Text

Amends 128,123 & 99 to Licenses DPR-33,DPR-52 & DPR-68, Respectively,Revising Tech Specs to Correct Inconsistencies & Typos & Add New Surveillance Requirements
	ML18030B234
Person / Time
Site:	Browns Ferry
Issue date:	03/31/1986
From:	Muller D; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML18030B235	List: ML18030B234; ML18030B236;
References
	DPR-33-A-128, DPR-52-A-123, DPR-68-A-099 NUDOCS 8604100469
	Download: ML18030B234 (37)
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UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-259 BROWNS FERRY NUCLEAR PLANT UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.

128 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 October 1, 1985, 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:

86041004b9 860331 PDR ADOCH, 05000259 P

PDR

0 1

'1 I

(2)

Technical S ecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.

128, are hereby incorporated in the license.

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

3.

This license amendment is effective 90 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION BMR Project Directorate ¹2 Division of BMR Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance:

March 31, 1985

0 I

ATTACHI'1ENT TO LICENSE AtiENDI'1ENT NO.

128 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.

108 157 179 185 227 262 2.

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

Section Pa "e '.lo.

6.2 Review and Audit 333 6.3 Procedures 338 6.4 Actions to be Taken in he Ev.nt of a Reportable Occurrence in Plan= Operation 346 6.5 Act'ons ro be taken in the Limit is Exceeded 6.6 Stat'on Op rating Records Event a Safety 346 346 6.7 Reporting Requiretents 6.6 Minimum Plant Staffing 349 359 Amendment No. 9g,

128,

4 J

~

SAFETY LIMIT FUEL CLADDING INTEGRITY LIMITING SAFETY SYSTEM SETTING 2.1 FUEL CLADDING INTEGRITY b,

For no combination of loop recirculation flow rate and core thermal power shall the APRM flux scram trip setting be allowed to exceed 1204 of rated thermal power.

(Note:

These settings assume operation within the. basic thermal hydraulic design criteria.

These criteria are LHGR<13.4 kw/ft for BxB, BxBR, and PBxBR fuel, MCPR limits of Spec 3.5.k. If it is hetermi.ned that either of these design criteria is being violated during operation, action shall be initiated within 15 minutes to restore operation within prescribed limit~.

Surveillance requirements for APRM scram setpoint are given in specification 4.5.L.

Co The APRM Rod block trip setting shall be:

S~<

(0.66M +425)

I where:

SR

= Rod block setting in percent of rated thermal power (3293 Mwt)

H

= Loop recirculation flow rate in percent of rated (rated loop recirculation flow rate equals.

34.2 x 10~ ib/hr)

Amendment No.

7t%, N, 128,

Table 4.2.J'eismic Monitori.

Instrunert Surveillance Re uirements INSTRUMENT CHANNEL CHECK CHANNEL CHANNEL FUNCTIONAL TEST CALIBRATION b.

Ce Unit l reactor bldg. floor slab Ei. 621.25 Diesel-generator bldg base slab da Monthl+

~Monthly+

TRIAXIALTDE HISTORY ACCELMRAPHS a.

Unit l reactor bl

. base slab El.

51.0)

Monthly" 6 months 6 months 6 months I/A BIAXIA1, SEIc:C

'id~TCHES Unit 1 rea~tnr bid ba e

lab Jilt I e.

Uni~treaet r bid ba e

lab Monthly

!4onthlp-Monthlg 6 months 6 months 6 months once/operating cycle once/operating cycle once/operating cycle THIk~r PrMk AC.EL 'GPJ.PHS a.

U-l PSCGf IO" i oe i1.. 62 b.

U l MR';!1 16' e

EL.

80.0l NA 12 months 12 months 12 months N/A N/A N/A

Except seismic switches

LIHITIN<I CONI)ITI(!NS< FOil O'I'.IIA'I'li)N 3.5.F R<<actor Cora! Isolation Coolio~

2. If chc RCICS is inoperable, the reactor may remain in operation for a period not to exceed 7 days if the HFCIS fs operable during such time.

SURVEILI <<NCE.,EQIJI Rl'.MFNTS 4.5.F Reactor Core Isolation Coolln 2.

When it. is determined chat the RCICS in inoperable, t.he HPCIS shall be demonstrated co be operable i<<<mediately.

3. If specifications 3.5.F.1 or 3.5'.F.2 arc noc mat, an order)y sh<<tdoun shall be initiat<<d and cha reactor shall bc dupzrssurizced to less than 122 psig uithin 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ".

G.

Automat lc Dc~rcssuricatinn

~Ss tcm~niil)~S C.

Automatic Da rassurization 1.

Four of th<! six valves of thc A<<c<<<nn!.Ir Dcl<rcssuri-cntic<< System shall be operable:

(1) prfor to a startup from a Cold Condition, or, (2) uhcnaver there is irra-dint.cd fuel. in the reactor v<.ssel and the reactor

.vessel prcssure is greater than 105 prig, except as spccfffcd I.n '3.5.C.2 and 3.5.0.3 below.

2 ~ lf cbr<<<

~f the::I.. Al)S valves are knouo to br.

in<.<pabla of automat)<:

>pot;<< I <<<<,

tl<e rcaci.or m <y ram.<l>> ln npcra-Cfon for a p<<rin<I <<<<I.

Lo

< x<<.

<<<t

<< d<<)"<, pr<<v I.dc<1 the liPCI sy::t<m I ) <<p<;r,<l>le.

(Note tl<at tha I<r<!n::urc rcl.icf function of. these vnlvrs is;<ss<<rcd by section 3.6.l) ol'hasa speal.l'frat ious a<<<l that this aper f fleo<:ion on) y

<)pp iles co cb<<

ADS f<<ncti<<<<.)

If mora than thrra ol'b<< six ADS valves arc knoun to ba fnrap-nblc of outomacic oprration, an fmm<!<If'<cc <<r<l<<rly sbutdnun sh;<ll b<! fnfcl.<c<<<l< <:lth thc res<)cor fn a ho<. sb<<ld<>un con-dition ln 6 I<<<<<rn '<<<0.005 hours 2.97619e-5 weeks 6.849e-6 months .

157 1

During each operating cycle the follouing tests shall. be performed on the ADS:

a.

A simulated automatic actuation test shall be performed prfor to scarcup after each refueling ouc-age.

Manual surveillance of the relief valves is covered, in 4.6.D.2.

2.

When it is dctcrmincd that three of the six ADS valves arc incapable of automatic operation, the HPCIS shall be drmonscrated to bc opcrablc lmmrdfaccly and daily thcreaftcr as long as Specification 3.5.C.Z applies.

Amendment No.

$9, 128,

LIHITING CONDITIONS FOR OPERATION SURVEILLANCE RFQUIRE~NTS 3 '

PRIMARY SYSTKH BOUNDARY 4 ~ 6 PRIMARY SYSTEM BOUNDARY 6.

Mhenever the reactor is critical, the limits on activity conc'entra-tions in the reactor coolant shall not exceed the equilibrium value of 3.2 uc/gm of dose equivalent*

I-131.

This limit may be exceeded following power transients for a maximum of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

Ouring this activity transient the iodine concentrations shall not exceed 26 uCi/gm whenever the reactor is critical.

The reactor shall not be operated more than 5 percent of its yearly power operation under this exception for the equilibrium activity limits.

1f the iodine concentration in the coolant exceeds 26 >Ci/gm, the reactor shall be shut

down, and the steam line isolation valves shall be closed immediately.

That concentration of k-131 which alone would produce the same thyroid dose as the quantity of total iodines actually present.

6.

Addi tional coolant samples sha)1 be taken whenever the reactor activity exceeds one percent of the equili-brium concentration

,specified in 3.6.8. 6 and one of the following conditions are met:

a.

Ouring startup b.

Following a significant power change**

c.

Following an increa e

in the equilibrium off-gas level exce<<ding 10,000 uCi/sec (at the steam Jet air ejector) within a 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months period.

d.

whenever the equilibrium iodine limit specified in 3.6.8. 6 is exceeded.

The additional coolant liquid samples shall be taken at 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months intervals for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , or unti 1 a stable iodine concentration below the limiting value (3.2 Ci/

gm) is established.

However, at

'east 3 consecutive samples ahall be taken in all cases.

An isotopic ana1ysis shall be performed for each

sample, and quantitative measurements made to determine the dose equiva1ent 1-131 concentration.

If the total iodine activity of tne sample is below 0.32 uci/gm, an isotopic analysis to determine equivalent 1-131 is not required.

8 fre For the purpose of tt 's s<<cti "n o sa-

~'n sa. p i ing

requency, a significan: power excharge is defined as a change exceeding 15" o, rated power in less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

179 Amendment No. 128,

LXMXTXNG CONDITIONS FOR OPEPATXON SURVEILLANCE REQUIREMENTS 3.6 PRI.lARY SYSTEM BOUNDARY H.

Seismic Restraints, Sup orts, and Snubbezs 1.

During all modes of operation except Cold Shutdown and Re-fuel, and seismic zestraints,

supports, and snubbers shall be operable except as noted in 3.6.H.2 and 3.6.H.3 below.

All safety-related snubbers aze listed in Surveillance Instr'uction BF SI 4.6.H 1

g -2, 2.

With one or more seismic restraint

, support, or snubber inoperable; within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months replace or restore the inoper-able seismic restraint(s),

support(s).,

or snubber(s),

to OPERABLE status and perform an engineering evaluation on the attached component or dec are the attached system inopezable and follow the appzopriate LIMITING CONDITION statement for that system.

3.

If a seismic restraint,

support, or snubber (SRSS) is determined to be inoperable while the reactor is 'n the shutdown or refuel mode, that SRSS shall be made opera".e or replaced prioz to reactor startup.

If the inoperab' SRSS is attached to a system that is required OPERABLE during the shutdown or refuel mode, the appropriate LIMITING CONDITIONS statement for that system shall be followed.

4. 6 PRIMARY SYSTEM BOUNDARY H.

Seismic Restraints Suooorts and Snubbers The surveillance requirements of paragraph 4.6.G are the only zequirements that app'y to any seismic restraint oz support other than snubbers.

Fach safety-related snubber shal1 be demonstrated OPERABLE BY pe rformanco of the following augumented inservice inspection program and the requirements of Soecif ication 3. 6. H/4. 6.H.

These snubbers are listed in Surveillance Instructions BF SI 4.6.H-1 and -2.

l.. Inspection Groups The rnubbers may be cate-gorized into two ma)or groups based on whether t'e snubbers are accessible or inaccessible during zeactor operation.

These major groups may be further subdivided into groups based on design, envir-onment., or other features which may be exoected to affect the opezability of the snubbers within the group.

Each group may be inspected independently in accordance with 4.6.H.2 through 4. 6.H.9.,

2.

Visual Insnection Schedule.

and Lot Siac The fizst insezvice visual inspection of snubbers not previously included in these technical specifications and whose visual inspection has not been performed and documented previously, shall be performed within six months for accessible snub-bers and before resuming power after the first refueling outage Amendment No. 75, gH, 128, 185

LINITIHG CONDITIONS FOR OPERATION SVRVEILLANCE REQVZREtKNTS 7

COfP'P I Hh1 EAT SYST 15 3 ~

mm Q, 7 COHTAI HMEHT 5 YST E."IS

~ab i \\ ca ~bi iit Applies to the operating status o f the pr imar y and second a" y con alnment systems.

Ob>>eCt 1 v' To assure the ln egri+y of the pr mary and secondary COnt>>blnmen>>

SyStemS

~

SC'e Cl e l Ca t l On A

1icabi lit Applies to the primary and second a ry containment integrity.

0~beet iv To verify the inte-,'rity of the pr'mary and secondary conta inme nt.

Soecl flcatlon A ~

P rimar Cont ainment At any time that the irradiated fuel is in the reactor vessel, and the nuclear system is pressurized above atmosphezic pressure or work is being done which has the potential to drain the vessel, the pressure suporession pool water level and temperature shall be main ained Mlthin the follming limits.

a.

M1nimum water level ~

-6.25" (differential pressure control

>0 psid) 1 ~

Pressure Suppress cn Chamber a.

The suppression chamber water level be checked once per day.

Ilhenever heat is added to the'uppressionoool by testing of the ECCS or relief valves the pool temperature shall be continually monitored and shall be observed and logged every 5

minutes until the heat addition is terminated.

-7'5" (0 psid differen-tial pressure control) b.

Naximum water level ~

llI 227 Amendment No. 75, 128,

I

~

TABLE 3.7.E Fi'I? WlY CollTAIi".8:IT ZSOLATI il VALVES MlZCH MLS TIE SUFrrZSSIO'L POOL l'lA~ MVEL T~ PIRATE Valve 12-738 12-741 43-20A 43-2SB 43-2gA 43-293 2>>1143

,1-14 71-32 71-580 71-592 73"23 73-24 73-o03

3-60/

"4-722 75-57 75-58 Valve Zdentii'ication AuxiliaryBoiler to RCZC AuxiliaryBoiler 4o RCZC RHR Suppression Chamber Sample Lines RHR Suppression Chamber Sample Lines R:K Suppression Chamber Sample Lines RHR Suppression Chamber Sample Lines Demineralized Mater RCIC Turbine Exhaust RCIC Vacuum Pump Discharge RCZC Turbine Exhaust RCZC Vacuum Pump Discharge HEI Turbine Exhaust HPCZ Turbine Exhaust Drain ITurbine Exl:aust HPCZ Exhaust Drain RHR Suppression Chamber Drain Suppression'hamber Drain 262 Amendment No. gf, 128

'AS Rfoy (4

~o Cy 0O g

o~

n ++*++

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-260 BROGANS FERRY NUCLEAR PLANT UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.

123 License No.

DPR-52 1.

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

A.

The application. for amendment by Tennessee Valley Authority (the licensee) dated October 1, 1985, 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 Technical Specifications contained in Appendices A and 8, as revised through Amendment No. 123, are hereby incorporated in the license.

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

3.

This license amendment is effective 90 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

Attachment:

Changes to the Technical Specifications Date of Issuance:

March 31, 1986 Daniel R. Muller, Director BWR Project Directorate 82 Division of BWR Licensing

ATTACHNENT TO LICENSE ANENDt1ENT NO. 123 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.

108 157 179 185 227 253a 260 262 2.

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

Sect.i on Pau.e iVo.

6.3 Procedures

~

6.4 Actions to be Taken in the Event R portable Occurrence in Plant Opc. ation

~

338 of' 6.5 Actions to he Taken in the Event Saf'ety Limit is Exceederl 6.6 Station Operating Hecorhs 6.7 Reporting Requirements 6.8 ifinimum Plant Staf'fin@

346 346 349 358 Amendment No. g$, 123,

Table 4.2.J Seismic Monitori Instrunent Surveillance Re uirements INSTRIKNT QIANNEL CHECK CHANNEL CHANNEL FUNCTIONAL TEST CALIBRATION TRIAXIALTIME HISTORY ACCEUSRAPHS Diesel-generator bldg base slab

~Monthl~

a.

Unit l reactor.bl

. 'base slab Zl, 51.0)

Mont~

Unit 1 reactor bldg. floor slab b.

El. 621.25 Monthl~

6 months 6 months 6 months BIAXIAL SEISMIC SWITCHES a.

Unit l reacto bid ba e

lab b.

Monthl+

. Monthl+

Monthl+

6 months 6 months 6'onths once/operating cycle once/operating cycle once/operating cycle c.

U-I core s rav " stem 14" e (U..

4l;.O TIUlCJ'AL Pf~ ACCELOGBAi~iL'3

.~6 b.

U-l K~C~S';1, 16"~ice (B..583.0'2 months 12 months 12 months N/A N/A N/A

Except seismic switches

LIHITIN><tdow>> shall be initiated and tha reactor shall bc duprrssurizced to 1css than 122 psig within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

G Automatic I)c~ressuritatinn

~Ss ram~<I))~S 1.

Four>>E t.hu six valves of thc hut,<>>aat.ir Dcprrssuri-satir>> System shall bc operable:

G.

Automatic Dc rcssuriaation 1.

During each operating cycle the following tests shall be performed on the ADS:

(1) prfor to a staztup from a Cold Condition, or, (2) whenever there is irra" diat.rd fuel. in the reactor vessel and the reactor vassal prcssure is greater than 105 pslg, except as specified I<) '3.5.G,2 and 3.5.0.3 below.

2.

If t.hre<

~f th<>>I.'.. Al)S valves ara know>> to bn In< apabla of autnmatI

<: !pc'r;<< I<a),

tl<e rracl.>>r m<<y rcm.>

i<) npcra-tlon for a prrin<l >><>t. I.o

< xi <<.<I 1

<I:<y <,

pr<><>t.dc~I the ltPCI sy>>t< m I'! up>;r.<hie

~

(Nota tl<at the I>ran>>>>re ral taf function of. I:hesa valvrs Is;<soured by scctln>>

3.6

~ I) nl'hese speal t'lent,i>>>>s a<<<l that this sl>col flc<<<.ion>>n) y applies t>> t hr ADS fu>>ct, I <>>>. )

If more titan thrrc of tbr six ADS valve n arc. k>><v~ to br. Incap-able of automatic oprration, an lmm<<: II;<t,c <>rd<'rl >> shutdown sh;<ll b<) l>>lt.l.<tr<l, with t:hc zra<:t<!r I>> a ho<.

"hutd<>wn co>>-

dlt lou I>> 6 lu>><r>>:>><d in a cold al>><tdnw>> con<lit. l>>>> tn the foilowl>>g lg hours.

157 a.

A simulatrd automatic actuation test shall be performed prior to startup after each r<<fueling out-age.

Manual surveillance of the relief valves is covered, in 4.6.D.2.

2.

When it is determined that three of Lhe six ADS valves azc incapable of automatic operation, the HPCIS shall be demonstrated to bc operable immrdlatcly and daily thereafter as long as Specification 3.5.0.2 applies.

Amendment No. g$, 123,

LIMITING CONDITIONS FOR OPERATZON SURVEILLANCE REQUIREMENTS 3 ~ 6 PRI33ARY sYsTFH l303)NRARY 4 ~ 6 PRIMARY SYSTEM BO33NDARY 6.

Whenever the reactor is critical, the limits on activity concentra-tions in the reactor coolant shall not exceed the equilibrium value of 3.2 pc/gm of dose equivalent*

1-131.

This limit may be exceeded following power transients for

.a maximum of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

During this activity transient the iodine concentrations shall not exceed 26 pCi/gm whenever the reactor is critical.

The reactor shall not be operated more than 5 percent of its yearly power operation under this exception for the equilibrium activity limits. If the iodine concentration in the coolant exceeds 26 pCi/gm. the reactor shall be shut down, and the steam line isolation valves shall be closed immediately.

That concentration of I-131 which alone would produce the same thyroid dose as the quantity of total iodines actually present.

6.

Additional coolant samples shall be taken whenever the reactor activity exceeds one percent of the equili-brium concentration specified in 3.6.8.6 and one of the following conditions are met:

a.

During startup b.

Following a significant power change**

c.

Following an increase in the equilibrium off-gas level exceeding 10,000 33Ci/sec (at the steam jet air ejector) with'in a 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months period.

d.

Whenever the equilibrium iodine limit specified in 3.6.8..6 is exceeded.

The additional coolant liquid samples shall be taken at 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months intervals for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , qr until a stable iodine concentration below the limiting value (3.2 Ci/

gm) is established.

However, at least 3 consecutive samples ahall be taken in all cases.

An isotopic analysis shall be performed for each

sample, and quantitative measurements made to determine the dose equivalent 1-131 concentration.
If the total iodine activity of the sample is below 0.32 pci/gm; an isotopic analysis to determine equivalent I-131 is not required.
For the purpose of this section frequency a 51gnificant power exchange ls defined as a change exceeding
)5Z of rated power in less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .
179 Amendment Ho. 123.
~
~
LIMITING CONDITIONS FOR OPERATION 3.6 PRIMARY SYSTEM BOUNDARY H.
Seismic Restraints, Sup orts, and Snubbers 1.
During all modes of operation cxcepr. Cold Shutdown and Re-fuel, and seismi.c restraints,
supports, and snubbers shall be operable except as noted in 3.6.H.2 and 3.6.H.3 below.
All sa fety-related snubbers aze listed in Surveillance Insrruction BF SI 4.6.H.
2, With one or morc seismic zestraint
upport, oz snubber inoperable; ui.thin 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months replace or zestore thc inoper-able seismic restraint(s),
support(s),
or snubber(s),
to OPERABLE status and perform an engineering evaluation on the atrached component or declare the attached system inoperable and follow the appropriate LIMITING CONDITION statement for t,hat system.
3.
If a seismic restraint,
support, or snubber (SRSS) is detcrnined to be inoperable uhile the reactor is in the shutdown or refuel mode, that SRSS shall be made opcrablc or replaced prior to reactor srartup.
If the inoperable SRSS is artached to a system that is zequired OPERABLE during the shutdown or rei'uel mode, the appropriate LIMITING CONDITIONS statement for that system shall be followed.
SURVEILLANCE REQUIREME.':TS
4. 6 PRIMARY SYSTEM BOUNDARY H.
Seismic Restraints
. Supports and Snubbers The surveillance zequirements of paragraph 4.6.G are rhe only requirements that apply to any seismic restzaint or support other than snubbers.
Fach safety-related snubber sh D be demonstrated OPERABLE SY pcrfnrmancc of thc following augumcntcd in"crvicc in:. scctior:
program and thu rtquLrcm uts v'.
Specification 3.6.H/4.6.H.
These snubbcrs arc listc! in Surveillance Instzuct iona BF SI 4 ~ 6 H -1 and -2.
1.
Inspection Groups The snubbers may be cate-gory.ted into two major groups based on whether the snubbers aze accessible or inaccessible during reactor operation.
These ma)or groups may be further subdivided into groups based on design, envir-
onment, or other ieatur uhich may bc exoccrcd ro affect the operability of t'e snubbers wi hin the group.
Each group may bv inspected independently in accordance with 4.6.H.2 through 4.6.H.9.
2.
Visual Ins ection Schedule.
and Lot Size The first insczvice visual inspection of snubbcrs not previously included in rhcse tcchnical specifications and who..c visual. inspecr.ion has not been performed and documented previously, shall be periorncd within six months for accessible snub-bers and before resuming po~er after the first refueling outage 185 At'amendment No. 87, 123,
L'HITINC CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3 ~ 7 Cot~g IHHENT SY. TE.IS 4, 7 COHTAI NHENT 5 YST EHS
~A iiica ~bi ]it Applies to the operatin'~ status of the primary and secondary con ainment systems.
Cb~ective To assure th': in egrity of the pz mary and secondary containment systems.
Soeci icat1 on 4 olicabilit Applies to the primary and secondary containment integri ty.
O~bect iv.
To verify the inte",'rity of "the prima y and secondary.
containment.
Soeci f s cation 4
At any time that the irradiated fuel is in the reactor vessel, and the nuclear syst,em is pres sur ized above atmospheric pressure or work is being done which has the potential to drain the vessel, the pressure suporess'on pool ~a~er level and temperature shall be main ained within the fo lowing limits a.
Minimum water level ~
-6.25" (differential pressure control
>0 psid)
A P rimar Cont a inment P essure Suppress cn Chamber a.
The suopression chamber water level be checked once per day.
Whenever heot is added to the suppression oool by testing of the fCCS or relief valves the pool temperature shall be continual ly mon) tored and shall be observed and logged every 5
minutes unt$ 1 the heat addition is terminated.
-7 25" (0 psid differen-tial pressure control) b.
Maximum water level ~
ill 227 Amendment No. gl, g$, 123,
TABLE 3.7.A (Continued)
~Grou Valve Identification Number of Power 0 crated Valves Inboard
. Outboard Maximum Operating Time (sec.)
Normal Position Action op Initiating
~S1 nal Drywell hP air compressor suction valve (FCV-64-139)
Drywell hP air compressor discharge valve (FCV-64-140)
Drywell CAM suction valves (FCV-90-254A and 254B) 10 10 10 SC SC Drywell CAM discharge valves (FCV-90-257A and 257B) 10 GC Drywell CAM suction valve (FCV-90-255) 10
li,ilve eo-2"i>le n) r>>r no-257 A n8- ">7B TABL"" 3.7.D (Continued)
Valve identification Radiation Monitor Suction Radiation Monitor Suction Radiation Monitor Discharye Radiation Monitor Discharge Amendment No. Ns 82>
HS 260
~
~
TABLE 3.7.E ESI1~JLRY COCLE'K."E ISOLATZO/1 VAL".ES H1IICif HQEIlATE B LUi T1E SUPPRESSIO?J POOL M~ LEVEL Valve 12-738 12-""1 43-2"A 43-2",3 43-2)A 43-2oB 2-1143 71-14 71""2 7] (j,lo 71-gc)2 73-23 73-24 73-6O3 73-609 74-722 7J"27 Vr-5'3 Valve Identification Au"iliaryBoiler to PCIC Auxiliary BoQer to RCIC R1$ Suppression Chamber Sample Lines RliR Suppression Cha~:er Sa-.iple Lines P Gl Q<ypression Chamber Sample Lines RiiR Suppression Ch mber Sample Lines Deminerali.@ed Hater RCIC Turbine Exhaust RCIC Vacuum Pump Discharge RC C Turbine Exhaust RCIC Vacuum Pump Dfscharpe IGCI Turbine Exhaust HEX Turbine Exhaust Drain HPCZ. Turbine Exhaust 1GCZ Exhaust Drain RHR Suppression Chamber Drain Suppression Chamber Drain 262 Amendment No.
5$, 123,
I
'p,S AEgy P0 0O
/yIP +**++
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-296 BROWNS FERRY NUCLEAR PLANT UNIT 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 99 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 October 1, 1985, 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-68 is hereby amended to read as follows:
(2)
Technical S ecifications The Technical Specifications contained in Appendices A and 8, as revised through Amendment No.
99
, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective 90 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Daniel R. Muller, Director BWR Project Directorate ¹2 Division of BWR Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance:
March 31, 1986
ATTACHMENT TO LICENSE AMENDMENT NO.
FACILITY OPERATING LICENSE NO.
DPR-68 DOCKET NO. 50-296 Revise Appendix A as follows:
1.
Remove the following pages and replace with identically numbered pages.
105 120 161 188 198 231 264A 279 2.
The marginal lines on these pages denote the area being changed.
Table
~.2 'J seismic Nonitori nst ent Surveillance R
irements INSTR UNEÃl'RIAXIAL TINE HISTORy ACCEIDGRAPBS a.
Unit 1 reactor bldg base slab fEl 519-0) b.
Unit 1 reactor bldg. floor slab (El. 621.25) c Qiesel-generator bldg. base slab (El. 565 5)
BIAXIALSEISNIC SNITCHES a.
Unit 1 reactor bldg. base slab b.
Unit 1 reactor bldg. base slab c.
Unit 1 reactor bldg. base slab CHANNEL CHECK Honthly~
Monthly~
Monthly
Nonthly~
l5onthlye CHANNEL PU?KTIONAL TEST 6 months 6 months 6 months 6 months 6 months 6 months CHANNEL CALIBRATICH NA once/operating cycle once/operating cycle once/operating cycle ti TRTPJI.!
)'l.'AK ACCE).KiRAPilS c.
'! a. c:.'~cv sr"tcc
. e a.
14" n; e
El.
i.~,): 0
?%
12 months 12 months 12 months N/A N/A N/A
Except seismic switches
LIHITIHG COHDITIOHS FOR OPERATIOH SURVEILLAHCE RE'QUIRZHEHTS 3.3 REACTIVITY CONTROL 4.3 REACTIVITY CONTROL Control rods with scram times greater than those permitted by Specification 3.3.C.3 are inoperable, but if they can be inserted with control rod drive pressure they need not be disarmed electrically.
Control rods with a failed
<<Full-in<< or
<<Full-Out <<
position switch may be bypassed in the Rod Sequence Control System and considered operable if the actual rod position is known.
These rods must, be moved in sequence to their correct, positions (full in on insertion or full out on withdrawa1).
C ~
When it is initially determined that a
control rod is incap-able of normal insertion a test shall be con-ducted to demonstrate that the cause of the malfunction is not a failure in the control rod drive mechanism.
If this can be demonstrated an attempt to fully insert the control rod shall be made.
If the control rod cannot be inserted and an investigation has demonstrated that the cause of failure is not a failed control rod drive mechanism collet housing, a shut-down margin test shall be made to demonstrate under this condition that the core can be made subcrit ical for any reactivity condition during the remainder of the operating cycle with the analytically determined, highest worth control rod capable of withdrawal>-
fully withdrawn, and all other control rods capable of insertion fully inserted.
d.
The control rod accumulators shall be determined operable. at least once per 7
days by verifying that the pressure and level detectors axe not in the alarmed condition.
120 Amendment No-
$5, 99
~
LIMITING CONDITIONS FOR OPEPATION SURVEILLANCE REQUIREMENTS l ~ 5 CORE AND CONTAINMENT COOLING S YSTEMS 4
5 CORE AND CONTAINMENT COOLING SYSTEMS G.
Automatic De ressurization G.
Automatic De ressurizacion 1.
Pour of the six.v'alves of the Automatic'Depressuri-zation System shall be operable:
(1) prior to a startup from a Cold Condition, ory (2) whenever there is irra-diated fuel in the reactor vessel and the reactor vessel pressure is greater than 105 psig, except as specified in 3.5.G.2 and 3.5.G.3 below.
2. If three of the six ADS valves are known to be incapable of automatic operation, the reactor may remain in opera-tion for' 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 thac this specification only applies to the ADS function.) If more than three of the six ADS valves are known to be incap-able of automatic operation, an immediate orderly shutdown shall be initiated, with che reactor in a hot shutdown con-di:ion in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in a cold shutdown condition in the following 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months .
1.
During each operating cycle the following tests shall be performed on the ADS:
a.
A simulaced automatic actuation test shall be performed prior to scartup after each refueling out-age.
Hanual surveillance of the relief valves is covered, in 4.6.D.2.
2.
When it is determined that three of the six ADS valves are incapable of automatic operation the HPCIS shall be demonstrated to be operable immediately and daily thereafter as long as Specification 3.5.G.2 applies.
161 Amendment No. 3, jig. 99.
LIMITING CONDITXONB FOR OPERATION SURVEILLANCE RFQUZREMENTS t.6 PRIMARY SYSTEM BOUNDARY 4 ~ 6 PRIMARY SYSTEM BOUNDARY 3 ~
At steaming rates greater than 100 F 000 lb/hr, the reactor water quality may exceed specification 3.6.B.2 only for the time limits specified below.
Exceeding these time limits of the following maximum quality limits shall be cause for placing the reactor in the cold shutdown condition.
Conductivxty time above 2 umho/cm825~C-4 weeks/year.
.,Maximum Limit 10 gmho/cm$ 25oC b.
Chloride concentration time above 0.2 ppm-4 we< "..s/year.
Maximum Limit-0 ~ 5 ppm ~
3 ~
a ~
b.
During startup Following a significant power change<<<<
ci d ~
Following an increase in the equilibrium off-gas level exceeding 10,000 uci/sec (at the steam jet air ejector) within a 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months period.
whenever the equilibrium iodine limit specified in 3 ~ 6 ~ B ~ 5 is exceeded.
Additional coolant samples shall be taken whenever the reactor activity exceeds one percent of the equilibrium concentration specified in 3.6.8.5 and one of the following conditions are met:
<<<<For the purpose of this section on sampling frequency, a significant power exchange is defined as a
change exceeding 15% of rated power in less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .
188 Amendment No.
99,
LIMITING CONDITIONS FOR OPERATION SURVEIL ANCE REQUIREMENTS 3.6 PRLfARY SYSTEM BOUNDARY H.
Seismic Restraints
Supports, and Snubbers
4. 6 PRIMARY SYSTEM BOUNDARY I
H.
Seismic Restraints Supports.
and Snubbers 1
During all modes of operation except Cold Shutdown and Re-fuel, and seismic restraints,
supports, and snubbcrs shall-be operable except as noted in 3.6.H.2 and 3.6.H.3 below.
All safety-related snubbcrs are listed in Surveillance Instruction BF SI 4.6.H.
2.
With one or more seismic restraint
, support, oz snubber inoperable; within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months replace or restore the inoper-able seismic restraint(s),
support(s),
or snubber(s),
to OPERABLE status and perform an engineering evaluation on the attached component or declare the attached system inoperable and follow the appropriate LIMITING CONDITION statement for that system.
3.
If a seismic restraint,
support, or snubber (SRSS) is determ.'ned to be inoperable while the reactor is in the shutdown or refuel mode. that SRSS shall be made opera=.e or replaced prior to re.ctor startuo.
If t'..~ inoper-='e SRSS is attached tc a system that is required OPERABLE during the shutdown or refuel mode, the appropriate LIMITING CONDITIONS statement for that system shall be followed.
The survcil'ance requirements of paragzaph 4.6.G are the onlv requirements tha app'y to anv seismic restza'nt or support other than snubbers.
Each safety-related snubber shal-be demonstrated OPERABLE BY performance of the following augumented inservice inspect'cn program and the requirements oi Specification
3. 6.H/4.6.H.
These snubbers aze listed in Surveillance Instructions BF SI 4.6.H-1 and -2..
1.
Inspection Groups The snubbers mav be cate-gorized into two major groups based on wnether t'e snubbers are accessible or inaccess'ble during reactor operation.
These major groups may be further subdivided into groups based on design, envir-
onment, or other features which may be exoected to affect the operability o:
the snubbers within the group.
Each grouo may bc inspected independ<<nt) v in accordance with 4.6.H.2 through 4.6.H.9.
2.
V'sual Inspection.
Schecule.
and Lot Size 198 Thc first inservice v'sual inspection of snubbers not previously included in these technical specifications and whose visual inspection has not been performed and documented previous).y, shall be oerformed within six months for acccssiblc snub-bcrs and before resuming power after the f'rst refueling outage Amendment Wo. 8,
$5, 99,
0 LIMITIHG COHDITIOHS FOR OPERATION SURVEILLANCE REOVZ REAGENTS 3m 7 CO!P~Q I HREHT SYnaT ~AS 4 ~ 7 COHTAI HH EHT SYSTEMS
~ao isea~bilit Applies to the operatin" status of the primary and secondary containment systems.
Ob~>c TO aSSure the In egrzty Of the p
mary and secondary containment systems.
Soeci fication A olicab'it Applies to the primary and secondary containment integrity.
O~b'eet ive To verify the integrity of the primary and secondary conta inme nt.
Speci fication A ~
~pimar.Contain.. nt At any time that, the i rradiated fuel is in the reactor vessel, and the nuclear system is pressurized aLove atmospheric pressure or work is being done which has the p~tential to drain the vessel, the pressure suporession pool ~ater level and temperature shall be maIntained Mithin the folloMing limits.
a.
Minimum water level
-6.?5" (differentinl pressure control
>0 psid)
-7.25" (0 psid differ entiol pressure contr< I)
P rimar Cont a inment 1
preeaure a
uoroeaeor, Chamber a.
The suppress)on chamber Hater level be checked once per Whenever heat
<s added to the suppress)on oool by testing of the ECCS or re1)ef valves the pool temperature shall be continually monitored and shall be observed and logged every 5
minutes unt) I the heat addft1on
<s terminated.
b.
Maximum water level
~
-1" 231 Amendment No.
$g, 99,
0 I
iv J
TABLE 3.7.A (Continued)
Gro~u Valve Identification Number of Power Operated Valves Inboard Outboard Haximum Operating Normal Time (Sec.)
Position Action on Initiating
~Sf nal o
6 Torus Oxygen Sample Line Valves-Analyzer B
(FSV-76-63, 64)
NA Note 1
SC 6
'U Drywell Hydrogen Sample Line Valves-Analyzer B
(FSV-76-59, 60)
NA Note 1
SC 6
'V Drywell Oxygen Sample Line Valves-Analyzer B
(FSV-76-61, 62)
NA Note 1
SC.
Sample Return Valves-Analyzer B (FSV-76-67, 68)
NA GC RCIC Steamline Drain (FSV 6A, 6B)
GC RCIC Condensate Pump Drain (FCV-71-7A, 7B)
SC HPCI Hotwell pump discharge isola-tion valves (FCV-73-17A, 178)
HP'CI steamline drain (FCV-73-6A, 6B)
TIP Guide Tubes (5) 1 per guide tube NA SC GC GC NOTE:
1:
Analyzers are such that one is sampling drywell hydrogen and oxygen (valves from drywell open valves from torus closed) while the other is sampling torus hydrogen and oxygen (valves from torus open - valves from drywell closed)
PR
'JJ';E COi."FAIi".=,'.T ISOLAiTIZ<<VALVESM)IC)f T~U!I)D'iTZ BiLC') T<<Z SUPx ~SSIO:<<POOL )1A'M LRVL
'1.>l.vn 12
$ 33 12-. <<il 43-2BA 43-2'-?A 43-2W 71-14 71-32
~1-5".O 1-5/2 73-23
'7Q
( 3-603
".3-'>Og
<<i--22
5"57 "5-5 "3 Vcl'e Tr)cnti.fioat(on Auxiliary ))oiler to RCXC Auxiliary Boiler 4o FiCIC R))R Suppre" ion Chamber Sample Lines R)B Suppression Cham'r Sample Linc=
R:B Suppression Chamber Sample Lines RhR Suppression Chamber Sample Lines F~CIC uro inc Pxhoust RCIC Vacu"a Pump Dischor'e RCIC urb inlc ZYJlau t RCIC Vncu.n Bx".p DischcrSe IKITurb).ne K~".oust 1)ZCI Turbine exhaust Drcin
.iCI Turb).nc i '~ou
))FCI Exhaust Drain R!B Suppression Chamber Drain Suppression Chamber Drain 279 Amendment No. g, 75, 99,

ML18030B234

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