Text

Amends 180,190 & 152 to Licenses DPR-33,DPR-52 & DPR-68, Respectively,Revising Sys Operability Requirements for Cold Shutdown Conditions
	ML20067E288
Person / Time
Site:	Browns Ferry
Issue date:	02/07/1991
From:	Hebdon F; Office of Nuclear Reactor Regulation
To:	;
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ML20067E291	List: ML20067E288;
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	NUDOCS 9102140141
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o, UNITED STATES 8"

NUCLEAR REGULATORY COMMISSION n

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,a WASHINGTON, D. C 20555

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TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-259 BROWNS FERRY NUCLEAR PLANT, UNIT 1, AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.180 License No. OPR-33 1.

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

A.

The application for amendment by Tennessee Valley Authority (the licensee) dated May 18, 1990, as superseded by your letter of October 30, 1990, complies with the standards and requirements of theAtomicEnergyActof1954,asamended(theAct),andthe Commission's rules and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 (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.

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

9102140141 910207 I

PDR ADOCK 05000259 l

P PDR

. o 2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendnent 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.180, 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 30 days from the date of issuance.

FOR THE NUCLEAR REGULATORY C0f7,ISSION

- gMC. h Fredcrick J. Hbdon, Director Project Directorate 11-4 Division of Reactor Projects - 1/II Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical

~ Specifications Date of Issuance: February 7, 1991 m

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ATTACHMENTTOLICENSEAfEj@HEl4TNO.180 FACILITY OPEPATING 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 are identified by the captioned amendment number and contain marginal lir,es indicating the area of change. Overleaf

and spillover** pages are provided to maintain document completeness.

RD10VE INSERT 3.2/4.2-14 3.2/4.2-14

.3.2/4.2-15 3.2/4.2-15 3.2/4.2-23 3.2/4.2-23*

3.2/4.2-24 3.2/4.2-24 3.5/4.5-7 3.5/4.5-7 3.5/4.5-8 3.5/4.5-8*

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 7

3.5/4.5-17 3.5/4.5-17**

3.6/4.6-9 3.6/4.6-9*

3.6/4.6-10 3.6/4.6-10.

t 3.6/4.6-30 3.6/4.6-30*

i j

3.6/4.6-31 3.6/4.6 i.

3.6/4.6.3.6/4.6-32**

3.6/4.6-33 3.6/4.6-33**

I-.~..-,..-..~._._.a._._._

i i

i TABLE 3.2.5 i

y IMSTRUMENTATION THAT INITATES OR CONTROLS THE CORE AND CONTAINMENT COOLING SYSTEMS

' Minimum No.

i c to Operable Per j-EE Trio Sys(11 Function Trio Level Settine Action Remarks e,

.2 Instrument Channel l 470* above vessel zero A'

1.

Below trip setting initiates Reactor Low Water Le-el HPCI.

2 Instrument Channel' -

2, 470" above vessel zero.

A 1.

Multiplier relays initiate Reactor Low Water Level RCIC.

i 2

Instrument Cha w l -

1 378 above vessel zero.

A 1.

Below trip setting initiates

. Reactor Low Water Level CSS.

(LIS-3-58A-D, SW #1)-

Multiplier relays initiate I

LPCI.

i 2.

Multiplier relay from CSS initiates accident sirutl, (15).

2(16)

Instrument Channel'-

1 378" above vessel zero.

A 1.

Below trip settings, in Reactor Low Water Level conjunction with drywell (LIS-3-58A-0, SW #2) high pressure, low water

.w).

level pemissive.120 sec.

i

+

delay timer and CSS or

)

y RHR pump running. initiates t

ADS.

[

-^

p 1(16)

Instrument Channel -

1 544" above vessel zero.

A 1.

Below trip setting p* missive i

Reactor Low Water Level for initiating signals on ADS.

Permissive (LIS-3-184 &

j 185 SW #1)

'l Instrument Channel -

- 2 312 5/16* above vessel zero. A 1.

Below trip setting prevents

- Reactor Low Water Level (2/3 core height) inadvertent operation of 3

(LITS-3-52 and 62, SW #1) containment spray during i

,E3 accident condition.

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i TABLE 3.2.3 (Continued) c txt Minimum No.

E. y Operable Per r

Trio Syst1) r nction Trio Level settino Acti on Remark s u

2(18)

Instrument Channel -

11 p12.5 psig A

1.

Below trip setting prevents f

~~

Drywell High Pressure inadvertent operation of (PS-64-58 E-H) containment spray during accident conditions.

2(18)

Instrument Channel -

1 2.5 psig A

1.

Above trip setting in con-Drywell High Pressure junction with low reactor (PS-64-58 A-0, SW F2) pressure initiates CSS.

Multiplier relays initiate HPCI.

2.

Multiplier relay from CSS initiates accident signal. (15) 2(18)

Instrument Channel -

1 2.5 psig A

1.

Abeve trip setting in g

Drywell High Pressure conjunction with low (PS-64-5BA-0, $W #1) reactor pressure initiates Id LPCI.

2(16)(18)

Instrument Channel -

1 2.5 psig A

1.

Above trip setting, in l

Drywell High Pressure conjunction with low reactor (PS-64-57A-0) water level, drywell high pressure,120 sec. delay timer and CSS or RHR pump running, initiates ADS.

<a 3C1 B

(D DO

NOTES FOR TABLE 3.2.B 1.

Whenever any CSCS System is required by Section 3.5 to be OPER13LE, there shall be two OPERABLE trip systems except as noted.

If a requirement of the first column is reduced by one, the indicated action shall be taken.

If the same function is inoperable in more than one trip system or the first coltum reduced by more than one, action B shall be taken.

l Action!

A.

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

If the function la not OPERABLE in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , take action D.

B.

Declare the system or component inoperable.

C.

Innediately take action B until power is verified on the trip system.

D.

No action required; indicators are considered redundant.

2.

In only one trip system.

3.

Not considered in a trip system.

4.

Requires one channel from each physical location (there are 4 locations) in the steam line' space.

5.

With diesel power, each RHRS pump is scheduled to start immediately and each CSS pump is seguenced to start about 7 sec. later.

6.

With norn4a1 power, one CSS and one RHRS pump is scheduled to start instantaneously, one CSS and one RHRS pump is sequenced to start after about 7 sec.. With similar pumps starting after about 14 sec. and 21 sec.,

at which time the full complement of CSS and RHRS pumps would be operating.

7.

The RCIC and HPCI steam line high flow trip level settings are given in terms of differential pressure. The RCTAS setting of 450" of water corresponds _to at least 150: percent above maximum steady state steam flow to assure tit.at'apprious isolation does not occur while ensuring the initiation of isolation' following a postulated steam line break.

Similarly, the HpCIS setting of 90 psi corresponds to at least 150

~

. percent above xaximum steady state flow while also ensuring the initiation of isolation following a postulated break.

8.

Note 1 doen 7 ot apply to this item.

9.

The head t0r.h ir designed to assure that the discharge piping from the CS and RHR purpo sre full. The pressure chall be maintained at or above the values lititad in 3.5.H, which ensures water in the discharge piping and up to the bc4 tank, BFN 3.2/4.2-23 Unit 1

l.

1 NOTES FOR TABLE 3.2.B (Cont'd) 10.

Only one trip system for each cooler fan.

11.

In only two of the four 4160-V shutdown boards.

See note 13.

12.

In only one of the four 4160-V shutdown boards.

See note 13.

13.

An emergency 4160-V shutdown board is considered a trip system.

14.

RHRSW pump would be inoperable.

Refer to Section 4.5.C for the requirements of a RHRSW pump being inoperable.

15. The accident signal is the satisfactory completion of a one-out-of-two taken twice logic of the drywell high pressure plus low reactor pressure or the vessel low water icvel (1 378" above vessel zero) originating in the core spray system trip system.

16.

The ADS circuitry is capable of accomplishing its protective action with i

one OPERABLE trip system. Therefore, one trip system may be taken out of service for functional testing and calibration for a period not to exceed eight hours.

17. Two RPT systems exist, either of which will trip both recirculation pumps.- The systems will be individually functionally tested monthly.

If the test period for one RPT system exceeds two consecutive hours, the system will be declared inoperable.

If both RPT systems are inoperable or if one RPT system is inoperable for more than 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , an orderly l

power reduction shall be initiated and reactor p(wer shall be less than i

30 percent within four hours.

l

19. Not required to be OPERABLE in the COLD SifDTDOWN CONDITION.

L BFN 3.2/4.2-24 Amendment 180 Unit 1

L5/4.5 CORE AND CONTAI}l MENT _ COOLING SYSTEMS LIMITINC CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.B R111 dual Heat Removal System 4.5.B Residual Heat Removal System fRHRS) (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 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 vascel i

pressure is atmospheric and pressure is atmospheric, irraciated fuel is in the the RHR pumps and valves reactor vessel, at least one

' chat are required to be

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

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 is not in the COLD SHUTDOWN cross-connect capability CONDITION, 2 RHR pumps and shall be demonstrated to be associated heat exchangers and OPERABLE per Specification valves on an adjacent unit 1.0.MM when the cross-must be OPERABLE and capable connect capability of supplying cross-connect is required, capability except as specified in Specification 3.5.B.12 below.

(Notes 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 .)

9 9

BFN 3.5/4.5-7 Amendment 180 Unit 1

1 3.5/4.5 CORE AND CONTAINMENT COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.B Rtaidual Heat Removal System 4.5.B Residual Heat Removal System IRHRS) (LPCI and Containment IRHRS) (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 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 RilR 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 70 be tested for OPERABILITY STARTUP (or closed if during any period of permitted elsewhere COLD SlfUTDOWN CONDITION in these specifications),

exceeding 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , if OPERABILITY tests have not been performed during the preceding 31 days.

BFN 3.5/4.5-8 Unit 1

3.5/4.5 CORE AND CONTAIMMENT COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5 C RHR Service Wgler and Emergency 4.5.C RHR Service Water and Emergency Eaulement Coolinn Water Systems Ecgingent Coolina Water SysLima

- (EECWS) (Continued)

(EECWS) (Continued) 4.

One of the D1 or D2 RHRSW 4.

No additional surveillance pumps assigned to the RHR is required.

heat exchanger supplying the standby coolant supply connection may be h

inoperable for a period not to exceed 30 days provided the OPERABLE pump is aligned to supply the RiiR 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 j

unit placed in the COLD S!!UTDOWN 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 Rl!RSW pumps, associated with the selected EllR pumps, aligned for RHR heet exchanger service (or each reactor vess21 containing irradiated fuel.

BFN 3.5/4.5-12 AMENDMERT No.16 9 Unit 1

.__m_.-_- _ _ _ _.

3.5/4.5 CORE AND CONTAINMENT COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.D Eautoment Area Coolers 4.5.D Eautoment 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 Intec112D E.

Elah Pressure Coolant System (HPCIS)

Iniection System (HPCIS) 1.

The HPCI system shall be 1.

HPCI Subsystem testing OPERABLE whenever there is shall be performed as 1rradiated fuel in the follows:

reactor vessel and the t

reactor vessel pressure a.

Simulated-Once/18 is greater than 150 psig, Automatic months except in the COLD SHUTDOWN Actuation j

CONDITION or as specified-in Test Specification 3.5.E.2.

OPERABILITY shall be deter-b.

Pump Per j

mined within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after OPERA-Specification reactor steam pressure BILITY 1.0.MM reaches 150 pais from a COLD CONDITION, or alternatively c.

Motor Oper-Per PRIOR TO STARTUP by using an ated Valve Specification

-i auxiliary steam supply.

OPERABILITY 1.0.MM t

d.

Flow Rate at Once/3 normal months reactor vessel

+

operating pressure

. BFN 3.5/4.5-13 Amendment 180 Unit 1-l

3.5/4.5 CORE AND CONTAINMENT COOLING SYSTEMS LIMITING CONDI210NS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.E lilah Pressure Coolant In11AliRD 4.5.E lifeh Prigsure Coolant Iniection System (HpCIS) gyltem (HPCIS) 4.5.E.1 (Cont'd) e.

Flow Rate at Once/18 1

150 psig months The HPCI pump shall deliver et 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 correcta position.

2.- If the !!PCI system is 2.

No additional surveillances inoperable, the reactor may are required, remain in operation for a period not to exceed 7 days, provided the ADS, CSS, RHRS 4

(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 an orderly shutdown shall automatic return to its be initiated and the ECCS position when an reactor vessel pressure ECCS signal is present shall be reduced to 150 may be in a position for psig or less within 24 another mode of

hours, operation.

F.

Reactor Core Isolation Coolina F.

Reactor Core Isolation CooliDA Evstem (RCICS)

System (RCICS) 1.

The RCICS shall be OPERABLE 1.

RCIC Subsystem testing shall

-whenever there is irradiated be performed as follows:

fuel in the reactor vessel and the reactor vessel a.

Simulated Auto-Once/18 pressure is above'150 psig, matic Actuation months except-in the COLD SHUTDOWN.

Test CONDITION or as specified in 3.5.F.2.

OPERABILITY ahall BFN 3.5/4.5-14 Amendment 180 Unit 1 l.

3,5/4.5 CORE AND CONTAINHENT COOLING SYSIIMS LIMITIl1G CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS

\\

3.5.F Reactor Core Isolation Coolina 4.5.F R.eactor Core Isolation Coolina System (RCICS)

System (RCICS) 3.5.F.1 (Cont'd) 4.5.F.1 (Cont'd) be determined within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

b. Pump Per after reactor steam pressure OPERABILITY Specifi-reaches 1b0 psig from a COLD cation CONDITION or alternatively 1.0.MM PRIOR TO STARTUP by using an auxiliary steam supply,

c. Motor-Operated Per l

Valve Specifi-OPERABILITY cation 1.0.MM d.

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

Flow Rate at Once/18 150 psig months 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 wise secured in orderly. shutdown shall be position, is in its initiated and the reactor correct

position, shall be depressurized to less than 150 psig within

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

are required.

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

BFN 3.5/4.5-15 AMEN 0 MENT NO.17 3 Unit 1

3.5/4.5 CORLAND CONTAINMENT COOLING SYSTElig -

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE, REQUIREMENTS 3.5.G

&gtomatic Deoressurization 4.5.C Automatic Depressurization Eyalgm ( ADS) system (ADEl 1.

Four of the six valves of 1.

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

on the ADS:

(1) PRIOR TO STARTUP from a.

A simulated automatic a COLD CONDITION, or, actuation test shall be performed PRIOR.TO (2) whenever there is STARTUP after each irradiated fuel in the refueling outage.

reactor vessel and the Manual surveillance reactor vessel pressure of the relief valves is greater than 105 psig, is covered in except in the COLD SilUT-4.6.D.2.

DOWN CONDITION or as specified in 3.5 G.2 and 3.5.G 3 below.

2.

If three of the six ADS 2.

No additional surveillances 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 or 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 SKUTDOWN CONDITION in the following 18 hourt.

3.

If Specifications 3.5.G.1 and 3.5.G.2 cannot be met, an orderly shutdown will be BFN 3.5/4.5-16 Amendment 180 Unit 1

a i

3.5/4.5 CORE AHD CONIAINMENT CQD(ING SYSTEMS LIMITINC CONDITIONS FOR OPERATION SURVEILLANCE REQUIREKENTS 3.5.G Automatie Degrinagrizatlan 4.5.C Automatic De1Jessurization System thD Svetem (ADS) 3.5.G.3 (Cont'd) initiated and the reactor vesse!' pressure shall be reduced to 10S psig or less within 24 hours.J / H. Maintenance of Filled Discharr.e H. Maintenance of Filled Dischargg ElRA P_its 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 pi) 9g from the pump dischstge piping of the core spray of onese systems to the last systems, LPCI, HPCI, and RCIC block valve shall be filled. are filled: The auction 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 suppreselon chamber spray system, the discharge head tank shall normally be piping of these systems shall aligned to serve the discharge be vented from the high point piping of the RHR and CS pumps, and water flow determined. The condensate head tank may Le used to serve the RHR and CS 2. Following any period where the discharge piping if the PSC head LPCI or core spray systema tank is unavailable. The have not been required to be pressure indicators on the OPERABLE, the discharge piping discharge of the RHR and CS of the inoperable system shall pumps shall indicate not less be vented from the high point than listed below. prior to the return of the system to service. PI-75-20 48 psig P1-75 48 48 psig 3. Whenever the HPCI or RCIC P1-74-51 48 psig system is lined up to take F1-74-65 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. 3.5/4.5-17l Amendment 180 BFN Unit 1

1 3.6/4.6 PRIMARY SYSTEM BOUNDARY t ;: LINITING CONDITIONS FOR OPERATION-StDNIGLLANCE REQUIREMENTS - 3.6.C. Coolant Leakang 4.6.C. GJL2 &nt Leakane 1 1. a. Any time irradiated T. IwdetoT coolart fuel is in the system leakage 2 hall reactor vessel and be checked by the reactor coolant susp and air sampling temperature is above (;ystem and recorded 212'F, reactor coolant at least once per leakage into the 4 hours. primary containment from unidentified sources shall not exceed 5 spu. - In addition, I l$ the total reactor-coolant System-1eakage into the primary containment shalA not exceed 25 spm. b. Anytime the reactor is in l RUN MODE,2 reactor coolant i leakage into the primary containment from. unidentified sources shall not increase by more than 2 spa averaged over any 24-hour period in which the reactor is in the RUN MODE except as defined in 3.6.C.1.c below. c. During the first 24 hours in the RUN MODE following. STARTUP, an increase in reactor coolant leakage into the primary containment of >2 spa is acceptable as long as the requirewenta of 3.6.C.1.a are met. 1 BFN 3.6/4.6-9 Unit 1-J'..

3.6/4.t PRIMARY SYSTEM BOUNDARY l LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.6.C faolant Leakagt 4.6.C C921ATA Leakane i 2. Both the sump ascd air sampling 2. With the air sampling systems shall be OPERA 2LE system inoperable, grab during REACTOR POWEP f*sERATION. samples shall be From and after the date that-obtained and analyzed at one of these systems is made least once every 24 or found to be inoperable for

hours, any reason, REACTOR POWER OPERATION is permissible only during the succeeding 24 hours for the sump system or 72 hours for the air sampling system.

The air sampling system may be removed from service for a . period of 4 hours for calibration, function testing, and maintenance without providing a temporary monitor. 3. If the condition in 1 or 2 above cannot be met, an 5 orderly shutdown shall be initiated and the reactor shall be placed in the COLD SHUTDOWN CONDITION within 24 hours. D. Relief Valyga D. Relief Valveg 1. When more than one relief valve 1. Approximately one-half is known to be failed, an of all relief valves orderly shutdown shall be shall be bench-checked initiatad'and tha reactor or replaced with a depressurized to less than 105 bench-checked valve psig within 24 hours. The each operating cycle, rel.ief valves are not required All 13 valves vill have to be OPERABLE in the COLD been checked or replaced- -SHUTDOWN CONDI' 7N. upon the completion of every second cycle. 2. In accordance with Specification 1.0.MM, each relief valve shall be manually opened until q thermocouples and acoustic monitors t downstream of the valve indicate steam is -flowing fror the valve. BFN 3.6/4.6-10 Amendment 180 Unit 1 l l

C; 3.6/446 BASES
3.6.C/4'6.C (Cont'd) suggest 'a reasonable margin of safety: that such leakage magnitude-would R

not result from'a crack-approaching the critical size for_ rapid 4 propagation.m. Leakage 1ess than the magnitude _specified can be detected j ~ reasonably-_in's matter-of a few hours utilizing-the available leakage j 'I detection schsaes, and if the origin cannot be determined.in'a reasonably -short time,--the unit should be shut down to allow further investigation" i and corrective action. The two spa limit for-coolant-leakage rate increase over any 24 hour ' period-is-a_ limit specified by the NRC-(Reference _2). -This' limit applies' j only during the RUN_ mode to avoid being penalized for the expected 1 J' coolant _leakagetincrease during pressurization._ The total leaks 3e. rate _ consists of all leakage, identified and unidentified, which flows to the drywell floor drain and equipment drain sumps. The capacity!of the drywell floor sump pump is 50 gpm and the capacity of ~ i the=drywell equipment sump pump is alsor50 spm.- Removal-of 25 spw from either of these sumps can be -accomplished with considerat'e margin.. REFERENCE

1. Nuclear System Leakage Rate Limits (BFNP FSAR Subsection'4.10)

2. Safety Evaluation _ Report-(SER) on IE Bulletin-82-03

-3.6.D/4.6.D = Relief Valves To' meet--.the_ safety basis, 13 relief valves-have been installed on the ) unit with-a total' capacity-of 84.1 percent.of nuclear boiler rated steam: 1 tflow atna.. reference pressure.cf'(1=,105 + 1 percent) psig. The analysis of: the worst -overprissure transient,:(3-second closure of all main steam. -line isolation: valves) neglecting the direct scram (valve position scram)= results;in a maximum' vessel pressure which,-if a' neutron-flux scram is assumed considering'12 valves operable, results in adequate margin'toLthe code' allowable-overpressure limit of 1,375 psig. To meet operational. design, the analysis 1of the plant isolation transient-(generator load:re. ject with-bypass valve-failure to open) shows that -12_of the=13 relief valves limit peak system pressure to a-value which ja well below the allowed vessel overpressure'of 1,375 psig.- Experience in relief. valve operation showsithat=a testing of 50 percent 7' of.the-valves per-year is. adequate to_ detect. failures or deterioraticas.- 1 >The: relief. valves are benchtested.every~second operating.. cycle to er.sure that.their setpoints-are within the i~1 percent tolerance. The relief valves.are. tested in place:in accordance with Specification 1.0.MM to establish that they'will2 open and pass steam.. A M 80' I 7 = BFN 3.6/4.6-30 Unit 1 l

9.6/4.6 BASES 3.6.D/4.6.D-(Cont'd) -The requirements established above apply when the nuclear system can be pressurized above ambient conditions. These requirements are applicable at nuclear system pressures below normal operating pressures because abnormal operational transients could possibly start at these conditions such that eventual overpressure relief would be:needed. However, these i transients are much less severe, in terms of pressure, than those starting.at rated conditions.' The valves need not be functional when the vessel head is removed, since the nuclear system cannot be pressurized. l The relief valves are not required to be OPERABLE-in the COLD SHUT 00WN CONDITION. Overpressure protection is provided during hydros..ciu tests by two of the relief valves whose relief setting has been established in conformance with ASME Section XI code requirements. The capacity of one relief velve exceeds the charging capacity of the pressurization cource used during hydrostatic. testing. Two relief valves-are used to provide redundancy. REFERENCES 1. ' Nuclear System Pressure Relief System (BFNP FSAR Subsection 4.4) i I 2. Amendment 22 in response to AEC Question 4.2 of December 6, 1971. 3. " Protection Against Overpressure" (ASME Boiler and Pressure Vessel Code, Section III, Article 9) 4. Browns Ferry Nuclear Plant Design Deficiency Report--Target Rock Safetv-Relief Valves, transmitted by J._E. Gilliland to_F. E. Kruesi, August 29,'1973 5. Generic Reload Fuel Application, Licensing Topical Report, NEDE-24014-P-A and. Addenda 3.6.E/4.6.E -Jet Pumos Failure of a jet pump nozzle assembly holddown mechanism, nozzle assembly and/or riser, would increase the crosa-sectional flow area for blowdovn following the, design basis double-ended line break.: Also, failure of the diffuser ~would eliminate'the capability to reflood the core to two-thirds height level following a-recirculation line break. Therefore, if a failure-occurred, repairs must be made. The detection technique is as follows. With the two recirculation pumps balanced'in speed to within 1 5 percent, the flow rates in both recirculation loops will be verified by control room monitoring instruments.- ~If the-two flow-rate values do not differ by more than 10 percent,' riser and nozzle assembly integrity has been verified. BFN 3.6/4.6-31 Amendment 180 Unit 1

3.6/4.6 Mifji 3.6.E/4.6.E (Cont'd) If they do differ by 10 percent or more, the core flow rate measure by d the jet pump diffuser differential pressure system must be checked against the core flow race derived from the measured values of loop flow to core flow correlation. If the difference between measured and derived core flow rate is 10 percent or more (with the derived value higher) diffuser measurements will be taken to define the location within the vessel of failed jet pump nozzle (or riser) and the unit shut down for repairs. If the potential blowdown flow area is increased, the system resistance to the recirculation pump is also reduced; hence, the affected drive pump will "run out" to a substantially higher flow rate (approximately 115 percent to 130 percent for a single nozzle failure). If the two loops are balanced in flow at the same pump speed, the resistance characteristics cannot have changed. Any imbalance between drive loop flow rates would be indicated by the plant process instrumentation. In addition, the affected jet pump would provide a leakage path past the core thus reducing the core flow rate. The reverse flow through the inactive jet pump would still be indicated by a positive differential pressure but the net effect would be a slight decrease (3 percent to 6 percent) in the total core flow measured. This decrease, together with the loop flow increase, would result in a lack of correlation between measured and derived core flow rate. Finally, the affected jet pump diffuser differential pressure signal would be reduced because the backflow would be less than the normal forward flow. A nozzle-riser system failure could also gcaerate the coincident failure of a jet pump diffuser body; however, the converse is not true. The lack of any substantial stress in the jet pump diffuser body makes failure impossible without an initial nozzle-riser system failure. 3.6.F/4.6 F RecirculALisa Pumo Operation Steady-state operation without forced recirculation will not be permitted for more than 12 hours. And the start of a recirculation pump from the natural circulation condition will not be permitted unless the temperature difference between the loop to be started and the core coolant temperature is less than 75'F. This reduces the positive reactivity insertion to an acceptably low value. Requiring the discharge valve of the lower speed loop to remain closed unt31 the speed of the faster pump is below 50% of its rated speed provides assurance when going from one-to-two pump operation that excessive vibration of the jet pump risers will not occur. 3.6.G/4.6.G St ructural Integrity The requirements for the reactor coolant systems inservice inspection program have been identified by evaluating the need for a sampling examination of areas of high stress and highest probability of failure in the system and the need to meet as closely as possible the requirements of Section XI, of the ASME Boiler and Pressure Vessel Code. 3.6/4.6-32l Amendment 180 BFN Unit 1 l

3.6/4.6 BASES l '3.6.C/4.6.G (Cont'd) The program reflects.the built-in limitations of access to the reactor coolant systems. It is intended _that-the' required examinations and inspection be completed 1 -during each 10-year interval. The periodic examinations are to be done during refueling outages or other extended plant shutdown periods. Only proven nondestructive testing techniques will be used. More. frequent inspections shall be performed on certain circumferential pipe welds as listed in Section 4.6.G.4 to provide additional protection against pipe _ whip. These welds were selected in respect to their distance from hangers.or supports wherein a failure of the weld would permit the unsupported segments of pipe to strike the_ drywell vall or i nearby auxiliary systems or control systems. Selection was based on . judgment from actual plant observation of hanger and support locations and review of drawings. Inspection of all these weld:s during each 10-year inspection interval will result in three additional examinations above the requirements of Section XI of ASME Code. An augmented inservice surveillance program is required to determine whether any stress corrosion has occurred in any stainless steel piping, stainless components, and highly-stressed alloy steel such as hanger springs, as a result of environmental conditions associated with the March 22, 1975 fire. REFERENCES

1. - Inservice' Inspection and Testing (BFNP FSAR Subsection 4.12) 2.

Inservice Inspection of Nuclear Reactor Coolant Systems, Section XI, ASME Boiler and Pressure Vessel Code 3. ASME Boiler and Pressure Vessel Code, Section III (1968 Edition) 4. American Society for Hondestructive Testing No. SNT-TC-1A (1968 Edition) 5. Mechanical' Maintenance Instruction 46 (Mechanical Equipment, Concrete, and Structural Steel Clean ( g Procedure.for' Residue From Plant Fire - Units I and 2) 4 6. Mechanical Maintenance Instruction 53 (Evaluation of Corrosion Damage-of Piping Components Which Were Exposed to Residue From March 22, 1975 Fire)-

7.. Plant Safety-Analysis (BFNP'FSAR Subsection 4.12)

- BFN: 3.6/4.6 33 Amendment 180 Unit 1

/ g-UNITED STATES 8, s NUCLEAR REGULATORY COMMISSION 3' , j: WA$HINGTON. D. C. 20656. \\...../ TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-260 BROWNS' FERRY NUCLEAR PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING' LICENSE Amendment No.190 License No. DPR-52 1. - The Nuclear Regulatory Comission (the Comission) has found that: . A.: The application for amendment by Tennessee Valley Authority (the. ' licensee) dated May. 18, 1990, as superseded by your letter of. October 30, 1990,: 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 4 provisions of the.Act, and the rules and regulations of the Commission;- C. LThere-isreasonableassurance(1)thattheactivitiesauthorizedLby this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be 2 s conducted.in compliance with the Comission's regulations; D. The issu'ance of this amendment wiil not be inimical to the common ~ defense and security or to the health and safety of the public; and E. The issuance of this amendment'is in accordance with '10 CFR Part 51! of.the Comission's regulations and all' applicable requirements'have .been satisfied. 1 i 1 >wv sp w r e -e,, v a ,, - ~., e -n--

.s 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 Operatirg License No. DPR-52 is hereby amended to read as follows: (2) Technical Specifications-The Technical Specifications contained in Appendices A_and B, as revised through Amendment No.190, are hereby incorporated in the license. The licensee shall-operate the facility in accordance with the Technical Specifications. 3. This license t.mendment is effective as of its date of issuance and shall be.implerented within 30 days from the date of issuance. FOR THE NUCLEAR REGULATORY COMMISSION

0

~ Frederick J. Heb on, Director Project Directorate 11-4 Division of Reactor Projects - 1/II Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: February 7, 1991 I I t [ 1

_... -.- - -. -._ - ---.._- ~ 1 ATTACHMENT TO LICENSE AMENDMENT N0.190 FACILITY OPERATIMG LICENSE NO. DPR-52 a 1 ' DOCKET H0. 50-260 4 Revise the Appendix A Technical Specifications by removing the pages -identified belcw and inserting the enclosed pages. The revised pages -are identified byithe -captioned amendment number and contain marginal lines indicatire the area of change. Overleaf

and spillover** pages

are provided to m6intain document conipleteness.

REMOVE INSERT = 3.2/4.2-14 3.2/4.2-14 '3.2/4.2-15 3.2/4.2-15 '3.2/4.2 3.2/4.2-16* 3.2/4.2-17 3.2/4.2-17 3.2/4.2-23 3.2/4.2-23* ' 3.2/4.2-24 3.2/4.2-24 3;5/4.5-7. 3.5 /4.5 -7 3.5/4.5 -8 3.5/4.5-8* ~ 'l 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 4 13.5/4.5-17' '3.5/4.5-17*= 3.6/4.6-9 3.6/4.6-9* 3.6/4.6-10 3.6/4.6 _ 3.6/4.6 3.6/4.6-30* 3.6/4.6-31 3.6/4.6-31 3.6/4.6-32 3.6/4.6-32** 3.6/4.6-33 3.6/4.6-33* l l l sm-

TABLE 3.2.8 ' INSTR'JMENTATION THAT INITIATES OR CONTROLS THE CORE AND CONTAINMENT COOLING SYSTEMS - . f If, Minimum No.' .- z Operable Per .. Trio Sys(1) Function Trio Level Settino Action Remarks N '2 Instrument Channel - 1470" above vessel zero. .A 1. Below trip setting initiates l Reactor low Water Level HPCI. -(LIS-3-58A-D) 2 Instrument Channel - 1 470" above vessel zero. A 1. Multiplier relays initiate. Reactor Low Water Level RCIC. (LIS-3-58A-0)' 2 Instrument Channel - 1398" above vessel zero. A 1. Below trip setting initiates Reactor Low Water Level CSS. (LS-3-58A-D) -Multiplier relays initiate LPCI. 2. Multiplier relay from CSS initiates accident signal (15). 2(16) Instrument Channel - 1398" above vessel zero. A 1. Below trip settings, in v Reactor Low Water Level conjunction with drywell (LS-3-58A-D) Z high pressure, low water 1evel permissive, ADS timer o . timed out and CSS or RHR w pump running, initiates ADS. Oo 2. Below trip settings, in conjunction with low reactor-water level permissive, . ADS timer timed out. ADS high drywell pressure bypass timer tisied out, g CSS or RHR pump running, e initiates ADS. 5 =a 1(16) Instrument Channel - 1 544" above vessel zero. A 1. Below trip setting pemissive Reactor Low Water Level for initiating signals on ADS. c+ Pemissive (LIS-3-184, 185)- m 1 ' Instrument Channel - 1 312 5/16" above vessel zero. A 1. Below trip setting prevents Reacter Low Water Level (2/3 core height) inadvertent operatioe of '(LIS-3-52 and LIS-3-62A) containment spray during accident condition.

,.s 4 . TABLE 3.2.B (Continued) IMinimum No. c to

*n Operable Per_

O* Trio Sys(1) Function Trio Level Settine' Action Remarks 2(18) Instrument Channel -. 11 p12.5 psig 'A 1. Belee trip setting prevents l Drywell High Pressure inadvertent operation of. (PIS44-58 E-H) containment spray during. accident conditions. " + -2(18) Instrument Channel-- 1 2.5 psig A 1.. Above trip setting in con-l Drywell High Pressure (PIS-64-58 A-D) junction with low reactor pressure initiates CSS. Multiplier relays initiate - HPCI. I 2. Multiplier relay free CSS initiates accident signal. (15) 2(18) Instrument Channel - 1 2.5 psig A 1. Above trip setting in - l Drywell High Pressure (PIS44-58A-0) conjunction with low P reactor pressure initiates LPCI. "} 2(15)(18) Instrument Channel - 1 2.5 psig A 1. Above trip setting, in .. l Drywell High Pressure (PIS-64-57A-D) conjunction with'iow reactor N water level, low reactor water level permissive, ADS timer timed out, and CSS or Rim pump running, initiates ADS. N n 2or+ .a O 1 0 ~ L -. = - -+.. =~s ~ y .g ~. _ -.,:.-

I TABLE 3.2.B (Continued) C: to Minimum No. 5E Operable Per Trio Sys(11 ' Function Trie Level Settino Action Remarks w 2 Instrument Channel -- 450 psig 15-A 1. Below trip setting permissive Reactor Low Pressstre for opening CSS and LPCI . (PIS-3-74 A & B) admission valves. (PIS-68-95, %) 2 -Instrument Channel.- 230 psig 15 A 1. Recirculation discharge valve Reactor Low Pressure actuation. (PS-3-74 A & B) (PS-68-95, %) 2 Core Spray Auto Sequencing 61 t 18 sec. B 1. With diesel power Timers (5) 2. One per motor 2 LPCI Auto Sequencing 01 t 11 sec. B 1 mth diesel power Timers ' (5) 2. Cne per motor F 1 RHR$W A1, 53, C1 and D3 : 131 t 115 sec. A 1. With diesel power R Timers 2. One per pump r. L 2 Core Spray and LPCI Auto -01 t il sec. B 1. With normal power -Sequencing Timers (6) 61 t 18 sec. 2. One per CSS motor g 121 t 116 sec. 3. Two per RHR motor 181 t 124 sec. 1 RHRSW A1, 83, C1, and D3 -271 t i 29 sec. A 1. With normal power Timers 2. One per pump Y- &a x C3 .rn 29~ >4 G3 C

-x ^ " TABLE 3.2.5 (Continued) em Minlauer No. . o -c. . Operable Per. Trio Sys(1)' ' Function Trio Level Settinc Action Remarks 2 ~knstrumentChannel-100~ 10 psig ~ A 1 Below trip setting defers ADS -

RHR Discharge Pressure actuation.

2 Instrument Channel: 185 10 psig A 1. Below trip setting defers ADS : CSS Pump Discharge Pressure actuation. 1(3) Core Spray Sparger to 2 psid 0.4 A 1. Alarm to detect core sparger Reactor Pressure Vessel d/p pipe break. 1 ' RHR (LPCI) Trip System bus N/A C 1. Monitors availability of - power monitor power to logic systems. ~ 1 ' Core Spray Trip System bus N/A C 1. Monitors availability of Power montter power to logic systems.- 1 ADS Trip System bus power N/A C 1. Monitors availability of. monitor power to logic systems and valves. ra l C N <o c1 a b. c a O i i 4 4 b , _..... ~ --~ ~ _. _. ~ _ -.. -..,

NOTES F0k TABLE 3.2.B 1. .Whenever any CSCS System is required by Section 3.5 to be OPERABLE, there shall be two:0PERABLE trip _ systems except as noted. If a requirement of the:first column is reduced by one, the indicated action shall be taken. Ifi he same function is inoperable in more than one trip system or the t first column reduced by more than one, action B shall be taken. Action: A. Repair in 24 hours. If the function is not OPERABLE in 24 hours, take action B. B. Declare the system or component inoperable. C. Immediately take action B until power is verified on the trip system. D. No. action required; indicators are considered redundant. E. Within 24 hours restore the inoperable channel (s) to OPERAB1.E status or. place the inoperable channel (s) in the tripped condition. 2. In only one trip system. 3.- Not considered in a trip system. 4. Deleted.- 5. With diesel power, each RHRS pump is scheduled to start immediately and -each CSS pump is sequenced to start about 7 seconds later. 6. With normal power, one CSS and one RHRS pump is scheduled to start instantaneously, one CSS and one RHRS pump is sequenced to start after about.7 sec. with similar pumps starting after about 14 sec. and 21 sec., at which' time the full ccmplement of CSS and RHRS pumps would' be operating. 7. The-RCIC and HPCI-steam line high flow trip level actcings are given in terms of differential pressure. The RCICS setting of 450" of water corresponds to at least 150 percent above maximum steady state steam flow to assure that spurious isolation does not occur while ensuring the initiation of isolation following a postulated steam line break. Similarly, the HPCIS setting of 90 psi corresponds to at least -150 percent above maximum steady state: flow while also ensuring the initiation of isolation following a postulated break. 8. Note 1.does not apply to this item. 9. The head tank is designed to assure that the discharge piping from the CS and RHR pumps are full.-LThe pressure shall be naintained at or above the values. listed in 3.5.H, which ensures water in'the discharge piping and up to the head tank. BFN 3.2/4.2-23 Unit 2

NOTES FOR TABLE 3.2.B' (Cont'd)

10..Only one trip system for each cooler fan.

11. In only two of the four 4160-V shutdown boards. See note 13. 12. In only one of the four 4160-V shutdown boards. See note 13. 13. An emergency 4160-V shutdown board is considered a trip system.- 14. RERSW pump would be inoperable. Refer to Section 4.5.C for the requirements of a RHRSW pump being inoperable. ~15. The accident signal is the satisfactory completion of a one-out-of-two taken twice' logic of the drywell high pressure plus low reactor pressure or the vessel low water. level (1 398" above vessel zero) originating in the core spray system trip system. ' 16. - The ADS circuitry-is capable of. accomplishing its protective action with one OPERABLE trip system. Therefore, one trip system may be taken out of service for functional testing and calibration for a period not to exceed eight hours. 17. Two RPT systems exist, either of which will trip both recirculation pumps.. The systems will be individually functionally tested monthly. If the test period for one RPT system exceeds two consecutive hours, the system will be declared inoperable. If both RPT systems are inoperable or if one RPT system is inoperable for more than 72 hours, an orderly-l power reduction shall be initiated and reactor power shall be less than 30 percent within four hours. l 18. Not required to be OPERABLE in the COLD SHUTDOWN CONDITION. l l 1 a I L BFN - 3.2/4.2-24 Amendment 190 Unit 2 l~ t

4 3.5/4.5 CORE AND CONTAINMENT COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMINTS 3.5.B Residual Heat Removal System 4.5.B Residual Heat Removal System LEHRE) (LPCI and Containment tRHRS) (LPCI and Containment Cooling) Cooling) 8. If Specifications 3.5.B.1 8. No additional surveillance through 3.5.B.7 are not met,

required, an orderly shutdown shall be initiated and the reactor shall be placed in the COLD SHUTDOWN CONDITION within 24 hours.

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 that are required to be RHR loop with two pumps or two OPERABLE shall be loops with one pump per loop demonstrated to be OPERABLE shall be OPERABLE. Theupumps' per Specification 1.0.PM. 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 is not in the COLD SHUTDOWN cross-connect capability CONDITION, 2 RHR pumps and shall be demonstrated to be associated heat exchangers and OPERABLE per Specification valves on an adjacent unit 1.0.MM when the cross-must be OPERABLE and capable connect capability of supplying cross-connect is required, capability except as specified in Specification 3.5.B.12 below. (Notes 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 hours.) BFH 3.5/4.5-7 Amendment 190 Unit 2

3.5/4.5 CORE AND CONTAINHENT COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION-SURVEILLAIJCE REQUIREMENTS 3.5.B Residual Heat Removal Svatem 4.5.B Rgsidual Heat Removal System (RHRS) (LPCI and Containment (RHRSJ (LPCI'and Containment Cooling) Cooling) 12.- If three Ri!R pumps or associated 12. No additional surveillance heat exchangers located required. on the unit cross-connection in the adjacent unite are inoperable for any reason (inciv41ng valve inoperability, pipe break, etc.), the reactor may remain in operation for a period not to' exceed 30 days provided the remaining RilR pump and associated diesel generator are OPERABLE. 13. If RilR cross-connection flow ir 13. No additional surveillance heat removel' 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 T0 be tested for OPERABILITY STARTUP (or closed if during any period of permitted elsewhere COLD SHUTDOWN CONDITION in these specifications)..

exceeding 48 hours, if OPERABILITY testa-have not been performed-d during the preceding 31 days. BFH 3.5/4.5-8 ' Unit 2 i

3.5/4.5 CORE AND CONTAINMENT COOLIEG SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3. 5 '. C RHR Service Water and Emergency 4.5.C RHR Service Water and Emergency Eaulement Coolina Water Systema Eautoment Coolina Water Systems (EECWS) (Continued) (EECWS) (Continued) 4. Three of the D1, 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.0PERABLE. 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 hours. 7. There shall be at least 2 Ri.. 4 pumps, associated with the selected RHR pumps, aligned for RHR heat exchanger service for each reactor vessel containing irradiated fuel. AMENDMENT N0.16 Q BFN-3.5/4.5-12 Unit 2

3.5/4.5 CORE _AND,_ffjEIAINMENT COOLING SYSTEMS LIMITING CONDITI,0HS F0k OPERATION SURVEILLANCE REQUIREMENTS 3.5.D Eau t oment_AIAA.Cc ol e rs 4.5.D Eautoment Area Coolers-1. The equipment area cooler 1. Each equipment area cooler l associated with each RHR is operated in cos. junction pump and the equipment with the equipment served area cooler associated by that particular cooler; with each set of cere therefore, the equipment spray pumps (A and C area coolers are tested at or.E 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. Talen an equipment area , cooler is not OPERABLE, the pump (s) served by that cooler must-be considered inoperable for Technical Specification purposes. E. jllgh Presagre Coolant Injection E..Elah Pressure Coolant Svetem (HPCIS) Iniection System (HPCIS) 1. The IIPCI system shall be 1. HPCI Subsystem testing OPERABLE whenever there is shall be performedsas irradiated fuel in the follows: reactor vessel and the reactor vessel pressure a. Simulated Once/18 is greater than 150 psig, Automatic months -except in the COLD SHUTDOWN Actuation CONDITION or as specified Test in Specification 3.5.E.2. OPERABILITY ahall be deter-b. Pump Per mined within 12 hours.after OPERA-Specification reactor steam pressure BILITY 1.0.MM reaches 150 psig from a COLD CONDITION, or alter-c. Motor Oper-Per natively PRIOR TO STARTUP ated Valve Specification by using an auxiliary steam OPERABILITY 1.0.MM

supply, d.

Flow Rate at Once/3 normal months reactor vessel operating pressure BFN 3,5/4.5-13 Amendment 190 Unit 2

M /4.5 ' CORE AND-CONTAINMENT C00LINC SYSTEMS -LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.E-High-Pressure Coolant Iniection 4.5.E lilah Pressure Ca.glant Iniection System (HPCIS) System (HPCIS) 4.5.E.1 (Cont'd) e. Flow Rate at Once/18 150 psig months The F'CI 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 UPCI system is 2. No additional surveillances inoperable, the reactor are required. may remain in operation for a period not to exceed 7 days,-provided the ADS, CSS, RHRS(LPCI), and RCICS are OPERABLE. 3 '3. If Specifications 3.5.E.1

Except that an automatic or-3.5.E.2 are not met, valve capable of automatic i

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 150 position for another mode psig or less within 24 of' operation. . hours. F. Rgaetor Core Isolation Cooling F. Etastor Core Isolation Coolina System (RCICS)~ System (RCICS) 1. The 'RCICS shall in OPERABLE 1. RCIC Subsystem testing shall whenever there is irradiated be performed as follows: fuel in the reactor vessel and the reactor vessel a.- Simulated Auto-Once/18 pressure is above 150 psig, matic Actuation months except in the COLD SHUTDOWN Test CONDITION or as specified in 3.5.F;2. OPERABILITY shall BFN 3.5/4,3 ;q Amendment 190 Unit 2 _ _ _ _ = _ _ _ _

3.5/4.5-CORE AND CONTAINMENT C0QLIEG SYSTEMS ' LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3;5.F. Reactor Core Isolation Coolina 4.5.F Reactor Core Isolation Coolina System (PCICS) System (RCICS) 3.5.F.1 (Cort'd) 4.5.F.1 (Cont'd) be determined within 12 hours L. Pump Per after reactor steam pressure OPERABILITY .Specifi-reaches 150 psig from a COLD cation i CONDITION or alternatively '1.0.MM-PRIOR TO STARTUP.by using an auxiliary steam supply. .c. Motor-Operated Per Valve Specifi-OPERABILITY cation 1.0 MM' d. Flow Rate at once/3 normal reactor months ~ vessel operating pressure e. Flow Rate at Once/18 l 150 psig months 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- -to1 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_ wise secured in orderly shutdown shall be -position,_-is in its initiated and the reactor. correct

position, chall be depressurized to less than-150 psig within-

2. No additional surveillances 24 hours.

are: required. t

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.

AMENDMENT N0.17 6 BFN 3.5/4.5-15 Unit 2

3.5/4.4 CORE AND CONTAINNINT COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.G Automatic Deoressurization 4.5.G Automatic Deortganrization System (ADS) System (ADS) 1. Six valves of 1. During each operating the Automatic cycle the following Depressurization System testa shall be performed shall be OPERABLE: 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 pressure is covered in is greater than 105 psig, 4.6.D.2. except in the COLD SHUT-DOWN CONDITION or as specified in 3.5.C.2 and 3.5.G.3 below. 2. With one of the above 2. No additional-surveillances required ADS valves are required, inoperable, provided the HPCI system, the core spray system, and the LPCI system are OPERABLE, restore the inoperable ADS valve to OPERABLE status-within 14-days or be-in at least a HOT SHUTDOWN CONDITION within the next 12 hours and -reduce reactor steam dome pressure to 1105 psig within 24 hours. 3. With two or more of the above required ADS valves inoperable, be in at least a HOT SHUTDOWN CONDITION within 12. hours and reduce reactor steam dome pressure to 1105 psig within 24 hours. BFN 3.5/4.5-16 Amendment 190 Unit 2

3.5/4.5 CORE AND CONTAINMENT C00LlHC SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.H. Maintenance of Filled Discharme 4.5.H. Maintenance of Pilled Dis.harge ElRg ElRg Whenever the core spray systems, The following surveillance LPCI, HPCI, or RCIC are required requirements shall be adhered to be OPERABLE, the discharge to assure that the discharge piping from the pump discharge piping of the core spray of these systems to the last systems, LPCI, HPCI, and RCIC block valve shall be filled, are filled: testing of the RHRS (LPCI and' l The auction of the RCIC and HPCI 1. Every month and prior to the . pumps shall be aligned to the 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, cordensate head tank may be used to serve the RHR and CS discharge 2. Following any period where piping if the PSC head tank the 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 -PI-75-20 48 psig system to service. P1-75-48 48 pois P1-74-51 48 psig 3. Whenever the HPCI or RCIC' PI-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. Jhen the RHRS and.the CSS are required to be OPERABLE, the pressure indicators which monitor the discharge lines shall be monitored daily and tbc pressure recorded. BFN 3.5/4.5-17 Unit 2 t l

l I=" 3.6/4.6' ~ PRIMARY SYSTEM BOUNDARY LIMITING CONDITIONS FOR OPERATION -SURVEILLANCE REQUIREMENTS-3.6.C. Coolant Leakane 4.6.C. Coolant Leakare. -1. -a. 1Any time irradiated 1.- Reactor-coolant fue1Lis in-the system leakage shall reactor vessel and be checked by the reactor coolant sump and air sampling temperature is above system and recorded 212*F, reactor coolant at least once per ~ leakage into the 4 hours. primary containment from= unidentified sources shall not exceed. 5 gpm. In-addition, the total reactor coolant system leakage into the primtcy containment shall-not exceed 25 spm. b. Anytime the reactor is in RUN mode, reactor coolant leakage into the primary containment from unidentified sources shall not increase by 1 more than 2 spm averaged over any 24-hour period in which the reactor is in the-RUN mode except as defined-in 3.6.C.1.c below, c. Durin6 the first 24 hours-in the RUN mode following STARTUP, an increase in reactor coolant leakage sinto the primary containment of >2 gpm is accentabl as long as the requirements -of.3.6.C.1.a are met, n-BFN 3.6/4.6-9 Unit 2 1

1 -3.6/4.6 PRIMARY SYSTEM BOUNDARY. l LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.6.C Coolant Leakage 4.6.C Coolant Leakagg 2.= Both the sump and air sampling 2. With the air sampling systems shall be OPERABLE . system inoperable, grab .g during REACTOR POWER-OPERATION. samples shall be obtained -From and after the date that and analyzed at least one of these systems is made once every 24 hours, or found to be inoperable for' any reason, REACTOR POWER OPERATION is permissible only during the succeeding 24 hours for the sump system or 72 hours for the air sampling system. The air sampling system may be removed from service for a . period of 4 hours for calibration, function testing, l and. maintenance without. L providing a temporary l monitor. ] 3.- If the condition in l'or 2 l' above cannot be met, an orderly shutdown shall be initiated L and the reactor shall be placed in the COLD SHUTDOWN CONDITION within 24 hours.- c D.- Relief Valves D. Relief Valves l. 1. When morel than one relief valve 1. Approximately one-half of is known to be failed, an all relief valves chall E orderly shutdown shall be be bench-checked or initiated and the reactor replaced with a-depressurized to less than 105 bench-checked-valve l psig within 24 hours. The -each operating cycle. relief valves are not required All 13 valves will have to be OPERABLE in the COLD been checked or replaced-SHUTDOWN CONDITION.- upon the completion of every second. cycle. j -2. In accordance with Specification 1.0.MM, each relief valve ah311 be. manually opened until L thermocouples and l-acoustic monitors L downstream of the valve indicate steam is flowing from the valve. L BFN 3.6/4.6-10 Amendment 190 Unit 2 l

P

13. 6/4.6.; $ASES 1

3.6.B/4!6.C~(Cont'd).

in 3.6.C, the experimental and analytical data- .{ five gpm,'as specifiedt . suggest. a reasonable _ margiu of safety that'such leakage magnitude would F g i not result from a crack appcoaching the critical size for rapid o,; _ propagation.. Leakage less then the magnitude specified can be detected ~ t reasonably in a matter of a few hours utilizing-the available leakage L - detection schemes,; and if the origin' cannot be determined in a reasonably short time,-the unit siould be shut down to allow further investigation and corrective action. t

The ~2 spa limit for coolant: 1eakage rate increases over any 24-hour period-is'a' limit specified-by'the-NRC (Reference 2).

This limit applies only-during;the.RUN mode _to. avoid being penalized for the-E expected coolant = leakage: increase -during pressurization.. t. The tota 111eakage = rate consists-of all leakage,-identified and .unidentifie6 which flows to the -drywell floor drain and equiprnent drain uumps. lThe' capacity of the drywell floor-sump pump is 50 spa and the' capacity of the drywell equipment sump pump is also;50 spa. Removal of'25 spm from either of' these sumps can be accomplished with considerable margin. REFERENCE l'. Nuclear System Leakage Rate Limits (BFNP FSAR-Subsection'4.10) m j7

2. Safety Evaluation Report (SER) on IE' Bulletin 82-03 L3.6.D/4.6.D:-' Relief Valves

.To meet-the -safety basis,13: relief valves have been installed on the-unit"with_a4 total; capacity;of 14.1Vpercent of nuclear boiler rated steam flow. Theianalysis of the worst overpressure transient,.(3-second- -g closure-of'all: main-' steam:line-isolation. valves)' neglecting:the direct - scram (valve _ position scram) results_in a maximum vessel pressure which, if a: neutron: flux scram is assumed considering 12' valves OPERABLE, = results 1n; adequate margin to'the code allowable overpressure limit of + 1',375 psig. 1 To meet _. operational' design, the analysis of the plant. isolation transient (generator.Ioad-reject-.with bypass valve failure toiopen). shows'that 12 -of the 13. relief valves limit' peak systemipressure to a .value'which is well below~the' allowed vessel overpressure of.1,375 psig.. Experience in' relief valve operation shows _'that:a testing: of 30 percente t of_the;vsives per year is adequate to detect failures or ideteriorations. -The; relief-valves are benchtested every'second operating cycle to: ensure that'their.setpoints are_within'the'i ll percent _ tolerance. The relief valves are tested-in place in accordance with-Specification =1.0.MM to establish ~that they will open and_ pass steam. AMENDMENT NO.170 BFN 3.6/4.6-30 -Unit 2 J - ~..

.7 __--._.r. t 3.6/4.6 BASES "3. 6.D/4'. 6.Dj (Cont 'd) L The requirements established above apply when the nuclear syste.i can be pressurized above ambient conditions. These requirements'are applicable

A at; nuclear systea pressures below normal operating pressures be.cause abnormal _-operational transients could-possibly. start et these conditions ^

such that eventual overpressure relief would be needed..However, these transients.are much less severe, in terms of pressure, than those starting'at' rated conditions. The valves need not be functional _when the vessel head is removed, since the nucitar system cannot be_ pressurized. 1 The-relief valves are not required to bel OPERABLE in the, COLD SmlTDOWN. CONDITION.' Overpressure protection.is-provided during hydrostatic tents by two ofothe relief valves whose relief setting:ha's been established.in-conformance with ASME Section XI code requirements.- The capacity of one relief valve exceeds the charging capacity of the pressuriestion source used during hydrostatic testing. Two relief _ valves are used'to provide, redundancy. REFERENCES 1.1 Nuclear' System Passure Relief System (BTNP FSAR Subsection 4.4)

n 3

2. -Amendment 22 in. response:to-AEC Question 4.2 of De: ember 6, 1971.

i 3. _" Protection Against Overpressure" (ASME Boiler and Pressure Vessel Code, Section.III, Article 9) 4. Browns Ferry Nuclear Plant Design Deficiency Report--Target Rock 7 Safety-Relief Valvesi. transmitted by J. E. G111 eland ^to. F. E. Eruest, August 29,.1973 5. Generic: Reload-Fuel! Application, Licensing Topical ~ Report, 'NEDE-24011-P-A and: Addenda _ 3.61E/4.6.E Jet Pumos

Failure of a-jet pump nozzle-assembly holddown: mechanism,tnozcle assembly-

-and/or riser, would increase the cross-sectional-flow urea for blowdown Lfollowing-the design basis. double-endedzline break. Also,- failure of the

diffuser would-eliminate the capability to:reflood.the_ core to two-thirds height level following a-recirculation linefbreak ' Therefore,.if a failure occurred, repairs must-be made.

b .The-detection technique is as follows. With-the two recirculation pumps balanced in speed to withi_n i 5 percent,_the flow rates in.both recirculation loops will-be verified byscontrol. room monitoring-

instruments.

If.the:' two ilow rate values do not_ dif fer hy more than 110'percentp riser and nozzle assembly _ integrity has been verified. 7 g i. l[ T8FN 3.6/4.6-31 Amendment 190 J -Unit;2-1- i o L

w...

y L 3.6/4.6 BASES 3.6.E/4.6.E (Cont'd) If they do differ by 10 percent or more, the core flow rate measured by the jet pump diffuser differential pressure system aust be checked against the core flow rate delived from the measured values of loop flow to core flow correlation. If the difference between measured and derived { core flow rate is 10 percent or more (with the derived value higher) diffuser measurements will be taken to define the location within the veseel of failed jet pump nozzle (or riser) and the unit shut down for repairs. If the potential blowdown flow area is increased, the system resistance to the recirculation pump is also reduced; hence, the affected drive pump will "run out" to a substantially higher flow rate f (approximately 115 percent to 120 percent for a single nozzle failure). f If the two loops are balance 4 *n flow at the same pump speed, the } resintance characteristica e ans. t hate ;Senged. Any imbalance between drive loop flow rates would te indicated t; the plant process } instrumentation. In addition, the affecte. jet pump would provide a ) leakege path pa st the core t aus reducing he core flow rate. The reverse l flow through the inactive jet pump Scuid still be indicated by a positive i I differential pressure but the net effect would be a slight decrease i (3 percent to 6 percent) in the total core flow measured. This decrease, l [ together with the loop flow increase, would result in a lack of l correlation between measured and derived core flow rate. Finally, the ) effected jet pump diffuser differential pressure signal would be reduced because the backflov would be less than the normal forward flow. l A nozzle-riser system failure could also generare the coincident failure i of a jet pump diffuser body; however, the converse is not true. The lack { of any substantial stress in the jet pump diffuser body makes failure l l impossible without an initial nozzle-riser system failure. 3.6.F/4.6 F Recirculation Pumo Operation l Operation without forced recirculation la pt:rmitted for up to 12 hours when the reactor is not in the RUN mode. And the start of a recirtu. tion pump from the natural circulation condition will not be permitted unless the temperattre difference between the loop to be started and the core coolan't temperature is less than 75'F. This reduces the positive reactivity insertion to an acceptably low value. l l Requiring at least one recirculation pump to be operable while in the RUN l mode provides protection against the potential occurrence of core l thermal-hydraulic instabilities at low flow conditions. l Requiring the discharge valve of the lower speed loop to remain closed until the speed of the faster pump is below 50% of its rated speed l provides assurance when going from one-to-two pump operation that excessive vibration of the jet pump risers will not occur. l BFN 3.6/4.6-32 Amendment 190 Unit 2 l

3.6/4.6 BASES 3.6.G/4.6.G Structural Integrity The requirementa for the reactor crolant systems inservice inspection program have been identified by evaluating the need for a sampling examination of areas of high stress and highest probability of failure in the system and the need to moet as closely as possible the requirements of Section XI, of the ASME Boiler and Pressure Vessel Code. The program reflecta the built-in limitations of access to the reactor coolant systems. It is intended that the required examinations and inspection be completed during each 10-year interval. The periodic examinations are to be done during refueling outages or other extended plant shutdown periods. Only proven nondestructive testing techniques will be used. More frequent inspections shall be performed on certain circumferential pipe welds as listed in Section 4.6.G.4 to provide additional protection against pipe whip. These welds were selected in respect to their distance from hangers or supports wherein a failure of the weld would permit the unsupported negments of pipe to strike the drywell vall or nearby auxiliary systems or control systems. Selection was based on judgment from c:tual plant observ.ation of hanger and support locations and review of drawings. Inspection of all these welds during each 10-year inspection interval will result in three additional examinations above the requirements of Section XI of ASME Code. An augmented inservice surveillance program is required to determine whether any stress corrosion has occurred in any stainless steel piping, stainless components, and highly-stressed alloy steel such as hanger springs, as a result of environmental conditions associated with the March 22, 1975 fire. REEERENCES 1. Inservice Inspection and Testing (BFNP FSAR Subsection 4.12) 2. Inservice Inspection of Nuclear Reactor Coolant Systems, Section XI, ASME Boiler and Pressure Vessel Code 3. ASME Boilar and Pressure Vessel Code, Section III (1968 Edition)

4.. American Society for Nondestructive Testing No. SNT-TC-1A (1968 Edition) 5.

Mechanical Maintenance Instruction 46.(Mechanical Equipnent, Concrete, and Structural Steel Cleaning Procedure for Residue From Plant. Fire - Units 1 and 2) 6. Mechanical Maintenance Instruction 53 (Evaluation of Corrosion Damage of Piping Components Which Were Exposed to Residue From March 22, 1975 Fire) 7. Plant Safety Analysis (BFNP FSAR Subsection 4.12) BFN 3.6/4.6-33 Unit 2 l

/ 'g UN11ED STATES ) 8, NUCLEAR RECULATORY COMMISSION e 3 4 WASHINoTON, D. C. 206% TENNESSEE VALLEY AUTHORITY , DOCKET NO. 50 296 BROWNS FERRY NUCLEAR PLANT, UNIT _3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.152 License No. DPR-68 1. The Nuclear Regulatory Commission (the Comission) has found that: A. The application for amendment by Tennessee Valley Authority (the licensee)-dated May 18 1990, as super:eded by your letter of October 30, 1990, complies with the-standaids and requirements of the Atomic Energy Ac', of 1954, as amended (tte Act), and the Comission's rules and regulations set forth in 10 CFR Chapter 1; B. The facility will operate in conformity with the applica' ion, the o provisions of the Act, and the rules and regulations of the Comission; C. Thereisreasonableassurance(1)thattheactivitics thorized by I this amendment can be conducted without endangering ths health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D. The issuance of this amendment will not be inimical to the 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 h-of.the Comission's regulations and all applicable requirements have been satisfied. I 3 i

i. 2. Accordingly, the license is amended by changes to the Technical Specif 0 cations as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Fecility Operating License No. DPR48 is hereby aver.ded to read as follows: (2) Technica) SpecH; cations The Technicel Specifications contained in Appendices A and B, as revised through Amendment No.102, are hereby incorporated in the license. The licensee shall o the Technical Specifications. perate the facility in accerdance with 3. This. license amendment is effective as of its date of issuance and shall be imp?ccented within 30 days from the date of issuance. FOR THE NUCLEAR REGULATORY C041SS10N n Frederick J. Heb on. Director i Project Directorate 11-4 Division of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation j

Attachment:

i Ch6nges to the Technical Specifications Date of Issuance: February 7, 1991

t A17Aci' MENT TO LICENSE AMEN 0 MENT NO.152 FACILITY OPERAT1kG LICENSE NO. DPR-68 DOCKET NO. 50 296 Revise the Appendix A Technical Specifications by reroving the pages identified below and inserting the enclosed pages. The revised pages are identified by the captioned amenda.ent number and contain marginal lines indicating the area of change. 0Yerleaf* and spilloVer** pages are provided to raintain document completeness. REMOVE INSERT 3.2/4.2-14 3.2// 14 3.2/4.2-15 3.2/4.2-15 ?.2/4.2-22 3.2/4.2-22* 3.2/4.2-23 3.2/4.2-23 3.5/4.5-7 3.5/4.5-7 3.5/4.5-8 3.5/4.5-8* 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** 3.6/4.6-9 3.6/4.6-9* 3.6/4.6-10 3.6/4.6-10 3.6/4.6-30 3.6/4.6-30* 3.6/4.6-31 3.6/4.6-31 3.6/4.6-32 3.6/4.6-32** 3.6/4.6-33 3.6/4.6-33*

TABLE 3.2.8 INSTRimDITATION THAT INITIATES OR CONTROLS THE CORE AND CONTAIMMENT CDDLING SYSTEMS - e to ' Minimum No. E2

Operable Per n

r Trio Sysf1) Function Trio Level Settine Actim Remarks u 2 Instrument Channel - 1 47D* above vessel zero. Reactor Low Water Level A-1. Below trip setting initiates l HPCI. 2 . Instrument Channel - 1 470" above vessel zero. A 1. N1tiplier relays initiate Reactor Low water Level RCIC. 2 Instrument Channel - 1 378" above vessel mere. A 1. Below trip setting initiates Reactor Low Water Level CSS. (LIS-3-58A-0, SWal) N1tiplier relays initiate LPCI. l 2. Multiplier relay from CSS initiates accident signal (15). 2(16) Instrument Channel 1 378* above vessel aero. A 1. Below trip settings, in Reactor Low Water Level conjunction with drywell w (LIS-3-58A-0. swr 2) high pressure, low water 2 level permissive. 120 sec. delay timer and CSS or RHR pump running, initiates AnS. 1(16) Instrument Channel -. 1 544 above vessel aero. A 1. Below trip setting permissive Reactor Low Water Level for initiating signals on ADS. Fermissive (LIS-3-184 8. g 185, SWal) 1 Instement Channel - 1 312 5/16" above vessel zero. A 1. Below 1949 setting prevents Reactor Low Water Level (2/3 core height) inadvertut operation of e (LITS-3-52 and 62, SWW1) containment spray during 3 accident condition. - -- = - -~ -

4' TABLE 3.2.8 (Continued) i ' slum No. ey. O M i. sble Per. O*' i d Sys(1) Function Trie Level Settine Action Remarks 2(18) . Instrument Channel - 11 p12.5 psig A 1. ~ (PS-64-58 E-M)- Below trip setting prevents ' l Drywell High Pressure inadvertent operation ef r containment spray during ~ accident conditions. 2(18) Instrument Channel - 1 2'.5 psig-A 1. Above trip settir*g in con-l Drywell High Pressure-(PS-64-58 A-0. Sw2) junction with low reactor pressere initiates CSS. Hultiplier relays initiate HPCI. 2. N1tiplier relay from CSS initiates accident signal. (15) 2(18) Instrument Channel - 1 2.5 psig A 1. Above trip setting in l Drywell High Pressure (PS-64-58A-0, Sw1) conjunction with low reactor pressure initiates LPCI. k 2(16)('18) Instrument Channel - 1 2.5 psig A 1. Above trip setting, in l [ t d Drywell High Pressere (PS-64-57A-0) conjunction with low reacter t water level, drywell high pressure. 120 sec. (,eTay I i timer and CSS er RHR pe=qp i running, initiates ADS. I 4 i b, CL E3 i (D J r+ 1 U1 1 N. 1 i 7 l = I ) b l t t r

30TES FOR TABLE _J M 1. Whenever any CSCS System is required by Section 3.5 to be OPERABLE, there shall be two OPEkABLE trip systems except as noted. If a requirement of the first column is reduced by one, the indicated action shall be taken. If the same function is inoperable in more than one trip system or the b first column reduced by more than one, action B shall be taken. j Action: A. Repair in 24 hours. If the function is not OPERABLE in 24 hours, take action B. B. Declare the system or component inoperable. C. Immediately take action B until power is verified on the trip system. D. No action required; indicators are considered redundant. 2. In only one trip system. 3. Not considered in a trip system. 4. Requires one channel from each physical location (there are 4 locations) in the steam line space. 5. With diesel power, each RHRS pump is scheduled to start immediately and each CSS pump is sequenced to start about 7 seconds later. 6.. With normal power, one CSS and one Ri!RS pump is scheduled to start instantaneously, one CSS and one RHRS pump is sequenced to start after about 7 seconds with similar pumps starting after about 14 seconds and 21 seconds, at which time the full complement of CSS and RHRS pumps would be operating. 7. The RCIC and HPCI steam line high flow trip level settings are given in terms of differential pressure. The RCICS setting of 450" of water corresponds to at least 150 percent above maximum steady state steam flow to assure that spurious isolation does not occur while ensuring the initiation of isolation following a postulated sters line break. Similarly, the HPCIS setting of 90 psi corresponds to at least 150 percent above maximum steady state flow while also ensuring the initiation of isolation following a postulated break. 8. Note 1 does not apply to this item. 9. The head tank is designed to assure that the discharge piping from the CS and RilR pumps are full. The pressure shall be maintained at or above the values. listed in 3.5.H, which ensures water in the discharge piping and up to the head tank. BFN 3.2/4.2-22 Unit 3

HOTES FOR TAELE_162tB (C:ntinued) 10. Only one trip cystem ic.r each cooler fan. 11. In only two of the four 4160-V shutdown boards. See note 13. 12. In only one of the four 4160-V shutdovn boards. See note 13. 13. An emergency 4160-V shutdown board is considered a trip system. 14. RHRSW pump would be inoperable. Refer to Section 4.5.C for the requirements of a RHRSW pump being inoperable.

15. The accident signal is the satisfactory completion of a one-out-of-two taken twice logic of the dryvell high pressure plus low reactor pressure or the vessel low water level (1 378" above vessel sero) originating in the core spray system trip system.

16. The ADS circuitry is capable of accomplishing its protective action with one OPERABLE trip system. Therefore, one trip system may be taken out of i

service for functional testing and calibration for a period not to exceed eight hours.

17. Two RPT systems exist, either of which will trip both recirculation pumps. The systems will be individually functionally tested monthly.

If the test period for one RPT system exceeds two consecutive hours, the system will be declared inoperable. It both RPT systems are inoperable or if one RPT system is inoperable for more than 72 hours, an orderly: l power reduction sha.; be initiated and reactor power shall be less than 30 percent within four hours. l 18. Not required to be OPERABLE in the COLD SHUTDOWN CONDITION. i l BFN-3.2/4.2-23 Amendment 152 Unit 3 ~

._._m_____ idl4.S CORE AND C0f(Ij1NMENT COOLING MSTEli$ ' LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.B Ruidual Heat Removal System 4.5.B Ruidual Heat Removal System (kHksi (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. anLorderly. shutdown shall be l initiated and the reactor shall be placed in the . COLD SIMTDOWN CONDITICH within 24 hours. 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 that are required to be RHR loop with two pumps or two OPERABLE shall be loops with one pump per loop demonstrated to be shall be OPERABLE. The pumps' OPERABLE per associated diesel generators Specification 1.0.MM. 'must also be OPERABLE. ' 1'. If the conditions of 10. No additional surveillance O . 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 is not in the COLD SHUTDOWN cross-connect capability. CONDITION, 2.RHR pumps ar.d shall be demonstrated to associated heat exchangers and be OPERABLE per valves on an~ adjacent unit Specification 1.0.MM when must be OPERA 8LE and capable the cross-connect of supplying cross-connect capability is required. capability except as specilled 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 $ hours.)- BFN 3.5/4.5-7 Amendment 152 Unit-3 4

m. - - -, -,, - -..,.

1 i 225/4.5 CORE AND CONTAINMENT COOLING SYSTEMS I LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIEEMENTS 3.5.B Residual Heat Removal Svatem 4.5.B Residual Heat Removal System LEHR11 (LPCI and Containment (RHRS) (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 unit 2 is inoperable for any reason (including valve inoperability, pipe break, etc.), the reactor may remain in operation l

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 70 be tested fcr OPERABILITY STARTUP (or closed if during any period of permitted'elsewhere COLD SHUTDOWN CONDITION in these specifications). exceeding.48 hours, if OPERABILITY tests have not been performed during the preceding 31 days. i BFN 3.5/4.5-8 AMENDMENT NO.14 0 Unit 3

3.5/4.5 CORE AND CONTAINMENT COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.C RilR Sgrvice Water and Emerzenev 4.5.C RHR Service Water and Emeraency Eculoment Coolinn Water Systems Eguloment Coolina, Water Systems IIECWS) (Continued) (EECWS) (Continued) 4. One of the B1 or B2 RHRSW 4. No additional surveillance pumps assigned to the RHR is required, heat exchanger supplying the standby coolant supply connection may be inoperable for a period 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 hours.

7. There shalllae at least 2 RHRSW pumps, associated, with the selected RHR pumps, 1 aligned for RHR heat i exchanger service for each reactor vessel containing irradiated fuel. l ENDMERT NO.14 6 BFN 3.5/4.5-12 Unit 3 .1

L,$fALgEE AND CONTAINMENT COOLINO SYSTEMS LIM 1TINC CONDITIONS FOR 3PERATION SURVEILLANCE REQUIREl.TNTS 3.5.D Eau 1PEgnt Area Coolerg 4.5.D Eaulement Area Coolgtg 1 1. The equipment area cooler 1. Each equipment area cooler associated with each RHR is operated in conjunction lump and the equipment with the equipment served i 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 4 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 onsidered inoperable for 'rechnical Specification purposes. E. His.h Pressure Coolant In' E. _High Pressure Coolani Eygtem (HPCIS) Iniection System (HPCIS) I 1. The HPCI system shall be 1. HPCI Subsystem testing OPERABLE whenever there is shall be perfwemed as irradiated fuel in the follows: reactor vessel and the reactor vessel pressure a. Simulated Once/18 is greater than 150 psig, Automatic months except in the COLD SHUTDOWN Actuation CONDITION or as specified in Test Specification 3.5.E.2. OPERABILITY ahall be deter-b. Pump Per mined within 12 hours after OPERA-Specification re.ctor steam pressure BILITY 1.0.MM reaches 150 psig from a COLD CONDITION, or alternatively c. Motor Oper-Per PRIOR TO STARTUP by using an ated Valve Specification auxiliary steam supply. OPERABILITY 1.0.MM d.- Flow Rate at once/3 normal months reactor vessel operating pressure BFN 3.5/4.5-13 Amendment 152 Unit 3 l

3.5/4.5. CORE AND CONTAINMENT COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.E Rich Pressure Coolant iniection 4.5.E Ulth Pressure Coolant In_iection lyJJem (HPCIS) $3Atem (HPCIS) 4.5.E.1 (Cont d) i e. Flow Rate at Once/18 150 psig months 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, scaled, or otherwise secured in position, is in its correcta position. 2. If the HPCI system is 2. No additional surveillances inoperable, the reactor may are required, remain in operation for a period not to exceed 7 days, provided the J.DS, 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 shusdown 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 150 position for another mode psig or less within 24 of operation.

hours. F. Ef JQL C.2re Is21ation Coolina F. Reactor Core Isolation C2911Ag E221.m (RCICS) System (RCICS) 1. The RCICS shall be OPERABLE 1. RCIC Subsystem testing shall whenever there is irradiated be performed as follows: fuel in the reactor vessel and the reactor vessel a. Simulated Auto-Once/18 pressure is above 150 psig, matic Actuation months except in the COLD SHUTDOWN Test CONDITION or as specified in 3.5.F.2. OPERABILITY shall BFN 3.5/4.5-14 Amendment 152 Unit 3 0

3.5/4.5 CORE AND CQEIALNBENT COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.F Reastor' Core Isolation Coolina 4.5.F Reactor Core Isolation Coolina Svatem (kCICS) System (RCICS) i 3.5.F.1 (Cont'd) 4.5.P.1 (Cont'd) be determined within 12 hours

b. Pump Per after reactor steam pressure OPERABILITY Specifi-reaches 150 psig from a COLD cation CONDITION or alternatively 1.0.MM PRIOR TO STARTUP by using an auxiliary steam supply.

c. Motor-Operated Per Valve Specifi-OPERABILITY cation 1.0.MM d.

Flow Rate at Once/3 normal reactor monthe vessel operating pressure e. Flow Rate at once/18 150 psig months 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 .g less than 150 psig within

2. No additional surveillances i

24 hours, are required.

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

BFN 3.5/4.5-15 AMENDMENT NO.14 4 -Unit 3

3.5/4.5 CORE AND CQHIAIMMENT COOLING SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.0 Automatic Depressuriration 4.5.G Automatic Deoressuriration Eystem (ADS) System (ADS) 1. Four of the six valves of 1. During each operating the Automatic. cycle the following Depressurization System tests shall be performed shall be OPERABLE on the ADS (1) PRIOR TO STARTUP from a. A simulated automatic a COLD CONDITION, or, actuation test shall be performed PRIOR TO (2) whenever there is STARTUP after each irradiated fuel in the refueling outage, reactor vessel and the Manual surveillance reactor vessel. pressure of the relief valves is greater than 105 psig, is covered in except in the COLD SHUT-4.6.D.2. DOWN CONDITION or as specified in 3.5.G.2 and 3.5.G.3 below.

2. -If three of the six ADS 2.

No additional surveillances 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 knovn to be incapable of automatic operation, an immediate orderly shutdown shall be initiated, with the reactor in a HOT SHUTDOWN CONDITION in 6 hours, and in a COLD SHUTDOWN CONDITION in the following 18 hours. 3. If: Specifications =3.5.G.1 and 3.5 G.2 cannot-be met, an orderly shutdown will be initiated and the reactor BFN 3.5/4.5-16 Amendment 152 Unit 3

l L5/4.5 COFE AND _ CONTAINMENT C00LINCJY$1Q18 LIMITING CONDITIONS POR OPERATION SURVEILLANCE REQUIREMENTS 3.5.G AuLquuttic Depressurization 4.5.G Automatic _Deorcea n t ation System (ADS) Sys t em (ADLJ 3.5.G.3 (Cont'd) vessel pressure shall be reduced to 105 psig or less within 24 hours. 11. lialnt enance of Filled Discharr.c H. Maintenance of Filled Dischnitt l' LAC l'ipl Whenever the core spray systems, The following surveillance LPCI, !!PCI, 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 llPCI 1. Every month and prior to the pumps shall be aligned to the testing of the RilRS (LPCI and condensate storage tank, and Containment Spray) and core the pressure suppression chamber spray systems, the discharge head tank shall normally be piping of these systems shall aligned to serve the discharge be vented from the high point piping of the RilR and CS pumps. and water flow determined. The condensate head tank may be used to serve the RilR and CS 2. Following any period where the discharge piping if the PSC head LPCI or core spray systems tank is unavailable. The have not been required to be pressure indicators on the OPERABLE, the discharge piping discharge of the RilR and CS pumps of the inoperable system shall 'shall indicate not less than be vented from the high point listed below, prior to the return of the system to service. P1-75-20 48 psig P1-75-48 48 psig 3. Whenever the liPCI 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 llPCI and RCIC shall be vented from the high point of the system and water flow observed on a monthly basis. 4. When the Ri!RS and the CSS are required to be OPERABLE, the pressure indicators which monitor thc discharge lines shall be monitored daily and the pressure recorded. BPN 3.5/4.5-17 Amendment 152 linit 3 u

3.6/4.6 PRIMARY SYSTEM BOUNDARY LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.6.C Coolant Leakaze 4.6.C Coolant Lankane 1. a. Any time irradiated 1. Reactor coolant fuel is in the system leakage shall reactor vessel and be checked by the reactor coolant sump and air sampling temperature is above system and recorded 212*F, reactor coolant at least once per leakage into the. 4 hours. primary containment from unidentified sources shall not exceed 5 spm, In addition, the total reactor coolant system l leakage into the primary containment shall not exceed 25 apm. i b. Anytime the reactor is in RUN mode, reactor coolant leakage into the primary containment from i unidentified sources shall not increase by more than 2 spm averaged over any 24-hour period in which the reactor is in the RUN mode except as defined in 3.6.C.I.c below, c. During the first 24 hours in the RUN mode following STARTUP, an increase in reactor coolant leakage into the primary containment of >2 spm is acceptable as long as the requirements of 3.6.C.I.a are met. i i AMEN 0 MENT NO.10 8 BFN 3.6/4.6-9 ' Unit 3

2 6/4.6. PRIMARY SY1IEN EDUNDARI LIMITING CONDITIONS FOR OPERATION SURVEILLANCE kEQUIREMENTS 3. f.. C Coolant..Leakane 4.6.C Coolant Leakan; 2. Both the sump and air sampling 2. With the air sampling systems shall be OPERABLE system inoperable, grab during REACTOR POWER OPERATION. samples shall be obtained From and after the date that and analyzed at least one of these systems is made or once every 24 hour:: found to be inoperable for any reason, REACTOR POWER OPERATION is permissible only during the succeeding 24 hours for the sump system or 72 hours for the air sampling system. The air sampling system may be removed from service for .a period of 4. hours for calibration, function testing, and maintenance without providing a temporary monitor. 3. If the condition in 1 or 2 above cannot be met, an orderly shutdown shall be initiated and the reactor shall be placed in the COLD SHUTDOWN CONDITION within 24 hours. D. Reller Valvea D. Relief Valves

1. - When more than one relief valve 1.

Approximately one-half of is known to be failed, an all relief valves shall orderly shutdown shall be be bench-checked or initiated and the reactor replaced with a depressurized to less than 105 bench-checked valve psig within 24 hours. The each operating cycle. relief valves are not required All 13 valves will have to be OPERABLE in the COLD been checked or replaced SHUTDOWN CONDITION. Upon the completion of every second cycle. 2. In accordance with Specification 1.0.MM,. each relief valve shc11 be manually opened until 1 thermocouples and acoustic monitors downstream of the valve indicate steam is flowing from the valve. BFN 3.6/4.6-10 Amendment 152 Unit 3

3.6/0.6 AASIS j i 3.6.0/4.6.C (Cont'd) l i suggest a reasonable margin of safety that such leakage magnitude would i not result from a crack approaching the critical size for rapid 1 propagation. Leakage less than the magnitude specified can be detected reasonably in a matter of a few hours utilizing the available leakage i detection schemes, and if the origin cannot be determined in a reasonably short time, the unit should be shut down to allow further investigation and corrective action. The two spm limit for coolant leakage rate increase over any 24 hour period is a limit specified by the NRC (Reference 2). This limit applies only during the EUN mode to avoid being penalized for the expected coolant leakage increase during pressurizat. ion. The total leakage rate consists of all leakage, identified and unidentified, which flows to the dryvell floor drain and equipment drain sumps. The capacity of the drywell floor aump pump la 50 gpm and the capacity of the drywell equipment sump pump is also 50 gpm. Removal of 25 spm from either of these sumps can be accomplished with considerable margin. Etitznntan 1. Nuclear System Leakage Rate Limits (BFNP FSAR Subsection 4.10) 2. Safety Evaluation Report (SER) on IE Bulletin 82-03 L 6.D/4. 6 D..._ R elitLY1111a 4 To meet the safety basis, 13 relief valves have been installed on the -unit with a total capacity of 83.77 percent of nuclear boiler rated steam flow. The analysis of.the worst overpressure transient, (3-second closure of all main steam line isolation valves) neglecting the direct scram (valve position scram) results in a maximum vessel pressure which, if a neutron flux scram is assumed considering 12 valves OPERABLE, results in adequate margin to the code allowable overpressure limit of 1,375-psig. To meet operational design, the analysis of the plant isolation transient (generator load reject with bypass valve failure to open)'shows that 12 of the 13 relief valves limit peak system pressure to a value which is well'below the allowed vessel overpressure of 1,375 pais. Experience in relief and safety valve operation shows that a testing of 50 percent of the valves per year is adequate to detect failures or deteriorations. The relief and safety. valves are benchtested every second operating' cycle to ensure that their setpoints are within the 1 1 percent tolerance. The relief valves are tested-in place in accordance with Specification 1.0.MM to establish that they will open and pass steam. i BFN 3.6/4.6-30 AMEN 0 MENT NO.141 Unit-3 ~

3.6/4.6 EASES 3.6.D/4.6.D (Cont'd) The requirements established above apply when the nuclear system can be pressurized above ambient conditions. These requirements are applicable at nuclear system pressures below normal operating pressures because abnormal operational transients could possibly start at these conditions such that eventual overpressure relief would be needed. However, these transients are much less severe, in terms of pressure, than those starting at rated conditions. The valves need not be functional when the vessel head is removed, since the nuclear system cannot be pressurized. The relief valves are not required to be OPERABLE in the COLD SHUTDOWN CONDITION. Overpressure protection is provided during hydrostatic tests by two of the relief valves whose relief setting has been established in conformance with ASME Section XI code requirements. The capacity of one relief valve exceeds the charging capacity of the pressurization source used during hydrostatic testing. Two relief valves are used to provide redundancy. Etferences 1. Nuclear System Pressure Relief System (BFNP FSAR Subsection 4.4) 2. " Protection Against Overpreesure" (ASME Boiler and Pressure Vessel Code, Section III, Article 9) 3. Browns Ferry Nuclear Plant Design Deficiency Repost--Target Rock Safety-Relief Valves, transmitted by J. E. Gilliland to F. E. Kruesi, August 29, 1973 3.6.E/4.6.E Jet Pumos Failure of a jet pump nozzle assembly holddown mechanism, nozzle assembly and/or riser, would increase the cross-sectional flow area for blowdown following the design basis double-ended line break. Also, failure of the diffuser would eliminate the capability to reflood the core to two-thirds height level following a recirculation line break. Therefore, if a failure occurred, repairs must be made. The detection technique is as follove. With the two recirculation pumps balanced in speed to within i 5 percent, the flow rates in both recirculation loops vill be verified by control room monitoring instruments. If the two flow rate values do not differ by more than 10 percent, riser and nozzle assembly integrity has been verified. If they do differ by 10 percent or more, the core flow rate measured by the jet pump diffuser differential pressure system must be checked against the core flow rate derived from the measured values of loop flow to core flow correlation. If the difference between measured and derived core flow rate is 10 percent or more (with the derived value higher) diffuser measurements will be taken to define the location within the vessel of failed jet pump nozzle (or riser) and the unit shut down for repairs. If the potential blowdown flov BFN 3.6/4.6-31 Amendment 152 Unit 3

3.6/4.6 BASES 3.6.E/4.6.E (Cont'd) area is increased, the system resistance to the recirculation pump is also reduced; hence, the affected drive pump will "run out" to a substantially higher flow rate (approximately 115 percent to 120 percent for a single nozzle failure). If the two loops are balanced in flow at the same pump speed, the resistance characteristics cannot have changed. Any imbalance b1 tween drive loop flow rates would be indicated by the plant process instrumentation. In addition, the affected jet pump would provide a leakage path past the core thus reducing the core flow rate. The reverse flow through the inactive jet pump would still be indicated by a positive differential pressure but the net effect would be a slight decrease (3 percent to 6 percent) in the total core flow measured. This decrease, together with the loop flow increase, would result in a lack of correlation between measured and derived core flow rate. Finally, the affected jet pump diffuser differential pressure signal would be reduced because the backflow would be less than the normal forward flow. A nozzle-riser system failure could also generate the coincident failure of a jet pump diffuser body; however, the converse is not true. The lack of any substantial stress in the jet pump diffuser body makes failure impossible without an initial nozzle-riser system failure. 3.6.F/4.6.F Recirculation Pumo OoeratiSD Steady-state operation without forced recirculation will not be permitted for more than 12 hours. And the start of a recirculation pump from the natural circulation condition will not be permitted unless the temperature difference between the loop to be started and the core coolant temperature is less than 75'F. This reduces the positive reactivity insertion to an acceptably low value. Requiring the discharge valve of the lower speed loop to remain closed until the speed of the faster pump is below 50 percent of its rated speed provides assurance when going from one-to-two pump operatfor. that excessive vibration of the jet pump risers will not occur. 3.6.C/4.6.0 Structural Integrity The requirements for the reactor coolant systems inservice inspection program have been identified by evaluating the need for a sampling examination of areas of high stress and highest probability of failure in the system and the need to meet as closely as possible the requirements of Section XI, of the ASME Boiler and Pressure Vessel Code. The program reflects the built-in limitations of access to the reactor coolant systems. It is intended that the required examinations and inspection be completed during each 10-year interval. The periodic examinations are to be done during refueling outages or other extended plant shutdown periods. BFN 3.6/4.6-32 Amendment 152 Unit 3

3.6/4.6 RMIS 3.6.G/4.6.0 (Cont'd) Only proven nondestructive cesting techniques will be used. More frequent inspections shall be performed on certain circumferential pipe velds as listed in Section 4.6.0.4 to provide additional protection against pipe whip. These velds were selected in respect to their distance from hangers or supports wherein a failure of the veld would permit the unsupported segments of pipe to strike the dryvell vall or nearby auxiliary systems or control systems. Selection was based on judgment from actual plant observation of hanger and support locations and review of drawings. Inspection of all these velds during each 10-year inspection interval vill result in three additional examinations above the requirements of Section XI of ASKE Code. Hererences 1. Inservice Inspection and Testing (BFNP TSAR Subsection 4.12) 2. Inservice Inspection of Nuclear Reactor Coolant Systems, Section XI, ASME Boiler and Pressure Vessel Code 3. ASME Boiler and Pressure Vessel Code, Section III (1968 Edition) 4. American Society for Nondestructive Testing No. SNT-TC-1A (1968 Edition) l y 1 i l BFN 3.6/4.6-33 Unit 3 _}}

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