Text

Amends 169,169 & 140 to Licenses DPR-33,DPR-52 & DPR-68, Respectively,Deleting Certain Surveillance Testing Requirements on Redundant Emergency Core Cooling & Reactor Core Isolation Sys During Limiting Condition for Operation
	ML20245F645
Person / Time
Site:	Browns Ferry
Issue date:	08/02/1989
From:	Black S; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20245F648	List: ML20245F645; ML20245F650;
References
	NUDOCS 8908150022
	Download: ML20245F645 (117)
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e UNITED STATES

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NUCLEAR REGULATORY COMMISSION p

WASHINGTON, D. C 20555

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. TENNESSEE VALLEY AUTHORITY 00CKET NO. 50-259 BROWNS FERRY NUCLEAR PLANT, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.169 License No. DPR-33 l'.

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

A.

The application for amendment by Tennessee Valley Authority ;the licensee) dated January 13, 1989, complies 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; 8.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and egulations of the Comission;

.C.

There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (11) that such activities wfll be conducted in compliance with the Comission's regulations; D.

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

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

8908150022 890630 DR ADOCK O.*000259.

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c.

l 1. 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. DPR-33 is hereby l

amended to read as follow!,:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.169, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective es of its date of issuance and shall be implemented within 60 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION h

Suzanneklack,AssistantDirector for Projects TVA Projects Division Office of Nuclear Reactor Regulation A*tachmert:

1 Chu.'.ces to the Technical Specifications Date of Issuance: August 2, '.989 l

i l

I

__...m_________.___.__.

n

ATTACHMENT TO LICENSE AMENDMENT NO. 169 FACILITY OPERATING LICENSE NO. DPR-33 DOCKET NO. 50-259 Revise the Appen'ix A Tech.ical Specifications by removing the pages d

identified below and inserting the enclosed pages.

The revised pages are identified by the carcioned amendment number and contain marginal f

. lines-indicating the area of change. Table of Contents' and overleaf pages*

are provided tc maintain document completeness.

REMOVE INSERT i

i 11 11 j

iii its iv iv v

v vi vi vii v11 viii viii 3.5/4.5-1 3.5/4.5-1*

3.5/4.5-2 3.5/4.5-2 3.5/4.5-3 3.5/4.5-3*

3.5/4.5-4 3.5/4.5-4 3.5/4.5-5 3.5/4.5-5 3.5/4.5-6 3.5/4.5-6 3.5/4.5-7 3.5/4.5-7 3.5/4.5-8 3.5/4.5-8 3.5/4.5-9 3.5/4.5-9 3.5/4.5-10 3.5/4.5-10 3.5/4.5-11 3.5/4.5-11*

3.5/4.5-11a 3.5/4.5-12 3.5/4.5-12 3.5/4.5-13 3.5/4.5-13*

3.5/4.5-14 3.5/4.5-14 3.5/4.5-15 3.5/4.5-15 1

3.5/4.5-16 3.5/4.5-16 l

3.5/4.5-17 3.5/4.5-17 3.5/4.5-26 3.5/4.5-26 3.5/4.5-27 3.5/4.5-27 3.5/4.5-28 3.5/4.5-2E 3.5/4.5-29 3.5/4.5-29*

3.5/4.5 3.5/4.5-30 3.5/4.5-31 3.5/4.5-31*

3.5/4.5-32 3.5/4.5-32*

3.5/4.5-33 3.5/4.5-33 3.5/4.5-34 3.5/4.5-34*

3.5/4.5-35 3.5/4.5-75 1

2.sillh>

TABLE OF CONTENTS Eiction Page No.

1.0 Definitions.

1.0-1 SAFETY LIMITS AND LIMITING SAFETY SYSTEM EETTINGS

'1.1/2.1 Fuel Cladding Integrity.

1.1/2.1-1

'1.2/2.2 Reactor Coolant System Integrity.

1.2/2.2-1 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3.1/4.1 Reactor Protection System.

3.1/4.1-l' 3.2/4.2 Protective Instrumentation.............

3.2/4.2-1 A.

Primary Containment and Reactor Building Isolation Functions.

3.2/4.2-1 B.

Core.and Containment Cooling Systems -

Initiation and Control.

3.2/4.2-1 C.

Control Rod Block Actuation.

3.2/4.2-2 D.

Radioactive Liquid Effluent Monitoring Instrumentation..

3.2/4.2-3 E.

Drywell Leak Detection.........

3.2/4.2-4 F.

Surveil 2ance Instrumentation......

3.2/4.2...

G.

Control Room Isolation........

3.2/4.2-4 H.

Flood Protection.

3.2/4.2-4 I.

Meteorological Monitoring Instrumentation.

3.2/4.2-4 J.

Seismic Monitoring Instrumentation..

3.2/4.2-5 K.-

Radioactive Gaseous Effluent Monitoring Instrumentation 3.2/4.2-6 L.

ATWS-Recirculation Pump Trip.

3.2/4.2-6a 3.3/4.3 Reactivity Control.

3.3/4.3-1 A.

Reactivity Limitations.

3.3/4.3-1 B.

Control Rods.

3.3/4.3-5 e

C.

Scram Insertion Times.

3.3/4.3-10 i

Amendment No. 129, 161, BFN Unit 2

Sectien Pane No.

D.

Reactivity Anomalies.

3.3/4.3-11 E.

Reactivity Control..

3.3/4.3-12 F.

. Scram Discharge Volume.

3.3/4.3-12

.3.4/4.4

. Standby Liquid Control System.

3.4/4.4-1 A.

Normal System Availability.

3.4/4.4-1 B.

Operation with Inoperable Components.

3.4/4.4-3 C.

Sodium Pentaborate Soluticn.....

3.4/4.4-3 3.5/4.5 Core and Containment Cooling Systems.

3.5/4.5-1 A.

Core Spray System (CSS).

3.5/4.5-1 B.

Residual Heat Removal System (RHRS)

(LPCI and Containment Cooling)..

3.5/4.5-4 C.

RHR Service Water and Emergency Equipment Cooling Water Systems (EECWS)...

3.5/4.5-9 D.

Equipment Area Coolers.

3.5/4.5-13

.E.

h.'eh. Pressure Coolant Injection System (2"CIS).

3.5/4.5-13 F.

Reactor om't Isolation Cooling System (RCICS).

3.5/4.5-14 G.

Automatic Depressurization System (ADS)..

3.5/4.5-16 H.

Maintenance of Filled Discharge Pipe.

3.5/4.5-17 I.

Average Planar Linear Heat Generation Rate..

3.5/4.5-18 J.

Linear Heat Generation Rate (LHGR).

3.5/4.5-18 K.

Minimum Critical Power Ratio (MCPR).

3.5/4.5-19 L.

APRM Setpoints.

3.5/4.5-20 3.6/4.6 Primary System Boundary.

3.6/4.6-1 A.

Thermal and Pressurization Limitations.

3.6/4.6-1 E.

Coolant Chemistry.

3.6/4.6-5 C.

Coolant Leakage.

3.6/4.6-9 D.

Relief Valves.

3.6/4.6-10 11 Amendment No. 129, 143, 161, 169 BFN Unit 2

U lbh l

E.5 l.

1Section.

Pane No.

l E.

Jet Pumps.

3.6/4.6-11 F.

Recirculation Pump Operation.

.3.6/4.6-12 G.

Structural Integrity.

3.6/4.6-13

,I H.

Snubbers.

3.6/4.6-15 l

3.7/4.7' Containment Systems.

3.7/4.7-1 A.

Primary Containment.

3.7/4.7-1 B.

. Standby Gas Treatment System.

3.7/4.7-13

)

..)

C.

Secondary Containment.

3.7/4.7-16 j

D.

Primary Containment Isolation Valves.

3.7/4.7-17 1

E.

Control Roora Emergency Ventilation......

3.7/4.7-19 I

F.

Primary Containment Purge System.

2.7/4.7-21 G.

Containment Atmosphere Dilution System (CAD).

3.7/4.7-22 H.

Containment Atmosphere Monitoring (CAM)

System H2 Analyzer............

3.7/4.7-24 3.8/4.8 Radioactive Materials...

3.8/4.8-1 A.

Liquid Effluents.

3.8/4.8..............

B.

Airborne Effluents.

3.8/4.8-3 C.

Radioactive Effluents - Dose.

3.8/4.8-6 D.

Mechanical Vacuum Pump.

3.8/4.8-6 E.

Miscellaneous Radioactive Materials Sources..

3.8/4.8-7 T.

Solid Radwaste.

3.8/4.8-9 3.9/4.9 Auxiliary Electrical System.

3.9/4.9-1 A.

Auxiliary Electrical Equipment 3.9/4.9-1 B.

Operation with Inoperable Equipment.

3.9/4.9-8 C.

Operation in Cold Shutdown.

3.9/4.9-15 3.10/4.10 Core Alterations.

3.10/4.10-1 A.

Refueling Interlocks.

3.10/4.10-1 B.

Core Monitoring.

3.10/4.10-4 Amendment No. 128, 160, 169 iii BEN Unit 2

_ - _ _ - - - - - - -. - - - - - - - -. - _ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ~ - - - - -

.d e.

Section Page Wo.

C.

Spent Fuel Pool Water.

3.10/4.10-7 D.

Reactor Building Crane.

3.10/4.10-8 E.

Spent Fuel Cask.

3.10/4.10-9 F.

Spent Fuel Cask Handling-Refueling Floor.

3.10/4.10-10 3.11/4.11 Fire Protection Systems 3.11/4.11-1 A.

Fire Detection Instrumentation...

3.11/4.11-1 B.

Fire Pumps and Water Distribution Mains 3.11/4.11-2 C.

Spray and/cr Sprinkler Systems.

3.11/4.11-7 D.

CO2 Systems.

3.11/4.11-8 E.

Fire Hose Stations.

3.11/4.11-9 F.

Yard Fire Hydrants and Hose Houses.

3.11/4.11-11 G.

Fire-Rated Assemblies 3.11/4.11-12 H.

Open Flames, Welding and Burning in the Cabic Spreading Room.

3.11/4.11-13 l

5.0 Major Design Features 5.0-1 5.1 Site Features 5.0-1

)

5.2 Reactor 5.0-1 5.3 Reactor Vessel.

5.0-1 5.4 Containment 5.0-1 5.5 Fuel Storage.

5.0-1 5.6 Seismic Design.

5.0-2 iv BFN Unit 2 Amendment No. 134,159,169

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ADMINISTRATIVE CONTROLS SECTION PAGE 121 RESPONSIBILITY..........................................

6.0-1 122 ORGANIZATION............................................

6.0-1 6.2.1 Of fsite and Onsite Organizations........................

6.0-1 6.2.2 Plant Staff.............................................

6.0-2 12]

PLANT STAFF 0UALI FI CATI ONS..............................

6. 0-5 124 TR A I NI N G................................................ 6. 0- 5 121 PLANT REVIEW AND AUDIT..................................

6.0-5 1 ~

6.5.1 Plant Operations Review Committee (P0RC)................

6.0-5 l

l 1

6.5.2 Nuclear Safety Review Board (N3RB)......................

6.0-11 6.5.3 Technical Review and Approvcl of Procedures.............

6.0-17 l

121 REPORT ABLE EVENT A CTIONS................................

6. 0-18 6.2 SAFETY LIMIT VIOLATION..................................

6.0-19 121 PROCEDURES / INSTRUCTIONS AND PROGRAMS....................

6.0-20 6.8.1 Procedures..............................................

6.0-20 6.8.2 Drills..................................................

6.0-21 i

6.8.3 Radiation Control Procedures............................

6.0-22 6.8.4 Quality Assurance Procedures - Effluent and Environmental Monitoring............................

6.0-23 123 PEPORTING RE0UIREMENTS..................................

6.0-24 6.9.1 Routine Reports.........................................

6.0-24 S t a r t up R e p o r t s......................................... 6. 0-2 4 Annual Operating Report.................................. 6.0-25

{

Monthly Operating Report................................. 6.0-26 Reportable Events.......................................

6.0-26

]

Radioactive Effluent Release Report...................... 6.0-26 l

1 Source Tests............................................

6.0-26 6.9.2 Special Reports.........................................

6.0-27

)

1 6.10 STATION OPERATING RECORDS AND RETENTION................

6.0-29 i

6.11 PR O CE S S CO NTR OL PR0 GRAM.................................

6. 0-3 2 6.12 0FFSITE DOSE CALCULATION MANUAL.........................

6.0-32 6.13 RADIOLOGICAL EFFLUENT MANUAL............................

6.0-33 V

BFN Unit 2 Amendment No. 124,161,169

we i.

1 LIST OF TABLES s

Table Title Pane No.

1 1.1 Surveillance Frequency Notation.

1.0-13 l

3.1.A Reactor Protection System (SCRAM)

Instrumentation Requirements.

3.1/4.1-3 4.1.A Reactor Protection System (SCRAM) Instrumentation l

Functional Tests Minimum Functional Test Frequencies for Safety Instr. and Control Circuits.

3.1/4.1-8 1

4.1.3 Reactor Protection System (SCRAM) Instrumentation

]

Calibration Minimum Calibration Frequencies for Reactor Protection Instrument Channels.

3.1/4.1-11 3.2.A Primary Containment and Reactor Building Isolation Instrumentation.

3.2/4.2-7 3.2.B Instrumentation that Initiates or Controls the l

Core and Containment Cooling Systems.

3.2/4.2-14 S.2.C Instrumentation that Initiates Rod Blocks.

3.2/4.2-25 3.2.D Radioactive Liquid Effluent Monitoring Instrumentation.

3.2/4.2-28 3.2.E Instrumentation that Monitors Leakage Into Drywell.

3.2/4.2-30 3.2.F Surveillance Instrumentation.

3.2/4.2-31 3.2.G Control Room Isolation Instrumentation.

3.2/4.2-34 v

e-3.1.H Flood Protection Instrumentation.

3.2/4.2-35 3.2.I Meteorological Monitoring Instrumentation.

3.2/4.2-36 3.2.J Seismic Monitoring Instrumentation..

3.2/4.2-37 3.2.V Radioactive Gaseous Effluent Monitoring l

Instrumentation.

3.2/4.2-38 3.2.L ATWS-Recirculation Pump Trip (RPT) Surveillance Instrumentation.

3.2/4.2-39b

.I -

4.2.A Surveillance Requirements for Primary Containment and Reactor Building Isolation Instrumentation.

3.2/4.2-40 4

4.2.B Surveillance Requirements nor Instrumentation that Initiate or Control the CSCS.

3.2/4.2-44 4.2.C Surveillance Requirements for Instrumentation that Initiate Rod Blocks.

3.2/4.2-50 4

4.2.D Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements.

3.2/4.2-51 vi Amendment No. 128, 161, 169 BTN Unit 2

.b !

Le>:

LIST OF_lAELES-(Cont'd)

I1111 Eagg No.

^ Iable Minimum Test and Calibration Frequency for Drywell 4.2.E 3.2/4.2-53 Leak "etection Instrumentation...

.'4. 2.F Minimum Test and Calibration Frequency for 3.2/4.2-54 Surveillance Instrumentation.

Surveillance Requirements for Control Room 4.2.G 3.2/4.2-56 Isolation Instrumentation.

Minimum Test and Calibration Frequency for 4'.2.H 3.2/4.2-57 Flood Protection Instrumentation.

i Seismic Monitoring Instrument Surveillance 4.2.J 3.2/4.2-58 Requirements...................

Radioactive Gaseous Effluent Instrumentation 4.2.K 3.2/4.2-62

-Surveillance.

ATWS-Recirculation Pump Trip (RPT) 4.2.L 3.2/4.2-63a

~ Instrumentation Surveillance.

3.5/4.5-11 3.5-1 Minimum RHRSW and EECW Pump Assignment.

3.5/4.5-21 3.5.I MAPLHGR Versus Average Planar Exposure.......

3.7/4.7-25 3.7.A Primary Containment Isolation Valves.

3.7/4.7-32 Testable Penetrations with Double 0-Ring Seals.

3.7.B 3.7/4.7-33 Testable Penetrations with Testable Bellows.

3.7.C 3.7/4.7-34 3.7.D Air Tested Isolation Valves.....

Primary Containment Isolation Valves which

.3.7.E 3.7/4.7-37 Terminate below the Suppression Pool Water Level.

Primary Containment Isolation Valves Located in 3.7.F 3.7/4.7-38 Water Sealed Seismic Class 1 Lines.

3.7/4.7-39 3.7.H Testable Electrical Penetrations.

3.9/4.9-16 4.9.A Diesel Ocnerator Reliability.

3.9/4.9-18 4.9.A.4.C Voltage Relay Setpoints/ Diesel Generator Start.

f 3.11/4.11-14 3.11.A Fire Detection Instrumer.tation.

3.11/4.11-17 3.11.B Spray / Sprinkler Systems..

3.11/4.11-18 1

3.11.C Hose Stations..

3.11/4.11-20 Yard Fire Hydrants and Fire Hose Houses.

3.11.D 6.0-4 6 2.A Minimum Shift Crew Requirements.

vii Amendment No. 149. P, 169 BFN Unit 2

l l

we 4

1 I

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LIST OF ILLUSTRATIONS Figure Title Pane No.

2.1.1 APRM Flow Reference Scram and APRM Rod Block Settings.

1.1/2.1-6 2.1-2 APRM Flow Bias Scram Vs. Reactor Core Flow.

1.1/2.1-7 4.1-1 Graphical Aid in the Selection of an Adequate Interval Between Tests.

3.1/4.1-13 4.2-1 System Unavailability.

3.2/4.2-64 3.5.K-1 MCPR Limits.

3.5/4.5-22 3.5.2 Kf Factor.

3.5/4.5-23 3.6-1 Minimum Temperature *F Above Change in Transient Temperature.

3.6/4.6-24 3.6-2 Change in Charpy V Transition Temperature Vs.

Neutron Exposure.

3.6/4.6-25 4.8.1.a Gaseous Release Po'nts and Elevations 3.8/4.8-10 4.8.1.b Land Site Boundary.

3.8/4.8-11 l

i viii Amendment No. 169 BFN Unit 2

..t:

'h>

3.5/4.5 CORE AND CONTAINMENT COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5 CORE AND CONTAINMENT COOLING 4.5 QQfE AND CONTAINMENT COOLING SYSTEMS ETSTEMS Aeolicability Aeolicability Applies to the operational Applies to the surveillance status of the core and requirements of the core and containment cooling systems.

containment cooling systems when the corresponding limiting condi-tion for operation is in effect.

Objective Obiective To assure the OPERABILITY of To verify the OPERABILITY of the the core and containment cooling core and containment cooling systems under all conditions for systems under all conditions for which this cooling capability is which this cooling capability is an essential response to plant an essential response to plant abnormalities.

abnormalities.

Specification Specification A.

Core Soray System (CSS)

A.

Care Soray System (CSS) 1.

The CSS shall be OPERABLE:

1.

Core Spray System Testing.

(1) PRICR TO STARTUP Liam Frecuency from a COLD CONDITION, or a.

Simulated Once/

Automatic Operating (2) when there is irradiated Actuation Cycle fuel in the vessel test and when the reactor vessel pressure b.

Pump Opera-Per Specifi-is greater than bility cation 1.0.MM atmospheric pressure, except as specified c.

Motor Per Specifi-in Specification Operated cation 1.0.M14 3.5.A.2.

Valve OPERABILITY d.

System flow Once/3 rate: Each months loop shall deliver at least 6250 gpm against a system head corres-ponding to a BFN 3.5/4.5-1 Amendment No. 155 Unit 2 Y-______-_____________

<s

'3.5/4.5' CORE AND CONTAINMENT COOLING SYSTJMS' LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5 A Core Sorav System (CSS)'

4.5. A ' Core Sorav' System (CSS) l_

i 4.5.A.1.d (Cont'd) 1 L

l l<

105 psi

)

differential pressure between the reactor vessel l-and the primary l

containment.

e.

Check Valve Per-Specification 1.0.MM i

2.

If one CSS loop is inoperable, f.

Verify that Once/ Month the reactor may remain in each valve operation for a period not to (manual, power-exceed 7 days providing operated, or all active components in automatic) in the the other CSS loop and the injection flowpath RHR system (LPCI mode) that is not locked, and the diesel generators sealed, or other-are OPERABLE.

vise secured in I

position, is in its correct

position.

3.

If Specification 3.5.A.1 or 2.

No additional surveillance Specification 3.5.A.2 cannot is required.

I be met, the reactor shall be placed in the COLD SEUIDOWN CONDITION within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

4.

When the reactor vessel pressure is atmospheric and irradiated fuel is in the reactor vessel at least one core spray loop with one OPERABLE pump and associated Except that an automatic diesel generator shall be valve capable of automatic OPERABLE, except with the return to its ECCS position reactor vessel head removed when an ECCS signal is as specified in 3.5.A.5 or present may be in a PRIOR TO STARTUP as position for another mode specified in~3.5.A.1.

of operation.

BFN 3.5/4.5-2 knendment No. 149, 155, 169 Unit 2

^ g f 33/4.5 CORE ' AND C014TAINMEfft_f00 LING Syji7JJ11 LIMITING CONDITIONS FOR OPI: RATION SURVEILLANCE REQUIREMENTS 3.5.A Core Soray System (CSS) 5.

When. irradiated fuel is in the reactor vessel and the reactor vessel head is removed, core spray is not required to be OPERABLE provided the cavity is flooded, the fuel pool J

. gates are open and the J

fuel pool water level is maintained above the low level alarm point, and provided one RHRSW pump and associated valves supplying the standby coolant supply are 7

OPERABLE.

k When work is in progress which has the potential to drain the vessel, manual initiation capability of either 1 CSS Loop or 1 RHR pump, with the capability of injecting water into the reactor vessel, and l

the associated diesel I

generator (s) are required.

l l

I

)

3.5/4.5-3 Amendment No. 158 BFN Unit 2

um 2

3.5/4.5 CORE AND CONTAINMENT COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.B Eggidual Heat Removal System 4.5.B. Residual Heat Removal System (RHRS) (LPCI and Containment (RHRS) (LPCI and Containment Cooling)

Cooling) 1.

The RHRS shall be OPERABLE:

1. a.

Simulated Once/

Automatic Operating (1) PRIOR TO STARTUP Actuation Cycle from a COLD Test CONDITION; or (2) when there is b.

Pump OPERA-Per irradiated fuel in BILITY Specification the reactor vessel 1.0.MM and when the reactor vessel pressure is c.

Motor Opera-Per greater than ted valve Specification atmospheric, except as OPERABILITY 1.0.MM specified in Specifications 3.5.B.2, d.

Pump Flow Once/3 through 3.5.B.7.

Rate months e.

Testable Per Check Specification Valve 1.0.MM f.

Verify that Once/ Month each valve (manual, power-operated, or automatic) in the injection flow-path that is not locked, sealed, or otherwise secured in posi-tion, is in its correct

position, g.

Verify LPCI Once/ Month subsystem cross-tie valve is closed and power removed from velve operator.

Except that an automatic valve capable of auto-matic return to its ECCS position when an ECCS signal is present may be in a position for another mode of operation.

BFN 3.5/4.5-4

7---_-____

i%n

^ me.

l' 3.5/4.5 CORE AND CONTAINMENT COOLING SYSTEMS SURVEILLANCE REQUIREMENT LIMITING CONDITIONS FOR OPERATION 4.5.B. Residual Heat Removal._ System 3.5.B Rggidual Heat Removal System (RHRS) (LPCI and Containment LEEB11-(LPCI and Containment Cooling)-

Cooling) 4.5.B.1 (cont'd)

Each LPCI pump shall deliver 2.

With the reactor vessel 9000 gpm against an indicated pressure less'than 105 psig, system pressure of 125 psig.

the RHRS may be removed Two LPCI pumps in.the same from service (except that two loop shall deliver 12000 gpm RHR pumps-containment cooling against an indicated system mode and associated heat pressure of 250 psig.

exchangers must remain OPERABLE) for a period not An air test on the drywell 2.

to exceed 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months while and torus headers and nozzles being drained of shall be conducted once/5 suppression chamber quality A water test may be years.

water and filled with performed on the torus header primary coolant quality in lieu of the air test.

water provided that during cooldown two loops with one pump per loop or one loop with two pumps, and associated diesel generators, in the core spray system are OPERABLE.

3.

No additional surveillance If one RHR pump (LPCI mode)

required, 3.

is inoperable, the reactor may remain in operation for a period not to exceed 7 days provided the' remaining RHR pumps (LPCI mode) and both access paths of the RHRS (LPCI mode) and the CSS and the diesel generators remain OPERABLE.

4 No additional surveillance 4.

If any 2 RHP pumps (LPCI required.

mode) become inoperable, the reactor shall be placed in the COLD SHUTDOWN CONDITION within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Amendment No. 149, 169 3.5/4.5-5 BEN Unit 2 1

=a 1 5/4.5 CORE AND CONTAINMENT COOLING SYSTEMS LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5 B.

Residual Heat Removal System 4.5 B.

Residual Heat Removal System (RHRS) (LPCI and Containment (RHRS) (LPCI and Containment Cooling)

Cooling) 5.

If one RER pump (containment 5.

No additional surveillance cooling mode) or associated required.

heat exchanger is inoperable, the reactor may remain in operation for a period not to exceed 30 days provided the remaining RHR pumps (containment cooling mode) and associated heat exchangers and diesel generators and all access paths of the RHRS (containment cooling mode) are OPERABLE.

6.

If two RHR pumps (containment 6.

No additional surveillance cooling mode) or associated required.

heat exchangers are inoperable, the reactor may remain in operation for a period not to exceed 7 days provided the remaining RHR pumps (containment cooling mode), the associated heat exchangers, diesel generators, and all access paths of the RHES (containment cooling mode) are OPERABLE.

7.

If two access paths of the 7.

No additional surveillance RHRS (containment cooling required.

mode) for each phase of the mode (drywell sprays, suppression chamber sprays, and suppression pool cooling) are not OPERABLE, the unit may remain in operation for a period not to exceed 7 days provided at least one path for each phase of the mode remains OPERABLE.

BFN 3.5/4.5-6 Amendment No. 149, 169 Unit 2

b en -

3.5/4. 5 ' CORE AWD CONTAINMEi4T COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.B Residual Heat Removal System 4.5.B Residual Heat Removal System (RHRS1 (LPCI and Containment (RHRSF (LPCI and Containment Cooling)

Cooling)

8. If Specifications 3.5.B.1 8.

No additional surveillance through 3.5.B.7 are not met,

required, an orderly shutdown shall be initiated and the reactor shall be placed in the COLD SHUTDOWN CONDITION within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

9.

When the reactor vessel 9.

When the reactor vessel pressure is atmospheric and pressure is atmospheric, irradiated fuel is in the the RHR pumps and valves reactor vessel, at least.one RHR that are required to be loop with.two pumps or two loops OPERABLE shall be with one pump per loop shall demonstrated to be OPERABLE be OPERABLE.

The pumps' per Specification 1.0.MM.

associated' diesel generators must also be OPERABLE.

10. If the conditions of 10.

No additional surveillance Specification 3.5.A.5 are met, required.

LPCI and containment cooling are not required.

11. When there is irradiated fuel 11.

The RHR pumps on the in the reactor and the reactor adjacent units which supply vessel pressure is greater.than cross-connect capability atmospheric, 2 RHR pumps and shall be demonstrated to be

- associated heat exchangers and OPERABLE per Specification valves on an adjacent unit must 1.0.MM when the cross-be OPERABLE and capable of connect capability supplying cross-connect is required.

capability except as specified in Specification 3.5.B.12 below.

(Note: Because cross-connect capability is not a short-term requirement, a component is not considered inoperable if cross-connect capability can be restored to service within 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months .)

BFN 3.5/4.5-7 Amendment No. 155, 169 Unit 2

)

ap 1,3/4.5-CORE AWD CONTAXWMEMT COOLING SYSTltLS '

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.B. Residual Heat Removal System 4.5.B Residual Heat Removal System (RHRS) (LPCI and Containment (RHRS).(LPCI and Containment p

Cooling)

12. If three RER pumps or associated

12. No additional surveillance heat exchangers located required.

on the unit cross-connection in the adjacent units are inoperable for any reason (including valve inoperability, pipe break, etc.), the reactor may remain.in operation:

for a period not to exceed

30 days provided the remaining RHR pump;and associated diesel generator are OPERABLE.

13.

If RER cross-connection flow or 13.

No additional surveillance heat removal capability is lost,

required, the unit may. remain in operation for a period not to exceed 10 days unless such capability is restored.

14.

All recirculation pump 14.

All recirculation pump discharge valves shall discharte valves shall be OPERABLE P710R TO be tested for OPERABILITY STARTUP (or closed if during any period of permitted elsewhere COLD SHUTDOWN CONDITION in these specifications) exceeding 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , if OPERABILITY tests have not been performed during the preceding 31 days.

l i

BFN 3.5/4.5-8 Amendment No. 149, 155, 169

(

Unit 2 j

__o

4

/_4J CORE AWD CONTAINMENT CQ0LJWG SYSTEMS

'ITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS

.C RER Service Water and Emergenc" 4.5.C BHR Service Water and Eh.erRenCV E.gpirment Cooling Water Systems Eauipment CoolinR Water Systems

[EECWS)

(EECWS) 1.

PRIOR TO STARTUP from 1.

a.

Each of the RHRSW pumps a COLD CONDITION, 9 RHRSW normally assigned to pumps must be OPERABLE, with automatic service on 7 pumps (including one of the EECW headers will pumps D1, D2, B2 or B1) be tested assigned to RHRSW service automatically each time and 2 automatically starting the diesel generators pumps assigned to EECW are tested.

Each of service.

the RHRSW pumps and all associated essential control l

valves for the EECW I

headers and RHR heat exchanger headers shall be demonstrated to be OPERABLE in accordance with Specification 1.0.MM.

b.

Annually each RHRSW pump shall be flow-rate tested. To be considered OPERABLE, each pump shall pump at least 4500 gpm through its normally assigned flow path.

c.

Monthly verify that each valve (manual, power-operated, or automatic) in the flowpath servicing safety-related equipment in the affected unit that is not locked, sealed, or otherwise secured in position, is in its correct position.

I l

l l

3.5/4.5-9 Amendment No. 155, 169

-2 l

i A.5/4.5 CORE AND CONTAINMENT COOLI_NG SXS7 EMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.C RHR Service Water and'Emernency 4.5.C RHR Service Water and EmeragILgy L

Eouioment Coolina Water Systems Eautoment Coolina Water Systema (EECWS) (Continued)

(EECWS) (Continued) 2.

During REACTOR POWER

2. No additional surveillance OPERATION, RHRSW pumps is required.

must be OPERABLE and assigned to service as indicated in Table 3.5-1 for the specified time limits.

3.

During Unit 2 REACTOR

3. Routine surveillance for POWER OPERATION, any two these pumps is specified RHRSW pumps (DI, D2, B1, in 4.5.C.1.

and B2) normally cr alternately assigned to the RER heat exchanger header supplying the standby coolant supply connection must be OPERABLE except'as specified in 3.5.C.4 and 3.5.C.5 below.

BFN 3.5/4.5-10 Unit 2 Amendment No. 149,169

..o TABLE 3.5-1 Time Minimum Limit Service Assigncent (Days)

RHRSW EECW(2)

(4)

(1)

Indefinite 7

3 (3)(4)

(1)

(3) 30 7

cr 6 2

or 3 (4)

(1) 7 6

2 (1)

At least one OPERABLE pump must be assigned to each header.

(2)

Only automatically starting pumps may be assigned to EECW header service.

(3)

Nine pumps must be OPERABLE. Either configuration is acceptable: 7 and 2 or 6 ind 3.

(4)

Requirements may be reduced by two for each unit with fuel unloaded.

BFN 3.5/4.5-11 Unit 2

.i THIS PAGE INTENTIONALLY LEFT BLANK l

l l

BFN 3.5/4.5-11a Amendment No. 169 I

Unit 2

,l l

L____ --_ _

}.i

e 1

.l L

.h5/4.5 'CORZ AND C07frA2PMNT COOLING SYSM

'}iIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS-3.5.C

,RHR Service Water and Emergency 4.5.C EHE_Sr.ryice Water and Emergency p

Eauipment Coolina Water Systema Eauioment Coolina Water Systems.

(EECWS)-(Continued)

4. -Three of the D1, D2, B1, B2

4. No additional surveillance RHRSW pumps assigned to the is required.

l L

RHR heat exchanger supplying the standby coolant supply connection may be inoperable for a period not to exceed 30 days provided the OPERABLE pump is aligned to supply the RHR heat exchanger header and'the associated diesel generator and essential control valves are OPERABLE.

5.

The standby. coolant supply

. capability may be inoperable for a period not to exceed 10 days.

6.

If Specifications 3.5.C.2 through 3.5.C.5 are not met, an orderly shutdown shall be initiated and the unit placed in the COLD SHUTDOWN CONDITION within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

7.

There shall be at least 2 RHRSW pumps, associated with the selected RHR pumps, aligned for RHR heat exchanger service for each reactor vessel containing irradiated fuel.

Amendment No. 149, 169 BTN 3.5/4.5-12 Unit 2

yo.-

'e i

'3;5/4.5 ^ CORE AND CONTAINMENT COOLING SYSTEMS LIMITING CONDITIONS FOR.0PERATION SURVEILLANCE REQUIREMENTS i

3.5.D Eaulement Area Coolers 4.5.D Eaulement Area Coolers 1.

The equipment area. cooler

1. Each equipment area cooler

' associated with each RHR is operated in conjunction pump and the equipment with the equipment served area cooler associated by that paiticular cooler; with each set of core therefore, the equipment spray pumps (A and C area coolers are tested at or B and D) must be the same frequency as the OPERABLE at all times pumps which they serve.

when the pump or pumps served by that specific cooler is considered to be OPERABLE.

2.

When an equipment area cooler is not OPERABLE, the pump (s) cerved by that cooler must be considered inoperable for Technical Specification purposes.

E.

High Pressure Coolant Iniection E. High Pressure Coolant System (HPCIS)

Iniection System (HPCIS) 1.

The HPCI system shall be 1.

HPCI Subsystem testing OPERABLE:

shall be performed as follows:

(1) PRIOR TO STARTUP from a a.

Simulated once/

COLD CONDITION; or Automatic operating Actuation cycle Test (2) whenever there is b.

Pump Per irradiated fuel in the OPERA-Specification reactor vessel and the BILITY 1.0.MM reactor vessel pressure is grcater than 122 psig, c.

Motor Oper-Per except as specified in ated Valve Specification Specification 3.5.E.2.

OPERABILITY 1.0.MK d.

Flow Rate at Once/3 normal months reactor i

vessel operating pressure BFN 3.5/4.5-13 Amendment No. 155 Unit 2

,{

~

j.5/4.5 CORE AND CONTAINMENT COOLING SYSTEMS-l:

)

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5!.E High Pressure Coolant Iniection 4.5.E High Pressure Coolant Iniection System (HPCIS)

System (HPCIS) 4.5.E.1 (Cont'd) e.

Flow Rate at Once/

150 psig operating.

cycle The HPCI pump shall deliver ~at least 5000 gpm during each flow rate test.

f.

Verify that once/ Month each valve (manual, power-operated, or automatic) in the injection flow-path that is not locked, sealed, or otherwise secured in position, is in its correct

position.

2.

If the HPCI system is

2. No additional surveillance inoperable, the reactor may are required.

remain in operation for a period not to exceed 7 days, provided the ADS, CSS, RHRS (LPCI), and RCICS are OPERABLE.

3.

If Specifications 3.5.E.1 Except that an automatic or 3.5.E.2 are not met, valve capable of automatic an orderly shutdown shall return to its ECCS position be initiated and the when an ECCS signal is reactor vessel pressure present may be in a shall be reduced to 122 position for another mode of psig or less within 24 operation.

hours.

F.

Reactor Core Isolation Coolina F.

Reactor Core Isolation Coolina System (RCICS)

System (RCICS) 1.

The RCICS shall be OPERABLE:

1. RCIC Subsystem testing shall be performed as follows:

,(1) PRIOR TO STARTUP from a COLD CONDITION; or

a. Simulated Auto-Once/

)

matic Actuation operating Test cycle BFN 3.5/4.5-14 Amendment No. 155, 169 Unit 2

c

.s

-J.5/4.5-QQ E AND CONTAINMENT COOLING SYSUliS.

' LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5 F.

Reactor Core Isolation Coolina 4.5.F Reactor Core Isolation Cooling System (RCICS)

System (RCICS) 3.5.F.1 (Cont'd)

-4.5.F.1 (Cont'd) 1 C.) whenever there is

b. Pump Per irradiated fuel in the OPERABILITY Specifi-reactor vessel and the cation.

reactor vessel pressure 1.0.MM is above 122 psig, except as specified in

c. Motor-Operated Per 3.5.F.2.

Talve Specifi-OPERABILITY cation 1.0.MM d.

Flow Rate at Once/3 normal reactor months vessel-operating pressure e.

Flow Rate at Once/

150 psig operating cycle The RCIC pump shall deliver at least 600 gpm during each flow test, 2.

If the RCICS is inoperable, f.

Verify that Once/ Month the reactor may remain in each valve operation for a period not (manual, power-to exceed 7 days if the.

operated, or HPCIS is OPERABLE during automatic) in the suen time.

injection flowpath that is nee locked, 3.

If Specifications 3.5.F.1 sealed, or other-or 3.5.F.2 are not met, an wise secured in orderly shutdown shall be position, la in its initiated and the reactor correct

position.

shall be depressurized to less than 122 psig within

2. No additional surveillance 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , are required.

Except that an automatic valve capable.of automatic return to its normal position when a signal is present may be in a position for another mode of cperation.

BFN 3.5/4.5-15 Amendment NL. 169

-Unit 2

3.5/4.5 CORE _AND CONTATHMENT Q_00' LING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.G Automatic Depressurization 4.5.G Automatic Depressurization System (ADS)

System LADS) i 1

Four of the six valves of 1.

During each operating the Automatic cycle the following Depressurization System tests shall be performed shall be OPERABLE:

on the ADS:

j l

(1) PRIOR TO STARTUP a.

A simulated automatic from a COLD CONDITION, actuation test shall or, be performed PRIOR TO l

STARTUP after each

)

(2) whenever there is refueling outage.

irradiated fuel in the Manual surveillance reactor vessel and the of the relief valves reactor vessel pressure is covered in is greater than 105 psig, 4.6.D.2.

except as specified in 3.5.G.2 and 3.5.G.3 below.

2.

If three of the six ADS 2.

No additional surveillance valves are known to be are required, incapable of automatic operation, the reactor may remain in operation for a period not to exceed 7 days, provided the HPCI system is OPERABLE. (Note that the pressure relief function of these valves is assured by Section 3.6.D of these specifications and that this specification only applies to the ADS function.) If more than three of the six

]

ADS valves are known to be I

incapable of automatic operation, an immediate orderly shutdown shall b1 initiated, with the reactor in a HOT SHUTOOWN CONDITION in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in a COLD SHUTDOWN CONDITION in the following 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months .

3.

If Specifications 3.5.G.1 and 3.5.G.2 cannot be met, an orderly shutdown will be initiated and the reactor vessel pressure shall be reduced to 105 psig or less withir, sa hours.

BFN 3.5/4.5-16 Unit 2 Amendment No. 169

e.

3.5/4.5 CORE AND_@lGUNMG7 COOLING SKSZQiS

_,_ LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.H.

Maintenance of Filled Discharge 4.5.H. Maintenance of Filled Discharge f.ipl P.1pJl Whenever the core' spray systems, The following surveillance LPCI, HPCI, or RCIC are required requirements shall be adhered to be OPERABLE, the discharge to assure that the discharge piping from the pump discharge piping of the core spray of these systems to the last systems, LPC1, HPCI, and RCIC block valve shall be filled, are filled:

The suction of the RCIC and HPCI

1. Every month and prior to the pumps shall be aligned to the testing of the RHRS (LPCI and condensate storage tank, and Containment Spray) and core the pressure suppression chamber spray system, the discharge head tank shall normally be aligned piping of these systems shall to serve the discharge piping of be vented from the high point the RHR and CS pumps. The and water flow determined.

condensate head tank may be used to serve the RHR and CS discharge

2. Following any period where the piping if the PSC head tank LPCI or core spray systems is unavailable. The pressure have not been required to be indicators on the discharge of the OPERABLE, the discharge piping RHR and CS pumps shall indicate of the inoperable system shall not less than listed below, be vented from the high point prior to the return of the F1-75-20 48 psig system to service.

P1-75-48 48 poig P1-74-51 48 psig

3. Whenever the HPCI or RCIC P1-74-65 48 psig system.is ?ined up to take auction from the condensate storage tank, the discharge piping of the HPCI and RCIC shall be vented from the high point of the system and water flow observed on a monthly basis.

4. When the RERS and the CSS are required to be OPERABLE, the pressure indicators which monitor the discharge lines shall be monitored daily tnd the pressure recorded.

Amendment No. 169 BFN 3.5/4.5-17 Unit 2 j

<e 3.5' BASES Removal System (RHRS)

Core Serav System (CSS) and 3.5.B Residual Heat 3.5.A.-

Analyses presented in the FSAR* and analyses presented in conformance i

with 10'CFR 50, Appendix K, demonstrated that the core spray system n conjunction with two LPCI pumps provides adequate cooling to the core to cident and to dissipate the energy associated with the loss-of-coolant ac limit fuel clad temperature to below 2,200'T which assures that core geometry remains intact and-to limit the core average clad metal-waterCo reaction to less than 1 percent.

in tests.of systems similar to design to BFNP to exceed the m h heater at less than half.the rated flow in simulated fuel assem requirements.

The RHRS (LPCI mede) is designed to provide emergency cooling to the core This system is by flooding in the event of a loss-of-coolant accident.

completely independent of the core spray system;.howev

,I.

The LPCI mode of the RHRS and the core spray system provide adequate cooling for break areas of fuel clad temperature.

without assistance from the high-pressure emergency core cooling subsystems.

The intent of the CSS and IWRS specifications is to not allow startup However, during operation, certain components may The allowable repair times have the specified allowable repair times.

been selected using engineering judgment based on experiences and supported by availability analysis.

l ii ore spray

.Should one core spray loop become inoperable, the rema n ng c loop, the RHR System, and the diesel generators are required to be-.These p OPERABLE should the need for core cooling arise.

margin over the OPERABLE equipment needed for adequata core cooling.

i. time of seven days With due regard for this margin, the allowable repc was chosen.

Should one RHR pump (LPCI mode) become inoperable, three RHR pumps Since adequate core (LPCI mode) and the core spray system are available. coo i

period is justified.

Should two RHR pumps (LPCI mode) become inoperable, there remains no Therefore, the reserve (redundant) capacity within the RHRS (LPCI m FSAR.

A detailed funct2onal analysis is given in Section 6 of the BFNP l

l l

l' l-3.5/4.5-24 Amendment No. 169 BFN Unit 2 l

L

t c,

3.5' Eldfd (Cont

d)

Should.one RHR pump.(containment cooling mode) become inoperable, a complementlof three full capacity containment heat removal systems is still available. Any two of the remaining pumps / heat exchanger combinations would provide more than adequate containment cooling for any abnormal.or postaccident situation. Because of the availability of equipment in excess of normal redundancy requirements, a 30-day repair period is justified.

Should two RHR pumps (containment cooling mode) become inoperable, a full heat removal' system is still available. The remaining pump / heat exchanger combinations would provide cdequate containment cooling for any abnormal postaccident situation. Because of the availability of a full complement of heat removal equipment, a 7-day repair period is justified.

l Observation of the stated requirements for the containment cooling mode assures that the suppression pool and the drywell will be sufficiently cooled, following a loss-of-coolant accident, to prevent primary containment overpressurization. The containment cooling function of the RHRS is permitted only after the core has reflooded to the two-thirds core height level. This prevents inadvertently diverting water needed for core flooding to the less urgent task of containment cooling. The two-thirds core height level interlock may be manually bypassed by a keylock switch.

Since the RHRS is filled with low quality water during power operation, it is planned that the system be filled with demineralized (condensate) water before using the shutdown cooling function of the RHR System.

Since it is desirable to have the RHRS in service if a " pipe-break" type of accident should occur, it is permitted to be out of operation for only a restricted amount of time and when the system pressure is low. At least one-half of the containment cooling function must remain OPERABLE during this time period. Requiring two OPERABLE CSS pumps during cooldown allows for flushing the RHRS even if the shutdown were caused by inability to meet the CSS specifications (3.5.A) on a number of OPERABLE pumps.

When the reactor vessel pressure is atmospheric, the limiting conditions for operation are less restrictive. At atmospheric pressure, the minimum requirement is for one supply of makeup water to the core. Requiring two OPERABLE RHR pumps and one CSS pump provides redundancy to ensure maV.sup water availability.

Should one RHR pump or associated heat <xchanger located on the unit cross-connection in the adjacent unit become inoperable, an equal capabilit'/ for long-term fluid makeup to the reactor and for cooling of the containment remains OPERABLE. Because of the availability of an equal makeup and cooling capability, a 30-day repair period is justified.

BFN 3.5/4.5-25 Amendment No. 169 Unit 2

3.5 Bases (Cont'd)

The suppression chamber can be drained when the reactor vessel pressure is atmospheric, irradiated fuel is in the reactor vessel, and work is By in progress which has the potential to drain the vessel.

requiring the fuel pool gate to be open with the vessel head removed, not d

the combined water inventory in the fuel pool, the reactor cavity, an d the the separator / dryer pool, between the fuel pool low leve This will provide adequate low-pressure cooling in lieu of CSS and RER (LPCI and containment cooling mode) as currently required inThe additional Specifications 3.5.A.4 and 3.5.B.9.

providing standby coolant supply available will ensure a redundantCo supply nf coolant supply.'

s period provided no more than one drive is removed 1 ad flanges are installed during the period of time CRDs are during unless not in place.

Should the capability for providing flow through the cross-connect a 10-day repair time is allowed before shutdown is This repair time is justified based on the very small lines be lost, probability for ever needing RHR pumps and heat exchangers to supply an required.

ad,iacent unit.

REFERENCES Residual Heat Removal System (BFNP FSAR subsection 4.8) 1.

Core Standby Cooling Systems (BFNP FSAk Section 6) 2.

Cooling Wate_r System 3.5.C. RHR Service Water System and Emergency Eoulument (EECWS)

There are two EECW headers (north and south) with four automatic All components requiring starting RERSW pumps on each header.

emergency cooling water are fed from both headers thus assuring Each header continuity of operation if either header is OPERABLE.

Two RHRSW pumps can alone can handle the flows to all components.

supply the full flow requirements of all essential EECW loads for any abnormal or postaccident situation.

There are four RHR heat exchanger headers (A, B, C, & D) with one RHR There are two RHRSW heat exchanger from each unit on each header.

pumps on each header; one normally assigned to each header (A2, B2, C2, (A1, B1, C1, or D1). One RHR or D2) and one on alternate assignmentcan adequately deliver the flow supplied by bo hect exchanger headet RHRSW pumps to any two of the three RHRSW heat excha header.

exchangers can more than adequately heat exchanger. Two RHR heat handle the cooling requirements of one unit in any abnormal or yestaccident situatien.

l-3.5/4.5-26 BFN Unit 2

=

L

e.

e U

3.5L BASES-(Cont'd)

The RHR Service Water System.was designed as.a shared system for_three units. 'The specification, as written, is conservative when consideration is given to particular pumps being out of service and to possible valving arrangements.

If unusual operating conditions arise such that more pumps are out of service than:allop:d by this

spec 4fication, a'special case request may be made to the NRC to allow continued operation if the actual system cooling requirements can be assured.

l Should one of the two RHRSW pumps normally or alternately assigned to the'RHR heat exchanger header supplying the standby coolant supply connection become inoperable, an equal capability for long-term fluid makeup.to the unit reactor and for cooling of the unit: containment remains OPERABLE. Because of the availability of an equal makeup and cooling capability, a 30-day repair period is justified. Should the i

capability to provide standby coolant supply be' lost, a 10-day repair time is justified based on the low probability for ever needing the standby coolant supply. Verification that the LPCI subsystem cross-tie valve is closed and power to its operator is disconnected ensures that each LPCI subsystem remains independent and a failure of the flow path I;

in one subsystem will not affect the flow path of the other LPCI.

subsystem.

-3.5.D EEauipment Area Coolers There is an equipment area cooler for each RHR pump and an equipment

~

area. cooler for each set (two pumps, either the A and C or B and D pumps) of core spray pumps..The-equipment area coolers take suction near the cooling air discharge of the motor of the pump (s) served and discharge air near theJcooling air suction of the motor.of the pump (s)'

served. This ensures that cool air is' supplied for cooling the pump motors.

The' equipment area coolers also remove the pump, and equipment waste heat from the basement rooms housing the engineered safeguard equipment. The various conditions under which the operation of the equipment air coolers is required have been identified hy evaluating-the normal and abnormal operating transients and accidents over the full range of planned operations. The surveillance and testing of the equipment area coolers in each of their various modes is accomplished during the testing of the equipment served by these coolers. This testing is adequate to assure the OPERABILITY of the equipment area

. coolers.

REFERENCES 1.

Residual Heat Removal System (BFN FSAR Section 4.8) 2.

Core Standby Cooling System (BFN FSAR subsection 6.7)

BFN 3.5/4.5-27 Amendment No. 169 Unit 2

l l

3.5 ESIS (Cont'd) 3.5.E. Hirh Pressure Coolant Iniection System (HPCIS) l The EPCIS is provided to assure that the reactor core is adequately j

cooled to limit fuel clad temperature in the event of a small break in tbc nuclear system and loss of coolant which does not result in rapid depressurization of the reactor vessel. The HPCIS permits the reactor to be shut down while maintaining sufficient reactor vessel water level inventory until the vessel is depressurized. The HPCIS continues to l

operate until reactor vessel pressure is below the pressure at which I

LPCI operation or core spray system operation maintains core cooling.

)

The capacity of the system is selected to provide this requirei core cooling. The HPCI pump is designed to pump 5,000 gpm at reactor pressures between 1,120 and 150 psig. Two sources of water are i

available.

Initially, water from the condensate storage tank is used

{

instead of injecting water from the suppression pool into the reactor, j

l When the HPCI System begins operation, the reactor depressurizes more rapidly than would occur if HPCI was not initiated due to the i

condensation of steam by the cold fluid pumped into the reactor vessel by the HPCI system. As the reactor vessel pressure continues to decrease, the HPCI flow momentarily reaches equilibrium with the flow through the break. Continued depressurization caused the break flow to l

decrease below the HPCI flow and the liquid inventory begins to rise.

)

This type of response is typical of the small breaks. The core never I

uncovers and is continuously cooled throughout the transient so that no core damage of any kind occurs for breaks that lie within the capacity range of the HPCI.

The minimum required NPSH for HPCI is 21 feet. There is adequate elevation head between the suppression pool and the HPCI pump, such that the required NPSH is available with a suppression pool temperature up to 140*F with no containment back pressure.

The HPCIS serves as a backup to the RCICS as a source of feedwater makeup durinE primary system isolation conditions. The ADS serves as a backup to the HPCIS for reactor depressurization for postulated transients and accidt The CSS and RHRS (LPCI) provide adequate core cooling at low reactor pressure when RCICS and ADS are no longer necessary.

Considering the redundant systems, an al.twable repair time of seven days was selected.

The HPCI and RCIC as well as all other Core Standby Cooling Systems must be OPERABLE when starting up from a Cold Condition.

It is realized that the HPCI is not designed to operate at full capacity until reactor pressure exceeds 150 psig and the steam supply to the HPCI turbine is automatically isolated before the reactor pressure decreases below 100 psig. It is the intent of this specification to assure that when the reactor is being started up from a Cold Condition, the HPCI is not known to be inoper6ble.

I BFN 3.5/4.5-28 Amendment No. 169 Unit 2 l

l

3.5 BASES (Cont'd) 3.5.T Leactor_Sgre Isolation Coolina System (RCICS)

The various conditions under which the RCICS plays an essential role in providing makeup water to the reactor vessel have been identified by evaluating the various plant events over the full range of planned operations.

The specifications ensure that the function for which the ECICS was designed will be available when needed. The minimum required NPSH for RCIC is 20 feet. There is adequate elevation head between the suppression pool and the RCIC pump, such that the required NPSH is available with a suppression pool temperature up to 140'T with no containment back pressure.

Because the low-pressure cooling systems (LPCI and core spray) are capable of providing all the cooling required for any plant event when nuclear system pressure is below 122 psig, the RCICS is not required Between 122 psig and 150 psig the RCICS need not below this pressure.

provide its design flow, but reduced flow is required for certain RCICS design flow (600 gpm) is sufficient to maintain water events.

level above the top of the active fuel for a complete icss of feedwater flow at design power (105 percent of rated).

Consideration of the availability of the RCICS reveals that the average risk associated with failure of the RCICS to cool the core when i

increased if the RCICS is inoperable for no longer than required is not seven days, provided that the HPCIS is OPERABLE during this period.

REFERENCE Reactor Core Isolation Cooling System (BFNP FSAR Subsection 4.7) 1.

3.5.G Automatic Deoressurization System (ADS)

This specification ensures the OPERABILITY of the ADS under all conditions for which the depressurization of the nucicar system is an essential response to station abnormalities.

The nuclear system pressure relief system provides automatic nuclear system depressurization for small breaks in the nuclear system so that the low-pressure coolant injection (LPCI) and the cote spray subsystems can operate to protect the fuel barrier. Note that t his specification applies only to the automatic feature of the pressure relief system.

Specification 3.6.D specifies the requirements for the pressure relief function of the valves. It is possible for any number of the valves assigned to the ADS to be incapable of performing their ADS functions because of instrumentation failures yet be fully capable of performing their pressure relief function.

3.5/4.5-29 BEN

' Unit 2

3.5 BASES (Cont'd)

Because the automatic depressurization system does not provide makeup to the reactor primary vessel, no credit is taken for the steam cooling of the core caused by the system actuation to provide further conservatism to the CSCS.

With two ADS valves known to be incapable of automatic operation, four valves remain OPERABLE to perform their ADS function. The ECCS loss-of-coolant accident analyses for small line breaks assumed that four of the six ADS valves were OPERABLE. Reactor operation with three ADS valvee inoperable is allowed to continue for seven days provided that the HPCI system is OPERABLE. Operaticn with more than three of the six ADS valves inoperable is not acceptable.

H.

Maintenance of Filled Discharge Pipe If the discharge piping of the core spray, LPCI, HPCIS, and RCICS are not filled, a water hammer can develop in this piping when the pump and/or pumps are started. To minimize damage to the discharge piping and to ensure added margin in the operation of these systems, this Technical Specification requires the discharge lines to be filled whenever the system is in an OPERABLE condition.

If a discharge pipe is not filled, l

the pumps that supply that line must be assumed to be inoperable for Technical Specification purposes.

The core spray and RHR system discharge piping high point vent is visually checked for water flow once a month and prior to testing to ensure that the lines are filled. The visual checking will avoid starting the core spray or RHR system with a discharge line not filled.

In addition to the visual observation and to ensure a filled discharge line other than prior to testing, a pressure suppression chamber head tank is located approximately 20 feet above the discharge line high point to supply makeup water for these systems. The condensate head tank located approximately 100 feet above the discharge high point serves as a backup charging system when the pressure suppression chamber head tank is not in service. System discharge pressure indicators are used to determine the water level above the discharge line high point. The indicators will reflect approximately 30 psig for a water level at the high point and 45 psig for a water level in the pressure suppression chamber head tank and are monitored daily to ensure that the discharge lines are filled.

When in their normal standby condition, the suction for the HPCI and RCIC pumps are aligned to the condensate storage tank, which is physically at a higher elevation than the HPCIS and RCICS piping. This assures that the HPCI and RCIC discharge piping remains filled. Further assurance is provided by observing water flow from these systems' high points monthly.

I.

Maximum Averare Planar Linear Heat Generation Rate (MAPLHGR)

This specification assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the limit'specified in the 10 CFR 50, Appendix K.

BFN 3.5/4.5-30 Unit 2 Amendment No. 169

e

.o 3.5 EAEEE (Cont'd)

The peak cladding temperature following a postulated loss-of-coolant accident is primarily a function of the average heat generation rate of all the rods of a fuel a --mbly et er.y axial location and is only dependent secondarily on e rod-to-rod power distribution within an assembly.

Since expectes seal variations in power distribution within a fuel assembly affect the calculated peak clad temperature by less than 20*F relative to the peak temperature for a typical fuel design, the limit on the average linear heat generation rate is sufficient to assure that calculated temperatures are within the 10 CFR 50 Appendix K limit. The limiting value for MAPLHGR is shown ir, Tables 3.5.I-1 and

-2.

The analyses supporting these limiting values are presented in Reference 1.

3.5.J. Linear Heat Generation Rate (LEGR)

This specification assures that the linear heat generation rate in any rod is less than the design linear heat generation if fuel pellet l

densification is postulated.

I The LHGR shall be checked daily during reactor operation at i 25 percent power to determine if fuel burnup, or control rod movement has caused changes in power distribution. For LHGR to be a limiting value below 25 percent rated thermal power, the R factor would have to be less than 0.241 which is precluded by a considerable margin when employing any permissible control rod pattern.

3.5.K. Minimum Critical Power Ratio (MCPR)

At core thermal pow /er levels less than or equal to 25 percent.

reactor will be operating at minimum recirculation pump speed ca6 ;he moderator void content will be very small. For all designated control rod patterns which may be employed at this point, operating plant experience and thermal hydraulic analysis indicated tbat the resulting MCPR value is in excess of requirements by a considerable margin. With this low void content, any in.?dvertent core ficw increase would only place operation in a more conservative mode relative to MCPR. The daily requirement for calculating MCPR above 25 percent rated thermal power is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes. The requirement for calculating MCPR when a limiting control rod pattern is approached ensures that MCPR will be known following a change in power or power shape (regardless of magnitude) that could place operation at a thermal limit.

3.5.L. APRM Setooints Operation is constrained to a maximum LFGR of 18.5 kW/ft for 7x7 fuel and 13.4 kW/ft. This limit is reached J.en core maximum fraction of limiting power density (CMFLPD) equals 1.0.

For the case where CMFLPD exceeds the fraction of rated thermal power, operation is permitted only at less than 100-percent rated power and only with APRM scram settings as required by Specification 2.5.L.1.

The scram trip setting and rod block trip setting are adjusted to ensure that no combination BFN 3.5/4.5-31 Unit 2

e.-

m 4

3.5 BASES (Cont'd) of CMFLPD and FRP will increase the LHGR transient peak beyond that allowed by'the 1-percent plastic strain limit. A 6-hour time period to achieve this condition is justified since the additional margin gained by the setdown adjustment is above and beyond that ensored by.the safety analysis.

3.5.M. References i

1.

Loss-of-Coolant Accident Analysis for Browns Ferry Nuclear Plant Unit 2, NEDO - 24088-1 and Addenda.

2.

"BWR Transient Analysis Model Utilizing the RETRAN Program,"

TVA-TRB1-Ol-A.

3.

Generic Reload Fuel Application, Licensing Topical Report, NEDE - 24011-P-A and Addenda.

BFN 3.5/4.5-32 Unit 2 Amendment No. 143 m

M__

l

.g 4.5 Goge and Containment Cooling Systems Surveillance Frequencies The testing interval for the core and containment cooling systems is based on industry practice, quantitative reliability analysis, judgment and practicality. The core cooling systems have not been designed to be fully testable during operation. For example, in the case of the HPCI, automatic initiation during power operation would result in pumping cold water into the reactor vessel which is not desirable.

Complete ADS testing during power operation causes an undesirable loss-of-coolant inventory. To increase the availability of the core and containment cooling system, the components which make up the

system, i.e.,

instrumentation, pumps, valves, etc., are tested frequently. The pumps and motor operated injection valves arm also tested in accordance with Specification 1.0.MM to assure their OPERABILITY. A simulated automatic actuation test once each cycle combined with testing of the pumps and injection valves in accor'snce with Specification 1.0.MM is deemed to be adequate testing of these sys t enas. Monthly alignment checks of valves that are not locked or sealed in position which affect the ability 3f the systemr to perform their intended safety function are also verified to be in the proper position. Valves which automatically reposition themselves on an initiation signal are permitted to be in a position other than normal to facilitate other operational modes of the system.

When components and subsystems are cut-of-service, overall core and containment cooling reliability la maintained by OPERABILITY of the remaining redundant equipment.

Whenever a CSCS system or loop is made inoperable, the other CSCS systems or loops that are required to be OPERABLE shall be considered OPERABLE if they are within the required surveillance testing frequency and there is no reason to suspect they are inoperaole.

If the function, system, or loop under test or calibration is found inoperable or exceeds the trip level setting, the LCO and the required surveillance testing for the system or loop shell apply.

Maximum Aver 8Re Planar LHCR. LHGR. and MCPR The MAPLHGR, LHGR, and MCPR shall be checked daily to determine if fuel burnup, or control rod movement has caused changes in power distribution.

Since changes due to burnup are slow, and only a few control rods are moved daily, a daily check of power distribution is adequate.

BFN 3.5/4.5-33 Unit 2 Amendment No. 155,169

ff

$ REQq%

UNITED STATES y=

,w NUCLEAR REGULATORY COMMISSION

'j.

WASHING TON, D. C. 20555

\\...../

TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-260 BROWNS FERRY NUCLEAR PLANT, UNIT 2 AMENDMENT TO FACILITY 0PERATING LICENSE Amendment No.169 License No. DPR-52 1.

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

A..

The application for amendment by Tennessee Valley Authority (the licer.see) dated January 13, 1989, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),

and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, 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 Commission's irgulations; 1

D.

The issuance of this amendment will not be inimical to the common rafense 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 i

of the Commission's regulations and all applicable requirements have l

been satistied.

~

l l

..- 2.-

Accordingly, the license is amended by changes to u a Technical Specifications as indicated in the attachment to tlas license amendment and paragraph 2.C.(2) of Facility Operating License No. DPR-52 is hereby amended to read as follows:

-(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 169, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 60 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION h

L Y

Suzanne Black, Assistant Director for Projects TVA Projects Division Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

August 2, 1989 l

l I

6 w

ATTACHMENT TO LICENSE AMENDMENT NO.169 FACILITY OPERATING LICENSE NO. DPR-52 DOCKET N0. 50-260 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.

Table of Contents and overleaf pages*

are provided to maintain document completeness.

REMOVE INSER i

1*

11 11 iii iii iv iv v

v vi vi vii vii viii viii 3.5/4.5-1 3.5/4.5-1*

3.5/4.5-2 3.5/4.5-2 3.5/4.5-3

3. 5/4. 5-3

3.5/4.5-4 3.5/4.5-4 3.5/4.5-5 3.5/4.5-5 3.5/4.5-6 3.5/4.5-6 3.5/4.5-7 3.5/4.5-7 3.5/4.5-8 3.5/4.5-8 3.5/4.5-9 3.5/4.5-9 3.5/4.5-10 3.5/4.5-10 3.5/4.5-11 3.5/4.5-11*

3.5/4.5-11a 3.5/4.5-12 3.5/4.5-12 3.5/4.5-13 3.5/4.5-13*

3.5/4.5-14 3.5/4.5-14 3.5/4.5-15 3.5/4.5-15 3.5/4.5-16 3.5/4.5-16 3.5/4.5-17 3.5/4.5-17 3.5/4.5-24 3.5/4.5-24 3.5/4.5-25 3.5/4.5-25 3.5/4.5-26 3.5/4.5-26*

3.5/4.5-27 3.5/4.5-27 3.5/4.5-28 3.5/4.5-28 3.5/4.5-29 3.5/4.5-29*

3.5/4.5-30 3.5/4.5-30 3.5/4.5-31 3.5/4.5-31*

i 3.5/4.5-32*

3.5/4.5-32*

3.5/4.5-33 3.5/4.5-33 1

~

1 i

TABLE OF CONTENTS Section Pare No.

1.0 Definitions 1.0-1 SAFETY LIMITS AND LIMITING SAFETY SYSTEM EXTTINGS 1.1/2.1 Fuel Cladding Integrity.

1.1/2.1-1 1.2/2.2 Reactor Coolant System Integrity.

1.2/2.2-1 l

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3.1/4.3 Reactor Protection System..

3.1/4.1-3 l

3.2/4.2 Protective Instrumentation..

3.2/4.2-1 A.

Primary Containment and Reactor Building Isolation Functions..

3.2/4.2-1 B.

Core and Containment Cooling Systems -

Initiation and Control.

3.2/4.2-1 C.

Control Rod Block Actuation...

3.2/4.2-2 D.

Radioactive Liquid Effluent Monitoring Instrumentation..

3.2/4.2-3 E.

Drywell Leak Detection..

3.2/4.2-4 F.

Surveillance Instrumentation.....

3.2/4.2-4 G.

Control Room Isolation.

3.2/4.2-4 H.

Flood Protection.

3.2/4.2-4 I.

Meteorological Monitoring Instrumentation.

3.2/4.2-4 J.

Seismic Monitoring Instrumentation..

3.2/4.2-5 K.

Radioactive Gaseous Effluent Monitoring Instrumentation 3.2/4.2-6 L.

ATWS Recirculation Pump Trip.

3.2/4.2-6a 3.3/4.3 Reactivity Control.

3.3/4.3-1 A.

Reactivity Limitations.

3.3/4.3-1 B.

Control Rods.

3.3/4.3-5 C.

Scram Insertion Times...

3.3/4.3-10 i

BFN UNIT 1 Amendment Nos. 133, 164, 169 l

Section Pane No.

D.

Reactivity Anomalies.

3.3/4.3..-..........

E.

Reactivity Control'.......

3.3/4.3-12 F.

_ Scram Discharge Volume....

.3.3/4.3-12 3.4/4.4 Standby Liquid Control System.......

3.4/4.4-1 A.

Normal System Availability..

3.4/4.4-1 B.'

Operation with Inoperable Components.....

3.4/4.4-3 C.

Sodium Pentaborate Solution.....

3.4/4.4-3

'3.5/4.5 Core and Containment Cooling Systems.

3.5/4.5-1 A.

. Core Spray System (CSS).

3.5/4.5-1 B.

Residual Heat Removal System (RHRS)

(LPCI and' Containment Cooling).

-3.5/4.5-4 C.

RHR Service Water and Emergency Equipment Cooling Water Syste:as (EECWS).

3.5/4.5-9 D.

Equipment Area Coolers.

3.5/4.5-13 E.

High Pressure Coolant Injection System (HPCIS).

~3.5/4.5-13

- T.

Reactor Core Isolation Cooling System

.(RCICS).

3.5/4.5-14 G.

Automatic Depressurization System (ADS).

3.5/4.5-16 H..

Maintenance of Filled Discharge Pipe.

3.5/4.5-17 I.

Average Planar Linear Heat Generation Rate.

3.5/4.5-18 J.

Linear Heat Generation Rate (LHGR).

3.5/4.5-18 K.

Minimum Critical Power Ratio (MCPR)...

3.5/4.5-19 L.

APRM Setpoints.

3.5/4.5-20 3.6/4.6 Primary System Boundary.

3.6/4.6-1 J

A.

Thermal and Pressurization Limitations.

3.6/4.6-1 l

B.

Coolant Chemistry.

3.6/4.6-5 C.

Coolant Leakage.

3.6/4.6-9 D.

Relief Valves.

3.6/4.6-10 ii Amendment No. 133, 147, 164, 169

-BFN

- -UNIT 1-

Sectica Pane No.

3.6/4.6-11 E.

Jet Pumps.

F.

Recirculation Pump Operation.....

3.6/4.6-12 G.

Structural Integrity.

3.6/4.6-13 H.

Snubbers 3.6/4.6-15 3.7/4.7 Containment Systems....

3.7/4.7-1 A.

Primary Containment.

3.7/4.7-1 3.

Standby Gas Treatment System.

3.7/4.7-13 C.

Secondary Containment.

3.7/4.7-16 D.

Primary Containment Isolation Valves.

3.7/4.7-17 E.

Control Room Emergency Ventilation.

3.7/4.7-19 F.

Primary Containment Purge System.

3.7/4.7-21 G.

Containment Atmosphere Dilution System (CAD) 3.7/4.7-22 H.

Containment Atmosphere Monitoring (CAM)

System H2 Analyzer..

3.7/4.7-24 3.8/4.8 Radioactive Materials.

3.8/4.8-1 A.

Liquid Effluents.

3.8/4.6-1 B.

Airbo:ne Effluents.

3.8/4.8-3 C.

Radioact ve Effluents - Dose.

3.8/4.8-6 D.

Mechanical Vacuum Pump.

3.8/4.8-6 E.

Miscellaneous Radioactive Materials Sources.

3.8/4.8-7 F.

Solid Radwaste.

3.8/4.8-9 3.9/4.9 Auxiliary Electrical System.

3.9/4.9-1 A.

Auxiliary Electrical Equipment 3.9/4.9-1 B.

Operation with Inoperable Equipment.

3.9/4.9-8 C.

Operation in Cold Shutdown.

3.9/4.9-15 3.10/4.10 Core Alterations 3.10/4.10-1 A.

Refueling Interlocks.

3.10/4.10-1 B.

Core Monitoring.

3.10/4.10-4 111 Amendment Nos. 132, 163, 169 BFN UNIT I l

Sectio.D Par.e No.

C.

Spent Fuel Pool Water.

3.10/4.10-7 D.

Reactor Building Crane.

3.10/4.10-8 E.

Spent Fuel Cask.

3.10/4.10-9 j

F.

Spent Fuel Cask Handling-Refueling Floor......

3.10/4.10-10 3.11/4.11 Fire Protectior. Systems 3.11/4.11-1 A.

Fire Detection Instrumentation.

3.11/4.11-1 B.

Fire Pumps and Water Distribution Mains 3.11/4.11-2 C.

Spray and/or Sprinkler Systems.

3 11/4.11-7 D.

CO2 Systems.

3.11/4.11-8 E.

Fire Hose Stations.

3.11/4.11-9 F.

Yard Fire Hydrants and Hose Houses.

3.11/4.11-11 G.

Fire-Rated Assemblies 3.11/4.11-12 H.

Open Flames, Welding and Burning in the Cable Spreading Room.

3.11/4.11-13 5.0 Major Design Features.

5.0-1 5.1 Site Features.

5.0-1 5., 2 Reactor.

5.0-1 5.3 Reactor Vessel....

5.0-1 5.4 Containment.

5.0-1 5.5 Fuel Storage.

5.0-1 5.6 Seismic Design.

5.0-2 iv BFN UNIT 1 Amendment Nos. 138, 162, 169

.pn re-ADMINISTRATIVE CONTROLS T

SECTION PAGE l

- hil RESPONSIBILITY..........................................

6.0-1 hil ORG4NIZATIOH............................................

6.0-1 6'.2.1 offsite and Onsite Organizations........................

6.0-1 h

6.2.2 Plant Staff.............................................

6.0-2

' 623 PLANT STAFF 0 QUALIFICATIONS..............................

6.0-5 his T RA I N I NG................................................

6. 0- 5 6.0-5 121 PLANT REVIEW AND AUDIT........

............s..............

. 6.5.1 Plant Operations Review Committee (P0RC).................

6.0-5 6.5.2 Nuclear Safety Revit* Board (NSRB)......................

6.0-11

- 6.5.3 Technical Review and Approval of Procedures.............

6.0-17 his REPORTABLE EVENT ACTI0ES................................

6.0-18 hil SAFETY LIMIT VIOLATION...........................,.......

6.0-19 421

.EROCEDURES/ INSTRUCTIONS AND PROGRAMS....................

6.0-20 6.8.1 Procedures..............................................

6.0-20 6.8.2 Dri11s..................................................

6.0-21 6.8.3 Radiation Control Procedures............................

6.0-22 6.8.4 Quality-Assurance Procedures - Efflu nt and Environmental Monitoring.............................

6.0-23 422 REPORTI NG RE0UIREME NTS..................................

6. 0-24 6.9.1 Routine Reports.........................................

6.0-24 Startup Reports.......................................... 6.0-24 Annual Operating Report.................................. 6.C-25 Monthly Operating Report................................. 6.0-26

. Reportable Events.......................................

6.0-26 Radioactive Effluent Release Report...................... 6.0-26 Source Tests............................................

6.0-26 6.9.2 Special Reports.........................................

6.0-27 kilD STATION OPERATING RECORDS AND RETENTION.................

6.0-29 6.11 EE0CESSz[0NTROL PR0 GRAM......................<,..........

6.0-32 6.12 0FFSITE DOSE _ CALCULATION MANUAL.........................

6.0-32 6.13 RADIOLOGI CAL EFFLUENT MANUAL............................

6. 0-3 3 v

BFN UNIT 1 Amendment Nos. 138, 169

LIST OF TABLES

[

IAhlt Title Page No.

1.1 Surveillance Frequency Notation..

1.0-13 3.1.A Reactor Protection System (SCRAK)

Instrumentation Requirements.

3.1/4.1-3 4.1.A Reactor Protection System (SCRAM) Instrumentation Functional Tests Minimum Functional Test Frequencies for Safety Instr. and Cont.rol Circuits...

3.1/4.1-8 4.1.B Reactor Protection System (S". RAM) Instrumentation Calibration Minimum Calibration Frequencies for Reactor Protection Instrument Channels..

3.1/4,1-11 3.2.A Primary Containment and Reactor Building Isolation 3.2/4.2-7 Instrumentation.................

3.2.B Instrumentation that Initiates or Controls the Core and Containment Cooling Systems.

3.2/4.2-14 3.2.C Instrumentation; that Initiates Rod Blocks....

3.2/4.2-25 3.2.D Radioactive Liquid Effluent Monitoring 3.2/4.2-28 Instrumentation..................

3.2.E Instrumentation that Monitors Leakage Into Drywell.

3.2/4.2-30 3.2.F Surveillance Instrumentation.

3.2/4.2-31 3.2.G Control Room Isolation Instrumentation.

3.2/4.2-34 3.2.H Flood Protection Instrumentation.

3.2/4.2-35 3.2.I Meteorological Monitoring Instrumentation.

3.2/4.2-36 3.1.J Seismic Monitoring Instrumentation.....

3.2/4.2-37 3.2.K Radioactive Gaseous Effluent Monitoring Instrumentation.............

3.2/4.2-38 3.2.L ATWS - Recirculation Pump Trip (RPT)

Surveillance Instrumentation.

3.2/4.2-39b 4.2.A Surveillance Requirements for Primary Containment and Reactor Building Isolation Instrumentation.

3.2/4.2-40 4.2.B Surveillance Requirements for Instrumentation that Initiate er Control the CSCS.

3.2/4.2-44 4.2.C Surveillance Requirements for Instrumentation that Initiate Rod Blocks.

3.2/a 2-50 4.2.D radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements....

3.2/4.2-51 vi Amendment No. 132, 164, 169 BFN Unit 1

.a

.w.

LIST OLIA1}1,M (Cont'd)

Table Ilth Page No.

l 4.2.E Minimum Test and Calibration Frequency for Drywell l

Leak Detection Instrumentation.....

3.2/4.2-53 4.2.F Minimum Test and Calibration Frequency for Surveillance Instrumentation.

3.2/4.2-54 4.2.G Surveillance Requirements for Control Room 1sclation Instrumentation.

3.2/4.2-56 4.2 H Minimum Test and Calibration Frequency for Flood Protection Instrumentation.

3.2/4,2-57 4.2.J Seismic Monitoring Instrument Surveillance Requirements.

3.2/4.2-58 4. 2.'K Radioactive Gaseous Effluent Instrumentation Surveillance.

3.2/4.2-62 4.2.L ATWS-Recirculation Pump Trip (RPT)

Instrumentation Surveillance.

3.2/4.2-63a 3.5-1 Minimum RHRSW and EECW Pump Assignment 3.5/4.5-11 3.5.I MAPLHGR Versus Average Planar Exposure.

3.5/4.5-21 3.7.A Primary Containment Isolation Valves..

3.7/4.7-25 3.7.B Testable Pen 9trations with Double 0-Ring Seals.

3.7/4.7-32 3.7.C Testable Penetrations with Testable Bellows.

3.7/4.7-33 3.7.D Air Tested Isolation Valves.

3.7/4.7-34 3.7.E Primary Containment Isolation Valves which Terminate below the Suppression Pool Water Level.

3.7/4.7-37 3.7.F Primary Containment Isolation Valves Located in Water Sealed Seismic Class 1 Lines.

3.7/4.7-38 3.7.H Testable Electrical Penetrations.

3.7/4.7-39 4.9.A Auxiliary Electrical Systems.

3.9/4.9-16 4.9 A.4.C Voltage Relay Setpoints/ Diesel Generator Start 3.9/4.9-18 3.11.A Fire Detection Instrumentation.

3.11/4.11-14 3.11.B Spray / Sprinkler Systems 3.11/4.11-18 j

..o 2.11.C Hose Stations 3.11/4.11-20 3.ll.D Yard Fire Hydrants and Fire Hose Houses.

3.11/4.11-22 6.2.A Minimum Shi'ft Crew Requirements.

6.0-4 vii BFN Unit 1 Amendment Nos. 153,162,164,169

LIST OF ILLUSTRATIONS Ficure Title Pane No.

2.1.1 APRM Flow Reference Scram and APRM Rod Block Settings.

1.1/2.1-6

~

2.1-2

.APRM Flow Bias Scram Vs. Reactor Core Flow.

1.1/2.1-7 4.1-1 Graphical Aid in the Selection of an Adequate Interval Between Tests.

3.1/4.1-13 4.2-1 System Unavailability.

3.2/4.2-64 3.5.K-1 MCPR Limits.

3.5/4.5-24 3.5.2 Kf Factor..

3.5/4.5-25 3.6-1 Minimum temperature OF.Above Change in 3.6/4.6-24 Transient Temperature.

3.6-2 Change in Charpy V Transition Temperature Vs.

Neutron Exposure.

3.6/4.6-25 4.8.1.a Gaseous Release Foints and Elevations.

3.8/4.8-10 4.S.I.b Land Site Boundary.

3.8/4.8-11 Amendment No. 169 viii BFN Unit 1

n,,

o.

r 3,5I/i.5' CORE AND CONTAINMENT COOLING SYSTEMS t

SURVEILLANCE REQUIREMENTS l

LIMITING CON _DITIONS FOR OPERATION 4.5 CORE AND CONTAINMENT COOLING 3.5. [0RE AND CONTAINMENT C001 dG SYSTEMS EYSTEMS Aeolicability Aeolicability Applies to the operational' Applies to the surveillat e status of the core and-requirements of the. core and containment cooling systems when containment cooling systems.

the corresponding limiting condi-tion for operation is in effect.

Obiective P

Obiective To verify the OPERABILITY of the h

To assure the OPERABILITY of core and containment cooling the core and containment cooling systems under all conditions for systems under all conditions for which this cooling capability is which this cooling capability is an essential response to plant an essential response to plant abnormalities.

abnormalities, Specification Specification

'A.

Core Sorav System (CSS)

A.

Core Soray System (CSS) 1.

The CSS shall be OPERABLE:

1.

Core Spray System Testing.

lism Frecuency (1) PRIOR TO STARTUP from a a.

Simulated Once/

COLD CONDITION, or L

Automatic Operating Actuation Cycle (2) when there is irradiated test fuel in the vessel and when the reactor b.

Pump OPERA-Per Speciff-vessel pressure BILITY cation 1.0.MM is greater than atmospheric pressure, c.

Motor Per Specifi-except as specified Operated cation 1.0.MM in Specification Valve 3.5.A.2.

OPERABILITY d.

System flow Once/3 rate: Each months loop shall deliver at least 6250 gpm against a system head corres-ponding to a 3.5/4.5-1 A,nendment No.159 BFN Unit 1

(7( -

n s

-3.3/4.5 CORE AND CONTAI'NMENT COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.A-Core Sorav System (CSil 4.5.A Core Sorav System (CSS) 4.5.A.1.d (Cont'd) 105 psi

. differential pressure between the reactor vessel and the primary containment.

e.

Check Valve Per Specification 1.0.MM 2.

If one CSS loop is inoperable, f.

Verify that Once/ Month the reactor may remain in each valve operation for a period not to (manual, power-exceed 7 days providing operated, or all active components in automatic) in the the other CSS loop and the injection flowpath RHR system (LPCI mode) that is not locked, and the diesel generators sealed, or other-are' OPERABLE.

wise secured in position, is in its correct

position.

3.

If Specification 3.5.A.1 or 2.

No additional surveillance Specification 3.5.A.2 cannot is required.

be met, the reactor shall be placed in the COLD SHUTDOWN CONDITION within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , 4.

When the reactor vessel pressure is atmospheric and irradiated fuel is in the reactor vessel at least one core spray loop with one I

Except that an automatic OPERABLE pump and ansociated dfasel generator shall be valve capable of automatic OPERABLE, except with the return to its ECCS position reactor vessel head removed when an ECCS signal is as specified in 3.5.A.5 or present may be in a PRIOR.TO STARTUP as position for another mode specified'in 3.5.A.1.

of operation.

f i

t i

3.5/4.5-2 Amendment No. 153, 159, 169

{

BTN l

Unit 1 1

I I

L5f4.5 CORE AND CONTAINMEiff COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREliENTS 3 5.A Gare Sorav System (CSS) 5.

When irradiated fuel is in the reactor vessel e.nd the reactor vessel head is removed, core spray is not required to be OPERABLE provided the cavity is flooded, the fuel pool gates are open and the fuel pool water level is maintained above the low level alarm point, and provided one RERSW pump and associated valves supplying the standby coolant supply are OPERABLE.

l When work is in progress which has the potential to drain the vessel, manual initiation capability of either 1 CGS Loop or 1 RHk pump, with the capability of injecting water into the reactor vessel, and the associated diesel generator (s) are required.

BFN 3.5/4.5-3 A,nendment No.161 Unit 1 i

3.5/4.5 CORE AND CONTAINMENT COOLING SYSTEMS-LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.B-Residual Heat Removal System 4.5.B. Residual Heat Removal System (RHRS) (LPCI and Containment (RHRS) (LPCI and. Containment Cooling)-

Cooling) 1.

The RHRS shall be OPERABLE:

1. a.

Simulated Once/

Automar5' Operating (1) PRIOR TO STARTUP Actuaticu Cycle from a COLD Test CONDITION; or (2) when there is b.

Pump OPERA-Per irradiated fuel in BILITY Specification the reactor vessel 1.0.MM 7-and when the reactor l.

vessel pressure is c.

Motor Opera-Per l

greater than ted valve Specification i

atmospheric, except as OPERABILITY 1.0.MM specified in Specifientions 3.5.B.2, d.

Pump Tiow Once/3 through 3.5.B.7.

Rate months e.

Test Chcek Per Valve Specification 1.0.MM f.

Verify that Once/ Month each valve (manual, power-operated, or automatic) in the injection flow-path that is not locked, sealed, or otherwise secured in posi-tion, is in its correct

position.

g.

Verify LPCI Once/ Month subsystem cross-tie valve is closed and power removed from valve operator.

Except that an automatic valve capable of auto-matic return to its ECCS position when an ECCS signal is

)

l present may be in a position for another mode of operation.

BFN 3.5/4.5-4 Unit 1 Amendment Nos. 159,169

.e.

3.5/4.5 CORE AND CONTAINMENT COOLING SYSTEMS LIMITING CONDITIONS FOR' OPERATION SURVEILLANCE REQUIREMENT I

3.5.B Residual Heat Removal System 4.5.B. Residual Heat Removal System l

(RHRS) (LPCI and Containment (RHRS) (LPCI and Containment Cooling)

Cooling) 4.5.B.1 (cont'd) 2.

With'the reactor vessel Each LPCI pump shall deliver i

pressure.less than 105 psig, 9000 gpm against an indicated the RHRS may be removed system pressure of 125 psig.

4 from service (except that two Two LPCI pumps in the same-RHR pumps-containment cooling loop shall deliver 12000 gpm 1

mode and associated heat against'an indicated system exchangers must remain pressure of 250 psig.

OPERABLE) for a period not to exceed 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months while 2.

An air test on the drywell.

l being drained of-and torus headers and nozzles suppression chamber quality shall be conducted once/5 vater and filled with years. A water test may be primary coolant quality performed on the torus header water provided that during in lieu of the air test, cooldown two loops with one pump per loop or one loop with two pumps,-and associated diesel generators, in the core spray system are OPERABLE.

3.

If one RHR pump (LPCI mode) 3.

No additional surveillance is inoperable, the reactor

required, may remain in operation for a period not to exceed 7 days provided the remaining RHR pumps (LPCI mode) and both access paths of the RHRS (LPCI mode) and the CSS and the diesel generators remain OPERABLE.

4 If any 2 RHR pumps (LPCI 4.

No additional surveillance mode) become inoperable, the required.

reactor shall be placed in the COLD SHUTDOWN CONDITION within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l' l

BFN 3.5/4.5-5 Amendment No. 153, 169 Unit 1

~~__-_ ___-_

1

7.

1

..a 3.5/4.5 CORE ' ANDiCONTdINMENT C00L8NG SYSE2iE -

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENTS.

3.5 B.

Residual Heat Removal System 4.5 B.

Residual Heat Removal System-IRHRS) (LPCI and Containment (RHRS) (LPCI and Containment Cooling)

Cooling) 5.

If one RHR pump (containment 5.

No additicual surveillance l

cooling mode) or associated required.

heat exchanger is: inoperable,.

the reactor may remain in operation for a period not to exceed 30 days provided the remaining RER pumps (containment. cooling mode) and associated heat exchangers and diesel generators and all access paths of the RHRS (containment cooling mode) are OPERABLE.

6.

If two RHR pumps (containment 6.

No additional surveillance' cooling mode) or associated required.

heat exchangers are inoperable, the reactor may remain in operation for a period not to exceed 7 days provided the remaining RHR pumps (containment cooling mode), the associated heat exchangers, diesel generators, and all access paths of the RHRS (containment cooling mode)

R are OPERABLE.

7.

If two access paths of the 7.

No additional surveillance RHRS (containment cooling required.

mode) for each phase of the mode (drywell sprays, suppression chamber sprays, and suppression pool cooling) are not OPERABLE, the unit may remain in operation for a period not to exceed 7 days provided at'least one path for each phase of the mode remains OPERABLE.

BEN 3.5/4.5-6 Amendment No. 153, 169 Unit 1 L____________-________-_____-___--___-_-

3.5/4.5 CORE AND, CONTAINMENT COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.B Residual Heat Removal SystIm 4.5.B Residual Heat Removal System (RHRS) (LPCI and Containment (RHRS) (LPCI and Containment Cooling)

Cooling)

8. If Specifications 3.5.B.1 8.

No additional surveillance through 3.5.B.7 are not met, required.

an orderly shutdovn shall be l

initiated and the reactor shall be placed in the COLD SHUTDOWN CONDITION l

vithin 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

9.

When the reactor vessel 9.

When the reactor vessel pressure is atmospheric and pressure is atmospheric, irradiated fuel is in the the RHR pumps and valves l

reactor vessel, at least one RHR that are required to be

{

loop with two pumps or two loops OPERABLE shall be with one pump per loop shall demonstrated to be OPERABLE be OPERABLE. The pumps' per Specification 1.0.MM.

associated diesel generators must also be OPERABLE.

10. If the conditions of 10.

No additional surveillance Specification 3.5.A.5 are met, required.

LPCI and containment cooling are not required.

11. When there is irradiated fuel 11.

The RHR pumps on the in the reactor and the reactor adjacent units which supply vessel pressure is greater than cross-connect capability atmospheric, 2 RHR pumps and shall be demonstrated to be associated heat exchangers and OPERABLE per Specification valves on an adjacent unit must 1.0.MM when the cross-be OPERABLE and capable of connect capability supplying cross-connect is required.

capability except as specified in Specification 3.5.B.12 below.

(Note:

Because cross-connect capability is not a short-term requirement, a component is not considered inoperable if cross-connect capability can be restored to service within 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months .)

BFN 3.5/4.5-7 Amendment No. 159, 169 Unit 1

i.

L3 /4. 5 CORE AWD C0iUAINMEfE COOLING ~ SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.B Residual Heat Removal System 4.5.B Residual Heat Removal System J

(RHRS1 (LPCI and Containment (RHRS) (LPCI and Containment f

Cooling)

Cooling) i I

12. If one RHR pump or associated

12. No additional surveillance j

heat exchanger located required.

on the unit cross-connection i

i in the adjacent unit is j

j inoperable for any reason (including valve inoperability, pipe break, etc.), the reactor j

may remain in operation J

for a period not to exceed 30 days provided the remaining RHR pump and associated diesel generator are OPERABLE.

j 13.

If RHR cross-connection flow or 13.

No additional surveillance heat removal capability is lost,

required, the unit may remain in operation for a period not to exceed 10 days unless such capability is j

restored.

l l

14 All recirculation pump 14.

All recirculation f$rp discharge valves shall discharge valves shall be OPERABLE PRIOR TO be tested for OPERABILITY STARTUP (or closed if during any period of permitted elsewhere COLD SKUTDOWN CONDITION in these specifications) exceeding 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , if OPERABILITY tests havc not been performed during the preceding 31 days.

i l

Amendment No. 158, 169 BFN 3.5/4.5-8 Unit 1 i

I

46.

3.5/4.5 CORE AND CONTAINMENT'C00 LING SYSTEMS LIMITING CONDITIONS FOR' OPERATION

' SURVEILLANCE REQUIREMENTS 3.5.C RHR Service Water and Emergency 4.5.C RHR Service Water and EmerRency Ecuipment Cooling Water Systems Eauipment Coolina Water Systems (EECWS)

(EECWS) 1.

PRIOR TO STARTUP.from 1.

a.

Each of the RHRSW pumps a COLD CONDITION, 9 RHRSW normally assigned to pumps must be OPERABLE,.with automatic service.on 7 pumps (including pump D1 the EECW headers will or D2) assigned to RHRSW be tested service and 2 automatically automatically each time starting pumps assigned to the diesel generators EECW service.

are tested. Each of the RHRSW pumps and all associated essential control valves for the EECW headers and RHR heat exchanger headers shall be demonstrated to be OPERABLE in accordance with-Specification 1.0.MM.

b.

Annually each RHRSW pump shall be flow-rate tested. To be considered OPERABLE, each pump shall pump at least 4500 gpm through its normally assigned flow

path, c.

Monthly verify that each valve (manual, power-operated, or automatic) in the flowpath servicing safety-related equipment in the affected unit that is not locked, sealed, or otherwise secured in position, is in its correct position.

BFN 3.5/4.5-9 Amendment No. 159, '69 i

Unit 1

Im /d. 5-CORE AND CONTAINMENT COOLING SYSTEMS s

' LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS

'3.5.C-

-RHR Service Water and Emeraency 4.5.C RHR Service Water and Emergency

~

Eauipmen? Coolina Water Systems Eauipment Coolina Water Systems LEECWS) (Continued):

(EECWS) (Continued)

'2.

During REACTOR POWER

2. No additional surveillance OPERATION, RHR3W pumps is required.

must be OPERABLE and assigned to service as indicated in Table 3.5-1 for the specified time limits.

3.

During REACTOR POWER

3. Routine surveillance for OPERATION, both RHRSW.

these pumps is specified pumps D1 and D2 normally in 4.5.C.1.

or alternately assigned.

to the RER heat exchanger header supplying the standby coolant supply connection must be OPERABLE except as specified in 3.5.C.4 and 3.5.C.5 below.

i BFN 3.5/4.5-10 Amendment No. 153, 169 l

Unit 1 I

v TABLE 3.5-1 Minimum Time Service Assignment Limit EF_Qy(2)

RHRSW (Days)

(1)

(4) 3 7

Indefinite (3)(4)

(1)

(3) 7 ott 6 2

or 3 30 (1)

(4) 2 6

7 one OPERABLE pump must be assigned to each header, i

(1)

At least Only automatically starting pumps may be assigned to EECW header (2) service.

Either configuration is Nine pumps must be OPERABLE.

(3) acceptable: 7 and 2 or 6 and 3.

I Requirements may be reduced by two fer each unit with fuel (4) unloaded.

I 3.5/4.5-11 BFN Unit 1

.a 'e

'b, THIS PAGE INTENTIONALLY LEFT BLANK l

s BFN 3.5/4.5-11a Amendment No. 169 Unit 1 lL.

o 3.5/4.5 CORE AND CONTAINMENT COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.C RHR Service Water and Emergency 4.5 C RHR Service Water and Emergency Eculoment Cooline Water Systems Eauioment Coolina Water Systems (EECW Q (Continued _1 (EECWST LContipuRO 4

One of the D1 or D2 RHRSW

4. No addi'tional surveillance pumps assigned to the RHR is required.

heat exchanger supplying the standby coolant supply connection may be inoperable for a period not to exceed 30 days provided the OPERABLE pump is aligned to supply the RER heat exchanger header and the associated diesel generator and essential control valves are OPERABLE.

5.. The standby coolant supply capability may be inoperable for a period not to exceed 10 days.

6.

If Specifications 3.5.C.2 through 3.5.C.5 are not met, an orderly shutdown shall be initiated and the unit placed in the COLD SHUTDOWN CONDITION within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

7.

There shall be at least 2 RERSW' pumps, assorauted with the selected RHR pumps, aligned for RHR heat exchanger service for each reactor vessel containing irradiated fuel.

BFN 3.5/4.5-12 Amendment No. 153, 169 Unit 1

3.5/4.5 CORE AND CONTAINMEin C0061NG SYSTEMS i

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS _

\\

3,5.D Eauipment Area Coolers 4.5.D Eauipment Area Coolers 1.

The equipment area cooler

1. Each equipment area cooler associated with each RHR is operated in conjunction pump and the equipment with the equipment served area cooler associated by that particular cooler; with each set of core therefore, the equipment spray pumps (A snd C area coolers are tested at or B and D) must be the same frequency as the OPERABLE at all times pumps which they serve.

when the pump or pumps served by that specific cooler is considered to be OPERABLE.

2.

When an equipment area cooler.is not OPERABLE, the pump (s) served by that cooler must be considered inoperable for Technical Specification purposes.

E.

High Pressure Coolant In_i e c': ion E. High Pressure Coolant System (HPCIS)

Iniection System (HPCIS) 1.

The HPCI system shall be 1.

HPCI Subsystem testing OPERABLE:

sha}l be performed as follows:

(1) PRIOR TO STARTUP from a a.

Simulated Once/

COLD CONDITION; or Automatic operating Actuation cycle Test (2) whenever there is b.

Pump Per irradiated fuel in the OPERA-Specification reactor vessel and the BILITY 1.0.MM reactor vessel pressure is greater than 122 peig, c.

Motor Oper-Per except as specified in ated Valve Specification Specification 3.5.E.2.

OPERABILITY 1.0.MM d.

Flow Rate at Once/3 I

normal months reactor vessel operating pressure 1

BFN 3.5/4.5-13 Amendment No. 159 Unit 1

223/4i5 ' CORE'AND' CONTAINMENT COOLTNG SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.E High Pressure Coolant'Iniection 4.5.E High Pressure Coolant Iniecti2D

System (HPQLEl System (HPCIS) 4.5 E.1 (Cont'd) e.

Flow Rate at Once/

150 psig operating cycle The HPCI pump shall deliver at least 5000 gpm during each flow rate test.

f.

Verify that Once/ Month each valve (manual, power-operated, or automatic) in the d

injection flow-path that is not locked, sealed, or otherwise secured in position,.is in its correct

position.

2.

If the HPCI system is

2. No additional surveillance inoperable, the reactor may are required.

remain in operation for a period not to exceed 7 days, provided the ADS, CSS, RHRS (LPCI), and RCICS are OPERABLE.

3.. If Specifications 3.5.E.1 Except that an automatic or 3.5.E.2 are not met, valve capable of automatic an orderly shutdown shall return to its ECCS position be initiated and the when an ECCS signal is reactor vessel pressure present may be in a shall be reduced to 122 position for another mode of psig or less within 24 operation.

hours.

F.

Reactor Core Isolation Coolina F.

Reactor Core Isolation Coolina System (RCICS)

System (RCICS) 1.

The RCICS shall be OPERABLE:

1. RCIC Subsystem testing shall be performed as follows:

(1) PRIOR TO STARTUP from a COLD CONDITION; or

a. Simulated Auto-Once/

matic Actuation operating Test cycle BFN 3.5/4.5-14 Amendment No. 159, 169 Unit 1

_ - _ _ - - - _ = - -

s 16,

3.5/4.5' CORE AND CONTAINMENT COOLING SYSTEMS 1

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS l

q 3.5.F.-

Reactor Core Isolation Coolina 4.5.F Reactor Core Isolation Coolina

-System (RCICS)

System (RCICSi 3.5.F.1 (Cont'd) 4 5.F.1 (Cont'd)

]

l (2) whenever there is

b. Pump Per_

J irradiated fuel in the OPERABILITY Specifi-

)

reactor vessel and the cation reactor vessel pressure 1.0.MM is above 122 psig, 3

except as specified in

c. Motor-Operated Per q

3.5.F.2.

Valve Specifi-j OPERABILITY cation'

,1.0.NM 3

I d.

Flow Rate at Once/3 ~

normal reactor months vessel. operating j

pressure q

i e.

Flow Rate at Once/

)

150 psig operating-l cycle j

The RCIC pump shall i

deliver at least 600 gpm during each flow test.

2.

-If the RCICS is inoperable, f.

Verify that Once/ Month the reactor may remain in ee.ch valve operation for a period not (manual, power-to exceed 7 days if the operated, or HPCIS is OPERABLE during automatic) in the such time.

injection flovpath that is not locked, 3.

If Specifications 3.5.F.1 sealed, or other-or 3.5.F.2 are not met, an vise secured in orderly shutdown shall be position, is in its initiated and the reactor correct

position.

shall be depressurized to less than 122 psig within

2. No additional surveillance 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

are required.

Except that an automatic valve capable of automatic return to its normal position when a signal is present may be in a position for another mode of operation.

BFN 3.5/4.5-15 Unit 1 Amendment No. 169 I

I 3.5/4.5: ' CORE AND CONTAINMENT-COOLING tYSTEMS j

i LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS j

3.5.G Automatic Deoressurizatian 4.5.G Automatic Deeressurization System (ADS)

System (ADS) i c.

l.

Four of the six. valves of 1.

.During each operating the Automatic cycle.the following Depressurization System tests shall be performed shall be OPERABLE:

cn the ADS:

(1) PRIOR TO STARTUP a.

A simulated automatic from a COLD CONDITION, actuation test shall or, be performed PRIOR TO I

STARTUP after each (2) whenever there is refueling outage.

. irradiated fuel in the Manual surveillance reactor vessel and the of the relief valves reactor vessel pressure is covered in is greater than 105 psig, 4.6.D.2.

except as specified in 3.5.G 2 and 3.5 G.3 below.

2.

If.three of the six ADS 2.

No additional surveillance valves are known to be are required.

incapable of automatic operation, the reactor may remain'in operation for a period not to exceed 7 days, provided the HPCI system is OPERABLE. (Note that the pressure relief function of these valves is assured by Section 3.6.D of these specifications and that this specification only applies to the ADS function.)

If-more than three of the six ADS valves are kncwn to be incapable of automatic operation, an immediate orderly shutdown shall be initiated, with the reactor in a HOT SHUTDOWN CONDITION in 6 houre, and in a COLD SHUTDOWN CONDITION in the following 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months .

3.

If Specifications 3.5.G.1 and 3.5.G.2 cannot be met, an orderly shutdown will be initiated and the reactor vessel pressure shall be reduced to 105 psig or less within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

BFN 3.5/4.5-16 Unit 1 Amendment No. 169

k,;_

I 3 5/4.5 CORE AND CONTAINMENT COOLING SYSTEMS

' LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.H.

Maintenance of Filled Discharge 4.5.H.' Maintenance of Filled Discharge Elpl Eip3 Whenever the core spray systems, The following surveillance LPCI, HPCI, or RCIC are required requirements shall be adhered to be OPERABLE, the discharge to assure that the discharge piping from the pump discharge-piping of the core spray of.these systems to the last systems, LPCI, HPCI, and RCIC block valve shall be filled.

are filled:

The suction;of the RCIC and HPCI

1. Every month and prior to the l

pumps shall be aligned to the testing of the RHRS (LPCI and condensate storage tank, and Containment Spray) and core.

the pressure suppression chamber spray system, the discharge head tank shall normally be aligned piping of'these systems shall to serve the discharge piping of be vented from the high point the RHR and CS pamps. The and water flow determined.

condensate head tank may be used to serve the RHR and CS discharge

2. Following any period where the piping if the PSC head tank LPCI or core spray systems is unavailable. The pres;ure have not been required,to be indicators on the discharge of the OPERABLE, the discharge piping RHR and CS pumps shall indicate of the inoperable system shall not less than-listed below.

he vented from the high point prier to the return of the F1-75-20 48 psig srstem to service.

P1-75-48 48 psig P1-74-51 48 psig

3. Vaenever the HPCI or RCIC P1-74-65 48 psig system is lined up to take suction from the condensate storage tank, the discharge piping of the HPCI and RCIC shall be vented from the high point of the system and water flow observed on a monthly basis.

4. When the RHRS and the CSS are required to be OPERABLE, the pressure indicators which monitor the discharge lines shall be monitored daily and the pressure recorded.

l BPN 3.5/4.5-17 Amendment No. 159 Unit 1

3.5 E&EXS 3.5.A.

Core Spray System (CSE1 and 3.5.B Residual Heat Removal System (RHES1 Analyses presented in the FSAR* and analyses presented in conformance with 10 CFR 50, Appendix K, demonstrated that the core spray system in conjunction with two LPCI pumps provides adequate cooling to the core to dissipate the energy associated with the lous-of-enolant accident and to limit fuel clad temperature to below 2,200*F which assures that core geometry remains intact and to limit the core average clad metal-water reaction to less than 1 percent. Core spray distribution has been shown in tests of systems similar to design to BFNP to exceed the minimum requirements. In addition, cooling effectiveness has been demonstrated at less than half the rated flow in simulated fuel assemblies with heater rods to duplicate the decay heat characteristics of irradiated fuel.

The RHRS (LPCI mode) is designed to provide emergency cooling to the core by flooding in the event of a less-of-coolant accident. This system is completely independent of the core spray system; however, it does function in combination with the core spray system to prevent excessive fuel clad temperature. The LPCI mode of the RHRS rnd the core spray system provide adequate cooling for break areas of approximately 0.2 square feet up to and including the double-ended recirculation line break without assistance from the high-pressure emergency core cooling subsystems.

The intent of the CSS and RHRS specifications is to not allow startup from the cold condition without all associated equipment being OPERABLE.

However, during opera *. en, certain components may be out of service for the specified allowable repair times. The allowable repair times have been selected using engineering judgment based on experiences and supported by availability analysis.

Should one core spray loop become inoperable, the remaining core spray loop, the RHR System, and che diesel generators are required to be OPERAELE should the need for core cooling arise. These provide extensive margin over the OPERABLE equipment needed for adequate core cooling.

With due regard for this margin, the allowable repair time of seven days was chosen.

Should one RHR pump (LPCI mode) become inoperable, three RHR pumps q

(LPCI mode) and the core spray system are available.

Since adequate core cooling is assured with this complement of ECCS, a seven day repair period is justified.

Should two RHR pumps (LPCI mode) become inoperable, there remains no reserve (redundant) capacity within the RHRS (LPCI mode).

Therefore, the affected unit shall be placed in cold shutdown within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

A detailed functional analynis is given in Section 6 of the BFNP FSAR.

BFN 3.5/4.5-26 Amendment No. 169 Unit 1

3.5 EA151 (Cont'd)

Should one RHR pump (containment cooling mode) become inoperable, a complement of three full capacity containment heat removal systems is still available. Any two of the remaining pumps / heat exchanger

. combinations would provide more than adequate containment cooling for any

?

abnormal or postaccident situation. Because of the availability of equipment in excess of normal redundancy requirements, a 30-day repair l

period is justified.

Should two RHR pumps (containment cooling mode) become inoperable, a full heat removal system is still available. The.renaining pump / heat exchanger combinations would provide adequate containment cooling for any abnormal postaccident situation. Because of the availability of a full

)

complement of heat removal equipment, a 7-day repair period is justified.

l Observation of the stated requirements for the containment cooling mode I

assures that the suppression pool and the drywell will be sufficiently coolid, following a loss-of-coolant accident, to prevent primary containment overpressurization. The containment cooling function of the RHRS is permitted only after the core has reflooded to the two-thirds core height level. This prevents inadvertently diverting water needed for core flooding to the less urgent task of containment cooling. The two-thirds core height level interlock may be manually bypassed by

keylock switch.

3 k

Since the RERS is filled with low quality water during power operation, l

it is planned that the system be filled with demineralized (condensate) water before using the shutdown cooling function of the RHR System.

Since it,is desirable to have the RHRS in service if a " pipe-break" type of accident should occur, it is permitted to be out of operation for only a restricted amount of time and when the system pressure is low. At least one-half of the containment cooling function must remain OPERABLE during this time period. Requiring two OPERABLE CSS pumps during cooldown allows for flushing the RHRS even if the shutdown were caused by inability to meet the CSS specifications (3.5. A) on a number of OPERABLE pumps.

When the reactor vessel pressure is atmospheric, the limiting conditions for operation are less restrictive. At atmospheric pressure, the minimum requirement is for one supply of makeup water to the core.

Requiring two OPERABLE RHR pumps and one CSS pump provides redundancy to ensure makeup water availability.

Should one RHR pump or associated heat exchanger located on the unit cross-cannection in the adjacent unit become inoperable, an equal capability for long-term fluid makeup to the reactor and for cooling of the containment remains OPERABLE. Because of the availability of an equal makeup and cooling capability, a 30-day repair period is justified.

l BFN 3.5/4.5-27 Amendment No. 169 Unit 1

3.5 Bases (Cont'd)

The suppression chamber can be drained when the reactor vessel pressure is atmospheric, irradiated fuel is in the reactor vessel, and work is not in progress which has the potential to drain the vessel. By requiring the fuel pool gate to be open with the vessel head removed, the combined water inventory in the fuel pool, the reactor cavity, and the separator / dryer pool, between the fuel pool low level alarm and the reactor vessel flange, is about 65,800 cubic feet (492,000 gallons).

This will provide adequate low-pressure cooling in lieu of CSS and RHR (LPCI and containment cooling mode) as currently required in Specifications 3.5.A.4 and 3.5.B.9.

The additional requirements for providing standby coolant supply available will ensure a redrndant supply of coolant supply. Control rod drive maintenance may continue during this period provided no more than one drive is removed at a time unless blind flanges are installed during the period of time CRDs are not in place.

Sheuld the capability for providing flow through the cross-connect lines be lost, a 10-day repair time is allowed before shutdown is required.

This repair time is justified based on the very small probability for ever needing RHR pumps and heat exchangers to supply an adjacent unit.

REFERENCES 1.

Residual Heat Renoval System (FFNP FSAR subsection 4.8) 2.

Core Standby Cooling Systems (BFNP FSAR Section 6) l 3.5 C.

RHR Service Water Systen_and Emergency Eauiement Coolina Water System (EECWS)

There are two EECW headers (north and south) with four automatic starting RPRSW pumps on each header. All components requiring emergency cooling water are fed from both headers thus assuring continuity of operation if either header is OPERABLE. Each header alone can handle the flows to all components. Two RHRSW pumps can supply the full flow requirements of all essential EECW loads for any abnormal or postaccident situation.

There are four RHR heat exchanger headers (A, B, C, & D) with one RHR heat exchanger from each unit on each header. There are two RHRSW pumps on each header; one normally assigned to each header (A2, B2, C2, or D2) l ano one on alternate sssignment (A1, B1, C1, or D1). One RHR heat l

exchanger header cat adequately deliver the flow supplied by both RHRSW j

pumps to any two of the three RHRSW heat exchangers on the header. One RHRSW pump can supply the full flow requirement of one RHR heat exchanger. Two RHR heat exchangers can mnre than adequately handle the l

cooling requirements of one unit in any aonormal or postaccident j

situation.

l l

BFN 3.5/4.5-28 Unit 1

.c

=>-

-3.5 MSJJ (Cont'd)

The RHR Service Water System was designed as a shared system for three units. The specification, as written, is conservative when consideration is given to particular pumps being out of service and to possible valving arrangements. If unusual operating conditions arise such that more pumps are out of service than allowed by this specification, a special case request may be made to the NRC to allow continued operation-if the actual system cooling requirements can be assured.

Should one of the two RURSW pumps normally or alternately assigned to the RHR heat exchanger header supplying the standby coolant supply connection become inoperable, an equal capability for long-term fluid makeup to the unit reactor and for cooling of the unit containment remains OPERABLE. Because of the availability of an equal makeup and cooling capability, a 30-day repair period is justified. Should the capability to provide standby coolant supply be lost, a 10-day repair time is justified based on the low probability for ever needing the standby coolant supply. Verification that the LPCI subsystem cross-tie valve is closed and power to its operator is disconnected ensures that each LPCI subsystem remains independent and a failure of the flow path j

in one subsystem will not affect the flow path of the other LPCI subsystem.

l 3.5.D Eauiement Area Coolers There is an equipment area cooler for each RHR pump and an equipment area cooler for each set (two pumps, either the A and C or B and D pumps) of core spray pumps. The equipment area coolers take suction near the cooling air discharge of the motor.of the pump (s) served and discharge air near the cooling air suction of the motor of the pump (s) served. This ensures that cool air is supplied for cooling the pump motors.

The equipment area coolers also remove the pump, and equir c waste heat from the basement rooms housing the engineered safegut.d equipment. The various conditions under which the operation of the equipment air coolers is required have been identified by evaluating the normal and abnormal operating transients and accidents over the full range of planned operations. The surveillance and testing of the equipment area coolers in each of their various modes is accomplished during the testing of the equipment served by these coolers. This testing is adequate to assure the OPERABILITY of the equipment area coolers.

REFERENCES 1.

Residual Heat Removal System (BFN FSAR Section 4.8) 2.

Core Standby Cooling System (BFN FSAR subsection 6.7) l BFN 3.5/4.5-29 Amendment No. 169 Unit 1 i

l

n j

3.5' EASIS (Cont'd) 3~5.E. Hinh Pressure Coolant Iniection System (HPCIS)

The HPCIS is provided to assure that the reactor core is adequately.

cooled to limit fuel clad temperature -in the event of a small break in i

the nuclear system and loss of coolant which does not result in rapid

-)

depressurization of the _ reactor ressel. The HPCIS permits the reactor to be shut down while maintaining sufficient reactor vessel water level inventory until the vessel is depressurized. The HPCIS continues to operate until reactor vessel pressure is below the pressure at which-LPCI operation or core spray system operation maintains core cooling.

The capacity of the system is selected to provide this required core cooling. The HPCI pump is designed to pump 5,000 gpm at reactor pressures between 1,120 and 150 psig. Two sources of water are available.

Initially, water from the condensate storage tank is used instead of injecting water from the suppression pool into the reactor.

When the HPCI ;; stem begins operation, the reactor depressurizes more rapidly than would occur if HPCI was not initiated due to the-condensation of steam by the cold fluid pumped into the reactor vessel by the HPCI system. As the reactor vessel pressure ccatinues to decrease, the MPCI flow momentarily reaches equilibrium with the flow through the break.

Continued depressurization caused the break flow to decrease below the HPCI flow and the liquid inventory begins to rise.

This. type of response is. typical of the small_breaXs. The core never uncovers and is continuously cooled throughout the transient so that no core damage of any kind occurs for breaks that lie within the capacity L

range of the HPCI.

The minimum required NPSH for HPCI is 21 feet. There is adequate elevation head between the suppression pool and the HPCI pump, such that the required NPSH is available with a suppression pool temperature up to 140*F with no containment back pressure.

l l

The HPCIS serves as a backup to the RCICS as a source of feedwater makeup during primary system isolation conditions. The ADS serves as a backup to the HPCIS for reactor depressurization for postulated transients and accident. The CSS and RHRS (LPCI) provide adequate core cooling at low reactor pressure when RCICS and ADS are no longer necessary. Considering the redundant systems, an allowable repair time of seven days was selected.

The HPCI and RCIC as well as all other Core Standby Cooling Systems must be OPERABLE when starting up from a Cold Condition.

It is realized that the HPCI is not designed to operate at full capacity until reactor pressure exceeds 150 psig and the steam supply to the HPCI turbine is automatically isolated before the reactor pressure decreases below 100 psig.

It is the intent of this specification to assure that when the reactor is being started up from a Cold Condition, the HPCI is not known to be inoperable.

BFN 3.5/4.5-30 Amendment No. 169 Unit 1 i

u.

+.

y

.l 3.5

. BASES (Cont'd)

" 3. 5. F: Reactor Core Isolation Cooling System (RCICS)-

The various conditions under which the RCICS plays an essential role in providing makeup water to the reactor vessel have been identified by evaluating the various plant events over the full range of planned j

operations.- The specifications ensure that the function for which the RCICS was designed will be available when needed. The minimum required I

NPSH for RCIC is 20 feet. There is adequate elevation head between the suppression pool and the RCIC pump, r.uch that the required NPSH is available with a suppression pool temperature up to 140*F with no containment back pressure.

Because the low-pressure cooling systems (LPCI and core spray) are capable of providing all the cooling required for any plant event when nuclear system pressure is below 122 psig, the RCICS is not required below this pressure. Between 122 psig and 150 psig the RCICS need not provide its design flow,-but reduced flow is required for certain events. RCICS design flow (600 gpm) is sufficient to maintain water level'above the top of the active fuel for a complete loas of feedwater i

flow at design power (105 percent of rated).

Consideration of the availability of the RCICS reveals that the average risk associated with failure of the RCICS to cool the core when required is not increased if the RCICS is inoperable for no longer than seven days provided that the HPCIS is OPERABLE during this period.

REFERENCE 1.

Reactor Core Isolation Cooling System (BFNP FSAR Subsection 4.7) 3.5.G Automatic Depressurization System _[AHS)

This specification ensures the OPERABILITY of the ADS under all conditions for which the depressurization of the nuclear system is an essential response to station abnormalities.

The nuclear system pressure relief system provides automatic nuclear system depressurization for small breaks in the nuclear system so that the low-pressure coolant injection (LPCI) and the core spray subsystems car operate to protect the fuel barrier. Note that this specification applies only to the automatic feature of the pressure relief syatem.

Specification 3.6.D specifies the requirements for the pressure relief function of the valves. It is possible for any number of the valves assigned to the ADS to be incapable of performing their ADS functions because of instrumentation failures yet be fully capable of performing their pressure relief function.

3FH 3.5/4.5-31 l

Unit 1

,a 3IS BASES'(Cont'd).

Because;the automatic depressurization. system does not. provide makeup to the reactor primary vessel, no credit is taken for the steam cooling of the core caused by the system actuation to provide further conservatism to the CSCS.

With two ADS valves known to be incapable of automatic operation, four valves remain 0PERABLE to perform their ADS function. The ECCS loss-of-coolant accident analyses.for small line breaks assumed that.four of the six ADS valves were OPERABLE. Reactor operation with three ADS valves inoperable is allowed to continue for seven days provided that the HPCI system is OPERABLE. Operation with more than three of the six ADS valves inoperable is not acceptable.

H.

Maintenance of Filled Discharge Pine If the discharge piping of the core spr.ay, LPCI, HPCIS, and RCICS are not filled, a water hammer can develop in this piping when the pump and/or pumps are started. To minimize damage to the discharge piping and to ensure added margin in the operation of these systems, this Technical Specification requires the discharge lines to be filled whenever the system is in an OPERABLE condition.

If a discharge pipe is not filled, the pumps that supply that line must be assumed to be inoperable for Technical Specification purposes.

The core spray and RHR r,ystem discharge piping high point vent is visually checked for water flow once a month and prior to testing to ensure that the lines are filled. The visual checking will avoid starting the core spray or RHR system with a discharge line not filled.

In addition to the visual observation and to ensure a filled discharge line other than prior to testing, a pressure suppression chamber head tank is located approximately 20 feet above the discharge line high point to supply makeup water for these systems. The condensate head tank located approximately 100 feet above the discharge high point serves as a backup charging system when the pressure suppression chamber head tank is not.in service. System discharge pressure indicators are used to determine the water level above the discharge line high point. The indicators will reflect approximately 30 psig for a water level at the high point and 45 psig for a water level in the pressure suppression chamber head tank and are monitored daily to ensure that the discharge lines are filled.

When in their normal standby condition, the suction for the HPCI and RCIC pumps are aligned to the condensate storage tank, which is physically at a higher elevation than the HPCIS and RCICS piping. This assures that the HPCI and RCIC discharge piping remains filled. Further assurance is provided by observing water flow from these systems' high points monthly.

I.

Maximum Averace Planar Linear Heat Generation Rate (MAPLHGR)

This specification assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the limit specified in the 10 CFR 50, Appendix K.

^

nt1 Amendment No. 169

4 :

[3.5-BASES (Con: a):

Ite peak cladding temperature following a postulated. loss-of-coalant accident is primarily e function of the average heat generation Trite of all.the rods of a fuel assembly at any axial location and is only dependent secondarily on the rod-to-rod power distribution within an assembly.- Since expected local variations in power distributf.on within I,

a fuel assembly affect the calculated peak clad temperature by less i

than t 20*F relative to the peak temperature for a typical fuel design, the limit on the average linear heat sencration rate is sufficient to assure that calculated temperatures are within the 10 CFR 50 Appendix K limit. The limiting value for MAPLHGR is shown in Tables 3.5.I-1,

-2,

-3.

-4,

-5, and -6.

The analyses supporting these limiting values are p:2sented in Reference 4.

1 3.5.J. Linear Heat Generation Rate (LHGR)

This specification assures that the linear heat generation rate in any rod is less than the design linear heat generation if fuel pellet densification is postulated.

The LHGR shall be checked daily during reactor operation at i 25 percent power to determine if fuel burnup, or centrol rod movement has caused changes in power distribution. For LHGR to be a limiting value below 25 percent rated thermal power, the MTPF would have to be greater than 10 which is precluded by a considerable margin when employing any permissible control rod pattern.

3.5.K. Minimum Critical Power Ratio (MCPR)

At core thermal power levels less than or equal to 25 percent, the reactor will be operating at minimum recirculation pump speed and the moderator void content will be very small. For all designated control rod patterns which may be employed at this point, operating plant experience and therms 1 hydraulic analysis indicated that the resulting MCPR value is in excess of requirements by a considerable margin. With this low void content, any inadvertent core flow increase would only place operation in a more conservative mode relative to MCPR. The daily requirement for calculating MCPR above 25 percent rated thermal power is sufficicnt since power distribution shifts are very slow when there have not been significant power or control rod changes. The requirement for calculating MCPR when a limiting control rod pattern is approached ensures that MCPR will be known following a change in power or power shape (regardless of magnitude) that could place operation at a thermal limit.

3.5.L. APRM Setooints The fuel cladding integrity safety limits of Section 2.1 were based on a total peaking factor within design limits (FRP/CMFLPD 1 1.0).

The APRM instruments must be adjusted to ensure that the core thermal limits are not exceeded in a degraded situation when entry conditions are less conservative than design assumptions.

BFN 3.5/4.5-33 Unit I

3.5 BASES (Cont'd) 3.5.M. References

" Fuel Densification Effects on General Electric Boiling Water Reactor Fuel," Supplements 6, 7, and 8, NEIM-10735, August 1973.

1.

I to Technical Report on Densification of General 2.

Supplement Electric Reactor Fuels, December 14,1974 (USA Regulatory Staff).

V. A. Moore to I. S. Mitchell, " Modified GE Model Communication:

for Fuel Densification," Docket 50-321, March 27, 1974 3.

Generic Reload Fuel Application, Licensing Topical Report, 4

NEDE-24011-P-A and Addenda.

Letter from R. H. Buchholz (GE) to P. S. Check (NRC), " Response to NRC Request For Information On ODYN Computer Model," September 5, 5.

1980.

l l

l Amendment No. 147 3.5/4.5-34 BFN I

Unit 1

4.5-Core and Containment Coolina Systems Surveillance Frequencies The testing interval for the core and containment cooling systems is based on industry practice, quantitative reliability analysis, judgment and practicality. The core cooling systems have not been designed te be fully testable during operation. For example, in the case of the HPCI, automatic initiation during power operation would result in pumping cold water Into the reactor vessel which is not desirable. Complete ADS testing during power operation causes an undesirable loss-of-coolant To increase the availability of the core and containment inventory.

cooling system, the components which make up the system, i.e.,

instrumentation, pumps, valves, etc., are tested frequently. The pumps and motor operated inject ion valves are also tested in accordance with A simulated automatic Specification 1.0.MM to sssure their OPERABILITY.

actuation test once each cycle combined with testing of the pumps and injection valves in accordance with Specification 1.0.MM is deemed to be Monthly alignment checks of valves adequate testing of these systems.

that are not locked or sealed in position which affect the ability of the systems to perform their intended safety function are also verified to be Valves which automatically reposition themselves in the proper position.

on an initiation signal are permitted to be in a position other than normal to facilitate other operational modes of the system.

k* hen con ponents and subsyttems are out-of-service, overall core and cooling relietility is maintained by G?ERABILITY of the containment remaining redundant equipnent.

Whenever a CSCS system or loop is made inoperable, this other CSCS systems are required to be OPERABLE shall be censidered OPERABLE if or loops that they are within the required surveillance testing frequency and there is If the function, system, or no reason to suspect they are inoperable.

loop under test or calibration is found inoperable or exceeds the trip level setting, the LCO and the required surveillance testing for the system or loop shall apply, Maximum Averare Planar LEGR. LEGR. and MCPR The MAPLHGR, LHGR, and MCPR shall be checked daily to determine if fuel burnup, or control rod movement has caused changes in power Since changes due to burnup are slov, and only a few distribution.

control rods are moved dcily, a daily check of power distribution is adequate.

Amendment No. 159, 169 3.5/4.5-35 BTN Unit 1

t 1,,.

th" ".I G k

+1

,c4 UNITED STATES

['

NUCLEAR REGULATORY COMMISSION

. q ~ N,f,j/( )t. J

' WASWNGTON, D, C. 20555 '

N..x [i TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-296 BROWNS. FERRY NUCLEAR PLANT, UNIT 3

. AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.140 License No. DPR-68 1.

The Nucleer Regulatory Comission (the Commission) has found that:

A.

The application for amendment by Tennessee Valley Authority (the licensee) dated January 13, 1989, complies with the standards 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; B..

The facility will operate in conformity 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 amenoment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to tl3e 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.

i

t t

. 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. DPR-68 is hereby anended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 140, are hereby~ incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of iss. lance and shall be implemented witnin 60 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Suzanne Black, Assistant Director for Projects TVA Projects Division O'fice of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: August 2, 1989

[

o I'

ATTACHMENT TO LICENSE AMENDMENT NO.140 FACILITY OPERATING LICENSE NO. DPR-68

. DOCKET NO. 50-296 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages.

The revised pages are identified by the captioned anendment number and contain marginal lines indicating the. area of change. Table of Contents and overleaf pages*

are provided to maintain document completeness.

REMOVL INSERT i

i ii ii iii 111 iv iv v

v vi.

vi vii vii viii viii 3.5/4.5-1 3.5/4.5-1*

-3.5/4.5-2 3.5/4.5-2 l

L 3.5/4.5-3

3. 5/4. 5-3

3.5/4.5-4 3.5/4.5-4 3.5/4.5-5 3.5/4.5-5 3.5/4.5-6

.3.5/4.5-6 3.5/4.5-7 3.5/4.5-7 3.5/4.5-8 3.5/4.5-8 3.5/4.5-9 3.5/4.5-9 3.5/4.5-10 3.5/4.5-10 3.5/4.5-11 3.5/4.5-11*

3.5/4.5-11a 3.5/4.5-12 3.5/4.5-12 3.5/4.5-13 3.5/4.5-13*

3.5/4.5-14 3.5/4.5-14 3.5/4.5-15 3.5/4.5-15 3.5/4.5-16 3.5/4.5-16 3.5/4.5-17 3.5/4.5-17 3.5/4.5-27 3.5/4.5-27 3.5/4.5-28 3.5/4.5-28 3.5/4.5-29 3.5/4.5-29*

3.5/4.5-30 3.5/4.5-30 3.5/4.5-31 3.5/4.5-31 3.5/4.5-32 3.5/4.5-32*

.3.5/4.5-33 3.5/4.5-33 3.5/4.5-34 3.5/4.5-34*

3.5/4.5-35 3.5/4.5-35*

3.5/4.5-36 3.5/4.5-36

o 5

TABLE OF CONTENTS Erftien Page No.

1.0 Definitions 1.0-1 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 1.1/2.1 Fuel Cladding Integrity.

1.1/2.1-1 1.2/2.2 Reactor Coolant System Integrity.

1.2/2.2-1 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3.1/4.1 Reactor Protection System.

3.1/4.1-1 3.2/4.2 Protective Instrumentation.

3.2/4.2-1 A.

Primary Containment and Reactor Building Isolation Functions....

3.2/4.2-1 B.

Core and Containment Cooling Systems -

Initiation and Centrol.

3.2/4.2-1 C.

Control Rod Block Actuation.

3.2/4.2-2 D.

Radioactive Liquid Effluent Monitoring Instrumentation.

3.2/4.2-3 E.

Drywell Leak Detection.

3.2/4.2-4 F.

Surveillance Instrumentation.

3.2/4.2-4 G.

Control Room Isolation.

3.2/4.2-4 H.

Flood Protection...

3.2/4.2-4 I.

Meteorological Monitoring Instrumentation.

3.2/4.2-4 J.

Seismic Monitoring Instrumentation.

3.2/4.2-5

)

K.

Radioactive Gaseous Effluent Monitoring I

Instrumentation 3.2/4.2-6 L.

ATWS-Recirculation Pump Trip.

3.2/4.2-6a l

l 3.3/4.3 Ecactivity Control.

3.3/4.3-1 A.

Reactivity Limitations.

3.3/4.3-1 B.

Control Rod.s 3.3/4.3-5 l

C.

Scram Insertion Times.

3.3/4.3-10 i

Amendment N.. 104, 135 Unit 3 I

e.

u Sectien Page go.

D.

Reactivity Anomalies.

3.3/4.3-11 k

E.

Reactivity Control.

3.3/4.3-12 F.

Scram Discharge Volume.

3.3/4.3-12 3.4/4.4 Standby Liquid Control System.

3.4/4.4-1 l

A.

Normal System Availability.

3.4/4.4-1 B.

Operation with Inoperable Corponents.

3.4/4.4-3 C.

Sodium Pentaborate Solution.

3.4/4.4-3 3.5/4.5 Core and Containment Cooling Systems.

3.5/4.5-1 A.

Core Spray System (CSS).

3.5/4.5-1 B.

Residual Heat Removal System (RHRS)

(LPCI and Containment Cooling).

3.5/4.5-4 C.

RHR Service Water and Emergency l

Equipment Cooling Water Systems (EECWS).

3.5/4.5-9 D.

Equipment Area Coolers.

3.5/4.5-13 E.

High Pressure Coolant Iniection System (HPCIS).

3.5/4.5-13 F.

Reactor Core Isolation Cooling System (RCICS).

3.5/4.5-14 G.

Automatic Depressurization System (ADS).

3.5/4.5-16 H.

Maintenance of Filled Discharge Pipe.

3.5/4.5-17 I.

Average Planar Linear Heat Generation Rate.

3.5/4.5-18 J.

Linear Heat Generation Rate (LHGR).

3.5/4.5-10 K.

Minimum Critical Power Ratio (MCPR).

3.5/4.5-19 L.

APRM Setpoints..

3.5/4.5-20 3.6/4.6 Primary System Boundary.

3.6/4.6-1 A.

Thermal and Pressurization Limitations 3.6/4.6-1 B.

Coolant Chemistry.

3.6/4.6-5 C.

Coolant Leakage.

3.6/4.6-9 D.

Relief Valves.

3.6/4.6-10 E.

Jet Pumps.

3.6/4.6-11 11 BFN Unit 3 Amendment Nos. 104,118,135,140

l

.9 Ser i;L Pace No.

F.

Recirculation Pump Operation.

3.6/4.6-12 G.

Structural Integrity.

3.6/4.6-13 H.

Snubbers.

3.6/4.6-15 3.7/4.7 Containment Systems.

3.7/4.7-1 A.

Primary Containment.

3.7/4.7-1 i

B.

Standby Gas Treatment System.

3.7/4.7-13 C.

Secondary Containment.

3.7/4.7-16 i

D.

Primary Containment Isolation Valves.

3.7/4.7-17 E.

Control Room Emergency Ventilation.

3.7/4.7-19 F.

Primary Containment Purge System.

3.7/4.7-21 G.

Containment Atmosphere Dilution System (CAD) 3.7/4.7-22 E.

Containmer.t Atmosphere Monitoring (CAM)

System E2 Analyzer.

3.7/4.7-23 3.8/4.8 Radioactive Materials.

3.8/4.8-1 A.

Liquid Effluents.

3.8/4.8-1 B.

Airborne Effluents.

3.8/4.8-2 C.

Radioactive Effluents - Dose.

3.8/4.8-6 D.

Mechanical Vacuum Pump.

3.8/4.5-6 E.

Miscellaneous Radioactive Materials Sources 3.8/4.8-7 F.

Solid Radwaste.

3.8/4.8-9 3.9/4.9 Auxiliary Electrical System.

3.9/4.9-1 l

l l

A.

Auxiliary Electrical Equipment 3.9/4.9-1 B.

Operation with Inoperable Equipment.

3.9/4.9-8 C.

Operation in Cold Shutdown.

3.9/4.9-14 1

3.10/4.10 Core Alterations.

3.10/4.10'1 A.

Refueling Interlocks.

3.10/4.10-1

)

i B.

Core Monitoring.

3.10/4.10-4 1

C.

Spent Fuel Pool Water.

3.10/4.10-7 l

iii Amendment No. 103, 134, 140 BFN Unit 3 L____________________.---

i

v.

~

Fection Eate No.

D.

. Reactor Building Crane.

3.10/4.10-8 E.

Spent Fuel Cask.

3.10/4.10-9 F.

Spent Fuel Cask Handling-Refueling Floor.....

3.10/4.10-9 l

3.11/4.11 Fire Protection Systems 3.11/4.11-1 A.

Fire Detection Instrumentation.

3.11/4.11-1 B.

Fire Pumps and Water Distribution Mains 3.11/4.11-2 C.

Spray and/or Sprinkler Systems.

3.11/4.11-7 D.

CO2 System.

3.11/4.11-8 l

l E.

Fire Hose Stations.

3.11/4.11-9 F.

Yard Fire Hydrants and Hose Houses.

3.11/4.11-11 G.

Fire-Rated Assemblies 3.11/4.11-12 l

H.

Open Flames, Welding and Burning in the Cable.

3.11/4.11-13 Spreading Room 5.0 Major Design Features.

5.0-1 l

5.1 Site Features.

5.0-1 5.2 Reactor.

5.0-1 5.3 Reactor Vessel.

5.0-1 5.4 Containment.

5.0-1 5.5 Fuel Storage.

5.0-1 5.6 Seismic Design.

5.0-2 iv BFN Unit 3 Amendment Nos. 109,133,140 i

.= *-

' ADMINISTRATIVE CONTROLS SECTION PAGE M

RESPONSIBILITY......................................

ORGANIZATION........................................

6.0-1 t_d 6.0-1 6.2.1-Offsite end Onsite Organizations.....................

Plant Staff..........................................

6.0-1 6.2.2 6.0-2 M

. PLANT STAFF 0 QUALIFICATIONS...............................

6.0-5 id TRAINING................................................

6 0-5 M

PL A NT REVI EW A ND AUD IT........................

6 0 5

.6.5.1 Plant Operations Review Committee (P0RC)................

6 0-5 6.5.2 Nuclear Safety Review Board (NSRB)......................

. 6.0-11 6.5.3 Technical Review and Approval of Procedures.............

6 0-17 M

REPORTABLE EVENT ACTIONS................................

6 0-18 M

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

6 0-19 fad PROCEDURES / INSTRUCTIONS AND PROGRAMS....................

6 0-20 6.8.1 Procedures.....................................

Dri11s.........................................

6.0-20 6.8.2 6.0-21 6.8.3 Radiation Control Procedures............................

6.0-22 6.8.4 Quality Assurance Procedures - Effluent and Environmental Monitoring.............................

6.0-23 M

REPORTING RE0UIREMENTS..................................

6 0-24 6.9.1 Routine Reports.........................................

6 0-24 Startup Reports.........................................

6 0-24 Annual Operating Report..................................

6.0-25 Monthly Operating Report................................. 6.0-26 Reportable Events......................................

6 0-26 Radioactive Effluent Release Report......................

6.0-26 Source Tests............................................

6 0-26 6.9.2 Special Reports.........................................

6 0-27 6.10 STATION OPERATING RECORDS ANDRETENTION.................

6 0-29 6.11 PROCESS CONTROL PR0 GRAM.................................

6 0-32 6.12 0FFSITE DOSE CALCULATION MANUAL..

6.0-32 6.13 RA D I OLOGI CAL EFFLUENT MANUAL.....................

6. 0-3 3 l'

BFN y

Amendment No. 109, 140 Onit 3

o.

LIST OF TABLES Table Title Page No.

1.1 Surveillance Frequency Notation....

1.0-13

'3.1.A Reactor Protection System (SCRAM)

Instrumentation Requirements.

3.1/4.1-2 4.1.A Reactor Protection System (SCRAM) Instrumentation Functional Tests Minimum Functional Test Frequencies for Safety Instr. and Control Circuits.

3.1/4.1-7 4.1.B Reactor Protection System (SCRAM) Instrumentation Calibration Minimum Calibration Frequencies for Reactor Protection Instrument Channels.

3.1/4.1-10 3.2.A Primary Containment and Reactor Building Isolation Instrumentation.

3.2/4.2-7 3.2.B Instrumentation that Initiates or Controls the Core and Containment Cooling Systems.

3.2/4.2-14 3.2.C Instrumentation that Initiates Rod Blocks.

3.2/4.2-24 3.2.D Radioactive Liquid Effluent Monitoring Instrumentation.

3.2/4.2-27 3.2.E Instrumentation that Mcnitors Leakage Into Drywell.

3.2/4.2-29 3.2.F Surveillance Instrumentation..

3.2/4.2-30 3.2.G Control Room Isolation Instrumentation.

3.2/4.2-33 3.2.H Flood Protection Instrumentation.....

3.2/4.2-34 3.2.I Meteorological MonitoriL3 Instrumentation.

3.2/4.2-35 3.2.J Seismic Monitoring Instrumentation.

3.2/4.2-36 3.2.K Radioactive Gaseous Effluent Monitoring Instrumentation.

3.2/4.2-37 3.2.L ATWS-Recirculation Pump Trip (RPT)

Surveillance Instrumentation.

3.2/4.2-38a 4.2.A Surveillance Requirements for Primary Containment and Reactor Building Isolation Instrumentation.

3.2/4.2-39 4.2.B Surveillance Requirements for Instrumentation that Initiate or Control the CSCS.

3.2/4.2-43 4.2.C Surveillance Requirements i't Instrumentation that Initiate Rod Blocks.

3.2/4.2-49 4.2.D Radios tive Liquid Effluent Monitoring Instrumentation Surveillance Requirements.

3.2/4.2-50 vi Amendment No. 103, 135,140 BPN Unit 3 L______-________-________--_____-

g LIST OF TABLES (Cont *4.)

Table Title Page No.

4.2.E Minimum Test and Calibration Frequency for Drywell Leak Detection Instrumentation.....

3.2/4.2-52 4.2.T Minimum Test and Calibration Frequency for l

Surveillance Instrumentation..

4 3.2/4.2-53 4,2.G Surveillance Requirements for Control Room Isolation Instrumentation............

3.2/4.2-55 4.2.H Minimum Test and Calibration Frequency for Flood Protection Instrumentation........

3.2/4.2-56 4.2.J Seismic Monitoring Instrument Surveillance Requirements....

3.2/4.2-57 4.2.K Radioactive Gaseous Effluent Monitoring Instrumentation...

3.2/4.2-61 4.2.L ATWS-Recirculation Pump Trip (RPT)

Instrumentation Surveillance.

3.2/4.2-62a 3.5-;

Minimum RERSW and EECW Pump Assignment.

3.5/4.5-11 3.5.I MAPLHGR Versus Average Planar Exposure.

3.5/4.5-21 3.7.A Primary Containment Isolation Valves.

3.7/4.7-24 3.7.B Testable Penetrations with Double 0-Ring Seals.

3.7/4.7-31 3.7.C Testable Penetrations with Testable Bellows..

3.7/4.7-32 3.7.D Air Tested Isoittien Valves..

3.7/4.7-33 3.7.E Primary Containment Isolation Valves which Terminate below the Suppression Pool Water Level.

3.7/4.7-36 3.7.F Primary Containment Isolation Valves Located in Water Sealed Seismic Class 1 Lines.

3.7/4.7-37 3.7.H

)

Testable Electrical Penetrations.

3.7/4.7-38 4.9.A Auxiliary Electrical System.

3.9/4.9-15 j

4.9.A.4.C Voltage Relay Setpoints/ Diesel Generator Start.

3.9/4.9-17 3.11.A Fire Detection Instrumentation...

3.11/4.11-14 3.11.B Spray / Sprinkler Systems.

3.11/4.11-18 3.11.C Hose Stations............

3.11/4.11-20

)

3.11.D Yard Fire Hydrants and Fire Hose Houses.

3.11/4.11-22 6.2.A Minimum Shift Crew Requirements.

6.0-4 ETH vii Unit 3 Amendment Nos. 124,135,140

LIST OF TLLUSTRATIOWS Figure Title Page No.

2.1.1 APRM Flow Reference Scram and APRM Rod Block Settings.

1.1/2.1-6 2.1-2 APRM Flow Bias Scram Vs. Reactor Core Flow.

1.1/2.1-7 4.1-1 Graphical Aid in the Selection of an Adequate Interval Between Tests.

3.3/4.1-12 i

4.2-1 System Unavailability.

3.2/4.2-63 3.5.K-1 MCPR Limits.

3.5/4.5-25 1

3.5.2 Kf Factor.

3.5/4.5-26 3.6-1 Minimum Temperature CF Above Change in Transient Temperature..

3.6/4.6-24 3.6-2 Change in Charpy V Transition Temperature Vs.

Neutron Exposure.

3.6/4.6-25 4.8.1.a Gaseous Release Points and Elevation.

3.8/4.8-10 4.8.1.b Land Site Boundary.

3.8/4.8-11 viii Amendment tio. 140 BFN Unit 3

'.i '

e's t

3.5/4.5-CORE AND' CONTAINMENT COOLING' SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS i

1 3.5 CORE AND CONTAINMENT COOLING 4.5 C.QEE AND CONTAINMENT COOLING l

SYSTEMS SYSTEME l

'Aeolicability Aeolicability Applies to the operational Applies.to.the surveillance-status of the core and requirements of.the core and containment cooling systems.

conte.inment cooling systems when the corresponding limiting condi-tion for operation is in effect.

Obiective Obiective To assure the OPERABILITY of To verify the OPERABILITY of the the core and containment cooling core and containment cooling systems under all, conditions for systems under all conditions for which this cooling capability is which this cooling capability is an essential response to plant an essential response to plant abnormalities.

abnormalities.

Ep_ edification Specification A.

Core Soray System (CSS)

A.

Core Sorav System (CSS) 1.

The CSS shall be OPERABLE:

1.

Core Snray System Testing.

(1) PRIOR TO STARTUP Ltfm Frecuency from a COLD CONDITION, or a.

Simulated Once/

Automatic Operating (2) when there is irradiated Actuation Cycle fuel in the vessel test and when the reactor vessel pressure b.

Pump OPERA-Per Specif1-is greater than BILITY cation 1.0.MM atmospheric pressure, except as specified c.

Motor Per Specifi-in Specification Operated cation 1.0.MM 3.5.A.2.

Valve 0?ERABILITY d.

System flow Once/3 rate: Each months loop shall deliver at least 6250 gpm against a system he,nd corres-ponding to a BFN 3.5/4.5-1 Unit 3 Amendment No. 130

gn l

CORE AND CONTAINMENT COOLING SYSTEMS 3.5/4.5 SURVEILLANCE REQUIREMENTS LIMITING' CONDITIONS:FOR OPERATION 4.5.A Core Sorav System (CSS) 3.5.A Core Sprav System (CSS)

I 4.5.A.I.d (Cont'd).

105 psi differential l'

pretsure between the reactor vessel and the primary containment.

e.

Testable Per Check _ Valve. Specification 1.0.MM f.

Verify that Once/ Month 2.

If one CSS loop.is inoperable, each valve the. reactor may remain in (manual, power-

+

operation for-a period not to operated,' or exceed 7 days'providing

--automatic) in the all active components in injection flowpath the other CSS loop and the that is not locked, RHR system (LPCI mode) sealed, or'other-and the diesel generators wise secured in are OPERABLE.

position, is in its correct

position.

2.

No additional surveillance

.3.

If Specification 3.5.A.1 or is required.

Specification 3.5.A.2 cannot be met, the reactor shall be placed in the COLD SHUTDOWN CONDITION within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

4.

When the reactor vessel pressure is atmospheric and irradiated fuel is in the reactor-vessel at least one core spray loop with one Except that an automatic OPERABLE pump and associated valve cepable of automatic diesel generator shall be return to its ECCS position OPERABLE, except with the when an ECCS signal is reactor vessel head removed present may be in a as specified in 3.5.A.5 or position for another mode PRIOR TO STARTUP as of operation.

specified in 3.5.A.l.

Amendment No. 124, 130, 140 3.5/4.5-2 BPN Unit 3 2

f 3.5/4.5 CORE AND COTTTAINMENT COOLING SYSTEMS SURVEILLANCE REQUIREMENTS LIMITINO CONDITIONS FOR OPERATION 3.5.A Core Spray System (CSS 1 When irradiated fuel is in 1

5.

the reactor vessel and the reactor vessel head is I

removed, core spray is not required to be OPERABLE provided the cavity is flooded, the fuel pool gatcc are open and the fuel pool water level is maintained above the low level alarm point, and provided one RHRSW pump and associated valves supplying the standby coolant sur'.y are OPERABLE.

When work is in progress which has the potential to drain the vessel, manual initiation capability of either 1 CSS Loop or 1 RHR pump, with the capability of injecting water into tha reactor vessel, and the associated diesel generator (s) are required, Amendmant No. 132 3.5/4.5-3 BFN Unit 3

'3;5/4.5 -CORE AND CONTAINMENT COOLING SYSTEMS g,

k LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS

~3.5.B Residual Heat Removal System ~

4.5.B.' Residual Heat Remvval System (RHRS) (LPCI and Containment

'(RHRS) (LPCI and Containment-Cooling)

Cooling) 1.

.The RHRS.shall be OPERABLE:

1. a.

-Simulated Once/

Automatic Operating L

(1)- PRIOR TO STARTUP Actuation Cycle from a COLD

. Test CONDITION; or (2) when there is-b.

Pump OPERA-Per irradiated-fuel in BILITY.

Specification

.the reactor' vessel 1.0.MM and when the reactor vessel pressurc is c.

Motor-Opera-Per greater than ted' valve Specification; atmospheric, except as OPERABILITY. 1.0,MM specified in Specifications.3.5 B.2, d.

Pump Flow Once/3 through 3.5.B.7.

Rate.

months-e.

. Testable Per Check-Specification-Valve.

1.0.MM f.

Verify that Once/ Month each valve (manual, power-operated, or automatic) in the injection flow-path that is not locked, sealed, or otherwise se ured in posi-tion, is in its correct

position.

g.

. Verify LPCI Once/ Month cubsystem cross-tie valve is closed and power removed from valve operator.

Except that an automatic valve capable of auto-matic return to its ECCS position when an ECCS signal is present may be in a position for another mode of operation.

BFN 3.5/4.5-4 Unit 3 Amendment Nos. 130,140

M M CORE AND COE AINMENT COOLING SYSTEMS SURVEILLANCE REQUIREMENT LIMITING CONDITIONS FOR OPERATION 4.5.B. Residual Heat Removal System Residual Heat Removal System (RHRS) (LPCI and Containment 3.5.B l

(RERS) (LPCI and Containment Cooling)

Cooling) i 4.5.B.1 (cont'd)

Each LPCI pump shall deliver With the reactor vessel 9000 gpm against an indicated 2.

pressure less than 105 psig, system pressure of 125 psig, the RHRS may be removed Two LPCI pumps in the same from service (except that loop shall deliver 12000 gpm two cooling RHR pumps-containment against an indicated system mode and associated heat pressure of 250 psig.

exchangers must remain OPERABLE) for e period not An air test on the drywell 2.

to exceed 24 bours while and torus headers and norzles being drained of shall be conducted once/5 suppression chamber quality A water test may be years.

water and filled with performed on the torus header primary coolant qua ity in lieu of the air test.

water provided that during cooldown two loops with one pump per loop or one loop with two pumps, and associated diesel generators, in the core spray system are OPERABLE.

No additional surveillance 3.

If one RHR pump (LPCI mode) required.

3.

is inoperable, the reactor may remain in operation for a period not to exceed 7 days provided the remaining RHR pumps (LPCI mode) and both access paths of the RERS (LPCI mode) and the CSS and the diesel generators remain OPERABLE.

4.

No additional surveillance If any 2 RHR pumps (LPCI required.

4.

the mode) become inoperable, reactor shall be placed in the COLD SHUTDOWN CONDITION within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

3.5/4.5-5 Amendment No. 124, 140

)

I BFN

)

Unit 3 I

5 4.5 CORE AND CONTAINMENT-COOLING SYSTEMS LIMITING CONDITION FOR OPERATIOI:

SURVEILLANCE' REQUIREMENTS 3.5 P Residual' Heat Removal System 4.5 B.

(RHRS) (LPCI and Containment-Residual Heat Removal Syst1m Cooling)

IRHRS) (LPCI and Containment Cooling)

.5.

If one RHR pump (containment 5.

cooling mode) or associated No additional surveillance heat exchanger is inoperable, required.

the reactor may remain in operation for a period not to 1'

exceed 30 days provided the remaining RHR pumps (containment cooling mode) and associated heat exchargers and diesel generators and all access paths of the RHRS

.(containment cooling mode) are OPERABLE.

6.

If two RHR pumps (containment 6.

cooling mode) or associated No additional surveillance heat exchangers are required.

inoperable, the reactor may remain in operation.for a period not to exceed 7 days provided the remaining RHR pumps (containment cooling mode), the associated heat exchangers, diesel generators, and all access paths of the RHRS (containment coo 31ng mode) are OPERABLE.

7.

If two_ access paths of the 7.

RHRS (containment cooling No additional surveillance mode) for each phase of the required.

mode (drywell sprays, suppression chamber sprays, and suppression pool cooling) are not OPERABLE, the unit may remain in operation for a period'not to exceed 7 days provided at least one path for each phase of the mode remains OPERABLE.

3rn Unit 3 3.5/4.5-6 Amendment No. 124, 140

____~~N-~"

~~

z.c 3.5/4.5' CORE AND CONTAINMENT COOLING SYSTEMS-

. LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS.

l 3.5.B Residual' Heat Removal System 4.5.B Residual Heat Removal System (FHRS) (LPCI and Containment (RHRS) (LPCI and Containment Cooling)

Cooling)

8. If Specifications 3.5.B.1 8.

No additional surveillance through 3.5.B.7 are not met, required.

j an orderly. shutdown shall be initiated'and the reactor' shall be placed in the

-COLD SHUTDOWN CONDITION within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

9.

When the reactor vessel 9.

When the reactor vessel pressure is atmospheric and pressure is atmospheric, irradiated fuel is'in the the RHR pumps and valves reactor vessel, at least one RHR that are required to be loop with two pumps or two loops OPERABLE shall be with one pump per loop shall demonstrated to be OPERABLE be OPERABLE. The pumps' per Specification 1.0.MM.

associated diesel generators must also be OPERABLE.

10. If the conditions of-10.

No additional surveillance-Specification 3.5.A.5 are met, required.

LPCI and containment. cooling are not required.

11. When there is irradiated fuel 11.

The B and D RHR pumps on in the reactor and the reactor unit 2 which supply vessel pressure is greater than cross-connect capability atmospheric, 2 RHR pumps and shall be demonstrated to be associated heat exchangers and OPERABLE per Specification valves on an adjacent unit must 1.0.MM when the cross-be OPERABLE and capable of connect capability supplying cross-connect is required.

capability except as specified in Specification 3.5.B.12 below.

(Note: Because cross-connect capability is not a short-term requirement, a component is not l

I considered inoperable if cross-connect capability can be-restored to service within 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months .)

BFN 3.5/4.5-7 Amandment No. 130, 140 Unit 3 I

3.5/4.5 CORE AWD CONTAINMENT _000LIWG' SYSTEMS LIMITING CDNDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.B Residual Heat ~ Removal System 4.5.B Residual Heat Removal System (RHRS) (LPCI and Containment (RHRS) (LPCI and Containment Cocling)

Cooling)

12. If one RHR pump or associated

12. No additional surveillance heat exchanger located required.

on the unit cross-connection in unit 2 is inoperable for any reason (including (valve inoperability, pipe break, etc.), the reactor may remain in operation fer a period not to exceed 30 days provided the remaining RHR pump ~and associated diesel generator are OPERABLE.

13.

If'RHR cross-connection flow or 13.

No additional surveillance heat removal capability is lost, required.

the unit may remain in operation for a period net to exceed 10 days unless such capability is restored.

14 All recirculation pump 14.

All recirculation pump discharge valves shall discharge valves shall be OPERABLE PRIOR T0 be tested for OPERABILITY STARTUP (or closed if during any period of permitted elsewhere COLD SEUTDOWN CONDITION in these specifications).

exceeding 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , if OPERABILITY tests have not been performed during the preceding 31 days.

BFN 3.5/4.5-8 Unit 3 Amendment Nos. 124,140

_____________________________________o

os 1 5/4.3 CORE AND CONTAIWMEWT COOLIWG SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.C RHR Service Water and Emergency 4.5.C RHR Service Water and Emergency Eauipment' Cooling Water Systems Eaulement Coolinz Water Systema I

(EECWS)

(EECWS) 1.

~ PRIOR TO STARTUP from 1.

a.

Each of the RHRSW pumps-a COLD CONDITION, 9 RHRSW normally assigned to pumps must be OPERABLE, with automatic service on 7 Instps- (including pump B1 the EECW' headers will or B2) assigned to RHRSW be tested service and 2 automatically automatically each time starting pumps assigned to the diesel generators EECW service, are tested. Each of the RHRSW pumps and all associated essential control valves for the EECW headers and RHR heat exchanger headers shall be demonstrated to be OPERABLE in accordance with Specification 1.0.MM.

b.

Annually each RHRSW pump shall be flow-rate tested. To be considered OPERABLE, each pump shall pump at least 4500 gpm through its normally assigned flow

path, c.

Monthly verify that each valve (manual, power-operated, or automatic) in the flowpath servicing safety-reitted equipment in the affected unit that is not locked, sealed, or otherwise secured in position, is in its j

correct position.

l

)

BFN 3.5/4.5-9 Amendment No. 130, 140 Unit 3 l

t l!

s<

u 3.5I4.5 CORE'AND CONTAINMENT'C00 LING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS

3. 5. C-RHR Service Water and Emerzency 4.5.C RHR Service Water and Emernency Eguipment Cooling Water Systems Eculoment Coolina Water Systems (EECWS) (Continued)

(EECWS)-(Continued) 2.

During' REACTOR POWER-

2. No additional surveillance OPERATION, RHRSW pumps is required.

must be OPERABLE und assigned to service as indicated in Table 3.5-1 l

for the specified time limits.

3.

During REACTOR POWER

'3. Routine surveillance for L

-OPERATION, both RHRSW' these pumps is specified pumps B1 and B2 normally in 4.5.C.l.

or alternately assigned to the RHR heat exchanger header supplying the standby coolant supply connection r.ust be OPERABLE except as specified in 3.5.C.4 and 3.5.C.5 below.

1 l

I l

BFN 3.5/4.5-10 Amendment No. 124, 130, 140 Unit 3 J

1

.1

e, -

o. s TABLE 3.5-1 Time Minimum Limit Service Assignment (Days)-

RHRSW EECW{2) _,

(4)

(1)

Indefinite 7

3 (3)(4)

(1)

(3) 30 7

or 6 2

or 3 (4)

(1) 7 6

2 (1)

At least one OPERABLE pump must be assigned to each header.

(2)-

Only automatically starting pumps may be assigned to EECW header service.

(3)

Nine pu'r.ps must be OPERABLE. Either configuration is acceptable: 7 and 2 or 6 and 3.

(4)

Requirements may be reduced by two for each unit with fuel

unloaded, l

s 1

l l

EFH 3.5/4.5-11 l

Unit 3 1

i l

i

i'

.THIS PAGE 1NTENIIONALLY LEFT BLANK

~

i BFN 3.5/4.5-11a Amendment fjo. 140 Unit 3

r.

s j

_3.5/4.5 CORE AND CONTAINMEIR COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.C RHR Service Water and Emergency 4.5.C RHR Service Water and Emergency Eauirment Cooling Water Systems Eauipment Coolina Water Systems (EECWS) (Continued)

(EECWS) (Continued) 4 One of the B1 or B2 RHRSW

4. No additional surveillance pumps assigned to the RHR is required.

heat exchanger supplying the standby coolant supply connection may be inoperable for a period net to exceed 30 days provided the OPERABLE pump is aligned to supply the RHR heat exchanger header and the associated diesel generator and essential control valves are OPERABLE.

5.

The standby coolant supply capability may be inoperable for a period not to exceed 10 days.

6.

If Specifications 3.5.C.2 through 3.5.C.5 are not I

met, an orderly shutdown shall be initiated and the unit placed in the COLD SHUTDOWN CONDITION within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

7.

There shall be at least 2 RHESW pumps, associated with the selected RER pumps, aligned for RHR heat exchanger service for each reactor vessel containing irradiated tuel.

I BrN 3.5/4.5-12 Amendment No. 124, 130, 140 Unit 3

]

3. 5 / t.. ?

CORE AND CONTAINMENT COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEI5 LANCE REQUIREMENTS 3.5.D Equioment Area Coolers 4.5.D Eculoment AreaCoojers 1.

The equipment area cooler

1. Each equipment area cooler associated with each RHR is operated in conjunction pump and the equipment with the equipment served area eccler associated by that particular cooler; with em.n set of core therefore, the equipment spray pumps (A and C area coolers are tested at or B and D) must be the same frequency as the OPERABLE at all times pumps which they serve, when the pump or pumps served by that specific cooler is considered to be OPERABLE.

2.

When an equipment area cooler is not OPERABLE, I

the pump (s) served by that cooler must be considered inoperable for Technical Specification purposes.

E.

Hich Pressure Coolant Iniection E. Mich Pressure Coolant System (HPCISl Iniection System (HPCIS1 1.

The HPCI system shall be 1.

HPCI Subsystem testing OPERABLE:

shall be performed as follows:

(1) PRIOR TO STARTUP from a a.

Simulated Once/

COLD CONDITION; or Automatic operating Actuation cyc1t Test (2) whenever there is b.

Pump Per irradiated fuel in the OPERA-Specification reactor vessel and the BILITY 1.0.MM reactor vessel pressure is greater than 122 psig, c.

Motor Oper-Per except as specified in ated Valve Specification Specification 3.5.E.2.

OPERABILITY 1.0.MM d.

Flow Rate at' Once/3 normal months reactor vessel operating pressure BTN 3.5/4.5-13 Amendment No. 130 Unit 3 i

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

s.

1&d_,,_ CORE AND CONTAINMENT COOLI?iG SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS

? !.E Hjrh Pressure Coolant Iniection System (HPCIS) 4.5.E High Pressure Coolant Iniection System (HPCIS) 4.5.E.1 (Cont'd) e.

Flow Rate at Once/

150 psig operating cycle The HPCI pump shall deliver at least 5000 gpm during each flow rate test, f.

Verify that once/ Month each valve (manual, power-operated, or automatic) in the injection flow-path that le not locked, sealed, or otherwise secured in position, is in its correct

position.

2.

If the HPCI system is inoperable, the reactor may

2. No additional surveillance remain in operation for a are required.

I period not to exceed 7 days, provided the ADS, CSS, RERS (LPCI), and RCICS are OPERABLE.

3.

If Specifications 3.5.E.1 or 3.5.E.2 are not met, Except that an automatic an orderly shutdown shall valve capable of automatic be initiated and the return to its ECCS position reactor vessel pressure when an ECCS signal is shall be reduced to 122 present may be in a psig or less within 24 position for another mode of hours.

operation.

F.

Reactor Core Isolatien Cooling F.

System (RCICS)

Reactor Core Isolation CooJing System (RCICS) 1.

The RCICS shaII be OPERABLE:

1. RCIC Subsystem testing shall (1) TRIOR TO STARTUP from a be performed as follows:

COLD CONDITION; or

a. Simulated Auto-Once/

matic Actuation operating Test cycle ETN Unit 3 3.5/4.5-14 Amendment fio. 130, 140

ea

'3.5/4.5 CORE AND CONTAINMENT COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS-3.5.T.

Reactor Core Isolation Coolina 4.5.F Reactor Core Isolation Coolinr.

Syst em (RCICS)

System (RCIC3) 3.5.F.I (Cont'd) 4.5.F.1 (Cont'd)

(2) whenever there is

b. Pump Per irradiated fuel in the OPERABILITY Specifi-reactor vessel and the cation reactor vessel pressure 1.0.MM.

is above 122 psig, except as specified in

c. Motor-Operated Per 3.5.F.2.

Valve Spr.ifi-OPERABILITY cation 1.0.MM d.

Flow Rate at Once/3 normal reactor months vessel operating pressure e.

Flow Rate at Once/

150 psig operating cycle The RCIC pump shall deliver at least 600 gpm during each flow test.

2.

If the RCICS is inoperable, f.

Verify that Once/ Month the reactor may remain in each valve operation for a period not (manual, power-to exceed 7 days if the operated, or HPCIS is OPERABLE during automatic) in the such time, injection flowpath that is not locked, 3.

If Specifications 3.5.F.1 sealed, or other-or 3.5.F.2 are not met, an vise secured in orderly shutdown shall be position, is in its initiated and the reactor correct

position.

shall be depressurized to less than 105 psig within

2. No additional surveillance 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , are required.

Except that an automatic valve capable of automatic return to its normal position when a signal is present may be in a position for another mode of operation.

BFN 3.5/4.5-15 Unit 3 Amendment No. 140

3 5 /t. 5 - CORE AWD CONDIWMERT COLLEG_flSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOTTIREMENTS 3.5.G Lylptptic Depressurfzation 4.5.G Automatic Deoressurization Evfter (ADS)

System (ADS) 1.

Four of the six valves of 1.

During each operating the Automatic cycle the following Depressurization System tests shall be performed shall be OPERABLE:

on the ADS.

1) PRIOR TO STARTUP a.

A simulated automat'.c from a COLD CONDITION, actuation test shall or, be performed PRIOR TO STARTUP after each (2) whenever there is refueling outage.

irradiated fuel in the Manual surveillance reactor vessel and the of the relief valves reactor vessel pressure is covered in is greater than 105 psig, 4.6.D.2.

except as specified in 3.5.G.2 and 3.5.G.3 below.

2.

If three of the six ADS 2.

No additional surveillance valves are known to be are required.

incapable of automatic operation, the reactor may remain in operation for a period not to exceed 7 days, provided the HPCI system is OPERABLE. (Note that the pressure relief function of these valves is assured by Section 3.6.D of these specifications and that this i

specification only applies to the ADS function.) If more than three of the six ADS valves are known to be incapable of automatic operation, an immediate orderly shutdovn shall be initiated, with the reactor in a HOT SHUTDOWN CONDITION in 6 houro, and in a COLD l

SHUTDOWN CONDITION in.he following 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months .

l l

3.

If Specifications 3.5.G.1 and 3.5.G.2 cannot be met, an orderly shutdown vill be initiated and the reactor vessel pressure shall be i

reduced to 105 psig or less l

vithin 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l BFN 5/4.5-16 Amendment No. 140 Unit 3 l

l d

i l

I 3,5/5.5 CORE AND CONTAINMENT COQLING SYSTEMS SURVEILLANCE REQUIREMENTS i'

LIMITIEG. CONDITIONS FOR OPERATION I

4.5.H. Maintenance of Filled Discharce-

.3.5.H.

Maintenance of Filled Discharge EipJll EiE!!

The following surveillance Whenever the core spray systems,.

requirements shall be adhered LPCI, HPCI, or RCIC are required to assure that the discharge to be OPERABLE, the discharge piping of the core spray piping from the pump discharge systems, LPCI, HPCI, and RCIC of these systems to the last are filled:

block valve shal1~be filled.

1. Every month and prior to the The suction of the RCIC and HPCI testing of the RHRS (LPCI and pumps shall be aligned to the.

Containment Spray) and core condensate storage tank, and spray system, the discharge the pressure suppression chamber piping of these systems shall head tank shall.normally be aligned be vented from the high point to serve the discharge piping of and water flow determined.

the RHR and CS pumps. The condensate head tank may be used

2. Following any period where the-to serve the RHR and CS discharge LPCI or core spray systems piping if the PSC head tank have not been required to be is unavailable. The pressure OPERABLE, the discharge piping indicators on the discharge of the of the inoperable system shall RHR and CS pumps shall indicate be vented from the high point not less than listed calow, prior to the return of the system to service.

F1-75-20 48 psig P1-75 48 48 psig

3. Whenever the HPCI or RCIC P1-74-51 48 psig system is lined up to take PI-74-65 48 psig suction from the condensate otorage tank, the discharge piping of the HPCI and RCIC shall be vented from the high point of the system and water flow observed on a monthly basis.

4. When the RHRS and the CSS are required to be OPERABLE, the pressure indicators.which monitor the discharge lines shall be monitored daily and the pressure recorded.

3.5/4.5-17 Amendment tio.140 BFN Unit 3

1

{

3.5.

PASES and 3.5.B Residual _Hr.AL__ Removal System (RHRS)

.3;5;A. Core Sorav' System (CSS)

Analyses presented in the FSAR* and analyses presented in conformance stem

-with 10 CFR 50, Appendix K, demonstrated that the core spray sy provides adequate cooling to the core to dissipate the'e l clad try remains temperature to below 2,200*F which assures that core geome intact and to limit the core average clad metal-water reaction to less Core spray distribution has been shown in tests of i ments.

systems similar in_ design to BENP to exceed the min than'1 percent.

l h

ds to half the rated-flow in simulated fuel assemblies with heater ro duplicate the decay heat characteristics of irradiated fuel.

h The RHR3 (LPCI mcde) is designed to provide emergency cooling to t e This core by flooding in the event of a loss-of-coolant eccident.

however, it system is completely independent of the core spray excessive fuel clad temperature.

f core spray system provide adequate cooling for break areas o double-ended approximately 0.2 square feet up to and including there emergency core cooling subsystems.

The intent of the CSS and RHRS specifications is to not allow startup from the cold condition without all associated equipmen OPERABLE.

service for the specified allowable repair times.

based on times have been selected using engineering judgnent experiences and supported by availability analysis.

y Should one core spray loop become inoperable, the remaining core spra loop, the RHR System, and the diesel generators are required to be i

These provide OPERABLE should the need for core cooling arise.

d quate core l

extensive margin over the OPERABLE equipment needed f cooling.

seven days was chosen.

Should one RHR pump (LPCI mode) become inoperable, three RHR pum Since c.dequate (LPCI mode) and the core spray system are available.

day repair period is justified.

Should teo RHR pumps (LPCI mode) become inoperable, there remain Therefore, i

reserve (redundant) capacity within the RHRS (LPCI mode).

d within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

the affected unit shall be placed in cold shut own FSAR.

A detailed functional analysis is given in Section 6 of the BFNP 3.5/4.5-27 Amendment f40, 140 BFN Unit 3 L

~-'----e-

e<

3.5 EASES (Cont'd)

'Should one RHR pump-(containment cooling mode) become inoperable, a complement.of three full capacity containment heat removal systems is still available. Any two of the remaining pumps / heat exchanger l

combinations would provide more than adequate containment cooling for any abnormal or postaccident situation. Because of the availability of equipment in excess of normal redundancy requirements, a 30-day repair.

[

I period is justified.

Should two RHR pumps (containment cooling mode) become inoperable, a full heat removal system is still available. The remaining pump / heat exchanger combinations would provide adequate containment cooling for any abnormal postaccident situation. Because of the availability of a full complement of heat removal equipment, a 7-day repair period is justified.

Observation of the stated requirements for the containment cooling mode assures that the suppre.ssion pool and the-drywell will be sufficiently cooled, following a loss-of-coolant accident, to prevent primary containment overpressurization. The containment cooling function of the RHRS is permitted only after the core has reflooded to the two-thirds core height level.. This prevents inadvertently diverting water needed for core flooding to the less urgent task of containment The two-thirds core height leve1' interlock may be manually cooling.

bypassed by a keylock switch.

Since the RHRS is filled with low quality water during power operation, it is planned that the system be filled with demineralized (condensate) water before using the shutdown cooling function of the RHR System.

Since it is desirable to have the RHRS in service if a " pipe-break" should occur, it is permitted to be out of operation type of accident for only a restricted amount of time and when the system pressure is remain low. 'At least one-half of the containment cooling function must OPERABLE during this time period. Requiring two OPERABLE CSS pumps during cooldown allows for flushing the RHRS even if the shutdown were caused by inability to meet the CSS specifications (3.5.A) on a number of OPERABLE pumps.

When the reactor vessel pressure is atmospheric, the limiting conditions for operation are less restrictive. At atmospheric pressure, the minimum requirement is for one supply of makeup water to Requiring two OPERABLE RHR pumps and one CSS pump provides.

the core.

redundancy to ensure makeup water availability.

Should one RHR pump or associated heat exchanger located on the unit cross-connection in the adjacent unit become inoperable, an equal capability for long-term fluid makeup to the reactor and for cooling of the containment remains OPERABLE. Because of the availability of an equal makeup and cooling capability, a 30-day repair period is justified.

Amendment No. 140 3.5/4.5-28 BFN Unit 3

c- -

m.

3.5 Bases (Cent'd)

[

The suppression chamber can be drained when the reactor vessel pressure is atmospheric, irradiated fuel is in the reactor vessel, and work is in progress.which has the potential to drain the vessel.

By not requiring the fuel pool gate to be open with the vessel head removed, the combined water inventory in the fuel pool, the reactor cavity, and

-the separator / dryer pool, between the fuel pool low level alarm and the

.. actor vessel flange, is about 65,800 cubic feet (492,000 gallons).

This will provide adequate low-pressure cooling in lieu of. CSS and RHR (LPCI and containment cooling mode) as currently required in Specifications 3.5.A.4 and 3.5.B.9.

The additional requirements for j

providing standby coolant supply available will ensure a redundant, supply of coolant supply.

Control rod drive maintenance may continue during this period provided'no more than one drive is removed at a time unless blind flanges are installed during the period of time CRDs are not in place.

Should the capability for providing flow through the cross-connect lines be lost, a 10-day repair time is allowed before shutdown is required. This repair time is justified based on the very small probability for ever needing RHR pumps and heat exchangers to supply an adjacent unit.

REFERENCES 1.

Residual Heat Removal System (BFNP FSAR subsection 4.8) 2.

Core Standby Cooling Systems (BFNP FSAR Section 6) 3.5.C. RHR Service Water System and Emergency Eauipment Cooling Water System (EECWS)

There are two EECW headers (north and south) with four automatic starting RHRSW pumps on each header. All components requiring emergency cooline, water are fed from both headers thus assuring continuity of optration if either header is OPERABLE. Each header alon can handle the flows to all components. Two RHRSW pumps can supply the full flow requirements of all essential EECW loads for any abncrmal or postaccident situation.

There are four RHR heat exchanger headers (A, B, C, & D) with one RHR heat exchanger from each unit on each header. There are two RHR5W l

pumps on each header; one normally assigned to each header (A2, B2, C2, er D2) and one on alternate assignment (A1, B1, C1, or D1). One RHR heat exchanger header can adequately deliver the flow supplied by both RHRSW pumps to any two of the three RHR heat exchangers on the header.

One RHRSW pump can supply the full flow requirement of one RHR heat exchanger. Two RHR heat exchangers can more than adequately handle the cooling requirements of one unit in any abnormal or poetaccident situation.

ETU 3.5/4.5-29 Unit 3

O*

315' BASES (Cont'd)

The RHR Service Water System was designed as a shared system for three The specification, as written, is conservative when consideration is given to particular pumps being out of service and to units.

If unusual operating conditions arise possible valving arrangements.such that more pumps are out of service than a NRC to allow specification, a special case request may be made to thacon assured.

Should one of the two RHRSW pumps normally or alternately assigned to l

the RHR heat exchanger header supplying the standby c makeup to the unit reactor and for cooling of the unit containmen Should the remains OPERABLE.

cooling capability, a 30-day repair period is justified.a 10-day repair capability to' provide standby coolant supply be lost, ding the time is justified based on the low probability for ever neeVerification that the standby coolant supply.

valve is closed and power to its operator is disconr_4cted ensures that each LPCI subsystem remains independent.and a failure of the flow path in one subsystem will not affect the flow path of the other LPCI subsystem.

3.5.D Eguioment Area Coolers There is an equipment area cooler for each RHR pump and an equipment (two pumps, either the A and C or B and D area cooler for each set The equipment area coolers take suction near the cooling air discharge of the motor of the pum pumps) of core spray pumps.

This ensures that cool air is supplied for cooling the pump served.

motors.

The equipment area coolers also remove the pump, and equipment vaste from the basement rooms housing the engineered safeguard The various conditions under which the operation'of the heat equipment air coolers is required have been identified by evaluating equipment.

the the normal and abnormal operating transients and accidents overThe surveillan full range of planned operations.

equipment area coolers in each of their various modes is accomplished This during the testing of the equipment served by these coolers.

testing is adequate to assure the OPERABILITY of the equipment area coolers.

REFERENCES Residual Heat Removal System (BFN FSAR Section 4.8) 1.

Core Standby Cooling System (BTN FSAR subsection 6.7) 2.

3.5/4.5-30 Amendment No. 140 BFN Uu.t 3

.a.

PASES (Cont'd) 3.5 Inicetion Sys_ tem (HPCIS) 3.5.E. High Pressure Coolant is adequately The HPCIS is provided to assure that the reactor corefuel clad tem hich does not result in rapid cooled to limit the nuclear system and loss of coolant w The HPCIS permits the reactor depressurization of the reactor vessel.fficient reactor vessel water level to be shut down while maintaining suinventory until the vessel i CIS continues to operate until reactor vessel pressure is below ti ore cooling.

LPCI operation or core spray system operation mainta ns c i

quired core The capacity of the system is selected to provide th s re Two sources of water are cooling.

is used pressures between 1,120 and 150 psig. Initially, water from the conde l into the reactor.

instead of injecting water from the suppression poo available.

depressurizes more When the HPCI System begins oper.ation, the reactor d due to the

~

rapidly than would occur if HPCI was not initiate the reactor vessel condensation of steam by the cold. fluid pumped intoAs the react i

decrease, the HPCI flow momentarily reaches equilibr umCont by the HPCI system.

begins to rise.

through the break.

decrease below the HPCI flow and the liquid inventory The core never This type of response is typical of the small breaks.hout the transient so uncovers and is continuously cooled throug lie within the capacity core damace of any kind occurs for breaks that range of the HPCI.

There is adequate The minimum required NPSH for HPCI is 21 feet. eleva HPCI pump, such that the required NPSH is available with a su up to 140*r with no containment back pressure.

f feedwater The HPCIS serves as a backup to the RCICS as a source o The ADS serves as a makeup during primary system isolation conditions.

tulated transients and accident. The CSS and RHR d ADS are no longer cooling at low reactor pressure when RCICS anConsidering th i

ime necessary.

of seven days was selected.

ling Systems The HPCI and RCIC as well as all other Core Standby Coo d Condition.

It is must be OPERABLE when starting up from e Colrea t full capacity until reactor pressure exceeds 150 psig and tHPCI t pressure It is the intent d up from a Cold Condition, decreases telow 100 psig.

assure that when the reactor is being starte l

the HPCI is not known to be inoperab e.

Amendment No. 140 3.5/4.5-31 BFN Unit 4 L

3.t MSIS (Cont 'd) 3.5.F Reactor Core Isolation Cooline Svstem (RCICS)

The'various conditions under which the RCICS plays an essential rol providing makeup water to the reactor vessel have been identified by e in evaluating the various plant events over the full range of planned operations.

RCICS was designed will be available when needed.'The spec NPSH for RCIC is 20 feet.

The minimum required L

suppression pool and the RCIC pump, such that the required NP l

available with a suppression pool temperature up to 140*F with no containment back pressure.

Because the low-pressure cooling systems (LFCI and core spray) are nuclear system pressure is below 122 psig, the RCI below this pressure.

provide its design flow, but reduced flow is required for certain events.

RCICS design flow (600 gpm) is sufficient to maintain water level above the top of the active fuel for a complete loss of feedwat flow at design power (105 percent of rated).

er Consideration of the availability of the RCICS reveals that the ave risk associated with failure of the RCICS to cool the core when rage required is not seven days, provided thatincreased if the RCICS is inoperable for no longer than the HPCIS is OPERABLE during this period.

REFERENCI 1.

Reactot Core Isolation Cooling System (BENP FSAR Subsection 4 7) 3.5.G Automatic Deeressurization System (ADSl This specification ensures the OPERABILITY of the ADS under all essential response to station abnormalities. conditions for whic The nuclear system pressure relief system provides automatic nucl system depressurization for small breaks in the nuclear system ear the low-pressure coolant so that can operate to protect the fuel barrier. injection (LPCI) and the core spray subsyst applies caly to the automatic feature of the pressure relief systemNote tha Specification 3.6.D specifies the requirements for the pressure r lt f j

function of the valves.

It e a in possible for any number of the valves assigned to the ADS to be incapable of performing their ADS functions because of instrumentation failures yet be fully capable of perf their pressure re?ief function.

orming BFN i

Unit 3 3.5/4.5-32 k

.3.

3.5, BAIES (Cont *d)

Because.the automatic depressurization system does not provide makeup to the reactor primary vescel, no credit is taken for the' steam cooling of the core caused by the system' actuation to provide further conservatism to the CSCS.

With two ADS valves known to be incapable of automatic operation, four valves remain OPERABLE to perform their ADS function. The ECCS loss-of-coolant accident analyses for small line breaks assumed that four of the six ADS' valves were OPERABLE.

Reactor operation with three ADS valves inoperable is allowed to continue for seven days provided that the HPCI system is'0PERABLE. Operation with more than three of the six ADS valves inoperable is not acceptable.

H.

Maintenance of Filled Discharge Pine If the discharge piping of the core' spray, LPCI, HPCIS, and RCICS are not filled, a water hammer can develop in this piping when the pump and/or pumps are started. To minimize damage to the discharge piping and to ensure added margin in the operation of these systems, this Technical Specification requires the discharge lines to be filled whenever the system is in an OPERABLE condition.

If a discharge pipe is not filled, the pumps that supply that line must be assumed to be inoperable for Technical Specification p::rposes.

The core spray and RHR system discharge piping high point vent is visually checked for water flow once a month and prior to testing co ensure that the lines are filled. The visual checking will avoid starting the core spray or RHR system with a discharge line not filled.

In addition to the visual observation and to ensure a filled discha se line other than prior to testing, a pressure suppression chamber heaa tank is located approximately 20 feet above the discharge line high point to supply makeup water for these systems. The condensate head tank located approximately 100 feet above the discharge high point serves as a backup charging system when the pressure suppression chamber head tank is not in service.

System discharge pressure indicators are used to determine the water level above the discharge line high point. The indicators will reflect approximate 1y'30 psig for a water level at the high point and 45 psig for a water level in the pressure suppression chamber head tank and are monitored daily to ensure that the discharge lines are filled.

When in their normal standby condition, the suction for the HPCI and RCIC pumps are aligned to the condensate storage tank, which is physically at a higher elevation than the HPCIS and RCICS piping. This assures that j

the HPCI and RCIC diccharge piping remains filled. Further asturance is provided by observing water flow from these systema' high points monthly.

~ I.

MLximum Averate Planar Linear Heat GeneratipjLJJkte (MAPLHGR)

This specification assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the limit specified in the 10 CFR 50, Appendix K.

bFh 3.5/4.5-33 Unit 3 Amendment No. 140 L_______________

ee n.

3.5 HASJ;f -(Cont'd)

The peak cladding temperature following a postulated los

)

l accident l

{

all the rods of a fuel assembly at any axial location and is on y j

dependent secondarily on the rod-to-rod power distribution w i

the calculated peak clad temperature by less than i assembly.

20'F relative to the peak temperature for a typical fuel design, the fuel assembly affect limit on the average linear heat generation rate is sufficient to assure dix K limit.

that calculated' temperatures are within the 10 CFR 50 Appen i

The The limiting value for MAPLHGR is shown in Tables 3.5.I-1 through 7.

f 1.

analyses supporting these limiting values are presented in Re erence 1

Generation Rate (LHGR) 3.5.J. Linear Heat the linear heat generation rate in any This specification. assures that l pellet rod is less than the design linear heat generation if fue densification is postulated.

25 percent The LHGR shall be checked daily during reactor operation at 1 d

power to determine if fuel burnup, or control rod movement has causeFo changes in power distribution, rated thermal power, the MTPF would have to be which is precluded by a considerable margin when employing any percent permissible control rod pattern.

3.5.E. Minitur Critical Power Ratio (MCPR) the At core thermal power levels less than or equal to 25 pe moderator void content will be very small.

operating plant rod patterns which may be employed at this point, experienc i

With MCPR value is in excess of requirements by a considerable margin.

any inadvertent core flow increase would only The daily this low void content, place operation in a more conservative mode relative to MCP sufficient since power distribution shifts are very slow when there have requirement The requirement for not been significant power or control rod changes. calculating M ensures that MCPR will be known following a change in power or power l

shape (regardless of magnitude) that could place operation at a therma limit.

3.5.L. APPE Setpoinu for 7x7 fuel and Operation is constrained to a maximum LHGR of 18.5 kW/f t This limit is reached when core 13.4 kW/ft for Ex8, 8x8R, and P8x82.

For the maximum fraction of limiting power density (CMFLPD) equals 1.0.

case where CMFLPD exceeds the fraction of rated therma is pt.rmit ted only at The scram trip scram settings as required by Specification 3.5.L.1.

setting and rod block. trip setting are adjusted to ensur 3.5/4.5-34 BFN Unit 3

ee 3.5 BASES (Cont'd) beyond that allowed by the one-percent plastic strain limit. A six-hour time period to achieve this condition is justified since the additional margin gained by the setdown adjustment is above and beyond that ensured by the safety analysis.

3.5.M Rtrerences 1.

L?ss-of-Coolant Accident Analysis for Browns Ferry Nuclear Plant Unit 3, NEDO-24194A and Addenda.

2.

"BWR Transient Analysis Model Utilizing the RETRAN Program,"

TVA-TR81-01-A.

3.

Generic Reload Fuel Application, Licensing Topical Report, NEDE-24011-P-A and Addenda.

k l

l

)

BFN 3.5/4.5-35 Unit 3 Amendment Nos. 118, 140 f

L_----- -

l j

i Cooline Systems Surveillance Frequencies 4.5 Core end Containment I

The testing interval for the core'and containment cooling systems is based on industry practice, quantitative reliability analysis, judgment The core cooling systems have not been designed to be and practicality.

fully testable during operation. For example, in the case of the HPCI, automatic initiation during power operation would result in pumping cold L

Complete ADS water into the reactor vessel which is not desirable.

testing during power operation causes an undesirable loss-of-coolant To increase.the availability of the core and containment I

inventory.

cooling system, the components which make up the system, i.e.,

The pumps instrumentation, pumps, valves, etc., ate tested frequently.

and motor operated-injection valves are also tested in accordance with A simulated automatic Specification 1.0.MM to assure their OPERABILITY.

actuation test once each cycle combined with testing of the pumps end injection valves in accordance with Specification 1.0.MM is deemed to be Monthly alignment checks of valves adequate testing of these systems. locked or sealed in position which affect the ability of the 1

I that are not systems to perform their intended safety function are also verified to be Valves which automatically reposition themselves in the proper position.

on an initiation signal are permitted to be in a position other than normal to facilitate other operational modes of the system.

When components and subsystems are out-of-service, overall core and cooling reliability is maintained by OPERABILITY.of the containment remaining redundant equipment.

Whenever a CSCS system or loop is made inoperable,'the other CSCS systems are required to be OPERABLE shall be considered OPERABLE if or loops that they are within the required surveillance testing frequency and there is If the function, system, or no reason to saspect they are inoperable.

or calibration is found inoperable or exceeds the trip loop under ter.t level setting, the LCO and the required surveillance testing for the system or loop shall apply.

Maximum AveraRe Planar LHGR. LHGR. and MCPR The MAPLHGR, LHGR, and MCPR shall be checked daily to determine if fuel burnup, or control rod movement 'us caused changes in power Since changes due to burnup are slow, and only a few distribution.

control rods are moved daily, a daily check of power distribution is adequate.

3.5/4.5-36 Amendment flo. 130, 140 BrN Unit 3

. _ _ = _

v t e Letter Sequence Request
TAC:69888, (Open)
Initiation Request Acceptance... Administration Questions Answers Results Approval... Other: ML20005E882, ML20005G526, ML20236D142, ML20246M497, ML20248A145, ML20248G602
MONTHYEAR1989-05-08 [Table View]

ML20245F645

Text

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