Text

Amends 169,169 & 140 to Licenses DPR-33,DPR-52 & DPR-68, Respectively,Deleting Certain Surveillance Testing Requirements on Redundant ECCS & RCIC Sys During Limiting Condition of Operation
	ML20245G212
Person / Time
Site:	Browns Ferry ; (DPR-33-A-169-01, DPR-33-A-169-1, DPR-52-A-169, DPR-68-A-140)
Issue date:	08/02/1989
From:	Black S; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20245G217	List: ML20245G212; ML20245G219;
References
	NUDOCS 8908150274
	Download: ML20245G212 (118)
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UNITED STATES NUCLEAR REGULATORY COMMISSION e

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TENNESSEE VALLEY AUTHORITY DOCKET NO.-G0-259 BROWNS FERRY NUCLEAR PLANT, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.169 License No. DPR-33 1.

The Nuclear 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 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 Commission; 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 will be j-conducted in compliance with the Commission's regulations; D.

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

E.

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

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According?y, che license is amended by changes to the Technical i

Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. DPR-33.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 Suzanne sistant Director for Projects TVA Projects Division Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: August 2, 1989 6

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ATTACHMENT TO LICENSE AMENDMENT NO. 169 FACILITY OPERATING LICENSE NO. DPR-33 DOCKET NO. 50-259 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages.

The revised pages l

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 INSERT 1

i 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 3.5/4.5-26 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*

3.5/4.5-30 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 i

e.--------__-.__a-- - _ - - - - - - _. - - _ _. -

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_w-, -

-_cm- - -..-

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D_ET,5 0F C0ifrRLqS l

Section 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 A5p SURVEILLANCE REQUIREMENTS 3.1/4.1 Reactor Protecticn 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 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.

Flced Protection..

3.2/4.2-4 I.

Meteorological Monitoring Instrumentation.

3.2/4.2-4 J.

Seismic Monitorirg 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

Amendment No. 129, 161, BFN Unit 2 l

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' StL% Ann Pane No, g

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 /omponents.....

3.4/4.4-3 C.

Sodium Pentaborate Solution..

3.4/4.4-3 3d5/4.5 Core and Containment Cooling Systemr 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.,

High Pressure Coolant Injection 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).

3.5/4.5-18 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 B.

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 i

BFN Unit 2

). ".

L

~1.*

Section * '

Pere No.

. Jet Pumps;.

3.6/4.6 E.

F.

Recirculation Pump Operation.

3.6/4.6-12 3.6/4 5 'G.

Structural Integrity.

i H.-

Snubbers.

3.6/4.6-15 3.7/4~7' Containment Systems..

3.7/4. -1 3.7/4.7 A.

Primary Containment.

B.-

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................

3.8/4.8-1 B.

' Airborne Effluents............

3.8/4.8-3 C.

Radioactive Effluents - Dose.

3.8/4.8-6 3.8/4.8-6 D.

Mechanical Vacuum Pump.

li 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 3.9/4.9-15 C.

Operation in Cold Shutdown.

I' 3.10/4.10-1 l-3.10/4.10 Core Alterations..................

3.10/4.10-1 A.

Refueling Interlocks....

3.10/4.10-4 B.

Core Monitoring.

Amendment No. 128, 160, 169 iii BFN Unit 2 l

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lEggsian

. Pane No.

Spentfruel' Pool' Water.

C.

3.10/4.10-7 D.

Reactor Building. Crane.

........1 3.10/4.10-8 E.

Spent Fuel Cask..

3.10/4.10-9 F.

Spent Fuel Cash Handling-Refueling Floor.......

3.10/4.10-10 3.11/4.11 Fire Frctection Systems 3.11/4.11-1

.A.

Fire Detection Instrumentation...........

3.11/4.11-1 B.

Fire Purps 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 Vesse1~......-.............

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

r ADMINISTRATIVE CQNTROLS EECTION PAGE 121 RESPONSIBILITY..........................................

6.0-1 h2 QFGANIZATI0N............................................

6.0-1 6.2.1 Offsite and Onsite Organizations........................

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

6.0-2 h22 PLANT STAFF 0 QUALIFICATIONS..............................

6.0-5 1.4 T RA I N I N G................................................ 6. 0 - 5 h21 PLANT REVIEW AND AUDIT..................................

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 (26 EEPORTABLE EVENT ACTIONS................................

6.0-18 h2 SAFETY LIMIT VIOLATION..................................

6.0-19 hh PROCEDU.EES/ 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 - Effluent and Environmental Monitoring.............................

6.0-23 h22 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 l

l 6.10 STATION OPERATING RECORDS AND RETENTION.................

6.0-29 6.11 PROCESS CONTROL PR0 GRAM.................................

6.0-32 6.12 0FFSITE_DQSE CALCULATION MANUAL.........................

6.0-32 6.13 RADI O LO GI C AL E FFLUE NT MA NU AL c............................ 6.0-33

(

1 I

v BFN Unit 2 Amendment No. 124,161,169

l'

't LIST'0F TABLES Table Title Pare No.

11.1.

Surveillance' Frequency Notation..........

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

-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 Control Circuits........

3.1/4.1-8 4.1.B

' Reactor Protection System-(SCRAM) Instrumentation Calibration Minimum Calibration Frequencies for Reactor P;otection Instrument Channels.

3.1/4.1-11

'3.2.A Primarv Containment and Reactor Building Isolation Instrumentation,...

3.2/4.2-7 3.2.B Instrumentation that Initiates or Controis the Core and. Containment Cooling Systems.

'3.2/4.2-14 l

3.2.C

' Instrumentation that Initiates Rod Blocks..-...

3.2/4.2-25' p

3.2.D Radioactive Liquid Effluent Monitoring Instrumentation..................

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

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 i

3.2.H Flood Protection Instrumentation..........

3.2/4.2-35 1

l 3.2.I Meteorological Monitoring Instrumentation.....

3.2/4.2-36 3.2.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 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 4.2.D Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements.

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

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$[

4-s LIST OF TABLES'(Cont'd) l L

Pane No.

Title Table J4.2.E' Minimum Test and Calibration Frequency for Drywell 3.2/4.2-53 Leak Detection Instrumentation..........

' Minimum Test and Calibration Frequency.for 3.2/4.2-54

' 4. 2. F, Surveillance Instrumentation...........

Surveillance Requirements for Control Room 3.2/4.2-56

'4.2.G isolation Instrumentation....

Minimum Test and Calibration Frequency for I

Flood Protection-Instrumentation.

3.2/4.2-57 4.2.H Seismic Monitoring Instrument Surveillance 3.2/4.2-58 L

-4.2.J

. Requirements...................

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

ATWS-Recirculation.Puk., Trip (RPT) 3.2/4.2-63a 4.2.L instrumentation Surveillance............

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

3.5-1 3.5/4.5-21 MAPLHGR Versus Average Planar Exposure.......

3.5.I 3.7/4.7-25 Primary Containment Isolation Valves.

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

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

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

3.7.H 1

3.9/4.9-16 Diesel Generator Reliability...........

4.9.A 3.9/4.9-18 Vo'itage Relay Setpoints/ Diesel Generator Start.

1 4.9.A.4.C 3.11/4.11-14 Fire Detection Instrumentation...........

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

3.11/4.11-18 3.11.C Hose ftations.

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, 161, 169 BFN Unit 2

t I

LIST OF ILLUSTRATIONS figurg' Title Paae 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 l

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 K _ Factor...

3.5/4.5-23 f

3.6-1 Minimum Temperature *F Above' Change in Transient Ten:perature...

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 Elevations 3.8/4.8-10

4. 8.1. b -

Land Site Boundary.

3.8/4.8-11 i

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viii Amendment fjo.169 BFN Unit 2 I

4.

4 3.5/4.5-CORE AND CONTAINMENT COOLING SYSTEMS I

-LIMITING ~ CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5' CORE AND' CONTAINMENT COOLING 4.5 CORE AND CONTAINMENT COOLING 4

SYSTEMS SYSTEMS Aeolicability Aeolicghility 1

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.

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.

Specification Specification 1

A.

Core Sorav System (CSS)

A.

Core Sorav System (CSS) 1.

The CSS shall be OPERABLE:

1.

Core Spray System Testing.

(1) PRIOR TO STARTUP ligm 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.MM 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

,~

a 3'5/4.5 ' CORE'AND CONTAINMENT COOLING SYSTEMS w

t-LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5 A fore Sorav System (CSS) 4.5.A Core Sorav System (CSS) 4.5.A.1.d.(Cont'd) i.

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.

vise 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 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 Amendment No. 149, 155, 169 Unit 2 l^

j' i

i

4~

4 3.5/4'.5' CORE AND CONTAINMENT COOLING SYSTEMS.

' LIMITING CONDITIONS FOR OPERATION

~ SURVEILLANCE REQUIREMENTS f 3.5. A' Core Sorav System (CSS)

5. -When irradiated fuel is in

.the reactor vessel and '.he.

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 RHRSW pump-and associated' valves supplying the standby coolant supply 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 the reactor vessel, and the associated diesel generator (s) are required.

1 1

3.5/4.5-3 Amendment No. 158 BFN

. Unit 2' L_ _ __. _ _ _ _ _ _ __ _

L3.5/4.5 CORE'AND CONTAINMENT COOLING SYSTEM 3 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 RERS 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

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

BFN 3.5/4.5-4 Unit 2 Amendment No. 155,169

'd b pg l' e-e.

D

/3.5/4.5 CORE AND CONTAINMENT COOLING SYSTEMS SURVEILLANCE REQUIREMENT LIMITING CONDITIONS FOR OPERATION 4.5.B. Residual Hest Removal _ System 13.5.B: Residual Heat Removal System

~

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

Cooling) 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 two loop shall deliver 12000 spm RHR pumps-containment cooling against an indicated systeu 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.0PERABLE.

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 RHRS (LPCI mode) and the CSS and the diesel generators remain OPERABLE.

4.

No additional surveillance If any 2 RHR pumps (LPCI required.

4.

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 BFN

-Unit'2

- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - _ = _ - _ _

^~

3.5/4.5.00RE 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 RHR 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 'irys provided the remaining PhR 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) 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.

l t

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

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3.5/4.5 GORE AND CONTAINMENT COOLING SYSTEMS

)

' LIMITING CONDITIONS FOR OPERATICN SURVEILLANCE REQUIREMENTS

)

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s f

~3.5.B Residual Heat Removal System 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 shutdown shall be

)

initiated and the reactor j

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 af 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 .)

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

=.

s A Jill._,1_fl0RE ANQ_EQ3IA$ Mein COOLXNG SXSTEMS.

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)

12. If three RHR pumps or associated

12. No additional surveillance heat exchaneers located.

required.

s 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 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 restored.

14.

All recirculation pump 14.

All recirculation pump 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 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 BFN 3.5/4.5-8 Amendment No. 149, 155, 169 Unit 2

/4.( CORE AWD C0iTTAISMENT COOLING SYSTEMS ITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS

.C FHR Service Water and Emergency 4.5.C RHR Service Water and Emernency Eauipment CoolinR Water Sv;tems Eautoment Cooline Water Systems (EECWS_1 (EECWS1 1.

PRIOR TO STARIVP 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 Dl, 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 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 safsty-related equipment in the affected unit that is not locked, sealed, or otherwise secured in position, is in its correct position.

l 3.5/4.5-9 Amendment No. 155, 169 2

I i

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ll r

-3l5/4.5 CORE AND CONTAINMENT COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION-SURVEILLANCE REQUIREMENTS

.3.5.C RHR Service Water and Emernency 4.5.C RHR Service Water and Emernency Eauioment Coolina Water Systems Eauiement Coolina Water Systems (EECWS) (Continued)

/EECWS) (Continued) 2.

During REACTOR POWER

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

i must be OPERABLE and 4

assigned to service as l

indicated in Table 3.5-1 R

for the specified time limits.

3.

During Unit 2 REACTOR

3. Routine surveillance for POWER OPERATION, any two these pumps is specified RHRSW pumps (D1, D2,_B1, in 4.5.C.1.

and B2) normally or alternateJJ assigned to the-RHR heat exchanger header supplying the standby coolant l

supply connection must be OPERABLE except as specified in 3.5.C.4 and 3.5.C.5 below, l

i i

1 l

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

)

l

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i TABLE 3.5-1 1

Time Minimum Limit Service Assignment EECW(2) l (Days)

RHRSW (4)

(1) l Indefinite 7

3 1

(3)(4)

(1)

(3)

)

30 7

or 6 2

or 3 l

(4)

(1) 7 6

2 (1)

At Ir.ast one OPERABLE pump must be assigned to each header.

t (2)

Only automatically starting pumps may be assigned to EECW header l

service.

(3)

Nine pumps 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.

3.5/4.5-11 BFN Unit 2

i.. :. 4 -

7;.

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THIS PAGE INTENTIONALLY LEFT BLANK BFN 3.5/4.5-11a Amendment No 169 Unit 2 L._____u.________.

3 5/4.5 CORE AND CONTAINMENT COOLING SYSUES l

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.C RHR Service Water and Emergency 4.5.C RER Service Water and Emergency Eauioment Coolina Water Systems Eaulement Cooling Water Systems (EECWS) (Continued)

(EECWS) (Continued) 4.

Three of the Dl, D2, B1, B2

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

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.

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

d 3.5/t. 5 CORE Ai4D C014TAXNMEV4T C00LIi4G SLeJEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.D Ecuirment 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 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) served by that cooler must be considered inoperable for Technical Specification purposes.

E.

High Pressure Coolant Iniection E. Hich 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 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 l

normal months reactor vessel operating pressure l

l EFN 3.5/4.5-13 Amendment No. 155 71 nit 2 I

1 32}/4i5' ' CORE AND CONTA8NMENT COOLING 'SXSTEMS E

LIMITING CONDITIONS FOR OPERATION-SURVEILLANCE REQUIREMENTS i

High' Pressure Coolant Iniection 4.5.E High Pressure Coolant'Iniection l

~ 3.5.E System (HPCIS)

System (HPCIS)

,.5.E.1 (Cont'd) e.

Flow Rate at Oncz '

150 psig

operatin, cycle The HPCI pump shall deliver at least 5000 gpm during each flow rats 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 Coolinz F.

Reactor Core Isolativ 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' I

e

~3.5/4'5 CORE AND CONTAINMENT COOLING SYSTEMS l-LIMITING' CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.F.

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

System (RCICS)

I f3.5.F.1 '(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 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 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 thar. 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 1

position when a signal is i

present may be in a position for another node of operation.

BFN 3.5/4.5-15 Amendment No. 169 Unit 2

s.

?3.5/4;5 CORE AND CONTA8NMENT COOLING pYSTEMS.

_ LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.G Automatic ~Deoressurization 4.5.G Automatic Deoressurization System'(ADS)

System (ADS)

),

1.

Four of the six valves of 1.

During each operating the Automatic cycle the following L

Depressurization System tests shall be performed l

shall be OPERABLE:

on the ADS:

(1) PRIOR TO STARIC?

a.

A simulated automatic from'a COLD CONDITION, actuation test shall or, be performed PRIOR TO STARTUP after etch (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 incapable of automatic operation, an immediate orderly shutdown shall be initiated, with the reactor in a HOT SHUTDOWN 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 Specification.s 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 reduced to 105 psig or leas within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

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

I l

J.5/4.5' CORE AND CONTAINMENT COOLING SYSIgMS t

' LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.H.

Maintenance of Filled Discharr.e 4.5.H. Maintenance of Filled Discharge Eip_t Pipe Whenever the-core spray systems, The following surveillance i

l 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 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 P1-75-20 48 psig system to service.

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

3. Whenever 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.

Amendment tio. 169 BFN 3.5/4.5-17 Unit 2

I

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(.,

1 J

l l

j 4;

I

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4 j

[

3.5' BASES and 3.5.B Residual Heat Removal System (RHRS)

Core Sorav System (CSS)

3.5.A.

Analyses presented in the FSAR* and analyses presented in conformance in with 10 CFR 50, Appendix K, demonstrated that the core spray eystem conjunction with two LPCI pumps provides adequate c limit fuel clad temperature to below 2,200*F which assures that core i

l ter geometry remains intact and to limit the core average clad meta -waCore s reaction to less than 1 percent.

i in tests of systems similar to design to BFNP to exceed the mi less than half the rated flow in simulated fuel assemblies with requirements.

l rods to duplicate the decay heat characteristics of irradiated fue.

at The RHRS (LPCI mode) is designed to provide emergency. cooling to the cor

-This system is by flooding in the event of.a loss-of-coolant accident.

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

k ling without assistance from the high-pressure emergency core coo subsystems.

The intent of the CSS and RERS specifications is to not allow startup E

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.

Should one core spray loop become inoperable, the remaining core spray l

loop, the RHR System, and the diesel generators are required to beThese p OPERABLE should the need for core cooling arise. margin days With due regard for this margin, the allowable repair time of seven 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

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

1 1

1 3.5/4.5-24 Amendment No. 169 BFN l

Unit 2

3.,5 EAIES (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 period is justified.

Should two RHR pumps (conte.inment cooling mode) become inoperable, a full "at removal system is still available. The remaining pump / heat

' hanger combinations would provide adequate enrtainment cooling for any recermal 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 vill be sufficiently cooled, following a loss-of-coolant accident, to prevent primary containment crerpressurization. The containment cooling function of ths 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 Systes.

Since it is desirable to have the RHRS in service if a " pipe-break" type of accident should occur, it is prrmitted 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 '.his 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 af 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 l

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 repait period is justified.

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

o E

a L3.5 Bases-(Cont'U)

The suppression chamber can be drained when the reactor vessel pressure is atmospheric, irradiated fuel is in the reactor vessel, and work is EBy not' in progress which has the potential to drain the vessel.

requiring the fuel pool gate to be open Sith the vessel head removed, the combined water inventory in the fue? pool, the reactor cavity, and the' separator / dryer pool, between the f el pool low level alarm and the reactor vessel flange, is abot;. t?,800 cubic feet (492,000 gallons).

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

providing standby coolant supply available will ensure a redundant Control rod drive maintenance may continue supply of coolant supply.

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 a 10-day repair time is allowed before shutdown is lines be lost, This repair time is justified based on the very small required.

probability for ever needing RHR pumps and heat exchangers to supply an adjacent unit.

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

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

Cooline Water System 3.5.C. RHR Service Water System and Emergency Eaulement

_EECWS)

(

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

emergency cooling water are fed from both headers thus assuring Each header contin.ity of operation if either hemeer 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 exchanger from each unit on each header.

pumps on each header; one normally assigned to each header (A2, B2, C2, heat or 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 RHRSW heat exchangers on the One RHRSW pump can supply the full flow requirement of one RHR header.

Two RHR heat exchangers can more than adequately heat exchanger.

handle the cooling requirements of one unit in any abnormal or postaccident situation.

3.5/4.5-26 BFN Unit 2

..* ?

A 3.5 BASF5 (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 i

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 RHRSW pumps normally or alternately assigned to the RHR heat exchanger header supplying the standby coolant supply.

connection become inoperable, an equel 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 repa r period is justified. Should the i

capab511ty 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 operstor is disconnected 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 Eculement 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 equipment vaste 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-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)

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

.3.5 BASES (Cont'd)-

3.5.E. High Pressure Col ant Iniection System (HPCIS) l 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 the nuclear syctem and loss of coolant which does not re1 ult 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 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 System begins operation, the reactor depressurizes more l'

rapidly than would occur if HPCI was not initiated due to the condensation of steam by the enld fluid pumped into the reactor vessel by the HPCI system. As the reactor vessel pressure continues to decraase, the HPCI 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 breaks. 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 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, auch 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 feedvater makeup during primary system isolation conditions. The ADS server 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 realited 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 axiomatically isolated before the reactor pressure decreases below 100 psig.

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

I ETN 3.5/4.5-28 Amendment No. 169

= Unit 2

325' HMM (Cont'd)

.3.5.F Egantpr Core Isolatien Cooling System (RCICS)

The various conditions under which the RCICS plays an ebsential role in providing makeup water to the reactor vessel have keen identified by evaluating the various plant events over the ful1 range of planned

~

operations. The specifications ensure that the function for which the RCICS 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 te.mperature up to 140*F vith no containment back pressure.

Because the low-pressure cooling systems (LPCI cud 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 loss of feedwater flow at design power (105 percent of rated).

Consideration of.the availability of the RCICS reveals that the average i

risk associated with failure of the RCICS to cool the core when 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 nuclear system is an essential response.7 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 Note that this specification can operate to protect the fuel barrie>.

applies only to the automatic feature of the pressure relief system.

Specification 3.6.D specifies the requirements for the pressure relief i

It is possible for any number of the valves function of the valves.

assigned to the ADS to be incapable of performing their ADS functions because or ' instrumentation failures yet be fully capable of performing their aressure relief function.

I 3.5/4.5-29

{

BFn i

Unit 2 i

i

)

m

_ _ ~ _ _ _ _. _ _ _ _..

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

3.5 BASES (Cont'd)

Because the automatic depressurization cystem 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.

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 Discharme Pine If the discharge piping of the core spray, LPCI, HPCIS, and RCICS are not filled, a water hammer can develop in this piping phen the pump and/or

'l 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 bc 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 in>.,perable for Technical Specification purposes.

The'coxe 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 ard RCIC discharge piping remains filled. Further assurance is provided by observing water flow from these systems' high points monthly.

I.

Maximum Average Planar Linear Heat Generation Rate (MAPLEGR) 1 j

1 This spec! fication assures that the peak cladding temperature following i

thc postulated design basis loss-of-coolant accident will not exceed the limit'specified in the 10 CFR 50, Appendix K.

}

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

s 3.5 BASIS (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 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 distribution within a fuel assembly affect the calculated peak clad temperature by less than i 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 in 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 (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 LHCR shall be checked daily during reactor operation at 1 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 power levels less than or equal to 25 percent, the reactor will be operating at minimum recirculation pump speed and the moderator void centent 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 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 l

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 LHGR of 18.5 kW/ft for 7x7 fuel and 13.8 kW/ft.

This limit is reached when 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 3.5.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 l

.3.5 ' BASES.(Cont'd),

of CHFLPD 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 ensured.by the safety analysis.

3 5.M. References 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-TR81-01-A.

3.

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

I BFN 3.5/4.5-32 Unit-2 Amendment No. 143 1

I

'y'

j 4.5-Core and Containment Coo 1Rnz 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 1

loss-of-coolant inventory. To increase the availability of the core l

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 tijection valves are also tested in accordance with Specification 1.0.MM to assure their OPERABILITY. A simulated automatic actuation test once each cycle combined with testing os the pumps and injection valves in accordance with' Specification 1.0.tlh is deemed to be adequate testing of these systems. Monthly alignment checks of valves 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 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 out-of-service, overall core and containment cooling reliability is 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 inoperable.

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 shall apply.

Maximum Average Planar LHGR. LHGR. and MCPR The MAPLEGR, 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.

l l

i i

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

. [J =<lu,%,

UNITED STATES

'i NUCLEAR. REGULATORY COMMISSION

%, ~. ]p

' WASHINGTON, D. C. 20555 l

TENNESSEE VALLEY AUTHORITY 1

DOCKET NO. 50-260 BROWNS FERRY NUCLEAR PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.169 License No. OPR-52 1.

The Nuclear Regulatory Commission (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 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 Commission; C.

There is reesonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

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

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

1

. l 2.

Accordingly, the license.is amended by changes to the Technical Specifications as indicated in the attachment to this license arandment and paragraph 2.C.(2) of Facility Operating License No. DPR-52 is'hereby.

dmended to read as follows:

(2) Technical Specifications

.The Technical-Specifications contained in Appendices A and B, as j

revised through Amendment ik 169, are hereby incorporated in the j

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

. C 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

.--_-_____--.--.-.m.

/"I

11,

. ~.

ATTACHMENT TO LICENSE AMENDMENT N0.169 FACILITY OPERATING LICENSE NO. OPR -52 DOCKET NO. 50-260 Rev'.se the Appendix. A Technical Specifications by removing the pages it entified below and inserting the enclosed pages. The revised pages a/e identified by the captioned amendment number and contain marginal line s indicating the area of change.

Table of Contents and overleaf pages*

are provided to maintain document completeness, 1

REMOVE INSERT i

1*

l 11 ii 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 3.5/4.5-24 3.5/4.5-24 3.5/4.5-25 3.5/4.5-25 4

3.5/4.5-26 3.5/4.5-26*

i 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*

q 3.5/4.5-30 3.5/4.5-30 1

3.5/4.5-31 3.5/4.5-31*

3.5/4.5-32*

3.5/4.5-33 3.5/4.5-33

1 o:1 '*

TIBLE OF CONTENTS

'Section Pane No.

+

'1.0-Definitions'....................

1. 0-l '

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 OPERATICN AND SURVEILLANCE REQUIREMENTS 3.1/4.1 Reactor Protection System...

3.1/4.1-1 3.2/4.2 Protective Instrumentatier.............

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..

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

Eeetien Eac t_HL.

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 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 Systems (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 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 l

J.

Linear Heat Generation Rate (LHGR) 3.5/4.5-18 1

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 1

8 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 11 Amendment No. 133, 147, 164, 169 BTN UNIT 1 l

1

s-

.s

,Section Egne 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 3.6/4.6-15 H.

Snubbers 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 D.

Primary Containment Isolation Valves.

3.7/4.7-17 E.

' Control Room Emergency Ventilation......

3.7/4.7-19 F.

Primary Contain. tent 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.

3.8/4.8-1 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 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 l

3.10/4.10-4 l

B.

Core Monitoring.

iii Amendment Nos. 132, 163, 169 273 I

UNIT 1 i

s Secti2D Page 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 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 Fire Pumps and Water Distribution Mains 3.11/4.11-2 B.

C.

Spray and/or Sprinkler Systems.

3.11/4.11-7 D.

CO2 Systems.

3.11/4.11-8 E.

Fire Hase 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 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

s

~.4 ADMINISTRATIVE CONTROLS i

SECTICH PAGE I

l 621 RESPONSIBILITY..........................................

6.0-1 622 ORGANIZATION............................................. b.0-1 6.2.1 Offsite and Onsite Organizations........................

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

6.0-2 612.

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

6.0-5 624 TRAINING................................................

6.0-5 hil PLANT REVIEW AND AUDIT..................................

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.............

6.0-17 111 RE PO RTABLE EVE NT A CT I O N S................................

6. 0-18 6.7 S A FETY LI MI T VI OLATION..................................

6. 0-19 111 PROCEDURES / INSTRUCTIONS AND PROGRAMS....................

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

6.0 6.8.2 Dri11s..................................................

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 h22 REPORTING RE0UIREMINTS..................................

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 AND RETENTION.................

6.0-29 6.11 PROCESS CONTROL PR0 GRAM.................................

6.0-32 l

6.12 0FFSITE DOSE CALCULATION MANUAL.........................

6.0-32 1

k213 RADIOLOGICAL EFFLUENT MANUAL............................

6.0-33 v

' BMN Amendment Nos. 138, 169 UNIT 1

LIST Of TABLES Table Title Page No.

1.1 Surveillance Frequency Notation..........

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

Instrumentation Requirements.

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

3.1/4.1-8 4.1.B 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 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 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

)

l 3.2.G Control Room Isolation Instrumentation.

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

3.2/4.2-35 3.2.1 Meteorological Monitoring Instrumentation.

3.2/4.2-36 3.2.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 j

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

3.2/4.2-4C 4.2.B Surveillance Requirements for Instrumentation that Initiate or Control the CSCS.

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

.2/4.2-50 Initiate Rod Blocks.

4.2.D Radioactive Liquid Effluent Monitoring f.

Instrumentation Surveillance Requirements.

3.2/4.2-51 vi Amendment No. 132, 164, 169 I

BFN f

Unit 1 I

l

f

~. -

L LIST OF TABLES (Cont'd)

Tible Title Pare N22 4.2.E Minimum Test and Calibration Frequency for Drywell 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 Isolation 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 Fffluent 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 Penetrations 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 Isc.Jaticn 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 j

4.9.A Auxiliary Electrical Systems.

3.9/4.9-16 l

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 1

I 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 vii l

BFN j

Unit 1 Amendment Nos. 153,162,164,169 l

l

t.

LISJ OF TLLUSTRA710RE

- Eiaure Title Page No.

2.1.1 -

'APRM Flow Reference Scram and APRM Rod Block 1.1/2.1-6

-Settings.

2.1-2 APRM Flow Bias Scram Vs. Reactor Core Flow.

1.1/2.1-7 f

4.1-1:

Graphical Aid.in.the Selection of an Adequate

. Interval'Between Tests...........~.

3.1/4.1-13 14.2 System Unavailability.

3.2/4.2-64 l

MCPR Limits.....................

3.5/4.5-24 L

.3.5 K-l'

'3.5.2

-Kf Factor.

3.5/4.5-25

3. 6-l'

' Minimum Temperature CF Above Change in

. Transient Temperature...

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

Neutron Exposure.

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

3.8/4.8-10 4.8.1.b Land Site Boundary.

3.8/4.8-11 1

J

)

Amendment No. 169 viii BFN j

Unit 1

)

..-1

~

j

,.; 7

=g

,a

.,.,y.

. J [ 5 / 4'. 5 CORE AND CO?rrAf fM;I[L,pAQLTNG SYSTEMS SURVEILLANCE REQUIREMENTS

' LIMITING r,0NDITIONS FOR g ERATION 3.5. CORE AND CONTAINMEN7 COOLING 4.5 00RE AND CONTAINMENT COOLING SYSTEMS SYSTEMS Aeolicability

'Arolicability Applies to the operational Applies to the surveillance requirements of the core and status of the core and containment cooling systems when containment cooling systems.

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

^

Obiective Obiective

~

To verify the OPERABILITY of the 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 Soray System (CEE)

A.

Core Sprav System (CSS) 1.

Core Spray System Testing.

1.

The CSS shall be OPERABLE:

11gm Freouency (1) PRIOR TO STARTUP from a a.

Simulated Once/

COLD CONDITION, or Automatic Operating Actuation Cycle (2) when there is irradiated test

, fuel in the vessel and when the reactor b.

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

Motor Per Specifi-except as specified Operated cation 1.0.MM i

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 Amendment No. 159 BFN

' Unit 1

= - _ _ _ _ - _ _ - _ _ _ _

0 3.5/4.5 CORE AND CONTA7WMENT COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS

-3.5.A G. ore Sorav System (CSS) 4.5.A G2re 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.

'f one CSS loop is inoperable, f.

Verify that Once/ Month che reactor may remain in each valle operation for a period not to (manual, power-exceed, days providing operated, or all active components in automatic) in the the other CSS loop and the injection flovpath 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 Except that an automatic OPERABLE pump and associated dies:1 generator shall be valve capable of autematic.

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 Amendment No. 153, 159, 169 Unit 1

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o 3.5/4.5 CORE AND CONTAINMENT COOLIRG SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS i

l 3.5.A Core Sorav System (CSS) 5.

When irradiated fuel is in the reactor vessel and the reactor vessel head is

{

removed, core spray is not j

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 RHRSW pump and associated valves supplying the standby

{

coolant supply are OPERABLE.

When work is in progress which has the potential to drain the vessel, manual initiation capability of either 1 CSS Loop or 3 RHR pump, with the capability of injecting water into the reactor vessel, and the associated diesel generator (s) are required.

1 l

l BFN 3.5/4.5-3 Amendment No. 161 Unit 1 l

t

i 3.5/i.5 CORE AND C0ifTAlf4MEWT COOLING SYSTLtiS LIMITING CONDITIONS FOR-CPERATION 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/

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 EILITY 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.

Test Check 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 ADA 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.

BPN 3.5/4.5-4 Unit 1 Amendment Nos. 159,169 0

~=

3.5/4.5 C01% AND CONTAINMENT COOLING SYSTEMS

' LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENT l

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

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

Cooling)

I 4.5.B.1 (cont'd) i l

2.

With the reactor vessel Each LPCI pump shall deliver pressure less than 105 psig,

-9000 gpm against an indicated the RHRS may be removed system pressure of 125 psig.

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

-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 being drained of and torus headers and nozzles suppression chamber quality shall be conducted once/5 water 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 .

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

e o

3 5/4.5 CORE AND CONIA7NMENT 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 RHR 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 survelliance 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 I-may remain in operation for a period not to exceed 7 days provided at least one path for each phase of the mode i

remains OPERABLE.

BFN 3.5/4.5-6 Amendment No. 153, 169 Unit 1

_____E_-_--_-_

J.$/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.1.(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 j

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 .)

a l

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

.o

225/4.5 CORE AND CONTAINMEETT C00LIWG SYSTEMS i

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS f3.5.B Residual Heat Removal System 4.5.B Residual Heat Removal System (RHRS) (LPCI and Containment (RHRSS (LPCI and Containment Cooling)

Cooling)

.12. If one RHR. pump or associated

12. No. additional surveillance heat exchanger-located required.

on the unit cross-connection in the adjacent unit is

. inoperable for any reason

-(including valve inoperability, pipe break, etc.), the reactor l '

may remain'in operation I

for 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 not 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 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.

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

-n

o 3.5/4.5 CORE AND CONTAINMENT COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS l

3.5.C RHR Service Water and Emerrency 4.5.C RHR Service Water and Emeraency Eauipment Cooline Water Systems Eauipment Cooline Water Systems (EECWS)

(EECWS) 1.

PRIOR TO STARTUP from 1.

a.

Each of the RHRSW pumps l

a COLD CONDITION, 9 RHRSW normally assigned to pumps must be OPERABLE, with automatic service en 7 pumps (including pump D1 the EECW headers will or D2) assigned to RHESW 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 cont.rol valves for thn EECW headers and ElR heat exchanger headers shall be demonstrated to be OPERABLE in accordance with Specification 1.0.MM.

b.

Annually each PJiRSW pump shall be flow-rate tested. To 5e 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

,mg position, is in its correct position.

5 BFN 3.5/4.5-9 Amendment No. 159, 169 Unit 1 a'

---~_-__

,u p

-3.5/4.5 CORE AND CONTAINMENT COOLING' SYSTEMS h

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS l3.5.C RHR-Service Water and Emermency 4.5.C RHR Service Water and Emergency Eauioment Coolina Water Systems Engippgni Coolina Water Systems R

LEECWS) (Continyrdl IEECWS) (Continued) 2.

During REACTOR POWER

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

must be 9PERABLE 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 RHR 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.

BTN 3.5/4.5-10 Amendment tio. 153, 169

' Unit 1-en

~. - - - - _ _

4 o

i

)

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

RHRSW (Days)

(1)

(4) 3 7

Indefinite (1)

(3)

(3)(4) 7 or 6 2

or 3 30 (1)

(4) 2 6

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

(1)

At Only automatically starting pumps may be assis,ned to EECW header (2) service.

Either configuration is Nine pumps must be OPERABLE.

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

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

l I

3.5/4.5-11 BFN Unit 1

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4

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THIS PAGE INTENTIONALLY LEFT BLANK BFN 3.5/4.5-11a Amendment No. 169 Unit 1

'3.5/4.5 CORE AND CONTAINMENT COOLING SY3TEMS LIMITING CONDITIONS FOR OPERATION-SURVEILLANCE REQUIREMENIJ 3.5 ' C RPR Service Water and Emergency 4.5.C RHR-Service Water and Emeraency Eauloment Coolina Water Systems Eauipment Coolina Water Systems' (EECWS) (Continutd)

(EECWS) (Continued) 4.

One of the El or D2 RHRSW

4. No additional survef?. lance pumps assigned to the RHR is required.

heat exchanger supplying tne standby coolant aspply 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 inoperatie 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 RiiR 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

c 11/4.5 (01E AND COWTAI?tMENT COOLYWG SYST?MS LIMITING CONDITIONS FOR OPERATION SURW!LLANCE 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 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 la not OPERABLE, the pump (s) served by that cooler must be considered inoperable for Technical Specification purposes.

E.

H.ich 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 SIARTUP from a a.

Simulated Once/

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

Pur p 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 et Once/3 normal months reactor vessel operating pressure l

l BFN 3.5/4.5-13 Amendment Po. 159 f

Unit 1 I

_-__-______ a

a

~

3.5/4.5 ' CORE AND CONTAINMENT COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.E High Pressure Coolant IniectiRD 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 cy-le 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, RH2S (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 mey be in a sha13 be reduced to 122 position for another mode of psia or less within 24 operation.

hours.

F.

Reactar Core Isolation Coolina F.

Reactor Core Isolation Coolina System (RqlgEl 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 BEN 3.5/4.5-14 Amendment No. 159, 169 Unit 1 m_

____.----------m-

3.5/4.5 Q.4"E AND CONTAINMENT COOLING SYSTEMS

, LIMITING MDITIONS' FOR OPERATION.

SURVEILLANCE REQUIREMENTS 3.5.F.

Reactor Core Isolation Coolina 4.5.F-Reactor Core Isolation Coolina System (RCICS,1 System (RCICS) i 3.5 F.1' (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 Specifi-OPERABILITY cation 1.5 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 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 i 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 l

valve capable of automatic i

return to its normal I

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 J______-______

3.5/4.5 CORE ARD CONTAINMENT COOLING SYSIL'11 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.G Automatic Deoressurization 4.5.G Automatic DeDressurization System (ADS)

System (ADS) 1.

Four of the six valves of 1.

During each operating the Automatic cycle the following Depressurization System teste shall be performed shall be OPERAELE:

on the ADS:

(1) PRIOR TO STARTUP a.

A simulated automatic 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 prersure is covered in 4

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, t he reactor may remain in operation for a period not to exceed 7 days, provided the HPCI system is OPERALLE. (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 incapable of automatic operation, an immecitte orderly shutdown shall be initiated, with the reactor in a HOT SHUTDOWN 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 snall be reduced to 105 psig or less within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

g BFN 3.5/4.5-16 l

Unit 1 Amendment No. 169 l

j

s.

el

.L.,

e' 23.5/4.5 CORE AND COITTAINMENT COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS-Maintenance of Filled Discharae 4.5.H. Maintenance of Filled Discharge 3.5 H.

E121 EiPA I

-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 Lore 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 j'

pumpe shall be aligned to the testing of the RHRS (LPCI and conderAate 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 psic P1-74-51 48 psig

3. Whenever 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.

i l

BFN 3.5/4.5-17 Amendment No. 169 Unit 1

l t

35 HASIS E3.5,A.

Core Sprav System (CSS} and 3.5.B Residual Heat Removal System (RHRS)

Analyses presented in the FSAR* at 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 associatei with the loss-of-coolant 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 syst.tms similar to design to BFNP to exceed the miniaum L

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 loss-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 and the core spray system provide adequate cooling for break areas of approximately 0.2 square feet up to and including the double-,nded 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 operation, 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 the diesel generators are required to be OPERABLE 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 RER pu=p (LPCI mode) become inoperable, three RHR pumps (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 .

i

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

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

+

4

.35 MSIS -(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 remaining pump / heat exchanger combinations would provide adequate containment cooling for any abnormal postaccident situation. Because of the availar, 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 vill 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-thiros core height level. This prevents inadvertent'y 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 kHR 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 i

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 RER pump or associated heat exchangar located on the unit I

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.

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

3.5' Eases (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 l

the fuel pool gate to be open with the vessel head removed, the combined I

water inventory in the fuel pool, the reactor cavity, and the i

separator / dryer pool, between the fuel pool low level alarm and the j

reactor vessel flange, is about 65,800 cubic feet (492,000 gallons).

l This will provide adequate low-pressure cooling in lieu of CSS and RHR l

l (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 redundant supply i

of coolant supply. Control rod drive maintenance may continue during this period provided no more than one drive is removed at a time unless l

blind flanges are installed during the period of time CRDs are not in

)

place, j

Should the capability for providing flow through the cross-connect lines i

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 Coolina Water System (EECWS)

There are two EECW headers (north and south) with four automatic starting RHRSW 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 tandle 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) 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 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 more than adequately handle the cooling requirements of one unit in any abnormal or postaccident situation.

BFN 3.5/4.5-28 Unit 1

)

4-3.3 E! AIM (Cent'd)

The RHR Service Water System was designed as a shared system for three units. The specification, as written, is conservative when consideration ir 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 tbe NRC to allow continued operation if the actual system co711ng requirements can be assured.

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

3.5.D Equipment Area CooJerg 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 cr 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 l

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 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 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 This during the testing of the equipment served by these coolers.

testing is adequate to assure the OPERABILITY of the equipment area coolers.

REFERENCES 1.

Residual Heat' Removal System (BFN FSAR Section 4.8)

Core Standby Cooling System (BFN FSAR subsection 6.7) 2.

3.5/4.5-29 Amendment No. 169 BFN Unit 1

3.5 BASES (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 the 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 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 l

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 System 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 continues to decrease, the HPCI 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 breaks. 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 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 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 capa'ity c

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

_-y.

q

)

3.3 ;

EAEEE (Cont'd) 3.5.F. Reactor Core Isolation CoolinR SVstem (RCICS) j i

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

operations. The specifications ensure that the function for which the RCICS 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'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 fue.1 for a complete loss 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 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 Deeressurization System (ADS)

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 coolans injection (LPCI) and the core spray subsystems can operate to protect the fuel barrier. Note that this 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.

3.5/4.5-31 BFN Unit 1

7

.~

3.5s 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 l

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 OPERABLE.

Operation with more than three of the six ADS valves inoperable is not acceptable.

H.

Maintenance of Filled Discharae Pine If the discharge piping of ' he core spray, LPCI, HPCIS, and RCICS are not t

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 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 Average 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.

i BFN 3.5/4.5-32

- Unit 1 Amendment No. 169 1

_m_-_._______

._________.._____.___-_.m___.._.____m______m

o

C

.Ol e

3.5-BASES (Cont'd) p_

The peak cladding temperature following a postulated loss-of-coolant j

accident is primarily a function of the average heat generation rate of-l 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 distribution within a fuel assembly affect the calculated peak clad temperature by less than i 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 in Tablen 3.5.I-1,

-2,

-3

-4,

-5, and -6.

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

3.5.J. Linear Heat Ggnerstion 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 1 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 ifTPF 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 thermal 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 it. crease 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 Setpoints 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 1

~ l

3.5~ ' BASES (Cont'd) 0.5.M. preferences

" Fuel Densification Effects on General Electric Boiling Water

' Reactor Fuel," Supplements 6, 7, and 8, NEIM-10735, August-1973.

1.

Supplement 1 to Technical Report on Densification of General 14, 1974 (USA Regulatory Staff).

2.

Electric Reactor Fuels, December 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.

Amendment No. 147 3.5/4.5-34 BFN Unit 1

n.

4.5 Core and Containment Cooling Systems Surveillance Frequencies i

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 LPCI, automatic initiation during power operation would result in pumping cold water into the reactor vessel which is not desirable. Complete ADS testing dur!_g 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' injection valves are 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 accordance with Specification 1.0.MM is deemed to be adequate testing of these systems. Monthly alignment checks of valves 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 in the proper position. Valves which automatically reposition themselves on an initiation signal are permitted to be in a position other than normal to facil' tate other operational modes of the system.

When components and subsystems are out-of-service, overall core and containment. cooling reliability is 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 inoperable. If the function, system, or loop under test or calibration is found inoperable or excacds the trip level setting, the LCO and'the required surveillance testing for the system or loop shall apply.

Maximum Average Planar LHGR. LHGR. and MCPR The MAPLHGR, LEGR, 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.

1

{

Amendment No. 159, 169 j

3.5/4.5-35 BFN Unit 1

3

. w>wc,h u

/

UNITED STATES

.[ '* :, (-i NUCLEAR REGULATORY COMMISSION 7

wassiucToN. o c 20sss

>,/

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 Commission (the Commission) has found that:

A.

The application for amendmen6 by Tennessee Valley Authority (the licensee) dated Janue y 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.1; B.

The facility will operate in conformity with the application, the provisions of the Act, and the' rules and regulations of the Commission; C.

There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; l

D.

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

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

l l'

l l

L _

7o.. -.

. 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 amended 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 issuance and shall be implemented within 60 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION 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 i

___.__._._________J

TT; -):

1 c

,O

)

4.

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 amendment number and contain marginal lines indicating the area of change. Table of Contents and overleaf pages*

are provided to maintain document completeness.

REMOVE INSERT-1 i

ii 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 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 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

U

.n 4

4 I

TABLE OF CONTEETTS Lt; iz Pane No.

l '. 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 COFTDITIONS FOR OPERATION AND CUEVEILLANCE 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 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 E.

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

Amendnent f! - 104, 135 Unit 3

Stction ZaRe 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 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 Systems (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 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).

3.5/4.5-18 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 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 li l

BFN Unit 3 Amendment Nos. 104,118,135,140 l

t

s E121.123 '

Pace No.

i F.

Recirculation Pump Operation.

3.6/4.6-12 C.

Struttural Integrity.

3.6/4.6-13 I

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 B.

Standby Gas Treatment System.

3.7/4.7-13 C.

Secondary Containment.

3.7/4.7-16 L.

Irin.ary 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 (CA'd).

3.7/4.7-22 H.

Containment Atmosphere Monitoring (CAM)

System Hy Analyzer.

3.7/4.7-23 3.b/4.b 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.

3.8/4.8-6 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.

3.9/4.9-14 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 C.

Spent Fuel Fool Water.

3.10/4.10-7 iii Amendment No. 103, 134, 140 BFN Unit 3 l

w__ - - -

Saetien Page h D.

Reactor Building Crane.

3.10/4.10-8 E.

Spent Tuel Cack.

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 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.

3.11/4.11-13 Spreading Room 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 Storsge.

5.0-1 5.6 Seismic Design.

5.0-2 l

iv BFN Unit 3 Amendment Nos. 109.133,140

g 4

ADMINISTRATIVE CONTROLS _.

SECTION PAGE 122 RESPONSIBILITY....................................,..

122 ORGANIZATION............................................

6.0-1 6.2.1

..... 6.0-1 Offsite'and'Onsite Organizations...................

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

6.0-1 6.0-2 12]

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

s2s TRAINING......................................

6.0-5 6.0-5 621 PLA NT REVIEW A ND AUD I T.......................

6.

. 6.0-5

' 5.1 Plant Operations Review Committee (P0RC)................

6.5.2

. 6.0-5 Nuclear Safety Review Board (NSRB).....................

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

12h 6.0-17 REPORTA BL E EVENT A CTI ONS.....................

6.0-18' 121 SA FETY LIMIT VIOLATIO N........................

121-

. 6.0 PROCEDURES / INSTRUCTIONS AND PR0 GRAMS............

'6.8.1

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

6.8.2 Dri11s......................................

6.0-20 6.0-21 6.8.3 Radiation Control Procedurea............................

6 0 22 6.8.4 Quality Assurance Procedures - Effluent and Environmental Monitoring.............................

6 0-23 629 -

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 AND RETENTION...

6.11 PROCESS CONTROL _ PROGRAM.................................

6 0-32 6.12 0FFSITE DOSE CALCULATION MANUAL...................

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

6 0-3 3 Amendment No. 109, 140 BFN y

Unit 3 l

l

i 6

LJST OF IAELES

+

Table Title Pane No.

1.I Surveillance irequency Notation.

1.0-13 I

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 Monitors Leakage Into Drywell.

3.2/4.2-29 3.2.F Surveillance Instrumentation..

3.2/4.2-30 l

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 Monitoring Instrumentation.

3.2/4.2-35 3.2.J Seismic Monitoring Instrumentation.

3.2/4.2-36 3.2.K Radiot

, re Gaseous Effluent Monitoring Instx mentation.

3.2/4.2-37 3.2.L ATWS-Recirculation Pump Trip (RPT)

Surveillance Instrumentation..

3.2/4.'-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 for Instrumentation that Initiate Rod Blocks.

3.2/4.2-49 4.2.D Radioactive Liquid Eff2uent Monitoring Instrumentation Surveillance Requirements.

3.2/4.2-50 vi Amendment No. 103, 135,140 BFN Unit 3

LIST OF TABLES (Cont 'd)

IAtl'e Title EaRe No.

4.2.E Minimum Test and Calibration Frequency for Dr Leak Detection Instrumentation.......ywell 3.2/4.2-52 4.2.F Minimum Test and Calibration Frequency for Surveillance Instrumentation.

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....

1 3.2/4.2-56 4.2.J Seismic Monitoring Instrument Surveillar.ae Requirements.

3.2/4.2-57 4.2.K Radioactive Gaseous Effluent Monitoring Instrumentation.................

2/4.2-61 4.2.L ATWS-Recirculation Pump Trip (RPT)

Instrumentation Surveillance.

3.2/4.2-62a 3.5-1 Minimum RHESW and EECW Pump Assignment.

3.5/4.5-11 3.5.I MAPLHCR Versus Average Planar Exposure.

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

3.7/4.7-24 i

3.7.B Testable Penetrations with Double 0-Ring Seals.

3.7/4.7-32

3. 7.. C Testable Penetrations with Testable Bellows.

3.7/4.7-32 3.7.D Air Tested Isolation Valves.

3.7/4.7-33 3.7.E Prir.ary Containment Isolation Valves which Terminate below the Suppression Pool Water Level.

3.7/4.7-36 3.7.T Primary Containment Isolation Valves Located in Water Sealed Seismic Class 1 Lines.

3.7/4.7-37 3.7.H Testable Electrical Penetrations..

i 3.7/4.7-38 4.9.A Auxiliary Electrical System.

3.9/4.9-15 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 l

Spray / Sprinkler Systems.

l 3.11.B l

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 Minimuu Shift Crew Requirements.

6.0-4 BFN vii Unit 3 l

Amenoment Nos. 1?4,135,140 l

LIST OF IU,USTRATIOi4S Etrure Title Pace No.

2.1.1 APRM Plow Reference Scra'n and APkM 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.

3.1/4.1-12 4.2-1 System Unavailability.

3.2/4.2-63 3.5.Y-1 MCPP Limits.

3.5/4.5-25 3.5.2 Kf factor.

3.5/4.5...................

3. 6 -2 Minimum Temperature OF Above Change 1:a Transient Temperature.

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

i Nsutron 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/ 8-11 1

\\

i l

l viii Amendment NO. 140 BFN Unit 3 lL_ __ - - - - - - - -

y;<

, 3.s

-3.5/4.5 CORE AND CONTM1MEEC COOLING SYSTEMS 9

LIMITING,CO'NDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5 CORE AND CONTAINMENT COOLING 4.5 CORE AND CONTAINMENT COOLING SYSTEMS SYSTEMS Arolicability' AppliEA)_ilil2 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.

Obiettive 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 issential ' response to plant

- an essential response to plant abnormalities.

abnormalities.

Specification-Specificatl2D A.

Core'Sprav Ssstem (CS11 A.

Core Sorav System (CSS) 1.

The CSS shall be OPERABLE:

1.

Core Spray System Testing.

(1) PRIOR TO STARTUP 11gm Ergavency from a e

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.MM 3.5.A.2.

Valve OPERABILITY d.

System flow Once/$

rate: Each. months loop shall deliver at least 6250 gpm against a system head corres-ponding to a BFN 3.5/4.5-1 Unit 3 Amendment No. 130

_ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ - _ _ _ _ - - _ _ _ _ - _ _ - _ _ _ - - - - _ _ _ _ _ _ _ _ _ - _ _ _ - _ _ _ - - _ _ - - _ _ - - _ _ _ _ - _ _ _ _ _ _ - _ _ _ _ _ _ _ - _ _ _ _ ~

sq s

CORE AND CONTAINMENT' COOLING SYSTEMS

- 3 5/t. 5 SURVEILLANCE REQUIREMENTS-ILIMITING CONDITIONS FOR OPERATION-4.5 A Core Serav System _(CSS) 3.5.A

_C_ ore Soray System (CSE1 4.5.A.1.d-(Cont'd)-

105 psi differential pressure between the reactor vessel

'and the primary s

1 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-to operation for a period not operated, or exceed 7 days providing a;tomatic) in the all active cceponents 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 capable 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 pasition for another mode PRIOR TO STARTUP as of operation.

specified in 3.5 A.1.

I Amendment No. 124, 130, 140 3.5/4.5-2 BPN Unit 3

~

L_

8 4

CORE AND CQ1JAlhVENT COOLING SYSTEMS 3.5/4.5 SURVEILLANCE REQUIREMENTS LIMITING CONDITIONS FOR OPERATION Core Sorav System (CSS 1 3.5.A When irradiated fuel is in 1

5.

the reactor vessel and the reactor vessel head is removed, core spray is not required to be OPERABLE provided the cavity is l

flooded, the fuel pool gates 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 supply are OPERABLE.

l 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 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

~

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/

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 i

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 se ured in posi-tion, is in its correct

position.

g.

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

valve operator.

Except that an I

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 l

1 l

J

= - - - _.

iM1 u...,;.

3.5/4c5 CORE AND CONTAINMENT COOLING SYSTEMS SURVEILLANCE REQUIREMENT

' LIMITING CONDITIONS'FOR OPERATION 4.5.B. Residual Heat Removal Svstem 3.5.B Ersidual Heat Removal System (RHRS1 (LPCI and Containment LEHR1J (LPCI and Containment Cooling)

Cooling) 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 RHR pumps-containment cooling against an indicated system mode and associated heat pressure of 250 psig.

exchangers must rer.ain 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 he conducted once/5 suppression chamber quality A~ water test may be years.

water and filled with performed on the torus-header prima ry ; 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.

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 RHRS (LPCI mode) and the CSS and the diesel generators remain OPERABLE.

4.

No additional surveillance If any 2 RHR pumps (LPCI required.

4 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 .

l 3.5/4.5-5 Amendment No. 124, 140 BFN

. Unit 3

1_5/4.5 C' ORE AND CONTAIPCENT'C00 LING SYSTEMS 4

LIMITING CONDITIOi1 FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5 B.

. Residual Heat Removal System 4.5 B.

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

IRHRST (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 exceed 30 days provided the remaining RHR pumps

-(containment cooling mode) and associated heat exchangers and diesel generators and a11' 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 cooling mode) are OPERABLE.

7.

If two access paths of the 7.

RERS (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.

i BFN Unit 3 3.5/4.5-6 Amendment No. 124, 140 1

c o-M5/4.5 CORE AND-CONTAINMENT COOLING SYSTEMS '

f LIMITING' CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS l

(

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

. LPCI and Containment (RHRS) (LPCI and Containment

(

(RHRS) i.

Cooling)

_8. LIf 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

0LD 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 valver reactor vessel, at least one RHR that are required to be loop with two. pumps or two loops OPERABLE shall be with one pu=p 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 OPFRABLE 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 .)

l i

l BFN 3.5/4.5-7 finendment No. 130, 140 Unit 3

r,

'3.5/4.5 COR'E ARD CONTA!NMENT'COOLXNG' 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 l

Cooling)

!=

l'

12. If one RHR pump or associated 12.-No additional surveillance l~

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 for 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 not 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 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, i

EHi 3.5/4.5-8 Unit 3 Amendment Nos. 124,140

h 5/4.5 CORE AND CONTAINMENT C00 LING' SYSTEMS LIMITING CONDITIONS FOR OPERATION ltXVEILLANCE REQUIREMENTS 3.5.C EHR Service Water and' Emergency 4.5.C RHP Service Water and Emergency

' Eauirn.gSt Coolina Watf-Systems Eauipment Coolina Water Systems (EECWSY (EECWS)

I.

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 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 gps through.its normally assigned flow

path, c.

Monthly verify that cach 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. 130, 140 Unit 3

I L

6 y

3.5/4.5 ' CORE AWD CONTAINMENT COOLING SYSTEMS 1

iT LIMITING CONDITIONS FOR OPERATION' SURVEILLANCE REQUIREMENTS

3. 5. C.

RHR' Service Water and Emerzenty 4.5.C'RHR Service Water and Emergency Eaulement Cooline Water Systems Eauipment Coolina Water Systems

-(EECWS) (Continued)

(EECWS) (Continued)

2. jDuring REACTOR POWER

2. No additional surveillance OPERATION, RHRSW pumps

.is required.

must be OPERABLE and assigned to serviceoas 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 B1 and B2 normally in 4.5.C.1.

or alternately assigned to the RHR heat exchanger header supplying the standby coolant supply connection-must be 0PERABLE except as specified in 3.5.C.4 and 3.5.C.5 below.

.BTN 3.5/4.5-10 Amendment No. 124, 130, 140 Unit 3

c:

l:

s I*

i TABLE 3.5-1 Time Minimum Limit Service.issignment

( D ay_s )

RHRSW EECWC2)

(4)

(1)

.i 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 pumps 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.

BFN 3.5/4.5-11 Unit 3 I

THIS PAGE INTENTIONALLY LEFT BLANK I

i

'l I

i i

,i I

l l

l BFS 3.5/4.5-11a Amendment No. 140 Unit 3

)

l-

-.______________________j

3,.5 /4. 5 CORE __ AND CONTAINMENT COOLING SYSTRtLS j

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.C FHR Service Water and EmerRency 4.5.C RHR Service Water and Emerzenev Eguipment Cooling Water Systems Eauipment Cooling Water Systems j

(EECWS) (Continued)

IEECWS) (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 OPEEABLE 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, en 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.

l BFN 3.5/4.5-12 Amendment No. 124, 130, 140 Unit 3

3.,5 /4. 5 CORE AND C0fiTAINMENT COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5,D Eauipment Area Coolerg 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 particular 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 cocler is not OPERABLE, the pump (s) served by that cooler must be considered inoperable for Technical Specification purposes.

E.

Hinh Pressure Coolant Iniection E. High Pressure Coolant System (HPCISl 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 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 Rata at Once/3 normal months reactor l

vessel operating pressure l

l 3.5/4.5-13 Amendment No. 130 BFN Unit 3

{

L_____________.________________________.__

E, ~

.Jii/4.5' CORE AND CONTAINMENT COOLING SYSTEMS 4

L'IMITING C0NDITIONS-FOR OPERATION

~

J SURVEILLANCE REQUIREMENTS p

3.5.E p

Hinh' Pressure Coolant Iniection System (HPCIS)_

4.5.E High Pressure Coolant Iniection System (HPCISS 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 l

flow rate test.

f 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 inoperable, the reactor may

2. No additional surveillance remain in operation ~for a are required, period not-to exceed 7 days, provided the ADS, CSS, RHRS (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 Isolation Coolinz F.

System (RCICS).

Reactor Core Isolation Coolina System (RCICS) 1.

The'RCICS shall be OPERABLE:

1. RCIC Subsystem testing shall (1) PRIOR TO STARTUP from a be performed-as follows:

COLD CONDITION; or

a. Simulated Auto-Once/

matic Actuation operating Test cycle BFN 3.5/4.5-14 Amendment No. 130, 140

- Unit 3

m m-f' a,.y n

. *. u, ;

93.5/4;5' CORE-AND CONTAINMENT COOLING SYSTEMS

" LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS

3. 5. F. -

Reactor Core Isplation'Coolf.ng

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

System (RCICS)

'3.5.F.1

'(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 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 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. Ne additional sur'veillances 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.

EFN 3.5/4.5-15 Unit 3 Amendment No. 140

3 5/4.5 CORE _gD C0ffTAINMENT C00LIWG SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.G Automatic Deoressurization 4.5.G Automatic Deeressurization l

System (ADS)

System (ADS) 1.

Fou-of the six valves of 1.

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

on the ADS:

1) PRIOR TO STARTUP a.

A simulated automatic from a COLD CONDITION, actuation test shall be performed PRIOR TO or, 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 incapable of automatic operation, an immediate orderly shutdown shall be J

initiated, with the reactor in a HOT SHUTDOWN 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 vill be j

initiated and the reactor 1

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 Amendment No. 140 i

Unit 3

a 9

COPE AND CONTAINMENT COOLIRG SYSTEMS 3,$ /.L. 5 SURVEILLANCE REQUIREMENTS LIMITING CONDITIONS FOR OPERATION 4.5.H. Maintenance of Filled Discharge 3.5.H.

Maintenance of Filled Discharge Pipe Eine The following surveillance Whenever t he 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 shall 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 i

to serve the RHR and CS discharge LPCI or core spray systems l

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 below.

prior to the return of the system to service.

F1-75-20 48 psig PI-75-48 48 psig

3. Whenever the HPCI or RCIC P1-74-51 48 psig system is lined up to take P1-74-65 48 psig 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 1

3.5/4.5-17 Amendnent No. 140 BrN Unit 3 l

pi

)

3.5

~k&SEE (RHRS) 3.5.A. Core Sprav' Syster.~ (CSSS and 3.5.B Residual Heat Removal System Analyses presented in the FSAR* and analyses presented in conformance tem with 10 CPR 50, Appendix K, demonstr.ated that the core spray sys

)

provides adequate cooling'to the core to dissipate the ene d

)

t y remains temperature to below 2,200*F which assures that core geome r intact'and to limit the core average clad metal-water reaction to less Core spray distribution has been shown in tests of systems similar in design to BFNP to exceed the min than 1 percent.

half the rated flow in simulated fuel assemblies with heater ro duplicate the decay heat characteristics of irradiated fuel.

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

it system is completely independent of the core spray system; however, does function in combination with the core spray system to preventThe excessive fuel clad. temperature. core spray system provide adequate coo ble-ended approximately 0.2 square feet up to and including the do emergency core cooling subsystems.

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

service for the specified allowable repair times.

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 the diesel generators are required to be These provide OPERABLE should the need for core cooling arise.

extensive margin over the OPERABLE equipment neede cooling.

seven days was chosen.

Should one RHR pump (LPCI mode) become inoperable, three RHR pumps Since adequate (LPCI mode) and the core spray system are avail &ble. c day repair period is justified.

Should two RHR pumps (LPCI mode) become inoperable, there remains n Therefore, reserve (redundant) capacity within the RHRS (LPCI mode).sh the affected unit SAR.

A detailed functional analysis is given in Section 6 of the BFNP F 3.5/4.5-27 Amendment flo,140 EFN Unit 3

4 9

3.5 SASES (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 in excess of normal redundancy requirements, a 30-day repair equipment

. 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 suppression pool and the drywell will be sufficiently cooled,:following a loss-of-coolant accident, to prevent primary containment overpressurization. The containment cooling fur.ction of the RHRS is permitted only after the core has reflooded to.the l

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, is planned that the system be filled with demineralized (condensate) it 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 least one-half of the containment cooling function must remain low. At OPERABLE'during this time period. Requiring two OPERABLE CSS pumps during cooldown allows for ficshing 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 j

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 j

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 L_______________

~.

l k

b-Eap.ff (Cont'd)

.The suppression chamber.can be drained when.the reactor vessel pressure 4

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 reactor vessel-flange, is about 65,800. cubic feet (492,000 gallons).

-This will provide adequate low-pressure cooling in lieu of CSS and RER L(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 redundant l

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-connset 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.

EEEERENCES 1.

Residual Heat Remova3 System (BFNP FSAR subsection 4.8) 2.

Core Standby Cooling Systems (BFNP FSAR Section 6) 3.5.C.RHRServiceWalrISyster.andEmerzencyEcukomentCoolingWaterSystem (EECWS)

There are two EECW headers (north and south) with four automatic starting RHRSW pumps on each header. All c-omponents 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) and one on alternate ossignment (A1, B1, C1, or D1). One RER 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 RHR5W 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 postaccident situation.

BFN-3.5/4.5-29 Unit 3

O 3.5 P4fD (Cont'd)

The RER Service Water System was designed as a shared system for three i

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 allowed by this l

specification, a special case requert may be made to the NRC to allow continued operation if the actual system cooling requirements can be assured.

Should one of the two RHRSW pumps normally or alternately assigned to l

the RHR heat exchanger header supplying the standby co 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 disconnected ensures that h

each LPCI subsystem remains independent and a failure of the flow pat affect the flow path of the other LPCI in one subsystem will not subsystem.

3.5.D Eauipment Area Coolers area cooler for each RHR pump and an equipment There is an equipment (two pumps, either the A and C or B and D area cooler for each set area coolers take suction The equipment 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.

area coolers also remove the pump, and equipment vaste from the basement rooms housing the engineered safeguard The equipment The various conditions under which the operation of the heat equipment air coolers is required have been identified by evaluating equipment.

h the normal and abnormal operating transients and accidents over t eThe s full range of planned operations. area coolers in each of their various modes is equipment This during the testing of the equipment served by these coolers.

testing is adequate to assure the OPERABILITY of the equipment area coolers.

REFEFENCES Residual Heat Removal System (BFN FSAR Section 4.8) 1.

Core Standby Cooling System (BFN FSAR subsection 6.7) 2.

3.5/4.5-30 Amendment No. 140 BFN Unit 3

w

[

l PASES (Cont'd)

L 3 '.- 5 Iniection System (HPCIS) 0.5.E. Hir.h Pressure Coolant is adequately The'HPCIS is provided to assure that the reactor core in fuel clad temperature in the event of a small breakhich i

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 The HPCIS continues to to be shut down while maintaining suinventory until the vesse i

i s core cooling.

eperate until reactor vessel pressureLPCI operation or core s

~

his required core The capacity'of the system is selected to provide tThe HP Two sources of water are d

cooling.

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 Sy. stem begins operation, the reactorinitiated due to the rapidly than would occur if HPCI was notl id pumped into the reactor vessel condensation of steam by the cold f uAs the reactor vessel pressure continues i

decrease, the EPCI flow momentarily reaches equilibr umCon by the HPCI' system.

begins to rise.

decrease below the HPCI flow and the liquid inventory through the break.

k The core never This type of response is typical of the'small brea s.he transient so that no uncovers and is continuously cooled throughout tlie within the capacity core damage of any kind occurs for breaks that range of the HPCI.

There is adequate The minimum required NPSH for HPCI is 21 feet.

d the HPCI pump, such elevation head between the suppression pool anthe requ ion pool temperature up to_140*F with no containment back pressure.

that of feedwater The HPCIS serves as a backup to the RCICS as a sourceThe ADS serves 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 t i time necessary.

of seven days was selected.

Cooling Systems The HPCI and RCIC as well as all other Core Standby dition.

It is must be OPERABLE when starting up from a Cold Conth full capacity realized that

-until reactor pressure exceeds 150 psig an sure HPCI turbine is automatically isolated before the reactor pres d up from a Cold Condition, It decreases below 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 BIN Unit 3

-~~

3.5 t

BASES;(Cont'd) 3.5.F Reactor Core Isolation Cooling System (RCICS)

The'various conditions under which the RCICS plays an essenti l evaluating.the various plant events over the full ran a role in y

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

the NPSH for RCIC is 20 feet.

The minimum required suppression pool and the RCIC pump, such that the required l'

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) a

capable of providing all the cooling required.for any plant event when re nuclear system pressure is below 122 psig,.the RCICS is not required below this pressure.

provide its design flow, but reduced flow is required'for certai events.

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

eedwater risk associated with failure of the RCICS to cool the erage

, required is not increased if the RCICS is inoperable for-no longer than seven days, provided that the HPCIS is OPERABLE during this period REFEREllCE 1.

Reactor Core Isolation Cooling System (BFNP FSAR Subsection 4 7) 3.5.G. Automatic Deoressurization System (ADS 1 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 nucl the low-pressure coolant injection (LPCI) and the core sp ear can operate to protect the fuel barrier.

ay subsystems applies only to the automatic feature of the pressure relief systemNote tha Specification 3.6.D specifies the requirements for the pressure function of the valves.

relief 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 perfor i their pressure relief function.

m ng BTN Unit 3 3.5/4.5-32

p t..;*

3.5 MSIS '(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 valves inoperabic 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.

11. Maintenance of Filled Discharge Ploe l

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, j

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 flos 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 Averane Planar Linear Heat Generation Rate (MAPLHGR1 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 3.5/4.5-33 Unit 3 Amendment No. 140

,,m

' A Y 13.5? E6115 (Cont'd) l t

The peak; cladding temperature following.a postulated loss-o d is only accident

.11 the rods of a fuel assembly at any axial location an dependent' secondarily on the rod-to-rod-power distribution by less than 1 assembly fuel asLembly af fect the calculated! peak clad terpr*:t;;al design, the 20*F-relative to tbe peak temperature for a typical fue limit on the average linear heat generation rate is sufficient to assure limit.

that calculated temperatures are within the 10 CFR 50 Appendix K The The limiting value for MAPLHGR is shown in Tables 3.5.I-1 through 7.

1.

l analfses supporting these limiting values are presented in Reference 3.5.J. Linear Heat staeration Rate (LMEEl 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.

25 percent The LHGR shs2.1 be checked daily during reactor operation at 1 power to determine if fuel burnup, or control rod movement has caus changes in power distribution. rated thermal power, the MTPF would have to be g which is precluded by a considerable margin when employing any percent

-permissitte control rod pattern.

3.5.K. Minirst_gj tical Power Ratio (MCPll the Ac core thermal power levels less than or equal to 25' percent, reactor will be operatina at minimum recirculation pump speed and theFor a moderator void content will be very small.

operating plant rod patterns which may be employed at this point, experience iderable margin. With

.MCPR value is in excess of requirements by a consany inadvertent core flow The daily this-low void content, place operation in a more conservative mode relative,to MCPR.

requirement for calculating MCPR above 25 percent The requirement for not been significant powar or control rod changes.

calculating MCFR when a limiting control rod pattern is approached wer ensures that MCFR will be known follo;ing a change in power or poration at a thermal l

. shape (regardless of mcgnitude) that tould p ace ope limit.

,L. 6ERM_fettseinta Operation is constrained to a maximum LHGR cf 18.5 xW/f t for 7x7 fuel 13.4 kW/ft for 8x8, 8x8R, and pbx 8R.

This limit is reached when core For the maximum fraction of limiting power density (CMFLPD) equals 1.0.

case where CMFLPD exceeds the fraction of rated thermal power, operation l

is permitted only at less than 100-percent rated power and only with APR The scram trip scram settings as requirGd by Specification 3.5.L.1.

setting and rod block, trip setting are adjusted to ensure that no i nt peak

.. combination of CMFLPD and FRP will increase the LHGR trans e 3.5/4.5-34 BFN l

l Unit'3 l

I I

3.$

BASES (Cont'd) i beyond that allowed by the one-percent plastic strain limit. A six-hour time period to achieve this condition is justified since the f

additional margin gained by the setdown adjustment is above and beyond l

that ensured by the safety analysis.

j 3. 5,. M References 1.

Loss-of-Coolant Accident Analysis for Browns Ferry Nuclear Plant i

Unit 3, NEDD-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.

BFN 3.5/4.5-35 Unit 3 Amendment Nos. 118, 140

+

l Cooline Systems Surveillance Frequencies 4.5 Core and Containment The tetting 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 Complete ADS water into the reactor vessel which is not desirable.

j E

testing during power operation causes an undesirable loss-of-coolant To increase the availability of the core end containment inventory.

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

The pumps instrumentation, pumps, valves, etc., are 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 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 scaled in position which affect the ability of the systems to perform their intended safety function are also verified to be i

Valves which automatically reposition themselves on an initiation signal are permitted to be in a position other than f

in the proper position.

normal to facilitate other operational modes of the system.

When coepenents and subsystems are out-of-service, overall core and containment cooling reliability is 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 If the function, system, or no reason to suspect they are inoperable.

or calibration is found inoperable or txceeds the trip loop under test level setting, the LCO and.the required surveillance testing for the system or loop shall apply.

Maximum Averas;e Planar LEGB. LHGR. and MCPR The MAPLHGR, LHGR, and MCPR shall be checked daily to determine if fuel burnup, or control rod movt.nent has caused changes in power Since changes due to burnup are slow, and only a few distribution.

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

3.5/4.5-36 Amendment to. 130, 140 BEN Unit 3

__ _ _ -._______.__:

Person / Time
ML20245G212
Site:	Browns Ferry (DPR-33-A-169-01, DPR-33-A-169-1, DPR-52-A-169, DPR-68-A-140)
Issue date:	08/02/1989
From:	Black S Office of Nuclear Reactor Regulation
To:
Shared Package
ML20245G217	List: ML20245G212 ML20245G219
References
NUDOCS 8908150274
Download: ML20245G212 (118)
v • d • e

ML20245G212

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