Text

Amend 135 to License NPF-3,correcting Typos & Making Minor Word Changes to Achieve Consistency Between Tech Specs & Plant Nomenclature & Deleting Unnecessary Statements Due to Elapsed Time &/Or Completion of Specified Actions
	ML20245H946
Person / Time
Site:	Davis Besse
Issue date:	08/04/1989
From:	Hannon J; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20245H951	List: ML20245H946; ML20245H953;
References
	NUDOCS 8908170369
	Download: ML20245H946 (161)
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UNITED STATES

[

p, NUCLEAR REGULATORY COMMISSION O

(j WASHINGTON, D. C,20555

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TOLEDO EDISON COMPANY AND THE CLEVELAND ELECTRIC ILLUMINATING COMPANY DOCKET NO. 50-346 DAVIS-BESSE NUCLEAR POWER STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.135 License No. NPF-3

-l 1.

The Nuclear Regulatory Comission (the Comission) has found that:

1 A.

The applications for amendment by the Toledo Edison Company and j

The Cleveland Electric Illuminating Company (the licensees) dated November 2,1987 and January 5,1989 comply with the standaros and requirements of the Atomic Energy Act of 1954, as amended (the Act),

and the Comission's rules and regulations set forth in 10 CFR Chapter I;

(

1 B.

The facility will operate in' conformity with the applications, the provisions of the Act, and the rules and regulations of the Comission; 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 Comission's regulations; D.

The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.

The issuance of this amendment ir in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-3 is hereby amended to read as follows:

8908170369 890804 PDR ADOCK 05000346.

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(a) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.135, are hereby incorporated in the license. The Toledo Edison Company shall operate the facility in-accordance with the Technical Specifications.

i 3.

This license amendment is effective as of its date of issuance and shall-be implemented not later than September 18, 1989.

FOR THE NUCLEAR REGULATORY COMMISSION i

f.

p.=.

John N. Hannon, Director Project Directorate III-3 Division of Reactor Projects - III, IV, V,'& Special Projects Office of~ Nuclear Reactor Regulation l

Attachment:

Changes to the Technical Specifications Date of Issuance:

Au9ust 4, 1989 i

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I-ATTACHMENT TO LICENSE AMENDMENT NO.135 FACILITY OPERATING LICENSE NO. NPF-3 DOCKET NO. 50-346 Replace the following pages' of the Appendix "A" Technical Specifications with -

the attached pages. The revised.pages are identified by amendment number and contain vertical lines indicating the area of change. The corresponding overleaf' pages are also provided to maintain document completeness.

Remove Insert Remove.

Insert Remove.

Insert-IV IV 3/4 3-26

- 3/4 3-26 3/4 9-6 3/4 9-6 V*

V*

3/4 3-27 3/4 3-27 3/4 9-12 3/4 9-12 i

VI VI 3/4.3 3/4 3-30 3/4 10-3 3/4 10 VII VII 3/4 3-30a 3/4 3-30a 3/4 11-9 3/4 11-9 1

IX*

3/4 3-30b 3/4 3-30b 3/4 12-1 3/4 12-1 XII XII 3/4 3-30c 3/4 3-30c 3/4 12-2 3/4 12-2 XIII XIII 3/4 3-30d 3/4 3-30d B 3/4 0-1b B 3/4 0-1b XV XV

-3/4 3-32 3/4 3-32 B 3/4 3-la B 3/4 3-la XVI XVI 3/4 3-33 3/4 3-33 B 3/4 4-la B 3/4 4-la-1 1-1 1-1 3/4 3-45 3/4 3-45 B 3/4 4-2 B 3/4 4-2 1

1-2 1-2 3/4 3-49 3/4 3-49 B 3/4 6-2 B 3/4 6-2.-

1-6 1-6 3/4 3-54 3/4 3 B 3/4 6-3 B 3/4 6-3 3/4 0-1 3/4 0-1 3/4 3-55 3/4 3-55 B 3/4 6-5 B 3/4 6-5 3/4 1-7 3/4 1 3/4 3-56 3/4 3-56 B 3/4 7-1 B 3/4 7-1 3/4 1-10 3/4 1-10 3/4 3-58 3/4 3-58 8 3/4 7-3 B 3/4 7-3 3/4 1-20 3/4 1-20 3/4 3-59 3/4 3-59 B 3/4 7-6 B 3/4 7-6 3/4 1-37*

3/4 1-37*

3/4 3-63 3/4 3-63 B 3/4 9-2 B 3/4 9-2.

3/4 2-8 3/4 2-8 3/4 3-64 3/4 3-64 B 3/4 9-3*

B 3/4 9-3*'

3/4 2-10 3/4 2-10 3/4 3-65 3/4 3-65 B 3/4 11-2 B 3/4 11-2 3/4 2-14*

3/4 2-14*

3/4 4-1

- 3/4 4-1 5-4 5-4 3/4 3-2 3/4 3-2 3/4 4-2a.

3/4 4-2a 6-1 6-1 3/4 3-4 3/4 3-4 3/4 4-3 3/4 4-3 6-la*

.6-la*

3/4 3-5 3/4 3-5 3/4 4-4 3/4 4-4 6-2*

6-2*

3/4 3-Sa 3/4 3-5a 3/4 4-15 3/4 4-15 6-3*

6-3*

i 3/4 3-7 3/4 3-7 3/4 4-16 3/4 4-16 6-5 6-5 3/4 3-8 3/4 3-8 3/4 4-23 3/4 4-23 6-15 6-15 3/4 3-10 3/4 3-10 3/4 4-32 3/4 4-32 6-16 6-16 3/4 3-11 3/4 3-11 3/4 5-4 3/4 5-4 6-18 6-18 3/4 3-12 3/4 3-12 3/4 6-10 3/4 6-10 3/4 3-12a 3/4 3-12a 3/4 6-22 3/4 6-22 6

3/4 3-15*

_3/4 6-29 3/4 6-29 3/4 3-17*

3/4 3-17*

3/4 6-30 3/4 6-30 i

3/4 3-19*

3/4 7-13 3/4 7-13 3/4 3-21 3/4 3 3/4 7-17 3/4 7-17

.3/4 3-22 3/4 3-22 3/4 7-18 3/4 7-18 l

3/4 3-24 3/4 3-24 3/4 7 3/4 7-38 3/4 3 3/4 3-25 3/4 7-39 3/4 7-39 3/4 7-44 3/4 7-44

Typographical errors corrected from previously-issued amendments (omission.

of amendment number from pages, etc.).

These pages were not part of the applications for this amendment.

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l INDEX I

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.2 POVER DISTRIBUTION LIMITS 3/4.2.1 AXIAL POWER IMBALANCE-...

3/4 2-1 3/4.2.2 NUCLEAR HEAT FLUX HOT CHANNEL FACTOR - F0.......

3/4 2-5 3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR - Fg.....

3/4 2-7 3/4.2.4 QUADRANT POWER TILT..................

3/4 2-9 l

3/4.2.5 DNB PARAMETERS.....................

3/4 2-13 l

l 3/4.3 INSTRUMENTATION l

3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION.......

3/4 3-1 I

3/4.3.2 SAFETY SYSTEM INSTRUMENTATION Safety Features Actuation System.

3/4 3-9 Steam and Feedwater Rupture Control System......

3/4 3-23 Anticipatory Reactor Trip System.

3/4 3-30a 3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation..........

3/4 3-31 Incore Detectors....................

3/4 3-35 Seismic Instrumentation....

3/4 3-37 Meteorological Instrumentation.

3/4 3-40 Remote Shutdown Instrumentation.

3/4 3-43 Post-Accident Instrumentation.

3/4 3-46 i

Fire Detection Instrumentation.

3/4 3-52 i

Radioactive Liquid Effluent Moritoring Instrumentation.

3/4 3-57 Radioactive Gaseous Effluent Monitoring Instrumentation. 3/4 3-62 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 COOLANT LOOPS AND COOLANT CIRCULATION Startup and Power Operation..............

3/4 4-1 Shutdown and Hot Standby................

3/4 4-2 3/4.4.2 SAFETY VALVES - SHUTDOVN.

3/4 4-3

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1 3/4.4.3 SAFETY VALVES AND PILOT OPERATED RELIEF VALVE -

1 OPERATING......................

3/4 4-4 I

DAVIS-BESSE, UNIT 1 IV Amendment No. 28, AV,/>A4,135 I

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INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE l

3 /4. 0 APPL I CAB I LITY '.......................................... 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 1

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3/4.1.1 BORATION CONTROL i

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Shu tdown Ma rg in...................................... 3/4 1 -1 Bo ro n D i l ut i on.............'................. -.......

3/4 1-3 Modera tor Temperature Coef ficient....................

3/4 i-4 Mininum Tempera ture for Cri tical ity.................. 3/4 1-5 3/4.1.2 BORATION SYSTEMS Fl ow P a ths - Shutd own................................ 3/4 1-6 Fl o w P a't hs - Op e r a t i n g...............................

3/4 1-7 Ma k e u p Pu mp - Shu td own............................'.. '.

3/4 1-9 Ma k eup Pumps - Op era ti ng.............................

3/4 1-10 i

Decay Heat Removal Pump - Shutdown...................

3/4 1-11 i

Bo ri c Ac i d Pump - Shutdown...........................

3/4 1-12' Beric Acid Pumps - Operating.........................

3/4 1-13 Bera ted Wa ter Sources - Shutdown.....................

3/4 1-14 Bora ted Wa ter Sources - Opera ti ng.............,.......

3/4 1-17 3/4.1.3 MOVASLE CONTROL ASSEF2 LIES Group Height - Safety and Regulating Rod Groups......

3/4 1-19 Group Height - Axial Power Shaping Rod Group.........

3/4.1-21 Po s i tion Indica tor Cha nnel s........................... 3/4i,-22 Ro d Dr o p T i me.............................'..........

3/4 1-24 Safety Rod Insertion Limit...........................-

3/4 1-25 Regul a ting Red Ins ertion Limits......................

3/4 1-26 Ro d P ro gram..........................................

3/4.1-30 Xenon, Reactivity.....................................

3/4 1-33 Axi al Powe r Sha pi ng Red Ins e rti on Li =f ts.............".' 3/4 1-34 DAVIS-BESSE, UNIT 1 III Amendment No. 38 O

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e INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.4.4 PRESSURIZER..............................................

3/4 4-5 3/4.4.5 STEAM GENERATORS.........................................

3/4 4-6 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems................................

3/4 4-13 Operational Leakage......................................

3/4 4-15 3/4.4.7 CH EM I ST RY...............................................

3/ 4 4 - 17 3/4.4.8 SPECI FI C ACTIV ITY.......................................

3/4 4-20 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Reactor Cool ant System...................................

3/4 4-24 P re s s u ri z e r..............................................

3/ 4 4-2 9 3/4.4.10 STRUCTU RAL I NTEGRITY.....................................

3/4.4-30 3/4.4.11 REACTOR COOLANT SYSTEM VENTS.............................

3/4 4-32 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4. 5.1 CORE FLOO DI NG T AN KS......................................

3/ 4 5-1 3/4.5.2 ECCS SUBSYSTEMS - T,yg > 280*F...........................

3/4 5-3 3/4.5.3 ECCS SUBSYST EMS - T,yg < 280'F...........................

3/4 5-6 3/4.5.4 BO RATED WATER STO RAGE TAN K...............................

3/ 4 5-7

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DAVIS-BESSE, UNIT 1 V

Amendment No.135

8

,1 INDEX LIMITING CONDITIONS POR OPERATION AND SURVEILLANCE REQUIREMENTS 4

SECTION PAGE 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Containment Integrity...............

3/4 6-1 Containment Leakage................

3/4 6-2 Containment Air Locks.......

3/4 6-6 i

Internal Pressure........

3/4 6-7 Air Temperature........

3/4 6-8 Containment Vessel Structural Integrity......

3/4 6-9 Containment Ventilation System 3/4 6-10 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS Containment Spray System 3/4 6-11 Containment Cooling System 3/4 6-13 3/4.6.3 CONTAINMENT ISOLATION VALVES 3/4 6-14 3/4.6.4 COMBUSTIBLE GAS CONTROL Hydrogen Analyzers 3/4 6-23 Deleted......................

3/4 6-24 Containment Hydrogen Dilution System 3/4 6-25 Hydrogen Purge System.

3/4 6-26 3/4.6.5 SHIELD BUILDING l

l Emergency Ventilation System 3/4 6-28 Shield Building Integrity..

3/4~6-31 l

Shield Building Structural Integrity 3/4 6-32 DAVIS-BESSE, UNIT 1 VI Amendment No. AE, 44.135 L

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INDEX LIMITING CONDITIONS FOR OPERATION *AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE Safety Valves...................

3/4 7-1

-Auxiliary Feedvater System 3/4 7-4 condensate Storage Tank..............

3/4 7-6 Activity 3/4 7-7 Main Steam Line Isolation Valves 3/4 7-9 l

Motor Driven Feedvater Pump System.

3/4 7-12a 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION..

3/4 7-13 3/4.7.3 COMPONENT COOLING VATER SYSTEM 3/4 7-14 3/4.7.4 SERVICE VATER SYSTEM 3/4 7-15 3/4.7.5 ULTIMATE HEAT SINK 3/4 7-16 3/4.7.6 CONTROL ROOM EMERGENCY VENTILATION SYSTEM.

3/4 7-17 3/4.7.7 SNUBBERS 3/4 7-20 l

3/4.7.8 SEALED SOURCE CONTAMINATION..

3/4 7-36 3/4.7.9 FIRE SUPPRESSION SYSTEMS Fire Suppression Vater System...........

3/4 7-38 Spray and/or Sprinkler System...........

3/4 7-42 Fire Hose Stations 3/4 7-44 3/4.7.10 FIRE BARRIERS.................

3/4 7-47 3/4.8 ELECTRICAL POVER SYSTEMS 3/4.8.1 A.C. SOURCES Operating.....................

3/4 8-1 Shutdown 3/4 8-5 J

3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEMS l

A.C. Distribution - Operating...........

3/4 8-6 I

A.C. Distribution - Shutdovn 3/4 8-7 D.C. Distribution - Operating...........

3/4 8-8 D.C. Distribution - Shutdown 3/4 6-11 l

l DAVIS-BESSE, UNIT 1 VII Amendment No. 38, AQS,A06,135 j

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i INDEX j

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LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.9 REFUELING OPERATIONS l

3/4.9.1 BORON CONCENTRATION......................................

3/4 9-1 i

3/4.9.2 INSTRUMENTATION..........................................

3/4 9-2 3/4.9.3 DEC..Y TIME...............................................

3/4 9-3

{

3/4.9.4 CONTAINMENT PEN ETRATIONS.................................

3/4 9-4 3/4.9.5 COMMUN I C AT I ONS...........................................

3/ 4 9-5 3/4.9.6 FUEL HANDLING BRIDGE OPERABILITY.........................

3/4 9-6 3/4.9.7 CRANE TRAVEL - FUEL HANDLING BUILDING....................

3/4 9-7 3/4.9.8 DECAY HEAT REMOVAL AND COOLANT CIRCULATION All Water Leve1s.........................................

3/4 9-8 Low Water Leve1........................................... 3/4 9-8a l

3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM...........

3/4 9-9 1

1 3/4.9.10 WATER LEVEL - REACTOR VESSEL.............................

3/4 9-10 3/4.9.11 STO RAGE P0OL WAT ER L EVEL.................................

3/ 4 9-11 l

3/4.9.12 STORAGE P0OL VENTILATION.................................

3/4 9-12 I

3/4.9.13 SPENT FUEL POOL FUEL ASSEMBLY ST0 RAGE....................

3/4 9-13 l

3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTI ON LIMITS....................................

3/4 10-1 3/4.10.2 PHYSICS TESTS............................................

3/4 10-2 3/4.10.3 REACTOR COOLANT L00PS....................................

3/4 10-3 3/4.10.4 SHUTDOWN MARGIN..........................................

3/4 10-4 1

DAVIS-BESSE, UNIT 1 VIII Amendment No. JE, 130

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i INDEX L.

l BASES SECTION PAGE 3/4.0 APPLICABILITY.............................................. B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 B0 RATION CONTR0L........................................ B 3/4 1-1 3/4.1.2 B0 RATION SYSTEMS........................................ B 3/4 1-2 3/4.1.3 MOVABLE CONTROL ASS EMBLIES.............................. B 3/4 1-3 3/4.2 POWER DISTRIBUTION LIMITS.................................. B 3/4 2-1 3/4.3 INSTRUMENTATION 3/4.3.1 and 3/4.3.2 REACTOR PROTECTION SYSTEM AND SAFETY SYSTEMS INSTRUMENTATION................................. B 3/4 3-1 l

3/4.3.3 MONITORING INSTRUMENTATION.............................. B 3/4 3-2 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT L00PS................................... B 3/4 4-1 3/4.4.2 and 3/4.4.3 SAFETY VALVES................................ B 3/4 4-1 3

3/4.4.4 PRESSURIZER............................................. B 3/4 4-2 3/4.4.5 STEAM GENERATORS........................................ B 3/4 4-2 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 4-6 3/4.4.10 STRUCTU RAL INTEGRITY.................................... B 3/4 4-13 3/4.4.11 HIGH POINT VENTS........................................ B 3/4 4-13 l

DAVIS-BESSE, UNIT 1 IX Amendment No.135 5

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_a INDEX.

BASES i

SECTION PAGE 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 CORE FLOODING TANKS..................................

B 3/4 5-1

~

3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS..........................

B 3/4~5-1 3/4.5.4 BORATED WATER STORAGE TANK..........................

B 3/4 5-2.

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i DAVIS-BESSE, UNIT 1 X

Amendment No. 38

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$l INDEX

-l l-BASES SECTION PAGE 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 6-2 3/4.6.3 CONTAINMENT ISOLATION VALVES............

B 3/4 6-3 1

3/4.6.4 c0MBUSTIBLE GAS CONTROL....

B 3/4 6-4

)

3/4.6.5 SHIELD BUILDING..

B 3/4 6-4 i

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DAVIS-BESSE, UNIT 1 II I

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INDEX 1

1 BASES

{

l SECTION PAGE j

3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE...................

B 3/4 7-1 i

3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION.

B 3/4 7-4 l

3/4.7.3 COMPONENT COOLING VATER SYSTEM...........

.B 3/4 7-4 3/4.7.4 SERVICE VATER SYSTEM...........

B 3/4 7-4 3/4.7.5 ULTIMATE HEAT SINK.......

B 3/4 7-4 3/4.7.6 CONTROL ROOM EMERGENCY VENTILATION SYSTEM...

B 3/4 7-4 3/4.7.7 SNUBBERS..........

B 3/4 7-5

-l 3/4.7.8 SEALED SOURCE CONTAMINATION............

B 3/4 7-6 3/4.7.9 FIRE SUPPRESSION SYSTEMS..............

B 3/4 7-6 3/4.7.10 FIRE BARRIERS.........

B 3/4 7-6 3/4.8 ELECTRICAL POVER SYSTEMS....

B 3/4 8-1 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 PENETRATIONS.

B 3/4 9-1 3/4.9.5 COMMUNICATIONS.........

B 3/4 9-1

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DAVIS-BESSE, UNIT 1 XII Amendment No./Jg,IG4,135 U

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INDEX I!

BASES SECTION PAGE 3/4.9.6 FUEL HANDLING BRIDGE OPERABILITY.......................... B 3/4 9 3/4.9.7 CRANE TRAVEL - FUEL HANDLING BUILDING..................... B 3/4 9-2 3/4.9.8 COOLANT CIRCULATION....................................... B 3/4 9-2 3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM............ B 3/4 9-2 3/4.9.10 and 3/4.9.11 WATER LEVEL - REACTOR VESSEL AND i

STORAGE P00L'............................................ B 3/4 9 3/4.9.12 STORAGE P0OL VENTILATION;................................. B 3/4 9-3 3/4.9.13 SPENT FUEL POOL FUEL ASSEMBLY ST0 RAGE..................... B 3/4 9-3 1

3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS..... B 3/4 10-1 3/4.10.2 PHYS I CS TESTS............................................. B 3/4 10- 1 3/4.10.3 REACTOR COOLANT L00PS..................................... B 3/4 10-1 3/4.10.4 SHUTDOWN MARGIN........................................-... B 3/4 10 3/4.11 RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS.......................................... B 3/4 11-1 I

3/4.11.2 GASE0US EFFLUENTS......................................... B 3/4 11-3 3/4.11.3 SOLID RADI0 ACTIVE WASTE................................... B 3/4 11-6 3/4.11.4 TOTAL D0 S E................................................ B 3/ 4 1 1 - 7 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MON I TO R I N G P R0 GR AM........................................ B 3 / 4 12-1 i

1 3/4.12.2 LAND USE CENSUS........................................... B 3/4 12 l 3/4.12.3 INTERLABORATORY COMPARISON PR0 GRAM........................ B 3/4 12-1 l

i DAVIS-BESSE, UNIT 1 XIII Amendment No. E6,JJD,135

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INDEX DESIGN FEATURES SECTION PAGE.

5.1 SITE E x c i n i o n A rea.................................................

5 - 1 Low Population Zone............................................._5 5. 2 CONTAINMENT Configuration..................................................

5-1 Design Pressu re and Temperature................................

5-4 5.3 REACTOR CORE _

F u e l As s emb l i e s................................................

5-4 1 C o n t ro l Rod s..................................^................

5 5.4 REACTOR COOLANT SYSTEM Design Pressure and Temperature................................

5-4 Vo1ume..........................................................

5-5 5.5 METEOROLOGY CAL TOWER L0C ATION..................................

5-5 5.6 FUEL STORAGE C ri t i c a l i ty..................................................... 5-5 D ra i n a g e........................................................ 5.

C a p a c i ty.......................................................

5-6 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT............................

5-6, DAVIS-BESSE, UNIT 1 XIV Amendment No. 38

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. INDEX ADMINISTRATIVE CONTROLS J

SECTION PAGE 6.l' RESPONSIBILITY................................................

6-1

6. 2 ORGANIZATION L

Offsi te and Ons i te Organi zati ons.........................

6-1 l

Fa c i l i ty S ta f f........................................... 6-l a Facility Staff Overtime...................................-6-4a.

I 6.3 ' FACILITY STAFF QUALIFICATIONS.................................

6-5

6. 4 T RA I N I N G......................................................

6-5

6. 5 REVIEW AND AUDIT 6.5.1 STATION REVIEW BOARD Function.................................................

6-5 C om p o s i ti o n..............................................

6-6 Al t e rn a t e s...............................................

6-6 Meeting Frequency........................................

6-6 Quorum...................................................

6-6 Responsibilities.........................................

6-6 l

Au t h o ri ty................................................

6-8 l

Re c o rd s...................................................

6-8 6.5.2 COMPANY NUCLEAR REVIEW BOARD Function..................................................

6 C ompo s i ti on............................................... 6-9 Al t e rn a t e s...............................................

6-9 Co n s u l t a n ts..............................................

6-9 l

DAVIS-BESSE, UNIT 1 XV Amendment No. 38.135 l

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INDEX ADMINISTRATIVE CONTROLS SECTION PAGE Mee ti ng F req u e n cy........................................

6-9 Quorum...................................................

6-9 Review...................................................

6-10 Au d i t s...................................................

6-1 1 Au tho r i ty................................................

6-12 I

Records..................................................

6-12 6.5.3 Technical Revi ew and Contro1.............................

6-12 I

6.6 REPORTABL E EVENT ACT I0N.......................................

6-12a l

6.7 SAFETY LIMIT VIOLATION........................................

6-13

6. 8 PROCEDURES AND PR0 GRAMS.......................................

6-13 l

6.9 REPORTING REQUIREMENTS 6.9.1 Routine Reports..........................................

6-14a l

6.9.2 Special Reports..........................................

6-18 6.10 RECORD RETENTION.............................................

6-18a

,6.11 RADIATION PROTECTION PR0 GRAM.................................

6-20 i

e I

6.12 H I GH RAD I AT I ON ARE A..........................................

6-20 t

6.13 ENVIRONMENTAL QUALIFICATION..................................

6-21 6.14 PROCESS CONTROL PROGRAM (PCP )................................

6 6.15 0FFSITE DOSE CALCULATION MANUAL (0DCM).......................

6-22 6.16 MAJOR CHANGES TO RADI0 ACTIVE LIQUID, GASEOUS AND SOL I D WAST E T REATMENT SYSTEMS..........................

6-23 DAVIS-BESSE. UNIT 1 XVI Amendment No. 38,105 l

l

1 1.0 DEFINITIONS DEFINED TERMS 1.1 The DEFINED TERMS of this section appear in capitalized type and are applicable throughout these Technical Specifications..

THERMAL POVER l

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

RATED THERMAL POVER 1.3 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 2772 MVt.

OPERATIONAL MODE 1.4 An OPERATIONAL MODE shall correspond to any one inclusive combination of core reactivity condition, power level and average reactor coolant temperature specified in Table 1.1.

ACTION 1.5 ACTION shall be those additional requirements specified as corollary statements to each principal specification and shall be part of the l

specifications.

OPERABLE - OPERABILITY 1.6 A system, subsystem, train, component or device shall be OPERABLE or have f

OPERABILITY vhen it is capable of performing its specified function (s). Implicit i

in this definition shall be the assumption that.all necessary attendant instrumentation, controls, normal and emergency electrical power sources, cooling or seal vater, 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).

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DAVIS-BESSE, UNIT 1 1-1 Amendment No. $f.135 l

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DEFINITIONS REPORTABLE EVENT 50 73 o OCR a 0.

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CONTAINMENT INTEGRITY 1

1.8 CONTAINMENT INTEGRITY shall exist when:

a.

All penetrations required to be closed during accident conditions are either:

1.

Capable of being closed by the Safety Features Actuation System, or 2.

Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except as provided in Table 3.6-2 of Specification 3.6.3.1.

b.

All equipment hatches are closed and sealed, c.

Each airlock is OPERABLE pursuant to Specification 3.6.1.3, d.

The containment leakage rates are within the limits of Specification 3.6.1.2, and e.

The sealing mechanism associated with each penetration (e.g., velds, bellows or 0-rings) is OPERABLE.

CHANNEL CALIBRATION 1.9 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds with necessary range and accuracy to known values of the parameter which the channel monitors. The CHANNEL CALIBRATION shall encompass the entire channel-including the sensor and alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST. CHANNEL CALIBRATION may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is calibrated.

CHANNEL CHECK 1.10 A CHANNEL CHECK shall be the qualitative assessment cf channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.

DAVIS-BESSE, UNIT 1 1-2 Amendment No./97.135

____--__.__------__------.___--_--__-----_e_.-_--_

(

DEFINITIONS per disintegration (in MeV) for isotopes, other than iodines, with half lives greater than 15 minutes, making up at least 95% of the total non-1 iodine activity in the coolant.

STAGGERED TEST BASIS 1.21 A STAGGERED TEST BASIS shall consist of:

1 a.

A est schedule for n systems, subsystems, trains or designated conponents obtained by dividing the specified test interval.

int) n equal subintervals, i

b.

The testing of one system, subsystem, train or designated l

components at the beginning of each subinterval.

FREQUENCY NOTATION 1.22 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2.

AXIAL POWER IMBALANCE 1.23 AXIAL POWER IMBALANCE shall be the THERMAL POWER in the top half of the core expressed as a percentage of RATED THERMAL POWER minus the THERMAL POWER in the bottom half of the core expressed as a percentage of RATED THERMAL POWER.

SHIELD BUILDING INTEGRITY 1.24 SHIELD BUILDING INTEGRITY shall exist when:

a.

The airtight doors and the blowout panels listed in Table 4.6-1 are closed except the airtight doors may be used for normal transit entry and exit.

b.

The emergency ventilation system is OPERABLE.

c.

The sealing mechanism associated with each penetration (e.g.,

welds, bellows or 0-rings) is OPERABLE.

REACTOR PROTECTION SYSTEM RESPONSE TIME 1.25 The REACTOR PROTECTION SYSTEM RESP 0NSE TIME shall be that time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until power interruption at the control rod drive breakers.

DAVIS-BESSE, UNIT 1 1-5

l DEFINITIONS I

SAFETY FEATURE RESPONSE TIME 1.26 The SAFETY FEATURE RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its SFAS actuation setpoint at the channel sensor until the safety features equipment is capable of performing its safety l

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.

i PHYSICS TESTS 1.27 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related-instrumentation and

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.

STEAM AND FEEDWATER RUPTURE CONTROL SYSTEM RESPONSE TIME 1.28 The STEAM AND FEEDWATER RUPTURE CONTROL SYSTEM RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its SFRCS actuation setpoint at the channel sensor until the equipment is capable of performing its safety function (i.e., the valves travel to their required positior.s. pump discharge pressures reach their required values, etc.).

1

' I DAVIS-BESSE, UNIT 1 1-6 Amendment No.135

_ - _ _ _ __ _ __ _ _ ___ _ _ _ _ J

R!l 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMEIUS l

3/4.0 APPLICABILITY LIMITING CONDITION FOR OPERATION l

3.0.1 Limiting Conditions for Operation and ACTION requirements shall be applicable during the OPERATIONAL MODES or other conditions specified for each specification.

3.0.2 Adherence to the requirements of the Limiting Condition for Operation and/or associated ACTION within the specified time interval shall constitute l

compliance with the specification. In the event the Limiting Condition for Operation is restored prior to expiration of the specified time interval, completion of the ACTION statement is_not required.

1 1

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

1.

At least HOT STANDBY vithin 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , 2.

At least HOT SHUTD0VN within the following G hours, and 3.

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

i Vhere corrective measures are completed that permit operation under the ACTION l

requirements, the ACTION may be taken in accordance with the specified time limits as measured from the time of failure to meet the Limiting Condition for Operation. Exceptions to these requirements are stated in the individual Specifications.

3.0.4 Entry into an OPERATIONAL MODE or other specified applicability condition shall not be made unless the conditions of the Limiting Condition for Operation are met without reliance on provisions contained in the ACTION statements unless otherwise excepted. This provision shall not prevent passage through OPERATIONAL MODES as required to comply with ACTION statements.

3.0.5 When a system, subsystem, train, component or device is determined to be l

inoperable solely because its emergency power source is inoperable, or solely because its normal power source is inoperable, it may be considered OPERABLE for the purpose of satisfying the requirements of its applicable Limiting Condition for Operation, provided: (1) its corresponding normal or emergency power source is OPERABLE; and (2) all of its redundant system (s), subsystem (s), train (s),

component (s) and device (s) are OPERABLI, or likewise satisfy the requirements of this specification. Unless both conditions (1) and (2) are satisfied, within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months action shall be initiated to place the unit in a MODE in which the applicable Limiting Condition for Operation does not apply by placing it as applicable in:

1.

At least HOT STANDBY vithin 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , 2.

At least HOT SHUTD0VN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and 3.

At least COLD SHUTD0VN vithin the subsequent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

This Specification is not applicable in MODES 5 or 6.

DAVIS-BESSE, UNIT 1 I

3/4 0-1 Amendment No. 7/V, 135

^'

'APPLICABfLRTY SURVEILLANCE REQUIREMENTS.

l 4.0.1 Surveillance Requirements shall be applicable during the OPERA-TIONAL MODES or other conditions specified-for individual Limiting Conditions for Operation unless othenvise stated in an individual Sur-veillance Requirement.

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4.0.2 Each Surveillance Requirement shall be perfomed within the d

specified time interval with:

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a.

A maximum allowable extension not to exceed 25t of the surveil-lance interval, and b.

A total maximum combined interval time for any 3 consecutive tests not to exceed 3.25 times the specified surveillance interval.

1 4.0.3.

Failure to perform a Surveillance Requirement within the specified time' interval shall constitute a failure to meet the OPERABILITY requirements

' for a Limiting Condition for Operation.

Exception to these requirements are stated in the individual Specifications. -Surveillance Requirements do not have to be perfomed on inoperable equipment.

4.0.4 Entry into an OPERATIONAL MODE or other specified applicability -

conditiert shall not be made unless the Surveillance Requirement (s) i associated with the Limiting Condition for Operation have been performed within the stated surveillance interval or as.otherwise specified.

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

a.

During the time period:

1.

From issuance of the Facility Operating License to the start of facility conrnercial operation, inservice testing j

of ASME Code Class 1, 2 and 3 pumps and valves shall be i

performed in accordance with Section XI of the ASME Boiler and Pressure Vesel Code 1974 Edition, and Addenda through Summer 1975, except where specific written relief has been granted by the Commission.

1 2.

Following start of facility commercial operation, inservice inspection of ASME Code Class 1, 2 and 3 components and inservice testing of ASME Code Class 1, 2 and 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g),

except where specific written relief hes been granted by the Comission pursuant to 10 CFR 50, Section l

50.55a(g)(6)(i).

b.

Surveillance intervals specified in Section XI of the ASME i

Boiler and Pressure Vessel Code and applicable. Addenda for the I

inservice inspection and testing activities required by the j

ASME Boiler andaPressure Yessel Code and. ~ applicable Addenda-shall be applicible as follows in these Technical Specifications:

DAVIS-BESSE, UNIT 1 3/4 0-2 Amendment No. 71

1 i

1.

REACTIVITY CONTROL SYSTEMS l

i FLOV PATHS - OPERATING LIMITING CONDITION FOR OPERATION I

3.1.2.2 Each of the follotting boron injection flow paths shall be OPERABLE:

A flow path from the concentrated boric acid storage system via a boric a.-

acid pump and makeup or decay heat removal (DHR) pump to the Reactor Coolant System, and b.

A flow path from the borated water storage tank via makeup or DHR pump to the Reactor Coolant System.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With the flow path from the concentrated boric acid storage system a.

inoperable, restore the inoperable flow path to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least BOT STANDBY and borated to a SHUTDOVN MARGIN equivalent to 1% Ak/k at 200'F vithin the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore the flow path to OPERABLE status within the next 7 days or be in COLD SHUTDOVN vithin the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b.

With the flow path from the borated vater storage tank inoperable, i

restore the flow path to OPERABLE status within one hour or be in at least HOT STANDBY vithin the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOVN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

l SURVEILLANCE REQUIREMENTS 4.1.2.2 Each of the above required flow paths shall be demonstrated OPERABLE:

a.

At least once per 7 daysII) by verifying that the pipe temperature of the heat traced portion of the flow path from the concentrated borie

.i acid storage system is 1 105'F.

l '.

l If the 7 day verification falls during transfers of makeup water or dilute i

boron solutions (fluid source concentration of less than 5000 ppmB), the verification period may be extended up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months after the addition of dilute boron solution has been stopped for a period of at least 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

(!

DAVIS-BESSE, UNIT 1 3/4 1-7 Amendment No. //7/,

135

_ _ _m. - -. - _ _ _ _ _.

a

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q 1

REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

]

b.

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

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4 e

=

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@VIS-BESSE.UNU1 3/4 1,g an-.

--.--_~-___.__.._-_--_----_.u--__________-__

.a

' REACTIVITY CONTROL SYSTEMS MAKEUP PUMP - SHUTDOWN LIMITING CONDITION FOR OPERATION i

3.1. 2. 3 At least one makeup pump in the boron injection flow path-required by Specification 3.1.2.1 shall be. 0PERABLE and capable of being powered from an OPERABLE essential bus.

r.

APPLICABILITY: MODE 5*.

ACTION:

With no makeup pump OPERABLE, suspend all operations involving positive reactivity changes until at least one makeup pump is' restored to'0PERABLE status.

SURVEILLANCE REQUIREMENTS 4.1.2.3 In addition to the Surveillance Requirements of Specification 4.0.5 at least the above makeup pump shall be demonstrated OPERABLE at least once per 31 days by:

a.

Starting (unless already operating) the pump from the control room.

b.

Verifying, that on recirculation flow, the pump develops a discharge pressure of > 2400 psig.

c.

Verifying that the pump operates for at least 15 minutes.

d.

Verifying that the pump is aligned to receive electrical power i

from an OPERABLE essential bus.

1 With RCS pressure > 150 psig.

DAVIS-BESSE, UNIT 1 3/4 1-9 4

4 l

m.

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1 l

REACTIVITY CONTROL SYSTEMS MAKEUP PUMPS - OPERATING LIMITING CONDITION FOR OPERATION I

3 1.2.4 Tuo makeup pumps shall be OPERABLE.

l APPLICABILITY: MODES 1, 2, 3 and 4*.

l l

ACTION:

)

Vith only one makeup pump OPERABLE, restore the inoperable pump to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months er be in at least HOT STANDBY and borated'to'a SHUTDOVN MARGIN equivalent to 1% ok/k at 200'F vithin the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore two pumps to OPERABLE status within the next 7 days or be in COLD SHUTDOVN vithin the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months , j

I SURVEILLANCE REQUIREMENTS 4.1.2.4 In addition to the requirements of Specification 4.0.5, at l

1 east two makeup pumps shall be demonstrated OPERABLE at least once per 31 days by:

4 a.

Starting (unless already operating) each pump from the control room.

b.

Verifying, that on recirculation flow, each pump develops a discharge i

pressure of > 2400 psig.

I c.

Verifying that each pump operates for at least 15 minutes.

d.

Verifying that each pump is aligned to receive electrical power from separate OPERABLE essential busses.

I

Vith RCS pressure > 150 psig.

DAVIS-BESSE, UNIT 1 3/4 1-10 Amendment No.135

REACTIVITY CONTROL SYSTEMS 3/4.1.3 MOVABLE CONTROL ASSEMBLIES GROUP HEIGHT - SAFETY AND REGULATING ROD GROUPS LIMITING CONDITION FOR OPERATIONS 3.1.3.1 All control (safety and regulating) rods shall be OPERABLE and positioned within + 6.5% (indicated position) of their group average height.

APPLICABILITY: MODES 1* and 2*.

ACTION:

a.

With one or more control rods inoperable due to being immovable as a result of excessive friction or mechanical interference or known to be untrippable, determine that the SHUTDOWN MARGIN l

requirement of Specification 3.1.1.1 is satisfied within one hour and be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

l b.

With more than one control rod inoperable or misaligned from its group average height by more than + 6.5% (indicated l

position), be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

l c.

With one control rod inoperable due to causes other than 3

addressed by ACTION a, above, or misaligned from its group average height by more than + 6.5% (indicated position), POWER OPERATION may continue provided that within one hour either:

1 1.

The control rod is restored to OPERABLE status within the above alignment requirements, or 2.

The control rod is declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied.

POWER OPERATION may then continue provided that:

a)

An analysis of the potential ejected rod worth is performed within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and the rod worth is deter-mined to be < 1.0% ak at zero power and < 0.65%

ak at RATED THERMAL POWER for the remainFer of the fuel cycle.

l b)

The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is determined at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

See Special Test Exceptions 3.10.1 and 3.10.2.

DAVIS-BESSE, UNIT 1 3/4 1-19 4

REACTIVITY CONTROL SYSTEMS ACTION:

(Continued) c)

A power distribution map is obtained from the incore O F"gare verified to be within their limits detectors and F within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months d)

Either the THERMAL POWER level is reduced to < 60% of the THERMAL POVER allowable for the reactor coolant pump combination within one hour and within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months the High Flux Trip Setpoint is reduced to < 70% of the THERMAL POWER allovable for the reactor coolant pump combination, or e)

The remainder of the rods in the group with the inoperable rod are aligned to within 6.5% of the inoperable rod within one hour while maintaining the rod sequence, insertion and overlap limits of Figure 3.1-2 and 3.1-3; the THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation.

SURVEILLANCE REQUIREMENTS 4.1.3.1.1 The position of each control rod shall be determined to be within the group average height limit by verifying the individual rod positions at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the Asymmetric Rod Fault Circuitry is inoperable, then verify the individual rod position (s) of the rod (s), with inoperable Fault Circuitry at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

l 4.1.3.1.2 Each control rod not fully inserted shall be determined to be OPERABLE by movement of at least 2% in any one direction at least once every 31 days.

DAVIS-BESSE, KNIT 1 3/4 1-20 Amendment No.135

8 4

Figure 3.1-Sc APSR Position Limits, O to 3251 10 EFPD, Three RC Pumps -- Davis-Besse 1, Cycle 6

)

i 100 1

i 80 RESTRICTED REGION

-l

=

)

g (0,77)

(100,77) a 4

a l

E E 60 S

Ea PERMISSIBLE 40 OPERATING REGION bS b

8 20 0

O 10 20.

30 40 50 60 70 80 90 100 APSR Position (% Withdrawn)

DAVIS-BESSE, UNIT 1 3/4 1-37 Amendment (2$ TT, //I, 6/,

U I

W, w, 135 l

l i

c Figure 3.1-5d AFSR Position Limits, 330 10 to 390 10 EFPD, 2

Three or Four RC Pumps, APSRs Vithdravn--Davis-Besse le cycle 5 DELETED

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DAVIS-BESSE, UNIT 1 3/4 1-38 Amendment No. W.

H, hp, bA, he, ho 123 s

.j i

j iI iI POWER DISTRIBUTION LIMITS NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR - F g i

LIMITING CONDITION FOR OPERATION 3.2.3 F shall be limited by the following relationship:

g a

P s i.7i ri. 0.eci-P33 s

THERMAL POWER where P = RATED THERMAL POWER and P 1 1.0 APPLICABILITY: MODE 1.

ACTION:

With F exceeding its limit:

H Reduce THERMAL POWER at least 1% for each 1% tha exceeds the

[

a.

limit within 15 minutes and similarly reduce the Hi Setpoint and Flux - aFlux - Flow Trip Setpoint within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

Demonstrate through in-core mapping that F" is within its limit b.

within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months af ter exceeding the limit bHr reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

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

prior to increasing THERMAL POWER above the reduced limit required by a or b, aboge; subsequent POWER OPERATION may proceed provided that F is demonstrated through in-core mappingtobewithinit$glimit at a nominal-50% of RATED THERMAL POWER prior to exceeding this THERMAL POWER, at a nominal 75% of RATED THERMAL POWER prior to exceeding this THERMAL POWER and within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after attaining 95% or greater RATED THERMAL POWER.

l DAVIS-BESSE, UNIT 1 3/4 2-7 t

PCWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS N

4.2.3.1 F

shall be determined to be within its limit by using the incore l

AH detectors to obtain'a power distribution maps a.

Prior to operation above 75 percent of RATED THERMAL POWER after each fuel loading, and b.

At least once per 31 Effective Full Power Days.

The provisions of Specification 4.0.4 are not applicable.

c.

4.2.3.2 The measured F f

a ve, shall be increased by 5% for AH measurement uncertainty, i

l l

I DAVIS-BESSE, UNIT 1 3/4 2-8 Amendment No.135

J POVER DISTRIBUTION LIMITS 3

OUADRMRT POWER TIUT i

LIMITING CONDITION POR OPERATION j

3.2.4 THE QUADRANI POVER TILT shall not exceed the Steady State Limit of Table 3.2-1.

[

APPLICABILITY: MODE I above 15% of RATED THERMAL POVER.*

ACTION:

Vith the QUADRANT POVER TILT determined to exceed the Steady a.

State Limit but less than or equal to the Transient Limit of Table 3.2-1.

l 1.

Vithin 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months s:

a)

Either reduce the OUADRANT POVER TILT to within its Steady State Limit, or b)

Reduce THERMAL POVER so as not to exceed THERMAL POVER, including power level cutoff, allovable for the reactor coolant pump combination less at least 2%

for each 1% of OUADRANT POVER TILT in excess of the Steady State Limit and within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , reduce the High Flux Trip Setpoint and the Flux-4 Flux-Flow Trip Setpoint at least 2% for each 1% of OUADRANT POVER TILT in excess of the Steady State Limit.

2.

Verify that the OUADRANT POVER TILT is within its Steady State Limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding the Steady State Limit or reduce THERMAL POVER to less than 60% of THERMAL POVER allovable for the reactor coolant pump combination within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the High Flux Trip Setpoint to 3 65.5% of THERMAL POVER allovable for the reactor coolant pump combination within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

3.

Identify and correct the cause of the out of limit j

condition prior to increasing TEERMAL POVER; subsequent POVER OPERATION above 60% of TEERMAL POVER allevable for the reactor coolant pump combination may proceed provided that the OUADRANT POVER TILT is verified within its Steady State Limit at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified acceptable at 95% or greater RATED ' THERMAL POVER.

i

See Special Iest Except2cn 3.10.1 DAVIS-BESSE, UNIT 1 3/4 2-9 Amendment No. 123

i POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION (Continued) i ACTION:

(Continued) l b.

With the QUADRANT POWER TILT determined to exceed the Transient Limit but less than the Maximum Limit of Table 3.2-1, due to misalignment of either a safety, regulating or axial power shapir.g rod:

1.

Reduce THERMAL POWER at least 2% for each 1% of indicated QUADRANT l

POVER TILT in excess of the Steady State Limit within 30 minutes.

2.

Verify that the OUADRANT POWER TILT is within its Transient Limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after exceeding the Transient Limit or reduce THERMAL POWER to less than 60% of THERMAL POVER allovable for the reactor coolant pump combination within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the High Flux Trip Setpoint to < 65.5% of THERMAL POVER allovable for the reactor coolant pump combination within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

3.

Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 60%

of THERMAL POVER allovable for the reactor coolant pump combination may proceed provided that the OUADRANT POWER TILT is verified within its Steady State Limit at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified acceptable at 95% or greater RATED THERMAL POWER.

c.

With the OUADRANT POVER TILT determined to exceed the Transient Limit i

but less than the Maximum Limit of Table 3.2-1, due to causes other I;

than the misalignment of either a safety, regulating or axial power shaping rod:

1.

Reduce THERMAL POWER to less than 60% of THERMAL POWER allowable for the reactor coolant pump combination within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the High Flux Trip Setpoint to 1 65.5% of THERMAL POWER allowable I

for the reactor coolant pump combination within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

2.

Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 60%

of THERMAL POVER allovable for the reactor coolant pump combination may proceed provided that the QUADRANT POWER TILT is verified within its Steady State Limit at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified at 95% or greater RATED THERMAL POWER.

DAVIS-BESSE, UNIT 1 3/4 2-10 Amendment No./J73,135

._____---______-_-___L

t k

POVER DISTRIBUTION LIMITS DNB PARAMETERS LIMITING CONDITION POR OPERATION 3.2.5 The following DNB related parameters shall be maintained within the limits shown on Table 3.2-2.

l a.

Reactor Coolant Hot Leg Temperature b.

Reactor Coolant Pressure c.

Reactor Coolant Flow Rate APPLICABILITY: MODE 1 ACTION:

If parameter a or b above exceeds its limit, restore the parameter.to within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POVER to less than 5%

of RATED THERMAL POVER vithin the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

If parameter e exceeds its limit, either:

1.

Restore the parameter to within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or 2.

Limit THERMAL POVER st least 2% below RATED THERMAL POVER for each 1% parameter e is outside its limit for four pump operation within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , or limit THERMAL POVER at least 2% belov 75% of RATED THERMAL POWER for each 1% parameter e is outside its limit for 3 pump operation within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS 4.2.5.1 Each of the parameters of Table 3.2-2 shall be verified to be l

vithin their limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.2.5.2 The Reactor Coolant System total flow rate shall be determined to be within its limit by measurement at least once per 18 months.

DAVIS-BESSE, UNIT 1 3/4 2-13 Amendment No. 4/. 123

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TABLE 3.3-1 (Continued)

TABLE NOTATION

With the control. rod drive trip breakers in the closed position and the control rod drive system capable of rod withdrawal.

- When Shutdown Bypass is actuated.

The provisions of Specification 3.0.4 are not applicable.

1. High voltage to detector may be de-energized above 10-10 amps on both Intermediate Range channels.

(a) Trip may be manually bypassed when RCS pressure 1 1820 psig by actuating Shutdown Bypass provided that:

(1) The High Flux Trip Setpoint is 1 5% of RATED THERMAL

POWER, 1

(2) The Shutdown. Bypass High Pressure Trip Setpoint of $ 820 psig is imposed, and (3) The Shutdown Bypass is removed when RCS pressure > 1820 psig.

l (b) Trip may be manually bypassed when Specification 3.10.3 is in effect.

(c) The minimum channels OPERABLE requirement may be reduced to one when Specification 3.10.1 or 3.10.2 is in effect.

l ACTION STATEMENTS With the number of channels OPERABLE one less than required I

ACTION 1 by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and/or open the control rod drive trip breakers.

I With the number of OPERABLE channels one less than the ACTION 2 Total Number of Channels STARTUP and/or POWER OPERATION may proceed provided all of the following conditions are satisfied:

a.

The inoperable channel is placed in the tripped condition within one hour.

b.

The Minimum Channels OPERABLE requirement is met; however, one additional channel may be bypassed for I

up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing-per

)

Specification 4.3.1.1.1, i

DAVIS-BESSE, UNIT 1 3/4 3-3

TABLE 3.3-1 (Continued)

. ACTION STATEMENTS (Continued) and the. inoperable channel above may be bypassed for up to 30 minutes in any 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period when necessary to test.the i

trip breaker associated with the logic of the channel being tested per Specification 4.3.1.1.1, and c.

Either,. THERMAL F0VER is restricted to < 75% of RATED-

~

THERMAL POWER'and the High Flux Trip Setpoint'is~ reduced.to

< 85% of RATED THERMAL POWER within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or the QUADRANT j

F0VER TILT is monitored at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

i ACTION 3 -

Vith the number of OPERABLE channels one less.than the Total Number of Channels STARTUP'and. POWER OPERATION may proceed provided both of the follovng conditions are satisfied:-

a.

The inoperable channel is placed in the tripped condition 3

vithin one hour.

1 b.

The Minimum Channels OPERABLE requirement is met; however, one additional channel may.be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing-per. Specification 4.3.1.1.1, and the inoperable channel above may be bypassed for up to 30 minutes in any 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ~ period when necessary to test the trip breaker associated with the logic of the channel being tested per Specification 4.3.1.1.1.

ACTION 4 -

Vith the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement and with the THERMAL POWER levels l

i a.

< 5% of RATED THERMAL POVER restore the inoperable channel to OPERABLE status prior to increasing THERMAL POVER above 5% of RATED THERMAL POWER.

b.

> 5% of RATED THERMAL POVER,~. POWER OPERATION may continue.

1 Ia i

i 4

DAVIS-BESSE, UNIT 1 3/4 3-4 Amendment No.135 i

i

l TABLE 3.3-1 (continued)

ACTION STATEMENTS (Continued)

ACTION 5 -

Vith the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement and with the THERMAL POWER level f 10-10 amps on the Intermediate Range (IR) instru-a.

mentation,restoretheinoperablechanneltoOPERAggE status prior to increasing THERMAL POWER above 10-amps on the IR instrumentation.

b.

> 10-10 amps on the IR instrumentation, opera. tion may continue.

ACTION 6 -

With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, verify compliance with the SHUTDOVN MARGIN requirements of Specification 3.1.1.1 within one hour and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter.

ACTION 7 -

With the number of OPERABLE channels one less than the Total Number of Channels STARTUP and/or POVER OPERATION may proceed provided all of the following conditions are' satisfied:

a.

Within 1 hour:

1.

Place the inoperable channel in the tripped condition, or 2.

Remove power supplied to the control rod trip device associated with the inoperative channel.

b.

One additional channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.1.1.1, and the inoperable channel above may be bypassed for up to 30 l

minutes in any 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period when necessary to test the j

trip breaker associated with the logic of the channel being i

tested per Specification 4.3.1.1.1.

The inoperable channel j

1 above may not be bypassed to test the logic of a channel of the trip system associated with the inoperable channel.

I i

k DAVIS-BESSE, UNIT 1 3/4 3-5 Amendment No. AQ8,135 (Next page is 3/4 3-Sa.)

TABLE 3.3-1 (Continued)

ACTION STATEMENTS (Continued)

ACTION 8 -

With one of the Reactor Trip Breaker diverse trip features I

(undervoltage or shunt trip devices) inoperable, restore it to OPERABLE status in 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or. place the breaker in trip in the next hour.

ACTION 9 -

Vith one or both channels of SCR Relays inoperable, restore the

.l channels to OPERABLE status during the next COLD SHUTDOVN exceeding 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l

\\

DAVIS-BESSE, UNIT 1 3/4 3-Sa Amendment No. Jpp'135 (Next page is 3/4 3-6.)

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TABLE 4.3-1 (Continued)

NOTATION (1) -

If not performed in previous 7 days.

(2) -

Heat balance only, above 15% of RATED THERMAL POWER.

(3) -

When THERMAL POWER [TP] is above 50% of RATED THERMAL POWER [RTP),

and at steady state, compare out-of-core measured AXIAL POWER IMBALANCE [ API,] to incore measured AXIAL POWER IMBALANCE [ API ] as y

follows:

RTP [ API, - API ] = Offset Error j

y TP Recalibrates if the absolute value of the Offset Error is > 2.5%

(4) -

AXIAL POWER IMBALANCE and loop flow indications only.

(5) -

Verify at least one decade overlap if not verified in previous 7 days.

(6) -

Neutron detectors may be excluded from CHANNEL CALIBRATION.

l (7) -

Plow rate measurement sensors may be excluded from CHANNEL l

CALIBRATION. However, each flow measurement sensor shall be calibrated at least once per 18 months.

I (8) -

The CHANNEL FUNCTIONAL TEST shall independently verify the l

)

OPERABILITY of both the undervoltage and shunt trip devices of the Reactor Trip Breakers.

Vith any control rod drive trip breaker closed.

i When Shutdown Bypass is actuated.

j l

l DAVIS-BESSE, UNIT 1 3/4 3-8 Amendment No. A3, AAB,A4% 135 l

l

4 l!

l-INSTRUMENTATION

'I 3/4.3.2 SAFETY SYSTEM INSTRUMENTATION SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2.1 The Safety Features Actuation System (SFAS) functional units shown in Table 3.3-3 shall be OPERABLE with their trip setpoints set I

consistent with the values shown in the Trip Setpoint column of Table 3.3-4 and with RESPONSE TIMES as shown in Table 3.3-5.

APPLICABILITY: As shown in Table 3.3-3.

ACTION:

a.

With a SFAS functional unit trip setpoint less conservative than the value shown in the Allowab'e Values column of Table 3.3-4. declare the functional unit inoperable and apply the 1

applicable ACTION requirement of Table 3.3-3. until the func-l tional unit is restored to OPERABLE status with the trip setpoint adjusted consistent with the Trip Setpoint value.

b.

With a SFAS functional unit inoperable, take the action shown in Table 3.3-3.

l l

SURVEILLANCE REQUIREMENTS 1

4. 3. 2.1.1 Each SFAS functional unit shall be demonstrated OPEP/BLE by the performance of the CHANNEL CHECK. CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST during the MODES and at the frequencies shewn in Table 4.3-2.

4.3.2.1.2 The logic for the bypasses shall be demonstrated OPERABLE a

i during the at power CHANNEL FUNCTIONAL TEST of functional units affected by bypass operation. The total bypass function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each functional unit affected by bypass operation.

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

j Each test shall include at least one functional unit per function such that all functional units are tested at least once every N times 18 months where N is the total number of redundant functional units in a specific SFAS function as shown in the " Total No. of Units" Column of Table 3.3-3.

DAVIS-BESSE. UNIT 1 3/4 3-9 e

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TABLE 3.3-3 (Continued)

TABLE NOTATION l

Trip function may be bypassed.in this MODE vith RCS pressure belov 1800 psig. Bypass shall be automatically removed when RCS pressure exceeds 1800 psig.

Trip function may be bypassed in this MODE vith RCS pressure belov 600 psig. Bypass shall be automatically removed when RCS pressure exceeds 600 psig.

One must be in SFAS Channels #1 or #3, the other must be in Channels

2 or #4.

This instrumentation must be OPERABLE during CORE ALTERATIONS or l

movement of irradiated fuel within the containment to meet the-requirements of Tech. Spec 3.9.4.

I All functional units may be bypassed for up to one minute when starting each Reactor Coolant Pump or Circulating Water Pump.

j 1

The provisions of Specification 3.0.4 are not applicable.

ACTION STATEMENTS ACTION 10 -

Vith the number of OPERABLE functional units orie less than the l

1 Total Number of Units, STARTUP and/or POWER OPERATION may l

proceed provided both of the following conditionL are satisfied:

a.

The inoperable functional unit is placed in the tripped condition within one hour.

For functional-unit 4a the 8

sequencer channel shall be placed in the tripped condition by physical removal of the sequencer module.

b.

The Minimum Units OPERABLE requirement is met; however, one I

additional functional unit may be bypassed for up to 2 l

hours for surveillance testing per Specification 4.3.2.1.1.

j ACTION 11 -

With any component in the Output Logic inoperable, trip the l

l associated components within one hour or be in at least HOT i

STANDBY vithin the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN vithin the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

ACTION 12 -

Vith the number of OPERABLE Units one less than the Total Number I

of Units, restore the inoperable functional unit to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT STANDBY vithin the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOVN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

ACTION 13 -

a.

With less than the Minimum Units OPERABLE and reactor l

coolant pressure > 438 psig, both Decay Heat Isolation Valves (DH11'and DH12) shall be verified closed.

\\

DAVIS-BESSE, UNIT 1 3/4 3-12 Amendment No. 28, 27, 32, 197, 135

l TABLE 3.3-3'(Continued)

{

' ACTION STATEMENTS

'i b.

With Less then the Minimum Units OPERABLE and reactor coolant pressure < 438 psig operation may continue; however, the functional unit shall be OPERABLE prior.-.to increasing reactor coolant pressure above 438 psig.

ACTION 14 -

.Vith less than the Minimum Units OPERABLE and reactor coolant-l pressure <-438 psig, operation may. continue; however, the functional unit shall be OPERABLE prior to increasing reactor-coolant pressure above'438 psig, or the inoperable functional unit shall be placed in the tripped _ state.

ACTION 15 -

Vith the number of OPERABLE functional units one less than the I.

Total Number of Units operation may proceed:provided both of the following conditions are satisfied The inoperable section of a functional unit is placed in.

a.

the tripped condition within one hour.

b.

The Minimum Units OPERABLE requirement is met; however, the inoperable section of a functional unit may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.2.1.1.

I 1

i l

J j

-l I

1 l

l DAVIS-BESSE, UNIT 1 3/4 3-12a Amendment No. 28, 52, AQA135 i

l TABLE 3.3-5 (Continued)

SAFETY FEATURES SYSTEM RESPONSE TIES

INITIATING SIGNAL AND FUNCTION RESPONSE TIE IN SECONDS i.

Containment Isolation Valves (cont'd) 5.

Pressurizer Sample NA 6.

Service Water to Cooling Water NA 7.

Vent Header NA B.

Drain Tank NA 1

9.

Core Flood Tank Vent NA l

10.

Core Flood Tank Fill NA I

11.

Steam Generator Sample

12. Quench Tank NA l

NA

.l

13. Emergency Sump NA I

14. RCP Seal Return NA

15. Air Systems NA

16. N System NA 2

17. Quench Tank Sample NA

18. RCP Seal Inlet NA

)

19.

Core Flood Tank Sample NA

20. RCP Standpipe Demin Water Supply NA 21.

Containment H Dilution Inlet NA 2

4 22.

Containment H Dilution Outlet NA I

^

l j.

BWST Outlet Valves NA j

k.

Low Pressure Injection 1.

Decay Heat Pumps NA 3

2.

Low Pressure Injection Valves NA i

3.

Decay Heat Pump Suction Valves NA I

4.

Decay Heat Cooler Outlet Valves NA 5.

Decay Heat Cooler Bypass Valves NA 1.

Containment Spray Pump NA m.

Component Cooling Isolation Valves 1.

Inlet to Containment NA 2.

Outlet from Containment NA 3.

Inlet to CRDM's NA 4.

CRDM Booster Pump Suction NA 5.

Component Cooling from Decay Heat Coolers NA i

l l

DAVIS-BESSE, UNIT 1 3/4 3-15 Amendment No. Iff, MA135 l

___ _ = =

TABLE 3.3-5 (Continued)

SAFETY FEATURES SYSTEM RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS I

2.

Containment Pressure - High I

a.

Fans 1.

Emergency Vent Fans

< 25*

2.

Containment Cooler Fans 345*

b.

HV & AC Isolation Valves 1.

ECCS Room

< 75*

2.

Emergency Ventilation 575*

3.

Containment Air Sample, 5 30*

4.

Containment Purge

< 15*

5.

Penetration Room Purge j[75*

c.

Control Room HV &'AC Units

,'10*

d.

High Pressure Injection 1.

High Pressure Injection Pumps 5 30*

2.

High Pressure Injection Valves

_ 30*

e.

Component Cooling Water 1.

Component Cooling Water Pumps 5 180*

2.

Component Cooling Aux. Equip. Inlet Valves 1 180*'

3.

Component Cooling to Air Compressor Valves 1

~< 180*

l f.

Service Water System 1.

Service Water Pumps 1 45*

2.

Service Water From Component Cooling l

Heat Exchanger Isolation Valves

,, NA*

l g.

Containment Spray Isolation Valves 5 80*

1 h.

Emergency Diesel Generator

< 15*

l i

DAVIS-BESSE, UNIT 1 3/4 3-16 Amendment No. 114

-__._______.__m

TABLE 3.3-5 (Continued)

SAFETY FEATURES SYSTEM RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS 2.

Containment Pressure - High (Continued) 1.

Containment Isolation Valves 1.

Vacuum Relief

< 30*

2.

Normal Sump 525*

3.

RCS Letdown Delay Coil Outlet 5 30*

4.

RCS Letdown High Temperature 1 30*

5.

Pressurizer Sample 5 48*

6.

Service-Water to Cooling Water 1 45*

7.

Vent Header

< 15*

8.

' Drain Tank

< 15*

9.

Core flood Tank Vent i 15*

10 Core Flood Tank Fill i 15*

11.

Steam Generator Sample.

515*

12.

Quench Tank 13.

Emergency Sump

-< 15*

NA*

14.

RCP Seal Return

< 45*

15.

Air System i 15*

16.

N2 System i 15*

17.

Quench Tank Sample 535*

18.

RCP Seal Inlet

< 17*

19.

Core Flood Tank Sample 515*

20.

RCP Standpipe Demin Water Supply 5 15*

21.

Containment H2 Dilution Inlet 5 75*

22.

Containment H2 Dilution Outlet 5 75*

j.

BWST Outlet Valves

.NA*

k.

Low Pressure Injection 1.

Decay Heat Pumps

< 30*

2.

Low Pressure Injection Valves i NA*

3.

Decay Heat Pump Suction Valves I NA 4.

Decay Heat Cooler Outlet Valves i NA*

5.

Decay Heat Cooler Bypass Valves 2_ NA*

3.

Containment Pressure--High-High j

l a.

Containment Spray Pump 5 80*

i b.

Component Cooling Isolation Valves 1

i 1.

Inlet to Containment 1 25*

)

2.

Outlet from Containment

_ 25*

1 DAVIS-BESSE, UNIT 1 3/4 3-17 Amendment No. Yff,174.

l

,,a TABLE 3.3-5 (Continued)

SAFETY FEATURES SYSTEM RESPONSE TIMES' INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN' SECONDS b.

Component Cooling Isolation Valves (Continued) 3.

Inlet to CRDM's

< 35*

' 4.-

CRDM Booster Pump Suction 535*

5.

Component Cooling from Decay Heat Cooler 5 NA*

4.

RCS-Pressure-Low a.

Fans 1.

Emergency Vent Fans 5 25*

l 2.

Containment Cooler Fans 5 45*

q b.

HV & AC Isolation Valves 1.

ECCS Room

< 75*

2.

Emergency Ventilation 2 75*

3.

Containment Air Sample 530*

4.

Containment Purge-5 15*

j 5.

Penetration Room Purge

< 75*

c.

Control Room HV & AC Units

< 10*

d.

High Pressure Injection 1.

High Pressure Injection Pumps 5 30*

2.

High Pressure Injection Valves

_ 30*

e.

Component Cooling Water l

1.

Component Cooling Water Pumps 5 180*

2.

Component Cooling Aux. Equipment' Inlet Valves

< 180*

3.

Component Cooling to Air Compressor Valves 5180*

f.

Service Water System l

.h 1.

Service Water Pumps 5 45*

i 2.

Service Water from Component Cooling Heat Exchanger Isolation Valves

< NA*

i i

g.

Containment Spray Isolation Valves 5 80*

)

h.

Emergency Diesel Generator

_ 15*

l DAVIS-EESSE, UNIT 1 3/4 3-18 Amendment No.

114

l TABLE 3.3-5 (Continued) 1 SAFETY FEATURES SYSTEM RESPONSE TIMES I

INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS 4.

RCS Pressure-Low (continued) 1.

Containment Isolation Valves 1.

Vacuum Relief

< 30*

2.

Normal Sump 325*

3.

RCS Letdown Delay Coil Outlet

< 30*

4.

RCS Letdown High Temperature 1 30*

5.

Pressurizer Sample 5 45*-

6.

Service Water to cooling Water 5 45*

7.

Vent Header

< 15*

8.

Drain' Tank 2 15*

9.

Core Flood Tank Vent 515*

a 10.

Core Flood-Tank Fill 5 15*

I l

11.

Steam Generator Sample 5 15*

j 12.

Quench Tank

~< 15*

13.

Emergency Sump NA*

14 Air Systems

< 15*

15.

N2 System 2 15*

16.

Quench Tank Sample 335*

i 17.

Core Flood Tank Sample

< 15*

18.

RCP Standpipe Demin Water Supply 515*

19.

Containment H2 Dilution Inlet

< 75*

20.

Containment H2 Dilution Outlet

_7 75*

j.

BWST Outlet Valves NA*

5.

RCS Pressure--Low-Low a.

Low Pressure Injection 1.

Decay Heat Pumps 5 30*

2.

Low Pressure Injection Valves 1 NA*.

3.

Decay Heat Pump Suction Valves 5 NA*

4.

Decay Heat Cooler Outlet Valves 5 NA*

5.

Decay Heat Cooler Bypass Valves

_ NA*

b.

Component Cooling Isolation Valves 1.

Auxiliary Equipment Inlet 5 90*

2.

Inlet to Air Compressor 5 90*

3.

Component Cooling from Decay Heat Cooler

_ NA*

c.

Containment Isolation Valves 1.

RCP Seal Return

< 45*

2.

RCP Seal Inlet 5 17 I

DAVIS-BESSE, UNIT 1 3/4 3-19 Amendment No A4.AA/ 444,135 l l

m.

l 1

i

-l l

]

TADLE 3.3-5 (Continued)

SAFETY FEATURES SYSTEM RESPONSE T!!1ES 1N1T1ATING $1GNAL AND FUNCTION RESPONSE T!!1E IN SECONDS 6.

Containment Radiation - High a.

Emergency Yent Fans

< 25*

b.

HV & AC ! solation Yalves 1.

ECCS Room

< 75*

2.

Emergency Ventilation 7 75*

3.

Containment Air Sample 7 30*

4 Containment Purge 7 15*

5.

Penetration Room Parge 175*

c.

Control Room HV & AC Units

< 10' TABLE NOTATION Diesel generator starting and sequence loading delays included when applicable.

Response time limit includes movement of valves and attainment of pump or blower discharge pressure.

i I

' DAVIS-BESSE, UNIT 1 3/4 3-20 Amendment No.40 I

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1 INSTRUMENTATION STEAM AND FEEDWATER RUPTURE CONTROL SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION-3.3.2.2 The Steam and Feedwater Rupture Control System (SFRCS) instrumen-tation channels shown in Table 3.3-11 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-12 and with RESPONSE TIMES as shown in Table 3.3-13.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

a.

With a SFRCS instrumentation channel trip setpoint less con-servative than the value shown in the Allowable Values column' of Table 3.3-12, declare the channel inoperable and apply the applicable ACTION requirement of Table 3.3-11, until the channel is restored to OPERABLE status with the trip setpoint adjusted consistent with the Trip Setpoint value.

b.

With a SFRCS instrumentation channel inoperable, take the j

action shown in Table 3.3-11.

j SURVEILLANCE REQUIREMENTS

4. 3. 2. 2.1 Each SFRCS instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST during the MODES and at the frequencies shown in Table 4.3-11.

4.3.2.2.2 The logic for the bypasses shall be demonstrated 0PERABLE during the at power CHANNEL FUNCTIONAL TEST of channels affected by bypass operation. The total bypass function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing l

of each channel affected by bypass operation.

4.3.2.2.3 The STEAM AND FEEDWATER RUPTURE CONTROL SYSTEM RESPONSE TIME of each SFRCS function shall be demonstrated to be within.the' limit at least once per 18 months.

Each test shall include at least one channel per function such that all channels are tested at least once every N times 18 months where N is the total rumber of redundant channels in a specific SFRCS function as shown in the " Total No. of Channels" Column of Table 3.3-11.

DAVIS-BESSE, UNIT 1 3/4 3-23 m

1

C A

SE NLL UEB NNA INR 2

NAE IHP N

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TABLE 3.3-11 (Continued)

TABLE NOTATION May be bypassed when steam pressure is belov 700 psig.

Bypass shall be automatically removed when the steam pressure exceeds 750 psig.

The provisions of Specification 3.0.4 are not applicable.

ACTION STATEMENTS ACTION 16 With the number of OPERABLE Channels one less than the I

Total Number of Channels, STARTUP and/or POWER OPERATION l

1 may proceed until performance of the next required CHANNEL FUNCTIONAL TEST provided the inoperable section of the channel is placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .-

l ACTION 17 Vith the number of OPERABLE Channels one less than the i

1 Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT STANDBY vithin the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOVN vithin the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

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i TABLE 3.3-13 STEAM AND TEEDVATER RUPTURE CONTROL SYSTEM RESPONSE TIMES i

ACTUATED EQUIPMENT RESPONSE TIME IN SECONDS I

1.

Auxiliary Feed Pump f 40 2.

Main Steam Isolation Valves

a.

Main Steas Low Pressues channels f6 b.

Teodwater/ Steam Generator Eigh Differential Pressure Channels 5 6.3 3.

Main Feedvater Val'res a.

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<8

b. 'Startup control 3 13 c.

Stop Valve

$ 16 4.

Turbine Stop Valves **

.31 l

l The response time is to be the time elapsed from the monitored variable exceeding the trip setpoint until the MSIV is fully closed.

The response time is to be the time elapsed from the main steam line low pressure trip condition until the TSV is fully closed.

DAVIS-BESSE UNIT 1 3/4 3-29 Amendment No. 119, 125

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l INSTRUMENTATION ANTICIPATORY REACTOR TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION l

3.3.2.3 The Anticipatory Reactor Trip System instrumentation channels of

'lable 3.3-17 shall be OPERABLE.

l APPLICABILITY: As shown in Table 3.3-17 l

ACTION: As shown in Table 3.3-17 g

SURVEILLANCE REQUIREMENTS l

4.3.2.3 The Anticipatory Reactor Trip System shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST for the modes and at the frequencies shown in Table 4.3-17.

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DAVIS-BESSE, UNIT 1 3/4 3-30a Amendment No. 77,135

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TABLE 3.3-17 (CONTINUED) l ACTION STATEMENTS ACTION 18 Vith the number of' channels OPERABLE one less than required l

I by the Minimum Channels OPERABLE requirements, restore the l

inoperable channel to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or reduce reactor power to less than 45 percent of RATED THERMAL POVER vithin the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION li With the number of channels OPERABLE one less than requ2 red l

by the Minimum Channels OPERABLE requirements, restore the inoperable channel to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY vithin the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 20 With the number of OPERABLE channels one less than the Total' l

Number of Channels, STARTUP and/or POWER OPERATION may l

proceed provided both of the following conditions are satisfied a)

The control rod drive trip breaker associated with the inoperable channel is placed in the tripped condition within one hour.

b)

The Minimum Channels OPERABLE requirement is met; however, one additional control rod drive trip breaker associated with another channel may be tripped for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.2.3, after reclosing the control rod drive trip breaker opened in a) above.

l l

1 DAVIS-BESSE, UNIT 1 3/4 30c Amendment No. 73,128,135

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INSTRUMENTATION 1

3/4.3.3 MONITORING INSTRUMENTATION RADIATION MONITORING INSTRUMENTATION l

LIMITING CONDITION FOR OPERATION 3.3.3.1 The radiation monitoring instrumentation channels shown in Table 3.3-6 shall be OPERABLE with their alarm / trip setpoints within the specified limits.

l APPLICABILITY: As shown in Table 3.3-6.

I ACTION:

a.-

With a radiation monitoring channel' alarm / trip setpoint exceeding the value shown in Table 3.3-6, adjust the setpoint to within~the limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or declare the channel inoperable.

b.

With one or more radiation monitoring channels inoperable, take the ACTION shown in Table 3.3-6.

c.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.1 Each radiation monitoring instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the modes at the frequencies shown in Table 4.3-3.

4 DAVIS-BESSE, UNIT 1 3/4 3-31 i

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TABLE 3.3-6 (Continued)

TABLE NOTATION ACTION 21 Vith the number of channels OPERABLE less than required by I

the Minimum Channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.4.6.1.

ACTION 22 Vith-the number of channels OPERABLE less than required by I

the Minimum Channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.9.12.

DAVIS-BES k, UNIT 1 3/4 3-33 Amendment No.135

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INSTRUMENTATION POST-ACCIDENT INSTRUMENTATION 1

LIMITING CONDITION FOR OPERATION 3.3.3.6 The post-accident monitoring instrumentation channels shown in Table 3.3-10 shall be OPERABLE.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

a.

With the number of OPERABLE post-accident monitoring channels less than required by Table 3.3-10, either restore the inoperable channel to OPERABLE status within 30 days,' or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b.

The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.6 Each post-accident monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-10.

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i Shxt 1 cf 4 TAB 1.E 3.3-14 FIRE DETECTION INSTRUMENTS j

J C

MINIMUM INSTRUMENTS OPERAB1.E l,

INSTRUMENT LOCATION HEAT TLAME SHOKE 1.

Contalament da. TDI-RCP 1 Elev. 603' O

O 1*

ib. TDZ-RCP 2 Elev. 603' 0

0 1~ *

c. TDZ-RCP 3 Elev. 603' 0

0 1*

d. TDZ-RCP 4 Elev. 603' 0

0 1*

i 0

0 1*

fe.

TDZ-FIR Elev. 603'

f. TDZ 214 - Oore Flooding Tank 1-1 Area -

Elev. 565' 0

0 3*'

1 fg. TDI 215 - Ctat. Leedown Cooler Arsa Elev. 565' 0

0

., 2 *

~

ih. TDZ 220'- Incore Instrumeat' Trench Area j

Elev. 565' 0

0

-4*

fi. TDZ 317 - Batch Area - Elev. 565' 0

0 20

fj. TDI 410 - East Fassage - Elev. 603'/657' 0

'0 9*

2.

Containment Annulus fa.

TDZ-A208 Elev. 590' 0

0 10 e

lb. TDZ-236E Elev. 774' O

O 3

fe. TDE-2361.

Elev. 590' 0

0 9

3.

Auxiliary Building TDZ 402 - #1 Electrical Penetration Rs.

a.

Elev. 603' 0'

0 12 b.

TDI 40$ - Auxiliary Building Storage Rs.

Elev. 603' O

O 1

FDZ 427 - #2 Electrical Penetration Rs.

Elev. 603' O

O 7

c.

d.

TDI 303 - #3 Mechanical Pentration Rs.

Elev. 585' O

0 12 TDZ 304 - Corridor to Hech. Pent Ras 364 e.

Elev. 585' 0

0 4

f.

TDI 310 - Passage to BA Mix Tank 8

Elev. 585' O

0 FDZ 312 - Spent Tuel Fool Fump Rs.

3 Elev. 385' 0

0 4

h.

FDZ 314 - #4 Mech. 7ent. Room Elev. 585' 0

0 17 1.

TDZ 300 - Fuel Handling Area Elev. 585' 0

0 5

Oo g # Tire Detectors in high radiation areas which are NOT accessible.

t DAVIS-BESSE, UNIT 1 3/4 3-53 Amendment No..#,34

j TABLE 3.3-14 (Continued) l l

sheet 2 of 4 i

INSTRUME!E LOCATION MINIMUM INSTRUMENTS OPERABLE REAT FLAME SMOKE 1.

Auxiliary Building (Continued) j.

FDZ 209 - Corridor to il Mech. Pent. Rs.

Elev. 565' O

O 3,

k.

FDZ 227 - Boric Acid Evap Passagevay Elev. 565' 0

0 6

1.

FDZ 208 - #1 Mechanical Penetration Ra.

Elev. 565' 0

0 10 a.

FDZ 231 - Clean Vaste Booster Pump Rs.

I Elev. 565' O

O 1

n.

FDZ 232 - Primary A Deborating Demin Viv Ra.

Elev. 565' O

O 1

o.

FDZ 234 - Boric Acid Evaporator Rs 1-2 Elev. 565' O

O p.

FDZ 235 - Boric Acid Evaporator Rs 1-1 1

Elev. 565' O

O 1

q.

FDZ 236 - #2 Mechanical Penetration Rs.

Elev. 565' O

O 4

r.

FDZ 240 - Boric Acid Addition Tank Ra.

Elev. 565' 0

0 5

s.

FDZ 241 - Passage to B.A. Addition Tk. Ra.

Elev. 565' O

O 2

t.

FDZ 101 - Equipment and Pipe Tunnel Elev. 545' O

O 1

u.

FDZ 105 - ECCS Pump Room 1-1 Elev. 545' O

O 4

v.

FDZ 110 - Corridor to Central Area of Aux Bldg. - Elev. 545' O

O 5

v.

FDZ 113 - Decay Heat Exchanger Pit Elev. 545' O

O 1

x.

FDZ 115 - ECCS Pump Koom 1-2 Elev. 545' O

O 2

y.

FDZ 124 - Clean Vaste Receiver Tank Ra. 1-1 Elev. 545' O

O 4

4.

Auxiliary Building Fan Rooms a.

FDZ 500 - Radvaste & Fuel Handling Area and Air Supply Area -

0 0

20 Elev. 623' b.

FDZ 501 - Radvaste Exhaust Equipment and Main Station Exhaust Fan Room Elev. 623' O

O 22 c.

FDZ 515 - Purge and Exhaust Equipment Ra.

Elev. 623' -

0 0

22 d.

FDZ 516 - Non-rad Air and Exhaust Equip.

Rm. - Elev. 623' O

O 5

{

DAVIS-BESSE, UNIT 1 3/4 3-54 Amendment No. 9, 24, 135

1 TABLE 3.3-14 (Continued) l Sheet 3 of 4 INSTRUMENT LOCATION MININUM INSTRUMENTS OPERABIE BEAT FLAME SMOKE 5.

Control Room Complex a.

FDZ 505 - Main control Board cabinets Elev. 623' O

O 3

b.

FDZ 505 - Control Cabinet Roon Elev. 623' 0

0 5

c.

FDZ 505 - Computer Room - Elev. 623' O

O 1

6.

Cable Spreading Room a.

FDZ 422A - Elev. 613' O

O' 5

7.

A/C Equipment Room a.

FDZ 603 - Elev. 643' O

O 11 8.

Diesel Generator Rooms

- a.

FDZ 318 - Diesel Generator Ra. 1-1 Elev. 585' O

O 5

- b.

FDZ 319 - Diesel Generator Rs. 1-2 Elev. 585' O

O 4

c.

FDZ 321A - Diesel Generator Day Tank Room 1 Elev. 5 0

0 1

d.

FDZ 320A - Diesel Generator Day Tank Room 1 Elev. 5 0

0 1

i 9.

Battery Rooms a.

FDZ 428A - Battery Room B - Elev. 603' O

O 2

b.

FDZ 429B - Battery Room A - Elev. 603' O

O 2

10.

Component Cooling Vater Pump Room a.

FDZ 328 - Elev. 585' O

O 9

11. Feed Pump Rooms a.

FDZ 237 - Auxiliary Feed Fusp 1-1

~

Elev. 565' O

O 3

b.

FDZ 238 - Auxiliary Feed Pump 1-2 Elev. 565' O

O 3

s i

DAVIS-BESSZ, UNIT 1 3/4 3-55 Amendment No. 9,34, 135

TABLE 3.3-14 (Continued) l Sheet 4 of 4 INSTRUMENT LOCATION MINIMUM INSTRUMENTS OPERABLE I

REAT FLAME SMOKE j

j

12. Svitchgear Rooms a.

FDZ 324 - CD Eigh Voltage Switchgear 0

0 3

Elev. 585' b.

FDZ 325 - A High Voltage Switchgear Elev. 585' O

O 8

c.

FDZ 323 - B High Voltage Switchgear Elev. 585' O

O 11 d.

FDZ 428 - F High Voltage Switchgear 12 i

Elev. 603' 0-

'O I

e.

FDZ 429 - E High Voltage Switchgear Elev. 603' 0

0 6

13. Intake Structure a.

FDZ 052 - Diesel Fire Fump Room i

Elev. 576' O

O 1

b.

FDZ 052 - Service Water Fump Roon l

Elev. 576' 0

0 3

c.

FDZ 053 - Service Water Vlv. Room Elev. 565' O

O 6

i 4

The fire detection instruments located within the Containment are not required to be OPERABLE during the performance of Type A Containment Leakage Rate Tests.

- These detectors automatically actuate fire suppression systems.

l l

DAVIS-BESSE, UNIT 1 3/4 3-56 Amendment No. 9, 34, 135 g

~

INSTRUMENTATION

.I RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION O

LIMITING CONDITION FOR OPERATION 3.3.3.9 The radioactive liquid affluent monitoring instroentation channels shown in Table 3.3-15 shall be OPERABLE with their alars/ trip setpoints set to ensure that the Itaits of Specification 3.11.1.1 are not exceeded. The alarm / trip setpoints of these channels shall be determined and adjusted in occordance with the methodology and parameters in the OFFSITE DOSE CALCULA-TION MANUAL (00CM).

APPLICABILITY: At all times.

ACTION:

a.

With a radioactive liquid effluent monitoring instrumentation channel alars/ trip setpoint less conservative than required by the above l

specification, without delay suspend the release of radioactive l

liquid effluents menitored by the affected channel, or declare the channel inoperable, or change the setpoint so it is acceptably conservative.

j i

b.. With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3-15.

Exert best efforts to return the instruments to OPERABLE status within 30 days and, if unsuccessful, explain in j

the next Semiannual Radioactive Effluent Release Report why the inoperability was not corrected in a timely manner.

c.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.9 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION AND CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3-15.

s DAVIS-BESSE, UNIT 1 3/4 3 57 Amendment No. 85

(

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TABLE 3.3-15 (Continued)

TABLE NOTATION (1) During radioactive releases via this pathway ACTION 23 Vith the number of channels OPERABLE less than required by the l

Minimum Channels OPELABLE requirement, effluent releases may be resumed, provided that prior to initiating a release:

1.

At least two independent samples are analyzed in accordance with Specification 4.11.1.1.1 for analyses performed with each batch; 1

2.

At least two independent verifications of the release rate-l calculations are performed; 3.

At least two independent verifications of the discharge valving are performed; otherwise, suspend release of radioactive effluents via this pathway.

ACTION 24 Vith the number of channels OPERABLE less than required by the l

Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during actual releases.

Pump curves may be used to estimate flow.

ACTION 25 Vith the number of channels OPERABLE less than required by the l

Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , grab samples are collected and analyzed for gross radioactivity (beta or gamma) at a lover limit of detection no greater than 10~7 pCi/ml.

l l

DAVIS-BESSE, UNIT 1 3/4 3-59 Amendment No. 55, 135

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TABLE 3.3-16 (Continued)

TABLE NOTATION (1) During radioactive vaste gas releases via this pathvay.

(2) During additions to the vaste gas surge tank ACTION 26 Vith the number of channels OPERABLE less than required by I

the Minimum Channels OPERABLE requirement, the contents of the tank any be released to the environment provided that prior to initiating the release:

1.

At least two independent samples are analyzed in accordance with Specification 4.11.2.1.1 for analyses performed with each batch; 2.

At least two independent verifications of the release rate calculations are performed; 3.

At least two independent verifications of the discharge valving are performed.

ACTION 27 Vith the number of channels OPERABLE less than required by l

the Minimum Channels OPERABLE -requirement, effluent releases via this pathway may continue provided the flov l

rate is estimated at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 28 Vith the number of channels OPERABLE less than required by I

the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided grab samples are taken at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and these samples are analyzed for gross activity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 29 Vith the number of channels OPERABLE less than required by l

the Minimum Channels OPERABLE requirement, additions to the vaste gas surge tank may continue provided another method i

for ascertaining oxygen concentrations, such as grab sample analysis, is implemented to provide measurements at least once per four (4) hours during degassing and daily during j

other operations.

ACTION 30 Vith the number of channels OPERABLE less than required by l

the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided samples are continuously collected with auxiliary sampling equipment, as required in Table 4.11-2.

DAVIS-BESSE, UNIT 1 3/4 3-65 Amendment No. S5,135

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i 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 COOLANT LOOPS AND COOLANT CIRCULATION STARTUP AND POWER OPERATION LIMITING CONDITION FOR OPERATION 3.4.1.1 Both reactor coolant loops and both reactor coolant pumps in each loop shall be in operation.

1 APPLICABILITY: MODES 1 and 2*.

ACTION:

With one reactor coolant pump not in operation. STARTUP and POWER '

a.

OPERATION may be initiated and may proceed provided THERMAL POWER is restricted to less than 80.6% of RATED THERMAL POWER and within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months the setpoints for the following trips have been reduced in accordance with Specification 2.2.1 for operation with three reactor l

coolant pumps operating:

]

1.

High Flux 2.

Flux-aFlux-Flow SURVEILLANCE REQUIREMENTS a

i 4.4.1.1.1 The above required reactor coolant loops shall be verified to be l

in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.4.1.1.2 The Reactor Protection System trip setpoints for the instruments-l tion channels specified in the ACTION statement above shall be verified to l

be in accordance with Specification 2.2.1 for the applicable number of reactor coolant pumps operating either:

a.

Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after switching to a three pump combination if I

the switch is made while operating, or b.

Prior to reactor criticality if the switch is made while shut-

down, i

See Special Test Exception 3.10.3.

DAVIS-BESSE. UNIT 1 3/4 4-1 Amendment No. 76.33.28.M.EO.

133.135

" 3/4.4 PJJ.CTOR COOLANT SYSTEM

' SHUTDOW AC HCT STANDET LIMITINC CONDITION FOR OPERATION 3.4.1[2 At least two of the coolant loops listed below shall be

~

a.

OPERA 3 2:

1.

Reactor " Coolant Isop 1 and its associated steam

]

generator.

i 2.

Reactor Coolant Loop 2 and its associated staan j

generator, 1

3.

Decay Esat Removal !aop 1,*

4.. Decay Beat Removal Ioop 2.*

l b.

At least one of the above coolant loops shall be in I

operation.**

l c.

Not more than one decay heat removal pump may be operated l

vith the sole suction path through DE-11 and DE-12 unless the control power has been removed from the DE-11 and DE-12 valve operator, or manual valves DE-21 and DE-23 are opened.

C d.

The provisions of Speci.fications 3.'0.3 and 3.0.4 are not (W

applicable.

APPLICA3ILITY: HODES 3, 4 and 5 l

ACTION:

)

l 4

I a.

With le.ss than the above required coolant loops CPERt.32, i=ediately initiate corrective action to return the required coolant icops to OPERA 3LE status as seen as possible, or be in COLD SEUIDOW vithin 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months .

l b.

With nous of the above required coolant loops in oper: tion, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System a:;d imediately initiate corrective action to return the required coolant loop to operation.,.

The normal or emergency power source may be inoperable in MODE 5.

This loop may not be selected in MODE 3 unless the primary side temperature and pressure are within the decay heat removal system's design conditions.

- The decay heat removal pumps may be de-energized for up to I hour provided (1) no operations are permitted that would cause dilution of the reactor coolant system boron concentration, and (2) core outlet OG temperature is maintained at least 10*F below saturation temperature.

g DAVIS-BESSE UNIT 1 3/4 4-2 Amendment No. 4, 8, 48,

88. 92

3/4.4 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.1.2.1 The required decay heat removal loop (s) shall be determined OPERABLE per Specification 4.0.5.

4.4.1.2.2 The required steam generator (s) shall be determined OPERABLE by verifying secondary side level to be greater than or equal to (a) 18 inches above the lower tube sheet once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months if an associated reactor coolant pump is operating, or, (b) 35 inches above the lower tube sheet once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months if no reactor coolant pumps are operating.

4.4.1.2.3 At least one coolant loop shall be verified to be in operation and circulating reactor coolant at least.once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

l DAVIS-BESSE, UNIT 1 3/4 4-a Amendment No. Jp,135 l

l l

REACTOR COOLANT SYSTEM

}I SATETY VALVES - SHUTDOVN LIMITING CONDITION FOR OPERATION 3.4.2 Decay Heat Removal System relief valve DH-4849 shall be OPERABLE vith a lift setting of f 330 psig* and isolation valves DH-ll and DH-12 open and l

control power to their valve operators removed.

APPLICABILITY: MODES 4 and 5.

ACTION:

A.

Vith DB-4849 not OPERABLE:

1.

Make the valve OPERABLE vithin eight hours; or-2.

a.

Vithin next one hour, disable the capability of both high pressure injection (HPI) pumps to inject water into the reactor coolant system; and b.

Within next eight hours:

I 1.

Disable the automatic transfer of makeup pump suction to the j

borated water storage tank on low makeup tank level; and j

l 2.

Reduce makeup tank level to f 73 inches and reduce reactor coolant system pressure and pressurizer level within the acceptable region on Figures 3.4-2a (in MODE 4) and 3.4-2b (in MODE 5).

I j

With DH-11 or DH-12 closed, open DB-21 and DH-23 within one hour.

j B.

C.

Vith the control power not removed from DH-11 and DH-12, remove the power to the valve operators at the Motor Control Centers within one hour.

i SURVEILLANCE P. REQUIREMENTS 4.4.2 Decay Heat Removal System relief valve DH-4849 shall be determined OPERABLE:

a.

per the surveillance requirements of Specification 4.0.5.

b.

at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying either:

1 1.

isolation valves DB-11 and DH-12 open with control power removed from their valve operators; or 2.

valves DH-21 and DB-23 open.

The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

DAVIS-BESSE, UNIT 1 3/4 4-3 Amendment No. 37,779.135 i

i i

i REACTOR COOLANT SYSTEM SAFETY VALVES AND PILOT OPERATED RELIEF VALVE - OPERATING l

LIMITING CONDITION FOR OPERATION

~

l 3.4.3 All pressurizer code safety valves shall be OPERABLE with a lift setting l

of < 2525 psig.* When not isolated, the pressurizer pilot operated relief valve l 4

shall have a trip setpoint of 1 2435 psig and an allovable value of 1 2435 psig.**

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

i Vith'one pressurizer code safety valve inoperable, either restore the inoperable I

valve to OPERABLE status within 15 minutes or be in BOT SHUTD0VN vithin 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

1 1

SURVEILLANCE REQUIREMENTS I

4.4.3 For the pressurizer code safety valves, there are no additional Surveillance Requirements other than those required by Specification 4.0.5.

For the pressurizer pilot operated relief valve a CHANNEL CALIBRATION check shall be l

performed every 18 months.

l The lift setting pressure shall corresnond to ambient conditions of the valve at nominal operating temperature and pressure.

Allovable value for CHANNEL CALIBRATION check.

l DAVIS-BESSE, UNIT 1~

3/4 4-4 Amendment No. 77, /F0/77F,135 l

l 1

i 4 i

' l REACTOR COOLANT SYSTDI OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION i

l 3.4.6.2 Reactor Coolant System leakage shall be limited tot j

a.

No PRESSURE BOUNDARY LEAKAGE, b.

1 GPM UNIDENTIFIED LEAKAGE, 1 GPM total primary-to-secondary leakage through steam generators, c.

d.

10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System,~

e.

10 GPM CONTROLLED LEAKAGE, and f.

5 GPM leakage from any Reactor Coolant System Pressure Isolation Valve as specified in Table 3.4-2.

APPJ.ICABILITY: MODES 1, 2, 3 and 4 ACTION:

a.

With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY vithin 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOVN vithin the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b.

Vith any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE, reduce the leakage rate to within '*,mits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STANDBY vithin the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOVN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months except as permitted by paragraph c below.

In the event that integrity of any pressure isolation valve specified c.

in Table 3.4-2 cannot be demonstrated, POVER OPERATION may continue, l

provided that at least two valvas in each high pressure line having a non-functional valve are in and remain in, the mode corresponding to the isolated condition.(a) d.

The provisions of Sections 3.0.4 and 4.0.4 are not applicable for entry into MODES 3 and 4 for the purpose of testing the isolation valves in Table 3.4-2.

(a) Motor operated valves shall be placed in the closed position and power supplies deenergized.

DAVIS-BESSE, UNIT 1 3/4 4-15 MMt #4. A/ID/M, Amendment No. 135 t.

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.6.2.1 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by Monitoring the containment atmosphere particulate radioactivity monitor a.

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

b.

Monitoring the containment sump inventory and discharge at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Measurement of the CONTROLLED LEAKAGE from the reactor coolant pump l

c.

seals to the makeup system when the Reactor Coolant System pressure is 2185 t 20 psig at least once per 31 days.

d.

Performance of a Reactor Coolant System vater inventory balance at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months during steady state operation.

4.4.6.2.2 Each Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-2 shall be individually demonstrated OPERABLE by verifying leakage testing (or the equivalent) to be within its limit prior to entering MODE 23 a.

After each refueling outage, b.

Whenever the plant has been in COLD SHUTDOVN for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or more, and if leakage testing has not been performed in the previous 9 months, and Prior to returning the valve to service following maintenance, repair c.

or replacement work on the valve.

4.4.6.2.3 Vhenever integrity of a pressure isolation valve listed in Table 1

3.4-2 cannot be demonstrated, the integrity of the remaining pressure isolation valve or the integrity of the remaining pressure isolation valve in series with the motor-operated containment isolation valve in each high pressure line having a leaking valve shall be determined and recorded daily.

In addition, the position of the closed motor-operated containment isolation valve located in the high pressure piping shall be recorded daily.

DAVIS-BESSE, UNIT 1 3/4 4-16 pp/g ///. g/p/pyi Amendment No. ff,135

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DOSE EQUIVALENT l-131 Primary Coolant Specific Activity Limit Versus Percent of RATED THERMAL POWER with the Primary Coolant Specrfic ActMty > 1.0gCl/ gram DOSE EQUIVALENT I-131 l

i

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DAVIS-BESSE, UNIT 1 3/4 4 23

.bendment No.135

q 1

REACTOR COOLANT SYSTEM 4

3/4.4.9 PRESSURE / TEMPERATURE LIMITS f

s REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION I

1 3.4.9.1 The Reactor Coolant system (except the pressurizer) temperature j

and pressure shall be limited in accordance with the limit lines shown on j

Figures 3.4-2, 3.4-3 and 3.4-4 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:

i A maximum heatup of 50'F in any one hour period, and l

a.

b.

A maximum cooldown of 100'F in any one hour period with cold leg temperature > 270'F an'd a maximum cooldown of 50'F in any one hour period with cold leg temperature <270'F.

!JPLICABILITY: At all times.

ACTION:

With any of the above limits exceeded, restore the temperature and/or pressure to within the limits within 30 minutes; peri'orm an engineering i

evaluation to determine the effects of the out-of-limit condition on the integrity of the Reactor Coolant System; determine that the Reactor i

Coolant System remains acceptable for continued operation or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

l SURVEILLANCE REQUIREMENTS l

4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operations.

4.4.9.1.2 The reactor vessel material irradiation surveillance specimens representative of the vessel materials shall be removed and examined, to determine changes in material properties, at the intervals defined in BAW 1543A.

The results of these examinations shall be used to update Figures 3.4-2, 3.4-3 and 3.4-4.

1 k

DAVIS-BESSE, UNIT 1 3/4 4-24 Amendment No. E Z,116

4 1

REACTOR COOLANT SYSTEM

(-

SURVEILLANCE REQUIREMENTS (C'ontinued) s b.

Each internals vent valve shall be demonstrated OPERABLE at least once per 18 months during shutdown,* by:

1.

Verifying through visual inspection that the valve body and valve disc exhibit no abnormal degradation, 2.

Verifying the valve is not stuck in an open position, and 3.

Verifying through manual actuation that the valve is fully open when a force of 5 400 lbs. is applied vertically ~ upward.

l

For Cycle 5 Operation, performance of this Surveillance Requirement may be deferred to coincide with the next reactor vessel head removal but no later than the Cycle 5 refueling outage.

DAVIS-BESSE, UNIT 1 3/4 4-31 Amendment No. 23, 95 1

__z______

1 _ _._ _ __ _ _

I l

REACTOR COOLANT SYSTEM REACTOR COOLANT SYSTEM VENTS LIMITING CONDIVION FOR OPERATION 3.4.11 The following reactor coolant system vent paths shall be operable l

a.

Reactor Coolant System Loop 1 vith vent path through valves RC 4608A and RC 4608B.

b.

Reactor Coolant System Loop 2 with vent path through valves RC 4610A and RC 4610B.

i c.

Pressurizer; with vent path through EITHER valves RCll and RC 2A (PORV). l OR valves RC 239A and RC 200.

I APPLICABILITY: Modes 1, 2 and 3

{

ACTION:

With one of the above vent paths inoperable, restore the inoperable a.

vent path to OPERABLE status within 30 days, or, be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOVN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

l b.

With two of the above vent paths inoperable, restore at least one of the inoperable vent paths to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT STANDBY vithin 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOVN within the following 30

hours, i

With three of the above vent paths inoperable, restore at least two of c.

the inoperable vent paths to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT STANDBY vithin 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTD0VN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

d.

The provisions of specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.4.11 Each reactor coolant system vent path shall be demonstrated OPERABLE at least once per 18 months by:

1.

Verifying all manual isolation valves in each vent path are locked in the open position, and 2.

Cycling each valve in the vent path through at least one complete cycle of full travel from the control room during COLD SHUTD0VN or REFUELING, and 3.

Verifying flow through the reactor coolant vent system vent paths during COLD SHUTD0VN or REFUELING.

DAVIS-BESSE, UNIT 1 3/4 4-32 Amendment No. N1,135

ti i

(

EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS - T....

280*F i

LIMITING CONDITION FOR OPERATION 3.5.2 Two independent ECCS subsystems shall be OPERABLE with each subsystem comprised of:

a.

OneOPERABLEhighpressureinjection(HPI) pump, l

b.

One OPERABLE low pressure injection (LPI) pump, c.

One OPERABLE. ' decay heat cooler, and d.

An OPERABLE flow path capable of taking suction from the borated water storage tank (BWST) on a safety injection signal and manually transferring suction to the containment sump during the recirculation phase of operation.

I APPLICABILITY: MODES 1, 2 and 3.

f ACTION:

a.

With one ECCS subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT SHUTDOWN within the next !? hours.

l b.

In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Comission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date.

SURVEILLANCE REQUIREMENTS 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE:

a.

At least once per 31 days by verifying that each valve (manuel, power operated or automatic) in the flow path that is not lodked, sealed or otherwise secured in position, is in its correct position.

DAVIS-BESSE, UNIT 1 3/4 5-3 Amendment No. 36 1

l l

1 l

SURVEILLANCE REQUIREMENTS (Continued) b.

At least once per 18 months, or prior to operation after ECCS piping has been drained by verifying that the ECCS piping is full of water by venting the ECCS pump casings and discharge piping high points.

c.

By a visual inspection which verifies that no loose debris (rags, trash, clothing, etc.) is present in the containment which could be transported to the containment emergency sump and cause restriction of the pump suction during LOCA conditions. This visual inspection shall be performed:

1.

For all accessible areas of the containment prior to establish-ing CONTAINMENT INTEGRITY, and 2.

Of the areas affected within containment at the completion of each containment entry when CONTAINMENT INTEGRITY is established.

d.

At least.once per 18 months by:

1.

Verifying that the interlocks:

a)

Close DH-11 and DH-12 and deenergize the pressurizer heaters, if either DH-11 or DH-12 is open and a simulated reactor coolant system pressure which is greater than the trip setpoint (<438 psig) is applied. The interlock to close DH-11 and/or DH-12 is not required if the valve is closed and 480 V AC power is disconnected from its motor operators.

b) Prevent the opening of DH-11 and DH-12 when a simulated or actual reactor coolant system pressure which is greater than the trip setpoint (<438 psig) is applied.

2.

a) A visual inspection of the containment emergency sump which verifies that the subsystem suction inlets are not restricted by qbris and that the sump components (trash racks, screens, 6.) show :s evidence of structural distress or corrosion.

b) Verifying that on a Borated Water Storage Tank (BWST) Low-Low Level interlock trip, the BWST Outlet Valve HV-DH7A (HV-DH78) automatically close in g 5 seconds after the operator manually pushes the control switch to open the Containment Emergency Sump Valve HV-DH9A (HV-DH9B) which should be verified to open in g 5 seconds.

3.

Verifying a total leak rate,< 20 gallons per hour for the LPI system at:

a) Normal operating pressure or hydrostatic test pressure of

>150 psig for those parts of the system downstream of the

_pump suction isolation valve, and b) >45 psig for the piping from the containment emergency sump 4

Tsolation valve to the pump suction isolation valve.

j DAVIS-BESSE. UNIT 1 3/4 5-4 Amendment No. 3,25,28,fD,77,135

CONTAINMENT SYSTEMS

(

CONTAINMENT VESSEL STRUCTURAL INTEGRITY LIMITING CONDITIONS FOR dPERATION

3. 6.1. 6 The structural integrity of the containment vessel shall be maintained at a level consistent with the acceptance criteria in Specification 4.6.1.6.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

dith the structural integrity of the containment vessel not conforming to-the above requirements, restore the structural integrity to within the limits prior to increasing the Reactor Coolant System temperature above-200*F.

l SURVEILLANCE REQUIREMENTS l

4.6.1.6 The structural integrity of the containment vessel shall be

' l determined during the shutdown for each Type A containment leakage rate test (reference Specification 4.6.1.2) by a visual inspection of the accessible interior and exterior surfaces of the vessel and verifying 2

no apparent changes in appearance of the surfaces-or other abnormal degradation. Any abnormal degradation of the containment vessel detected during the above required inspections shall be reported to the Commission.

i DAVIS-BESSE, UNIT 1 3/4 6-9 Amendment No. g3 B

e

CONTAINMENT SYSTEMS l

CONTAINMENT VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.6.1.7 The containment purge supply and exhausttisolation valves shall be closed.

APPLICABILITY: MODES 1, 2, 3 and 4.

I ACTION:

With any containment purge supply and/or exhaust isolation valve open and providing access to the outside atmosphere, operation may continue, orovided that the accumulated time is < 90 hours0.00104 days 0.025 hours 1.488095e-4 weeks 3.4245e-5 months for the preceding 365 days; otherwise, j

be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

1 SURVEILLANCE REQUIREMENTS 4.6.1.7 The accumulated time any containment purge supply and/or exhaust valve i

is open and provides access to the outside atmosphere shall be determined at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l L

i 1

i DAVIS _BESSE, UNIT 1 3/4 6-10 Amendment No.135 l

TABLE 3.6-2 CONTAINMENT ISOLATION VALVES (Continued)

.\\

l PENETRATION VALVE ISOLATION I

NUMBER NUMBER FUNCTION TIME (seconds)

29 9 DH11 Decay Heat Pump Suction Line N/A

29 DB23 Decay Heat Pump Suction Line N/A 29 9 PSV4849 Decay Beat Pump Suction Line N/A 35 i AF599 Auxiliary Feedvater Line N/A 36 i AF608 Auxiliary Feedvater Line N/A 37 i FV601 Main Feedvater Line N/A 38 4 FV612 Main Feedvater Line N/A

- 39 i M5100 Main Steam Line N/A

- 39 9 ICS11A Main Steam Line N/A 39 9 MS375 Main Steam Line N/A -

39 i MS100-1 Main Steam Line N/A

39 9 MS107 Main Steam Line N/A

39 #

MS107A Main Steam Line N/A

40 i MS106 Main Steam Line N/A

40 t MS106A Main Steam Line N/A

- 40 6 MS101 Main Steam Line N/A i

- 40 t ICS11B Main Steam Line N/A 40 t MS394 Main Steam Line N/A 40 t MS101-1 Main Steam Line N/A j

41 RC113 Pressurizer Quench Tank Inlet Line N/A 42A SA502 Service Air Supply Line N/A 42B CV124 Containment Vessel Air Sample Return N/A 43A IA501 Service Air Supply Line N/A 43B CV125 Containment Vessel Air Sample Return N/A 44A CF15 Core Flood Tank Fill and Nitrogen Supply Line N/A 44B NN58 Pressurizer Quench Tank Inlet Line N/A J

47A CF2A Core Flood Tank Sample Line N/A

47A CF2B Core Flood Tank Sample Line N/A

478 CF5A Core Flood Tank Vent Line N/A

478 CF5B Core Flood Tank Vent Line N/A D

DAVIS-BESSE, UNIT 1 3/4 6-21 Amendment No. 3,114, 127

TABLE 3.6-2 CONTAINMENT ISOLATION VALVES (Continued)

PENETRATION VALVE ISOLATION NUMBER NUMBER FUNCTION TIME (seconds) i 49 DB87 Refueling Canal Fill Line N/A l

49 D888 Refueling Canal Fill Line N/A 1

50 9 HP48 High Pressure Injection Line N/A 50 $

HP2C High Pressure Injection Line N/A j

50 MU6421 RCS Makeup Line N/A

)

l I

52 MU242 RCP Seal Vater Supply N/A 53 MU243 RCP Seal Vater Supply N/A 54 MU244 RCP Seal Vater Supply N/A 55 MU245

.RCP Seal Vater Supply N/A.

59 Plange Secondary Side Cleaning (Inside Containment)

N/A 59 Plange Secondary Side Cleaning (Outside Containment)

N/A 57 i MS603 Steam Generator Blowdown Line N/A

-]

60 t MS611 Steam Generator Blowdown Line N/A 67 CV209 Bydrogen Dilution System Supply N/A 69 CV210 Hydrogen Dilution System Supply N/A 71A 9 CV20008 Containment Pressure Sensor N/A 71C CP16 Core Plood Tank Nitrogen Pill Line N/A 72A #

CV2001B Containment Pressure Sensor N/A 72C #

CV624B Containment Pressure Differential l

Transmitter N/A l

73A #

CV2002B Containment Pressure Sensor N/A 73C #

CV645B Containment Pressure Differential Transmitter N/A 74A 4 CV2003B Containment Pressure Sensor N/A 074C DB2735 Pressurizer Auxiliary Spray N/A

74C DH2736 Pressurizer Auxiliary Spray N/A

May be opened on an intermittent basis under administrative control.

GNot subject to Type' C leakage tests.

I

- Surveillance testing not required prior to entering MODE 4 but shall be i

I performed prior to entering MODE 3.

flProvisions of Specification 3.0.4 a e not applicable provided the valve is in the closed position and deactiva?rd.

I DAVIS-BESSE, UNIT 1 3/4 6-22 Amendment No. 3, 31, 18, 111,114,111, 135

l.

1 CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 2.

Verifying that the system satisfies the in-place testing acceptance criteria and uses the test procedures of Regulatory Positions C.S.a, C.S.c and C.S.d of Regulatory Guide 1.52, Revision 1, July 1976, and the system flow rate is 8,000 cfm t10%.

3.

Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 1, July 1976, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 1, July 1976.*

4.

Verifying a system flow rate of 8,000 cfm t10% during system operation when tested in accordance with ANSI N510-1975.

After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation by verifying c.

within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 1. July 1976, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 1, July 1976.*

d.

At least once per 18 months by:

1.

Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is < 6 inches Vater Gauge while operating the system at a flow rate of 8,000 cfm i 10%.

2.

Verifying that the system starts automatically on any containment isolation test signal.

I l

3.

Verifying that the filter cooling bypass valves can be manually opened.

I

Representative samples of used activated carbon from the EVS shall pass the laboratory test given in Table 3 for an activated carbon bed depth of 2 inches (i.e., the two 2 inch filter beds in series shall be tested per Test 5.b in Table 2 at a relative humidity of 70% for a methyl iodide penetration of less than 1%). The pre-and post-loading sweep medium temperature shall be 80'c for Test 5.b of Table 2, Regulatory Guide 1.52, Revision 1, July 1976.

DAVIS-B SSE, UNIT 1 3/4 6-29 Amendment No. A8,135

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

)

4.

Verifying that each system produces a negative pressure of > 0.25 inches Vater Gauge in the annulus within 4 seconds after the fan l

attains a flow rate of 8000 cfm i 10%. This test is to be performed with the flow path established prior to starting the EVS fan, and the other dampers associated with the negative pressure boundary closed.

e.

After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove 1 99% of the DOP when they l

are tested in-place in accordance with ANSI N510-1975 while operating i

the system at a flow rate of 8000 cfm i 10%.

f.

After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove 2 99% of a halogenated l

hydrocarbon refrigerant test gas when they are tested in-place in I

accordance with ANSI N510-1975 while operating the system at a flov l

rate of 8000 cfm 10%.

l i

l F

i i

0 DAVIS-BESSE, UNIT 1 3/4 6-30 Amendment No. /(135 E_____.._.._____.______

PLANT SYSTEMS 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION LIMITING CONDITION FOR OPERATION 3.7.2.1 The temperature of the secondary coolant in the steam generators shall be > 110'F vhen the pressure of the secondary coolant in the steam generator is

> 237 psig.

APPLICABILITY: At all times.

ACTION:

With the requirements.of the above specification not satisfied:

a.

Reduce the steam generator pressure to < 237 psig within 30 minutes, and b.

Perform an engineering evaluation to determine the effect of overpressurization on the structural integrity of the steam generator.

Determine that the steam generator remains acceptable for continued operation prior to increasing its pressure above 237 psig.

l l

l SURVEILLANCE REQUIREMENTS 4.7.2.1 The temperature of the secondary coolant in each steam generator shall l

be determined to be > 110'F at least once per hour when secondary pressure in i

the steam generator is > 237 psig and T is < 200'F.

ave DAVIS-BESSE, UNIT 1 3/4 7-13 Amendment No. 135

_____.__.__.-______m_

PLANT SYSTEMS l

3/4.7.3 COMPONENT COOLING WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.3.1 Two independent component cooling water loops shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With one component cooling water loop inoperable, restore the inoperable loop to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.7.3.1 Each component cooling water loop shall be demonstrated OPERABLE:

At least once per 31 days by verifying that each valve (manual, a.

power operated or automatic) servicing safety related equipment that is not locked, sealed or otherwise secured in position, is ira its correct position.

b.

At least once per 18 months, during shutdown, by:

1.

Verifying that each automatic valve in the flow path actuates to its correct position on an SFAS test signal.

2.

Verifying that each component cooling water emergency pump starts automatically on an SFAS test signal.

I l

DAVIS-BESSE, UNIT 1 3/4 7-14

t i

PLANT SYSTEMS 3/4.7.6 CONTROL ROOM EMERGENCY VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.7.6.1 Two independent control room emergency ventilation systems shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With one control room emergency ventilation system inoperable, restore the inoperable system to OPERABLE status within 7 days or be in at least BOT STANDBY vithin the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOVN within.the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.7.6.1 Each control room emergency ventilation system shall be demonstrated OPERABLE:

a.

At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that the control room air l

temperature is < 110'F vhen the control room emergency ventilation system is operating.

b.

At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the system operates for at least 15 minutes.

c'.

At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the system by:

DAVIS-BESSE, UNIT 1 3/4 7-17 Amendment No.135

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 1.

Verifying that with the system operating st a flow rate of 3300 cfm i 10% and exhausting through the BEPA filters and charcoal adsorbers, the total bypass flow of the system is f 1% when the system is tested by admitting DOP at the system intake.

2.

Verifying that the system satisfies the in-place testing acceptance criteria and uses the test procedures of Regulatory Positions C.5.a, C.5.c and C.5.d of Regulatory Guide 1.52, Revision 1, July 1976, and the system flow rate is 3300 efs t10%.

3., Verifying within 31 days after removal that a laboratory analysis f of a representative carbon sample obtained in accordance with l

Regulatory Position.C.6.b of Regulatory Guide 1.52, Revision 1, July 1976, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 1, July 1976.*

4.

Verifying a system flow rate of 3300 cfm 110% during system operation when tested in accordance with ANSI N510-1975.

d.

After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 1, July 1976, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 1, July 1976.

At least once per 18 months by e.

1.

Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is < 4.4 inches Vater Gauge while l

l operating the system at a flow rate of 3300 cfm t 10%.

I 2.

Verifying that the control room normal ventilation system is isolated by a SFAS test signal, Control Roon Ventilation Air Intake chlorine Concentration - High test signal, and a Station Vent Radiation High test signal.

The pre-and post-loading sweep medium temperature shall be 80'C for Test 5.b' of Table 2, Regulatory Guide 1.52, Revision 1, July 1976.

DAVIS-BESSE, UNIT 1 3/4 7-18 Amendment No.135 4

PLANT SYSTEMS l

SURVEILLANCE REQUIREMENTS (Continued) 1.

With a half-life greater than 30 days (excluding Hydrogen

3) and 2.

In any form other than gas.

b.

Stored sources not in use - Each sealed source and fission detector shall be tested prior to use or transfer to another.

licensee unless tested within the previous six months. Sealed i) sources and fission detectors transferred without a certificate indicating the last test date shall be tested prior to being placed into use.

Startup sources and fission detectors - Each sealed startup c.

source and fission detector shall be tested within 31 days prior to being subjected to core flux or installed in the core and following repair or maintenance to the source.

4.7.8.1.3 Reports - A report shall be prepared and submitted to the Commission on an annual basis if sealed source or fission detector leakage tests reveal the presence of > 0.005 microcuries of removable contamination.

I DAVIS-BESSE, UNIT 1 3/4 7-37

_ _ = _.

)

i j

PLANT SYSTEMS I

3/4.7.9 FIRE SUPPRESSION SYSTEMS

~

FIRE SUPPRESSION WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.9.1 The fire suppression water system shall be OPERABLE with:

a.

Two high pressure pumps, each with a capacity of 2500 gpm, with i

their discharge aligned to the fire suppression header, b.

Separate water supplies, each with a minimum contained volume of 250,000 gallons, and c.

An OPERABLE flow path capable of taking suction from the Intake Forebay and the Fire Water Storage Tank and transferring the water through distribution piping with OPERABLE sectionalizing l

control or isolation valves to the yard hydrant curb valves and

{

the first valve ahead of the water flow alarm device on each l

sprinkler hose standpipe or spray system riser required to be i

GPERABLE per Specifications 3.7.9.2 and 3.7.9.3.

APPLICABILITY: At all times.

I ACTION:

a.

With one pump and/or one water supply inoperable, restore the I

inoperable equipment to OPERABLE status within 7 days or prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 30 days outlining the plans and procedures to be used to provide for the loss of redundancy in this system. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

b.

With the fire suppression water system otherwise inoperable:

1.

Establish a backup fire suppression water system within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and 2.

Submit a Special Report in accordance with Specification 6.9.2:

a) By telephone within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b) Confirmed by telegraph, mailgram or facsimile tians-mission no later than the first working day following the event, and DAVIS-BESSE, UNIT 1 3/4 7-38 Amendment No. 7,93.J0J,135 l

11 PLANT SYSTEMS ll LIMITING CONDITION FOR OPERATION (Continued) c)

In writing within 14 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

SURVEILLANCE REQUIREMENTS 1

4.7.9.1.1 The fire suppression water system shall be demonstrated OPERABLE:

At least once per 7 days by verifying the contained water supply a.

volume.

b.

At least once per 31 days on a STAGGERED TEST BASIS by starting each pump and operating it for at least 15 minutes on recircu-1ation flow.

+

l At least once per 31 days by verifying' that each valve (manual.

I c.

power operated or automatic) in the flow path that is accessible is in its correct position.

d.

At least once per 12 months by performance of a system flush.

e.

At least once per 12 months by cycling each testable valve in the flow path through at least one complete cycle of full travel.

f.

At least once per 18 months by performing a system functional test which includes simulated automatic actuation of the system throughout its operating sequence, and:

1.

Verifying that each automatic valve in the flow path actuates to its correct position.

l 2.

Verifying that each pump develops at least 2375 gpm at a system head of 250 feet.

3.

Cycling each pump valve in the flow path that is not testable during plant operation through at least one complete cycle of travell.and~

l 4.

Verifying that each high pressure pump starts (sequentially) to maintain the fire suppression water system pressure

> 95 psig.

l DAVIS-BESSE, UNIT 1 3/4 7-39 Amendment No. 9,135 l

f PLANT SYSTEMS l

SURVEILLANCE REQUIREMENTS (Continued)

I g.

At least once per 3 years by performing a flow test 'of the I

~

system in accordance with Chapter 5. Section 11 of the Fire Protection Handbook,14th Edition, published by the National Fire Protection Association.

4.7.9.1.2 The fire pump diesel engine shall be demonstrated OPERABLE.

a..

At least once per 31 days by verifying; 1.

The Fire Pump Diesel Day Tank contains at least 300 gallons of fuel, and s

2.

The diesel starts from ambient conditions and operates for at least 20 minutes.~

b.

At least once per 92 days by verifying that a sample of diesel fuel from the fuel storage tank, obtained in accordance with ASTM-D270-65, is within the acceptable limits specified irr t

Table 1 of ASTM 0975-74 when checked for viscosity, water content, and sediment.

c.

At least once per 18 months, during shutdown, by:

1.

Subject-ing the diesel to an inspection in accordance with procedures prepared in conjuction with its manufacturer's l

recommendations for the class of service, and

)

2.

Verifying the diesel starts from ambient conditions on the auto-start signal and operates for > 20 minutes while loaded

)

with the fire pump.

4.7.9.1.3 The fire pump diesel starting 24 volt battery bank and charger shall be demonstrated OPERABLE:

a.

At least once per 7 days by verifying that:

1.

The electrolyte level of each cell is above the plates, and 2.

The overall battery voltage when not discharging is > 24 vol ts.

DAVIS-BESSE, UNIT 1 3/4 7-40 Amendment No. 9

PLANT SYSTEMS f

(

SURVEILLANCE REQUIREMENTS (Continued) b)

Cycling each valve in the flow path that is not testable during plant oneration through at least one complete cycle of full travel.

2.

By inspection of each nozzle to verify no blockage.

I 1

I I

l I

DAVIS-BESSE, UNIT 1 3/4 7-43 Amendment No. 9

1 PLANT SYSTEMS FIRE HOSE STATIONS LIMITING CONDITION FOR OPERATION 3.7.9.3 The fire hose stations shown in Table 3.7-4 shall be OPERABLE.

APPLICABILITY: Whenever equipment in the areas protected by the fire I

hose stations is required to be OPERABLE.

j ACTION:

a.

With'one or more of the fire hose stations in Table 3.7-4 inoperable, route an additional equivalent capacity fire hose to the unprotected area (s) from an OPERABLE hose station within l

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

b.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS l

4.7.9.3 Each of the fire hose stations shown in Table 3.7-4 shall be demonstrated OPERABLE:

a.

At least once per 31 days by visual inspection of the station

)

to assure all required equipment is at the station.

j l

b.

At least once per 18 months by:

j 1.

Removing the hose for inspection and re-racking, and j

j 2.

Replacement of all degraded gaskets in couplings.

c.

At least once per 3 years by:

1.

Partially opening each hose station valve to verify valve OPERABILITY and no flow blockage.

l 2.

Conducting a hose hydrostatic test at a pressure at least 50 psig greater than the maximum pressure avail-i able at that hose station.

t l

DAVIS _BESSE, UNIT 1 3/4 7-44 Amendment No. 9.135 4

e a

_._m.___._

_. _ _ _ _ _ _ ____________ _ ____.J

i REFUELING OPERATIONS

[

C0!NUNICATIONS LIMITING CONDITION FOR OPERATION P

3.9.5 Direct communications shall be. maintained between the control-room and personnel at the refueling station.

APPLICABILITY:

During CORE ALTERATIONS.

ACTION:

When direct communications between the control room and personnel at the refueling station cannot be maintained, suspend all CORE ALTERATIONS.

The provisions of Specification 3.0.3 are not applicable.

t SURVEILLANCE REQUIREMENTS 4.9.5 Direct communications between the control room and personnel at the refueling station shall be demonstrated within one hour prior to the start of and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during CORE ALTERATIONS.

l l

DAVIS-BESSE, UNIT 1 3/4 9-5 l

REFUELING OPERATIONS FUEL HANDLING BRIDGE OPERABILITY LIMITING CONDITION FOR OPERATION 1

3.9.6 The control rod hoist and fuel assembly hoist of the fuel handling bridge shall be used for movement of control rods or fuel assemblies and shall be OPERABLE with:

a.

The control rod hoist having:

1.

A minimum capacity of 3000 pounds, and

'.- m 2.

An overload cutoff limit of 1 2650 pounds.

b.

The fuel assembly hoist having:

1.

A minimum capacity of 3000 pounds, and 2.

An overload cutoff limit of 1 2700 pounds.

APPLICABILITY:

During movement of control rods or fuel assemblies within the reactor pressure vessel.

ACTION:

With the requirements for control rod hoist and/or fuel assembly hoist OPERABILITY not satisfied, suspend use of any inoperable control rod hoist and/or fuel assembly hoist from operations involving the movement of control rods or fuel assemblies within the reactor pressure vessel. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.6.1 Each control rod hoist used for movement of control rods or fuel assemblies within the reactor pressure vessel shall be demonstrated OPERABLE within 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to the start of such operations by performing a hoist load test of at least 3000 pounds and demonstrating an automatic load cutoff when the control rod hoist load exceeds 2650 pounds.

1 4.9.6.2 Each fuel assembly hoist used for movement of control rods or fuel l

assemblies within the reactor pressure vessel shall be demonstrated OPERABLE within 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to the start of such operations by performing a load test of at least 3000 pounds and demonstrating an automatic load cutoff when the fuel assembly hoist load exceeds 2700 pounds.

I DAVIS-BESSE. UNIT 1 3/4 9-6 Amendment No.135

j

!1 REFUELING OPERATIONS C

U,

! O STORAGE POOL WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.11 As a minimum, 23 feet of water shall be maintained over the top of irradiated fuel assemblies seated in the storage racks.

APPLICABILITY: Whenever irradiated fuel assemblies are in the storage pool.

ACTION:

With the requirement.of the specification not satisfied, suspend.all movement of fuel and crane operations with loads in the fuel storage area and restore the water level to within its limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

The provisions of Specification 3.0.3 are not applicable.

l' t

\\.

~~

SURVEILLANCE REQUIREMENTS 4.9.11 The water level in the storage pool shall be determined to be at least its minimum required depth at least once per 7 days when irradiated fuel assemblies are in the fuel storage pool.

i l

l l

I t

DAVIS-BESSE, UNIT 1 3/4 9 11 1-

______:.__2

REFUELING OPERATIONS STORAGE POOL VENTILATION LIMITING CONDITION FOR OPERATION 3.9.12 Two independent emergency ventilation systems servicing the storage pool area shall be OPERABLE.

APPLICABILITY: Whenever irradiated fuel is in the storage pool.

1 ACTION:

Vith one emergency ventilation system servicing the storage pool area a.

inoperable, fuel movement within the storage pool or crane operation with loads over the storage pool may proceed provided the OPERABLE emergency.

ventilation system servicing the~ storage pool area is in operation and discharging through at least one train of HEPA filters and charcoal adsorbers.

b.

With no emergency ventilation system servicing the storage pool area OPERABLE, suspend all operations involving movement of fuel within the storage pool or crane operation with loads over the storage pool until at least one system is restored to OPERABLE status.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

c.

SURVEILLANCE REQUIREMENTS 4.9.12.1 The above required emergency ventilation system servicing the storage pool area shall be demonstrated OPERABLE per the applicable Surveillance Requirements of 4.6.5.1, and at least once per 18 months by verifying that the emergency ventilation system servicing the storage pool area maintains the storage pool area at a negative pressure of > 1/8 inches Vater Gauge relative to the outside atmosphere during system operation.

l 4.9.12.2 The normal storage pool ventilation system shall be demonstrated OPERABLE at least once per 18 months by verifying that the system fans stop automatically and that dampers automatically divert flow into the emergency ventilation system on a fuel storage area high radiation test signal.

l l

Davis-Besse, Unit 1 3/4 9-12 Amendment No.135 1

I7 l

SPECIAL TEST EXCEPTION REACTOR COOLANT LOOPS LIMITING CONDITION POR OPERATION I

3.10.3 The limitations of Specification 3.4.1 may be suspended during the performance of STARTUP and PHYSICS TESTS provided:

l a.

The THERMAL POWER does not exceed 5% of RATED THERMAL POWER, and b.

The reactor trip setpoints on the OPERABLE High Flux channels are set

< 25% of RATED THERMAL POWER.

APPLICABILITY: MODE 2.

ACTION:

With the THERMAL POWER greater than 5% of RATED THERMAL POWER, immediately open the control rod drive trip breakers.

SURVEILLANCE REQUIREMENTS 4.10.3.1 The THERMAL POWER shall be determined to be < 5% of RATED THERMAL POVER at least once per hour during STARTUP and PHYSIC 5 TESTS.

l 4.10.3.2. Each High Flux Channel shall be subjected to a CHANNEL, FUNCTIONAL TEST vithin 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months prior to initiating startup or PHYSICS TESTS.

DAVIS-BESSE, UNIT 1 3/4 10-3 Amendment No.135

SPECIAL TEST EXCEPTIONS SHUTDOWN MARGI_N LIMITING CONDITION FOR OPERATION 3,10.4 The SHUTDOWN MARGIN requiregnt of Specification 3.1.1.1 may be l

uspended for measurement of control rod worth and shutdown margin' provided:

Reactivity equivalent to at least the highest estimated a.

control rod worth is available for trip insertion from OPER-i ABLE control rod (s), and I

b.

All axial power shaping rods are withdrawn to at least 35%

.(indicated position) a.nd OPERABLE.

APPLICABILITY: MODE 2.

i ACTION:

l i

With any safety or regulating control rod not fully inserted-i a.

and with less than the above reactivity equivalent available 1

for trip insertion or the axial power shaping rods not within their withdrawal limi.ts, imedie+.ely initiate and continue boration'at > 18 gpm of 7875 ppm boric acid solution or its equivalen_t until the SHUTDOWN MARGIN required by Specifi-cation 3.1.1.1 is restored.

i b.

With all safety or regulating control rods fully inserted and the reactor subcritical by less than the above reactivity equivalent, immediately initiate had continue boration at >-

18 gpm of 7875 ppm boric acid solution or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.

I SURVEILLANCE REQUIREMENTS

]

i d,

4.10.4.1 The position of each safety, regulating, and axial power shaping rod either partially or fully withdrawn shall be determined at i

least once per 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

4.10.4.2 Each safety or regulating control rod not fully inserted shall l

be demonstrated capable of full insertion when tripped from at least the 50% withdrawn position within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to reducing the SHUTDOWN iARGIN to less than the limits of Specification 3.1.1 1.

i

,0 AVIS-BESSE, UNIT 1 3/4 10-4

=_

d RADI0 ACTIVE EFFLUENTS

.3/t 11.2 GASEOUS EFFLUENTS DOSE RATE i

LIMITING CONDITION FOR OPERATION, 3.11.2.1' : The dose rate due to radioactive materials released in gaseous effluents from the site to areas at and beyond the SITE BOUNDARY (see Figure 3.11-2) shall be limited to the following:

a.

For noble gasest Less'than or equal'to 500 mrems/ year to the total l

body and less than or equal to 3000 arems/ year to the skin, and b.

For iodine-131, for tritium, and for all radionuclides in particulate-form with half lives greater than 8 dayss. Less than or equal to 1500 mrems/yr to any organ.

APPLICABILITY: At all times.

ACTION:

With the dose rate (s) exceeding the above limits, without delay restore l

a.

the release rate to within the above limit (s).

b.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.1.1 The dose rate due to noble gases in gaseous effluents shall be determined to be within the above limits in accordance with the methodology and parameters in the ODCM.

4.11.2.1.2 The dose rate due to iodine-131, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents shall' I

be determined to be within the above limits in accordance with the methodology 3

and parameters in the ODCM by obtaining representative samples and performing l

analyses in accordance with the sampling and analysis program specified in 1

Table 4.11-2.

l i

DAVIS-BESSE, UNIT 1 3/4 11-9 Amendment No. p%,135

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3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAMS LIMITING CONDITION FOR OPERATION l

3.12.1 The radiological environmental monitoring program shall be conducted as specified in Table 3.12-1.

APPLICABILITY: At all times.

ACTION:

a.

With the radiological environmental monitoring program not being conducted as specified in Table 3.12-1, in lieu of a Licensee Event Report, prepare and submit to the Commission, in the Annual Radiological Environmental Operating Report required by Specification 6.9.1.10, a l

description of the reasons for not conducting the program as required and the plans for preventing a recurrence, b.

With the level of radioactivity as the result of plant effluents in an environmental sampling medium at a specified location exceeding the reporting levels of Table 3.12-2 when averaged over any calendar quarter, in lieu of a Licensee Event Report, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit (s) and defines the corrective actions to be taken to reduce radioactive effluents so that the potential annual dose to A MEMBER OF THE PUBLIC is less than the calendar year limits of Specification 3.11.1.2, 3.11.2.2, and 3.11.2.3.

When more than one of the radionuclides in Table 3.12-2 are detected in the sampling medium, this report shall be submitted if:

concentration (1) concentration (2) reporting level (1). reporting level (2) +... -> 1.0 When radionuclides other than those in Table 3.12-2 are detected and.are the result of plant effluents, this report shall be submitted if the potential annual dose to A MEMBER OF THE PUBLIC is equal to or greater than the calendar year limits of Specifications 3.11.1.2, 3.11.2.2 and 3.11.2.3.

This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report.

c.

Vith milk or fresh leafy vegetable samples unavailable from one or more of the sample locations required by Table 3.12-1, identify locations for obtaining replacement samples and if practical add them to the j

radiological environmental monitoring program within 30 days. The j

locations from which samples were unavailable may then be deleted from the monitoring program. In lieu of a Licensee Event Report and pursuant to Specification 6.9.1.11, identify the cause of the unavailability of samples and identify the new locations (s) for obtaining replacement samples in the next Semiannual Radioactive Effluent Release Report and also include in the report a gevised figure (s) and table for the ODCM reflecting the new locations (h).

DAVIS-BESSE, UNIT 1 3/4 12-1 Amendment No. SS,135 i

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RADIOLOGICAL ENVIRONMENTAL MONITORING LIMITING CONDITION FOR OPERATION l

ACTION: (Continued) l l

d.

With specimens unobtainable due to hazardous conditions, seasonal unavailability, malfunction of automatic sampling equipment and~

other legitimate reasons, every effort will be made to complete corrective action prior to the end of the next sampling period.

All deviations from the sampling schedule will be documented in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.10.

l e.

The provisions of Specifications 3.0.3 and 3.0.4 are not.

applicable.

SURVEILLANCE REQUIREMENTS 4.12.1.1 The radiological environmental monitoring samples shall be collected

)

pursuant to Table 3.12-1 from the specific' locations given in the table and j

figure (s) in the ODCM and shall be analyzed pursuant to the requirements of i

Table 3.12-1, and the detection capabilities required by Table 4.12-1.

j!

4.12.1.2 Cumulative potential dose contributions for the current calendar year from radionuclides detected in environmental samples shall be determined 1

in accordance with the methodology and parameters in the ODCM.

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DAVIS-BESSE, UNIT 1 3/4 12-2 Amendment No. E6,135

- - _ _ _ _ _ _ _ = _ _ - _ _ _ _ _ _ _ - _ _ _ - _ _ _

i 1

1 APPLICABILITY _

BASES other specified conditions are satisfied.

In this case, this would mean that i

for one division the emergency power source must be OPERABLE (as must be the j

components supplied by the emergency power source) and all redundant systems, subsystems, trains, components and devices in the other_ division must be OPERABLE, or likewise satisfy Specification 3.0.5 (i.e., be capable of per-forming their design functions and have an. emergency power source OPERABLE).

In other words, both emergency power sources must be OPERABLE and all r j

redundant systems, subsystems, trains, components and devices in both divisions must also be OPERABLE.

If these conditions are not satisfied, action is required in accordance with this specification.

In MODES 5 or 6, Specification 3.0.5 is not applicable, and thus the individual ACTION statements for each applicable Limiting Condition for Operation in these MODES must be adhered to.

I DAVIS-BESSE, UNIT 1 B 3/4 0-lb Amendment No. 77,135 l

- ________ _ _ _ _ _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ - _ _ _ - _ _ _ = _ - - _.

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3/4.3 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 REACTOR PROTECTION SYSTEM AND SAFETY SYSTEM INSTRUMENTATION (Continued) q l

\\

l Safety-grade anticipatory reactor trip is initiated by a turbine trip (above 45 l

percent of RATED THERMAL POVER) or trip of both main feedvater pump turbines.

This anticipatory trip will operate in advance of the reactor coolant system high pressure reactor trip to reduce the peak reactor coolant system pressure and thus reduce challenges to the pilot operated relief valve. This l

anticipatory reactor trip system was installed to satisfy Item II.K.2.10 of

(,

NUREG-0737. The justification for the ARTS turbine trip arming level of 45% is

~

given in BAV-1893, October, 1985.

J' DAVIS-BESSE, UNIT 1 B 3/4 3-la Amendment No. 72.

775, ygg, 135

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]

3/4.4 REACTOR COOLANT SYSTEM BASES 1

inject water into the reactor coolant system is. disabled to-ensure operation within reactor coolant system pressure-temperature limits.

j Demonstration of the safety valves' lift settings vill occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Code.

i The pressurizer code safety valves must be set such that the peak Reactor Coolant System pressure does not exceed 110% of design system pressure (2500

{

psig) or, 2750 psig. The control rod group withdrawal accident will result in

}

the most limiting high pressure in the RCS. The analysis assumes RPS high pressure trip at 2355 psig and the code safety valves open at 2500 psig. The tolerance on the RPS instrument accuracy is 30 psi and, it is +1% for the code safety valve settings. The pressurizer pilot operated relief valve was assumed l~

not to open for this transient. The resulting system peak pressure was calculated to be 2700 psig. Therefore, the code safety valve setpoint is conservatively set at 3 2525 psig which is the maximum pressure of 2500 psig +1%

for tolerance.

The pressurizer pilot operated relief valve should be set such that it will open l

before the code safety valves are opened.

However, it should not.open on any anticipated transients.

BAV-1890, September 1985 identified that the turbine trip from full power would cause the largest overpressure transient. This report demonstrated that with a RPS high pressure trip setpoint of 2355 psig the i

resulting overshoot in RCS pressure vould be limited to 50 psi.

Consequently, the minimum PORV setpoint needs to accommodate both the RCS pressure overshoot and the RPS instrument string error of 30 psi.

k DAVIS-BESSE, UNIT 1 B 3/4 4-la Amendment No. 33,Ah N24 135

i REACTOR COOLANT SYSTEM BASES 3/4.4.4 PRESSURIZER A steam bubble in the pressurizer ensures that the RCS is not a hydraulically solid system and is capable of accommodating pressure surges during operation.

The steam bubble also protects the pressurizer code safety valves and pilot operated relief valve against water relief.

The low level limit is based on providing enough water volume to prevent a reactor coolant system low pressure condition that vould actuate the Reactor Protection System or the Safety Feature Actuation System. The high level limit is based on providing enough steam volume to prevent a pressurizer high' level'as a result of any transient.

The pilot operated relief valve and steam bubble function to relieve RCS pressure during all design transients. Operation of the pilot operated relief valve minimizes the undesirable opening of the spring-loaded pressurizer code safety valves.

3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained.

The program for inservice inspection of steam generator tubes is based'on a modification of Regulatory Guide 1.83, Revision 1.

Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion. Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.

The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes. If the secondary coolant chemistry is not maintained within these chemistry limits, localized corrosion may likely result in stress corrosion cracking. The extent of cracking during plant operation vould be limited by the limitation of steam generator tube leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage - 1 GPM).

Cracks having a primary-to-secondary leakage less than this limit during operation vill have an adequate margin of safety to withstand the loads l

DAVIS-BESSE, UNIT 1 B 3/4 4-2 Amendment No.135 t_ _

m__-

____m_________

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3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 CONTAINMENT INTEGRITY Primary CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakao paths and associated leak rates assumed in the safety analyses.

Thir. restriction, in conjunction with the leakage rate limitation, will limit the site boundary radiation doses to within the limits of 10 CFR 100 during accident conditions.

3/4.6.1.2 CONTAINMENT LEAKAGE The limitations on containment leakage rates ensure that the total containment leakage volume will not exceed the value assumed in the safety analyses at the peak accident pressure of 38 psig, P. As an added conservatism, the measured overall integrated leakage ra'te is a

further limited to < 0.75 L, during perfonnance of the periodic tests a

to account for possible degradation of the containment leakage barriers between leakage tests.

The surveillance testing for measuring leakage rates are consistent with the requirements of Appendix "J" of 10 CFR 50.

t The special test for the containment purge and exhaust isolation t

valves is intended to detect gross degradation of seals on the valve seats.

The special test is performed in addition to the Aopendix J requirements.

3/4.6.1.3 CONTAINMENT AIR LOCKS The limitations on closure and leak rate for the containment air locks are required to meet the restrictions on CONTAINMENT INTEGRITY and containment leak rate. Surveillance testing of the af-lock seals I

provide assurance that the overall air lock leakage v til not become excessive due to seal damage during the intervals brtween air lock leakage tests.

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DAVIS-BESSE, UNIT 1 B 3/4 6-1 Amendment No. 90 1

CONTAINMENT SYSTEMS BASES 3/4.6.1.4 INTERNAL PRESSURE The limitations on containment internal pressure ensure that 1) the containment structure is prevented from exceeding its design negative pressure differential with respect to the annulus atmosphere of 0.5 psi and 2) the containment peak pressure does not exceed the design pressure of 40 psig during LOCA conditions.

The maximum peak pressure obtained from a LOCA event is 37 psig.

The limit of 1 psig for initial positive containment pressure will limit the total pressure to 38 psig which is less than the design pressure and is consistent with the safety analyses.

3/4.6.1.5 AIR TEMPERATURE The limitations on containment average air temperature ensure that j

the overall containment average air temperature does not exceed the initial temperature condition assumed in the accident analysis for a LOCA.

3/4.6.1.6 CONTAINMENT VESSEL STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the contain-ment steel vessel will be maintained comparable to the original design standards for the life of the facility.

Structural integrity is required to ensure that the vessel will withstand the maximum pressure of 38 psig in the event of a LOCA. A visual inspection in conjunction with Type A leskage tests is sufficient to demonstrate this capability.

3/4. 6.1. 7 CONTAINMENT VENTILATION SYSTEM The limitation on use of the Containment Purge and Exhaust System limits the time this system may be in operation with the reactor coolant system temperature above 200*F. This restriction minimizes the time that a direct open path would exist from the containment atmosphere to the outside atmosphere and consequently reduces the probability that an accident dose would exceed 10 CFR 100 guideline values in the event of a LOCA occurring coincident with purge system operation. The use of this system is therefore restricted to non-routine usage not to, exceed 90 I

hours in any consecutive 365 day period which is equivalent to approximately 1% of the total possible yearly unit operating time.

3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS 3/4.6.2.1 CONTAINMENT SPRAY SYSTEM The OPERABILITY of the containment spray system ensures that contain-ment depressurization and cooling capability will be available in the event of a LOCA. The pressure reduction and resultant lower containment DAVIS-BESSE, UNIT 1 E 3/4 6-2 Amendment No.135

i s

CONTAINMENT SYSTEMS 1

i BASES leakage rate are consistent with the assumptions used in the safety analyses.

The leak rate surveillance requirements assure that the leakage assumed for the system during the recirculation phase vill not be exceeded.

3/4.6.2.2 CONTAINMENT COOLING SYSTEM The OPERABILITY of the containment cooling system ensures that 1) the containment air temperature vill be maintained within limits during normal operation, and 2) adequate heat removal capacity is available when operated in conjunction with the containment spray systems during post-LOCA conditions.

3/4.6.3 CONTAINMENT ISOLATION VALVES The OPERABILITY of the containment isolation valves ensures that the containment atmosphere vill be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressurization of the containment. Containment isolation within the time limits specified ensures that the release of radioactive material to the environment vill be consistent with the assumptions used in the analyses for a LOCA.

DAVIS-BESSE, UNIT 1 B 3/4 6-3 Amendment No.155 l

{l CONTAINMENT SYSTEMS BASES 3/4.6.4 COMBUSTIBLE GAS CONTROL l

1 The OPERABILITY of the Hydrogen Analyzers, containment Hydrogen Dilution System and Hydrogen Purge System ensures that this equipment I

will be available t6 maintain the maxiinum hydrogen concentration within the containmen't vessel at or Telow three volume percent following a LOCA.

I The two redundant Hydrogen Analyzers determine the content of hydro-gen within the containment vessel.

The Containment Hydrogen Dilution (CHD) System consists of two full capacity, redundant, rotary, positive displacement type blowers to supply -

1 air to the containment. The CHD System controls the hydrogen concentra-l tion by the addition of air to the containment vessel, resulting in a pressurization of the containment and suppression of the hydrogen volume

fraction, i

The Containment Hydrogen Purge System Filter Unit functions as a backup to the CHD System and is designed to release air from the con-tainment atmosphere through a HEPA filter and charcoal filter prior to discharge to the station vent.

3/4.6.5 SHIELD BUILDING I

3/4.6.5.1 EMERGENCY VENTILATION SYSTEM i

The OPERABILITY of the emergency ven.tilation systems ensures that containment vessel leakage occurring during LOCA conditions ~ into the annulus will be filtered through the HEPA filters and charcoal adsorber trains prior to discharge to the atmosphere. This requirement is neces-sary to meet the assumptions used in the safety analyses and limit the site boundary radiation doses to within the limits of 10 CFR 100 during LOCA conditions.

CAVIS-EESSE, UNIT 1 B 3/4 6-4 Amendment No. 56

___.._.__________________-___.__________________w

s CONTAINMENT SYSTEMS BASES 3/4.6.5.2 SHIELD BUILDING INTEGRITY SHIELDING BUILDING INTEGRITY ensures that the release of radioactive material from the containment vessel vill be restricted to those leakage paths and associated leak rates assumed in the safety analysis. The closure of the airtight doors and blevout panels listed in Table 4.6-1 ensure that the Emergency Ventilation System (EVS) can provide a negative pressure between 0.25 and 1.5 inches Vater Gauge within the annulus between the shield building and l

containment vessel and within the interconnecting mechanical penetration rooms after a loss-of-coolant accident (LOCA). This restriction, in conjunction with l

the operation of the EVS, vill limit the site boundary radiation doses to within the limits of 10 CFR 100 during accident conditions.

l I

3/4.6.5.3 SHIELD BUILDING STRUCTURAL INTEGRITY l

This limitation ensures that the structural integrity of the containment shield building vill be maintained comparable to the original design standards for the I

life of the facility.

Structural integrity is required to provide 1) protection for the steel vessel from external missiles, 2) radiation shielding in the vent of a LOCA, and 3) an annulus surrounding the steel vessel that can be maintained at a negative pressure during accident conditions.

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DAVIS-BESSE, UNIT 1 B 3/4 6-5 Amendment No.135

3/4.7 PLAhT SYSTEMS

.i BASES

'i

,1 3/4.7.1 -TURBINE CYCLE 3/4.7.1.1.-SAFETY VALVES The OPERABILITY of the main steam line code safety valves ensures.that the

'f secondarysystempressurevillbelimitedtowithin110%ofitsdesignpressure-l of 1050 psig during the most severe anticipated system operational transient.

The maximum relieving capacity is associated with a turbine trip from 100% RATED THERMAL POVER coincident with an assumed loss of condenser heet sink (i.e., no steam bypass to the condenser).

The safety valve set pressures and relieving capacities are in accordance with Section III of ASME Boiler and Pressure Vessel Code, 1971 Edition. The code.

requires the following:

1.

At least two pressure-relief valves are required to provide relieving capacity for steam systems.

2.

The capacity of the smallest pressure-relief valve shall not be less

~

than 50 percent of that of the largest pressure-relief' device.

3.

The set pressure of one of the pressure-relief _ devices shall not be greater than the maximum allovable working pressure _of~the system at design temperature.

l 4.

Total rated relieving capacity of the pressure-relief devices shall l

prevent a rise of more than 10 percent above system design pressure at l

design temperature under any pressure transients anticipated to arise.

These requirements are, respectively, met as follows:

1.

Nine safety valves are installed per steam generator.

i

(

2.

The relief capacity of two of the nine safety valves per steam generator is 583,574 lbs/hr each, and the capacity of the remaining seven is 845,759 lbs/hr each.

3.

A minimum of two OPERABLE safety valves per steam generator, with a combined total relief capacity of at least 1,167,148 lbs/hr, one with a setpoint not greater than 1050 psig (+/-l%), and one with a setpoint not greater than 1100 psig (+/-l%).

4.

The total relieving capacity of all safety.. valves on both main' steam lines is 14.175,000 lbs/hr which is 120 percent'of the total s'econdary-system flow of 11,760,000 lbs/hr at 100 percent of RATED THERMAL POWER-.

A maximum safety valve setpoint pressure of 1100 psig (+/-1%) assures main steam system pressure remains belov 110 percent, or 1155 psig.

DAVIS-BESSE, UNIT 1 B 3/4 7-1

. Amendment No'. pp, JJ/,

4 135 I

3/4.7 PLANT SYSTEMS BASES 3/4.7.1.1 SAFETY VALVES (Continued)

STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis of the reduction in secondary system steam flow and THERMAL POWER required by the reduced reactor trip settings of the High Flux channels. The reactor trip setpoint reductions are derived on the following bases:

(X) - (Y)(V) 3p, xy Z

where:

SP = reduced Trip Setpoint in percent of RATED THERMAL POWER l

(Not to Exceed W)

V = maximum number of inoperable safety valves per steam generator W = High Flux Trip Setpoint for four pump operation as specified in Table 2.2-1 X = Total relieving capacity of all safety valves per steam OO generator in 1bs/ hour, 7,087,500 lbs/ hour Y = Maximum relieving capacity of any one safety valve in 1bs/ hour, 845,759 lbs/ hour Z = Required relieving capacity per steam generator in 1bs/hr, l

6,585,600 lbs/hr.

3/4.7.1.2 AUXILIARY FEEDWATER SYSTEMS The OPERABILITY of the Auxiliary Feedwater Systems ensures that the Reactor Coolant System can be cooled down to less than 280 F from normal operating conditions in the event of a total loss of offsite power. The OPERABILITY of the Auxiliary Feed Pump Turbine Inlet Steam Pressure Interlocks is required only for high energy line break concerns and does not affect Auxiliary Feedwater System OPERABILITY.

Each steam driven auxiliary feedwater pump is capable of delivering a total feedwater flow of 600 gpm at a pressure of 1050 psig to the entrance of the steam generators. This capacity is sufficient to ensure that adequate feedwater flow is available to remove decay heat and reduce the Reactor Coolant System temperature to less than.280'F where the Decay Heat Removal System may be placed in operation.

1 (Y

DAVIS-BESSE, UNIT 1 B 3/4 7-la Amendment No. JJ7,J22, 131 g

Next page is B 3/4 7-2 l

1 i

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PLANT SYSTEMS j.

BASES positive reactivity effects of the Reactor Coolant System cooldown associated with the blowdown, and 2)-limit the pressure rise within containment in the event the steam line rupture occurs within containment. The OPERABILITY of the main steam isolation valves'within the closure times of the surveillance requirements are consistent with the assumptions used in the safety analyses.

3/4.7.1.6 SECONDARY WATER CHEMISTRY - Deleted 4

i 3/4.7.1.7 MOTOR DRIVEN FEEDWATER PUMP SYSTEM The OPERABILITY of the Motor Driven Feedwater Pump System entures that the Reactor Coolant System can be cooled down from normal operating conditions in the event of the total loss of Main Feedwater and Auxiliary Feedwater Pumps.

The Motor Driven Feedwater Pump flow capability ensures that adequate feedwater flow is available to remove Decay Heat and reduce the Reactor.

1 Coolant System temperature to where the Decay Heat System may be placed l

into operation.

When at 40% RATED THERMAL POWER or less and in MODES 1, 2, or 3, the Motor Driven Feedwater Pump System may be aligned to provide a flow path from the Deaerator Storage Tank through the Motor Driven Feedwater Pump to the Main Feedwater System.

During this Motor Driven Feedwater Pump mode of operation, a flow path from the Condensate Storage Tanks through the Motor Driven Feedwater Pump to the Auxiliary Feedwater System shall be maintained with the ability for manual positioning of valves such that DAVIS-BESSE, UNIT 1 B 3/4 7-3 Amendment No. J03, 135

__-___-__-.- - _ D

PLANT SYSTEMS

_ BASES the flow path can be established. The ability for local, manuel j

operation is demonstrated by verifying the presence of the handwheels for all manual valves and the presence of either handvheels or available power supply for actor operated valves.

_3/4.7.2 STEAM CENERATOR PRESSURE /TDGERATURE LIMITATION The limitation on steam generator pressure and temperature ensures that the pressure induced stresses in the steam generators do not exceed the maximum allowable fracture toughness stress limits. The limitations of Il0*T and 237 psig are based on a steam generator RT of 40*F and are sufficient to prevent brittle fracture.

NDT 3/4.7.3 COMPONENT COOLING WATER SYSTEM j

The OPERABILITY of the component cooling water system ensures that sufficient cooling capacity is available for continued operation of safety related equipment during normal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the safety analyses.

3/4.7.4 SERVICE WATER SYSTEM The OPERABILITY of the service veter system ensures that sufficient cooling espacity is available for continued operation of safety related equipment during normal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the safety analyses.

3/4.7.5 ULTIMATE REAT SIhT The limitations en the ultimate heat sink level and temperature ensure that sufficient cooling capacity is available to either 1) provide normal cooldown of the f acility, or 2) to af tigate the effects of accident conditions within acceptable limits.

The limitations on minimum water level and taximum temperature are based on providing a 30 day cooling watcr supply to safety related equipment without exceedits their design basis temperature and is consistent with the recoc=endations of Regulatory Guide 1.27, " Ultimate Beat Sink for Nuclear Plants" March 1974 3/4.7.6 CONTROL R00M EMERCENCY VEh7ILATION SYSTEM 1

The OPERABILITY of the control room emergency ventilation system ensures that 1) the ambient air temperature does not exceed the allowable temperature for continuous duty rating for the equipment and instrumentation cooled by this systen and 2) the control room vill remain habitable for operations personnel during and following all credible accident conditions. The OPERABILITY of this system in conjunction with control roen design provisions is based on liatting the radiation exposure to personnel occupying the control room to 5 rea or less whole body, or its equivalent. This limitation is consistent with the requirements of General Design Criterion 19 of Appendix "A".

10 CTR 50.

DAVIS-BESSE, UNIT 1 B 3/4 7 4 Amendment No.103

i PLANT SYSTEMS BASES l

3/4.7.8 SEALED SOURCE CONTAMINATION The limitations on removable contamination for sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(c) limits for plutonium.

This limitation vill ensure that leakage from by-product, source, and special l

nuclear material sources will not exceed allowable intake values.

3/4.7.9 FIRE SUPPRESSION SYSTEMS The OPERABILITY of the fire suppression systems ensures that adequate fire suppression capability is available to confine and'extinquish fires occurring in l

any portion of the facility where safety related equipment is located. The fire suppression system consists of the vater system, spray and/or sprinklers, and fire hose stations. The collective capability of the fire suppression systems

{

is adequate to minimize potential damage to safety related equipment and is a major element in the facility fire protection program.

i In the event that portions of the fire suppression systems are inoperable, alternate backup fire fighting equipment is required to be made available in the affected areas until the inoperable equipment is restored to service.

In the event the fire suppression water system becomes inoperable, immediate corrective measures must be taken since this system provides the major-fire suppression capability of the plant. The requirement for a twenty-four hour I

report to the Commission provides for prompt evaluation of the acceptability of l the corrective measures to provide adequate fire suppression capability for the continued protection of the nuclear plant.

i 1

3/4.7.10 FIRE BARRIERS The OPERABILITY of the fire barrier ensures that fires vill be confined or adequately retarded from spreading to adjacent fire areas or to portions of redundant safe shutdown systems required in the event of a fire within the fire l

This design feature min'imizes the possibility of a single fire rapidly area.

involving several fire areas of the facility prior to detection and extinguishment. The fire barriers are passive elements in the facility fire protection program.

Fire barriers, including cable penetration barriers, fire doors and dampers, are considered OPERABLE vhen the visually observed condition is the same as the as-designed condition. The as-designed condition of each fire barrier is based on a tested configuration or a configuration analyzed to withstand the fire hazards associated with the fire area.

DAVIS-BESSE, UNIT 1 B 3/4 7-6 Amendment No. 9, 106, 135

3/4.9 REFUELING OPERATIONS p

BASES 3/4.9.1 BORON CONCENTRATION The limitations on reactivity conditions during REFUELING ensure that:

~

1) the reactor will remain subcritical' during CORE ALTERATIONS, and 2) a uniform boron concentration is maintained for reactivity control in the water volumes having direct access to the reactor vessel. These limita-tions are consistent with the initial conditions assumed-for the boron dilution incident in the accident analysis.

3/4.9.2 INSTRUMENTATION The OPERABILITY of source range neutron flux monitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core.

3/4.9.3 DECAY TIME The minimum requirement for reactor subcriticality prior to movement of irradiated fuel assemblies in the reactor pressure vessel ensures that sufficient time has elapsed to allow the radioactive decay of the short lived fission products. This decay time is consistent with the assumptions used in the safety analyses.

3/4.9.4 CONTAINMENT PENETRATIONS The requirements on containment penetration closure and OPERABILITY ensure that a release of radioactive material within containment will be restricted from leakage to the environment. The OPERABILITY and closure requirements are sufficient to restrict radioactive material release from a fuel element rupture based upon the lack of containment pressuriza-tion potential while in the REFUELING MODE.

3/4.9.5 COMMUNICATIONS Therequirementforcommunicationscapabilityensuresthathefueling station personnel can be promptly infonned of si'gnificant changes in the 4

facility status or core reactivity condition during CORE ALTERATIONS.

l DAVIS-BESSE, UNIT 1 B 3/4 9-1 S

REFUELING OPERATIONS BASES 3/4.9.6 FUEL HANDLING BRIDGE OPERABILITY The OPERABILITY requirements of the hoist bridges used for movement of fuel assemblies ensures that:

1) fuel handling bridges vill be used for movement of control rods and fuel assemblies, 2) each hoist has sufficient load capacity to l

lift a fuel element, and 3) the core internals and pressure vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations.

3/4.9.7 CRANE TRAVEL - FUEL HANDLING BUILDING The restriction on movement of loads in excess of the nominal weight of a fuel l

assembly in a failed fuel container over other fuel assemblies in the storage l

pool ensures that in the event this load is dropped (1) the activity release vill be limited to that contained in a single fuel assembly, and (2) any possible distortion of fuel in the storage racks vill not result in a critical array. This assumption is consistent with the activity release assumed in the accident analyses.

l 3/4.9.8 COOLANT CIRCULATION The requirement that at least one decay heat removal loop be in operation

)

ensures that (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel belov 140'F as required during the REFUELING MODE, and (2) sufficient coolant circulation is maintained through the reactor core to minimize the effect of a boron dilution incident and prevent boron stratification.

The requirement to have two DHR loops OPERABLE vhen there is less than 23 feet of water above the core ensures that a single failure of the operating DHR loop i

vill not result in a complete loss of decay heat removal capability. With the reactor vessel head removed and 23 feet of water above the core, a large heat

{

sink is available for core cooling. Thus, in the event of a failure of the operating DHR loop, adequate time is provided to initiate emergency procedures l

to cool the core.

3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM The OPERABILITY of this system ensures that the containment purge and exhaust penetrations vill be automatically isolated upon detection of high radiation levels within the containment. The OPERABILITY of this system is required to I

restrict the release of radioactive material from the containment atmosphere to the environment.

3/4.9.10 and 3/4.9.11 VATER LEVEL - REACTOR VESSEL AND STORAGE POOL l

The restrictions on minimum vater level ensure that sufficient vater depth is available to remove 99% of the assumed 10% iodine gap activity released from the i

rupture of an irradiated fuel assembly. The minimum vater depth is consistent I

with the assumptions of the safety analysis.

DAVIS-BESSE, UNIT 1 B 3/4 9-2 Amendment No. 36/ 135

5 REFUELING OPERATIONS i

BASES 3/4.9.12 STORAGE P0OL VENTILATION The requirements on the emergency ventilation system servicing the storage pool area to be operating or OPERABLE ensure that all radioactive material released from an irradiated fuel assembly will be filtered through the HEPA filters and charcoal adsorber prior to discharge to the atmosphere.

The OPERABILITY of this system and the resulting iodine removal capacity are consistent with the assumptions of the safety analyses.

3/4.9.13 SPENT FUEL POOL FUEL ASSEMBLY STORAGE The restrictions on the placement of fuel assemblies within the spent fuel pool, as dictated by Figure 3.9-1, ensure that the k-effective of the spent fuel pool will always remain less than 0.95 assuming the pool to be flooded with non-borated water. The restrictions delineated in Figure 3.9-1 and the action statement are consistent with the criticality safety analysis performed for the spent fuel pool.

DAVIS-BESSE, UNIT 1 B 3/4 9-3 Amendment No. lid.135 l

RADI0 ACTIVE EFFLUENTS

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BASES 3/4.11.1 LIQUID EFFLUENTS 3/4.11.1.1 CONCENTRATION This specification is provided to ensure that the concentration of radioactive materials released in liquid waste effluents from the site to unrestricted areas will be less than the concentration levels specified in 10 CFR Part 20.106. This Ifaitation as implemented by the ODCM pro-vides additional assurance that the levels of radioactive materials in bodies of water outside the site should not result in exposures exceeding (1) the Section II.A design objectives of Appendix I, 10 CFR Part 50, to an individual and (2) the limits of 10 CFR Part 20.106(e) to the population.

The concentration limit for noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in. International Commission on Radiological Protection (ICRP) Publication 2.

3/4.11.1.2 005E This specification is provided to implement the requirements of Sections II.A. III.A and IV.A of Appendix I, 10 CFR Part 50. The Limiting condition for Operation implements the guides set forth in Section II.A of Appendix I.

The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in liquid effluents will be kept "as low as is reasonably achievable". Also, for fresh water sites with drinking water supplies which can be potentially affected by plant operations, there is reasonable assurance that the operation of the facility will not result

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in radionuclides concentrations in the finished drinking water that are in excess of the requirements of 40 CFR 141. The dose calculations in the ODCM implement the requirements in Section III. A of Appendix I that conformance with the guides of Appendix I is to be shown by cal-culational procedures based on modes and data such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I", Revision 1, October 1977.

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DAVIS-BESSE, UNIT 1 g 374 11-1 Amendment No. 86

a RADIOACTIVE EFFLUENTS BASES 3/4.11.1.3 LIOUID VASTE TREATMENT The requirements that the appropriate portions of this system be used when specified provides assurance that the releases of radioactive materials in liquid effluents will be kept "as lov as is reasonably achievable". This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and design objective Section II.D of Appendix I to 10 CFR Part 50.

Based on a cost analysis of treating liquid

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radvaste, the specified limits governing the use of appropriate portions of the liquid radvaste treatment system were specified as the dose design objectives as set forth in Section II.A of Appendix I, 10 CFR Part 50, for, liquid effluents.

3/4.11.1.4 LIQUID HOLDUP TANKS 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 vould 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.

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DAVIS-BESSE, UNIT 1 B 3/4 11-2 Amendment No. 26,135

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7 DAVIS-BESSE NUCLEAR POWER STATON LOW POPULATON ZONE FIGURE 5.1-2 DAVIS-BESSE, UNIT 1 5-3 I

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DESIGN FEATURES DESIGN PRESSURE AND TEMPERATURE 5.2.2 The reactor containment building is designed and shall be maintained for a maximum internal pressure of 40 psig and a temperature of 264'F.

5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 177 fuel assemblies with each fuel assembly containing 208 fuel rods clad with Zircaloy -4.

Each fuel rod shall have a nominal active fuel length of 144 inches and contain a maximum total i

weight of 2500 grams uranium. Reload fuel shall be similar in physical design l

to the initial core loading and shall have a maximum enrichment of 3.8 veight percent U-235.

CONTROL RODS 5.3.2 The reactor core shall contain 53 safety and regulating and 8 axial power i

shaping (APSR) control rods. The safety and regulating control rods shall contain a nominal 134 inches of absorber material. The nominal values of absorber material shall be 80 percent silver, 15 percent indium and 5 percent cadmium. All control rods shall be clad with stainless steel tubing. The APSRs shall contain a nominal 63 inches of absorber material at their lover ends. The absorber material for the APSRs shall be 100 percent Inconel-600.

5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The reactor coolant system is designed and shall be maintained:

a.

In accordance with the code requirements specified in Section 5.2 of the FSAR, with allovance for normal degradation pursuant to applicable Surveillance Requirements.

b.

For a pressure of 2500 psig, and c.

For a temperature of 650'F, except for the pressurizer and pressurizer i

l surge line which is 670'F.

DAVIS-BESSE, UNIT 1 5-4 Amendment No. II, M,M/3,MO, 135

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I l 6.0 ADMINISTRATIVE CONTROLS

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I 6.1 RESPONSIBILITY l

6.1.1 The Plant Manager shall be responsible for overall facility operation I

and shall delegate in writing the succession to this responsibility during his absence.

6.2 ORGANIZATION j

j 6.2.1 0FFSITE AND ONSITE ORGANIZATIONS Onsite and offsite organizations shall be established for facility operation and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting the safety of the nuclear power plant.

I a.

Lines of authority, responsibility, and communication shall be l

established and defined for the highest management levels through intermediate levels to and including all operating organization positions. These relationships shall be documented and updated, as appropriate, in the form of organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent i

forms of documentation.

These requirements shal. be documented in l

the Updated Safety Analysis Report.

b.

The Vice President, Nuclear shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety.

c.

The Plant Manager shall be responsible for overall unit safe operation and shall have control over those onsite activities necessary for safe operation and maintenance of the plant.

d.

The individuals who train the operating staff and those who carry out health physics and quality assurance functions may report to the appropriate orsite manager; however, they shall have sufficient organizational freedom to casure their independence from operating pressures.

DAVIS-BESSE, UNIT 1 6-1 Amendment No. 9,26,98,115,13$

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6.0 ADMINISTRATIVE CONTROLS l'

6.2.2 FACILITY STAFF a.

Each on duty shift shall be composed of at least the

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minimum shift crew composition shown in Table 6.2-1.

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b.

At least one licensed Operator shall be in the control room when fuel is in the reactor.

c.

At least two licensed Operators shall be present in the control room during reactor start-up, scheduled reactor shutdown and during recovery from reactor trips.

d.

An individual qualified in radiation protection procedures l

shall be on site when fuel is in the reactor #.

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e.

All CORE ALTERATIONS shall be directly supervised by either a licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling who has no other con-current responsibilities during this operation.

f.

A site Fire Brigade of at least 5 members shall be maintained onsite at all times #.

The Fire Brigade shall not include 3 members of the minimum shift crew necessary l

1 for safe shutdown of the unit and any personnel required for other essential functions during a fire emergency.

g.

The Assistant Plant Manager-Operations shall hold a senior l

I reactor operator license.

The Operations Superintendent shall hold a senior reactor operator license.

The individual qualified in radiation protection procedures and the Fire Brigade Composition may be less than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accommodate unexpected absence, provided immediate action is taken to fill the required positions.

DAVIS-BESSE, UNIT 1 6-la Amendment No. 9,J$,98,JJ5,135 (next page is 6-2)

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Davis-Besse Nuclear Power Station Offsite/Onsite Organization DELETED j

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l DAVIS-BESSE, UNIT 1 6-2 Amendment No. H,E+, r+,M,(14 135 c_-_-_-_-_-__x

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Davis-Besse Nuclear Power Station Operations organization DELETED-1 1

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.l DAVIS-BESSE, 1: NIT 1 6-3

- Amendment No. e,%.,w,gy).135 -

i TABLE 6.2-1 MINIMUM SHIFT CREW COMPOSITION #

LICENSE APPLICABLE MODES CATEGORY 1, 2, 3 & 4 5&6 SOL 1

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J OL 2

1 Non-Licensed 2

1 Shift. Technical Advisor 1

None Required

Does not include the licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling supervising CORE ALTERATIONS.

Shift crew composition may be less than the ninfr.:n requirements for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accommodate unerprated 2

absence of on duty shift crew members provided immediate action is :?len

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to restore the shift crew composition to within the minimum requirements j

of Table 6.2-1.

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DAVIS-BESSE, UNIT 1 6-4 Amendment No. 37 i

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ADMINISTRATIVE CONTROLS

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6.3 FACILITY STAFF QUALIFICATIONS 6.3.1 Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except for (1) the Chemistry and Health Physics General Superintendent who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975 and (2) the Shift Technical Advisor who shall have a bachelor's degree or equiv-alent in a scientific or engineering discipline with specific training in plant design, and response and analysis of the' plant for transients and accidents.

P 6.4 TRAINING 6.4.1 A retraining and replacement training program for the facility staff shall be maintained under the direction of the Nuclear Training Director and shall meet or exceed the requireti.ents and recommendations of Section 5.5 of ANSI N18.1-1971 and of 10 CFR 55.59.

l 6.4.2 A training program for the Fire Brigade shall be maintained under-the direction of the Nuclear Training Director.

6.5 REVIEW AND AUDIT 6.5.1 STATION REVIEW BOARD (SRB)

FUNCTION 6.5.1.1 The Station Review Board (SRB) shall function to advise the Plant Manager on all matters related to nuclear safety.

DAVIS-BESSE, UNIT 1 6-5 Amendment No. 9,37,E9,9E,J06,135

ADMINISTRATIVE CONTROLS COMPOSITION 6.5.1.2 The Station Review Board shall be composed of the:

Chairman:

Station Review Board Chairman

Member:

Assistant Plant. Manager, Operations Member:

Assistant Plant Manager, Maintenance Member:

Technical Support Manager..

Member:

Chemistry and Health Physics General Superintendent Member:

Operations Engineering Supervisor -(Plant)

Member:

An Engineering Director or Performance Engineering Manager l

Member:

Operations Quality Assurance Manager Member:

Operations Superintendent (Plant)-

Designated in writing by the Plant Manager. The Chairman will be drawn from SRB members.

ALTERNATES 6.5.1.3 All alternate members shall be appointed in writing by the SRB

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Chairman; however, no more than two alternates shall participate as voting l

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members in SRB activities at any one time.

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1 MEETING FREOUENCY

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6.5.1.4 The SRB shall meet at least once per calendar month and as convened by the SRB Chairman or his designated' alternate.

OUORUM 6.5.1.5 A quorum of the SRB shall consist of the Chairman or his desig-nated alternate and four members including alternates.

RESPONSIBILITIES i

6.5.1.6 The Station Review Board shal3 be responsible for:

a.

Review of plant administrative procedures and changes thereto, b.

Review of the safety evaluation for 1) procedures, 2) changes to procedures, equipment or systems and 3) tests or experiments completed under the provisions of 10 CFR 50.59, to verify that i

l such actions do not constitute an unreviewed safety question.

1 Review of proposed procedures and changes to procedures and c.

equipment determined to involve an unreviewed safety question as defined in 10 CFR 50.59.

DAVIS-BESSE, UNIT 1 6-6 Amendment No. 12,76,98, 109

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i ADMINISTRATIVE CONTROLS i

power operation). supplementary reports shall be submitted at least every three months until all three events have been completed.

ANNUAL OPERATING REPORT j

6.9.1.4 Annual reports covering the activities of the unit during the previous calendar year shall be submitted prior to March 31 of each year.

l 6.9.1.5 Reports required on an annual basis shall include:

a.

A tabulation on an annual basis of the number of station, utility and other personnel (including contractors) receiving exposures greater than 100 mrem /yr and their associated man rem exposure according to work and job functions 3/, e.g., reactor l

operations and surveillance, inservice inspection, routine maintenance, special maintenance (described maintenance), waste processing, and refueling.

The dose assignment to various duty functions may be estimates based on pocket dosimeter, TLD, or film badge measurements.

Small exposures totalling less than 20% of the individual total dose need not be accounted for.

In the aggregate, at least 80% of the total whole body dose received from external sources shall be assigned to specific major work functions.

b.

The complete results of steam generator tube inservice inspections (Specification 4.4.5.5.b).

c.

The results of specific activity analysis 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 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded; (2) Results of the last isotopic analysis for radioiodine performed prior to exceeding the limit, results of analysis while limit was exceeded and results of one analysis after the radiciodine activity was reduceo to less than limit.

Each result should include date and time of sampling and the radioiodine concentrations; (3) Clean-up system flow history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded; (4) Graph of the I-131 concen-tration and one other radioiodine isotope concentration in 1

1/ This tabulation supplements the requirements of 620.407 of 10 CFR I

Part 20.

i DAVIS-BESSE, UNIT 1 6-15 Amendment No. 9,J2,fJ,E2,73.

B7,10 A,135

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ADMINISTRATIVE CONTROLS microcuries per gram as a function of time for the duration of the j

specific activity above the steady-state level; and (5) The time duration when the specific activity of the primary coolant exceeded

.the radioiodine limit.

MONTHLY OPERATING REPORT 6.9.1.6 Routine reports of operating statistics, shutdown experience and challenges to the Pressurizer Pilot Operated Relief Valve (PORV) and the l

Pressurizer Code Safety Valves shall be submitted on a monthly basis to arrive no later than the 15th of each month following the calendar month covered by the l

report, as follows:

the signed original to the Nuclear Regulatory Commission, Document Control Desk, Washington, D.C. 20555, and one copy each to the Region III Administrator and the Davis-Besse Resident Inspector.

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i DAVIS-BESSE, UNIT 1 6-16 Amendment No. 8,12, 93,Jpf,135

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1 ADMINISTRATIVE CONTROLS I

l SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the U.S. Nuclear Regulatory Commission in accordance with 10 CFR 50.4 within the time period specified for each report. These reports shall be submitted covering the activities l

identified below pursuant to the requirements of the applicable reference specification:

a.

ECCS Actuation, Specifications 3.5.2 and 3.5.3.

b.

Inoperable Seismic Monitoring Instrumentation, Specification 3.3.3.3.

c.

Inoperable Meteorological Monitoring Instrumentation, Specification 3.3.3.4.

d.

Seismic event analysis, Specification 4.3.3.3.2.

l e.

Fire Detection Instrumentation, Specification 3.3.3.8.

f.

Fire Suppression Systems, Specifications 3.7.9.1 and 3.7.9.2.

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g.

Dose or dose commitment exceedences to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released to UNRESTRICTED AREAS (Specification 3.11.1.2).

h.

The discharge of radioactive liquid vaste without treatment and in excess of the limits in 3 specification 3.11.1.3.

i.

The calculated air dose from radioactive gases exceeding the limits in Specification 3.11.2.2.

j.

The calculated dose from the release of iodine-131, tritium, and radionuclides in particulate form with half-lives greater than 8 days, in gaseous effluents exceeding the limits in Specification 3.11.2.3.

k.

The discharge of radioactive gaseous vaste without treatment and in excess of the limits in Specification 3.11.2.4.

1.

The calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding the limits in Specification 3.11.4.

The level of radioactivity as the result of plant effluents in an m.

environmental sampling medium exceeding the reporting levels of Table 3.12-2 (Specification 3.12.1).

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I DAVIS-BESSE, UNIT 1 6-18 Amendment No. 9, 12, 65, 86, 106,135

r ADMINISTRATIVE CONTROLS 6.10 RECORD RETENTION 6.10.1 The following records shall be retained for at least five years:

Records and logs of facility operation covering time interval a.

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.

I DAVIS-BESSE, UNIT 1 6-18a Amendment Fo. 86,93 (Next pags is 6-19)

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