ML18029A192

Amends 112,106 & 80 to Licenses DPR-33,DPR-52 & DPR-68, Respectively,Revising Group 1 Isolation Valve Reactor Water Level Setpoint to 370 Inches Above Vessel Zero

ML18029A192
Person / Time
Site:	Browns Ferry
Issue date:	09/19/1984
From:	Vassallo D Office of Nuclear Reactor Regulation
To:	Tennessee Valley Authority
Shared Package
ML18029A193	List: ML18029A192 ML18029A194
References
DPR-33-A-112, DPR-52-A-106, DPR-68-A-080 NUDOCS 8410260171
Download: ML18029A192 (31)

Text

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UNITED STATES NUCLEAR REGULATORY COMIVllSSION WASHINGTON, O. C. 20555 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-259 BROWNS FERRY NUCLEAR PLANT, UNIT I AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 112 License No. DPR-33 1.

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

A.

The application for amendment by Tennessee Valley Authority (the licensee) dated February 22, 1984, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),

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

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

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

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

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

2.

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

DPR-33 is hereby amended to read as follows:

(2)

Technical S ecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 112, are hereby incorporated in the license.

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

84i026017i 8409i9 PDR ADOCK 05000259 P

PDR

,0 S

3.

This license amendment is effective as of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

~<~@+/~

Domenic B. Vassallo, Chief Operating Reactors Branch

//2 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance:

September 19, 1984

ATTACHMENT TO LICENSE AMENDMENT NO. 112 FACILITY OPERATING LICENSE NO. DPR-33 DOCKET NO. 50-259 Revise Appendix A as follows:

1.

Remove the following 'pages and replace with identically numbered pages.

11, 55, 111, 112, 254, 277 2.

The marginal lines on these pages denote the area being changed.

SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING I.l FUEL CLADDING INTEGRITY

2. 1 FUEL CLADDING INTEGRITY Power Trans I e>>t.

D. Power Transient Tri Settin s

To ensure that thc Safety Limits established in Specification

1. 1.A arc nn c receded, each required scram shall bc initiated by its expected scram sil;nal.

Thc Safety Limit shall bc assumed to be exceeded when scram is accomplished by means other than thc expected scram sil;nal.

1.

Scram and isola-tion (PCIS groups 2,3,6) re~ctor low water level

2. Scram--turbine stop valve c1osure Scram--turbine cont.rol valve fast clo"ura or turbine trip 2 538 in.

above vessel zero 5

10 per-cent valve closure 550 psig 4.

Scram--low con-denser vacuua!

5.

Scram--main steam line isolation 2 23 inches Hg vacu'~l 5

10 per-cent valve

.. =losur~

6.

Main steam isola-tion valve closure

--nuclear system low pressure

~ 825 'sig C.

lh actnr Vessel 14atcr Level C. Natcr Level Tri Scttin s

<tl>>>>ever t hc ro is irrad Ia tcd ital in the reactor vc'snc'1

~

thc wnter level shall not be

)css than 17.7 in. above thc top of the normal active fuel xone.

Core spray and LPCI actuation

'reactor low water level 2,

HPCI and RCIC actuation reac-tor low water level

? 378 in above vessel xero 2

470 in.

above vessel zero 3,

Main steam isola tion valve closure--reactor low wa ter level 2 378 in.

above vessel zero Amendment No

Minixua yo.

Instrunent Channels Operable CABLE 3 2 A PRIllAlly CotfTAIfofEPI'ND RECTOR BUlLOKfiG ISOKATIO!4 INSTRUHKS ATIOR Remz rks Instrument Channel-Reactor Lov Mater Level

(.6)

Instrument Channel-Reactor 1)fqh Pressure Instrument Channel Reactor Loy Mater Level (LIS 3-56$ -Dr SM a 1) 538" above vessel zero A or CB and r) 100

~

15 psig a 378" above vessel zero h

Below trip setting does the folio+ing:

Initiates Reactor Building Isolation b.

Inftfatea Primary Conzafnment Isolation (Groups 2, 3, and 6)

Initiates SGTS Co 1.

Above trip setting isolates the shutdovn cooling suction valves of the RHR system l.

Below'rip setting initiates

.".zfn Steam Lfne Iso la tion Instrument Channe 1 Sigh Dryvell Pressure (6]

(PS-64 56A D) 2.5 pals A or IB and EJ l.

Above trip setting does the follovfng 1 Znftiates Reactor Building Isol ation b

Initiates Prfmar:r Contain>>nt Isolation

. c.

Initiates SQTS Instruaent Channel-Bfgh Radiation Nafn Steam Kine Tunnel t6)

Instrument Channel-Lov Pressure Hain Stear Line Instrument Channel-Sigh tlow Hain Steam Line 3 times normal rated Cull poMer background (g3) 2 ra25 psfg (t) 6 feoj OC rated steam fice B

1 Above trip setting tnftfates wfn Steam Line Isolation l.

Befog trip setting initiates Hain Steam Line Isolation 1.

Above trip settfng fnftfates Hain Steam Line Isolation

3.2 BASES ln addition to reactor protection instrumentation which initiates a

reactor scram, protective instrumentation has been provided which initiates action to mitigate the consequences of accidents which are beyond the operator's ability to control, or terminates operator errors before they result in serious consequences.

This set of specifications provides the limiting conditions of operation for the primary system isolation function, initiation of the core cooling systems,"

control rod block and standby gas treatment systems.

The objectives of the Specifi-cations are (i) to assure the effectiveness of the protective instru-mentation when required by preserving its capability to tolerate a

single failure of any component of such systems even during periods when portions of such systems are out of service for maintenance, and (ii) to prescribe the trip settings required to assure adequate per-formances~

When necessary, one channel may be made inoperable for brief intervals to conduct required functional tests and calibrations.

Some of the settings on the instrumentation that initiate or control cor'e and containment cooling have tolerances explicitly stated where the high and low values are both critical and may have a substantial effect on safety.

The set points of other instrumentation, where only the high or low end of the setting has a direct bearing on safety, are chosen at a

'evel away from the normal operating range to prevent i'nadvertent actua-tion of the safety system involved and exposure to abnormal situations.

Actuation of primary containment valves is initiated by protective instru-mentation shown in Table 3.2.A which senses the conditions for which iso-lation is required.

Such instrumentation must be available whenever primary containment integrity is required.

The instrumentation which initiates primary system isolation is connected in a dual bus arrangement.

The low water level instrumentation set to trip at 177.7" (538" above vessel zero) above the top of the active fuel closes isolation valves in the RHR System, Drywell and Suppression Chamber exhausts and drains and Reactor Water Cleanup Lines (Groups 2 and 3 isolation valves).

The low reactor water level instrumentation that is set to trip when reactor water level is 109.7" (470" above vessel zero) 'above the top of the active fuel (Table 3.2.B)'rips the recirculation pumps and initiates the RCIC and HPCI systems.

The RCIC and HPCI system initiation opens the turbine steam supply valve which in turn initiates closure of the respective drain valves (Group 7).

The, low water level instrumentation set to trip at 17.7" (378" above vessel

'ero) above the active fuel (Table 3.2.B) closes the Main Steam Isolation Valves, the'ain Steam Line Drain Valves, and the Reactor Water Sample Valves (Group 1).

Details of valve grouping nnd required closing times are given in Specil ica-tion 3.7.

These trip settings are adequate to prevent core. uncovery in the case of a break in the ] argest line assuming the maximum closing time.

Amendment No. '

112

3..2 BASES The low reactor water level instrumentation that is et to trip when reactor water level is 17.7" (378" above vessel zero) above the top of the active fuel (Table 3.2.8) initiates the LPCI, Core Spray

Pumps, contributes to ADS initiation; and starts the diesel generators.

These trip setting levels wer e chosen to be high enough to prevent spur ious actuation but low enough to initiate CSCS operation so that postaccident cooling can be accomplished and the guidelines of 10 CFR 100 will not be violated.

For large breaks up to the complete circumferential break of a 28-inch recirculation line and with the trip setting given above, CSCS initiation is initiated in time to meet the above criteria.

The high dr ywell pressure instrumentation is a diverse signal to the water level instrumentation and in addition to initiating CSCS, it causes isolation of Groups 2 and 8 isolation valves.

For the breaks discussed above, this instrumentation wi3 1 initiate CSCS operation at about, the same time as the low water level instrumentation; thus the results given above are applicable here also.

~

'I'enturis are provided in the main steam lines as a means of measuring steam flow and also limiting the loss of mass inventory from the vessel during a steam line break accident.

The primary function of the inst'ruraentation is to detect a break i'n the main steam line.

For the worst case accident, main steam line break outside the drywell, a trip setting of 140$ of rated steam flow in con)unction with the flow limiters and main steam line valve closure, limits the mass inventory loss such that fuel is not uncovered, fuel cladding temperatures remain below 1000 F and release of radioactivity to the environs is well below 10 CFR 100 guidelines.

Reference Section-14.6.5 FSAR.

Temperature monitoring instrumentation is provided in the main steam line tunnel to detect leaks in these areas.

Trips are provided on this instrumentation and when exceeded, cause closure of isolation valves.

The setting of 200oF for the main steam line tunnel detector'is low enough to detect leaks of the order of 15 gpm; thus, it is capable of coverin~ the entire spectrum of breaks.

For large breaks, the high steam flow instrumentation is. a backup to the temperature instrumentation.

In the event of a loss of the reactor building ventilation system, radiant heating in the vicinity of the main steam lines raises the ambient temperature above 200 degrees F.

The temperature increases can cause an unnecessary main steam line isolation and reactot scram.

Permission is

,provided to bypass the temperature trip for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to avoid an unnecessary plant transient and allow performance of the secondary containment leak rate'est or make repairs necessary to regain nor mal ventilation.

Hi-h radiation monitors in the rain team ne tunnel have been g

provided to detect gross fuel, failure as-, 1n"the control rod drop accident.

Mith the establishqd nominal setting of 3 times nodal background and main steam line isolation valve closure, fission product release is limited so that 10, CFR 100 guidelines are not exceeded for this accident.

Reference Section 14.6.2 FSAR.

An alarm with a nominal setpoiat of 1.5 x normal full-power background is provided also.

Pressure instrumentation is provided to close the main steam a~via

~..

valves in Run Mode when the main steam li'ne pressure drops below 825 psig.

112-Amendment No.

1 1 0, 112'

~

S ~ 1<<V'el

~ 4 ~

~ ~,

i

~ ~4 HOTES FOR TABLE 3.7.A Key'.

0 Open C

Cl,osad SC << Stays Closed GC << Coca Closed Note.'Isolation groupings are as follovs.':oup 1:

The valvea in Croup 1 are actuated by any one of the follovfng condf tfone:

I ~

Reactor Vos ~c)

Lov Ms ter I.evol (378")

2.

Mein Stcanl fne Hfgh Red iatfen 3.

Mafn Steanifne High Plov 4.

Mein StoaeLlfne Space High Teaperature 5.

Msfn Steanlino Lov Pressure Group 2:

The valves in Croup 2 are actuated by any of the follovfng conditions:

Croup 3:

8 e a c t0r Ve o s e I Lov Ma t or Level

( 5 3 8")

2.

High Dryvell Pressure The valves in Croup 3 are actuated by any of the iollovfng conditions:

I ~

Renctor Lnv Mater Level (538")

Rehc tor Mater Cleanup 5 ye tea Xigh T empers tur e 3,

Reactor Mater Cleanup Systen High Drain Teaporaturo Croup 4:

The valvcr jn Croup 4 are actuated by any of the folloving condi t fons HPC1 Steer lino Specs High Tenperature 2.

KPCI S t ear 1 f n ~ High Plov 3.

HPCI Stra+line Lov Pressure Croup 5:

The valves in Croup S are actuated by any of the folloving condition:

RCI" Stea~lfne Space High Taapersturo 2.

RCIC Stoeolin>>

High Fiov 3.

RCIC 5 t caal fns l.ov Pros sure Group 6:

The valves in Croup 6 aro actuated by any of tha follovfng conditions:

1

~

Reactor Vessel Lov Mater Lavsi (538")

2.

High Dryvell Prcssure 3.

Reactor guilding Ventilation High kadfatfon 254 Amendment No.

, 112

BASES 0

~Grou l - Process lines are isolat'ed by reactor vessel low water level (378") in order to allow for removal of decay heat subsequent to a scram, yet isolate in time for proper operation of the core standby cooling systems.

The valves in group l, except the reactor water sample line valves, are also closed when process instru-mentation detects excessive main steam line flow, high radiation, low

pressure, or main steam space high temperature.

The reactor water sample line valves isolate only on reactor low water level at 378" or main steam line high radiation.

~Gzou 2 Isolation valves are closed by reactor vessel low water level (538") or high drywell pressure.

The group 2 isolation signal also "iso-late>>" the reactor building and starts the standby gas treatment system.

It is not desirable to actuate the group 2 isolation signal by a transient or spurious signai.

1Q6.

)R-52

~Grou 3 Process lines are normally in use and it is therefore t

r ore not esizable to cause spurious isolation due to high drywell pressure resulting from non-safety related causes.

To protect the reactor from a possible pipe brcak in the system, isolation is provided by high temperature in the clean-up system area or high flow through the inlet to the cleanup system.

Also, since the vessel could potentially be drained through the cleanup

system, a low level isolation is provided.

Gzou s

4 and 5 Process lines are designed to remain operable and mitigate the consequences of an accident which results in the isolation of other process lines.

The signals which initiate isolation of Groups 4'nd 5

process lines are therefore indicative of a condition which would render them inoperable.

~Grnu 6 Lines ara conaected to the primary containment but not directl to the reactor vessel.

These valves are isolated on reactor low water 1 evul (538" h '

ev<<(

),

iigh drywull piessure, or reactor building ventilation hi h radiation which would indicate a possible accident and necessitate, primary containment isolation.

~Gzou 7 - Process lines are closed only on the respective turbine steam supply valve not fully closed.

This assures that the valves are not open when 11PCI or RCIC action is required.

\\

Griiup 8 - l.inL ((ravi 1 i>>1, in-curn probL ) is iso iatcd on high dryweii pressure or rencltyr iow water level (538").

This is to <<ssuze that this 1i>>e

<ious not provide a

1'aknge path when containment pressure i>r reactor water level indicates a possible accident condition.

nd 51 ave nt by The maximum closure time for the automatic isolation vlaves of the s

o t>e pr mary containment and reactor vessel isolation control system have been selected in consideration of the design intent to prevent core uncovering following pipe breaks outside the primary containment and the need to contain released fission products following pipe breaks inside the primary containment.

In satisfying this design intent, an additional margin has been included in specifying maximum closure times.

This margin permits identification of degraded valve performance pri'or to exceeding the design closure times.

Amendment No.

1

, 112 277

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UNITEDSTATES NUCLEAR REGULATORY COMIVIISSION WASHINGTON, D. C. 20555 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-260 BROMNS FERRY NUCLEAR PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. l06 License No. DPR-52 1.

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

A.

The application for amendment by Tennessee Valley Authority (the licensee) dated February 22, 1984, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),

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

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

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

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

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

2.

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

DPR-52 is hereby amended to read as follows:

(2)

Technical S ecifications The Tqchnical Specifications contained in Appendices A and B, as revised through Amendment No. 106, are hereby incorporated in th) license.

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

3.

This license amendment is effective as of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

~+4P=s~c-Domenic B. Vassallo, Chief Operating Reactors Branch 82 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance:

September 19, 1984

ATTACHMENT TO LICENSE AMENDMENT NO. '106 FACILITY OPERATING LICENSE NO. DPR-52 DOCKET NO. 50-260 Revise Appendix A as follows:

1.

Remove the following pages and replace with identically numbered pages.

11, 55, 111, 112, 254, 277 2.

The marginal lines on these pages denote the area being changed.

SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING 1.1 FUEL CLADDING INTEGRITY

2. 1 FUEL CLADDING INTEGRITY PoM<<r Trans 1 unu D. Power Transient Tri Settin s

To cnsurc that the Safety Limits cstablishcd in Specification

1. 1.A arc Ant <<xccedud

~

each rcquircd scram shall bc initiated by its expcctcd scram s innal.

Thc Sa Ic ty Limit shall bc assumed to be exceeded when scram is accompli hcd by means other than thc,expected scram sibnal.

l. Scram and isola tion (PCIS groups 2,3,6) reactor lou watc r level

2. Scram--turbine stop valve closure Scram--turbine control valve fast clo"ura or turbine trip 538 in.

above vessel xexo 5

10 per-cent valve closure 550 psig 4.

Scramlow con-densex vacuum 5,

Scram--main steam line isolat ion 2 23 inches Qg vacu'~l 5

10 per-cent valve

.. =losur~

6.

Hain steam isola-tion valve closure nuclear system low pressure

~ 825 psi; C.

Ib actor Vessel Water Level C. Water Level Tri Settin s

~thin<<vur ther<< is irradiated intel in t.hc reactor vc snc 1, the un'<<r 1 eve 1 shall nut be less than 17.7 in. above thc top of the normal active fuel xone.

Core spray and LPCI actuation reactor low eater level 2.

HPCI and RCIC actuation"-reac-tor low water level 378 in.

above.

vessel xero 470 above vessel xero 3.

Main stcam isola-tion valve closure--reactor low water level h '378 in.

above vessel xero Amendment No.

. 106

Hinixun Ko.

Instrument Channels Operable PASLE 3

2 'A PRINARV COHTAIHHEpr AHD REACTO!l BUILD1146 ISOIATION IHSTRUNEJTATIO14 Leve Sett i e.

rks Instrument Channel-Reactor Lou Mater Level (6)

Instrument Channel-Reactor High Pressure Instrument Channel Reactor Lov Mater Level (LIS-3-56A Di'M 01) 538" above vessel zero A or fa and E) 100

~ )5 psig

? 378" above vessel zero 1.

Belov trip setting does the tolloving:

.a.

Initiates Reactor Building Isolation b

Initiatea Primary containment Isolation ((:roups 2, 3, and 6) c Initiates SOTS l.

Above trip sat'ting isolates the shutdovn coolinsuction valves of the RHR system I ~

Belov trip setting initiates sain Steam Line Isolation Instrument Channel Bigh Dryvell Pr'essure

.(6)

(PS-6a-56h D) 5 I.S psfs I or IB an4 t) l.

Above trip setting does the fol loving I Initiates Reactor Building Isolation'~Initiates Pr imar'r Conta inguen Isolition

. c Ini.tiates SOTS

> l3)

Instrument Channel-Bigh Radiation Rain Steam Line Tunnel

{6)

Instrument Channel-Lov Pressure Hain Steam Line Instrument Channel-Qigh Plov Hain Steam Line 3 times normal rate4 B

full poMer background (]3) qp psig f4)

$ loOt of rate4 steam flow B

t ~

Above trip setting initiates uin Steam Line Isolation 1.

Belov trip setting initiates 1ai'n Steam Line Isolation 1.

Above trip setting initiazes Hain steam Line Isolation

3.2 BASES In addition. to reactor protection instrumentation which initiates a

reactor scram, protective instrumentation has been provided which initiates action to mitigate the consequences of accidents which are beyond the operator's ability to control, or terminates operator errors before they result in serious consequences.

This set of specifications provides the limiting conditions of operation for the primary system isolation function, initiation of the core cooling systems, control rod

. block and standby gas treatment systems.

The objectives of the Specifi-cations are (i) tc assure the effectiveness of the protective instru-mentation when required by preserving its capability to tolerate a

single failure of any component of such systems even during periods when portions of such systems are out of service for maintenance, and (ii) to prescribe the trip settings required to assure adequate per-formance.

When necessary, one channel may be made inoperable for brief intervals to conduct required functional tests and calibrations.

Some of the settings on the instrumentation that initiate or control core and containment'ooling have tolerances explicitly stated where the high and low values are both critical and may have a substantial effect on safety.

The set points of other instrumentation, where only the high or low end of the setting has a direct bearing on safety, are chosen at a level away from the normal operating range to prevent inadvertent actua-tion of the safety system involved and exposure to abnormal situations.

Actuation of primary containment valves is initiated by protective instru-mentation shown in Table 3. 2.A which, senses the conditions for which iso-lation is required.

Such instrumentation must be available whenever primary containment "ntegrity is required.

The instrumentation which initiates primary system isolation is connected in a dual bus arrangement.

The low water level instrumentation set to trip at 177.7" (S38" above vessel zero) above the top of the active fuel closes isolation valves in the RHR System, Drywell and Suppression Chamber exhausts and drains and Reactor Water Cleanup Lines (Groups 2 and 3 isolation valves).

The low reactor water level instrumentation that is set to trip when reactor water level is 109.7" (470" above vessel zero) above the top of the active fu 1 (Table 3.2.B) trips the recirculation pumps and initiates the RCIC and HPCI systems.

The RCIC and HPCI system initiation opens the turbine steam supply valve which in turn initiates closure oi the respective drain valves (Group 7).

The low water level instrumentation set to trip at 17.7" (378" above vessel zero) above the active fuel (Table 3.2.B) closes the Main Steam Isol.ation Valves, the Main Steam Line Drain Valves, and thu Reactor Water Sampl.e Valves (Group 1).

Details of valve grouping and required closing times are given in Specifica-tion 3.7.

These trip settings are adequate to prevent core uncovery in the case of a break. in the largest line assuming the maximum closing time.

Amendment No.

, 106

- ~2BASES

[

. The low reactor water level instrumentation that is set to trip when reactor water level is 17.7" (378" above vessel zero) above the top of'he active fuel (Table 3.2.B) initiates the LPCI, Core Spray

Pumps, contributes to ADS initiation, and starts the diesel generators.

These trip setting levels were chosen to be high enough to prevent spurious actuation but low enough to initiate CSCS operation so that postaccident cooling can be accomplished and the guidelines of 10 CFR 100 will not be violated.

For large breaks up to the complete circumferential break of a 28-inch recirculation line and with th trip setting given above, CSCS initiation is initiated in time to meet w

e the above criteria.

The high drywell pressure instrumentation is a diverse signal to the water level instrumentation and in addition to initiating CSCS, it causes solation of Groups 2 and 8 isolation valves.

For the breaks discussed above, this instrumentation will initiate CSCS operation at about the same time as the low water level instrumentation; thus the results given above are applicable here also.

Uenturis are provided in the main steam lines as a means of measuring steam flow and also limiting the loss of mass inventory from the vessel during a steam line break accident.

The primary function of the instrumentation is to detect a break in the main steam line.

For the worst case accident.,

main steam line break outside-the drywell, a tri

,a rp.

setting of 140$ of rated steam flow in con)unction with the flow limiters and main steam, line valve closure, limits the mass inventory loss such that fuel is not uncovered, fuel cladding temperatures remain below 1000 F and release of radioactivity to the environs is well below 10 CFR '100 guidelines.

Reference Section 14.6.5 FSAR.-

Temperature monitoring instrumentation is provided in the main steam line tunnel to detect leaks in these areas.

Trips are provided on this instrumentation and when exceeded, cause closure of isolation valves.

.The setting of 200 F for the main steam line tunnel detector is low enough to detect leaks of the order of.15 gpm; thus, it is capable of'overing the entire spectrum of breaks.

For large "brea1rs, the high steam flow instrumentation is a backup to the temperature instrumentation.

Zn the event of a loss of the reactor building ventilation system,-radiant heating in the vicinity of the main steam lines raises the ambient temperature above 200 degrees F.

The temperature increases can cause an unnecessary main steam line isolation and reactor scram.

Permission is provided to bypass the temperature trip for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to avo'id an unnecessary plant transient and allow performance of the secondary containment leak rate test or make repairs necessary to regain normal ventilation.

High radiation monitors in the main steam ne tunnel have Ween provided to detect gross f'uel failure as ln the control rod dr op accidents Mith the established nominal setting of' times nodal background and main steam line isolation valve closure, fission product release is limited so that 10 CFR 100 guidelines are not exceeded f'r this accident.

Ref'erence Sec-ion 14.6.2 FSAR.

An alarm with a nominal setpoint of 1.5 x normal full-power background is provided also.

Pressure instrumentation is provided to c'lose the main steam isolation valves in Run Mode when the main steam line pressure drops below 825 psig.

112 Amendment 1lc.~ lg t i

HOTES FOR TABLE 3 loA Key':

0 ~ Open C

Closed SC ~ Stay's Closed CC ~ Goes Closed Hote.'solation groupings are as foliose.

Croup 1:

The valved in Group 1 are actuatod by any one of thc folloving condi tions:

1, Roactor Ydsde)

Lcni Mater Level (378")

2.

Hain Stcamline High Radiation 3.

Hain Stenmline High Flov 4,

Hain Stoamlknd Space High Temperature 5,

.Hain Steamlina Lov Presdure

. Group 2:

Thc valves in Group 2 are actuated by any of the foiling condi ions:

1, Reactor Veooel Lov Mater Level (538")

2.

High Drywall Pressure Croup 3:

The valves in Croup 3 are actuated by any of the following conditions:

1 Rene CO, Lnv Mater Level {538")

2.

Reactor Mater Cleanup System High Temperstur

~

3, Reactor Mater Cleanup Systen High Drain Tempdraturn Croup 4:

The valves Jn Group 4 are actuated by sny of thd folloving conditions:

1.

HPCI Steaml.ine Space High Temperature 2.

HPCI Steamli~d High FIOv 3.

HPCI S teamlind Lov Pressure Croup 5:

The valves in Croup 5 are actuated by any of tha folloving cond' ion-:

1.

RCI" S t ca~line Space High Tempera cure 2.

RCIC Stnseline High Flov 3.

RCIC Stedmlind Lov Proasure Croup 6:

The valves in Croup 6 are actuated by any of thd following conditions:

1.

Reactor Vessel Lov Mater Level (538")

2, High Dryvell Pressure 3.

Reactor Building Ventila tion High Radiation

)54 Amendment No.

, 106

~Grou l - Process

.lines are isolated by reactor vessel low water level (378") in order to allow for removal of decay heat subsequent to a scram, yet isolate in time for proper operation of the core standby cooling systems.

The valves in group 1, except t.

rt.nr ton w:)ter sn>>>pl<<

line valves, are also closed when process ins tru-mentation detects excessive main steam line flow, high radiation, low

pressure, or main steam space high temperature.

The reactor water sample line valves isolate only on reactor low water level at 378" or main steam line high radiation.

~Grou 2 - isolation valves are closed by reactor vessel low water level (538") or high drywell pressure.

The group 2 isolation signal also "iso-lates" the reactor building and starts the standby gas treatment system.

It is not desirable to actuate the group 2 isolation signal by a transient or spurious signal.

Grn~u3 Process lines are normally in use, and it is tl)erefore not desirable to cause spurious isolation due to high dryw<<1 l pressuri r<>1 ti>>>;

f ri)>>> no>> "of<<ty r<< la teil cn>>ses.

To pro tee t the rr ar tor from a por s il> 1 <<p i pi

~

br<<nk in the system, isolation is provided by high temperature in the cion>>-

up syst<<m area or high flow through the inlet to thc cleanup system.

hl o, since the vessel could potentially be drained through the cleanup system, a low level isolation is provided.

the consequences of an accident which results in the isolation of other

, process lines.

The signals which initiate isolation of Groups 4 and 5

process lines are therefore indicative of a condition which would render them inoperable.

v

~Grou 6 Lines are connected to the primary containment but not directly to the reactor vessel.

These valves are isolated on reactor low water level (538"), high drywell pressure, or reactor bu'ilding ventilation high radiation which would indicate a possible acc'ident and necessitate primary containment isolation.

~Grou 7

Process lines are closed only on the respective turbine stcam supply valve not fully closed.

This assures that the valves are not open when HFCI or RCIC action is required.

I

~Gr u>

6 i.in (trave ting in-rura probe>

is i:ulated un high dryueil pressure 6>r reacti>r luw water level (538").

This is ti) <<ssure that this line dues nut provide a'eak'>ge path when containment prcssure vr reactor water level indicates a possible accident condition.

The maximum closure time for the automatic isolation vlaves of the primary containment and reactor vessel isolation control system have been selected in consideration of the design intent to prevent core uncovering following pipe breaks outside the primary containment and the need to contain released fission products following pipe breaks inside the primary containment.

In satisfying this design intent, an additional margin has been included in specifying maximum closure times.

This margin permits identification of degraded valve performance prior to exceeding the design closure times.

Amendment No.

, l06 277

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O. C. 20555 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-296 BROWNS FERRY NUCLEAR PLANT, UNIT 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.

8Q, License No. DPR-68 1.

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

A.

The application for amendment by Tennessee Valley Authority (the licensee) dated February 22, 1984, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),

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

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

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

The issuance of this amendment will not he 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.

2.

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

DPR-68 is hereby amended to read as follows:

(2)

Technical S ecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 80., 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 the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

~~p/~~~-('!>>

Domenic B. Vassallo, Chief Operating Reactors Branch 82 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance:

September 19, 1984

ATTACHMENT TO LICENSE AMENDMENT NO.

80.

FACILITY OPERATING LICENSE NO. DPR-68 DOCKET NO. 50-296 Revise Appendix A as follows:

l.

2.

Remove the following pages and replace with identically numbered pages.

13, 57,

108, 109, 266, 294 l

The marginal lines on these pages denote the area being changed.

SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING l.l FllEL CLADDING INTEGRITY 2 l FUEL CLADDING INTEGRITY Bi Power Translc<<t D. Power Transient Tri Settin<'s To ensure tlrat thc Safety Limits cstablishcd Cn Specification

l. l.A are Ant cxr c<!<lcd ~

each required scram shall bc initiated by its expected scram signai.

The Safety Liml.t shall be assumed to be cxcecded when scram is accomplished by means other than thc expcctcd scram signal.

l. Scram and isola-ticn (PCIS groups 2,3,6) rc seto,r low wats r level

2. Scram--turbine stop valve closure 3, Scram--turbine control valve fast clo ura or turbine trip 538 in.

above vessel zerLJ 10 per-cent val;e closure 550 psip 4.

Scram--).ov con-denser vacuum 5.

Scram--main steam line isolation 6,

Main steam isola<<.

tion valve closure

--nuclear system low pressure 2 23 inche5 Hg vacu a 5

10 per-cent valve "losur'~

~ 825 psi-,

C. Reactor Vessel

'i<at.cr Level C. Water Lev<!l Tri Settings N<<a<ever there is irradiat<!<l fuel iA the run<:Lor vessel,

~ thc water level shall not be loss thari 17. 7 in. above thc top of the normal active fuel Z.otic ~

Co re a pr ay and LFCI actuation--

'reactor lov 'vater level HFCI and RCIC actuation--reac-tor low water level 37S in above vessel'ero 2

470 in.

above vessel r.ero Main stcam isola-tion valve closure"-reactor low water level

? 378 in.

above vessel xero l3 Amendment No.

, 80

.'1ini-..u=.

Instr..Ont Cl~hnnels Ct:arable per TT.- 5vs ('.)' l )

nction Lave Settf ctfo 1

TABLE 3 ~ 2 A

PRIHARY coRTAINHEHT AHD RKAcToR BUILDIHG IsozATzoH zHSTRUHERTATzoH emrks Instrument Channel-Reactor Low Mater Level

{6)

Instrument Channel Reactor Hiqh Pressure Instrument Channel Reactor Lo~ Mater Level

[LIS-3-56A-Di SM l1) 5364 above vessel sero A or fB and Rl 100 4

15 psig 2 378" above vessel sero A

1.

Belov trip setting does the folloving:

a.

Initiates Reactor Building Isolation b.

Znitfates Primary Containment Isolatfon (firoups 2, 3, and 6)

c. 'Initiates SGTS Above trip setting isolates the shutdown cooling suction valves of the RHR system.

1.

Bel~ trip setting initiates Hain Steam Line Isolation Instrument Channel-Hfc.h Dryvell Pressure

{6}

{PS-64-56A-D) 5 2.S pIZS A or fB and Ej Above tr}p setting does the follovfng'nitiates Reactor Building Isolatf on b.

Initiates Primary Containment Isolation c.

Initiates SGTS Instrument Channel-Hiqh Radiation Hain Steam Line Tunnel f6)

Instrument Channel-LoM Pressure Hain Steam Line Instrument Charm~ l Biqh Pfou hain sc 4~m Line 2 ( f 2) instrument Channel Nafn Steam Line Tunnel High Temperatur' times normal rated B

full pover background (l3) 2 BlS psig fa}

'la0$ of rated steam flav B

2004 F Above trip setting fnitiates Hain Steam Line Isoiatfon 1.

Befog trip setting initiates Hain Steam Line Isolation 1.

Above trip setting initiates Hain Steam Z.ine Isolation 1.

Above trip set ting fnitfates Hain S earn Line Isolation

3'. 2 BASES In addition to reactor protection instrumentation which initiates a

ruac tor scram, pro tec tive ins trumenta tion has been provided which initiates action to mitigate the conse'quences of accidents which are beyond the operator's ability to control, or terminates operator errors before they result in serious consequences.

This set of specifications provides the limiting conditions of operation for the primary system isolation function, initiation of the core cooling systems, control rod block and standby gas treatment systems.

The obgectives of the Specifi-cations are (i) to assure the effectiveness o

the protective instru-mentation when required by preserving its capability to tolerate a

single failure of any component of such systems even during periods whe>> portions of such systems are out of service for mai>>tenance, and (ii) to prescribe the trip settings required to assure adequate per-formance.

When necessary, one channel may be made inoperable for brief intervals to conduct required functional tests and calibrations.

Some of the settings on the instrumentation that initiate or control core and containment cooling have tolerances explicitly stated where the high and low values are both critical and may have a substantial effect on safety.

The set points of other instrumentation, where only the high or'ow end of the setting has a direct bearing on safety, are chosen at a level away from the normal operating range to prevent inadvertent actua-tion of the safety system involved and exposure to abnormal situations.

Actuation of primary containment valves is initiated by protective instru-mei>tation shown in Table 3.2.A which senses the conditions for which iso-lation is required.

Such instrumentation must be available whenever r primary containment Integrity is required.

The instrumentation which initiates primary system isolation is connected in a dual bus arrangement.

The low water level instrumentation set to trip at 177.7" (538" above vessel ero) above the top of the active fuel closes isolation valves in the RHR System, Drywell and Suppression Chamber exhausts and drains and Reactor Water Cleanup Lines (Groups 2 and 3 isolation valves).

The low reactor water level instrumentation that is set to trip when reactor water level is 109.7" (470" above vessel zero) above the top of the active fu l (Table 3. 2.B) trips the recirculation pumps and initiates the RCIC and HPCI systems.

The RCIC and HPCI system initiation opens the turbine steam supply valve which in turn initiates closure of the respective drain valves (Group 7).

The low water level instrumentation set to trip at 17.7" (378" above vessel zero) above the active fuel (Table 3.2.B) closes the Main Steam Isolation Valves, the Main Steam Line Drain Valves, and the Reactor Water Sample Valves (Group 1).

Details of valve grouping and required closing times are given in Specifica-tion 3.7.

These trip settings are adequate to prevent core uncovery in the case of a break in the largest line assuming the maximum closing time.

108 Amendment No.

, 80

.3. 2 BASES 0

The low reactor water level instrumentation that is set to trip when reactor water level is 17.7" (378" above vessel zero) above the top of the active fuel (Table 3.2.B) initiates the LPCI, Core Spray 'Pumps, contributes to ADS initiation, and starts the diesel generators.

These trip setting levels were chosen to be high enough to pr event spurious actuation but low enough to initiate CSCS operation so that postaccident cooling can be accomplished and the guidelines of 10 CFR 100 will not be violated.

For large breaks up to the complete circumferential break of a 28-inch recirculation line and with the trip setting given above, CSCS initiation is initiated in time to meet the above criteria.

The high drywell pressure instrumentation is a diverse signal to the water level'instrumentation and in addition to initiating CSCS, it causes isolation of Groups 2 and 8 isolation valves.

For the breaks discussed above, this instrumentation will initiate CSCS operation at about the same time as the low water level instrumentation; thus the results given above are applicable here also.

I Venturis are provided in the main steam lines as

a. means of measuring steam flow and also limiting the loss 'of mass inventory from the vessel during a steam line break accident.

The primary function of the instrumentation is to detect a break in the main steam line.

For the

~orst case accident, main steam line break outside the drywell, a trip setting of 140$ of rated steam flow in conJunction with the flaw limiters and main steam line valve closure, limits the mass inventory loss such that fuel is not uncovered, fuel cladding temperatures remain below 1000 F and release of radioactivity to the environs is well below 10 CFR 100 guidelines.

Reference Section 14.6.5 FSAR.

Temperature monitoring instrumentation is provided in the main steam line tunnel.to detect leaks in these areas.

Trips are provided on this instrumentation and when exceeded,

'cause closure of isolation valves.

The setting of 200oF for the main steam line tunnel detector is low enough to detect leaks of the order of 15 gpm; thus, it is capable of covering the entire spectrum of breaks.

For large breaks, the high steam flow instrumentation is a backup to the temperature instrumentation.

In the event of a loss of the reactor building ventilation system, radiant heating in the vicinity of the main steam lines raises the ambient.

temperature above 200 degrees F.

The temperature increases can cause an unnecessary main steam line isolation and reactor scram.

Permission is provided to bypass the temperature trip for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to avoid an unnecessary

'plant transient'nd allow performance of the secondary containment leak rate test or make repairs necessary to regain normal ventilation.

High radiation monitors in the main steam I'ne tunnel have been provided to detect gross fuel failure as in the control rod drop accident.

Mith the established nominal setting o1 3 times normal background and main steam line isolation valve closure, fission product release is limited so that 10 CFH 100 guidelines are not exceeded for this accident.

Reference Sec ion 14.6.2 FSAR.

An alarm with a nominal setpoint'f 1.5 x normal full-power bacpgr ound is provided also..

109 Amendment No.

. 7

, 80

NOTES FOR TABLE 3.7 '

Key:

0 C

sc GC

~ Open

~ Closed Stays Closed

~ Goes Closed Note:

Isolation groupings are as follows:

Group The valves in Group 1 are actuated by any of the folloving conditions:

Group Group Group 3:

1.

Reactor Vessel Low Mater Level (378~~)

2.

Main Steamline High Radiation 3.

Main Steamline High Flow'.

Main Steamline Space High Temperature 5.

Main Steamline Lov Pressure The valves in Group 2 are actuated by any of the following conditions:

1 ~

Reactor Vessel Low Water Level (538'~)

2.

High Dryvell Pressure The valves in Group 3 are actuated by any of the following conditions:

1. 'Reactor Lov Water Level (538>)

2.

Reactor Mater Cleanup System High Temperature 3.

Reactor Water 'Cleanup System High Drain Temperature The valves in Group 4 are actuated by any of the follo~ing conditions:

1.

HPCI Steamline Space High Temperature 2.

HPCI Steamline High Flov 3.

HPCZ Steamline Lov Pressure Group 5:

The valves in Group 5 are actuated by any of the folloving conditions:

Group Group 6:

7 0 1 ~

RCIC Steamline Space High Temperature 2.

RCIC Steamline High Flow 3.

RCIC Steamline Lov Pressure The valves in Group 6 are actuated by any of the folloving conditions:

1.

Reactor Vessel Low Mater Level (538>)

2.

High Dzyvell Pressure 3.

Reactor'uilding Ventilation High Radiation The valves in Group 7 are automatically actuated by Amendment No. gf, 80

4 3.7.D/4.7.D Primar Containment Isolation Valves Double isolation valves are provided on lines penetrating the primary containment and open to the free space'f the containment.

Closure of one of the valves in each line would be sufficient to maintain the integrity of the pressure suppression system.

Automatic initiation is required to minimize the potential leakage paths from the containment in the event of a loss of coolant accident.

Grou~1 - Process lines ar e isolated by reactor vessel low water level.

(378") in order to allow for removal of decay heat subsequent to a scram yet isolate in time for proper operation of the core standby cooling systems.

The valves in group 1p except the reactor water sample line

valves, are also closed when process instrumentation detects excessive main steam line flow, high radiation, low pressur e, or main steam space high

'emperature.

The reactor water. sample line valves isolate only on reactor low water level at 378" or main steam line high radiation.

e

~Grou 2 - Isolation valves are closed by reactor vessel low water level (538") or high drywell pressure.

The group 2 isolation signal also "isolates" the reactor building and starts the standby gas treatment system.

Zt is not desirable to actuate the group 2 isolation signal by a transient or, spurious signal.

~Qrou 3 - Process lines are normally in use, and it is therefore

not, desirable to cause spurious isolation due to high drywell pressure r esulting from nonsafety-related causes.

To protect the reactor from a possible pipe break in the system, isolation is provided by high temperature in the cleanup system area or high flow through the inlet to the cleanup system.

Also, since the vessel could potentially be drained thr ough the cleanup system, a low level isolation is provided.

the consequences of an accident which results in the isolation of other process lines.

The signals which initiate isolation of groups 4 and 5

process lines are ther efore indicative of a condition which would render them inoperable.

~Grou 6 - Lines are connected to the primary containment but not directly.

to the reactor vessel.

These valves are isolated on reactor low water level (538"), high drywell pressur e, or reactor building ventilation high radiation which would indicate a possible accident and necessitate primary containment isolation.

~Gr ou 7 - process lines are closed only on the respective turbine steam supply valve not fully closed.

This ensures that the valves are not open when HPCIS or RCZCS action is required.

~Gr ou 2 - Line (traveling in-core probe) is isolated on high dr ywell pressure or reactor low water level (538").

This is to assure that this line does not provide a leakage path when containment pressure or reactor water level indicates a possible accident condition.

294 Amendment No.

. 80