ML20206N387
| ML20206N387 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 08/22/1986 |
| From: | Noonan V Office of Nuclear Reactor Regulation |
| To: | Little W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| References | |
| NUDOCS 8608260245 | |
| Download: ML20206N387 (512) | |
Text
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f, urog'o UNITED STATES g
!" o NUCLEAR REGULATORY COMMISSION
{ q WASHINGTON, D. C. 20655
\ , , , , . *# M]fi 2 2 O MEMORANDUM FOR: William S. Little, Chief Projects Cranch #3 Division of Reactor Project, Region III FROM: Vincent S. Noonan, Director PWR Project Directorate #5 Division of PWR licensing-A
SUBJECT:
FINAL DRAFT TECHNICAL SPECIFICATIONS FOR BRAIDWOOD STATION, UNITS 1 AND 2 Enclosed are the final draft Technical Specifications (TS) for Braidwood Station, Units 1 and 2, By letter dated June 10, 1986, we issued the Proof and Review Copy of the Braidwood Station, Units 1 and 2, Technical Specifications. Only the following four changes have been made to the Proof and Review Copy: (1) added RPR pump curve on pa corrected Specification c.1) on page 3/4 6-3, (ges 3/4Specification
- 3) dropped 5-5 and 5-6a, (2)
. 3.8.2.lc on page 3/4 8-10, and (d) corrected addressee in Specification 6.9.1.9 on page 6-21.
A few comments remain outstanding and the staff is awaiting the applicant response to those items. The items include: 1) valve listings in Technical Specification Table 3.8-2a and 3.8-2b, 2) verification of the Containment leakage Specification value for l by preoperational test, 3) and updating theorganizationalchartsinSection6.
In addition, two other issues remain outstanding regarding high energy line breaks and seismic monitoring instrumentation. The applicant and staff are continuing to discuss these issues to reach resolution.
Except as they may be modified by the resolution of remaining issues discussed above, the Technical Specifications, to be issued a_s Appendix A to the license for Braidwood Station, Units 1 and 2 are expected to be essentially identical to the enclosed final draft. The applicant has been requested to certify under oath and affirmation that the enclosed Technical Specifications are consistent with the Final Safety Analysis Report, the Safety Evaluation Report, and the as-built facility. The applicant has also been requested to document in some detail the resources and process used to review and certify the Technical Specifications. NRC Region III should also review and comment on the enclosed Technical Specifications. These responses must be provided by September 16, 1986.
8603260245 860822 PDR A ADOCK 05000456 PDR
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7" I" If you have any questions or if we can be of further assistance, please contact the Braidwood Project Manager, J. A. Stevens, at (301) 492-7702.
, ,J/ --
set S. na , Director PWR Projec Direc orate #5 Division of PWR licensing-A
Enclosure:
Final Draft of Braidwood Technical Specifications cc: See next page
qa .
s Mr. Dennis L. Farrar Braidwood Station l Commonwealth Edison Company Units I and 2 l
cc:v/o enclosures ifr. William Kortier Ms. Lorraine Creek l.tomic Power Distribution Route 1 Box 182 Westinghouse Electric Corporation Manteno. Illinois 60950 Post Office Box 355 Pittsburgh, Pennsylvania 15230 Douglass Cassel, Esq.
109 N. Dearborn Street Joseph Gallo, tsq. Chicago, Illinois 60602 Isham, Lincoln & Beale 1150 Connecticut Ave., N. W. Elena 7. Kezelis, Esq.
Sulte 1100 Isham, Lincoln & Beale Washington, D. C. 20036 Three First National Plaza Suite 5200 C. Allen Bock, Esq. Chicago, Illinois 60602 Post Offices Box 342
, Urbana, Illinois 61801 Mr. Charles D. Jones, Director Illinois Emergency Services Thomas J. Gordon, Esq. and Disaster Agenc?
- Waaler, Evans & Gordon 110 East Adams Street 2503 S. Neil Springfield. Illinois 62706 Champaign, Illinois 61820 George L. Edgar Ms. Bridget little Rorem Newman & Poltzinger, P.C.
Appleseed Coordinator 1615 L Street, N.W.
117 North Linden Street Washington, D.C. 20036 Essex. Illinois 60935
. Michael Miller, Esq.
Mr. Edward R. Crass Isham, Lincoln & Beale Nuclear Safeguards and One first National Plaza licensing Division 42nd Floor Sargent & Lundy Engineers Chicago, Illinois 60603 55 East Monroe Street
. Chicago, Illinois 60603 U. S. Nuclear Regulatory Commission Resident Inspectors Office RR#1, Box 79 Braceville, Illinois 60407 Regional Administrator, Region III U. S. Nuclear Regulatory Commission 799 Roosevelt Road Glen Ellyn, Illinois 60137
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- N If you have any questions or if we can be of further assistance, please contact the Braidwood Pro. ject Manager, J. A. Stevens, at (301) 492-7702.
Vincent S. Noonan, Director PWR Pro. ject Directorate #5 Division of PWR l.icensing-A
Enclosure:
Final Draft of Braidwood Technical Specifications cc: See next page Distribution g Tugeset f1Tei+ '
NRC~PDR l.ocal PDR PD#5 R/F T. Novak OGC in Bethesda V. S. Noonan J. A. Stevens E. Jordan J. Partlow B. Grfres ARCS (10)
M. Rushbrook.
PD#5 DIR:PD#5 JASte ns:ss VSNoona 8/22/86 84,1/86
FIEL DRAFT AUG 191986 TECHNICAL SPECIFICATIONS FOR BRAIDWOOD STATION UNIT NO. 1 & NO. 2 DOCKET NO. STN 50-456 & 50-457 s
G 9
9 FINAL DRAFT AUG 19 gg G
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INDEX l
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DEFINITIONS FINAL AUg 19 ggg DRAFT SECTIDN PAGE
- 1. 0 DEFINITIONS 1.1 ACTI0N........................................................ 1-1 1.2 ACTUATION LOGIC TEST.......................................... 1-1 1.3 ANALOG CHANNEL OPERATIONAL TEST............................... 1-1 1.4 AXIAL FLUX DIFFERENCE......................................... 1-1
- 1. 5 CHANNEL CALIBRATION........................................... 1-1 1.6 CHANNEL CHECK................................................. 1-1
- 1. 7 CONTAINMENT INTEGRITY......................................... 1-2 1.8 CONTROLLED LEAKAGE............................................ 1-2
- 1. 9 CORE ALTERATION............................................... 1-2 1.10 OIGITAL CHANNEL OPERATIONAL TEST............................. 1-2 1.11 DOSE EQUIVALENT I-131........................................ 1-2 1.12 E-AVERAGE DISINTEGRATION ENERGY.............................. 1-3 1.13 ENGINEERED SAFETY FEATURES RESPONSE TIME..................... 1-3 1.14 FREQUENCY N0TATION........................................... 1-3 1.15 IDENTIFIED LEAKAGE........................................... 1-3 1.16 MASTER RELAY TEST............................................ 1-3 1.17 MEMBER (S) 0F THE PUBLIC...................................... 1-3 1.18 0FFSITE DOSE CALCULATION MANUAL. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4 i 1.19 OPERABLE - OPERABILITY....................................... 1-4 1.20 OPERATIONAL MODE - M0DE...................................... 1-4 1.21 PHYSICS TESTS................................................ 1-4 1.22 PRESSURE BOUNDARY LEAKAGE.................................... 1-4 1.23 PROCESS CONTROL PR0 GRAM...................................... 1-4 1.24 PURGE - PURGING.............................................. 1-5 1.25 QUADRANT POWER TILT RATI0.................................... 1-5 1.26 RATED THERMAL P0WER.......................................... 1-5 1.2 REACTOR TRIP SYSTEM RESPONSE TIME............................ 1-5 1.2l1 REPORTABLE EVENT............................................. 1-5 1.29 SHUTDOWN MARGIN.............................................. 1-5 1.30 SITE B0VNDARY................................................ 1-5 1.31 S LAV E R E LAY T E ST . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5 i
l BRAIDWOOD - UNITS 1 & 2 I
DEFINITIONS FINAL DRAFT AUG 19 m SECTION PAGE 1.32 SOLIDIFICATION............................................... 1-6 1.33 SOURCE CHECK................................................. 1-6 1.34 STAGGERED TEST BASIS......................................... 1-6 1.35 THERMAL P0WER................................................ 1-6 1.36 TRIP ACTUATING DEVICE OPERATIONAL TEST....................... 1-6 1.37 UNIDENTIFIED LEAKAGE......................................... 1-6 1.38 UNRESTRICTED AREA............................................ 1-6 1.39 VENTILATION EXHAUST TREATMENT SYSTEM......................... 1-7 1.40 VENTING...................................................... 1-7 1.41 WASTE GAS HOLDUP SYSTEM...................................... 1-7 TABLE 1.1 FREQUENCY N0TATION...................................... 1-8 TABLE 1.2 OPERATIONAL M0 DES....................................... 1-9 BRAIDWOOD - UNITS 1 & 2 II
SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SECTION PAGE 2.1 SAFETY LIMITS 2.1.1 REACTOR C0RE................................................ 2-1 2.1.2 REACTOR COO LANT SYSTEM PRESSURE. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-1 FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION.. 2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETP0INTS............... 2-3 TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS.... 2-4 BASES SECTION PAGE 2.1 SAFETY LIMITS 2.1.1 REACTOR C0RE................................................ B 2-1 2.1.2 REACTOR COO LANT SYSTEM PRESSURE. . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 2-2
- 2. 2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETP0INTS............... B 2-3 l
, BRAIDWOOD - UNITS 1 & 2 III 1
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FINAL HAFT LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS E 10 E SECTION PAGE i
3/4.0 APPLICABILITY............................................... 3/4 0-1 3/4.1 REACTIVITY CONTROL .[" STEMS 3/4.1.1 B0 RATION CONTROL Shutdown Margin - T,yg > 200*F........................... 3/4 1-1 Shutdown Margin - T,yg <_ 200*F........................... 3/4 1-3 Moderator Temperature Coefficient........................ 3/4 1-4 Minimum Temperature for Criticality...................... 3/4 1-6 3/4.1.2 BORATION SYSTEMS Flow Path - Shutdown..................................... 3/4 1-7 Flow Paths - Operating................................... 3/4 1-8 Ch.arging Pump - Shutdown................................. 3/4 1-9 Charging Pumps - 0perating............................... 3/4 1-10 Borated Water Source - Shutdown.......................... 3/4 1-11 Borated Water Sources - Operating........................ 3/4 1-12 3/4.1.3 MOVABLE CONTROL ASSEMBLIES Group Height............................................. 3/4 1-14 TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT OF AN INOPERABLE FULL-LENGTH R0D.............. 3/4 1-16 Position Indication Systems - Operating.................. 3/4 1-17 Position Indication System - Shutdown.................... 3/4 1-18 Rod Drop Time............................................ 3/4 1-19 Shutdown Rod Insertion Limit............................. 3/4 1-20 Control Rod Insertion Limits..... ....................... 3/4 1-21 FIGURE 3.1-1 R0D BANK INSERTION LIMITS VERSUS THERMAL POWER FOUR LOOP 0PERATION........................... 3/4 1-22 l
1 BRAIDWOOD - UNITS 1 & 2 IV l
FINAL MAFT LIMITING CONDITIONS FOR OPERATI'N O AND SURVEILLANCE REQUIREMENTS AUS 191986 SECTION PAGE 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE.................................... 3/4 2-1 FIGURE 3.2-1 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL P0WER................................ 3/4 2-3 3/4.2.2 HEAT FLUX HOT CHANNEL FACT 0R............................. 3/4 2-4 FIGURE 3.2-2 K(Z)-NORMALIZED Fq (Z) AS A FUNCTION OF CORE HEIGHT... 3/4 2-5 3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACT 0R................................................. 3/4 2-8 3/4.2.4 QUADRANT POWER TILT RATI0................................ 3/4 2-10 3/4.2.5 DNB PARAMETERS........................................... 3/4 2-13 TABLE 3.2-1 DNB PARAMETERS........................................ 3/4 2-14 3/4.3 INSTRL' MENTATION 3/4.3.1 REACTOR TRIP SYSTEM I.'!STRUMENTATI0N. . . . . . . . . . . . . . . . . . . . . . 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 3-7 TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS........................................ 3/4 3-9 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION........................................ 3/4 3-13 TABLE 3.3-3 ENGINEER?.D SAFETY FEATURES ACTUATION SYSTEM i
l INSTRUMENTATION..................................... 3/4 3-15 TA.BLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETP0INTS...................... 3/4 3-23 TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES............. 3/4 3-30 TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS........... 3/4 3-34 BRAIDWOOD - UNITS 1 & 2 V
LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS FINAL HAFT Alin 19 iTE SECTION PAGE 3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitoring for Plant Operations................ 3/4 3-39 TABLE 3.3-6 RADIATION MONITORING INSTRUMENTATION FO R P LANT O P ERATIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 3-40 TABLE 4.3-3 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS SURVEILLANCE REQUIREMENTS........................................ 3/4 3-42 Movable Incore Detectors................................. 3/4 3-43 Seismic Instrumentation.................................. 3/4 3-4'4 TABLE 3.3-7 SEISMIC MONITORING INSTRUMENTATION.................... 3/4 3-45 Meteorological Instrumentation........................... 3/4 3-47 TABLE 3.3-8 METEOROLOGICAL MONITORING INSTRUMENTATION............. 3/4 3-48 TABLE 4.3-5 METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS........................... 3/4 3-49 Remote Shutdown Instrumentation.......................... 3/4 3-50 TABLE 3.3-9 REMOTE SHUTDOWN MONITORING INSTRUMENTATION............ 3/4 3-51 TABLE 4.3-6 REMOTE SHUTDOWN MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS........................... 3/4 3-52 Accident Monitoring Instrumentation...................... 3/4 3-53 TABLE 3.3-10 ACCIDENT MONITORING INSTRUMENTATION.................. 3/4 3-54 TABLE 4.3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS........................... 3/4 3-55 Fire Detection Instrumentation........................... 3/4 3-56 TABLE 3.3-11a FIRE DETECTION INSTRUMENTS (UNIT 1)................. 3/4 3-58 TABLE 3.3-11b FIRE DETECTION INSTRUMENTS (UNIT 2)................. 3/4 3-61 l
l Loose-Part Detection System.............................. 3/4 3-64 Radioactive Liquid Effluent Monitoring Instrumentation... 3/4 3-65 TABLE 3.3-12 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION.................................... 3/4 3-66 BRAIDWOOD - UNITS 1 & 2 VI
FINAL HAFT LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS AUG 1 SECTION PAGE TABLE 4.3-8 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.......... 3/4 3-68 Radioactive Gaseous Effluent Monitoring Instrumentation. 3/4 3-70 TABLE 3.3-13 RADI0 ACTIVE GASE0US EFFLUENT MONITORING
{
INSTRUMENTATION.................................... 3/4 3-71 TABLE 4.3-9 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCI REQUIREMENTS........... 3/4 3-75 i Chlorine Detection Systems............................... 3/4 3-79 3/4.3.4 TURBINE OVERSPEED PROTECTION............................. 3/4 3-80 -
3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION Startup and Power 0peration.............................. 3/4 4-1 Hot Standby.............................................. 3/4 4-2 Hot Shutdown............................................. 3/4 4-3 Cold Shutdown - Loops Filled............................. 3/4 4-5 Cold Shutdown - Loops Not Fi11ed......................... 3/4 4-6 Loop Isolation Valves-0peration.......................... 3/4 4-7 Loop Isolation Valves-Shutdown........................... 3/4 4-8 3/4.4.2 SAFETY VALVES Shutdown............................................... 3/4 4-9 0perating.............................................. 3/4 4-10 3/4.4.3 PRESSURIZER.............................................. 3/4 4-11 3/4.4.4 RELIEF VALVES............................................ 3/4 4-12 3/4.4.5 STEAM GENERATORS......................................... 3/4 4-13 TABLE 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION......................... 3/4 4-18 TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION....................... 3/4 4-19 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems................................ 3/4 4-20 Operational Leakage...................................... ~3/4 4-21 BRAIDWOOD - UNITS 1 & 2 VII
FINAL DRAFT LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE'UIREMENTS ,ming 3 o 377 SECTION PAGE 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 COOLANT SYSTEM CHEMISTRY LIMITS............... 3/4 4-25 TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY SURVEILLANCE REQUIREMENTS........................................ 3/4 4-26 3/4.4.8 SPECIFIC ACTIVITY........................................ 3/4 4-27 FIGURE 3.4-1 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL
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POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY
>l pCi/ GRAM DOSE EQUIVALENT I-131.................. 3/4 4-2E TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM.................................... 3/4 4-30 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System................................... 3/4 4-32 FIGURE 3.4-2a REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE UP TO 32 EFPY (UNIT 1)...... 3/4 4-33 FIGURE 3.4-2b REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE UP TO 16 EFPY (UNIT 2)...... 3/4 4-34 FIGURE 3.4-3a REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS APPLICABLE UP TO 32 EFPY (UNIT 1)...... 3/4 4-35 FIGURE 3.4-3b REACTOR COOLANT SYSTEM C00LD0VN LIMITATIONS APPLICABLE UP TO 16 EFPY (UNIT 2)...... 3/4 4-36 TABLE 4.4-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM - WITHDRAWAL SCHEDULE....................... 3/4 4-37 Pressurizer.............................................. 3/4 4-38 Overpressure Protection Systems.......................... 3/4 4-39 FIGURE 3.4-4 NOMINAL PORV PRESSURE RELIEF SETPOINT VERSUS RCS TEMPERATURE FOR THE COLD OVERPRESSURE PROTECTION SYSTEM APPLICABLE UP TO 10 EFPY......... 3/4 4-40 3/4.4.10 STRUCTURAL INTEGRITY..................................... 3/4 4-42 3/4.4.11 REACTOR COOLANT SYSTEM VENTS'............................. 3/4 4-43 BRAIDWOOD - UNITS 1 & 2 VIII
FINAL DRAT LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS n ,m . , , _ _ _
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SECTION PAGE 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS............................................. -
3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - T,yg > 350*F........................... 3/4 5-3 FIGURE 4.5-1 RESIDUAL HEAT REMOVAL PUMP MINIMUM ACCEPTABLE PERFORMANCE CURVE....................... 3/4 5-6a 3/4.5.3 ECCS SUBSYSTEMS - T,yg < 350*F........................... 3/4 5-7 3/4.5.4 REFUELING WATER STORAGE TANK............................. 3/4 5-9 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Containment Integrity.................................... 3/4 6-1 Containment Leakage...................................... 3/4 6-2 Containment Air Locks.................................... 3/4 6-4 Internal Pressure........................................ 3/4 6-6 Air Temperature.......................................... 3/4 6-7 Containment Vessel Structural Integrity.................. 3/4 6-8 Containment Purge Ventilation System..................... 3/4 6-11 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS Containment Spray System................................. 3/4 6-13 Spray Additive System.................................... 3/4 6-14 Containment Cooling System............................... 3/4 6-15 3/4.6.3 CONTAINMENT ISO LATION VALVES. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6-16 TABLE 3.6-1 CONTAINMENT ISOLATION VALVES.......................... 3/4 6-18 3/4.6.4 COMBUSTIBLE GAS CONTROL l
l Hydrogen Monitors........................................ 3/4 6-25 Electric Hydrogen Recombiners............................ 3/4 6-26 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE i l
Safety Va1ves............................................ 3/4 7-1 '
BRAIDWOOD - UNITS 1 & 2 IX
FINAL DRAFT LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS an ,
m..n . .n om moo ,
SECTION PAGE TABLE 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE t
STEAM LINE SAFETY VALVES DURING FOUR LOOP 0PERATION........................................... 3/4 7-2 TABLE 3.7-2 S1EAM LINE SAFETY VALVES PER L00P..................... 3/4 7-3 Auxil i ary Fee dwate r Sys tem. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-4 Condensate Storage Tank.................................. 3/4 7-6 Specific Activity........................................ 3/4 7-7 TABLE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM......................... 3/4 7-8 Main Steam Line Isolation Va1ves......................... 3/4 7-9 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION.......... 3/4 7-10 3/4.7.3 COMPONENT COOLING WATER SYSTEM...........................
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3/4 7-11 3/4.7.4 ESSENTIAL SERVICE WATER SYSTEM........................... 3/4 7-12 3/4.7.5 ULTIMATE HEAT SINK....................................... 3/4 7-13 3/4.7.6 CONTROL ROOM VENTI LATION SYSTEM. . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-14 3/4.7.7 NON-ACCESSIBLE AREA EXHAUST FILTER PLENUM VENTILATION SYSTEM....................................... 3/4 7-17 3/4.7.8 SNUBBERS................................................. 3/4 7-20 FIGURE 4.7-1 SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST........... 3/4 7-25 3/4.7.9 SEALED SOURCE CONTAMINATION.............................. 3/4 7-26 l
l l
BRAIDWOOD - UNITS 1 & 2 X l
FINAL DRAFT LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS atyg 3 o 7aae SECTION PAGE 3/4.7.10 FIRE SUPPRESSION SYSTEMS Fire Suppression Water Supply System..................... , 3/4 7-28 Water Systems............................................ 3/4 7-31 CO 2 Systems.............................................. 3/4 7-33 Halon Systems............................................ 3/4 7-34 Fire Hose Stations....................................... 3/4 7-35 TABLE 3.7-Sa FIRE HOSE STATIONS (UNIT 1).......................... 3/4 7-36 TABLE 3.7-5b FIRE HOSE STATIONS (UNIT 2).......................... 3/4 7-40 3/4/7.11 FIRE RATED ASSEMBLIES.................................... 3/4 7-45 3/4.7.12 AREA TEMPERATURE MONITORING.............................. 3/4 7-46 TABLE 3.7-6 AREA TEMPERATURE M0NITORING........................... 3/4 7-47 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES 0perating................................................ 3/4 8-1 TABLE 4.8-1 DIESEL GENERATOR TEST SCHEDULE........................ 3/4 8-8 Shutdown................................................. 3/4 8-9 0perating................................................ 3/4 8-9a-3/4.8.2 D.C. SOURCES 0perating................................................ 3/4 8-10 TABLE 4.8-2 BATTERY SURVEILLANCE REQUIREMENTS..................... 3/4 8-12 Shutdown................................................. 3/4 8-13 t 3/4.8.3 ONSITE POWER DISTRIBUTION 0perating................................................ 3/4 8-14 Shutdown................................................. 3/4 8-16 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES Containment Penetration Conductor Overcurrent Protective Devices..................................... 3/4 8-17 TABLE 3.8-la CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES (UNIT 1)............. ,3/4 8-19 i
TABLE 3.8-lb CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES (UNIT 2)............. 3/4 8-29 l
l BRAIDWOOD - UNITS 1 & 2 XI
FINAL HAFT LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS AUG 191996 SECTION PAGE Motor-0perated Valves Thermal Overload Protection 0evices................................................
3/4 8-39 TABLE 3.8-2a MOTOR OPERATED VALVES THERMAL OVERLOAD PROTECTION DEVICES (UNIT 1).................................... 3/4 8-40 TABLE 3.8-2b MOTOR OPERATED VALVES THERMAL OVERLOAD PROTECTION DEVICES (UNIT 2).................................... 3/4 8-44 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION...................................... 3/4 9-1 3/4.9.2 INSTRUMENTATION.......................................... 3/4 9-2 3/4.9.3 DECAY TIME............................................... 3/4 9-3 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS........................ 3/4 9-4 3/4.9.5 COMMUNICATIONS........................................... 3/4 9-6 3/4.9.6 REFUELING MACHINE........................................ 3/4 9-7 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE FACILITY............... 3/4 9-8 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION High Water Level......................................... 3/4 9-9 Low Water Leve1.......................................... 3/4 9-10 3/4.9.9 CONTAINMENT PURGE ISOLATION SYSTEM....................... 3/4 9-11 3/4.9.10 WATER LEVEL - REACTOR VESSEL............................. 3/4 9-12 3/4.9.11 WATER LEVEL - STORAGE P00L............................... 3/4 9-13 3/4.9.12 FUEL HANDLING BUILDING EXHAUST FILTER PLENUMS............ 3/4 9-14 1
BRAIDWOOD - UNITS 1 & 2 XII
FINAL HAFT LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS Aug 1 g yre 1
SECTION PAGE i
3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN.......................................... ' 3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS... 3/4 10-2 3/4.10.3 PHYSICS TESTS............................................ 3/4 10-3 3/4.10.4 REACTO R COO LANT L00PS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 10-4 3/4.10.5 POSITION INDICATION SYSTEM - SHUTD0WN.................... 3/4 10-5 3/4.11 RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS Concentration............................................ 3/4 11-1 TABLE 4.11-1 RADI0 ACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PR0 GRAM................................... 3/4 11-2 Dose..................................................... 3/4 11-6 Liquid Radwaste Treatment System......................... 3/4 11-7 Liquid Holdup Tanks...................................... 3/4 11-8 3/4.11.2 GASEOUS EFFLUENTS Dose Rate................................................ 3/4 11-9 TABLE 4.11-2 RADI0 ACTIVE GASE0US WASTE SAMPLING AND ANALYSIS PR0 GRAM................................... 3/4 11-10 Dose - Noble Gases....................................... 3/4 11-13 Dose - Iodine-131 and 133, Tritium, and Radioactive Material in Particulate Form........................... 3/4 11-14 Gaseous Radwaste Treatment System........................ 3/4 11-15 Explosive Gas Mixture.................................... 3/4 11-16 Gas Decay Tanks.......................................... 3/4 11-17 BRAIDWOOD - UNITS 1 & 2 XIII ;
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FINAL HAFT LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS AUG 19 M SECTION PAGE 3/4.11.3 SOLID RADIOACTIVE WASTES................................. 3/4 11-18 3/4.11.4 TOTAL 00SE............................................... , 3/4 11-19 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1' MONITORING PR0 GRAM....................................... 3/4 12-1 TABLE 3.12-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM........ 3/4 12-3 TABLE 3.12-2 REPORTING LEVELS FOR RADI0 ACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES........................... 3/4 12-9 TABLE 4.12-1 DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSIS.................................... 3/4 12-19 3/4.12.2 LAND USE CENSUS.......................................... 3/4 12-13 3/4.12.3 INTERLABORATORY COMPARISON PR0 GRAM....................... 3/4 12-14 BRAIDWOOD - UNITS 1 & 2 XIV
FINAL DRAFT BASES Al]G 191MR
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SECTION PAGE ;
3/4.0 APPLICABILITY............................................... B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 B0 RATION CONTR0L.......................................... B 3/4 1-1 3/4.1.2 BORATION SYSTEMS.......................................... B 3/4 1-2 3/4.1.3 MOVABLE CONTROL ASSEMBLIES................................ B 3/4 1-3 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE..................................... B 3/4 2-1 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR....... B 3/4 2-2 .
FIGURE B 3/4.2-1 TYPICAL INDICATED AXIAL FLUX DIFFERENCE VERSUS THERMAL P0WER.................................. B 3/4 2-3 3/4.2.4 QUADRANT POWER TILT RATI0................................. B 3/4 2-5 3/4.2.5 DNB PARAMETERS............................................ B 3/4 2-6 3/4.3 INSTRUMENTATION 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION............... B 3/4 3-1 3/4.3.3 MONITORING INSTRUMENTATION................................ B 3/4 3-3 3/4.3.4 TURBINE OVERSPEED PROTECTION.............................. B 3/4 3-7 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION. . . . . . . . . . . . . B 3/4 4-1 3/4.4.2 SAFETY VALVES............................................. B 3/4 4-2 3/4.4.3 PRESSURIZER............................................... B 3/4 4-2 3/4.4.4 RELIEF VALVES............................................. B 3/4 4-2 BRAIDWOOD - UNITS 1 & 2 XV
FINAL HAFT ;
BASES 1 a r m , o_ v.m SECTION PAGE l
3/4.4.5 STEAM GENERATORS.......................................... B 3/4 4-3 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE............................ B 3/4 4-4 3/4.4.7 CHEMISTRY.................................................' B 3/4 4-5 3/4.4.8 SPECIFIC ACTIVITY'........................................
B 3/4 4-5 3/4.4.9 PRESSURE / TEMPERATURE LIMITS............................... B 3/4 4-7
. TABLE B 3/4.4-la REACTOR VESSEL TOUGHNESS (UNIT 1)................ B 3/4 4-11 TABLE B 3/4.4-Ib REACTOR VESSEL TOUGHNESS (UNIT 2)................ B 3/4 4-12 FIGURE B 3/4.4-1 FAST NEUTRON FLUENCE (E>1MeV) AS A FUNCTION OF FULL POWER SERVICE LIFE........................ B 3/4 4-13 FIGURE B 3/4.4-2 EFFECT OF FLUENCE AND COPPER ON SHIFT OF RT NDT FOR REACTOR VESSEL STEELS EXPOSED TO IRRADIATION AT 550 F........................... B 3/4 4-14 3/4.4.10 STRUCTURAL INTEGRITY..............................:...... B 3/4 4-16 3/4.4.11 REACTOR VESSEL HEAD VENTS................................ B 3/4 4-17 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS.............................................. B 3/4 5-1 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS............................... B 3/4 5-1 3/4.5.4 REFUELING WATER STORAGE TANK.............................. B 3/4 5-2 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT....................................... B 3/4 6-1 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS...................... B 3/4 6-3 3/4.6.3 CONTAINMENT ISOLATION VALVES.............................. B 3/4 6-4 3/4.6.4 COMBUSTIBLE GAS CONTR0L................................... B 3/4 6-4 BRAIDWOOD - UNITS 1 & 2 XVI
BASES l
AUG 191986 i SECTION PAGE 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE............................................. ~ B 3/4 7-1 3/4.7/2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION........... B 3/4 7-3' 3/4.7.3 COMPONENT COOLING WATER SYSTEM............................ B 3/4 7-3 3/4.7.4 ESSENTIAL SERVICE WATER SYSTEM............................ B 3/4 7-3 3/4.7.5 ULTIMATE HEAT SINK........................................ B 3/4 7-3 3/4.7.6 CONTROL ROOM VENTILATION SYSTEM........................... B 3/4 7-4 3/4.7.7 NON-ACCESSIBLE AREA EXHAUST FILTER PLENUM VENTILATION SYSTEM ................................................. B 3/4 7-4 3/4.7.8 SNUBBERS.................................................. B 3/4 7-4 3/4.7.9 SEALED SOURCE CONTAMINATION............................... B 3/4 7-6 3/4.7.10 FIRE SUPPRESSION SYSTEMS...... ........................... B 3/4 7-6 3/4.7.11 FIRE RATED ASSEMBLIES .................................... B 3/4 7-7 3/4.7.12 AREA TEMPERATURE MONITORING............................... B 3/4 7-7 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1, 3/4.8.2 and 3/4.8.3 A.C. SOURCES, D.C. SOURCES, AND ONSITE POWER DISTRIBUTION............................... B 3/4 8-1 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES................... B 3/4 8-3 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 BRAIDWOOD - UNITS 1 & 2 XVII
FINAL RAFT BASES AUG 191986 SECTION PAGE 3/4.9.6 REFUELING MACHINE......................................... B 3/4 9-2 l 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE FACILITY................ ' B 3/4 9-2 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION............. B 3/4 9-2 3/4.9.9 CONTAINMENT PURGE ISOLATION SYSTEM........................ B 3/4 9-2 3/4.9.10 and 3/4.9.11 WATER LEVEL - REACTOR VESSEL and STORAGE P00L............................................ B 3/4 9-3 3/4.9.12 FUEL HANDLING BUILDING EXHAUST FILTER PLENUM SYSTEM....... B 3/4 9-3 3/4.10 -SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN........................................... B 3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS.... B 3/4 10-1 ,
3/4.10.3 PHYSICS TESTS............................................. B 3/4 10-1 3/4.10.4 REACTOR COOLANT L00PS..................................... B 3/4 10-1 3/4.10.5 POSITICN INDICATION SYSTEM - SHUTD0WN..................... B 3/4 10-1 3/4.11 RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS......................................... B 3/4 11-1 3/4.11.2 GASEOUS EFFLUENTS........................................ B 3/4 11-3 3/4.11.3 SOLID RADI0 ACTIVE WASTES................................. B 3/4 11-7 3/4.11.4 TOTA L 0 0 S E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 11-7 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PR0 GRAM....................................... B 3/4 12-1 3/4.12.2 LAND USE CENSUS.......................................... B 3/4 12-1 3/4.12.3 INTERLABORATORY COMPARIS0N PR0 GRAM....................... B 3/4 12-2 BRAIDWOOD - UNITS 1 & 2 XVIII
OESIGN FEATURES AUG 191986 SECTION PAGE 5.1 SITE 5.1.1 EXCLUSION AREA.............................................. 5-1 <
5.1.2 LOW POPULATION Z0NE......................................... 5-1 l
5.1.3 MAP DEFINING UNRESTRICTED AREAS AND SITE BOUNDARY FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS.............. 5-1 FIGURE 5.1-1 EXCLUSION AREA AND UNRESTRICTED AREA FOR RADI0 ACTIVE GASEOUS AND LIQUID EFFLUENTS............. 5-2 FIGURE 5.1-2 LOW POPULATION 20NE.................................. 5-3
- 5. 2 CONTAINMENT 5.2.1 CONFIGURATION............................................... 5-1 5.2.2 DESIGN PRESSURE AND TEMPERATURE.............................' 5-1 5.3 REACTOR CORE 5.3.1 FUEL ASSEMBLIES............................................. 5-4 5.3.2 CONTROL R0D ASSEMBLIES...................................... 5-4 5.4 REACTOR COOLANT SYSTEM 5.4.1 DESIGN PRESSURE AND TEMPERATURE............................. 5-4 5.4.2 V0LUME...................................................... 5-4 l
5.5 METEOROLOGICAL TOWER LOCATION................................. 5-4 l
5.6 FUEL STORAGE 5.6.1 CRITICALITY................................................. 5-5 5.6.2 DRAINAGE.................................................... 5-5 5.6.3 CAPACITY.................................................... 5-5 l
5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT........................... -
5-5 TABLE 5.7-1 COMPONENT CYCLIC OR TRANSIENT LIMIT................... 5-5 I
i BRAIDWOOD - UNITS 1 & 2 XIX
ADMINISTRATIVE CONTROLS FINAL HAFT AUG I 91996 SECTION PAGE 6.1 RESPONSIBILITY................................................ 6-1 6.2 ORGANIZATION.................................................. 6-1 FIGURE 6.2-1 0FFSITE ORGANIZATION................................. 6-3 FIGURE 6.2-2 UNIT ORGANIZATION.................................... 6-4 TABLE 6.2-1 MINIMUM SHIFT CREW COMPOSITION........................ 6-5 TABLE 6.2-la MINIMUM SHIFT CREW COMPOSITION....................... 6-Sa 6.3 UNIT STAFF QUALIFICATIONS..................................... 6-6 6.4 TRAINING...................................................... 6-6 6.5 REVIEW INVESTIGATION AND AUDIT................................ 6-6 6.5.1 0FFSITE..................................................... 6-7 Offsite Review and Investigative Function................... 6-7 Audit Function.............................................. 6-8 Authority................................................... 6-9 Records..................................................... 6-10 Procedures.................................................. 6-10 Personne1................................ .................. 6-10 6.5.2 0NSITE...................................................... 6-12 Onsite Review and Investigative Functions................... 6-12 Responsibility.............................................. 6-12 l Authority................................................... 6-13 Records..................................................... 6-14 Procedures.................................................. 6-14 Personne1................................................... 6-14 6.6 REPORTABLE EVENT ACTI0N....................................... . 6-15 BRAIDWOOD - UNITS 1 & 2 XX
ADMINISTRATIVE CONTROLS AUG 191996 SECTION PAGE 6.7 SAFETY LIMIT VIOLATION........................................ 6-15 6.8 PROCEDURES AND PR0 GRAMS....................................... 6-15 6.9 REPORTING REQUIREMENTS........................................ 6-18 6.9.1 ROUTINE REP 0RTS.........................................'.... 6-18 Startup Report.............................................. 6-18 Annual Reports.............................................. 6-18 Annual Radiological Environmental Operating Report.......... 6-19 Semiannual Radioactive Effluent Release Report.............. 6-20 Monthly Operating Report.................................... 6-21 Radial Peaking Factor Limit Report.......................... 6-21 6.9.2 SPECIAL REP 0RTS............................................. 6-22 6.10 RECORD RETENTION............................................. 6-22 6.11 RADIATION PROTECTION PR0 GRAM................................. 6-23 6.12 HIGH RADIATION AREA.......................................... 6-23 6.13 PROCESS CONTROL PROGRAM (PCP)................................ 6-25 6.14 0FFSITE DOSE CALCULATION MANUAL (0DCM)....................... 6-25 6.15 MAJOR CHANGE 5 TO LIQUID, GASEOUS, AND SOLID RADWASTE TREATMENT SYSTEMS........................................... 6-26 j l
BRAIDWOOD - UNITS 1 & 2 XXI
- -%A._ ma__ - .-2.'
FINAL DRAFT AUG 191996 G
e l
, SECTION 1.0 DEFINITIONS I
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- 1. 0 DEFINITIONS AUG 191996 The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications.
A_CTION 1.1 ACTION shall be that part of a Technical Specification which prescribes remedial measures required under designated conditions.
ACTUATION LOGIC TEST 1.2 An ACTUATION LOGIC TEST shall be the application of various simulated input combinations in conjunction with each possible interlock logic state and verification of the required logic output. The ACTUATION LOGIC TEST shall include a continuity check, as a minimum, of output devices.
ANALOG CHANNEL OPERATIONAL TEST 1.3 An ANALOG CHANNEL OPERATIONAL TEST shall be the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY of alarm, interlock and/or trip functions. Tne ANALOG CHANNEL OPERATIONAL TEST shall include adjustments, as necessary, of the alarm, inter-lock and/or Trip Setpoints such that the Setpoints are within the required range and accuracy.
AXIAL FLUX DIFFERENCE 1.4 AXIAL FLUX DIFFERENCE shall be the difference in normalized flux signals between the top and bottom halves of a two section excore neutron detector.
CHANNEL CALIBRATION 1.5 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the
, channel such that it responds within the required range and accuracy to known l values of input. The CHANNEL CALIBRATION shall encompass the entire channel l including the sensors and alarm, interlock and/or trip functions and may be performed by any series of sequential, overlapping, or total channel steps such that the entire channel is calibrated.
CHANNEL CHECK 1.6 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other i
indications and/or status derived from independent instrument channels measuring the same parameter.
1 BRAIDWOOD UNITS 1 & 2 1-1
DEFINITIONS AUG 191996 CONTAINMENT INTEGRITY
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1.7 CONTAINMENT INTEGRITY shall exist when:
- a. All penetrat. ions required to be closed during accident conditions are either: .
- 1) Capable of being closed by an OPERABLE containment automatic isolation valve system, or .
- 2) Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except as provided in Table 3.6-1 of Specification 3.6.3.
- b. All equipment hatches are closed and sealed, I c. Each air lock is in compliance with the requirements of Specification 3.6.1.3,
- d. The containment leakage rates are within the limits of Specification 3.6.1.2, and
- e. The sealing mechanism associated with each penetration (e.g., welds,-
bellows, or 0-rings) is OPERABLE.
CONTR01 LED LEAKAGE 1.8 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals.
CORE ALTERATION
- 1. 9 CORE ALTERATION shall be the movement or manipulation of any component within the reactor vessel with the vessel head removed and fuel in the vessel.
Suspension of CORE ALTERATION shall not pret.lude complei. ion of movement of a component to a safe conservative position.
DIGITAL CHANNEL OPERATIONAL TEST
( 1.10 A DIGITAL CHANNEL OPERATIONAL TEST shall consist of exercising the digital computer hardware using data base manipulation and injecting simulated process' data to verify OPERABILITY of alarm and/or trip functions.
DOSE EQUIVALENT I-131 1.11 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie /grem) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. Tha thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites."
BRAIDWOOD UNITS 1 &'2 1-2
DEFINITIONS AUG 19 ngs E - AVERAGE DISINTEGRATION ENERGY 1.12 E shall be the average (weighted in proportion to the concentration of each radionuclide in the sample) of the sum of the average beta and gamma energies per disintegration (MeV/d) for the radionuclides in the sample.
ENGINEERED SAFETY FEATURES RESPONSE TIME 1.13 The ENGINEERED SAFETY FEATURES (ESF) RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF Actuation Setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable.
FREQUENCY NOTATION 1.14 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.
IDENTIFIED LEAKAGE 1.15 IDENTIFIED LEAKAGE shall be:
- a. Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or
- b. Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of Leakage Detection Systems or not to be PRESSURE BOUNDARY LEAKAGE, or
- c. Reactor Coolant System leakage through a steam generator to the Secondary Coolant System.
MASTER RELAY TEST 1.16 A MASTER RELAY TEST shall be the energization of each master relay and verification of OPERABILITY of each relay. The MASTER RELAY TEST shall include a continuity check of each associated slave relay.
MEMBER (S) 0F THE PUBLIC 1.17 MEMBER (S) 0F THE PUBLIC shall include all persons who are not occupationally associated with the plant. This category does not include imployees of the licensee, its contractors or vendors and persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational, or other purposes not associated with the plant.
BRAIDWOOD UNITS 1 & 2 1-3
DEFINITIONS AUG 19 Iglg 0FFSITE DOSE CALCULATION MANUAL 1.18 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm / Trip Setpoints and in the conduct of the Environ-mental RadioloCical Monitoring Program.
OPERABLE - OPERABILITY 1.19 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s),
and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its function (s) are also capable of performing their related support function (s).
t OPERATIONAL MODE - MODE 1.20 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive combination of core reactivity condition, power level, and average reactor coolant temperature specified in Table 1.2.
PHYSICS TESTS 1.21 PHYSICS TESTS shall be those tests performed to measure the fundamentai nuclear characteristics of the core and related instrumentation: (1) described '
in Chapter 14.0 of the FSAR, (2) authorized under the provisions of 10 CFR 50.59, or (3) otherwise approved by the Commission.
PRESSURE BOUNDARY LEAKAGE 1.22 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tude leakage) through a nonisolable fault in a Reactor Coolant System component body, pipe wall, or vessel wall.
PROCESS CONTROL PROGRAM 1.23 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, i
sampling, analyses, tests, and determinations to be made to ensure that l processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplishad in such a way as to assure compliance with 10 CFR Parts 20,,61, 71 and Federal and State regulations, burial ground requirements, and other requirements governing the disposal of radioactive wastes.
I BRAIDWOOD UNITS 1 & 2 1-4
DEFINITIONS AUG 19 %
PURGE - PURGING 1.24 PURGE or PURGING shall be any controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.
QUADRANT POWER TILT RATIO 1.25 QUADRANT F0WER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector cali-brated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater. With one excore detector inoperable, the remaining three detectors shall be used for computing the average.
RATED THERMAL POWER 1.26 RATED THERMAL POWER shall be a total core heat transfer rate to the -
reactor coolant of 3411 MWt.
REACTOR TRIP SYSTEM RESPONSE TIME 1.27 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its Trip Setpoint at the channel sensor until loss of stationary gripper coil voltage.
REPORTABLE EVENT 1.28 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 of 10 CFR Part 50.
SHUTDOWN MARGIN 1.29 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming all full-length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn.
SITE BOUNDARY l
1.30 The SITE BOUNDARY shall be that line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee.
SLAVE RELAY TEST >
1.31 A SLAVE RELAY TEST shall be the energization of each slave relay and verification of OPERABILITY of each relay. The SLAVE RELAY TEST shall include a continuity check, as a minimum, of associated testable actuation devices.
BRAIDWOOD UNITS'1 & 2 1-5
l DEFINITIONS SOLIDIFICATION 1.32 SOLIDIFICATION shall be the conversion of wet wastes into a form that meets shipping and burial ground requirements.
SOURCE CHECK
- 1.33 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity.
STAGGERED TEST BASIS 1.34 A STAGGERED TEST BASIS shall consist of:
- a. A test schedule for systems, subsystems, trains, or other designated components obtained by dividing the specified test interval into n equal subintervals, and
- b. The testing of one system, subsystem, train, or other designated
. component at the beginning of each subinterval.
THERMAL POWER 1.35 THERMAL POWER shall.be the total core heat transfer rate to the reactor coolant.
TRIP ACTUATING DEVICE OPERATIONAL TEST 1.36 A TRIP ACTUATING DEVICE OPERATIONAL TEST shall consist of operating the Trip Actuating Device and verifying OPERABILITY of alarm, interlock and/or trip functions. The TRIP ACTUATING DEVICE OPERATIONAL TEST shall include adjustment, as necessary, of the Trip Actuating Device such that it actuates at the required Setpoint within the required accuracy.
UNIDENTIFIED LEAKAGE 1.37 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE or CONTROLLED LEAKAGE.
UNRESTRICTED AREA L 38 An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials, or any area within the SITE BOUNDARY used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes.
BRAIDWOOD UNITS 1 & 2 1-6
DJFINITIONS AUG 191986 VENTILATION EXHAUST TREATMENT SYSTEM 1.39 A VENTILATION EXHAUST TREATMENT SYSTEM shall be any system designed and ;
installed to reduce gaseous radioiodine or radioactive material in particulate I form in effluents by passing ventilation or vent exhaust gases through charcoal I adsorbers and/or HEPA filters for the purpose of removing iodines or partic- !
ulates from the gaseous exhaust stream prior to the release to the environment.
Such a system is not considered to have any effect on noble gas effluents. !
Engineered Safety Features Atmospheric Cleanup Systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.
VENTING 1.40 VENTING shall be any controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is not provided or required during VENTING. Vent, used in system names, does not imply a VENTING process.
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WASTE GAS HOLDUP SYSTEM 1.41 A WASTE GAS HOLDUP SYSTEM shall be any system designed and installed to redace radioactive gaseous effluents by collecting Reactor Coolant System off-gases from the Reactor Coolant System and providing for delay or holdup for the purpose of reducing the total radioactivity prior to' release to the environment.
BRAIDWOOD UNITS 1 & 2 1-7
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TABLE 1.1 FINALg3 y g g DRAFT FREQUENCY NOTATION I
NOTATION FREQUENCY S At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
D At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. -
W At least once per 7 days.
M At least once per 31 days.
Q At least once per 92 days.
SA At least once per 184 days.
R At least once per 18 months.
S/U Prior to each reactor startup.
N.A. Not applicable.
P Completed prior to each release.
I e
BRAIDWOOD UNITS 1 & 2 1-8 i
TABLE 1.2 OPERATIONAL MODES AljG 1 g g REACTIVITY
% RATED AVERAGE COOLANT CONDITION, K HODE eff THERMAL POWER
- TEMPERATURE
- 1. POWER OPERATION > 0.99 > 5% > 350*F ,
- 2. STARTUP > 0.99 $ 5% > 350*F j
' > 350 F
- 3. HOT STANDBY <. 0.99 0 I
- 4. HOT SHUTDOWN < 0.99 0 350*F > T
> 200*F avg
- 5. COLD SHUTOOWN < 0.99 0 $ 200*F
- 6. REFUELING ** $ 0.95 0 $ 140 F
- Excluding decay heat.
- Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.
e BRAIDWOOD UNITS 1 & 2 1-9
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FINAL DRAFT ,
AUG 191996 1
SECTION 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS i
i
2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS
, AUG 191996 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temperature (T,yg) shall not exceed the limits shown in Figure 2.1-1 for four loop operation.
APPLICABILITY: MODES 1 and 2.
ACTION:
Whenever the point defined by the combination of the highest operating loop average temperature and THERMAL POWER has exceeded the appropriate pres-surizer pressure line, be.in HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and comply with the requirements of Specification 6.7.1.
REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant Syster. pressure shall not exceed 2735 psig.
APPLICABILITY: MODES 1, 2, 3, 4, and 5.
ACTION:
MODES 1 and 2:
Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANDBY with the Reactor Coolant System pressure within this limit within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and comply with the requirements of Specification 6.7.1.
MODES 3, 4 and 5:
Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within this limit within 5 minutes, and comply with.the requirements of Specification 6.7.1.
BRAIDWOOD - UNITS 1 & 2 2-1
FINAL DRAFT 670
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BRAIDWOOD - UNITS 1 & 2 2-2
SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS
. AUG 191986
- 2. 2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The Reactor Trip System Instrumentation and Interlock Setpoints shall be set consistent within the Trip Setpoint values shown in Table 2.2.-l.
APPLICABILITY: As shown for each channel in Table 3.3-1. ACTION:
- a. With a Reactor Trip System Instrumentation or Interlock Setpoint less conservative than the value shown in the Trip Setpoint column but more conservative than the value shown in the Allowable Value column of Table 2.2-1, adjust the Setpoint consistent with the Trip Setpoint value,
- b. With the Reactor Trip System Instrumentation or Interlock Setpoint less conservative than the value shown in the Allowable Values column of Table 2.2-1, either:
- 1. Adjust the Setpoint consistent with the Trip Setpoint value of Table 2.2-1 and determine within 12 hours that Equation 2.2-1 was satisfied for the affected channel, or
- 2. Declare the channel inoperable and apply the applicable ACTION statement requirement of Specification.3.3-1 until the channel is restored to OPERABLE status with its Setpoint adjusted consistent with the Trip Setpoint value.
Equation 2.2-1 Z + RE + SE 1 TA Where: 2= The value for Column Z of Table 2.2-1 for the affected channel, RE = The "as measured" value (in percent span) of rack error for the affected channel, SE = Either the "as measured" value (in percent span) of the sensor error, or the value for Column SE (Sensor Error) of Table 2.2-1 for the affected channel, and TA = The value for Column TA (Total Allowance) of Table 2.2-1 for the affected channel. BRAIDWOOD - UNITS 1 & 2 2-3
TABLE 2.2-1 l g REACTOR T.9IP SYSTEM INSTRUMENTATION TRIP SETPOINTS 8 SENSOR E TOTAL ERROR FUNCTIONAL UNIT ALLOWANCE (TA) Z (SE) TRIP SETPOINT ALLOWABLE VALUE ] g 1. Manual Reactor Trip N.A. N.A. N.A. N.A. N.A. e-m 2. Power Range, Neutron Flux
- a. High Setpoint 7.5 4.56 0 $109% of RTP* $111.1% of RTP*
- b. Low Setpoint 8.3 4.56 0 $25% of RTP* $27.1% of RTP*
- 3. Power Range, Neutron Flux, 1.6 0.5 0 <5% of RTP* with <6.3% of RTP* with High Positive Rate i time constant i time constant 7 22 seconds 22 seconds a
- 4. Power Range, Neutron Flux, 1.6 0.5 0 15% of RTP* with <6.3% of RTP* with High Negative Rate a time constant i time constant 22 seconds 22 seconds
- 5. Intermediate Range, 17.0 8.4 0 $25% of RTP* $30.9% of RTP*
Neutron Flux
- 6. Source Range, Neutron Flux 17.0 10.0 0 $105 cps $1.4 x 105 cps
- 7. Overtemperature AT 27.7 5.38 See See Note 1 See Note 2 q Note 5 mi. -
4.3 1.3 5 E
- 8. Overpower AT 1.2 See Note 3 See Note 4 s y
- 9. ,
Pressurizer Pressure-Low 5.0 2.21 1.5 21885 psig 21871 psig
- 10. Pressurizer Pressure-High 3.1 0.71 1. 5 52385 psig $2396 psig- O
- 11. Pressurizer Water Level-High 5. 0 2.18 1. 5 $92% of instrument $93.8% of instrument span span
*RTP = RATED THERMAL POWER
E TABLE 2.2-1 (Continued) b E REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 8 SENSOR TOTAL ERROR E FUNCTIONAL UNIT ALLOWANCE (TA) Z (SE) TRIP SETPOINT ALLOWABLE VALUE d 12. Reactor Coolant Flow-Low 2.5 1.77 0.6 >90% of loop mini- >89.2% of loop mini-r mum measured flow
- mum measured flow
- e.
m 13. Steam Generator Water Level Low-Low
- a. Unit 1 27.1 18.28 1.5 >40.8% of narrow >39.1% of narrow range instrument Fange instrument span span
- b. Unit 2 17.0 14.78 1.5 317% of narrow >15.3% of narrow roi ge instrument Fange instrument 4 span span
- 14. Undervoltage - Reactor 12.0 0.7 0 >5268 volts -
>4728 volts -
Coolant Pumps each bus each bus
- 15. Underfrequency - Reactor 14.4 13.3 0 >57.0 Hz >56.5 Hz Coolant Pumps
- 16. Turbine Trip
- a. Emergency Trip Header N.A. N.A. N.A. >540 psig ->520 psig Pressure
- b. Turbine Throttle Valve N.A. N.A. N.A. >1% open >1% open Z.
Closure p
- 17. Safety Injection Input N.A. N.A. N.A. N.A. N.A.
C m P from ESF g
- 18. Reactor Coolant Pump Breaker Position Trip N.A. N.A. N.A. N.A. N.A. g
- Minimum measured flow = 97,600 gpm
% TABLE 2.2-1 (Continued) E E REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 8 o SENSOR E TOTAL ERROR g FUNCTIONAL UNIT ALLOWANCE (TA) Z (SE) TRIP SETPOINT ALLOWABLE VALUE g 19. Reactor Trip System , Interlocks
- a. Intermediate Range N.A. N.A. N.A. El x 10 10 amp 16 x 10 12 amp Neutron Flux, P-6
- b. Low Power Reactor Trips Block, P-7
- 1) P-10 input N.A. N.A. N.A. 510% of RTP* 17.9% to $12.1% of RTP*
- 2) P-13 input N.A. N.A. N.A. <10% RTP* Turbine $12.1% RTP* Turbine Impulse Pressure Impulse Pressure Equivalent Equivalent
- c. Power Range Neutron N.A. N.A. N.A. 530% of RTP* $32.1% of RTP*
Flux, P-8
- d. Power Range Neutron N.A. N.A. N.A $10% of RTP* 17.9% to $12.1% of RTP*
Flux, P-10
- e. Turbine Impulse Chamber N.A. N.A. N.A. <10% RTP* Turbine <12.1% RTP* Turbine q Pressure, P-13 Tapulse Pressure Impulse Pressure -
Equivalent Equivalent 2
- 20. Reactor Trip Breakers N.A. N.A. N.A N.A. N .. A . E
- 21. Automatic Trip and Interlock N.A. N.A. N.A. N.A. N.A. E' Logic *O
*RTP = RATED THERMAL POWER
y ' TABLE 2.2-1 (Continued) E TABLE NOTATIONS 5 8 i NOTE 1: OVERTEMPERATURE AT AT t (1 r3 5) 5 AT, {K3 - K 2 h [T'(1 4 S) - T'] + K 3(P - P') - f (AI)} 3 w [ Where: AT = Measured AT by RTD Manifold Instrumentation, 1 = Lead-lag compensator on measured AT, T1, T2 = Time constants utilized in lead-lag compensator for AT, r3 = 8 s, T2=3s, , 1+153
= Lag compensator on measured AT, r3 = Time constants utilized in the lag compensator for AT, r3 = 0 s,
= Indicated AT at RATED THERMAL POWER, AT, ,
Ki = 1.164, K2 = 0.0265/*F, 1 = The function generated by the lead-lag compensator for T,yg dynamic compensation, ,,
=
T4, Ts Time Ts = 4constants s, utilized in the lead-lag compensator for Tavg, T 4 = 33 s, p T = Average temperature, F, E r-1
- y. 3
=
Lag compensator on measured T,yg, g W
!g TABLE 2.2-1 (Continued)
TABLE NOTATIONS (Continued) NOTE 1: (Continued) E rs = Time constant utilized in the measured T,yg lag compensator, Ts = 0 s, T' 5 588.4 F (Nominal T,yg at RATED THERMAL POWER),
" K3 = 0.00134, m
P = Pressurizer pressure, psig, P', = 2235 psig (Nominal RCS operating pressure), S = Laplace transform operator, s 1, , and f3 (AI) is a function of the indicated difference between top and bottom detectors of the 7 power-range neutron ion chambers; with gains to be selected based on measured instrument
- response during plant STARTUP tests such that:
(i) for'qt 9b between -=% and +10%,t f (AI) = 0, where qt and qb are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and qt
- 9b s total THERMAL POWER in percent of RATED THERMAL POWER; (ii) for each percent that the magnitude of qt ~9b exceeds +10%, the AT Trip Setpoint shall be automatically reduced by 2.0% of its value at RATED THERMAL POWER.
T NOTE 2: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 3.9% of AT span. M ge o Eh! 21
g TABLE 2.2-1 (Continued) B g TABLE NOTATIONS (Continued) 8 7 NOTE 3: OVERPOWER AT g (1 + tsS) y ( tyS ) y ( 1 ) AT (1 + T2 5) (1 + IsS} I o {K 4 K3 y gt3 y T3 6
) T - Ks IT(y y 7s3
- T9 - f 2(al}}
w e. Where: AT = As defined in Note 1, 1 + ins
= As defined in Note 1, 1+TS2 i
II, 12 = As defined in Note.1,
- 1 y
= As defined in Note 1, 13 = As defined in Note 1, i
AT, = As defined in Note 1, K4 = 1.072, Ks = 0.02/*F for increasing average temperature and 0 for decreasing average temperature, 157
= The function generated by the rate-lag compensator for Tavg dynamic 1 + tyS compensation, 7g
=
p r7 Time constants utilized in the rate-lag compensator for T,yg, t7 = 10*s , ggm 1
- y. 3
= As defined in Note 1, [g i
g i ts = As defined in Note 1, M
I co TABLE 2.2-1 (Continued) b O TABLE NOTATIONS (Continued) 6 o NOTE 3: (Continued) E Ks = 0.00170/*F for T > T" and Ks = 0 for T $ T", Z T = As defined in Note 1, e--
** T" = Indicated T avg at RATED THERMAL POWER (Calibration temperature for AT m
i instrumentation, 5 588.4*F), i' S = As defined in Note 1, and f 2(AI) = 0 for all AI.
'? NOTE 4: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 0 2.6% of AT span.
NOTE 5: The sensor error for temperature is 1.2 and for pressure is 1.0. 1 i i P o E C-
! :l 2
i e FINAL DRAFT AUG 191996 BASES FOR SECTION 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS e 9 , , _ . , _ , , - , - - - , - . , - - - - , , . , - _ - - -- .-,-,,--w------- -,<n,-- - - - - - - ,e -- - -- - '- - ' ' - ~
1 FINAL DRAFT AUG 191996 i i 1 I NOTE The BASES contained in succeeding pages summarize the reasons for the Specifications in Section 2.0, but in accordance with 10 CFR 50.36 are not part of these Technical Specifications. 0
2.1 SAFETY LIMITS AUG 191986 BASES 2.1.1 REACTOR CORE The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature. Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB through the WRB-1 correlation. The WRB-1 DNB correlation has been developed to predict the DNB flux and the location of DNB for axially uniform and nonuniform heat flux distributions. The local DNB heat flux ratio (DNBR) is defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, and is indicative of the margin to DNB. The DNB design. basis is as follows: there must be at least a 95 percent probability that the minimum DNBR of the limiting rod during Condition I and II events is greater than or equal to the DNBR limit of the DNB correlation being used (the WRB-1 correlation in this application). The correlation DNBR limit is established based on the entire applicable experimental data set such that there is a 95 percent probability with 95 percent confidence that DNB will not occur when the minimum DNBR is at the DNBR limit. In meeting this design basis, uncertainties in plant operating parameters, nuclear and thermal parameters, and fuel fabrication parameters are considered statistically such that there is at least a 95 confidence that the minimum DNBR for the limiting rods is greater than or equal to the DNBR limit. The uncer-tainties in the above plant parameters are used to determine the plant DNBR uncertainty. This DNBR uncertainty, combined with the correlation DNBR limit, establishes a design DNBR value which must be met in plant safety analysis using values of input parameters without uncertainties. I The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum design DNBR is no less than 1.34 for a typical cell and 1.32 for a thimble cell, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid. These curves are based on an enthalpy hot channel factor, F g, of 1.49. An allowance is included for an increase in F H at reduced power basect on the expression: BRAIDWOOD - UNITS 1 & 2 8 2-1
SAFETY LIMITS BASES - AUG191%6 REACTOR CORE (Continued) F H = 1.49 [1+ 0.3 (1-P)] Where P is the fraction of RATED THERMAL POWER. These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance is within the limits of the f1 (AI) function of the Overtemperature trip. When the axial power imbalance is not within the tolerance, the axial power imbalance effect on the Overtem-perature AT trips will reduce the Setpoints to provide protection consistent with core Safety Limits. 2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolar.t System (RCS) from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere. The reactor vessel, pressurizer, and the RCS piping, valves, and fittings are designed to Section III of the ASME Code for Nuclear Power Plants which permits a maximum transient pressure of 110% (2735 psig) of design pressure. The Safety Limit of 2735 psig is therefore consistent with the design criteria and associated Code requirements. The entire RCS is hydrotested at 3110 psig,125% of design pressure, to demonstrate integrity prior to initial operation. l BRAIDWOOD - UNITS 1 & 2 B 2-2
2.2 LIMITING SAFETY SYSTEM SETTINGS FINAL HAFT M 19 286 BASES 2.2.1 REACTOR TRIP SYSTEM INSTF.UMENTATION SETPOINTS The Reactor Trip Setpoint Limits specified in Table 2.2-1 are the nominal values at which the Reactor trips are set for each functional unit. The Trip Setpoints have been selected to ensure that the core and Reactor Coolant System are prevented from exceeding their Safety Limits during normal operation and design basis anticipated operational occurrences and to assist the Engi-neered Safety Features Actuation System in mitigating the consequences of accidents. The Setpoint for a Reactor Trip System or interlock function is considered to be adjusted consistent with the nominal value when the "as measured" Setpoint is within the band allowed for calibration accuracy. To accommodate the instrument drift assumed to occur between operational tests and the accuracy to which Setpoints can be measured and calibrated, Allowable Values for the Reactor Trip Setpoints have been specified in Table 2.2-1. Operation with Setpoints less conservative than the Trip Setpoint but within the Allowable Value is acceptable since an allowance has been made in the safety analysis to accommodate this error. An optional provision has been included for determining the CPERABILITY of a channel when its Trip Setpoint is found to exceed the Allowable Value. The methodology of this option utilizes the "as measured" deviation from the specified calibration point for rack and sensor components in conjunction with a statistical combin-ation of the other uncertainties of the instrumentation to measure the process variable and the uncertainties in calibrating the instrumentation. In Equa-tion 2.2-1, Z + RE + SE < TA, the interactive effects of the errors in the rack and the sensor, and the Was measured" values of the errors are considered. Z, as specified in Table 2.2-1, in percent span, is the statistical summation of errors assumed in the analysis excluding those associated with the sensor and rack drift and the accuracy of their measurement. TA or Total Allowance is the difference, in percent span, between the Trip Setpoint and the value used in the analysis for Reactor trip. RE or Rack Error is the "as measured" devia-tion, in percent span, for the affected channel from the specified Trip Setpoint. SE or Sensor Error is either the "as measured" deviation of the sensor from its calibration point or the value specified in Table 2.2-1, 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. l The methodology to derive the Trip Setpoints is based upon combining all l of the uncertainties in the channels. Inherent to the determination of the Trip Setpoints are the magnitudes of these channel uncertainties. Sensors and other instrumentation utilized in these channels are expected to be capable of operating within the allowances of these uncertainty magnitudes. Rack drift j in excess of the Allowable Value exhibits the behavior that the rack has not I met its allowance. Being that there is a small statistical chance that this I will happen, an infrequent excessive drift is expected. Rack or sensor drift, l in excess of the allowance that is more than occasional, may be indicative of more serious proble:ns and should warrant further investigation. BRAIDWOOD - UNITS 1 & 2 B 2-3
FINAL HAFT l LIMITING SAFETY SYSTEM SETTINGS AUG 191986 BASES REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS (Continued) The various Reactor trip circuits automatically open the Reacto.r trip breakers whenever a condition monitored by the Reactor Trip System reaches a preset or calculated level. In addition to redundant channels and trains, the design approach provides a Reactor Trip System which monitors numerous system variables, therefore, providing Trip System functional diversity. The func-tional capability at the specified trip setting is required for those antici-patory or diverse Reactor trips for which no direct credit was assumed in the accident analysis to enhance the overall reliability of the Reactor Trip System. The Reactor Trip System initiates a Turbine trip signal whenever Reactor trip is initiated. This prevents the reactivity addition that would otherwise result from excessive Reactor Coolant System cooldown and thus avoids unnecessary actuation of the Engineered Safety Features Actuation System. Manual Reactor Trip The Reactor Trip System includes manual Reactor trip capability. Power Ran'ge, Neutron Flux In each of the Power Range Neutron Flux channels there are two ir. dependent bistables, each with its own trip setting used for a High and Low Range trip setting. The Low Setpoint trip provides protection during subcritical and low power operations to mitigate the consequences of a power excursion beginning from low power, and the High Setpoint trip provides protection during power operations to mitigate the consequences of a reactivity excursion from all power levels. The Low Setpoint trip may be manually blocked above P-10 (a power level of approximately 10% of RATED THERMAL POWER) and is automatically reinstated below the P-10 Setpoint. , Power Range, Neutron Flux, High Rates The Power Range Positive Rate trip provides protection against rapid flux increases which are characteristic of a rupture of a control rod drive housing. Specifically, this trip complements the Power Range Neutron Flux High and Low trips to ensure that the criteria are met for rod ejection from mid power. BRAIDWOOD - UNITS 1 & 2 B 2-4
LIMITING SAFETY SYSTEM SETTINGS BASES Power Range, Neutron Flux, High Rates (Continued) The Power Range Negative Rate trip provides protection for control rod drop accidents. At high power a single or multiple rod drop accident could cause local flux peaking which could cause an unconservative local DNBR to exist. The Power Range Negative Rate trip will prevent this from occurring by tripping the reactor. No credit is taken for operation of the Power Range Negative Rate trip for those control rod drop accidents for which DNBRs will be greater than the limit value. Intermediate and Source Range, Neutron Flux The Intermediate and Source Range, Neutron Flux trips provide core protection during reactor STARTUP to mitigate the consequences of an uncontrolled rod cluster control assembly bank withdrawal from a subcritical condition. These trips provide redundant protection to the Low Setpoint trip of the Power Range, Neutron Flux channels. The Source Range channels will initiate a Reactor trip at about 105 counts per second unless manually blocked when P-6 becomes active. The Intermediate Range channels will initiate a Reactor trip at a current level equivalent to approximately 25% of RATED THERMAL POWER unless manually blocked when P-10 becomes active. Overtemperature aT The Overtemperature AT trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power distri'bution, provided that the transient is slow with respect to piping transit delays from the core to the temperature detectors (about 4 seconds), and pressure is within the range between the Pressurizer High and Low Pressure trips. The Setpoint is automatically varied with: (1) coolant temperature to correct for temperature induced changes in density and heat capacity of water and includes dynamic compensation for piping delays from the core to the loop temperature detectors, (2) pressurizer pressure, and (3) axial power distribution. With normal axial power distribution, this Reactor trip limit is always below the core Safety Limit as shown in Figure 2.1-1. If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the Reactor trip is automatically reduced according to the notations in Table 2.2-1. BRAIDWOOD - UNITS 1 & 2 B 2-5 a
LIMITING SAFETY SYSTEM SETTINGS AUG 191986 BASES Overpower AT
. The Overpower A' Reactor trip provides assurance of fuel integrity (e.g., no fuel pellet melting and less than 1% cladding strain) under all possible overpower conditions, limits the required range for Overtemperature AT trip, and provides a backup to the High Neutron Flux trip. The Setpoint is automatically varied with: (1) coolant temperature to correct for temperature
( induced changes in density and heat capacity of water, and (2) rate of change of temperature for dynamic compensation for piping delays from the core to the loop temperature detectors, to ensure that the allowable heat generation rate (kW/ft) is not exceeded. The Overpower AT trip provides protection to mitigate the consequences of various size steam breaks as reported in WCAP-9226, " Reactor Core Response to Excessive Secondary Steam Releases." 7 Presst.izer riessure ' In each of the pressure channels, there are two independent bistables, each with its own trip setting to provide for a High and Low Pressure trip thus lim ting the pressure range in which reactor operation is permitted. The i 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 reiief and safety valves to protect the Reactor Coolant System against system t 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 approximately 10% of RATED THERMAL POWER with a turbine impulse chamber , pressure at approximately 10% of full power equivalent); and on increasing power, automatically reinstated by P-7. BRAIDWOOD - UNITS 1 & 2 B 2-6
LIMITING SAFETY SYSTEM SETTINGS BASES 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 90% of nominal full loop flow. Above P-8 (a power level of approximately 30% of RATED THERMAL POWER) an automatic Reactor trip will occur if the flow in any single loop drops below 90% of nominal full loop flow. Conversely on decreasing power between P-8 and 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. 5 team 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 resulting from loss of normal feedwater. The specified Setpoint provides allowances for starting delays of the Auxiliary Feedwater System. Undervoltage and Underfrequency - Reactor Coolant Pump Busses l The Undervoltage and Underfrequency Reactor Coolant Pump Bus trips provide ! core protection against DNB as a result of complete loss of forced coolant flow. The specified Setpoints assure a Reactor trip signal is generated before the Low Flow Trip Setpoint is reached. For undervoltage, the delay is set so that the time required for a signal to cause a reactor trip after the Under-voltage Trip Setpoint is reached shall not exceed 1.5 seconds. For under-f equency, the delay is set so that the time required for a signal to cause a reactor trip after the Underfrequency Trip Setpoint is reached shall not exceed 0.6 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. BRAIDWOOD - UNITS 1 & 2 8 2-7
LIMITING SAFETY SYSTEM SETTINGS ' AUC 19 E06 BASES Turbine Trip A Turbine trip initiates a Reactor trip. On decreasing power the Turoine trip is 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. Safety Injection Input from ESF If a Reactor trip has not already been generated by the Reactor Trip System instrumentation, the ESF automatic actuation logic channels will initiate a Reactor trip upon any signal which initiates a Safety Injection. The ESF instrumentation channels which initiate a Safety Injection signal are shown in Table 3.3-3. Reactor Coolant Pump Breaker Position Trip The Reactor Coolant Pump Breaker Position trips are anticipatory trips which provide core protection against DNB. The Open/Close Position trips assure a Reactor trip signal is generated before the Low Flow Trip Setpoint is reached. No credit was taken in the accident analyses for operation of these trips. Their functional capability at the open/close position settings is required to enhance the overall reliability of the Reactor Trip System. 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 l automatic Reactor trip will occur if more than one reactor coolant pump breaker ! is opened. Below P-7 the trip function is automatically blocked. O BRAIDWOOD - UNITS 1 & 2 B 2-8
LIMITING SAPITY SYSTEM SETTINGS g
. AUG 191986 3
Reactor Trip System Interlocks The Reactor Trip System Interlocks perform,the following functions: P-6 On increasing power, P-6 allows the manual block of the Source Range Reactor trip (i.e., prevents premature block of Source Range trip), provides an satomatic backup block for Source Range Neutron Flux doubling, and the manual block that de-energizes the high voltage to the Source Range detectors. On decreasing power, Source Range Level trips and Neutron Flux doubling circuits are automatically reactivated and high voltage restored. P-7 On increasing power, P-7 automatically enables Reactor trips on low flow in more than one reactor coolant loop, more than one reactor coolant pump breaker open, reactor coolant pump bus undervoltage and underfrequency, Turbine trip, pressurizer low pressure and pressurizer high level. On decreasing power, the above listed trips are automatically blocked. P-8 On increasing power, P-8 automatically enables Reactor trips on low flow in one or more reactor coolant loops. On decreasing power, the P-8 automatically blocks the single loop low flow trip. P-10 On increasing power, P-10 allows the manual block of the Intermediate Range Reactor trip and the Low Setpoint Power Range Reactor trip; and automatically blocks the Source Range Reactor trip and provides an automatic backup function to de-energize the Source Range high voltage power. On decreasing power, the Intermediate Range Reactor trip and the Low Setpoint Power Range Reactor trip are automatically reactivated and Source Range high voltage to the detectors is restored if power decreases below the P-6 setpoint. Provides input to P-7. P-13 Provides input to P-7. 1 BRAIDWOOD - UNITS 1 & 2 B 2-9
FINAL DRAFT l AUG 191986 0 6 SECTIONS 3.0 AND 4.0 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS e
- . . ~ . - _ _ . _ _ _ _ _ _ _ _ . _ _ _ _ _ _
FINAL HAFT 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREME b I O E 3/4.0 APPLICABILITY LIMITING CONDITION FOR OPERATION 3.0.1 Compliance with the Limiting Conditions for Operation contained in the succeeding specifications is required during the OPERATIONAL MODES or other conditions specified therein; except that upon failure to meet the Limiting Conditions for Operation, the associated ACTION requirements shall be met. 3.0.2 Noncompliance with a specification shall exist when the requirements of the Limiting Condition for Operation and associated ACTION requirements are not met within the specified time intervals. If the Limiting Condition for Operation is restored prior to expiration of the specified time intervals, completion of the ACTION requirements is not required. 3.0.3 When a Limiting Condition for Operation is not met, except as provided in the associated ACTION requirements, within 1 hour action shall be initiated to place the unit in a MODE in which the specification does not apply by placing it, as applicable, in:
- a. At least HOT STANDBY within the next 6 hours,
- b. At least HOT SHUTDOWN within the following 6 hours, and
- c. At least COLD SHUTDOWN within the subsequent 24 hours.
Where corrective measures are completed that permit operation under the ACTION requirements, the action may be taken in accordance with the specified time limits as measured from the time of failure to meet the Limiting Condition for Operation. Exceptions to these requirements are stated in the individual specifications. This specification is not applicable in MODE 5 or 6. 3.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made unless the conditions for the Limiting Condition for Operation are met l without reliance on provisions contained in the ACTION requirements. This j provision shall not prevent passage through or to OPERATIONAL MODES as required i to comply with ACTION requirements. Exceptions to these requirements are ; stated in the individual specifications. 3.0.5 Limiting Conditions for Operation including the associated ACTION require-ments shall apply to each unit individually unless otherwise indicated as follows:
- a. Whenever the Limiting Conditions for Operation refers to systems or components which are shared by both units, the ACTION requirements will apply to both units simultaneously.
- b. Whenever the Limiting Conditions for Operation applies to only one unit, this will be identified in the APPLICABILITY section o.f the specification; and
- c. Whenever certain portions of a specification contain operating param-eters, Setpoints, etc., which are different for each unit, this will be identified in parentheses, footnotes or body of the requirement.
BRAIDWOOD - UNITS 1 & 2 3/4 0-1
APPLICABILITY AUG 191986 SURVEILLANCE REQUIREMENTS 4.0.1 Surveillance Requirements shall be met during the OPERATIONAL MODES or other conditions specified for individual Limiting Conditions for Operation unless otherwise stated in an individual Surveillance Requirement. 4.0.2 Each Surveillance Requirement shall be performed within the specified time interval with:
- a. A maximum allowable extension not to exceed 25% of the surveillance interval, but
- b. The combined time interval for any three consecutive surveillance intervals shall not exceed 3.25 times the specified surveillance interval.
4.0.3 Failure to perform a Surveillance Requirement within the specified time interval shall constitute a failure to meet the OPERABILITY requirements for a Limiting Condition for Operation. Exceptions to these requirements are stated in the individual specifications. Surveillance Requirements do not have to be performed on inoperable equipment. 4.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made unless the Surveillance Requirement (s) associated with the Limiting Condition for Operation have been performed within the stated surveillance interval or as otherwise specified. 4.0.5 Surveillance Requirements for inservice inspection and testing of ASME Code Class 1, 2, and 3 components shall be applicable as follows: I a. Inservice inspection of ASME Code Class 1, 2, and 3 components and i inservice testing of ASME Code Class 1, 2, and 3 pumps and valves j shall be performed in accordance with Section XI of the ASME Boiler I and Pressure Vessel Code and applicable Addenda as required by 10 CFR Part 50, Section 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50, Section 50.55a(g)(6)(i);
- b. Surveillance intervals specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda for the inservice inspec-tion and testing activities shall be applicable as follows in these Technical Specifications:
BRAIDWOOD - UNITS 1 & 2 3/4 0-2
FINAL DRAFT APPLICABILITY AUG 19 EE6 SURVEILLANCE REQUIREMENTS (Continued) ASME BOILER AND PRESSURE VESSEL REQUIRED FREQUENCIES FOR CODE AND APPLICABLE ADDENDA PERFORtING INSERVICE TERMIN0 LOGY FOR INSERVICE INSPECTION AND TESTING ~ INSPECTION AND TESTING ACTIVITIES ACTIVITIES Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days . i c. The provisions of Specification 4.0.2_are applicable to the above required frequencies for performing inservice inspection and testing activities;
- d. Performance of the above inservice inspection and testing activities shall be in addition to other specified Surveillance Requirements; and .
- e. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any Technical Specification.
4.0.6 Surveillance Requirements shall apply to each unit individually unless otherwise indicated as stated in Specification 3.0.5 for individual specifica-tions or whenever certain portions of a specification contain surveillance parameters different for each unit, which will be identified in parentheses, footnotes or body of the requirement. O BRAIDWOOD - UNITS 1 & 2 3/4 0-3
3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL , SHUT 00WN MARGIN - T,y >200*F LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to 1.3% Ak/k for four loop operation. APPLICABILITY: MODES 1, 2*, 3, and 4. ACTION: With the SHUTDOWN MARGIN less than 1.3% Ak/k, immediately initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored. SURVEILLANCE REQUIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be c'etermined to be greater than or equal to 1.3% Ak/k:
- a. Within 1 hour after detection of an inoperable control rod (s) and at least once per 12 hours thereafter while the rod (s) is inoperable.
If the inoperable control rod (s) is immovable or untrippable, the above ,i required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable ! control rod (s); l
- b. When in MODE 1 or MODE 2 with K,ff greater than or equal to 1 at least once per 12 hours by verifying that control bank insertion is within the limits of Specification 3.1.3.6;
- c. When in MODE 2 with K,ff less than 1, within 4 hours prior to achieving reactor criticality by verifying that the predicted critical control rod position is within the limits of Specification 3.1.3.6;
- d. Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors of Specification 4.1.1.1.le. below, with the control banks at the maximum insertien limit of Specification 3.1.3.6; and
*See Special Test Exceptions Specification 3.10.1.
~
l BRAIDWOOD - UNITS 1 & 2 3/4 1-1
REACTIVITY CONTROL SYSTEMS FINAL DRAFT AUG 191936 SURVEILLANCE REQUIREMENTS (Continued)
- e. When in MODE 3 or 4, at least once per 24 hours by consideration of the following factors:
- 1) Reactor Coolant System boron concentration, .
- 2) Control rod position,
- 3) Reactor Coolant System average temperature,
- 4) Fuel burnup based on gross thermal energy generation,
- 5) Xenon concentration, and
- 6) Samarium concentration.
4.1.1.1.2 The overall core reactivity balance shall be compared to predicted values to demonstrate agreement within i 1% Ak/k at least once per 31 Effective Full Power Days (EFPD). This comparison shall consider at least those factors stated in Specification 4.1.1.1.le. above. The predicted reactivity values shall be adjusted (normalized) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 EFPD after each fuel loading. BRAIDWOOD - UNITS 1 & 2 3/4 1-2
REACTIVITY CONTROL SYSTEMS 0 SHUTOOWN MARGIN - T,yg < 200*F 1 LIMITING CONDITION FOR OPERATION l l 3.1.1.2 The SHUTDOWN MARGIN shall be greater than or equal to 1% Ak/k. I 1 gPLICABILITY: MODE 5. l ACTION: With the. SHUTDOWN MARGIN less than 1% ak/k, immediately initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored. SURVEILLANCE REQUIREMENTS 4.1.1.2 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1% Ak/k:
- a. Within 1 hour after detection of an ino;1erable control rod (s) and at least once per 12 hours thereafter while the rod (s) is inoperable.
If the inoperable control rod (s) is immovable or untrippable, tne SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod (s); and
- b. At least once per 24' hours by consideration of the following factors:
- 1) Reactor Coolant System boron concentration,
- 2) Control rod position,
- 3) Reactor Coolant System average temperature,
- 4) Fuel burnup based on gross thermal energy generation, l 5) Xenon concentration, and
- 6) Samarium concentration.
1 L l
. I BRAIDWOOD - UNITS 1 & 2 3/4 1-3
REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION 3.1.1.3 The moderator temperature coefficient (MTC) shall be:
- a. Less positive than 0 ak/k/ F for the all rods withdrawn, beginning of cycle life (BOL), hot zero THERMAL POWER condition, or
- b. Less negative than -4.1 x 10 4 Ak/k/*F for the all rods withdrawn, end of cycle life (E0L), RATED THERMAL POWER condition.
APPLICABILITY: Specification 3.1.1.3a. - MODES 1 and 2* only#. Specification 3.1.1.3b. - MODES 1, 2, and 3 only#. ACTION:
- a. With the MTC more positive than the limit of Specification 3.1.1.3a.
above, operation in MODES 1 and 2 may proceed provided:
- 1. Control rod withdrawal limits are established and maintained sufficient to restore the MTC to less positive than 0 ak/k/ F within 24 hours or be in HOT STANDBY within the next 6 hours.
These withdrawal lintits shall be in addition to the insertion limits of Specification 3.1.3.6;
- 2. The control rods are maintained within the withdrawal limits established above until a subsequent calculation verifies that the MTC has been restored to within its limit for the all rods withdrawn condition; and
- 3. A Special Report is prepared and submitted to the Commission pursuant to Specification 6.9.2 within 10 days, describing the value of the measured MTC, the interim control rod withdrawal limits, and the predicted average core burnup necessary for restoring the positive MTC to within its limit for the all rods withdrawn condition.
- b. With the MTC more negative than the limit of Specification 3.1.1.3b.
above, be in HOT SHUTDOWN within 12 hours. "With K,ff greater than or equal to 1.
- See Special Test Exceptions Specification 3.10.3.
BRAIDWOOD - UNITS 1 & 2 3/4 1-4
REACTIVITY CONTROL SYSTEMS AUG 19 KB6 ;
~
SURVEILLANCE REQUIREMENTS 4.1.1. 3 The MTC shall be determined to be within its limits during each fuel cycle as follows:
- a. The MTC shall be measured and compared to th'e BOL limit of Specification 3.1.1.3a. , above, prior to initial operation above 5%
of RATED THERMAL POWER, after each fuel loading, and
- b. The MTC shall be measured at any THERMAL POWER and compared to
-3.2 x 10 4 Ak/k/ F (all rods withdrawn, RATED THERMAL POWER condi-tion) within 7 EFPD after reaching an equilibrium boron concen-tration of 300 ppm. In the event this comparison indicates the MTC is more negative than -3.2 x 10 4 Ak/k/ F, the MTC shall be remeasured, and compared to the E0L MTC limit of Specification 3.1.1.3b. , at least once per 14 EFPD during the remainder of the fuel cycle.
BRAIDWOOD - UNITS 1 & 2 3/4 1-5
REACTIVITY CONTROL SYSTEMS MINIMUM TEMPERATURE FOR CRITICALITY AUG 19 M LIMITING CONDITION FOR OPERATION 3.1.1.4 shall beThe Reactor greater thanCoolant System or equal to 550lowest F. operating loop temperatur'e (Tavg) APPLICABILITY: MODES 1 and 2#*. ACTION: With a Reactor Coolant System operating loop temperature (T,yg) less than 550*F, restore T,yg to within its limit within 15 minutes or be in HOT STANDBY within the next 15 minutes. SURVEILLANCE REQUIREMENTS 4.1.1.4 The Reactor Coolant System temperature (Tavg) shall be determined tc be greater than or equal to 550 F:
- a. Within 15 minutes prior to achieving re~ actor criticality, and
- b. At least once per 30 minutes when the reactor is critical and the Reactor Coolant System T,yg is less than 557*F with the T,yg-T ref Deviation Alarm not reset.
#With K,7f greater than or equal to 1. *See Special Test Exceptions Specification 3.10.3.
I I BRAIDWOOD - UNITS 1 & 2 3/4 1-6 I
REACTIVITY CONTROL SYSTEMS 3/4.1.2 80 RATION SYSTEMS FLOW PATH - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.1 As a minimum, one of the following boron injection flow paths shall be OPERABLE and capable of being powered from an OPERABLE emergency power source:
- a. A flow path from the Boric Acid Storage System via a boric acid transfer pump and a centrifugal charging pump to the Reactor Coolant System if the Boric Acid Storage System is OPERABLE as given in Specification 3.1.2.5a. for MODES 5 and 6 or as given in Specification 3.1.2.6a.
for MODE 4; or
- b. The flow path from the refueling water storage tank via a centrifugal charging pump to the Reactor Coolant System if the refueling water storage tank is OPERABLE as given in Specification 3.1.2.5b. for MODES 5 and 6 or as given in Specification 3.1.2.6b. for MODE 4.
APPLICABILITY: MODES 4*, 5 and 6. ACTION: With none of the above flow paths OPERABLE or capable of being powered from an OPERABLE emergency power source, suspend all operations involving CORE ALTERATIONS or positive reactivity changes. SURVEILLANCE REQUIREMENTS 4.1.2.1 At least one of the above required flow paths shall be demonstrated OPERABLE:
- a. At least once per 7 days by verifying that the temperature of the heat traced portion of the flow path is greater than or equal to 65*F when a flow path from the Boric Acid Storage System is used, and
- b. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.
- A maximum of one charging pump shall be OPERABLE, and that pump sball be a centrifugal charging pump, whenever the temperature of one or more of the RCS cold legs is less than or equal to 330*F.
BRAIDWOOD - UNITS 1 & 2 3/4 1-7
REACTIVITY CONTROL SYSTEMS FLOW PATHS - OPERATING AUG 1 g ]ggg LIMITING CONDITION FOR OPERATION 3.1.2.2 At least two of the following three boron injection flow pa,ths shall be OPERABLE:
- a. The flow path from the Boric Acid Storage System via a boric acid transfer pump and a charging pump to the Reactor Coolant System, and
- b. Two flow paths from the refueling water storage tank via charging pumps to the Reactor Coolant System.
APPLICABILITY: MODES 1, 2, and 3. ACTION: With only one of the above required boron injection flow paths to the Reactor Coolant System OPERABLE, restore at least two boron injection flow paths to the Reactor Coolant System to OPERABLE status within 72 hours or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least 1% Ak/k at'200 F within the next 6 hours; restore at least two flow paths to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours. SURVEILLANCE REQUIREMENTS 4.1.2.2 At least two of the above required flow paths shall be demonstrated OPERABLE:
- a. At least once per 7 days by verifying that the temperature of the heat traced portion of the flow path from the boric acid tanks is .
greater than or equal to 65*F when it is a required water source;
- b. At least once per 31 days by verifying that each valve (manual, ;
power operated, or automatic) in the flow path that is not locked, I sealed, or otherwise secured in position, is in its correct position;
- c. At least once per 18 months during shutdown by verifying that each I
automatic valve in the flow path actuates to its correct position on a Safety Injection test signal; and
- d. At least once per 18 months by verifying that the flow path required by Specification 3.1.2.2a. delivers at least 30 gpm to the Reactor Coolant System.
l
. i BRAIDWOOD - UNITS 1 & 2 3/4 1-8
REACTIVITY CONTROL SYSTEMS CHARGING PUMP - SHUTDOWN Agg I g ggg LIMITING CONDITION FOR OPERATION 3.1.2.3 One charging pump in the boron injection flow path required by Specification 3.1.2.1 shall be OPERABLE and capable of being powered from an OPERABLE emergency power source. APPLICABILITY: MODES 4*, 5, and 6. ACTION: With no charging' pump OPERABLE or capable of being powered from an OPERABLE emergency power source, suspend all operations involving CORE ALTERATIONS or positive reactivity changes. SURVEILLANCE REQUIREMENTS 4.1.2.3.1 The above required charging pump shall be demonstrated OPERABLE by verifying, on t ecirculation flow, that a differential pressure acrosr, the pump of greater than or equal to 2396 psid is developed when tested p'Jrsuant to Specification 4.0.5. 4.1.2.3.2 Whenever the temperature of one or more of the RCS cold legs is less than or equal to 330*F, all charging pumps, exclu,1ing the above required OPERABLE pump, shall be demonstrated inoperable ** at least once per 31 days, except when the reactor vessel head is removed, by verifying that the motor circuit breakers are secured in the open position.
*A maximum of one charging pump shall be OPERABLE, and that pump shall be a centrifugal charging pump, whenever the temperature of one or more of the RCS cold legs is less than or equal to 330 F. **An inoperable pump may be energized for testing provided the discharge of the pump has been isolated from the RCS by a closed isolation valve with power removed from the valve operator, or by a manual isolation valve secured in the closed position.
BRAIDWOOD - UNITS 1 & 2 3/4 1-9
REACTIVITY CONTROL SYSTEMS CHARGING PUMPS - OPERATING AUG 19 586 LIMITING CONDITION FOR OPERATION 3.1.2.4 At least two charging pumps shall be OPERABLE. - APPLICABILITY: MODES 1, 2, and 3. ACTION: With only one charging pump OPERABLE, restore at least two charging pumps to OPERABLE status within 72 hours or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least 1% Ak/k at 200*F within the next 6 hours; restore at least two charging pumps to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours. SURVEILLANCE REQUIREMENTS
,4.1.2.4. At least two charging pumps shall be demonstrated OPERABLE by verifying, on recirculation flow, that a differential pressure across each pump of greater than or equal to 2396 psid is developed when tested pursuant to Specification 4.0.5.
BRAIDWOOD - UNITS 1 & 2 3/4 1-10
REACTIVITY CCNTROL SYSTEMS BORATED WATER SOURCE - SHUTDOWN AUG 191986 LIMITING CONDITION FOR OPERATION 3.1.2.5 As a minimum, one of the following borated water sources sh'all be OPERABLE:
- a. A Boric Acid Storage System with:
- 1) A minimum contained borated water level of 7.0%,
- 2) A minimum boron concentration of 7000 ppm, and
- 3) A minimum solution temperature of 65*F.
- b. The refueling water storage tank (RWST) with:
- 1) A minimum contained borated water level of 9.0%,
- 2) A minimum boron concentration of 2000 ppm, and j 3) A minimum solution temperature of 35*F.
l APPLICABILITY: MODES 5 and 6. ACTION: With no borated water source OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes. SURVEILLANCE REQUIREMENTS 4.1.2.5 The above required borated water source shall be demonstrated OPERABLE:
- a. At least once per 7 days by:
l
- 1) Verifying the boron concentration of the water,
- 2) Verifying the contained borated water level, and
- 3) Verifying the boric acid storage tank solution temperature when it is the source of borated water.
- b. At least once per 24 hours by verifying the RWST temperature when it is the source of borated water and the outside air temperature is less than 35*F. ~
BRAIDWOOD - UNITS 1 & 2 3/4 1-11
l REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES - OPERATING AUG 191986 LIMITING CONDITION FOR OPERATION 3.1.2.6 As a minimum, the following borated water source (s) shall b.e OPERABLE as required by Specification 3.1.2.2 for MODES 1, 2 and 3 and one of the following borated water sources shall be OPERABLE as required by Specification 3.1.2.1 for MODE 4:
- a. A Boric Acid Storage System with:
- 1) A minimum contained borated water level of 40%,
- 2) A minimum boron concentration of 7000 ppm, and
- 3) A minimum solution temperature of 65 F.
- b. The refueling water storage tank (RWST) with:
- 1) A minimum contained borated water level of 89%,
- 2) A minimum boron concentration of 2000 ppm,
- 3) A minimum solution temperature of 35 F, and
- 4) A maximum solution temperature of 100 F.
APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:
- a. With the Boric Acid Storage System inoperable and being used as one of the above required borated water sources in MODE 1, 2, or 3, restore the system to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and borated to a SHUTDOWN MARGIN equivalent to at least 1% Ak/k at 200*F; restore the Boric Acid Storage System to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours.
- b. With the RWST inoperable in MODE 1, 2, or 3, restore the tank to OPERABLE status within 1 hour or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- c. With no borated water source OPERABLE in MODE 4, restore one borated water source to OPERABLE status within 6 hours or be in COLD SHUTDOWN witb*n the following 30 hours.
BRAIDWOOD - UNITS 1 & 2 3/4 1-12
s REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS 4.1.2.6 Each borated water source shall be demonstrated OPERABLE:
- a. At least once per 7 days by:
- 1) Verifying the boron concentration in the water,
- 2) Verifying the contained borated water level of the water source, and
- 3) Verifying the Boric Acid Storage System solution temperature when it is the source of borated water.
- b. At least once per 24 hours by verifying the RWST temperature when the outside air temperature is either less than 35 F or greater than 100 F, and
- c. At least once per 24 hours by verifying the RWST vent path temperature to be greater than or equal to'35*F when the outside air temperature is less than 35*F.
e BRAIDWOOD - UNITS 1 & 2 3/4 1-13
REACTIVITY CONTROL SYSTEMS 3/4.1.3 MOVABLE CONTROL ASSEMBLIES EG .I 91986 GROUP HEIGHT LIMITING CONDITION FOR OPERATION 3.1.3.1 All full-length shutdown and control rods shall be OPERABLE.and positioned within i 12 steps (indicated position) of their group step counter demand position. APPLICABILITY: MODES 1* and 2*. ACTION:
- a. With one or more full-length rods inoperable due to being immovable as a result of excessive friction or mechanical interference or known to be untrippable, determine that the SHUTDOWN MARGIN require-ment of Specification 3.1.1.1 is satisfied within 1 hour and be in HOT STANDBY within 6 hours.
- b. With more than one full-length rod inoperable or misaligned from the group step counter demand position by more than i 12 steps (indicated position), be in HOT STANDBY within 6 hours.
- c. With one full-length rod trippable but inoperable due to causes other than addressed by ACTION a. above, or misaligned from its group step counter demand height by more than i 12 steps (indicated position), POWER OPERATION may continue provided that within 1 hour:
- 1. The rod is restored to OPERABLE status within the above alignment requirements, or
- 2. The rod is declared inoperable and the remainder of the rods in the group with the inoperable rod are aligned to within i 12 steps of~the inoperable rod while maintaining the rod sequence and ir,sertion limits of Figure 3.1-1. The THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation, or
- 3. The rod is declared inoperable and the SHUTOOWN MARGIN requirement of Specification 3.1.1.1 is satisfied. POWER OPERATION may then continue provided that:
a) The THERMAL POWER level is reduced to less than or equal to 75% of RATED THERMAL POWER within the next hour and within the following 4 hours the High Neutron Flux Trip Setpoint is reduced to less than or equal to 85% of RATED THERMAL POWER. b) The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is determined at least once per 12 hours;
- See Special Test Exceptions Specifications 3.10.2 and 3.10.3. -
BRAIDWOOD - UNITS 1 & 2 3/4 1-14
REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION N 10 I986 ACTION (Continued) c) A power distribution map is obtained from the movable N incore detectors and F9(Z) and F are verified to be within their limits within 72 hours; and d) A reevaluation of each accident analysis of Table 3.1-1 is performed within 5 days; this reevaluation shall confirm that the previously analyzed results of these accidents remain valid for the duration of operation under these conditions; SURVEILLANCE REQUIREMENTS 4.1.3.1.1 The position of each full-length rod shall be determined to be within the group demand limit.by verifying the individual rod positions at least once per 12 hours except during time intervals when the rod position deviation monitor is inoperable, then verify the group positions at least once per 4 hours. 4.1.3.1.2 Each full-length rod not fully inserted in the core shall be determined OPERABLE by movement of at least 10 steps in any one direction at least once per 31 days. BRAIDWOOD - UNITS 1 & 2 3/4 1-15
TABLE 3.1-1 AUG 19 506 ACCIDENT ANALYSES REQUIRING FEEVALUATION IN THE EVENT 0F AN IN0PERABLE FULL-LENGTH R00 Rod Cluster Control Assembly Insertion Characteristics. Rod Cluster Control Assembly Misalignment. ; Loss of Reactor Coolant from Small Ruptured Pipes or fron t.r.cks in large ' Pipes Which Actuates the Emergency Core Cooling Syste:n. Single Rod Cluster Control Assembly Withdrawal at Full Power. Major Reactor Coolant System Pipe Ruptures (Loss of Coolant Accident). Major Secondary Coolant System Pipe Rupture. Rupture of a Control Rod Drive Mechanism Housing (Rod Cluster Control Assembly Ejection). e BRAIDWOOD - UNITS 1 & 2 3/4 1-16
REACTIVITY CONTROL SYSTEMS 3 ftl, POSITION INDICATION SYSTEMS - OPERATING AU818 000 LIMITING CONDITION FOR OPERATION 3.1.3.2 The Digital Rod Position Indication System and the Demand Rosition Indication System shall be OPERABLE and capable of det'ermining the control rod positions within i 12 steps. APPLICABILITY: MODES 1 and 2. ACTION:
- a. With a maximum of one digital rod position indicator per bank inoper-able either:
- 1. Determine the position of the nonindicsting rod (s) indirectly by the movable incore detectors at least once per 8 hours and immediately after any motion of the non-indicating rod which exceeds 24 steps in one direction since the last determination of the rod's position, or
- 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 8 hours.
- b. With a maximum of one bank demand position indicator inoperable either:
- 1. Verify that all digital rod position indicators for the affected bank are OPERABLE and that the most withdrawn rod and the least withdrawn rod of the bank are within a maximum of 12 steps of each other at least once per 8 hours, or
- 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 8 hours.
SURVEILLANCE REQUIREMENTS 4.1.3.2 Each digital rod position indicator shall be determined OPERABLE by l i verifying that the Bank Demand Position Indication System and the Digital Rod i Position Indication System agree within 12 steps at least once per 12 hours except during time intervals when the rod position deviation alarm is inoper- l able, then compare the Demand Position Indication System and the Digital Rod ' Position Indication System at least once per 4 hours. BRAIDWOOD - UNITS 1 & 2 3/4 1-17
REACTIVITY CONTROL SYSTEMS POSITION INDICATION SYSTEM - SHUTDOWN ggyg LIMITING CONDITION FOR OPERATION 3.1.3.3 One digital rod position indicator (excluding bank demand position indication) shall be OPERABLE and capable of determining the control rod position within i 12 steps for each shutdown or control rod not fully inserted. APPLICABILITY: MODES 3*#, 4*# and 5*#. ACTION: With less than the above required position indicator (s) OPERABLE, immediately open the Reactor Trip System breakers. SURVEILLANCE REQUIREMENTS 4.1.3.3 Each of the above required digital rod position indicator (s) shall be determined OPERABLE by verifying that the digital rod position indicator agrees with the demand position indicator within 12 steps when exercised over the full-range of rod travel at least once per 18 months.
*With the Reactor Trip System breakers in the closed position. #See Special Test Exceptions Specification 3.10.5.
1 BRAIDWOOD - UNITS 1 & 2 3/4 1-18
REACTIVITY CONTROL SYSTEMS ROD OROP TIME
.AUG 191986 LIMITING CONDITION FOR OPERATION 3.1.3.4 The individual full-length shutdown and control rod drop time from th.e fully withdrawn position shall be less than or equal to 2.4 seconds from beginning of decay of stationary gripper coil voltage to dashpot entry with:
- a. T,yg greater than or equal to 550"F, and b'. All reactor coolant pumps operating.
APPLICABILITY: MODES 1 and 2. ACTION:
- a. With the rod drop time of any full-length rod determined to exceed the above limit, restore the rod drop time to within the above limit prior to proceeding to MODE 1 or 2.
- b. With the rod drop time within limits but determined with three reactor coolant pumps operating, operation may proceed provided THERMAL POWER is restricted to less than or equal to 66% of RATED THERMAL POWER.
SURVEILLANCE REQUIREMENTS 4.1.3.4 The rod drop time of full-length rods shall be demonstrated through measurement prior to reactor criticality:
- a. For all rods following each removal of the reactor vessel head,
- b. For specifically affected individual rods following any maintenance on or modification to the Control Rod Drive System which could affect the drop time of those specific rods, and
- c. At least once per 18 months.
l BRAIDWOOD - UNITS 1 & 2 3/4 1-19
REACTIVITY CONTROL SYSTEMS SHUTDOWN ROD INSERTION LIMIT - AUG 191986 LIMITING CONDITION FOR OPERATION 3.1.3.5 All shutdown rods shall be fully withdrawn. . APPLICABILITY: MODES 1* and 2*#. ACTION: With a maximum of one shutdown rod not fully withdrawn, except for surveillance testing pursuant to Specification 4.1.3.1.2, within 1 hour either:
- a. Fully withdraw the rod, or
- b. Declare the rod to be inoperable and apply Specification 3.1.3.1.
SURVEILLANCE REQUIREMENTS 4.1.3.5 Each shutdown rod shall be determined fully withdrawn:
- a. Within 15 minutes prior to withdrawal of any rods in Control Bank A, B, C, or D during an approach to reactor criticality, and
- b. At least once per 12 hours thereafter.
- See Special Test Exceptions Specifications 3.10.2 and 3.10.3.
- With K,77 greater than or equal to 1.
BRAIDWOOD - UNITS 1 & 2 3/4 1-20
REACTIVITY CONTROL SYSTEMS CONTROL ROD INSERTION LIMITS
. AUG I 9198s LIMITING CONDITION FOR OPERATION 3.1.3.6 The control banks shall be limited in physical insertion as.shown in Figure 3.1-1.
APPLICABILITY: MODES 1* and 2*#. ACTION: With the control banks inserted beyond the above insertion limits, except for surveillance testing pursuant to Specification 4.1.3.1.2:
- a. Restore the control banks to within the limits within 2 hours, or
- b. Reduce THERMAL POWER within 2 hours to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the bank position using the above figure, or c ._ Be in at least HOT STANDBY within 6 hours.
SURVEILLANCE REQUIREMENTS 4.1.3.6 The position of each control bank shall be determined to be within the insertion limits at least once per 12 hours except during time intervals when the Rod Insertion Limit Alarm is inoperable, then verify the individual rod positions at least once per 4 hours. l 1 l "See Special Test Exceptions Specifications 3.10.2 and 3.10.3.
#With K,ff greater than or equal to 1.
I BRAIDWOOD - UNITS 1 & 2 3/4 1-21
t e FINAL DRAFT AUG 19 BB6 228 py~ /(79%,228)L..
~~
(29%,228) I 220 f f
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I 160 .,0%,162)! ( .f E 140 h iM; BANK C / a =: -; _r i 120 f f g -
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- - - - - 07 20 " - -~-~ 40 ' - ~60 80 100 RELATIVE POWER (Percent) l FIGURE 3.1-1 ROD BANK INSERTION LIMITS VERSUS THERMAL POWER FOUR LOOP OPERATION l BRAIDWOOD - UNITS 1 & 2 3/4 1-22
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3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE LIMITING CONDIT'ON FOR OPERATION 3.2.1 The in< ated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within the following target band (flux difference units) about the target flux difference: l
- a. 1 5% for Cycle 1 core average accumulated burnup of less than or i equal to 5000 MWD /MTU,-and l
- b. + 3%, -9% for Cycle 1 core average accumulated burnup of greater than 5000 MWD /MTU, and
- c. + 3%, -12% for each subsequent cycle.
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 cumulative penalty deviation time does not exceed 1 hour during the previous 24 hours. 7heindicatedAFDmaydeviateoutsidetheaboverequiredtargetbandatgreater than 15% but less than 50% of RATED THERMAL POWER provided the cumulative penalty deviation tin.e does not exceed 1 hour during the previous 24 hours. 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 above required 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 more than 1 hour of cumulative penalty deviation time during the previous 24 hours or outside the Acceptable Operation Limits of Figure 3.2-1 and with THERMAL POWER less than 90% but equal to or greater than 50% of RATED THERMAL POWER, reduce:
. 1. THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes, and
- 2. The Power Range Neutron Flux - High Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours.
"See Special Test Exceptions Specification 3.10.2. #Surveillance testing of the Power Range Neutron Flux channel may be performed pursuant to Specification 4.3.1.1 provided the indicated AFD is maintained within the Acceptable Operation Limits of Figure 3.2-1. A total of 16 hours operation may be accumulated with the AFD outside of the above required target band during testing without penalty deviation.
BRAIDWOOD - UNITS 1 & 2 3/4 2-1
LIMITING CONDITION FOR OPERATION FINAL BAFT AUG 191036 ACTION (Continued)
- c. With the indicated AFD outside of the above required target band for more than 1 hour of cumulative penalty deviation time during the previous 24 hours and with THERMAL POWER less than 50% but greater than 15% of RATED THERMAL POWER, the THERMAL POWER shall npt be increased equal to or greater than 50% of RATED THERMAL POWER until the indicated AFD is within the above required target band.
SURVEILLANCE REQUIREMENTS 4.2.1.1 The indicated AFD shall be determined to be within its limits during POWER OPERATION above 15% of RATED THERMAL POWER by:
- a. Monitoring the indicated AFD for each OPERABLE excore channel:
- 1) At least once per 7 days when the AFD Monitor Alarm is OPERABLE, and
- 2) At least once per hour for the first 24 hours after restoring the AFD Monitor Alar:n to OPERABLE status.
- b. Monitoring and logging the indicated AFD for each OPERABLE excore channel at least once per hour for the first 24 hours and at least once per 30 minutes thereafter, when the AFD Monitor Alarm is inoperable. The logged values of the indicated AFD shall be assumed to exist during the interval preceding each logging.
4.2.1.2 The indicated AFD shall be considered outside of its target band when two or more OPERABLE excore channels are indicating the AFD to be outside the target band. Penalty deviation outside of the above required target band shall be accumulated on a time basis of:
- a. One minute penalty deviation for each 1 minute of POWER OPERATION outside of the target band at THERMAL POWER levels equal to or above 50% of RATED THERMAL POWER, and
- b. One-half minute penalty deviation for each 1 minute of POWER OPERATION outside of the target band at THERMAL POWER levels between 15% and 50% of RATED THERMAL POWER.
4.2.1.3 The target flux difference of each OPERABLE excore channel shall be determined by measurement at least once per 92 Effective Full Power Days. The provisions of Specification 4.0.4 are not applicable. 4.2.1.4 The target flux difference shall be updated at least once per i 31 Effective Full Power Days by either determining the target flux difference ! pursuant to Specification 4.2.1.3 above or by linear interpolation between the most recently measured value and 0% at the end of the cycle life. The provisions of Specification 4.0.4 are not applicable. BRAIDWOOD - UNITS 1 & 2 3/4 2-2 l
l 1
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FINAL DRAFT AUG 191986 l 1 l FIGURE 3.2-1 ; AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER Ein: :
- :w .
- :3 ;
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-$(i:4-:2
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- UNACCEPTABLE li(11,90);_ '-~(11,,90)5 UNACCEPTABLE VPERATION --,-_-i- ----s OPERATION 80 f. '.
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- I ' .-
g _ rE ACCEPTABLEF: OPERATION M, t _ 1 ( 31,50) ; ; (31,50) 40 l y -- l i i 0 60 -40 -30, ~ 20 -10 0 10 20 30 40 50 FLUX DIFFERENCE (AI) % - BRAIDWOOD - UNITS 1 & 2 3/4 2-3
~
l 3/4.2.2 HEATFLUXHOTCHANNELFACTOR-Fg FINAL DRAFT LIMITING CONDITION FOR OPERATION 3.2.2 F q(Z) shall be limited by the following relationships: FO (Z) 5 [2.32] [K(Z)] for P > 0.5, and . l P Fq (Z) 5 [4.64] [K(Z)] for P 5 0.5. 1 Where p _ THERMAL POWER , RATED THERMAL POWER and K(Z) is 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% q F (Z) exceeds the limit within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours; POWER OPERATION may proceed for up to a total of 72 hours; subsequent POWER OPERATION may proceed provided the Overpower AT Trip Setpoints have been reduced at least 1% for each 1% qF (Z) exceeds the limit; and !
l
- b. Identify and correct the cause of the out-of-limit condition prior '
to increasing THERMAL POWER above the reduced limit required by ACTION a., above; THERMAL POWER may then be increased provided F (Z) 1 is demonstrated through incore mapping to be within its limit. 9 l BRAIDWOOD - UNITS 1 & 2 3/4 2-4
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8 10 12 14 16 - 0 2 4 6 CORE HEIGHT (FT) : 2 mP s FIGURE 3.2-2 h @@ K(Z)- NORMALIZED Fo(Z) AS A FUNCTION M OF CORE HEIGHT -
POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS 10 E 4.2.2.1 The provisions of Specification 4.0.4 are not applicable. 4.2.2.2 F xy shall be evaluated to determine if F (Z) is within its limit by: 9 Using the movable incore detectors to obtain a power distribution a. map at any THERMAL POWER greater than 5% of RATED THERMAL POWER;
- b. Increasing the measured F xy C mPonent of the power distribution map by 3% to account for manufacturing tolerances and further increasing the value by 5% to account for measure.?ent uncertainties;
- c. Comparing the F xy computed (F ) obtained in Specification 4.2.2.2b.,
above, to:
- 1) The F limits for RATED THERMAL POWER (FRTP) for the appropriate xy x measured core planes given in Specifications 4.2.2.2e. and f.,
below, and
- 2) The relationship:
F =FRTP x [1+0.2(1-P)] Where F is the limit for fractional THERMAL POWER operation expressed as a function of F,RTP and P is the fraction of RATED xy was measured. THERMAL POWER at which F xy according to the following schedule:
- d. Remeasuring F P
- 1. When F is greater than the F x limit for the appropriate measured core plane but less than the F relationship, additional power distribution maps shall be taken and F compared to F RTP x
and F : a) Within 24 hours after exceeding by 20% of RATED THERMAL C POWER or greater, the THERMAL POWER at which F*Y was last determined, or b) At least once per 31 EFPD, whichever occurs first, IIRAIDWOOD - UNITS 1 & 2 3/4 2-6
a POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) 10 E
- 2) When the F is less than or equal to the F limit for the appropriate measured core plane, additional power distribution maps shall be taken and F ccmpared to F and F l -at least x xy once per 31 EFPD.
- e. The F xy limits for RATED THERMAL POWER x (FRTP) shall be 1.71 for all core planes containing Bank "D" control rods and 1.55 for all unrodded core planes;
- f. The F xy limits of Specification 4.2.2.2e., above, are not applicable in the following core planes regions as measured in percent of core height from the bottom of the fuel:
- 1) Lower core region from 0 to 15%, inclusive,
- 2) Upper core region from 85 to 100%, inclusive,
- 3) Grid plane regions at 17.5 1 2%, 31.8 1 2%, 46.0 2%, 60.3 1 2%
and 74.6 1 2%, inclusive, and
- 4) Core plane regions within i 2% of core height (i 2.88 inches) about the bank demand position of the Bank "D" control rods.
L
- g. With F exceeding Fxy, the effects of Fxy on Fq (Z) shall be evaluated to determine ifqF (Z) is within its limits.
4.2.2.3 When Fq (Z) is measured for other than F xy determinations, an overall measured F (Z) shall be obtained from a power distribution map and increased 9 by 3% to account for manufacturing tolerances and further increased by 5% to account for measurement uncertainty. BRAIDWOOD - UNITS 1 & 2 3/4 2-7 O
POWER DISTRIBUTION LIMITS FINAL DRAFT AUG 1 S 1985 3/4.2.3 RCS FLOW RATE AND NUCLEAR.ENTHALPY RISE HOT CHANNEL FACTOR LIMITING CONDITION FOR OPERATION 3.2.3 IndicatedReactorCoolantSystem(RCS)totalflowrateandFh.shallbe maintained as follows for four loop operation.
- a. RCS Total Flowrate > 390,400 gps, and b.
Fh51.55[1.0+0.3(1.0-P)] where: Measured values of F H are obtained by using the movable incore detectors. An appropriate uncertainty of 4% (nominal) or greater shall then be applied to the measured value of F H before it is compared to the requirements, and THERMAL POWER p _ RATED THERMAL POWER APPLICABILITY: MODE 1. ACTION: With RCS total flow rate or Fh outside the region of acceptable operation:
- a. Within 2 hours either:
N
- 1. Restore RCS total flow rate and F AH to within the above limits, or
- 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux-High Trip Setpoint to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours.
9 BRAIDWOOD - UNITS 1 & 2 3/4 2-8 3
POWER DISTRIBUTION LIMITS g LIMITING CONDITION FOR OPERATION ACTION (Continued)
- b. Within 24 hours of initially being outside the above limits, verify through incore flux mapping and RCS total flow rate comparison that the combination of Fh and RCS total f' low rate are restored to within the above limits, or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours; and
- c. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced THERMAL POWER limit required by ACTION a.2. and/or b. above; subsequent POWER OPERATION mayproceedprovidedthatthecombinationofFhandindicatedRCS -
l total flow rate are demonstrated, through incore flux mapping and l RCS total flow rate comparison, to be within the region of acceptable operation defined by Specification 3.2.3 prior to exceeding the fol-lowing THERMAL POWER levels:
- 1. A nominal 50% of RATED THERMAL POWER,
- 2. A nominal 75% of RATED THERMAL POWEP., and
- 3. Within 24 hours of attaining greater than or equal to 95% of RATED THERMAL POWER.
SURVEILLANCE REQUIREMENTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicable. 4.2.3.2 ThecombinationofindicatedRCStotalflowrateandFhshallbedeter-mined to be within the region of acceptable operation of Specification 3.2.3:
- a. Prior to operation above 75% of RATED THERMAL POWER after each fuel loading, and
- b. At least once per 31 Effective Full Power Days.
4.2.3.3 The indicated RCS total flow rate shall be verified to be within the region of acceptable operation of Specification 3.2.3 at least once per 12 hours when the most recently obtained value of F g, obtained per Specification 4.2.3.2, is assumed to exist. 4.2.3.4 The RCS total flow rate indicators shall be subjected to a CHANNEL CALIBRATION at least once per 18 months. 4.2.3.5 The RCS total flow rate shall be determined by precision heat balance measurement at least once per 18 months. The measurement instrumentation shall be calibrated within seven days prior to the performance of the calorfmetric flow measurement. Prior to the precision heat balance measurement, at least two of the four feedwater flow meter venturis shall be visually inspected and, if fouling is found, all venturis shall be cleaned. BRAIDWOOD - UNITS 1 & 2 3/4 ?-9
POWER DISTRIBUTION LIMITS 3/4.2.4 QUADRANT POKER' TILT RATIO AUG 19 ges LIMITING CONDITION FOR OPERATION 3.2.4 The QUADRANT POWER TILT RATIO shall not exceed 1.02 above 50% of RATED THERMAL POWER. APPLICABILITY: MODE 1*. ACTION:
- a. With the QUADRANT POWER TILT RATIO determined to exceed 1.02 but less than or equal to 1.09:
- 1. Calculate the QUADRANT POWER TILT RATIO at least once per hour until either:
a) The QUADRANT POWER TILT RATIO is reduced to within its limit, or b) THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER.
- 2. Within 2 hours either:
a) Reduce the QUADRANT POWER TILT RATIO to within its limit, or b) Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1 and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours.
- 3. Verify that the QUAD; TANT POWER TILT RATIO is within its limit within 24 hours after exceeding the limit or reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within the next 2 hours and reduce the Power Range Neutron Flux-High Trip Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours, and
- 4. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours or until verified acceptable at 95%
or greater RATED THERMAL POWER.
- See Special Test Exceptions Specification 3.10.2.
BRAIDWOOD - UNITS 1 & 2 3/4 2-10
POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION N 10 M ACTION (Continued)
- b. With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to misalignment of either a shutdown or control rod:
- 1. Calculate the QUADRANT POWER TILT RATIO at least once per hour until either:
a) The Ql'ADRANT POWER TILT RATIO i.i reduced to within its limit, or b) THERMAL POWER is reduced to less than'50% of RATED THERMAL l POWER.
- 2. Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1, within 30 minutes;
- 3. Verify that the QUADRANT POWER TILT RATIO is within its limit within 2 hours after exceeding the limit or reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within the next 2 hours and reduce the Power Range Neutron Flux-High Trip Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours; and
- 4. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours or until verified acceptable at 95%
or greater RATED THERMAL POWER.
- c. With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to causes other than the misalignment of either a shutdown or control rod:
l
- 1. Calculate the QUADRANT POWER TILT RATIO at least once per hour until either:
a) The QUADRANT POWER TILT RATIO is reduced to within its limit, or b) THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER. BRAIDWOOD - UNITS 1 & 4 3/4 2-11
l POWER DISTRIBUTION LIMITS u mI1fl e AUG 191906 LIMITING CONDITION FOR OPERATION ACTION (Continued)
- 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 2 hours and reduce the Power Range Neutron Flux-High Trip Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours; and
- 3. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours or until verified at 95% or greater RATED THERMAL POWER.
- d. The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.2.4.1 The QUADRAh' F0WER 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 hours during steady-state operation when the alarm is inoperable.
4.2.4.2 The QUADRANT POWER TILT RATIO shall be determined to be within the limit when above 75% of RATED THERMAL POWER with one Power Range channel inoperable by using the movable incore detectors to confirm that the normalized symmetric power distribution, obtained from two sets of four symmetric thimble locations or a full-core flux map, is consistent with the indicated QUADRANT POWER TILT RATIO at least once per 12 hours. BRAIDWOOD - UNITS 1 & 2 3/4 2-12
POWER DISTRIBUTION LIMITS 3/4.2.5 DNB PARAMETERS - AUG 191986 LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB related parameters shall be maintained with.in the limits shown cn Table 3.2-1:
- a. Reactor Coolant System T,yg, and
- b. Pressurizer Pressure.
AFPLICABILITY: MODE 1. ACTION: With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours. SURVEILLANCE REQUIREMENTS 4.2.5 Each of the parameters of Table 3.2-1 shall be verified to be within their limits at least once per 12 hours. BRAIDWOOD - UNITS 1 & 2 3/4 2-13
TABLE 3.2-1 DNB PARAMETERS 5 O PARAMETER LIMITS I E Indicated Reactor Coolant System T avg -< 591.2*F 7 w Indicated Pressurizer Pressure -> 2219 psig* ro l R.
- '?
I 5 1 i i
- Limit not applicable during either a THERMAL POWER ramp in 2 excess of 5% of RATED THERMAL POWER pen mirute or a THERMAL POWER step in excess of 10% of RATED THERMAL POWER. -
3 g ymmans s k l
i 1 1 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION AUG 19 566 LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the Reactor Trip System instrumentation channels and interlocks of Table 3.3-1 shall be OPERABLE with RESPONSE TIMES as shown in Table 3.3-2. APPLICABILITY: As shown in Table 3.3-1. ACTION: As shown in Table 3.3-1. SURVEILLANCE REQUIREMENTS 4.3.1.1 Each Reactor Trip System instrumentation channel and interlock and the automatic trip logic shall be demonstrated OPERABLE by the performance of the Reactor Trip System Instrumentation Surveillance Requirements specified in Table 4.3-1. 4.3.1.2 The REACTOR TRIP SYSTEM RESPONSE TIME of each Reactor trip function shall be demonstrated to be within its limit at least once per 18 months. Each test shall include at least one train such that both trains are tested at least once per 36 months and one channel per function such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific Reactor trip function as shown in the
" Total No. of Channels" column of Table 3.3-1.
l l BRAIDWOO3 - UNITS 1 & 2 3/4 3-1 i l
TABLE 3.3-1 E i"; REACTOR TRIP SYSTEM INSTRUMENTATION E 8 MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION E Q 1. Manual Reactor Trip 2 1 2 1, 2 1
- 3*, 4*, 5*
2 1 2 10 w
- 2. Power Range, Neutron Flux to
- a. High Setpoint 4 2 3 1, 2 2#
- b. Low Setpoint 4 2 3 1###, 2 2#
- 3. Power Range, Neutron Flux 4 2 3 1, 2 2# M in-sum High Positive Rate , g t' 4. Power Range, Neutron Flux, 4 2 3 1, 2 2#
- High Negative Rate w
E 5. Intermediate Range, Neutron Flux 2 1 2 1###, 2 3 O
- 6. Source Range, Neutron Flux
- a. Startup 2 1 2 2##** 4
- b. Shutdown 2 1 2 3,4,5 5
- 7. Overtemperature AT 4 2 3 1, 2 6# g m
- 8. Overpower AT 4 2 3 1, 2 6# s e
- 9. Pressurizer Pressure-Low g (Above P-7) 4 2 3 1 6#*** m.
, TABLE 3.3-1 (Continue.)
REACTOR TRIP SYSTEM INSTRUMENTATION c 8 MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION
] 10. Pressurizer Pressure-High 4 2 3 1, 2 6#
- 11. Pressurizer Water Level-High (Above P-7) 3 2 2 1 6#
[
- 12. Reactor Coolant Flow-Low
- a. Single Loop (Above P-8) 3/ loop 2/ loop in 2/ loop in 1 6#
any oper- each oper-ating loop ating loop . R
- b. Two Loops (Above P-7 and 3/ loop 2/ loop in 2/ loop in 1 6#
below P-8) two oper- each oper- { ating loops ating loop
- 13. Steam Generator Water 4/stm. gen. 2/sta. gen. 3/stm. gen. 1, 2 6#***
Level-Low-Low in any each operating operating
- stm. gen. stm. gen.
- 14. Undervoltage-Reactor Coolant d
. Pumps (Above P-7) 4-1/ bus 2 3 1 6#***
- 15. Underfrequency-Reactor Coolant 2 p
Pumps (Above P-7) 4-1/ bus 2 3 1 6#
- 16. Turbine Trip w
*g
.(Above P-7)
- a. Emergency Trip Header Pressure 3/ Train 2/ Train 2/ Train 1 6#
- b. Turbine Throttle Valve Closure 4 4 1 1 6#
TABLE 3.3-1 (Continued) REACTOR TRIP SYSTEM INSTRUMENTATION E 8 o MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION E Q 17. Safety Injection Input
" from ESF 2 1 2 1, 2 9 s " 18. Reactor Coolant Pump N
Breaker Position Trip Above P-7 1/ breaker 2 1/ breaker 1 11# per operating loop .
- 19. Reactor Trip Systa Interlocks .
R
- a. Intermediate Range Neutron Flux, P-6 2 1 2 2## 8 T
- b. Low Power Reactor Trips Block, P-7 P-10 Input 4 2 3 1 8 or l P-13 Input 2 1 2 1 8 [
- c. Power Range Neutron Flux, P-8 4 2 3 1 8 h
- d. Power Range 9 nummme Neutron Flux, P-10 4 2 3 1, 2 8 Z
- e. Turbine Impulse Chamber ,
M y
. Pressure, P-13 2 1 2 1 8
- 20. Reactor Trip Breakers 2 1 2 1, 2 9 2 1 2 3*, 4*, 5* 10 E l
l 21. Automatic Trip and Interlock Logic 2 1 2 1, 2 9 l 2 1 2 3*, 4*, 5* 10
TABLE 3.3-1(Continued), TABLE NOTATIONS AUG 191986
*With the Reactor Trip System breakers in the closed position and the Control Rod Drive System capable of rod withdrawal. **The boron dilution flux doubling signals may be blocked during reactor startup. ***These. channels also provide inputs to ESF.AS. The Action Statemen't for the channels in Table 3.3-3 is more conservative and, therefore, controlling. #The provisions of Specification 3.0.4 are not applicable. ##8elow the P-6 (Intcrmediate Range Neutron Flux Interlock) Setpoint. ###Below the P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.
ACTION STATEMENTS ACTION 1 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours or be in HOT STANDBY within the next 6 hours. , ACTION 2 - With the number-of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:
- a. The inoperable channel is placed in the tripped condition within 6 hours;
- b. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours for surveillance testing of other channels per Specification 4.3.1.1; and
- c. Either, THERMAL POWER is restricted to less than or equal to 75% of RATED THERMAL POWER and the Power Range Neutron Flux Trip Setpoint is reduced to less than or equal to 85% of RATED THERMAL POWER within 4 hours; or, the QUADRANT POWER TILT RATIO is monitored at least once per 12 hours per Specification 4.2.4.2.
ACTION 3 - With the number of channels OPERABLE one less than the Minimum Channels OPERABLE requirement and with the THERMAL POWER level:
- a. Below the P-6 (Intermediate Range Neutron Flux Interlock)
Setpoint, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the P-6 Setpoint; and
- b. Above the P-6 (Intermediate Range Neutron Flux Interlock)
Setpoint but below 10% of RATED THERMAL POWER, restore the inoperable channel to OPERABLE status prior to l increasing THERMAL POWER above 10% of RATED THERMAL ( POWER. BRAIDWOOD - UNIfS 1 & 2 3/4 3-5
t i TABLE 3.3-1 (Continued) , ACTION STATEMENTS (Continued) AUG 191986 l ACTION 4 - With the number of OPERABLE channels one less than the Minimum i Channels OPERABLE requirement suspend all operations involving positive reactivity changes. ACTION 5 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement restore the inoperable channel to ! OPERABLE status within 48 hours or within the next hour open the reactor trip breakers, suspend all operations involving positive reactivity changes, and verify valves CV-111B, CV-8428, CV-8439, 2 CV-8441 and CV-8435 are closed and secured in position. With ' no channels OPERABLE verify compliance with the SHUTDOWN MARGIN requirements of Specification 3.1.1.1 or 3.1.1.2, as applicable, and take the actions stated above within 1 hour and verify compliance at it c once per 12 hours thereafter. ACTION 6 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:
- a. The inoperable channel is placed in the tripped condition within 6 hours; and
- b. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours for surveillance testing of other channels per Specification 4.3.1.1.
ACTION 7 - Deleted ACTION 8 - With less than the Minimum Number of Channels OPERABLE, within 1 hour determine by observation of the associated permissive annunciator window (s) that the interlock is in its required state for the existing plant condition, or apply Specification 3.0.3. ACTION 9 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours; however, one channel may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.1.1, provided the other channel is OPERABLE. l l ACTION 10 - With the r.Jmber of OPERABLE channels one less than the Minimum l Channels OPERABLE requirement, restore the inoperable channel l to OPERABLE status within 48 hours or open the Reactor trip breakers within the next hour. ACTION 11 - With the number of OPERABLE channels less than the Total Number of Channels, operation may continue provided the inoperable channels are placed in the tripped condition within 1 hour. BRAIDWOOD - UNITS 1 & 2 3/4 3-6 _ _ . _ . . _ . _ o _ _ _ _ _ _ _ _ _ _ .
TABLE 3.3-2 E R REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES E 8 FUNCTIONAL UNIT RESPONSE TIME E 1. Manual Reactor Trip N.A. Z
- 2. Power Range, Neutron Flux 10.5 second*
[
** 3. Power Range, Neutron Flux, " High Positive Rate N.A.
- 4. Power Range, Neutron Flux, High Negative Rate 105 second*
- 5. Intermediate Range, Neutron Flux N.A. ,
y 6. Source Range, Neutron Flux N.A.
" 7. Overtemperature AT 4 14.0 seconds *#
- 8. Overpower AT N.A.
M
- 9. Pressurizer Pressure-Low 12.0 seconds P (Above P-7)
O
- 10. Pressurizer Pressure-High 12.0 seconds g
- 11. Pressurizer Water Level-High N.A.
(Above P-7)
- Neutron detectors are exempt from response time testing. Response time of the neutron flux signal port' ion of'the channel shall be measured from detector output or input of first electronic component in channel. y ee
#Therral lag and RTD bypass manifold delay times are not included. g G
i
TABLE 3.3-2 (Continued) i"; REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES E 8 o FUNCTIONAL UNIT RESPONSE TIME E 12. Low Reactor Coolant Flow - Low Z
'^
- a. Single Loop (Above P-8) 11.0 second
- b. Two Loops (/bove P-7 and below P-8) e 11.0 second N
j 13. Steam Generator Water Level-Low-Low $2.0 seconds
- 14. Undervoltage-Reactor Coola>t Pumps (Above P-7) 51.5 seconds
- 15. Underfrequency-Reactor Cociant Pumps (Above P-7) 10.6 second
- 16. Turbine Trip (Above P-7) s* a. Emergency. Trip Header Pressure N.A.
- b. Turbine Throttle Valve Closure N.A.
- 17. Safety Injection Input from ESF N.A.
j 18. Reactor Coolant Pump Breaker Position Trip (Above P-7) N.A.
- 19. Reactor Trip System Interlocks N.A.
1 j
- 20. Reactor Trip Breakers N.A.
mamma
- 21. Automatic Trip and Interlock Logic N.A. 3 i
NM
=
- r===
l
'" tlllll3 1
E ll23 2:= 1 f
co TABLE 4.3-1 b REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS E 8 TRIP ANALOG ACTUt il.5'G MODES FOR CHANNEL DEV!CE bMICH E CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED
] FUNCTIONAL UNIT
- 1. Manual Reactor Trip N.A. N.A. N.A. R N.A. 1, 2, 3* , 4* , 5*
[
- 2. Power Range, Neutron Flux
- a. High Setpoint S O(2, 4), Q N.A. N.A. 1, 2 M(3, 4),
Q(4, 6), R(4, 5)
- b. Low Setpoint S R(4) Q N.A. N.A. 1###, 2
$ 3. Power Range, Neutron Flux, N.A. R(4) Q N.A. N.A. 1, 2 w High Positive Rate S
- 4. Power Range, Neutron Flux, N.A. R(4) Q N.A. N.A. 1, 2 High Negative Rate
- 5. Intermediate Range, S R(4, 5) Q N.A. N.A. 1###, 2 Neutron Flux
- 6. Source Range, Neutron Flux S R(4, 5, 12) Q(9) N.A. N.A. 2##, 3, 4, 5
- 7. Overtemperature AT S R(13) Q N.A. N.A. 1, 2
- 8. Overpower AT S R Q N.A. N.A. 1, 2 g
- 9. Pressurizer Pressure-Low S R Q** N.A. -N.A. 1 m" i
*(Above P-7) w P
- 10. Pressurizer Pressure-High S R Q N.A. N.A. 1, 2 E l 11. Pressuriz.er Water Level-High S R Q N.A. N.A. 1
' (Above P-7) l
TABLE 4.3-1 (Continued) E
?; REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS E
8 o TRIP ANALOG ACTUATING MODES FOR CHANNEL DEVICE WHICH E CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE g FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED
~ 12. Reactor Coolant Flow-Low S R Q N.A. N.A. 1 e.
N 13. Steam Generator Water Level- S R Q** N.A. N.A. 1, 2 Low-Low
- 14. Undervoltage-Reactor Coolant N.A. R N.A. Q** N.A. 1 Pumps (Above P-7) w 15. Underfrequency-Reactor N.A. R N.A. Q N.A. 1 i
} Coolant Pemps (Above P-7) $ 16. Turbine Trip (Above P-7) o
- a. Emergency Trip Header N.A. R N.A. S/U(1, 10) N.A. 1 Pressure
- b. Turbine Throttle Valve N.A. R N.A. S/U(1, 10) N.A. 1 Closure
- 17. Safety Injection Input from N.A. N.A. N.A. R N.A. 1, 2 s ESF
- 18. Reactor Coolant Pump Breaker N.A. N.A. N.A. R N.A. 1 Position Trip (Above P-7) 2
- 19. Reactor Trip System Interlocks Intermediate Range
~
M
- a.
- Neutron Flux, P-6 N.A. R(4) Q N.A. N.A. 2##
- b. Low Power Reactor ~
IE m Trips Block, P-7 N.A. R(4) . Q (8) N.A. N.A. 1
- c. Power Range Neutron Flux, P-8 N.A. R(4) Q (8) N.A. N.A. 1
- TABLE 4.3-1 (Continued) b o REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS b TRIP ANALOG ACTUATING MODES FOR CHANNEL DEVICE WHICH
, E CHANNEL CHANNEL GPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED
] FUNCTIONAL UNIT
- 19. Reactor Trip System Interlocks (Continued) e.
N
- d. Low Setpoint Power Range Neutron Flux, P-10 N.A. R(4) Q (8) N.A. N.A. 1, 2
- e. Turbine Impulse Chamber Pressure, P-13 N.A. R Q (8) N.A. N.A. I w 20. Reactor Trip Breaker N.A. N.A. N.A. M (7, 11) N.A. 1, 2', 3*, 4*, 5*
A w 21. Automatic Trip and Interlock H.A. N.A. N.A. N.A. M (7) 1, 2, 3*, 4*, 5*
,'. Logic s
l i A .1
- s. =lg3
~m
$es i
i II
TABLE 4.3-1 (Continued) TABLE NOTATIONS AUG 19191E
*With the Reactor Trip System breakers closed and the Control Rod Drive System capable of rod withdrawal. **These channels also provide inputs to ESFAS. The Operational Test Frequency for these channels in Table 4.3-2 is more conservative and, therefore, controlling. ##Below P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.
- Below P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.
(1) If not performed in previous 7 days. (2) Comparison of calorimetric to excore power indication above 15% of RATED THERMAL POWER. Adjust excore channel gains consistent with calorimetric power if absolute difference is greater than 2%. The provisions of Speci-fication 4.0.4 are not applicable for entry into MODE 2 or 1. (3) Single point comparison of incore to excore AXIAL FLUX DIFFERENCE above 15% of RATED THERMAL POWER. Recalibrate if the absolute difference is greater than or equal to 3%. The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1. (4) Neutron detectors may be excluded from CHANNEL CALIBRATION. (5) Initial plateau curves shall be measured'for each detector. Subsequent plateau curves shall be obtained, evaluated and compared to the initial curves. For the Intermediate Range and Power Range Neutron Flux channels the provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1. (6) Incore - Excore Calibration, above 75% of RATED THERMAL POWER. The prov1-sions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1. (7) Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS. (8) With power greater than or equal to the interlock Setpoint the required ANALOG CHANNEL OPERATIONAL TEST shall consist of verifying that the inter-lock is in the required state by observing the permissive annunciator window. (9) Surveillance in MODES 3*, 4*, and 5* shall also include verification that permissives P-6 and P-10 are in their required state for existing plant conditions by observation of the permissive annunciator window. Surveil-lance shall include verification of the Boron Dilution Alarm Setpoint of less than or equal to an increase of twice the count rate within a 10-minute period. - (10) Setpoint verification is not applicable. (11) At least once per 18 months and following maintenance or adjustment of the Reactor trip breakers, the TRIP ACTUATING DEVICE OPERATIONAL TEST shall include independent verification of the Undervoltage and Shunt trips. (12) At least once per 18 months during shutdown verify that on a simulated Boron Dilution Doubling test signal CVCS valves 112D and E open and 1128 and C close within 30 seconds. (13) CHANNEL CALIBRATION shall include the RTD bypass loops flow rate.' BRAIDWOOD - UNITS 1 & 2 3/4 3-12
l INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION M61! !!B6 LIMITING CONDITION FOR OPERATION 3.3.2 The Engineered Safety Features Actuation System (ESFAS) instrumentation channels and interlocks shown in Table 3.3-3 shall be OPERABLE with their Trip Setpoints set consistent with the values shown in the Trip Setpoint. column of Table 3.3-4 and with RESPONSE TIMES as shown in Table 3.3-5. APPLICABILITY: As shown in Table 3.3-3. ACTION:
- a. With an ESFAS Instrumentation or Interlock Trip Setpoint less con-servative than the value shown in the Trip Setpoint column but more conservative than the value shown in the Allowable Value column of Table 3.3-4 adjust the Setpoint consistent with the Trip Setpoint value.
- b. With an ESFAS Instrumentation or Interlock Trip Setpoint less con-servative than the value shown in the Allowable Values column of Table 3.3-4, either:
- 1. Adjust the Setpoint consistent with the Trip Setpoint value of Table 3.3-4 and determine within 12 hours that Equation 2.2-1 was satisfied for the affected channel, or
- 2. Declare the channel inoperable and apply the applicable ACTION statement requirements of Table 3.3-3 until the channel is restored to OPERABLE status with its Setpoint adjusted consistent with the Trip Setpoint value.
Equation 2.2-1 Z + RE +-SE 1 TA Where: Z = The value from Column Z of Table 3.3-4 for the affected channel, RE = The "as measured" value (in percent span) of rack error for the affected channel, 1 SE = Either the "as measured" value (in percent span) of the l sensor error, or the value for Column SE (Sensor Error) of l Table 3.3-4 for the affected channel, and TA = The value from Column TA (Total Allowance) of Table 3.3-4 for the affected channel.
- c. With an ESFAS instrumentation channel or interlock inoperable, take the i ACTION shown in Table 3.3-3. .
l l l BRAIDWOOD - UNITS 1 & 2 3/4 3-13 1 . - _ _ _ _ _ _ _ _ _ _ _ _ . _ _ - _ -
INSTRUMENTATION FINAL AUG 19 E66 DRAFT SURVEILLANCE REQUIREMENTS 4.3.2.1 Each ESFAS instrumentation channel and interlock and the automatic actuation logic and relays shall be demonstrated OPERABLE by the performance of the ESFAS Instrumentation Surveillance Requirements specified in Table 4.3-2. 4.3.2.2 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function shall be demonstrated to be within the limit at least once per 18 months. Each test shall include at least one train such that both trains are tested at least once per 36 months and one channel per function such that all channels are tested at least once per N times 18 months where N is the total number of redundant channels in a specific ESFAS function as shown in the " Total No. of Channels" Column of Table 3.3-3. BRAIDWOOD - UNITS 1 & 2 3/4 3-14 I i
TABLE 3.3-3 O ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION MINIMUM
' TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION E
I Z
- 1. Safety Injection (Raactor Trip, Feedwater Isolation.
Start Diesel Generators, Containment Cooling Fans, Control Room Isolation, Phase "A" Isolation, Turbine Trip, Auxiliary Feedwater, Containment Vent Isolation, and Essential Service Water). w a. Manual Initiation 2 1 2 1,2,3,4 18 1 w b. Automatic Actuation 2 1 2 1,2,3,4 14 I
, O Logic and Actuation
- ' Relays
- c. Containment 3 2 2 1,2,3 IS* -
Pressure-High-1 l
- d. Pressurizer Pressure- 4 2 3 1, 2, 3# 19*
Low (Above P-11) i
- e. Steam Line Pressure- 3/ steam line 2/ steam line 2/ steam line 1, 2, 3# 15*
i Low (Above P-11) any steam T line g 2. Containment Spray
, [M
-r-
- a. Manual Initiation 2 pair 1 pair 2 pair 1, 2, 3, 4 18
- b. Automatic Actuation 2 1 2 1,2,3,4 14 Logic and Actuation Relays
- c. Containment Pressure- 4 2 3 1,2,3 16 High-3
TABLE 3.3-3 (Continued) ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION k MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE i FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION j Z u,
- 3. Containment Isolatian
! a. Phase "A" Isolation e-
- 1) Manual Initiation 2 1 2 1,2,3,4 18
- 2) Automatic Actuation 2 1 2 1,2,3,4 14 Logic and Actuation Relays
- 3) Safety Injection See Item 1. above for all Safety Injection initiating functions and requirements. ,
w 1 b. Phase "B" Isolation
- 1) Manual Initiation 2 pair 1 pair 2 pair 1,2,3,4 18
- 2) Automatic Actuation 2 1 2 1,2,3,4 14 Logic and Actuation Relays
- 3) Containment 4 2 3 1,2,3 16 -
Pressure-High-3
- c. Containment Vent Isolation
- 1) Automatic Actuation 2 1 2 1,2,3,4 17 T
Logic and Actuation Relays ! 2) Manual Phase "A" See Item 3.a.1 for all manual Phase "A" Isolation initiating Isolation functions and requirements.
~
- 3) Manual Phase "B" See Item 3.b.1 for all manual Phase "B" Isolation initiating Isolation functions and requirements. M
- 4) Safety Injection See Item 1. above for all Safety Injection initiating functions and requirements.
I
TABLE 3.3-3 (Continued) o ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION E o i ' MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE E FUNCTIONAL UNIT OF CHANNELS TO TRIP .0PERABLE MODES ACTION Z s.-
- 4. Steam Line Isolation
- e. ~a. Manual Initiation m 1) Individual 1/ steam line 1/ steam line 1/ operating 1, 2, 3 23 steam line l 2) System 2 1 2 1,2,3 22
- b. Automatic Actuation 2 1 2 1,2,3 21 Logic and Actuation Relays ,
; t' c. Containment Pressure- 3 2 2 1,2,3 15* ~
High-2 w
- d. Steam Line 3/ steam line 2/ steam line 2/ steam line 1, 2, 3# 15*
' " Pressure-Low any steam (above P-11) line
- e. Steam Line Pressure - 3/ steam line 2/ steam line 2/ steam line 3## 15*
Negative Rate-High any steam j (below P-11) line
- 5. Turbine Trip &
Feedwater Isolation
; a. Automatic Actuation 2 1 2 1, 2 24 "T1 1 Logic and Actuation m j Relays 2 ! b. Steam Generator 4/sta. gen. 2/sta. gen. 3/sta. gen. 1, 2 19* M Water Level- in any oper in each oper- '
w P High-High (P-14) ating stm. ating stm. *
- c. Safety Injection gen. gen.
See Item 3. above for all Safety Injection initiating functions and gQ requirements.
TABLE 3.3-3 (Continued) I E B O ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION 6 O
' MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE -
E FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES - ACTION d 6. Auxiliary Feedwater o.. a. Manual Initiation 2 1 2 1,2,3 22 m b. Automatic Actuation Logic 2 1 2 1,2,3 21 and Actuation Relays
- c. Stm. Gen. Water Level-Low-Low i
- 1) Start Motor-Driven Pump 4/sta. gen. 2/sta. gen. 3/sta. gen. 1, 2, 3 19* -
R in any opera- in each ting sta gen. operating Y stm. gen, i 5 1
- 2) Start Diesel-I Driven Pump 4/sta. gen. 2/sta. gen. 3/sta. gen. 1, 2, 3 19*
I in any in each ! operating operating l sta.' gen. stm. gen. g m l d. Undervoltage - RCP 4-1/ bus 2 3 1, 2 19* s Bus-Start Motor-
- Driven Pump and "F1 l
I Diesel-Driven Pump g - 2 l e. Safety I. ejection - M y-- ! Start Motor-Driven Pump See Item 1. above for all Safety Injection initiating functions and j and Diesel-Driven Pump requirements. g
- f. Division 11 for Unit 1 (Division 21 for Unit 2)
ESF Bus Undervoltage- < j Start Motor-Driven Pump (Start as part 2 2 - 2 1,2,3,4 2Sa* i of DG sequencing)
TABLE 3.3-3 (Continued) o ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION 75 O o MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE E FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION "e
- 6. Auxiliary Feedwater (Continued)
- g. Auxiliary Feed-i water Pump Suction Pressure-Low (Transfer to Essential Service Water) 2 2 2 1,2,3 15*
- 7. Automatic Opening of i
m Containment Sump Suction
) Isolation Valves , 4 a. Automatic Actuation Logic 2 1 2 1,2,3,4 14
!
- and Actuation Relays
- b. RWST Level - Low-Low 4 2 3 1,2,3,4 16 Coincident With Safety Injection See Item 1. above for Safety Injection initiating functions and requirements.
l 8. Loss of Power l a. ESF Bus Undervoltage 2/ Bus 2/ Bus 2/ Bus 1, 2, 3, 4 25a* M
======
- b. Grid Degraded Voltage 2/ Bus 2/ Bus NE 2/ Bus 1, 2, 3, 4 25b* p F"""
NQ E ZI
TARLF 3.3-3 (Continued) ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION E 8 o MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE g FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 4
- 9. Engineered Safety Features W Actuation System Interlocks N a. Pressurizer Pressure, 3 2 2 1,2,3 20 P-11
- b. Reactor Trip, P-4 4-2/ Train 2/ Train 2/ Train 1, 2, 3 22
- c. Low-Low T,yg, P-12 4 2 3 1,2,3 20 ,
$ d. Steam Generator Water Level, 4/sta. 2/sta. gen. 3/sta. 1, 2 20 m P-14 (High-High) gen. in any gen. in 4 operating each o stm. gen. operating stm. gen.
- 10. Steam Generator Blowdown ( ) ( ) ( ) ( ) ( )
System HELB Isolation
- 11. Auxiliary Steam System ( ) ( ) ( ! ( ) ( )
HELB Isolation Y i THIS PAGE OPEN PENDNG REN OF gg NFORMADON FROM WE APPLICANT - [P
$ :C
- cs3
FINAL HAFT TABLE 3.3-3 (Continued) AUG 191986 TABLE NOTATIONS l
# Trip function may be blocked in this MODE below the P-11 (Pressurizer Pressure Interlock) Setpoint.
## Trip function automatically blocked above P-11 and may be blocked below P-11 when Safety Injection on low steam line pressure is not blocked.
*The provisions of Specification 3.0.4 are not applicable.
ACTION STATEMENTS ACTION 14 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours; however, one channel may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.2.1, provided the other channel is OPERABLE. ACTION 15 - With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed until performance of the next required ANALOG CHANNEL OPERATIONAL TEST provided the inoperable channel is placed in the tripped condition within 1 hour. ACTION 16 - With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed provided the inoperable channel is placed in the bypassed condition and the Minimum Channels OPERABLE requirement is met. One additional channel may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.2.1. ACTION 17 - With less than the Minimum Channels OPERABLE requirement, operation may continue provided the containment purge supply and exhaust valves are maintained closed. ACTION 18 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following i 30 hours. BRAIDWOOD - UNITS 1 & 2 3/4 3-21
TABLE 3.3-3 (Continued) ACTION STATEMENTS (Continued) AUG 19 m ACTION 19 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:
- a. The inoperable channel is placed in the tripped condition within 1 hour, and ,
- b. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 2 hours for sur-veillance testing of other channels per Specification 4.3.2.1.
ACTION 20 - With less than the Minimum Number of Channels OPERABLE, within 1 hour determine by observation of the associated permissive annunciator window (s) that the interlock is in its required state for the existing plant condition, or apply Specification 3.0.3. ACTION 21 - With the number of OPERABLE Channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours and in at least HOT SHUTDOWN within the fcellowing 6 hours; however, one channel may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.2.1 provided the other channel is OPERABLE. ACTION 22 - With the number ~of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours or be in at least HOT STANDBY within 6 hours and in at least HOT SHUTDOWN within the following 6 hours. ACTION 23 - With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE
. status within 48 hours or declare the associated valve inoperable and take the ACTION required by Specification 3.7.1.5.
ACTION 24 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement be in at least HOT STANDBY within 6 hours; however, one channel may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.2.1 provided the other channel is OPERABLE. ACTION 25 - a. With the number of OPERABLE channels one less than the Minimum Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the inoperable channel is placed in the tripped condition within 1 hour. The inoperable channel may be bypassed for up to 2 hours for surveillance testing of the OPERABLE channel per Specification 4.3.2.1.
- b. With the number of OPERABLE channels one less than the Mini-mum Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the inoperable channel is placed in the tripped condition within 1 hour. .
BRAIDWOOD - UNITS 1 & 2 3/4 3-22
TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS TOTAL SENSOR TRIP ALLOWABLE o FUNCTIONAL UNIT ALLOWANCE (TA) Z ERROR (SE) SETPOINT VALUE E 1. Safety Injection Q (Reactor Trip, Feedwater Isolation, Start Diesel Generators, Containment
- Cooling Fans, Control Room Isolation, Phase "A" Isolation, Turbine Trip, Auxiliary Feedwater, Containment Vent Isolation and Essential Service Water)
$ a. Manual Initiation N.A. N.A. N.A. N.A. N.A.
k" b. Automatic Actuation Logic and Actuation N.A. N.A. N.A. N.A. N.A. Relays
- c. Containment Pressure-High-1 5.7 0.71 1.5 5 3.4 psig 5 5.8 psig
- d. Pressurizer Pressure-l Low (Above P-11) 16.1 14.41 1.5 t 1829 psig 1 1823 psig
- e. Steam Line Pressure-Low (Above P-11) 21.2 14.81 1.5 1 640 psig* -> 617 psig*
M
- 2. Containment Spray > J2
, gp
- a. Manual Initiation N.A. w t="
N.A. N.A. N.A. N.A. as
- b. Automatic Actuation .
Logic and Actuation Relays N.A. N.A. N.A. N.A. N.A.
- c. Containment Pressure-High-3 8.0 0.71 1.5 5 20.0 psig 5 21.0 psig
TABLE 3.3-4 (Continued) E R ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS TOTAL SENSOR TRIP ALLOWABLE FUNCTIONAL UNIT ALLOWANCE (TA) Z, ERROR (SE) SETPOINT VALUE E 3. Containment Isolation 53 a. Phase "A" Isolation *
'^
- 1) Manual Initiation N.A. N.A. N.A. N.A. N.A.
- 2) Automatic Actuation Logic and Actuation Relays N.A. N.A. N.A. N.A. N.A.
- 3) Safety Injection See Itee 1. above for all Safety Injection Trip Setpoints and Allowable Values.
~
m b. Phase "B" Isolation l w 1) Manual Initiation N.A. N.A. N.A. N.A. N.A. A
- 2) Automatic Actuation N.A. N.A. N.A. N.A. N.A.
Logic and Actuation Relays
- 3) Containment Pressure-High-3 8.0 0.71 1.5 5 20.0 psig $ 21.0 psig
- c. Containment Vent Isolation
- 1) Automatic Actuation M
=m====
Logic and Actuation 2 Relays N.A. N.A. N.A. N.A. N.A. gg
- 2) Manual Phase "A" N.A. N.A. N.A. N.A. N.A.
- Isolation *
- 3) Manual Phase "B" N.A. N.A. N.A. N.A. N.A.
Isolation
- 4) Safety Injection See Ites 1 above for all Safety Injection Trip Setpoints and Allowable Values. "
TABLE 3.3-4 (Continued) E
$ ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS E
o
~
TOTAL SENSOR TRIP ALLOWABLE FUNCTIONAL UNIT ALLOWANCE (TA) Z ERROR (SE) SETPOINT VALUE E 4. Steam Line Isolation M [ a. Manual Initiation N.A. N.A. N.A. N.A. N.A.
> b. Automatic Actuation m Logic and Actuation Re1ays N.A. N.A. N.A. N.A. M.A.
- c. Containment Pressure- '
High-2 7.7 0.71 1.5 18.2 psig 19.2 psig
- d. Steam Line Pressure- 21.2 14.81 1.5 >640 psig* >617 psig*
, Low (Above P-11) -
1 e. Steam Line Pressure w Negative Rate-High 8.0 0.5 0 1100 psi"* 1111.5 psi ** l 4 w (Below P-11) l S. Turbine Trip and Feedwater Isolation l a. Automatic Actuation Logic and Actuation Relays N.A. N.A. N.A. N.A. N.A. i b. Steam Generator Water Level-High-High (P-14) 6.0 M
- 1) Unit 1 4.28 1. 5 181.4% of 182.7% of narrow range narrow range E instrument instrument 3 M span span gF
- 2) Unit 2 5.0 2.18 1.5 178.1% of $79.9% of y
) narrow range instrument narrow range instrument eo e span span R
TABLE 3.3-4 (Continued) ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS TOTAL SENSOR TRIP ALLOWABLE o FUNCTIONAL UNIT ALLOWANCE (TA) Z ERROR (SE) SETPOINT VALUE E 5. Turbine Trip and g Feedwater Isolation (continued)
- c. Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and
- e. Allowable Values.
m
- 6. Auxiliary Feedwater
- a. Manual Initiation N.A. N.A. N.A. N.A. N.A.
- b. Automatic Actuation Logic and Actuation '
w Relays N.A. N.A. N.A. N.A. N.A. s
- c. Steam Generator Water Y Level-Low-Low-Start g Motor-Driven Pump and Diesel-Driven Pump
- 1) Unit 1 27.1 18.28 1. 5 140.8% of 139.1% of narrow range narrow range instrument instrument span span
- 2) Unit 2 17.0 14.78 1.5 117% of 115.3% of narrow range narrow range instrument instrument span span
- d. Undervoltage-RCP Bus- N.A. N.A. N.A. ->5268 volts ->4728 volts -
Start Motor Driven Pump Z and Diesel-Driven Pump gg
- e. Safety Injection-
[ F""'" Start Motor- e Driven Pump and Diesel-Driven Pump See Item 1. above for all Safety Injection Trip Setpoints and Allowable Values. m Q R m
TABLE 3.3-4 (Continued) ENGINEERED SAFETY FEATURES ACTUATION SYSTEN INSTRUMENTATION TRIP SETPOINTS TOTAL SENSOR TRIP ALLOWABLE FUNCTIONAL UNIT ALLOWANCE (TA) Z ERROR (SE) SETPOINT VALUE E
~
- 6. Auxiliary Feedwater (Continued)
-e ~
- f. Division 11 for Unit 1 (Division 21 for Unit 2)
ESF Bus Undervoltage-Start Motor-Driven Pump N.A. N.A. N.A. 2870 volts 2730 volts Auxiliary Feedwater g. Pump Suction Pressure-Low (Transfer to . w Essential Service 1 Water) N.A. N.A. N.A. 1.22" Hg vac 2" Hg vac Y m
" 7. Automatic Opening of Containment Sump Suction Isolation Valves
- a. Automatic Actuation N.A. N.A. N.A. N.A. N.A.
Logic and Actuation Relays T
- b. RWST Level-Low-Low N.A. N.A. N.A. 46. 3 44. A Coincident with Safety Injection See Item 1. above for Safety Injection Trip Setpoints and Allowable Values. M P
R@ y NE Q
TABLE 3.3-4 (Continued) E R ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS TOTAL SENSOR TRIP ALLOWABLE o FUNCTIONAL UNIT ALLOWANCE (TA) Z ERROR (SE) SETPOINT VALUE E 8. Loss of Power Z
- a. ESF Bus Undervoltage N.A. N.A. N.A. 2870 volts >2730 volts
~
- w/1.8s delay w/51.9s delay
- b. Grid Degraded N
Voltage N.A. N.A. N.A. 3804 volts 13728 volts w/310s delay w/310 1 30s delay
- 9. Engineered Safety Feature Actuation System Interlocks w
} a. Pressurizer Pressure, m P-11 N.A. N.A. N.A. 51930 psig $1936 psig i
m b. Reactor Trip, P-4 N.A. N.A. N.A. N.A. N.A.
- c. Low-Low T,yg, P-12 N.A. N.A. N.A. 1550*F 1547.6*F ,
- d. Steam Generator Water See Item 5.b. above for all Steam Generator Water Level Trip
! Level, P-14 Setpoints and Allowable Values. (High-High)
- 10. Steam Generator Blowdown ( ) ( ) ( ) ( ) ( )
System HELB Isolation
- 11. Auxiliary Steam System HELB Isolation
( ) ( ) ( ) ( ) ( ) 3 g M F e THIS PAGE OPEN PENDING RECEIPT OF gmy 950RMATION FROM THE APPUCANT A
TABLE 3.3-4 (Continued) TABLE NOTATIONS
- Time constants utilized in the lead-lag con'. roller for Steam Line Pressure-Low are I, > 50 seconds and t2 < 5 seconds. CHANNEL CALIBRATION shall ensure that these time constants are adjusted to those values.
**The time constant utilized in the rate-lag gon: roller for Steam Line Pressure -
Negative Rate - High is greater than or equal to 50 seconds. CHANNEL CALIBRA-TION shall ensure that this time constant is adjusted to this value.
- l BRAIDWOOD - UNITS 1 & 2 3/4 3-29
FINAL DRAFT TABLE 3.3-5 10 % ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS
- 1. Manual Initiation
- a. Safety Injection (ECCS) N.A.
- b. Containment Spray N.A.
- c. Phase "A" Isolation . N.A. -
- d. Phase "B" Isolation N.A.
- e. Containment Vent Isolation N.A.
- f. Steam Line Isolation N.A.
- g. Feedwater Isolation N.A.
- h. Auxiliary Feedwater N.A.
- i. Essential Service Water N.A.
j Containment Cooling Fans N.A.
- k. Start Diesel Generator N.A.
- 1. Control Room Isolation N.A.
- m. Turbine Trip N.A.
- 2. Containment Pressure-High-1
- a. Safety Injection (ECCS) < 27(1)/12(5)
- 1) Reactor Trip 52
- 2) Feedwater Isolation 17(3)
- 3) Phase "A" Isolation 12(6)
- 4) Containment Vent Isolation 17
- 5) Auxiliary Feedwater 160
- 6) Essential Service Water 142(1)
- 7) Containment Cooling Fans 140(1)
- 8) Start Diesel Generator 112
- 9) Control Room Isolation N.A.
- 10) Turbine Trip. N.A.
- 3. Pressurizer Pressure-Low
- a. Safety Injection (ECCS) 127(1)/12(5)
- 1) Reactor Trip 12
- 2) Feedwater Isolation _7(3)
- 3) Phase "A" Isolation 12(6) .
- 4) Containment Vent Isolation <
_7 BRAIDWOOD - UNITS 1 & 2 3/4 3-30 l
FINAL DRAFT TABLE 3.3-5 (Continued) AUGI9 g ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS
- 3. Pressurizer Pressure-Low (continued)
- 5) Auxiliary Fcedwater <60 ,
- 6) Essential Service Water 142(1)
- 7) Containment Cooling Fans 140(1)
- 8) Start Diesel Generator 112
- 9) Control Room Isolation N.A.
- 10) Turbine Trip N.A.
- 4. Steam Line Pressure-Low
- a. Safety Injection (ECCS) 122(4)/12(5)
- 1) Reactor Trip 12
- 2) Feedwater Isolation _7(3)
- 3) Phase "A" Isolation 12(6)
- 4) Containment Vent Isolation <
_7
- 5) Auxiliary Feedwater 160
- 6) Essential Service Water 142(1)
- 7) Containment Cooling Fans 140(1)
- 8) Start Diesel Generator 112
- 9) Control Room Isolation N.A.
- 10) Turbine Trip N.A.
- b. Steam Line Isolation 17
- 5. Containment Pressure-High-3
- a. Containment Spray <45(1)
- b. Phase "B" Isolation 122(1)/12(2)
- 6. Steam Generator Water Level-High-High
- a. Turbine Trip 12.5
- b. Feedwater Isolation 17(3)
BRAIDWOOD - UNITS 1 & 2 3/4 3-31
FINAL DRAFT TABLE 3.3-5 (Continued) ) ENGINEERED SAFETY FEATURES RESPONSE TIMES AUG 19 m INITIATING SIGNAL AND FUNCTION -
. RESPONSE TIME IN SECONOS
- 7. Steam Generator Water Level-Low-Low
- a. Motor-Driven Auxiliary Feedwater Pump -
<60
- b. Diesel-Driven Auxiliary
- Feedwater Pumps 160
- 8. Containment Pressure-High-2 Steam Line Isolation 17
- 9. RWST Level-Low-Low Coincident with Safety Injection -
Automatic Opening of Containment <100 Sump Suction Isolation Valves
- 10. Undervoltage RCP Bus
- a. Motor-Driven Auxiliary Feedwater Pump $60
- b. Diesel-Driven Auxiliary Feedwater Pump .
106
- 11. Division 11 for Unit 1 (Division 21 for Unit 2) ESF Bus Undervoltage Motor-Driven Auxiliary Feedwater Pump 106
- 12. Loss of Power
- a. ESF Bus Undervoltage <1.9
- b. Grid Degraded Voltage 5310*30 delay
- 13. Steam Line Pressure - Negative Rate-High (Below P-11)
Steam Line Isolation 17 o*
- 14. Phase "A" Isolation 9:
D3a. Containment Vent Isolation 17 55 % l ea
- 15. Auxiliary Feedwater Pump Suction E6 Pressure-Low-Low
@0$
w Automatic Switchover to ESW N.A. F6 2
- 16. Area Temoerature O9 m
- Steam Generator Blowdown System a.
HELB Isolation ( )
'O I f
- b. Auxiliary Steam System HELB Isolation ( ) $S 53 <
BRAIDWOOD - UNITS 1 & 2 3/4 3-32
FINAL DRAFT TABLE 3.3-5 (Continued) AUG 191986
, TABLE NOTATIONS (1) Diesel generator starting and sequence loading delays included.
(2) Diesel generator starting and sequence loading delay not included. Offsite power available. (3) Hydraulic operated valves. (4) Diesel generator starting and sequence loading delay included. Only centrifugal charging pumps included. (5) Diesel generator starting and sequence loading delays not included. Offsite power available. Only centrifugal charging pumps included. (6) Does not include valve closure time. e 5 BRAIDWOOD - UNITS 1 & 2 3/4 3-33
TABLE 4.3-2 ' ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRIMENTATION g SURVEILLANCE REQUIREMENTS
- O TRIP ANALOG ACTUATING MODES E CHANNEL DEVICE MASTER SLAVE FOR WHICH Z CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE
FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED
- 1. Safety Injection (Reactor Trip,
" Feedwater Isolation, Start Diesel Generators, Containment Cooling Fans, Control Room Isolation, Phase "A" Isolation, Turbine Trip, Auxiliary Feedwater, Containment Vent Isolation and '
g Essential Service Water)
- a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1, 2, 3, 4
- b. Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q 1,2,3,4 Logic and Actuation Relays
- c. Containment Pressure- S R M N.A. N.A. N.A. N.A. 1, 2, 3 High-1
- d. Pressurizer Prcssure- S R M N.A. N.A. N.A. N.A. 1,2,3 <
Low (Above P-11)
- e. Steam Line Pressure- S R M N.A. N.A. N.A. N.A. 1, 2, 3 Low (Above P-11)
- 2. Containment Spray -
l a. Manual Initiation N.A. N.A. N.A. R N.A. N. A,. N.A. 1, 2, 3, 4
- b. Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q 1,2,3,4 i
Logic and Actuation Relays
- c. ntd nment Pressure- S R M N.A. N.A. . ,
1 l
D m TABLE 4.3-2 (Continued) b ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS TRIP ANALOG ACTUATING MODES E CHANNEL DEVICE MASTER SLAVE FOR WHICH Q CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE '^ FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED
- 3. Containment Isolation m
- a. Phase "A" Isolation
- 1) Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A 1, 2, 3, 4
- 2) Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q 1,2,3,4 Logic and Actuation Relays
$ 3) Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements. Y b. Phase "B" Isolation U 1) Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A 1,2,3,4
- 2) Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q 1,2,3,4 Logic Actuation Relays
- 3) Containment S R M N.A. N.A. N.A. N.A. 1, 2, 3 Pressure-High-3
- c. Containment Vent Isolation
- 1) Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q 1,2,3,4 Logic and Actuation Relays 0
="" FINAL DRAFT
TABLE 4.3-2 (Continued) E R ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS TRIP ANALOG ACTUATING MODES E CHANNEL DEVICE MASTER SLAVE FOR WHICH Q CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED [ FUNCTIONAL UNIT 3.c. Containment Vent Isolation (Continued)
- 2) Manual Phase "A" See Item 3.a.1 above for all manual Phase "A" Isolation Surveillance Requirements.
Isolation
- 3) Manual Phase "B" See Item 3.b.1 above for all manual Phase "B" Isolation Surveillance Requirements.
Isolation
- 4) Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.
R
- 4. Steam Line Isolation .
- a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1,2,3
- b. Automatic Actuation N.A. N.A N.A N.A. M(1) M(1) Q 1,2,3 Logic and Actuation Relays
- c. Containment Pressure- S R M N.A. N.A. N.A. N.A. 1,2,3 High-2
- d. Steam Line Pressure- S R M N.A. N.A. N.A. N.A. 1, 2, 3 Low (Above P-11)
- e. Steam Line Pressure S R M N.A. N.A. N.A. N.A. 3E
- Negative Rate - High
- E (Below P-11) $%
- 5. Turbine Trip and Feedwater h
~
Isolation g
- a. Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q 1, 2 Logic and Actuation Relay
TABLE 4.3-2 (Continued) E D ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS
' TRIP ANALOG ACTUATING MODES E CHANNEL DEVICE MASTER SLAVE FOR WHICH Z CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE
[ FUNCTIONAL UNIT CHECK CALIBRATION TEST - TEST LOGIC TEST TEST TEST IS REQUIRED l [ 5. Turbine Trip and Feedwater (Continued)
- b. Steam Generator Water S R M N.A. N.A. N.A. N.A. 1, 2 Level-High-High (P-14)
- c. Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.
- 6. Auxiliary Feedwater R a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1,2,3 w b. Automatic Actuation N.A. N.A N.A. N.A. M(1) M(1) Q 1,2,3
- g Logic and Actuation Relay I
- c. Steam Generator Water S R M N.A. N.A. N.A. N.A. 1,2,3 Level-Low-Low
- d. Undervoltage-RCP Bus N.A. R N.A. M N.A. N.A. N.A. 1, 2
- e. Safety Injection See Ites 1. above for all Safety Injection Surveillance Requirements.
) f. Division 11 for Unit 1 N. A. R N.A. M(2) N.A. N.A. N.A 1,2,3,4 (Division 21 for Unit 2) ESF Bus Undervoltage
- g. Auxiliary Feedwater S R M N.A. N.A. N.A. N.A. 1, 2, 3 Pump Suction Pressure-j Low I 7. Automatic Opening of Containment Sump Suction
-- FINAL DRAFT
TABLE 4.3-2 (Continued) ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS E
?; TRIP o ANALOG ACTUATING MODES k
o CHANNEL DEVICE MASTER SLAVE FOR WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE e FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST ' TEST TEST IS REQUIRED
- 7. Automatic Opening (Continued)
- a. Automatic Actuation
[ Logic and Actuation
, Relays N.A. N.A. N.A. N.A. M(1) M(1) Q 1, 2, 3, 4
- b. RWST Level-Low-Low S R H N.A. N.A. N.A. N.A. 1,2,3,4 Coincident With Safety Injection See Item 1. above for all Safety Injection Surveillance Requiremente
- 8. Loss of Power
, a. ESF Bus Undervoltage N.A. R N.A. M(2) N.A. N.A. N.A. 1,2,3,4 D b. Grid Degraded Voltage N.A. R N.A. M(3) N.A. N.A. N.A. 1,2,3,4 Y 9. Engineered Safety Feature
$ Actuation System Interlocks
- a. Pressurizer Pressure, N.A. R M N.A. N.A. N.A. N.A. 1,2,3 yg P-11 3 Ui b. Reactor Trip, P-4 N.A. N.A. N.A. R N.A. N.A. N.A. 1,2,3 g c. Low-Low T,yg, P-12 N.A. R H N.A. N.A. N.A. N.A. 1, 2, 3
- d. Steam Generator Water S R M N.A. M(1) M(1) Q 1, 2 e
O *Q Level, P-14 Zg (High-High) l8;y g S 10. Steam Generator Blowdown System HELB Isolation ( ) ( ) ( ) ( ) ( ) ( )( )(
)g y2 mO > n
- 11. Auxiliary Steam System HELB Isolation
( ) ( ) ( ) ( ) ( ) ( -) ( )(
)[ O e %R C=
TABLE NOTATION g j (1) Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS. O (2) Undervoltage relay operability is to be verified independently. An inoperable channel may be bypassed n for up to 2 hours for surveillance testing of the OPERABLE channel per Specification 4.3.2.1. (3) Verify undervoltage relay operability and alarm annunciation only. Do not maintain an undervoltage condition beyond the setpoint for the time delay relay.
INSTRUMENTATION 3/4.3.3 MONITORING INSTRUMENTATION AUG 191986 RADIATION HONITORING FOR PLANT OPERATIONS LIMITING CONDITION FOR OPERATION 3.3.3.1 The radiation monitoring instrumentation channels for plant operations shown in Table 3.3-6 shall be OPERABLE with their Alarm / Trip Setpoints within the specified limits. APPLICABILITY: As shown in Table 3.3-6. ACTION:
- a. With a radiation monitoring channel Alarm / Trip Setpoint for~ plant operations exceeding the value shown in Table 3.3-6, adjust the Setpoint to within the limit within 4 hours or declare the channel inoperable.
- b. With one or more radiation monitoring channels for plant operations inoperable, take the ACTION shown in Table 3.3-6.
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.3.3.1 Each radiation monitoring instrumentation channel for plant operations shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and DIGITAL CHANNEL OPERATIONAL TEST for the MODES and at the frequencies shown in Table 4.3-3. BRAIDWOOD - UNITS 1 & 2 3/4 3-39
TABLE 3.3-6 .E R RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS E 8 o MINIMUM CHANNELS CHANNELS APPLICABLE ALARM / TRIP E FUNCTIONAL UNIT TO TRIP / ALARM OPERABLE MODES SETPOINT ACTION 4 m 1. Fuel Building Isolation-H Radioactivity-High and { Criticality (ORE-AR055/56) 1 2
<5 mR/h 29
- 2. Containment Isolation-Containment Radioactivity-High a) Unit 1 (1RE-AR011/12) 1 2 All ** 26 b) Unit 2 (2RE-AR011/12) 1 2 All ** 26 .
$ 3. Gaseous Radioactivity-w RCS Leakage Detection i a) Unit 1 (IRE-PR011B) N.A. I 1,2,3,4 N.A. 28 o
b) Unit 2 (2RE-PR0118) N.A. 1 1,2,3,4 N.A. 28
- 4. Particulate Radioactivity-RCS Leakage Detection a) Unit 1 (1RE-PR011A) N.A. 1 1,2,3,4 N.A. 28 b) Unit 2 (2RE-PR011A) N.A. 1 1,2,3,4 N.A. 28 augmg
====
- 5. Main Control Room Isolation-Outside Air Intake-Gaseous 2 Radioactivity-High g y
(0RE-PR031B/328 and ORE-PR0338/348) 1 2 per All < 2 mR/h 27 m intake g 8Q 21
FINAL DRAFT TABLE NOTATIONS AUG 191986
*With new fuel or irradiated fuel in the fuel storage areas or fuel building. ** Trip Setpoint is to be established such that the actual submersion dose rate would not exceed 10 mR/hr in the containment building. For containment purge or vent the Setpoint value may be increased up to twice the maximum concentra-tion activity in the containment determined by the sample analysis performed prior to each release in accordance with Table 4.11-2 provided the value does not exceed 10% of the equivalent limits of Specification 3.11.2.1.a in accord-ance with.the methodology and parameters in the ODCM. 1 1
ACTION STATEMENTS ACTION 26 - With less than the Minimum Channels OPERABLE requirement, operation may continue provided the containment purge valves are maintained closea. ACTION ~27 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, within 1 hour isolate the Control Room Ventilation System and initiate operation of the Control Room Make-up System. ACTION 28 - Must satisfy the ACTION requirement for Specification 3.4.6.1. ACTION 29 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, ACTION a. of Specification 3.9.12 must be satisfied. With both channels inoperable, provide an appropriate portable continuous monitor with the same Alarm Set-point in the fuel pool area with one Fuel Handling Building Exhaust filter plenum in operation. Otherwise satisfy ACTION b. of Specification 3.9.12. BRAIDWOOD - UNITS 1 & 2 3/4 3-41
cm TABLE 4.3-3 k RADIATION MONITORING INSTRUMENTATION FOR PLANT g OPERATIONS SURVEILLANCE REQUIREMENTS 8 DIGITAL E CHANNEL Q CHANNEL CHANNEL OPERATIONAL MODES FOR WHICH FUNCTIONAL UNIT CHECK CALIBRATION TEST
] SURVEILLANCE IS REQUIRED 8*
N
- 1. Fuel Building Isolation-Radioactivity-High and Criticality (ORE-AR055/56) S R M *
- 2. Containment Isolation-Containment Radioactivity-High ,
w a) Unit 1 (1RE-AR011/12) S R M All b b) Unit 2 (2RE-AR011/12) S R M All w 1 N
- 3. Gaseous Radioactivity-RCS Leakage Detection a) Unit 1 (IRE-PR0118) S R M 1,2,3,4 b) Unit 2 (2RE-PR0118) S R M 1,2,3,4
- 4. Particulate Radioactivity-
- RCS Leakage Detection a) Unit 1 (IRE-PR011A) S R ,M 1,2,3,4 b) Unit 2 (2RE-PR011A) S R M 1,2,3,4
- 5. Main Control Room Isolation- M Outside Air Intake-Gaseous 2= E Radioactivity-High (ORE-PR031B/328 and ORE-PR033B/348) S ,
R H All , EM s genus
*with new fuel or irradiated fuel in the fuel storage areas or fuel building. kM l
i
FINAL HAFT INSTRUM$NTATION AUG1g g , MOVABLE INCORE DETECTORS , LIMITING CONDITION FOR OPERATION 3.3.3.2 The Movable Incore Detection System shall be OPERABLE with:
- a. At least 75% of the detector thimbles,
- b. A minimum of two detector thimbles per core quadrant, and l
- c. Sufficient movable detectors, drive, and readout equipment to map these thimbles.
APPLICABILITY: When the Movable Incore Detection System is used for:
- a. Recalibration of the excore neutron flux detection system, or
- b. Monitoring the QUADRANT POWER TILT RATIO, or
- c. MeasurementofFh,F(Z)andF q xy*
ACTION: With the Movable Incore Detection System inoperable, do not use the system for the above applicable monitoring or calibration functions. The provisions of Spacifications 3.0.3 and 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS 4.3.3.2 The Movable Incore Detection System shall be demonstrated OPERABLE at least once per 24 hours by normalizing each detector output when required for:
- a. Recalibration of the Excore Neutron Flux Detection System, or
- b. Monitoring the QUADRANT POWER TILT RATIO, or
- c. Measurement of F g, Fq(Z), and Fxy' BRAIDWOOD - UNITS 1 & 2 3/4 3-43 l
l INSTRUMENTATION FINAL HAFT AUG 191986 SEISMIC INSTRUMENTATION LIMITING CONDITION FOR OPERATION ! 3.3.3.3 The seismic monitoring instrumentation shown in Table 3.3-7 shall be OPERABLE. APPLICABILITY: At all times. ACTION:
- a. With one or more of the above required seismic monitoring instruments inoperable for more than 30 days, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 10 days outlining the cause of the malfunction and the plans for restoring the instrument (s) to OPERABLE status.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not. applicable.
SURVEILLANCE REQUIREMENTS 4.3.3.3.1 The seismic monitoring instrumentation shall be determined OPERABLE:
- a. At least once per 31 days by verifying operable status indications of '
the seismic monitoring instrumentation.
- b. At least once per 92 days by verifying that:
- 1) The triaxial acceleration sensors and the time-history accelero-graphs properly process the equipment internal test signals.
- 2) The response spectrum analyzer properly executes its diagnostic routine.
- c. At least once per 184 days by verifying that the-triaxial acceleration sensors and the time-history accelerographs properly record the equipment internal test signals. The test may be performed in lieu of the test required by Specification 4.3.3.3.1.b.1), and
- d. At least once per 18 months, during shutdown, by:
- 1) Verifying the electronic calibration of the time-history accelerographs.
- 2) Installing fresh magnetic recording plates in the triaxial peak accelerographs.
4.3.3.3.2 Upon actuation of the seismic monitoring instruments, the equipment listed in Table 3.3-7 shall be restored to OPERABLE status within 24 hours following the seismic event. Data shall be retrieved from actuated instruments and analyzed to determine the magnitude of the vibratory ground motion. A Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 14 days describing the magnitude, frequency spectrum and resultant effect upon facility features important to safety. , BRAIDWOOD - UNITS 1 & 2 3/4 3-44
TABLE 3.3-7 SEISMIC MONITORING INSTRUMENTATION AUG 191986 MINIMUM MEASUREMENT INSTRUMENTS INSTRUMENTS AND SENSOR LOCATIONS RANGE OPERABLE
- 1. Time - History Accelerographs
~
- a. Aux. Elect. Rm, OPA02J N.A. 1
- 2. Triaxial Peak Accelerographs
- a. Cont./ Reactor Eq. Accumulators -2 g to +2 g 1
- b. Cont./ Reactor piping -2 g to +2 g 1
- c. Aux. Bldg./ Cat. I piping -2 g to +2 g 1
- 3. Response-Spectrum Analyzer Aux Elect Rm, OPA02J None 1
- 4. Triaxial Acceleration Sensors
- a. Cont./10W - 377' -2 g to +2 g 1
- b. Cont /10W - 502' -2 g to +2 g -
1
- c. Cont./10X - 426' -2 g to +2 g 1
- d. Free Field /39 + 00E, 41 + OOS -2 g to +2 g 1
- e. Aux. Bldg./18N - 426' -2 g to +2 g 1
- f. Braidwood Pond Screen House -2 g to +2 g 1 i
1 1 l THIS PAGE OPEN PENDING RECEIPT OF INFORMATION FROM THE APPLICANT BRAIDWOOD - UNITS 1 & 2 3/4 3-45
FINAL DRAFT AUG 1~91986 . 1 This Page Intentionally Left Blank C e BRAIDWOOD - UNITS 1 & 2 3/4 3-46
' ~ ~ - -- - - __._._.__ _._..__ _ _ ___
INSTRUMENTATION METEOROLOGICAL INSTRUMENTATION , AUG 19 m LIMITING CONDITION FOR OPERATION 3.3.3.4 The meteorological monitoring instrumentation channels given in Table 3.3-8 shall be OPERABLE. APPLICABILITY: At all times. ACTION:
- a. With one or more required meteorological monitoring channels inoperable for more than 7 days, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 10 days outlining the cause of the malfunction and the plans for restoring the channel (s) to OPERABLE status.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.3.3.4 Each of the above meteorological monitoring instrumentation channels shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies given in. Table 4.3-5. l BRAIDWOOD - UNITS 1 & 2 3/4 3-47 l
TABLE 3.3-8 METEOROLOGICAL MONITORING INSTRUMENTATION AUG 191986 MINIMUM INSTRUMENT LOCATION OPERABLE i
- 1. Wind Speed Nominal Elev. 34 ft 1 l
Nominal Elev. 203 ft 1
- 2. Wind Direction Nomical Elev. 34 ft 1 Nominal Elev. 203 ft 1
- 3. Air Temperature - AT Nominal Elev. 30 ft/199 ft 1 6
BRAIDWOOD - UNITS 1 & 2 3/4 3-48
TABLE 4.3-5 METEOROLOGICAL MONITORING INSTRUMENTATION IO E SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL INSTRUMENT CHECK CALIBRATION
- 1. Wind Speed -
- a. Nominal Elev. 34 ft D SA
- b. Nominal Elev. 203 ft D SA 9
- 2. Wind Direction
- a. Nominal Elev. 34 ft D SA
- b. Nominal Elev. 203 ft D SA
- 3. Air Temperature - AT Nominal Elev. 30 ft/199 ft D SA l
i l l BRAIDWOCD - UNITS 1 & 2 3/4 3-49 l
a INSTRUMENTATION a mE Ef th w REMOTE SHUT 00WN INSTRUMENTATION AUG 191986 LIMITING CONDITION FOR OPERATION 3.3.3.5 The remote shutdown monitoring instrumentation channels given in Table 3.3-9 shall be OPERABLE with readouts displayed external to the control room. APPLICABILITY: MODES 1,.2, and 3. ACTION:
- a. With the number of OPERABLE remote shutdown monitoring channels less than 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 at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
- b. The provisions of Specification 3.0.4 are not applicable.
s SERVEILLANCE REQUIREMENTS 4.3.3.5 Each remote shutdown monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies given in Table 4.3-6. l BRAIDWOOD - UNITS 1 & 2 3/4 3-50
TABLE 3.3-9 REMOTE SHGTDOWN MONITORING INSTRUMENTATION s 8 TOTAL NO. MINIMUM .
. READOUT OF CHANNELS e INSTRUMENT LOCATION CHANNELS OPERABLE 5
d 1. Intermediate Range Neutron Flux PLO6J 2 1
- e. 2. Source Range Neutron Flux PLO6J 2 1 to
- 3. Reactor Coolant Temperature -
Wide Range
- a. Hot Leg PLO5J 1/ loop 1/ loop
- b. Cold Leg PLO5J 1/ loop 1/ loop g 4. Pressurizer Pressure PLO6J 1 1 T 5. Pressurizer Level PLO6J 2 1
- 6. Steam Generator Pressure PLO4J/PLO5J 1/sta gen 1/sta gen
- 7. Steam Generator Level PLO4J 1/sta gen 1/ste gen
- 8. RHR Temperature LOCAL 2 1
- 9. Auxiliary Feedwater Flow Rate PLO4J/PLO5J 2/sta gen 1/sta gen I
M i e l3> [ F"""
=l
TABLE 4.3-6 b REMOTE SHUTDOWN MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL INSTRUMENT CHECK CALIBRATION E
- 1. Intermediate Range Neutron Flux M R
] ". 2. Source Range Neutron Flux M* R e " 3. Reactor Coolant Temperature - Wide Range M R
- 4. Pressurizer Pressure M R
- 5. Pressurizer Level M R w 6. Steam Generator Pressure M R 1
w 7. Steam Generator Level M R u,
" 8. RHR Temperature M R 3
- 9. Auxiliary Feedwater Flow Rate M R I
*When below P-6.
9 to
- $ l23 D
y t
INSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATION AUG 191986 LIMITING CONDITICN FOR OPERATION 3.3.3.6 The accident monitoring instrumentation channels shown in Table 3.3-10 shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3. ACTION:
- a. With the number of OPERABLE accident monitoring instrumentation chan-nels less than the Requirec' Number of Channels shown in Table 3.3-10, restore the inoperable channel (s) to CPERABLE status within 7 days; otherwise, be in at least HOT STf.NDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
- b. With the number of OPERABLE accident monitoring instrumentation channels, except the containment high range area radiation monitor, main steam line radiation monitor, and the auxiliary building vent stack wide range noble gas monitor, less than the Minimum Channels OPERABLE requirements of Table 3.3-10, restore the inoperable channel (s) to OPERABLE status within 48 hours; otherwise, be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
- c. With the number of OPERABLE channels for the containment high range area radiation monitor, or main steam line radiation monitor, or the auxiliary building vent stack wide range noble gas monitor less than the Minimum Channels OPERABLE requirements of Table 3.3-10, initiate an alternate method of monitoring the appropriate parameter (s) within 72 hours and either restore the inoperable channel to OPERABLE status within 7 days, or prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 14 days that provides actions taken, cause of the inoperability and plans and schedule for restoring the channels to OPERABLE status.
- d. The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.3.3.6 Each accident monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-7. - BRAIDWOOD - UNITS 1 & 2 3/4 3-53
TABLE 3.3-10 ACCIDENT MONITORING INSTRUMENTATION
=
g REQUIRED MINIMUM E; NO. OF CHANNELS g INSTRUMENT CHANNELS OPERABLE o 1. Containment Pressure 2 1
- 2. Reactor Coolant Outlet Temperature - THOT (Wide Range) 2 1
- 3. Reactor Coolant Inlet Temperature - TCOLD (Wide Range) 2 1
]
g 4. Reactor Coolant Pressure - Wide Range 2 1
** 5. Pressurizer Water Level 2 1 m
- 6. Steam Line Pressure 2/ steam generator 1/ steam generator
- 7. Steam Generator Water Level - Narrow Range 1/ steam generator 1/ steam generator
- 8. Steam Generator Water Level - Wide Range '1/ steam generator 1/ steam generator
- 9. Refueling Water Storage Tank Water Level 2 1 R 10. Auxiliary Feedater Flow Rate 2/ steam generator 1/ steam generator
$ 11. PORV Position Indicator * (0 pen / Closed) 1/ Valve 1/ Valve h 12. PORV Block Valve Position Indicator ** (0 pen / Closed) 1/ Valve 1/ Valve
, 13. Safety Valve Position Indicator (0 pen / Closed) 1/ Valve 1/ Valve
- 14. Containment Floor Drain Sump Water Level (Narrow Range) 2 1
- 15. Containment Water Level (Wide Range) 2 1
- 16. In Core Thermocouples 4/ core quadrant 2/ core quadrant
- 17. Containment High Range Area Radiation N.A. 1
- 18. Auxiliary Building Vent Stack - Wide Range Noble Gas N.A. 1/ stack
- 19. Main Steam Line Radiation N.A. 1/sts line 2
- 20. Reactor Vessel Water Level 2 1
- 21. , Reactor Coolant Subcooling Margin Monitor 2*** 1*** E P e
*Not applicable if the a'ssociated block valve is in the closed position.
**Not applicable if the block valve is verified in the closed position and power is removed. h
***Use monitoring channels (10 highest average core exit temperatures) in conjunction with RCS Pressure (item 4 above) to determine the subcooling margin.
TABLE 4.3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS h INSTRUMENT CHANNEL CHECK CHANNEL CALIBRATION f8 1. 2. Containment Pressure Reactor Coolant Outlet Temperature - TH0T (Wide Range) M M R, R
- 3. Reactor Coolant Inlet Temperature - TCOLD (Wide Range) M R .
5 -e
- 4. Reactor Coolant Pressure - Wide Range M R
- 5. Pressurizer Water Level . M R
- e. 6. Steam Line Pressure M R
" 7. Steam Generator Water Level - Narrow Range M R
- 8. Steam Generator Water Level - Wide Range M R
- 9. Refueling Water Storage Tank Water Level M R
- 10. Auxiliary Feedater Flow Rate M R w 11. PORY Position Indicator * (0 pen / Closed) M N.A.
- 12. PORV Block Valve Position Indicator ** (0 pen / Closed) M N.A.
J, 13. Safety Valve Position Indicator (0 pen / Closed) M N.A. q
- 14. Containment Floor Drain Sump Water Level (Narrow Range) M R
- 15. Containment Water Level (Wide Range) M R p
- 16. In Core Thermocouples M R P""
- 17. Containment High Range Area Radiation M R*** [
- 18. Auxiliary Building Vent Stack - Wide Range Noble Gas M R g
- 19. Main Steam Line Radiation
- M R
- 20. Reactor Vessel Water Level M R
*Not applicable if the associated block valve is in the closed position. **Not applicable if the block valve is verified in the closed position and power is removed. *** CHANNEL CALIBRATION may consist of an electronic calibration of the channel, not including the detector, for range decades above 10R/h and a one point calibration check of the detector below 10R/h with an installed or portable gamma source.
INSTRUMENTATION FIRE DETECTION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.7 As a minimum, the fire detection instrumentation for each fire detection zone shown in Table 3.3-11a for Unit 1 (Table 3.3-11b for Unit 2) sh'all be OPERABLE. APPLICABILITY: Whenever equipment protected by the fire detection instrument is required to be OPERABLE. ACTION:
- a. With any, but not more than one-half the total in any fire zone, fire detection instruments shown in Table 3.3-11 inoperable, restore the inoperable instrument (s) to OPERABLE status within 14 days or within the next 1 hour establish a fire watch patrol to inspect the zone (s) with the inoperable instrument (s) at least once per hour, unless the instrument (s) is located inside the containment, then inspect that containment zone at least once per 8 hours or
- monitor the containment air temperature at least once per hour at the locations listed in Specification 4.6.1.5.
- b. With more than one-half of the fire detection instruments in any fire zone shown in Table 3.3-11 inoperable or with any fire suppression instruments shown in Table 3.3-11, inoperable, or with any two or more adjacent fire detection instruments shown in Table 3.3-11 inoperable, within 1 hour establish a fire watch patrol to inspect the zone (s) with the inoperable instrument (s) at least once per hour, unless the instrument (s) is located inside the containment, then inspect that containment zone at least once per 8 hours or monitor the containment air temperature at least once per hour at the locations listed in Specification 4.6.1.5.
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.3.3.7.1 Each of the above required fire detection instruments which are accessible during plant operation shall be demonstrated OPERABLE at least once per 6 months by performance of a TRIP ACTUATING DEVICE OPERATIONAL TEST. Fire detectors which are not accessible during plant operation shall be demonstrated OPERABLE by the perfor:nnce of a TRIP ACTUATING DEVICE OPERATIONAL TEST during each COLD SHUTDOWN exceeding 24 hours unless performed in the previous 6 months. BRAIDWOOD - UNITS 1 & 2 3/4 3-56
INSTRUMENTATION SURVEILLANCE REQUIREMENTS (Continued) AUG 1 S IS66 4.3.3.7.2 The NFPA Standard 72D supervised circuits supervision associated with the detector alarms of each of the above required fire detection instruments shall be demonstrated OPERABLE at least once per 6 months. x BRAIDWOOD - UNITS 1 & 2 3/4 3-57
TABLE 3.3-11a FIRE DETECTION INSTRUMENTS (UNIT 1) AUG 19 M l INSTRUMENT LOCATION INSTRUMENT TYPE
- TOTAL NUMBER OF INSTRUMENTS lleat Flame Smoke
- 1. Containment ***
Zone 11 Elev 426 Suppression 1** Zone 12 Elev 426 Suppression 1 ** ~ Zone 2 Elev 401 Detection 2 Zone 3 Elev 401 Detection 2 Zone 4 Elev 401 Detection 2 Zone 5 Elev 401 Detection 2 Zone 6 Elev 426 Detection 6 Zone 76 Elev 426 Detection 13 Zone 7 Elev 414 Detection 7 Zone 24 Elev 414 Detection 16
~
- 2. Control Room Zone 29 Elev 383 Detection 4 Zone 68 Elev 451 Detection 3 Zone 69 Elev 451 Detection 12 Zone 75 Elev 451 Detection 20 Zone 68 Elev 451 (Unit 2) Detection 3
- 3. Switchgear Rooms Zone 66 Elev 451 Detection 10 Zone 77 Elev 426 Detection 21 Zone 78 Elev 426 Detection 19 -
- 4. Upper Cable Spreading Room Zone 41 Elev 463 Detection 4 Zone 42 Elev 463 Detection 4 Zone 43 Elev 463 Detection 8 Zone 44 Elev 463 Detection 8 Zone 45 Elev 463 Detection 10 l Zone 46 Elev 463 Detection 10 I
Zone 47 Elev 463 Detection 5 Zone 48 Elev 463 Detection 5 , Lower Cable Spreading Room l Zone 49 Elev 439 Detection 23 l Zone 50 Elev 439 Detection 23 l Zone 51 Elev 439 Detection 13 Zone 52 Elev 439 Detection 13 Zone 53 Elev 439 Detection 9 Zone 54 Elev 439 Detection 9 Zone 55 Elev 439 Detection 6 Zone 56 Elev 439 Detection 6
- 5. Remote Shutdown Panel
, Zone 13 Elev 383 Detection 7 1 BRAIDWOOD - UNITS 1 & 2 3/4 3-58
TABLE 3.3-11a (Continued) FIRE DETECTION INSTRUMENTS AUG 191986 (UNIT 1) INSTRUMENT LOCATION INSTRUMENT TYPE
- TOTAL NUMBER OF INSTRUMENTS Heat Flame Smcke
- 6. Station Battery Room Zone 67 Elev 451 Detection 13
- 7. Diesel Generator Room Zone 37 Elev 401 Suppression 6 Zone 38 Elev 401 Suppression 6 Zone 71 Elev 401 Detection 1 Zone 72 Elev 401 Detection 1
- 8. Diesel Fuel Storage Zone 39 Elev 401 Suppression 1 Zone 40 Elev 401 Suppression 1 Zone 27 Elev 383 Suppression 3 Zone 28 Elev 383 Suppression 3 Zone 10 Elev 383 Detection 6
- 9. Safety Related Pumps Zone 41 Elev 383 Suppression 2 Zone 42 Elev 383 Suppression 1 Zone 16 Elev 364 Detection 2 Zone 18 Elev 364 Detection 10 Zone 19 Elev 364 Detection 3 Zone 20 Elev 346 Detection 3 Zone 21 Elev 346 Detection 3 Zone 11 Elev 330 (Unit 2) Detection 23'
- 10. Fuel Storage Zone 39 Elev 401 Detection 26 Zone 38 Elev 426 Detection 3
- 11. Auxiliary Building Zone 70 Elev 451 Detection 11 Zone 17 Elev 364 Detection 37 Zone 40 Elev 364 Detection 7 Zone 11 Elev 383 Detection 37 Zone 8 Elev 401 Detection 36 Zone 23 Elev 426 Detection 6 Zone 25 Elev 467 Detection 7 Zone 64 Elev 401 Detection 5 Zone 65 Elev 401 Detection 5 Zone 14 Elev 401 Detection 4 Zone 15 Elev 401 Detection . 4 BRAIDWOOD - UNITS 1 & 2 3/4 3-59
TABLE 3.3-11a (Continued) j FIRE DETECTION INSTRUMENTS (UNIT 1) AUG 191986 INSTRUMENT LOCATION INSTRUMENT TYPE
- TOTAL NUMBER OF INSTRUMENTS Heat Flame Smoke
- 11. AuxDicry Building (continued)
Zone 17 Elev 346 (Unit 2) Detection - 32 Zone 40 Elev 346 (Unit 2) Detection 12-Zone 12 Elev 459 (Unit 2) Detection 11 Zone 22 Elev 467 (Unit 2) Detection 18 Zone 62 Elev 451 (Unit 2) Detection 11 Zone 61 Elev 439 (Unit 2) Detection 7 Zone 75 Elev 426 (Unit 2) Detection 33 TABLE NOTATIONS
*A single detector in a zone marked " Detection" will alarm in the Main Control Room. A single detector in a zone marked " Suppression" will initiate suppression and alarm in the Main Control Room. **Thrte are Containment Ventilation temperature switches. Upon receipt of a Hi-Hi tes3erature, suppression must be manually initiated. These switches are not 720 supervised. ***The fire detection instruments located within the containment are not required to be D ERABLE during the performance of Type A containment leakage rate tests.
I BRAIDWOOD - UNITS 1 & 2 3/4 3-60
TABLE 3.3-11b FIRE DETECTION INSTRUMENTS AUG I g 1986
- (UNIT 2)
INSTRUMENT LOCATION INSTRUMENT TYPE
- TOTAL NUMBER OF INSTRUMENTS Heat Flame Smoke
- 1. Containment ***
Zone 11 Elev 426 Suppression 1 ** Zone 12 Elev 426 Suppression 1 ** Zone 2 Elev 401 Detection 2 Zone 3 Elev 401 Detection 2 Zone 4 Elev 401 Detection 2 Zone 5 Elev 401 Detection 2 Zone 6 Elev 426 Detection 6 Zone 76 Elev 426 Detection 9 Zone 7 Elev 414 Detection 6 Zone 24 Elev 414 Detection 10
- 2. Control Room Zone 29 Elev 383 (Unit 1) Detection 4 Zone 68 Elev 451 Detection 3 Zone 69 Elev 451 Detection 14 Zone 75 Elev 451 (Unit 1) Detection 20 Zone 68 Elev 451 (Unit 1) Detection 3
- 3. Switchgear Rooms Zone 66 Elev 451 Detection 6 Zone 77 Elev 426 Detection 9 Zone 78 Elev 426 Detection 9
- 4. Upper Cable Spreading Room Zone 41 Elev 463 Detection 4 Zone 42 Elev 463 Detection 4 Zone 43 Elev 463 Detection 8 Zone 44 Elev 463 Detection 8 Zone 45 Elev 463 Detection 9 Zone 46 Elev 463 Detection 9 Zone 47 Elev 463 Detection 5 Zone 48 Elev 463 Detection 5 Lower Cable Spreading Room Zone 49 Elev 439 Detection 8 Zone 50 Elev 439 Detection 8 Zone 51 Elev 439 Detection 19 Zone 52 Elev 439 Detection 19 Zone 53 Elev 439 Detection 11 Zone 54 Elev 439 Detection 11 Zone 55 Elev 439 Detection '
6 Zone 56 Elev 439 Detection 6 l l BRAIDWOOD - UNITS 1 & 2 3/4 3-61
TABLE 3.3-11b (Continued) FIRE DETECTION INSTRUMENTS
)
(UNIT 2) AUG 191986 INSTRUMENT LOCATION INSTRUMENT TYPE
- TOTAL NUMBER OF INSTRUMENTS j Heat Flame Smoke
- 5. Remote Shutdown Panel Detection Zone 13 Elev 383 (Unit 1) 7
- 6. Station Battery Room Zone 67 Elev 451 Detection 6
- 7. Diesel Generator Room Zone 37 Elev 401 Suppression 4 Zone 38 Elev 401 Suppression 4 Zone 71 Elev 401 Detection 1 Zone 72 Elev 401 Detection 1
- 8. Diesel Fuel Storage Zone 39 Elev 401 Suppression 1 Zone 40 Elev 401 Suppression 1 Zone 27 Elev 383 Suppression 3 Zone 28 Elev 383 Suppression 3 Zone 10 Elev 383 Detection 6
- 9. Safety Related Pumps Zone 41 Elev 383 Suppression 2 Zone 42 Elev 383 Suppression 1 Zone 16 Elev 364 Detection 2 Zone 18 Elev 364 Detection 3 Zone 19 Elev 364 Detection 2 Zone 20 Elev 346 Detection 3 Zone 21 Elev 346 Detection 3 Zone 11 Elev 330 Detection 23
- 10. Fuel Storage Zone 39 Elev 401 (Unit 1) Detection 29 Zone 38 Elev 426 (Unit 1) Detection 3 l 11. Auxiliary Building Zone 17 Elev 346 Detection 32 Zone 40 Elev 346 Detection 12 Zone 22 Elev 467 Detection 18 Zone 62 Elev 451 Detection 11 Zone 14 Elev 401 Detection 4 Zone 15 Elev 401 Detection 4 Zone 64 Elev 401 Detection 3
Zone 65 Elev 401 Detection 3 BRAIDWOOD - UNITS 1 & 2 3/4 3-62
TABLE 3.3-11b (Continued) FIRE DETECTION INSTRUMENTS AUG 191986 (UNIT 2) , INSTRUMENT LOCATION INSTRUMENT TYPE
- TOTAL NUMBER OF INSTRUMENTS Heat Flane Smoke
- 11. Auxiliary Building (continued)
Zone 61 Elev 439 Detection . 7 Zone 70 Elev 451 Detection 11 Zone 75 Elev 426 Detection ' 33 Zone 25 Elev 467 Detection 7 Zone 17 Elev 364 (Unit 1) Detection 37 Zone 40 Elev 364 (Unit 1) Detection 7 Zone 11 Elev 383 (Unit 1) Detection 37 Zone 8 Elev 401 (Unit 1) Detection 36 Zone 23 Elev 426 (Unit 1) Detection 6 Zcne 12 Elev 459 Detection 11 TABLE NOTATIONS .
*A single detector in a zone marked " Detection" will alarm in the Main Control Room. A single detector in a zone marked " Suppression" will initiate suppression and alarm in the Main Control Room. **These are Containment Ventilation temperature switches. Upon receipt of a Hi-Hi temperature, suppression must be manually initiated. These switches are not 72D supervised. ***The fire detection instruments located within the containment are not required to be OPERABLE during the performance of Type A containment leakage rate tests.
l BRAIDWOOD - UNITS 1 & 2 3/4 3-63
INSTRUMENTATION LOOSE-PART DETECTION SYSTEM AUG 191986 LIMITING CONDITION FOR OPERATION 3.3.3.8 The Loose-Part Detection System shall be OPERABLE. APPLICABILITY: MODES 1 and 2. ACTION:
- a. With one or more Loose-Part Detection System channels inoperable for more than 30 days, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 10 days outlining the cause of the malfunction and the plans for restoring the channel (s) to OPERABLE status,
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.3.3.8 Each channel of the Loose-Part Detection Systems shall be demonstrated OPERABLE by performance of:
- a. A CHANNEL CHECK at least once per 24 hours,
- b. An ANALOG CHANNEL OPERATIONAL TEST except for verification of setpoint at least once per 31 days, and
- c. A CHANNEL CALIBRATION at least once per 18 months.
l BRAIDWOOD - UNITS 1 & 2 3/4 3-64
INSTRUMENTATION f RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 19 000 3.3.3.9 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 (00CM). 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 l in Table 3.3-12. Restore the inoperable instrumentation to OPERABLE status within the time specified in the ACTION, or explain in the next Semiannual Radioactive Effluent Release Report pursuant to Specifica-tion 6.9.1.7 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.
SURVEILLANCE REQUIREMENTS 4.3.3.9 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and DIGITAL and ANALOG CHANNEL OPERATIONAL TEST at the frequencies shown in Table 4.3-8. BRAIDWOOD - UNITS 1 & 2 3/4 3-65
TABLE 3.3-12 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION o k o MINIMUM CHANNELS e INSTRUMENT OPERABLE ACTION C 3 1. Radioactivity Monitors Providing Alarm and d Automatic Termination of Release w e- Liquid Radwaste Effluent Line (ORE-PR001) 1 31
~
- 2. Radioactivity Monitors Providing Alarm But Not Providing Automatic Termination of Release w
1 a. Essential Service Water w - h 1) Unit 1 a) RCFC 1A and 1C Outlet (1RE-PR002) 1 32 b) RCFC IB and ID Outlet (1RE-PR003) 1 32
- 2) Unit 2 a) RCFC 2A and 2C Outlet (2RE-PR002) 1 32 b) RCFC 28 and 2D Outlet (2RE-PR003) 1 32
- b. Station Blowdown Line (0RE-PR010) 1 32 ==yng
- 3. Flow Rate Measurement Devices E c) Z
- a. Liquid Radwaste Effluent Line (Loop-WX001) 1 -
33 ,7
- b. Station Blowdown Line (Loop-CWO32) 1 33 h
TABLE 3.3-12 (Continued) ACTION STATEMENTS ACTION 31 - 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 14 days provided that prior to initiating a release: s
- a. At least two independent samples are analyzed in accordance with Specification 4.11.1.1.1, and
- b. At least two technically qualified members of the facility staff independently verify the release rate calculations and discharge line valving.
Otherwise, suspend release of radioactive effluents via this pathway. ACTION 32 - With the number of channels OPERABLE less than re" quired by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided that, at least once per 12 hours, grab samples are collected and analyzed for ' radioactivity at a lower limit of detection of no more than 10 7 microcurie /ml. ACTION 33 - With the number of channels OPERABLE less than required'by the Minimum Channels OPERABLE requirement, efflucnt releases via this pathway may continue for up to 30 days provided the flow rate is estimated at least once per 4 hours during actual aleases. Pump performance curves generated in place may be used u estimate - flow. a BRAIDWOOD - UNITS 1 & 2 3/4 3-67
T_ABLE 4.3-8 . O RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS h o DIGITAL ANALOG - CHANNEL CHANNEL CHANNEL SOURCE CHANNEL OPERATIONAL OPERATIONAL g INSTRUMENT CHECK CHECK CALIBRATION TEST TEST d 1. Radioactivity Monitors Providing w Alarm and Autc;ratic Termination 2 of Release ro Liquid Radwaste Effluent Line (0RE-PR001) D P R(3) Q(1) N.A.
- 2. Radioactivity Monitors Providing Alarm But y Not Providing Automatic Termination
- of Release .
w - h a. Essential Service Water
- 1) Unit 1 -
a) RCFC 1A and 1C Outlet (IRE-PR002) D M R(3) Q(2) N.A. b) RCFC IB and 10 Outlet (1RE-PR003) D M R(3) Q(2) N.A.
- 2) Unit 2 a) RCFC 2A and 2C Outlet (2RE-PR002) D M R(3) Q(2) N.A.
b) RCFC 28 and 20 Outlet (2RE-PR003) D M R(3) Q(2) N.A. , b. Station Blowdown Line (0RE-PR010) D M R(3) Q(2) N.A. 9 nummme E
- 3. Flow Rate Measurement Devices , y
- a. Liquid Radwaste Effluent Line .
(Loop-WX001) D(4) N.A. R N.A. QU
- b. Station Blowdown Line (Loop-CWO32) D(4) N.A. R N.A. Q l
TABLE 4.3-8 (Continued) TABLE NOTATIONS AUG 191900 (1) The DIGITAL CHANNEL OPERATIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occur if any of the following conditions exists:
- a. Instrument indicates measured levels above the Alarm / Trip Setpoint, or
- b. Circuit failure (monitor loss of communications - alarm only, detector loss of counts, or monitor loss of power), or
- c. Detector check source test failure, or
- d. Detector channel out-of-service, or
- e. Monitor loss of sample flow.
(2) The DIGITAL CHANNEL OPERATIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exists:
- a. Instrument indicates measured levels above the Alarm Setpoint, or
- b. Circuit failure (monitor loss of communications - alarm only, detector loss of counts, or monitor loss of power), or
- c. Detector check source test failure, or
- d. Detector channel out-of-service, or
- e. Monitor loss of sample flow.
(3) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards (NBS) or using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS. These standards shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used. (4) CHANNEL CHECK shall consist of verifying indication of flow during periods of release. CHANNEL CHECK shall be made at least once per 24 hours on days on which continuous, periodic, or batch releases are made. BRAIDWOOD - UNITS 1 & 2 3/4 3-69
INSTRUMENTATION RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION M IO E LIMITING CONDITION FOR OPERATION 3.3.3.10 The radioactive gaseous effluent monitoring instrumentation channels shown in Table 3.3-13 shall be OPERABLE with their Alarm / Trip Setpoints set to ensure that the limits of Specifications 3.11.2.1 and 3.11.2.5 are not exceeded. The Alarm / Trip Setpoints of these channels meeting Specification 3.11.2.1 shall be determined and adjusted in accordance with the methodology and parameters in the ODCM. APPLICABILITY: As shown in Table 3.3-13 ACTION:
- a. With a radioactive gaseous effluent monitoring instrumentation channel Alarm / Trip Setpoint less conservative than required by the above specification, immediately suspend the release of radioactive gaseous effluents monitored by the affected channel, or declare the channel inoperable.
- b. With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels CPERABLE take the ACTION shown in Table 3.3-13. Restore the inoperable instrumentation to OPERABLE status within the time specified in the ACTION, or explain in the next Semiannual Radioactive Effluent Release Report pursuant to Specification 6.9.1.7 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.
SURVEILLANCE REQUIREMENTS 4.3.3.10 Each radioactive gaseous effluent monitoring instrumentation channel l shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and DIGITAL CHANNEL OPERATIONAL TEST at the fre-quencies shown in Table 4.3-9. BRAIDWOOD - UNITS 1 & 2 3/4 3-70
- 1
TABLE 3.3-13 cn 5 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION o k o INSTRUMENT MINIMUM CHANNELS OPERABLE APPLICABILITY ACTION i g 1. Plant Vent Monitoring System - Unit 1
- a. Noble Gas Activity Monitor-H Providing Alarm e.
m 1) High Range (1RE-PR0280) 1
- 39
- 2) Low Range (1RE-PR0288) 1
- 39
- b. Iodine Sampler (IRE-PR028C) 1
- 40
- c. Particulate Sampler (1RE-PR028A) 1
- 40 sa
- O d. Effluent System Flow Rate Measuring Device (LOOP-VA019) 1
- 36
- e. Sampler Flow Rate Measuring Device 1
- 36 (IFT-PR165)
- 2. Plant Vent Monitoring System - Unit 2
- a. Noble Gas Activity Monitor-Providing Alarm
- 1) High Range (2RE-PR0280) 1
- 39
- 2) Low Range (2RE-PR0288) 1
- 39 q mm-me
- b. Icdine Sampler (2RE-PR028C)
- 1 40 , Z
- c. Particulate Sampler (2RE-PR028A) 1
- 40 E H
P
- d. Effluent System Flow Rate Measuring Device (LOOP-VA020) hO 1
- 36
- e. Sampler Flow Rate Measuring Device 1
- 36 (2FT-PR165)
TABLE 3.3-13 (Continued) cn 5 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS O INSTRUMENT OPERABLE APPLICABILITY ACTION g 3. Gaseous Waste Management System U a. Hydrogen Analyzer (OAT-GW8000) 1 ** 38 w e- b. Oxygen Analyzer (OAIT-GW004 and m 0AT-GW8003) 2 ** 38
- 4. Gas Decay Tank System ca a. Noble Gas Activity Monitor - Providing A Alarm and Automatic Termination of y Release (0RE-PR002A and 28) 2 35 -
- 5. Containment Purge System
- a. Noble Gas Activity Monitor - Providing Alarm,(RE-PR001B)
- 1 37
- b. Iodine Sampler (RE-PRC01C) 1 40
- c. Particulate Sampler (RE-PR001A) 1 40
- 6. Radioactivity Monitors Providing Alarm and -
Automatic Closure of Surge Tank Vent-Component 3 g Cooling Water Line (ORE-PR009 and RE-PR009) 2
- 41 g
$P
$ c:3 E
21
TABLE 3.3-13 (Continued) TABLE NOTATIONS
*At all times.
- 0uring WASTE GAS HOLDUP SYSTEM operation.
All instruments required for Unit 1 or Unit 2 operation. ACTION STATEMENTS ACTION 35 - With the number of channels OPERABLE less than required by the Hinimum Channels OPERABLE requirement, the contents of the tank (s) may be released to the environment for up to 14 days provided that prior to initiating the release:
- a. At least two independent samples of the tank's contents are analyzed, and
- b. At least two technically qualified members of the facility staff independently verify the release rate calculations and discharge valve lineup.
Otherwise, suspend release of radioactive effluents via this pathway. ACTION 36 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided the flow rate is estimated at least once per 4 hours. ACTION 37 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, immediately suspend PURGING of radioactive effluents via this pathway. ACTION 38 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, operation of this system may continue provided grab samples are taken and analyzed at least once per 24 hours. With both channels inoperable, operation may continue provided grab samples are taken and analyzed at least once per 4 hours during degassing operations and at least once per 24 hours during other operations. ACTION 39 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided grab samples are taken at least once per 12 hours and these samples are analyzed for radioactivity within 24 hours. BRAIDWOOD - UNITS 1 & 2 3/4 3-73
TABLE 3.3-13 (Continued) g2gg ACTION STATEMENTS (Continued) ACTION 40 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via the affected pathway may continue for up to 30 days provided samples are continuously collected with auxiliary sampling equipment as required in Table 4.11-2. . ACTION 41 - 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 that, at least once per 12 hours, grab samples are collected and analyzed for radioactivity at a lower limit of detection of no more than 10 7 microcurie /ml. BRAIDWOOD - UNITS 1 & 2 3/4 3-74
TABLE 4.3-9 cn o RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS E DIGITAL 8 CHANNEL MODES FOR WHICH l e CHANNEL SOURCE CHANNEL OPERATIONAL SURVEILLANCE FUNCTIONAL UNIT CALIBRATION g CHECK CHECK TEST IS REQUIRED
- 1. Plant Vent Monitoring System - Unit 1 e- a. Noble Gas Activity Monitor -
ro Providing Alarm
- 1) High Range (1RE-PRJ28D) D M R(3) Q(2)
- 2) Low Range (1RE-PR0288) D M R(3) Q(2)
- b. Iodine Sampler (IRE-PR028C) D M R(3) Q(2)
- c. Particulate Sampler (IRE-PR028A) D M R(3) Q(2)
- d. Effluent System Flow Rate D N.A. R Q Measuring Device (LOOP-VA019)
- e. Sampler Flow Rate Measuring D N.A. R Q i
Device (IFT-PR165)
- 2. Plant Vent Monitoring System - Unit Two
- a. Noble Gas Activity Monitor-Providing Alarm
- 1) High Range'(2RE-PR0280)
- 2) Low Range (2RE-PR028B)
D D M M R(3) R(3) Q(2) Q(2) hE> w
- b. Iodine Sampler (2RE-PR028C) D M R(3) Q(2) *
- c. Particulate Sampler (2RE-PR028A) D M R(3) Q(2)
- O 21
TABLE 4.3-9 (Continued) RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS o 6 DIGITAL 8 CHANNEL MODES FOR WHICH i CHANNEL SOURCE CHANNEL OPERATIONAL SURVEILLANCE g FUNCTIONAL UNIT CHECK CHECK CALIBRATION TEST IS REQUIRED
- 2. Plant Vent Monitoring System - Unit Two e* (continued) e m d. Effluent System Flow Rate D N.A. R Q Measuring Device (LOOP-VA020)
- e. Sampler Flow Rate Measuring D N.A. R Q y
a Device (2FT-PR165) y 3. Gaseous Waste Management System -
- a. Hydrogen Analyzer (0AT-GW8000) D N.A. Q(4) M **
- b. Oxygen Analyze.r (OAIT-GW004 and 0AT-GW8003) D N.A. **
Q(5) M
- 4. Gas Decay Tank System
- a. Noble Gas Activity Monitor - P P R(3) Q(1)
Providing Alarm and Automatic Termination of Release (ORE-PR002A and 28)
- 5. Containment Purge System M
m
- a. Noble Gas Activity Monitor - >
Providing Alarm (RE-PR0018)
- g p' D P R(3) Q(2) -
g y co
- b. Iodine Sampler (RE-PR001C)
P P R(3) N.A. *m g g
- c. Particulate Sampler P P R(3) N.A. *
(RE-PR001A)
TABLE 4.3-9 (Continued) RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS o 5 DIGITAL 8 CHANNEL MODES FOR WHICH i CHANNEL SOURCE CHANNEL OPERATIONAL SURVEILLANCE g FUNCTIONAL UNIT CHECK CHECK CALIBRATION TEST IS REQUIRED
- 6. Radioactivity Monitors Providing
- Alarm and Automatic Closure of
- e. Surge Tank Vent Component Cooling m Water Line (ORE-PR009 and RE-PR009) D M R(3)
- Q(1)
R. O C 0
>r -
4
' E
- I
TABLE 4.3-9 (Continued) TABLE NOTATIONS
*At all times.
- During WASTE GAS HOLDUP SYSTEM operation.
All instruments required for Unit 1 or Unit 2 operation. (1) The DIGITAL CHANNEL OPERATIONAL TEST shall also demonstrate tha't automatic isolation of this pathway and control room alarm annunciation occur if any of the following conditions exists:
- a. Instrument indicates measure.1 levels above the Alarm / Trip Setpoint, '
or
- b. Circuit failure (monitor loss of communications - alarm only, detector loss of counts, or monitor . loss of power), or
- c. Detector check source test failure, or
- d. Detector channel out-of-service, or
- e. Monitor loss of sample flow.
(2) The DIGITAL CHANNEL OPERATIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exists:
- a. Instrument indicates measured levels above the Alarm Setpoint, or
- b. Circuit failure (monitor loss of communications - alarm only, detector loss of counts, or monitor loss of power), or
- c. Detector check source test failure, or
- d. Detector channel out-of-service, or
- e. Monitor loss of sample flow.
(3) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards (NBS) or using standards that hwe been obtained from suppliers that participate in measurement assurance activities with NBS. These standards shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used. (4) The CHANNEL CALIBRATION shall include the use of standard gas samples containing hydrogen and nitrogen. (5) The CHANNEL CALIBRATION shall include the use of standard gas samples containing oxygen and nitrogen. , BRAIDWOOD - UNITS 1 & 2 3/4 3-78
INSTRUMENTATION CHLORINE DETECTION INSTRUMENTATION AUG 19193$ LIMITING CONDITION FOR OPERATION 3.3.3.11 Two independent chlorine detoction instrumentation channel.s, with their Alarm / Trip setpoints adjusted to actuate at a chlorine concentration of less than or equal to 5 ppm, shall be OPERABLE. APPLICABILITY: All MODES. ACTION:
- a. With one chlorine detection instrumentation channel inoperable, re-store the inoperable instrumentation channel to OPERABLE status within 7 days or within the next 6 hours initiate and maintain operation of the Control Room Emergency Ventilation System in the recirculation mode of operation.
- b. With both chlorine detection instrumentation channels inoperable, within 1 hour initiate and maintain operation of the Control Room Emergency Ventilation System in the recirculation mode of operation.
- c. The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.3.3.11 Each chlorine detection instrumentation channel shall be demonstrated OPERABLE by performance of a functional test in accordance with the manufac-turer's recommendations at least once per 31 days and a CHANNEL CALIBRATION at least once per 12 months. l l l l BRAIDWOOD - UNITS 1 & 2 3/4 3-79
INSTRUMENTATION FINALDRAFT 3/4.3.4 TURBINE OVERSPEED PROTECTION AUG 191986 LIMITING CONDITION FOR OPERATION 3.3.4 At least one Turbine Overspeed Protection System shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3. , ACTION:
- a. With one throttle valve or one governor valve per high pressure turbine steam line inoperable and/or with one reheat stop valve or one reheat intercept valve per low pressure turbine steam line inoperable, restore the inoperable valve (s) to OPERABLE status within 72 hours, or close at least one valve in the affected steam line(s) or isolate the turbine from the steam supply within the next 6 hours.
- b. With the above required Turbine Overspeed Protection System otherwise inoperable, within 6 hours isolate the turbine from the steam supply.
SURVEILLANCE REQUIREMENTS 4.3.4.1 The provisions of Specification 4.0.4 are not applicable. 4.3.4.2 The above required Turbine Overspeed Protection System shall be demonstrated OPERABLE:
- a. During turbine operation at least once per 31 days by direct obser-vation of the movement of the valves below through one complete cycle from the running position:
- 1) Four high pressure turbine throttle valves,
- 2) Four high pressure turbine governor valves,
- 3) Six turbine reheat stop valves, and
- 4) Six turbine reheat intercept valves,
- b. Within 7 days prior to entering MODE 3 from MODE 4, by cycling each of the 12 extraction steam nonreturn check valves from the closed position,
- c. During turbine operation at least once per 31 days by direct obser'va-tion, of freedom of movement of each of the 12 extraction steam non-return check valve weight arms,
- d. At least once per 18 months by performance of CHANNEL CALIBRATION on the Turbine Overspeed Protection Systems, and
- e. At least once per 40 months by disassembling at least one of each of the valves given in Specifications 4.3.4.2a. and b. above, and per-forming a visual and surface inspection of valve seats, disks and ster:.s and verifying no unacceptable flaws or corrosion. .
BRAIDWOOD - UNITS 1 & 2 3/4 3-80
FINAL DRAFT 3/4.4 REACTOR COOLANT SYSTEM AUG 19 506 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION STARTUP AND POWER OPERATION l LIMITING CONDITION FOR OPERATIONS 3.4.1.1 All reactor coolant loops shall be in operation. APPLICABILITY: MODES I and 2.* ACTION: With less than the above required reactor coolant loops in operation, be in at least HOT STANDBY within 6 hours. SURVEILLANCE REQUIREMENTS 4.4.1.1 The above required reactor coolant loops shall be verified in operation and circulating reactor coolant at least once per 12 hours. l "See Special Test Exceptions Specification 3.10.4. BRAIDWOOD - UNITS 1 & 2 3/4 4-1
REACTOR COOLANT SYSTEM FINAL MAR g yg g HOT STAN0BY LIMITING CONDITION FOR OPERATION 3.4.1.2 At least two of the reactor coolant loops listed below shall be OPERABLE 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
- d. Reactor Coolant Loop D and its associated steam generator and reactor coolant pump.
APPLICABILITY: MODE 3.** ACTION:
- a. With less than the above required reactor coolant loops OPERABLE, restore the required loops to OPERABLE status within 72 hours or be in HOT SHUTDOWN within the next 12 hours.
- b. With only one reactor coolant loop in operation and the Reactor Trip System breakers in the closed position, within 1 hour open the Reactor 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 return the required reactor coolant loop to operation.
SURVEILLANCE REQUIREMENTS 4.4.1.2.1 At least the above required reactor coolant pumps, if not in operation, shall be determined OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability. 4.4.1.2.2 The required steam generators shall be determined OPERABLE by verifying secondary side narrow range water level to be greater than or equal to 41% for Unit 1 (18% for Unit 2) at least once per 12 hours. 4.4.1.2.3 The required coolant loops shall be verified in operation and circu-lating reactor coolant at least once per 12 hours.
*All Reactor Coolant pumps may be deenergized for up to 1 hour provided:
(1) no. operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10*F below saturation temperature.
- See Special Test Exceptions Specification 3.10.4.
BRAIDWOOD - UNITS 1 & 2 3/4 4-2
REACTOR COOLANT SYSTEM FINAL ggy9 DRAFT HOT SHUTOOWN ' LIMITING CONDITION FOR OPERATION 3.4.1.3 At least two of the loops listed below shall be OPERABLE and at least one of these 1Jops shall be in operation:* l
- a. Reactor Coolant Loop A and its associated steam generator and reactor coolant pump,**
- b. Reactor Coolant Loop 8 and its associated steam generator and reactor coolant pump,**
- c. Reactor Coolant Loop C and its associated steam generator and reactor coolant pump,**
- d. Reactor Coolant Loop D and its associated steam generator and reactor coolant purp,**
- e. RHR_ Loop A, and
- f. RHR Loop B.
APPLICABILITY: MODE 4 ACTION:
- a. With less than the above required reactor coolant and/or RHR loops OPERABLE, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible; if the remaining OPERABLE loop is an RHR loop, be in COLD SHUTOOWN within 24 hours,
- b. With no reactor coolant or 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 coolant loop to operation.
*All reactor coolant pumps and RHR pumps may be deenergized for up to 1 hour provided: (1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10*F below saturation temperature. **A reactor coolant pump shall not be started with one or more of the Reactor Coolant System cold leg temperatures less than or equal to 350*F unless the ;
secondary water temperature of each steam generator is less than 50*F above ' each of the Reactor Coolant System cold leg temperatures. BRAIDWOOD - UNITS 1 & 2 3/4 4-3
REACTOR COOLANT SYSTEM AUG 19 ggg SURVEILLANCE REQUIREMENTS 4.4.1.3.1 The required reactor coolant pump (s) and/or RHR pumps, if not in operation, shall be determined OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability. - 4.4.1.3.2 The required steam generator (s) shall be determined OPERABLE by verifying secondary side narrow range water level to be greater than or equal to 41% for Unit 1 (18% for Unit 2) at least once per 12 hours. 4.4.1.3.3 At least one reactor coolant or RHR loop shall be verified in operation and circulating reactor coolant at least once per 12 hours. BRAIDWOOD - UNITS 1 & 2 3/4 4-4
\
j REACTOR COOLANT SYSTEM FINAL DRAFT i COLD SHUTDOWN - LOOPS FILLED , LIMITING CONDITION FOR OPERATION 3.4.1.4.1 At least one residual heat removal (RHR) loop shall be OPERABLE and in operation *, and either:
- a. One additional RHR loop shall be OPERABLE #, or
- b. The secondary side narrow range water level of at least two steam generators shall be greater than 41% for Unit 1 (18% for Unit 2).
APPLICABILITY: MODE 5 with reactor coolant loops filled ##. ACTION:
- a. With one of the RHR loops inoperable and with less than the required steam generator level, immediately initiate corrective action to return the inoperable RHR loop to OPERABLE status or restore the required steam generator level as soon as possible.
- b. With no RHR loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to operation.
SURVEILLANCE REQUIREMENTS 4.4.1.4.1.1 The secondary side water level of at least two steam generators when required shall be determined to be within limits at least once per 12 hours. 4.4.1.4.1.2 At least one RHR loop shall be determined to be in operation and circulating reactor coolant at least once per 12 hours.
*The RHR pu?p may be deenergized for up to 1 hour provided: (1) no operations are permitted that would cause dilution of the Reactor Coolant System boron
- concentration, and (2) core outlet temperature is maintained at least 10*F below saturation temperature.
#0ne RHR loop may be inoperable for up to 2 hours for surveillance testing provided the other RHR loop is OPERABLE and in operation. . ##A reactor coclant pump shall not be started with one or more of the Reactor Coolant System cold leg temperatures less than or equal to 350 F unless the secondary water temperature of each steam generator is less than 50 F above each of the Reactor Coolant System cold leg temperatures.
BRAIDWOOD - UNITS 1 & 2 3/4 4-5
l REACTOR COOLANT SYSTEM COLD SHUT 00WN - LOOPS NOT FILLED IO M LIMITING CONDITION FOR OPERATION 3.4.1.4.2 Two residual heat removal (RHR) loops shall be OPERABLE
- and at least one RHR loop shall be in operation.**
APPLICABILITY: MODE 5 with reactor coolant loops not filled. ACTION:
- a. With less than the above required RHR loops OPERABLE, immediately initiate corrective action to return the required RHR loops to OPERABLE status as soon as possible.
- b. With no RHR loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and
- immediately initiate corrective action to return the required RHR loop to operation.
SURVEILLANCE REQUIREMENTS 4.4.1.4.2 At least one RHR loop shall be determined to be in operation and circulating reactor coolant at least once per 12 hours.
*0ne RHR loop may be inoperable for up to 2 hours for surveillance testing provided the other RHR loop is OPERABLE and in operation.
- The RHR pump may be deenergized for up to 1 hour provided: (1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10*F below saturation temperature.
BRAIDWOOD - UNITS 1 & 2 3/4 4-6
I l REACTOR COOLANT SYSTEM LOOP ISOLATION VALVES MG 191986 OPERATION LIMITING CONDITION FOR OPERATION 3.4.1.5.1 All RCS loop isolation valves (hot leg and cold leg stop valves) shall be open and power removed from the isolation valve operators. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With any RCS loop isolation valve closed, suspend startup of the iso-lated loop and be in at least HOT SHUTDOWN within 6 hours and in at least COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.4.1.5.1 All RCS loop isolation valves shall be verified open and power removed from the isolation valve operators at least once per 31 days. BRAIDWOOD - UNITS 1 & 2 3/4 4-7' ,
REACTOR COOLANT SYSTEM LOOP ISOLATION VALVES AUG 191986 SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.1.5.2 If an RCS loop is isolated, maintain the hot leg and cold leg stop valves closed until:
- a. The boron concentration of the isolated loop is greater than or equal to the boron concentration of the operating loops, and
- b. The temperature of the cold leg of the isolated loop is within 20*F of the highest cold leg temperature of the operating loops.
APPLICABILITY: MODES 5 and 6. ACTION: With the requirements of tre above specification not satisfied, do not open either the hot leg or cold leg stop valves. SURVEILLANCE REQUIREMENTS 4.4.1.5.2.1 The isolated loop cold leg temperature shall be determined to be within 20 F of the highest cold leg temperature of the operating loops within 30 minutes prior to opening the cold leg stop valve. 4.4.1.5.2.2 The boron concentration of an isolated loop shall be determined to be greater than or equal to the boron concentration of the operating loops within 2 hours prior to opening either the hot leg or cold leg stop valves of an isolated loop. BRAIDWOOD - UNITS 1 & 2 3/4 4-8 f
REACTOR COOLANT SYSTEM 3/4.4.2 SAFETY VALVES == M AUG 19 gygg SHUT 00WN LIMITING CONDITION FOR OPERATION 3.4.2.1 A minimum of one pressurizer Code safety valve shall be OPERABLE with a lift setting of 2485 psig i 1%.* APPLICABILITY: MODES 4 and 5. ACTION: With no pressurizer Code safety valve OPERABLE, immediately suspend all opera-tions involving positive reactivity changes and place an OPERABLE RHR loop into operation in the shutdown cooling mode. SURVEILLANCE REQUIREMENTS 4.4.2.1 No additional requirements other than those required by Specification 4.0.5. "The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure. BRAIDWOOD - UNITS 1 & 2 3/4 4-9
REACTOR COOLANT SYSTEM
, AUG 191986 OPERATING LIMITING CONDITION FOR OPEdATION 3.4.2.2 All pressurizer Code safety valves shall be OPERABLE with a lift setting of 2485 psig i 1%.*
APPLICABILITY: MODES 1, 2, and 3. ACTION: With one pressurizer Code safety valve inoperable, either restore the inoperable valve to OPERABLE status within 15 minutes or be in at least HOT STANDBY within 6 hours and in at least HOT SHUTOOWN within the following 6 hours. SURVEILLANCE REQUIREMENTS 4.4.2.2 No additional requirements other than those required by Specification 4.0.5. "The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure. BRAIDWOOD - UNITS 1 & 2 3/4 4-10
REACTOR COOLANT SYSTEM 3/4.4.3 PRESSURIZER , AUG 191986 LIMITING CONDITION FOR OPERATION 3.4.3 The pressurizer shall be OPERABLE with at least two groups of pressurizer heaters each having a capacity of at least 150 kW and a water level of less than or equal to 92%. APPLICABILITY: MODES 1, 2, and 3. ACTION:
- a. With less than two groups of pressurizer heaters OPERABLE, restore at least two groups of pressurizer heaters to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours,
- b. With the pressurizer otherwise inoperable, be in at least HOT STANDBY with the Reactor trip breakers open within 6 hours and in HOT SHUTDOWN within the following 6 hours. ,
SURVEILLANCE REQUIREMENTS 4.4.3.1 The pressurizer water level shall be determined to be within its limit at least once per 12 hours. 4.4.3.2 The capacity of each of the above required groups of pressurizer heaters shall be verified by energizing the heaters and measuring circuit current at least once per 92 days. 4.4.3.3 The cross-tie for the pressurizer heaters to tie ESF power supply shall be demonstrated OPERABLE at least once per 18 months by energizing the heaters. BRAIDWOOD - UNITS 1 & 2 3/4 4-11 l
REACTOR C00LANT' SYSTEM 3/S.4.4 RELIEF VALVES , ) LIMITING CONDITION FOR OPERATION l 3.4.4 All power-operated relief valves (PORVs) and their associated. block valves shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3. ACTION:
- a. With one or more PORV(s) inoperable because of excessive seat leakage, within 1 hour either restore the PORV(s) to OPERABLE status or close the associated block valve (s); otherwise be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours.
- b. With one PORV inoperable due to causes other than excessive seat leakage, within 1 hour either restore the PORV to OPERABLE status or close the associated block valve and remove power from the block valve; restore the PORV to OPERABLE status within the following 72 hours or be in HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- c. With both PORV(s) inoperable due to causes other than excessive seat leakage, within 1 hour either restore each of the PORV(s) to OPERABLE status or close their associated block valve (s) and remove power from the block valve (s) and be in HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- d. With one or more block valve (s) inoperable, within I 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 the PORV and remove power from its associated solenoid valve; and 2) apply the ACTION of b. or c. above, as appropriate for the isolated PORV(s).
- e. The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.4.4.1 In addition to the requirements of Specification 4.0.5, each PORV shall be demonstrated OPERABLE at least once per 18 months by:
- a. Performance of a CHANNEL CALIBRATION, and
- b. Operating the valve through one complete cycle of full travel.
4.4.4.2 Each block valve shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel unless the block valve is closed with power removed in order to meet the requirements of ACTION b. or c. of Specification 3.4.4. BRAIDWOOD - UNITS 1 & 2 3/4 4-12
REACTOR COOLANT SYSTEM 3/4.4.5 STEAM GENERATORS AUO 10 E LIMITING CONDITION FOR OPERATION 3.4.5 Each steam generator shall be OPERABLE. - APPLICABILITY: MODES 1, 2, 3 and 4. MTION: With one or more steam generators inoperable, restore the inoperable steam generator (s) to OPERABLE status prior to increasing T,yg above 200*F. SURVEILLANCE REQUIREMENTS 4.4.5.0 Each steam generator shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the requirements of Specification 4.0.5. l 4.4.5.1 Steam Generator Sample Selection and Inspection - Each steam generator l shall be determined OPERABLE during shutdown by selecting and inspecting at least the minimum number of steam generators specified in Table 4.4-1. 4.4.5.2 Steam Generator Tube Sample Selection and Inspection - The steam generator tube minimum sample size, inspection result classification, and the corresponding action required shall be as specified in Table 4.4-2. The inservice inspection of steam generator tubes shall be performed at the fre-quencies specified in Specification 4.4.5.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 4.4.5.4. The tubes selected for each inservice inspection shall include at least 3% of the total number of tubes in all steam generators; the tubes selected for these inspections shall be selected on a random basis except:
- a. Where experience in similar plants with similar water chemistry indicates critical areas to be inspected, then at least 50% of the tubes inspected shall be from these critical areas;
- b. The first sample of tubes selected for each inservice inspection (subsequent to the preservice inspection) of each steam generator shall include:
BRAIDWOOD - UNITS 1 & 2 3/4 4-13
~
REACTOR COOLANT SYSTEM AUG 191986 SURVEILLANCE REQUIREMENTS (Continued)
- 1) All nonplugged tubes that previously had detectable wall penetrations (greater than 20% of wall thickness),
- 2) Tubes in those areas where experience has indicated potential problems, and
- 3) A tube inspection (pursuant to Specification 4.4.5.4a.8) shall be performed on each selected tube. If any selected tube does not permit the passage of the eddy current probe for a tube inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection.
- c. The tubes selected as the second and third samples (if required by Table 4.4-2) during each inservice inspection may be subjected to a partial tube inspection provided:
- 1) The tubes selected for these samples include the tubes from those areas of the tube sheet array where tubes with imperfections were previously found, and
- 2) The inspections include those portions of the tubes where imperfections were previously found.
The results of each sample inspection shall be classified into one of the following three categories: Category Inspection Results C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective. C-2 One or more tubes, but not more than 1% of the total tubes inspected are defective, or between 5% and 10% of the total tubes inspected are degraded tubes. C-3 More than 10% of the total tubes inspected are degraded tubes or more than 1% of the inspected tubes are defective. Note: In all inspections, previously degraded tubes must exhibit significant (greater than 10% of wall thickness) further wall penetrations to be included in the above percentage calculations. l BRAIDWOOD - UNITS 1 & 2 3/4 4-14
REACTOR COOLANT SYSTEM E EB15 th M AUG 19 ES6 SURVEILLANCE REQUIREMENTS (Continued) , 4.4.5.3 Inspection Frequencies - The above required inservice inspections of steam generator tubes shall be performed at the following frequencies:
- a. The first inservice inspection shall be performed after 6 Effective Full Power Months but within 24 calerdar months of initial criticality.
Subsequent inservice inspections shall be performed at intervals of , not less than 12 nor more than 24 calendar months after the previous inspection. If two consecutive inspections, not including the pre-service inspection, result in all inspection results faf fing into the C-1 category or if two consecutive inspections demonstrate that pre-viously observed degradation has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of once per 40 months;
- b. If the results of the inservice inspection of a steam generator conducted in accordance with Table 4.4-2 at 40-month intervals fall
- in Category C-3, the inspection frequency shall be increased to at least once per 20 months. The increase in inspection frequency shall apply.until the subsequent inspections satisfy the criteria of Specification 4.4.5.3a.; the interval may then be extended to a maximum of once per 40 months; and
- c. Additional, unscheduled inservice inspections sha?l be perfor ed on-each steam generator in accordance with the first sample inspection specified in Table 4.4-2 during the shutdown subsequent to any of the following conditions:
- 1) Reactor-to-secondary tube leaks (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.4.6.2c., or
- 2) A seismic occurrence greater than the Operating Basis Earthquaks, or
- 3) A Condition IV loss-of-coolant accident requiring actuation of the Engineered Safety Features, or
- 4) A Conditiot IV main steam line or feedwater line break.
1
)
l 1 BRAIDWOOD - UNITS 1 & 2 3/4 4-15
, REACTOR COOLANT SYSTEM FINAL DRAFT l SUkVEILLANCE REQUIREMENTS (Continued) 1 4.4.5.4 Acceptance Criteria
- a. As used in this specification: -
- 1) Imperfection means an exception to the dimensions, finish or contour of a tube from that required by fabrication drawings or specifications. Eddy-current testing indications below 20% of the nominal tube wall thickness, if detectable, may be l
considered as imperfections; "e 2) Degradation means a service-induced cracking, wastage, wear or general corrosion occurring on either inside or outside of a tube;
- 3) , Degraded Tube means a tube containing imperfections greater than or equal to 20% of the nominal wall thickness caused by degradation;
- 4) % Degradation means the percentage of the tube wall thickness affected or removed by degradation;
- 5) Defect means an imperfection of such severity that it exceeds the plugging limit. A tube containing a defect is defective;
- 6) Plugging Limit means the imperfection depth at or beyond which l the tube shall be removed from service and is equal to 40%
of the nominal tube wall thickness; .
- 7) Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integ-rity in the event of an Operating Basis Earthquake, a loss-of-ccolant accident, or a steam line or feedwater line break as specified in 4.4.5.3c., above;
- 8) Tube Inspection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg; and I
O BRAIDWOOD - UNITS 1 & 2 3/4 4-16
REACTOR COOLANT SYSTEM AUG 191986
~
SURVEILLANCE REQUIREMENTS (Continued)
- 9) Preservice Inspection means an inspection of the full length of 1 each tube in each steam generator performed by eddy current techniques prior to service to establish a baseline condition of the tubing. This inspection shall be performed prior to l initial POWER OPERATION using the equipment and techniques I expected to be used during subsequent inservice inspections.
- b. The steam generator shall be determined OPERABLE after completing the corresponding actior.s (plug all tubes exceeding the plugging limit and all tubes containing through-wall cracks) required by Table 4.4-2.
4.4.5.5 Reports i
- a. Within 15 days following the completion of each inservice inspection of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the Commission in a Special Report pursuant to Specification 6.9.2;
- b. The complete results of the steam generator tube inservice inspection shall be submitted to the Commission in a Special Report pursuant to Specification 6.9.2 within 12 months following the completion of the inspection. This Special Report shall include:
- 1) Number and extent of tubes inspected, l 2) Location and percent of wall-thickness penetration for each indication of an imperfection, and
( 3) Identification of tubes plugged.
- c. Results of steam generator tube inspections which fall into Category C-3 shall be reported in a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days and prior to resumption of plant operation. This report shall provide a description of investi-gations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.
BRAIDWOOD - UNITS 1 & 2 3/4 4-17
TABLE 4.4-1 FINAL DRAFT AUG 191986 l MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION Preservice Inspection Yes No. of Steam Generators per Unit Four 1 l First Inservice Inspection Two 1 l Second & Subsequent Inservice Inspections One TABLE NOTATION
- 1. The inservice inspection may be limited to one steam generator on a rotating schedule encompassing 3 N % of the tubes (where N is the number of steam generators in the plant) if the results of the first or previous inspections indicate that all steam generators are performing in a like manner. Note that under some circumstances, the operating conditions in one or more steam generators may be found to be more severe than those in other steam generators. Under such circumstances the sample sequence shall be modified to inspect the most severe conditions. Each of the other two steam generators not inspected during the first inservice inspections shall be inspected during the second and third inspections. The fourth and subsequent inspections shall follow the instructions described above.
BRAIDWOOD - UNITS 1 & 2 3/4 4-18
g TABLE 4.4-2 2 E3 6 - STEAM GENER ATOR TUBE INSPECTION 8 IST SAMPLE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTION Sample Size Result Action Required Result Action Required Result Action Required
]
H A minimum of C-1 None N. A. N. A. N. A. N. A.
- e. S Tubes per ,
o S. G. C-2 Plug defective tubes C-1 No'ne N. A. N. A. and inspect additional Plug defective tubes C-1 None 2S tubes in this S. G. C-2 and inspect additional C-2 Plug defective tubes 4S tubes in this S. G. p,, ,,;,,,,, C-3 C-3 result of first sample R Perform action for i C-3 C-3 result of first N. A. N. A. , U sample C-3 inspect all tubes in All other this S. G., plug de- S. G.s are None N. A. N. A. fective tubes and . C-1
" S me S. G.s Perform action for eac oth N. A. N. A.
C-2 but no C-2 result of second additional sample Notification to NRC S. G. are pursuant to 650.72 C-3 "Pl (b)(2) of 10 CFR Additional Inspect all tubes in p Part 50 S. G. is C-3 each S. G. and plug c defective tubes. [ M Notification to NRC pursuant to 550.72 N. A. N. A. . F (b)(2) of 10 CFR Part 50 hg Where N is the number of steam generators in the unit, and n is the number of steam generators inspected S=3 % during an inspection n
REACTOR COOLANT SYSTEM FINAL HAFT g yg 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE l LEAKAGE DETECTION SYSTEMS ' LIMITING CONDITION FOR OPERATION 3.4.6.1 The following Reactor Coolant System Leakage Detection Systems shall be OPERABLE:
- a. The Containment Atmosphere Particulate Radioactivity Monitoring System,
- b. The Containment Floor Drain and Reactor Cavity Flow Monitoring System, and
- c. The Containment Gaseous Radioactivity Monitoring System.
APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:
- a. 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 containment atmosphere are obtained and analyzed for gaseous and particulate radioactivity at least once per 24 hours when the required Gaseous or Particulate Radioactivity Monitoring System is inoperable; otherwise, be in at least HOT STANDBY within the next 6 hours and ic. COLD SHUTDOWN within the following 30 hours.
- b. With b. of the above required Leakage Detection Systems inoperable be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- c. With a. and c. of the above required Leakage Detection Systems inoperable:
- 1) Restore either Monitoring System (a. or c.) to OPERABLE status within 72 hours and
- 2) Obtain and analyze a grab sample of the containment atmosphere for gaseous and particulate radioactivity at least once per 24 hours, and
- 3) Perform a Reactor Coolant System water inventory balance at least once per 8 hours.
Otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.4.6.1 The Leakage Detection Systems shall be demonstrated OPERABLE by:
- a. Containment Atmosphere Gaseous and Particulate Monitoring System-performance of CHANNEL CHECK, CHANNEL CALIBRATION, and DIGITAL CHANNEL OPERATIONAL TEST at the frequencies specified in Table 4.3-3,
- b. Containment Floor Drain and Reactor Cavity Flow Monitoring System-performance of CHANNEL CALIBRATION at least once per 18 months, and
- c. Verify the oil separator portion of the containment floor drain collection sump has been filled to the level of the overflow to the containment floor drain unidentified leakage collection weir box once per 18 months, following refueling, and prior to initial startup.
BRAIDWOOD - UNITS 1 & 2 3/4 4-20
REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE D 19 E LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to:
- a. No PRESSURE B0UNDARY LEAKAGE,
- b. 1 gpm UNIDENTIFIED LEAKAGE,
- c. 1 gpm total reactor-to-secondary leakage through all steam generators not isolated from the Reactor Coolant System and 500 gallons per day through any one steam generator,
- d. 10 gpm IDENTIFIED LEAKAGE from the Reactor Coolant System,
- e. 40 gpm CONTROLLED LEAKAGE at a Reactor Coolant System pressure of 2235 1 20 psig, and f.
1 gpm leakage at a Reactor Coolant System pressure of 2235 1 20 psig from any Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1.* APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:
- a. With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours.
- b. With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE and leakage from Reactor Coolant System Pressure Isolation Valves, reduce the leakage rate to within limits within 4 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- c. With any Reactor Coolant System Pressure Isolation Valve leakage greater than the above limit, reduce the leakage rate to within limits within 4 hours, or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- Test pressures less than 2235 psig but greater than 350 psig are allowed.
Observed leakage shall be adjusted for the actual test pressure up to 2235 psig assuming the leakage to be directly proporational to pressure differential to the one-half power. BRAIDWOOD - UNITS 1 & 2 3/44-21
REACTOR COOLANT SYSTEM FINAL MAFT g yg g SURVEILLANCE REQUIREMENTS 4.4.6.2.1 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by:
- a. " Monitoring the containment atmosph'ere gaseous and particulate radioactivity monitor at least once per 12 hours;
- b. Monitoring the reactor cavity sump discharge, and the containment floor drain sump discharge and inventory at least once per 12 hours;
- c. Measurement of the CONTROLLED LEAKAGE from the reactor coolant pump seals when the Reactor Coolant System pressure is 2235 1 20 psig at least once per 31 days with the modulating valve fully open. The provisions of Specification 4.0.4 are not applicable for-entry into MODE 3 or 4;
- d. Performance of a Reactor Coolant System water inventory balance at '
least once per 72 hours; and
- e. Monitoring the Reactor Head Flange Leakoff System at least once per 24 hours.
4.4.6.2.2 Each Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1 shall be demonstrated OPERABLE by verifying leakage to be within its limit:
- a. At least once per 18 months,
- b. Prior to entering MODE 2 whenever the plant has been in COLD SHUTOOWN for 72 hours or more and if leakage testing has not been performed in the previous 9 months,
- c. Prior to returning the valve to service follow *ng maintenance, repair or replacement work on the valve, and
- d. Within 24 hours following valve actuation due to automatic or manual action or flow through the valve except for valves RH 8701 A and B l and RH 8702 A and B.
The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4. BRAIDWOOD - UNITS 1 & 2 3/4 4-22
TABLE 3.4-1 AlfG 11 are REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES VALVE NUMBER FUNCTION / SI8900A,B,C,D CHG/SI Check Valve SI8815 CHG/SI Backup Check Valve SI8948A,B,C,0 Accumulator Check Valve SI8956A,B,C,D Accumulator Backup Check Valve SI8818A,B,C,0 RHR Cold Leg Check Valve SI8819A,B,C,D SI Cold Leg Check Valve SI8949A,B,C,D SI Hot Leg Check Valve SI8905A,B,C,D SI Hot Leg Backup Check Valve SI8841A,B RHR Hot Leg Check Valve
*RH8701A,8 RHR Suction MOV's
*RH8702A,8 RHR Suction M0V's
- NOTE:
- 1. Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm are considered acceptable if the latest measured rate has not exceeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of 5.0 gpm by 50% or greater,
- 2. Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm are considered unacceptable if the latest measured rate exceeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of 5.0 gpm by 50% or greater, and
- 3. Leakage rates greater than 5.0 gpm are considered unacceptable.
l BRAIDWOOD - UNITS 1 & 2 3/4 4-23
REACTOR COOLANT SYSTEM 3/4.4.7 CHEMISTRY , l LIMITING CONDITION FOR OPERATION l 3.4.7 The Reactor Coolant System chemistry shall be maintained within the limits specified in Table 3.4-2. APPLICABILITY: At a'11 times. ACTION: MODES 1, 2, 3, and 4:
- a. With any one or more chemistry parameter in excess of its Steady-State Limit but within its Transient Limit, restore the parameter to within its Steady-State Limit within 24 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours; and -
- b. With any one or more chemistry parameter in excess of its Transient Limit, be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours. ,
At All Other Times: With the concentration of either chloride or fluoride in the Reactor Coolant System in excess of its Steady-State Limit for more than 24 hours I or in excess of its Transient Limit, reduce the pressurizer pressure to ) less than or equal to 500 psig, if applicable, and perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation prior j to increasing the pressurizer pressure above 500 psig or prior to 1 proceeding to MODE 4. l SURVEILLANCE REQUIREMENTS 4.4.7 The Reactor Coolant System chemistry shall be determined to be within the limits by analysis of those parameters at the frequencies specified in Table 4.4-3. BRAIDWOOD - UNITS 1 & 2 3/4 4-24
TABLE 3.4-2 REACTOR COOLANT SYSTEM NUE I O l900 CHEMISTRY LIMITS STEADY-STATE TRANSIENT PARAMETER LIMIT LIMIT Dissolved Oxygen
- 5 0.10 ppm 5 1.00 ppm Chloride 5 0.15 ppm i 1.50 ppm Fluoride 5 0.15 ppm 5 1.50 ppm
" Limit not applicable with T,yg less than or equal to 250 F.
l l l BRAIDWOOD - UNITS 1 & 2 3/4 4-25
TABLE 4.4-3 FINAL HAFT ! AUG 191986 REACTOR COOLANT SYSTEM CHEMISTRY SURVEILLANCE REQUIREMENTS SAMPLE AND PARAMETER ANALYSIS FREQUENCY Dissolved Oxygen
- At least once per 72 hours Chloride At least once per 72 hours Fluoride At least once per 72 hours l
"Not required with T,yg less than or equal to 250*F l
BRAIDWOOD - UNITS 1 & 2 3/4 4-26
REACTOR COOLANT SYSTEM FINAL MAFT 3/4.4.8 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.8 The specific activity of the reactor coolant shall be limited'to:
- a. Less than or equal to 1 microcurie per gram DOSE EQUIVALENT I-131, and
- b. Less than or equal to 100/E microcuries per gram of gross radioactivity.
APPLICABILITY: MODES 1, 2, 3, 4, and 5. ACTION: MODES 1, 2 and 3*:
- a. With the specific activity of the reactor coolant greater than 1 microcurie per gram DOSE EQUIVALENT I-131 but within the allowable limit (below and to the left of the line) shown on Figure 3.4-1, .
operation may continue for up to 48 hours provided that the cumula-tive operating time under these circumstances does not exceed 800 hours in any consecutive 12-month period. The provisions of Specification 3.0.4 are not applicable;
- b. With the total cumulative operating time at a reactor coolant specific activity greater than 1 microcurie per gram DOSE EQUIVALENT I-131 exceeding 500 hours in any consecutive 6-month period, prepare and submit a Special Report to the Commission within 30 days, pursuant to Specification 6.9.2, indicating the nt;mber of hours above this limit. The provisions of Specification 3.0.4 are not applicable;
- c. With the specific activity of the reactor coolant greater than 1 microcurie per gram DOSE EQUIVALENT I-131 for more than 48 hours during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in at least HOT STANDBY with T less than 500 F within 6 hours; and avg l
f d. With the specific activity of the reactor coolant greater than 100/E i microCuries per gram of gross radioactivity, be in at least HOT l STANDBY with T,yg less tnan 500 F within 6 hours.
- With T,yg greater than or equal to 500 F. ,
BRAIDWOOD - UNITS 1 & 2 3/4 4-27
REACTOR COOLANT SYSTEM AUG 191986 LIMITING CONDITION FOR OPERATION ACTION (Continued) MODES 1, 2, 3, 4, and 5: With the specific activity of the reactor coolant greater :than 1 mir.roCurie per gram DOSE EQUIVALENT I-131 cr greater than 100/E microCuries per gram of gross radioactivity, 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. Prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days with a copy to the Director, Nuclear Reactor Regulation, Attention: Chief, Core Performance Branch, and Chief, Accident Evaluatior. Branch, U.S. Nuclear Regulatory Commission, Washington, D.C., 20555. This report shall contain the results of the specific activity analyses together with the following information:
- a. Reactor power history starting 48 hours prior to the first sample in which the limit was exceeded;
- b. Results of: (1) the last isotopic analysis f6r radioiodine performed prior to exceeding the limit, (2) analysis while limit was exceeded, and (3) one analysis after the radioiodine was reduced to less than the limit, including for each isotopic analysis, the date and time of sampling and the radioiodine concentration;
- c. Clean-up flow history starting 48 hours prior to the first sample in which the limit was exceeded,
- d. History of degassing operations, if any, starting 48 hours prior to the first sample in which the limit was exceeded, and
- e. The time duration when the specific activity of the primary coolant I exceeded 1 microcurie per gram DOSE EQUIVALENT I-131.
l I SURVEILLANCE REQUIREMENTS l 4.4.8 The specific activity of the reactor coolant shall be determined to be ' within the limits by performance of the sampling and analysis program of Table 4.4-4. BRAIDWOOD - UNITS 1 & 2 3/4 4-28
l FINAL DM I s A U G 1 S B lfr 3 l, g ROG 15 Ilus E 250 \i b \ E a h i
\i 4
200 h 3 4 h UNACCEPTABLE 9 \ OPERATION ; e ( l o i T E
" \
$ 150 \'
5 o ' o ' ' T o g t 4 T E g 100 \
- a. ' t h
L 11
>- T Z ACCEPTABLE \
$ OPERATION s 50 s
C w 4 o 0 0 20 30 40 50 60 70 80 90 100 PERCENT OF RATED THERMAL POViER . FIGURE 3.4-1 4 l DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY - ! LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY > 1pCi/ GRAM DOSE EQUIVALENT I-131 ; i BRAIDWOOD - UNITS 1 & 2 3/4 4-29 l
TABLE 4.4-4 E D REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE y AND ANALYSIS PROGRAM O TYPE OF MEASUREMENT SAMPLE AND ANALYSIS
' MODES IN WHICH SAMPLE AND ANALYSIS FREQUENCY AND ANALYSIS REQUIRED 5 1. Gross Radioactivity At least once per 72 hours 1, 2, 3, 4 d Determination **
- 2. Isotopic Analysis for DOSE EQUIVA-8' Once per 14 days 1 LENT I-131 Concentration to
- 3. Radiochemical for E Determination *** Once per 6 months
- 1
- 4. Isotopic Analysis for Iodine a) Once per 4 hours, 1#, 2#, 3#, 4#, 5#
Including I-131, I-133, and I-135 whenever the specific activity exceeds 1 pCi/ gram DOSE ' R
- EQUIVALENT I-131 or 100/E pCi/ gram t of gross radioactivity, 8 and
'b) One sample between 2 1,2,3 and 6 hours following a THERMAL POWER change exceeding 15%
of the RATED THERMAL POWER within a 1-hour period. - 5
=
3>
' 91 C:ll3 E
21
TABLE 4.4-4 (Continued) TABLE NOTATIONS
#Until the specific activity of the Reactor Coolant System is restored within its limits.
- Sample to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since reactor was last subcritical for 48 hours or longer.
**A gross radioactivity analysis shall consist of the quantitative measurement of the total specific activity of the reactor coolant except for radionuclides with half-lives less than 10 minutes and all radiciodines. The total specific activity shall be the sum of the degassed beta gamma activity and the total of all identified gaseous activities in the sample within 2 hours after the sample is taken and extrapolated back to when the sample was taken. Determination of the contributors to the gross specific activity shall be based upon those energy peaks identifiable with a 95% confidence level. The latest available data may be used for pure beta-emitting radio-nuclides. . ***A radiochemical analysis for I shall consist of the quantitative measurement of the specific activity for each radionuclide, except for radionuclides with half-lives less than 10 minutes and all radio-iodines, which is identified in the reactor coolant. The specific activities for these individual radionuclides shall be used in the determination of E for the reactor coolant sample. Determination of the contributors to E shall be based upon these energy peaks identifiable with a 95% confidence level.
l BRAIDWOOD - UNITS 1 & 2 3/4 4-31
REACTOR COOLANT SYSTEM AUG 19 % 3/4.4.9 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION l 1 l 3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and l pressure shall be limited in accordance with the limit lines shown on Figures 3.4-2a and 3.4-3a for Unit 1 (Figures 3.4-2b and 3.4-3b for Unit 2) during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:
- a. A maximum heatup of 100 F in any 1-hour period,
- b. A maximum cooldown of 100 F in any 1-hour period, and
- c. A maximum temperature change of less than or equal to 10 F in any 1-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.
APPLICABILITY: At all times. ACTION: With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation or be in at least HOT STANDBY within the next 6 hours and reduce the RCS T,yg and pressure to less than 200 F and 500 psig, respectively, within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operations. 4.4.9.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, as required by 10 CFR Part 50, Appendix H, in accordance with the schedule in Table 4.4-5. The results of these examinations shall be used to update Figures 3.4-2, 3.4-3, and 3.4-4. BRAIDWOOD - UNITS 1 & 2 3/4 4-32
- FINAL DRAFT
. AUG1ggyg 3000 LEAR TEST LIMIT
/ ) I i / /
/ ; i 2000 r f
g a / / E / / E ) ) 5 '
/ f 5 / /
E / / e / t a / / , o EEATUP RATES j i 5 1000 UP 70 60*F/ER z
/
- CRITICALITT LIMIT .
j/ BASED OR INSERVICE *
/ NYDROSTATIC TEST
# TEMPERATURE (267'F) .
FOR TEE SERVICE PER10D UP 70 32 IFFY
- 0-,- -
0 100 , 200 300 400 1 INDICATED TEMPERATURE ('F) l FIGURE 3.4-2a REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE UP TO 32 EFPY (UNIT 1) i l BRAIDWOOD - UNITS 1 & 2 3/4 4-33
FINAL HAFT AUG 19 m; 3000 f IZAR Test LIMIT ! W / / / 1 f
/ /
I awo / / / uj / /
/ /
g } } g / / ( f
- / /
nAnr uns / / 1000 -- UP To 60*F/ n / /
/ '
l 11 \ l
! CRITICALITT LIMIT
/ BASED ON INSERV12 7 5'F)
_/ FOR TIE SERVI G PERIOD UP TO 16 EFFT O, 0 100 200 300 400 NDDCATED TERFERATURE - DEG F i FIGURE 3.4-2b REACTOR COOLANT SYSTEM HEATUP LIMITATIONS . APPLICABLE UP TO 16 EFPY (UNIT 2) l BRAIDWOOD - UNITS 1 & 2 3/4 4-34
FINAL DRAFT AUG 191986 3000 l 1 l j I 2000 g l f O } f LU
)
l l000 l C00LDOW RATE *T/ER r r r g ww
*% - - 5&&/
, 40 D ECCF/
-- 100
- o. .
0 100 200 300 400 liODCATED TEMPERATURE - DEG F i l FIGURE 3.4-3a REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS APPLICABLE UP TO 32 EFPY (UNIT 1) BRAIDWOOD - UNITS 1 & 2 3/4 4-35
FINAL DRAFT AUG 191986
'L' 3000 -
}
'h a.
- l. i i
I 2000 I w / u) l
/
/
/
< 1000 \\\\ >: e
.--- coouas un *nn #
,J '
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$ =E=y MN 20o s .-
0 0 100 200 300 400 50DOATED. TEMPERATURE - DEG F , FIGURE 3.4-3b REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS APPLICABLE UP TO 16 EFPY (UNIT 2) 3 BRAIDWOOD - UNITS 1 & 2 3/4 4-36 ? - . _ _ _ _ . - . _ _ _ _ _ _ . .-- . _. . . - - _ -.
~
\ TABLE 4.4-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM - WITHDRAWAL SCHEDULE o h O UNIT-1 I E CAPSULE VESSEL LEAD Q vs NUMBER LOCATION FACTOR WITH0RAWAL TIME (EFPY)* ,
" U 58.5* 3.6 1st Refueling e.
^) X 238.5* 3.6 6 V 61* 3.0 10 Y 241* 3.0 15 w W 121.5* 3.6 Standby 1
. Z 301.5* 3.6 Standby b .
UNIT-2 U 58.5* 4.05 1st Refueling W 121.5 4.05 4.5 Y X 238.5* 4.05 8.0
) E V 61* 3.37 15.O Epr-Y 241* 3.37 Standby M
. Z 301.5 4.05 Standby tll=3 l
l3:=
- Withdrawal time may be modified to coincide with those refueling outages or reactor shutdowns most closely approaching the withdrawal schedule.
REACTOR COOLANT SYSTEM PRESSURIZER , 010IS00 LIMITING CONDITION FOR OPERATION 3.4.9.2 The pressurizer temperature shall be limited to:
- a. A maximum heatup of 100*F in any 1-hour period,
- b. A maximum cooldown of 200 F in any 1-hour period, and
- c. A maximum spray water temperature differential of 320 F.
APPLICABILITY: At all times. ACTION: With the pressurizer temperature limits in excess of any of the above limits, restore the temperature to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the pressurizer; determine that the pressurizer remains acceptable for continued operation or be in at least HOT STANDBY within the next 6 hours and reduce the pressurizer pressure to less than 500 psig within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.4.9.2 The pressurizer temperatures shall be determined to be within the limits at least once per 30 minutes during system heatup or cooldown. The spray water temperature differential shall be determined to be within the limit at least once per 12 hours during auxiliary spray operation. O BRAIDWOOD - UNITS 1 & 2 3/4 4-38
REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS A[Jg y g g LIMITING CONDITION FOR OPERATION 3.4.9.3 At least one of the following Overpressure Protection Systems shall be OPERABLE:
- a. Two residual heat removal (RHR) suction relief valves each with a Setpoint of 450 psig i 1%, or
- b. Two power-operated relief valves (PORVs) with lift Setpoints that vary with RCS temperature which do not exceed the limit established in Figure 3.4-4, or
- c. The Reactor Coolant System (RCS) depressurized with an RCS vent of greater than or equal to 2 square inches.
APPLICABILITY: MODES 4 and 5, and MODE 6 with the reactor vessel head on. i ACTION:
- a. With one PORV and one RHR suction relief valve inoperable, either restore two PORVs or two RHR suction relief valves to OPERABLE status within 7 days or depressurize and vent the RCS through at least a 2 square inch vent within the next 8 hours.
- b. With both PORVs and both RHR suction relief valves inoperable,
[ depressurize and vent the RCS through at least a 2 square inch vent I within 8 hours. l l
- c. In the event the PORVs, or the RHR suction relief valves, or the RCS vent (s) are used to mitigate an RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days. The report shall describe the circumstances initiating the transient, the effect of the PORVs, or the RHR suction relief valves, or RCS vent (s) on the transient, and any corrective action necessary to prevent recurrence.
- d. The provisions of Specification 3.0.4 are not applicable.
BRAIDWOOD - UNITS 1 & 2 3/4 4-39 l l
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T. NTD LOWWST COMS NTO TEMPERATURE (T) . FIGURE 3.4-4 NOMINAL PORV PRESSURE RELIEF SETPOINT VERSUS RCS TEMPERATURE FOR THE COLD OVERPRESSURE PROTECTION SYSTEM APPLICABLE UP TO 10 EFPY BRAIDWOOD - UNITS 1 & 2 3/4 4-40
FINAL DRAFT REACTOR COOLANT SYSTEM g yg g SURVEILLANCE REQUIREMENTS 4.4.9.3.1 Each PORV shall be demonstrated OPERABLF by:
- a. Performance of an ANALOG CHANNEL OPERATIONAL TEST on the PORV actuation channel, but excluding valve operation, within 3'1 days prior to entering a condition in which the PORV is required OPERABLE and at least once per 31 days thereafter when the PORV is required OPERABLE;
- b. Performance of a CHANNEL CALIBRATION on the PORV actuation channel at least once per 18 months; and
- c. Verifying the PORV isolation valve is open at least once per 72 hours when the PORV is being used for overpressure protection.
4.4.9.3.2 Each RHR suction relief valve shall be demonstrated OPERABLE when . the RHR suction relief valves are being used for cold overpressure protection as follows:
- a. For RHR suction relief valve 8708B:
- 1) By verifying at least once per 31 days that RHR RCS Suction Isolation Valve RH8702A is open with power to the valve cperator removed, and
- 2) By verifying at least onca per '2 hours that RH8702B is open.
- b. For RHR suction relief valve 8708A:
- 1) By verifying at least once per 31 days that RH8701B is open with power to the valve operator removed, and
- 2) By verifying at least once per 12 hours that RH8701A is open.
- c. Testing pursuant to Specification 4.0.5.
4.4.9.3.3 The RCS vent (s) shall be verified to be open at least once per 12 hours
- when the vent (s) is being used for overpressure protection.
"Except when the vent pathway is provided with a valve which is locked, sealed, or otherwise secured in the open position, then verify these valves open at least once per 31 days. BRAIDWOOD - UNITS 1 & 2 3/4 4-41
REACTOR COOLANT SYSTEM FINR HAFT 3/4.4.10 STRUCTURAL INTEGRITY l 1 LIMITING CONDITION FOR OPERATION
. 3.4.10 The structural integrity of ASME Code Class 1, 2, and 3 comp'onents shall be maintained in accordance with Specification 4.4.10.
APPLICABILITY: All MODES. ACTION:
- a. With the structural integrity of any ASME Code Class 1 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature above 200*F.
- b. With the structural integrity of any ASME Code Class 2 component (s) not conforming to the above requirements, restere the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature above 200*F.
- c. With the structural integrity of any ASME Code Class 3 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) from service.
- d. The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.4.10 In addition to the requirements of Specification 4.0.5, each reactor coolant pump flywheel shall be inspected as follows: !
- a. Volumetric examination of the areas of higher stress concentration at the bore and keyways will be performed each 40 month period during i refueling or maintenance shutdowns coinciding with the service inspection schedule as required by Section XI of the ASME Code.
- b. Visual examination of all exposed surfaces will be performed and a surface examination of the bore and keyway surfaces will be per-formed whenever the flywheels are removed for maintenance purposes, but not more frequently than once each 10 year interval.
BRAIDWOOD - UNITS 1 & 2 3/4 4-42 l
REACTOR COOLANT SYSTEM { 3/4.4.11 REACTOR COOLANT SYSTEM VENTS AUG 191986 LIMITING CONDITION FOR OPERATION 3.4.11 At least one reactor vessel head vent path consisting of two ~ valves in series powered from emergency busses shall be OPERABLE and closed. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With the above reactor vessel head vent path inoperable, STARTUP and/or POWER OPERATION may continue provided the inoperable vent path is main-tained closed with power removed from the valve actuator of all the valves in the inoperable vent path; restore the inoperable vent path.to OPERABLE status within 30 days, or, be in HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS l 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 valve in the vent path through at least one complete cycle of full travel from the control room during COLD SHUTDOWN or REFUELING, and
- c. Verifying flow through the reactor vessel head vent paths during venting operations at COLD SHUTDOWN or REFUELING.
l BRAIDWOOD - UNITS 1 & 2 3/4 4-43
3/4.5 EMERGENCY CORE COOLING SYSTEMS W . 3/4.5.1 ACCUMULATORS . AUG1S1!26 I l LIMITING CONDITION FOR OPERATION
't 3.5.1 Each Reactor Coolant System accumulator shall be OPERABLE with:
- a. The isolation valve open and power removed,
- b. A contained borated water level of between 31% and 63%,
- c. A boron concentration of between 1900 and 2100 ppm, and
- d. A nitrogen cover pressure of between 602 and 647 psig.
~
APPLICABILITY: MODES 1, 2, and 3*. 3 A_CT C g:
- a. With one accumulator inoperable, except as a result of a closed isolation valve, restore the inoperable accumulator to OPERABLE status within 1 hour or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
l b. With one accumulator inoperable due to the isolation valve being closed, either immediately open the isolation valve or be in at least HOT STANDBY within 6 hours and in HOT SHUTDOWN within the following 6 hours. SURVEILLANCE REQUIREMENTS 4.5.1.1 Each accumulator shall be demonstrated OPERABLE:
- a. At least once per 12 hours by:
- 1) Verifying, by the absence of alarms, the contained borated water level and nitrogen cover pressure in the tanks, and
- 2) Verifying that each accumulator isolation valve is open.
l i
" Pressurizer pressure above 1000 psig.
BRNIDWOOD-UNITS 1&2 3/4 5-1
1 EMERGENCY CORE COOLING SYSTEMS ** "" AUG 19 E06 SURVEILLANCE REQUIREMENTS (Continued)
- b. At least once per 31 days and within 6 hours after each solution volume increase of greater than or equal to 70 gallons by verifying the boron concentration of the accumulator solution, and !
- c. At least once per 31 days when the RCS pressure is above 1000 psig by verifying that the MCC compartment is open and tagged out of service.
4.5.1.2 Each accumulator water level and pressure channel shall be demonstrated OPERABLE at least once per 18 months by the performance of a CHANNEL CALIBRATION. i BRAIDWOOD - UNITS 1 & 2 3/4 5-2
l l i EMERGENCY CORE COOLING SYSTEMS FIN D 3/4"5.2 ECCS SUBSYSTEMS - T,yg > 350'F IO E LIMITING CONDITION FOR OPERATION l l 3.5.2 Two independent Emergency Core Cooling System (ECCS) subsystems shall be OPERABLE with each subsystem comprised of:
- a. One OPERABLE centrifugal charging pump,
- b. One OPERABLE Safety Injection pump,
- c. One OPERABLE RHR heat exchanger,
- d. One OPERABLE RHR pump, and .
- e. An OPERABLE flow path
- capable of taking suction from the refueling water storage tank on a Safety Injection signal and automatic opening of the containment sump suction valves.
APPLICABI'LITY: MODES 1, 2, and 3. l ACTION:
- a. With one ECCS subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in NOT SHUTDOWN within the following 6 hours.
- b. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describ-ing the circumstances of the actuation and the total accumulated actuation cycles to date. The current value of the usage factor for each affected Safety '.njection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.
"During MODE 3, the discharge paths of both Safety Injection pumps may,be isolated by closing SI 8835 for a period of up to 2 hours to perform surveil-lance testing as required by Specification 4.4.6.2.2.
BRAIDWOOD - UNITS 1 & 2 3/4 5-3
n EMERGENCY CORE COOLING SYSTEMS AUG 191986 SURVEILLANC2 REQUIREMENTS 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE:
- a. At least once per 12 hours by verifying that the following valves are in the indicated positions with power to the valve operators removed:
Valve Number Valve Function Valve Position. MOV SI8806 Suction to the SI Open rumps MOV SI8835 SI Pump Discharge Open To RCS Cold Legs MOV SI8813 SI Pump Recirculation Open To The RWST M0V SI8809A RHR Pump Discharge to Open RCS Cold Legs MOV SI88098 RHR Pump Discharge to Open RCS Cold Legs MOV SI8840 RHR Pump Discharge to Closed RCS Hot Legs MOV SI8802A SI Pump Discharge to Closed RCS Hot Legs MOV SI8802B SI Pump Discharge to Closed RCS Hot Legs
- b. At least once per 31 days by:
- 1) Verifying that the ECCS piping is full of water by venting the ECCS pump casings and accessible discharge piping high points, and
- 2) Verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.
- c. By a visual inspection which verifies that no loose debris (rags, trash, clothing, etc.) is present in the containment which could be transported to the containment sump and cause restriction of the pump suctions during LOCA conditions. This visual inspection shall be performed:
BRAIDWOOD - UNITS 1 & 2 3/4 5-4
l EMERGENCY CORE COOLING SYSTEMS AUG 191986 SURVEILLANCE REQUIREMENTS (Continued) *
- 1) For all accessible areas of the containment prior to establishing CONTAINMENT INTEGRITY, and
- 2) Of the areas affected within containment at the completion of each containment entry when CONTAINMENT INTEGRITY is established.
- d. At least once per 18 months by:
- 1) Verifyin,q automatic isolation and interlock action of the RHR System from the Reactor Coolant System by ensuring that:
a) With a simulated or actual Reactor Coolant System pressure signal greater than or equal to 360 psig the interlocks prevent the valves from being opened, and b) With a simulated or actual Reactor Coolant System pressure signal greater than or equal to 662 psig the interlocks will cause the valves to automatically close.
- 2) A visual inspection of the containment semp and verifying that -
the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens, etc.) show no evidence of structural distress or abnormal corrosion.
- e. At least once per 18 months, during shutdown, by:
- 1) Verifying that each automatic valve in the flow path actuates to its correct position on a Safety Injection test signal and on a RWST Level-Low-Low test signal, and
- 2) Verifying that each of the following pumps start automatically upon receipt of a Safety Injection actuation test signal:
a) Centrifugal charging pump, b) Safety Injection pump, and c) RHR 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) Centrifugal charging pump 1 2396 psid,
- 2) Safety Injection pump 2 1412 psid, and
- 3) RHR pump In accordance with Figure' 4.5-1 BRAIDWOOD - UNITS 1 & 2 3/4 5-5
EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
- g. By verifying the correct position of each mechanical position stop for the following ECCS throttle valves:
- 1) Within 4 hours following completion of each valve stroking operation or maintenance on the valve when the ECCS subsystems are required to be OPERABLE, and
- 2) At least once per 18 months.
High Head SI System SI System Valve Number Valve Number SI8810 A,B,C,D SI8822 A,B,C,D SI8816 A,B,C,0
- h. By performing a flow balance test, during shutdown, following com-pletion of modifications to the ECCS subsystems that alter the subsystem flow characteristics and verifying that:
- 1) For centrifugal charging pump lines, with a single pump running:
a) The sum of the injection line flow rates, excluding the highest flow rate, is greater than or equal to 330 gpm, and b) The total pump flow rate is less than or equal to 550 gpm, including a simulated seal injection flow of 80 gpm.
- 2) For Safety Injection pump lines, with a single pump running:
a) The sum of the injection line flow rates, excluding the highest flow rate, is greater than or equal to 439 gpm, and b) The total pump flow rate is less than or equal to 655 gpm.
- 3) For RHR pump lines, with a single pump running, the sum of the injection line flow rates is greater than or equal to 3804 gpm.
BRAIDWOOD - UNITS 1 & 2 3/4 5-6
--- - - - - - - - - , , - - , - , , - - ,, ,n. - , -- , - - - --w, w n -- -w-ww,e
-"- FINAL DRAFT
_ _m-- ---
- n. -- -
--+ Selected Points on the Curve: -2 Flow Head 0 187 500 181 200 1000 177 - - - -
1500 171
- 2000 164 190 - - = ~~
' 2500 159 C -
3000 155 -- 3500 149 180
----= -
4000 139
'- 4500 127
~
E 112
'_-^
170 5
~
S, '% H- . E w 160 '_ _
! ACCEPTABLE OPERATION -_,,_
150 - g E 140 - lll UNACCEPTABLE 'F . r 130 OPERATION 120
~ ~
110
-- t - ---
100 --- - ' - - C - _ _ p -i 0 1000 2000 3000 4000 5000 RHR PUMP FLOW (GPM) FIGURE 4.5-1 RESIOUAL HEAT REMOVAL PUMP ' MINIMUM ACCEPTABLE PERFORMANCE CURVE BRAIDWOOD - UNITS 1 & 2 3/4 5-6a
l EMERGENCY CORE COOLING SYSTEMS 3/4.5.3 ECCS SUBSYSTEMS - T 3y < 350 F g yg LIMITING CONDITION FOR OPERATION l 3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:
- a. One OPERABLE centrifugal charging pump,*
- b. One OPERABLE RHR heat exchanger,
- c. One OPERABLE RHR pump, and
- d. An OPERABLE flow path capable of taking suction from~the refueling water storage tank upon being manually realigned and transferring suction to the containment sump during the recirculation phase of operation.
APPLICABILITY: MODE 4. ACTION:
- a. With no ECCS subsystem OPERABLE because of the inoperability of either the centrifugal charging pump or the flow path from the refueling water storage tank, restore at least one ECCS subsystem to OPERABLE status within 1 hour or be in COLD SHUTDGWN within the next 20 hours.
- b. With no ECCS subsystem OPERABLE because of the inoperability of either the RHR heat exchanger or RHR pump, restore at least one ECCS subsystem to OPERABLE status or maintain the Rettctor Coolant System T,yg less than 350 F by use of alternate heat removal methods.
- c. In the event the ECCS is actuated and injects water into the Reactor Ccolant System, a Special Report shall be prepared and submitted to the Commission within 90 days, pursuant to Specification 6.9.2, describing the circumstances of the actuation and the total accumulated
[ actuation cycles to date. The current value of the usage factor for l each affected Safety Injection nozzle shall be provided in this
- Special Report whenever its value exceeds 0.70.
"A maximum of one charging pump shall be OPERABLE, and that pump shall be a centrifugal charging pump, whenever the temperature of one or more of the RCS cold legs is less than or equal to 330 F.
l - l t BRAIDWOOD - UNITS 1 & 2 3/4 5-7
EMERGENCY CORE COOLING SYSTEMS 8 SURVEILLANCE REQUIREMENTS 4.5.3.1 The ECCS subsystem shall be demonstrated OPERABLE per the applicable requirements of Specification 4.5.2. 4.5.3.2 All charging pumps and Safety Injection pumps, except the above allowed OPERABLE pumps, shall be demonstrated inoperable
- by verifying that the motor circuit breakers are secured in the open position at least once per 12 hours whenever the temperature of one or more of the RCS cold legs is less than or equal to 330*F.
- An inoperable pump may be energized for testing or for filling accumulators provided the discharge of the pump has been isolated from the RCS by a closed isolation valve with power removed from the valve operator, or by a manual isolation valve secured in the closed position.
BRAIDWOOD - UNITS 1 & 2 3/4 5-8
EMERGENCY CORE COOLING SYSTEMS 3/4.5.4 REFUELING WATER STORAGE TANK AUG 191956 LIMITING CONDITION FOR OPERATION 3.5.4 The refueling water storage tank (RWST) and the heat traced portion of the RWST vent path shall be OPERABLE with:
- a. A minimum contained borated water level of 89%,
- b. A minimum boron concentration of 2000 ppm,
- c. A minimum water temperature of 35 F, and
- d. A maximum water temperature of 100 F.
APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With the RWST inoperable, restore the tank to OPERABLE status within 1 hotfr or be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS l 4.5.4 The RWST shall be demonstrated OPERABLE:
- a. At least once per 7 days by:
- 1) Verifying the contained borated water level in the tank, and
- 2) Verifying the boron concentration of the water.
- b. At least once per 24 hours by verifying the RWST temperature when the outside air temperature is either less than 35*F or greater than 100*F, and
- c. At least once per 24 hours by verifying the RWST vent path temperature to be greater than or equal to 35*F when the outside air temperature is less than 35*F. ,
l BRAIDWOOD - UNITS 1 & 2 3/4 5-9 l
a me at 3/4.6 CONTAINMENT SYSTEMS
~
3/4.6.1 PRIMARY CONTAINMENT AUG 1 S 55 CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.1 Primary CONTAINMENT INTEGRITY shall be maintained. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY within I hour or be in at least HOT STAN0BY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.6.1.1 Primary CONTAINMENT INTEGRITY shall be. demonstrated:
- a. At least once per 31 days by verifying that all penetrations
- not capable of being closed by OPERABLE containment automatic isolation valves and required to be closed during accident conditions are i closed by valves, blind flanges, or deactivated automatic valves secured in their positions, except as provided in Table 3.6-1 of Specification 3.6.3;
- b. By verifying that each containment air lock is in compliance with the requirements of Specification 3.6.1.3; and
- c. After each closing of each penetration subject'to Type B testing, except the containment air locks, if opened following a Type A or B test, by leak rate testing the seal with gas at a pressure not less than P , 44.4 psig, and verifying that when the measured leakage rate forth$sesealsisaddedtotheleakageratesdeterminedpursuantto Specification 4.6.1.2d. for all other Type B and C penetrations, the combined leakage rate is less than 0.60 L,.
"Except valves, blind flanges, and deactivated automatic valvos which are located inside the containment and are locked, sealed or otherwise secured in the closed position. These penetrations shall be verified closed during ,
each COLD SHUTDOWN except that such verification need not be performed more l often than once per 92 days. , BRAIDWOOD - I' NITS 1 & 2 3/4 6-1
l CONTAINMENT SYSTEMS FINAL MAFT
, AUG 191986 CONTAINMENT LEAKAGE LIMITING CONDITION FOR OPERATION 3.6.1.2 Containment leakage rates shall be limited to: l
- a. An overall integrated leakage rate of:
- 1) Less than or equal to L,, 0.10% by weight of the gontainment air per 24 hours at P,, 44.4 psig, or
- 2) Less than or equal to L t, 0.07% by weight of the containment air per 24 hours for Unit 1 (Later % by weight cf the contain-ment air per 24 hours for Unit 2) at Pt , 22.2 psi0.
- b. A combined leakage rate of less than 0.60 L, for all penetrations and valves subject to Type B and C tests, when pressurized to P,.
APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With either the measured overall integrated containment leakage rate exceeding 0.75 L, or 0.75 L t, as applicable, or the measured combined leakage rate for all penetrations and valves subject to Types B and C tests exceeding 0.60 L,, restoretheoverallintegratedleakageratetolessthan0.75L,orley,sthan 0.75L,asapplicable,andthecombinedleakagerateforallpenetragions t subject to Type B and C tests to less than 0.60 L, prior to increasing the Reactor Coolant System temperature above 200*F. i
\
i SURVEILLANCE REQUIREMENTS 4.6.1.2 The containment leakage rates shall be demonstrated at the following test schedule and shall be determined in conformance with the criteria speci-fied in Appendix J of 10 CFR Part 50 using the methods and provisions of ANSI N45.4-1972: l a. Three Type A tests (Overall Integrated Containment Leakage Rate) shall be conducted at 40 i 10 month intervals during shutdown at a pressure not less than P,, 44.4 psig, or Pt. 22.2 psig, during each 10 year service period. The third test of each set shall be . conducted during the shutdown for the 10 year plant inservice inspection; BRAIDWOOD - UNITS 1 & 2 3/4 6-2
CONTAINMENT SYSTEMS AUG 191986 SURVEILLANCE REQUIREMENTS (Continued) b. If any periodic Type A test fails to meet either 0.75 L, or 0.75 Lt ' the test schedule for subsequent Type A tests shall be revfewed 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,, 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 supplemental test result, Lc , is in accordance with the follow-ing equation:
Lc ~(l am * 'c) 1 0.25 L, where L, is the .neasured Type A test leakage and L, is the superimposed leak;
- 2) Has a duration sufficient to establish accurately the change in leakage rate between the Type A test and the supplemental test; and
- 3) Requires that the rate at which gas is injected into the con-tainment or bled from the containment during the supplemental test is between 0.75 L, and 1.25 L,.
- d. Type B and C tests shall be conducted with gas at a pressure not less than P,, 44.4 psig, at intervals no greater than 24 months except for tests involving:
- 1) Air locks, and
- 2) Purge supply and exhaust isolation valves with resilient material seals.
- e. Air locks shall be tested and demonstrated OPERABLE by the require-ments of Specification 4.6.1.3;
- f. 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; and
- g. The provisions of Specification 4.0.2 are not applicable.
BRAIDWOOD - UNITS 1 & 2 3/4 6-3
l CONTAINMENT SYSTEMS AUG 1 9 1986 CONTAINMENT AIR LOCKS LIMITING CONDITION FOR OPERATION 3.6.1.3 Each containment air lock shall be OPERABLE with: l
- a. Both doors closed except when the air lock is being used for normal transit entry and exits through the containment, then at least one air lock door shall be closed, and
- b. An overall air lock leakage rate of less than or equal to 0.05 L a at P,, 44.4 psig.
APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:
- a. With one containment air lock door inoperable:
- 1. Maintain at least the OPERABLE air lock door closed *and either restore the inoperable air lock door to OPERABLE status within 24 hours or lock the OPERABLE air lock door closed;
- 2. Operation may then continue until performance of the next required overall air lock leakage test provided that the OPERABLE air lock door is verified to be locked closed at least once per 31 days;
- 3. Otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours; and
- 4. The provisions of Specification 3.0.4 are not applicable.
- b. With the containment air lock inoperable, except as the result of an inoperable air lock door, maintain at least one air lock door closed; restore the inoperable air lock to OPERABLE status within 24 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
{ BRAIDWOOD - UNITS 1 & 2 3/4 6-4
1 CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS l 4.6.1.3 Each containment air lock shall be demonstrated OPERABLE:
- a. Within 72 hours following each closing, except when the air lock is being used for multiple entries, then at least once per 72 hours, by verifying seal leakage is less than 0.01 L when the volume between the door seals is pressurized to greater tRan or equal to 10 psig, for at least 15 minutes;
- b. By conducting overall air lock leakage tests at not less than P,,
44.4 psig, and verifying the overall air lock leakage rate is within its limit:
- 1) At least once per 6 months,* and
- 2) Prior to establishing CONTAINMENT INTEGRITY when maintenance has been performed on the air lock that could affect the air lock sealing capability.**
- c. At least once per 6 months by verifying that only one door in each air lock can be opened at a time.
*The provisions of Specification 4.0.2 are not applicable. **This represents an exemption to Appendix J of 10 CFR Part 50, Paragr'aph III D.2(b)(fi).
BRAIDWOOD - UNITS 1 & 2 3/4 6-5
l CONTAINMENT SYSTEMS i INTERNAL PRESSURE AUG 19 ilmi ! LIMITING CONDITION FOR OPERATION 3.6.1.4 Primary containment internal pressure shall be maintained between
-0.1 and +1.0 psig.
APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With the containment internal pressure outside of the limits above, restore the internal pressure to within the limits within 1 hour or be in'at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.6.1.4 The primary containment internal pressure shall be determined to be within the limits at least once per 12 hours. BRAIDWOOD - UNITS 1 & 2 3/4 6-6
CONTAINMENT SYSTEMS AIR TEMPERATURE . LIMITING CONDITION FOR OPERATION 3.6.1.5 Primary containment average air temperature shall not exceed 120*F. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With the containment average air temperature greater than 120*F, reduce the average air temperature to within the limit within 8 hours, or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.6.1.5 The primary containment average air temperature shall be the arithmetical average of the temperatures of the running fans at the following. locations and shall be determined at least once per 24 hours: Location A RCFC Dry Bulb Inlet Temperature B RCFC Dry Bulb Inlet Temperature l C RCFC Dry Bulb Inlet Temperature D RCFC Dry Bulb Inlet Temperature. BRAIDWOOD - UNITS 1 & 2 3/4 6-7
FINAL HAFT CONTAINMENT SYSTEMS AUG 19 HE CONTAINMENT VESSEL STRUCTURAL INTE'GRITY LIMITING CONDITION FOR OPERATION 3.6.1.6 The structural integrity of the containment vessel shall be maintained at a level consistent with the acceptance criteria in Specifications 4.6.1.6.1, 4.6.1.6.2, and 4.6.1.6.3. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:
- a. With more than one tendon with an observed lift-off force between the predicted lower limit and 90% of the predicted lower limit or with one tendon below 90% of the predicted lower limit, restore the tendon (s) to the required level of integrity within 15 days and perform an engineering evaluation of the containment and provide a Special Report to the Commission within 30 days in accordance with Specificaticn 6.9.2 or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- b. With any abnormal degradation of the structural integrity other than ACTION a. at a level below the acceptance criteria of Specifications 4.6.1.6.1, 4.6.1.6.2, and 4.6.1.6.3, restore the containment vessel to the required level of integrity within 72 hours and perform an engi-neering evaluation of the containment and provide a Special Report to the Commission within 15 days in accordance with Specification 6.9.2 or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE REQUIREMENTS 4.6.1.6.1 Containment Vessel Tendons. The containment vessel tendons' struc-tural integrity shall be demonstrated at the end of 1, 3, and 5 years following the initial containment vessel structural integrity test and at 5 year intervals thereafter. The tendons' structural integrity shall be demonstrated by:
- a. Determining that a random but representative sample of at least 19
, tendons (5 dome, 6 vertical, and 8 hoop) each have an observed lift-l off force within predicted limits for each. For each subsequent inspection one tendon from each group may be kept unchanged to develop a history and to correlate the observed data. If the observed lift-off force of any one tendon in the original sample population lies between the predicted lower limit and 90% of the predicted lower limit, two tendons, one on each side of this tendon should be checked for their lift-off forces. If both of these adjacent tendons are found to be within their predicted limits, all three tendons should be restored to the required level of integrity. This single deficiency may be considered unique and acceptable. Unless there is ab' normal degradation of the containment vessel during the first three inspec-tions, the sample population for subsequent inspections shall include at least 10 tendons (3 dome, 3 vertical, and 4 hoop); BRAIDWOOD - UNITS 1 & 2 3/4 6-8 1
FINAL HAFT CONTAINMENT SYSTEMS AUG 19 m SURVEILLANCE REQUIREMENTS (Continued)
- b. Performing tendon detensioning, inspections, and material tests on a previously stressed tendon from each group (dome, vertical, 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 or strand that:
- 1) The tendon wires or strands are free of corrosion, cracks, and damage,
- 2) There are no changes in the presence or physical appei.rance 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 or strand samples (one from each end and one at mid-length) cut from each removed wire or strand. Failure of any one of the wire or strand samples to meet the minimum tensile strength test is evidence.of abnormal degradation of the containment vessel structure.
- c. Performing tendon retensioning of those tendons detensioned for inspection to their observed lift-off force with a tolerance limit of +6%. During retensioning of these tendons, the changes in load and elongation should be measured simultaneously at a minimum of three approximately equally spaced levels of force between zero and the seating 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 adjusted to account for elastic losses exceed the average minimum design value given below:
Dome 143 ksi Vertical 144 ksi Hoop 140 ksi
- e. Verifying the OPERABILITY of the sheathing filler grease by assuring:
- 1) No voids in excess of 5% of the net duct volume,
- 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. .
1 BRAIDWOOD - UNITS 1 & 2 3/4 6-9
FINAL HAFT CONTAINMENT SYSTEMS AUG 191986 SURVEILLANCE REQUIREMENTS (Continu'ed) 4.6.1.6.2 End Anchorages and Adjacent Concrete Surfaces. The structural integrity of the end anchorages of all tendons inspected pursuant to Specifi-cation 4.6.1.6.1 and the adjacent concrete surfaces shall be demonstrated by determining through inspection that no apparent changes have occurred in the visual appearance of the end anchorage or the concrete crack patterns adjacent to the end anchorages since last inspected. Inspections of the concrete shall be performed during the containment vessel tendon tests (reference Specifica-tion 4.6.1.6.1). 4.6.1.6.3 Containment Vessel Surfaces. The structural integrity of the exposed accessible interior and exterior surfaces of the containment vessel, including the liner plate, shall be determined during the shutdown for each Type A containment leakage rate test (reference Specification 4.6.1.2) by a visual inspection of these surfaces. This inspection shall be performed prior to the Type A containment leakage rate test to verify no apparent changes in appearance or other abnormal degradation. 9 BRAIDWOOD - UNITS 1 & 2 3/4 6-10
I CONTAINMENT SYSTEMS AUG 191986 CONTAINMENT PURGE VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.6.1.7 Each containment purge supply and exhaust isolation valves shall be OPERABLE and:
- a. Each 48-inch containment shutdown purge supply and exhaust isolation valve shall be closed and power removed, and
- b. The 8-inch containment purge supply and exhaust isolation valve (s) may be open for up to 1000 hours during a calendar year provided no more than one line is open at one time.
APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:
- a. With a 48-inch containment purge supply and/or exhaust isolation valve open and/or powered, close'and remove power to isolate the penetration (s) within 4 hours, otherwise be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- b. With the 8-inch containment purge supply and/or exhaust isolation valve (s) open for more than 1000 hours during a calendar year, close the open 8-inch valve (s) or isolate the penetration (s) within 4 hours, otherwise be in at least HOT STANDBY within the next 6 hours, and in COLD SHUTDOWN within the following 30 hours,
- c. With a containment purge supply and/or exhaust isolation valve (s) having a measured leakage rate in excess of the limits of Specifi-cations 4.6.1.7.3 and/or 4.6.1.7.4, restore the inoperable valve (s) to OPERABLE status within 24 hours, otherwise be in at least HOT STANDBY within the next 6 hours, and in COLD SHUTDOWN within the following 30 hours.
I BRAIDWOOD - UNITS 1 & 2 3/4 6-11
FINAL DRAFT CONTAINMENT SYSTEMS AUG 191986 SURVEILLANCE REQUIREMENTS , 1 4.6.1.7.1 Each 48-inch containment purge supply and exhaust isolation valve (s) ' shall be verified closed and power removed at least once per 31 days. 4.6.1.7.2 The cumulative time that'all 8-inch containment purge supply and/or I exhaust isolation valves have been open during a calendar year shall be determined at least once per 7 days. 4.6.1.7.3 At least once per 6 months on a STAGGERED TEST BASIS, the inboard and outboard valves with resilient material seals in each 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 pressurized to at least P,, 44.4 psig. 4.6.1.7.4 At least once per 3 months, each 8-inch containment purge supply and exhaust isolation valve with resilient material seals shall be demonstrated OPERABLE by verifying that the measured leakage rate is less than 0.01 L, when pressurized to at least P,, 44.4 psig. BRAIDh000 - UNITS 1 & 2 3/4 6-12
)
l CONTAINMENT SYSTEMS FINAL DRAFT Aug i g jggg 3/4.6.2 DEPRESSURIZATION AND C00 LING SYST$MS CONTAINMENT SPRAY SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.1 Two independent Containment Spray Systems shall be OPERABLE with each Spray System capable of taking suction from the RWST and transferring suction to the containment sump. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With one Containment Spray System inoperable, restore the inoperable Spray System to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours; restore the inoperable Spray System to OPERABLE status within the next 48 hours or be in COLD SHUTOOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.6.2.1 Each Containment Spray System shall be demonstrated OPERABLE:
- a. At least once per 31 days by verifying that each. valve (manual, power-cperated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position;
- b. By verifying, that on recirculation flow, each pump develops a discharge pressure of greater than or equal to 265 psig when tested pursuant to Specification 4.0.5;
- c. At least once per 18 months during shutdown, by:
- 1) Verifying that each automatic valve in the flow path actuates to its correct position on a Containment Spray Actuation test signal, and
- 2) Verifying that each spray pump starts automatically on a Containment Spray Actuation test signal.
- d. At least once per 5 years by performing an air or smoke flow test through each spray header and verifying each spray nozzle is unobstructed.
BRAIDWOOD - UNITS 1 & 2 3/4 6-13
CONTAINMENT SYSTEMS FINAL DRAFT AUG 191986 SPRAY ADDITIVE SYSTEM
- LIMITING CONDITION FOR OPERATION 3.6.2.2 The Spray Additive System shall be OPERABLE with: ,
- a. A spray additive tank containing a level of between 78.6% and 90.3%
of between 30% and 36% by weight NaOH solution, and
- b. Two spray additive eductors each capable of adding NaOH solution from the spray additive tank to a Containment Spray System pump flow.
APPLICABILITY: H0 DES 1, 2, 3, and 4. ACTION: With the Spray Additive System inoperable, restore the system to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours; restore the Spray Additive System to OPERABLE status within the next 48 hours or be in COLD SHUTDOWN within the following 30 hours. I SURVEILLANCE REQUIREMENTS 4.6.2.2 The Spray Additive System shall be demonstrated OPERABLL':
- 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 level in the tank, and
- 2) Verifying the concentration of the NaOH solution by chemical analysis.
- c. At least once per 18 months during shutdown, by verifying that each automatic valve in the flow path actuates to its correct position on a Containment Spray Actuation test signal; and
- d. At least once per 5 years by verifying each water flow rate equivalent to 55(+5,-0) gallons per minute for 30% NaOH from the eductor test connections in the Spray Additive System:
+6
- 1) CS26A 68 -0 gpm (Train A), and
+6
- 2) CS268 68 -0 gpm (Train B).
BRAIDWOOD - UNITS 1 & 2 3/4 6-14
FINAL HAFT CONTAINMENT COOLING SYSTEM LIMITING CONDITIONS FOR OPERATION 3.6.2.3 Two electrically independent systems af containment cooling fans shall be OPERABLE with two fans to each system. MODES 1, 2, 3, and 4. APPLICABILITY: ACTION:
- a. With one system of the above required containment cooling fans inoperable and both Containment Spray Systems OPERABLE, restore the inoperable system of cooling fans to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- b. With two systems of the above required containment cooling fans inoperable and both Containment Spray Systems OPERABLE, restore at least one system of cooling fans to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. Restore both above required
. systems of cooling fans to OPERABLE status within 7 days of initial loss or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- c. With one system of the above required containment cooling fans inoperable and one Containment Spray System inoperable, restore the inoperable Spray System to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. Restore the inoperable system of containment cooling fans to OPERABLE status within 7 days of initial loss or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE REQUIREMENTS 4 4.6.2.3 Each system of containment cooling fans shall be demonstrated OPERABLE:
- a. At least once per 31 days by:
- 1) Starting each fan system in slow speed from the control room, and verifying that each fan system operates for at least 15 minutes, and
- 2) Verifying an essential service water flow rate of greater than or equal to 2660 gpm to each cooler,
- b. At least once per 18 months by verifying that each fan system starts l l
automatically on a Safety Injection test signal. - l BRAIDWOOD - UNITS 1 & 2 3/4 6-15 1 1
FINAL DRAFT CONTAINMENT SYSTEMS 3/4.6.3 CONTAINMENT ISOLATION VAL'VES LIMITING CONDITION FOR OPERATION 1 3.6.3 The containment isolation valves specified in Table 3.6-1 shall be 1 OPERABLE with isolation times as shown in Table 3.6-1. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:
- a. With one or more of the isolation valve (s) specified in Table 3.6-1 inoperable, maintain at least one isolation valve OPERABLE in each affected penetration that is open and within 4 hours:
- 1. Restore the inoperable valve (s) to OPERABLE status, or
- 2. Isolate each affected penetration by use of at.least one deactivated automatic valve secured in the isolation position, or
- 3. Isolate each affected penetration by use of at least one closed manual valve or blind flange.
Otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- b. The provisions of Specification 3.0.4 are not applicable provided that within 4 hours the affected penetration is isolated in accordance with ACTION a.2 or a.3 above, and provided that the associated system, if applicable, is declared inoperable and the appropriate ACTION statements for that system are taken.
SURVEILLANCE REQUIREMENTS 4.6.3.1 The isolation valves specified in Table 3.6-1 shall be demonstrated OPERABLE prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or its associated actuator, control or power circuit by performance of a cycling test, and verification of isolation time. BRAIDWOOD - UNITS 1 & 2 3/4 6-16
FINAL DRAFT SURVEILLANCE RECUIREMENTS (Continued) IE E 4.6.3.2 Each isolation valve specified in Table 3.6-1 shall be demonstrated OPERABLE during the COLD SHUTOOWN or REFUELING MODE at least once per 18 months by:
- a. Verifying that on a Phase "A" Isolation test signal, each Phase "A" isolation valve actuates to its isolation position;
- b. Verifying that on a Phase "B" Isolation test signal, each Phase "B" isolation valve actuates to its isolation position; and
- c. Verifying that on a Containment Vent Isolation test signal, each purge ar.d exhaust isolation valve actuates to its isolation position.
4.6.3.3 The isolation time of each power operated or automatic valve of Table 3.6-1 shall be determined to be within its limit when tested pursuant to Specification 4.0.5. l BRAIDWOOD - UNITS 1 & 2 3/4 6-17
TA8LE 3.6-1 ggg i g jggg CONTAINME'NT ISOLATION VALVES MAXIMUM. PENETRATION VALVE NO. FUNCTION ISOLATION TIME (SEC)
- 1. Phase "A" Isolation ,
28 CV8100 RCP Seal Water Return 10 28 CV8112 RCP Seal Water Return 10 41 CV8152 RCS Letdown 10 41 CV8160 RCS Letdown 10 5 W9020A Chilled Water 50 5 W9056A Chilled Water 50 6 W9006A Chilled Water 50 8 . W90208 Chilled Water 50 8 W9056B Chilled Water 50 10 W9006B Chilled Water 50 22 CC9437B* Excess Ltdn HX Return 10 48 CC9437A* Excess Ltdn HX Supply 10 34 FP010* Fire Protection 12 39 IA065 Instrument Air 15 39 IA066 Instrument Air 15 13 9G079 Hydrogen Recombiner 60 13 9G080 Hydrogen Recombinor 60 13 9G082 Hydrogen Recombiner 60 13 9G084 Hydrogen Recombiner 60 23 9G081 Hydrogen Recombiner 60 23 9G085 Hydrogen Recombiner 60 69 9G057A Hydrogen Recombiner 60 69 9G083 Hydrogen Recombiner 60 56 SA032 Service Air 4.5 56 SA033 Service Air 4.5 80 SD002C Steam Generator Blowdown 7.5 80 SD005B Steam Generator Blowdown 3.0 81 SD0020 Steam Generator Blowdown 7.5 82 SD002A Steam Generator Blowdown 7.5 82 50005A Steam Generator Blowdown 3.0 83 500028 Steam Generator Blowdown 7.5 , 88 SD002E Steam Generator Blowdown 7.5 l 88 SD005C Steam Generator Blowdown 3.0 ! 89 SD002F Steam Generator Blowdown 7.5 l 90 S0002G Steam Generator Blowdown 7.5 90 SD005D Steam Generator Blowdown 3.0 91 SD002H Steam Generator Blowdown 7. 5 BRAIDWOOD - UNITS 1 & 2 3/4 6-18 l l
FINAL DRAFT TABLE 3.6-1 (Continued) AUG i s 1986 CONTAINMENT ISOLATION VALVES MAXIMUM PENETRATION VALVE NO. FUNCTION ISOLATION TIME (SEC)
- 1. Phase "A" Isolation (Continued) .
52 PR001A Process Radiation 4.5 52 PR001B Process Radiation 4.5 52 PR066 Process Radiation 5.0 12 PS228A Hydrogen Monitor N.A.** 12 PS229A Hydrogen Monitor N.A.** 12 PS230A Hydrogen Monitor N.A.** 31 PS228B Hydrogen Monitor N.A.** 31 PS2298 Hydrogen Monitor N.A.** 31 PS230B Hydrogen Monitor N.A.** 70 PS9354A Primary Process Sampling 10 70 PS93548 Primary Process Sampling 10 70 PS9355A Primary Process Sampling 10 70 PS93558 Primary Process Sampling 10 70 PS9356A Primary Process Sampling 10 70 PS93568 Primary Process Sampling 10 70 PS9357A Primary Process Sampling 10 70 PS9357B Primary Process Sampling 10 11 RE9170 Reactor Bldg Equip Drains 10 11 RE1003 Reactor Bldg Equip Drains 10 65 RE9157 Reactor Bldg Equip Drains 10 65 RE9159A Reactor Bldg Equip Drains 10 65 RE9159B Reactor Bldg Equip Drains 10 65 RE9160A Reactor Bldg Equip Drains 10 65 RE91608 Reactor Bldg Equip Drains 10 27 RY8025 PRT Nitrogen 10 27 RY8026 PRT Nitrogen 10 27 RY8033 PRT Nitrogen 10 44 RY8028 PRT Make-up 10 55 SI8964 Accumulator Fill 10 55 SI8880 Nitrogen Supply to Accumulator 10 55 SI8871 Accumulator Fill 10 55 SI8888 Hot Leg Safety Injection 10 47 RF026 Reactor Building Floor Drains 15 47 RF027 Reactor Building Floor Drains 15 BRAIDWOOD - UNITS 1 & 2 3/4 6-19
FINAI. DRAFT TABLE 3.6-1 (Continued) AUG 191986 ! CONTAINM5NTISOLATIONVALVES MAXIMUM PENETRATION VALVE NO. FUNCTION ISOLATION TIME (SEC)
- 2. Phase "B" Isolation 21 CC9414 RCP Mtr Brng Return 10 21 CC9416 RCP Mtr Brna Return 10 24 CC685 RCP Thermal Barrier Return 10 24 CC9438 RCP Thermal Barrier Return 10 25 CC9413A RCP Cooling Wtr Supply 10
- 3. Safety Injection 71 CV8105* CVCS Charging 10 -
71 CV8106* CVCS Charging 10 7 SX016B* Essential Service Water N.A. 9 SX027B* Essential Service Water N.A. 14 SX027A* Essential Service Water N.A. 15 SX016A* Essential Service Water N.A. 26 SI6801A* Cold Leg Safety Injection N.A. 26 SI8801B* Cold Leg Safety Injection N.A. 92 SI8811A* Containment Recirc. Sump N.A. 93 5I8811B* Containment Recire. Sump N.A.
- 4. Containment Ventilation Isolation 94 VQ003 Mini-Flow Purge Exhaust 5 94 VQ005A Mini-Flow Purge Exhaust '5 94 VQ00EB Mini-Flow Purge Exhaust 5 94 VQ005C Mini-Flow Purge Exhaust 5 95 VQ002A Purge Exhaust 5 95 VQOO2B Purge Exhaust 5 96 VQ004A Mini-Flow Purge Supply 5 96 VQ0048 Mini-Flow Purge Supply 5 97 VQ001A Purge Supply 5 97 VQ001B Purge Supply 5
- 5. Containment Spray Actuation '
1 CS007A Containment Spray 30 16 CS0078 Containment Spray 30
- 6. Main Steam Isolation 77 MS101D* Main Steam 6 78 MS101A* Main Steam 6' BRAIDWOOD - UNITS 1 & 2 3/4 6-20
FINAL HAFT TABLE 3.6-1 (Continued) CONTAINME'NT ISOLATION VALVES MAXIMUM PENETRATION VALVE NO. FUNCTION ISOLATION TIME (SEC)
- 6. Main Steam Isolation (Continued) .
85 MS101B* Main Steam 6 86 MS101C* Main Steam 6
- 7. Feedwater Isolation 76 FW009D* Main Feedwater 5 76 FWO43D* Main Feedwater 6 79 FWOO9A* Main Feedwater 5 79 FWO43A* Main Feedwater 6 84 FWOO9B* Main Feedwater -5 84 FWO438* Main Feedwater 6 87 FWOO9C* Main Feedwater 5 87 FWO43C* Main Feedwater 6 59 FWO35D* Main Feedwater 6 99 FWO390* Main Feedwater 6 100 FWO35A* Main Feedwater 6 -
100 FWO39A* Main Feedwater 6 101 FWO35B* Main Feedwater 6 101 FWO39B* Main Feedwater 6 102 FWO35C* Main Feedwater 6 102 FWO39C* Main Feedwater 6
- 8. Remote Manual 68 RH8701A*,# RH Suction N.A.
68 RH8701B",# RH Suction N.A. 75 RH8702A*,# RH Suction N.A. 75 RH87028*,# RH Suction N.A. 59 SI8881* Hot Leg Safety Injection N.A. 73 SI8824* Hot Leg Safety Injection N.A. 66 SI8825* Hot Leg RH Injection N.A. 60 SI8823* Cold Leg Safety Injection N.A. 50 SI8890A* Cold Leg RH Injection N.A. 51 SI8890B* Cold Leg RH Injection N.A. 26 SI8843* Cold Leg Safety Injection N.A. 33 CV8355A* RCP Seal Injection N.A. 33 CV83550* RCP Seal Injection N.A. 53 CV8355B* RCP Seal Injection N.A. 53 CV8355C* RCP Seal Injection N.A. BRAIDWOCD - UNITS 1 & 2 3/4 6-21
FINAL HAFT TABLE 3.6-1 (Continued) AUG 191986 CONTAINME'NT ISOLATION VALVES MAXIMUM PENETRATION VALVE NO. FUNCTION ISOLATION TIME (SEC)
- 8. Remote Manual (Continued) .
59 SI8802A* Hot Leg Safety Injection N.A. 73 SI88028* Hot Leg Safety Injection N.A. 60 SI8835" Hot Leg Safety Injection N.A. 50 SI8809A* RH Cold Leg Injection N.A. 51 SI8809B* RH Cold Leg Injection N.A. 66 SI8840* Hot Leg Safety Injection N.A. 100 AF013A* Feedwater N.A. 100 AF013E* Feedwater N.A. 101 AF0138* Feedwater N.A. 101 AF013F* Feedwater N.A. 102 AF013C* Feedwater N.A. 102 AF013G* Feedwater 3 N.A. 99 AF013D* Feedwater N.A. 99 AF013H* Feedwater N.A.
- 9. Manual 37 CV8346* RCS Loop Fill N.A.
I3 VQOl6 Instrument Penetration N.A. 13 VQ017 Instrument Penetration N.A. I3 VQ018 Instrument Penetration N.A. 1 13 VQ019 Instrument Penetration N.A. 15 RYO75 Instrument Penetration N.A. 30 WM190 Make-Up Demin N.A. 57 FC009 Spent Fuel Pool Cleaning N.A. 57 FC010 Spent Fuel Pool Cleaning N.A. 32 FC011 Spent Fuel Pool Cleaning N.A. 32 FC012 Spent Fuel Pool Cleaning N.A. 77 MS021D*,# Main Steam N.A. 78 MS021A*,# Main Steam N.A. 85 MS021B*,# Main Steam N.A. 86 MS021C*,# Main Steam N.A. AL PR002E# Process Radiation N.A. AL PR033A# Process Radiation N.A. AL PR0338# Process Radiation N.A. AL PR002F# Process Radiation N. A'. AL PR033C# Process Radiation N.A. AL PR033D# Process Radiation N.A. l l BRAIDWOOD - UNITS 1 & 2 3/4 6-22
~
FINAL HAFT l TABLE 3.6-1 (Continued) AUG 191986 CONTAINME'NT ISOLATION VALVES MAXIMUM PENETRATION VALVE NO. FUNCTION ISOLATION TIME (SEC)
- 9. Manual (Continued) ,
99 FWO15D*,# Feedwater N.A. 100 FW015A*,# Feedwater N.A. 101 FWO15B*,# Feedwater N.A. 102 FW015C*,# Feedwater N.A.
- 10. Check 28 CV8113 RCP Seal Water Return N.A.
37 CV8348* RCS Loop Fill N.A. 6 W9007A Chilled Water N.A. 10 . W9007B Chilled Water N.A. 21 CC9534 RCP Mtr Brng Return N.A. 24 CC9518 RCP Thermal Barrier Return N.A. 25 CC9486 RCP Cooling Wtr Supply N.A. 1 CS008A Containment Spray N.A. 16 CS008B Containment Spray N. A. 39 IA091 Instrument Air N.A. 30 WM191 Make-Up Demin N.A. 52 PR032 Process Radiation N.A. AL PR002G Process Radiation N.A. AL PR002H Process Radiation N.A. 12 PS231A Hydrogen Monitor N.A. 31 PS231B Hydrogen' Monitor N.A. 27 RY8047 PRT Nitrogen N.A. 44 RY8046 PRT Make-Up N.A. 26 SI8815* Safety Injection N.A. 50 SI8818A* Safety Injection N.A. 50 SI8818D* Safety Injection N.A. 51 SI88188* Safety Injection N.A. 51 SI8818C* Safety Injection N.A. 59 SI8905A* Safety Injection N.A. 59 SI88050* Safety Injection N.A. 60 SI8819A* Safety Injection N . A'. l 60 SI88198* Safety Injection N.A. BRAIDWOOD - UNITS 1 & 2 3/4 6-23
FINAL DRAFT TABLE 3.6-1 (Continued)
. AUG 191986 CONTAINMENT ISOLATION VALVES MAXIMUM PENETRATION VALVE NO. FUNCTION ISOLATION TIME (SEC)
- 10. Check (Continued) 60 SI8819C* Safety Injection N.A.
60 SI8819D* Safety Injection N.A. 66 SI8841A* Safety Injection N.A. 66 SI8841B* Safety Injection N.A. 73 SI8905B* Safety Injection N.A. 73 SI8905C* Safety Injection N.A. 55 SI8968* Safety Injection N.A. 34 FP345* Fire Protection N.A. 33 CV8368A* RCP Seal Injection N. A. 33 CV8368D* RCP Seal Injection N. A. 53 CV8368B" RCP Seal Injection N.A. 53 CV8368C* RCP Seal Injection N.A.
- 11. S/G Safeties /PORVs 77 MS013D* Main Steam N.A.
77 MS0140* Main Steam N.A. 77 MS015D* Main Steam N.A. 77 MS0160* Main Steam N.A. 77 MS017D* Main Steam N.A. 78 MS013A* Main Steam N.A. 78 MS014A" Main Steam N.A. 78 MS015A* Main Steam N.A. 78 MS016A* Main Steam N.A. 78 MS017A* Main Steam N.A. 85 M50138* Main Steam N.A. 85 MS014B* Main Steam N.A. 85 MS015B* Main Steam N.A. 85 MS0168* Main Steam N.A. 85 MS0178* Main Steam N.A. 86 MS013C* Main Steam N.A. 86 MS014C* Main Steam N.A. 86 MS015C* Main Steam N.A. I 86 MS016C* Main Steam N.A. 86 MS017C* Main Steam N.A. 77 MS0180* Main Steam 20 78 M5018A* Main Steam 20 85 MS018B* Main Steam 20 86 MS018C* Main Steam 20 "Not subject to Type C leakage tests. ,
** Proper valve operation will be demonstrated by verifying that the valve strokes to its required position. #May be opened on an intermittent basis under administrative control.
BRAIDWOOD - UNITS 1 & 2 3/4 6-24
FINAL DRAFT CONTAINMENT SYSTEMS AUG 191986 3/4.6.4 COMBUSTIBLE GAS CONTROL ' HYOR0 GEN MONITORS LIMITING CONDITION FOR OPERATION 3.6.4.1 Two independent containment hydrogen monitors shall be OPERABLE.* APPLICABILITY: MODES 1 and 2. ACTION:
- a. With one hydrogen monitor inoperable, restore the inoperable monitor i to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 6 hours.
- b. With both hydrogen monitors inoperable, restore at least one monitor to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours.
SURVEILLANCE REQUIREMENTS 4.6.4.1 Each hydrogen monitor shall be demonstrated OPERABLE by the performance of a CHANNEL CHECK and a check that the monitor is in standby mode at least once per 12 hours, an ANALOG CHANNEL OPERATIONAL TEST at least once per 31 days, and at least once per 92 days by per+orming a CHANNEL CA.LIBRATION using five gas vanples which shall cover the range from zero volume percent hydrogen (100% N2 ) to greater than 20 volume percent hydrogen, balance nitrogen. l
*The monitors must be in standby mode to meet the requirement in NUREG-0737, Item II.F.1.6. -
BRAIDWOOD - UNITS 1 & 2 3/4 6-25
CONTAINMENT SYSTEMS FINAL DRAFT AUG 191986 , ELECTRIC HYDROGEN RECOMBINERS - l LIMITING CONDITION FOR OPERATION 3.6.4.2 Two independent Hydrogen Recombiner Systems shall be OPERAB.LE. 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 STANO3Y within the next 6 hours. SURVEILLANCE REQUIREMENTS 4.6.4.2 Each Hydrogen Recombiner System shall be demonstrated OPERABLE:
- a. At least once per 6 months by verifying, during a Recombiner System functional test that the minimum heater sheath temperature increases to greater than or equal to 1200*F within 90 minutes. Upon reaching 1200*F, increase the temperature controller to maximum setting for 2 minutes and verify that the power is greater than or equal to 38 kW, and
- b. At least once per 18 months by:
- 1) Performing a CHANNEL CALIBRATION of all recombiner instrumen-tation and control circuits, l 2) Verifying through a visual examination that there is no evidence l
of abnormal conditions within the recombiners enclosure (i.e., loose wiring or structural connections, deposits of foreign materials,etc.),and
- 3) Verifying the integrity of all heater electrical circuits by performing a resistance to ground test following the above required functional test. The resistance to ground for any heater phase shall be greater than or equal to 10,000 ohms.
I i BRAIDWOOD - UNITS 1 & 2 3/4 6-26 l
3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE ggg y g ygy; SAFETY VALVES LIMITING CONDITION FOR OPERATION ; 3.7.1.1 All main steam line Code safety valves associated with each steam generator shall be OPERABLE with lift settings as specified in Table 3.7-2. APPLICABILITY: MODES 1, 2, and 3. - ACTION:
- a. With four reactor coolant loops and associated steam generators in operation and with one or more main steam line Code safety valves inoperable, operation in MODES 1, 2, and 3 may proceed provided, that within 4 hours, either the inoperable valve is restored to OPERABLE status or the Power Range Neutron Flux High Trip Setpoint is reduced per Table 3.7-1; otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- 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 l l 1 BRAIDWOOD - UNITS 1 & 2 3/4 7-1 i
TABLE 3.7-1 FINAL HAFT AUG 191986 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING FOUR LOOP OPEP.ATION . MAXIMUM NUMBER OF INOFERABLE MAXIMUM ALLOWABLE POWER RANGE SAFETY VALVES ON ANY NEUTRON FLUX HIGH SETPOINT OPERATING STEAM GENERATOR (PERCENT OF RATED THERMAL POWER) 1 87 2 65 3 43 e 9 BRAIDWOOD - UNITS 1 & 2 3/4 7-2
FINAL DRAFT TABLE 3.7-2 gg g STEAM LINE SAFETY VALVES PER LOOP VALVE NUMBER LIFT SETTING (*1%)* ORIFICE SIZE M5013(A-D) 1235 psig 16 in2 MS014(A-D) 1220 psig 16 in2 MS015(A-D) 1205 psig 16 in2 MS016(A-D) 1190 psig 16 ins MS017(A-D) 1175 psig 16 in2
*The lift setting pressure shall correspond to anbient conditions of the valve at nominal operating temperature and pressure.
e BRAIDWOOD - UNITS 1 & 2 3/4 7-3
PLANT SYSTEMS AUXILIARY FEEDWATER SYSTEM AUS is rges LIMITING CONDITION FOR OPERATION 3.7.1.2 At least two independent steam generator auxiliary feedwater pumps and associated flow paths shall be OPERABLE with:
- a. One motor-driven auxiliary feedwater pump capable of being powered from an ESF Bus, and
- b. One direct-driven diesel auxiliary feedwater pump capable of being powered from a direct-drive diesel engine and an OPERABLE Diesel Fuel Supply System consisting of a day tank containing a minimum level of 71% (420 gallons) of fuel.
APPLICABILITY: MODES 1, 2, and 3. ACTION:
- a. With one auxiliary feedwater pump inoperable, restore the required auxiliary feedwater pumps to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
- b. With both auxiliary feedwater pumps inoperable, be in at least HOT STANDBY within 6 hours and in HOT SHUTDOWN within the following 6 hours.
SURVEILLANCE REQUIREMENTS 4.7.1.2.1 Each auxiliary feedwater pump shall be demonstrated OPERABLE:
- a. At least once per 31 days on a STAGGERED TEST BASIS by:
- 1) Verifying that the pump develops a differential pressure of greater than or equal to 1825 psid at a flow of greater than or equal to 85 gpm on the recirculation flow when tested pursuant to Specification 4.0.5; i
j BRAIDWOOD - UNITS 1 & 2 3/4 7-4 l t l
1 1 PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) AU610 l000
- 2) Verifying by flow or position check that each valve (manual, power-operated, or automatic) valve in the flow path that is not locked, sealed, or otherwise secured in position is in its cor-rect position.
- b. At least once per 18 months during shutdown by:
- 1) Verifying that each automatic valve in the flow path actuates to its correct position upon receipt of an Auxiliary Feedwater Actuation test signal, and
- 2) Verifying that the motor-driven pump and the direct-driven diesel pump start automatically upon receipt of each of the following test signals:
a) SI or b) Steam Ger orator Water Level Low-Low from one steam generatoi, or c) Undervoltage on Reactor Coolant Pump 6.9 kV Buses (2/4), or d) ESF Bus 141 for Unit 1 (Bus 241 for Unit 2) Undervoltage (motor-driven pump only). 4.7.1.2.2 An auxiliary feedwater flow path to each steam generator shall be demonstrated OPERABLE following each COLD SHUTDOWN of greater than 30 days prior to entering MODE 2 by verifying normal flow to each steam generator. 4.7.1.2.3 The auxiliary feedwater pump diesel shall be demonstrated OPERABLE:
- a. At least once per 31 days by verifying the fuel level in its day tank;
- b. At least once per 92 days by verifying that a drain sample of diesel fuel from its day tank, obtained in accordance with ASTM-04057-1981 is ,
within the acceptable limits specified in Table 1 of ASTM-0975-1977 when checked for viscosity, water, and sediment; and
- c. At least once per 18 months, during shutdown, by subjecting the l diesel to an inspection in accordance with its manufacturer's recommendations for this class of service.
I BRAIDWOOD - UNITS 1 & 2 3/4 7-5
a PLANT SYSTEMS u aE H lb av s CONDENSATE STORAGE TANK . MS2ffor: LIMITING CONDITION FOR OPERATION l 3.7.1.3 The condensate storage tank (CST) shall be OPERABLE with a contained water level of at least 40%. APPLICABILITY: MODES 1, 2, and 3. ACTION: With the CST inoperable, within 4 hours either:
- a. Restore the CST to OPERABLE status or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours, or
- b. Demonstrate the OPERABILITY of the Essential Service Water System as a backup supply to the auxiliary feedwater pumps and restore the CST to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
SURVEILLANCE REQUIREMENTS 4.7.1.3.1 The CST shall be demonstrated OPERABLE at least once per 12 hours by verifying the contained water level is within its limits when the tank is the supply source for the auxiliary feedwater pumps. 4.7.1.3.2 The Essential Service Water System shall be demonstrated OPERABLE at least once per 12 hours by performing the surveillance specified in Specifica-tion 4.7.4a. whenever the Essential Service Water System is the supply source for the auxiliary feedwater pumps. BRAIDWOOD - UNITS 1 & 2 3/4 7-6 \
PLANT SYSTEMS FINAL DRAFT l SPECIFIC ACTIVITY . AUG is q LIMITING CONDITION FOR OPERATION i 3.7.1.4 The specific activity of the Secondary Coolant System shall be less than or equal to 0.1 microcurie / gram DOSE EQUIVALEilT I-131. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With the specific activity of the Secondary Coolant System greater than 0.1 microcurie / gram DOSE EQUIVALENT I-131 be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.7.1.4 The specific activity of the Secondary Coolant System shall be determined to be within the limit by performance of the sampling and analysis program of Table 4.7-1. , BRAIDWOOD - UNITS 1 & 2 3/4 7-7
FINAL MAFT TABLE 4.7-1 l AUG191995 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY i SAMPLE AND ANALYSIS PROGRAM TYPE OF MEASUREMENT SAMPLE AND ANALYSIS AND ANALYSIS FREQUENCY.
- 1. Gross Radioactivity At least once per 72 hours.
Determination *
- 2. Isotopic Analysis for DOSE a) Once per 31 days, when-EQUIVALENT I-131 Concentration ever the gross radio-activity determination indicates concentrations greater than 10% of the allowable limit for radioiodines.
b) Once per 6 months, when-ever the gross radio-activity determination indicates concentrations less than or equal to 10% of the allowable limit for radiofodines.
*A gross radioactivity analysis shall consist of the quantitative measurement of the total specific activity of the secondary coolant except for radio-nuclides with half-lives less than 10 minutes. Determination of the contri-butors to the gross specific activity shall be based upon those energy peaks identifiable with a 95% confidence level.
1 BRAIDWOOD - UNITS 1 & 2 3/4 7-8
FINAL HAFT PLANT SYSTEMS MAIN STEAM LINE ISOLATION VALVES AUG 191986 LIMITING CONDITION FOR OPERATION 3.7.1.5 Each main steam line isolation valve (MSIV) shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3. ACTION: - MODE 1: With one MSIV inoperable but open, POWER OPERATION may continue provided the inoperable valve is restored to OPERABLE status within 4 hours; otherwise be in HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours. MODES 2 and 3: - With one MSIV inoperable, subsequent operation in MODE 2 or 3 may proceed provided the isolation valve is maintained closed. The provisions of Specification 3.0.4 are not applicable. Otherwise, be in HOT STANDBY within the next 6. hours and in HOT SHUTOOWN within the following 6 hours. SURVEILLANCE REQUIREMENTS 4.7.1.5 Each MSIV shall be demonstrated OPERABLE by verifying full closure within 5 seconds when tested pursuant to Specification 4.0.5. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3. BRAIDWOOD - UNITS 1 & 2 3/4 7-9
PLANT SYSTEMS FINAL MAFT 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION AUG 191985 LIMITING CONDITION FOR OPERATION j 3.7.2 The temperatures of both the reactor and secondary coolants 1n the , steam generator shall be greater than 70 F when the pressure of either coolant in the steam generator is greater than 200 psig. APPLICABILITY: At all times. ACTION: With the requirements of the above specification not satisfied:
- a. Reduce the steam generator pressure of the applicable side to less than or equal to 200 psig within 30 minutes, and
- b. Perform an engineering evaluation to determine the effect of the overpressurization on the structural integrity of the steam generator.
Determine that the steam generator remains acceptable for continued operation prior to increasing its temperatures above 200*F. SURVEILLANCE REQUIREMENTS 4.7.2 The pre.;sure in each side of the steam generator shall be determined to be less than 200 psig at least once per hour when the temperature of either the reactor or secondary coolant is less than 70*F. I 1 l BRAIDWOOD - UNITS 1 & 2 3/4 7-10 i t
FINAL HAFT PLANT SYSTEMS 3/4.7.3 COMPONENT COOLING WATER SYSTEM IO E LIMITING CONDITION FOR OPERATION 3.7.3 The Component Cooling Water System shall be OPERABLE with:
- a. Two safety loops serving the RH pumps and RH heat exchangers.
- b. Two component cooling water pumps powered from 4 kV Busses'141 and 142 for Unit 1 (Busses 241 and 242 for Unit 2), and
- c. Two component cooling water heat exchangers.
APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:
- a. With only one safety loop OPERABLE, restore at least two loops to t OPERABLE status within 72 hours or be in at least HOT STANDBY within l the next 6 hours and in COLD SHUTDOWN within the following 30 hours,
- b. With only one component cooling water pump OPERABLE, restore at least two pumps to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the follow-ing 30 hours.
- c. With only one heat exchanger OPERABLE, restore at lea'st two heat exchangers to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within'the fol-lowing 30 hours.
SURVEILLANCE REQUIREMENTS 4.7.3.1 At least two component cooling water loops shall be demonstrated OPERABLE 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. 4.7.3.2 At least two component cooling water pumps shall be demonstrated OPERABLE by performing the following:
- a. The component cooling water pumps shall be operated each month. Per-formance will be acceptable if the pump starts upon actuation, oper-ates for at least 4 hours, and satisfies the cooling requirements for the routine operation of the component cooling water system, and
- b. Verifying at least once per 18 months during shutdown that each component cooling water pump starts automatically on a SI test signal.
This will include a test of the common component cooling water pump while powered from 4 kV Busses 141 and 142 for Unit 1 (Busses 241 and 242 for Unit 2). 4.7.3.3 At least two component cooling water heat exchangers shall be verified OPERABLE at least once per 31 days by:
- a. Verifying that each component cooling water heat exchanger 1-nlet and outlet valve is OPERABLE, and
- b. Verifying the Essential Service Water is available to each component cooling water heat exchanger.
BRAIDWOOD - UNITS 1 & 2 3/4 7-11
PLANT SYSTEMS FINAL DRAFT - 3/4.7.4 ESSENTIAL SERVICE WATER SYSTEM AUG 191986 LIMITING CONDITION FOR OPERATION 3.7.4 At least two independent Essential Service Water Systems shal.1 be OPERABLE. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:
- a. With only one Essential Service Water System OPERABLE, restore at least two Essential Service Water Systems to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE REQUIREMENTS 4.7.4 At least two Essential Service Water Systems 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. OSX165A and OSX165B shall be verified open and power removed from the valve operators at least once per 31 days.
- c. 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-related portion of the system actuates to its correct position on a Safety Injection test signal, and
- 2) Each Essential Service Water System pump starts automatically on a Safety Injection test signal.
BRAIDWOOD - UNITS 1 & 2 3/4 7-12
PLANT SYSTEMS FINAL BAFT
'M 191066 3/4.7.5 ULTIMATE HEAT SINK (Essential Service Cooling Pond)
LIMITING CONDITION FOR OPERATION 3.7.5 The essential service cooling pond (ESCP) shall be OPERABLE with: a. A minimum water level at or above elevation 590 ft. Mean S,ea Level, USGS datum, and
- b. An essential service water pump discharge water temperature of less than or equal to 98 F.
APPLICABILITY: MODES 1, 2, 3, and 4 ACTION: With the requirements of the above specification not satisfied, be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS . 4.7.5.1. The ESCP shall be determined OPERABLE at least once per 24 hours by verifying the average water temperature and water level to be within their limits. 4.7.5.2 The ESCP shall be determined OPERABLE at least once per 18 months by the performance of a hydrographic survey to verify:
- a. That the ESCP bottom elevation is less than or equal to 584 feet.
- b. That the ESCP slopes exhibit no excess degradation.
BRAIDWOOD - UNITS 1 & 2 3/4 7-13
FINAL DRAFT ' P'_ ANT SYSTEMS AUG 191966 3/4.7.6 CONTROL ROOM VENTILATION. SYSTEM LIMITING CONDITION FOR OPERATION 3.7.6 Two independent Control Room Ventilation Systems shall be OPERABLE. APPLICABILITY: All MODES. ACTION: MODES 1, 2, 3 and.4: With one Control Room Ventilation System inoperable, restore the inoperable system to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. MODES 5 and 6:
- a. With one Control Room Ventilation System inoperable, restore the inoperable system to OPERABLE status within 7 days or initiate and maintain operation of the remaining OPERABLE Control Room Ventilation System in the makeup mode..
- b. With both Control Room Ventilation Systems inoperable, or with the OPERABLE Control Room Ventilation System, required to be in the makeup mode by ACTION a. not capable of being powered by an OPERABLE emer-gency power source, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.
SURVEILLANCE REQUIREMENTS 4.7.6 Each Control Room Ventilation System shall be demonstrated OPERABLE:
- a. At least once per 12 hours by verifying that the control room air temperature is less than or equal to 90'F;
- b. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the Emergency Makeup System HEPA filters and charcoal adsorbers and verifying that the system operates i for at least 10 continuous hours with the heaters operating; and flow through the recirculation charcoal adsorber for 15 minutes.
- c. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communi-cating with the Emergency Makeup System filter plenum by:
- 1) Verifying that the cleanup system satisfies the in plac,e penetra-tion testing acceptance criteria of less than 0.05% 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 Emergency Makeup System; BRAIDWOOD - UNITS 1 & 2 3/4 7-14
FINAL MAFT PLANT SYSTEMS E 191966 SURVEILLANCE REQUIREMENTS (Continued)
- 2) Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample from the Emergency i Makeup System 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 thaa 0.175% when tested at a tempera-ture of 30*C and a relative humidity of 70%; and
- 3) Verifying a system flow rate of 6000 cfm + 10% for the Emergency Makeup System and 49,500 cfm 110% for the Recirculation System when tested in accordance with ANSI N510-1980.
- d. After every 720 hours of Emergency Makeup System operation by verify-ing 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 0.175% 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 6.0 inches Water Gauge while operating the Emergency Makeup System at a flow rate of 6000 cfm i 10%,
- 2) Verifying that on a Safety Injection or High Radiation-Control Room Outside Air Intake test signal, the system automatically switches into a makeup mode of control room ventilation with flow through the Emergency Makeup System HEPA filters and charcoal adsorber banks and the recirculation charcoal adsorber;
- 3) Verifying that the Emergency Makeup System maintains the control room at a positive nominal pressure of greater than or equal to 1/8 inch Water Gauge relative to ambient pressure in areas adjacent to the control room area when operating an Emergency Makeup System at a flowrate of 6,000 cfm 110% and the recircula-tion charcoal adsorber at a flowrate of 49,500 cfm i 10%.
- 4) Verifying that the heaters dissipate 27.2
- 2.7 kW when tested in accordance with ANSI N510-1980.
- 5) Verifying that the Emergency Makeup System maintains the Upper Cable Spreading Area at a positive nominal pressure of greater than or equal to 0.02 inches Water Gauge relative to the ambient pressure in areas adjacent to the upper cable spreading area (except for adjacent control room areas pressurized as specified above) when operating an Emergency Makeup System at a flow rate of 6,000 cfm i 10% and the recirculation charcoal adsorber at a flowrate of 49,500 cfm i 10%. '
BRAIDWOOD - UNITS 1 & 2 3/4 7-15 i
FINAL DRAFT PLANT SYSTEMS AUG 191985 SURVEILLANCE REQUIREMENTS (Continued)
- f. After each complete or partial replacement of a HEPA filter bank, by verifying that the cleanup system satisfies the in place penetration j testing acceptance criteria of less than 0.05% in accordance with l ANSI N510-1980 for a D0P test aerosol while operating the, Emergency ,
Makeup System at a flow rate of 6000 cfm + 10%; and 1
- g. After each complete or partial replacement of a charcoal adsorber bank in the Emergency Makeup System by verifying that the cleanup system satisfies the in place penetration testing acceptance criteria of less than 0.05% in accordance with ANSI N510-1980 for a halogenated hydro-carbon refrigerant test gas while operating the system at a flow rate of 6000 cfm i 10%.
- h. At least once per 18 months or (1) after any structural maintenance on the charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the recirculation charcoal adsorber by:
(1) Verifying that the recirculation charcoal adsorber satisfies the
. in place penetration testing acceptance criteria of less than 2%
total bypass and uses the test procedure guidance in Regulatory Positions C.S.a, and C.S.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is 49,500 cfm i 10% for the recirculation charcoal adsorber; (2) Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample from the recirculation charcoal adsorber 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% when tested at a temperature
- of 30*C and a relative humidity of 70%; and (3) Verifying a system flow rate of 49,500 cfm i 10% for the Recircula-tion Charcoal Adsorber when tested in accordance with ANSI N510-1980.
- 1. After each complete or partial replacement of a charcoal adsorber bank in the Recirculation Charcoal Adsorber System by verifying that the cleanup system satisfies the in place penetration testing acceptance criteria of less than 0.1% in accordance with ANSI N510-1980 for a halogenated hydrocarbon refrigerant test gas while operating at a system flowrate of 49,500 cfm i 10%.
- j. After every 720 hours of Recirculation Charcoal Adsorber operation by l
verifying within 31 days after removal, that a laboratory analysis of l a representative carbon sample obtained in accordance with Regulatory l Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978 meets the laboratory testing criteria of Regulatory Guide 1.52, Rivision 2, March 1978 for a methyliodide penetration of less than 1% when tested at a temperature of 30*C and a relative humidity of 70%. BRAIDWOOD - UNITS 1 & 2 3/4 7-16
PLANT SYSTEMS FINAL. RAFT AUG 191996 3/4.7.7 NON-ACCESSIBLE AREA EXHAUST FILTER PLENUM VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.7.7 Three independent non-accessible area exhaust filter plenums (50% capacity each) shall be OPERABLE. - APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With one non-accessible area exhaust filter plenum inoperable, restore the inoperable plenum to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. l SURVEILLANCE REQUIREMENTS 4.7.7 Each non-accessible area exhaust filter plenum shall be demonstrated OPERABLE:
- a. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that operation occurs for at least 15 minutes;
- b. At least once per 18 months, or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone communi- .
cating with the exhaust filter plenum by:
- 1) Verifying that the exhaust filter plenum satisfies the in place penetration and bypass leakage testing acceptance criteria of less than 1% when using the test procedure guidance in Regulatory Positions C.S.a, C.S.c and C.S.d of Regulatory Guide 1.52, Revi-sion 2, March 1978, and the flow rate is 66,900 cfm i 10% for the train and 22,300 cfm 110% per bank;
- 2) Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample from each bank of adsorbers of the train 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 methyl idodide penetration of less than 1% when tested at the temperature of 30 C and a relative humidity of 70%; ,
BRAIDWOOD - UNITS 1 & 2 3/4 7-17
l FINAL DRAFT PLANT SYSTEMS gjg2ggg SURVEILLANCE REQUIREMENTS (Continued)
- 3) Verifying a system flow rate of 66,900 cfm i 10% through the train and 22,300 cfm 110% per bank through the exhaust filter i plenum during operation when tested in accordance with ANSI N510-1980; and
- 4) Verifying that with the system operating at a flow rate of 66,900 cfm 110% through the train and 22,300 cfm 110% per bank and exhausting through the HEPA filter and charcoal adsorbers, the total bypass flow of the system and the damper leakage is less than or equal to 1% when the system is tested by admitting cold 00P at the system intake and the damper leakage rate is determined by either direct measurements or pressure decay measurements at a test pressure of 2 inches of water and the auxiliary building exhaust fans are operating at their rated flow.
- c. After every 720 hours of charcoal adsorber operation, by verifying, within 31 days after removal, that a laboratory analysis of a repre-sentative carbon sample obtained from each bank of adscrbers of the train in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing cri-teria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, when the average for a methyl iodide penetration of less than 1% when tested at a temperature of 30*C and a relative humidity of 70%.
- d. At least once per 18 months by:
- 1) Verifying for each filter bank of the train that the pressure drop across the combined HEPA filters and charcoal adsorber banks of less than 6.0 inches Water Gauge while operating the exhaust filter plenum at a flow rate of 66,900 cfm + 10% through the train and 22,300 cfm 110% per bank;
- 2) Verifying that the exhaust filter plenum starts on manual initiation or Safety Injection test signal; and
- 3) Verifying that the system maintains the ECCS equipment rooms at a negative pressure of greater than or equal to 1/4 in. Yater Gauge relative to the outside atmosphere during system operation while operating at a flow rate of 66,900 cfm 110% through the train and 22,300 cfm 110% per bank.
- e. After each complete or partial replacement of a HEPA filter bank, by verifying that the exhaust filter plenum satisfies the in place penetration testing acceptance criteria of less than 1% in accord-ance with ANSI N510-1980 for a 00P test aerosol while operating at a flow rate of 66,900 cfm + 10% through the train and 22,300 cfm 110%
per bank; and - I l BRAIDWOOD - UNITS 1 & 2 3/4 7-18 l l
FINAL DRAFT PLANT SYSTEMS AUG 191986 SURVEILLANCE REQUIREMENTS (Continued) l
- f. After each complete or partial replacement of a charcoal adsorber l bank, by verifying that the exhaust filter plenum satisfies the l in place penetration testing acceptance criteria of less than 1%
in accordance with ANSI N510-1980 for a halogenated hydrocarbon refrigerant test gas while operating the system at a flow rate of 66,900 cfm + 10% through the train and 22,300 cfm 110% per bank. l l l 1 I l l l l BRAIDWOOD - UNITS 1 & 2 3/4 7-19 l l i
~
PLANT SYSTEMS-FINAL HAFT ! 3/4.7.8 SNU8BERS AUG 191966 LIMITING CONDITION FOR OPERATION 3.7.8 All snubbers shall be OPERABLE. Snubbers excluded from this require-ment are those installed on nonsafety-related systems and then only if their failure or failure of the system on which they are installed would have no adverse effect on any safety-related system. APPLICABILITY: MODES 1, 2, 3, and 4. MODES 5 and 6 for snubbers located on systems required OPERABLE in those MODES. ACTION: With one or more snubbers inoperable, within 72 hours replace or restore the inoperable snubber (s) to OPERABLE status and perform an engineering evaluation per Specification 4.7.8g. on the attached component or declare the attached system inoperable and follow the appropriate ACTION statement for that system. SURVEILLANCE REQUIREMENTS 4.7.8 Each snubber shall be demonstrated OPERABLE by performance of the fallowing augmented inservice inspection program and the requirements of Specification 4.0.5.
- a. Inspection Types As used in this specification, type of snubber shall mean snubbers of the same design and manufacturer, irrespective of capacity.
- b. Visual Inspections Snubbers are categorized as inaccessible or accessible during reactor operation. Each of these groups (inaccessible and accessible) may be inspected independently according to the schedule below. The first inservice visual inspection of each type of snubber shall be performed after 4 months but within 10 months rf commencing POWER OPERATION and shall include all hydraulic and W htlical snubbers. If all snubbers of each type are found OPERAPM ut b : the first inservice visual inspection, the second insewie v".xil inspection of that type shall be performed at the first retaeling octage. Otherwise, subsequent visual inspections of a given type shall be performed in accordance with the following schedule:
BRAIDWOOD - UNITS 1 & 2 3/4 7-20
FINAL MAFT ! l PLANT SYSTEMS AUg 1 g toog SURVEILLANCE REQUIREMENTS (Continued) No. of Inoperable Snubbers of Each Subsequent Visual Type per Inspection Period Inspection Period *# 0 18 months i 25% 1 12 months i 25% 2 6 months i 25% 3,4 124 days i 25% 5,6,7 62 days i 25% 8 or more 31 days i 25%
- c. Visual Inspection Acceptance Criteria Visual inspections shall verify that: (1) there are no visible indications of damage or impaired OPERABILITY, (2) attachments to the foundation or supporting structure are functional and (3) fasteners for attachment of the snubber to the component and to the snubber anchorage are functional. Snubbers which appear inoperable as a result of visual inspections may be determined OPERABLE for the purpose of establishing the next visual inspection interval, provided that:
(1) the cause of the rejection is clearly established and remedied for that particular snubber and for other snubbers irrespective of type that may be generically susceptible; and (2) the affected snubber is functicnally tested in the as-found condition and determined OPERABLE per Specification 4.7.8f. All snubbers connected to an in-operable common hydraulic fluid reservior shall be counted as inoper-able snubbers.
- d. Transient Event Inspection An inspectio~n shall be performed of all snubbers attached to sections of systems that have experienced unexpected, potentially damaging transients as determined from a review of operational data and a visual inspection of the systems within 6 months following such an event. In addition to satisfying the visual inspection acceptance criteria, freedom of-motion of mechanical snubbers shall be verified using at least one of the following: (1) manually induced snubber movement; or (2) evaluation of in place snubber piston setting; or (3) stroking the mechanical snubber through its full range of travel.
*The inspection interval of each type of snubber shall not be lengthened more than one step at a time unless a generic problem has been identified and cor-rected; in that event the inspection interval may be lengthened one step the first time and two steps thereafter if no inoperable snubbers of that type are found. #The provisions of Specification 4.0.2 are not applicable.
BRAIDWOOD - UNITS 1 & 2 3/4 7-21
PLANT SYSTEMS FINAL DRAFT AUG 1919as SURVEILLANCE REQUIREMENTS (Continued)
- e. Functional Tests During the first refueling shutdown and at least once per 18 months thereafter during shutdown, a representative sample of snubbers of each type shall be tested using one of the following sampl,e plans.
The sample plan shall be selected prior to the test period and cannot be changed during the test period. The NRC Regional Administrator shall be notified in writing of the sample plan selected for each snubber type prior to the test period or the sample plan used in the prior test period shall be implemented:
- 1) At least 10% of the total of each type of snubber shall be functionally tested either in place or in a bench test. For each snubber of a type that does not meet the functional test acceptance criteria of Specification 4.7.8f., on additional 10%
of that type of snubber shall be functionally tested until no more failures are found or until all snubbers of that type have been functionally tested; or
- 2) A representative sample of each type of snubber shall be func-tionally tested in accordance with Figure 4.7-1. "C" is the total number of snubbers of a type found not meeting the accep-tance requirements of Specification 4.7.8f. The cumulative number of snubbers of a type tested is denoted by "N". At the end of each day's testing, the new values of "N" and "C" (pre-vious day's total plus current day's increments) shall be plotted on Figure 4.7-1. If at any time the point plotted falls in the " Reject" region, all snubbers of that type shall be functionally tested. If at any time the point plotted falls in the " Accept" region, testing of snubbers of that type may be terminated. When the point plotted lies in the " Continue Test-ing" region, additional snubbers of that type shall be tested until the point falls in the " Accept" region or the " Reject" region, or all the snubbers of that type have been tested; or
- 3) An initial representative sample of 55 snubbers shall be function-ally tested. For each snubber type which does not meet the functional test acceptance criteria, another sample of at l
l l l BRAIDWOOD - UNITS 1 & 2 3/4 7-22
PLANT SYSTEMS FINAL DRAFT SURVEILLANCE REQUIREMENTS (Continued) E N '7eM least one-half the size of the initial sample shall be tested until the total number tested is equal to the initial sample size multi-plied by the factor, 1 + C/2, where "C" is the number of snubbers found which do not meet the functional test acceptance criteria. The results from this sample plan shall be plotted using an." Accept" line which follows the equation N = 55(1 + C/2). Each snubber point should be plotted as soon as the snubber is tested. If the point plotted falls on or below the " Accept" line, testing of that type of snubber may be terminated. If the point plotted falls above the
" Accept" line, testing must continue until the point falls in the
" Accept" region or all the snubbers of that type have been tested.
Testing equipment failure during functional testing may invalidate that day's testing and allow the day's testing to resume anew at a later time provided all snubbers tested with the failed equipment during the day of equipment failure are retested. The representative sample selected for the functional test sample plans shall be randomly selected from the snubbers of each type and reviewed before beginning the testing. The review shall ensure, as far as practicable, that they are representative of the various configurations, operating en-vironments, range of size, and capacity of snubbers of each type. Snubbers placed in the same location as snubbers which failed the previous functional test shall be retested at the time of the next functional test but shall not be included in the sample plan. If during the functional testing, additional sampling is required due to failure of only one type of snubber, the functional test results shall be reviewed at that time to determine if additional samples should be limited to the type of snubber which has failed the functional testing.
- f. Functional Test Acceptance Criteria The snubber functional test shall verify that:
- 1) Activation (restraining action) is achieved within the specified range in both tension and compression;
- 2) Snubber bleed, cr release rate where required, is present in both tension and compression, within the specified range;
- 3) For mechanical snubbers the force required to initiate or main-tain motion of the snubber is,within the specified range in both directions of travel; and
- 4) For snubbers specifically required not to displace under continuous load, the ability of the snubber to withstand load without displacement.
Testing methods may be used to measure parameters indirectly- or parameters other than those specified if those results can be correlated to the specified parameters through established methods. BRAIDWOOD - UNITS 1 & 2 3/4 7-23
PLANT SYSTEMS
' AUG 19 N SURVEILLANCE REQUIREMENTS (Continued)
- g. Functional Test Failure Analysis An engineering evaluation shall be made of each failure to meet the functional test acceptance criteria to. determine the cause of the 1
failure. The results of this evaluation shall be used, if applicable, in selecting snubbers to be tested in an effort to determine the OPERABILITY of other snubbers irrespective of type which may be subject to the same failure mode. c For the snubbers found inoperable, an engineering evaluation shall be performed on the components to which the inoperable snubbers are attached. The purpose of this engineering evaluation shall be to determine if the components to which the inoperable snubbers are attached vere adversely affected by the inoperability of the snubbers in order to ensure that the component remains capable for meeting the designed service. If any snubber selected for functional testing either fails to lock up or fails to move, i.e., frozen-in place, the cause will be evaluated and, if caused by manufacturer or design deficiency, all l snubbers of the same type subject to the same defect shall be func-l tionally tested. This testing requirement shall be independent of the requirements stated in Specification 4.7.8e. for snubbers not meeting the functional test acceptance criteria.
- h. Functional Testing of Repaired and Replaced Snubbers Snubbers which fail the visual inspection or the functional test acceptance criteria shall be repaired or replaced. Replacement snubbers and snubbers which have repairs which might affect the functional test results shall be tested to meet the functional test l
criteria before installation in the unit. Mechanical snubbers shall I have met the acceptance criteria subsequent to their most recent service, and the freedom-of-motion test must have been performed within 12 months before being installed in the unit.
- i. Snubber Service Life Program The service life of hydraulic and mechanical snubbers shall be moni-tored to ensure that the service life is not exceeded between sur-veillance inspections. The maximum expected service life for various seals, springs, and other critical parts shall be determined and established based on engineering information and shall be extended or shortened based on monitored test results and failure history.
Critical parts shall be replaced so that the maximum service life will not be exceeded during a period when the snubber is required to be OPERABLE. The parts replacements shall be documented and the documentation shall be retained in accordance with Specifica-tion 6.10.2. BRAIDWOOD - UNITS 1 & 2 3/4 7-24
FINAL DRAFT AUGH5MSSE AUG 191986 10 1 \ S 8 7 REJECT j e#*xfY CONTINUE 2 3 r" I" - ACCEPT 1 F 0 10 20 30 40 50 60 70 80 90 100 N FIGURE 4.7-1 -
- SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST l
BRAIDWOOD - UNITS 1 & 2 3/4 7-25
1 PLANT SYSTEMS 3/4.7.9 SEALED SOURCE CONTAMINATION AUG 1919i LIMITING CONDITION FOR OPERATION l 3.7.9 Each sealed source containing radioactive material either in excess of 100 mi'crocuries of beta and/or gamma emitting material or 5 microcuries of alpha emitting material shall be free of greater than or equal to 0.005 microcurie of removable contamination. APPLICABILITY: At all times. ACTION:
- a. With a sealed source having removable contamination in excess of the above limits, immediately withdraw the sealed source from use and either:
- 1. Decontaminate and repair the sealed source, or -
- 2. Dispose of the sealed source in accordance with Commission Regulations.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.7.9.1 Test Requirements - Each sealed source shall be tested for leakage and/or contamination by: l a. The licensee, or
- b. Other persons specifically authorized by the Commission or an Agreement State.
The test method shall have a detection sensitivity of at least 0.005 microcurie per test sample. 4.7.9.2 Test Frequencies - Each category of sealed sources (excluding startup sources and fission detectors previously subjected to core flux) shall be tested at the frequency described below.
- a. Sources in use - At least once per 6 months for all sealed sources containing radioactive materials: ,
- 1) With a half-life greater than 30 days (excluding Hydrogen 3), )
and
- 2) In any form other than gas. ~
BRAIDWOOD-UNITS 1&2 3/4 7-26
PLANT SYSTEMS FINAL DRAFT gg y g g SURVEILLANCE REQUIREMENTS (Continued)
- b. Stored sources not in use - Each sealed scurce and fission detector shall be tested prior to use or transfer to another licensee unless tested within the previous 6 months. Sealed sources and fission detectors transferred without a certificate indicating the'last test date shall be tested prior to being placed into use; and
- c. Startup sources and fission detectors - Each sealed startup source and fission detector shall be tested within 31 days prior to being subjected to core flux or installed in the core and following repair or maintenance to the source.
4.7.9.3 Reports - A report shall be prepared and submitted to the Commission l on an annual basis if sealed source or fission detector leakage tests reveal the presence of greater than or equal to 0.005 microcurie of removable contamination. G 4 BRAIDWOOD - UNITS 1 & 2 3/4 7-27
PLAN 7 SYSTEMS 3/4.7.10 FIRE SUPPRESSION SYSTEMS. AUG 191986 FIRE SUPPRESSION WATER SUPPLY SYSTEM LIMITING CONDITION FOR OPERATION 3.7.10.1 T.be Fire Suppression Water Supply System shall be OPERABLE with:
- a. Two fire suppression pumps with their discharge aligned to the fire suppression header, and
- b. An OPERABLE flow path capable of taking suction from the pond and transferring the water through distribution piping with OPERABLE sectionalizing control or isolation valves to:
- 1) The yard hydrant isolation valves (for hydrants near buildings containing safety-related equipment),
- 2) The last valve ahead of each hose standpipe as required by Specification 3.7.10.5,
- 3) The last valve ahead of the deluge valve (on the diesel generator fuel oil storage room foam system and manual contain-ment charcoal filter deluge systems), or
- 4) Flow alarm valves (on sprinkler systems) as required by Specification 3.7.10.2. .
APPLICABILITY: At all times. ACTION:
- a. With one pump and/or one water supply inoperable, restore the inoperable equipment to OPERABLE status within 7 days or provide an alternate backup pump or supply. The provisions of Specifica-tions 3.0.3 and 3.0.4 are not applicable.
- b. With the Fire Suppression Water Supply System otherwise inoperable establish a backup Fire Suppression Water Supply System within 24 hours.
SURVEILLANCE REQUIREMENTS 4.7.10.1.1 The Fire Suppression Water Supply System shall be demonstrated OPERABLE: BRAIDWOOD - UNITS 1 & 2 3/4 7-28
l PLANT SYSTEMS
, SURVEILLANCE REQUIREMENTS (Continued) M 191966
- a. At least once per 31 days by starting the electric motor-driven pump and operating it for at least 15 minutes on recirculation flow,
- b. At least once per 31 days by verifying that each valve (manual, power- 1 operated, or automatic) in the flow path is in its correct position,
- c. At least once per 6 months by performance of a system ring header flush,
- d. At least once per 12 months by cycling each testable valve in the flow path through at least one complete cycle of full travel,
- e. At least once per 18 months by performing a system functional test which includes simulated automatic actuation of the system throughout its operating sequence, and:
- 1) Verifying that each fire pump develops a discharge of 150% of rated capacity at 65% of rated pressure (3750 gpm i 10% gpm at 107 1 10% psig), and recording measured performance at minimum and rated loads.
- 2) Cycling each valve in the flow path that is not testable during plant operation through at least one complete cycle of full travel, and
- 3) Verifying that each fire suppression pump starts (sequentially) to maintain the fire suppression water system pressure greater than or equal to 125 psig.
- f. At least once per 3 years by performing a ficw test of the system in accordance with Chapter 8, Section 16 of the Fire Protection Handbook, 15th Edition, published by the National Fire Protection Association.
4.7.10.1.2 The fire pump diesel engine shall be demonstrated OPERABLE:
- a. At least once per 31 days by verifying:
l
- 1) The fuel storage tank contains at least 325 gallons of fuel, and
- 2) The diesel starts from ambient conditions and operates for at least 30 minutes on recirculation flow.
BRAIDWOOD - UNITS 1 & 2 3/4 7-29
PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) AUG 191906
- b. At least once per 92 days by verifying that a drain sample of diesel fuel from the fuel oil day tank, obtained in accordance with ASTM-04057-1981, is within the acceptable limits specified in Table 1 of ASTM 0975-1977 when checked for viscosity, water, and sediment; and
- c. At least once per 18 months, during shutdown, by su'bjecting the diesel to an inspection in accordance with procedures prepared in conjunction with its manufacturer's recommendations for the class of service.
4.7.10.1.3 The fire pump diesel starting 24-volt battery bank shall be demonstrated OPERABLE:
- a. At least once per 7 days by verifying that:
- 1) The electrolyte level of each battery is above the plates, and
- 2) The overall battery voltage is greater than or equal to 24 volts.
- b. At least once per 92 days by verifyir.g that the specific gravity is appropriate for continued service of the battery, and
- c. At least once per 18 months, by verifying that:
- 1) The batteries, cell plates, and battery racks show no visual indication of physical damage or abnormal deterioration, and
- 2) The battery-to-battery and terminal connections are clean, tight, free of corrosion, and coated with anticorrosion material.
l BRAIDWOOD - UNITS 1 & 2 3/4 7-30
PLANT SYSTEMS AUG 191986 WATER SYSTEMS . LIMITING CONDITION FOR OPERATION 3.7.10.2 The Water Systems including the Foam Systems in the diesel generator fuel oil storage tank room, the Sprinkler Systems in the Auxiliary Building, l and the Containment Charcoal Filter Manual Deluge System shall be OPERABLE. ; APPLICABILITY: Whenever equipment protected by the above systems is required to be OPERABLE. ACTION:
- a. With one or more of the above required systems inoperable, within 1 hour establish a continuous fire watch with backup fire suppression equipment for those, areas in which redundant systems or components could be damaged; for other areas, establish an hourly fire watch patrol.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.7.10.2.1 The above required Foam 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 is in its correct position,
- b. At least once per 12 months, by cycling each testable valve in the flow path through at least one complete cycle of full travel,
- c. At least once per 18 months:
- 1) By performing a system functional test which includes simulated actuation of the system, and:
a) Verifying a " Fire Trouble" alarm is received when the isolation valve is closed, and b) Verifying a " Fire" alarm is received on actuation of the alarm switch.
- 2) By a visual inspection of the dry pipe deluge headers to verify their integrity, and
- 3) By a visual inspection of each nozzle's spray area to verify the spray pattern is not obstructed.
- d. At least once per 3 years by performing an air flow test through each open deluge nozzle and verifying each nozzle is unobstructed.
BRAIDWOOD - UNITS 1 & 2 3/4 7-31
PLANT SYSTEMS g WATER SYSTEMS ggg y g SURVEILLANCE REQUIREMENTS (Continued) 4.7.10.2.2 The required Sprinkler System shall be de::tonstrated OPERABLE:
- a. At least once per 92 days: -
- 1) By discharging water out of the inspectors test connections and verifying a " Fire" alarm.
- b. At least once per 12 months:
- 1) By a visual inspection of the sprinkler header to verify integrity and that the head spray pattern is not obstructed.
~
- 2) By verifying water flow through the 2-inch test drain at the riser.
4.7.10.2.3 The required Manual Deluge System shall be demonstrated OPERABLE:
- a. At least once per 18 months:
- 1) B cycling the isolation valve and verifying the " Fire Trouble" alarm.
- 2) By cycling the deluge valve.
- 3) By a visual inspection of the deluge header to verify integrity and that the spray heads are not obstructed.
O BRAIDWOOD - UNITS 1 & 2 3/4 7-32
PLANT SYSTEMS = = ' ' "" CO2 SYSTEMS . AG 8 !900 LIMITING CONDITION FOR OPERATION 3.7.10.3 The following CO2 Systems shall be OPERABLE:
- a. Diesel generator rooms and day tank rooms, -
- b. Lower cable spreading room,
- c. Auxiliary feedwater dicsol room and day tank room, and
- d. Cable tunnel room.
APPLICABILITY: Whenever equipment protected by the CO2 systems is required to j be OPERABLE: l ACTION:
- a. With one or more of the above required CO2 systems inoperable, within 1 hour establish a continuous fire watch with backup fire suppression equipment for those areas (Lower Cable Spreading Room) in which redundant systems or components could be damaged; for other areas, establish an hourly fire watch patrol,
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
' SURVEILLANCE REQUIREMENTS 4.7.10.3.1 Each of the above required CO 2 Systems shall be demonstrated OPERABLE at least once per 31 days by verifying that each valve (manual, power operated or automatic) in the flow path is in its correct position.
4.7.10.3.2 Each of the above required CO 2 Systems shall be demonstrated OPERABLE:
- a. At it.ast once per 7 days by verifying the plant CO2 storage tank level to be greater than 96% (9.6 tons) and pressure to be greater than 275 and less than 357 psig, and
- b. At least once per 18 months by verifying:
l 1) The system, including associated ventilation system fire dampers, actuates both automatically upon receipt of a simulated actuation signal, and manually, and i 2) Flow from each nozzle during a " Puff Test." L BRAIDWOOD - UNITS 1 & 2 3/4 7-33 '
PLANT SYSTEMS HALON SYSTEMS - AUG 191986 LIMITING CONDITION FOR OPERATION 3.7.10.4 The Halon Systems in the upper cable spreading room shall be OPERABLE. APPLICABILITY: Whenever equipment protected by the Halon System is required to be OPERABLE. ACTION:
- a. With one or more of the above required Halon systems inoperable,
( within 1 hour establish a continuous fire watch with backup fire l suppression equipment for thosa areas in which redundant systems or components could be damaged; for other areas, establish an hourly fire watch patrol.
~
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.7.10.4 Each of the above required Halon Systems shall be demonstrated OPERABLE:
- a. At least once per 31 days, by verifying that each solenoid valve in the flow path is in its correct position,
- b. At least once per 6 months by verifying Halon storage cylinder weight to be at least 95% of full charge weight and pressure to be at least 90% of full charge pressure, and
- c. At least once per 18 months by:
- 1) Verifying the system, including associated ventilation dampers, actuates automatically, upon receipt of a simulated actuation signal, and
- 2) Performance of a flow test through headers and nozzles to assure no blockage.
l BRAIDWOOD - UNITS 1 & 2 3/4 7-34
0 PLANT SYSTEMS FINR HAFT FIRE HOSE STATIONS LIMITING CONDITION FOR OPERATION 3.7.10.5 The fire hose stations given in Table 3.7-Sa for Unit 1 (Table 3.7-5b for Unit 2) shall be OPERABLE. APPLICABILITY: Whenever equipment in the areas protected by the fire hose stations is required to be OPERABLE. ACTION:
- a. With one or more of the fire hose stations given in Table 3.7-5 inoperable, provide gated wye (s) on the nearest OPERABLE hose station (s). One outlet of the wye shall be connected to the stan-dard. length of hose provided for the hose station. The second outlet of the wye shall be connected to a length of hose sufficient to provide coverage for the area left unprotected by the-inoperable hose station. Where it can be demonstrated that the physical routing of the fire hose would result in a recognizable hazard to operating personnel, plant equipment, or the hose itself, the fire hose shall be stored in a roll at the outlet of the operable hose station.
Signs shall be mounted above the gated wye (s) to identiD the proper hose to use. The above action shall be accomplished within 1 hour if the inoperable fire hose is the primary means of fire suppression; otherwise route the additional hose within 24 hours.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.7.10.5 Each of the fire hose stations given in Table 3.7-5 shall be demonstrated OPERABLE:
- a. At least once per 31 days, by a visual inspection of the fire hose stations accessible during plant operations to assure all required equipment is at the station,
- b. At least once per 18 months, by:
- 1) Visual inspection of the stations not accessible during plant operations to assure all required equipment is at the station,
- 2) Removing the hose for inspection and re-racking, and
- 3) Inspecting all gaskets and replacing any degraded gaskets in the couplings.
- c. At least once per 3 years, by:
- 1) Partially opening each hose station valve to verify valve OPERABILITY and no flow blockage, and
- 2) Conducting a hose hydrostatic test at a pressure of 150 psig or at least 50 psig above maximum fire main operating pres'sure, whichever is greater.
BRAIDWOOD - UNITS 1 & 2 3/4 7-35
TABLE 3.7-Sa (Unit 1) g FIRE HOSE STATIONS 2 y LOCATION ELEVATION HOSE RACK REEL ANGLE VALVE y Aux. Roof
. L-10: North wall U-1 of safety valve penthouse 481 1 0FP331 g L-26: South wall U-2 of safety valve penthouse 481 2 0FP338 Aux. Bldg. ] ~ 5-18: By dumb waiter 480 233 0FP458
- e. 5-15: By U-1 prefilters (near stairs) 471 176 0FP329 m S-21: By U-2 prefilters (near stairs) 471 177 0FP334 Q-17: Wall by elevator in upper cable room 469 244* 0FP469 Q-19: Wall by stairs in upper cable room 469 252* 0FP477 L-11: Outside northwest corner of upper cable spreading room A-1 467 240 0FP465 R
L-14: By the northwest door of UCSR C-1 467 241* OFP466 M-13: By the southeast corner of UCSR A-1 467 242* OFP467 Y Q-13: In the southeast corner of UCSR B-1 467 243* OFP468 M P-18: Southeast corner of UCSR C-1 467 245* OFP470 M-18: South wall of UCSR C-1 467 246* OFP471 M-18: North wall of UCSR C-2 467 247* OFP472 L-25: Outside southwest corner of UCSR A-2 467 248 0FP473 L-22: In the southwest corner of UCSR C-2 467 249* OFP474 M-23: In the northeast corner of UCSR A-2 467 250* OFP475 P-20: East wall of UCSR C-2 467 251* OFP476 Q-23: In the northeast corner of UCSR B-2 467 253* OFP478 "T1 5-21: By U-2 VA Filters (U-2 side) 5-15: By U-1 VA Filters (U-1 side) 464 464 232 234 0FP457 0FP459 g m S-21: By VA Filters (U-2 side) 456 231 0FP456 . ymme S-IS: By VA Filters (U-1 side) 456 235 0FP460 Q-24: Radwaste drumming Station 387 110 0FP375 - hg L-10: By Control room refrig. units 387 106 0FP385 L-12: By blowdown after filters 387 107 0FP384 1 5-19: By 2A RHR heat exchanger room 387 112 0FP378
- Fire hoses that do not supply the primary means of fire suppression.
4 l
TABLE 3.7-Sa (Continued) (Unit 1) m FIRE HOSE STATIONS 5 5 LOCATION ELEVATION HOSE RACK REEL ANGLE VALVE 6 g Aux. Bldg. (Continued)
. M-26: Radwaste control panel 387 109 0FP377 c- M-18: By Aux. Feedwater motor driven pump 1A 387 108 0FP383 5 N-23: By remote shutdown panel U-1 387 111 0FP376 d Q-15: By 480V MCC 132X3 387 113 0FP382
- V-18: By letdown heat exchanger 387 114 0FP379
- e. P-7: East wall 6.9 kV switchgear room 455 20 0FP324 to L-11: In UC HVAC Rm OA of LCSR C-1 455 22 0FP332 L-25: By OBVC HVAC room 455 27 0FP335 M-8: North wall of battery room 451 279 0FP638 M-26: North wall of battery room 451 280 0FP639 M-18: South wall U-1 AB by door 444 238* OFP463 L-7: West wall LCSR A-1 443 207* OFP330 ~
w M-10: In the northwest corner of LCSR B-1 443 208* OFP327 1 P-10: In the northeast corner of LCSR B-1 443 209* OFP325 w M-13: North wall of LCSR C-1 443 210* OFP326 a P-13: East wall of LCSR D-1 443 211* OFP328 M-18: North wall of LCSR U-2 by door 443 239* OFP464 L-29: West wall LCSR A-2 443 212* OFP336 M-26: Southwest corner of LCSR B-2 443 213* OFP337 P-26: Southeast corner of LCSR B-2 443 214* OFP340 M-23: South wall of LCSR C-2 443 215* OFP341 P-21: West wall of LCSR D-2 443 216* 0FP333 5-21: By cabinet 2RYO1EC (elec. pen. area) 431 229 0FP454 5-24: By U-2 cont. shield wall (elec pen. area) 431 230 0FP455 ""T1 S-15: By Pzr htr. transformer (elec pen. area) 431 236 0FP461 """ S-12: By U-1 cent. shield wall (elec pen. area) 431 237 0FP462 2n Z M-18: Rad chem offices 430 57 0FP323 Q-10: Back of Div. 11 swgr room P-11: Outside Laundry Room 430 430 283 52 0FP640 0FP313
$P g
Q-19: By U-2 VCT valve aisle 430 54 0FP342 g Q P-24: By radwaste evaporator 430 55 0FP343 V-17: By west door to decon/ change area 430 58 0FP319 V-17: By east door to decon/ change area 430 61 0FP320
- Fire hoses that do not supply the primary means of fire suppression.
TABLE 3.7-50 (Centinued) (Unit 1) g FIRE HOSE STATIONS 2 E LOCATION ELEVATION HOSE RACK REEL ANGLE VALVE 6 g Aux. Bldg. (Continued) e S-15: By Pzr. htr. transformer (elec pen. area) 419 174 0FP322 e Q-10: By electrical penetration area 419 205 0FP321 5 L-11: By waste oil tank room 405 90 0FP315 d P-18: By elevator 405 91 0FP318
- P-23: By spent resin pumps 405 92 0FP349
- e. Q-11: By laundry tanks 405 93 0FP314 m S-21: West of U-2 hydrogen recombiner 405 94 0FP348 V-21: East of U-2 hydrogen recombiner 405 95 0FP345 V-15: East of U-1 hydrogen recombiner control panel 405 96 0FP316 5-15: West of U-1 hydrogen recombiner 405 97 0FP317 N-11: By the recycle holdup tanks 368 130 0FP373 R
M-13: By the U-1 stairs 368 131 0FP374 P-13: By panel IPL84JB 368 132 0FP369 Y L-20: By the U-2 stairs 368 133 0FP355
$ P-21: By the blowdown condenser 368 134 0FP356 L-25: By the PW M/U pumps 368 135 0FP361 N-25: By chemical drain tank 368 136 0FP357 S-18: By panel IPL86J 368 138 0FP362 Q-11: By Aux. Bldg. floor drain tanks 368 139 0FP368 U-15: By U-1 spray, add tank 368 140 0FP372 V-18: By U-2 cent. chg. pump room 368 141 UFP366 "T1
- P-11
- By recycle evaporator feed pumps 350 151 0FP381 -
i M-13: By U-1 stairs 350 152 0FP370 @ Z l N-23: Q-19: By gas decay tanks By "B" Aux. Bldg. Equip. drain tank 350 350 154 155 0FP352 0FP365 m p Q-17: By "A" Aux. Bldg. Equip. drain tank By collection sump pumps 350 156 0FP371 $ Q-13: 5-18: Between moderating heat exchangers 350 350 157 158 0FP380 0FP354 '
!B
- g V-18: Between BR chiller units 350 161 0FP353 W-15: By CS pump 1A 350 163 0FP367 L-19: By OB recycle evap. room 350 153 0FP363
TABLE 3.7-Sa (Crntinued) (Unit 1) co FIRE HOSE STATIONS 5 E LOCATION ELEVATION HOSE RACK REEL ANGLE VALVE 5 j g Aux. Bldg. (Continued) M-13: By leak detection sump 334 165 0FP448 g P-18: By elevator pit 334 166 0FP449 q P-18: By IB SX pump room 334 167 0FP351 , m M-23: By IB SX pump room 334 168 0FP350 Fuel Hand. Bldg. [ na Z-15: North of decon. area 430 170 0FP389 X-21: South of spent fuel pool 430 171 0FP386 Z-15: By 480V MCC 134X6 405 172 0FP388 AA-19: Outside FC pump room 405 173 0FP387 Cont. #1 R-17: By reactor head assembly area 430 62 IFP163 R-2: By accumulator tank 1C 430 63 1FP154 { u de R-7: By equipment hatch 430 64 1FP160 R-12: By charcoal filter 1A 430 65 1FP157 R-17: By north stairs 403 98 1FP164 R-2: By RCFC 1C 403 99 1FP155 R-7: By pressurizer (outside missile shield) 403 100 IFP161 R-12: By panel IPL69J 403 101 1FP158 R-12: By PRT 381 143 IFP159
- R-17: By north stairs 381 144 1FP162 R-2: By RCFC IC 381 145 1FP156 R-7: By panel IPL52J 381 146 1FP165 2n Z Turbine Bldg. E H P K-14: By the control room 451 16 1FP194 e L-6: Outside 18 D/G room 405 87 1FP183 ' g g K-10: Outside 1A D/G room 405 281 IFP275 cn L-7: Basement of Turbine Bldg. 361 129 1FP184 L-7: Outside Div. 12 SWGR 430 47 1FP181 L-10: Outside Div. 11 SWGR 430 51 1FP182 L-5: Outside non-ESF SWGR 455 18 1FP180 L-10: Outside battery room 455 21 1FPl79 l
TABLE 3.7-5b (Unit 2) - as FIRE HOSE STATIONS - 5 G LOCATION ELEVATION HOSE RACK REEL ANGLE VALVE 6 Aux. Roof g
. L-10: North wall U-1 of safety valve penthouse 481 1 0FP331 L-26: South wall U-2 of safety valve penthouse 481 2 0FP338 d Aux. Bldg.
S-18: By dumb waiter 480 233 0FP458
" S-15: By U-1 prefilters (near stairs) 471 176 0FP329 S-21: By U-2 prefilters (near stairs) 471 177 0FP334 Q-17: Wall by elevator in upper cable room 469 244* OFP469 Q-19: Wall by stairs in upper cable room 469 252* OFP477 L-11: Outside northwest corner of upper cable spreading room A-1 467 240 0FP465 w L-14: By the northwest door of UCSR C-1 467 241* OFP466 i M-13: By the southeast corner of UCSR A-1 _467 242* OFP467 y Q-13: In the southeast corner of UCSR B-1 '467 243* OFP468 i P-18: Southeast corner of UCSR C-1 467 245* 0FP470 c'
M-18: South wall of UCSR C-1 467 246* 0FP471 M-18: North wall of UCSR C-2 467 247* 0FP472 L-25: Outside southwest corner of UCSR A-2 467 248 0FP473 L-22: In the southwest corner of UCSR C-2 467 249* OFP474 M-23: In the northeast corner of UCSR A-2 467 250* OFP475 P-20: East wall of UCSR C-2 467 251* OFP476 Q-23: In the northeast corner of UCSR B-2 467 253* OFP478 S-21: By U-2 VA Filters (U-2 side) 464 232 0FP457 T mummum S-15: By U-1 VA Filters (U-1 side) 464 234 0FP459 ,2 i 5-21: S-15: By VA Filters (U-2 side) 456 231 0FP456 5M By VA Filters (U-1 side) 456 235 0FP460 s F"""" Q-24: Radwaste Drumming Station 387 110 0FP375
- L-10
L-12: By control room refrig. units By blowdown after filters 387 387 106 0FP385 $O 107 0FP384 S-19: By 2A RHR Heat Exchanger Room 387 112 0FP378
- Fire hoses that do not supply the primary means of fire suppression, t - - - - - - -
TABLE 3.7-5b (Continued) (Unit 2) oo FIRE HOSE STATIONS E , y LOCATION ELEVATION HOSE RACK REEL ANGLE VALVE y Aux. Bldg (Continued) e M-26: Radwaste Control Panel 387 109 0FP377 c- M-18: By Aux. Feedwater motor driven pump 1A 387 108 0FP383 5 N-23: By remote shutdown panel U-1 387 111 0FP376 U Q-15: By 480V MCC 132X3 387 113 0FP382 l w V-18: By letdown heat exchanger ~387 114 0FP379
- e. P-29: East wall 6.9kV SWGR Rm U-2 455 29 0FP339 L-11: In UC HVAC Rm OA of LCSR C-1 455 22 0FP332 L-25: By OBVC HVAC Room 455 27 0FP335 M-8: North wall of battery room 451 279 0FP638 M-26: North wall of battery room 451 280 0FP639 M-18: South wall U-1 AB by door 444 238* OFP463 .
L-7: West wall LCSR A-1 443 207* OFP330 R M-10: In the northwest corner of LCSR B-1 443 208* OFP327 P-10: In the northwest corner of LCSR B-1 443 209* OFP325 Y M-13: North wall of LCSR C-1 443 210* OFP326 0 P-13: East wall of LCSR D-1 443 211* OFP328 M-18: North wall of LCSR U-2 AB By Door 443 239* OFP464 L-29: West wall LCSR A-2 443 , 212* OFP336 M-26: In southwest corner of LCSR B-2 443 213* OFP337 P-26: In southeast corner of LC.cR B-2 443 214* 0FP340 M-23: South wall of LCSR C-2 443 215* 0FP341 P-21: West wall of LCSR D-2 443 216* OFP333 9 S-21: By cabinet 2RY01EC (elec. pen. area) 431 229 0FP454 5-24: By U-2 cont. shield wall (elec pen. area) 431 230 0FP455 3 5-15: By Pzr htr. transformer (elec pen area) 431 236 0FP461 C S-12: By U-1 cont shield wall (elec pen area) 431 237 0FP462 g P M-18: Rad Chem Offices 430 57 0FP323 as P-11:. Outside Laundry Room 430 52 0FP313' Q-19: By U-2 VCT valve aisle 430 54 0FP342 { P-24: By radwaste evaporator 430 55 0FP343 V-17: By west door to decon/ change area 430 58 0FP319 V-17: By east door to decon/ change area 430 61 0FP320
- Fire hoses that do not supply the primary means of fire suppression.
TABLE 3.7-Sb (Continued)
<=
(Unit 2) FIRE HOSE STATIONS C LOCATION ELEVATION HOSE RACK REEL ANGLE VALVE g Aux. Bldg. (Continued)
' V-19: By East Door to Oecon Pad & Storage 430 59 0FP344 g Q-26: Back of Div. 21 SWGR Room 430 284 0FP641 3 S-21: By U-2 Pzr. HTR. Transformer (elec. pen. area) 419 175 0FP347 y Q-26: By U-2 Elect. Penetration Area 419 206 0FP346
- L-11: By waste oil tank room 405 90 0FP315 N P-18: By elevator 405 91 0FP318 P-23: By spent resin pumps 405 92 0FP349 Q-11: By laundry tanks 405 93 0FP314 S-21: West of U-2 hydrogen recombir.er 405 94 0FP348 V-21: East of U-2 hydrogen recombiner 405 95 0FP345 V-15: East of U-1 hydrogen recombiner control -
i y panel 405 96 0FP316
- S-15: West of U-1 hydrogen recombiner 405 97 0FP317
[ N-11: By the recycle holdup tanks 368 130 0FP373 N M-13: By the U-1 stirs 368 131 0FP374 P-13: By panel IP1.84JB 368 132 0FP369 . L-20: By the U-2 stairs 368 133 0FF355 l P-21: By the blowdown condenser 368 134 0FP356 L-25: By the PW M/U pumps 368 135 0FP361 N-25: By chemical drain tank 368 136 0FP357 S-18: By panel IPL86J 368 138 0FP362 q Q-11: By Aux. Bldg. floor drain tanks 368 139 0FP368 m Q-25: By spent resin flushing pump 368 137 0FP360 2=
- U-21
- By U-2 spray add tank 368 142 0FP358 E5 V-18: By U-2 cent. chg. pump room 368 141 0FP366 [
P-11: By recycle evaporator feed pumps 350 151 0FP381 - M-13: , By U-1 stairs N-23: By gas decay tanks 350 350 152 154 0FP370' 0FP352
$ O.
Q-19: By "B" Aux. Bldg. Equip. drain tank 350 155 0FP365 Q-17: By "A" Aux. Bldg. Equip. drain tank 350 156 0FP371
TABLE 3.7-5b (Continued) (Unit 2) E FIRE HOSE STATIONS 2: g LOCATION ELEVATION , HOSE RACK REEL ANGLE VALVE 8 Aux. Bldg. (Continued) S-18: Between moderating heat exchangers 350 158 0FP354 E V-18: Between BR chiller units 350 161 0FP353 3 Q-21: By U-2 collection sump pumps 350 159 0FP364 s W-21: By CS Pump 2A 350 164 0FP359 e* L-19: By 08 Recycle Evap Rn 350 153 0FP363
" M-13: By leak detection sump 334 165 0FP448 P-18: By elevator pit 334 166 0FP449 P-18: By IB SX pump room 334 167 0FP351 M-23: By IB SX pump room 334 168 0FP350 Fuel Hand. Bldg.
$ Z-15: North of decon. area 430 170 0FP389 y X-21: South of spent fuel pool 430 171 0FP386 1
'd Z-15: By 480V MCC 134X6 405 172 0FP388 AA-19: Outside FC pump room 405 173 0FP387 Cont. - #2 R-26: By personnel hatch 430 66 2FP163
, R-31: By east stairs 430 67 2FP154 R-37: By reactor head assembly area 430 68 2FP160 R-42: By charcoal filter 2A 430 69 2FP157 R-26: By pressurizer (outside missile shield) 403 R-31: By east stairs 403 102 103 2FP164 2FP155
*g R-37: By south stairs 403 104 2FP161 E R-42: Outside missile barrier by Loop D MS pipe 403 105 2FP158 w P R-26: By panel 2Pt52J 381 147 2FP165
- I R-31: By east stairs R-37:. By south stairs 381 381 148 149 2FP156 - @ @
' 2FP162 R-42: By PRT 381 150 2FP159
TABLE 3.7-5b (Continued) (Unit 2) g FIRE HOSE STATIONS f8 LOCATION Turbine Bldg. (Continued) ELEVATION HOSE RACK REEL ANGLE VALVE L-30: Outside Unit 2 non-ESF swgr 455 14 2FP232 l E L-24: In IM shop 455 15 2FP237 Z L-26: Outside Unit 2 battery room 455 28 2FP233 L-29: Outside Div. 22 swgr 430 42 2FP231 L-26: Outside Div. 21 swgr 430 56 2FP230
- L-29: Outside 28 D/G room 405
" 82 2FP229 K-28: Outside 2A D/G room 405 282 2FP275 L-29: By 2C cond/cond booster pump 364 124 2FP228
- 2 m
3a
=
- >==
r= 21
PLANT SYSTEMS FINAL MAFT 3/4.7.11 FIRE RATED ASSEMBLIES AUG 191986 LIMITING CONDITION FOR OPERATION 3.7,11 All fire rated assemblies (walls, floor / ceilings, cable tray enclosures and other fire barriers) separating safety-related fire areas or separating portions of redundant systems important to safe shutdown within a fire area and all sealing devices in fire-rated assembly penetrations (fire doors,. fire windows, fire dampers, cable, piping and ventilation duct penetration seals) shall be OPERABLE. APPLICABILITY: At all times. ACTION:
- a. With one or more of the above required fire rated assemblies and/or sealing devices inoperable, within 1 hour either establish a contin-uous fire watch on at least one side of the affected assembly, or l verify the OPERABILITY of fire detectors on at least one side of the l inoperable assembly and establish an hourly fire watch patrol.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.7.11.1 At least once per 18 months the above required fire barrier pe'netra-tions and penetration sealing devices shall be verified OPERABLE by:
- a. Performing a visual inspection of:
- 1) The exposed surfaces of each fire rated assembly,
- 2) Each fire window / fire damper and associated hardware,
- 3) At least 10% of each type of sealed penetration. If apparent changes in appearance or abnormal degradations are found, a visual inspection of an additional IC% of each type of sealed penetration shall be made. This inspection process shall con-tinue until a 10% sample with no apparent changes in appear-ance or abnormal degradation is found. Samples shall be selected such that each penetration seal will be inspected every 15 years.
- b. Performing a functional test of at least 10% of the fire dampers. If any nonconforming dampers are found, an additional 10% will be func-tionally tested. This process will continue until an acceptable sample is found.
4.7.11.2 Each of the above required fire doors shall be verified OPERABLE by inspecting the release and closing mechanism and latches at least once per 6 months, and by verifying:
- a. The OPERABILITY of the Fire Door Supervision System for each electri-cally supervised fire door by performing a TRIP ACTUATING DEVICE OPERATIONAL TEST at least once per 31 days,
- b. That each locked closed fire door is closed at least once per 7 days, and
- c. That each unlocked fire door without electrical supervision is closed at least once per 24 hours.
BRAIDWOOD - UNITS 1 & 2 3/4 7-45
PLANT SYSTEMS 3/4.7.12 AREA TEMPERATURE MONITORING AUG 1919% LIMITING CONDITION FOR OPERATION 3.7.12 The temperature limit of each area shown in Table 3.7-6 shal1 not be exceeded for more than 8 hours, or by more than 30*F. APPLICABILITY: Whenever the equipment in an affected area is required to be OPERABLE. ACTION:
- a. With one or more areas exceeding the temperature Ifmit(s) shown in Table 3.7-6 for more than 8 hours, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that provides a record of the cumulative time and the amount by which the temperature in the affected area (s) exceeded the limit (s) and an analysis to demonstrate the continued OPERABILITY of the affected equipment.
- b. With one or more areas exceeding the temperature' limit (s) shown in Table 3.7-6 by more than 30*F, prepare and submit a Special Report as required by ACTION a. above, and within 4 hours either return the area (s) to within the temperature limit (s) or declare the equipment in the affected area (s) inoperable.
SURVEILLANCE REQUIREMENTS 4.7.12 The temperature in each of the areas shown in Table 3.7-6 shall be determined to be within its limit at least once per 12 hours. S BRAIDWOOD - UNITS 1 & 2 3/4 7-46
TABLE 3.7-6 FINAL DRAFT l l AREA TEMPERATURE MONITORING IO E TEMPERATURE AREA LIMIT (*F) j
- 1. Misc. Electric Equipment and Battery Rooms 108 .
- 2. ESF Switchgear Rooms 108
- 3. Division 12 for Unit 1 (Division 22 for Unit 2) Cable Spreading Room 108
- 4. Upper Cable Spreading Rooms 90
- 5. Diesel-Generator Rooms 132
- 6. Diesel Oil Storage Rooms 132
- 7. Aux. Building Vent Exhaust Filter Cubicle 122
- 8. Centrifugal Charging Pump Rooms 122
- 9. Containment Spray Pump Rooms 130
- 10. RHR Pump Rooms 130
- 11. Safety Injection Pump Room 130
- 12. Control Room 90
- 13. Lower Cable Spreading Rooms ' 108 BRAIDWOOD - UNITS 1 & 2 3/4 7-47
I 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES . AUG 191986 OPERATING LIMITING CONDITION FOR OPERATION 3.8.1.1 As a minimum, the following A.C. electrical power sources shall be i OPERABLE: l
- a. Each Class 1E 4160 volt bus capable of being powered from: l
- 1) Either transformer of a given units normal System Auxiliary Transformer bank, and
- 2) Either transformer of the other units System Auxiliary Transformers bank, with Each units System Auxiliary Transformer bank energized from an independent transmissica circuit.
- b. Two separate and independent diesel generators, each with:
- 1) A separate day tank containing a minimum volume of 450 gallons of fuel,
- 2) A separate Fuel Oil Storage System containing a minimum volume of 44,000 gallons of fuel, and
- 3) A separate fuel transfer pump.
APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:
- a. With either an offsite circuit or diesel generator of the above required A.C. electrical power sources inoperable, demonstrate the OPERASILITY of the remaining A.C. sources by performing Specification 4.8.1.1.la within 1 hour and at least once per 8 hours thereafter and Specification 4.8.1.1.2.a.4 within 24 hours; restore at least two offsite circuits and two diesel generators to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- b. With one offsite circuit and one diesel generator of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C. sources by performing Specifications 4.8.1.1.la within 1 hour and at least once per 8 hours thereafter and Specification 4.8.1.1.2.a.4 within 8 hours; restore at least one of the inoperable sources to OPERABLE status within 12 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. Restore at least two offsite circuits and two diesel generators to OPERABLE status within 72 hours from the time of initial loss or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
BRAIDWOOD - UNITS 1 & 2 3/4 8-1
FINAL MAFI ELECTRICAL POWER SYSTEMS AUG 191986 LIMITING CONDITION FOR OPERATION ACTION (Continued)
- c. With one diesel generator inoperable in addition to ACTION a. or b.
above, verify that:
- 1. All required systems, subsystems, trains, components and devices that depend on the remaining OPERABLE diesel generator as a source of emergency power are also OPERABLE, and
- 2. When in MODE 1, 2, or 3, the diesel-driven auxiliary feedwater pump is OPERABLE and the other Unit's A Diesel Generator is OPERABLE,* if the inoperable diesel generator is the emergency power supply for the motor-driven auxiliary feedwater pump.
If these conditions are not satisfied within 2 hours be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- d. With two of the above required offsite A.C. circuits inoperable, demonstrate the OPERABILITY of two diesel generators by performing Specification 4.8.1.1.2a.4) within 8 hours, unless the diesel gener-ators are already. operating; restore at least one of the inoperable offsite sources to OPERABLE status within 24 hours or be in at least HOT STANOBY within the next 6 hours. With only one offsite source restored, restore at least two offsite circuits to OPERABLE status within 72 hours from time of initial loss or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- e. With two of the above required diesel generators inoperable, demonstrate the OPERABILITY of two offsite A.C. circuits by perform-ing Specification 4.8.1.1.la. within 1 hour and at least once per 8 hours thereafter; restore at least one of the inoperable diesel generators to OPERABLE status within 2 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 bcurs. Restore at least two diesel generators to OPERABLE status within 72 hours 1 rom time of initial loss or be in least HOT STANDBY within the f. ext 6 hours and in COLD SHUTDOWN within the following 30 hours.
- Until 2 years after issuance of an operating license for Unit 1 the Unit 2, A diesel generator must be capable of providing power to bus 141, and the LCO, ACTION and SURVEILLANCE requirements of Specifications 3/4.8.1.3 shall be applicable. Subsequently, LCO 3.8.1.1.b.1), 2), and 3), and Surveillance Requirements 4.8.1.1.2 shall be applicable to the Unit 2, A diesel as applicable for demonstrating that the Unit 2, A diesel is OPERABLE as an emergency power supply for the Unit 1 motor-driven auxiliary feedwater pump.
BRAIDWOOD - UNITS 1 & 2 3/4 8-2
ELECTRICAL POWER SYSTEMS FINAL HAFT ' AUG 191986 SURVEILLANCE REQUIREMENTS 4.8.1.1.1 Each of the above required independent circuits between the offsite transmission network and the Onsite Class 1E Distribution System shall be:
- a. Determined OPERABLE at least once per 7 days by verifying correct breaker alignments, indicated power availability, and .
l
- b. Demonstrated OPERABLE at least once per 18 months during shutdown by l transferring manually unit power supply from the normal circuit to !
the alternate circuit. 4.8.1.1.2 Each diesel generator shall be demonstrated OPERABLE:
- a. In accordance with the frequency specified in Table 4.8-1 on a STAGGERED TEST BASIS by:
- 1) Verifying the fuel level in the day tank,
- 2) Verifying the fuel level in the fuel storage tank,
- 3) Verifying the fuel transfer pump starts and transfers fuel from the storage system to the day tank,
- 4) Verifying the diesel starts from ambient condition and accelerates to at least 600 rpm in less than or equal to 10 seconds.* The generator voltage and frequency shall be 4160 + 420 volts and 60 + 1.2 Hz within 10 seconds
- after the start signal. The diesel generator shall be started for this test by using one of the following signals:
a) Manual, or b) Simulated loss of ESF bus voltage by itself, or c) Simulated loss of ESF bus voltage in conjunction with an ESF actuation test signal, or d) An ESF actuation test signal by itself.
- 5) Verifying the generator is synchronized, loaded to greater than or equal to 5500 kW in less than or equal to 60 seconds *, operates with a load greater than or equal to 5500 kW for at least 60 minutes, and
- 6) Verifying the diesel generator is aligned to provide standby power to the associated ESF busses.
- b. At least once per 31 days and after each operation of the diesel where the period of operation was greater than or equal to 1 hour by checking for and removing accumulated water from the day tanks;
- The diesel generator start (10 sec) from ambient conditions shall be performed at least once per 184 days in these surveillance tests. All other engine starts for the purpose of this surveillance testing may be preceded by an engine pre-lube period and/or other warmup procedures recommended by the manufacturer so that nechanical stress and wear on the diesel engine is minimized.
BRAIDWOOD - UNITS 1 & 2 3/4 8-3
ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) M II
- c. At least once per 31 days by checking for and removing accumulated water from the fuel oil storage tanks;
- d. By sampling new fuel oil in accordance with ASTM-D4057 prior,to addition to storage tanks and:
- 1) By verifying in accordance with the tests specified in ASTM-D975-81 prior to addition to the storage tanks that the sample has:
a) An API Gravity of within 0.3 degrees at 60*F, or a specific gravity of within 0.0016 at 60*F, when compared to the supplier's certificate, or an absolute specific gravity at 60*F of greater than or equal to 0.83 but less than or equal to 0.89, or an API gravity of greater than or equal to 27 degrees but less than or equal to 39 degrees; b) A kinematic viscosity at 40*C of greater than or equal to 1.9 centistokes, but less than or equal to 4.1 centistokes, if the gravity was not determined by comparison with the supplier's certification; c) A flash point equal to or greater than 125*F; and d) A clear and bright appearance with proper color when tested in accordance with ASTM-D4176-82.
- 2) By verifying within 30 days of obtaining the sample that the other properties specified in Table 1 of ASTM-0975-81 are met when tested in accordance with ASTM-0975-81 ex::ept that the analysis for sulfur may be performed in accordance with ASTM-D1552-79 or ASTM-D2622-82.
- e. At least once every 31 days by obtaining a sample of fuel oil from the storage tank, in accordance with ASTM-D2276-78, and verifying that total particulate contamination is less than 10 mg/ liter when checked in accordance with ASTM-D2276-78, Method A.
- f. At least once per 18 months, during shutdown, by:
- 1) Subjecting the diesel to an inspection in accordance with procedures prepared in conjunction with its manufacturer's recommendations for this class of standby service,
- 2) Verifying the generator capability to reject a load of greater than or equal to 1034 kW while maintaining voltage at 4160 + 420 volts and frequency at 60 + 4.5 Hz, (transient state 7,60+1.2Hz(steadystate).
BRAIDWOOD - UNITS 1 & 2 3/4 8-4
ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS.(Continued) 10 E
- 3) Verifying the diesel generator capability to reject a load of 5500 kW without tripping. The generator voltage shall not exceed 4784 volts during and following the load rejection,
- 4) . Simulating a loss of ESF bus voltage by itself, and: -
a) Verifying de-energization of the ESF busses and load shedding from the ESF busses, and I b) Verifying the diesel starts on the auto-start signal, ' energizes the ESF busses with permanently connected loads within 10 seconds, energizes the auto-connected safe shutdown loads through the load sequencing timer and operates for greater than or equal to 5 minutes while its generator is loaded with the shutdown loads. After energization, the steady-state voltage and frequency of the ESF busses shall be maintained at 4160 + 420 volts and 60 + 1.2 Hz during this test.
~
- 5) Verifying that on an ESF Actuation test signal without loss of ESF bus voltages, the diesel generator starts on the auto-start signal and operates on standby for greater than or equal to 5 minutes. The generator voltage and frequency shall be 4160 + 420 volts and 60 + 1.2 Hz within 10 seconds after the auto-start signal; the generator steady state generator voltage and frequency shall be maintained within these limits during this test;
- 6) Simulating a loss of ESF bus voltage in conjunction with an ESF Actuation test signal, and
~
a) Verifying deenergization of the ESF busses and load shedding from the ESF busses; b) Verifying the diesel starts on the auto-start signal, energizes the ESF busses with permanently connected loads within 10 seconds, energizes the auto-connected emergency (accident) loads through the LOCA sequencer and operates for greater than or equal to 5 minutes while its generator is loaded with emergency loads. After l energization, the steady-state voltage and frequency of the ESF busses shall be maintained at 4160 + 420 volts and 60 + 1.2 Hz during this test; and c) Verifying that all automatic diesel generator trips, except engine overspeed and generator differential, are automatically bypassed upon loss-of-voltage on the emergency bus concurrent with a Safety Injection Actuation signal. BRAIDWOOD - UNITS 1 & 2 3/4 8-5
l FINAL MAFT ELECTRICAL POWER SYSTEMS AUG 191986 SURVEILLANCE REQUIREMENTS (Continued)
- 7) Verifying the diesel generator operates for at least 24 hours.
During the first 2 hours of this test, the diesel generator lead-ing shall be equivalent to the 2-hour rating of 6050 kW* and during
. the remaining 22 hours of this test, the diesel generator shall be loaded to greater than or equal to 5500 kW. The generator voltage and frequency shall be 4160 1 420 volts and 60 1 1.2 Hz within 10 seconds after the start signal; the steady-state generator voltage and frequency shall be maintained within these limits during this test. Within 5 minutes after completing this 24-hour test, perform Specification 4.8.1.1.2f.6)b);**
- 8) Verifying that the auto-connected loads to each diesel generator
, do not exceed the 2000-hour rating of 5935 kW; i
- 9) Verifying the diesel generator's capability to:
a) Synchronize with the offsite power source while the generator is loaded with its emergency loads upon a simulated restoration of offsite power, b) Transfer its loads to the offsite power source, and c) Be restored to its standby status.
- 10) Verifying that wi,th the diesel generator operating in a test mode, connected to its bus, a simulated Safety Injection signal overrides the test mode by: (1) returning the diesel generator to standby operation and (2) automatically energizing the emergency loads with offsite power;
- 11) Verifying that the fuel transfer pump transfers fuel fron each fuel storage tank to the day tank of each diesel via the installed cross-connection lines;
- 12) Verifying that the automatic LOCA and Shutdown sequence timer is OPERABLE with the interval between each load block within i 10% of its design interval; and
- Instantaneous loads of 6050 kW (+0, -150) are acceptable as equivalent to the 2-hour rating provided voltage and frequency requirements and cooling system functioning requirements are verified to be within design limits at 6050 kW.
**If Specification 4.8.1.1.2f.6)b) is not satisfactorily completed, it is not necessary to repeat the preceding 24-hour test. Instead, the diesel generator may be operated at 5500 kW for 1 hour or until operating temperature Jias stabilized.
i l BRAIDWOOD - UNITS 1 & 2 3/4 8-6
ELECTRICAL POWER SYSTEMS l SURVEILLANCE REQUIREMENTS (Continued)
- 13) Verifying that the following diesel generator lockout features prevent diesel generator starting only when required:
a) Turning gear engaged, and b) Emergency stop.
- g. At least once per 10 years or after any modifications which could affect diesel generator interdependence by starting both diesel generators simultaneously, during shutdown, and verifying that both diesel generators accelerate to at least 600 rpm in less than or equal to 10 seconds;
- h. At least once per 10 years by:
- 1) Draining each fuel oil storage tank, removing the accumulated
, sediment and cleaning the tank using a sodium hypochlorite ! solution, and
- 2) Performing a pressure test of those portions of the diesel fuel oil system designed to Section III, subsection ND of the ASME Code at a test pressure equal to 110 percent of the system design pressure.
- i. "At least once per 31 days by:
#1) Verifying the capability of crosstieing the Unit.2, A diesel generator to Bus 141 by independently performing the following:
a) Synchronizing the Unit 2, A diesel generator to Bus 241. b) Closing breaker 1414, and. c) Closing breaker 2414.
##2) Verifying the capability of crosstieing the Unit 1 A diesel generator to Bus 241 by independently performing the following:
a) Synchronizing the Unit 1, A diesel generator to Bus 141, b) Closing breaker 1414, and c) Closing breaker 2414.
- j. *At least once per 18 months by:
#1) Crosstieing the 2A diesel generator to Bus 141.
##2) Crosstieing the 1A diesel generator to Bus 241.
4.8.1.1.3 Reports - All diesel generator failures, valid or non-valid, shall be reported to the Commission pursuant to Specification 6.9.2. Reports of diesel generator failures shall include the information recommended in Regulatory Position C.3.b of Regulatory Guide 1.108, Revision 1, August 1977. If the number of failures in the last 100 valid tests (on a per nuclear unit basis) is greater than or equal to 7, the report shall be supplemented to include the additional information recommended in Regulatory Position C.3.b of Regulatory Guide 1.108, Revision 1, August 1977.
"This surveillance only applies to MODES 1, 2, and 3 and is not applicable until 2 years after issuance of an operating license for Unit 1. - #0nly required for Unit 1 operation in MODES 1, 2, or 3. ##0nly required for Unit 2 operation in MODES 1, 2, or 3.
BRAIDWOOD - UNITS 1 & 2 3/4 8-7
Table 4.8-1 FINAL DRAFT g 7g DIESELGENERATORTESTSCHEDUM NUMBER OF FAILURES IN LAST 20 VALID TESTS
- TEST FREQUENCY 11 At least once per 31 days .
>2 At least once per 7 days **
1 l
- Criteria for determining number of failures and number of valid tests shall be in accordance with Regulatory Position C.2.e of Regulatory Guide 1.108, Revision 1, August 1977, where the number of tests and failures is determined on a per diesel generator basis. For the purposes of this test schedule, only valid tests conducted after the completion of the preoperational test requirements of Regulatory Guide 1.108, Rev 1, Aug 1977, shall be included in the computation of the "last 20 valid tests."
**This test frequency shall be maintained until seven consecutive failure free demands have been performed and the number of failures in the last 20 valid demands has been reduced to one or less.
BRAIDWOOD - UNITS 1 & 2 3/4 8-8 l l
~
ELECTRICAL POWER SYSTEMS FINAL. DRAFT AUG 191986 A.C. SOURCES . SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.1.2 As a minimum, the following A.C. electrical power sources shall be OPERABLE. j
- a. One class 1E 4160 volt bus capable of being powered from:
- 1) Either transformer of the associated units System Auxiliary Transformer bank, or
- 2) Either transformer of the other units System Auxiliary Trans-former bank, with the System Auxiliary Transformer bank supplying the 4160 volt bus energized from an off-site transmission circuit.
- b. One diesel generator with:
- 1) A day-tank containing a minimum volume of 450 gallons of fuel,
- 2) A fuel storage system containing a minimum volume of 44,000 gallons of fuel, and
- 3) A fuel transfer pump.
APPLICABILITY: MODES 5 and G. ACTION: With less than the abo'ev minimum required A.C. electrical power sources OPERABLE, immediately suspend all operations involving CORE ALTERATIONS, positive reactiv-ity changes, movement of irradiated fuel, or crane operation with loads over the spent fuel pool, and within 8 hours, depressurize and vent the Reactor Coolant System through at least a 2 square inch vent. In addition, when in MODE 5 with the reactor coolant loops not filled, or in MODE 6 with the water level less than 23 feet above the reactor vessel flange, immediately initiate corrective action to restore the required sources to OPERABLE status as soon as possible. SURVEILLANCE REQUIREMENTS 4.8.1.2 The above required A.C. electrical power sources shall be demonstrated OPERABLE by the performance of each of the requirements of Specifications 4.8.1.1.1, 4.8.1.1.2 (except for Specification 4.8.1.1.2a.5)), and 4.8.1.1.3. BRAIDWOOD - UNITS 1 & 2 3/4 8-9
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3/4.8 ELECTRICAL POWER SYSTEMS F1NAL HAFT 3/4.8.1 A. C. SOURCES . AUG 191986 OPERATING LIMITING CONDITION FOR OPERATION 3.8.1.3* The Unit 2, A Diesel Generator shall be capable of being manually started and crosstied to Bus 141. APPLICABILITY: MODES 1, 2, and 3. ACTION: With the Unit 2, A Diesel Generator incapable of being manually started and crosstied to Bus 141, restore the diesel generator to the required status within 7 days or be in HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.* SURVEILLANCE REQUIREMENTS 4.8.1.3* The Unit 2, A Diesel Generator shall be demonstrated capable of pro-viding power to Bus 141:
- a. At least once per day by:
- 1. Verifying the day tank level is greater than 450 gallons.
- 2. Verifying DC control power is available to the Unit 2, A Diesel Generator local control panel (2PLO7J)J
- 3. Verifying that at least one starting air receiver is at greater than 175 psig.
- 4. Verifying the Essential Service Water System is available to supply cooling requirements.
- b. At least once per 31 days by:
- 1. Verifying the diesel generater starts manually and operates with a load of greater than or equal to 5500 KW for one half hour.
~ 2. Verifying the capability of crosstieing the Unit 2, A diesel generator to Bus 141 by independently performing the following:
a) Synchronizing the Unit 2, A Diesel Generator to Bus 241. ( b) Closing breaker 1414. c) Closing breaker 2414.
- c. At least once per 18 months by:
- 1. Crosstieing the diesel generator to Bus 141.
- Applicable for only 2 years after issuance of an operating license for Unit 1.
Subsequently the requirements of Specification 3/4 8.1.1 shall be app,licable to the Unit 2, A Diesel Generator as an emergency power supply for the Unit 1 motor-driven auxiliary feedwater pump. BRAIDWOOD - UNITS 1 & 2 3/4 8-9a
ELECTRICAL POWER SYSTEMS EE 3/4.8.2 D.C. SOURCES
, AUG 19 506 OPERATING l
)
LIMITING CONDITION FOR OPERATION 3.8.2.1* As a minimum the following D.C. electrical sources shall be OPERABLE:
- a. 125-Volt D.C. Bus 111 fed from Battery 111 for Unit 1 (Bus 211 fed from Battery 211 for Unit 2), and its associated full capacity charger, and
- b. 125-Volt D.C. Bus 112 fed from Battery 112 for Unit 1 (Bus 212 fed from Battery 212 for Unit 2), and its associated full capacity charger.
APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:
- a. With one of the required battery banks and/or chargers inoperable, restore the inoperable battery bank and/or battery bus to OPERABLE status within 2 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- b. With the normal full capacity charger inoperable: 1) restore the affected battery and/or battery bus to operable status with the opposite units full capacity charger within 2 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours, and 2) restore the normal full capacity charger to operable status within 24 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE REQUIREMENTS 4.8.2.1.1 Each D.C. bus shall be determined OPERABLE and energized from its battery at least once per 7 days by verifying correct breaker alignment. 4.8.2.1.2 Each 125-volt battery bank and its associated charger shall be demonstrateo OPERABLE:
- a. At least once per 7 days by verifying that:
- 1) The parameters in Table 4.8-2 meet the Category A limits, and
- 2) The total battery terminal voltage is greater than or equal to 126 volts on float charge. ;
"This specification is only applicable prior to Unit 2 operation in MO.DE 4.
l l l BRAIDWOOD - UNITS 1 & 2 3/4 8-10
ELECTRICAL POWER SYSTEMS FINAL DRAff AUG 191986 SURVEILLANCE REQUIREMENTS (Continued)
- b. At least once per 92 days and within 7 days after a battery discharge with battery terminal voltage below 110 volts, or battery overcharge with battery terminal voltage above 145 volts, by verifying that: '
- 1) The parameters in Table 4.8-2 meet the Category B limits,
- 2) There is no visible corrosion at either terminals or connectors, or the connection resistance of these items is less than 150 x 10 s ohm *, and
- 3) The average electrolyte temperature of all connected cells is above 60*F.
- c. At least once per 18 months by verifying that:
- 1) The cells, cell plates, and battery racks show no visual indication of physical damage or abnormal deterioration,
- 2) The cell-to-cell and terminal connections are clean, tight, and coated with anticorrosion material,
- 3) The resistance of each cell-to-cell ar.d terminal connection is less than or equal to 150 x 10 s ohm *, and
- 4) The battery charger will supply a load equal to the manufacturer's rating for at least 8 hours.
- d. At least once per 18 months, during shutdown, by verifying that the battery capacity is adequate to supply and maintain in OPERABLE status all of the actual or simulated en:ergency loads for 240 minutes when the battery is subject to a battery service test;
- e. At least once per 60 months, during shutdown, by verifying that the battery capacity is at least 80% of the manufacturer's rating when subjected to a parformance discharge test. This performance discharge test may be performed in lieu of the battery service test required by Specification 4.8.2.1.2d.;
- f. At least once per 18 months during shutdown, by giving performance discharge tests of battery capacity to any battery that shows signs of degradation or has reached 85% of the service life expected for the application. Degradation is indicated when the battery capacity drops more than 10% of rated capacity from its average on previous performance tests, or is below 90% of the manufacturer's rating.
"Obtained by subtracting the normal resistance of: 1) the cross room rack connector (400 x 10.s ohm, typical) and 2) the bi-level rack connector (50 x 10 8 ohm, typical); from the measured cell-to-cell connection resistance.
BRAIDWOOD - UNITS 1 & 2 3/4 8-11 ;
I TABLE 4.8-2 ' BATTERY SURVEILLANCE REQUIREMENTS AUG 191986 CATEGORY A II) CATEGORY B(2) PARAMETER LIMITS FOR EACH LIMITS FOR EACH ALLOWABLE (3) DESIGNATED PILOT CONNECTED CELL VALUE FOR EACH CELL CONNECTED CELL Electrolyte > Minimum level > Minimum level Above top of Level indication mark, indication mark, plates, and < %" above and < %" above and not maxiium level maximum level overflowing indication mark indication mark Float Voltage > 2.13 volts 1 2.13 volts (6) > 2.07 volts Not more than 0.020 below the average of all Specific > 1.195 connected cells Gravity (4)
> 1.200(5)
Average of all Average of all connected cells connected cells
> 1.205
> 1.195(5)
TABLE NOTATIONS (1) For any Category A parameter (s) outside the limit (s) shown, the battery may be considered OPERABLE provided that within 24 hours all the Category B measurements are taken and found to be within their allowable values, and provided all Category A and B parameter (s) are restored to within limits within the next 6 days. (2) For any Category B parameter (s) outside the limit (s) shown, the battery may be considered OPERABLE provided that the Category 8 parameters are wi'.hin their allowable values and provided the Category B parameter (s) are restored to within limits within 7 days. (3) Any Category B parameter not within its allowable value indicates an l inoperable battery. 1 (4) Corrected for electrolyte temperature and level. l (5) Or battery charging current is less than 2 amps when on charge. . (6) Corrected for average electrolyte temperature. l BRAIDWOOD - UNITS 1 A 2 3/4 8-12
i ELECTRICAL POWER SYSTEMS Ea flh p D.C. SOURCES . AUG 19 EE6 SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.2.2 As a minimum, one 125-volt D.C. bus fed from its battery and its associated full-capacity charger shall be OPERABLE. APPLICABILITY: MODES 5 and 6. ACTION: 1 With the required battery bank and/or full-capacity charger inoperable, imme-diately suspend all operations involving CORE ALTERATIONS, positive reactivity changes or movement of irradiated fuel; initiate corrective action to restore the required battery bank and full-capacity charger to OPERABLE status as soon as possible, and within 8 hours, depressurize and vent the Reactor Coulant System through at least a 2 square inch vent. SURVEILLANCE REQUIREMENTS 4.8.2.2 The above required 125-volt D.C. bus fed from its battery and its I associated charger shall be demonstrated OPERABLE per Specifications 4.8.2.1.1 and 4.8.2.1.2. r b i BRAIDWOOD - UNITS 1 & 2 3/4 8-13
ELECTRICAL POWER SYSTEMS RNAL HAFT AtlG 191986 3/4.8.3 ONSITE POWER DISTRIBUTION OPERATING LIMITING CONDITION FOR OPERATION 3.8.3.1 The following electrical busses shall be energized in the specified manner for the applicable unit: ,
- a. A.C. ESF Busses consisting of:
UNIT 1 UNIT 2 Division 11 Division 21
- 1) 4160-Volt Bus 141, 1) 4160-Volt Bus 241
- 2) 480-Volt Bus 131X. 2) 480-Volt Bus 231X.
- b. A.C. ESF Busses consisting of:
UNIT 1 UNIT 2 l Division 12 Division 22
- 1) 4160-Volt Bus 142 1) 4160-Volt Bus 242
- 2) 480-Volt Bus 132X. 2) 480-Volt Bus 232X.
- c. 120-Volt A.C. Instrument Bus 111 for Unit 1 (Bus 211 for Unit 2) energized frort. Its associated inverter connected to D.C. Bus 111 for Unit 1 (Bus 211 for Unit 2),
- d. 120-Volt A.C. Instrument Bus 113 for Unit 1 (Bus 213 for Unit 2) energized from its associated inverter connected to D.C. Bus 111 for Unit 1 (Bus 211 for Unit 2),
- e. 120-Volt A.C. Instrument Bus 112 for Unit 1 (Bus 212 for Unit 2) energized from its associated inverter connected to D.C. Bus 112 for Unit 1 (Bus 212 for Unit 2), and f.
120-Volt A.C. Instrument Bus 114 for Unit 1 (Bus 214 for Unit 2) energized from its associated inverter connected to D.C. Bus 112 for Unit 1 (Bus 212 for Unit 2). APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:
- a. With one of the required divisions of A.C. ESF busses not fully energized, reenergize the division within 8 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours, l b. With one A.C. instrument bus either not energized from its associated inverter, or with the inverter not connected to its associated D.C.
bus: 1) reenergize the A.C. instrument bus within 2 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours, and 2) reenergize the A.C. instrument bus from its associated inverter connected to its associated D.C. bus within 24 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. - i i l l BRAIDWOOD - UNITS 1 & 2 3/4 8-14 l l
ELECTRICAL POWER SYSTEMS FINAL DRAFT AUG I g 1986 SURVEILLANCE REQUIREMENTS 4.8.3.1 The specified busses shall be determined energized in the required manner at least once per 7 days by verifying correct breaker alignment and indicated voltage on the ESF busses. 1 i l O BRAIDWOOD - UNITS 1 & 2 3/4 8-15
i ELECTRICAL POWER SYSTEMS ONSITE POWER DISTRIBUTION AUO 10 000 SHUTDOWN ' LIMITING CONDITION FOR OPERATION 3.8.3.2 As a minimum, the following A.C. electrical busses shall be' operable and energized in the specified manner:
- a. One 4160-Volt ESF Bus (141 or 142 for Unit 1) (241 or 242 for Unit 2),
- b. One 480-Volt ESF Bus (131X or 132X for Unit 1) (231X or 232X for Unit 2)
- c. Two of the 120-Volt A.C. instrument busses powered from their associated inverter with the inverter connected to its D.C. power supply.
APPLICABILITY: MODES 5 and 6. ACTION: With any of the above required A.C. busses inoperable or not energized, immediately suspend all operations involving CORE ALTERATIONS, positive reactivity changes, movement of irradiated fuel, or crane operation with loads over the spent fuel pool, and within 8 hours depressurize and vent the RCS through at least a 2 square inch vent. In addition, when in MODE 5 with the reactor coolant loops not filled or in MODE 6 with less than 23 feet of borated water covering the reactor vessel flange, immediately initiate corrective action to restore the required A.C. busses to OPERABLE status. l SURVEILLANCE REQUIREMENTS j 4.8.3.2 The specified busses shall be determined energized in the required l manner at least once per 7 days by verifying correct breaker alignment and l Indicated voltage on the busses. l t l BRAIDbV00 - UNITS 1 & 2 3/4 8-16 l
1 ELECTRICAL POWER SYSTEMS FINAL DRAM AUG 19 586 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES LIMITING CONDITION FOR OPERATION 3.8.4.1 All containment penetration conductor overcurrent protective devices given in Table 3.8-la for Unit 1 (Table 3.8-lb for Unit 2) shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With one or more of the above required containment penetration conductor overcurrent protective device (s) inoperable:
- a. Restore the protective device (s) to OPERABLE status or de energize the circuit (s) by tripping the associated circuit breaker or racking out or removing the inoperable circuit breaker within 72 hours, declare the affected system or component inoperable, and verify the circuit breaker to be tripped or the inoperable 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 circuit breakers tripped, their inoperable circuit breakers racked out, or removed, or
- b. Be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE REQUIREMENTS 4.8.4.1 All containment penetration conductor overcurrent protective devices given in Table 3.8-1 shall be demonstrated OPERABLE:
- a. At least once per 18 months:
- 1) By verifying that the 6.9 kV and the 4.16 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 to demonstrate that the overall penetration protection design rentains within operable limits. BRAIDWOOD - UNITS 1 & 2 3/4 8-17
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ELECTRICAL POWER SYSTEMS FINAL DRAFT AllG 191985 SURVEILLANCE REQUIREMENTS (Continued) c) For each circuit breaker found inoperable during these functional tests, an additional representative sample of at least 10% of all the circuit breakers of the inoperable type shall also be functionally tested until no more failures are found or all circuit breakers of that type have been functionally tested. By selecting and functionally testing a representative sample 2) of at least 10% of each type of 480-volt circuit breaker. Circuit breakers selected for functional testing shall be selected on a rotating basis. Testing of these circuit breakers shall consist of injecting a current in excess of the breakers nominal setpoint and measuring the response time. The measured response time will be compared to the manufacturers data to ensure that it is less than or equal to a value specified by the manufacturer. Circuit breakers found inoperable during functional testing shall be restored to OPERABLE status prior to resuming operation. For each circuit breaker found inoperable during these functional tests, an additional representative sample of at least 10% of all the circuit breakers of the inoperable ~ type shall also be functionally tested until no more failures are found or all circuit breakers of that type have been functionally tested; and
- 3) By selecting and functionally testing a representative sample of each type of fuse on a rotating basis. Each representative sample of fuses shall include at least 10% of all fuses of that type. The functional test shall consist of a nondestructive resistance measurement test which demonstrates that the fuse meets its manufacturer's design criteria. Fuses found inoper-able during these functional tests shall be replaced with OPERABLE fuses prior to resuming operation. For each fuse found inoperable during these functional tests, an additional representative sample of at least 10% of all fuses of that type shall be functionally tested until no more failures are found or all fuses of that type have been functionally tested.
- b. At least once per 60 months by subjecting each 6.9 kV and 4.16 kV circuit breaker to an inspection and preventive maintenance in accordance with procedures prepared in conjunction with its manu-l facturer's recommendations.
BRAIDWOOD - UNITS 1 & 2 3/4 8-18
l FINAL DRAFT TABLE 3.8-la AUG 191986 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES l (Unit 1) PROTECTIVE DEVICE NUMBER AND LOCATION DEVICE
- 1. 6.9 kV Switchgear IRC01PA-RCPA Primary Bus 157 Cub 1 Bus 157 Norm. Feed Backup ACB 1571 Bus 157 Emerg. Feed Backup ACB 1572 1RC01PB-RCPB Primary Bus 156 Cub 2 Bus 156 Norm. Feed Backup ACB 1561 Bus 156 Emerg. Feed Backup ACB 1562 1RC01PC-RCPC Primary Bus 158 Cub 5 Bus 158 Norm. Feed Backup ACB 1582 Bus 158 Emerg. Feed Backup ACB 1581 l
1RC01PD - RCPD Primary Bus 159 Cub 5 Bus 159 Norm. Feed Backup ACB 1592 Bus 159 Emerg. Feed Backup ACB 1,591
- 2. 480V Switchgear
, 1RYO3EA - Pzr. Primary i Htr. Backup Group A l Compt. Al-A6, B1 Backup 1RYO3EB - Pzr. Primary l Htr. Backup Group B Compt. 81-86, Al Backup BRAIDWOOD - UNITS 1 & 2 3/4 8-19
FINAL DRAFT l TABLE 3.8-la (Continued) ggyg CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES (Unit 1)
~
PROTECTIVE DEVICE NUMBER AND LOCATION DEVICE
- 2. 480V Switchgear (Continued) 1RYO3EC - Pzr. Primary Htr. Backup Group C Compt. Al-A6 Backup 1RYO3ED - Pzr. Primary Htr. Backup Group D Compt. B1-B6 Backup
- 3. 480V A.C. Ckt. Bkrs.
IVP01CA - RCFC Fan 1A Low Speed Feed Bkr Swgr 131X Primary Cub 4C Hi Speed Feed Bkr Primary l Swgr 131X Cub SC IVP01CC - RCFC Fan IC Low Speed Feed Bkr Swgr Primary 131X Cub 2C Hi Speed Feed Bkr Primary Swgr 131X Cub 3C Bus 131X Norm. Feed 141 Swgr., Backup Cub 19, ACB 1415X - IVP01CB - RCFC Fan 10 Low Speed Feed Bkr Primary Swgr 132X Cub 4C Hi Speed Feed Bkr Primary Swgr 132X Cub SC IVP01CD - RCFC Fan ID ' Low Speed Feed Bkr Primary Swgr 132X Cub 2C l BRAIDWOOD - UNITS 1 & 2 3/4 8-20
~
FINE DRAFT TABLE 3.8-la (Continued) Aus2s g CONTAINMENT P'ENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES (Unit 1) PROTECTIVE DEVICE NUMBER AND LOCATION DEVICE
- 3. 480V A.C. Ckt. Bkrs. (Continued)
Hi Speed Feed Bkr Primary Swgr 132X Cub 3C Bus 132X Norm. Feed Backup 142 Swgr., Cub 14, i ACB 1425X
- 4. 480V Molded Case Ckt. Bkrs. (MCCB)
MCC 133X4 1RC01PA-A Primary Cub B1 Backup 1RC01PA-B Primary Cub B2 Backup IHC22G Primary Cub B3 Backup IFH03G Primary Cub B4 Backup IVP05CA Primary Cub C1 Backup 1RF03P Primary Cub C2 Backup 1RC01PD-A Primary Cub D1 Backup 1RC01PD-B Primary Cub 02 Backup 1RF02PB Primary Cub D4 Backup 1RF01P Primary Cub D5 , Backup i BRAIDWOOD - UNITS 1 & 2 3/4 8-21
l FINAL DRE l l TABLE 3.8-la (Continued) CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES (Unit 1) PROTECTIVE DEVICE NUMBER AND LOCATION DEVICE
- 4. 480V Molded Case Ckt. Bkrs. (MCCB) (Continued)
MCC 133X4 1RE01PA Primary Cub D6 Backup IVP02CA Primary Cub El Backup IVP04CA Primary Cub E2 Backup IVPO4CC Primary Cub F1 Backup IEW11EA,B,C Primary Cub F3 Backup IIC02EA Primary Cub F5 Backup IIC02EB Primary i Cub G1 Backup IIC02EC Primary Cub G2 Backup MCC 134X5 IIC02EF - Primary Cub Al Backup IIC02EE Primary Cub A2 Backup IIC02ED Primary < Cub A3 Backup l BRAIDWOOD - UNITS 1 & 2 3/4 8-22
i TABLE 3.8-la (Continued) FINAL DRAFT CONTAINMENT PENETRATION CONDUCTOR AUG1g g OVERCURRENT PROTECTIVE DEVICES (Unit 1) PROTECTIVE DEVICE NUMBER AND LOCATION DEVICE
- 4. 480V Molded Case Ckt. Bkrs. (MCCB) (Continued)
MCC 134X5 1FH02J Primary Cub G1 Backup 1 1FH03J Primary Cub G2 Backup 1RC01PB-B Primary Cub B1 Backup 1RE01PB Primary Cub B3 Backup 1RC01PC-A Primary Cub C1 Backup 1RC01PC-B Primary Cub C2 Backup IVPOSCB Primary Cub J1 Backup 1RC01PB-A Primary Cub C3 Backup 1HC65G-A Primary Cub 03 Backup I IVP02CB Primary Cub F1 Backup
~
1RC01R-A Primary Cub F2 A&B Backup 1RF02PA Primary Cub G3 Backup 1EW12EA,B,C Primary Cub F3 A&B Backup
\
BRAIDWOOD - UNITS 1 & 2 3/4 8-23
FINAL DRAFT TABLE 3.8-la (Continued) AUG 191986 CONTAINMENT P'ENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES (Unit 1) PROTECTIVE DEVICE NUMBER AND LOCATION DEVICE
- 4. 480V Molded Case Ckt. Bkrs. (MCCB) (Continued)
MCC 134X5 IVPO4CB Primary Cub F4 Backup IVP04CD Primary Cub F5 Backup MCC 132X2A 1SI8808C Primary Cub A2 MCC 132X2 Backup Cub B2 ISIB8088 Primary Cub A3 MCC 132X2 Backup Cub B2 MCC 132X2 1RH8702B Primary Cub B1 Backup 1RH8701B Primary Cub B3 Backup 1CV8112 Primary Cub B4 Backup 10G079 Primary Cub C1 Backup IW0056A Primary Cub C2 Backup , 10G080 Primary Cub C3 Backup BRAIDWOOD - UNITS 1 & 2 3/4 8-24
TABLE 3.8-la (Continued) FINAL. DRAFT CONTAINMENT PENETRATION CONOUCTOR OVERCURRENT PROTECTIVE DEVICES (Unit 1) PROTECTIVE DEVICE NUMBER AND LOCATION DEVICE
- 4. 480V Molded Case Ckt. Bkrs. (MCCB) (Continued)
MCC 132X2 1RY80008 Primary Cub C4 Backup 1RC8003C Primary Cub D5 Backup 1RC80038 Primary Cub D4 Backup 1RC8002A Primary Cub G1 Backup 1RC8002B Primary Cub G2 Backup 1RC8002C Primary Cub G3 Backup 1RC80020 Primary Cub G4 Backup MCC 131X2A 15I88080 Primary Cub A2 i MCC 131X2 Backup Cub B2 ISI8808A Primary Cub A3 MCC 131X2 Backup Cub B2 MCC 131X2 , 1RC8001A Primary I Cub G1 Backup BRAIDWOOD - UNITS I & 2 3/4 8-25
RNAL DRAFT TABLE 3.8-la (Continued) E 19 E CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES (Unit 1) PROTECTIVE DEVICE NUMBER AND LOCATION DEVICE
- 4. 480V Molded Case Ckt. Bkrs. (MCCB) (Continued)
MCC 131X2 1RC8001B Primary Cub G2 Backup 1RC8001C Primary Cub G3 Backup 1RC8001D Primary Cub G4 Backup 1RH8701A Primary Cub B1 Backup 1RH8702A Primary Cub B4 Backup l ILL42J Primary Cub C1 Backup IVQ001A Primary Cub C3 ,, Backup IVQ002A Primary Cub F1 Backup 1RC8003D Primary Cub C4 Backup 1RC8003A Primary Cub C5 Backup l 10G057A Primary Cub D1 l Backup l i l ICC9416 Primary Cub D3 Backup ICC9438 Primary Cub D4 Backup 10G081 Primary Cub E2 Backup BRAIDWOOD - UNITS 1 & 2 3/4 8-26
I FINAL DRAFT 1 TABLE 3.8-la (Continued) CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES (Unit 1) PROTECTIVE DEVICE NUMBER AND LOCATION DEVICE
- 4. 480V Molded Case Ckt. Bkrs. (MCCB) (Continued)
MCC 133X6 1HC01G - Cub B2 Primary Cub B1 Backup ILLO4E - Cub C2 Primary Cub C1 Backup IVP03CA Primary Cub A3 Backup IVP03CD Primary Cub C4 Backup MCC 134X7 ILLO5E - Cub B2 . Primary Cub B1 Backup l IVP03CB Primary Cub A3 Backup IVP03CC Primary Cub B4 Backup MCC 131X2B 1 WOOS 6B Primary Cub A4 Backup 1RY8000A Primary Cub AS Backup l l l l l l , BRAIDWOOD - UNITS 1 & 2 3/4 8-27
FINAL DRAFT TABLE 3.8-la (Continued) CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES (Unit 1) PROTECTIVE DEVICE NUMBER AND LOCATION DEVICE
- 5. 260 VAC RCD Pcwer (53 rods, 5 panels)
Stationary Gripper Primary Coils (all panels) Backup Lift Coils Primary (all panels) Backup Movable Gripper Primary Coils (all panels) Backup 1 BRAIDWOOD - UNITS 1 & 2 3/4 8-28
FINAL DRAFT TABLE 3.8-1b gg y 9 g CONTAINMENT PENE1 RATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES (Unit M
~
PROTECTIVE DEVICE NUMBER AND LOCATION DEVICE )
- 1. 6.9 kV Switchgear 2RC01PA-RCPA Prirary Bus 257 Cub 7 Bus 257 Norm. Feed Backup ACB 2571 Bus 257 Emerg. Feed E.ackup AC8 2572 2RC01PB-RCPB Primary Bus 256 Cub 5
^
Bus 256 Norm. Feed Backup ACB 2561 Bus 256 Emerg. Feed Backup ACB 2562 2RC01PC-RCPC Primary Bus 258 Cub 3 Bus 258 Norm. Feed Backup AC8 2582 Bus 258 Emerg. Feed Backup ACB 2581 l 2RC01PD - RCPD Primary i Bus 259 Cub 3 ' 1 Bus 259 Norm. Feed Backup ; ACB 2592 Bus 259 Emerg. Feed Sackup ACB 2591
- 2. 480V Switchgear 2RYC3EA - Pzr. Primary Htr. Backup Group A Compt. B1-86, Al Backup 2RYO3EB - Pzr. Primary Htr. Backup Group B l
Compt. Al-A6, B1 Backup 1 BRAIDWOOD - UNITS 1 & 2 3/4 8-29
FINAL DRAFT TABLE 3.8-lb (Continued) AUG 191986 CONTAINMENT P'ENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES (Unit 2) PROTECTIVE DEVICE NUMBER AND LOCATION DEVICE
- 2. 480V Switchgear (Continued) 2RYO3EC - Pzr. Primary Htr. Backup Group C Compt. B1-86 Backup 2RYO3ED - Pzr. Primary Htr. Backup Group D Compt. Al-A6 Backup
- 3. 480Y A.C. Ckt. Bkrs.
2VP01CA - RCFC Fan 2A Low Speed Feed Bkr Swgr 231X Primary Cub 4C Hi Speed Feed Bkr Primary Swgr 231X Cub 3C 2VP01CC - RCFC Fan 2C Low Speed Feed Bkr Swgr Primary 231X Cub 2C Hi Speed Feed Bkr Primary Swgr 231X Cub 3C Bus 231X Norm. Feed 241 Swgr., Backup Cub 4, j ACB 2415X 2VP01CB - RCFC Fan 2B Low Speed Feed Bkr Primary Swgr 232X Cub 4C Hi Speed Feed Bkr Primary Swgr 232X Cub SC 2VP01CD - RCFC Fan 2D ~ Low Speed Feed Bkr Primary Swgr 232X Cub 2C BRAIDWOOD - UNITS 1 & 2 3/4 8-30 ) l 1
FINAL. DRAFT TABLE 3.8-lb (Continued) CONTAINMENT PENETRATION CONDUCTOR AUG 191986 OVERCURRENT PROTECTIVE DEVICES (Unit 2) l PROTECTIVE DEVICE NUMBER AND LOCATION DEVICE
- 3. 480V A.C. Ckt. Bkrs. (Continued)
Hi Speed Feed Bkr Primary Swgr 232X Cub 3C Bus 232X Norm. Feed Backup 242 Swgr., Cub 8, ACB 2425X
- 4. 480V Molded Case Ckt. Bkrs. (MCCB)
MCC 233X4 2RC01PA-A Primary Cub B1 Backup 2RC01PA-B Primary Cub B2 Backup 2HC22G Primary Cub B3 Backup 2FH03G Primary Cub B4 Backup 2VP05CA Primary Cub C1~- Backup 2RF03P , Primary Cub C2 Backup 2RC01PD-A Primary Cub D1 Backup 2RC01PD-B Primary Cub D2 Backup 2RF02PB Primary Cub D4 Backup 2RF01P Primary Cub DS Backup BRAIDWOOD - UNITS 1 & 2 3/4 8-31
l
~
FINAL DRAFT TABLE 3.8-lb (Continued) 1 9 1986 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES (Unit 2) PROTECTIVE DEVICE NUMBER AND LOCATION DEVICE
- 4. 480V Molded Case Ckt. Bkrs. (MCCB) (Continued)
MCC 233X4 2RE01PA Primary Cub D6 Backup 2VP02CA Primary Cub El Backup 2VP04CA Primary Cub E2 Backup 2VP04CC Primary Cub F1 Backup 2EW11EA,B,C Primary Cub F3 Backup 2IC02EA Primary : Cub F5 Backup 2IC02EB Primary ; Cub G1 Backup 2IC02EC Primary Cub G2 Backup , l MCC 234X5 2IC02EF Primary ' Cub Al Backup 2IC02EE Primary Cub A2 Backup 2IC02ED Primary Cub A3 Backup BRAIDWOOD - UNITS 1 & 2 3/4 8-32
1
~
FINAL MAFT TABLE 3.3-lb (Continued) CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES (Unit 2) PROTECTIVE DEVICE NUMBER AND LOCATION DEVICE
- 4. 480V Molded Case Ckt. Bkrs. (MCCB) (Continued)
MCC 234XS 2FH02J Primary Cub G1 Backup 2FH03J Primary Cub G2 Backup 2RC01PB-B Primary Cub B1 Bac.kup 2RE01PB Primary Cub.B3 Backup 2RC01PC-A Primary Cub C1 Backup 2RC01PC-8 Primary Cub C2 Backup 2VP05CB Primary Cub J1 Backup 2RC01PB-A Primary Cub C3 Backup 2HC65G-A Primary Cub D3 Backup 2VP02CB Primary Cub F1 Backup 2RC01R-A Primary Cub F2 A&B Backup 2RF02PA Primary Cub G3 Backup 2EW12EA,B,C Primary Cub F3 A&B Backup BRAIDWOOD - UNITS 1 & 2 3/4 8-33
~
FINAL DRAFT TABLE 3.8-lb (Continued) ggyg ' CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES l (Unit 2) PROTECTIVE DEVICE NUMBER AND LOCATION DEVICE 1
- 4. 480V Molded Case Ckt. Bkrs. (MCCB) (Continued)
MCC 234X5 2VP04CB Primary Cub F4 Backup 2VP04CD Primary Cub F5 Backup MCC 232X2A 2SI8808C Primary Cub A2 MCC 232X2 Backup Cub B2 2SI8808B Primary Cub A3 MCC 232X2 Backup Cub B2 MCC 232X2 2RH87028 Primary Cub B1 Backup 2RH8701B Primary Cub B3 Backup 2CV8112 Primary Cub B4 Backup 20G079 Primary Cub Cl Backup 2 WOOS 6A Primary Cub C2 Backup , 20G080 Primary Cub C3 Backup BRAIDWOOD - UNITS 1 & 2 3/4 8-34
TABLE 3.8-lb (Continued) CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES (Unit 2) PROTECTIVE DEVICE NUMBER AND LOCATION DEVICE
- 4. 480V Molded Case Ckt. Bkrs. (MCCB) (Continuea)
MCC 232X2 2RY80008 Primary Cub C4 Backup 2RCS003C Primary Cub D5 Backup 2RC8003B Primary Cub 04 Backup 2RC8002A Primary Cub G1 Backup 2RC8002B Primary Cub G2 Backup 2RC8002C Primary Cub G3 Backup 2RC8002D Primary Cub G4 Backup s MCC 231X2A 2SI8808D Primary Cub A2 fiCC 231X2 Backup Cub B2 2SI8808A Primary Cub A3 MCC 231X2 Backup Cub B2 MCC 231X2 , 2RC8001A Primary Cub G1 Backup 1 9 BRAIDWOOD - UNITS 1 & 2 3/4 8-35
TABLE 3.8-1b (Continued) FINAL. DRAFT g CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES (Unit 2) PROTECTIVE DEVICE - NUMBER AND LOCATION DEVICE
- 4. 480V Molded Case Ckt. Bkrs. (MCCB) (Continued)
MCC 231X2 2RC8001B Primary Cub G2 Backup 2RC8001C Primary Cub G3 Backup 2RC8001D Primary Cub G4 Backup 2RH8701A Primary Cub B1 Backup 2RH8702A Primary Cub B4 Backup 2LL42J Primary Cub Cl Backup 2VQ001A Primary Cub C3 Backup 2VQ002A Primary Cub F1 Backup 2RC8003D Primary Cub C4 Backup 2RC8003A Primary Cub C5 Backup 20G057A Primary Cub D1 Backup 2CC9416 Primary Cub D3 Backup 2CC9438 Primary Cub 04 Backup 20G081 Primary Cub E2 Backup BRAIDWOOD - UNITS 1 & 2 3/4 8-36
TABLE 3.8-16 (Continued) CONTAINMENT P'ENETRATION CONDUCTOR AUS 1 0 1006 OVERCURRENT PROTECTIVE DEVICES (Unit 2) PROTECTIVE DEVICE - NUMBER AND LOCATION DEVICE
- 4. 480V Molded Case Ckt. Bkrs. (MCCB) (Continued)
MCC 233X6 2HC01G - Cub B2 Primary Cub B1 Backup 2LLO4E - Cub C2 Primary Cub C1 Backup 2VP03CA Primary Cub A3 Backup 2VP03CD Primary Cub C4 Backup MCC 234X7 2LLOSE - Cub B2 Primary Cub B1 Backup 2VP03CB Primary Cub A3 Backup 2VP03CC Primary Cub B4 Backup MCC 231X2B 2 WOOS 6B Primary Cub A4 Backup 2RYB000A Primary Cub A5 Backup O BRAIDWOOD - UNITS 1 & 2 3/4 8-37
TABLE 3.8-1b (Continued) FINAL DRAFT CONTAINMENT P'ENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES (Unit 2) PROTECTIVE DEVICE - NUMBER AND LOCATION DEVICC
- 5. 260 VAC RCD Power (53 rods, 5 panels)
Stationary Gripper Primary - Coils (all panels) Backup Lift Coils Primary (all panels) Backup Movable Gripper Primary Coils (all panels) Backup
\
~
BRAIDWOOD - UNITS 1 & 2 3/4 8-38
ELECTRICAL POWER SYSTEMS FINAL DRAFT MOTOR-0PERATED VALVES THERMAL OVERLOAD PROTECTION DEVICES E 10 E LIMITING CONDITION FOR OPERATION 3.8.4.2 The thermal overload protection devices, integral with the motor starter of each valve listed in Table 3.8-2a for Unit 1 (Table 3.8-2b for Unit 2), shall be OPERABLE. APPLICABILITY: Whenever the motor-operated valve is required to be OPERABLE. ACTION: With one or more of the thermal overload protection devices inoperable, declare the affected valve (s) inoperable and apply the appropriate ACTION statement (s) for the affected valve (s). SURVEILLANCE REQUIREMENTS 4.8.4.2 The above required thermal overload protection devices shall be demonstrated 0FERABLE at least once per 18 months by the performance of a CHANNEL CALIBR/iTION of a representative sample of at least 25% of:
- a. All thermal overload devices, such that each device is calibrated at least once per 6 years, and
- b. All thermal overload devices such that each thermal overload is calibrated and each valve is cycled through at least ona complete cycle of full travel with the motor-operator when the thermal overload is OPERABLE, at least once per 6 years.
l BRAIDWOOD - UNITS 1 & 2 3/4 8-39
TABLE 3.8-2a FINAL DRAFT MOTOR-OPERAT D VALVES THERMAL dVERLOAD 10 E PROTECTION DEVICES UNIT 1 VALVE NUMBER FUNCTION 00G059 Unit 1 Suct Isol Viv H Recomb 2 00G060 Unit 1 Discharge Isol Viv H2 Recombiner 00G061 Unit Discharge Xtie for H2 Recombiner 00G062 Unit Xtie on Discharge of H2 Recombiner 00G063 Unit Suction Xtie for H Recombiner 2 00G064 Unit Suction Xtie for H Recombiners 2 00G065 08 H 2 Analyzer Inlet Isol V1v 00G066 OB H 2 Recomb Disch Isol Viv 10G057A OA H 2 Recomb Disch. Isol. Valve 10G079 H2 Recomb Disch. Cnmt. Isol. Valve 10G080 H2 Recomb Suct. Cnet. Isol. Valve 10G081 H2 Recomb Suction Cnmt. Isol. Valve 10G082 OA H 2 Recomb Disch Cnet Isol V1v . 10G083 OA H2 Recomb Disch Cnmt Isol V1v 10G084 OA H 2 Recomb Cnmt Outlet Isol Viv 10G085 H2 Recomb Cnmt Outlet Isol V1v 1AF006A 1A AF Pp SX Suct Isol Viv 1AF006B 1B AF Pp SX Suct Dwst Isol V1v 1AF013A AF Mtr Drv Pmp Disch Hdr Dwst Isol V1v 1AF013B I.F Mtr Drv Pmp Dsch Hdr Dwst Isol Viv 1AF013C AF Mtr Drv Pp Disch Hdr Dwst Isol V1v 1AF0130 AF Mtr Dry app Disch Hdr Dwst Isol Viv 1AF013E AF Ds1 Drv Pm Dsch Hdr Dwst Isol Viv 1AF013F AF Ds1 Drv Pp Dsch Hdr Dwst Isol Viv 1AF013G AF Ds1 Drv Pp Dsch Hdr Dwst Isol Viv 1AF013H AF Ds1 Drv Pp Dsch Hdr Dwst Isol Viv 1AF017A 1A AF Pp SX Suct Upst Isol Viv 1AF0178 IB AF Pp SX Suct Upst Isol V1v l ICC685 RCP Thermal Barrier Outlet Hdr Cnmt Isol Viv 1CC9412A CC to RH HX 1A Isol Viv ICC94128 CC to RH HX IB Isol Viv 1CC9413A RCP CC Supply Dwst CNMT Isol 1CC9413B RCPs CC Supply Upst CNMT Isol ICC9414 CC Water from RCPs Isol. Valve ICC9415 Unit 1 Serv. Loop Isol Viv ICC9416 CC Wtr from RCPS Isol. Valve
- ICC9438 CC Wtr from RC Pumps Thermal Bar Isol. Valve 1CC9473A Disch Hdr X-tie Isol Viv ICC94738 Disch Hdr X-tie Isol V1v i
BRAIDWOOD - UNITS 1 & 2 3/4 8-40
FINAL DRAFT TABLE 3.8-2a (Continued) MOTOR-OPER'AT5DVALVESTHERMALOVERLOAD AUG 1 0 1906 PROTECTION DEVICES l i UNIT 1 (Continued) VALVE NU'4BER FUNCTION ICS001A 1A CS Pp Suct from RWST 1CS001B IL CS Pp Suction from RWST 1CS007A CS Pp 1A Disch Line Dwst Isol Viv 1C50078 CS Pp 1B Disch Line Downstream Isol Viv 1CS009A 1A Pump Suction from 1A Recirc Sump 1CS0098 IB CS Cont Recirc Sump B Suct Isol Viv to CS 1CS019A CS Eductor 1A Suction Conn Isol V1v 1C50198 CS Eductor 1B Suction Conn Isol.Viv ICV 1128 MOV VCT Outlet Upstm Isol VCT Viv 1CV112C MOV VCT Outlet Dwstm Isol VCT V1v 1CV1120 MOV RWST to Chg Pp Suct Hdr ICV 112E MOV RWST to Chg Pp Suct Hdr ICV 8100 MOV RCP Seal Leakoff Hdr Isol 1CV8104 MOV Emerg Boration Viv ICV 8105 MOV Chrg Pps Disch Hdr Isol V1v 1CV8106 MOV Chrg Pps Disch Hdr Isol Viv 1CV8110 MOV A & B Chg. pp Recirc Downstream Isol ICV 8111 MOV A & B Chg Pp Recirc Upstream Isol ICV 8112 RC Pump Seal Water Return Isol. Valve ICV 8355A MOV RCP 1A Seal Inj Inlet to containment Isol ICV 8355B MOV RCP IB Seal Inj Inlet Isol 1CV8355C MOV RCP IC Seal Inj Isol ICV 83550 MOV RCP 1D Seal Inj Isol l 1CV8804A MOV RHR Sys X-Tie V1v to Chrgng Pump Suction Hdr A/B. , 1RH610 RH PP 1RH01PA Recirc, Line Isol. 1RH611 RH PP 1RH01PB Recire, Line Isol. 1RH8701A RC Loop 1A to RHR Pump Isol. Valve 1RH8702A RC Loop IC to RHR Pump Isol. Valve IRH87018 RC Loop 1A to RHR Pump Isol. Valve 1RH87028 RC Loop 1C to RHR Pump Isol. Valve 1RH8716A RH HX 1RH02AA Dwnstrm Isol Viv 1RH8716B RH HX 1RH02AB Dwnstra Isol Valve IRY8000A Prz. Relief Isol. Valve 1A l 1RY80008 Prz. Relief Isol. Valve IB l 1SI8801A SI Charging Pump Disch Isol Viv l 1518801B SI Charging Pump Disch Isol Viv ISI8802A SI PP 1A Disch Line Dwst Cont Isol V1v ' 15I8802B SI PP IB Disch Line Dwst Isol Viv BRAIDWOOD - UNITS 1 & 2 3/4 8-41
/
FINAL DRAFT TABLE 3.8-2a (Continued) MOTOR-OPERAT5DVALVESTHERMALOVERLOAD AUG 191986 PROTECTION DEVICES UNIT 1 (Continued) VALVE NUMBER FUNCTION 1SI88048 SI Pump 18 Suct X-tie from RHR HX 1518806 SI Pumps Upstream Suction Isol ISI8807A SI to Chg PP Suction Crosstie Isol Viv 15I88078 SI to Chg PP Suction Crosstie Isol Viv ISI8808A Accum. lA Disch. Isol. Valve ISI8808B Accum. 18 Disch. Isol. Valve ISI8808C Accum. 1C Disch. Isol. Valve 15I8808D Accum. 1D Disch. Isol. Valve ISI8809A SI RX HX 1A Dsch Line Dwst Isol Viv 1SI8809B SI RX HX 18 Dsch Line Dwst Isol V1v ISI8811A SI Cnat Sump A Outlet Isol Viv 15I8811B SI Cnmt Sump B Outlet Isol Viv ISI8812A SI Rwst to RH Pp 1A Outlet Isol Viv ISI88128 SI Rwst to RH Pp 1B Outlet Isol Viv 1SI8813 SI Pumps 1A-1B Recirc Line Dwst Isol 15I8814 SI Pump 1A Recirc Line Isol V1v ISI8835 SI Pumps X-tie Disch Isol Viv 15I8840 SI RHR HX Disch Line Upstrm Cont Pen Is1 V1v ISI8821A .SI PP 1A Disch Line X-tie Isol V1v ISI8821B SI Pump 18 Disch Line X-tie Isol V1v 1518920 SI Pump 1B Recirc Line Isol V1v ISI8923A SI PP 1A Suction Isol V1v ISI8923B SI Pump 18 Suct Isol Valve l 15I8924 SI Pump 1A Suction X-tie Dwnstrm Isoi Viv ISX0168 RCFC B&D Sx Supply MOV 1SX016A RCFC A&C SX Supply MOV 1SX027A RCFC A&C Return ISX027B RCFC B&D SX Return MOV OSX007 CC HX Outlet Viv OSX063A SX to Cont Rm Refrig Cdsr 0A OSX063B SX to Cont Rm Refrig Cdsr OB OSX146 CC Hx "0" Outlet Valve OSX147 CC Hx "0" Outlet Valve BRAIDWOOD - UNITS 1 & 2 3/4 8-42
I FINAL DRAFT TABLE 3.8-2a (Continued) - MOTOR-OPERATED VALVES THERMAL OVERLOAD AUG 191986 PROTECTION DEVICES UNIT 1 (Continued) VALVE NUMBER FUNCTION . OSX165A SX Train A Return Valve to Pond OSX165B - SX Train B Return Valve to Pond ISX001A 1A SX Pp Suct Viv MOV 1SX0018 IB SX Pp Suct V1v MOV . ISX004 U-1 SX Supply to U-1 CCW HX MOV ISX005 IB SX Pp Supply to O CCW HX MOV ISX007 CC HX Outlet Viv ISX010 U-1 Trn A return V1v AB l 1SX011 Trn A Trn B Unit I return X-tie V1v AB l 1SX033 1A SX Pp Disch X-tie MOV 1SX034 18 SX Pp Disch X-tie MOV ISX136 Unit 1 Trn B return V1v AB IWOOO6A Chilled wtr coils 1A & IC Supply Isol viv IWOOO68 Chilled wtr coils IB & 10 Supply Isol viv IWOO20A Chilled wtr coils 1A & IC Return Isol viv 1W00208 Chilled wtr coils IB & 10 Return Isol v1v 1W0056A Chilled Water Cnmt. Isol. Valve IW00568 Chilled Water Cnmt. Isol. Valve BRAIDWOOD - UNITS 1 & 2 3/4 8-43
FINAL DRAFT TABLE 3.8-2b MOTOR-OPERATED VALVES THERMAL OVERLOAD 1 PROTECTION DEVICES UNIT 2 VALVE NUMBER FUNCTION - 00G059 Unit 1 Suct Isol Viv H Recomb 2 ! 00G060 Unit 1 Discharge Isol V1v H2 Recombiner 00G061 Unit Discharge Xtie for H2 Recombiner 00G062 Unit Xtie on Discharge of H2 Recombiner 00G063 Unit Suction Xtie for H Recombiner 2 00G064 Unit Suction Xtie for H Recombiners 2 00G065 OB H2 Analyzer Inlet Isol Viv 00G066 OB H2 Recomb Disch Isol V1'v 20G057A OB H 2 Recomb Disch. Isol. Valve 20G079 H2 Recomb Disch. Cnmt. Isol. Valvo ' 20G080 H2 Recomb Suct. Cnmt. Isol. Valve 20G081 H2 Recomb Suction Cnmt. Isol. Valve 20G082 08 H2 Recomb Disch Cnmt Isol Viv 20G083 OB H2 Recomb Disch Cnmt Isol Viv 20G084 ' OB H 2 Recomb Cnet Outlet Isol V1v 20G085 'H 2 Recomb Cnmt Outlet Isol V1v 2AF006A 2A AF Pp SX Suct Isol Viv 2AF0068 28 AF Pp SX Suct Dwst Isol V1v 2AF013A AF Mtr Drv Pmp Disch Hdr Dwst Isol Viv 2AF013B AF Mtr Drv Pmp Dsch Hdr Dwst Isol Viv 2AF013C AF Mtr Drv Pp Disch Hdr Dwst Isol V1v 2AF013D AF Mtr Drv Pp Disch Hdr Dwst Isol Viv 2AF013E AF Ds1 Drv Pm Dsch Hdr Dwst Isol Viv 2AF013F FF Ds1 Drv Pp Dsch Hdr Dwst Isol Viv 2AF013G AF Ds1 Dry Pp Dsch Hdr Dwst Isol V1v 2AF013H AF Ds1 Drv Pp Dsch Hdr Dwst Isol V1v 2AF017A 2A AF Pp SX Suct Upst Isol Viv 2AF0178 28 AF Pp SX Suct Upst Isol Viv 20C685 RCP Thermal Barrier Outlet Hdr Cnmt Isol Viv 2CC9412A CC to RH HX 2A Isol V1v 2CC94128 CC to RH HX 2B Isol Viv 2CC9413A RCP CC Supply Dwst CNMT Isol 2CC94138 RCPs CC Supply Upst CNMT Isol 2CC9414 CC Water from RCPs Isol. Valve 2CC9415 Unit 2 Serv. Loop Isol V1v 2CC9416 CC Wtr from RCPS Isol. Valve 2CC9438 CC Wtr from RC Pumps Thermal Bar Isol. Valve - 2CC9473A Disch Hdr X-tie Isol Viv 2CC94738 Disch Hdr X-tie Isol Viv BRAIDWOOD - UNITS 1 & 2 3/4 8-44
FINAL DRAFT TABLE 3.8-2b (Continued) g MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION DEVICES UNIT 2 (Continued) VALVE NUMBER FUNCTION 2CS001A 2A CS Pp Suct from RWST 2CS001B 28 CS Pp Suction from RWST 2CS307A CS Pp 2A Disch Line Dwst Isol Viv 2CS0078 CS Pp 2B Disch Line Downstream Isol V1v 2CS009A 2A Pump Suction from 1A Recire Sump 2CS009B 2B CS Cont Recirc Sump B Suct Isol Viv to CS 2CS019A CS Eductor 2A Suction Conn Isol V1v 2CS019B CS Eductor 28 Suction Conn Iso! V1v 2CV1128 MOV VCT Outlet Upstm Isol VCT Viv 2CV112C MOV VCT Outlet Dwstm Isol VCT Viv 2CV1120 MOV RWST to Chg Pp Suct Hdr 2CV112E MOV RWST to Chg Pp Suct Hdr 2CV8100 MOV RCP Seal Leakoff Hdr Isol 2CV8104 MOV Emerg Boration Viv 2CV8105 MOV Chrg Pps Disch Hdr Isol Viv - 2CV8106 MOV Chrg Pps Disch Hdr Isol Viv 2CV8110 MOV A & B Chg. Pp Recirc Downstream Isol 2CV8111 MOV A & B Chg Pp Recirc Upstream Isol 2CV8112 RC Pump Seal Water Return Isol. Valve 2CV8355A MOV RCP 2A Seal Inj Inlet to containment Isol 2CV8355B MOV RCP 2B Seal Inj Inlet Isol 2CV8355C MOV RCP 2C Seal Inj Isol 2CV8355D MOV RCP 2D Seal Inj Isol 2CV8804A MOV RHR Sys X-Tie Viv to Chrgng Pump Suction Hdr A/B. 2RH610 RH PP 1RH01PA Recirc, Line Isol. 2RH611 RH PP 1RH01PB Recire, Line Isol. 2RH8701A RC Loop 2A to RHR Pump Isol. Valve 2RH8702A RC Loop 2C to RHR Pump Isol. Valve 2RH87018 RC Loop 2A to RHR Pump Isol. Valve 2RH87028 RC Loop 2C to RHR Pump Isol. Valve 2RH8716A RH HX 2RH02AA Dwnstrm Isol Viv 2RH8716B RH HX 2RH02AB Dwnstrm Isol Valve 2RY8000A Prz. Relief Isol. Valve 2A 2RY80008 Prz. Relief Isol. Valve 28 2SI8801A SI Charging Pump Disch Isol Viv 2SI88018 SI Charging Pump Disch Isol V1v 2SI8802A SI PP 2A Disch Line Dwst Cont Isol V1v . 25I88028 SI PP 28 Disch Line Dwst Isol Viv BRAIDWOOD - UNITS 1 & 2 3/4 8-45
FINAL DRAFT TABLE 3.8-2b (Continued) MOTOR-OPERATED VALVES THERMAL OVERLOAD AUG 191986 PROTECTION DEVICES UNIT 2 (Continued) VALVE NUMBER FUNCTION - 25I88048 SI Pump 2B Suct X-tie from RHR HX 2SI8806 SI Pumps Upstream Suction Isol 2SI8807A SI to Chg PP Suction Crosstie Isol Viv 2SI8807B SI to Chg PP Suction Crosstie Isol Viv 2 SIB 808A Accum. 2A Disch. Isol. Valve 2SI88088 Accum. 28 Disch. Isol. Valve 2SI8808C Accum. 2C Disch. Isol. Valve 2SI88080 Accum. 20 Disch. Isol. Valve 2SI8809A SI RX HX 2A Dsch Line Dwst Isoi Viv 25I88098 SI RX HX 28 Dsch Line Dwst Isol Viv 2SI8811A SI Cnet Sump A Outlet Isol Viv 2SI88118 SI Cnet Sump B Outlet Isol Viv 2SI8812A SI Rwst to RH Pp 2A Outlet Isol Viv 25188128 SI Rwst to RH Pp 2B Outlet Isol V1v 2SI8813 SI Pumps 2A-28 Recirc Line Dwst Isol 2SI8814 SI Pump 2A Recirc Line Isol V1v 2SI8835 SI Pumps X-tie Disch Isol Viv 2SI8840 SI RHR HX Disch Line Upstra Cont Pen Is1 Viv 2SI8821A SI PP 2A Disch Line X-tie Isol Viv 2SI8821B SI Pump 28 Disch Line X-tie Isol Viv 2SI8920 SI' Pump 28 Recire Line Isol V1v 2SI8923A SI PP 2A Suction Isol Viv 2SI89238 SI Pump 2B Suct Isol Valve 2SI8924 SI Pump 2A Suction X-tie Dwnstra Isol Viv 2SX016B RCFC B&D SX Supply MOV 2SX016A RCFC A&C SX Supply MOV 2SX027A RCFC A&C SX Return MOV 2SX0278 RCFC B&D SX Return MOV OSX146 CC HX "0" Outlet Valve OSX147 CC HX "0" Outlet Valve OSX165A SX Train A Return Valve to Pond OSX165B SX Train B Return Valve to' Pond 1 BRAIDWOOD - UNITS 1 & 2 3/4 8-46
FINAL DRAFT TABLE 3.8-2b (Continued) MOTOR-0PERAT50VALVESTHERMALOVERLOAD PROTECTION DEVICES UNIT 2 (Continued) VALVE NUMBER FUNCTION 2SX001A 2A SX Pp Suct V1v MOV 2SX001B 28 SX Pp Suct V1v MOV 2SX004 U-2 SX Supply to U-2 CCW HX MOV 2SX005 28 SX Pp Supply to O CCW HX MOV 2SX007 CC HX Outlet V1v 2SX010 U-2 Trn A return Viv AB 2SX011 Trn A Trn B Unit 2 return X-tie V1v AB 2SX033 2A SX Pp Disch X-tie MOV 2SX034 2B SX Pp Disch X-tie MOV 2SX136 Unit 2 Trn B return Viv AB 2W0006A Chilled wtr coils 2A & 2C Supply Isol viv 2WOOO68 Chilled wtr coils 2B & 2D Supply Isol viv 2 WOO 20A Chilled wtr coils 2A & 2C Return Isol viv 2 WOO 20B Chilled wtr coils 2B & 2D Return Isol viv 2 WOO 56A Chilled Water Cnmt. Isol. Valve ' 2 WOOS 6B Chilled Water Cnet. Isol. Valve OSX0Q7 CC HX Outlet Viv OSX063A SX to Cont Rm Refrig Cdsr 0A OSX0638 SX to Cont Rm Refrig Cdsr 08 6 BRAIDWOOD - UNITS 1 & 2 3/4 8-47
3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION AUG 191986 LIMITING CONDITION FOR OPERATION 3.9.1 The boron concentration of all filled portions of the Reactor' Coolant System and the refueling canal shall be maintained uniform and sufficient to I ensure that the more restrictive of the following reactivity conditions is I met:
- a. A K,ff of 0.95 or less, or
- b. A boron concentration of greater than or equal to 2000 ppm.
APPLICALILITY: 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 ppm boron or its equiv-alent 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 2000 ppm, whichever is the more restrictive. SURVEILLANCE REQUIREMENTS 4.9.1.1 The more restrictive of the above two reactivity conditions shall be determined prior to:
- a. Removing or unbolting the reactor vessel head, and
- b. Withdrawal of any full-length control rod in excess of 57 steps (approximately 3 feet) from its fully inserted position within the reactor vessel.
4.9.1.2 The boron concentration of the Reactor Coolant System and the refueling canal shall be determined by chemical analysis at least once per 72 hours. 4.9.1.3 Valves CV111B, CV8428, CV8441, CV8435, and CV8439 shall be verified closed and secured in position by mechanical stops or by removal of air or electrical pcwer 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.
BRAIDWOOD - UNITS 1 & 2 3/4 9-1 t i
REFUELING OPERATIONS 3/4.9.2 INSTRUMENTATION . O 10 E LIMITING CONDITION FOR OPERATION 3.9.2 As a minimum, two Source Range Neutron Flux Monitors shall be OPERABLE each with continuous visual indication in tha control room and one with audible indication in the containment and control room. APPLICABILITY: MODE 6. ACTION:
- a. With one of the above required monitors inoperable or not operating,
! immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes.
- b. With both of the above required monitors inoperable or not operating, determine the boron concentration of the Reactor Coolant System at least once per 12 hours.
SURVEILLANCE REQUIREMENTS 4.9.2 Each Source Range Neutron Flux Monitor shall be demonstrated OPERABLE by performance of:
- a. A CHANNEL CHECK at least once per 12 hours,
- b. An ANALOG CHANNEL OPERATIONAL TEST within 8 hours prior to the initial start of CORE ALTERATIONS, and
- c. An ANALOG CHANNEL OPERATIONAL TEST at least once per 7 days.
BRAIDWOOD - UNITS 1 & 2 3/4 9-2
REFUELING OPERATIONS 3/4.9.3 DECAY TIME LIMITING CONDITION FOR OPERATION 3.9.3 The reactor shall have been subcritical for at least the last 100 hours. APPLICABILITY: During movement of irradiated fuel in the reactor vessel. ACTION: With the reactor subcritical for less than 100 hours, suspend all operations involving movement of irradiated fuel in the reactor vessel. SURVEILLANCE REQUIREMENTS 4.9.3 The reactor shall be determined to have been subcritical for at least 100 hours by verification of the date and time of subcriticality prior to movement of irradiated fuel in the reactor vessel. BRAIDWOOD - UNITS 1 & 2 3/4 9-3
REFUELING OPERATIONS 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS AUS 19 % LIMITING CONDITION FOR OPERATION 3.9.4 The containment building penetrations shall be in the following status:
- a. The personnel hatch should have a minimum of one door closed at any one time and the equipment hatch shall be in place and held by a minimum of four bolts or the equipment hatch removed pursuant to Surveillance Requirement 4.9.4.2,
- b. A minimum of one door in the personnel emergency exit hatch is closed, and
- c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere shall be either:
- 1) Closed by an isolation valve, blind flange, or manual valve, or
- 2) Capable of being closed by an OPERABLE automatic containment purge isolation valve.
APPLICABILITY: During CORE ALTERATIONS or movement of irradiated fuel within the containment. ACTION: With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or movement of irradiated fuel in the containment building. SURVEILLANCE REQUIREMENTS 4.9.4.1 Each of the above required containment building penetrations shall be determined to be either in its closed / isolated condition or capable of being closed by an OPERABLE automatic containment purge isolation valve within 100 hours prior to the start of and at least once per 7 days during CORE ALTERATIONS or movement of irradiated fuel in the containment building by:
- a. Verifying the penetrations are in their closed / isolated condition, or
- b. Testing the containment purge isolation valves per the applicable l portions of Specification 4.6.3.2. .
BRAIDWOOD - UNITS 1 & 2 3/4 9-4
REFUELING OPERATIONS
. AUG 191986 SURVEILLANCE REQUIREMENTS (Continued) 4.9.4.2 Verifying that the fuel handling building exhaust filter plenums maintain the fuel building at a negative pressure of greater than or equal to 1/4 inch water gauge relative to the outside atmosphere with the equipment hatch removed, ~
- a. Prior to CORE ALTERATIONS or movement of irradiated fuel and
- b. At least once per 7 days during CORE ALTERATIONS or movement of irradiated fuel.
Verification of negative pressure will be performed with systems in the normal REFUELING MODE. O BRAIDWOOD - UNITS 1 & 2 3/4 9-5
REFUELING OPERATIONS 3/4.9.5 COMMUNICATIONS , g g LIMITING CONDITION FOR OPERATION 3.9.5 Direct communications shall be maintained between the control room and personnel at the containment refueling station. APPLICABILITY: During CORE ALTERATIONS. ACTION: When direct communications between the control room and personnel at the containment refueling station cannot be maintained, suspend all CORE ALTERATIONS. SU9VEILLANCE REQUIREMENTS 4.9.5 Direct communications between the control room and personnel at the containment refueling station shall be demonstrated within 1 hour prior to the start of and at least once per 12 hours during CORE ALTERATIONS. BRAIDWOOD - UNITS 1 & 2 3/4 9-6
i REFUELING OPERATIONS 3/4.9.6 REFUELING MACHINE ,
, AUG 19 $$g LIMITING CONDITION FOR OPERATION 3.9.6 The refueling machine shall be used for movement of drive rod.s or fuel assemblies and shall be OPERABLE with:
- a. The refueling machine used for movement of fuel assemblies having:
- 1) A capacity equal to or greater than 2850 pounds, and
- 2) An overload cutoff limit less than or equal to 2850 pounds.
- b. The auxiliary hoist used for latching and unlatching drive rods having:
- 1) A capacity equal to or greater than 2000 pounds, and
- 2) A load indicator which shall be used to prevent lifting loads in excess of 1000 pounds.
~
APPLICABILITY: During movement of drive rods or fuel assemblies within the reactor vessel. ACTION: With the requirements for refueling machine and/or hoist OPERABILITY not satis-fied, suspend use of any inoperable refueling machine and/or auxiliary hoist from operations involving the movement of drive rods and fuel assemblies within the reactor vessel. SURVEILLANCE REQUIREMENTS 4.9.6.1 Each refueling machine used for movement of fuel assemblies within the reactor vessel shall be demonstrated OPERABLE within 100 hours prior to the start of such operations by performing a load test of at least 3563 pounds and demonstrating an automatic load cutoff when the crane load exceeds 2850 pounds. 4.9.6.2 Each auxiliary hoist and associated load indicator used for movement of drive rods within the reactor vessel shall be demonstrated OPERABLE within 100 hours prior to the start of such operations by performing a load test of at least 2500 pounds. l i BRAIDWOOD - UNITS 1 & 2 3/4 9-7
REFUELING OPERATIONS 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE FACILITY AUG 19 tggg LIMITING CONDITION FOR OPERATION 3.9.7 Loads in excess of 2000 pounds shall be prohibited from travel over fuel assemblies in the spent fuel storage facility. - APPLICABILITY: With fuel assemblies in the spent fuel storage facility. ACTION:
- a. With the requirements of the above specification not satisfied, place the crane load in a safe condition.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLAN,CE REQUIREMENTS 4.9.7 Crane interlocks and physical stops which prevent crane travel with loads in excess of 2000 pounds over fuel assemblies shall be demonstrated OPERABLE within 7 days prior to crane use and at least once per 7 days thereafter during crane operation. BRAIDWOOD - UNITS 1 & 2 3/4 9-8
REFUELING OPERATIONS 3/4.9.8 RESIOUAL HEAT REMOVAL AND COOLANT CIRCULATION M 10 E l HIGH WATER LEVEL l LIMITING CONDITION FOR OPERATION 3.9.8.1 At least one residual heat removal (RHR) loop shall be OPERABLE and in operation.* i APPLICABILITY: MODE 6, when the water level above the top of the reactor vessel flange is greater than or equal to 23 feet. ACTION: I With no RHR loop OPERABLE and in operation, suspend all operations involving an increase in the reactor decay heat load or a reduction in boron concentra-tion of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to OPERABLE and operating status as soon as possible. Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours. SURVEILLANCE REQUIREMENTS 4.9.8.1 At least one RHR loop shall be verified in operstion and circulating reactor coolant at a flow rate of greater than or equal to 2800 gpm at least once per 12 hours. l "The RHR loop may be removed from operation for up to 1 hour per 8-hour period during the performance of CORE ALTERATIONS in the vicinity of the reactor
- vessel hot legs.
I 1 BRAIDWOOD - UNITS 1 & 2 3/4 9-9
REFUELING OPERATIONS LOW WATER LEVEL , AUG 19 506 LIMITING CONDITION FOR OPERATION 3.9.8.2 Two residual heat removal (RHR) loops shall be OPERABLE, and at least one RHR loop shall be in operation.*' APPLICABILITY: MODE 6, when the water level above the top of the reactor vessel flange is less then 23 feet. ACTION:
- a. With less than the required RHR loops OPERABLE, immediately initiate corrective action to return the required RHR loops,to OPERABLE status, or establish greater than or equal to 23 feet of water above the reactor vessel flange, as soon as possible.
- b. With no RHR loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to operation. Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours.
SURVEILLANCE REQUIREMENTS 4.9.8.2 At least one RHR loop shall be verified in operation and circulating reactor coolant at a flow rate of greater than or equal to 2800 gpm at least once per 12 hours.
" Prior to initial criticality, the RHR loop may be removed from operation for up to I hour per 2 hour period during the performance of CORE ALTERATIONS in the vicinity of the reactor vessel hot legs.
BRAIDh000 - UNITS 1 & 2 3/4 9-10
REFUELING OPERATIONS 3/4.9.9 CONTAINMENT PURGE ISOLATION SYSTEM AUG 191986 LIMITINGCONDITIONF.19 OPERATION 3.9.9 The Containment Purge Isolation System shall be OPEP.ABLE. - APPLICABILITY: During CORE ALTERATIONS or movement of irradiated fuel within the containment. ACTION:
- a. With the Containment Purge Isolation System inoperable, close each of the purge valves providing direct access from the containment atmosphere to the outside atm,osphere.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.9.9 The Containment Purge Isolation System shall be demonstrated OPERABLE within 100 hours prior to the start of and at least once per 7 days during , CORE ALTERATIONS by verifying that containment purge isolation occurs on an ESF l test signal from each of the containment radiation monitoring instrumentation channels. BRAIDWOOD - UNITS 1 & 2 3/4 9-11
l REFUELING OPERATIONS NA D 3/4.9.10 WATER LEVEL - REACTOR VESSEL LIMITING CONDITION FOR OPERATION j 3.9.10 At least 23 feet of water shall be maintained over the top of the reactor vessel flange. I APPLICABILITY: During movement of fuel assemblies or control rods within the containment when either the fuel assemblies being moved or the fuel assemblies seated within the reactor vessel are irradiated while in MODE 6. ACTION: With the requirements of the above specification not satisfied, suspend all operations involvi6g movement of fuel assemblies or control rods within the reactor vessel. SURVEILLANCE REQUIREMENTS ( 4.9.10 The water level shall be determined to be at least its minimum required depth within 2 hours prior to the start of and at least once per 24 hours thereafter during movement of fuel assemblies or control rods. BRAIDWOOD - UNITS 1 & 2 3/4 9-12
l REFUELING OPERATIONS 3/4.9.11 WATER LEVEL - STORAGE P0_0L #IIG 10 E LIMITING CONDITION FOR OPERATION 3.9.11 At least 23 feet of water shall be maintained over the top of irradiated fuel assemblies seated in the storage racks. APPLICABILITY: Whenever irradiated fuel assemblies are in the storage pool. ACTION:
- 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 hours.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.9.11 The water level in the storage pool shall be determined to be at least its minimum required depth at least once per 7 days when irradiated fuel assemblies are in the fuel storage pool. l l
)
! -\ BRAIDWOOD - UNITS 1 & 2 3/4 9-13
REFUELING OPERATIONS 3/4.9.12 FUEL HANDLING BUILDING EXHAUST FILTER PLENUMS AUG 191966 LIMITING CONDITION FOR OPERATION 3.9.12 Two independent Fuel Handling Building Exhaust Filter Plenums shall be OPERABLE. APPLICABILITY: Whenever irradiated fuel is in the storage pool ACTION:
- a. With one Fuel Handling Building Exhaust Filter Plenum inoperable, fuel movement within the storage pool, or crane operation with loads over the storage pool, may proceed provided the OPERABLE Fuel Handling Building Exhaust Filter Plenum is capable of being powered from an OPERABLE emergency power source and is in operation and taking suction from at least one train of HEPA filters and charcoal adsorbers.
- b. With no Fuel Handling Building Exhaust Filter Plenums OPERABLE, suspend all operations involving movement of fuel within the storage pool, or crane operation with loads over the storage pool, until at least one Fuel Handling Building Exhaust Filter Plenum is restored to OPERABLE status.
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.9.12 The above required Fuel Handling Building Exhaust Filter Plenums shall be demonstrated OPERABLE:
- a. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the system operates for at least 15 minutes;
- b. At least once per'18 months, or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone communicating with the system, by:
BRAIDWOOD - UNITS 1 & 2 3/4 9-14
1 I REFUELING OPERATIONS NA D ' AUG 19 M SURVEILLANCE REQUIREMENTS (Continued)
- 1) Verifying that the Fuel Handling Building Exhaust Filter Plenum satisfies the in place penetration testing acceptance criteria of less than 1% when using the test procedure guidance in Regula-tory Positions C.S.a, C.S.c, and C.S.d of Regulatory Guide 1.52, Revision 2, March 1978, and the flow rate is 21,000 cfm i 10%;
- 2) Ycrifying, 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, Revi-sion 2, March 1978, and by showing a methyl iodide penetration of less than 10% when tested at a temperature of 30*C and a relative l humidity of 95%.
- 3) Verifying a flow rate of 21,000 cfm i 10% through the Fuel Handling Building Exhaust Filter Plenum during operation when tested in accordance with ANSI N510-1980; and
- 4) Verifying that with the system operating at a flow rate of 21,000 .cfm 110% and exhausting through the HEPA filters and charcoal adsorbers, the total bypass flow of the sy' stem and the leakage is less than or equal to 1% when the system is tested by injecting cold DOP at the system intake and the damper leakage rate is determined by either direct measurements or pressure decay measurements at a test pressure of 2 inches of water and the auxiliary building exhaust fans are operating at their rated flow.
- c. After every 720 hours of charcoal adsorber operation, by verifying, within 31 days after removal, that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, by showing a methyl iodide penetration of less than 10% when tested at a temperature of 30*C and a relative humidity of 95%.
- 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 exhaust filter plenum at a flow rate of 21,000 cfm i 10%;
- 2) Verifying that on a Safety Injection or a High Radiation test signal, the system automatically starts (unless already operating) and directs its exhaust flow through the HEPA filters and charcoal adsorber banks; and BRAIDWOOD - UNITS 1 & 2 3/4 9-15
REFUELING OPERATIONS FINAL DRAFT SURVEILLANCE REQUIREMENTS (Continued)
- 3) Verifying that the Fuel Handling Building Exhaust Filter Plenum maintains the fuel building at a negative pressure of greater than or equal to 1/4 inch Water Gauge relative to the outside atmosphere during operation.
- e. After each complete or partial replacement of a HEPA filter bank, by verifying that the Fuel Handling Building Exhaust Filter Plenum satis-fics the in place penetration testing acceptance criteria of less than 1% in accordance with ANSI N510-1980 for a 00P test aerosol while operating the system at a flow rate of 21,000 cfm i 10%; and
- f. After each complete or partial replacement of a charcoal adsorber bank, by verifying that the Fuel Handling Building Exhaust Filter Plenum satisfies the in place penetration testing acceptance criteria of less than 1% in accordance with ANSI N510-1980 for a halogenated hydrocarbon refrigerant test gas while operating the system at a flow rate of 21,000 cfm i 10%.
1 BRAIDWOOD - UNITS 1 & 2 3/4 9-16
FINAL DRAFT 3/4.10 SPECIAL TEST EXCEPTIONS AUG 191986 3/4.10.1 SHUTDOWN MARGIN LIMITING CONDITION FOR OPERATION __ 3.10.1 The SHUTOOWN MARGIN aquirement of Specification 3.1.1.1 may be suspended for measurement o' control rod worth and SHUTDOWN MARGIN provided reactivity equivalent to at least the highest estimated control rod worth is available for trip insertion from OPERABLE control rod (s). APPLICABILITY: MODE 2. ACTION:
- a. With cny full-length control rod not fully inserted and with less than the above reactivity equivalent available for trip insertion, immediately initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm boron or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.
- b. With all full-length control rods fully inserted and the reactor subcritical by less than the above reactivity equivalent, immediately initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm boron or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.
SURVEILLANCE REQUIREMENTS 4.10.1.1 The position of each full-length control rod either partially or fully withdrawn shall be determined at least once per 2 hours. i 4.10.1.2 Each full-length control rod not fully inserted shall be demoristrated capable of full insertion when tripped from at least the 50% withdrawn position within 24 hours prior to reducing the SHUTDOWN MARGIN to less than the limits of Specification 3.1.1.1. BRAIDWOOD - UNITS 1 & 2 3/4 10-1 1
FINAL HAFT 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITfijg ig 99gg LIMITING CONDITION FOR OPERATION 1 1 3.10.2 The group height, insertion, and power distribution limits of Specifications 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.1, and 3.2.4 may be suspended during the performance of PHYSICS TESTS provided:
- a. The THERMAL POWER is maintained less than or equal to 85% of RATED THERMAL POWER, and
- b. The limits of Specifications 3.2.2 and 3.2.3 are maintained and determined at the frequencies specified in Specification 4.10.2.2, below.
APPLICABILITY: MODE 1. ACTION: With any of the limits of Specifications 3.2.2 or 3.2.3 being exceeded while the reqairements of Specifications 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.1, and 3.2.4 are suspended, either: l
- a. Reduce THERMAL POWER sufficient to satisfy the ACTION requirements of Specifications 3.2.2 and 3.2.3, or
- b. Be in HOT STANDBY within 6 hours.
SURVEILLANCE REQUIREMENTS 4.10.2.1 The THERMAL POWER shall be determined to be less than or equal to 85% of RATED THERMAL POWER at least once per hour during PHYSICS TESTS. 4.10.2.2 The requirements of the below listed specfications shall be performed at least once per 12 hours during PHYSICS TESTS:
- a. Specifications 4.2.2.2 and 4.2.2.3, and
- b. Specification 4.2.3.2.
BRAIDWOOD - UNITS 1 & 2 3/4 10-2
~
SPECIAL TEST EXCEPTIONS FINAL DRAFT g yg 3/4.10.3 PHYSICS TESTS LIMITING CONDITION FOR OPERATION 3.10.3 The limitations of Specifications 3.1.1.3, 3.1.1.4, 3.1.3.1, 3.1.3.5, and 3.1.3.6, may be suspended during the performance of PHYSICS TESTS provided:
- a. The THERMAL POWER does not exceed 5% of RATED THERMAL POWER,
- b. The Reactor Trip Setpoints on the OPERABLE Intermediate and Power Range channels are set at less than or equal to 25% of RATED THERMAL POWER, and c.
The Reactor is greater Coolant than System or equal lowest operating loop temperature (Tavg) to 530*F. APPLICABILITY: MODE 2. ACTION:
- a. With the THERMAL POWER greater than 5% of RATED THERMAL POWER, immediately open the Reactor trip breakers.
b. With a Reactor Coolant System operating loop temperature (T,yg) less than 530*F, restore T,yg to within its limit within 15 minutes or be in at least HOT STANDBY within the next 15 minutes. SURVEILLANCE REQUIREMENTS 4.10.3.1 The THERMAL POWER shall be determined to be less than or equal to 5% of RATED THERMAL POWER at least once per hour during PHYSICS TESTS. 4.10.3.2 Each Intermediate and Power Range channel shall be subjected to an ANALOG CHANNEL OPERATIONAL TEST within 12 hours prior to initiating PHYSICS TESTS. 4.10.3.3 The Reactor Coolant System temperature (T,yg) shall be determined to be greater than or equal to 530*F at least once per 30 minutes during PHYSICS TESTS. BRAIDWOOD - UNITS 1 & 2 3/4 10-3 l
1 l
~
SPECIAL TEST EXCEPTIONS FINAL DRAFT 3/4.10.4 REACTOR COOLANT LOOPS MG .I S @ LIMITING CONDITION FOR OPERATION 3.10.4 The limitations of the following requirements may be suspended:
- a. Specification 3.4.1.1 - During the performance of startup and PHYSICS TESTS in MODE 1 or 2 provided:
- 1) The THERMAL POWER does not exceed the P-7 Interlock Setpoint, and l 2) The Reactor Trip Setpoints on the OPERABLE Intermediate and Power l Range channels are set less than or equal to 25% of RATED THERMAL POWER.
- b. Specification 3.4.1.2 - During the performance of hot rod drop time measurements in MODE 3 provided at least two reactor coolant loops as listed in Specification 3.4.1.2 are OPERABLE..
APPLICABILITY: During operation below the P-7 Interlock Setpoint or performance of hot rod drop time measurements. ACTION:
- a. With the THERMAL POWER greater than the P-7 Interlock Setpoint during the performance of startup and PHYSICS TESTS, immediately open the Reactor trip breakers.
- b. With less than the above required reactor coolant loops OPERABLE dJring performance of hot rod drop time measurements, immediately open the reactor trip breakers and comply with the provisions of the ACTION statements of Specification 3.4.1.2.
SURVEILLANCE REQUIREMENTS I 4.10.4.1 The THERMAL POWER shall be determined to be less than P-7 Interlock Setpoint at least once per hour during startup and PHYSICS TESTS. l l 4.10.4.2 Each Ir.termediate and Power Range channel, and P-7 Interlock shall be subjected to an ANALOG CHANNEL OPERATIONAL TEST within 12 hours prior to initiating startup and PHYSICS TESTS. 4.10.4.3 At least the above required reactor coolant loops shall be determined OPERABLE within 4 hours prior to initiation of the hot rod drop time measure-cents and at least once per 4 hours during the hot rod drop time measurements by verifying correct breaker alignments and indicated power availability and l by verifying secondary side narrow range water level to be greater than or equal to 41% for Unit 1 (18% for Unit 2). I BRAIDWOOD - UNITS 1 & 2 3/4 10-4
FINAL DEFT SPECIAL TEST EXCEPTIONS AUG 1 g y \ 3/4.10.5 POSITION INDICATION SYSTEM - SHUTDOWN LIMITING CONDITION FOR OPERATION i 3.10.5 The limitations of Specification 3.1.3.3 may be suspended during the performance of individual full-length shutdown and control rod drop time measurements provided only one shutdown or control bank is withdrawn from the fully inserted position at a time. APPLICABILITY: MODES 3, 4, and 5 during performance of rod drop time measurements and during surveillance of digital rod position indicators for OPERABILITY. ACTION: With the Position Indication System inoperable and with more than one bank of rods withdrawn, immediately open the Reactor trip breakers. SURVEILLANCE REQUIREMENTS 4.10.5 The above required Position Indication Systems shall be determined to be OPERABLE within 24 hours prior to the start of and at least. once per 24 hours thereafter during rod drop time measurements by verifying the Demand Position Indication System and the Digital Rod Position Indication System agree:
- a. Within 12 steps when the rods are stationary, and
- b. Within 24 steps during rod motion.
t BRAIDWOOD - UNITS 1 & 2 3/4 10-5
6 3/4.11 RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS AUG 1 g 19gg l l CONCENTRATION ' LIMITING CONDITION FOR OPERATION 3.11.1.1 The concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS (see Figure 5.1-1) shall be limited to the concentrations specified in 10 CFR Part 20, Appendix B, Table II, Column 2, for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained r' noble gases, the concentration shall be limited to 2 x 10 4 microcurie /ml total activity. APPLICABILITY: At all times. ACTION:
- a. With the concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS exceeding the above limits, immediately restore the concentration to within the above limits.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.1.1.1 Radioactive liquid wastes shall be sampled and analyzed according to the sampling and analysis program of Table 4.11-1. i 4.11.1.1.2 The results of the radioactivity analysis shall be used in accord-l ance with the methodology and parameters in the ODCM to assure that the con-centraT..ans at the point of release are maintained within the limits'of Specification 3.11.1.1. l BRAIDWOOD - UNITS 1 & 2 3/4 11-1
TABLE 4.11-1 RNAL DRAFT AUG 191986 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM LOWER LIMIT MINIMUM OF DETECTION LIQUID RELEASE SAMPLING ANALYSIS TYPE OF ACTIVITY (LLD)(1) TYPE FREQUENCY FREQUENCY ANALYSIS (pCi/ml)
- 1. Batch P P Release Each Batch Each Batch Principal Gamma 5x10
-7 Tanks (2) Emitters (3)
I-131 1x10
-6 P M Dissolved and 1x10
-5 One Batch /M Entrained Gases (Gamma Emitters)
P M H-3 1x10
-5 Each Batch Composite (4)
Gross Alpha ' _7 1x10 P Sr-89, Sr-90 5x10
-8 Q
Each Batch Composite (4) Fe-55 1x10
-6
- 2. Continuous W Principal Gamma 5x10
-7 Continuous (6) Composite (6)
Releases (5) Emitters (3) l Circulating I-131 h10 Water Blowdown M M Dissolved and 1x10
-5 Grab Sample Entrained Gases (Gamma Emitters)
M H-3 1x10
-5 l
Continuous (6) Composite (6)
~7 Gross Alpha 1x10 Sr-89, Sr-90 5x10
-8 Q
Continuous (6) Composite (6) Fe-55 .1x10
-6 l
BRAIDWOOD - UNITS 1 & 2 3/4 11-2
FINAL DRAFT TABLE 4.11-1 (Continued) bil010 I9E RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM LOWER LIMIT MINIMUM OF DETECTION LIQUID RELEASE SAMPLING ANALYSIS TYPE OF ACTIVITY TYPE (LLD)(1) FREQUENCY FREQUENCY ANALYSIS (pCi/ml)
- 3. Continuous W C7) W C7) Principal Gamma 5x10 7 i Release (5) Grab Sample Emitters (3)
Essential Service Water Reactor Coolant Fan Cooler (RCFC) l Outlet Line I-131 1x10 8 Dissolved and Entrained Gases (Gamma Emitters) 1x10 s H-3 1x10 s BRAIDWOOD - UNITS 1 & 2 3/4 11-3
FM DRMI TABLE 4.11-1 (Continued) WG 19 lggg TABLE NOTATIONS (1) The LLD is defined, for purposes of these specifications, as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a "real" signal. ~ For a particular measurement system, which may include radiochemical separation: 4' $ D LLD = E V 2.22 x 108 Y exp (-Aat) Where: LLD = the lower limit of detection (microCuries per unit mass or volume), sb = the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (counts per minute), E = the counting efficiency (counts per disintegration), V = the sample size (units of mass or volume), 2.22 x 10s = the number of disintegrations per minute per microcurie, Y = the fractional radiochemical yield, when applicable, A = the radioactive decay constant for the particular radionuclide (sec 1), and at = the elapsed time between the midpoint of sample collection and the time of counting (sec). Typical values of E, V, Y, and at should be used in the calculation. It should be recognized that the LLD is defined as a before the fact limit representing the capability of a measurement system and not as an after the fact limit for a particular measurement. (2) A batch release is the discharge of liquid wastes of a discrete volume. Prior to sampling for analyses, each batch shall be isolated, and then thoroughly mixed by a method described in the ODCM to assure representative sampling. BRAIDWOOD - UNITS 1 & 2 3/4 11-4
FINAL. DNFT TABLE 4.11-1 (Continued) AUG 1gtggg TABLE NOTATIONS (Continued) (3) The principal gamma emitters for which the LLD specification applies include the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144. This list does not mean that only these nuclides are to be considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Semiannual Radioactive Effluent Release Report pursuant to Specification 6.9.1.7 in the format outlined in Regulatory Guide 1.21, Appendix B, Revision 1, June 1974. (4) A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen that is representative of the liquids released. (5) A continuous release is the discharge of liquid wastes of a nondiscrete volume, e.g., from a volume of a system that has an input flow during the continuous release. (6) To be representative of the quantities and concentrations of radioactive matecials in liquid effluents, samples shall be collected continuously in proportion to the rate of flow of the effluent stream. Prior to analyses, l all samples taken for the composite shall be thoroughly mixed in order for the composite sample to be representative of the effluent release. (7) Not required unless the Essential Service Water RCFC Outlet Radiation Monitors RE-PR002 and RE-PR003 indicates measured levels greater than 1 x 10 8 pCi/ml above background at any time during the week. l l BRAIDWOOD - UNITS 1 & 2 3/4 11-5
RADI0 ACTIVE EFFLUENTS E EE15 th w AUG 11 B6# . DOSE LIMITING CONDITION FOR OPERATION 3.11.1.2 The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released, from each unit, to UNRESTRICTED AREAS (see Figure 5.1-1) shall be limited:
- a. During any calendar quarter to less than or equal to 1.5 mrems to the whole body and to less than or equal to 5 mrems to any organ, and
! b. During any calendar year to less than or equal to 3 mrems to the whole body and to less than or equal to 10 mrems to any organ. APPLICABILITY: At all times. ACTION:
- a. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specifica-tion 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit (s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.1.2 Cumulative dose contributions from liquid effluents for the current calendar quarter and the current calendar year shall be determined in accordance with the methodology and parameters in the ODCM at least once per 31 days. BRAIDWOOD - UNITS 1 & 2 3/4 11-6 l l
RADI0 ACTIVE EFFLUENTS AUG 191918 LIQUID RADWASTE TREATMENT SYSTEM ' LIMITING CONDITION FOR OPERATION 3.11.1.3 The Liquid Radwaste Treatment System shall be OPERABLE and appropriate portions of the system shall be used to reduce releases of radioactivity when the projected doses due to the liquid effluent, from each unit, to UNRESTRICTED AREAS (see Figure 5.1-1) wo^uld exceed 0.06 mrem to the whole. body or 0.2 mrem to any organ in a 31-day period. APPLICABILITY: At all times. ACTION:
- a. With radioactive liquid waste being discharged without treatment and in excess of the at,0ve limits and any portion of the Liquid Radwaste Treatment System not in operation, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that includes the following information:
- 1. Explanation of why liquid radwaste was being discharged without treatment, identification of any inoperable equipment or subsystems, and the reason for the,inoperability,
- 2. Action (s) taken to restore the inoperable equipment to OPERABLE status, and
- 3. Summary description of action (s) taken to prevent a recurrence.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS l 4.11.1.3.1 Doses due to liquid releases from each unit to UNRESTRICTED AREAS shall be projected at least once per 31 days in accordance with the methodology and parameters in the ODCM when the Liquid Radwaste Treatment System is not being fully utilized. 4.11.1.3.2 The installed Liquid Radwaste Treatment System shall be considered OPERABLE by meeting Specifications 3.11.1.1 and 3.11.1.2. BRAIDWOOD - UNITS 1 & 2 3/4 11-7
RADI0 ACTIVE EFFLUENTS
. AUG 191986 LIQUID HOLDUP TANKS LIMITING CONDITION FOR OPERATION 3.11.1.4 The quantity of radioactive material, excluding tritium and dissolved
.s or entrained noble gases, contained in any outside tanks shall be limited to the following:
- a. Primary Water Storage Tank 's 2000 Curies, and
- b. Outside Temporary Tank 5 10 Curies.
APPLICABILITY: At all times. ACTION:
- a. With the quantity of radioactive material in any of the above listed tanks exceeding the above limit, immediately suspend all additions of radioactive material to the tank, within 48 hours reduce the tank contents to within .the limit, and describe the events leading to '
this condition in the next Semiannual Radioactive Effluent Release Report, pursuant to Specification 6.9.1.7.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.1.4 The quantity of radioactive material contained in each of the above tanks shall be determined to be within the above limit by analyzing a representative sample of the tank's contents at least once per 7 days when radioactive materials are being added to the tank. BRAIDWOOD - UNITS 1 & 2 3/4 11-8
RADIOACTIVE EFFLUENTS FINAL DRAFT
~
3/4.11.2 GASEOUS EFFLUENTS 10 E DOSE RATE LIMITING CONDITION FOR OPERATION 3.11.2.1 The dose rate due to radioactive materials released in gaseous effluents from the' site to areas at and beyond the SITE B0UNDARY (see Figure 5.1-1) shall be limited to the following:
- a. For noble gases: Less than or equal to 500 mrems/yr to the whole body and less than or equal to 3000 mrems/yr to the skin, and
- b. For Iodine-131 and 133, for tritium, and for all radionuclides in particulate form with half-lives greater than 8 days: Less than or equal to 1500 mrems/yr to any organ.
APPLICABILITY: At all times. - ACTION:
- a. With the dose rate (s) exceeding the above limits, immediately restore the release rate to within the above limit (s).
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.2.1.1 The dose rate due to noble gases in gaseous effluents shall be determined to be within the above limits in accordance with the methodology and parameters in the ODCM. 4.11.2.1.2 The dose rate due to Iodine-131 and 133, tritium, and all radio-nuclides in particulate form with half-lives greater than 8 days in gaseous effluents shall be determined to be within the above limits in accordance with the methodology and parameters in the ODCM by obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified in Table 4.11-2. l l l BRAIDWOOD - UNITS 1 & 2 3/4 11-9 l i
TABLE 4.11-2 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM o 5 MINIMUM LOWER LIMIT OF SAMPLING ANALYSIS TYPE OF DETECTION (LLD)(1) E GASEOUS RELEASE TYPE FREQUENCY FREQUENCY ACTIVITY ANALYSIS (pCi/ml) Principal Gamma Emitters (2) ~4
- 1. Waste Gas Decay Each Tank Each Tank 1x10
- Tank Grab Sample Principal Gamma Emitters (2) ~4
- 2. Containment Purge Each PURGE (3) Each PURGE (3) 1x10 Grab Sample
~7 H-3 1x10 Principal Gamma Emitters (2) ~4
- 3. Auxiliary Bldg M(4)(5) 1x10 Vent Stack Grab Sample M
-7 R (Units 1 and 2) H-3 1x10 Continuous (6) y(7) I-131 1x10
-12
[ Charcoal Sample I-133 1x10
-10 g(7) Principal Gamma Emitters (2) ~11 Continuous (6) 1x10 Particulate Sample Continuous (6) M Gross Alpha 1x10-11, Composite Particulate Sample
-11 g
Continuous (6) 9 Sr-89, Sr-90 1x10 Composite Particulate Sample Continuous Noble Gas Monitor Noble Gases Gross Beta or ~ 1x10
-6 g
, Gamma M
tD E
TABLE 4.11-2 (Continued) TABLE NOTATIONS (1) The LLD is defined, for purposes of these specifications, as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a "real" signal. For a particular measurement system, which may include radiochemical separation: 4 s b LLD = E V 2.22 x 10s y . exp (-Mt) Where: LLD = the lower limit of detection (microcuries per unit mass or volume), sb = the standard deviation of the background counting rate or of the counting rate of a blanli sample as appropriate (counts per minute), E = the counting efficiency (counts per disintegration), V = the sample size (units of mass or volume), 2.22 x 10e = the number of disintegrations per minute per microcurie, Y = the fractional radiochemical yield, when applicable, A = the radioactive decay constant for the particular radionuclide (sec 1), and at = the elapsed time between the midpoint of sample collection and the time of counting (sec). Typical values of E, V, Y, and at should be used in the calculation. It should be recognized that the LLD is defined as a before the fact limit representing the capability of a measurement system and not as an after the fact limit for a particular measurement. BRAIDWOOD - UNITS 1 & 2 3/4 11-11 l
TABLE 4.11-2 (Continued) TABLE NOTATIONS (Continued) g l (2) The principal gamma emitters for which the LLD specification applies 1 include the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 in noble gas releases and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Ho-99, I-131, Cs-134, Cs-137, Ce-141, and Ce-144 in - iodine and particulate releases. This list does not mean that only these nuclides are to be considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Semiannual Radioactive Effluent Release Report pursuant to Specification 6.9.1.7, in the format outlined in Regulatory Guide 1.21, Appendix 8, Revision 1, June 1974. (3) Sampling and analysis shall also be performed following shutdown, startup, or a THERMAL POWER change exceeding 15% of RATED THERMAL POWER within a 1-hour period. (4) Tritium grab samples shall be taken at least once per 24 hours when the refueling canal is flooded. (5) Tritium grab samples shall be taken at least once per 7 days from the spent fuel pool area, whenever spent fuel is in the spent fuel pool. (6) The ratio of the sample flow rate to the sampled stream flow rate shall be.known for the time period covered by each dose or dose rate calculation made in accordance with Specifications 3.11.2.1, 3.11.2.2, and 3.11.2.3. (7) Samples shall be changed at least once per 7 days and analyses shall be completed within 48 hours after changing, or after removal from sampler. Sampling shall also be performed at least once per 24 hours for at least 7 days following each shutdown, startup or THERMAL POWER change exceeding 15% of RATED THERMAL PCWER within a 1-hour period and analyses shall be completed within 48 hours of changing. When sample's collected for 24 hours are analyzed, the corresponding LLDs may be increased by a factor of 10. This requirement does not apply if: (1) analysis shows that the DOSE EQUIVALENT I-131 concentration in the reactor coolant has not increased more than a factor of 3, and (2) the noble gas monitor shows that effluent activity has not increased more than a factor of 3. BRAIDWOOD - UNITS 1 & 2 3/4 11-12
RADIOACTIVE EFFLUENTS DOSE - NOBLE GASES SN78 I S fogy LIMITING CONDITION FOR OPERATION 3.11.2.2 The air dose due to noble gases released in gaseous effluents, from each unit, to areas at and beyond the SITE BOUNDARY (see Figure 5.1-1) shall be limited to the following:
- a. During any calendar quarter: Less than or equal to 5 mrads for gamma radiation and less than or equal to 10 mrads for beta radiation, and
- b. During any calendar year: Less than or equal to 10 mrads for gamma radiation and less than or equal to 20 mrads for beta radiation.
APPLICABILITY: At all times. ACTION:
- a. With the calculatea air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit (s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.2.2 Cumulative dose contributions for the current calendar quarter and current calendar year for noble gases shall be determined in accordance with the methodology and parameters in the ODCM at least once per 31 days. G l BRAIDWOOD - UNITS 1 & 2 3/4 11-13
RADI0 ACTIVE EFFLUENTS FINAL DRAFT ~ AUG 19 g5 DOSE - 10 DINE-131 AND 133, TRITIUM, AND RADI0 ACTIVE MATERIAL IN PARTICULATE FORM LIMITING CONDITION FOR OPERATION 3.11.2.3 The dose to a MEMBER OF THE PUBLIC from Iodine-131 and 133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released, from each unit, to areas at and beyond the SITE BOUNDARY (see Figure 5.1-1) shall be limited to the following:
- a. During any calendar quarter: Less than or equal to 7.5 mrems to any organ, and
- b. During any calendar year: Less than or equal to 15 mrems to any organ.
APPLICABILITY: At all times. . ACTION:
- a. With the calculated dose from the release of Iodine-131 and 133, tritium, and radionuclides in particulate form with half-lives greater than 8 days, in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limits and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.2.3 Cumulative dose contributions for the current calendar quarter and current calendar year for Iodine-131 and 133, tritium, and radionuclides in particulate form with half-lives greater than 8 days shall be determined in accordance with the methodology and parameters in the ODCM at least once per 31 days. BRAIDWOOD - UNITS 1 & 2 3/4 11-14
RADIOACTIVE EFFLUENTS AUG 19 gggg GASEOUS RADWASTE TREATMENT SYSTEM LIMITING CONDITION FOR OPERATION 3.11.2.4 The VENTILATION EXHAUST TF.EATMENT SYSTEM and the WASTE GAS HOLDUP SYSTEM shall be OPERABLE and appropriate portions of these systems shall be used to reduce releases of radioactivity when the projected doses in 31 days due to gaseous effluent releases, from each unit, to areas at and beyond the SITE BOUNDARY (see Figure 5.1-1) would exceed:
- a. 0.2 mrad to air from gamma radiation, or
- b. 0.4 mrad to air from beta radiation, or
- c. 0.3 mrem to any organ of a MEMBER OF THE PUBLIC.
APPLICABILITY: At a11 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
- 3. Summary description of action (s) taken to prevent a recurrence.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.2.4.1 Doses due to gaseous releases from each unit to areas at and beyond the SITE BOUNDARY shall be projected at least once per 31 days in accordance with the methodology and parameters in the ODCM when Gaseous Radwaste Treatment Systems are not-being fully utilized. 4.11.2.4.2 The installed VENTILATION EXHAUST TREATMENT SYSTEM and WASTE GAS HOLDUP SYSTEM shall be considered OPERABLE by meeting Specification 3.'11.2.1 and 3.11.2.2 or 3.11.2.3. i BRAIDWOOD - UNITS 1 & 2 3/4 11-15
RADI0 ACTIVE EFFLUENTS EXPLOSIVE GAS MIXTURE D 191986 LIMITING CONDITION FOR OPERATION 3.11.2.5 The concentration of oxygen in the WASTE GAS HOLDUP SYSTEM shall be limited to less than or equal to 2% by volume whenever the hydrogen concentration exceeds 4% by volume. APPLICABILITY: At all times. ACTION,:
- a. With the concentration ef oxygen in the WASTE GAS HOLDUP SYSTEM greeter than 2% by volume but less than or equal to 4% by volume, reduce the oxygen concentration to the above limits within 48 hours.
~
- b. With the concentration of oxygen in the WASTE GAS HOLDUP SYSTEM greater than 4% by volume and the hydrogen concentration greater than 4% by volume, immediately suspend all additions of waste gases to the system and reduce the concentration of oxygen to less than or equal to 4%.by volume, then take ACTION a. above.
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
l I SURVEILLANCE REQUIREMENTS 4.11.2.5 The concentrations of hydrogen and oxygen in the WASTE GAS HOLDUP SYSTEM shall be determined to be within the above limits by continuously monitoring the waste gases in the WASTE GAS HOLDUP SYSTEM with the hydrogen and cxygen monitors required OPERABLE by Table 3.3-13 of Specification 3.3.3.11. l l ? . BRAIDWOOD - UNITS 1 & 2 3/4 11-16 i
RADIOACTIVE EFFLUENTS FINAL DRAFT
, AUG 191986 GAS DECAY TANKS LIMIT,ING CONDITION FOR OPERATION
. i 3.11l2.6 The quantity of radioactivity contained in each gas decay tank shall be limited to less than or equal to 5x104 Curies of noble gases (considered as Xe-133 equivalent).
APPLICABILITY: At all times. ACTION: l a. With the quantity of radioactive material in any gas decay tank exceeding the above limit, immediately suspend all additions of radioactive material to the tank and, within 48 hours, reduce the tank contents to within the limit, and describe the events leading to this condition in the next Semiannual Radioactive Effluent Release Report, pursuant to Specification 6.9.1.7.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.2.5 The quantity of radioactiva material contained in each gas decay tank shall be determined to be within the above limit at least once per 24 hours when radioactive materials are being added to the tank. BRAIDWOOD - UNITS 1 & 2 3/4 11-17
RADI0 ACTIVE EFFLUhNTS i 3/4.11.3 SOLID RADI0 ACTIVE WASTES' "S 19 IS00 LIMITING CONDITION FOR OPERATION 3.11.3 Radioactive wastes shall be solidified or dewatered in accordance with the PROCESS CONTROL PROGRAM to meet shipping and transportation requ'irements during transit, and disposal site requirements when received at the disposal site. APPLICABILITY: At all times. ACTION:
- a. With SOLIDIFICATION or dewatering not meeting disposal site and shipping and transportation requirements, suspend shipment of the inadequately processed wastes and correct the PROCESS CONTROL PROGRAM, the procedures and/or the solid waste system as necessary to prevent recurrence,
- b. With SOLIDIFICATION or dewatering not performed in accordance with the PROCESS CONTROL PROGRAM, test the improperly processed waste in each container to ensure that it meets burial ground and shipping requirements and take appropriate administrative action to preven,t recurrence.
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. '
SURVEILLANCE REQUIREMENTS 4.11.3 SOLIDIFICATION of at least one representative test specimen from at least every tenth batch of each type of wet radioactive wastes (e.g., filter sludges, spent resins, evaporator bottoms, boric acid solutions and sodium sulfate solutions) shall be verified in accordance with the PROCESS CONTROL PROGRAM:
- a. If any test specimen fails to verify SOLIDIFICATION, the SOLIDIFICATION of the batch under test shall be suspended until such time as additional test specimens can be obtained, alternative SOLIDIFICATION parameters can be determined in accordance with the PROCESS CONTROL PROGRAM, and a subsequent test verifies SOLIDIFICATION. SOLIDIFICATION of the batch may then be resumed using the alternative SOLIDIFICATION parameters determined by the PROCESS CONTROL PROGRAM;
- b. If the initial test specimen from a batch of waste fails to verify SOLIDIFICATION, the PROCESS CONTROL PROGRAM shall provide for the collection and testing of representative test specimens from each consecutive batch of the same type of wet waste until at least 3 consecutive initial test specimens demonstrate SOLIDIFICATION. The PROCESS CONTROL PROGRAM shall be modified as required, as provided in Specification 6.13, to assure SOLIDIFICATION of subsequent batches of waste; snd
- c. With the installed equipment incapable of meeting Specification 3.11.3 or declared out-of-service, restore the equipment to operable status or provide for contract capability to process wastes as necessary to satisfy all applicable transportation and disposal requirements.
BRAIDWOOD - UNITS 1 & 2 3/4 11-18
RADI0 ACTIVE EFFLUENTS 3/4.11.4 TOTAL DOSE LIMITING CONDITION FOR OPERATION 3.11.4 The annual (calendar year) dose or dose commitment to any ME'BER M OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources shall be limited to less than or equal to 25 mrems to the whole body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrems. APPLICABILITY: At all times. ACTION:
- a. With the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Specification 3.11.1.2a., 3.11.1.2b., 3.11.2.2a., 3.11.2.2b.,
3.11.2.3a., or 3.11.2.3b., calculations should be made including direct radiation contributions from the units and from outside storage tanks to determine whether the above limits of Specification 3.11.4 have been exceeded. If such 1s the case, prepare and submit to the Commission within 30 days, pur,suant to Specification 6.9.2, a Special Report that defines the corrective action to be taken to reduce sub-sequent releases to prevent recurrence of exceeding the above limits and includes the schedule for achieving conformance with the above limits. This Special Report, as defined in 10 CFR 20.405c, shall include an analysis that estimates the radiation exposure (dose) to a MEMBER OF THE PUBLIC from uranium fuel cycle sources, including all effluent pathways and direct radiation, for the calendar year that includes the release (s) covered by this report. It shall also describe levels of radiation and concentrations of radioactive material involved, and the cause of the exposure levels or concen-trations. If the estimated dose (s) exceeds the above limits, and if the release condition resulting in violation of 40 CFR Part 190 has not already been corrected, the Special Report shall include a request for a variance in accordance with the provisions of 40 CFR Part 190. Submittal of the report is considered a timely request, and a variance is granted until staff action on the request l is complete.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.4.1 Cumulative dose contributions from liquid and gaseous effluents ! shall be determined in accordance with Specifications 4.11.1.2, 4.11.2.2, and l 4.11.2.3, and in accordance with the methodology and parameters in the ODCM. l 4.11.4.2 Cumulative dose contributions from direct radiation from the units and from radwaste storage tanks shall be determined in accordance with the methodology and parameters in the ODCH.
~
This requirement is applicable only under conditions set forth in ACTION a. of Specification 3.11.4. BRAIDWOOD - UNITS 1 & 2 3/4 11-19
3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM . LIMITING CONDITION FOR OPERATION 3.12.1 The Radiological Environmental Monitoring Program shall be conducted as specified in Table 3.12-1. . APPLICABILITY: At all times. ACTION:
- a. With the Radiological Environmental Monitoring Program not being conducted as specified in Table 3.12-1, prepare and submit to the Commission, in the Annual Radiological Environmental Operating Report required by Specification 6.9.1.6, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence.
- b. With the level of radioactivity as the result of plant effluents in an environmental sampling medium at a specified location exceeding the reporting levels of Table 3.12-2 when averaged over any calendar quarter, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that .idantifies the cause(s) for exceeding the limit (s) and defines the corrective actions to be taken to reduce radioactive effluents so that the potential annual dose
- to a MEMBER OF THE PUBLIC is less than the calendar year limits of Specifications 3.11.1.2, 3.11.2.2, or 3.11.2.3. When more than one of the radionuclides in Table 3.12-2 are detected in the sampling medium, this report shall be submitted if:
concentration (1) , concentration (2) + ...> 1.0 reporting level (1) reporting level (2) - When radionuclides other than those in Table 3.12-2 are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose
- to A MEMBER OF THE PUBLIC from all radio-nuclides is equal to or greater than the calendar year limits of Specifications 3.11.1.2, 3.11.2.2, or 3.11.2.3. This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report required by Specification 6.9.1.6.
"The methodology and parameters used to estimate the potential annual tiose to a MEMBER OF THE PUBLIC shall be indicated in this report. BRAIDWOOD - UNITS 1 & 2 3/4 12-1
RADIOLOGICAL ENVIRONMENTAL MONITORING FINAL HAFT l AUG 19 f986 LIMITING CONDITION FOR OPERATION ~ ACTION (Ccatinued)
- c. With milk or fresh leafy vegetable samples unavailable from one or ;
more of the sample locations required by Table 3.12-1, identify ! specific locations for obtaining replacement samples and add them . within 30 days to the Radiological Environmental Monitoring Program ; given in the ODCM. The specific locations from which samples were unavailable.nay then be deleted from the monitoring program. Submit ; controlled version of the ODCM within 180 days including a revised ' figure (s) and table reflecting the new location (s) with supporting information identifying the cause of the unavailability of samples and justifying the selection of the new location (s) for obtaining samples.
- d. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
1 SURVEILLANCE REQUIREMENTS 4.12.1 The radiological environmental monitoring samples shall be collected pursuant to Table 3.12-1 from the specific locations given in the table and figure (s) in the ODCM, and shall be analyzed pursuant to the requirements of Table 3.12-1 and the detection capabilities required by Table 4.12-1. l l i BRAIDWOOD - UNITS 1 & 2 3/4 12-2
TABLE 3.12-1 E R O RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM k NUE ER OF REPRESENTATIVE EXPOSURE PATHWAY SAMPLES AND SAMPLING AND TYPE AND FREQUENCY E AND/OR SAMPLE SAMPLE LOCATIONS II) COLLECTION FREQUENCY OF ANALYSIS Z
- 1. Direct Radiation (2) Forty routine monitoring stations Quarterly. Gamma dose quarterly.
H
" either with two or more dosimeters or with one instrument for measur-ing and recording dose rate -
continuously, placed as follows: An inner ring of stations, one in each meteorological sector in the general area of the SITE BOUM0ARY; , 5 An outer ring of stations, one in g each meteorological sector in 7 the 6- to 8-km range from the site; and The balance of the stations to be placed in special . interest areas such as popu-i
~
lation centers, nearby residences, schools, and in one or two areas to serve as control stations. 2n - b
. e m
qllll3 3ma M
TABLE 3.12-1 (Continued) y RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM E NUMBER OF E REPRESENTATIVE 8 EXPOSURE PATHWAY SAMPLES AND SAMPLING AND TYPE AND FREQUENCY e AND/0R SAMPLE SAMPLE LOCATIONS (1) COLLECTION FREQUENCY OF ANALYSIS E 2. Airborne d Radioiodine and Samples from,fivo locations: Continuous sampler Radioiodine Cannister: s Particulates operation with sample I-131 analysis weekly.
- e. Three samples from close to collection weekly, or m the three SITE BOUNDARY loca- more frequently if tions, in different sectors, required by dust Particulate Sampler:
of the highest calculated loading. Gross beta radioactivity annual average ground level D/Q; analysisfollogg filter change; and One sample from the vicinity of gamma isotopic analysis (4) s a community having the highest of composite (by location) T
- calculated annual average ground- quarterly.
level D/Q; and One sample from a control-location, as for example 10 to 30 km distant and in the least prevalent wind direction.
- 3. Waterborna
- a. Surface (5) One sample upstream. Composite sample over Gamma isotopic analysis (4)
One sample downstream. 1 month period by monthly. Composite for weekly grab samples. tritium analysis quarterly.
- b. Ground Samples from one or two sourc Quarterly. Gamma isotopic (4) and trit,ium only if likely to be affected analysis quarterly.
- c. Drinking One sample of each community Composite sample I-131 analysis on each drinking water supply within over 2 week period (6) compos,ite when the dose
, 10 miles downstream of the when I-131 analysis calculated for the consump-discharge. is performed, monthly tion of the water is greater composite otherwise. than 1 mrem per year.(8) Com-One sample from a control positeforgrossbetaag gamma isotopic analyses location. monthly. Composite for AUG 191986 tritium analysis quarterly.
TABLE 3.12-1 (Continued) h RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM o NUMBER OF k o REPRESENTATIVE EXPOSURE PATHWAY SAMPLES AND SAMPLING AND TYPE AND FREQUENCY
' AND/OR SAMPLE SAMPLE LOCATIONS ( ) COLLECTION FREQUENCY OF ANALYSIS E 3. Waterborne (Continued) 3 d. Sediment One sample from downstream area Semiannually. Gamma isotopic analysis I4) g from with existing or potential semiannually.
,, shoreline ~ recreational value.
N 4 Ingestion
- a. Milk Samples from milking animals Semimonthly when Gamma isotopicI4) and I-131 in three locations within animals are on analysis semimonthly when 5 km distance having the highest pasture, monthly at animals are on pasture; dose potential. If there are other times. monthly at other times.
none, then, one sample from milking - R animals in each of three areas
- between 5 to 8 km distant g where be J, greaterdoses than 1are calculated mrem per yr (
One sample from milking animals at a control location, 15 to 30 km distant and in the least prevalent wind direction.
- b. Fish and Representative samples of Three times per year Gamma isotopic enslysis(4)
Inverte- commercially and recreationally (spring, summer and on edible portions. brates important species in vicinity fall). of, plant discharge area. Representative samples of commercially and recreationally important species in areas not
- influenced by plant discharge.
- c. Food Representative samples of the At time of harvest.(9) Gamma isotopic analyses I4)
Products principal classes of food products on edible portion. from any area within 10 miles of the plant that is irrigated by , water in which liquid plant AUG 191986 wastes have been discharged.
, TABLE 3.12-1 (Continued)
[ RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM o NUMBER OF REPRESENTATIVE
' EXPOSURE PATHWAY SAMPLES AND SAMPLING AND TYPE AND FREQUENCY AND/0R SAMPLE SAMPLE LOCATIONS II) COLLECTION FREQUENCY OF ANALYSIS
- 4. Ingestion (Continued) s c. Food Samples of three different kinds Monthly whesi Gamma isotopic (4) and I-131
- c. Products of broad leaf vegetation grown available. analysis, to (continued) nearest each of two different .
offsite locations of highest predicted annual average ground-level D/Q if milk sampling is not performed. One sample of each of the similar Monthly when Gamma isotopic (4) and I-131 R
- broad leaf vegetation grown available. analysis.
15 to 30 km distant in the least U prevalent wind direction if milk a sampling is not performed.
, :a:
= 3m=
r==
~
m g es gi A 4
TABLE 3.12-1 (Continued)
. T.ABLE NOTATIONS SUS I g ggp6 (1) Specific parameters of distance and direction sector from the centerline of one unit, and additional description where pertinent, shall be provided for each and every sample location in Table 3.12-1 in a table and figure (s) in the ODCM. Refer to NUREG-0133, " Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants," October 1978, and to Radiological Assessment Branch Technical Position, Revision 1, November 1979. Deviations are permitted from the required sampling schedule if specimens are unobtainable due to hazardous conditions, seasonal unavailability, malfunction of automatic sampling equipment and other legitimate reasons. If specimens are unobtainable due to sampling equip-ment malfunction, every effort shall be made to complete corrective action prior to the end of the next sampling period. All deviations from the sampling schedule shall be documented in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.6. It is recognized that, at times, it may not be possible or practicable to l continue to obtain samples of the media of choice at the most desired location or time. In these instances suitable specific alternative media and locations may be chosen for the particular pathway in question and .
appropriate substitutions made within 30 days in the Radiological Environmental Monitoring Program given in the ODCM. Submit controlled revisions of the ODCM within 180 days including a revised figure (s) and table reflecting the new location (s) with supporting information identifying the cause of the unavailability of samples for that pathway and justifying the selection of the new location (s) for obtaining samples. (2) One or more instruments, such as a pressurized ion chamber, for measuring and recording dose rate continuously may be used in place of, or in addition to, integrating dosimeters. For the purposes of this table, a thermoluminescent dosimeter (TLD) is considered to be one phosphor; two or more phosphors in a packet are considered as two or more dosimeters. Film badges shall not be used as dosimeters for measuring direct radiation. The 40 stations is not an absolute number. The number of direct radiation monitoring stations may be reduced according to geographical limitations; e.g., at an ocean site, some sectors will be over water so that the number of dosimeters may be reduced accordingly. The frequency of analysis or i readout for TLD systems will depend upon the characteristics of the specific system used and should be selected to obtain optimum dose information with minimal fading. (3) Airborne particulate sample filters shall be analyzed for gross beta radio-activity 24 hours or more after sampling to allow for radon and thoron daughter decay. If gross beta activity in air particulate samples is < greater than 10 times the yearly mean of control samples, gamma isotopic l analysis shall be performed on the individual samples. l (4) Gamma isotopic analysis means the identification and quantification of l gamma-emitting radionuclides that may be attributable to the effluents from the facility. , BRAIDWOOD - UNITS 1 & 2 3/4 12-7
TABLE 3.12-1 (Continued) FINAL DRAFT AUG 191986 TABLE NOTATIONS (Continued) (5) The " upstream sample" shall be taken at a distance beyond significant influence of the discharge. The " downstream" sample shall be taken in an area beyond but near the mixing zone. (6) A composite sample is one in which the quantity (aliquot) of liquid sampled is proportional to.the quantity of flowing liquid and in which the method of sampling employed results in a specimen that is representative of the liquid flow. In this program composite sample aliquots shall be collected at time intervals that are very short (e.g. , hourly) relative to the com-positing period (e.g., monthly) in order to assure obtaining a representa-tive sample. (7) Groundwater samples shall be taken when this source is tapped for drinking or irrigation purposes in areas where the hydraulic gradient or recharge prcperties are suitable for contamination. - (8) The dose shall be calculated for the maximum organ and age group, using the methodology and parameters in the CDCM. . (9) If harvest occurs more than once a year, sampling.shall be performed during each discrete harvest. If harvest occurs continuously, sampling shall be monthly. Attention shall be paid to including samples of tuberous and root food products. e 0 BRAIDWOOD - UNITS 1 & 2 3/4 12-8
TABLE 3.12-2 E R O REPORTING LEVELS FOR RADI0 ACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES h o REPORTING LEVELS [ WATER AIRBORNE PARTICULATE FISH MILK FOOD PRODUCTS 3 ANALYSIS (pCf/1) OR GASES (pC1/ma ) (pCi/kg, wet) (pCi/2) (pCi/kg, wet) d H H-3 20,000* e. m Mn-54 1,000 30,000 Fe-59 400 10,000 Co-58 1,000 30,000 m Co-60 300 10,000 s [ Zn-65 300 20,000 m E Zr-Nb-95 400 I-131 2 0.9 3 100 Cs-134 30 10 1,000 60 1,000 Cs-137 50 20 2,000 70 2,000 Ba-La-140 200 300 m-
*For drinking water samples. This is 40 CFR Part 141 value. If no drinking water pathway exists, a value of 30,000 pCi/t may be used. M 2m C
P W xs I
- >Z
TABLE 4.12-1 b DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSISII) LOWER LIMIT OF DECTECTION (LLD)(2)(3) h WATER AIRBORNE PARTICULATE FISH MILK FOOD PRODUCTS SEDIMENT 5 ANAL.YSIS (pci/2) OR GAS (pCi/ma ) (pCi/kg, wet) (pCi/1) (pCi/kg, wet) (pCi/kg, dry) d
- Gross Beta 4 0.01 e.
" H-3 2000*
Mn-54 15 130 Fe-59 30 260 w Co-58,60 15 130 1 g Zn-65 30 260 7 g Zr-Nb- 95 15 I-131 l I4) 0.07 1 60 Cs-134 0.05 15 130 15 60 150 Cs-137 18 0.06 150 18 80 180 Ba-La-140 15 15
]
*If no drinking water pathway exists, a value of 3000 pCi/E may be used. :a=
g N
=
P go I h l 21
TABLE 4.12-1 (Continued) TABLE NOTATIONS AUG I S 1986 (1) This list does not mean that only these nuclides are to be considered. Other peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Annual Radiological i Environmental Operating Report pursuant to Specification 6.9.1.6. ' (2) Required detection capabilities for thermoluminescent dosihieters used for environmental measurements shall be in accordance with the recommendations of Regulatory Guide 4.13. (3) The LLD is defined, for purposes of these specifications, as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a "real" signal. For a particular measurement system, which may include radiochemical separation: 4.66 s b E V + 2.22 Y exp (-Aat) Where: LLD = the "a priori" lower limit of detection (picoCuries per unit mass or volume), s b = the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (counts per minute), E = the counting efficiency (counts per disintegration), V = the sample size (units of mass or volume), 2.22 = the number of disintegrations per minute per picocurie, Y = the fractional radiochemical yield, when applicable, A = the radioactive decay constant for the particular radionuclide, (sec 1), and at = the elapsed time between sample collection, or end of the sample collection period, and time of counting (sec). Typical values of E, V, Y, and At should be used in the calculation. l l l l BRAIDWOOD - UNITS 1 & 2 3/4 12-11
TABLE 4.12-1 (Continued) TABLE NOTATIONS (Continued) RUG 1 $ q It should be recognized that the LLD is defined as a before the fact limit representing the capability of a measurement system and not as an after the fact limit for a particular measurement. Analyses shall be performed in such a manner that t!. stated LLDs will be achieved under routine conditions. Occasionally background fluctuations, unavoidable small sample sizes, the presence of interfering nuclides, or other uncon-tro11able circumstances may render these LLDs unachievable. In such cases, the contributing factors shall be identified and described in the Annual Radiological Environmental Operating Report pursuant to Specifica-tion 6.9.1.6. (4) LLD for drinking water samples. If no drinking water pathway exists, the LLD of gamma isotopic analysis may be used. 9 BRAIDWOOD - UNITS 1 & 2 3/4 12-12
RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.2 LAND USE CENSUS , LIMITING CONDITION FOR OPERATION 3.12.2 A Land Use Census shall be conducted and shall identify within a distance of 8 km (5 miles) the location in each of the 16 meteorological sectors of the nearest milk animal, and the nearest residence. For dose calculation, a garden will be assumed at the nearest residence. APPLICABILITY: At all times. ACTION:
- a. With a Land Use Census identifying a location (s) that yields a calculated dose or dose commitment greater than the values currently being calculated in Specification 4.11.2.3, identify the new location (s) in the next Annual Radiological Environmental Operating Report, pursuant to Specification 6.9.1.6.
- b. With a Land Use Census identifying a location (s) that yields a calculated dose or dose commitment (via the same exposure pathway) 20% greater than at a location from which samples are currently being obtained in accordance with Specification 3.12.1, add the new location (s) within 30 days to the Radiological Environmental Monitoring Program given in the ODCM.
The sampling location (s), excluding the control station location, having the lowest calculated dose or dose commitment (s), via the same exposure pathway, may be deleted from this monitoring program after October 31 of the year in which this Land Use Census was conducted. Pursuant to Specification 6.14, submit in the next Annual Radiological Environmental Operating Report documentation for a change in the ODCM including a revised figure (s) and table (s) for the ODCM reflecting the new location (s) with information supporting the change in sampling locations.
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.12.2 The Land Use Census shall be conducted during the growing season at least once per 12 months using that information that will provide the best results, such as by a door-to-door survey, aerial survey, or by consulting local agriculture authorities. The results of the Land Use Census shall be included in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.6. 6 BRAIDWOOD - UNITS 1 & 2 3/4 12-13
1 RADIOLOGICAL ENVIRONMENTAL MONITORING "~ ~
\
3/4.12.3 INTERLABORATORY COMPARISON PROGRAM l AUG I P 1986 LIMITING CONDITION FOR OPERATION 3.12.3 Analyses shall be performed on radioactive materials, supplied as part of an Interlaboratory Comparison Program that has been approved by the Commission, that correspond to samples required by Table 3.12-1. APPLICABILITY: At all times. ACTION:
- a. With analyses not being performed as required above, report the corrective actions taken to prevent a recurrence to the Commission in the Annual Radiological Environmental Operating Repurt pursuant to Specification 6.9.1.6.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.12.3 The Interlaboratory Comparison Program shall be describ;d in the ODCM. ; A summary of the results obtained as part of the above required Interlaboratory Comparison Program shall be included in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.6. BRAIDWOOD - UNITS 1 & 2 3/4 12-14
A a FINAL DRAFT
. AUG 1919g5 BASES FOR SECTIONS 3.0 AND 4.0 LIMITING CONDITIONS FOR OPERATION AND SURVEILL%NCE REQUIREMENTS
FINAL DRAFT l AUG 191986 i NOTE The BASES contained in succeeding pages summarize the reasons for the Specifications in Sections 3.0 and 4.0, but in accordance with 10 CFR 50.36 are not part of these Technical Specifications. 1 l l
3/4.0 APPLICABILITY . . BASES AUG 191986 The specifications of this section provide the general requirements applicable to each of the Limiting Conditions for an Operation and Surveil-lance Requirements within Section 3/4. In the event of a disagreement between the requirements stated in these Technical Specifications and those stated in an applicable Federal Regulation or Act, the requirements stated in the applicable Federal Regulation or Act shall take precedence and shall.be met. , 3.0.1 This specification defines the applicability of each specification in terms of defined OPERATIONAL MODES or other specified conditions and is provided to delineate specifically when each specification is applicable. 3.0.2 This specification defines those conditions necessary to constitute compliance with the terms of an individual Limiting Condition for Operation and associated ACTION requirement. 3.0.3 The specification delineates the measures to be taken for those l circumstances not directly provided for in tha ACTION statements and whose i occurrence would violate the intent of a specification. For example, Specifi-cation 3.5.2 requires two independent ECCS subsystems to be OPERABLE and provides explicit ACTION requirements if one ECCS subsystem is inoperable. Under the requirements of Specification 3.0.3, if both the required ECCS subsystems are inoperable, within 1 hour measures must be initiated to place the unit in at least HOT STANDBY within the next 6 hours, and in at least HOT SHUTDOWN within the following 6 hours. As a further example, Specification 3.6.2.1 requires two Containment Spray Systems to be OPERABLE and provides explicit ACTION requirements if one Spray System is inoperable. Under the requirements of Specification 3.0.3 if both the required Containment Spray Systems are inoperable, within 1 hour measures must be initiated to place the unit in at least HOT STAND 8Y within the next 6 hours, in at least HOT SHUTDOWN ; within the following 6 hours, and in COLD SHUTDOWN within the subsequent 24 hours. It is acceptable to initiate and complete a reduction in OPERATIONAL MODES in a shorter time interval than required in the ACTION statement and to add the unused portion of this allowable out-of-service time to that provided for operation in subsequent lower OPERATIONAL MODE (S). Stated allowable out-of-service times are applicable regardless of the OPERATIONAL MODE (S) in which the inoperability is discovered but the times provided for achieving a mode reduction are not cpplicable if the inoperability is discovered in a mode lower than the applicable mode. For example, if the Containment Spray System was discovered to be inoperable while in STARTUP, the ACTION Statement would allow up to 156 hours to achieve COLD SHUTDOWN. If HOT STANDBY is attained in 16 hours rather than the allowed 78 hours, 140 hours would still be available i before the plant would be required to be in COLD SHUTDOWN. However, if this system was discovered to be inoperable while in HOT STANDBY, the 6 hours provided to achieve HOT STANDBY would not be additive to the time available , to achieve COLD SHUTDOWN so that the total allowable time is reduced from ' 156 hours to 150 hours. 3.0.4 This specification provides that entry into an OPERATIONAL MODE or other specified applicability condition must be made with: (1) the full complement of required systems, equipment, or components OPERABLE, and (2) all other parameters as specified in the Limiting Conditions for Operation being. met without regard for allowable deviations and out-of-service provisions contained in the ACTION statements. BRAIDWOOD - UNITS 1 & 2 B 3/4 0-1
FINAL DRAF APPLICABILITY AUG 191986 BASES The intent of this provision is to ensure that facility operation is not initiated with either required equipment or systems inoperable or other specified limits being exceeded. Exceptions to this provision have been provided for a limited number of specifications when STARTUP with inoperable equipment would not affei::t plant safety. These exceptions are stated in the ACTION statements of the appropriate specifications. 3.0.5 This specification delineates the applicability of each specifica-tion to Unit 1 and Unit 2 operation. 4.0.1 This specification provides that surveillance activities necessary to ensure the Limiting Conditions for Operation are met and will be performed during the OPERATIONAL MODES or other conditions for which the Limiting Condi-tions for Operation are applicable. Provisions for additional surveillance activities to be performed without regard to the applicable OPERATIONAL MODES or other conditions are provided in the individual Surveillance Requirements. Surveillance Requirements for Special Test Exceptions need only be performed when the Special Test Exception is being utilized as an exception to an individual specification.
" 4.0.2 The provisions of this specification provide allowable tolera'nces for performing surveillance activities beyond those specified in the nominal surveillance interval. These tolerances are necessary to provide operational flexibility because of scheduling and performance considerations. The phrase "at least" associated with a surveillance frequency does not negate this allowable tolerance value and permits the performance of more frequent surveillance activities.
The tolerance values, taken either individually or consecutively over three test intervals, are sufficiently restrictive to ensure that the reliability associated with the surveillance activity is not significantly degraded beyond that obtained from the nominal specified interval. 4.0.3 The provisions of this specification set forth the criteria for determination of compliance with the OPERABILITY requirements of the Limiting Conditions for Operation. Under these criteria, equipment, systems, or components are assumed to be OPERABLE if the associated surveillance activities have been satisfactorily performed within the specified time interval. Nothing in this provision is to be construed as defining equipment, systems, or components OPERABLE, when such items are found or known to be inoperable although still meeting the Surveillance Requirements. Items may be determined inoperable during use, during surveillance tests, or in accordance with this specification. Therefore, ACTION statements are entered when the Surveillence Requirements should have been performed rather than at the time it is discovered that the tests were not performed. 4.0.4 This specification ensures that the surveillance activities asso-I ciated with a Limiting Condition for Operation have Men performed within the I specified time interval prior to entry into an OPERA. iONAL MODE or other appli-cable condition. The intent of this provision is to ensure that surveillance ! BRAIDWOOD - UNITS 1 & 2 B 3/4 0-2
APPLICABILITY FINAL DRAFT BASES 19 S6 activities have been satisfactorily demonstrated on a current basis as required to meet the OPERABILITY requirements of the Limiting Condition for Operation.
~
Under the terms of this specification, for example, during initial plant STARTUP or following extended plant outages, the applicable surveillance activities must be performed within the stated surveillance interval prior to placing or returning the system or equipment into OPERABLE status. 4.0.5 This specification ensures that inservice inspection of ASME Code Class 1, 2 and 3 components and inservice testing of ASME Code Class 1, 2 and 3 pumps and valves will be perfcrued in accordance with a periodically updated version of Section XI of the A W Poiler and Pressure Vessel Code and Addenda as required by 10 CFR 50.55a. Relief from any of the above requirements has been provided in writing by the Commission and is not a part of these Technical Specifications. This specification includes a clarification of the frequencies for performing the inservice inspection and testing activities required by Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda. This clarification is provided tr snsure cor,sistency in surveillance intervals throughout these Technical Spt c'w.. tons and to remove any ambiguities relative to the frequencies for perforr.,7 m required inservice inspection and testing
~
activities. Under the terms of this specification, the more restrictive requirements of the Technical Specifications take precedence over the ASME Boiler and Pressure Vessel Code and applicable Addenda. For example, the requirements of Specification 4.0.4 to perform surveillance activities prior to entry into an OPERATIONAL MODE or other specified applicability condition takes precedence over the ASME Boiler and Pressure Vessel Code provision which allows pumps to be tested up to 1 week after return to normal operation. And for example, the Technical Specification definition of OPERABLE does not grant a grace period before a device that is not capable of performing its specified function is declared inoperable and takes precedence over the ASME Boiler and Pressure Vessel Code provision which allows a valve to be incapable of performing its specified function for up to 24 hours before being declared inoperable. 4.0.6 This specification delineates the applicability of the surveillance l activities to Unit 1 and Unit 2 operations. l BRAIDWOOD - UNITS 1 & 2 B 3/4 0-3 l
l 3/4.1 REACTIVITY CONTROL SYSTEMS AUG 191986 BASES 3/4.1.1 BORATION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN . 1 A sufficient SHUTDOWN MARGIN ensures that: (1) the reactor can be made subcritical from all operating conditions, (2) the reactivity transients asso-ciated with postulated accident conditions are controllable within acceptable limits, and (3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition. SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS T,yg. The most restrictive condition occurs at EOL, with T,yg at no load operating temperature, and is associated with a postulated steam line break accident and resulting uncon-trolled RCS cooldown. In the analysis of this accident, a minimum SHUTDOWN MARGIN of 1.3% ak/k is required to control the reactivity transient. Accordingly, the SHUTDOWN MARGIN requirement is based upon this limiting condition and is consistent with FSAR safety analysis assumptions. With T,yg less than 200*F, the reactivity transients resulting from a postulated steam line break cooldown are minimal and a 1% Ak/k SHUTDOWN MARGIN provides adequate protection. The OPERABILITY of the four charging pump suction valves ensures adequate capability for negative reactivity insertion to prevent a transient caused by the uncontrolled dilution of the RCS. The functioning of the valves precludes the necessity of operator action to prevent further dilution by terminating flow to the charging pumps from possible unborated water sources and initiating flow from the RWST. Actions taken by the microprocessor if the neutron count rate is doubled will prevent return to criticality in these MODES. 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. i 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 conparison. BRAIDWOOD - UNITS 1 & 2 8 3/4 1-1
REACTIVITY CONTROL SYSTEMS BASES MODERATOR TEMPERATURE COEFFICIENT (Continued) l The most negative MTC value equivalent to the most positive moderator density coefficient (MDC), was obtained by incrementally correcting the MDC used in the FSAR analyses to nominal operating conditions. These corrections 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 -4.1 x 10 4 Ak/k/*F. The MTC
'value of -3.2 x 10 4 Ak/k/*F represents a conservative value (with corrections for burnup and soluble boron) at a core condition of 300 ppm equilibrium boron concentration and is obtained by making these corrections to the limiting MTC value of -4.1 x 10 4 Ak/k/*F.
The Surveillance Requirements for measurement of the MTC at the beginning and near the end of the fuel cycle are adequate to confirm that the MTC remains within its limits since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup. 3/4.1.1.4 HINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 550*F. This limitation is required to ensure: (1) the moderator temperature coefficient 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 status with a steam bubble, (4) the reactor vessel is above its minimum RT temperature, and (5) the plant is above the cooldown steam dump NDT permissive, P-12. 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 the 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 1.3% 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 require's 15,780 gallons of 7000 ppm borated water from the boric acid storage tanks or 70,450 gallons of 2000 ppm borated water from the refueling water storage tank. BRAIDWOOD - UNITS 1 & 2 B 3/4 1-2
REACTIVITY CONTROL SYSTEMS FINAL MAFT g yg BASES l BORATION SYSTEMS (Continued) A Boric Acid Storage System level of 40% ensures that there is a volume of greater than or equal to 15,780 gallons available. A RWST level of 89% ensures I that there is a volume of greater than or equal to 395,000 gallons available. With the RCS temperature below 350*F, one Boron Injection System is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single Baron Injection System becomes 11 operable. The limitation for a mcximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps except the required OPERABLE pump to be inoperable below 330*F provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV or an RHR Suction valve. ~ The boron capability required below 200*F is sufficient to provide a SHUTDOWN MARGIN of 1% Ak/k after xenon decay and cooldown from 200*F to 140*F. This condition requires either 2,652 gallons of 7000 ppm borated water from the boric acid storage tanks or 11,840 gallons of 2000 ppm borated water from the refueling water storage tank (RWST). A Boric Acid Storage System level of 7% ensures there is a volume of greater than or equal to 2652 gallons available. An RWST level of 9% ensures there is a volume of greater than or equal to 38,740 gallons available. 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 8.5 and 11.0 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. l The OPERABILITY of one Boron Injection System during REFUELING ensures l that this system is available for reactivity control while in MODE 6. 1 l 3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that: (1) acceptable power
- distribution limits are maintained, (2) the minimum SHUT 00WN MARGIN is main-tained, and (3) the potential effects of rod misalignment on associated accident analyses are limited. OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby. ensure compliance with the control rod alignment and insertion limits. Verification that the Digital Rod Position Indicator agrees with the demanded position within 112 steps at 24, 48, 120, and 228 steps withdrawn for the Control Banks and BRAIDWOOD - UNITS 1 & 2 8 3/4 1-3
REACTIVITY CONTROL SYSTEMS Ignb y BASES 1 MOVABLE CONTROL ASSEMBLIES (Continued) ! 18, 210, and 228 stept withdrawn 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 System does not indicate the actual shutdown rod position between 18 steps and 210 steps, only points in the indicated ranges are picked for. verification of agreement with demanded position. 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 measurement of peaking factors and a restriction in THERMAL POWER. These restrictions provide assurance of fuel rod integrity during continued operation. In addition, those safetv '" ' !ses affected by a misaligned rod are reevaluated to confirm that tt.< resuet.s 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 ecual to 550 F and with all reactor coolant pumps operating ensures 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 hours with more frequent verifications required if a rod position deviation monitor is inoper-able. These verification frequencies are adequate for assuring that the applicable LCOs are satisfied. BRAIDWOOD - UNITS 1 & 2 8 3/4 1-4 t t _
3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) ; events by: (1) maintaining the minimum Design DNBR in the core greater than or equal to 1.34 for a typical cell and 1.32 for a thimble cell during normal l operation and in short-term transients, and (2) limiting the fission gas release, fuel pellet temperature, and cladding mechanical properties to within assumed design criteria. In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumad for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200 F is 3 i not exceeded. ! The definitions of certain hot channel and peaking factors as used in these specifications are as follows: . F9 (Z) Heat fjux Hot Channel Factor, is defined as the maximum local heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods; Fh Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio.of the integral of linear power al power to the average rod power;,ong and the rod with the highest integrated Fxy(Z) Radial Peaking Factor, is defined as the ratio of peak power density to average power density in the horizontal plane at core elevation Z. 3/4.2.1 AXIAL FLUX DIFFERENCE The limits on AXIAL FLUX DIFFERENCE (AFD) assure that the Fq (Z) upper bound envelope of 2.32 times the normalized axial peaking factor is not exceeded during either normal operation or in the event of xenon redistribution following power changes. Target flux difference is determined at equilibrium xenon conditions. The full-length rods may be positioned within the core in accordance with their respective insertion limits and should be inserted near their normal position for steady-state operation at high power levels. The value of the target flux difference obtained under these conditions divided by the fraction of RATED THERMAL POWER is the target flux difference at RATED THERMAL POWER for the associated core burnup conditions. Target flux differences for other THERMAL POWER levels are obtained by multiplying the RATED THERMAL POWER value by the appropriate fractional THERMAL POWER level. The periodic updating of the target flux difference value is necessary to reflect core burnup considerations. BRAIDWOOD - UNITS 1 & 2 B 3/4 2-1
POWER DISTRIBUTION LIMITS FI , AUG 191986 AXIAL FLUX DIFFERENCE (Continued) Although it is intended that the plant will be operated with the AFD within the target band required by Specification 3.2.1 about the target flux difference, during rapid plant THERMAL POWER reductions, control rod motion will cause the AFD to deviate outside of the target band at reduced THERMAL POWER levels. This deviation will not affect the xenon redistribution sufficiently to change the envelope of peaking factors which may be reached on a subsequent return to RATED THERMAL POWER (with the AFD within the target band) provided the time duration of the deviation is limited. Accordingly, a 1-hour penalty deviation limit cumulative during the previous 24 hours is provided for operation outside of the target band but within the limits of Figure 3.2-1 while at THERMAL POWER levels between 50% and 90% of RATED THERMAL POWER. For THERMAL POWER levels between 15% and 50% of RATED THERMAL POWER, deviations of the AFD outside of the target band are less significant. The penalty of 2 hours actual time reflects this reduced significance. Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitor Alarm. The computer deter-mines the 1 minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for two or more OPERABLE excore channels are outside the target band and the THERMAL POWER is greater than 90% of RATED THERMAL POWER. During operation at THERMAL POWER levels between 50% and 90% and between 15% and 50% RATED THERMAL POWER, the computer outputs an alarm message when the penalty deviation accumulates beyond the limits of I hour and 2 hours, respectively. Figure B 3/4 2-1 shows a typical monthly target band. 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR, and RCS FLOWRATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR l l The limits on heat flux hot channel factor, RCS flowrate, and nuclear enthalpy rise hot channel factor ensure that: (1) the design limits on peak ! local power density and minimum DNBR are not exceeded, and (2) in the event of l a LOCA the peak fuel clad temperature will not exceed the 2200*F ECCS acceptance criteria limit. Each of these is measurable but will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3. This periodic surveillance is sufficient to insure that the limits are maintained provided:
- a. Control rods in a single group move together with no individual rod position differing by more than + 12 steps, indicated, from the group demand position,
- b. Control rod groups are sequenced with overlapping groups as described j in Specification 3.1.3.6, l BRAIDWOOD - UNITS 1 & 2 B 3/4 2-2 i
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FIGURE B 3/4 2-1 TYPICAL INDICATED AXIAL FLUX DIFFERENCE VERSUS THERMAL POWER BRAIDWOOD - UNITS 1 & 2 B 3/4 2-3
POWER DISTRIBUTION LIMITS FINAL HAFT gg BASES HEAT FLUX HOT CHANNEL FACTOR, and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued)
- c. The control rod insertion limits of Specification 3.1.3.6 are maintained, and
- d. The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.
FhwillbemaintainedwithinitslimitsprovidedtheConditionsa.through
- d. above are maintained.N The combination of the RCS flow requirement (390,400 gpm) and the requirement on Fg guarantee that the DNBR used in the safety analysis will be met.
ArodbowpenaltyisnotappliedtothefinalvalueofFhforthe following reason: Fuel rod bowing does reduce the value of the DNBR. However, predictions with the methods described in WCAP-8691, Revision 1, " Fuel Rod Bow Evaluation," July 1979 for the 17x17 Optimized Fuel Assemblies indicate that the fuel rod bow reduction on DNBR will be less than 3% at 33,000 MWD /MTU assembly average burnup. At higher burnups, the decrease in fissionable isotopes and the buildup of fission product inventory more. than compensate for the rod bow reduction in DNBR. There is a 11% margin available between the 1.32 and 1.34 design DNBR limits and the 1.47 and 1.49 safety analysis DNBR limit. Use of the 3% fuel rod bow DNBR margin reduction still leaves a 8% margin in DNBR between design limits and safety analysis limits. The RCS flow requirement is based on the loop minimum measured flow rate of 97,600 gpm which is used in the Improved Thermal Design Procedure described in FSAR 4.4.1 and 15.0.3. A precision heat balance is performed once each cycle and is used to calibrate the RCS flow rate indicators. Potential fouling of the feedwater venturi, which might not be detected, could bias the results from the precision heat balance in a non-conservative manner. Therefore, a penalty of 0.1% is assessed for potential feedwater venturi fouling. A maximum measurement uncertainty of 2.2% has been included in the loop minimum measured flow rate to account for potential undetected feedwater venturi fouling and the use of the RCS flow indicators for flow rate verification. Any fouling which might bias the RCS flow rate measurement greater than 0.1% can be detected by monitoring and trending various plant performance parameters. If detected, action shall be
- taken, before performing subsequent precision heat balance measurements, i.e.,
either the effect of fouling shall be quantified and compensated for in the RCS flow rate measurement, or the venturi shall be cleaned to eliminate the fouling. Surveillance Requirement 4.2.3.4 provides adequate monitoring to detect possible flow reductions due to any rapid core crud buildup. Surveillance Requirement 4.2.3.5 specifies that the measurement instrumen-tation shall be calibrated within seven days prior to the performance of the calorimetric flow measurement. This requirement is due to the fact that the drift effects of this instrumentation are not included in the flow mea'surement uncertainty analysis. This requirement does not apply for the instrumentaticn whose drift effects have been included in the uncertainty analysis. BRAIDWOOD - UNITS 1 & 2 B 3/4 2-4
POWER DISTRIBUTION LIMITS HR DMFT ) AUG 191986 l BASES HEAT FLUX HOT CHANNEL FACTOR, and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued) Thelimitof1.55forFhdoesnotassumeanyspecificuncertaintyonthe j measured value of F H. An appropriate uncertainty of 4% (nominal) or greater isaddedtothemeasuredvalueofFhbeforeitiscomparedwiththerequirement. When an qF measurement is taken, an allowance for both experimental error and manufacturing tolerance must be made. An allowance of 5% is appropriate for a full-core map taken with the Incore Detector Flux Mapping System, and a 3% allowance is appropriate for manufacturing tolerance. The Radial Peaking Factor, Fxy(Z) is measured periodically to provide assurance that the Hot Channel F (Z) remains within its limit. The F limit for RATED THERMAL POWER (F RTP)as q provided in Specification 3.2.2 wa determined from expected power control maneuvers over the full range of burnup conditions in the core. The 12-hour periodic surveillance of indicated RCS flow is sufficient to detect flow degradation which could lead to operation outside the acceptable limit. 3/4.2.4 QUADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that the radial power dis-tribution 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 but less than 1.09 is provided to allow identification and correc-tion of a dropped or misaligned control rod. In the event such action does not correct the tilt, the margin for uncertainty on Fqis reinstated by reduc ing 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 trie 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 sym-metric 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. BRAIDWOOD - UNITS 1 & 2 B 3/4 2-5
POWER DISTRIBUTION LIMITS AUG 19 $86 BASES 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 initial FSAR assumptions and have been analytically demonstrated adequate to maintain a design DNBR throughout each analyzed transient. The calculated valves of the DNB-related parameters will be nn average of the indicated values for the OPERABLE channels. 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. BRAIDWOOD - UNITS 1 & 2 B 3/4 2-6
3/4,3 INSTRUMENTATION BASES ' 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION The OPERABILITY of the Reactor Trip System and the Engineered Safety Features Actuation System instrumentation and interlocks ensures that: (1) the associated ACTION and/or Reactor trip will be initiated when the parameter monitored by each channel or combination thereof reaches its Setpoint, (2) the specified coincidence logic is maintained, (3) sufficient redundancy is main-tained to permit a channel to be out-of-service for testing or maintenance, and (4) sufficient system functional capability is available from diverse parameters. The OPERABILITY of these systems is required to provide the overall reliability, redundancy, and diversity assumed available in the facility design for the protection and mitigation of accident and transient. conditions. The integrated operation of each of these systems is consistent with the assumptions used in the safety analyses. The Surveillance Requirements specified for these systems ensure that the overall system functional capability is maintained ' comparable to the original design standards. The periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capability. The Engineered Safety Features Actuation System Instrumentation Trip Setpoints specified in Table 3.3-4 are the nominal values at which the bist&bles are set for each functional unit. A Setpoint is considered to be adjusted consistent with the nominal value when the "as measured" Setpoint is within the band allowed for calibration accuracy. To accommodate the instrument drift assumed to occur between operational tests and the accuracy to which Setpoints can be measured and calibrated, Allowable Values for the Setpoints have been specified in Table 3.3-4. Operation with Setpoints less conservative than the Trip Setpoint but within the Allowable Value is acceptable since an allowance has been made in the safety analysis to accommodate this error. An optional provision has been included for determining i the OPERABILITY of a channel when its Trip Setpoint is found to exceed the Allowable Value. The methodology of this option utilizes the "as measured" deviation from the specified calibration point for rack and sensor components j in conjunction with a statistical combination of the other uncertainties of I the instrumentation to measure the process variable and the uncertainties in calibrating the instrumentation. In Equation 3.3-1, Z + RE + SE < TA, the l interactive effects of the errors in the rack and the sensor, and the "as measured" values of the errors are considered. Z, as specified in Table 3.3-4, in percent span, is the statistical summation of errors assumed in the analysis excluding those associated with the sensor and rack drift and the accuracy of l their measurement. TA or Total Allowance is the difference, in percent span, l between the Trip Setpoint and the value used in the analysis for the actuation. RE or Rack Error is the "as measured" deviation, in percent span, for the affected channel from the specified Trip Setpoint. SE or Sensor Error.is either BRAIDWOOD - UNITS 1 & 2 8 3/4 3-1 l
INSTRUMENTATION FIN"" l BASES REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued) 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 3.3-1 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 of the uncertainties in the channels. Inherent to the determination of the Trip Setpoints are the magnitudes of these channel uncertainties. Sensor and rack instrumentation utilized in these channels are expected to be capable of operating within the allowances of these uncertainty magnitudes. Rack drift in excess of the Allowable Value exhibits the behavior that the rack has not met its allowance. Being that there is a small statisit.ical chance that this will happen, an infrequent excessive drift is expected. Rack or sensor drift, in excess of the allowance that is more than occasional, may be indicative of more serious problems and should warrant further investigation. 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 litit 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 demon-strated by any series of sequential, overlapping or total channel test measurements provided that such tests demonstrate the total channel response time as defined. Sensor response time verification may be demonstrated by either: (1) in place, onsite, or offsite test measurements, or (2) utilizing replacement sensors with certified response times. The Engineered Safety Features Actuation System senses selected plant parameters and determines whether or not predetermined limits are being exceeded. If they are, the signals are combined into logic matrices sensitive to combinations indicative of various accidents, events, and transients. Once the required logic combination is completed, the system sends actuation signals to those Engineered Safety Features components whose aggregate function best serves the requirements of the condition. As an example, the following actions may be initiated by the Engineered Safety Features Actuation System to mitigate the consequences of a steam line break or loss of coolant accident: (1) Safety Injection pumps start and automatic valves position, (2) Reactor trip, (3) feed-water isolation, (4) startup of the emergency diesel generators, (5) containment , spray pumps start and automatic valves position, (6) containment isolation, ) (7) steam line isolation, (8) Turbine trip, (9) auxiliary feedwater pumps ! start and automatic valves position, (10) containment cooling fans start and l automatic valves position, and (11) essential service water pumps start and j automatic valves position. BRAIDWOOD - UNITS 1 & 2 B 3/4 3-2 L
INSTRUMENTATION AUG 191986 BASES Engineered Safety Features Actuation System Interlocks The Engineered Safety Features Actuation System interlocks perform the following functions: l P-4 Reactor tripped - Actuates Turbine trip, closes main feedwater valves on T,yg below Setpoint, prevents the opening of the main feedwater valves which were closed by a Safety Injection or High Steam Generator Water , Level signal, allows Safety Injection block so that components can be reset or tripped. Reactor not tripped prevents manual block of Safety Injection. P-11 On increasing pressure P-11 automatically reinstates Safety Injection actuation on low pressurizer pressure and low steamline pressure and automatically blocks steamline isolation on negative steamline pressure rate. On decreasing pressure; P-11 allows the manual block of Safety Injection low pressurizar pressure and low steamline pressure and hilows steamline isolation on negative steamline pressure rate to become active upon manual block of low steamline pressure SI. 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 An increasing steam generator water level, P-14 automatically trips all feedwater isolation valves and inhibits feedwater control valve modulation. 3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 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 reaches its Setpoint and (2) suffi-cient redundancy is maintained to permit a channel to be out-of-service for testing or maintenance. The radiation monitors for plant operations senses radiation levels in selected plant systems and locations and determines whether or not predetermined limits are being exceeded. If they are, the system sends actuation signals to initiate alarms and automatic actuation of Emergency Exhaust or Ventilation Systems. The radiation monitor Setpoints given in the requirements are assumed to be values established above normal background radiation levels for the particular area. Radiation monitors ORE-AR055 and 56 serve a dual purpose for plant operations as criticality and fuel handling accident sensors. Although these monitors are designed primarily to detect fuel handling accident releases, they are capable of detecting an inadvertent criticality incident. The Setpoint given in the requirement is established for the fuel handling building isolation function but is also adequate for an inadvertent criticality. BRAIDWOOD - UNITS 1 & 2 B 3/4 3-3
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INSTRUMENTATION FINAL. RAFT 1
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BASES 3/4.3.3.2 MOVABLE INCORE DETECTORS l The OPERABILITY of the movable incore detectors with the specified minimum complement of equipment ensures that the measurements obtained from use of this system accurately represent the spatial neutron flux distributibn of the core. The OPERABILITY of this system is demonstrated by irradiating each detector used and determining the acceptability of its voltage curve. j ForthepurposeofmeasuringF(Z)orFhafullincorefluxmapisused. 9 Quarter-core flux maps, as defined in WCAP-8648, June 1976, may be used in recalibration of the Excore Neutron Flux Detection System, and full incore flux maps or symmetric incore thimbles may be used for monitoring the QUADRANT POWER TILT RATIO when one Power Range channel is inoperable. 3/4.3.3.3 SEISMIC INSTRUMENTATION The OPERABILITY of the seismic instrumentation ensures that sufficient capability is available to promptly determine the magnitude of a seismic event and evaluate the response of those features important to safety. This capability is required to permit comparison of the measured response to that used in the design basis for the facility to determine if plant shutdown is required pursuant to Appendix A of 10 CFR Part 100. The instrumentation consists of one time-history response spectrum analyzer, a playback unit, three peak recording accelerometers, and five triaxial acceler-ometers. The above-mentioned equipment, excluding the sensors, is located in the Auxiliary Electrical Room. The remaining sensors are located as follows: three in containment, one in the Auxiliary Building, and one at the free field location 39 + 00E, 41 + 005. The peak recording accelerometers are passive devices which have no interplay on the rest of the system and are located on reactor equipment, reactor piping, and outside containment on the Category I piping. The triaxial accelerometer is based on three orthogonal force-balanced servo-accelerometers which generate a voltage signal upon stimulation. The voltage signals are transmitted to the time-history recorder in the Auxiliary Electrical Room, digitized, and recorded on magnetic tape. The time-history recorder is the master control unit for all control timing signals and system data interface. It also contains the system triggers used to actuate the system. The master control unit continualiy monitors two of the sensor inputs, which are processed through the trigger circuits for comparison to the system actuation level. The time-history recorder also has the ability to record both pre- and post-seismic event data. The other key component ir d system is the response spectrum analyzer. This unit determines the variation in the maximum response of a single-degree-of-freedom systern system versus its natural frequency of vibration when either of two designated triaxial.accelero-meters is subjected to a time-history motion of the accelerometer. BRAIDWOOD - UNITS 1 & 2 B 3/4 3-4
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INSTRUMENTATION AUG 191986 BASES SEISMIC INSTRUMENTATION (Continued) The response spectrum analyzer computes the response spectrum of the event for two sensor locations, compares it to the design response spectra of the plant, and indicates whether the event exceeded the operating basis earthquake criteria or the safe shutdown earthquake criteria. This instrumentation is consistent with the recommendations of Regulatory Guide 1.12, " Instrumentation for Earthquake," April 1974. , 3/4.3.3.4 METEOROLOGICAL INSTRUMENTATION The OPERABILITY of the meteorological instrumentation ensures that sufficient meteorological data are available for estimating potential radiation doses to the public as a result of routine or accidental release of radioactive - materials to the atmosphere. This capability is required to evaluate the need for initiating protective measures to protect the health and safety of the public and is consistent with the recommendations of Regulatory Guide 1.23, "Onsite Meteorological Programs," February 1972. 3/4.3.3.5 REMOTE SHUTDOWN INSTRUMENTATIO.1 The OPERABILITY of the remote shutdown instrumentation ensures that sufficient capability is available to permit shutdown and maintenance of HOT STANDBY of the facility from locations outside of the control room. This capability is required in the event control room habitability is lost and is consistent with General Design Criterion 19 of 10 CFR Part 50. 3/4.3.3.6 ACCIDENT MONITORING INSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables following aa accident. This capability is consistent with the recommendations of Regulatory Guide 1.97, Revision 3, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident," May 1983 and NUREG 0737, " Clarification of TMI Action Plan Requirements," November 1980. 3/4.3.3.7 CHLORINE DETECTION SYSTEMS The OPERABILITY of the Chlorine Detection Systems ensures that sufficient capability is available to promptly detect and initiate protective action in the event of an accidental chlorine release. This capability is required to protect control room personnel and is consistent with the recommendations of Regulatory Guide 1.95, Revision 1, " Protection of Nuclear Power Plant . Control Room Operators Against an Accidental Chlorine Release," January 1977. l BRAIDWOOD - UNITS 1 & 2 B 3/4 3-5
INSTRUMENTATION AUG 191 BASES . 3/4.3.3.8 FIRE DETECTION INSTRUMENTATION OPERABILITY of the detection instrumentation ensures that both adequate warning capability is available for the prompt detection of fires and that Fire Suppression Systems, that are actuated by fire detectors, will discharge extinguishing agents in a timely manner. Prompt detection and suppression of fires will reduce the potential for damage to safety-related equipment and is an integral element in the overall facility fire protection program. The zone designations given in Table 3.3-11 are from electrical schematic diagrams. Fire detectors that'are used to actuate Fire Suppression Systems represent a more critically important component of a plant's fire protection program than detectors that are installed solely for early fire warning and notification. Consequently, the minimum number of operable fire detectors must be greater. The loss of detection capability for Fire Suppression Systems, actuated by fire detectors represents a significant degradation of fire protection for any area. As a result, the establishment of a fire watch patrol must be initiated at an earlier stage than would be warranted for the loss of detectors that provide only early fire warning. The establishment of frequent fire patrols in the affected areas is required to provide detection capability until the inoperable instrumentation is restored to OPERABILITY. 3/4.3.3.9 LOOSE-PART DETECTION SYSTEM The OPERABILITY of the loose part detection system ensures that sufficient capability is available to ietect loose metallic parts in the Reactor Coolant System and avoid or mitigate damage to Reactor Coolant System components. The allowable out-of-service times and Surveillance Requirements are consistent with the recommendations of Regulatory Guide 1.133, " Loose-Part Detection Program for the Primary System of Light-Water-Cooled Reactors," May 1981. 3/4.3.3.10 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The Alarm / Trip Setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50. 3/4.3.3.11 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The Alarm / Trip Setpoints for these instruments shall be calculated and adjusted j BRAIDWOOD - UNITS 1 & 2 B 3/4 3-6
INSTRUMENTATION FINAL DRAFT AUG19gggy BASES RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION (Continued) in accordance with the methodology and parameters in the 00CM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. . The instrumentation also includes provisions for monitoring (and controlling) the concentrations of potentially explosive gas mixtures in the WASTE GAS HOLDUP SYSTEM. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50. The sensitivity of any noble gas activity monitor used to show compliance with the gaseous effluent release requirements of Specification 3.11.2.2 shall be such that concentrations as low as 1x10 8 uCi/cc are measurable. 3/4.3.4 TURBINE OVERSPEED PROTECTION This specification is provided to ensure that the turbine overspeed protection instrumentation and the turbine speed control valves are OPERABLE and will protect the turbine from excessive overspeed. Protection from turbine excessive overspeed is required since excessive overspeed of the turbine could generate potentially damaging missiles which could impact and damage safety-related components, equipment, or structures. Specification 4.3.4.2a (High Pressure Turbine and Reheat Valves) These valves isolate large quantities of steam with high potential for delivering energy to the rotor system. The turbine design recognizes this potential in providing rapid action, dual shut off capability in each path, remote testing capability, and a flow path that reduces the effects of changes in flow distribution, load reductions, and thermal transients during testing. The testing intervals are in accordance with the latest manufacturer's recommendations: " Operation and Maintenance Memo 041," Steam Turbine Division, Westinghouse. Specification 4.3.4.2b and c (Extraction Steam Non-Return Check Valves) These valves are provided to protect the turbine from reflux of steam remaining in the feedwater heater shells and piping following the pressure reduction caused by the actuation of valves in Specification 4.3.4.2a. The quantities of stored steam controlled by these valves are smaller and are divided up into separate heater shells. The feedwater heating system design, including these valves, did not intend routine full stroke testing. The extraction steam check valves are self closing swing disk non-return valves which shut under the combined effect of gravity and reverse flow of steam. The weight of the disk is partly balanced by a counterweight and lever on the pivot shaft. A spring cylinder acting on the lever assists the start of the automatic closing, but is not intended to close the valve fully against normal steam flow and pressure. In normal operation the spring assist is held clear by air pressure acting on a piston under the spring. The turbine trip system releases the air pressure to assist the closing. BRAIDWOOD - UNITS 1 & 2 8 3/4 3-7
INSTRUMENTATION FIEL DMFT AUG 19 y BASES TURBINE OVERSPEED PROTECTION (Continued) Manual stroking of the extraction steam non return valves is 'possible under shutdown conditions by latching the turbine and applying the a'ir pressure to the spring cylinder. It is possible to hear and feel the disk contact the seat solidly. This manual stroking was not provided for in the design but will be done within 7 days prior to entering Mode 3 from Mode 4. The engineering specifications provided for testing the extraction steam non-return check valves during operation by equalizing the air pressure across the piston in the spring cylinder, permitting the spring to partially close the disk against the steam flow. The rotation of the shaft accompanying the disk closure can be observed by movement of the weight lever. The amount of movement observed in other stations has depended on the extraction steam conditions and valve size, but has been ample to indicate freedom of movement, and this will be verified during startup testing. Partial stroking demonstrates that the disk system is free at the beginning of the closing stroke where the steam closing forces are smallest. As the disk enters a reverse steam flow the closing forces build up rapidly with progressive closure. The design of the feedwater heating system is such that full stroke testing of the extraction steam non-return valves during turbine operation involves several penalties without significant additional advantages over partial stroke testing. The motor-operated isolating valve must be closed on an individual heater. Heater stages 1, 2, 3, and 4 are arranged in three parallel strings with cascaded drains in each string and heater stages 5, 6 and 7 are similarly arranged in two parallel strings. An entire string is taken out of service, isolated, and bypassed for maintenance. Isolating the extraction steam to a single intermediate heater involves several complications. The motor operated valves are too large for routine manual operation, do not have bypasses to allow controlled warmup conditions, stroke quickly (about 15 seconds), and are intended for turbine protection against heater flooding. A comparison of the thermal capacity of a heater and the rate of heat transfer to the flowing condensate or feedwater shows that cycling an extraction steam isolating valve would cause rapid cooling and heating transients. Isolating the steam to a top heater drops the feedwater temperature to the steam generators. Isolating the steam to an intermeolate heater causes the next heater to assume the heating load, approximately doubling the steam demand and drain flow, and nearly quadrupling the potential for erosion and vibration in the affected heater and piping. The shell pressure collapses in the isolated heater causing insufficient head to discharge the cascading BRAIDWOOD - UNITS 1 & 2 B 3/4 3-8
INSTRUMENTATION AUG 19 % BASES
)
TURBINE OVERSPEED PROTECTION (Continued) drains to the next lower heater. Rapid action of the emergency drain control is required to prevent flooding, with the potential for flashing in the drain cooler section from pressure decay. Isolating a heater degrades the cycle thermal performance, requiring a corresponding drop in electrical output for the same reactor thermal power. Partial closing of the extraction steam non-return check valves with the installed test provisions demonstrates freedom of movement while avoiding transient states. A 31-day interval will be adequate since it is likely that sticking conditions would develop during shutdown conditions rather than in operation. BRAIDWOOD - UNITS 1 & 2 B 3/4 3-9
\
m 3/4.4 REACTOR COOLANT SYSTEM FM g yg HAFT BASES 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with all reactor coolant loops in operation and maintain DNBR above the applicable design bases DNBR during all normal operations and anticipated transients. In MODES 1 and 2 with one reactor cool-ant loop not in operation this specification requires that the plant be in at least HOT STANDBY within 6 hours. In MODE 3, two reactor coolant loops provide sufficient heat removal capability for removing decay heat even in the event of a bank withdrawal accident; however, a single reactor coolant loop provides sufficient heat removal 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 sufficient heat removal espability 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 operation of one reactor coolant pump (RCP) or one RHR puhp provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System. The reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognition and control. The restrictions on starting a reactor coolant pump 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 l Appendix G by restricting starting of the RCPs to when the secondary water l temperature of each steam generator is less than 50 F above each of the RCS cold leg temperatures. 3 The requirement to maintain the boron concentration of an isolated loop greater than or equal to the boron concentration of the operating loops ensures that no reactivity addition to the core could occur during startup of an isolated loop. Verification of the boron concentration in an idle loop prior to opening the stop valves provides a reassurance of the adequacy of the boron concentration in the isolated loop. Startup of an idle loop will inject cool water from the loop into the core. The reactivity transient resulting from this cool water injection is minimized by delaying isolated loop startup until its temperature is within 20 F of the operating loops. l BRAIDWOOD - UNITS 1 & 2 B 3/4 4-1
REACTOR COOLANT SYSTEM '
- dk AUG 191986 BASES -
3/4.4.2 SAFETY VALVES The pressurizer Code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2735 psig. Each safety valve is designed to relieve 420,000 lbs per hour of saturated steam at the valve Setpoint. 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 RCS, provides overpressure relief capability and will prevent RCS overpressurization. In addition, the Overpressure Protection System provides a diverse means of protection against RCS overpressurization at low temperatures. During operation, all pressurizer Code safety valves must be OPERABLE to prevent the RCS from being pressurized above its Safety Limit of 2735 psig. The combined relief capacity of all of these valves is greater than the maximum surge rate resulting from a complete loss-of-load assuming no Reactor trip and also assuming no operation of the power-operated relief valves or steam dump valves. Demonstration of the safety valves' lift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Code. 3/4.4.3 PRESSURIZER The limit on the maximum water volume (1656 cubic feet) in the prassurizer assures that the parameter is maintained within the normal steady-stata envelope of operation assumed in the SAR. The limit is consistent with the initial SAR assumptions. The 12-hour periodic surveillance is sufficient to ensure that the parameter is restored to within its limit following expected transient operation. The maximum water volume also ensures that a steam bubble is formed and thus the RCS is not a hydraulically solid system. The requirement that a minimum number of pressurizer heaters be OPERABLE enhances the capability of the plant to control Reactor Coolant System pressure and establish natural circulation. 3/4.4.4 RELIEF VALVES The power-operated relief valves (PORVs) and steam bubble function to relieve RCS pressure during all design transients up to and including the design step load decrease with steam dump. Operation of the PORVs minimizes the undesirable opening of the spring-loaded pressurizer Code safety valves. Each PORV has a remotely operated block valve to provide a positive shutoff capability should a relief valve become inopera51e. BRAIDWOOD - UNITS 1 & 2 8 3/4 4-2
a m E U V.n. p REACTOR COOLANT SYSTEM MG 191986 BASES 3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be main-tained. The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1. Inservice inspection of steam generator tubing is essential in order to maintain surveil-lance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion. Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken. The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes. If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking. The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the Reactor Coolant System and the Secondary Coolant System (reactor-to-secondary leakage = 500 gallons per day per steam generator). Cracks having-a' reactor-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents. Operating plants have demonstrated that reactor-to-secondary leakage of 500 gallons per day per steam generator can readily be detected by radiation monitors of steam generator blowdown. Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged. Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant. However, even if a defect should develop in service, it will be found during scheduled inservice steam generator tube examinations. Plugging will be required for all tubes with imperfections exceeding the plugging limit of 40% of the tube nominal wall thickness. Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20% of the original tube wall thickness. Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be reported to the Commission pur-suant to Specification 6.9.2 prior to resumption of plant operation. Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary. I l BRAIDWOOD - UNITS 1 & 2 B 3/4'4-3 l
REACTOR COOLANT SYSTEM FINAL. N I 9 IS% HMT BASES 3/4.4.6 REACTOR COOLANT SYSTEMJ EAKAGE 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS The RCS Leakage Detection Systems required by this specification are provided to monitor and detect leakage from the reactor coolant pressure boundary. These Detection Systeras are consistent with the recommendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973. 3/4.4.6.2 OPERATIONAL LEAKAGE PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary. Therefore, the presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to be promptly placed in COLD SHUTDOWN. Industry. experience has shown that while a limited amount of leakage is expected from the RCS, the unidentified portion of this leakage can be reduced to a threshold value of less than 1 gpm. This threshold value is sufficiently low to ensure early detection of additional leakage. ' The total steam generator tube leakage limit of 1 gpm for all steam generators not isolated from the RCS ensures that the dosage contribution from the tube leakage will be limited to a small fraction of 10 CFR Part 100 dose guideline values in the event of either a steam generator tube rupture or steam line break. The 1 gpm limit is consistent with the assumptions used in the analysis of these accidents. The 500 gpd leakage limit per steam generator ensures that steam generator tube integrity is maintained in the event of a main steam line rupture or under LOCA conditions. The 10 gpm IDENTIFIED LEAKAGE limitation provides allowance for a limited amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the Leakage Detection Systems. The CONTROLLED LEAKAGE limitation restricts operation when the total flow supplied to the reactor coolant pump seals exceeds 40 gpm with the modulating valve in the supply line fully open at a nominal RCS pressure of 2235 psig. l This limitation ensures that in the cvent of a LOCA, the Safety Injection flow will not be less than assumed in the safety analyses. The 1 gpm leakage from any RCS pressure isolation valve is sufficiently low to ensure early detection of possible in-series check valve failure. It is apparent that when pressure isolation is provided by two in-series check valves and when failure of one valve in the pair can go undetected for a substantial length of time, verification of valve integrity is required. Since these valves are important in preventing overpressurization and rupture of the ECCS low pressure piping which could result in a LOCA that bypasses containment, those valves should be tested periodically to ensure low probability of gross failure. BRAIDWOOD - UNITS 1 & 2 8 3/4 4-4
REACTOR COOLANT SYSTEM MG19 Q BASES - OPERATIONAL LEAKAGE (Continued) The Surveillance Requirements for RCS pressure isolation valves provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA. Leakage from the RCS pressure isolation valves is IDENTIFIED LEAKAGE and will be considered a.s a portion of the allowed limit. 3/4.4.7 CHEMISTRY The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reactor Coolant System leakage or failure due to stress corrosion. Maintaining the chemistry within the Steady-State Lir.its 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 the structural integrity of the Reactor Coolant System. The time interval permitting continued operation within tha restrictions of the Transient Limits provides time for taking corrective acticas to restore the contaminant concen-trations to within the Steady-State Limiti. The Surveillance Requirements providt. adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action. 3/4.4.8 SPECIFIC ACTIVITY The limitations on the specific activity of the reactor coolant ensure that the resulting 2-hour doses at the SITE BOUNDARY will not exceed an appropriately small fraction of 10 CFR Part 100 dose guideline values following a steam generator tube rupture accident in conjunction with an assumed steady-state reactor-to-secondary steam generator leakage rate of 1 gpm. The values for the limits on specific activity represent limits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative in that specific site parameters of the Braidwood Station, 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, accommodatcs possible iodine spiking phenomenon which may J occur following changes in THERMAL POWER. Operation with specific activity i levels exceeding 1 microcurie / gram DOSE EQUIVALENT I-131 but within the ; limits shown on Figure 3.4-1 must be restricted to no more than 800 hours per ! year (approximately 10% of the' unit's yearly operating time) since the activity : levels allowed by Figure 3.4-1 increase the 2-hour thyroid dose at the SITE l B0UNDARY by a factor of up to 20 following a postulated steam generator tube 1 BRAIDWOOD - UNITS 1 & 2 8 3/4 4-5
REACTOR COOLANT SYSTEM FINAL HAFT AUG 191986 BASES SPECIFIC ACTIVITY (Continued) rupture. The reporting of cumulative operating time over 500 hours in any 6-month consecutive period with greater than 1 microcurie / gram DOSE EQUIVALENT I-131 will allow sufficient time for Commission evaluation of the circumstances prior to reaching the 800-hour limit. The sample analysis for determining the gross specific activity and E can exclude the radioiodines because of the low reactor coolant limit of 1 microcurie / gram DOSE EQUIVALENT I-131, and because, if the limit is exceeded, the radioiodine level is to be determined every 4 hours. If the gross specific activity level and radioiodine level in the reactor coolant were at their limits, the radioiodine contribution would be approximately 1%. In a release of reactor coolant with a typical mixture of radioactivity, the actual radiciodine contribution would be about 20%. The exclusion of radionuclides with half-lives less than 10 minutes from these determinations has 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 the reactor coolant to its release to the environ-ment and transport to the SITE BOUNDARY, which is relatable to at least 30 minutes decay time. The choice of 10 minutes for the half-life cutoff was made because of the nuclear characteristics of the typical reactor coolant radioactivity. The radionuclides in the typical reactor coolant have half-lives of less than 4 minutes or half-lives of greater than 14 minutes, which allows a distinct window for determination of the radionuclides above and below a half-life of 10 minutes. For these reasons the radionuclides that are excluded from consideration are expected to decay to very low levels before they could be transported from the reactor coolant to the SITE BOUNDARY under any accident condition. Based upon the above considerations for excluding certain radionuclides from the sample analysis, the allowable time of 2 hours between sample taking and completing the initial analysis is based upon a typical time necessary to perform the sampling, transport the sample, and perform the analysis of about 90 minutes. After 90 minutes, the gross count should be made in a reproducible geometry of sample and counter having reproducible beta or gamma self-shielding , properties. The counter should be reset to a reproducible efficiency versus l energy. It is not necessary to identify specific nuclides. The radio-l chemical determination of nuclides should be based on multiple counting of the sample with typical counting basis following sampling of less than 1 hour, about 2 hours, about I day, about 1 week, and about 1 month. 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 BRAIDWOOD - UNITS 1 & 2 8 3/4 4-6
REACTOR COOLANT SYSTEM BASES AUG 19 fong SPECIFIC ACTIVITY (Continued) take corrective action. Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomenon. A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data'obtained.
. 3/4.4.9 PRESSURE / TEMPERATURE LIMITS-The temperature and pressure changes during heatup and cooldown are limited to be consistent with the requirements given in the ASME Boiler and Pressure Vessel Code, Sed. ion 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:
- 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/hr and 200' F/hr respectively. The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 320*F, and
- 5. System preservice hydrotests and in-service leak and hydrotests shall be performed 't pressures in accordance with the requirm.ents of ASME Boiler and .'ressure Vessel Code, Section XI.
The fracture toughness properties of the ferritic materials in the reactor vessel are determined in accordance with the 1973 Summer Addenda to Section III
- of the ASME Boiler and Pressure Vessel and Code, l
l l ! BRAIDWOOD - UNITS 1 & 2 B 3/4 4-7
REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RTNDT, at the ed of l 32 effective full power years for Unit 1 (16 effective full pma years for Unit 2) of service life. The 32 EFPY for Unit 1 (16 EFPY for Unit 2) service life period is chosen such that the limiting RTNDT at the 1/4T location in t % core region is greater than the RT of the limiting unitradiatea material. NDT The selection of such a limiting RT assures that all components in the NDT Reactor Coolant System will be operated conservatively in accordance with applicable Code requirements. The reactor vessel materials have been tested to determine their initial RT ; the results of these tests are shown in Table B 3/4.4-1. Reactor ophItion and resultant fast neutron (E greater than 1 MeV) irradiation can cause an increase in the RT NDT. Therefore, an adjusted reference temperature, based upon the fluence, copper content and phosphorus content of the material in question, can be predicted using Figure B 3/4.4-1 and the largest value of ART NDT computed by either Regulatory Guide 1.99, Revision 1, " Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials" or the Westinghouse Copper Trend Curves shown in Figure B 3/4.4-2. The heatup and cooldown limit curves of Figures 3.4-2 and 3.4-3 include predicted adjustments for this shift in RT at the end of 32 EFPY for Unit 1 (16 EFPY HDT for Unit 2) as well as adjustments for possible errors in the pressure and temperature sensing instruments. Values of ART determined in this manner may be used until the results NDT from the material surveillance program, evaluated according to ASTM E185, are available. Capsules will be removed in accordance with the requirements of ASTM E185-73 and 10 CFR Part 50, Appendix H. The surveillance specimen with-drawal schedule is shown in Table 4.4-5. The lead factor represents the rela-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 the 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 NDT determined from the surveillance capsule exceeds the calculated ART NDT for the equivalent capsule radiation exposure. Allowable pressure-temperature relationships for various heatup and cooldown rates are calculated using methods derived from Appendix G in Section III of the ASME Boiler and Pressure Vessel Code as required by l Appendix G to 10 CFR Part 50, and these methods are discussed in detail in WCAP-7924-A, " Basis for Heatup and Cooldown Limit Curves," April 1975. BRAIDWOOD - UNITS 1 & 2 8 3/4 4-8
REACTOR COOLANT SYSTEM FINAL. R\FT BASES I9 $86 PRESSURE / TEMPERATURE LIMITS (Continued) 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 semi-elliptical 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 out' side of the vessel wall. The dimensions of this postulated crack, referred to in Appendix G of ASME Section III as the reference flaw, amply exceed the current capabilities of inservice inspection techniques. Therefore, the reactor opera-tion limit curves developed for this reference crack are conservative and provide sufficient safety margins for protection against non-ductile failure. To assure that the radiation embrittlement effects are accounted for in the calculation of the limit curves, the most limiting value of the nil-ductility reference temperature, RTNDT, is used and this includes the radiation-induced shift, ARTNDT, corresponding to the end of the period for which heatup and cocidown curves are generated. The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, Ky , for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, K IR' for the metal temperature at that time. K is obtained from the reference IR fracture toughness curve, defined in Appendix G to the ASME Code. The K IR curve is given by the equation: KIR = 26.78 + 1.223 exp [0.0145(T-RTNDT + 160)] (1) Where: K IR is the reference stress intensity factor as a function of the metal temperature T and the metal nil-ductility reference temperature RT Thus, NDT. the governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows: CKyg + kit 1Kyp (2) Where: Kyg = the stress intensity factor caused by membrane (pressure) stress, K It = the stress intensity factor caused by the thermal gradients, BRAIDWOOD - UNITS 1 & 2 B 3/4 4-9
REACTOR COOLANT SYSTEM FINAL DMT BASES PRESSURE / TEMPERATURE LIMITS (Continued) j KIR = constant provided by the code as a function of temperature relative ! to the RTNDT of the material, C= 2.0 for level A and B service limits, and C= 1.5 for inservice hydrostatic and leak test operations. At any tirre during the heatup or cooldown transient, K is determined by IR the metal temperature at the tip of the postulated flaw, the appropriate value for RTNDT, and the reference fracture toughness curve. The thermal stresses resulting from temperature gradierits through the vessel wall are. calculated and then the corresponding thermal str a intensity factor, KIT, for the reference flaw is computed. From Equation (2) the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated. COOLDOWN For the calculation of the allowable pressure versus coolant temperature
~
during cooldown, the Code reference flaw is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the wall because the thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates. Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations. From these relations, composite limit curves are constructed for each cooldown rate of interest. The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the 1/4T vessel location is at a higher temperature than the fluid adjacent to the vessel ID. This condition, of course, is not true for the steady-state situation. It follows that at any given reactor coolant temperature, the AT developed during
- cooldown results in a higher value of K 7g at the 1/4T location for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist such that the increase in K IR exceeds kit, the calculated allowable pressure during cooldown will be greater than the steady-state value, t
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. BRAIDWOOD - UNITS 1 & 2 8 3/4 4-10
TABLE B 3/4.4-la cri s REACTOR VESSEL TOUGHNESS - y (UNIT 1) o 8 Average
' Shelf Enercy T RT e MATERIAL Cu P NDT NDT NMWD* FWD **
5 COMPONENT Heat No. SPEC. % % *F F* ft-lbs ft-lbs d Closure Head Dome D1398-1 A533B, C1. 1 .06 .009 90 90 129 -
~ Closure Head Ring 49C1126-1-1 A508, C1. 3 .02 .009 -20 -20 123 -
- e. Closure Head Flange 2030-V-1 A508, C1. 2 .11 .009 -20 -20 163 -
to Vessel Flange 122N357VA1 A508, C1. 2 -
.010 -10 -10 106 -
Inlet Nozzle 21-3257 A508, C1. 2 .09 .008 -20 -20 144 - Inlet Nozzle 21-3257 A508, C1. 2 .09 .010 -10 -10 144 - Inlet Nozzle 22-3313 A508, C1. 2 .07 .008 -10 -10 130 - Ir.let Nozzle 22-3313 A508, C1. 2 .07 .010 0 0 115 - Outlet Nozzle 22-3025 A508, C1. 2 .13 .013 -10 -10 125 - Outlet Nozzle 4-3329 A508, C1. 2 .08 .009 -20 -20 156 - 5 Outlet Nozzle 4-3383 A508, C1. 2 .08 .008 -20 -20 147 - 1 Outlet Nozzle 11-5226 A508, C1. 2 .09 .007 -10 -10 125 -
= Nozzle Shell SP7016 A508, C1. 2 .04 .008 10 10 155 -
0 Upper Shell 490383/ A508, C1. 3 .05 .008 -30 -30 122 173 49C344-1-1 Lower Shell 490867/ A508, C1. 3 .03 .007 -20 -20 135 151 49C813-1-1 Bottom Head Ring 49D148-1-1 A508, C1. 3 .05 .008 -50 -50 147 - Bottom Head Dome C4882-1 A533B, C1. 1 .14 .010 -20 -20 123 - Upper Shell to WF-562 .04 .015 40 40 80 - Lower Shell Girth Weld Weld HAZ -70 <-10 151 - 9 mummum 2 N
- Normal to major working direction. g Major working direction.
O E A
TABLE B 3/4.4-1b E REACTOR VESSEL TOUGHNESS g (UNIT 2) 6 Average 8 Shelf Energy
. T RT MATERIAL Cu P NOT NDT NMWD" MWD **
c- COMPONENT HEAT NO. SPEC. % % *F F* ft-lbs ft-lbs 5 Closure Head Dome D9754-1 A5338, C1. 1 .16 .005 -60 -60 151 - d Closure Head Ring 50C478-1-1 A508, C1. 3 .05 .006 -30 -30 128 - g Closure Head Flange 2031-V-1 A508, C1. 2 -
.009 20 20 135 -
- e. Vessel Flange 124P455 A508, C1. 2 .07 .010 20 20 128 -
m Inlet Nozzle 41-5414 A508, C1. 2 .07 .008 -10 -10 137 - Inlet Nozzle 41-5414 A508, C1. 2 .07 .009 -10 -10 140 - Inlet Nozzle 42-5417 A508, Cl. 2 .09 .011 -10 -10 122 - Inlet Nozzle 42-5417 A508, C1. 2 .09 .009 -10 -10 116 - Outlet Nozzle 4-3502 A508, C1. 2 .09 .012 -10 -10 155 - Outlet Nozzle 11-5226 A508, C1. 2 .09 .009 -10 -10 116 - eo Outlet Nozzle 4-3481 A508, C1. 2 .07 .008 -10 -10 163 - w Outlet Nozzle 11-5266 A508, C1. 2- .09 .010 10 10 117 -
% Nozzle Shell SP7056 A508, C1. 2 .04 .005 30 30 115 - # Upper Shell 490963/ A508, C1. 3 .03 .007 -30 -30 119 147 4
49C904-1-1 Lower Shell 500102/ A508, C1. 3 .06 .006 -30 -30 144 168 50C97-1-1 Bottom Head Ring 47D1066-1-1 A508, C1. 3 .07 .008 -30 -30 156 - Bottom Head Dome D1429-1 A533B, C1. 1 .11 .010 -20 -20 120 - Upper Shell to WF-562 .04 .015 40 40 80 - Lower Shell Girth Weld Weld HAZ -30 -30 145 -
- Normal to major working direction.
** Major working direction. g
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REACTOR COOLANT SYSTEM FINAL. RAFT ) BASES PRESSURE / TEMPER- 2E LIMITS (Continued) The notch the cooldown curve of Figure 3.4-3 is due to the added con-straint on the vessel closure flange given in Appendix G of 10 CFR 50. This constraint requires that, at pressures greater than 20% of the preservice system hydrostatic test pressure, the flange regions that are highly stressed by the bolt preload must exceed the RT f the material by at least 120 F. The NOT flange R NDT + 120*F impinges on the cooldown curves and therefore the notch is required HEATUP l Three separate calculations are required to determine the limit curves l for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions ss well as finite heatup rate conditions assuming the presence of a 1/4T Mfect at the inside of the vessel wall. The thermal gradients during heatup
#0 duce compressive stresses at the inside of the wall that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the K f r the 1/4T crack IR during heatup is lower than the K for the 1/4T crack during steady-state IR 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 different K IR 's for steady-state and finite heatup rates do not offset each other and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/4T flaw is considered. Therefore, both cases have to 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 I 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. BRAIDWOOD - UNITS 1 & 2 B 3/4 4-15
REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) : 1 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. The OPERABILITY of two PORVs, or two RHR suction valves, or an RCS vent opening of at least 2 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 overpres-surization when the transient is limited to either: (1) the start of an idle RCP with the secondary water temperature of the steam generator less than or equal to 50*F above the RCS cold leg temperatures, or (2) the start of a centrifugal charging pump and its injection into a water solid RCS. These two scenarios are analyzed to determine the resulting overshoots assuming a single PORV actuation with a stroke time of 2.0 seconds from full closed to full open. Figure 3.4-4 is based upon this analysis and represents the maximum allowable PORV variable setpoint such that, for the two overpres-surization transients noted, the resulting pressure will not exceed the nominal 10 effective full power years (EFPY) Appendix G reactor vessel NDT limits. RHR RCS suction isolation valves 8701A and 8702A are interlocked with an "A" train wide range,oressure transmitter and valves 87018 and 8702B are inter-locked with a "B" train wide range pressure transmitter. Removing power from valves 8701B and 8702A, prevents a single failure from inadvertently isolating
- both RHR suction relief valves while maintaining RHR isolation capability for both RHR flow paths.
3/4.4.10 STRUCTURAL INTEGRITY ! The inservice inspection ar-i 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 th plant. These programs are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50.55a(g) except where specific written relief has been granted by the . Commission pursuant to 10 CFR 50.55a(g)(6)(1). l l BRAIDWOOD - UNITS 1 & 2 8 3/4 4-16
REACTOR COOLANT SYSTEM FINAL MAFT AUG 19 m BASES STRUCTURAL INTEGRITY (Continued) Components of the Reactor Coolant System were designed to provide access to permit inservice inspections in accordance with Section XI of the ASME Boiler and Pressure Vessel Code, 1974 Edition and Addenda through Summer 1975. 3/4.4.11 REACTOR VESSEL HEAD VENTS , Reactor Coolant System 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 a reactor vessel head vent path ensures the capability exists to perform this function. The valve redundancy of the Reactor Coolant System vent paths serves to minimize the probability of inadvertent or irreversible actuation while ensuring that a single vent valve power supply or control system does not prevent isolation of the vent path. The function, capabilities, and testing requirements of the Reactor Coolant System Vent Systems are consistent with the requirements of Item II.B.1 of NUREG-0737, " Clarification of TMI Action Plan Requirements," November 1980.
)
l l
)
i l BRAIDWOOD - UNITS 1 & 2 B 3/4 4-17
I a i 1 3/4.5 EMERGENCY CORE COOLING SYSTEMS I - - AUG 19 q BASES l 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 ' core through each of the cold legs in the event the RCS pressure falls below the pressure of the accumulators. This initial surge of water into the core provides the initial cooling mechanism during large RCS pipe ruptures. The limits on accumulator volume, boron concentration and pressure ensure that the assumptions used for accumulator injection in the safety analysis are met. A contained borated water level between 31% and 63% ensures a volume of greater than or equal to 6995 gallons but less than or equal to 7217 gallons. 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 permissbi conditions are not met. In addition, as these accumulator isolation valves fail to meet single failure criteria, removal of power to the valves is required. The limits for operation with an accumulator inoperable for any reason except an isolation valve closed minimizes the time exposure of the plant to a LOCA event occurring concurrent with failure of an additional accumulator which may result in unacceptable peak cladding temperatures. If a closed isolation valve cannot be immediately opened, the full capability of one accumulator is not available and prompt action is required to place the reactor in a mode where this capability is not required. The requirement to verify accumulator isolation valves shut with power removed from the valve operator when the pressurizer is solid ensures the accumulators will not inject water and cause a pressure transient when the Reactor Coolant System is on solid plant pressure control. 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS The OPERABILITY of two independent ECCS subsystems ensures that sufficient emergency core cooling capability will be available in the event of a LOCA assuming the loss of one subsystem through any single failure consideration. Either subsystem operating in conjunction with the accumulators is capable of supplying sufficient core cooling to limit the peak cladding temperatures l within acceptable limits for all postulated break sizes ranging from the i double ended break of the largest RCS cold leg pipe downward. In addition, each ECCS subsystem provides long-term core cooling capability in the recirculation mode during the accident recovery period. With the RCS temperature below 350*F, one OPERABLE ECCS subsystem is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the limited core cooling requirements. BRAIDWOOD - UNITS 1 & 2 B 3/4 5-1 1
l EMERGENCY CORE COOLING SYSTEMS RNAL MAFT MG 191996 BASES ECCS SUBSYSTEMS (Continued) l The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps and Safety Injection pumps except the required OPERABLE charging pump to be inoper-able below 330*F provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV. , l The Surveillance Requirements provided to ensure OPERABILITY of each l component ensures that at a' minimum, the assumptions used in the safety analyses are met and that subsystem OPERABILITY is maintained. Surveillance Requirements for throttle valve position stops and flow balance testing provide assurance that proper ECCS flows will be maintained in the event of a LOCA. Maintenance of proper flow resistance and pressure drop in the piping system to each injection point is necessary to: (1) prevent total pump flow from exceeding runout conditions when the system is in its minimum resistance configuration, (2) provide the proper flow split between injection points in accordance with the assumptions used in the ECCS-LOCA analyses, and (3) provide an acceptable level of total ECCS flow to all injection points equal to or above that assumed in the ECCS-LOCA analyses. The Surveillance Requirements for leakage testing of ECCS check valves ensures that a failure of one valve will not cause an intersystem LOCA. 3/4.5.4 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. 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, and (2) the reactor will remain subcritical in the cold condition following mixing of the RWST and the RCS water volumes with all control rods inserted except for the most reactive control assembly. These assumptions are consistent with the LOCA analyses. The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics. A minimum contained borated water level of 89% ensures a volume of greater than or equal to 395,000 gallons.
- The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between 8.5 and 11.'O 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.
BRAIDWOOD - UNITS 1 &.2 8 3/4 5-2
3/4.6 CONTAINMENT SYSTEMS
- t BASES I9 %
3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 CONTAINMENT INTEGRITY Primary CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the safety analyses. This restriction, in conjunction with the leakage rate limitation, will limit the SITE BOUNDARY radiation doses to within the dose guideline values of 10 CFE Part 100 during accident conditions. 3/4.6.1.2 CONTAINMENT LEAKAGE The limitations on containment leakage rates ensure that the total containment leakage volume will not exceed the value assumed in the accident analyses at the peak accident pressure, P,. As an added conservatism, the measured overall integrated leakage rate is further limited to less than or equal to 0.75 L, or 0.75 L t, as applicable, during performance of the periodic test to account for possible degradation of the containment leakage barriers between leakage tests. The surveillance testing for measuring leakage rates are consistent with the requirements of Appendix J of 10 CFR Part 50. 3/4.6.1.3 CONTAINMENT AIR LOCKS The limitations on closure and leak rate for the containment air locks are required to meet the restrictions on CONTAINMENT INTEGRITY and containment leak rate. Surveillance testing of the air lock seals provides assurance that the overall air lock leakage will not become excessive due to seal damage during the intervals between air lock leakage tests. 3/4.6.1.4 INTERNAL PRESSURE The limitations on containment internal pressure ensure that: (1) the containment structure is prevented from exceeding its design negative pressure differential with respect to the outside atmosphere of 0.1 psig, and (2) the containment peak pressure does not exceed the design pressure of 50 psig during steam line break conditions. The maximum increase in peak pressure expected to be obtained from a cold leg double-ended break event is 44.4 psig. The limit of 1.0 psig for initial positive containment pressure will limit the total pressure to 44.4 psig, which is higher than the FSAR Chapter accident analysis calculated peak pres-sure assuming a limit of 0.3 psig for initial positive containment pressure, but is considerably less than the design pressure of 50 psig. BRAIDWOOD - llNITS 1 & 2 B 3/4 6-1
CONTAINMENT SYSTEMS BASES AUG 19 7986 3/4.6.1.5 AIR TEMPERATURE The limitations on containment average air temperature ensure that the overall containment average air temperature does not exceed the initial temperature condition assumed in the accident analysis for a steam line break accident. Measurements shall be made at all of the listed running fan locations, whether by fixed or portable instruments, to determine the average air temperature. 3/4.6.1.6 CONTAINMENT VESSEL 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 withstand the maximum pressure of 44.4 psig in the event of a cold leg double ended break accident. The measurement of containment tendon lift-off force, the tensile tests of the tendon wires or strands, 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 reccamendations of proposed. Rev. 3 to Regulatory Guide 1.35, " Inservice Surveillance of Ungrouted Tendons in Prestressed Concrete Containment Structures," April 1979 and proposed Regulatory Guide 1.35.1, " Determining Prestressing Forces for Inspection of Prestressed Concrete Containments," April 1979. The required Special Reports from any engineering evaluation of containment abnormalties shall include a description of the tendon condition, the condition of the concrete (especially at tendon anchorages), the inspection procedure, the tolerances on cracking, the results of the engineering evaluation and the corrective actions taken. 3/4.6.1.7 CONTAINMENT PURGE VENTILATION SYSTEM The 48-inch containment purge supply and exhaust isolation valves are i required to be sealed closed (power removed) during plant operations since these j valves have not been demonstrated capable of closing during a LOCA or steam line break accident. Maintaining these valves sealed closed during plant operation ensures that excessive quantities of radioactive material will not be released via the Containment Purge System. To provide assurance that the 48-inch contain-ment valves cannot be inadvertently opened, the valves are sealed closed in accordance with Staadard 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 8-inch purge supply and exhaust isolation valves since, unlike the 48-inch valves, the 8-inch valves are capable of closing during a LOCA or steam line break accident. Therefore, the SITE BOUNDARY dose guideline values of 10 CFR Part 100 would not BRAIDWOOD - UNITS 1 & 2 B 3/4 6-2
CONTAINMENT SYSTEMS FINAL DRAFT BASES O CONTAINMENT PURGE VENTILATION SYSTEM (Continued) be exceeded in the event of an accident during containment purging operation. Operation with one line open will be limited to 1000 hours during a i:alendar - year. Leakage integrity tests with a maximure allowable leakage rate for containment purge supply and axhaust supply valves will provide early indica-tion of resilient material seal degradation and will allow opportunity for repair before gross leakage failures could develop. The 0.60 L leakage limit
*of Specification 3.6.1.2.b. shall not be exceeded when the leak $ge rates determined by the leakage integrity tests of these valves are added to the previously determined total for all valves and penetrations subject to Type B and C tests.
3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS 3/4.6.2.1 CONTAINMENT SPRAY SYSTEM The OPERABILITY of the Containment Spray System ensures that containment depressurization and cooling capability will be available in the event of a LOCA cr steam line break. The pressure reduction and resultant lower containment leakage rate are consistent with the assumptions used in the safety analyses. The Containment Spray System and the Containment Cooling System are redundant to each other in providing post-accident cooling of the containment atmosphere. However, the Containment Spray System also provides a mechanism ! for removing iodine from the containment atmosphere and therefore the time requirements for restoring an inoperable Spray System to OPERABLE status have been maintained consistent with that assigned other inoperable ESF equipment. 3/4.6.2.2 SPRAY ADDITIVE SYSTEM The OPERABILITY of the Spray Additive System ensures that sufficient NaOH is added to the containment spray in the event of a LOCA. The limits on NaOH ' volume and concentration ensure a pH value of between 8.5 and 11.0 for tne ! solution recirculated within containment after a LOCA. This pH band minimizes i the evolution of iodine and minimizes the effect of chloride and caustid stress ! corrosion on mechanical systems and components. The contained solution volume l limit includes an allowance for solution not usable because of tank discharge line location or other physical characteristics. These assumptions are consistent with the iodine removal efficiency assumed in the safety analyses. A spray additive tank level of between 78.6% and 90.3% ensures a volume of greater than or equal to 4000 gallons but less than or equal to 4540 gallons. BRAIDWOOD - UNITS 1 & 2 8 3/4 6-3
l CONTAINMENT SYSTEMS BASES 3/4.6.2.3 CONTAINMENT COOLING SYSTEM The OPERABILITY of the Containment Cooling System ensures that: (1) the containment air temperature will be maintained within linits during normal operation, and (2) adequate heat removal capacity is available when operated - in conjunction with the Containment Spray Systems during post-LOCA con'.itions. l The Containment Cooling System and the Containment Spray System are redundant to each other in providing post-accident cooling of the containment atmosphere. As a result of this redundancy in cooling capability, the allowable out-of-service time requirements for the Containment Cooling System have been appropriately adjusted. However, the allowable out-of service time require-I ments for the Containment Spray System have been maintained consistent with that assigned other inoperable ESF equipment since the Containment Spray System also provides a mechanism for removing iodine from the containment atmosphere. 3/4.6.3 CONTAINMENT ISOLATION VALVES The OPERABILITY of the containment isolation valves ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressurization of the containment and is consistent with the requirements of GDC 54 thru 57 of Appendix A to 10 CFR Part 50. Containment isolation within the time limits specified for those isolation valves designed to close auto-matically ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a LOCA. 3/4.6.4 COMBUSTIBLE GAS CONTROL The OPERABILITY of the equipment and systems required for the detection and control of hydrogen g s ensures that this equipment will be available to maintain the hydrogen concentration within containment below its flammable limit during post-LOCA conditions. Either recombiner unit (or the Purge System) is capable of controlling the expected hydrogen generation associated with: (1) zirconium-water reactions, (2) radiolytic decomposition of water, and (3) corrosion of metals within containment. These Hydrogen Control Systems l are consistent with the recommendations of Regulatory Guide 1.7, " Control of Combustible Gas Concentrations in Containment Following a LOCA," March 1971. The Hydrogen Mixing Systems are provided to ensure adequate mixing of the containment atmosphere following a LOCA. This mixing action will prevent localized accumulations of hydrogen from exceeding the flammable limit. BRAIDWOOD - UNITS 1 & 2 B 3/4 6-4 _-e- - - - - - - , - - - - -
d 3/4.7 PLANT SYSTEMS FINAL DRAFT l AUG 191986 BASES 3/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAFETY VALVES - The OPERABILITY of the main steam line Code safety valves ensures that the Secondary Coolant System pressure will be limited to within 110% (1320 psia) of its design pressure of 1200 psia during the most severe anticipated system operational transient. The maximum relieving capacity i:, associated with a turbine trip from 102% RATED THERMAL POWER coincident with an assumed loss of condenser heat sink (i.e., no steam dumps to the condenser). The specified valve lift settings and relieving capacities are in accordance with the requirements of Section III of the ASME Boiler and Pressure Code, 1971 Edition. The total relieving capacity for all valves on all of the steam lines is 17.958 x 10s lbs/h which is 119% of the total secondary steam flow of 15.135 x 108 lbs/h at 100% RATED THERMAL POWER. A minimum of two OPERABLE safety valves per steam generator ensures that sufficient relieving capacity is available for the allowable THERMAL POWER restriction in Table 3.7-1. ' STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis of the reduction in Secondary Coolant System steam flow and THERMAL POWER required by the reduced Reactor trip settings of the Power Range Neutron Flux channels. The Reactor Trip Setpoint reductions ere derived on the following bases: For four loop operation: SP = (X) - (Y)(V) x (109). Where: SP = Reduced Reactor Trip Setpoint in percent of RATED THERMAL POWER, V = Maximum number of inoperable safety valves per steam line. BRAIDWOOD - UNITS 1 & 2 8 3/4 7-1
PLANT SYSTEMS MG19 y
~
~
BASES l SAFETY VALVES (Continued) 109 = Power Range Neutron Flux-High Trip Setpoint for four loop operation, X = Total relieving capaci'ty of all safety valves per steam line in 1bs/ hour, and
=
Y Maximum relieving capacity of any one safety valve in lbs/ hour. 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. The motor-driven auxiliary feedwater pump is capable of delivering a total feedwater flow of 740 gpm at a pressure of 1450 psig to the entrance of the steam generators. The diesel-driven auxiliary feedwater pump is cap *able of delivering a total feedwater flow of 740 gpm at a pressure of 1450 psig to the entrance of the steam generators. This capacity is sufficient to ensure that adequate feedwater flow is available to remove decay heat and reduce the Reactor Coolant System temperature to less than 350*F when the RHR System may be placed into operation. 3/4.7.1.3 CONDENSATE STORAGE TANK The OPERABILITY of the condensate storage tank with the minimum water level of 40% ensures that sufficient water (200,000 gallons) is available to maintain the RCS at HOT STANDBY conditions for 9 hours with steam discharge to the atmosphere concurrent with total loss-of-offsite power. 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 break. This dose also includes the effects of a coincident 1 gpa reactor 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. BRAIDWOOD - UNITS 1 & 2 B 3/4 7-2
PLANT SYSTEMS FINAL. HAF
~
AUG 191986 BASES 3/4.7.1.5 MAIN STEAM LINE ISOLATION VALVES The OPERABILITY of the main steam line isolation valves ensures that no more than one steam generator will blowdown in the event of a steam line rupture. This restriction is required to: (1) minimize the positive reactivity 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 Requirements are consistent with the assumptions used in the safety analyses. 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION The limitation on steam generator pressure and temperature ensures that the pressure-induced stresses in the steam generators do not exceed the maximum allowable fractcre tougnness stress limits. The limitations of 70*F and 200 psig are based on a steam generator RT to prevent brittle fracture. NDT f 60 F and are sufficient 3/4.7.3 COMPONENT COOLING WATER SYSTEM The OPERABILITY of the Component Cooling Water System ensures that sufficient 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. 3/4.7.4 ESSENTIAL SERVICE WATER SYSTEM The OPERABILITY of the Essential Service Water System ensures that sufficient 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 accident conditions within acceptable limits. 3/4.7.5 ULTIMATE HEAT SINK The limitations on the ultimate heat sink level and temperature ensure that sufficient cooling capacity is available to either 1) provide normal cooldown of the facility, or 2) to mitigate the effects of accident conditions within acceptable limits. The limitations on minimum water level and maximum temperature are based on providing a 30-day cooling water supply to safety related equipment without exceeding their design basis temperature and is consistent with the recommenda-tions of Regulatory Guide 1.27, " Ultimate Heat Sink for Nuclear Plants," March 1974. BRAIDWOOD - UNITS 1 & 2 B 3/4 7-3
PLANT SYSTEMS FINAL DRAFT AUG 191985 BASES 3/4.7.6 CONTROL ROOM VENTILATION SYSTEM The OPERABILITY of the Control Room Ventilation System ensures that: (1) the ambient air temperature does not exceed the allowable temperature for continuous duty rating for the equipment and instrumentatioa cooled by this system, and (2) the control room will remain habitable for cperations personnel during and following all credible accident conditions. Operation of the system with the heaters operating for at least 10 continuous hours in a 31-day period is sufficient to reduce the buildup of moisture on the adsorbers and HEPA filters. The OPERABILITY of this system in conjunction with control room design provisions is based on limiting the radiation exposure to personnel occupying the control room to 5 rem or less whole body, or its equivalent. This limitation is consistent with the requirements of General Design Cri-terion 19 of Appendix A,10 CFR Part 50. ANSI N510-1980 will be used as a procedural guide for surveillance testing. 3/4.7.7 NON-ACCESSIBLE AREA EXHAUST FILTER PLENUM VENTILATION SYSTEM The OPERABILITY of the Non-Accessible Area Exhaust Filter Plenum Ventilation System ensures that radioactive materials leaking from the ECCS equipment within the pump rooms following a LOCA are filtered prior to reaching the environment. The operation of this system and the resultant effect on offsite dosage calcu-lations was assumed in the safety analyses. ANSI N510-1980 will be used as a procedural guide for surveillance testing. 3/4.7.8 SNUBBERS All snubbers are required OPERABLE to ensure that the structural integrity of the Reactor Coolant System and all other safety-related systems is main-tained during and following a seismic or other event initiating dynamic loads. Snubbers are classified and grouped by design and manufacturer but not by size. For example, mechanical snubbers utilizing the same design features of the 2-kip,10 ' 'p, and 100-kip capacity manufactured by Company "A" are of the same type. The same design mechanical snubbers manufactured by Company "B" for the purposes of this specification would be of a different type, as would hydraulic snubbers from either manufacturer. 1 A list of incividual snubbers with detailed information of snubber loca- l tion and size and of systems affected shall be available at the plant in accor-dance with Section 50.71(c) of 10 CFR Part 50. The accessibility of each snubber shall be determined and approved by the Onsite Review and Investigative Function. The determination shall be basea upon the existing radiation levels and the expected time to perform a visual inspection in each snubber location as well as other factors associated with accessibility during plant operations (e.g., temperature, atmosphere, location etc.), and the recommendations of Regu-latory Guides 8.8 and 8.10. The addition or deletion of any hydraulic or mechan-ical snubber shall be made in accordance,with Section 50.59 of 10 CFR Part 50. BRAIDWOOD - UNITS 1 & 2 B 3/4 7 ,4
PLANT SYSTEMS AUG 19 Iggg BASES SNUBBERS (Continued) The visual inspection frequency is based upon maintaining a constant level of snubber protection during an earthquake or severe transient. There-fore, the required inspection interval varies inversely with the observed snubber failures on a given type and is determined by the number of inoperable snubbers found during an inspection of each type. In order to establish the inspection frequency for each type of snubber on a safety related system, it was assumed that the frequency of snubber failures and initiating events is constant with time and that the failure of any snubber on that system could cause the system to be unprotected and to result in failure during an initiating event. Inspec-tions performed before that interval has elapsed may be used as a new reference point to determine the next inspection. 'However, the results of such early inspections performed before the original required time interval has elapsed (nominal time less 25%) may not be used to lengthen the reqaired inspection interval. Any inspection whose results require a shorter hapection interval will override the previous schedule. The acceptance criteria are to be used in the visual inspection to deter-mine OPERABILITY of the snubbers. For example, if a fluid port of a hydraulic snubber is found to be uncovered, the snubber shall be declared inoperable and shall not be determined OPERABLE via functional testing. To provide assurance of snubber functional reliability, one of three functional testing methods are used with the stated acceptance criteria:
- 1. Functionally test 10% of a type of snubber with an additional 10%
tested for each functional testing failure, or
- 2. Functionally test a sample size and determine sample acceptance or rejection using Figure 4.7-1, or
- 3. Functionally test a representative sample size and determine sample acceptance or rejection using the stated equation.
Figure 4.7-1 was developed using "Wald's Sequential Probability Ratio Plan" as described in " Quality Control and Industrial Statistics" by Acheson J. Duncan. Permanent or other exemptions from the surveillance program for individual snubbers may ba granted by the Commission if a justifiable basis for exemption is presented and, if applicable, snubber life destructive testing was performed to qualify the snubber for the applicable design conditions at either the com-pletion of their fabrication or at a subsequent date. Snubbers so exempted shall be listed in the list of individual snubbers indicating the extent of the l exemptions. The service life of a snubber is established via manufacturer input and information through consideration of the snubber service conditions and associated installation and maintenance records (nowly installed snubbers, seal , replaced, spring replaced, in high radiation area, in high temperature area, i BRAIDWOOD - UNITS 1 & 2 8 3/4 7-5
PLANT SYSTEMS AUG 19gggg BASES SNUBBERS (Continued) 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. 3/4.7.9 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. 3/4.7.10 FIRE SUPPRESSION SYSTEMS The OPERABILITY of the Fire Suppression Systems ensures that adequate fire suppression capability is available to confine and extinguish fires occurring in any portion of the facility where safety-related equipment is located. The Fire Suppression System consists of the water system, foam, spray, and/or sprinklers, CO2 , Halon, and fire hose stations. The collective capability of the Fire Suppression Systems is adequate to minimize potential damage to safety-related equipment and is a major element in the facility Fire Protection Program. In the event that portions of the Fire Suppression Systems are inoperable, l alternate backup fire-fighting equipment is required to be made available in the affected areas until the inoperable equipment is restored to service. When the inoperable fire-fighting equipment is intended for use as a backup means of fire suppression, a longer period of time is allowed to provide an I alternate means of fire fighting than if the inoperable equipment is the pri-mary means of fire suppression. 3750 gpm pump capacity is based on NFPA 20, 1981 standards which call for 150% of rated capacity (2500 gpm) at 65% of dis-charge pressure. The Surveillance Requirements provide assurance that the minimum OPERABILITY requirements of the Fire Suppression Systems are met. An allowance is made for ensuring a sufficient volume of Halon in the Halon storage tanks by verifying either the weight or the level of the tanks. Level measurements are niade by either a U.L. or F.M. approved method. BRAIDWOOD - UNITS 1 & 2 B 3/4 7-6
PLANT SYSTEMS BASES 3/4.7.10 FIRE, SUPPRESSION SYSTEMS (Continued) In the event the Fire Suppression Water System becomes inoperable, immediate corrective measures must be taken since this system provides the major fire suppression capability of the plant. 3/4.7.11 FIRE RATED ASSEMBLIES The functional integrity of the fire rated assemblies and barrier penetrations ensures that fires will be confined or adequately retarded from spreading to adjacent portions of the facility. These design features minimize the possibility of a single fire rapidly involving several areas of the facility prior to detection of and the extinguishing of the fire. The fire barrier penetrations are a passive element in the facility fire protection program and are subject to periodic inspections. Fire barrier penetrations, including cable penetration barriers, fire doors and dampers are considered functional when the observed condition is the same as the as-designed condition. For those fire barrier penetrations that are not in the as-designed condition, an evaluation shall be performed to show that the modification has not degraded the fire rating of the fire barrier penetration. During periods of time when a barrier is not functional, either: (1) a continuous fire watch is required to be maintained in the vicinity of the affected barrier, or (2) the fire detectors on at least one side of the affected barrier must be verified OPERABLE and an hourly fire watch patrol established, until the barrier is restored to functional status. 3/4.7.12 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. W BRAIDWOOD - UNITS 1 & 2 B 3/4 7-7
3/4.8 ELECTRICAL POWER SYSTEMS AUG 1919g5 BASES 3/4.8.1, 3/4.8.2 and 3/4.8.3 A.C. SOURCES, D.C. SOURCES, and ONSITE POWER DISTRIBUTION The OPERABILITY of the A.C. and D.C power sources and associated distribution systems during operation ensures that sufficient power will be available to supply the safety-related equipment required for: (1) the safe shutdown of the facility, and (2) the mitigation and control of accident condi-tions within the facility. The minimum specified independent and redundant A.C. and D.C. power sources and distribution systems satisfy the requirements of General Design Criterion 17 of Appendix A to 10 CFR Part 50. The ACTION requirements specified for the levels of degradation of the power sources provide restriction upon continued facility operation commensurate with the level of degradation. The OPERABILITY of the power sources are consistent with the initial condition assumptions of the safety analyses and are based upon maintaining at least one redundant set of onsite A.C. and D.C. power sources and associated distribution systems OPERABLE during accident conditions coincident with an assumed loss of-offsite power and single failure of the other onsite A.C. source. The A.C. and D.C. source allowable out-of-service times are based on Regulatory Guide 1.93, " Availability of Electrical Power Scurces," December 1974. When one diesel generator is inoperable, there is an additional ACTION requirement to verify that all required systems, subsystems, trains, components and devices, that depend on the remaining OPERABLE diesel generator as a source of emergency power, are also OPERABLE, and that the diesel-driven auxiliary feedwater pump is OPERABLE. This require-ment is intended to provide assurance that a loss-of-offsite power event will not result in a complete loss of safety function of critical systems during the period one of the diesel generators is inoperable. The term verify as used in this context means to administratively check by examining logs or other information to determine if certain components are out-of-service for maintenance or other reasons. It does not mean to perform the Surveillance Requirements needed to demonstrate the OPERABILITY of the component. The OPERABILITY of the minimum specified A.C. and D.C. power sources and associated distribution systems during shutdown and refueling ensures that: (1) the facility can be maintained in the shutdown or refueling condition for extended time periods, and (2) sufficient instrumentation and control capability i is available for monitoring and maintaining the unit status. ~ The Surveillance Requirements for demonstrating the OPERABILITY of the diesel generators are in accordance with the' recommendations of Regulatory Guides 1.9, " Selection of Diesel Generator Set Capacity for Standby Power Supplies," March 10, 1971, 1.108, " Periodic Testing of Diesel Generator Units Used as Onsite Electric Power Systems et Nuclear Power Plants," Revision 1, August 1977, and 1.137, " Fuel-011 Systems for Standby Diesel Generators," Revision 1, October 1979. The station chose its largest emergency load to be the SX pump. 'The maximum BHP of the SX pump is 1247 per FSAR Table 8.3-1. A BHP of 1247 corresponds to a load of 1034 kW. BRAIDWOOD - UNITS 1 & 2 B 3/4 8-1
ELECTRIC POWER SYSTEMS BASES I9 % A.C. SOURCES, D.C. SOURCES, AND ONSITE POWER DISTRIBUTION (Continued) The Surveillance Requirement for demonstrating the OPERABILITY of the station batteries is based on the recommendations of Regulatory Guide 1.129,
" Maintenance Testing and Replacement of Large Lead Storage Batteries for Nuclear Power Plants," February 1978, and IEEE Std 450-1980, "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Large Lead Storage Batteries for Generating Stations and Substations."
Verifying average electrolyte temperature above the minimum for which the battery was sized, total battery tarminal voltage on float charge, and the performance of battery service and discharje tests ensures the effectiveness of the charging system, the ability to handle high discharge rates and compares the battery capacity at that time with the rated capacity. Table 4.8-2 specifies the normal limits for each designated pilot cell and each connected cell for electrolyte level, float voltage and specific gravity. The limits for the designated pilot cells float voltage and specific gravity, greater than 2.13 volts and 0.015 below the manufacturer's full charge specific gravity or a battery charger current that had stabilized at a low value, is characteristic of a charged cell with adequate capacity. The normal limits for each connected cell for float voltage and specific gravity, greater than 2.13 volts and not more than 0.020 below the manufacturer's full charge specific gravity with an average specific gravity of all the connected cells not more than 0.010 below the manufacturer's full charge specific gravity, ensures the OPERABILITY and capability of the battery. Operation with a battery cell's parameter outside the normal limit but within the allowable value specified in Table 4.8-2.is permitted for up to 7 days. During this 7-day period: (1) the allowable values for electrolyte level ensures no physical damage to the plates with an adequate electron transfer capability; (2) the allowable value for the average specific gravity of all the cells, not more than 0.020 below the manufacturer's recommended full charge specific gravity, ensures that the decrease in rating will be less than the safety margin provided in sizing; (3) the allowable value for an individual cell's specific gravity, ensures that an individual cell's specific gravity will not be more than 0.040 below the manufacturer's full charge specific gravity and that the overall capability df the battery will be maintained within an acceptable limit; and (4) the allowable value for an individual cell's float voltage, greater than 2.05 volts, ensures the battery's capability to perform its design function. l j BRAIDWOOD - UNITS 1 & 2 B 3/4 8-2
ELECTRICAL POWER SYSTEMS i AUG 191986 BASES I l 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES Containment electrical penetrations and penetration conductors.are protected by either deenergizing circuits not required during reactor operation or by demonstrating the OPERABILITY of primary and backup overcurrent protective circuit breakers during periodic surveillance. The Surveillance Requirements applicable to lower voltage circuit breakers provide assurance of breaker reliability by testing at least one representative sample of each manufacturer's brand of circuit breaker. Each manufacturer's molded case and metal case circuit breakers are grouped into representative samples which are then tested on a rotating basis to ensure that all breakers are tested. If a wide variety exists within any manufacturer's brand of circuit breakers, it is necessary to divide that manufacturer's breakers into groups and treat each group as a separate type of breaker for surveillance purposes. The OPERABILITY of the motor-operated valves thermal overload protective devices ensures that these devices will not prevent safety-related valves from performing their function. The Surveillance Requirements for demonstrating the OPERABILITY of these devices are in accordance with Regulatory Guide 1.106,
" Thermal Overload Protection for Electric Motors on Motor Operated Valves,"
Revision 1, March 1977. I G BRAIDWOOD - UNITS 1 & 2 B 3/4 8-3 t
3/4.9 REFUELING OPERATIONS BASES 3/4.9.1 BORON CONCENTRATION The limitations on reactivity conditions during REFUELING ensure that: (1) tne reactor will remain subcritical during CORE ALTERATIONS, and (2) a uniform boron concentration is maintained for reactivity control in the water volume having direct access to the reactor vessel. The limitation on Keff of no greater than 0.95 is sufficient to prevent reactor criticality during refueling operations and includes a 1% ak/k conservative allowance for uncertainties. Similarly, the boron concentration value of 2000 ppm or greater includes a conservative uncertainty allowance of 50 ppm. These limitations are consistent with the initial cc 'itions assumed for the boron dilution incident in the safety analyses. ine locking closed of the required valves during refueling operations precludes the possibility of uncontrolled boron dilution of the filled portions of the RCS. This action prevents flow to the RCS of unborated water by closing flow paths from sources of unborated water. 3/4.9.2 INSTRUMENTATION The OPERABILITY of the Source Range Neutron Flux Mcnitors ensures that redundant monitoring capability is available to detect enanges in the reactivity condition of the core. 3/4.9.3 DECAY TIME The minimum requirement for reactor subcriticality prior to movement of irradiated fuel assemblies in the reactor vessel ensures that sufficient time has elapsed to allow the radioactive decay of the short-lived fission products. This decay time is consistent with the assumptions used in the safety analyses. 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS The requirements on containment building penetration closure and OPERABILITY ensure that a release of radioactive material within containment will be restricted from leakage to the environment. The OPERABILITY and closure restrictions are sufficient to restrict radioactive material release from a fuel element rupture based upon the lack of containment pressurization potential while in the REFUELING MODE. l The Braidwood Station is designed such that the containment opens into the fuel building through.the personnel hatch or equipment hatch. In the event of a fuel drop accident in the containment, any gaseous radioactivity escaping from the containment building will be filtered through the Fuel Handling Building Exhaust Ventilation System. [ 3/4.9.5 COMMUNICATIONS The requirement for comunications capability ensures that refuel.ing , station personnel can be promptly informed of significant changes in the facility status or core reactivity conditions during CORE ALTERATIONS. BRAIDWOOD - UNITS 1 & 2 B 3/4 9-1
REFUELING OPERATIONS FlHAL RAFTAUG 1 g jggg BASES 3/4.9.6 REFUELING MACHINE The OPERABILITY requirements for the refueling machine and auxiliary hoist ensure that: (1) refueling machines will be used for movement of drive rods and fuel assemblies, (2) each refueling machine has sufficient load capacity to lift a drive rod or fuel assembly, and (3) the core internals and reactor vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations. 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE FACILITY 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 areas ensures that in the event this load is dropped: (1) the activity release will be limited to that contained in a single fuel assembly, and (2) any possible distortion of fuel in the storage racks will not result in a critical array. This assumption is consistent with the activity release assumed in the safety analyses. 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION The requirement that at least one residual heat removal (RHR) loop be in operation ensures that: (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor vessel below 140'F as required during the REFUELING MODE, and (2) sufficient coolant circulation is maintained through the core to minimize the effect of a boron dilution incident and prevent boron stratification. The requirement to have two RHR loops OPERABLE when there is less than 23 feet of water above the reactor vessel flange ensures that a single failure of the operating RHR loop will not result in a complete loss of RHR capability. With the reactor vessel head removed and at least 23 feet of water above the reactor vessel flange, a large heat sink is available for core cooling. Thus, i 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 PURGE ISOLATION SYSTEM The OPERABILITY of this system ensures that the containment purge penetrations will be automatically isolated upon detection of high radiation levels within the containment. The OPERABILITY of this system is required to restrict the release of radioactive material from the containment atmosphere to the environment. BRAIDWOOD - UNITS 1 & 2 B 3/4 9-2
REFUELING OPERATIONS AUG 191986 BASES 3/4.9.10 and 3/4.9.11 WATER LEVEL - REACTOR VESSEL and STORAGE POOL 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 dapth is consistent with the assumptions of the safety analysis. , 1 3/4.9.12 FUEL HANDLING BUILDING EXHAUST FILTER PLENUM The limitations on the Fuel Handling Building Exhaust Filter Plenum ensure that all radioactive material released from an irradiated fuel assembly will be filtered through the HEPA filters and charcoal adsorber prior to dis-charge to the atmosphere. The OPERABILITY cf this ustem and the resulting iodine removal capacity are consistent with the assumptions of the safety analyses. ANSI N510-1980 will be used as a procedural guide for surveillance
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testing. l f BRAIDWOOD - UNITS 1 & 2 B 3/4 9-3
1 3/4.10 SPECIAL TEST EXCEPTIONS FINAL DRAFT AUG 191986 BASES 3/4.10.1 SHUTDOWN MARGD! This special test exception provides that a minimum amount of control rod worth is immediately available for reactivity control when tests are performed for control rod worth measurement and shutdown margin determination. This special test exception is required to permit the periodic verification of the actual versus predicted core reactivity condition occurring as a result of fuel burnup or fuel cycling operations. 3/4.10.2 GR0llP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS This special test exception permits individual control rods to be positioned outside of their normal group heights and. insertion limits during the performance of such PHYSICS TESTS as those required to: (1) measure control rod worth, and (2) determine the reactor stability index and dampening factor under xenon oscillation conditions. 3/4.10.3 PHYSICS TESTS This special test exception permits PHYSICS TESTS to be performed at less than or equal to 5% of RATED THERMAL PCddP. with the RCS T slightly lower thannormallyallowedsothatthefundamentalnuclearchaN8teristicsofthe core and related instrumentation can be verified. In order for various characteristics to be accurately measured, it is at times necessary to operate outside the normal restrictions of these Technical Specifications. For instance, to measure the moderator temperature coefficient at BOL, it is necessary to position the various control rods at heights which may not normally be allowed by belowSpecification the minimum3.1.3.6 which inofturn temperature may cause Specification the RCS T"V9 to fall slightly 3.1.1.4. 3/4.10.4 REACTOR COOLANT LOOPS This special test exception permits reactor criticality under no flow conditions and is required to perform certain startup and PHYSICS TESTS while at low THERMAL POWER levels. 3/4.10.5 POSITION INDICATION SYSTEM-SHUTDOWN This special test exception permits the Position Indication Systems to be inoperable during rod drop time measurements. The exception is required since the data necessary to determine the rod drop time is derived from the induced voltage in the position indicator coils as the rod is dropped. This induced voltage is small compared to the normal voltage and, therefore, cannot be observed if the Position Indication Systems remain OPERABLE. .
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BRAIDWOOD - UNITS 1 & 2 8 3/4 10-1
3/4.11 RADI0ACT_IVE EFFLUENTS BASES l l I 3/4.11.1 LIQUID EFFLUENTS 3/4.11.1.1 CONCENTRATION This specification is provided to ensure that the cencentration of radio-active materials released in liquid waste effluents to UNRESTRICTED AREAS will be less than the concentration levels specified in 10 CFR Part 20, Appendix B, Table II, Column 2. This limitation provides additional assurance that the levels of radioactive materials in bodies of water in UNRESTRICTED AREAS will result in exposures within: (1) the Section II.A design objectives of Appendix I, 10 CFR Part 50, to a MEMBER OF THE PUBLIC, and (2) the limits of 10 CFR Part 20.106(e) to the population. The concentration b .it for dissolved or entrained noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2. This specification applies to the release of radioactive materials in liquid effluents from all units at the site. The required detection capabilities for radioactive materials in liquid waste samples are tabulated in terms of the lower limits of detection (LLDs). Detailed discussion of the LLD, and other detection limits can be found in HASL Procedures Manual, HASL-300 (revised annually), Currie, L. A. , " Limits for Qualitative Detection and Quantitative Determination - Application to Radiochemistry," Anal. Chem. 40, 586-93 (1968), and Hartwell, J. K. , " Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975). 3/4.11.1.2 DOSE This' specification is provided to implement the requirements of Sec-tions II.A, III.A and IV.A of Appendiv I, 10 CFR Part 50. The Limiting Condi-tion for Operation implements the gusues set forth in Section II. A of Appen-dix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV. A of Appendix I to assure that the releases of radioactive material in liquid effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." The dose calculation niethodology and parameters in the ODCM implement the require-ments in Section III.A of Appendix I that conformance with the guides BRAIDWOOD - UNITS 1 & 2 8 3/4 11-1
6 RADIOACTIVE EFFLUENTS BASES IO E DOSE (Continued) of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from i Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance ' with 10 CFR Part 50, Appendix I" Revision 1, October 1977 and Regulatory Guide 1.113, " Estimating Aquatic Dispersion of Effluents from Accidental and Routine 1 Reactor Releases for the Purpose of Implementing Appendix I," April 1977. l This specification applies to the release of radioactive materials in liquid effluents from each reactor at the site. When shared Radwaste Treatment Systems are used by more than one unit on a site, the wastes from all units are mixed for shared treatment; by such mixing, the effluent releases cannot accurately be ascribed to a specific unit. An estimate should be made of the contributions from each unit based on input conditions, e.g., flow rates and radioact'ivity concentrations, or, if not practicable, the treated effluent releases may be allocated equally to each of the radioactive waste producing units sharing the Radwaste Treatment System. For determining conformance to LCOs, these allocations from shared Radwaste Treatment Systems are to be added to the releases specifically attributed to each unit to obtain the total releases per unit. 3/4.11.1.3 LIQUID RADWASTE TREATMENT SYSTEM The OPERABILITY of the Liquid Radwaste Treatment System ensures that this system will be available for use whenever liquid effluents require treatment prior to release to the environment. The requirement that the appropriate portions of this system be used when specified provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as is reasonably achievable". This specification implements the requirements of 10 CFR 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objective given in Section II.D of Appendix I to 10 CFR Part 50. l The specified limits governing the use of appropriate portions of the Liquid l Radwaste Treatment System were specified as a suitable fraction of the dose l design objectives set forth in Section II.A of Appendix I, 10 CFR Part 50, for liquid effluents. This specification applies to the release of radioactive materials in liquid effluents from each unit at the site. When shared Radwaste Treatment Systems are used by more than one unit on a site, the wastes from all units are mixed for shared treatment; by such mixing, the effluent releases cannot accurately be ascribed to a specific unit. An estimate should be made of the l contributions from each unit based on input conditions, e.g., flow rates and radioactivity concentrations, or, if not practicable, the treated effluent releases may be allocated equally to each of the radioactive waste producing i i BRAIDWOOD - UNITS 1 & 2 8 3/4 11-2
I 1 RADIOACTIVE EFFLUENTS BASES - AUG 191986 LIQUID RADWASTE TREATMENT SYSTEM (Continued) units sharing the Radwaste Treatment System. For determining conformance to LCOs, these allocations from shared Radwaste Treatment Systems are tb be added to the releases specifically attributed to each unit to obtain the total releases per unit. 3/4.11.1.4 LIQUID HOLDUP TANKS The tanks listed in this specification include all those outdoor radwaste tanks that are not surrounded by liners, dikes, or walls capable of holding the tank contents and that do not have tank overflows and surrounding area drains connected to the Liquid Radwaste Treatment System. Restricting the quantity of radioactive material contained in the specified tanks provides assurance that in the event of an uncontrolled release of the tanks' contents, the resulting concentrations would be less than the limits of 10 CFR Part 20, Appendix B, Table II, Column 2, at the nearest potable water supply and the nearest surface water supply in an UNRESTRICTED AREA. 3/4.11.2 GASEOUS EFFLUENTS 3/4.11.2.1 DOSE RATE This specification is provided to ensure that the dose at any time at and beyond the SITE BOUNDARY from gaseous effluents from all units on the site will be within the annual dose limits of 10 CFR Part 20 to UNRESTRICTED AREAS. The annual dose limits are the doses associated with the concentrations of 10 C'.-R Part 20, Appendix B, Table II, Column 1. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of a MEMBER OF THE PUBLIC in an UNRESTRICTED AREA, either within or outsida the SITE BOUNDARY, to annual average concentrations exceeding the limits specified in Appendix B, Table II of 10 CFR Part 20 (10 CFR 20.106(b)). For MEMBERS OF THE PUBLIC who may at times be within the SITE BOUNDARY, the occupancy of that MEMBER OF THE PUBLIC will usually be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the SITE BOUNDARY. Examples of calculations for such , MEMBERS OF THE PUBLIC, with the appropriate occupancy factors, shall be given l in the 00CM. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to a MEMBER OF THE PUBLIC at or beyond the SITE B0UNDARY to less than or equal to 500 mrems/ year to the whole body or to less than or equal to 3000 mrems/ year to the skin. These release rate limits also restrict, at all times the corresponding thyroid dose rate above background to a child via the inhalation pathway to less than or equal to 1500 mrems/ year. l This specification applies to the release of radioactive materials in gaseous effluents from all units at the site. BRAIDWOOD - UNITS 1 & 2 B 3/4 11-3
RADIOACTIVE EFFLUENTS a AUG I 91986 BASES DOSE RATE (Continued) The required detection capabilities for radioactive materials in liquid waste samples are tabulated in terms of the lower limits of detectio'n (LLDs). Detailed discussion of the LLD, and other detection limits can be 'found in HASL Procedures Manual, HASL-300 (revised annually), Currie, L. A., " Limits for Qualitative Detection and Quantitative Determination - Application to Radiochemistry," Anal. Chem. 40, 586-93 (1968), and Hartwell, J. K., " Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975). 3/4.11.2.2 DOSE - NOBLE GASES This specification is provided to implement the requirements of Sec-tions II.B, III.A and IV.A of Appendix I, 10 CFR Part 50. The Limiting Condi-tion for Operation implements the guides set forth in Section II.B of Appen-dix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in gaseous effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." The Surveillance Requirements implement the requirements in Section III.A of Appen-dix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The dose calculation methodology and parameters established in the ODCM for calculating the doses due to the actual release rates of radio-active materials in liquid effluents are consistent with the methodology pro-vided 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 I" 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. The ODCM equations provided for determining the air doses at and beyond the SITE BOUNDARY are based upon the historical average atmospheric conditions. This specification applies to the release of radioactive materials in gaseous effluents from each unit at the site. When shared Radwaste Treatment Systems are used by more than one unit on a site, the wastes from all units are mixed for shared treatment; by such mixing, the effluent releases cannot accurately be ascribed to a specific unit. An estimate should be made of the contributions from each unit based on input conditions, e.g., flow rates and radioactivity concentrations, or, if not practicable, the treated effluent releases may be allocated equally to each of the radioactive waste producing units sharing the Radwaste Treatment System. For determining conformance to LCOs, these allocations from shared Radwaste Treatment Systems are to be added to the releases specifically attributed to each unit to obtain the total releases per unit. BRAIDWOOD - UNITS 1 & 2 B 3/4 11-4
RADI0 ACTIVE EFFLUENTS AUG 191986 BASES I 3/4.11.2.3 DOSE - IODINE-131 AND 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 Operation are the guides set forth in Section II.C of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." The ODCM calculational 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 shown by calculational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The ODCM calculational methodology and parameters for calculating the doses due to the actual release rates of the subject materials are consistent with the methodology provided in Regulstory 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 I," 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 equations also provide for determining the actual doses based upon the historical average atmospheric conditions. The release rate specifications for Iodine-131 and 133, tritium, and radionuclides in particulate form with half-lives greater than 8 days are dependent upon the existing radionuclide pathways to man, in the areas at and beyond the SITE BOUNDARY. The pathways that were examined in the development of these calculations were: (1) indi-vidual inhalation of airborne 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. When shared Radwaste Treatment Systems are used by more than one unit on a site, the wastes from all units are mixed for shared treatment; by such mixing, the effluent releases cannot accurately be ascribed to a specific unit. An estimate should be made of the contributions from each unit based on input conditions, e.g., flow rates and radioactivity concentrations, or, if not practicable, the treated effluent releases may be allocated equally to each of the radioactive waste producing units sharing the Radwaste Treatment System. For determining conformance to LCOs, these allocations from shared Radwaste Treatment Systems are to be added to the releases specifically attributed to each unit to obtain the total releases per unit. , l BRAIDWOOD - UNITS 1 & 2 B 3/4 11-5 l
t RADI0 ACTIVE EFFLUENTS EWbm av AUG19 m BASES 3/4.11.2.4 GASEOUS RADWASTE TREATMENT SYSTEM The OPERABILITY of the WASTE GAS HOLDUP SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM ensures that the systems will be available for use whenever gaseous effluents require treatment prior to release to the environment. The requirement that the appropriate portions of this system be used when specified provides reasonable assurance that the releases of radioactive materials in gas-eous effluents will be kept "as low as is reasonably achievable". This specifi-cation implements the requirements of 10 CFR 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objectives given in Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the Gaseous Radwaste Treatment System were specified as a suitable fraction of the dose design objectives set forth in Section II.B and II.C of Appendix I,10 CFR Part 50, for <jaseous effluents. l This specification applies to the release of radioactive materials in gaseous effluents from each unit at the site. When shared Radwaste Treatment Systems are used by more than one unit on a site, the wastes from all units are mixed for shared treatment; by such mixing, the effluent releases cannot accurately be ascribed to a specific unit. An estimate should be made of the contributions from each unit based on input conditions, e.g., flow rates and radioactivity concentrations, or, if not practicable, the treated effluent releases may be allocated equally to each of the radioactive waste producing units sharing the Radwaste Treatment System. For determining conformance to LCOs, these allocations from shared Radwaste Treatment Systems are to be added to the releases specifically attributed to each unit to obtain the total releases per unit. 3/4.11.2.5 EXPLOSIVE GAS MIXTURE This specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the WASTE GAS HOLDUP SYSTEM is maintained below the flammability limits of hydrogen and oxygen. Automatic control features are included in the system to prevent the hydrogen and oxygen concentrations from reaching these flammability limits. These automatic control features include isolation of the source of hydrogen and/or oxygen, automatic diversion to recombiners, or injection of dilutants to reduce the concentration below the flammability limits. Maintaining the concentration of hydrogen and oxygen below their flammability limits provides assurance that the releases of radioactive materials will be controlled in conformance with the requirements t of General Design Criterion 60 of Appendix A to 10 CFR Part 50. , 3/4.11.2.6 GAS DECAY TANKS The tanks included in this specification are those tanks for which the quantity of radioactivity contained is not limited directly or indirectly by another Technical Specification. BRAIDWOOD - UNITS 1 & 2 B 3/4 11-6
RADI0 ACTIVE EFFLUENTS i BASES 0D GAS DECAY TANKS (Continued) Restricting the quantity of radioactivity contained in each gas storage tank provides assurance that in the event of an uncontrolled release. of the 3 tank's contents, the resulting whole body exposure to a MEMBER OF THE PUBLIC l at the nearest SITE B0UNDARY will not exceed 0.5 rem. This is consistent with Standard Review Plan 11.3, Branch Technical Position ETSB 11-5,
" Postulated Radioactive Releases Due to a Waste Gas System Leak or Failure,"
in NUREG-0800, July 1981. 3/4.11.3 SOLID RADI0 ACTIVE WASTES This specification implements the requirements of 10 CFR 50.36a and General Design Criterion 60 of Appendix A to 10 CFR Part 50. The process parameters included in establishing the PROCESS CONTROL PROGRAM may include, but are not limited to, waste type, waste pH, waste / liquid / SOLIDIFICATION agent / catalyst ratios, waste oil content, waste principal chemical constituents, and mixing and curing times. 3/4.11.4 TOTAL DOSE This specification is provided to meet the dose limitations of 40 CFR Part 190 that have been incorporated into 10 CFR Part 20 by 46 FR 18525. The specification requires the preparation and submittal of a Special Report whenever the calculated doses due to releases of radioactivity and to radiation from uranium fuel cycle sources exceed 25 mrems to the whole body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrems. For sites containing up to four reactors, it is highly unlikely that the resultant dose to a MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR Part 190 if the individual reactors remain within twice the dose design objectives of Appendix I, and if direct radiation doses from the reactor units and outside storage tanks are kept small. The Special Report will describe a course of action that should result in the limitation of the annual dose to a MEMBER OF THE PUBLIC to within the 40 CFR Part 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to the MEMBER OF THE PUBLIC from other uranium fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of 8 km must be considered. If the dose to any MEMBER OF THE PUBLIC is estimated to exceed the requirements of 40 CFR Part 190, the Special Report with a request for a variance (provided the release conditions resulting in violation of 40 CFR Part 190 have not already been corrected), in accordance with the provisions of 40 CFR 190.11 and 10 CFR 20.405c, is considered to be a timely request and fulfills the requirements of 40 CFR Part 190 until NRC staff action is completed. The variance only relates to the limits of 40 CFR Part 190, and does not apply in any way to the other requirements for dose limitation of 10 CFR Part 20, as addressed in Specifications 3.11.1.1 and 3.11.2.1. An individual is not considered a HEMBER OF THE PUBLIC during any period in which he/she is engaged in carrying out any operation that is part of the nuclear fuel cycle. BRAIDWOOD - UNITS 1 & 2 8 3/4 11-7
l 1 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING BASES 0N 3/4.12.1 MONITORING PROGRAM The Radiological Environmental Monitoring Program required by this specification provides representative measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides I that lead to the highest potential radiation exposures of MEMBERS OF THE PUBLIC resulting from the station operation. This monitoring program implements Section IV.B.2 of Appendix I to 10 CFR Part 50 and thereby supplements the < radiological effluent monitoring program by verifying that the measurable ' concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and the modeling of the environmental exposure pathways. Guidance for this monitoring program is provided by the Radiological Assessment Branch Technical Position on Environmental Monitoring. The initially specified monitoring program will be effective for at least the first 3 years of commercial operation. Following this period, program changes may be initiated based on operational experience. The required detection capabilities for env >onmental sample analyses are tabulated in terms of the lower limits of detection (LLDs). The LLDs required by Table 4.12-1 are considered optimum for routine environmental measurements in industrial laboratories. It should be recognized that the LLD is defined as a before the fact limit representing the capability of a measurement system and not as an after the fact limit for a particular measurement. Detailed discussion of the LLD, and other detection limits, can be found in HASL Procedures Manual, HASL-300 (revised annually), Currie, L. A., " Limits for Qualitative Detection and Quantitative Determination - Application to Radiochemistry," Anal. Chem. 40, 586-93 (1968), and Hartwell, J. K., " Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975). 3/4.12.2 LAND USE CENSUS This specification is provided to ensure that changes in the use of areas at and beyond the SITE BOUNDARY are identified and that modifications to the Radiological Environmental Monitoring Program given in the ODCM are made if required by the results of this census. The best inforniation from the door-to-door survey, from aerial survey, or from consulting with local agricul-tural authorities shall be used. This census satisfies the requirements of Section IV.B.3 of Appendix I to 10 CFR Part 50. An annual garden census will not be required since the licensee will assume that there is a garden at the nearest residence in each sector for dose calculations. I { BRAIDWOOD - UNITS 1 & 2 B 3/4 12-1 L
RADIOLOGICAL ENVIRONMENTAL MONITORING AUS 19 ngg BASES 3/4.12.3 INTERLABORATORY COMPARIS0N PROGRAM The requirement for participation in an approved Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and acevaacy of the measurements of radioactive material in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring in order to demonstrate that the results are valid for the purposes of Section IV.B.2 of Appendix I to 10 CFR Part 50. i r I BRAIDWOOD - UNITS 1 & 2 B 3/4 12-2 , e
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FINA;. DRAFT AUG ig g SECTION 5.0 DESIGN FEATURES i i l -- . _ _ . . - - - - _ _ . . _ . - - . _ - . _ - - _ - - . - _ - _ - - _ . - _ - - - . _ . _ _ _ . - . . --
5.0 DESIGN FEATURES 5.1 SITE , gy EXCLUSION AREA 5.1.1 The Exclusion Area shall be as shown in Figure 5.1-1. LOW POPULATION ZONE 5.1.2 The Low Population Zone shall be as shown in Figure 5.1-2. MAP DEFINING UNRESTRICTED AREAS AND SITE BOUNDARY FOR RADI0 ACTIVE GASE0US AND LIQUID EFFLUENTS 5.1.3 Information regarding radioactive gaseous and liquid effluents, which will allow identification of structures and release points as well as definition of UNRESTRICTED AREAS within the SITE BOUNDARY that are accessible to MEMBERS OF THE PUBLIC, shall be as shown in Figure 5.1-1. The definition of UNRESTRICTED AREA used in implementing these Technical Specifications has been expanded over that in 10 CFR 20.3 (a)(17). The UNRESTRICTED AREA boundary may coincide with the Exclusion (fenced) Area boundary, as defined in 10 CFR 100.3(a), but the UNRESTRICTED AREA does not include areas over water bodies. The concept of UNRESTRICTED AREAS, established at or beyond the SITE BOUNDARY, is utilized in the Limiting Conditions for Operation to keep levels of radioactive materials in liquid and gaseous effluents as low as is reasonably achievable, pursuant to 10 CFR 50.36a. For the Braidwood Station, the Exclusion Area lies within.the UNRESTRICTED AREA boundary. 5.2 CONTAINMENT CONFIGURATION 5.2.1 The containment building is a steel lined, reinforced concrete building of cylindrical shape, with a dome roof and having the following design features:
- a. Nominal inside diameter = 140 feet,
- b. Nominal inside height = 222 feet,
- c. Nominal thickness of concrete walls = 3.5 feet,
- d. Nominal thickness of concrete dome = 3 feet,
- e. Nominal thickness of concrete base slab = 12 feet,
- f. Nominal thickness of steel liner = 0.25 inch, and
- g. Het free volume = 2.8 x 108 cubic feet.
DESIGN PRESSURE AND TEMPERATURE 5.2.2 The containment building is designed and shall be maintained fdr a maximum internal pressure of 50 psig and a temperature of 250 F. BRAIDWOOD - UNITS 1 & 2 5-1
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DESIGN FEATURES u o ,n_ _ y 5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The core shall contain 193 fuel assemblies with each fuel assembly containing 264 fuel rods clad with Zircaloy-4. Each fuel rod shall have a nominal active fuel length of 144 inches and contain a maximum total weight of 1619 grams uranium. The initial core loading shall have a' maximum enrichment of less than 3.20 weight percent U-235. Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum enrichment of 4.0 weight percent U-235. CONTROL ROD ASSEMBLIES 5.3.2 The core shall contain 53 full-length and no part-length control rod assemblies. The full-length control rod assemblies shall contain a nominal 142 inches of absorber material. All control rods shall be hafnium, clad with stainless steel tubing. 5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The Reactor Coolant System is designed and shall be maintained:
- a. In accordance with the Code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements,
- b. For a pressure of 2485 psig, and
- c. For a temperature of 650*F, except for the pressurizer which is 680*F.
VOLUME 5.4.2 The total water and steam volume of the Reactor Coolant System is 12,257 cubic feet at a nominal T,yg of 588.4*F. 5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure 5.1-1. BRAIDWOOD - UNITS 1 & 2 5-4
DESIGN FEATURES AUG 191985
- 5. 6 FUEL STORAGE CRITICALITY 5.6.1.1 The spent fuel storage racks are designed and shall be maintained with: .
- a. A k,7f equivalent to less than or equal to 0.95 when flooded with unborated water, which includes a conservative allowance of 3.31%
Ak/k for uncertainties as described in Section 9.1 of the FSAR; and
- b. A nominal 14 inch center-to-center distance between fuel assemblies placed in the storage racks.
5.6.1.2 The k,77 for new fuel for the first core loading stored dry in the spent fuel storage racks shall not exceed 0.98 when aqueous foam moderation is assumed. DRAINAGE 5.6.2 The spent fuel storage pool'is designed and shall be maintained to ' prevent inadvertent draining of the pool below elevation 423 feet 0 inches. CAPACITY 5.6.3 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1060 fuel assemblies. 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT 5.7.1 The components identified in Table 5.7-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.7-1. BRAIDWOOD - UNITS 1 & 2 5-5
TABLE 5.7-1 O COMPONENT CYCLIC OR TRANSIENT LIMITS 6 o CYCLIC OR DESIGN CYCLE C0i90NENT TRANSIENT LIMIT OR TRANSIENT E Q Reacto:- Coolant System 200 heatup cycles at < 100*F/h Heatup cycle - T and 200 cooldown cycles at avg from 5 200*F to > 550*F. [ < 100*F/h. CooTdown cycle - T avg from m > 550*F to 5 200*F. 200 pressurizer cooldown cycles Pressurizer cooldown cycle at 5 200*F/h. temperatures from > 650*F to 5 100*F. 80 loss of load cycles, without > 15% of RATED THERMAL POWER to immediate Turbine or Reactor trip. D% of RATED THERMAL POWER. 40 cycles of loss-of-offsite Loss-of-offsite A.C. electrical A.C. electrical power. ESF Electrical System. 80 cycles of loss of flow in one Loss of only one reactor reactor coolant loop. coolant pump. 400 Reactor trip cycles. 100% to 0% of RATED THERMAL POWER. 10 auxiliary spray actuation Spray water temperature differential cycles. > 320*F. 50 leak tests. Pressurized to > 2485 psig. 5 hydrostatic pressure tests. Pressurized to > 3100 psig. 3 M ei r"~ Secondary Coolant System 1 large steam line break. Break in a > 6-inch steam line. % 5 hydrostatic pressure tests. Pressurized to > 1350 psig. e E A
9 FINAL DRAFT AUG1e a SECTION 6.0 ADMINISTRATIVE CONTROLS O
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ADMINISTRATIVE' CONTROLS FINAL RE 6.1 RESPONSIBILITY - IO M 6.1.1 The Superintendent, Braidwood Station, shall be responsible for overall unit operation and shall delegate in writing the succession to this responsi-bility during his absence. 6.1.2 The Shift Engineer (or during his absence from the control room, a designated individual) shall be responsible for the control room command function. A management directive to this effect, signed by the Division Vice President and General Manager-Nuclear Stations shall be reissued to all station personnel on an annual basis. 6.2 ORGANIZATION OFFSITE 6.2.1 The offsite organization for unit management and technical support shall be as shown in Figure 6.2-1. UNIT STAFF 6.2.2 The unit organization shall be as shown in Figure 6.2-2 and:
- a. Each on duty shift shall be composed of at least the minimum shift crew composition shown in Table 6.2-1; and
- b. At least one licensed Operator shall be in the control room when fuel is in the reactor. In addition, while the unit is in MODE 1, 2, 3, or 4, at least one licensed Senior Operator shall be in the control room;
- c. A Radiation Chemistry Technician,* qualified in radiation protection procedures, shall be on site when fuel is in the reactor;
- d. ALL CORE ALTERATIONS shall be observed and directly supervised by either a licensed Senior Operator or licensed Senior Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation;
- e. A site Fire Brigade of at least five members
- shall be maintained onsite at all times. The Fire Brigade shall not include the Shift Engineer, and the two other members of the minimum shift crew l
necessary for safe shutdown of.the unit and any personnel required for other essential functions during a fire emergency; and
*The Radiation Chemistry Technician and Fire Brigade composition may be less than the minimum requirements for a period of time not to exceed 2 hours in order to accommodate unexpected absence provided immediate action is taken to fill the required positions.
- BRAIDWOOD - UNITS 1 & 2 6-1
l ADMINISTRATIVE CONTROLS UNIT STAFF (Continued) ggyg
- f. Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safety-related functions; e.g. , licensed Senior Operators, licensed Operators, health physics personnel, equipment operators, and key maintenance personnel. .
The amount of overtime worked by Unit staff members performing safety-related functions shall be limited in accordance with the NRC Policy Statement on working hours (Generic Letter No. 82-12). 6.2.3 ONSITE NUCLEAR SAFETY GROUP (ONSG) FUNCTION 6.2.3.1 The ONSG serves as an independent safety engineering group and shall function to uamine plant operating characteristics, NRC issuances, industry advisories, REPORTABLE EVENTS and other sources of plant design and operating experience information, including plants of similar design, which may indicate areas for improving plant safety. The ONSG shall make detailed recommendations for revised procedures, equipment modifications, maintenance activities, opera-tions activities or other means of improving plant safety to the Manager of Nuclear Safety, and the Superintendent, Braidwood Station. CGMPOSITION 6.2.3.2 The ONSG shall be composed of at least four, dedicated, full-time engineers located on site. RESPONSIBILITIES 6.2.3.3 The ONSG shall be responsible for maintaining surveillance of plant activities to provide independent verification
- that th>ase activities are performed correctly and that human errors are reduced n much as practical.
RECORDS 6.2.3.4 Records of activities performed by the ONSG shall be prepared, main-tained, and forwarded each calendar month to the Manager of Nuclear Safety, and the Superintendent, Braidwood Station. 6.2.4 SHIFT TCCHNICAL ADVISOR The Station Control Room Engineer (SCRE) may serve as the Shift Technical Advisor (STA) during abnormal operating or accident conditions. During these conditions the SCRE or other on duty STA shall provide technical support to the Shift Supervisor in the areas of thermal hydraulics, reactor engineering and plant analysis with regard to the safe operation of the unit.
*Not responsible for sign-off function.
BRAIDWOOD - UNITS 1 & 2 6-2
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- TABLE 6.2-1 FINAL DRAFT AUG 191986 MINIMUM SHIFT CREW COMPOSITION POSITION NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION BOTH UNITS IN BOTH UNITS IN ONE UNIT IN MODE 1, 2, 3 OR 4 MODE 1, 2, 3, MODE 5 OR 6 AND .
OR 4 OR DEFUELED ONE UNIT IN MODE,5 OR 6 OR DEFUELED SE 1 1 1 SF 1 None## 1 R0 3 2 3 A0 3 3 3 STA or SCRE 1 None 1 SE - Shift Supervisor (Shift Engineer) with a Senior Operator license SF - Shift Foreman with a Senior Operator license RO - Individual with an Operator license AO - Auxiliary Operator STA - Shift Technical Advisor SCRE - Station Control Room Engineer with a Senior Operator License The Shift Crew Composition may be one less than the minimum requirements of Table 6.2-1 for a period of time not to exceed 2 hours in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the Shift Crew Composition to within the minimum requirements of Table.6.2-1. This provision does not permit any shift crew position to be unmanned upon shift change due to an oncoming shift crewman being late or absent. During any absence of the Shift Supervisor from the control room wnile the Unit is in MODE 1, 2, 3 or 4, an individual with a valid Senior Operator license shall be designated to assume the control room command function. During any absence of the Shift Supervisor from the control room while the Unit is in MODE 5 or 6, an individual with a valid Operator license shall bc designated to assume the control room command function.
#At least one of the required individuals must be assigned to the designated position for each unit.
- At least one licensed Senior Operator or licensed Senior Operator Limited to Fuel Handling must be present during Core Alterations on either unit, who has no other concurrent responsibilities.
- Table 6.2-1 will replace Table 6.2-la when Unit 2 receivos an operating license.
BRAIDWOOD - UNITS 1 & 2 6-5
- TABLE 6.2-la MINIMUM SHIFT CREW COMPOSITION 10 N POSITION NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION MODES 1, 2, 3 & 4 MODES 5 & 6 SE 1 1 SF 1 None R0 2 1 A0 2 1 SCRE 1 None or, whenever a SCRE (SRO/STA) is not included in the shift crew composition, the minimum shift crew shall be as follows:
POSITION NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION MODES 1, 2, 3 & 4 MODES S & 6 SE 1 1 SF 1 None R0 2 1 A0 2 1 STA 1 None SE - Shift Supervisor (Shift Engineer) with a Senior Operator license on Unit 1 SF - Shift Foreman with a Senior Operator license on Unit 1 RO - Individual with a Reactor Operator license on Unit 1 AO - Auxiliary Operator SCRE - Station Control Room Engineer with a Senior Reactor Operator's License on Unit 1 STA - Shift Technical Advisor The Shift Crew Composition may be one less than the minimum requirements of Table 6.2-la for a period of time not to exceed 2 hours in order to accommodate unexpected aba6.n e of on-duty shif t crew sembers provided innediate acLlusi I, taken to restore the Shift Crew Composition to within the minimum requirements of Table 6.2-la. This provision does not permit any shift crew position to be unmanned upon shift change due to an oncoming shift crewman being late or absent. During any absence of the Shift Supervisor from the control room while the Unit is in MODE 1, 2, 3 or 4, an individual with a valid Senior Operator license shall be designated to assume the control room command function. During any absence of the Shift Supervisor from the control room while the Unit is in MODE 5 or 6, an individual with a valid Operator license shall be designated to assume the control room command function. " Table 6.2-la will be replaced by Table 6.2-1 when Unit 2 receives an operating license. BRAIDWOOD - UNITS 1 & 2 6-Sa
FIEL DRAFT ADMINISTRATIVE CONTROLS 6.2.4 SHIFT TECHNICAL ADVISOR (Continued) To assure capability for performance of all STA functions: (1) The shift foreman (SRO) shall participate in the SCRE shift relief turnover. . (2) During the shift, the shift engineer and the shift foreman (SRO) shall be made aware of any significant changes in plant status in a timely manner by the SCRE. (3) During the shift, the shift engineer and the shift foreman (SRO) shall remain abreast of the current plant status. The shift foreman (SRO) shall return to the control room two or three times per shift, where practicable, to confer with the SCRE regarding plant status. Where not practicable to return to the control room, the shift foreman (SRO) shall peri.odically check with the SCRE for a plant status update. The shift foreman (SRO) shall not abandon duties original to reactor operation, unless specifically ordered by the shift engineer. 6.3 UNIT STAFF QUALIFICATIONS 6.3.1 Each member of the unit staff shall meet or exceed the minimum qualifi-cations of ANSI N18.1-1971, except for the Rad / Chem Supervisor or Lead Health Physicist, who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975, for a Radiation Protection Manager. The licensed Operators and Senior Operators shal1 also meet or exceed the minimum qualifications of the supplemental requirements specified in Sections A and C of Enclosure 1 of the March 28, 1980 NRC letter to all licensees. 6.4 TRAINING 6.4.1 A retraining and replacement training program for the unit staff shall be maintained under the direction of the Production Training Department and shall meet or exceed the requirements and recommendations of Section 5 of ANSI / ANS 3.1-1978 and /sppendix A of 10 CFR Part 55 and the supplemental requirements specified in Sect'.ons A and C of Enclosure 1 of the March 28, 1980 NRC letter to all licensees, and shall include familiarization with relevant industry i operational experience identified by the Office of Nuclear Safety. l 6.5 REVIEW INVESTIGATION AND AUDIT The Review and Investigative Function and the Audit Function of activities l affecting quality during facility operations shall be constituted and have I the responsibilities and authorities outlined below. l BRAIDWOOD - UNITS 1 & 2 6-6 i
ADMINISTRATIVE CONTROLS AUG 191986 6.5 REVIEW INVESTIGATION AND AUDIT (Continued) 0FFSITE 6.5.1 The Supervisor of the Offsite Review and Investigative Function shall be appointed by the Manager of Nuclear Safety responsible for nuclear activities. The audit function shall be the responsibility of the Manager of Quality Assurance and shall be independent of operations,
- a. Offsite Review and Investigative Function The Supervisor of the Offsite Review and Investigative Function shall: (1) provide directions for the review and investigative function and appoint a senior participant to provide appropriate direction, (2) select each participant for this function, (3) select a complement of more than one participant who collectively possess background and qualifications in the subject matter under review to provide comprehensive interdisciplinary review coverage under this function, (4) independently review and approve the findings and recommendations developed by personnel performing the review and investigative function, (5) approve and report in a timely manner all findings of non-compliance with NRC requirements to the Station Superintendent, Division Vice President and General Manager -
Nuclear Stations, Manager of Quality Assurance, and the Vice President - Nuclear Operations. During periods when the Supervisor of Offsite Review and Investigative Function is unavailable, he shall designate this responsibility to an established alternate, who satisfies the formal training and experience for the Supervisor of the Offsite Review and Investigative Function. The responsibilities of the per-sonnel performing this function are stated below. The Offsite Review and Investigative Function shall review:
- 1) The safety evaluations for: (1) changes to procedures, equip-ment, or systems as described in the safety analysis report, and (2) tests or experiments completed under the provision of 10 CFR 50.59 to verify that such actions did not constitute an unreviewed safety question. Proposed changes to the Quality Assurance Program description shall be reviewed and approved by the Manager of Quality Assurance;
- 2) Proposed changes to procedures, equipment or systems which involve an unreviewed safety question as defined in 10 CFR 50.59;
- 3) Proposed tests or experiments which involve an unreviewed safety question as defined in 10 CFR 50.59;
- 4) Proposed changes in Technical Specifications or this Operating License; .
BRAIDWOOD - UNITS 1 & 2 6-7
ADMINISTRATIVE CONTROLS OFFSITE (Continued) . AUG 191986
- 5) Noncompliance with Codes, regulations, orders, Technical Speci-fications, license requirements, or of internal procedures, or instructions having nuclear safety significance;
- 6) Significant operating abnormalities or deviation from normal and expected performance of plant equipment that affect nuclear safety as referred to it by the Onsite Review and Investigative Function;
- 7) All REPORTABLE EVENTS;
- 8) All recognized indications of an unanticipated deficiency in some aspect of design or operation of safety-related structures, systems, or components;
- 9) Review and report findings and recommendations regarding all changes to the Generating Stations Emergency Plan prior to implementation of such change; and
- 10) Review and report findings and recommendations regarding all items referred by the Technical Staff Supervisor, Station Superintendent, Division Vice President and General Manager -
Nuclear Stations, and Manager of Quality Assurance.
- b. Audit Function The audit function shall be the responsibility of the Manager of Quality Assurance independent of the Production Department. Such responsibility is delegated to the Director of Quality Assurance (Operating) and the General Supervisor Quality Assurance (Maintenance).
Either shall approve the audit agenda and checklists, the findings and the report of each audit. Audits shall be performed in accord-ance with the Company Quality Assurance Program and Procedures. Audits shall be performed to assure that safety-related functions are covered within the period designated below:
- 1) The conformance of facility operation to provisions contained within the Technical Specifications and applicable license conditions at least once per.12 months;
- 2) The adherence to procedure, training, and qualification of the station staff at least once per 12 months; j
- 3) The results of actions taken to correct deficiencies occurring in facility equipment, structures, systems, or methods of operation that affect nuclear safety at least once per 6 months;
- 4) The performance of activities required by the Operational Quality Assurance Program to meet the criteria of Appendix B, 10 CFR Part 50, at least once per 24 months; BRAIDWOOD - UNITS 1 & 2 6-8
ADMINISTRATIVE CONTROLS 5U6 L U 158b
.0FFSITE (Continued)
- 5) The Facility Emergency Plan and implementing procedures at least once per 12 months;
- 6) The Facility Security Plan and implementing procedures at least once per 12 months; .
- 7) Onsite and offsite reviews;
- 8) The Facility Fire Protection programmatic controls including the implementing procedures at least once per 24 months by qualified QA personnel;
- 9) The fire protection equipment and program implementation at least once per 12 months utilizing either a qualified offsite licensee fire protection engineer or an outside independent fire protection consultant. An outside independent fire pro-tection consultant shall be used at least every third year;
- 10) The Radiological Environmental Monitoring Program and the results thereof at least once per 12 months;
- 11) The OFFSITE DOSE CALCULATION MANUAL and implementing procedures at least once per 24 months;
- 12) The PROCESS CONTROL PROGRAM and implementing procedures for solidification of radioactive wastes at least once per 24 months; and
- 13) The performance of activities required by the Company Quality Assurance Program for effluent and environmental monitoring at least once per 12 months.
Report all findings of noncompliance with NRC requirements and recommendations and results of each audit to the Station Superin-tendent, Manager of Nuclear Safety, the Division Vice President and General Manager - Nuclear Stations, Manager of Quality Assurance, the Vice Chairman, and the Vice President - Nuclear Operations. 1 l
- c. Authority The Manager of Quality Assurance and the Manager of Nuclear Safety report to the Chairman and President. Either the Manager of Quality Assurance or the Manager of Nuclear Safety has the authority to order unit shutdown or request any other action which he deems necessary to avoid unsafe plant conditions.
BRAIDWOOD - UNITS 1 & 2 6-9
ADMINISTRATIVE CONTROLS 0FFSITE (Continued) AUG 2 9 506
- d. Records
- 1) Reviews, audits, and recommendations shall be documented and distributed as covered in Specification 6.5.la and 6.5.lb.;
and
- 2) Copies of documentation, reports, and correspondence shall be kept on file at the station.
- e. Procedures Written administrative procedures shall be prepared and maintained for the offsite reviews and investigative functions described in Specification 6.5.la. and for the audit functions described in Specification 6.5.lb. Those procedures shall cover the following:
- 1) Content and method of submission of presentations to the Supervisor of the Offsite Review and Investigative Function,
- 2) Use of committees and consultants,
- 3) Review and approval,
- 4) Detailed listing of items to be reviewed,
- 5) Method of: (1) appointing personnel, (2) performing reviews, investigations, (3) reporting findings and recommendations of reviews and investigations, (4) approving reports, and (5) distributing reports, and
- 6) Determining satisfactory completion of action required based on approved findings and recommendations reported by personnel performing the review and investigative function.
- f. Personnel
- 1) The persons, including consultants, performing the review and investigative function, in addition to the Supervisor of the Offsite Review and Investigative Function, shall have expertise in one or more of the following disciplines as appropriate for the subject or subjects being reviewed and investigated:
a) Nuclear power plant technology, b) Reactor operations, c) Utility operations, d) Power plant design, e) Reactor engineering, f) Radiological safety, . g) Reactor safety analysis, BRAIDWOOD - UNITS 1 & 2 6-10
ADMINISTRATIVE CONTROLS k , OFFSITE (Continued) . gIsg h) Instrumentation and control,
- 1) Metallurgy, and j) Any other appropriate disciplines required by unique characteristics of the facility.
- 2) Individuals performing the Review and Investigative Function shall possess a minimum formal training and experience as listed below for each discipline.
a) Nuclear Power Plant Technology Engineering graduate or equivalent with 5 years experience in the nuclear power field design and/or operation. b) Reactor Operations Engineering graduate or equivalent with 5 years experience in nuclear power plant operations. c) Utility Operations Engineering graduate or equivalent with at least 5 years of experience in utility operation and/or engineering. d) Power Plant Design Engineering graduate or equivalent with at least 5 years of experience in power plant design and/or operation. e) Reactor Engineering Engineering graduate or equivalent. In addition, at least 5 years of experience in nuclear plant engineering, operation,
.and/or graduate work in nuclear engineering or equivalent in reactor physics is required.
f) Radiological Safety Engineering graduate or equivalent with at least 5 years of experience in radiation control and safety. g) Reactor Safety Analysis Engineering graduate or equivalent with at least 5 years of experience in nuclear engineering. BRAIDWOOD - UNITS 1 & 2 6-11
ADMINISTRATIVE CONTROLS FINAL MAFT OFFSITE (Continued) . AUG 191986 h) Instrumentation and Control Engineering graduate or equivalent with at least 5 years of experience in instrumentation and control design and/or operation.
- 1) Metallurgy Engineering graduate or equivalent with at least 5 years of experience in the metallurgical field.
- 3) The Supervisor of the'Offsite Review and Investigative Function shall have experience and training which satisfy ANSI N18.1-1971 requirements for plant managers.
ONSITE 6.5.2 The Onsite Review and Investigative Function shall be supervised by the Station Superintendent.
- a. Onsite Review and Investigative Function The Station Superintendent shall: (1) provide directions for the Review and Investigative Function and appoint the Technical Staff Supervisor, or other comparably qualified individual as the senior participant to provide appropriate directions; (2) approve partici-pants for this function; (3) assure that at least two participants who collectively possess background and qualifications in the sub-ject matter under review are selected to provide comprehensive interdisciplinary review coverage under this function; (4) indepen-dently review and approve the findings and recommendations developed by personnel performing the Review and Investigative Function; (5) report all findings of noncompliance with NRC requirements, and t provide recommendations to the Division Vice President and General Manager - Nuclear Stations and the Supervisor of the Offsite Review and Investigative Function; and (6) submit to the Offsite Review and Investigative Function for concurrence in a timely manner, those items described in Specification 6.5.la which have been approved by the Onsite Review and Investigative Function.
- b. Responsibility The responsibilities of the personnel performing this function are:
- 1) Review of: (1) station specific portions of all procedures required by Specification 6.8.1 and changes thereto, (2) all programs required by Specification 6.8.4 and changes thereto, and (3) any other proposed procedures or changes thereto as determined by the Station Superintendent to affect nuclear safety; -
- 2) Review of all proposed tests and experiments that affect nuclear safety; BRAIDWOOD - UNITS 1 & 2 6-12
ADMINISTRATIVE CONTROLS ONSITE (Continued) . AUG 19 4
- 3) Review of all proposed changes to the Technical Specifications;
- 4) Review of all proposed changes or modifications to plant systems or equipment that affect nuclear safety; 5). Investigation of all violations of the Technical Specifications including the preparation and forwarding of reports covering evaluation and recommendations to prevent recurrence to the Division Vice President and General Manager - Nuclear Stations and to the Supervisor of the Offsite Review.and Investigative Function;
- 6) Review of all REPORTABLE EVENTS;
- 7) Performance of special reviews and investigations and reports thereon as requested by the Supervisor of the Offsite Review and Investigative Function;
- 8) Review of the Station Security Plan and implementing procedures and submittal of recommended changes to the Division Vice President and General Manager - Nuclear Stations; -
- 9) Review of the Emergency Plan and station implementing procedures
. and submittal of recommended changes to the Division Vice President and General Manager - Nuclear Stations;
- 10) Review of Unit operations to detect potential hazards to nuclear safety;
- 11) Review of any accidental, unplanned, or uncontrolled radioactive release including the preparation of reports covering evaluation, recommendations and disposition of the corrective action to prevent recurrence and the forwarding of these reports to the Division Vice President and General Manager - Nuclear Stations and the Supervisor of the Offsite Review and Investigative Function; and
- 12) Review of changes to the PROCESS CONTROL PROGRAM, the OFFSITE DOSE CALCULATION MANUAL, and the Radwaste Treatment Systems.
- c. Authority The Technical Staff Supervisor is responsible to the Station Superintendent and shall make recommendations in a timely manner in all areas of review, investigation, and quality control phases of plant maintenance, operation, and administrative procedures relating to facility operations and shall have the authority to request the action necessary to ensure compliance with rules, regulation,s, and proceduros when in his opinion such action is necessary. The Station Superintendent shall follow such recommendations or select a course BRAIDWOOD - UNITS 1 & 2 6-13
ADMINISTRATIVE CONTROLS FWLEAfl ONSITE (Continued) of action that is more conservative regarding safe operation of the facility. All such disagreements shall be reported immediately to the Division Vice President and General Manager - Nuclear Stations and the Supervisor of the Offsite Review and Investigative Function.
- d. Records
- 1) Reports, reviews, investigations, and recommendations shall be documented with copies to the Division Vice President and General Manager - Nuclear Stations, the Supervisor of the Offsite Review and Investigative Function, the Station Superintendent, and the Manager of Quality Assurance.
- 2) Copies of all records and documentation shall be kept on file at the station.
- e. Procedures Written administrative procedures shall be prepared and maintained for conduct of the Onsite Review and Investigative Function. These procedures shall include the following:
- 1) Content and method of submission and presentation to the Station Superintendent, Division Vice President and General Manager - Nuclear Stations, and the Supervisor of the Offsite Review and Investigative Function,
- 2) Use of committees,
- 3) Review and approval,
- 4) Detailed listing of items to be reviewed,
- 5) Procedures for administration of the quality control activities, and
- 6) Assignment of responsibilities.
- f. Personnel
- 1) The personnel performing the Onsite Review and Investigative Function, in addition to the Station Superintendent, shall consist of persons having expertise in:
a) Nuclear power plant technology, b) Reactor operations, c) Reactor engineering, < d) Chemistry . I e) Radiological controls, f) Instrumentation and control, and l g) Mechanical and electrical systems. BRAIDWOOD - UNITS 1 & 2 6-14
ADMINISTRATIVE CONTROLS FIR.DWT Uli 191986 ONSITE (Continued)
- 2) Personnel performing the Onsite Review and Investigative Function shall meet minimum acceptable levels as described in ANSI N18.1-1971, Sections 4.2 and 4.4. ;
i 6.6 REPORTABLE EVENT ACTION . 6.6.1 The following actions shall be taken for REPORTABLE EVENTS:
- a. The Commission shall be notified and a report submitted pursuant to the requirements of Section 50.73 to 10 CFR Part 50, and
- b. Each REPORTABLE EVENT shall be reviewed by the Onsite Review and Investigative Function and the results of this review shall be submitted to the Offsite Review and Investigative Function and the-t Division Vice President and General Manager - Nuclear Stations.
6.7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:
- a. The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within I hour. The Division Vice President
- and General Manager - Nuclear Stations and the Offsite Review and Investigative Function shall be notified within 24 hours;
- b. A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the Onsite Review and Investigative Function. This report shall describe: (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems or structures, and (3) corrective action taken to prevent recurrence;
- c. The Safety Limit Violation Report shall be submitted to the Commission, the Offsite Review and Investigative Function and the Division Vice President and General Manager - Nuclear Stations within 14 days of the violation; and
- d. Critical operation of the Unit shall not be resumed until authorized by the Commission.
6.8 PROCEDURES AND PROGRAMS 6.8.1 Written procedures shall be established, implemented, and maintained l covering the activities referenced below: l ! a. The applicable procedures recommended in Appendix A, of Regulatory Guide 1.33, Revision 2, February 1978, . BRAIDWOOD - UNITS 1 & 2 6-15
1 ADMINISTRATIVE CONTROLS FINAL DMFT l AUG li M PROCEDURES AND PROGRAMS (Continued)
- b. The emergency operating procedures required to implement the require-ments of NUREG-0737 and Supplement 1 to NUREG-0737 as stated in Sec-tion 7.1 of Generic Letter No. 82-33;
- c. Plant Security Plan implementation,
- d. Radiological Emergency Response Plan implementation,
- e. PROCESS CONTROL PROGRAM implementation,
- f. OFFSITE DOSE CALCULATION MANUAL implementation, and
- g. Quality Assurance Program implementation for effluent and environ-mental monitoring.
6.8.2 Each procedure of Specification 6.8.1 above, and changes thereto, shall be reviewed prior to implementation as set forth in Specification 6.5 above. 6.8.3 Temporary changes to procedures of Specification 6.8.1 above, may be made provided:
- a. .The intent of the original procedure is not altered;
- b. The change is approved by two members of the plant management staff, at least one of whom holds a Senior Operator license on the Unit affected; and
- c. The change is documented, reviewed by the Onsite Review and Investi-gative Function, and approved by the Station Superintendent within 14 days of implementation.
6.8.4 The following programs shall be established, implemented, and maintained:
- a. Reactor Coolant Sources Outside Containment A program to reduce leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. The systems include the recirculation portion of the Containment Spray System, Safety Injection System, Chemical and Volume Control System, and RHR System. The program shall include the following:
- 1) Preventive maintenance and periodic visual inspection requirements, and
- 2) Integrated leak test requirements for each system at refueling cycle intervals or less.
BRAIDWOOD - UNITS 1 & 2 6-16
d i ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)
- b. In-Plant Radiation Monitoring A program which will ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions. This program shall include the following: .
- 1) Training of personnel,
- 2) Procedures for monitoring, and
- 3) Provisions for maintenance of sampling and analysis equipment.
- c. Secondary Water Chemistry A program for monitoring of secondary water chemistry to inhibit steam generator tube degradation. This program shall include:
- 1) Identification of a sampling schedule for the critical variables and control points for these variables,
- 2) Identification of the procedures used to measure the values of the critical variables,
- 3) Identification of process sampling points, which shall include monitoring the discharge of the condensate pumps for evidence of condenser in-leakage,
- 4) Procedures for the recording and management of data,
- 5) Procedures defining corrective action for all off-control point chemistry conditions, and
, 6) A procedure' identifying: (a) the authority responsible for the i interpretation of the data, and (b) the sequence and timing of t administrative events required to initiate corrective action.
- d. Post-accident Sampling A program which will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant !
gaseous effluents, and containment atmosphere samples under accident conditions. The program shall include the following:
- 1) Training of personnel,
- 2) Procedures for sampling and analysis, and
- 3) Provisions for maintenance of sampling and analysis equipment.
BRAIDWOOD - UNITS 1 & 2 6-17
ADMINISTRATIVE CONTROLS 6.9 REPORTING REQUIREMENTS - AUG 191986 ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Regional Administrator of the NRC Regional Office unless otherwise noted. STARTUP REPORT 6.9.1.1 A summary report of plant startup and power escalation testing shall be submitted following: (1) receipt of an Operating License, (2) amendment to the license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant. l 6.9.1.2 The Startup Report shall address each of the tests identified in the Final Safety Analysis Report FSAR and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifica-tions. Any corrective actions that were required to obtain satisfactory opera-tion shall also be described. Any additional specific details required in license conditions based on other commitments shall be included in this report. 6.9.1.3 Startup Reports shall be submitted within: (1) 90 days following com-pletion of the Startup Test Program, (2) 90 days following resumption or com-mencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest. If the Startup Report does not cover all three events (i.e., initial criticality, completion of Startup Test Program, and resumption or commencement of commercial operation) supplementary reports shall be submitted at least every 3 months until all three events have been completed. I ANNUAL REPORTS 6.9.1.4 Annual Reports covering the activities of the unit as described below for the previous calendar year shall be submitted prior to March 1 of each year. The initial report shall be submitted prior to March 1 of the year i following initial criticality. 6.9.1.5 Reports required on an annual basis shall include tabulation on an annual basis of the number of station, utility, and other personnel (including contractors) receiving exposures greater than 100 mrems/yr and their associated man-re;n exposure according to work and job functions,* e.g. , reactor operations and surveillance, inservice inspection, routine maintenance, special mainte-nance (describe maintenance), waste processing, and refueling. The dose assignments to various duty functions may be estimated based on pocket dosimeter, TLD, or film badge measurements. Small exposures totalling less than 20% of the individual total dose need not be accounted for. In the aggregate, at least 80% of the total whole body dose received from external sources should l be assigned to specific major work functions. l ,
*This tabulation supplements the requirements of $20.407 of 10 CFR Part 20.
BRAIDWOOD - UNITS 1 & 2 6-18
I ADMINISTRATIVE CONTROLS RM*DWT hug .L y lwn REPORTING REQUIREMENTS (Continued). ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT
- 6.9.1.6 Routine Annual Radiological Environmental Operating Reports covering the operation of the Unit during the previous calendar year shall be submitted prior to May 1 of each year. The initial report shall be submitted prior to 1
i May 1 of the year following initial criticality. The Annual Radiological Environmental Operating Reports shall include summaries, interpretations, and an analysis of trends of the results of the radiological environmental surveillance activites for the report period, including a comparison with preoperational studies, with operational controls as appropriate, and with previous environmental surveillance reports, and an assessment of the observed impacts of the plant operation on the environment. The reports shall also include the results of the Land Use Census required by Specification 3.12.2. The Annual Radiological Environmental Operating Reports shall include the results of analysis of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the tables and figures in the ODCM, as well as summarized and tabulated results of these analyses and measurements in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979. In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted as soon as possible in a supplementary report. The reports shall also include the following: a summary description of the Radiological Environmental Monitoring Program; at least two legible maps ** covering all sampling locations keyed to a table giving distances and directions from the centerline of one reactor; the results of licensee participation in the Interlaboratory Comparison Program and the corrective actions being taken if the specified program is not being performed as required by Specification 3.12.3; ( reasons for not conducting the Radiological Environmental Monitoring Program as
. required by Specification 3.12.1, and discussion of all deviations from the sampling schedule of Table 3.12-1; discussion of environmental sample measure-ments that exceed the reporting levels of Table 3.12-2 but are not the result of plant effluents, pursuant to Specification 3.12.1; and discussion of all analyses in which the LLD required by Table 4.12-1 was not achievable.
l
*A single submittal may be made for a multiple unit station. **0ne map shall cover stations near the SITE BOUNDARY; a second shall include the more distant stations.
I BRAIDWOOD - UNITS 1 & 2 6-19
ADMINISTRATIVE CONTROLS REPORTING REQUIREMENTS (Continued). The Annual Radiological Environmental Operating Report shall also include an annual summary of hourly meteorological data collected over the previous year. This annual summary may be either in the form of an hour-by-hour listing on magnetic tape of wind speed, wind direction, atmospheric stability, and precipitation (if measured), or in the form of joint frequency distribu-tions of wind speed, wind direction, and atmospheric stability.* This same report shall include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the Unit or Station during the previous calendar year. This same report shall also include an assessment of the radiation doses from radioactive liquid and gaseous effluents to MEMBERS OF THE PUBLIC due to their activities inside the SITE BOUNDARY (Figure 5.1-1) during the report period. All assumptions used in making these assessments, i.e., specific activity, exposure time and location, shall be
. included in these reports. The meteorological conditions concurrent with the time of release of radioactive materials in gaseous effluents, as determined by sampling frequency and measurement, shall be used for determining the gaseous pathway doses. The assessment of radiation doses shall be performed in accordance with the methodology and parameters in the ODCM.
The Annual Radiological. Environmental Operating Report to be submitted prior to May 1 of each year shall also include an assessment of radiation doses to the most likely exposed MEMBER OF THE PUBLIC from reactor releases and other nearby uranium fuel cycle sources, including doses from primary effluent
' pathways and direct radiation, for the previous calendar year to show con-formance with 40 CFR Part 190, " Environmental Radiation Protection Standards for Nuclear Power Operation." Acceptable methods for calculating the dose contribution from liquid and gaseous effluents are given in Regulatory Guide 1.109, Rev. 1, October 1977.
SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT ** 6.9.1.7 Routine Semiannual Radioactive Effluent Release Reports covering the operation of the unit during the previous 6 months of operation shall be submitted within 60 days after January 1 and July 1 of each year. The period of the first report shall begin with the date of initial criticality. The Semiannual Radioactive Effluent Release Reports shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit as outlined in Regulatory Guide 1.21,
*In lieu of submission with the Annual Radiological Environmental Operating Report, the licensee has the option of retaining this summary of required meteorological data on site in a file that shall be provided to the NRC upon request.
**A single submittal may be made for a multiple unit station. The submittal should combine those sections that are common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.
BRAIDWOOD - UNITS 1 & 2 6-20
ADMINISTRATIVE CONTROLS FINAL DRAFT REPORTING REQUIREMENTS (Continued)
" Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants," Revision 1, June 1974, with data summarized on a quarterly basis following the format of Appendix B thereof. For solid wastes, the format for Table 3 in Appendix B shall be supplemented with three additional categories: class of solid wastes (as defined by 10 CFR Part 61),
type of container (e.g., LSA, Type A, Type B, Large Quantity), and SOLIDIFICA-TION agent or absorbent (e.g., cement, urea formaldehyde). The Semiannual Radioactive Effluent Release Reports shall include a list and description of unplanned releases from the site to UNRESTRICTED AREAS of radioactive materials in gaseous and liquid effluents made during the reporting period. The Semiannual Radioactive Effluent Release Reports shall include any changes made during the reporting period to the PCP, pursuant to Specifications 6.13, as well as any major changes to Liquid, Gaseous or Solid Radwaste Treat-ment Systems, pursuant to Specification 6.15. The Semiannual Radioactive Effluent Release Reports shall also include the following: an explanation as to why the inoperability of liquid or gaseous effluent monitoring instrumentation was not corrected within the time specified in Specifications 3.3.3.10 or 3.3.3.11, respectively; and description of the events leading to liquid holdup tanks or gas storage tanks exceeding the limits of Specification 3.11.1.4 or 3.11.2.6, respectively. MONTHLY OPERATING REPORT 6.9.1.8 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the PORVs or RCS safety valves, shall be submitted on a monthly basis to the Director, Office of Resource Management, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, with a copy to the Regional Administrator of the NRC Regional Office, no later than the 15th of each month following the calendar month covered by the report. RADIAL PEAKING FACTOR LIMIT REPORT 6.9.1.9 Changes to the F limits for Rated Thermal Power (FRTP) shall be xy x provided to the NRC Regional Administrator with a copy to Director of Nuclear Reactor Regulation, Attention: Chief, Reactor Systems Branch, DPL-A, U.S. Nuclear Regulatory Commission, Washington, D. C. 20555 for all core planes con-taining Bank "D" control rods and all unrodded core planes and the plot of pre-T dicted (F q .Pgg ) vs Axial Core Height with the limit envelope at least 60 days prior to cycle initial criticality unless otherwise approved by the Commission by letter. In addition, in the event that the limit should change requiring a new BPAIDWOOD - UNITS 1 & 2 6-21
ADMINISTRATIVE CONTROLS RADIAL PEAKING FACTOR LIMIT REPORT.(Continued) AUG 191986 submittal or an amended submittal to the Peaking Factor Limit Report, it shall be submitted 60 days prior to the date the limit would become effective unless otherwise approved by the Commission by letter. Any information needed to support F RTP will be by request from the NRC and need not be included in this report. U SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator of the NRC Regional Office within the time period specified for each report. 6.10 RECORD RETENTION In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated. 6.10.1 The following records shall be retained for at least 5 years:
- a. Records and logs of unit operation, covering time interval at each power level;
- b. Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety;
- c. All REPORTABLE EVENTS;
- d. Records of surveillance activities, inspections, and calibrations required by these Technical Specifications;
- e. Records of changes made to the procedures required by Specification 6.8.1;
- f. Records of radioactive shipments;
- g. Records of sealed source and fission detector leak tests and results; and
- h. Records of annual physical inventory of all sealed source material of record.
6.10.2 The following records shall be retained for the duration of the unit Operating License:
- a. Records and drawing changes reflecting unit design modifications made to systems and equipment described in the Final Safety Analysis Report; -
- b. Records of new and irradiated fuel inventory, fuel transfers and assembly burnup histories; BRAIDWOOD - UNITS 1 & 2 6-22
l
)
ADMINISTRATIVE CONTROLS l l RECORD RETENTION (Continued) . All,G 19 W
- c. Records of radiation exposure for all individuals entering radiation control areas;
- d. Records of gaseous and liquid radioactive material released to the environs; .
- e. Records of transient or operational cycles for those unit components identified in Table 5.7-1;
- f. Records of reactor tests and experiments;-
- g. Records of training and qualification for current members of the unit staff;
- h. Records of in-service inspections performed pursuant to these Technical Specifications;
- i. Records of Quality Assurance activities required by the QA Program;
- j. Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59;
- k. Records of meetings and results of reviews and audits performed by the Offsite Review and Investigative Function and the Onsite Review and Investigative Function;
- 1. Records of the service lives of all hydraulic and mechanical snubbers required by Specification 3.7.8 including the date at which the service life commences and associated installation and maintenance records;
- m. Records of secondary water sampling and water quality; and
- n. Records of analysis required by the Radiological Environmental Monitoring Program that would permit evaluation of the accuracy of the analysis at a later date. This should include procedures effective at specified times and QA records showing that these procedures were followed.
6.11 RADIATION PROTECTION PROGRAM Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure. 6.12 HIGH RADIATION AREA 6.12.1 Pursuant to Paragraph 20.203(c)(5) of 10 CFR Part 20, in lieu.of the
" control device" or " alarm signal" required by paragraph 20.203(c), each high radiation area, as defined in 10 CFR Part 20, in which the intensity of radia-tion is equal to or less than 1000 mR/hr at 45 cm (18 in.) from the radiation BRAIDWOOD - UNITS 1 & 2 6-23
l ADMINISTRATIVE CONTROLS E mE Vf Ram asa U ma u HIGH RADIATION AREA (Continued) AUG 19195 source or from any surface which the radiation penetrates shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP). Individuals qualified in radiation protection procedures (e.g., Rad / Chem Technician) or personnel continuously escorted by such individuals may be exempt from the RWP issuance requirement during the performance of their assigned duties in high radiation areas with exposure rates equal to or less than 1000 mR/h, provided they are otherwise following plant radiation protection procedures for entry into such high radiation areas. Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:
- a. A radiation monitoring device which continuously indicates the radiation dose rate in the area; or i "b. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel have been made knowledgeable of them; or
- c. An. individual qualified in radiation protection procedures with a radiation dose rate monitoring device, who is responsible for providing positive control over the activities within the area and l
shall perform periodic radiation surveillance at the frequency specified in the Radiation Work Permit. i 6.12.2 In addition to the requirements of Specification 6.12.1, areas accessible to personnel with radiation levels greater than 1000 mR/h at 45 cm (18 in.) from the radiation source or from any surface which the radiation penetrates shall be, provided with locked doors to prevent unauthorized entry, and the keys shall be maintained under the administrative control of the Shift Foreman on duty and/or health physics supervision. Doors shall remain locked except during periods of access by personnel under an approved RWP which shall specify the dose rate levels in the immediate work areas and the maximum allowable stay time for individuals in that area. In lieu of the stay time specification of the RWP, direct or remote (such as closed circuit TV cameras) continuous surveillance may be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities being performed within the area. During emergency situations which involve personnel injury or actions taken to prevent major equipment damage, continuous surveillance and radiation monitoring of the work area by a qualified individual may be substituted for the routine RWP procedure. For individual high radiation areas accessible to personnel with radiation levels of greater than 1000 mR/h that are located within large areas, such as PWR containment, where no enclosure exists for purposes of locking, and where no enclosure can be reasonably constructed around the individual area, that individual area shall be barricaded (by a more substantial obstacle than rope), conspicuously posted, and a flashing light shall be activated as a warning device. BRAIDWOOD - UNITS 1 & 2 6-24
l I ADMINISTRATIVE CONTROLS E 'h 6.13 PROCESS CONTROL PROGRAM (PCP.) 6.13.1 The PCP shall be approved by the Commission prior to implementation. 6.13.2 Licensee-initiated changes to'the PCP:
- a. Shall be submitted to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the change (s) was made. This submittal shall contain:
- 1) Sufficiently detailed information to totally support the rationale for the change wthout benefit of additional or supplemental information;
- 2) A determination that the change did not reduce the overall conformance of the solidified waste product to existing criteria for solid wastes; and
- 3) Documentation of the fact that the change has been reviewed and found acceptable by the Onsite Review and Investigative Function.
- b. Shall beccme effective upon review and acceptance by the Onsite Review and Investigative Function in accordance with Specification 6.5.2.
6.14 0FFSITE DOSE CALCULATION MANUAL (ODCM) 6.14.1 The ODCM shall be approved by the Commission prior to implementation. 6.14.2 Licensee-initiated changes to the ODCH:
- a. Shall be submitted to the Commission by inclusion in the Monthly Operating Report pursuant to Specification 6.9.1.8 within 90 days of the date the change (s) was made effective. This submittal shall contain:
- 1) Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information. Information submitted should consist of a package of those pages of the ODCM to be changed with each page numbered, dated and containing the revision number, together with appropriate analyses or evaluations justifying the change (s);
- 2) A determination that the change will not reduce the accuracy or reliability of dose calculations or Setpoint determinations; and
- 3) Documentation of the fact that the change has been reviewed and found acceptable by the Onsite Review and Investigative Function.
- b. Shall become effective upon review and acceptance by the Ons'ite Review and Investigative Function in accordance with Specification 6.5.2.
BRAIDWOOD - UNITS 1 & 2 6-25
ADMINISTRATIVE CONTROLS
. AUG 191966 6.15 MAJOR CHANGES TO LIQUID, GASEOUS, AND SOLID RADWASTE TREATMENT SYSTEMS
- 6.15.1 Licensee-initiated major changes to the Radwaste Treatment Systems (liquid, gaseous, and solid):
- a. Shall be reported to the Commission in the Semiannual Radioactive Effluent Release Report for the period in whicn the evaluation was reviewed by the Onsite Review and Investigative Function. The dis-cussion of each change shall contain:
- 1) A summary of the evaluation that led to the determination that the change could be made in accordance with 10 CFR 50.59;
- 2) Sufficient detailed info mation to totally support the reason for the change without benefit of additional and supplemental information;
- 3) A detailed description of the equipment, components, and processes involved and the interfaces with other plant systems.
- 4) An evaluation of the change which shows the predicted releases of radioactive materials in liquid and gaseous effluents and/or quantity of solid waste that differ from those previously predicted in the License application and amendments thereto; -
- 5) An evaluation of the change, which shows the expected maximum exposures to a MEMBER OF THE PUBLIC in the UNRESTRICTED AREA and to the general population that differ from those previously estimated in the License application and amendments thereto;
- 6) A comparison of the predicted releases of radioactive materials, in liquid and gaseous effluents and in solid waste, to the actual releases for the period prior to when the changes are to be made;
- 7) An estimate of the exposure to plant operating personnel as a result of the change; and
- 8) Documentation of the fact that the change was reviewed and l found acceptable by the Onsite Review and Investigative Function.
- b. Shall become effective upon review and acceptance by the Onsite Review and Investigative Function in accordance with Specification 6.5.2.
- Licensees may choose to submit the information called for in this specification as part of the annual FSAR update. ,
BRAIDWOOD - UNITS 1 & 2 6-26
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