ML20004D961

Amend 38 to License NPF-3,changing Tech Specs Dealing W/ Decay Heat Removal Capability

ML20004D961
Person / Time
Site:	Davis Besse
Issue date:	06/01/1981
From:	Stolz J Office of Nuclear Reactor Regulation
To:	Cleveland Electric Illuminating Co, Toledo Edison Co
Shared Package
ML20004D962	List: ML20004D961 ML20004D964
References
NPF-03-A-038 NUDOCS 8106100412
Download: ML20004D961 (28)

Text

_.

i O

0 f

UNITED STATES NUCLEAR REGULATORY COMMISSION y'

~,

g

.y WASHINGTON, D. C. 20555 t

%, *.... /

THE 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. 38 License No. NPF-3 1.

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

A.

The application for amendment by The Toledo Edison Company and The Cleveland Electric Illuminating Company (the licensees) dated' December 26, 1980, conplies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Com-mission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in confomity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.

There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health i

and safety of the public, and (ii) that such activities will be conducted in ecmpliance 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 is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

8106100Y/A;

12

(

I ATTACHMENT TO LICENSE AMENDMENT NO. 38 IA'ILITY OPERATING LICENSE NO. NPF-3 j

DOCKET NO. 50-346 t

Replace the following pages of the Appendix "A" Technical Specifications

[

with the enclosed pages as indicated. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.

j The corresponding overleaf pages are also provided to maintain document I

completeness.

r Pages j

Table of Contents pgs. I-XVI 3/4 4-1 f

3/4 4-2 l

[

3/4 4-2a (new) 3/4 9-8 3/4 9-8a (new)-

B3/4 4-1 t

B3/4 9-2

{

INDEX N.

DEFINITIONS SECTION PAGE 1.' O DEFINITIONS DEFINED TERMS......................'........................

1-1 THERMAL P0WER..............................................

1-1 RATED THERMAL P0WER........................................

1-l' OPERATIONAL M0DE........................................'..

1-1 ACTIO1.....................................................

l.-l OPERABLE - OPERABILITY.....................................

1-1 REPORTABLE OCCURRENCE.....................................

1.2 l

CONTAINMENTiN'TEGRITY......................................

1-2

~

CHANNEL CALIBRATION........................................

1-2 CHANNEL CHECK...........................................'...

1-2 CHANNEL FUNCTIONAL TEST....................................

1-3 CORE ALTERATION............................................

1-3 SHUTDOWN MARGIN......'......................................

1-3 IDENTIFIED LEAKAGE.........................................

1-3 UNIDENTIFIED LEAKAGE........................................

1-4 PRESSURE BOUNDARY LEAKAGE..................................

1-4 CONTROLLED LEAKAGE........................................

1-4 QUADRANT POWER TILT........................................

1-4 DOSE EQUIVALENT I-131......................................

1-4 E-AVERAGE DISINTEGRATION ENERGY............................

1-4 STAGGERED TEST BASIS........'...............................

1-5 FREQUENCY N0TATION.........................................

1-5 AXIAL POWER IMBALANCE......................................

1-5 e

SHIELD BUILDING INTEGRITY..................................

1-5 REACTOR PROTECTION SYSTEM RESPONSE TIME....................

1-5 SAFETY FEATURE RESPONSE TIME...............................

1-6 PHYSICS TESTS..............................................

1-6 STEAM AND FEEDWATER RUPTURE CONTROL SYSTEM RESPONSE TIME....

1-6 OPERATIONAL MODES (TABLE 1.1 )..............................

1-7 FREQUENCY NOTATION (TABLE 1.2).............................

1-8 DAVIS-BESSE, UNIT 1 I

Amendment No. 38-

^

~ ~~

^

n P

INDEX SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS P

SECTION PAGE-2.1 SAFETY LIMITS f

R e a c to r Co r e..............................................

2-1

{

Reactor Coolant System Pressure...........................

2-1 l

2.2 LIMITING SAFETY SYSTEM SETTINGS i

Reactor Protection System Setpoints......................

2-4 f

i

[

t I

i i

BASES s

SECTION PAGE t

2.1 SAFETY LIMITS Reactor Core..............................................

B 2-1

[

TeactorCoolantSystemPressure...........................

B 2-3

+

2.2 LIMITING SAFETY SYSTEM SETTINGS 7

i Reactor Protection Systern Instrumentation Setpoints.......

B 2-4 l

l t

I l

DAVIS-BESSE, UNIT 1 II Amendment No. 38

INDEX t

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS PAGE SECTION 3/4.0 APPLICABILITY...........................................

3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS l

3/4.1.1 BORATION CONTROL Shutdown Margin......................................

3/4 1-1 Boron Dilution...................................... 3/4 1-3 Moderator Temperature Coefficient....................

3/4 i-4 i

Minimum Temperature for Criticality..................

3/4 1-5 3/4.1.2 B0 RATION SYSTEMS Fl ow Pa ths - Shu tdown................................ 3/4 1-6 Fl ow Pa ths - Operati ng...............................

3/4 1-7 Makeup Pump - Shutdown...............................

3/4 1-9 Makeup Pumps - Operating.............................

3/4 1-10 Decay Heat Removal Pump - Shutdown...................

3/4 1-11 Bo ric Acid Pump - Shutdown...........................

3/4 1-12 Boric Acid Pumps - Operating.........................

3/4 1-13 l

Barated Water Sources - Shutdown.....................

3/4 1-14 Borated Wa ter Sources - Operating.................... 3/4 1-17 e

i 3/4.1.3 MOVABLE CONTROL ASSEMBLIES l

Group Height - Safety and Regulating Rod Groups......

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

3/41-2i l

Posi tion Indicator Channel s.......................... 3/4'l-22 Rod Drop Time.........................................

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

3/4 1-25

/

Regulating Rod Insertion Limits......................

3/4 1-26 i

Rod Program..........................................

3/4 1 30 E

a t

l Xenon, Reactivity.....................................

3/4 1 33 i

Axi al Powe r Shapi ng Rod Ins e rti on Li mi ts............. ".' 3/4 1-34 l

DAVIS-BESSE, UNIT 1 III Amendment No. 38

~

)

...-.. a l

l l

INDEX LIM: TING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL POWER IMBALANCE................................

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

CHANNEL FACTOR - F................................

3/42 9

'3/4.2.3 NUCLEAR ENTHALPY RISE N

HOT CHANNEL FACTOR - F 3/42-7 aH...........................

3/4.2.4 QUADRANT POWER TILT..................................

3/42-9,

~

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

........................ 3/4 2-13 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION............

3/4 3-1 3/4.3.2 SAFETY SYSTEMS INSTRUMENTATION Safety Features Actuation System.....................

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

3/4 3-23 3/4.3.3 MONITORING INSTRUMENTAT:0N Radiation Moni toring Ins trumentation................. 3/4 3 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 Pos t-Accident Ins trumentati on........................ 3/4 3-46 l

Chl ori ne Detection Sys tems,........................... 3/4 3-51 I

l Fi re Dete ction Ins trumenta tio n........................ 3/4 3-52 3/4.4 REACTOR COOLANT SYSTEM 5

3/4.4.1 COOLANT LOOPS AND COOLANT CIRCULATION

~

Startup and Powe r 0pe rati on........................... 3/4 4-1 Shutdewn and Hot Stan dby.............................. 3/4 4-2 3/4.4.2 SAFETY VALVES - S HUTD0WN.............................. 3/44-3 3/4.4.3 SAFETY VALVES AND ELECTROMATIC RELIEF VALVE - OPERATING 3/4 4-4 DAVIS'BESSE, UNIT 1 IV Amendment No. 38

INDEX LIMITING CONDITION FOR OPERATION AND SURVEILLANCE REQUIREMENTS i

SECTION PAGE 3/4.4.4 PRESSURIZER..........................................

3/4 4-5 i

3/4.4.5 STEAM GEN E RATORS.....................................

3/4 4-6 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE I

Lea kage Detection Sys tems............................ 3/4 4-13 Operational Leakage..................................

3/4 4-15 3/4.4.7 CH EM I STRY............................................ 3/4 4 -17 3/4.4,8 SPECIFIC ACTIVITY..................................... 3/4 4-?0 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System...............................

3/4 4-24 Pressurizer..........................................

3/4 4-29 3/4.4.10 STRUCTURAL INTEGRITY.................................. 3/4 4-30 3/4.5 EMERGENCY CORE COOLING SYSTEMS (EC' j[

i 3/4. 5.1 CORE FLOODING TANKS..................................

3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - Tavg - 280*F..................,....

3/4 5-3 3/4,5.3 ECCS SUBSYSTEMS - T,yg < 280*F.......................

3/4 5-6 j344.5.4 BORATED WATER STORAGE TANK...........................

3/4 5-7 I

l DAVIS-BESSE., UNIT 1 V

i

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Containment Integrity................................

3/4 6-1 l

Co n ta i nmen t Lea ka g e..................................

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

3/4 6-6 Internal Pressure....................................

3/4 6-7 Ai r T em p e ra tu re...................................... 3/4 6-8 l

Containnent -Vessel Structural Integri ty..............

3/4 6-9 l

i Conta inment Ventil a tion System.......................

3/4 6-10 i

3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS i

Co n ta i nmen t Sp ray Sy s tem.............................

3/4 6-11 Containment Cooling System...........................

3/4 6-13 l

3/4.6.3 CONTAINMENT ISOLATION VALVES......................... 3/4.6-14 3/4.6.4 COMBUSTIBLE GAS CONTROL Hydrogen Analyzers...................................

3/4 6-23 Containment Recirculation System.....................

3/4 6-24

[

Containment Hydrogen Dilution System.................

3/4 6-25 t

Hydrogen Purge System................................

3/4 6-26 3/4.6.5 SECONDARY CONTAINMENT Emergency Ventilation System................:........

3/4 6-28

'~~

Shield Building Integrity............................

3/4 6-31,

Shield Building Structural Integrity.................

3/4 6-32 l

l 7

i l

b l

i t

i b

t DAVIS-BESSE, UN'IT 1 VI Amendment No. 38 i

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS PAGE SECTION t

3/4.7 PLANT SYSTEMS i

3/4.7.1 TURBINE CYCLE Safety va1ves........................................

3/4 7-1 Auxil iary Feedwa ter System........................... 3/4 7-4 Condensate Storage Tank..............................

3/4 7-6 Ac t i v i ty............................................. 3/ 4 7 - 7 Main Steam Line I' solation Va1ves.....................

3/4 7e9 l

Seconda ry Water Chemistry............................ 3/4 7-10 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION......

3/4 7-13 3/4.7.3 COMPONENT COOLING WATER SYSTEM.......................

3/4 7-14 '

3/4.7.4 SERVICE WATER SYSTEM.................................

3/4 7-15 3/4.7.5 U LT I MATE H EAT S I N K................................... 3/ 4 7-16 3/4.7.6 CONTROL ROOM EMERGENCY VENTILATION SYSTEM............

3/4 7-17 i

3/4.7.7 HYDRAULIC SNUBSERS...................................

3/4 7-20 l

i 3/4.7.8 S EALED SOURCE CONTAMINATION........................... 3/4 7-36 j

r

--- 3/4. 7.9 FIRE SUPPRESSION SYSTEMS Fi re Supp ression Wate r Sys tem......................... 3/4 7-38 Sp ray and/or Sprinkl er Sys tem........................ 3/4 7-42 Fi re Hos e Stati ons.................................... : 3/4 7-44 3/4.7.10 PENETRATION FIRE BARRIERS.............................

3/4 7-47 3/4.8 EL'ECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES O pe ra t i n g...........................................

3/ 4 8-1

[

Shutdown.............................................

3/4 8-5 3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEMS A.C. Distribution - Operating........................

3/4 8-6*

A.C. Di stri bution - Shutdown.........................

3/4 8-7 f

D.C. Distribution - Operating........................

3/4 8-8 i

D.C. Distribution - Shutdown...........'..............

3/4 8-10 i

DAVIS-BESSE, UNIT 1 yII knendment No. 38

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON C0NCEdTRATION.................................

3/4 9-1 3/4.9.2 INSTRUMENTATION.....................................

3/4 9-2 3/4.9.3 DECAY TIME..........................................

3/4 9-3 3/4.9.4 CONTAINMENT PENETRATIONS............................

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

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 "

3/4.9.8 DECAY HEAT REMOVAL AND COOLANT CIRCULATION Al l Wa te r Le ve l s.....................................

3/4 9-8 Low W a t e r L e ve l......................................

3/4 9-8a

~

3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM...'... 3/49-9 3/4.9.16 WATER LEVEL - REACTOR VESSEL........................

3/4 9-10 3/4.9.11 STORAGE POOL WATER LEVEL............................

3/4 9-11 3/4.9.12 STORAGE POOL VENTILATION............................

3/4 9-12 3/4.10 SPECIAL TEST EXCEPTIONS

~

3/4.10.1 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS.........................

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

3/4 10.-2 e

3/4.10.3. REACTOR COOLANT L00PS...............................

3/4 10-3 3/4.10.4 S H UT CO WN MA R G I N.................'.................... 3/4 10-4 DAVIS-3 ESSE, UNIT 1 VIII Amendment No, 38 e

.m Ws

--.e

~e e

INDEX BASES s

PAGE SECTION 3/4.0 APPLICABILITY..........................................

B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS l

3/4.1.1 B O RAT I ON CONTR0L..................................... B 3/4 1-1 i

3/4.1.2 B0 RATION' SYSTEMS.....................................

B 3/4 1-2 3/4.1.3 MOVABLE CONTROL ASS EMBLI ES........................... B 3/4 1-3 3/4.2 POWER DISTRIBUTION LIMITS..............................

B 3/4 2-1 1

3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION............

B 3/4 3-1 SAFETY SYSTEMS INSTRUMENTATION.......................

B'3/4 3-1 3/4.3.2 i

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 i

3/4.4.2 and 3/4.4.3 SAFETY VALVES...........................

B 3/4 4-1 3/4.4.4 PRESSURI2ER.........................................

B.3/4 4-2 l

3/4.4.5 STEAM GENERATORS....................................

B 3/4 4-2 3/4 4.6 REACTOR COOLANT SYSTEM LEAKAGE......................

B 3/4 4-4 l

B 3/4 4-S i

3/4.4.7 CHEMISTRY...........................................

3/4.4.8 SPECIFIC ACTIVITY...................................

B 3/4 4-5 t

B 3/4 4-6 i

3/4.4.9.. PRESSURE / TEMPERATURE LIMITS.........................

3/4.4.10 STRUCTURAL INTEGRITY................................

B 3/4 4-13 r

DAVIS-BESSE, UNIT 1 IX i

l INDEX BASES SECTION PAGE t

3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 CORE FLOODING TAN KS.................................. B'3/4 5-1 l

i 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 l

f f

P b

t l

r i

i s

DAVIS-BESSE, UNIT 1 X

Amendment No. 38 l

1 i

INDEX BASES PAGE SECTION 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT...................................

B 3/4 6-1 7

3/4.6.2 DEPRESSURIZATION A'ND COOLING SYSTEMS.................. B 3/4 6-2 3/4.6.3 CONTAINMENT ISOLATION VALVES.......................... B 3/4 6-3 3/4.6.4 COMBUSTIBLE GAS CONTR0L............................... B 3/4 6,4 3/4.6.5 SHIELD BUILDING....................................... B 3/4 6-4 c

t P

i' t

t DAVIS-BESSE,-UNIT 1 XI t

?

• ~=*--=+-.---,s

INDEX i

BASES SECTION PAGE 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE.................................... B 3/4 7-1 t

3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION.. B 3/4 7-3 3/4.7.3 COMPONENT COOLING WATER SYSTEM................... B 3/4 7-3 3/4.7.4 SERVICE WATER SYSTEM............................. B 3/4 7-4 f

3/4.7.5 ULTIMATE HEAT SINK............................... B 3/4 7-4 3/4.7.6 CONTROL ROOM EMERGENCY VENTILATION SYSTEM........ B 3/4.-4 3/4.7.7 ' HYDRAULIC SNUBBERS............................... B 3/4 7-5 3/4.7.8 SEALED SOURCE CONTAMINATION........................B 3/4 7-6

[

,3/4.7.9 FIRE SUPPRESS ION SYSTEMS........................... B 3/4. 7-6 3/4.7.10 PENETRATION FIRE B ARRIERS......................... B 3/4 7-6

, /4.8 ELECTRICAL POWER SYSTEMS............................

B 3/4 8-1 3

3/4.9 REFUELING OPERATIONS i

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 I

i 3/4 9.4 CONTAINMENT PENETRATIONS........................ B 3/4 9-1 i

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

I DAVIS-BESSE, UNIT 1 XII Amendment No. 38 i

l l

i q

..c.

L INDEX BASES f

PAGE SECTION 3/4.9.6 FUEL HANDLING BRIDGE OPERABILITY................... B 3/4 9-2 3/4.9.7 CRANE TRAVEL - FUEL HANDLING BUILDING.............. B 3/4 9-2 1

3/4.9.B 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 STORAGE POOL WATER LEVEL...........................

B 3/4 9-2 3/4.9.12 STORAGE POOL VENTILATION...........................

B 3/4 9-3 3/4.10 SPECIAL TEST EXCEPTIONS

- t 3/4.10.1 GROUP. HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS........................

B 3/4 10-1 B 3/410-1 3/4.10.2 P,HYSICS TESTS......................................

3/4.10.3 REACTOR COOLANT L00PS..............................

B 3/4 10-1 3/4.10.4 SHUTDOWN MARGIN................................... B 3/410-1 h

e h

i e

i s

4

/

L t

l DAVIS-BESSE, UNIT 1 XIII i

.i

INDEX DESIGN FEATURES SECTION PAGE 5.1 SITE l

t ExclusionArea............................................

5-1 l,

Low Population Zone.......................................

5-1 l

l 5.2 CONTAINMENT q

Configuration.............................................

5-1

[

Design Pressure and Temperature...........................

5-4 I-P 5.3 REACTOR CORE i

Fu el A s s em bl i e s...........................................

5-4 Control Rods..............................................

5-4 5.4 REACTOR COOLANT SYSTEM i

Design Pressure and Temperature...........................

5-4 l

Vo1ume....................................................

5-5 r

5.5 METEORLOGICAL TOWER LOCATION..............................

5-5 5.6 FUEL STORAGE i

Criticality...............................................

5-5

- Drainage........................................'..........

5-5

(

Capacity..................................................

5-6 l

e 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT........................

5-6 t

i I

i t

l 1

DAVIS-BESSE, UNIT 1 XIV Amendment No. 38 a

I

+

INDEX ADMINISTRATIVE CONTROLS PAGE SECTION 6-1 6.1 RESPONSIBILITY..........................................

i I

6.2 ORGANIZATION 6-1 0ffsite.................................................

6-1 Facility Staff..........................................

6-5 l

6.3 FACILITY STAFF OUAL IFICATIONS...........................

l 6.4 TRAINING...............

6-5 t

6.5 REVIEW AND AUDIT 6.5.1 STATION REVIEW BOARD 6-5 i

Function..............................................

6-6 Composition...........................................

6-6 Al t e r n a t e s............................................

6-6 Mee ti n g Freoti ency.....................................

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

6-6 j

Respons i bi11 ti es......................................

6-7 l

Authority.............................................

6-8 Records...............................................

P 6.5.2 COMPANY NUCLEAR REVIEW BOARD c

6-8 Function.......................................'.......

6-9 Com po s i ti o n...........................................

6-9 A1tarnates............................................

I 6-9 Co'nsu1tants...........................................

i I

i DAVIS-BESSE, UNIT 1 XV Amendment No. 38 l

F n-.

n.

y_

r

INDEX ADMINISTRATIVE CONTROLC t

SECTION PAGE Me e t i ng F re c ue n cy.....................................

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

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

G-10 Audits.................'...............................

6-11 Authority.............................................

6-11 Records...............................................

6-12 i

6.6 REPORTABLE OCCURRENCE ACTI0N............................

6-12 6.7 SAFETY LIMIT VIOLATION..................................

6-13 i

6.8 PROCEDURES..............................................

6-13 r

6.9 REPORTING REQUIREMENTS t

I 6.9.1 ROUTINE REPORTS AND REPORTABLE ' 0CCURRENCES............ 6-14 6.9.2 SPECIAL REP 0RTS.......,...............................

6-18

[

6.10 RECORD RETEN1 ION.......................................

6-18 6.11 RADIATION PROTECTION PR0 GRAM...........................

6-20 6.12 HIGH RADIATION AREA....................................

6420 6.13 ENVIRONMENTAL: QUAL I F I CAT I O N...................'........ '

6-21 P

DAVIS-BESSE, UNIT 1 XVI Amendment No.38

3/4.4 REACTOR COOLANT SYSTEM I

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.

APPLICABILITY: MODES 1 and 2*.

ACTION:

l With one reactor coolant pump not in operation, STARTUP and a.

POWER OPERATION may be initiated and may pror.eed provided THERMAL POWER is restricted to less than 80.2% of RATED THERMAL POWER and within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> the setpoints for the following trips have been reduced to the values specified in Specification 2.2.1 for operation with three reactor coolant pumps operating:

1.

High Flux 2.

Flux-AFlux-Flow SURVEILLANCE REQUIREMENTS 4.4.1.1 The above required reactor coolant loops shall be verified to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

4.4.1.2 The Reactor Protective Instrumentation channels specified in the applicable ACTION statement above shall be verified to have had their trip setpoints changed to the values specified in Specification 2.2.1 for the applicable number of reactor coolant pumps operating either:

Within 4 houk after switching to a different pump combination if a.

the switch is made while operating, or b.

Prior to reactor criticality if the switch is made while shutdown.

TSee Special Test Exception 3.10.3.

r 1

g DAVIS-BESS'E, UNIT 1 3/4 4-1 Amendment No. JM )Y,' 38 i

l

.. ~ -

.-+-e.

[

i

' 3/4.4 REACTOR COOLANT SYSTEM SHUTDOWN AND H(fr_ STANDBY LTMITING CONDITION FOR OPERATION l

l 3.4.1. 2 a.

At least two of the coolant loops listed below shall be f

OPERABLE:

i 1.

Reactor Coolant Loop 1 and its associated steam' generator, i

2.

Reactor Coolant Loop 2 and its associated steam l

generator, i

3.

Decay Heat Removal Loop 1,*

4.

Decay Heat Removal Loop 2.*

b.

At least one of the above coolant loops shall be in operation.**

i l

Not more than one decay heat removal pump may be operated c.

with the sole suction path through DE-11 and DH-12 unless the control power has been rermved from the DR-11 and DE-12 valve operator, or manual valves DH-21 and DE-23 are l

j opened.

1 d.

The provisions of Specifications 3.0.3 and 3.0.4 are not i

applicable.

/

APPLICABILITY: MODES 3, 4 and 5 ACTION:

I With less than the above required caolant loops OPERABLE, a.

immediately initiate corrective action to return the required coolant loops to OPERABLE status as soon as possible, or be in COLD SHUTDOWN within 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />.

I I

b.

With none of the above required coolant loops in operation.

I suspend all operations involving a reduction in boron l

concentration of the Reactor Coolant System and immediately initiate corrective action to return the required coolant-i i

loop to operation.

l I

l

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

(

• • 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 concp tration, and (2) core outlet l

temperature is maintained at least 10 F below saturation temperature.

t i

3/4 4 Aracndment No..%.#,,2ff, 38 j

- DAVIS-BESSE UNIT 1 4

-.= ~ -*-..

. - pp

+*

3/4.4 REACTOR COOLANT SYSTEM i

SURVEILLANCE REQUIRE N S 4. 4.1.,2.1 The required dscay heat removal loop (s) shall be detemined OPERABLE per Specification 4.0'.5.

l

~

4.4.1.2.2 The required steam generator (s) shall be determined 7

i 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 <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> if an associated reactor coolant pump is operating, or, (b) 35 inches above the lower tube l

sheet once per ?2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> if no reactor coolant pumps are.

l operating.

i 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 <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

l l

t l

t I

[

I i

i I

l F

i

~

t' DAVIS-BESSE UNIT 1 3/4 4-2a i

l 1

REFUELING OPERATIONS CRANE TRAVEL - FUEL HANDLING BUILDING

[

LIMITING CONDITION FOR OPERATION I

i 3.9.7 Loads in excess of 2430 pounds shall be prohibited from travel over fuel assemblies in the spent fuel pool.

APPLICABILITY: With fuel assemblies and water in the spent fuel pool.

j l

ACTION:

With the requirements of the above specification not satisfied, place tha crane load in a safe condition. The provisions of Specification l

i 3.0.3 are not applicable.

l I

l l

1 SURVEILLANCE REQUIREMENTS I

I i

4.9.7 The weight of each load, other than a fuel assembly, shall be verified to be < 2430 pounds prior to moving it over fuel assemblies in the spent fuel pool.

l I

l t

l

[

l e

i-I i

1 DAVIS-BESSE, UNIT 1 3/4 9-7 9

~ * '

+

REFUELING OPERATIONS 3/4.9.8 DECAY fiEAT REMOVAL AND COOLANT CIRCULATION-ALL WATER LEVELS LIMITING CONDITION FOR OPERATION:

3.9.8.1 At least one decay heat removal loop shall be in operation.

APPLICABILITY: MODE 6 When the water level above the top of the irradiated fuel assemblies seated within the reactor pressure vessel is > 23 feet.

ACTION:

a.

With less than one decay heat renoval loop in operation, except as provided in b below, suspend all operations involving an increase in the reactor decay heat load or a reduction in boron ccncentration of the Reactor Coolant System. Close all containment penetrations providing direct access -from the containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, b.

The decay heat removal loop may be removed from operation for up to one hour per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period durino the performance of CORE ALTERATIONS in the vicinity of the reactor pressure vessel (hot) legs.

c.

The provisions c' Specification 3.0,3 are~ not applicable.

i SURVEILLANCE REQUIREMENTS 4.9.8.1 Surveillance at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shall verify at least one decay heat removal loop to be in operation and circulating reactor coolant through the reactor core:

a.

At a flow rate of > 2800 gpm, whenever a reduction in Reactor Coolant System boron concentration is being made.

b.

At a flow rate such that the core outlet temprature is maintained i

1140*F, provided no reduction in Reactor Coolant System' boron concentration is being made.

i i

e j

O me S

0 DAVIS-BESSE,. UNIT 1 3/4 9-8 Amendment No. 38

+

w

==-.m-

REFUELING OPERATIONS LOW WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.8.2 Two independent DHR loops shall be OPERABLE.*

APPLICABILITY: MDDE 6 when the water level above the top of the irradiated fuel assenblies seated within the reactor pressure vessel is less than 23 feet.

ACTION:

a.

With less than the required DHR loops OPERABLE, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible.

b.

The provisions of Specification 3.0.3 are not appifcable, i

SURVEILLANCE REQUIREMENTS 4.9.8.2 At least one DHR loop shall be determined to be in operation per Speci-fication 4.9.8.1.

The inactive loop shall be determined to be OPERABLE per S' ecification 4.0.5.

p I

• The normal or emergency power source nay be inoperable for each DHR loop.

j l

DAVIS-BESSE, UNIT t 3/4 9-8a Amendment No. 38

3/4.4 REACTOR C00 TRIT SYSTEM BASES 3/4.4.1 REACTOR COOLANT LOOPS The plant is designed to operate with both reactor coolant loops in operation, and maintain DNBR above 1.30.during all normal operations and anticipated transients. With one reactor coolant pump not in operation in one loop, THERMAL POWER is restricted by the Nuclear Overpower Based on RCS Flow and AXIAL POWER IMBALANCE, ensuring that the DNBR will be maintained above 1.30 at the maximum possible THERNAL POWER for the number of reactor coolant pumps in operation or the local quality at the point of minimum DNBR equal to 22%, whichever is more restrictive.

In HODES 3, 4 had 5, a single reactor coolant loop or DHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops be OPERABLE.

Thus, if the reactor coolant loops are not OPERABLE, this specification requires two DER loops to be OPERABLE.

r Natural circulation flow or the operation of one DER pump provides i

adequate flow to ensure mixing, prevent stratification and produce i

gradual reactivity changes during boron concentration reductions.in the I

Raac. tor Coolant System. The reactivity change rate associated with boron reduction will, therefore, be within the capacity of operator recognition and control.

3/4.4.2 and 3/4.4.3 SAFEir VALVES The pressuri:er code safety valves operate to prevent the RCS from being pressuri:ed above its Safety Limit of 2750 psig. Each safety valve is casigned to relieve 336,000 lbs per hour of saturated steam at the valve's setpoint.

The relief capacit/ of a single safety valve is adequate to relieve any overpressure condition which could occur during shutcown.

In the event that no safety valves are OPEMBLE, an operating CHR loop, con-nected to the RCS, provides overpressure relief' capability and will prevent RCS overpressuri:ation.

j During oceration, all pressuri:er code safety valves must be OPEMBLE to prevent tne RCS from being pressuri:ed above its safety limit of 2750 psig. The comoined relief capacity of all of these valves is greater than the maximum surge rate resulting from any transient.

Demonstratien of the safety valves' lift settings wi.ll oc:ur only curing snutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Coce.

i I

DAVIS-BESSE UNIT 1 B 3/4 4 I Amendnent No. #f, 38 P00ROR81M

l 3/4.9 REFUELING OPERATIONS BASES 3/4.9.1 BORON CONCENTRATION i

The limitations on reactivity conditions during REFUELING ensure that:

i

~

1) the reactor will remain subcritical' during CORE ALTERATIONS, and 2) a j

uniform baron concentration is maintained for reactivity control in the i

water volumes having direct access to the reactor vessel. These limita-l tions are consistent with the initial conditions assumed for the boron dilution incident in the accident analysis.

j i

3/4.9.2 INSTRUMENTATION The OPERABILITY of source range neutron flux monitors ensures that l

redundant monitoring capability is available to detect changes in the

[

reactivity condition of the core.

j t

3/4.9.3 DECAY TIME

}

The minimum requirement for reactor subcriticality prior to movecent l

of irradiated fuel assemblies in the reactor pressure vessel ensures that sufficient time has elapsad to allow the radioactive-decay of the i

short lived fission products. This decay time is consistent with the i

I assumptions used in the safety analyses.

3/4.9.4 CONTAINMENT PENETRATIONS J'

The requirements on containment penetration closure and OPERABILITY ensure that a release of radioactive material within containment will be r

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-i tion potential while in the REFUELING MODE.

l i

3/4.9.5 COMMUNICATIONS The requirement for comunications capability ensures that refueling i

station personnel can be promptly infomed of sfgnificant changos fr. the facility status or core reactivity condition during CORE ALTERATIONS.

i OAVIS-BESSE, UNIT 1 B 3/4 9-1 l

~

l i

i

REFUELING OPERA' IONS 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 will be used for movement of control rods and fuel assemblies, 2) each hoist has sufficient load capacity to 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 assembly in a failed fuel container over other fuel assemblies in the storage pool ensures that in the event this i

load is dropped (1) the activity release will be limited to that contained in a single fuel assembly, and (2) any possible distortion of fuel in the storage racks will not result in a critical array. This assumption is consistent with the activity release assumed in the accident analyses.

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 pressurc 1

vessel below 140'F as required during the REFUELING MODE, and (2) sufficient coolant circulation is maintained through the reactor core to minimize the ef feet of a boron dilution incident and prevent boron stratification.

l The requirement to have two DHR loops OPERABLE when there is less than 23 feet of water above the core ensums that a single failure of the operating DHR loop will not result in a complete loss of decay heat removal capability.

With the reactor vessel head removed and 23 feet of water above the com, a large heat sink is available for core cooling. Thus, in the event of a faile: of the operating DHR loop, adequate time is provided to initiate eme,rg;encyprocedurestocoolthecore.

3/4.9.[9.CONTAINMENTPURGEANDEXHAUSTISOLATIONSYSTEM The OPERABILITY of this system ensures that the containment purge and exhaust penetrations will be automatically isolated upon detection of high radiation levels within the containment. The OPERABILITY of

this system is required to restM ct the release of ra.dioactive material from the containment atmosphere to the environment.

3/4.9.10 and 3/(.9.11 WATER LEVEL - REACTOR VESSEL AND STORAGE POOL <

WATER LEVEL The restrictions on minimum water level ensure that sufficient water depth is available to remove 99% of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly. The minimum water depth is consistent with the assumptions ;,f the safety analysis..

DAVIS-BESSE, UNIT 1 B 3/4 9-2 Amendment No. 38 T