ . - . ~ ~ - , -
. o 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE JUL 7 587 SAFETY VALVES LIWITING CONDITION FOR OPERATION 3.7.1.1 All main steam line Code safety valves associated with each steam generator shall be OPERABLE with lift settings as specified in Table 3.7-2.
APPLICABILITY: MODES 1, 2, and 3.
l ACTION:
4
- 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 H0 DES 1, 2, and 3 may proceed provided that )
within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, either the inoperable valve is restored to OPERABLE (
status or the Power Range Neutron Flux High Trip Setpoint is reduced l per Table 3.7-1; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUT 00WN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
i
- b. The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.7.1.1 No additional requirements other than those required by Specification
- 4. 0. 5.
l l
SOUTH TEXAS - UNIT 1 3/4 7-1
_-_ __, , . . . . - . . .._ .,....-. .. _~._,_. _ - . .
PLANT SYSTEMS AU11LIARY FEEDWATER STORAGE TANK LIWITING CONDITION FOR OPERATION 3.7.1.3 The auxiliary feedwater storage tank (AFST) shall be OPERABLE with a contained water volume of at least 518,000 gallons of water.
APPLICABILITY: MODES 1, 2, and 3.
ACTION:
With the AFST inoperable, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> restore the AFST to OPERABLE status or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTOOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
SURVEILLANCE REQUIREMENTS _
4.7.1.3 The AFST shall be demonstrated OPERABLE at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying the contained water volume is within its limits.
t SOUTH TEXAS - UNIT 1 3/4 7-6
DRAFT PLANT SYSTEMS JUL 7 BSI 3/4.7.3 COMPONENT COOLING WATER SYSTEjj LIMITING CONDITION FOR OPERATION 3.7.3 At least three independent component cooling water loops shall be OPERABLE.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
With only two component cooling water loops OPERABLE, restore at least three l loops to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SI)RVEILLANCE RE0_VIREMENTS 4.7.3 At least three component cooling water loops shall be demonstrated OPERA 8LE:
- a. At least once per 31 days by verifying that each valve outside con-tainment (manual, power-operated, or automatic) servicing safety-related equipment that is not locked, sealed, or otherwise secured in position.is in its correct position; and
- 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 of Offsite Power, Containment Phase "B" Isolation, or Low Surge Tank test signal, as applicable, l
- 2) Each Component Cooling Water System pump starts automatically on a Safety Injection or loss of Offsite Power test signal, and l
1 3) The surge tank level instrumentation which provides automatic isolation of the non-safety-related portion of the system is demonstrated OPERABLE by pe'rformance of a CHANNEL CALIBRATION test.
- c. By verifying that each valve inside containment (manual, power-operated, or automatic) servicing safety-related equipment that is not locked, sealed, or otherwise secured in position is in its cor-rect position prior to entering MODE 4 following each COLD SHUTOOWN of greater than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if not performed within the previous 31 days. j f
l SOUTH TEXAS - UNIT 1 3/4 7-12 l
y 9 DRLFT ggt 7 1987 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, restore at least three loops to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE 1JE0VIREMENTS 4.7.4 At least three essential cooling water loops shall be demonstrated OPERABLE:
- a. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) servicing safety-related equipment that is not locked, sealed, or otherwise secured in position is in its correct position;
- 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, 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 TEXAS - UNIT 1 3/4 7-13
t l
PLANT SYSTEMS JUL 7 1987 3/4.7.5 ULTIMATE HEAT SINX LIMITING CONDITION FOR OPERATION 3.7.5 The ultimate heat sink shall be OPERABLE with:
- a. A minimum water level at or above elevation 25.5 feet Mean Sea Level, l USGS datum, and
- b. An Essential Cooling Water intake temperature of less than or equal ,
l to 99 F.
APPLICABILITY: MODES 1, 2, 3, and 4.
I ACTION:
With the requirements of the above specification not satisfied, be in at least !
HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COW SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. l l
SURVEILLANCE RE0VIREMENTS 4.7.5 The ultimate heat sink shall be determined OPERABLE at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying the intake water temperature and water level to be within their limits.
l l
SOUTH TEXAS - UNIT 1 3/4 7-14
. _ _ _ - m . . . _ . _ _ _. __. . _ , , _ . . . _ . . _ _.. .
um M +
DRAFT M
Pl>ST SYSTEMS l
SURVEILLANCE RE0VIREMENTS (ContinueID
- c. At least once per 18 months or (1) af ter any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following i painting, fire, or chemical release in any ventilation zone communi-cating with the system by:
- 1) Verifying that the makeup and cleanup systees 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.S.c, and C.S.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 - 1
- 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 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> 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 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%;
- 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 and 6.0 inches Water Gauge for the cleanup units while operating the systes at a flow rate of 6000 cfm i 10% for the cleanup units and 1000 cfm i 10% for the makeup units;
- 2) Verifying that on 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 filtration mode of operation with flow through the HEPA filters and charcoal adsorber banks of the cleanup and makeup units; SOUTH TEXAS - UNIT 1 3/4 7-17
~~ - ~ ^ - - -
r_r_2_rL__:rt_ _ _ _ __ __ : -~- _ _
1 DEFT PLANT SYSTEMS 7 1967 JUL SUWEILLANCE RE0VIREMENTS (Continued) _
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 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. i
- c. After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> 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 once per 18 months by:
- 1) Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 6 inches Water Gauge while operating the system at a flow rate of 29,000 cfm i 10%,
- 2) Verifying that the system starts on High Radiation and Safety Injection test signals and directs flow through the HEPA filters and charcoal adsorbers,
- 3) Verifying that the system maintains the FHB at a negative pressure of greater than or equal to 1/8 inch Water Gauge relative to the outside atmosphere, and
- 4) Verifyihg that the heaters dissipate 50 i 5 kW when tested in accordance with ANSI H510-1980.
- 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 operating the system at a flow rate of 29,000 cfm i 10%; and
- f. After each complete or partial replacement 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 H510-1980 for a halogenated hydrocarbon refrigerant test gas while operating the system at a flow rate of 29,000 cfm i 10%.
SOUTH TEXAS - UNIT 1 3/4 7-20
4 0 i
DRW 7 1987 PLANT SYSTEMS-3/4.7.13 AREA TEMPERATURE MONITORING LIMITING COND1110N FOR OPERATION 3.7.13 The temperature of each area shown in Table 3.7-3 shall not be exceeded for more than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or by more than 30 F.
APPLICABILITY: Whenever the equipment in an affected area is required to be OPERABLE.
ACTION:
- a. With the temperature inside any QDPS auxiliary processing cabinet exceeding 110*F for more than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, prepare an engineering evalu-ation within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to determine the temperature effects on QDPS OPERABILITY and service life. The provisions of Specifica-tion 3.0.3 and 3.0.4 are not applicable.
- b. With one or more areas exceeding the temperature limit (s) shown in Table 3.7-3 for more than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that provides a record of the cusulative time and the amount by which the temperature in the affected area (s) exceeded the limit (s) and an analysis to demonstrate the continued )
OPERABILITY of the affected equipment. The provisions of Specifi- 1 cations 3.0.3 and 3.0.4 are not applicable. l I
- c. With one or more areas exceeding the temperature limit (s) shown in Table 3.7-3 by more than 30 F, prepare and submit a Special Report as required by ACTION a. above and within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> either restore the area (s) to within the temperature limit (s) or declare the equip-ment in the affected area (s) inoperable.
SURVEILLANCE REQUIREMENTS 1 4.7.13 The temperature in each of the areas shown in Table 3.7-3 shall be determined to be within its limit at least: once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
I s
SOUTH TEXAS - UNIT 1 3/4 7-31
l DRAFT TABLE 3.7-3 g 7 g7 AREA TEMPERATURE MONITORING AREA TEMPERATURE LIMIT ( F)
- 1. Relay Room ~
< 78 (Electrical Auxiliary Building El. 35'0")
- 2. Switchgear Rooms
~
< 85 (Electrical Auxiliary Building E1.10'0",
35'0",60'0")
- 3. Electrical Penetration Spaces < 103
~
(Electrical Auxiliary Building E1.10'0",
35'0",60'0")
- 4. Safety Injection and Containment Spray -
< 101 Pump Cubicles (Fuel Handling Building E1. -29'0")
- 5. Component Cooling Water Pump Cubicles < 112
~
(Mechanical Auxiliary Building E1.10'0")
- 6. Centrifugal Charging Pump Cubicles < 132
~
(Mechanical Auxiliary Building E1. 10'0") i
- 7. Hydrogen Analyzer Room ~
< 102 (Mechanical Auxiliary B.uilding E1. 60'0")
- 8. Boric Acid Transfer Pump Cubicles < 101
~
(Mechanical Auxiliary Building El.10'0")
- 9. Standby Diesel Generator Rooms < 101*
(Diesel Generator Building E1. 25'0")
- 10. Essential Cooling. Water Pump Rooms < 101
~
(Essential Cooling Water Intake Structure E1. 34'0")
- 11. Isolation Valve Cubicles < 101 (Isolation Valve Cubicle E1.10' 0")
- 12. Qualified Display Processing System Rooms ~
< 94**
(Electrical Auxiliary Building E1.10'0")
l l
i
- Temperature limit is < 120 F when testing the standby diesel generator pursuant to Surveillance Requirement 4.8.1.1.2.e.7).
- Measurement inside QOPS auxiliary processing cabinets.
4 SOUTH TEXAS - UNIT 1 3/4 7-32
DRAFT PLANT SYSTEMS 7 1987 JJL 3/4.7.14 ESSENTIAL CHILLED WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.14 At least three independent Essential Chilled Water System loops shall be OPERABLE.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
With only two Essential Chilled Water System loops OPERABLE, restore three loops to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE RE0VIREMENTS 4.7.14 The Essential Chilled Water System shall be demonstrated OPERABLE by:
- a. Performance of surveillance as required by Specification 4.0.5, and
- b. At least once per 18 months by demonstrating that the system starts automatically on a Safety Injection test signal.
I I
l 1
l i
N SOUTH TEXAS - UNIT 1 3/4 7-33 l
DRAFT JUL 7 1987 ELECTRICAL POWER SYSTEMS D.C. SOURCES SHUTDOWN P LIMITING CONDITION FOR OPERATION l
3.8.2.2 As a minimum, Channel I and Channel IV 125-volt battery banks and their two associated chargers shall be OPERABLE.
APPLICABILITY: MODES 5 and 6.
i ACTION:
With the required battery banks and/or charger (s) inoperable, immediately sus-pend all operations involving CORE ALTERATIONS, positive reactivity changes, or movement of irradiated fuel; initiate corrective action to restore the required battery banks and/or chargers to OPERABLE status as soon as possible, and within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, depressurize and vent the Reactor Coolant System through a 2.0 square inch vent.
SURVEILLANCE REQUIREMENTS 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.
l l
l l
i SOUTH TEXAS - UNIT 1 3/4 8-13
._ _mm..__. .- __..__._-____-___m. __-__W
DRAFT JUL 7 1987 ELECTRICAL POWER SYSTEMS ONSITE POWER DISTRIBUTION SH"JTDOWN LIMITING CONDITI0B_f_0R OPERATION 3.8.3.2 As a minimum, the following electrical busses shall be energized in the specified manner:
I a. Train A and Train C of A.C. ESF busses 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, l
DP1201, DP002, and DP1204 energized from their associated inverter connected to its respective D.C. bus, and
- c. Channel I and Channel IV 125-voi.t D.C. busses energized from their associated battery banks.
APPLICABILITY MODES 5 and 6.
ACTION:
With any of the above required electrical busses not energized in the required manner, immediately suspend all operations involving CORE ALTERATIONS, positive reactivity changes, or movement of irradiated fuel, initiate corrective action to energize the required electrical busses in the specified manner as soon as possible, and within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, depressurize and vent the RCS through at least a 2.0 square inch vent.
SURVEILLANCE REQUIREMENTS 4.8.3.2 The specified busses shall be determined energized in the required manner at least once per 7 days by verifying correct breaker alignment and ,
indicated voltage on the busses.
b SOUTH TEXAS - UNIT 1 3/4 8-16 m _ _ , . . , _ __ _ _ _
, a l
DRAFT JUL 7 1987 ELECTRICAL POWER SYSTEMS 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES CONTAINMENT PENETRATION CONDUCTOR OVERCURkENT 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: H0 DES 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 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, 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; the provisions of Specification 3.0.4 are not applicable to overcurrent devices in circuits which have their backup circuit breakers tripped, their inoperable circuit breakers racked out or removed, or 1 l
- b. Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD l SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
l SURVEILLANCE REOUIREMENTi
- 4. 8. 4.1 Protective devices required to be OPERABLE as containment penetration conductor overcurrent protective devices shall be demonstrated OPERABLE:
- a. At least once per 18 months:
- 1) By verifying that the medium voltage 13.8 kV circuit breakers are OPERABLE by selecting, on a rotating basis, at least 10% 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 centrol circuits function as designed, and SOUTH TEXAS - UNIT 1 3/4 8-17
- *
- W++ pug e-posg - = ,,p . ,
.NN ,.
DRTT 3/4.9 REFUELING OPERATIONS 7 19E' JUL 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 unifors and sufficient to ensure that the more restrictive of the following reactivity conditions is met; either:
- a. A K,ff of 0.95 or less, or
- b. A boron concentration of greater than or equal to 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 continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppe 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 l the more restrictive.
1 ..
i SURVEILLANCE RE0VIREMENTS
- 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 Withdrawal of any full-length control rod in excess of 3 feet from l b.
its fully inserted position within the reactor vessel. l
- 4. 9.1. 2 The boron concentration of the Reactor Coolant System and the refueling canal shall be determined by chemical analysis at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. '
1 4.9.1.3 Valves FCV-1108, FCV-1113, CV0201A, and CV0221 shall be verified closed and secured in position by mechanical stops or by removal of air or electrical power at least once per 31 days.
- 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.
i 8
SOUTH TEXAS - UNIT 1 3/4 9-1 w,.,,n . e , wm e s nes,.1 w.--- - m ,-- - , ,
DRAFT REFUELING OPERATIONS JUL 7 1987 3/4.9.11 WATER LEVEL - STORAGE POOLS SPENT FUEL P00L LIWITING 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 requirements of the above specification not satisfied, suspend all movement of fuel assemblies and crane operations with loads in the fuel storage areas and restore the water level to within its limit within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. >
SURVEILLANCE REQUIREMENTS
- 4. 9.11.1 The water level in the spent fuel pool shall be determined to be at least its minimum required depth at least once per 7 days when irradiated fuel I assemblies are in the spent fuel pool.
i l
l SOUTH TEXAS - UNIT 1 3/4 9-12 i
j
-l I
REFUELING OPERATIONS JUL 7 1987 SLRyEILLANCE RE0VIREMENTS (Continued 1
- 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 penetration 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 Regulatory Positions C.5.a, C.5.c, and C.S.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is 29,000 cfm i 10%;
l
- 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, 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%; 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.
I
- c. After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by verifying, within 31 days af ter 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%. f
! d. At least once per 18 months by:
- 1) Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 6 inches Water Gauge while operating the system at a flow rate of ,
29,000 cfm i 10%, ;
- 2) Verifying that on a High Radiation test signal, the system j automatically starts (unless already operating) and directs its exhaust flow through the HEPA filters and charcoal adsorber l .
banks, 1
l SOUTH TEXAS - UNIT 1 3/4 9-15
. I o
- RADIOACTIVE EFFLUENTS GASE0US WASTE PROCESSING SYSTEM g ej 37 LIMITING CONDITION FOR OPERATION l
3.11.2.4 The GASEOUS WASTE PROCESSING SYSTEM shall be OPERABLE and appropriate l portions of this system shall be used to reduce releases of radioactivity when l I
the projected doses in 31 days due to gaseous effluent releases, from each unit, to areas at and beyond the SITE B0UNDARY (see Figure 5.1-3) would exceed:
- a. 0.2 mrad to air from gamma radiation, or
- b. 0.4 mrad to air from beta radiation, or
- c. 0.3 mrem to any organ of a MEMBER OF THE PUBLIC. !
APPLICABILITY: At all times.
ACTION:
- a. With radioactive gaseous waste being discharged without treatment and in excess of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that includes the following information:
- 1. Identification of any inoperable equipment or subsystems, and the reason for the inoperability,
- 2. Action (s) taken to restore the inoperable equipment to OPERABLE ,
status, and l
- 3. Summary description of action (s) taken to prevent a recurrence,
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
3RVEILLMCE RE0VIREMEkTS 4.11.2.4.1 Doses due to gaseous releases from each unit to areas at and beyond the SITE BOUNDARY shall be projected at least once per 31 days in accordance with the methodology and parameters in the ODCM when the GASEOUS WASTE PROCESSING SYSTEM is not being fully utilized.
4.11.2.4.2 The installed GASEOUS WASTE PROCESSING SYSTEM shall be considered OPERABLE by meeting Specifications 3.11.2.1 and 3.11.2.2 or 3.11.2.3.
SOUTH iEXAS - UNIT 1 3/4 11-8
~ ~" ~~
_ _-----__ l __ ___1 f_T ' ~ ~ ~ ~ - - ~'
i RA310 ACTIVE EFFLUENTS JUL 7N (
EXPLOSIVE GAS HIXTURE l
LIMITING _CONDIUON FOR OPERAUON 3.11.2.5 The concentration of oxygen in the GASE0US WASTE PROCESSING SYSTEM inlet shall be limited to less than or equal to 3% by volume.
APPLICABILITY: At all times.
ACTION:
- a. With the concentration of oxygen in the GASEOUS WASTE PROCESSING SYSTEM inlet exceeding the limit, restore the concentration to within the limit within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE RE0VIREMENTS . _ . _
4.11.2.5 The concentration of oxygen in the GASEOUS WASTE PROCESSING SYSTEM shall be determined to be within the above limits by continuously monitoring the waste gases entering the GASE0US WASTE PROCESSING SYSTEM with the oxygen monitor required OPERABLE by Table 3.3-13 of Specification 3.3.3.11.
SOUTH TEXAS - UNIT 1 3/4 11-9
. . . . . ~ . . _ . , _ . . _ . , . , . _ . . ._ . . _ . . ..__
e 1
3/4.1 REACTIVITY CONTROL SYSTEMS 7 1987 l JUL BASES 3/4.1.1 BORATION CONTROL i
3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN A suf ficient SHUTDOWN MARGIN ensures that: (1) the reactor can be made l subcritical from all operating conditions, (2) the reactivity transients asso- )
ciated with postulated accident conditions are controllable within acceptable !
i limits, and (3) the reactor will be maintained sufficiently subcritical to l
preclude inadvertent criticality in the shutdown condition.
l SHUTDOWN MARGIN requirements vary throughnut core life as a function of I In MODES 1 and 2, the !
fuel depletion, RCS boron concentration, and RCS Tavg.
most restrictive condition occurs at EOL, with T ,yg at no load operating temperature, and is associated with a postulated steam line break accident and resulting uncontrolled RCS cooldown. In the analysis of this accident, a l sinimum SHUTDOWN MARGIN of 1.75% ak/k is required to control the reactivity transient. The 1.75% Ak/k SHUTDOWN MARGIN is the design basis minimum for the 14-foot fuel using Hafnium control rods (Ref. FSAR Table 4.3-3). Accordingly, I the SHUTDOWN MARGIN requirement for MODES 1 and 2 is based upon this limiting condition and is consistent with FSAR safety analysis assumptions. In MODES 3, 4, and 5, the most restrictive condition occurs at BOL, when the boron concen-tration is the greatest. In these modes, the required SHUTDOWN MARGIN is com-posed of a constant requirement and a variable requirement, which is a function of the RCS boron concentration. The constant SHUTDOWN MARGIN requirement of 1.75% ak/k is based on an uncontrolled RCS cooldown from a steamline break accident. The variable SHUTDOWN MARGIN requirement is based on the results of a boron dilution accident analysis, where the SHUTDOWN MARGIN is varied as a function of RCS boron concentration, to guarantee a minimum of 15 minutes for operator action after a boron dilution alarm, prior to a loss of all SHUTDOWN j MARGIN.
The boron dilution analysis assumed a common RCS volume, and maximum dilution flow rate for MODES 3 and 4, and a different volume and flow rate for H0DE 5. The MODE 5 conditions 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 operating in MODE 3 or 4, then the SHUTDOWN MARGIN requirements for MODE 5 l shall apply.
j 3/4.1.1. 3 MODERATOR TEMPERATURE COEFFICIENT The limitations on moderator temperature coefficient (MTC) are provided to ensure that the value of this coefficient remains within the limiting !
condition assumed in the FSAR accident and transient analyses.
The MTC values of this specification are applicable to a specific set of plant conditions; accordingly, verification of MTC values at conditions other !
than those explicitly stated will require extrapolation to those conditions in order to permit an accurate comparison.
SOUTH TEXAS - UNIT 1 8 3/4 1-1 l
DRAFT RE. ACTIVITY CONTROL SYSTEMS JUL 7 1967 ggES KDERATOR TEMPERATURE COEFFICIENT _ (Continued)
The most negative MTC, value equivalent to the most positive moderator density coefficient (MDC), was obtained by incrementally correcting the MDC used in the FSAR analyses to nominal operating conditions. These corrections involved subtracting the incremental change in the MDC associated with a core condition of all rods inserted (most positive MDC) to an all rods withdrawn condition and, a conversion for the rate of change of moderator density with temperature at RATED THERMAL 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 boron) at 6 core condition of 300 ppm equilibrium boron concentration and is obtained by making these corrections to the Ifmiting MTC value of -4.0 x 10 4 ok/k/*F.
The Surveillance Requirements for measurement of the HTC at the beginning and near the end of the fuel cycle are adequate to confim that the MTC remains within its limits since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup.
3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY l This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 561*F. This limitation is required to ensure: (1) the moderator temperature coefficient l is within its analyzed temperature range, (2) the trip instrumentation is within !
its normal operating range, (3) the pressurizer is capable of being in an OPERABLE ststus with a steam bubDie, and (4) the reactor vessel is above its minimum RT NDT temperature. l 3/4.1. 2 BORATION SYSTEMS The Boron Injection System ensures that negative reactivity control is available during each, mode of facility operation. The components required to perform this function include: (1) borated water sources, (2) charging pumps, (3) separate flow paths, (4) boric acid transfer pumps, and (5) an emergency power supply from OPERABLE diesel generators.
With the RCS average temperature above 350*F, a minimum of two boron injection flow paths are required to ensure single functional capability in t.he event an assumed failure renders one of the flow paths inoperable. The boration capability of either flow path is sufficient to provide a SHUTDOWN MARGIN from expected operating conditions of 1.75% ak/k after xenon decay and cooldown to 200 F. The maximum expected boration capability requirement occurs at EOL from full power equilibrium xenon conditions and requires 27,000 gallons of 7000 ppm borated water from the boric acid storage system or 458,000 gallons of 2500 ppm borated water from the refueling water storage tank (RWST). The RWST volume is an ECCS requirement and is more than adequate for the required boration capability. ,
SOUTH TEXAS - UNIT 1 B 3/4 1-2
DRAFT REACTIVITY CONTROL SYSTEMS g 7 jgg7 BASES BORATION SYSTEMS (Continued)-
With the RCS temperature below 350*F, one boron injection flow path / source )
l 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, pumps except the required OPERABLE pump to be inoperable below 350*F provides assurance that a mass addi-l 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 SHUTOOWN MARGIN is varied as a function of RCS boron concen-tration after xenon decay and cooldown from 200*F to 140*F. This condition requires either 2900 gallons of 7000 ppm borated water from the boric acid storage system or 122,000 gallons of 2500 ppm borated water from the RWST for K)DE 5 and 33,000 gallons of 2500 ppm borated water from the RWST for MODE 6.
The contained water volume limits include allowance for water not available because of discharge line location and other physical characteristics.
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 i within containment after a LOCA. This pH band minimizes the evolution of l iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components. l The OPERABILITY of one Boron Injection System during REFUELING ensures that this system is available for reactivity control while in MODE 6.
3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that: (1) acceptable power distri-bution limits are maintained, (2) the minimum SHUTDOWN MARGIN is maintained, and (3) the potential effects of rod misalignment on associated accident analyses are i limited. OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits. Verification that the Digital Rod Position Indicator agrees with the demanded position within i 12 steps at 24, 48, 120, and 259 steps withdrawn for the Control Banks and 18, 234, and 259 steps with-drawn for the Shutdown Banks provides assurances that the Digital Rod Position Indicator is operating correctly over the full range of indication. Since the Digital Rod Position Indication System does 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.
SOUTH TEXAS - UNIT 1 8 3/4 1-3
DRAFT 1 REACTIVITY CONTROL SYSTEMS JUL 7 587 psES HCYABLE CONTROL 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 rod requires These measurement restrictions pro-of peaking factors and a restriction in THERMAL POWER. In addition, vide assurance of fuel rod integrity during continued operation.
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. Measurement with T,yg greater than or equal to 561*F and with all reactor coolant pumps operating ensure 3 that the measured drop times will be representative of insertion times experienced during a Reactor trip at operating conditions.
Control rod positions and OPERABILITY of the rod position indicators are required to be verified on a nominal basis of once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> with more fre-que.nt verifications required if an automatic monitoring channel is inoperable. .
These verification frequencies are adequate for assuring that the applicable LCDs are satisfied,
,i 9
SOUTH TEXAS - UNIT I B 3/4 1-4
DRLFT POWER DISTRIBUTION LIMITS JUL 7 1987 f 1
BASES HEAT FLUX H0T CHANNEL FACTOR and NUCLEAR ENTHALPY RISE HOT CHANNEL FACIOR (Continued)
When an F measurement is taken, an allowance for both experimental error q
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, Fxy(Z), is measured periodically to provide assurance that the Hot Channel Factor, F (Z), remains within its limit. The 9
F limit for RATED THERMAL POWER (F RTP) as provided in the Radial Peaking xy !
Factor Limit Report per Specification 6.9.1.6 was determined from expected l 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 DNB and linear heat generation rate protection with x y plane power tilts. A limit of 1.02 was selected to provide an allowance for the uncertainty 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 identification and correction of a dropped or misaligned control rod. In the event such action does not correct the tilt, the margin for uncertainty on Fq is reinstated by reducing the maximum allowed power by 3% for each percent of tilt in excess of 1.
For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the moveable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO. The incore detector monitoring is done with a full incore flux map or two sets of four symmetric thimbles. The two sets of four symmetric 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.
3/4.2.5 DNB PARAMETERS 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 t
SOUTH TEXAS - UNIT 1 B 3/4 2-5
PC%TR DISTRIBUTION LIMITS 7 1987 JUL 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 avg 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 flew 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 within their limits following load changes and other expected transient operation.
S SOUTH TEXAS - UNIT 1 B 3/4 2-6
~~ -
^
DRAFT INSTRUMENTATION 7 1987 JUL PASES REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTCM IN5TRUMENTATION (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 I
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 value for REPORTABLE EVENTS.
The methodology to derive the Trip Setpoints is based upon combining all ,
/
c f 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 those uncertainty magnitudes. Rack drift in excess of the Allowable Value exhibits the behavior that the rack has not met its allowance. Being that there is a small statistical chance that this will happen, an infrequent excessive drift is expected. Rack or sensor drift, in excess of the allowance that is more than occasional, say 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 taken in the analyses for those channels with ,
response times indicated as not applicable. Response time may be demonstrated i i
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 selectad plant para-meters and determines whether or not prede.termined 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) reactor containment fan coolers start, (11) essential cooling water pumps l
start and automatic valves position, (12) Control Room Ventilation Systems start, and (13) component cooling water pumps start and automatic valves position.
SOUTH TEXAS - UNIT 1 B 3/4 3-2 1 . . - .. - , ..
DRUT 7 1987 JUL INSTRUMENTATION BASES REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM IN5T RUMENTATION (Continued)
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 ]
I main feedwater valves which were closed by a safety Injection or High Steam Generator Water Level or Excessive Cooldown Protection signal, allows Safety Injection block so that components can be reset or tripped, and actuates P-15.
Reactor not tripped prevents manual block of Safety Injection.
P-11 On increasing pressurizer pressure, P-11 automatically reinstates i Safety Injection actuation on low pressurizer pressure or excessive cooldown signals, reinstates steamline isolation on excessive cool-down signals, and opens the accumulator discharge isolation valves.
On decreasing pressure, P-11 allows the manual block of Safety Injec-tion actuation on low pressurizer pressure or excessive cooldown sig-nals, allows the manual block of steamline isolation on excessive cooldown signals, and enables steam line isolation on high negative steam line pressure rate.
P-12 On increasing reactor coolant loop temperature, P-12 automatically provides an arming signal to the Steam Dump System. On decreasing 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 feedwater pumps, and closes all feedwater isolation valves and feedwater control valves.
P-15 When the reactor is tripped (P-4) or when below the power range neutron flux setpoint, P-15 is present and allcws Safety injection actuation and allows feedwater and main steamline isolation on Low-Low Tcold isolation an'd turbine trip from Low Compensated Tcold r high feed-water flow.
3/4.3.3 MONITORING 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 selected plant systems and locations and determine whether or not predetermined limits are being exceeded. If they are, the signals are combined into logic matrices sensitive to combinations indicative of various accidents and abnormal conditions. Once the required logic combination is
- completed, the system sends actuation signals to initiate alarms or automatic isolation action and actuation of Emergency Exhaust or Ventilation Systems.
SOUTH TEXAS - UNIT 1 B 3/4 3-3
3/4.4 REACTOR COOLANT SYSTEM JUL 7 1987 jLASES 3/4.4.1 REACTOR C0OLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with all reactor coolant loops in operation and maintain DNBR above 1.30 during all normal operations and anticipated transients. In MODES 1 and 2 with one reactor coolant loop not in operation this specification requires that the plant be in at least HOT STAND 8Y within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
In MODE 3, two reactor coolant loops provide sufficient he.at 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 opening the Reactor Trip System breakers. Single failure considerations require that two loops be OPERABLE at all times.
In MODE 4, and in MODE 5 with reactor coolant loops filled, a single reactor coolant loop or RHR loop provides sufficient heat removal capability 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 I
sufficient heat removal capability for removing decay heat; but single failure considerations, and the unavailability of the steam generators as a heat removing component, require that at least two RHR loops be OPERABLE.
The boron dilution analysis assumed a common RCS volume, and maximum di-lution flow rate for MODES 3 and 4, and a different volume and flow rate for MODE 5. The MODE 5 conditions 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 operating 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 reactivity changes during boron concentration reductions in the Reactor Coolant System. The reactivity change rate associated with boron reduction will, t.herefore, be within the capability of operator recognition and control. l l
The restrictions on starting an RCP with one or more RCS cold legs less than or equal to 350*F are provided to prevent RCS pressure transients, caused by energy additions from the Secondary Coolant System, which could exceed the limits of 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 l restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 50 F above each of the RCS cold leg temperatures.
i 3/4.4. 2 SAFETY VALVES j
The pressurizer 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 2500 psia. The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during shutdown. In the event that no safety valves are OPERABLE, an operating RHR loop, connected to the l
SOUTH TEXAS - UNIT 1 B 3/4 4-1 !
, o DRAFT l REACTOR COOLANT SYSTEM JUL 7 1987 l
BASES i
CHEMISTRY (Continued) i the chemistry within the Steady-State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant. The associated effects of exceeding the oxygen, )
chloride, and fluoride limits are time and temperature dependent. Corrosion studies show that operation may be continued with contaminant concentration levels in excess of the Steady-State Limits, up to the Transient Limits, for ,
the specified limited time intervals without having a significant effect on l the structural integrity of the Reactor Coolant 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 r that the resulting 2-hour doses at the SITE BOUNDARY will not exceed an l
appropriately small fraction of 10 CFR Part 100 dose guideline values following a steam generator tube rupture accident in conjunction with an assumed steady- l state reactor-to-secondary steam generator leakage rate of 1 gps. 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 meteorological conditions, were not considered in this evaluation.
The ACTION statement permitting POWER OPERATION to continue for limited time periods with the reactor coolant's specific activity greater than 1 microcurie / gram DOSE, EQUIVALENT I-131, but within the allowable limit shown on Figure 3.4-1, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER.
The sample analysis for determining the gross specific activity and E can exclude the radioiodines because of the low reactor coolant limit of 1 microcurfe/
gram DOSE EQUIVALENT I-131, and because, if the limit is exceeded, the radiciodine level is to be determined every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. If the gross specific activity level and radioiodine 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 20%. The exclusion of radio-nuclides with half-lives less than 15 minutes from these determinations has SOUTH TEXAS - UNIT 1 B 3/4 4-5 l
1 - . _ - - -. .-
. e DRAFT REACTOR COOLANT SYSTEM JUL 7 1987 BASES SPECIFIC ACTIVITY (Continued) been made for several reasons.
The first consideration is the difficulty to identify short-lived radionuclides in a sample that requires a significant time to collect, transport, and analyze. The second consideration is the ;
predictable delay time between the postulated release of radioactivity from l 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 l
in the typical reactor coolant have half-lives of less than 4 minutes or l
half-lives of greater than 14 minutes, which allows a distinction between the radionuclides above and below a half-life of 15 minutes. For these reasons the' radionuclides that are excluded from consideration are expected to decay to very low levels before they could be transported from the reactor coolant to the SITE BOUNDARY under any accident condition.
t Based upon the above considerations for excluding certain radionuclides from the sample analysis, the allowable time of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> between sample taking and completing the initial analysis is based upon a typical time necessary to perform the sampling, transport the sartple, and ?erform the analysis of about 90 minutes. After 90 minutes, the gross count s1ould be made in a reproducible i
.. geometry o'f sample and counter having reproducible beta or gamma self-shielding i properties. The counter should be reset to a reproducible efficiency versus energy. It is not necessary to identify specific nuclides. The radiochemical determination of nuclides should be based on multiple counting of the sample within typical counting basis following sampling of less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, about 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, about 1 day, about 1 week, and about 1 month.
l j
Reducing T,yg 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 relief valves.
The Surveillance Requirements provide adequate assurance that excessive specific ,
activity levels in the reactor coolant will be detected in sufficient time to take corrective action. A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.
3/4.4.9 PRESSURE / TEMPERATURE LIMITS l
The temperature and pressure changes during heatup and cooldown are limited to be consistent with the requirements given in the ASME Boiler and Pressure Vessel Code,Section III, Appendix G:
- 1. The reactor coolant temperature and pressure and system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with Figures 3.4-2 and 3.4-3 for the service period specified thereon:
SOUTH TEXAS - UNIT 1 B 3/4 4-6 l
. o DRAFT REACTOR COOLANT SYSTEM g 7 3937 l
BASES PRESSURE TEMPERATURE LIMITS (Continued)
- a. Allowable combinations of pressure and temperature for specific temperature change rates are below and to the right of the limit lines shown. Limit lines for cooldown rates between those presented may be obtained by interpolation; and
- b. Figures 3.4-2 and 3.4-3 define limits to assure prevention of non-ductile failure only. For normal operation, other inherent plant characteristics, e.g. , pump heat addition and pressurizer heater capacity, may limit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges.
- 2. These limit lines shall be calculated periodically using methods provided 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 accordance 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 accordance with the NRC Standard Review Plan, ASTM E185-73, and in accordance with additional reactor vessel requirements. These properties are then evaluated in accordance with Appendix G of the 1976 Summer Addenda to Section III.of the ASME Boiler and Pressure Vessel Code and the calculation methods described in WCAP-7924-A, " Basis for Heatup and Cooldown Limit Curves," April 1975.
Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RTNDT, at the end of 32 effective full power years (EFPY) of service life. The 32 EFPY service at the 1/4T location in life period is chosen such that the limiting RTf thUDIimiting unirradiated material.
the core region is greater than the RTNDT The selection of such a limiting RTHDT assures that all components in the Reactor Coolant System will be operated conservatively in accordance with applicable Code requirements.
The reactor vessel materials have been tested to determineReactor their initial opera-RT NDT; the results of these tests are shown in Table B 3/4.4-1.
tion and resultant fast neutron (E greater than 1 MeV) irradiation can cause SOUTH TEXAS - UNIT 1 B 3/4 4-7
. i. .
DMFI ;
i REACTOR COOLANT SYSTEM 7g BASES PRESSURE / TEMPERATURE LIMITS (Continued)
Therefore, an adjusted reference temperature, based an increase in the RTNDT.
upon the fluence, copper content, and phosphorus content of the material inCO*~
question, can be predicted using Figure B 3/4.4-1 and the value of ART NDT puted by Regulatory Guide 1.99, Revision 1, " Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials." The heatup and cool-down limit curves of Figures 3.4-2 and 3.4-3 include predicted adjustments for at the end of 32 EFPY as well as adjustments for possible this shift in RTNDT ;
errors in the pressure and temperature sensing instruments.
determined in this manner may be used until the results
, Values of ARTNDT j
from the material surveillance program, evaluated according to ASTM E185, are available. Capsules will be removed in accordance with the requirements of The surveillance specimen with-3 j
ASTM E185-73 and 10 CFR Part 50, Appendix H.The lead factor represents the rela- l drawal schedule is shown in Table 4.4-5. 4 tionship between the fast neutron flux density at the location of the capsule '
and the inner wall of the reactor vessel. Therefore, the results obtained from the surveillance specimens can be used to predict future radiation damage to the reactor vessel material by using the lead factor and the withdrawal time of the capsule. The heatup and cooldown curves must be recalculated when the ART determined from the surveillance capsule exceeds the calculated ARTHDT NDT I for the equivalent capsule radiation exposure.
f Allowable pressure-temperature relationships for various heatup and cool- !
down rates are calculated using methods derived from Appendix G in Section III l
of the ASME Boiler and Pressure Vessel Code as required by Appendix G to 10 CFR j
Part 50, and these methods are discussed in detail in WCAP-7924-A.
The general method for calculating heatup and cooldown limit curves is based upon the principles of the linear elastic fracture mechanics (LEFM) technology.
In the calculation procedures a semielliptical surface defect with a depth of one quarter of the wall thickness, T, and a length of 3/2T is assumed to exist .
at the inside of the vessel wall as well as at the outside of the vessel wall. l The dimensions of this postulated crack, referred to in Appendix G of ASME Sec-tion III as the reference flaw, amply exceed the current capabilities of inser-vice inspection techniques. Therefore, the reactor operation limit curves de-veloped for this reference crack are conservative and provide sufficient safety margins for protection against nonductile failure. To assure that the radiation i embrittlement effects are accounted for in the calculation of the limit curves, j the most limiting value of the nil-ductility reference temperature, RTNDT, is used and this includes the radiation-induced shift, ARTNDT, c rresponding to the l end of the period for which heatup and cooldown curves are generated.
The ASME approach for calculating the allowable limit curves for various l heatup and cooldown rates specifies that the total stress intensity factor, Ky , for the combined thermal and pressure stresses at any time during heatup l or cooldown cannot be greater than the reference stress intensity factor, KIR' for the metal temperature at that time. K IR is obtained from the reference SOUTH TEXAS - UNIT 1 B 3/4 4-8
DRLFT JUL 7 1987 4
2 ,/f 1
/ 1
'/ /
0; g /
/ I/4T-2.22 X 1019N/CM2
\
O EE 6 '
f i
o / l
\
Z 4
}
7 j
, s i
[ i b f / /
k, g ] /? / .
d
( /-
-7 3/4T-4.67 x lote N/cM2 l
z o i gl 8 >
o- g I e H r D 6 l to Z
/ - - -- -- ---
7 I
4
/
I r
2 -
i lOI7 0 IO 20 30 40 50 60 EFFECTIVE FULL POWER (YEARS)
FIGURE B 3/4.4-1 FAST NEUTRON FLUENCE (E>1MeV) AS A FUNCTION OF FULL POWER SERVICE LIFE 4
SOUTH TEXAS - UNIT 1 B 3/4 4-10
_______::_:_2-_________________-___-________
. o REACTOR COOLANT SYSTEM WL 7 M7 BASES PRESSURE / TEMPERATURE LIMITS (Continued) i fracture toughness curve, defined in Appendix G to the ASME Code. The K IR curve is given by the equation:
K IR = 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 RTHDT. Thus, the governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows:
(2)
CKIM + kit I IR l
Where: K IM = the stress intensity factor caused by membrane (pressure) stress, K It = the stress intensity factor caused by the thermal gradients, ]
i K IR = c nstant provided by thef Code as a function of temperature the material, l relative to the RT NDT C = 2.0 for level A and B service limits, and C = 1.5 for inservice hydrostatic and leak test operations.
At any time during the heatup or cooldown transient, KIR is determined by the metal temperature at the tip of the postulated flaw, the appropriate value The thermal stresses for RTNDT, and the reference fracture toughness curve.
resulting from temperature gradients through the vessel wall are calculated and then the corresponding thermal stress intensity factor, KIT, for the y reference flaw is computed. From Equation (2) the pressure stress intensity l' factors are obtained and, from these, the allowable pressures are calculated.
COOLDOWN I
For the calculation of the allowable pressure versus coolant temperature I during cooldown, the Code reference flaw is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the wall because the thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates. Allowable pressure-temperature relations are generated for both steady-state and finite ,
cooldown rate situations. From these relations, composite limit curves are j
constructed for each cooldown rate of interest. s The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on measurement of reactor j i
coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the 1 1
B 3/4 4-11 l SOUTH TEXAS - UNIT 1 m.__-.- --_________-
, s DR\FT JUL 7 1987 REACTOR COOLANT SYSTEM l
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 coolant temperature, at the AT the 1/4T location developed during cooldown results in a higher value of KIR for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist such that the increase in KIR exceeds kit, the calculated allowable pressure during cooldown will be greater than the steady-state value.
The above procedures are needed because there is no direct control on temperature at the 1/4T location; therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and assures conservative operation of the system for the entire cooldown period.
HEATUP Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4T defect at the inside of the vessel wall. The thermal gradients during heatup i
produce compressive stresses at the inside of the wall that alhviate the tensile stresses produced by internal pressure. The metal temperature at the ,
crack tip lags the coolant temperature; therefore, the KIR for the 1/4T crack during heatup is lower than the Kyg for the 1/4T crack during steady-state conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist such that the effects of compressive thermal stresses and di,fferent Kyg's for steady-state and finite heatup rates state do not offset each other and the pressure-temperature curve based on steadyite conditions no longer represents a lower bound of all similar curves for fin heatup rates when the 1/4T flaw is consider.ed. Therefore, both cases have to be analyzed in order to assure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained.
The second portion of the heatup analysis concerns the calculation of pressure-temperature limitations for the case in which a 1/4T deep outside surface flaw is assumed.
Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and thus tend to reinforce any pressure stresses present. These thermal stresses, of course, are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp. Furthermore, since the thermal stresses at the outside are tensile and SOUTH TEXAS - UNIT 1 B 3/4 4-12
. c DRAFT REACTOR COOLANT SYSTEM JUL 7 1987 ,
BASES PRESSURE / TEMPERATURE LIMITS (Continued) increase with increasing heatup rate, a lower bound curve cannot be defined.
Rather, each 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 composite curve is constructed based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from 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 condition switches from the inside to the outside and the pressure limit must at all times be based on analysis of the most critical criterion.
Finally, the composite curves for the heatup rate data and the cooldown rate data are adjusted for possible errors in the pressure and temperature sensing instruments by the values indicated on the respective curves.
Although the pressurizer operates in temperature ranges above those for which there is reason for concern of nonductile failure, operating limits are provided to assure compatibility of operation with the fatigue analysis performed in accordance with the ASME Code requirements.
LOW TEMPERATURE OVERPRESSURE PROTECTION The OPERABILITY of two PORVs or an RCS vent opening of at least 2.0 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal 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 secondary water temperature of the steam generator less thari or equal to 50*F above the RCS cold leg temperatures, or (2) the maximum credible mass injection flow rate due to the startup of a single HHSI pump plus 100 gpm net charging flow, while the RCS is in a water solid condition and the RCS temperature is between 350*F and 200 F.
For RCS temperatures less than 200 F, the maximum overpressure event con-sists of operating a centrifugal charging pump with complete termination of letdown and a failure of the charging flow control valve to the full flow ondition.
The Maximum Allowed PORY Setpoint for the Cold Overpressure Mitigation System (COMS) is derived by analysis which models the performance of the COMS assuming various mass input and heat input transients. Operation with a PORV Setpoint less than or equal to the maximum Setpoint ensures that Appendix G 1 criteria will not be violated with consideration for a maximum pressure SOUTH TEXAS - UNIT 1 B 3/4 4-13
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DRAFT REACTOR COOLANT SYSTEM g 7g l
BASES LOW TEMPERATURE OVERPRESSURE PROTECTION (Continued) overshoot beyond the PORV Setpoint which can occur as a result of time delays :
in signal processing and valve opening, instrument uncertainties, and single '
failure. To ensure that mass and heat input transients more severe than those assumed cannot occur, Technical Specifications require lockout of all high head !
safety injection pumps while in MODE 5 and MCDE 6 with the reactor vessel head on. All but one high head safety injection pump are required to be locked out in MODE 4. Technical Specifications also require lockout of all but one charging pump while in MODES 4, 5, and 6 with the reactor vessel head installed and disallow start of an RCP if secondary temperature is more than 50'F above primary temperature.
The Maximum Allowed PORV Setpoint for the COMS will be updated based on the results of examinations of reactor vessel material irradiation surveillance specimens performed as required by 10 CFR Part 50, Appendix H, and in accordance l with the schedule in Table 4.4-5.
3/4.4.10 STRUCTURAL INTEGRITY The inservice inspection and testing programs for ASME Code Class 1, 2, and 3 components ensure that the structural integrity and operational readiness of these components will be maintained at an acceptable level throughout the
- life of the plant. These programs are in accordance with Section XI of the ASME Boilcr and Pressure Vessel Code and applicable Addenda as required by 4 10 CFR 50.55a(g) except where specific written relief has been granted by l the Commission pursuant to 10 CFR 50.55a(g)(6)(1). l i
Components of the Reactor Coolant System were designed to provide access l
to permit inservice inspections in accordance with Section XI of the ASME Boiler and Pressure Vessel Code,1974 Edition and Addenda through Winter 1975.
3/4.4.11 REACTOR VESSEL HEAD VENTS Reactor vessel head vents are provided to exhaust noncondensible gases and/or steam from the Reactor Coolant System that could inhibit natural circulation core cooling. The OPERABILITY of at least two reactor vessel head vent paths ensures that the capability exists to perform this function.
The valve redundancy of the reactor vessel head vent paths serves to mini-mize the probability of inadvertent or irreversible actuation while ensuring that a single failure of a vent valve, power supply, or control system does not prevent isolation of the vent path.
The function, capabilities, and testing requirements of the reactor vessel head vents are consistent with the requirements of Item II.B.1 of NUREG-0737,
" Clarification of TMI Action Plan Requirements," November 1980.
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. o DRAFT 3/4.5 EMERGENCY CORE COOLING SYSTEMS JUL 7 1987 BASES 3/4. 5.1 ACCUMULATORS The OPERABILITY 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.
The limits on accumulator volume represent a spread about an average value used in the safety analysis and have been demonstrated by sensitivity studies to vary the peak clad temperature by less than 20*F. The limit on accumulator pressure ensures that the assumptions used for accumulator injec-tion in the safety analysis are met.
The accumulator power operated isolation valves are considered to be
" operating bypasses" in the context of IEEE Std. 279-1971, which requires that bypasses of a protective function be removed automatically whenever permissive conditions are not met. In addition, as these accumulator isolation valves I
fail to meet single failure criteria, removal of power to the valves is required.
l The limits for operation with an accumulator inoperable for any reason except an isolation valve closed minimizes the time exposure of the plant to a LOCA event occurring concurrent with failure of an additional accumulator which may result in unacceptable peak cladding temperatures. If a closed j
isolation valve cannot be opened within one hour, the full capability of one l accumulator is not available and prompt action is required to place the reactor in a mode where this capability is not required. -
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3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS !
The OPERABILITY of three independent ECCS subsystems ensures that sufficient emergency core cooling capability will be available in the event of a LOCA ,
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assuming the loss of one subsystem through any single failure consideration.
Each subsystem operating in conjunction with the accumulators is capable of l
l supplying sufficient core cooling to limit the peak cladding temperatures l i
' within acceptable limits for all postulated break sizes ranging from the double ended break of the largest RCS cold leg pipe downward. One ECCS is !
l assumed to discharge completely through the postulated break in the RCS loop.
Note Thus, three trains are required to satisfy the single failure criterion.
that the centrifugal charging pumps are not part of ECCS and that the RHR pumps ;
are not used in the injection phase of the ECCS. Each ECCS subsystem and the RHR :
pumps and heat exchanges provide long-term core cooling capability in the !
recirculation mode during the accident recovery period.
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When the RCS temperature is below 350 F, the ECCS requirements are balanced !
between the limitations imposed by the low temperature overpressure protection l
[
' and the requirements necessary to mitigate the consequences of a LOCA below !
350'F. At these temperatures, single failure considerations are not required because of the stable reactivity condition of the reactor and the limited core l cooling requirements. Only a single Low Head Safety Injectior, pump is required to mitigate the effects of a large-break LOCA in this mode. However, two are SOUTH TEXAS - UNIT 1 B 3/4 5-1 l __ -
. c DRAFT j EMERGENCY CORE COOLING SYSTEMS JUL 7 1987 BASES ECCS SUBSYSTEMS (Continued) provided to accommodate the possibility that the break occurs in a loop con-taining one of the Low Head pumps. Low Head Safety Injection pumps are not required inoperable below 350 F because their shutoff head 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 Injection pumps are locked out to prevent cold overpressure.
The Surveillance Requirements provided to ensure OPERABILITY of each l
component ensure that, at a minimum, the assumptions used in the safety analyses 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.
3/4. 5. 4 (Not used) j 3/4.5.5 REFUELING WATER STORAGE TANK The OPERABILITY of the refueling water storage tank (RWST) as part of the ECCS ensures that a sufficient supply of borated water is available for injection by the ECCS in the event of a LOCA or a steamline break. The limits on RWST minimum volume and boron concentration ensure that: (1) sufficient water is available within containment to permit recirculation cooling flow to the core, (2) the reactor will remain subcritical in the cold condition (68'F to 212 F) following a small break LOCA assuming complete mixing of the RWST, RCS, Spray Additive Tank, Containment' Spray System and ECCS water volumes with all con-trol rods inserted except the most reactive control rod assembly (ARI-1),
(3) the reactor will, remain subcritical in cnid condition following a large break LOCA (break flow area > 3.0 ft )2assuraing complete mixing of the RWST, RCS, Spray Additive Tank, Containment Spray System and ECCS water volumes and other sources of wqter that may eventually reside in the sump post-LOCA with all control rods assumed to be out (ARO), and (4) long term subcriticality following a steamline break assuming ARI-1 and preclude fuel failure.
The maximum allowable value for the RWST boron concentration forms the basis for determining the time (post-LOCA) at which operator action is required to switch over the ECCS to hot leg recirculation in order to avoid precipita-tion of the soluble boron.
The contained water volume limit includes an allowance for water not usable because of t' ink discharge line location or other physical characteristics.
The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between 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 corrosion on mechanical systems and components.
SOUTH TEXAS - UNIT 1 B 3/4 5-2
. o DRAFT Jyt 7 jgg7 EMERGENCY CORE COOLING SYSTEMS l l
BASES l
3/4. 5. 6 RESIDUAL HEAT REMOVAL (RHR) SYSTEM l The OPERABILITY of the RHR system ensures adequate heat removal capabili-1 ties for Long-Term Core Cooling in the event of a small-break loss-of-coolant accident (LOCA), 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 j loop is available for cooling including single active failure criteria.
The surveillance ensure that RHR system isolation valves close upon an l
overpressure protection system signal. l l
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SOUTH TEXAS - UNIT 1 B 3/4 5-3 )
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DRJT l CONTAINMENT SYSTEMS JUL 7 1987 MSES 3/4.6.1.5 AIR TEMPERATURE I The limitations on containment average air temperature ensure that the over-all containment average air temperature does not exceed the initial temperature I condition assumed in the safety analysis for a LOCA or steam line break l accident. Measurements shall be made by fixed instruments, prior to determin-ing the average air temperature.
l 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 containment will with-stand the maximum pressure of 37.5 psig in the event of a LOCA or steam line l 1
break accident. The measurement of containment tendon lift-off force, the tensile tests of the tendon wires, the visual examination of tendons, anchorages and exposed interior and exterior surfaces of the containment, and the Type A leakage test are sufficient to demonstrate this capability.
The Surveillance Requirements for demonstrating the containment's structural integrity are in compliance with the recommendations of proposed Regulatory Guide 1.35, " Inservice Surveillance of Ungrouted Tendons in Prestressed Concrete Containment Structures," April 1979, and proposed Regulatory Guide 1.35.1, " Deter-mining Prestressing Forces for Inspection of Prestressed Concrete Containments,"
April 1979.
The required Special Reports from any engineering evaluation of containment abnormalities shall include a description of the tendon condition, the condition of the concrete (especially at tendon anchorages), the inspection procedures, the t tolerances on cracking, the results of the engineering evaluation, and the correc- l tive actions taken.
3 /4. 6.1. 7 CONTAINMENT VENTILATION SYSTEM The 48-inch containment purge supply and exhaust isolation valves are required to be sealed closed during plant operations since these valves have not been demonstrated capable of closing duting a LOCA or steam line break accident. '
Maintaining these valves sealed closed during plant operation ensures that exces-sive quantities of radioactive materials will not be released via the Containment Purge System. To provide assurance that these containment valves cannot be inad-vertently opened, the valves are sealed closed in accordance with Standard Review Plan 6.2.4 which includes mechanical devices to seal or lock the valve closed, or prevents power 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 during a LOCA or steam line break accident. There-r SOUTH TEXAS - UNIT 1 B 3/4 6-2
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- 4 DRTI 3/4.7 PLANT SYSTEMS g 7g BASES 3/4.7.1 TUR81NE CYCLE 3/4.7.1.1 SAFETY VALVES The OPERABILITY of the main steam line Code safety 110% valves(1413.5 ensurespsig) that of the Secondary System pressure will be limited to within its design pressure of 1285 psig during the most severe anticipated system operational transient. The maximum relieving capacity is associated with a Turbine trip from 100% RATED THERMAL POWER coincident with an assumed loss of condenser heat sink (i.e. , no steam bypass to the condenser).
l The specified valve lift settings and relieving capacities are in accordance with the requirements of Section III of the ASE Boiler and Pressure Code, 1971 Edition. The total relieving capacity for all valves on all of the l
steam lines is 20.65 x 106 lbs/h which is 122% of the total secondary A minima steam flow of two OPERABLE safety of 16.94 x IOS lbs/h at 100% RATED THERMAL POWER.
valves per steam generator ensures that sufficient relieving capacity is available for the allowable THERMAL POWER restriction in Table 3.7-1.
l STARTUP and/or POWLR OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis of the reduction in Secondary Coolant System steam flow and THERMAL POWER required by theThe reduced Reactor trip settings of the Power Range Neutron Flux channels.
Reactor Trip Setpoint reductions are derived on the following bases:
SP = (X) - (Y)(V) x (109)
Where:
SP = Reduced Reactor Trip Setpoint in percent of RATED THERMAL POWER, V = Maximum number of inoperable safety valves per steam line, 109 = Power Range Neutron Flux-High Trip Setpoint for four loop operation,
)
X = Total relieving capacity of all safety valves per steam line in 1bs/ hour, and Y = Maximum relieving capacity of any one safety valve in 1bs/ hour l
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SOUTH TEXAS - UNIT 1 B 3/4 7-1
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DRAFT g 7 ;gg PLANT SYSTEM _5 MSES 3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM The OPERABILITY of the Auxiliary Feedwater System ensures that the Reactor Coolant System can be cooled down to less than 350 F from normal operating conditions in the event of a total loss-of-offsite power.
Each' auxiliary feedwater pump is capable of delivering a total feedwater flow of 540 gpm at a pressure of 1324 psig to the entrance of the steam generators. This capacity is sufficient to ensure that adt.quate feedwater flow is available to remove decay heat and reduce the Reactor Coolant System temperature to less than 350*F when the Residual Heat Removal System may be The AFW pumps are tested using the test line back to placed into operation.the AFST and the AFW isolation valves closed to prevent injec 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 valves to be verified in the correct position.
3/4.7.1.3 AUXILIARY FEE 0 WATER 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 HOT STANDBY conditions for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> with steam discharge to the atmosphere 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 characteristics.
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 effects of a coincident 1 gpm primary-to-secondary tube leak in the steam generator of the affected steam line. These values are consistent with the assumptions used in the safety analyses.
3/4.7.1.5 MAIN STEAM LINE ISOLATION VALVES The OPERABILITY of the main steam line isolation valves ensures that no more than one steam generator will blow down in the event of a steam line rupture. This restriction is required to: (1) minimize the positive reac-tivity effects of the Reactor Coolant System cooldown associated with the blowdown, and (2) limit the pressure rise within containment in the event the steam line rupture occurs within containment. The OPERABILITY of the main steam isolation valves within the closure times of the Surveillance Require-ments are consistent with the assumptions used in the safety analyses.
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SOUTH TEXAS - UNIT 1 B 3/4 7-2
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- 1 l DRAFT PLANT SYSTEMS _ JUL 7 1987 BASES SNUBBERS (Continued) i associated installation and maintenance records (newly installed snubbers, seal l replaced, spring replaced, in high radiation area, in high temperature area, etc.). The requirement to monitor the snubber service life is included to ensure that the snubbers 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 from Byproduct, Source, and Special Nuclear Material sources will not exceed allowable intake values.
Sealed sources are classified into three groups according to their use, with Surveillance Requirements commensurate with the probability of damage to a source in that group. Those sources which are frequently handled are required to be tested more often than those which are not. Sealed sources which are continuously enclosed within a shielded mechanism (i.e. , sealed sources within radiation monitoring or boron measuring devices) are considered to be stored and need not be tested unless they are removed from the shielded mechanism.
3/4.7.11 (Not Used) 3/4.7.12 (Not Used) l 3/4.7.13 AREA TEMPERATURE MONITORING :
The area temperature limitations ensure that safety-related equipment will {
not be subjected to temperatures in excess of their environmental qualification temperatures. Exposure to excessive temperatures may degrade equipment and can ,
cause a loss of its OPERABILITY. The temperature limits include an allowance for l
instrument error of i 3 F maximum.
3/4.7.14 ESSENTIAL CHILLED WATER SYSTEM The OPERABILITY of the Essential Chilled Water Systes ensures that suffi-cient cooling capacity is available for continued operation of safety-related equipment during normal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the safety analyses.
SOUTH TEXAS - UNIT 1 8 3/4 7-6
1 1
- 4 l DRAFI JUL 7 1987 REFUELING OPERATIONS 1
BASES 3/4.9.6 REFUELING MACHINE l
The OPERABILITY requirements for the refueling machine and auxiliary l hoist ensure that: (1) the refueling machine and auxiliary hoist will be used j for movement of drive rods and fuel assemblies, (2) the refueling machine has l sufficient load capacity to lift'a drive rod or fuel assecbly, and (3) the core internals and reactor vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations.
l 3/4.9.7 CRANE TRAVEL - FUEL HANDLING BUILDING _
The restriction on movement of loads in excess of the nominal weight of a fuel and control rod assembly and associated handling tool over other fuel assemblies in the storage pool, unless handled by the single-failure proof main hoist of the FHB 15-ton crane, ensures that in the event this load is dropped:
(1) the activity release will be limited to that contained in a single fuel
' assembly, and (2) any possible distortion of fuel in the storage racks will not result in a critical array. This assumption is consistent with the activity release assumed in the safety analyses.
3/4.9.8 RESIOUAL HEAT REMOVAL AND COOLANT CIRCULATION The requirement that at least one residual heat removal (RHR) loop be in operation ensures that: (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor vessel below 140'F as required j during the REFUELING MODE, and (2) sufficient coolant circulation is maintained l through the core to minimize the effect of a boron dilution incident and prevent l l boron stratification.
The requirement to have two RHR loops OPERABLE when there is less than f 23 feet of water above the reactor vessel flange ensures that a single failure l
of the operating RHR loop will not result in a complete loss of residual heat removal capability. With the reactor vessel head removed and at least 23 feet of water above the reactor pressure vessel flange, a large heat sink is avail-able for core cooling. Thus, in the event of a failure of the operating RHR loop, adequate time is provided to initiate emergency procedures to cool the Core.
3/4. 9.9 CONTAINMENT VENTILATION ISOLATION SYSTEM The OPERABILITY of this system ensures that the containment purge and exhaust penetrations will be automatically isolated upon detection of high radiation levels in the purge exhaust. The OPERABILITY of this system is l
required to restrict the release of radioactive material from the containment atmosphere to the environment.
SOUTH TEXAS - UNIT 1 B 3/4 9-2
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- 9 DRAFT REFUELING OPERATIONS JUL 71%7 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 depth is available to remove 99% of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly. The minimum water depth is consistent with the assumptions of the safety analysis.
3/4.9.12 FUEL HANDLING BUILDING EXHAUST AIR SYSTEM The limitations on the Fuel Handling Building Exhaust Air System ensure l
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 moisture on the adsorbers and HEPA filters. The OPERABILITY of this ,
system and the resulting iodine removal capacity are consistent with the assump-tions of the safety analyses. ANSI N510-1980 will be used as a procedural guide for surveillance testing.
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S SOUTH TEXAS - UNIT 1 B 3/4 9-3
. !b DRTI RADI0 ACTIVE EFFLUENTS g 7 ggg7 BASES ,
This specification applies to the release of radioactive materials in gaseous effluents from each unit at the site.
3/4.11.2.3 DOSE - IODINE-131,10 DINE-133, TRITIUM, AND RADI0 ACTIVE MATERIAL IN PARTICULATE FORM 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 Operaticn are the guides set forth in Section II.C of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV. A of Appendix I to assure that the releases of radioactive materials in gaseous effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." The ODCM calculational
' methods specified in the Surveillance Requirements implement the requirements in Section III. A of Appendix I that conformance with the guides of Appendix I be shosn by calculational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The 00CM calculational methodology arx1 parameters for calculating the doses due to the actual release rates of the subject materials are consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose.of Evaluating Compliance with 10 CFR Part 50, Appendix 1," Revision 1, October 1977 and Regulatory Guide 1.111. " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1, July 1977. These equa-
' tions also provide for determining the actual doses based upon the historical average atmospheric conditions. The release rate specifications for Iodine-131, Iodine-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days are dependent upon the existing radionuclides pathways to man, in the areas at and beyond the SITE BOUNDARY. The pathways that were examined in the development of these calculations were: (1) individual inhalation of air-borne radionuclides, (2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, (3) deposition onto grassy areas where milk animals and meat producing animals graze with consumption of the milk and meat by man, and (4) deposition on the ground with subsequent exposure to man.
This specification applies to the release of radioactive materials in gaseous effluents from each unit at the site.
3/4.11.2.4 GASEOUS WASTE PROCESSING SYSTEM The OPERABILITY of the GASE0US WASTE PROCESSING SYSTEM ensures that the systems will be available for use whenever gaseous effluents require treatment l
prior to release to the environment. The requirement that the appropriate por-tions of these systems be used, when specified, provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable". This specification implements the re-quirements of 10 CFR 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objectives given in Section 11.0 of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions SOUTH TEXAS - UNIT 1 B 3/411-4
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DRAFT JUL 7 1987 i l
ADMINISTRATIVE CDNTROLS 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 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 familiarization with relevant industry operational experience.
6.5 REVIEW AND AUDIT 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:
Technical Services Manager Member: Plant Operations Manager Member: Plant Engineering Manager Member: Maintenance Manager Member: Operations QA Manager The PORC Chairman shall be appointed in writing from among these members by l the Plant Manager. .
l l ALTERNATES
- 6. 5.1. 3 All alternate members shall be appointed in writing by the Plant Manager to serve on a temporary basis; however, no more than two alternates i
shall participate as voting members in PORC activities at any one time.
MEETING FREQUENCY 6.5.1.4 The PORC shall meet at least once,per calendar month and as convened by the PORC Chairman or his designated alternate.
OVORUM
- 6. 5.1. 5 The quorum of the PORC necessary for the performance of the PORC responsibility and authority provisions of these Technical Specifications shall consist of the Chairman or his designated alternate, the Operations QA Manager or his designated alternate, and two other members including alternates.
l l
i SOUTH TEXAS - UNIT 1 6-7
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