ML20095D010
| ML20095D010 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 04/13/1992 |
| From: | Barrett R Office of Nuclear Reactor Regulation |
| To: | Commonwealth Edison Co |
| Shared Package | |
| ML20095D015 | List: |
| References | |
| NPF-37-A-046, NPF-66-A-046, NPF-72-A-035, NPF-77-A-035 NUDOCS 9204240248 | |
| Download: ML20095D010 (106) | |
Text
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NUCLE AR REGULATORY COMMISSION t
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s, COMMONWEA_LTH EDISON COMPAy1 DOCKET NO. STN 50-454 BlRQ!LSIATI0fL SNIT N0J AMENDMENT TQ.LBCILITY OPERATING LICENSE Amendment No. 46 License No. NPF-37 1.
The Nuclear Regulatory Comm!sm;on (the Commission) has found that:
A.
The application for amendment by Commonwealth Edison Company (the licensee) dated June 10, 1991 as supplemented on October 17, 1991, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The fachity will operate in co.Jormity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly.the license is a.anded by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-37 is hereby amended to read as follows:
9204240248 92041T1 PDR ADOCK 05000454 P
. (2)
Techc :al Snecifiell10m The Technical Specifications contained in Appendix A as revised through Amendment No. 46 and the Environmental Protection Plan cretained in Appendix B, both of which are attached hereto, are hereby incorporated into this license.
The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Pian.
3.
This license amendme.t is effective as of the date of its issuance.
FOR 1HE NUCLEAR REGULATORY COMMISSION l Y, ?
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- Richa, JfBff're'tt, Director Project Directorate 111-2 Division of Reactor Projects - lil/IV/V Of fice of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
Acril 13, 1992
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NUCLE AR REGULATORY COMMISSION j
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[0MMONWEALTH EDISON 10MPANY DOCKET No. STN 50-455 BYRON STATION. UNIT N0. 2 AMENDHDLTJ(1 FACillTY OP(RATING LICENSE Amendment No. 46 License No. NPF-66 1.
The Nuclear Regulatory Cormission (the Commission)..:s found that:
A.
The application for amendment by Commonwealth Edison Company (the licensee) dated June 10, 1991 as supplemented on October 17, 1991, complies with the standards and rec,uirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter 1; 8.
The facility will operate in conformity with the applit.ation, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in tccordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specif t-cations'as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-66 is hereby amended to read as follows:
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(2) leihrlical Specifications.
The Technical Specifications contained in Appendix A (NUREG-lll3),
as revised through Amendment No. 46 and revised by Attachment 2 to NPF-66, and the Environmental Protection Plan contained in Appendix B. both of which were attached to License No. NPf-37, dated February 14, 1985, are hereby incorporated into this license. Attachment 2 contains a revision to Appendix A which is hereby incorporated into this license.
The licensee shall operate the facility in acccrdance with the Technical Specifications and the Environmental ^.otection Plan.
3.
This license amendment is effective as of the date of its issuance.
FORTHENUCLEARREGULpORYCOMMIS$100
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Rich r
. dhh tt, Director Projec Directorate 111-2 Division of Reactor Projects - l!!/IV/V Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: April 13, 1992 1
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ATTACHMENT TO LICENSE AMENM@l.JOS. 46 AND _46 EAcIllTY_ OPERATING LICENSE N05.-NPf-37 AND NPF-65 j
DCCKET N05. STN,30-454 AND STN $IL-4_H Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages.
The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change. Overleaf pages identified by an asterisk are provided for convenience, Remove Elggi insert Faces i
I 11 11 V*
V*
VI VI Vil Vl!
Vill Vill Xill Xill XIV XIV XV XV XVI XVI XVil XVII XVill XVill XIX-XIX XX XX XXI 1 3*
l-3*
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l-4 1-5 1-5 1-6 1-6 3/4 3-56 3/4 3-56 3/4 3-57 3/4 3-57 3/4 3-58 3/4 3-58 3/4 3-59 3/4 3-59 3/4 3-60 3/4 3-60 3/4 3-61 3/4 3-61 3/4 3-62 3/4 3-62 3/4.3-63 3/4 3-63 3/4 3-64 3/4 3-64 3/4 3-65 3/4 3-65 3/4 3-66 thru 73
+
3/4 11-1 3/4 11-1 3/4 11-2 3/4 11-2 3/4 11-3 3/4 11-3 3/4 11-4 thru 19 3/4 12-1 thru 14 B 3/4 3-6 B 3/4 3-6 B 3/4 3-7 8 3/4 3-7 B 3/4 3-8 B 3/4 3-8 8 3/4 3-9 B 3/4 7-7 8 3/4 7-7
? 3/4 11-1 0 3/4 11-1 B 3/4 11-2 B 3/4 11-2 l
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. Remove Paaes inserLPages 8 3/4 11-3 thru 8 3/4 11-7 8 3/4 12-1 thru 8 3/4 12-2 6-17*
6-17*
6-170 6-17b 6-18 6-18 6-18a 6-18a 6-19 6-19 6-20 6-20 6-21 6-21 6-22 6-22 6-23 6-23 6-24 6-24 i
6-25 thru 6-26 i
u l
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INDEX DEFINIT!UNS SECTION PAGE 1.0 DEFINITIONS 1.1 ACTION.......................................................
1-1 1.2 ACTUATION LOGIC TEST..............
1-1 1.3 ANALOG CHANNEL OPERATIONAL TEST...............................
1-1 1.4 AXIAL FLUX DIFFERENCE......................................
1-1 1.5 CHANNEL CALIBRATION................
1-1
- 1. 6 CHANNEL CHECK.................................................
1-1 1.7 CONTAINMENT INTEGRITY.........................................
1-2 1.8 C O N T R O L L E D L E A KAG E............................................
1-2 1.9 CORE ALTERATION............................
1-2 1.9.a CRITICALITY ANALYSIS OF BYRON AND BRAIDWOOD STATION FUEL STORAGE RACKS................................................
1-2 1.10 DIGITAL CHANNEL OPERATIONAL TEST.............................
1-2 1.11 DOSE EQUIVALENT I-131...........................
1-2a l
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 I D E NT I F I E D L E A KAG E...........................................
1-3 1.16 MAS T E R R E LAY T E ST..............................
1-3 1.17 MEMBER (S) 0F THE PUBLIC......................................
1-3 1.18 0 F F S I T E DO S E C A LCU LAT I O N MANU AL..............................
1-4 1.19 OPERABLE - OPERABILITY.......................................
1-4 1.19.a OPERATING LIMITS REP 0RT.....................................
1-4 l
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......................................
1-5 1.24 PURGE - PURGING..............................................
1-5 1.25 QUADRANT POWER TILT RAT 10....................................
1-5
' 26 RATED THERMAL P0WER................
1-5 1.27 REACTOR TRIP SYSTEM RESPONSE TIME............................
1-5 1.28 REPORTABLE EVENT........................................
1-5 l
BYRON - UNITS 1 & 2 I
AMENDMENT NO. 46 l
DEFINITIONS
\\
l SECTION PAGE i
1 1.29 SHUTOOWN MARGIN..............................................
1-5 1.30 SITE B00NDARY................................................
1-6 1.31 SLAVE RELAY TEST.............................................
1-6 1.32 DELETED......................................................
1-6 1.33 SOURCE CHECK.................................................
1-6 1.34 STAGGERED TEST BASIS.........................................
1-6 1.35 THERMAL POWER................................................
1-6 1.36 TRIP ACTUATING DEVICE OPERATIONAL TEST.......................
1-6 1.37 U N I D E NT I F I E D L E A KAG E.........................................
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 TASLE 1.2 OPERATIONAL M0 DES.......................................
1-9 i
BYRON - UNITS 1 &-2 II AMENDMENT NO. 46
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LIMITING CONDITIONS FOR OPERATION AND SURVE!LLANCE REQUIREMENTS SELTION 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 F (Z) AS A FUNCTION OF CORE HEIGHT...
3/4 2-5 q
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 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION......................
3/43-1 i
TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION...................
3/4 3-2 TABLE 3.3-2 (THIS TABLE IS NOT USE0)..............................
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 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTAT!0d.....................................
3/4 3-15 TABLE 3.3-4. ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETP0lNTS......................
3/4 3-23 TABLE 3.3-5 (THISTABLEISNOTUSE0)..............................
3/4 3-30 TABLE 4.3-2 -ENGINEERED SAFETY FEATURES ACTUATION-SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS...........
3/4 3-34 BYRON - UNITS 1 & 2 V
AMENDMENT NO.
23
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LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS r
SECTION PAGE 3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitoring for Plant Operations..
3/4 3-39 TABLE 3.3-6 RADIATION MONITORING INSTRUMENTATION FOR PLANT 0PERATIONS................................
3/4 3-40 l
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 3-44 TABLE 3.3-7 SEISMIC MONITORING INSTPUMENTATION....................
3/4 3-45 Meteorological Instrumentation...........................
3/4 3-47 TABLE 3.3-8 METEOROLOGICAL MONITOPING 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 T.\\BLE 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 TA3LE 4.3-7 -ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS...........................
3/4 3-55 TABLE 3.3-11 (This table number is not used.)......................
3/4 3-57 TABLE 3.3-12 (This table number is not used.)......................
3/4 3-57 Loose-Part Detection System..............................
3/4 3-58 Explosive Gas Monitoring Instrumentation.................
3/4 3-60 BYRON - UNITS 1 & 2 VI AMENDHENT NO. 46
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LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE TABLE 3.3-13 EXPLOSIVE GAS MONITORING INSTRUMENTATION..............
3/4 3-61 TA0LE 4.3-9 EXPLOSIVE GAS MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS...........................
3/4 3-62 High Energy Line Break Isolation Sensors.........
1 3/4 3-63 TABLE 3.3-14 HIGH ENERGY LINE BREAK INSTRUMENTATION...............
3/4 3-64 3/4.3.4 TURBINE OVERSPEED PROTECTION.............................
3/4 3-65 3/4.4 REACTOR COOLANT SYSTEM 4
3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION Startup and Power 0peration..............................
3/4 4-1 Hot Star.dby..............................................
3/4 4-2 Hot Shutdown..............................................
3/4 4-3 Cold Shutdown - Loops Fi11ed.............................
3/4 4-5 Cold Shutdown - Loops Not Filled.........................
3/4 4-6 Loop Isolation Valves-Operation..........................
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 REllEF. 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 l
l DURING INSERVICE INSPECT 10N.........................
3/4 4-18 TABLE 4.4-2 STEAM GENERATOR TUBE INSPECT 10N.......................
3/4 4-19 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE l
Leakage Detection Systems..-.............................
3/4 4-20 Operational Leakage......................................
3/4 4-21
-TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES......
3/4 4-23 3/4.4.7 CHEMISTRY..........................................
3/4 4-24 i
BYRON - UNITS 1-& 2-VII AMENDMENT N0. 46 l
1
... _ _,.. ~ -.. - - - -, -. _.,. -. -, _. -,, - - - _ -.. _. _, - - _.... _ _ _. _. _, ~... _.,
1 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 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 POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY
>l pCi/ GRAM DOSE EQUIVALENT I-131..................
3/4 4-29 TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY 5 AMPLE 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 F GURE 3.4-2b REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE UP TO 16 EFPY (lWIT 1,......
3/4 4-34 FIGURE 3.4-3a REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS APPLICABLE UP TO 32 EFPY (UNIT 1)......
3/4 4-35 FIGURE 3.4-3b REACTOR COOLANT SYSTEM C00LDOWN 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 3.4-4a NOMINAL PORV PRESSURE RELIEF SETPOINT VERSUS l
RCS TEMPERATURE FOR THE COLD OVERPRESSURE PROTECTION SYSTEM APPLICABLE UP TO 10 EFPY (UNIT 1).
3/4 4-40a FIGURE 3.4-4b NOMINAL PORY PRESSURE RELIEF SETPOINT VERSOS RCS TEMPERATURE FOR THE COLD OVERPRESSURE PROTECTION SYSTEM (UNIT 2).......
3/4 4-40b 3/4.4.10 STRUCTURAL INTEGRITY.....................................
3/4 4-42 3/4.4.11 REACTOR COOLANT SYSTEM VENTS.............................
3/4 4-43 BYRON - UNITS 1 & 2 VIII AMENDMENT NO. 46
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LIMITING CONDITIONS FOR OPERATION AND SURVEllLANCE REQUIREMENTS SEC110N PAGE 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN............
3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION, AND F0WER DISTRIBUTION LIMITS...
3/4 10-2 3/4.10.3 PHYSICS TESTS.........
3/4 10-3 3/4.10.4 REACTOR COOLANT L00PS...........................
3/4 10-4 3/4.10.5 POSITION INDICATION SYSTEM - SHUTDOWN..................
3/4 10-5 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS Liquid Holdup Tanks....................................
3/4 11-1 3/4.11.2 GASE0US EFFLUENTS Explosive Gas Mixture....................................
3/4 11-2 Gas Decay Tanks..................
3/4 11-3
=
BYRON - UNITS 1 & 2 XIII AMENDMENT NO. 46
1 I
BASES SECTION PAGE 3/4.0 APPLICABILITY...............................................
B 3/4 0-1 3/4.1 REACTIVIT) CONTROL SYSTEMS i
3/4.1.1 BORAT10N CONTR0L..........................................
B 3/4 1-1 3/4.1.2 BORAT10N SYSTEMS..........................................
B 3/4 1-2 3/4.1.3 MOVABLE CONTROL ASSEMBLIES................................
B 3/4 1-3 f
3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AX1AL 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 RAT 10.................................
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 3-6 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 l-l
-BYRON - UNITS 1 & 2 XIV AMENDMENT NO. 46 l
BASES SECTION PAGE 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-lb 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 EMERGEFCY CORE COOLING ;YSTEMS 3/4.5.1 A CvJ tV LA T 0 R S..............................................
B 3/4 5-1 3/4.5.2, 3/4.3.3 AND 3/4.5.4 ECCS SUBSYSTEMS.......................
B 3/4 5-1 3/4.5.5 REFUGLING WATER STORAGE TANK..............................
B 3/4 5-4 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..............................
8 3/4 6-4 3/4.6.4 COMBUSTIBLE GAS CONTR0L,..................................
B 3/4 6-4 l
l BYRON - UNITS 1 & 2 XV AMENDMENT NO. 46
. - -, _,, - -. ~
BASES i
SECTION PAGE e
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 7-5 3/4.7.8 SNVBBERS..................................................
B 3/4 7-5 3/4.7.9 SEALED SOURCE CONTAMINATION...............................
B 3/4 7-7
-3/4.7.10 (This specification number is not used)..................
B 3/4 7-7 3/4.7.11 (This specification number is not used)..................
B 3/4 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 BYRON - UNITS 1 & 2 XVI AMENDMENT NO. 46
1 BASES i
SECTION PAGE 3/4.9.6 REFUELING MACHINE.........................................
B 3/4 9-2 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 9-3 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..............
B 3/4 9-3 l
3/4.10 SPECIAL TEST EXCEPTIONS i
3/4.10.1 SHUTOOWN 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 C00LANi L00PS.....................................
B 3/4 10-1 3/4.10.5 POSITION INDICATION SYSTEM - SHUT 00WN.....................
B 3/4 10-1 3/4.11 RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS Liquid Holdup Tanks.................................
B 3/4 12 1 3/4.11.2 GASEOUS EFFLUENTS Explosive Gas Mixture...............................
B 3/4 11-2 Cas Decay Tanks.....................................
B 3/4 11-2 l
I BYRON - UNITS 1 & 2 XVII AMEN 0 MENT NO. 46
l DESIGN FEATURES 1
SECTION PAGE 5.1 SITE 5.1.1 EXCLUSION AREA..............................................
5-1 i
5.1. 2 LOW POPULATION 20NE.........................................
5-1 5.1.3 MAP DEFINING UNRESTRICTED AREAS AND SITE BOUNDARY FOR RADI0 ACTIVE GASEOUS AND LIQUID EFFLUENTS..............
5-1 FIGURE 5.1-1 EXCLUSION AREA AND UNRESTRICTED AREA FOR RADI0 ACTIVE GASE0US AND LIQUID EFFLUENTS.............
5-2 FIGURE 5.1-2 LOW POPULATION 20NE.................................
5-3 5.2 CONTAINMENT 5.2.1 C O N F I G U RAT I O N...............................................
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 ROD ASSEMBLIES......................................
5-4 5.4 REACTOR COOLANT SYSTEM 5.4.1 DESIGN PRESSURE AND TEMPERATURE.............................
5-4 b.4.2 V0LUME......................................................
5-4
- 5. 5 METEOROLOGICAL TOWER LOCATION.....................
5-4
- 5. 6 FUEL STORAGE
^
5.6.1 ' CRITICALITY.................................................
5-5 5.6,2 DRAINAGE....................................................
5-5 5.6.3 CAPACITY....................................................
5-b i
5.7 COMP 0NENT CYCLIC OR TRANSIENT LIMIT...........................
5-5 L
TABLE 5.7-1 COMPONENT CYCLIC OR TRANSIENT LIMITS..................
5-6 l
i BYRON - UNITS 1 & 2 XVIII AMENDMENT NO. 46 i
.,i,
ADMINISTRATIVE CONTROLS SECTION PAGE 6.1 RESPONSIBILITY...............................................
6-1 6.2 ORGANIZATION.................................................
6-1 FIGURE 6.2-1 (THIS FIGURE NOT USE0)...............................
6-3 FIGUPE 6.2-2 (THIS FIGURE NOT USE0)...............................
6-4 TARLE 6.2-1 MINIMUM SHIFT CREW COMPOSITION........................
6-5 6.3 UNIT STAFF QUALIFICATIONS....................................
6-6 6.4 TRAINING......................................................
6-6
- 6. 5 REVIEW INVESTIGATION AND AUDIT...........
6-6 6.5.1 0FFSITE.....................................................
6-6 Offsite Review and Investigative Function...................
6-6 Station Audit Function......................................
6-8 Authority...................................................
69 Records.....................................................
6-10 Procedures..................................................
6-10 Personnel...................................................
5-10 6.5.2 0NSITE......................................................
6-12 Onsite Review and Investigative Functions...................
6-12 Responsibility..............................................
6-12 Authority...................................................
6-13 i
Records.....................................................
6-14 Procedures.................................................
6-14 i
i Personnel...................................................
6-14 6.6 REPORTABLE EVENT ACTI0N.......................................
6-15 1.
l BYRON
. UNITS 1 & 2 XIX AMENDMENT NO. 46 o
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ADMINISTRATIVE CONTROLS 7.-
SECTION PAGE 6.7 SAFETY LIMIT V10LAT10N........................................
6-15 6.8 PROCEDURES AND PR0 GRAMS.......................................
6-15 6.9 REPORTlhG 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..............
6-19 Monthly Operating Report....................................
6-19 Operating Limits Report.....................................
6-19 Criticality Analysis of Byron and Braidwood Station fuel Storage Rocks................................
6-20 6.9.2 SPECIAL REP 0RTS.............................................
6-20 6.10 RECORD RETENT10N.............................................
6-20 6.11 RADIATION PROTECTION PR0 GRAM.................................
6-21 6.12 HIGH RADIATION AREA.......................................
6-22 6.13 PROCESS CONTROL PROGRAM (PCP)................................
6-23 6.14 0FFSITE DOSE CALCULATION HANUAL (0DCM).......................
6-23 BYRON - UNITS 1 & 2 XX AMEN 0 MENT NO. 46
DEFINITIONS t' - 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 avorAce bel 3 and gamma energies per disintegration (MeV/d) for the radionuclides in the sample.
ENGINEERED SAFETY FEATURES RESPONSE TIME 1.13 The ENGINEERED SAFETY FEA1URES (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 N01ATION specified fnr 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 80l'NDARY 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 REta, TEST shall include a continuity check of each associated slave relay.
MEMBER (S) 0F THE PUBLIC 1.17 MEMBER (5) 0F THE PUBLIC shall include all persons who are not occupationally associated with the plant.
This category does not include employees 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.
BYRON - UNITS 1 & 2 1-3
[
l DEFINITIONS OFFSITE DOSE CALCULATION MANUAL 1.18 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radio-active gaseous and liquid effluents, in the calculation of gastous and liquid effluent monitoring Alarm / Trip Setpoints, and in the conduct of the Environ-mental Radiological Monitoring Program.
The ODCM shall also contain n ) the Radioactive Effluent Controls and Radiological Environmental Monitori.s Programs required by Sections 6.8.4e and f, and (2) descriptions of the information that thould be included in the Anr,ual Radiological Environmental Operating and Semiannual Radioactive Effluent Release Reports required by Specifications 6.9.1.6 and 6.9.1.7.
g 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 attendcnt 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 f inction(s) are also capable (.f performing their related support function (s).
OPERATING L?MITS REPORT 1.19.a The OPERATING LIMITS REPORT is the unit-specific document that provides l
operating linits for the current operating reload cycle.
These cycle-specific operating limits shall be determined for each reload cycle in accordance with Specification 6.9.1.9.
Plant operation within these operating limits is addressed in individual specifications.
OPERATIONAL MODE - MODE 1.2n 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 fundamental 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 tube leakage) through a nonisolable fault in a Reactor Coolant System component body, pipe wall, or vessel wall.
BYRON - UNITS 1 & 2 1-4 AMENDMENT NO. 46
DEFINI110NS PROCESS CONTROL PROGRAM 1.23 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, tests, and determinations to be made to ensure thet processing and packliging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in suchawayastoassurecompliancewith10CFRParts20}reme,ntsgoverningthe 61 and 71, State regulations, burial ground requirements, and other requ disposal of solid radioactive waste.
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, i
QUADRANT POWER TILT RATIO 1.25 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average cf the upper excore detector cali-ibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore + u tor 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 SHUTOOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subtritical'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.
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BYRON - UNITS 1 & 2 1-5 AMEMMENT NO 46 p
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1 DEFINITIONS 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 c,f OPERABILITY of each relay._ The SLAVE RELAY TEST shall include a contf ruity check, as a rsinimum, of associated testable actuation devices.
SOLIDIFICATION 1.32 Deleted
-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 OPERAP"LITY 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.
-1.37 UNIDENTIFIED LEAKAGE shall be all. leakage which is not IDENTIFIED LEAKAGE or CONTROLLED LEAKAGE.
UNRESTRICTED AREA 1.39 An UNRESTRICTED AREA shall be any area at or beyond the SITE B0UNDARY 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 B0UNDARY used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes.
BYRON - UNITS 1 & 2 6 AMENDMENT NO. 46
INSTRUMENTATION MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.7 (This specification number is not used.)
c BYRON - UNITS 1-& 2 3/4 3-56 AMENDMENT NO. 46
TABLE 3.3-11 (This table ntrnber is not used.)
TABl.E 3.3-12 (This table number is not used.)
BYRON - UNITS 1 & 2 3/4 3-57 AMENDMENT NO. 46
4 INSTRUMENTATION LOOSE-PART DETECTION SYSTEM LIMITING CONDITION FOR OPERATION 3.3.3.8 The Loose-Part Detection System shall be OPERABLE.
APPLICABILITY:
MODES 1 and 2.
ACTION:
a.
With e or more Loose-Part Detection System-channels inoperable for more..an 30 days, prepare and submit a Special. Report to the Commission persuant to Specification 6.9.2 within the next 10 days outliaing the cause of the malfunction and the plans for restoring the channel (s) to OPERABLE status.
b.
The provisi,..s 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 cemonstrated OPERABLE by performance of:
a.
A CHANNEL CHECK at-least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, 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, BYRON - UNITS 1 & 2 3/4 3-58 AMENDMENT NO. 46
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INSTRUMENTATION RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION i lMITING CONDITION FOR OPERATION 3.3.3.9 (This specification number is not used.)
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BYRON - UNITS 1 & 2 3/4 3-59 AMENDMENT NO. 46
4 1
INSTRUMENTATION EXPLOSIVE GAS MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.10 The explosive gas monitoring instrumentation channels shown in Table 3.3-13 shall be OPERABLE with heir Alarm / Trip Setpoints set to ensure that the limits of Specification 3.11.2.5 are not exceeded.
APPLICABILITY:
As shown in Table 3 3-13 ACTION:
a.
With an explosive gas monitoring instrumentation channel Alarm / Trip Setpnint less conservative than required by the abose specif"Ication, declare the channel inoperable and take the ACTION shown in-Table 3.3-13.
b.
With less than the minimum nuniber of explosive gas monitoring instru-mentation channels OPERABLE, take the ACTION shown in Table 3.3-13.
Restore the inoperable instrumentation to OPERABLE status within 30 days and, if unsuccessful, prepare and submit a Special Report to the Commission within the next 30 days pursuant to Specification 6.9.2 to explain why this inoperability was not corrected in a timely manner.
c.
The provisions of Specifications 3.0.3, and 3.0.4 are not applicable.
SURVEILLANCE RE0UIREMENTS 4.3.3.10 Each explosive gas monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, CHANNEL CALIBRATION and DIGITAL CHANNEL OPERATIONAL TEST at the frequencies shown in Table 4.3-9.
DYRON - UNITS 1 & 2 3/4 3-60 AMENDHENT N0. 46
l TABLE 3.3-13 h
EXPLOSIVE GAS MONITORING INSTRUMENTATION MINIMUM CHANNELS k
INSTRUMENT OPERABLE APPLICABILITY ACTION ii 1.
(Not Used) g e
2.
(Not Used) l 3.
Gaseous Waste Management System Hydrogen Analyzer (CAT-GW8000) 1 38 a.
b.
Oxygen Analyzer (0AT-GW8003) 1 38 wg c,
Waste Gas Compressor Discharge Oxygen Analyzer (OAIT-Gw004) 1 38 h
TABLE NOTATIONS
- (not used)
- During WASTE GAS HOLDUP SYSTEM operation.
- 0uring Waste Gas Compressor Operation.
- All instruments required for Unit 1 or Unit 2 operation.
ACTION STATEMENTS I
ACTION 38 - With the number of channels OPERABLE less than required by the Minimum Channels GPERABLE requirement, operation of this system may continue provided grab samples are taken and g
analyzed at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during degassing operations and at least or.ce per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> gg during other operations.
2 5
u TABLE 4.3-9 m
35 S
EXPLOSIVE GAS MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS DIGITAL b
CHANNEL MODES FOR WHICH g
CHANNEL SOURCE CHANNEL OPERATIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CHECK CALIBRATION TEST IS REQUIRED g
b 1.
(Not Used) 2.
(Not Used) 3.
Gaseous Waste Management System Hydrogen Analyzer (CAT GW8000) 0 N.A.
Q(4)
M a.
R3 b.
Oxygen Analyzer (OAT-GW8003)
D N.A.
Q(5)
M b
c.
Waste Gas Compressor Discharge g
Oxygen Analyzer (OAIT-GW004)
D N.A.
Q(5)
M TABLE NOTATIONS l
- (Not used)
- During WASTE GAS HOLDUP SYSTEM operation.
- During Waste 6as Compressor Operation.
- All instruments required for Unit 1 or Unit 2 operation.
(1) (Not used)
(2) (Not used) 3 (3) (Not used)
(4) The CHANNEL CALIBRATION shall incle:2 the use of standerd gas samples containing hydrogen and nitrogen.
(5) The CHANNEL CALIBRATION shall include the use of standard gas samples containing oxygen and nitrogen.
5
m.
INSTRUMENTATION HIGH ENERGY LINE BREAK ISOLATION SENSORS LIMITING CONDITION FOR OPERATION 3.3.3.11 The high energy line break instrumentation shown in Table 3.3-14 shall be OPERABLE.
APPLICABILITY:
As shown in Table 3.3-14.
ACTION:
a.
With the number of OPERABLE auxiliary steam isolation instruments less than the Minimum Channels OPERABLE as required by Table 3.3-14, restore the inoperable instru'nent(s) tn OPERABLE status within 7 days, or suspend the supply of auxiliary steam to the Auxiliary Building, or establish a continuous watch in the affected area (s) until the inoperable sensors are restored to OPERAB' status.
b.
With the number of OPERABLE steam generator blowdown line isolation instru-ments less than the Minimum Channels OPERABLE as required by Table 3.3-14, restore the inoperable instrument (s) to OPERABLE status within 7 days, or limit the total steam generator blowdown flow rate to less than or equal to 60 gpm or establish a continuous watch in the affected area (s) until the inoperable sensors are restored to OPERABLE status.
c.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANDEREQUIREMENTS 4.3.3.-12 Each of the above high energy line break isolation instruments shall be demonstrated OPERABLE by the performance of an ANALOG CHANNEL OPERATIONAL TEST AND CHANNEL CALIBRATION at least once per 18 months.
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l BYRON - UNITS 1 & 2 3/4 3-63 AMENDMENT NO. 46 l
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Table 3.3-14 HIGH ENERGY LINE BREAK INSTRUMENTATION ISOLATION INSTRUMENT MINIMUM CHANNELS APPLICABl.E FUNCTION CHANNEL OPERABLE HODES 1.
Auxiliary Steam OTS-AS031A 1
Isolation OTS-AS032A OTS-AS031B 1
OTS-AS0328 OTS-AS031C 1
OTS-AS032C OTS-AS031D 1
OTS-AS0320 OTS-AS031E 1
OTS-AS032E OTS-A5031F 1
OTS-A5032F 2.
Steam Generator TS-50045A 1
1,2,3,4 Blowdown Line TS-SD045B Isolation TS-SD046A-1 1,2,3,4 TS-SD046B TS-SD045C 1
1,2,3,4 TS-SD045D TS-50046C 1
1,2,3,4 TS-SD0460 t
- Required when auxiliary steam is being supplied, from any source, to the Auxiliary Building.
t BYRON - UNITS 1 & 2 3/4 3-64 AMENDMENT-N0. 46
)
INSTRUMENTATION 3/4.3.4 TURBINE OVERSPEED PROTECTION LIMITING CONDITION FOR OFERATION 3.3.4 At least one Turbine Overspeed Protection System shall be OPERABLE.
APPLICABILITY; ' MODES 1, 2, and 3.
ACTION:
With'one throttle valve or one governcr valve per high pressure turbine a.
steam lino inoperable and/or with one reheat stop valve or one reheat intercept valve per low pressure turbine steam line inoperable, restore the inoperable valve (s) to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or close at least one valve in the affected steam line(s) or isolate the turbine from the steam supply withi, the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
b.
With the above required Turbine Overspeed Protection System otherwise inoperable, within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> isolate the turbine from the steam supply.
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-l
- position, l-
_c.
During turbine operation at least once per 31 days by direct observa-tion,c of freedom of movement of each of the 12 extraction steam non-return check valve weight arms, l
d.
At least once per 18 months by performance of CHANNEL CALIBRATION L
on the Turbine Overspeed Protection Systems, md 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 stems and verifying no unacceptable flaws or corrosion.
BYRON - UNITS 1 & 2 3/4 3-65 AMENDMENT N0. 46 l
i j
3/4.11 RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS LIQUID HOLOUP TANKS LIMITING CONDITION FOR OPERATION 3.11.1.1 Deleted 3.11.1.2 Deleted 3.11.1.3 Deleted 3.11.1.4 The quantity of radioactive material, excluding tritium and dissolved or entrained noble gases, contained in any outside tanks shall be limited to the following:
a.
Primary Water Storage Tank 5 2000 Curies, and b.
Outside Temporary Tank
$ 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 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> 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.
BYRON - UNITS 1 & 2 3/4 11-1 AMENDMENT NO. 46
RADI0 ACTIVE EFFLUENTS 3/4.11.2 GASEOUS EFFLUENTS EXPLOSIVE GAS MIXTURE LIMITING CONDITION FOR OPERATION 3.11.2.5 The concentration of oxygen in the WASTE GAS HOLDUP SYSTEM shall be limited to less than or equal to 2% by volume whenever the hydrogen concentration exceeds 4% by volume.
APPLICABILITY:
At all times.
ACTION:
a.
With the concentration of oxygen in the WASTE GAS HOLDUP SYSTEM greater thcn 2% by volume but less than or equal to 4% by volume, reduce the oxygen concentration to the above limits within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, b.
With the concentration of oxygen in the WASTE GAS HOLOUP SYSTEM greater than 4% by volume and the hydrogen concentration greater than 4% by volume, immediately su= pend all additions of wastt gases to the system and reduce the co.:entration of oxygen to less than or equal to 4% by volume, then take ACTION a. above.
c.~.The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.2.5 The concentrations of hydrogen and oxygen in the WASTE GAS HOLOUP SYSTEM shall be determined to be within the above limits by continuously monitoring the waste gases in the WASTE GAS HOLDUP SYSTEM with the hydrogen and oxygen monitors required OPERABLE by Table 3.3-13 of Specification 3.3.3.10.
BYRON - UNITS 1 & 2 3/4 11-2 AMENDMENT NO. 46
'RADI0 ACTIVE' EFFLUENTS GAS DECAY TANKS LIMITING CONDITION FOR OPERATION 3.11.2.6 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).
APPLICAB?.LITY:
At all times.
ACTION:
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 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, reduce the tank contents to within the limit, and describe the events leading to this condition in the next Semiannual Radioactive Effluent i
Release Report, pursuant to Specification 6.9.1.7.
l l
b.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS L
4.11.2.6 The quantity of radioactive material ccntained in each gas decay tank shall be determined to be within the above limit at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when radioactive materials are being added to the tank.
i I
BYRON - UNITS 1 & 2 3/4 11-3 AMENDMENT N0. 46
INSTRLHENTATION BASES 3/4.3.3.8 LOOSE PART DETECTION SYSTEM The OPERABILITY of-the loose part detection system ensures that sufficient capability is available to detect loose metallic parts in the Reactor Coolant System and avoid or mitigate dataage 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.
3/4.3.3.9 DELETED
-3/4.3.3.10- EXPLOSIVE GAS MONITORING INSTRUMENTATION The-instrumentation includes _ provisions for monitoring the concentrations of potentially explosive gas mixtures in the WASTE GAS HOLDUP SYSTEM.
3/4 3.3.11-HIGH ENERGY LINE BREAK ISOLATION SENSORS The OPERABILITY of the high energy line break isolation sensors ensures that the capability is available to promptly detect and initiate protective action-in the event of a-line break.
This capability is required to prevent the potential for damage to-safety-related systems and structures in the auxiliary building.
l 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 cuuld generate pote,ntially 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 tesign recognizes this potential in providing rapid' action, dual. shut off capability in each path,
-remote testing capability, and a Ylow path that reduces-the effects of changes i=
-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.
BYRON - UNITS 1 AND 2 B 3/4 3-6 AMENDMENT NO. 46
O INSTRUMENTATION BASES TURBINE OVERSPEED PROTECTION (continued)
Specification 4.3.4.2b and c (Extraction Steam Non-Return l' heck 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.
Manual stroking of the extraction steam non-return valves is possible under shutdown conditions by latching the turbine and applying the air 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 Ecior to entering Mode 3 from Mode 4.
The engineering specifications provided for testing the extraction steam non-return check valvas during operation by equalizing the air pressure across the piston in the spri'ig cylinder, permitting the spring to partially close the disk against tM 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 5 team 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.
BYRON - UNITS 1 AND 2 8 3/4 3-7 AMENDMENT N0. 46
INSTRUMENTATION
@ASES TURBINE OVERSPEED PROTECTION (continued)
The motor nperated valves are too large for routine manual operation, do not have bypasses to allow controlled warmup conditions, str9ke 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 intermediate 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 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.
BYRON - UNITS 1 AND 2 8 3/4 3-8 AMENDMENT N0. 46
PLANT SYSTEMS BASES 3/4.7.9 SEALED SOURC" 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 net 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 (This specification number is not used.)
3/4.7.11 (This specification number is not used.)
3/4.7.12 AREA TEMPERATURE MONITORING The arca 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.
m BYRON - UNITS 1 & 2 B 3/4 7-7 AMENDMENT NO. 46
3/4.11 RADI0 ACTIVE EFFLUENTS BASES 3/4.11.1 LIQUID EFFLUENTS 3/4.11.1.1 DELETED 3/4.11.1.2 DELETED 3/4.11.1.3 DELETED 3/4.11.1.4 LIQUID HOLDUP TANKS The tanks-listed in thic a,ecification 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 radiorctive material contained in the specified tanks provides assurance that in the ovent of an uncontrolled release of the tanks' contents, the resulting concen; rations 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.
BYRON - UNITS 1 & 2 B 3/4 11-1 AMENDMENT NO. 46
m.
RADIDACTIVE EFFLUENTS BASES 3.4.11.2 GASEOUS EFFLUENTS 3.4.11.2.1 DELETED 3.4.11.2.2 DELETED 3.4.11.2.3 DELETED-3/4.-11.2.4 DELETED 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 autom& tic control features include isolation of the source of hydrogen and/or oxygen, automatic diversion to recombiners, or ihjection 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 of General Design Criterion 60 of Appendix A to 10 CFR Part SU.
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.
Restricting the quantity of radioactivity contained in each gas storage tankprovidesassurancethatintheeventofanuncontrolledreleaseofthe tanks contents,:the resulting:whole body exposure to a-MEMBER OF THE PUBLIC 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.
BYRON - UNITS 1 & 2 B 3/4 11-2 AMENDMENT NO. -46
- ~...
ADMINISTRATIVE CONTROLS 4.
PROCEDURES'AND PROGRAMS (Continued)
I 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 interpretation of the data, and (b) the sequence and timing of administrative events required to initiate corrective actior..
d.
Post-accident Sampling A program which will ensure the capability to obta'n and analyze reactor coolant, radioactive iodines and particulaces 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.
BYRON - UNITS 1 & 2 6-17
ADMINISTRATIVE CONTROLS PROCEDURESANDPROGRAMS(Continuedj e.
Radioactive Effluent Controls Program A program shall be provided conforming with 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to MEMBERS OF THE PUBLIC from radioactive effluents as low as reasonably achievable.
The program (1) shall be contained in the ODCM, (2) shall be implemented by station procedures, and (3) shall include remedial actions to be taken whenever the program limits are exceeded.
The program shall include the-following elements:
1)
Limitations on the operability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM, 2)
Limitations on the concentrations of radioactive material released in liquid effluents to UNRESTRICTED AREAS conforming to 10 CFR Part 20, Appendix 8, Table II, Column 2, 3)
Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.106 and with the methodology and parameters in the ODCM, 4)
Limitations on the annual and quarterly doses or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released from each unit to UNRESTRICTED AREAS conforming to Appendix I to 10 CFR Part 50, 5)
Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 51 days, 6)
Limitations on the operability and use of the liquid and gaseous effluent treatment systems to ensure that the appropriate portions-of these systems are used to reduce releases of radioactivity when the projected doses in a 31-day period would exceed 2 percent of the guidelines for the annual dose or dose commitment conforming to Appendix I to 10 CFR Part 50, 7) limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the SITE BOUNDARY conforming to the doses associated with 10 CFR Part 20, Appendix B, Table II, Column 1, 8)
Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50, BYRON - UNITS 1 & 2 6-17a AMENDMENT NO. 46
ADMINISTRATIVE-CONTROLS PROCEDURES AND' PROGRAMS (Continued) i P
9)
Limitations on the annual and quarterly doses to a MEMBER 0F THE PUBLIC from Iodine-131, Iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released from each unit to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50, and 10)
Limitations on the annual dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources conforming to 40 CFR Part 190.
f.
Radiological Environmental Monitoring Program A program shall be provided to monitor the radiation and radionuclides in the environs of the plant.
The program shall' provide (1) representative measurements of radioactivity in the highest potential exposure pathways, and (2) verification of the accuracy of the effluent monitoring program and modeling of environmental exposure pathways.
The program shall (1) be contained in the ODCM, (2) conform to the guidance of Appendix I to 10 CFR Part 50, and (3) include the following:
1)
Monitoring, sampling, analysis, and reporting of radiation arm radionuclides in the environment in accordance with the methodology and parameters in the ODCM, 4
2)
A Land Use Census to ensure that changes in the use of areas at and beyond the SITE B0UNDARY are identified and that modifica-tions to the monitoring program are made if required by the results of this census, and-3)
Participation in a Interlaboratory Comparison Program to ensure that independent checks on the precision and accuracy of the measurements-of radioactive materials in environmental sample f
matrices are performed as part of the quality assurance program for environmental monitoring.
L L
I BYRON - UNITS 1 & 2 6-17b AMENDMENT N0. 46
ADMINISTRATIVE-CONTR01S 6_. 9 REPORTING REQUIREMENTS ROUTINE REPORfS 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) modificrtions that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant.
6.9.1.2 The Startup Report shall address each of the tests identified in the 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.
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 following initial criticality.
6.9.1.5 Reports required on an annual basis shall include:
a.
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-rem exposure according to work and job _ functions,* e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling.
The dose 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 n.eed not be accounted for.
In the aggregate, at least 80% of the total whole body dose received from external sources should be assigned to specific major work' functions.
- This tabulation supplements the requirements of 620.407 of 10 CFR Part 20.
BYRON - UNITS 1 & 2 6-18 AMENDMENT NO. 46
ADMINISTRATIVF CONTROLS REPORTING REQUIREMENTS (Continued) b.
The results of specific activity 'aalysis in which the primary coolant exceeded the limits of Specification 3.4.8.
The following information shall be included:
(1) Reactor power history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded; (2) Results of the last isotopic analysis for radiciodine performed prior to exceeding the limit, results of analysis while limit was exceeded and results of one analysis after the radiciodine activity was reduced to less than limit.
Each result should include date and time of sampling and the radiciodine concentrations; (3) Clean-up system flow history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded; (4) Graph of the I-131 concentration and one other radiciodine isotope concentration in microcuries per gram as a function of time for the duration of the specific activity above the steady-state level; and (S) The time duration when the specific activity of the primary coolant exceeded the radioiodine limit.
BYRON - UNITS 1 & 2 6-18a AMENDMENT NO. 46
ADMINISTRATIVE CONTROLS REPORTING RE0VIREMENTS (Continued)
ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT *
- 6. 9.1. 6 The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year.
The report shall include summaries, interpreta-tions, and analysis of trends of the result <
f the Radiological Environmental Monitoring Program for the reportinc) perioo.
The material provided shall be consistent with the objectives outlined in (1) the ODCM and (2) Sections IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR Part 50.
SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT **
6.9.1.7 The Semiannual Radioactive Effluent Release Report covering the opera-tion 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 report shall in-clude a summary of the cuantities of radioactive liquid.and gaseous effluents and solid waste releasec from the unit.
The material provided shall be (1) con-sistent with the objectives outlined in the ODCM and PCP and (2) in conformance with 10 CFR 50.36a and Section IV.B.1 of Appendix I to 10 CFR Part 50.
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.
OPERATING LIMITS REPORT 6.9.1.9 Operating limits shall be established and documented in the OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle.
The analytical methods used to determine the operating limits shall be those previouslyreviewedandapprovedbytheNRCinTopicalReports: 1) WCAP 9272-P-A Westinghouse Reload Safety Evaluations Methodology" dated July 1985,
- 2) WCAP-8385 " Power Distribution Control and Load Followiny Procedures" dated September 1974, 3) NFSR-0016 " Benchmark of PWR Nuclear Deslgn Methods" dated July 1983, and/or 4) NFSR-0081 " Benchmark of PWR Nutlear Design Methods Using the PH0ENIX-P and ANC Computer Codes" dated July 1990.
The operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.
The OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements.thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator.and Resident Inspector.
~
- A single submittal may be made for a multi-unit station.
- A single submittal may be made for a multi-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.
BYRON'- UNITS 1 & 2 6-19 AMENDMENT NO. 46
, _ ~, _
ADMINISTRATIVE CONTROLS CRITICALITY ANALYSIS OF BYRON AND BRAlbWOOD STATION FUEL STORAGE RACKS 6.9.1.10 Fuel enrichment limits for storage shall be established and documented in the CRITICALITY ANALYSIS OF BYRON AND BRAIDWOOD STATION FUEL STORAGE RACKS.
The analytical methods used to determine the maximum fuel enrichments shall be those previously reviewed and approved by the NRC in
" CRITICALITY ANALYSIS OF BYRON AND BRAIDWOOD STATION FUEL STORAGE RACKS" The fuel enrichment limits for storage shall be determined so that all applicable limits (e.g., fubcriticality) of the safety analysis are met.
The CRIT!JALITY ANALYSIS OF BYRON AND BRAIDWCOD STATION FUEL STORAGE RACKS report shall be provided upon issuance of any changes, to the NRC Document Control Desk, with copies to the Regional Administrator and the Resident Inspector.
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:
Records and drawing changes reflecting unit design modifications a.
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; BYP.ON - UNITS 1 & 2 6-20 AMENDMENT NO. 46
ADMINISTRATIVE CONTROLS RECORD RETENTION (Continued) c.
Records of radiation exposure for all individuals entering radiation control areas; d.
Records of gaseous and liquid radioactive material released to the environs; Records of transient or operational cycles for those unit components e.
identified in. Table 5.7-1; f.
Records of-reactor test.5 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 E.e to procedures or equipment or reviews of tests and experimencs 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; 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 ef_fective at specified times and QA records showing that these procedures-were followed,-and o.
Records of reviews performed for changes made to the OFFSITE DOSE CALCULATION MANUAL and the PROCESS CONTROL PROGRAM.
6.11 RADIATION PROTECTION PROGRAM Procedures for personnel-radiation protection shall be prepared c.onsistent with the requfrements of 10 CFR Part 20 and shall be approved, maintained and adhered _to for all operations involving personnel radiation exposure.
BYRON - UNITS 1 & 2 6-21 AMENDMENT N0. 46
~.
ADMINISTRATIVE CONTROLS 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 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 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 pro-vided 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 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 shall perform periodic radiation surveillance at the frequency specified in the Radiation Work Permit.
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.
BYRON - UNITS 1 & 2 6-22 AMENDMENT N0. 46
-ADMINISTRATIVE CONTROLS HIGH RADIATION AREA (Continued)
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.
6.13 PROCESS CONTROL PROGRA?1 (PCP) 6.13.1 Changes to the'PCP:
l a.
Shall be documented and records of reviews performed shall be retained-as required by Specification 6.10.20.
This documentation-shall contain:
1)
Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change (s)-
- and, 2)
A determination that the change will maintain the overall conformance of the solidified waste product to existing requirements of Federal, State, or other applicable regulations.
b.
Shall become effective after review and acceptance by the Onsite Review and Investigative Function (Onsite Review) and the approval of the Station Manager.
6.14 0FFSITE DOSE CALCULATION MANUAL (0DCM) 6.14.1 Changes to the ODCM:
4 a.
Shall be documented and records of reviews performed s bil be retained as required by Specification 6.10.20.
This documentation shall contain:
L
'1)
Sufficient information to support the change together with the l
appropriate analyses or evaluations justifying the change (s)_
l and,_
l 2)
A determination that the change will maintain the level of radioactive effluent control required by 10 CFR 20.106, 40 CFR Part 190, 10 CFR 50.36a, and Appendix I to 10 CFR Part 50 and.
not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations, b.
Shall become: effective after review and acceptance by the Onsite Review and Investigative Function and the approval of the Station Manager on the date specified by the Onsite Review and Investigative Function.
1 BYRON - UNITS 1 & 2 6-23 AMENDMENT N0. 46
~
ADMINISTRATIVE CONTROLS OFFSITE DOSE CALCULATION MANUAL (0DCM) (Continued)
Shall be submitted to the Commi3sion in the form of a complete, c.
legible copy of the entire 00CM as a part of or concurrent with the Semiannual Radioactive Effluent Release Report for +.he period of the report in which any change to the ODCM was made effective.
Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shail indicate the date (e.g., month / year) the change was implemented.
F BYRON - UNITS 1 & 2 6-24 AMENDMENT NO. 46 l
[o, UNITED STATES NUCLEAR REGULATORY COMMISSION n
4 WASHINoTON, D. C 20555 i
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COMMONWEALTH EDISON COMPANY DOCKET N0. STN 50-4M BRAIDWOOD STATION. UNIT NO. 1 AMENDMENT TO FAClllTY OPERAT?NG llCENSE Amendment No. 35 License No. NPF-72 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Commonwealth Edison Company (the licensee) dated June 10, 1991 as supplemented on October 17, 1991, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter 1; B.
The facility will operate in conformity with the application, the pmvisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety 4 the public, and (ii) that such activities will be conducted compliance with the Commission's regulations; D.
The issua <e of this amendment will not be inimical to the common defense a security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifi-t cations"as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-72 is hereby amended to read as follows:
l l
l
. (2)
Technical Spfcificallans The Technical Specifications contained in Appendix A as revised through Amendment No. 35 and the Environmental Protectice Plan contained in Appendix 0, both of which are attached her9to, ar.e herely incorporated into this license. The licensee f.hal) operate ths facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is offective as of the date of its issuancq.
FOR THE NUCLEAR REGULATORY COMMISSION 3
ll Richar/ ;/f' fi Barrett, Director Project Directorate 111-2 Division of Reactor Projects - lil/IV/V Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: April 13, 1992 4
'9 r
3Ma f
ug g
y, UNITED STATES e
vn i
NUCLE AR REGULATORY COMMISSION
$/
t WASHINGTON, D C. 20555
.....J COMMONWEALTH EDISON COMP &tiY D.0CKET fq. STN 50-457 EMIDWOOD STATION. UNIT NO. 2 l
6tiUiDMENT TO FAClllTY OPERATING llCENSE Amendment No. 35 License No. NPF-77 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The applicatien for amendment by Comraonwealth Edison Company (the licensee) dated June 10, 1991 as supplemented on October 17, 1991.
complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter 1; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and saiety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Concission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifi-cations'as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-77 is hereby amended to read as follows:
m._ _ __. _ _. _ _ _. _ _ -
d (2)
Tec hnical_.Sp_qci fic at iopJ The Technical Specifications contained in Appendix A as revised through Amendment No. 35 and the Environmental Protection Plan containad in Appendix 0, both of which were attached to License No. NPF-72, dated July 2. 1987, are hereby incorporated into this license The licensee shall operate the ft.c111ty in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of the date if its issuance.
FOR THE NUCLEAR REGULAT RY COMMISSION
)
l /QS~,
(/
/
Richar J Barrett, Director Project rectorate 111-2 Division of Reactor Projects - til/IV/V Office of Nuclear Reactor Regulation
Attachment:
Changes to the Iechnical Specifications Date of Issuance:
April 13, 1992 I
l
. -. ~..
ATTACHMENT TO LICENSE AMENDMENT NOS. 35 AND 35 FACILITY QPLRf 1NG LICENSE NOS. NPF-72 AND NPF-77 POCKET N05. STN 50-456 AND STN 50-457 Replace the following pages of the Appendix "A" Technical Specifications with the attached pagen.
The revised pages are identified by amendment number and contain vertical lines indicating the area of change.
Overleaf pages identified by an asterisk are provided for convenience.
Ermove Paaes Insert Pacci l
I 11 11 V*
V*
VI VI Vil VII XI*
XI*
XII XII-Xill Xill XIV XIV XV XV XVI XVI XVII XVII XVill XVill XIX XIX XX XX XXI l
1-3*
l-3*
1-4 1-4 1-5 1-5 1-6 1-6 3/4 3-57 3/4 3 57 3/4 3-58*
3/4 3-58*
3/4 3-59 3/4 3-59 i
3/4 3-60 3/4 3-60 3/4 3-61 3/4 3-61 3/4 3-62 3/4 3-62 1
3/4 3-63 3/4 3-63 3/4 3-64 3/4 3-64 3/4 3-65 3/4 3-65 3/4 3-66 thru 75 3/4 11-1 3/4 11-1 3/4 11-2 3/4 11-2 3/4 11-3 3/4 11-3 3/4 11-4 thru 19 l
3/4 12-1 thru 12-14 l
B 3/4 3-5*
B 3/4 3-5*
i B 3/4 3-6 B 3/4 3-6 B 3/4 3 B 3/4 3-7 8 3/4 3-8 8 3/4 3-8 8 3/4 11-1 B 3/4 11-1 B 3/4 11-2 B 3/4 11-2 B 3/4 11-3 thru B 3/4 11-7 B 3/4 12-1 thru B 3/4 12-2 1
Remove Patles insert pas M 6-Sa 6-17*
6-17*
6-17a 6-17b 6-19 6-19 6-20 6-20 6-21 6-21 6-22 6-22 6-23 6-23 6-24 6-24 6-25 thru 6-26 1
(
INDEX DEFINITIONS SECTION PAGE 1.0 DEJINITIONS 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 CONTRO L L E D L E A KAG E............................................
1-2 1.9 CORE ALTERATION...............................................
1-2 1.9.a CRITICALITY ANALYSIS OF BYRON AND BRAIDWOOD STATION FUEL STORAGE RACKS................................................
1-2 1.10 DIGITAL CHANNEL OPERATIONAL TEST.............................
1-2 1.11 DO S E E QU I VA LEN T I - 131........................................
1-2 1.12 5-AVERAGEDISINTEGRATIONENERGY..............................
1-3 1.13 ENGINEERED SAFETY FEATURES RESPONSE TIME.....................
1-3 1.14 FREQUENCY N0TATION...........................................
13 1.15 I D E NT I F I E D L E A KAG E...........................................
1-3 1.16 MASTER RELAY TEST............................................
1-3 l
1.17 MEMBER (S) 0F THE PUBLIC......................................
1-3 1.18 0FFSITE DOSE CALCULATION MANUAL..............................
1-4 1.19 OPERABLE - OPERABILITY..................................
1-4 1.19.a OPERATING LIMITS REP 0RT.....................................
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 CONT RO L P R0GR AM......................................
1-5 l
1.24 eURGE - euPGING..................
1-5 1.25 QUADRANT POWER TILT RATI0....................................
1-5 1.26 RATED THERMAL P0WEP..........................................,
1-5 1.27 REACTOR TRIP SYSTEM RESPONSE TIME............................
1-5 1.28 REPORYABLE EVENT.............................................
1-5 BRAIDWOOD - UNITS 1 & 2 I
AMENDMENT NO. 35
I 0,EFINITIONS SECTION PAGE 1.29 SHUTDOWN MARGIN..............................................
1-5 1.30 SITE 80VNDARY................................................
1-6 1.31 SLAVE RELAY TEST.............................................
1-6 1.32 0ELETED......................................................
1-6 1.33 SOURCE CHECK.................................................
1-6 1.34 STAGGERED TEST BASIS.........................................
1-6 1.35 THERMAL. POWER.......
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 SY3 TEM.........................
1-7 1.40 VENTING......................................................
1-7 1.41 WASTE GAS HOLOUP SYSTEM......................................
1-7 TABLE 1.1 FREQUENCY N0TATION......................................
1-8 TABLE 1.2 OPERATIONAL M00ES................................
1-9 I
BRAIDWOOD - UNITS 1 & 2 II AMEN 0 MENT NO. 35 1
_ -. - _ - - ~ _
_m-..
- ~. -
LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE....................................
3/4 2-1 FIGURE 3.2-1 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL P0WER................
3/4 2-3 3/4.2.2 HEAT FLUX HOT CHANNEL FACT 0R.............................
3/4 2-4 FIGURE 3.2-2 K(Z)-NORMALIZED F (Z) AS A FUNCTION OF CORE HEIGHT...
3/4 2-5 9
3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACT0R.................................................
3/4 2-8 3/4.2.4 QUADRANT POWER TILT RATIO........
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 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMEhTATION......................
3/4 3-1 TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION...................
3/4 3-2 TABLE 3.3-2 (THIS TABLE IS NOT USED)..............................
3/4 3-7 TABLE 4.3 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 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION.....................................
1/4 3-15 TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION 1 RIP SETP0lNTS......................
3/4 3-23 TABLE 3.3-5 (THIS TABLE-IS NOT USED)..............................
3/4 3 TABLE 4,3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS...........
3/4 3-34 l
BRAIDWOOD - UNITS 1 & 2 V
LIMITING CONDITIONS,FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.3.3 MONITORING INSTRUMEN1ATION Rcdiation Monitoring for Plant Operations.......
3/4 3-39 TABLE 3.3-6 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERAT10NS....................
3/4 3-40 TABLE 4.3-3 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS SURVEllLANCE REQUIREMENTS..............................
3/4 3-42 i
Movable Incore Detectors.............................
3/4 3-43 Seismic Instrumentation.................................
3/4 3-44 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 REQUIREPINTS...........................
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 INSTRUMENTAT10N SURVEILLANCE REQUIREMENTS...........................
3/4 3-55 TABLE 3.3-11 (This table number is not used)..
3/4 3-57 TABLE 3.3-12 (This table number is not used)..................
3/4 3-57 Loose-Part Detection System.............................
3/4 3-58 Explosive Gas Monitoring Instrumentation............
3/4 3-60 BRAIDWOOD - UNITS 1 & 2 VI AMENDMENT NO. 35
LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE TABLE 3.3 13 EXPLOSIVE GAS MONITORING INSTRUMENTATION.............
3/4 3-61 TABLE 4.3-9 EXPLOSIVE GAS MONITORING IN MRUMENTATION SURVEILLANCE REQUIREMENTS.........................
3/4 3-62 High Energy Line Break Isolation Sencors................
3/4 3-63 TABLE 3.3-14 HIGH ENERGY LINE BREAK INSTRUMENTAT10N...............
3/4 3-64 3/4.3.4 TURBINE OVERSPEED PROTECTION.............................
3/4 3-65 3/4.4 REACTOR CO0lANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION Startup and Power Operation..............................
3/4 4-1 Hot Standby..............................................
3/4 4-2 Hot Shutdown.............................................
3/4 4-3 Cold Shutdown - Loops Fi11ed............................,
3/4 4-5 Cold Shutdown - Loops Not filled.........................
3/4 4-6 Loop Isolation Valves 0peration..........................
3/4 4-7 Loop Isolation V61ves-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 E
TABLE 4.4-1 HINIMUM 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 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE P-Leakage Detection Systems...........
3/4 4-20 Operational Leakage......................................
3/4 4-21
-TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION-VALVES......
3/4 4-23 3/4.4.7 CHEMISTRY................................................
3/4 4 i BRAIDWOOD UNITS 1 & 2 VII AMENDMENT NO. 35
~. _. _. _ _ _. _ _ _
LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.7.10 (This specificatirn number is not used.)..........
3/4 7-28 TADLE 3.7-5 (This table number is not used.)......................
3/4 7-29 3/4.7.11 (This specifiation number is not used.)..................
3/4 7-30 3/4.7,12 AREA TEMPERATURE H0NITORING..............................
3/4 7-31 TABLE 3.7-6 AREA TEMPERATURC H0NITORING...........................
J/4 7-32 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.6.1 A.C. SOURCES 0persting................................................
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 8-1.2 Shutdown.................................................
3/4 8-13 3/4.8.3 ONSITE POWER DISTRIBUTI0d 0perating................................................
3/4 8-14 Shutdown.,
3/4 8-16 3/4.8.4 ELECTRICAL tQUIPMENT PROTECTIVE DEVICES Containment Penetration Conductor Overcurrent Protective Devices,....................................
3/4 0-17 TABLE 3.8-la CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES (UNIT 1).............
3/4 8-19 TABLE 3.8-lb CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES (UNIT 2).............
3/4 8-29 Motor-Operated Valves Thermal Overload Protection Devices................................................
3/4 8-39 BRAIDWOOD - UNITS 1 & 2 XI
LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 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 PfNETRATIONS........................
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 Leve1.........................................
3/4 9-9 L ow Wa t e r L e v e 1..........................................
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 P00f 3/4 9-13 3/4.9.12 FUEL HANDLING B'JILDING EXHAUST FILTER PLENUMS............
3/4 9-14 i
1 l
t BRAIDWOOD - UNITS 1 & 2 XII AMENDMENT NO. 35
LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.10 SPECIAL TEST EXCEPTIONS 3
3/4.10.1 SHUTOOWN MARGIN..........................................
3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUf!ON LIMITS...
3/4 10-2 3/4.10.3 PHYSICS TESTS...............
3/4 10-3 3/4.10.4 REACTOR COOLANT L00PS....................................
3/4 10-4
-3/4.10.5 POSITION INDICATION SYSTEM - SHUTDOWN...........
3/4 10-5 3/4.11 RADIDACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS L i q'J i d H o l d up T a n k s.....................................
3/4 11-1 3/4.11.2 GAsiOUS EFFLUENTS Explosive Gas Mixture...................................
3/4 11-2 Gas Decay Tanks..........................................
3/4 11-3 I
i BRAIDWOOD - UNITS 1 & 2 XIII AMENDMENT NO. 35
I BASES SECTION PAGE 3/4.0 APPLICABILITY...............................................
B 3/4 0 1 3/4.1 REACTIVITY CONTROL SYSTEM' 3/4.1.1 B0 RATION CONTR0L...................................
B 3/4 1 1 3/4.1.2 B0 RATION SYSTEMS..........................................
B 3/4 1-2 3/4.1.3 MOVABLE CONTROL ASSEMBLIES...............................
B 3/4 1-3 3/4.2 POWER 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 l
FEATURES ACTUATION SYSTEM INSTRUMENTATI0H...............
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 3-6 3/4.4 REACTOR COOLANT SYSTEM s
3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION.............
B 3/4 4 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 XIV AMENDMENT NO. 35
d BASES SECTION PAGE 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.B 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-lb 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, 3/4.5.3 and 3/4.5.4 ECCS SUBSYSTEMS.......................
B 3/4 5-1 3/4.5.5 REFUELING WATER STORAGE TANK..............................
B 3/4 5-4 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 CONTROL..................................
B 3/4 6-4 BRAIDWOOD - UNITS 1 & 2 XV AMENDMENT N0. 35
BASES SECTION PAGE 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE.............................................
B 3/4 7-1 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION..........
B 3/4 7-3 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 SYSTEF...........................
B 3/4 7-4 3/4.7.7 NON-ACCES$1BLE 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 (This specification number is not used.).................
B 3/4 7-6 3/4.7.11 (This specification number is not used.)..................
B 3/4 7-6 3/4.7.12 AREA TEMPERATURE MONITORING....................
B 3/4 7-6 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1, 3/4.8.2 and 3/4.8.3 A.C. SOURCES, D.C. SOURCES, AND ONSITE POWER DISTRIBUTION...............................
B 3/4 8-1 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES...................
B 3/4 B-3 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION......................................
B 3/4 9-1 I
3/4.9.2 INSTRUMENTATION...........................................
B 3/4 9-1 3/4.9.3 DECAY'T1ME................................................
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 XVI AMENDMENT N0. 35
~1 BASES ECTION PAGE 3/4.9.6 REFUELING HACHINE.........................................
B 3/4 9-2 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 9-3 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..............
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 POSITION INDICATION SYSTEM - SHUT 00WN.....................
B 3/4 10-1 3/4.11 RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS Liquid Holdup Tanks..................................
B 3/4 11-1 3/4.11.2 GASEOUS EFFLUENTS Explosive Gas Mixture.................................
B 3/4 11-2 Gas Decay Tanks.......................................
B 3/4 11-2 BRAIDWOOD - UNITS 1 & 2 XVII AMENDMENT NO. 35
O DESIGN FEATURES SECTION PAGE I
5.1 SITE
)
5.1.1 EXCLUSION AREA..............................................
5-1 5.1. 2 LOW POPULATION Z0NE.........................................
5-1 5.1. 3 MAP DEFINING UNRESTRICTED AREAS AND SITE BOUNDARY FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS.,............
5-1 FIGURE 5.1-1 EXCLUSION AREA AND UNRESTRICTED AREA FOR RADIOACTIVE 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 ROD ASSEMBLIES......................................
5-4 5.4 REACTOR COOLANT SYSTEM 5.4.1 DF"IGN PRESSURE AND TEMPERATURE.............................
5-4 5.4.2 V0LUME......................................................
5-4 5.5 METEOROLOGICAL TOWER L0 CATION.................................
5-4 5.6, FUEL STORACE 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 LIMITS..................
5-6 BRAIDWOOD - UNITS 1 & 2 XVIII AMENDMENT NO. 35
ADMINISTRATIVE CONTROLS SECTION PAGE C
l 6.1 RESPONSIBILITY................................................
6-1 6.2 ORGANIZATION....................................
6-1 FIGURE 6.2-1 (THIS FIGURE NOT USED)...............................
6-3 FIGURE 6.2-2 (THIS FIGURE NOT USED)............
6-4 TABLE 6.2-1 MINIMUM SHIFT CREW COMPOSITION........................
6-5 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 Station 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 Authority...................................................
6-13
..s; co Records.....................................................
6-14 Procedures..................................................
6-14 Personne1...................................................
6-14 6.5 REPORTABLE EVENT ACTI0N.......................................
6-15 BRAIDWOOD - UNITS 1 & 2 XIX AMENDMENT NO. 35
-. _ ~ -
4 ADMINISTRATIVE CONTROLS SECTION PAGE
- 6. 7 SAFETY LIMIT VIOLATION........................................
6-15 6.8 PROCEDURES AND PR0 GRAMS.......................................
6-15 6.9 REPORTIN3 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..............
6-19 Monthly Operating Report.................
6-19 Operating Limits Report.....................................
6-19 Criticality Analysis of Byron and Braidwood Station Fuel Storage Racks................................
6-20 6.9.2 SPECIAL REP 0RTS.............................................
6-20 6.10 RECORD RETENTION.............................................
6-20 6.11 RADIATION PROTECTION PR0 GRAM.................................
6-21 6.12 HIGH RADIATION AREA..........................................
6-22 6.13 PROCESS CONTROL PROGRAM (PCP)......................
6-23 6.14 0FFSITE DOSE CALCULATION MANUAL (00CM).......................
6-23 BRAIDWOOD - UNITS 1 & 2 XX AMENDMENT NO. 35 s
v w
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-i--gy-WWep erm N
wi gy
-sw--w-Y=
g'y-'yv' y-9 gi'-'u y
DEFINITIONS 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 l
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 Pnown either not to interfere with the operation of Leakage Detection Systems or not to be PRESSURE BOUNDARY LEAVAGE, 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.
Thi3 category does not include employees of the licensee, its contractors or ve7 dors 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 0FFSITE DOSE CALCULATION MANUAL 1.18 The OFFSITE DOSE CALCULATION MANUAL (00CM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radio-active gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm / trip setpoints, and in the conduct of the Environ-mental Radiological Monitoring Program.
The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Sections 6.8.4.e and f, and (2) descriptions of the information that should be included in the Annuci Radiological Environmental Operating and Semiannual Radioactive Effluent Release Reports required by Specification 6.9.1.6 and 6.9.1.7.
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).
OPERATING LIMITS REPORT 1.19.a The OPERATING LIMITS REPORT 1s the unit-specific document that provides operating limits for the current operating reload cycle.
These cycle-specific operating limits shall be determired for each reload cycle in accordance with Specification 6.9.1.9.
Plant Operation within these operating limits is addressed in individual specifications.
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 fundamental 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 S0.59, or (3) otherwise approved by the Commission.
PRESSURE BOUNDARY LEAKAGE 1.22 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube leakage) through a nonisolable fault in a Reactor Coolant System component body, pipe wall, or vessel wall.
BRAIDWOOD UNITS 1 & 2 1-4 AMENDMENT NO. 35
DEFINITIONS PROCESS CONTROL PROGRAM 1.23 The PROCFSS CONTROL PROGRAM (PCP) shall contain the current furmulas, sampling, analyses, tesis, and determinations to be made to ensure that processing and packaging,f solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71, State regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste.
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 RAf!O 1.25 QUADRANT POWER 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 SH'TDOWN MARGIN shall be the instantaneous amount of reactivity by which 1.29 J
l the reactor is subcritical or would be subcritical from its present condition l
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.
BRAIDWOOD UNITS 1 & 2 1-5 AMENDMENT NO. 35
DEFINITIONS SI1E BOUNDAR.
1.30 The SITE 6 JDARY shall be that line beyond which the land is neither owned, nor leasec, 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.
SOLIDIFICATION 1.32 Deleted 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 comoonents 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 OPEP:ATIONAL 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 1.38 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 l
of individuals from exposure to radiation and radioactive materials or any L
areawithintheSITEBOUNDARYusedforresidentialquartersorforIndustrial, l
commercial, institutional, and/or recreational purposes.
l 1 -
BRAIDWOOD UNITS 1 & 2 1-6 AMENDMENT NO. 35
^
TABLE 3.3-11 (This table number is not used.)
Table 3.3 12 (This table number is not used.)
l l
I l
BRAIDWOOD - UNITS 1 & 2 3/4 3-57 AMENDMENT NO. 35
INSTRUMENTATION LOOSE-PART DETECTION SYSTEM LIMITING CONDITION FOR OPERATION 3.3.3.8 The Loose-Part Detection System shall be OPED.ABLE.
APPLICABILITY:
HOCES I and 2.
ACTION:
With one or more Loose-Part Detection System channels inoperable for a.
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 REQU7REMENTS 4.3.3.8 Each channel of the Loose-Part Detection Systems shall be demonstrated OPERABLE by performance of:
A CHANNEL CHECK at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, a.
b.
An ANALOG CHANNEL OPERATIONAL TEST except for verification of setpoint at least once per 31 days, and A CHANNEL CALIBRATION at leest once per 18 months, c.
i BRAIDWOOD - UNITS 1 & 2 3/4 3-58
~
1 INSTRUMENTATION RADI0 ACTIVE LIQUID F.FFLUENT MONITORING 1*;STRUMENTAT10N LIMITING CONDITION FOR OPERATION 3.3.3.9 (This specification number is not used.)
'q,;
AI,jE i
i BRAIDWOOD - UNITS 1 & 2 3/4 3-59 AMENDMENT NO. 35
INSTRUMENTATION EXPLOSIVE GAS 60NITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.10 The explosive gas monitoring instrumentation chronels shown in Table 3.3-13 shall be OPERABLE with their Alarm / Trip Setpoints set to ensure that the limits of Specification 3.11.2.5 are not exceeded.
APPLICABILITY:
As shown in Table 3.3-13 ACTION:
a.
With an explosive gas monitoring instrumentation channel Alarm / Trip Setpoint less conservative than required by the above specification, declare the channel inoperable and take the ACTION shown in Table 3.3-13.
b.
With less than the minimum number of explosive gas monitoring instru-mentation channels OPERABLE, take the ACTION shown in Table 3.3-13.
Restore the inoperable instrumentation to OPERABLE status within 30 days and, if unsuccessful, prepare and submit a Special Report to the Cornission within the next 30 days pursuant to Specification 6.9.2 to explain why this inoperability was not corrected in a timely manner.
c.
The provisions of Specifications 3.0.3, and 3.0.4 are not apnlicable.
SURVLlLLANCE REQUIREMENTS 4.3.3.10 Each explosive gas monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, CHANNEL CALIBRATION and DIGITAL CHANNEL OPERATIONAL TEST at the frequencies shown in Table 4.3-9.
I i
l
-BRAIDWOOD - UNITS 1 & 2 3/4 3-60 AMENDMENT NO. 35
TABLE 3.3-13 cn
- =R EXDLOSIVE' GAS MONITORING INSTRUMENTATION #
1 a
I b
MINIMUM CHANNELS y
INSTRUMENT OPERABLE APPLICABILITY ACTION E
1.
(Not Used)
Z 2.
(Not Used) 3.
Gaseous Waste Management Systen m
a.
Hydrogen Analyz'er (04T-GW8000) 1 38 b.
Oxygen Analyzer (0AT-GW8603) 1 38 c.
Waste Gas Compressor Discharge Oxygen Analyzer (OAIT-GW004) 1 38 M
t Y
TABLE NOTATIONS
- (?!ot used)
- During WASTE GAS HOLDUP SYSTEM operation.
- 0uring Waste Gas Compressor Operation.
- All instruments required for Unit 1 or Unit 2 operation.
ACTION STATEMENTS l
ACTION 38 - With the number of channels OPERABLE less than required by the Minimum Channels OPERA 3LE raquirement, operation of this systes may continue providad grab samples are taken and analyzed at least once per 4. hours during deg6ssing operations and 3t %st once per 24 h:::rs k
during other operations.
5
~
m
-4 5
i
o 9
+
x-.
u_
y y
d TABLE 4.3-9 EXPL. "IVE GAS MONITORING INSTRTiENTATION 'SURVEILLACE REQUIREMENTS 8
DIGITAL CHANNEL NDE5 F0ii WHICH CHANNEL SOURCE CHANNEL OPERATIONAL; SURVEILLANCE FUNCTIONAL UNIT CHECK CHECK CM IBRATION TEST-15 REQUIRED w
[
1.
(Not Used)-
[
2.
(Not Used) 3.
Gaseous Waste Management System a.
Hydrogen Analyzer (0AT-GW8000)
D N.A.
Q(4)
M b.
Oxygen Analyzer (0AT-GW8003)
D N.A.
Q(5)
M T
c.
Waste Gas Compressor Discharge Oxygen Analyzer (OAll-GWOO4)
D N.A.
Q(5)
M O
TABLE NOTATIONS
- (Not used)
- During WASTE GAS HOLDUP SYSmi operation.
- During Waste Gas Compresso
,eration.
- All instruments required for Unit 1 or Unit 2 operation.
(1) (Not Used)
(2) (Not Used) g (3) (Not-Used) y (4) The CHANNEL CALIBRATION shall include the use of standard gas samples containing hydrogen and nitrogen.
(5) Th'e CHANNEL CALIBRATIOk shall include the use of standard gas samples containing oxygen and nitrogen.
.E
INSTRUMENTATION HIGH ENERGY LINE BREAK ISOLATION SENSORS LIMITING CONDITION FOR OPEPATION 3.3.3.11 The high energy line break instrumentation shown in Table 3.3-14 i
shall be OPERABLE.
APPLICABILITY:
As shown in Table 3.3-14 ACTION:
a.
With the number of OPERABLE auxiliary steam isolation instruments less than the Minimum Channels OPERABLE as required by Table 3.3-14, restore the inoperable instrument (s) to OPERABLE status within 7 days, or suspend the supply of auxiliary steam to the Auxiliary Buildirig, or establish a continuous watch in the affected area (s) until the inoperable sensors are restored to OPERABLE status, b.
With the number of OPERABLE steam generator blowdown line isolation instru-i imum Channels OPERABLE as required by Table 3.3-14, ments less than the n
restore the inoperat,.e instrument (s) to OPERABLE status within 7 days, or limit the total steam generator blowdewn flow rate to less than or equal to 60 gpm or establish a continuous watch in the affected area (s) until the inoperable sensors are restored to 0FERABLE status.
c.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicabla.
S_yRVEILLANCE REQUIREMENTS 4.3.3.12 Each of the above high energy line break isolation instruments t, hall be demonstrated OPERABLE by the performance of an ANALOG CHANNEL-0PERATIONAL TEST and CHANNEL CALIBRATION at least once per 18 months.
BRAIDWOOD UNITS 1 & 2 3/4 3-63 AMENDMENT NO. 35
.. _ _. ~
TABLE 3.3-14 HIGH ENERGY LINE BREAK INSTRUMENTATION ISOLATION INSTRUMENT MINIMUM CHANNELS APPLICABLE FbNCTION
-CHANNEL OPERABLE
.>0 DES 1.
Auxiliary Steam OTS-AS031A 1
Isolation OTS-AS032A OTS-AS031B 1
OTS-AS032B OTS-AS031C 1
OTS-AS032C OTS-AS031D 1
OTS-AS032D OTS-A5031E 1
OTS-AS032E OTS-A5031F 1
OTS-AS032F 2.
Steam Generator TS-50045A 1
1,2,3,4 Blowdown Line TS-SD045B Isolation TS-SD046A 1
1,2,3,4 TS-SD046B TS-SD045C 1
1,2,3,4 TS-SD045D TS-50046C 1
1,2,3,4 TS-SD046D
- Required when auxiliary steam is being supplied, from any source, to the
-Auxiliary Building.
BRAIDWOOD UNITS 1 & 2 3/4 3-64 AMENDMENT NO. 35 r
=
INSTRUMENTATION 3/4.3.4 Tl'RBINE OVERSPEED PROTECTION LI'tITING CONDITION FOR OPERATION 3.3.4 At laast one Turbine Overspeed Protection System shall be OPERABLE.
APPLICABILIT/:
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 rehear intercept valve per low pressure turbine steam line inoperable, restore the inoperable valve (s) to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or clcse at least one valve in the affected steam line(s) or isolate the tLrbine from the steam supply within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, b.
With the above required Turbine Overspeed Protection System otherwise inoptrable, within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> isolate the turbine from the steam supply, SURVEILLANCE RE0VIREMENTS 4.3.4.1 P e 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 g:vernor valves, 3)
Six turbine reheat stop valves, and 4)
Six turbirie 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 observa-tion,'cf 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 0/erspeed Protection Systems, and At least once per 40 months by disassembling at least one of each of e.
the valves given in Specifications 4.3.4.2a. and b. above, and per-forming a e:sual and surface inspection of valve seats, disks and stems and verifying no una~ eptable flaws or corrosion.
BRAIDWOOD UNITS 1 & 2 3/4 3-65 AMENDMENT N0. 35
6 3/4.11 RADIDACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS LIQUID HOLOUP TANKS LIMITING CONDITION FOR OPERATION 3.11.1.1 Deleted 3.11.1.2 Deleted 3.11.1.3 Deleted I
3.11.1.4 The _ quantity of radioactive material, excluding tritium and dissolved or entrained noble gases, contained in any outside tanks shall be limited to the following:
a.
Primary Water Storage Tank 1 2000 Curies, and b.
Outside Temporary Tank
$ 10 Curies.
APPLICABILITY:
At all times.
ACTION:
a.
With the quantity of radioactive material in any of the above lis:ed tanks exceeding the ebove limit, immediately suspend all additions of radioactive material to the tank, within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> reduce the tank contents to within the limit, and describe the events leading to this condition in the next Semiannual Radioactive Effluent Release l
Report, pursuant to Specification 6.9.1.7.
l b.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
l l
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 11-1 AMENDMENT NO. 35
l a
RADI0 ACTIVE EFFLUENTS 3/4.11.2 GASEOUS EFFLUENTS EXPLOSIVE GAS MIXTURE LIMITING CONDITION FOR OPERATION 3.11.2.5 The concentration of oxygen in the WASTE GAS HOLOUP SYSTEM shall be limited to less than or equal to 2% by volume whenever the hydrogen concentration exceeds 4% by volume.
APPLICABILITY:
At all times.
ACTION:
a.
With the concentration of oxygen in the WASTE GAS HOLDUP SYSTEM greater than 2% by volume but less than or equal to 4% by volume; reduce the oxygen concentration to the above limits within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, 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
+ sual to 4% by volume, then take ACTION a. above.
c.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.2.5 The concentrations of hydrogen and oxygen in the WASTE GAS 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 oxygen monitors required OPERABLE by -Table 3.3-13 of Specification 3.3.3,10.
l l
BRAIDWOOD - UNITS 1 & 2 3/4 11-2 AMENDMENT NO. 35
~_..
RADI0 ACTIVE EFFLUENTS GAS DECAY TANKS LIMITING CONDITION FOR OPERt. TION 3.11.2.6 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:
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 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, 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 4.11.2.6 The quantity of radioactive material contained in each gas decay tank shall be determined to be within the above limit at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when radioactive materials are being added to the tank.
BRAIDWOOD - UNITS 1 & 2 3/4 11-3 AMENDMENT NO. 35
l INSTRUMENTATION BASES SE!SMIC 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 earthquaka criteria.
This iastrumentation is consistent with the recommendations of Regulatory Guide 1.12, " Instrumentation for Earthquake," April 1974.
3/4.3.3.4 METEGR0 LOGICAL 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 fa r initiating protective measures to protect the health and safety of the pLblic and is consistent with the recommendations of Regulatory Guide 1.23, "Ot. ite Meteorological Programs," February 1972.
3/4.3.3.5 REMOTE SHUTDOWN INSTRUMENTATION 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 cutside 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 an accident.
This capability is consiscent 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 (This specification number is not used.)
BRAIDWOOD - UNITS 1 & 2 B 3/4 3-5 l
9 INSTRUMENTATION BASES 3/4.3.3.8 LOOSE-PART DETECTION SYSTEM The OPERABILITY of the loose part detection system ensures that sufficient capability is available to detect loose metallic parts in the Reactor Coolant System and avoid or mitigate damage to Reactor Coolant System components.
The allowable out-of-service tines and Surveillance Pequirements 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.
3/4.3.3.9 DELETED 3/4.3.3.10 EXPLOSIVE GAS HONITORING INSTRUMENTATION The instrumentation includes provisions for monitoring the concentrations of potentially explosive gas mixtures in the WASTE GAS HOLDUP SYSTEM.
3/4.3.3.11 HIGH ENERGY LINE BREAK ISOLATION SENSORS The OPERABILITY of the high energy line break isolation sensors ensures that the capability is available to promptly detect and initiate protective action in the event of a line break.
This capability is required to prevent the potential for damage to safety-related systems and structures in the auxiliary building.
3/4.3.4 ~ TURBINE OVERSPEED PROTECTION This specification is provided to ensure that the turbine overspeed protection instrunentation 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 intervels are in accordance with the latest manufacturer's l '
Division, Westinghouse.
recommendations:
" Operation and Maintenance Memo 041," Steam Turbine l
l BRAIDWOOD - UNITS 1 & 2 8 3/4 3-6 AMENDMENT NO. 35
e INSTRUMENTATION BASES TURBINE OVERSPEED PROTECTION (Continued)
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 Lp 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.
t4anual stroking of the extraction steam non-return valves is possible under shutdown conditions by latching the turbine and applying the air pressure to the spring cylinder.
It is possible to hear and feel the disk contact the seat solidly.
This n nual stroking-was not provided for in the design but will be done within 7 oays 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
'he 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 mo'ement 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 witi, progressive :losure.
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 l;
an individual heater.
Heater stages 1, 2, 3, and 4 are arrang:.d 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 tuen out of service, isolated, and bypassed for maintenance.
Isolating the extraction steam to a single intermediate heater involves several complications.
BRAIDWOOD - UNITS 1 & 2 B 3/4 3-7 AMENDMENT NO. 35 1
k INSTRUMENTATION BASES TURBINE OVERSPEED PROTECTION (Continued)
The motor operated valves are too large for routine manual operation, do not hate bypasses to allow controlled warmup conditions, stroke quickiy (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 intermediate heater causes the next heater to assume the heating load, approximately doubling the steam demand ano 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 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.
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1 BRAIDWOOD - UNITS 1 & 2 B 3/4 3-8 AMENDMENT NO. 35
3/4.11 RADI0 ACTIVE EFFLUENTS BASES 3 & 11.1 LIQUID EFFLUENTS 3/4.11.1.1 DELETED 3/4.11.1.2 DELETED 3/4.11.1.3 DELETED 3/4.11.1.4 LIQUID HOLDUP TANKS The tanks listed in this specification include all those outdoor radwaste tanks thht 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.
BRAIDWOOD - UNITS 1 & 2 B 3/411-1 AMENDMENT NO. 35
'3/4.11 RADI0 ACTIVE CFFLUENTS BASES 3/4.11.2 GASEOUS EFFLUENTS 3/4.11.2.1 DELETED 3/4.11.2.2 UELETED 3/4.11.2.3 DELETED 3/4.11.2.4 DELETED 3/4.11.2.5 EXPL0SIVE 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 nxygen 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 dilutant.s 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 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.
Restricting the quantity of radioactivity contained in each gas storage tank provides assurance thtt in the event of an uncontrolled release of the tanks' contents, the resulting whole body exposure to a MEMBER OF THE PUBLIC at the nearest SITE B0UNDARY will not exceed 0.5 rem.
This is consistent with Standard Rev1ew 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.
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l BRAIDWOOD - UNITS 1 & 2 8 3/4 11-2 AMENDMENT NO. 35 l
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ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued) b.
In-Plant Radiation Monitorino 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 Ctemistry A program for monitoring of secondary water chemistry to inhibit steam generator tube degradation.
This program shall include:
1)
Identification of a salopling 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 l
6)
A procedure identifying: (a) the authority responsible for the l
interpretation of the data, and (b) the sequence and timing of l
administrative event'.equired 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 l
gaseous effluents, and containment atmosphere samples under accident conditions.
The program shall include the following:
i 1
1)
Training of personnel,-
2)
Proceriures for sampling and analysis, and 3)
Provisions for maintenance of sampling and analysis equipment.
BRAIDWOOD - UNITS 1 & 2 6-17
O ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued) e, Radioactive Effluent Controls Prooram A program shall be provided conforming with 10 CfR 50.36a for the control of radioactive effluents and for maintaining the doses to MEMBERS OF THE PUBLIC from radioactive effluents as low as ODCM, (2)y achievable.The program (1) shall be contained in the reasonabl shall be implemented by station procedures, and (3) shall include remedial actions to be taken whenever the program limits are
- exceeded, The program shall include the following elements:
1)
Limitations on the operability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM, 2)
Limitations on the concentrations of radioactive material released in liquid effluents to UNRESTRICTED AREAS conforming to 10 CFR Part 20, Appendix B, Table II, Column 2, 3)
Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.106 and with the methodology and parameters in tha ODCM, 4)
Limitations on the annual and quarterly doses or dose commitment to a MEMBER 0F THE PUBLIC from radioactive materials in liquid effluents released from each unit to UNRESTRICTED G AS conforming to Appendix I to 10 CFR Part 50, 5)
Determinationofcumulativeandprojecteddosecontributions from radioactive effluents for the current calendar quarter and current calencar year in accordance with the methodology and parameters in the ODCM at least every 31 dr 3 6)
Limitations on the operability and use of the liquid and gaseous effluent treatment systems to ensure that the appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a 31-day period would exceed 2 percent of the guidelines for the annual dose or dose commitment conforming to Appendix I to 10 CFR Part 50, 7)
Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the SITE B0UNDARY conforming to the doses associated with 10 CFd Part 20, Appendix B, Table II, Column 1, 8)
Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50, BRAIDWOOD - UNITS 1 & 2 6-17a AMENDMENT NO. 35
._m ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued) 9)
Limitations on the annual and quarterly doses to a MEMBER OF THE PUBLIC from Iodine-131, Iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released from each einit to areas beyond thn SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50, and 10)
Limitations on the annual dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to radiation from uraniua fuel cycle sources conforming to 40 CFR Part 190.
f.
Radiological Environmental Monitoring Program A program shall be provided to monitor the radiation and radionuclides in the environs of the plant.
The program shall provide (1) representative measurements of radioactivity in the highest potential exposure pathways, and (2) verification of the accuracy of the effluent monitoring program and modeling of environmental exposure pathways.
The program shall (1) be contained in the ODCM, (2) conform to the guidance of Appendix I to 10 CFR Part 50, and (3) include the following:
1)
Monitoring, sampling, analysis, and reporting of radiation and radionuclides in the environment in accordance with the methodology and parameters in the ODCM, 2)
A Land Use Census to ensure that changes in the use of areas at and bevond the SITE B0UNDARY are identified and that modifica-tions to the monitoring program are made if required by the results of this census, and 3)
Participation in a Interlaboratory Comparison Program to ensure that independent checks on the precision and accuracy of the measurements of radioactive materials in environmental sample matrices are performed as part of the quality assurance program for envi onmental monitoring.
BRAIDWOOD - UNITS 1 & 2 6-17b AMENDMENT NO. 35
~ _
ADMINISTRATIVE CONTROLS REPORTING REQUIREMENTS (Continued)
ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT
- 6.9.1.6 The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year.
The report shall include summaries, interpreta-tions, and analysis of trends of the results of the Radiological Environmental Monitoring Program for the reporting period.
The material provided shall be consistent with the objectives outlined in (1) the ODCM and (2) Sections IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR Part 50.
SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT **
6.9.1.7 The Semiannual Radioactive Effluent Release Report covering the o; m e
}
tion of the unit during the previous 6 months of operation shall be submitto I
within 60 days af ter January 1 and July 1 of each year.
The report shall clude a cummary of the quantities of radioactive liquid and gaseous effle a i
and solid waste released from the unit.
The material provided shall be (1) con-sistent with-the objectives outlined in the 00CM and PCP and (2) in conformance with 10 CFR 50.36a and Section IV.B.1 of Appendix I to 10 CFR Part 50.
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,,'ith a copy to the Regional Administrator of the NRC Regional Office, no iater than the 15th of each month following the calendar month covereri by the report.
OPERATING LIMITS REPORT 6.9.1.9 Operating limits shall be established and documented in the OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle.
The analytical methods used to determine the operating limits shall be those previously reviewed and approved by the NRC in Topical Reports:
- 1) WCAF 9272-P-A " Westinghouse Reload Safety Evaluations Methodology" dated July 1985,
- 2) WCAP-8385 " Power Distribution Control and Load Following Procedures" dated September 1974, 3) NFSR-0016 " Benchmark of PWR Nuclear Design Methods" dated July 1983, and/or 4) NFSR-0081 " Benchmark of PWR Nuclear Design Methods Using the PHOENIX-P and ANC Computer Codes" dated July 1990.
The operating limits shall be determined so that all applicable limits (e.g., fuel thermal mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.
The OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.
^A single submittal may be made for a multi-unit station.
- A single submittal may be made for a multi-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 6-19 AMENDMENT NO. 35
ADMINISTRATIVE CONTROLS CRITICALITY ANALYSIS OF BYRON AND BRAIDWOOD STATION FUEL STORAGE RACKS 6.9.1.10 Fuel enrichment limits for storage shall be established and documented in the CRITICALITY ANALYSIS OF BYRON AND BRAIDWOOD STATION FUEL STORAGE RACKS.
The analytical methods used to determine the maximum fuel enrichments shall be those previously reviewed and approved by the NRC in
" CRITICALITY ANALYS15 0F BYRON AND BRAIDWOOD STATION FUEL STORAGE RACKS." The fuel enrichment limits for storage shall be determined so that all applicable limits (e.g., subtriticality) of the safety analysis are met.
The CRITICALITY ANALYSIS OF BYRON AND BRAIDWOOD STATION FUEL STORAGE RACKS report shall be providad upon issuance of any changes, to the NRC Document Control Desk, with copies to the Regional Administrator and the
{
Resident Inspector.
SPECIAL REPORTS 6.9.2 Special reports shall he 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 g
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 r:flecting unit design modifications made to systems and equipment described in the Final Safoty Analysis Report; b.
Records of new and irradiated fuel inventory, fuel transfers and assembly burnup histories;
ADMINISTRATIVE CONTROLS RECORD RETENTION (Continued) 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 Revier 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; 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, and o.
Records of reviews performed for changes made to the 0FFSITE DOSE CALCULATION MANUAL and the PROCESS CONTROL PROGRAM.
6.11 RADIATION PROTECTION PROGRAM i
Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained and L
adhered to for all operations involving personnel radiation exposure.
BRAIDWOOD - UNITS 1 & 2 6-21 AMENDMENT NO. 35
-ADMINISTRATIVE CONTROLS 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 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 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 pro-vided 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 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 shall perform periodic radiation surveillance at the frequency specified in the Radiation Work Permit.
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 nenetrates 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 isnmediate 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.
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l BRAIDWOOD - UNITS 1 & 2 6-22 AMENDMENT NO. 35
ADMINISTRATIVE CONTROLS HIGH RADIATION AREA (Continued)
For individual high radiation areas accestible 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.
6.13 PROCESS CONTROL PROGPAM (FCP) 6.13.1 Changes to the PCP:
a.
Shall be documented and records of reviews performed shall be retained as required by Specification 6.10.20.
This documentation shi ~
contain:
1)
Sufficient infcrmation to support the change together with the appropriate analyses or evaluations justifying the change (s)
- and, 2)
A determination that the change will maintain the overal' conformance of the solidified waste product to existing requirements of Federal, State, or other applicable regulations.
b.
Shall become effective af ter review and acceptance by the Onsite Review and Investigative Function (Onsite Review) and the approval of the Station Manager.
6.14 0FFSITE DOSE CALCULATION MANUAL (00CM) 6.14.1 Changes to the OCCM:
a.
Shall be documented and records of reviews performed shall be retained as required by Specification 6.10.20.
This documentation shall contain:
1)
Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change (s)
- and, 2)
A determination that the change will maintain the level of radioactive effluent control required by 10 CFR 20.106, 40 CFR
'Part 190, 10 CFR 50.36a, and Appendix I to 10 CFR Part 50 and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations, b.
Shall become effective after review and acceptance by the Onsite Review and Investigative Function and the approval of the Station Manager on the date specified by the Onsite Review and Investigative Function.
BRAIDWOOD - UNITS 1 & 2 6-23 AMENDMENT N0. 35
e ADMINISTRATIVE CONTROLS 0FFSITE DOSE CALCULATION MANUAL (0DCM) (Continued) c.
Shall be submitted to the Commission in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Semiannual Radioactive Effluent Release Report for the period of the report in which any change to the ODCM was made eflective.
Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g., month / year) the change was implemented.
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BRAIDWOOD - UNITS 1 & 2 6-24 AMENDMENT NO. 35
'