Text

Proposed Tech Specs,Deleting Sections 6.9, Environ Qualification, 6.10, Integrity of Sys Outside Containment & 6.11, Iodine Monitoring from Table of Contents
	ML18029A905
Person / Time
Site:	Browns Ferry
Issue date:	10/01/1985
From:	; TENNESSEE VALLEY AUTHORITY
To:	;
Shared Package
ML18029A904	List: ML18029A903; ML18029A905;
References
	TVA-BFNP-TS-213, NUDOCS 8510070340
	Download: ML18029A905 (54)
	v • d • e

ENCLOSURE 1 PROPOSED TECHNICAL SPECIFICATION REVISIONS BROWNS FERRY NUCLEAR PLANT UNITS 1, 2, AND 3 (TVA BFNP TS 213) 85100'70~ egfQOi aSOCW 0500{)959 ZZR pgR t'

l J

PROPOSED CHANGES UNIT 1

Seoaioo, ~Pa "e .'lo, 6e2 Review and Audit 333 6 ' Procedures e ~ . e ~ e . . ~ ~ 338 6.4 Actions to be Taken in he Event of a Reportable Occurrence in Plant Operation 346 6.5 Act'ons to be taken in the Event a Safety Limit is Exceeded 346 6.6 Sear.'on Operating Records 346 6e 7 Reporting Requirements 349 6.8 Minimum Plant Staffing 358

SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING FUEL CLADDING INTEGRITY 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 120$ of rated thermal power.

4 (Noter These settings as'sume operation within 'the, basic thermal hydraulic design criteria. These criteria are LHGR<13.4 kw/ft for &x&,

8x&R, and P8x8R fuel, MCPR limits 3.5.k. If it is determined thatof'pec 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.

C~ The APRM Rod block trip setting shall be:

S~< (0.66H +42%)

where:

Rod block setting in percent of rated thermal power (3293 M'Ht)

Loop recirculation flow rate in perce'nt of rated (rated loop recirculation flow rate equals.

34.2 x 10,6 ib/hr)'

Table 4.2.J Seismic Monitori. Instrunent Surveillance Re uirements CHANNEL INSTRMKhT QUU5EL CHECK ~ CHANNEL FUNCTIONAL TEST CALIBRATION TRIAXIAL TDE HISTORY ACCELMRAPHS Unit 1 reactor bl . base slab El. 51 .0) Mont~ 6 months Unit 1 reactor bldg. f1oor slab

b. El. 621,25 Monthl+ 6 months Diesel-genera or bldg base slab c, Monthly+ 6 months BIAXIAL SEIS"-.C ="i. z TCHES Unit 1 reactor bid ba e lab Monthl> 6 months once/operating cycle Bello Monthly. 6 months once/operating cycle

c. Vni~treecton bid bc e lnb Monthl+ 6 months once/operating cycle TiiKM FiUK AC Es ~'~PAH{S

a. U-l P33CCrf IG"' oe El. 62 12 month's N/A

b. U-1 RASE) 16" i oe EL. 80.0 l 12 months N/A

c. V-1 core s~r'-" s stern 14" i e Vl. 5i!'.0 inn 12 months N/A

Except seismic switches x

(/

LIHITIIC COHDITIOHS FOll Ul'FRATIOH SURVEIIJ ANC F..".1'ikEHES TS 3 ~ 5.F Reactot'oro Isolation Or<olin CD 5iF Reactor Core Isolation Coolin

2. If tlic RCICS is inoperable, 2, When it is deterrincd that the the react. or may remain in RCICS is inoperable, the HPCIg operation for a period not shall be demonstrated to bc Lo exceed 7 days if the opcrabla it<asdiatcly.

HPCIS ls operable during ouch time.

3r If opccifi<:scions 3.5.F.l or 3.5.Ã.2 sr<: nnr. nret, an

'ord<'r ly sh<<rdn<rn nlinll be initiated nnd clio ronctor shall bc depress<<rixerd to less than 122 psig uithin 2C hours'.

AutomaLCc Ur.<ircssuricntion C, Aiitomatic Dc zessurication

~5st ~AIIH . S stern AUS For<r <<f thn six valves of X. During ench operiting cycle the hucnmnLic Dcpr< nsuri- the follouing tests shall be zatinn System shall be performed oa the ADS:

opcrab)c:

a.' sf<<<ulated automatic (1) prior to a scnrtup actuation test sliall b<r from a Cold Condition, performed prior co scarrrin ol' after each refueling out.-

age. Hanual surveillance (2) uhcn< ver cher<< Is irta- of thc relief valves is dinr.<'d foci'I<i t<<e rt'actor covered, in C.b.D.2.

and th< teartor 'essel vessel prix<sure is greater titan 105 p'Ig, except as spcclficd Io 3.5.C.2 and 3.5.G ~ 3 belou. 2~ Mhcn it is detcrmlnc c!<r<t three of the six ADS valves are

2. If Lh<<'r <<f tlic stx AOS valves incapnhlc o aucora c or>> ~t!ni ace knou<t Lo be Incapable of thd III<CIS shall br de<conscrare<I auccmn tie <pre: c I on, thc to bu operable isc<cdiatcry <<iid reactor <wy rcrm<in iii opera- daily thereafter as leo"'as tion for n prrirrd ooc co Specificatioa 3.5.C.2 applies.

ex<<cod l days, prov<Ca~I Lhe III'L<l syst n la operab!e.

(Hot<< Lhnt tlic I>><<:rri>>rc re t<<f fi<<i:LI n . f t licnc 1

vnlvns is;issurcd bv sect ioi< 3.6. 0 nf tlicxo spec!fin<<t lnns nnd rhnc this hp<<r rc br Incap-nbl< of nr<<ox>>ci>> npr<<ntion, eii ]<eon !I<it<<rir<lc<'ly olin<.doun shall bt Inlti.i<id. "Irh the ,157 rcacrnr Io r< hnt sh"'.<Ir'a< con-dition Irr 6 lirriirr. n>>d Iri o cold chotdni"< conrli cirro in the follouiog 18 hnurs.

Amendment No. 59

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RFQUI REMENTS 3 ' PRIMARY SYSTEM BOUNDARY 4 6 PRIMARY SYSTEM BOUNDARY

6. Whenever the reactor is critical, 6. Additional coolant the limits on activity conc'entra- samples shall be taken tions in the reactor coolant shall whenever the reactor not exceed the equilibrium value activity exceeds one of 3.2 uc/gm of dose equivalent* percent of the equili-I-131. brium concentration

.specified in 3.6.8. 6 This limit may be exceeded and one of the following following power transients for conditions are met:

a maximum of 48 hours. Ouring this activity transient the a. Ouring startup iodine concentrations shall not b. Following a significant exceed 26 uCi/gm whenever the power change**

reactor is critical. The, c. Following an increa.e reactor shall not be operated in the equilibrium more than 5 percent of its yearly off-gas level exceedinr; power operation under this 10,000 uCi/sec (at the exception for the equi librium steam jet air ejector) activity limits. If the iodine within a 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months period.

concentration in the coolant d. Whenever the equilibrium exceeds 26 uCi/gm, the reactor iodine limit specified shall be shut down, and the in 3.6.8 . 6 is exceeded.

steam line isolation valves shall be closed immediately. 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 until a stable iodine concentra tion below the limiting value (3.2 Ci/

gm) is established. However, at

'east 3 consecutive samples ahall be taken in all cases. An isotopic analysis shall be perfoneed for each sample, and quantitative measurements made to determine the dose equivalent I-131 concentration. If the total iodine activity of tne sample is below 0.32 uci/gm, ar:

isotopic analysis to determine That concentration of I-131 equivalent I-r31 is not required.

which alone would produce the same thyroid dose as the quantity

of total iodines actual'ly present. For the purpose of this section or s ..

fre uenc y, a significant power excharge is defined as a dhange 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

.C LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.6 PRLVLRY SYSTEM BOUNDARY 4. 6 PRIMARY SYSTEM BODNDARY H. Seismic Restraints Sup ores, H. Seismic Restraints Supports and Snubbers and Snubbers

1. During all modes of operation The surveillance requirements except Cold Shutdown and Re- of paragraph 4.6.G are the fuel, and seismic zestraints, only requirements that app'y supports, and snubbers shall to any, seismic restraint oz be operable except as noted support other than snubbers.

in 3.6.H.2 and 3.6.H.3 below.

All safety-related snubbers Each safety-related snubber sha 1 aze listed in Surveillance be demonstrated OPERABLE BY Instruction BF SI 4.6.H j 6 -2, performance of thc following augumented insezvice inspec"ion

2. With one or more seismic program and the requirements of restraint , support, or snubber Specification 3.6.H/4 '.H.

inoperable; within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months These snubbers are listed 'n replace or restore the inoper- Surveillance Instructions able seismic restraint(s), BP SI 4.6.H-1 and -2.

support(s)., or snubber(s), to OPERABLE status and per'form 1. Inspection Groups an engineering evaluation on the attached component or The snubbezs nay be cate-dec are the attached system gorized into two major inoperable and follow the groups based on whether the appropriate LIMITING CONDITION snubbezs aze accessible or statement for that system. inaccessible during reactor operation. These major 3~ If a seismic restraint, suppozt, groups may be further subdivided into groups or snubber (SRSS) is determined to be inoperable while the based on design, envir-reactor is in the shutdown or onment, or other features refuel mode, that SRSS shall be which may be expected to made opezab e or replaced affect the operability of prior to reactor startup. If the snubbers within the the inoperable SRSS is attached group. Each group may be to a system that is required inspected independently in OPERABLE during the shutdown accordance with 4.6.H.2 or refuel mode, the appropriate through 4.6.H.9.,

LIMITING CONDITIONS statement for that system shall be 2. Visual Ins ection Schedule.

folio~ed. and Lot Size The first insezvice visual, inspection of snubbezs 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 po~er after the fir'st refueling outage 185

LIMITIHC COHDITIOHS FOR OPERATIOH SURVEILLANCE REQVIREI'ANTS 3 y Colr g I HblE.'4T SYST .'4S 4,7 COHTAIHHEiHT SYSTEMS

~Aoi tca~bl ii t A licabi lit Applies to the operating status Applies to the primary and of the primary and secondary secondary containment containment systems. integri ty .

Cb~moct 1 <'b O~b'ectira To assure the xn egriby of the j

pz ma= and s econda. y containment systems.

To verify the inte-,'rity of the pr'ary and secondary containment.

Soccl f 1 cat lan coact f i ca ti on A. P r omar Con t a a nment At any time that the irradiated fuel is in pressure Suooressicn the reactor vessels Chamber and the nuclear syst.em is pressuri=ed a. The s upp ress) on aLove atmospheric chamber water level pressure or cwork is be checked once per being done which has day. llhenever heat th potential to is added to drain the vessel, the pressur suporession oool by the'uppression pool ~ater level and testing of the ECCS t empe r at ur e s hall be or relief valves the maantained Mxthin the pool temperature shall f ol lm ing limits. be continually monitored end shall be observed and logged every 5 minutes until the heat

a. Minimum water level addition is terminated.

-6.25" (differential pressure control

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

b. Maximum water level ~

ill 227

TABIZ 3.7.E mDARV COnAI.,"..S.rr ISOLATIO;r VALVES WtICH mrrraTE BELQf TlK SUPPBZSSXON POOL UA~ ZZVZL Valve Valve Ident'.iieet ion 12-738 Auxiliary Boiler to RClC 12-741 Auxiliary Boiler Co RCIC 43-2QA 'HR Suppression Chamber Sample Lines 43-2SB RHR Suppression Chamber Sample Lines43-29A R:K Suppression Chamber Sample Lines43-29B RHR Suppression Chamber Sample Lines 2-1143 Daminerol.ized Water

,1-14 RCIC Turbine Exhaust 71>>32 RCIC Vacuum Pump Discharge 71-580 RCIC Turbine Exhaust 71-592 RCIC Vacuum Pump Discharge 73"23 HPCI Turbine Exhaust 73-24 HEX Turbine Exhaust Drain 73-603 HECI Turbine Ed:aust 73-609 HPCI Exhaust Drain "4-722 RHR 75-57 Suppression Chamber Drain 75-58 Suppression'hamber Drain 262

PROPOSED CHANGES UNIT 2

Sectian PBRC lO 6.3 Procediires 338 6.4 Actions to be Taken in the Event of' Reportable Occurrence in Plant Operation . . . . . . . . . . . . . . . 346 6.5 Actions to be Taken in the Event a Saf'ety Limit is Exceeded . . . . . . . 346 6.6 Station Operating Records . . . . . . . 346 6.7 Reporting Requirements . . . . . . . 34/

6.8 Hinimum Plant Staff'ing 358

Table 42.J Seismic Monitori Instnment Surveillance Re uirements CHANN1Z, INSTRBKNT QIANNEL CHECK CHANNEL FUNCTIONAL TEST CALIBRATION TRIAXIAL TIME HISTORY ACCELOGRAPHS

a. Unit 1 reactor,bl . base slab El, 51 .0) Mont~ 6 months NA Unit 1 reactor bldg. floor slab

b. El. 621.25 Monthly+ 6 months Diesel-generator bldg base slab C ~ Monthly+ 6 months O

BIAXIAL SEISIGC SWITCHES

a. Unit 1 reactor bid ba e lab Monthl+ 6 months onc%perating cycle

b. Monthl+ 6 months once/operating cycle Monthl+ 6 months once/operating cycle T1UAXIAL PEAK ACCELOGRAPiPS

a. U-1 RBC(M 10" nine (El. 62 .7 ) 12 months N/A

b. U-l L>PS'cl, 16" ni e Q.. 580.0'.

12 months N/A U-1 core s rav s stem 14" 'H.. V;.0 12 months N/A

Except seismic switches

I.IHITIN<t C(NVTTII>t>S Ft)lt 4'I'.>tATli)N SVRVRII.I.WCg ."Xqt)I RI:.)tRNTS

(

I<

3.5.F R<<actor Coro Isolat:ion Coolin~

2. If tlic RCICS is lnopcrablc, the reactor may remain in 4.5.F Reactor Core Isolation Coolln 2~ When RCICS ic is determined that the in inoperable, the"HPCIS opera t ion for a period not shell be demonstrated to be co exceed 7 days if thc operable immediaccly.

HPCIS ls operable during such t:imc.

3. 'If specifications 3.5.F.1 or 3.5tF.2 arc not mct, an orderly st>>it:dow>> shall be initiated and thc reactor shall bc d<<prcssiirizced to less than 122 psig within 24 hours'.

C. Automatic Dc~rcssurization C. Automatic Dc rcssurization

~Ss t em~At)~S

1. Four of thi. six valves of 1. During each operating cycle the A>>t<>>aat.i<: Dctircssuri- the following t'eats shall be zati<<, i System sl>all be performed on the ADS:

operable:

A simulated aucnmattc prior to a staztup actuation test shall be from a Cold Condition, performed prior to start>>p or, L after each refueling out-age. Manual surveillance (2) wIien<.ver there is irra- of the relief valves is diated fuel. in the reactor covered, in 4.6.D.2.

vessel and the reactor

.vessel prcssure is greater tlian 105 psig, <;xcept as specified L>>'3.5.0.2 and 3.5.0.3 below. 2. When it is determined chat three applies.

2. lf t,lire< nf chr>>I.:< AI>S valves o f the six ADS incapalile of automatic val "cs arc operation, are known to bo In<.ipablc of the HPCIS shall be drmonstrated automat I < )prri>i'I>, tlie to be opcrablc imm<.diatcly and r<.act.or m>iy rcm.il>> In opcra- daily chcreaftcr as long as cio<> for a p<<rio><<t. t.o Specification 3.5.0. 2

< x<<cd <>
d:>y <, pr>>vide<1 t: he HFCI sy>>t<!m l.)>>)><tr,>I<le.
(Ilotc t:liat the I<re."".iurc rel,ief function of. I:tiese vnlvrs is;isa>>red hy sc<:cion 3.6.I) oI'licsc sp<<r.lt lracin>>s a<>:ion on)y applies to t.h<<ADS fiii>ac I<<ii.) If more tlian thrrc'oI'li<<six ADS vol v< .. arc known tu bc Inr. ip-ablc of, outomacic operation; an Immc<II;>t,<<<<rd<<rly sliutdnwn sh:il.l. I<<<. i>>it.t.it.<<<I, ut th ttic 157 res<>cor I>> a ho<. shutd<<wn co>>-
dition I>> 6 I>>>i>r>>;>>>d in a cold rhiicd owl> ron>> in thc followlug 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months .
LIMITING CONDITIONS FOR OPERATION 'URVEILLANCE REQUIREMENTS 3~ 6 PRIMARY SYSTEM BOUNDARY 4 6 PRIMARY SYSTEM BOUNDARY
6. Whenever the reac tor is cri ca 1, the limits on activity concentra-ti 6. Addi tiona1 coolant samples shall be taken tions in the reactor coolant shall whenever the reactor not exceed the equilibrium value activity exceeds one of 3.2 pc/gm of dose equivalent* percent of the equili-I-131. brium concentration specified in 3.6.8.6 This limit may be exceeded and one of the following following power transients for conditions are met:
.a maximum of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . During this activity transient the a. During startup iodine concentrations shall not b. 'ollowing a significant exceed 26 uci/gm whenever the power change**
reactor is critical. The c. Following an increase reactor shall not be operated in the equilibrium more than 5 percent of its yearly off-gas level exceeding power operation under this 10,000 uCi/sec (at the exception for the equilibrium steam jet air ejector) act'ivity limits . If the iodine within a 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months period .
concentration in the coolant d. Whenever the equilibrium exceeds 26 Ci/gm, the reactor iodine limit specified shall be shut down, and the in 3.6.6.'6 is exceeded.
s team 1 i ne i sol a ti on va ves 1
shall be closed immediately. 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/
is established. However, at 'm) 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 I-131 concentration.
total iodine activity of the If the sample is below 0.32 uci/gm; an isotopic analysis to determine
That concentration of I-131 equivalent I-131 is not required.
which alone would produce the same thyroid dose as the of total iodines actually quantity For the purpose of this section present.
frequency, a significant power li defined as a change exceeding exchange power in less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . 15'f 179
i
( LIMITING CONDITIONS
3. 6 FOR OPERATION PRIMARY SYSTEM BOUNDARY 4.6 SURVEILLANCE REQUIRF 1E'.:TS PRIMARY SYSTEM BOUViDARY H. Seismic Restraints, Sup orts, H. Seismic Restraints, Suo11nrts and Snubbers and Snubbers
1. During. al,l modes of operation The surveillance zequizemcnts except Cold Shutdown and Re- of paragraph 4.6.G are the fuel, and seismic restraints, only requirements that apply supports, and snubbers shall to any seismic restraint oz be operable except as noted suppo.'t other than snubbers.
in 3.6.H.2 and 3.6.H.3 belov.
All safecy-related snubbers Fach safety-relateC snubber s1aaD are lisced in Surveillance be demonstrated OPE!1ABLE BY Instruction BF SI 4.6.H. pcrfnrmancc of thc following augumcntcd inscrvicc ins iec. inr:
~
2. With one oz morc seismic prngram and tb> requircm 11ts o'.
restraint , ,upport, or snubber Spec l f icat'l.on 3. 6. H/4. 6. H.
inoperable; within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months These snvbbers arc listc.'n.
replace or restore.thc inoper- Surveillance Instructions able seismic restraint(s), BF SI 4~6 H;1 and -2.
support(s),, oz snubber(s), to OPERABLE status and perform 1. Insuection Groups an engineering evaluation on I the attached component or The sn11bbers may bc cate-declare the attached system gorized into two ma)or inoperable and follov the groups based on whether the C- appropriate LIMITING CONDITION statement'or that svstem.
If a seismic zestraint, support, snubbers are accessible or inaccessible during reactor operation. These ma)or
3. groups may bc further or snubber (SRSS) is determined subdivided into groups to be inoperable while the based on design, envir-reactor is in the shutdown or onment, or ntbez features refuel mode, that SRSS shall be vhich may bc exocctcd to made operablc or replaced affect thc operability of prior to reactor startuo. If thc snubbers within the the inoperable SRSS is attached group. Each group may bi to a system that is required inspected independently in OPERABLE during the shutdown accordance with 4.6.H.2 or refuel mode, the appropriate through 4.6.H.9.
LIMITING CONDITIONS statement for that system shall be 2. Visual Ins ection Schedule.
followed. and Lot Size The first inservice risual inspection of snubbcrs not previously included in these tcchnical spec ificat fons and vhose visual. inspection has not been performed and documehted previously, shall be performed vithin six months for accessib)e snub-bers and before zesuming power after the first re fue ling outa F e 185
i LINI IHC CONDI TIOHS FOR OPERATION SVRVEILLANCE REQVZREHEHTS 3 ~ 7 CO Pa 8! NUGENT SYSTS!irs 4. 0 COHTAINHEHT SYSTEMS
~Al \ ca~bi iit A licabi lit Applies to the operating status Applies to the primary and o f the pr imar y and second a y secondary containment containment systems. integ=ity.
Ob~oct 1 Vo o~bect iv To assure the intogrity of the pr mary and secondary To verify the inte",rity of the containment systems. primary and secondary.
containment.
Soeci 1cat 1 &ll Soeci f i cation A P r i mar Containment At any time that the irradiated fuel is in 1 ~ P essure Suppress 'n the r actor vessel, Chamber and the nuclear system is pressurized a, The suppression above atmospheric chamber rrater level pressure or work is be checked once per being done which has day. Whenever heat the po ential to is added to the drain the vessel, the suppression oool by pressu e suppress'on testing of the ECCS pool water level and or relief valves the temperature shall be maintained within the pool temperature shall f 0 cwiilg llmltsr be continually monitored and shall be observed and logged every 5 minutes unti 1 the heat
a. Minimum water level addition is terminated.
-6.25" (differential pressure control
>0 psid)
-7.25" (0 psid differen-tial pressure control)
b. Maximum water level ~
ill 227
TABLE 3.7.A (Continued)
Number of Power Maximum Action on 0 crated Valves Operating Normal Initiating
~Grou Valve Identification Inboard . Outboard Time (sec.) Position ~Si nal Drywell 5P air compressor suction 10 C SC valve (FCV-64-139)
Drywell hP air compressor discharge 10 SC valve (FCV-64-140)
Drywell CAM suction valves 10 0 GC (FCV-90-254A and 254B)
Drywell CAM discharge valves 10 0 GC (FCV-90-257A and 257B) 6 Drywell CAM suction valve 10 0 (FCV-90-255)
TABL""- 3.7.D (Continued)
Y::l ve Valve Identification Radiation Monitor Suction Radiation Monitor Suction Radiation Monitor Discharge Radiation Monitor Discharge 260
TABLE 3.7.E Panaurr CHAI'4E."rr ISOLATIOn VA-VES ~erIC r ZmmaTE B LCH TlK SUPPRESSIO?L POOL WA~ LEVEL Valve Valve Identiiicotlon 12- 738 Au"iliary Boiler to PCIC 12--.41 Auxi1iary Boiler to RCIC 43-2"A RtlR Suppreesion Chamber Somcle Lines 43-2".3 RlR Suppression Cho~':er Sample Lines 43-r7A P Gl Suppression Chomber Sample Linea 43-29B KiR Suppression Chamber Somple Lines 2-1143 Dominerulizcd Hater 71-14 RCIC Turbine Exhaue" 71-"2 RCIC Vocuum Pump Diachorgc 71-y 0 RC C Turbine Exhaust 71-592 RCIC Vacuum Pump Diachaz'~c 73-23 IGCZ Turbine Exhaust 73-24 H CZ Turbine Exhaust Drain 73-6O3 H8:Z. Turbine Exhaust 73-6O9 HICZ Exhaust Drain 74-722 RHR 75-57 Suppression Chamber Drain 75-5'3 Suppression Chamber Drain 262
PROPOSED CHANGES UNIT 3
Table A.z b seismic Monitori Inst ent Surveillance R irenente INSTR 0 MENl'HANNEL TRIAXIAL TIME HISTORy ACCEQXRAPHS CHECK CHANNEL FUNCTIONAL TEST CHANNEL CAAZBBATZaI
a. Unit 1 reactor bldg. base slab (El 519 0) Monthly~ 6 months
b. Unit 1 reactor bldg. floor slab (El. 621 25) Monthly+ 6 months NA
c. Diesel-generator bldg. base slab (El. 565.5) Monthly+ 6 months BIAXIALSEISMIC SNITCHES
a. Unit 1 reactor bldg. base slab 6 months once/operating cycle
b. Unit 1 reactor bldg. base slab Monthly+ 6 months once/operating cycle
c. Unit 1 reactor bldg. base slab Monthly~ once/operating cycle TRMD.I ~ PL'..X AC EL5 RAH(S a 6
. 12 months N/A 12 months N/A c..'I- \ cs:~s."c sv" teIB 1V'.'ce El. lent 12 months N/A
Except seismic switches
LIHITING CONDITIONS FOR OPZRATIOH SURVEILLANCE REOUZRENENTS 3.3 REACTIVITY CONTROL 4.3 REACTIVITY CONTROL c~ Control rods C ~ When it is initially with scram times determined that a greater than control rod is incap-those permitted able of normal insertion by Specification a test shall be con-3.3.C.3 are ducted to demonstrate inoperable, but if they can be inserted with that the cause of the malfunction is not a control rod failure in the control'od drive pressure drive mechanism.
they need not be If this can be demonstrated an disarmed attempt to fully insert the electrically. control rod shall be made.
d~ Control rods If the control rod cannot be inserted and an investigation with a failed has demonstrated that the "Full-in<'r cause of failure Full-Out is not s failed control position switch rod drive mechanism may be bypassed in the Rod collet housing, a shut-Sequence Control down margin test shall System and be made to demonstrate considered under this condit'iori operable actual rod if the that the core made can be subcritical for position is any reactivity condition known. These during the remainder of rods must be the opera ting cycle wi th moved in the analytically determined,
'sequence to highest worth control correct'heir rod capable oF. withdrawal>.
positions (full fully withdrawn, and all in on insertion or full out on other control rods capable withdrawal) . of insertion fully inserted.
d. The control rod accumulators shall be determined operable. at least once per 7 days by verzfyxng that the pressure and level detectors are not in the alarmed condition.
120
Revised 11-18-78 LIMIT1NG CONDITIONS FOR OPEPATION SURVEILLANCE REQUIREMENTS 1 ~ 5 CORE AND CONTAINMENT COOLING S YSTEMS 5 CORE AND CONTAINMENT COOLING II SYSTEMS G. Automatic De ressurization G. Automatic De ressurization
1. Four of the six.Valves of the Automatic Depressuri" 1.
sation System shall During each operating be the following tests cycle operable: performed on the ADS: shall be (1) prior to a startup a. A simulated automatic from a Cold Condition, actuation test shall be org performed prior to (2) whenever there is after each refuelingstartup out-diated fuel in the irra- reactor age. Manual surveillance of the relief va'es vessel and the reactor vessel pressure is covered, in 4.6.D.2. is than 105 psig, except greater as specified in 3.5.G.2 and 3.5.G.3 below.
2. If three 2. When it is determined
~
are known to be of the six ADS valves of the six ADS valves*that are three automatic operation, incapable of incapable of automatic the HPCIS shall operation reactor may remain ':: the to be operable be demonstrated tion for a period not opera- daily thereafterimmediately and exceed 7 days, to as longg as provided the S pecification e a HP CI system is 3.5 .G.2 applies .
(Note that the operable.
relief function pressureof valves is assured these section 3.6.D of these by specifications and that this specification only applies to the ADS function.)
'han three of the If more
, valves are known tosixbc ADS able of automatic incap-an immediate operation, shall orderly shutdown be initiated, reactor in a hot shutdown thc with di:ion in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and con-shutdown condition in a cold following 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months . in the 161 Amendment No. 3, >8
1'l LIMITING CONDITIONS FOR OPERATION SURUEILLANCE RFOUIREMENTS 3.6 PRIMARY SYSTEM BOUNDARY 4 6 PRIMARY SYSTEM BOUNDARY 3 ~ At steaming rates 3 ~ Additional, coolant greater than 100,000 samples shall be lb/hr, the reactor taken when'ever the water quality may ~
reactor activity exceed specification exceeds one percent 3.6.B. 2 only for the of the equilibrium time limits specified concentration below. 'xceeding specified in 3.6.B.5 these time limits of and one of the the following maximum following conditions quality limits shall are met:
be cause for placing the reactor in the a~ During startup cold shutdown condition. b. Following a significant,
a. Conductivity power change*4 time above 2 prnho/cm9250C- ce Following an 4 weeks/year. increase in the Maximum Limit equilibrium off-,
10 pmho/cm8250C gas level exceeding 10,0QQ
b. Chloride uci/sec (at the concentration tim0 steam jet air above 0.2 ppm- ejector) within 4 wer ks/year. a 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months Maximum Limit- period.
0.5 pram.
d. Whenever the equilibrium iodine limit specified in 3 '.B 5 exceeded.
is
+For the purpose of this section on sampling fr equency, a signif icant power exchange is define'd 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
LIMITING CONDITIONS FOR OPERATION SURVEZ'ANCE REQUIREHENTS 3.6 PRIORY SYSTEH BOUNDARY 4.6 PRIMARY SYSTEN BOUNDARY I
H. Seismic'estraints Supports, H. Seismic Restraints Supports.
and Snubbers and Snubbers During all modes of operation The surveillance requirements except Cold Shutdown and Re- of paragzaph 4.6.G are the fuel, and seismic restraints, onlv zequircments that apply supports, and snubbcrs shall to any seismic restra'nt or be operable except as noted support other than snubbezs.
in 3.6.H.2 and 3.6.H.3 below.
All safety-related snubbcrs Each safety-related snubber shal are listed in Surveillance be demonstrated OPERABLE BY Instruction BF SI 4.6.H. performance of the following augumented insezvice inspect 'cn
2. With one or more seismic program and the requirements of, restraint , support, oz snubber Speci: ication 3. 6. H/4.6.H.
inoperable; within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months These snubbers are listed in replace or restore the inoper- Surveillance Instructions able seismic restraint(s), BF SI 4.6.H-1 and -2.
support(s), or snubber(s), to OPERABLE status and perform l. Inspection Groups an engineering evaluation on the attached component oz The snubbers may be cate-declare the attached system gorized into two major inoperable and follow the groups based on wnethe" the appropriate LZNZTZNG CONDITION snubbers are accessible or statement for that system. inaccess'ble duz'ng reactor operation. These major 3~ If a seismic restraint, support, gzoups may be further ot'nubber (SRSS) is determined subdivided into groups to be inopera'ole while the based on design, envir-reactor is in the shutdown or onment, or other features refuel mode. that SRSS shall bc which may be exoected to made opera'='.e or replaced affect the operability o:
prior to reactor startuo. If the snubbers within the t';.'. inoper- le SRSS is aetached group. Each gzouo mav ba tc a system thae is required inspected independent)v in OPERABLE during ehu ::hutdown accordance with 4.6.H.2 oz refuel mode, the appropriate through 4.6.H.9.
LZNZTING CONDITIONS s ea t omen t for that system shall be 2. Visual Inspection. Schecule.
followed. and Lot Size Thc first inseryice visual inspection of snuboezs,not previously included in these technical specificat ons and whose visual inspecrion has not been performed and documented pzeviouslv, shall be performed ~ithin six months for accessible snub-bars and bc'fore resuming power after the f'rst refueling outage 198
I.I NI 7 INC CONDITIONS rOR OPERATION SURVEILLANCE nEQVZAEMENTS 3.7 cotna,fvwENT sYsTKs5 4 ~ 7 CONTAINHENT. SYSTEMS
~ao i tea 1 it
~bi A olicabi i t Applies to the operating status Appl~es to the primary and of the primary and secondary secondary containment containment systems. integrity.
Ob~ec=ive 0~beet i re To assure the xntegrxty of the pr mary and secondary containment systems.
To verify the inte"rity of the primary and secondary containment Soy ci f scat l a~3 e
Speci f ication A. ~a".imar Contain.
Primar Containment At any time that the irradiated fuel is in Pressure Suppress cn the reactor vessel, Chamber and t.he nuclear system is pressurized a. The suppressfon above atmospheric pressure or work is chamber water Ievo1 being done Mhich has be checket'nce per the pvtential to day. Whenever heat-drain the vessel, the fs added to the pressure suporession suppressfon ooo1 by pool Mater level and testfng of the ECCS temperature shall be or reIfef vaIves the maantained Mithin the pool temperature sha11 following limits. be contfnua11y monf tored and sha11 be observed and Iogged every 5 mfnutes untf I the heat
a. Minimum water level ~
addftfon fs termfnated.
-6e?5" (differential control 'ressure
>0 paid)
-7.25" (0 paid differ ential pressure contre
b. Maximum water level ~
ill 231
TABLE 3. 7.A (Continued)
Number of Power hhximum Action on Operated Valves Operating Normal Initiating
~Gt au Valve Identif ication Inboard Outboard Time (Sec.) Position ~si nal Torus Oxygen Sample Line Valves-Analyzer B (FSV-76-63, 64) NA Note 1 SC Drywell Hydrogen Sample:
Line Valves-Analyzer B (FSU-76-59, 60) NA Note 1 SC 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) cc j
RCIC Condensate Pump Drain (FCV-71-7A% 7B) SC HPCI Hotwell pump discharge isola-tion valves (FCV-73-17A, 17B) SC HPCI steamline drain (FCV-73-6A, 6B) GC TIP Guide Tubes (5) 1 per guide tube NA 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)
T/Z~m ~.7.E PR JJ"C COi.""l Ii".Z:lT SOLATXQil VALVES MlXCl'. 2U!I14TE 9'r.'LC'l TIE SUPr ~SSIO:1 POOL l'1A~r M~rL
'1.1l.ve Val'e Xdcntl.fi.co ion 12 I 3'~ Auxill.ary l3oiler to RCXC 12-":41 Auxiliary Boiler Co RCIC 43-2OA RllR Suppression Chamber Sample Lines RliR Suppression Chamber Sample Linc" 43 2oA R:B Suppression Chamber Sample Lines 43-2W BhR Suppression Chamber Sample Lines 71-1 RC ' Turo inc Fxhaus 71-~2 HCIC Vacuum Pump Discharge RCIC .urbinc Kvi>aust "1-jg2 RCXC Vncua~ imp DischcrSe 73-23
'78 lKI Turbine K~~gust llcCX Turb ine Lchaus 4 Drain 7~-603 !'.iCI Turbine Z haus l(KI Exhaust Drain "4-"22 RlB
2"J7 Suppression Chamber Dr'ain 75-59 Suppression Chamber Drain 279
LIMITING CONDITIONS FOR OPERATION J~N 25 nu SURVEILLANCE REQUIREMENT 3.10 CORE ALTERATIONS 4 . 10 CORE ALTERATIONS B. Core Monitor in B. Core Monitorin
1. During core alterations, Prior to making any alterations except as in 3. 10.B.2, to the core the SRM's shall be two SRM's shall be functionally tested and checked for operable, in or adjacent neutron response. Thereafter, while to any quadrant where required to be operable, the SRM's fuel or control rods are will be checked daily for response being moved. For an SRM except as specified in 3. 10.B..1.b.2.
to be considered operable, the following shall be satisfied:
a ~ The SRM shall be inserted to the normal operating level. (Use of special moveable, dunking type detectors during initial fuel loading and ma)or core alterations in place of normal detectors is per-missible as long as the detector is connected to th normal SRM circuit.)
b.1 The SRM shall have a minimum= of 3 cps with all rods fully inserted in the core, if'one or more fuel assemblies are in the core, or b.2 During a full core reload where both irradiated and fresh fuel is being loaded, SRM's (FLC's) may have a count rate of <<3 cps provided that the SRM's are response checked at least once every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months with a neutron source until >3 cps can be main-tained, and provided also that the core is loaded in a spiral sequence only, or 336
ENCLOSURE 2 DESCRIPTION AND JUSTIFICATION TVA BFNP TS 213 BROWNS FERRY NUCLEAR PLANT
1. Units 1 and 2 a e v Delete sections 6.9, Environmental Qualification; 6.10, Integrity of Systems Outside Containment; 6.11, Iodine Monitoring from the Table of Contents. These sections were previously removed from the technical specifications but the Table of Contents was never updated.
This proposed change a change needed is considered administrative in nature because for consistency within the technical specifications, it is
2. Unit 1 a e 9 The present specification refers to 4.1.B as containing the surveillance requirements for APRM scram setpoints. Specification 4.1 B contains the
~
surveillance requirements for the Reactor Protection, System power monitoring system instrumentation. The proposed change is to refer to specification 4.5.L as containing the surveillance requirements for APRM scram setpoints, since that is the correct section.
This proposed change is considered administrative in nature because it is a change needed for consistency within and correction of the technical specifications.
3. Units 1 and 2 a e 108 'nits 3 a e 105 Presently, Table 3,2,J, Seismic Instrumentation, contains reference to Triaxial Peak Accelographs. However, Table 4.2.J containing the surveillance requirements for the seismic monitoring instsrument has no similar reference. The proposed change is addition of Triaxial Peak Accelographs to Table 4.2.J with appropriate surveillance requirements.
The surveillance proposed is an annual Channel Functional Test. This differs from BWR Standard Technical Specifications which has a channel calibration every 18 months, and "Not Applicable" for channel check and channel functional test. This is explained on the basis that the peak recording accelograph used at Browns Ferry, PRA-103, has no adjustments to be made. We, therefore, refer to our yearly surveillance as a Functional Test, rather than a calibration test.
This proposed change a change needed is considered administrative in nature because for consistency within the technical specifications.
it'is
Unit 3 a e 120 Specification 4.3.A.2.c discusses performance of a test to demonstrate that the cause of a control rod insertion failure is not a failure in the control rod drive mechanism. As presently written, the section makes no grammatical sense,, apparently because of typographical errors made when it was typed. This grammatically confusing specification was issued by Amendment No. 56 to Unit 3 License No. DPR-68 dated July 22, 1982. The
'proposed change will make the specification grammatically correct.
This proposed change is considered administrative in nature because it only serves to correct typographical errors in the section.
Units 1 and 2 a e 157 unit 3 a e 161 As presently written, specification 4.5.G.2 requires certain surveillance when "more than two" ADS valves are'incapable of automatic operation.
The associated specification 3.5.G.2 discusses action to 'be taken when "three of the six" ADS valves are incapable of automatic oper'ation. The proposed change to 4.5.G.2 is to make the wording identical to 3.5.G'.2 for when action is needed.
This proposed change a change needed is considered administrative in nature because for consistency within the technical specifications.
it is Units 1 and 2 a e 179 unit 3 a e 188 As presently written, specification 4.6.B.6 makes a reference to 3,6.B.4 as containing an equilibrium iodine limit. However, 3.6.B.4 contains reactor limits on conductivty, chloride, and pH, with no limits on equilibrium iodine. The equi. librium value of iodine is given in 3.6.B.6. The proposed change is to refer to the correct specification, i,e., 3.6.B.6, rather than 3.6.B.4.
Also, for units 1 and 2, it is proposed to correct a typographical error in a reference to the steam line isolation values in specification 3.6.B.6.
These changes are administrative since they are to correct wrong references.
Units 1 and 2 a e 185'nit 3 a e 198 Specifications 3.6.H.1 and 4.6.H, Seismic Restraints, Supports, and Snubbers, currently states that snubbers are listed in Surveillance Instruction BF SI 4.6.H. The proposed change is to correctly identify reissued surveillance instructions.
8. Units 1 and 2 a e 227'nit 3 a e 231 Specification 3.7.A.1 currently makes reference to 3.7,A.2 as containing exceptions to the requirements for maintaining torus water level.
However, specification 3.7.A.2 contains no discussion of conditions for exceptions to these requirements, and therefore, is not appropriate. The proposed change is deletion of the reference to 3.7.A.2 since it is inappropriate.
9. Unit 2 a e 253a Table 3.7,A, Primary Containment Isolation Valves, currently lists the drywell differential pressure air compressor suction valve (FCV-64-139) and air compressor discharge valve (FCV-64-140) as normal position of "open" and "goes closed" as Action on Initiating Signal. Actually the normal position of these valves is "closed" and the valves "stay closed" on an initiating signal. The proposed change is a revision to this table to reflect the correct posi.tion for these valves.
Also proposed is deletion of a footnote clarifying operation of valve FCV-64-139. The footnote is no longer needed.
10. Unit, 2 a e 260 Table 3.7.D, Air Tested Isolation Valves, currently lists valves 90-254B and 90-255 as Radiation Monitor Discharge valves. This is incorrect.
The proposed change is to correctly identify these valves as radiation monitor suction valves.
ll. Unit 3 a e 264A Table 3.7.A, Primary Containment Isolation Valves, currently lists the Reactor Core Isolation Cooling System (RCIC) steamline drain (FSV-71-6A,
-6B) normal position as "closed" and the action on initiating signal as "stays closed." The correct position of this valve is "open," and the valve "goes closed" on initiating, signal. The proposed change is to have Table 3.7.A reflect the correct valve position and action.
12. Units 1 and 2 a e 262'nit 3 a e 279 Table 3.7.E currently lists valves 75-57 and 75-58 as "core spray to auxiliary boiler." The proposed change is to correctly designate these valves as "Suppression Chamber Drain" valves.
ENCLOSURE 3 Determination of No Significant Hazards (TVA BFNP TS 213)
Description of Change This proposed amendment includes several changes to,the technical specifications, all of which are administrative in nature. The table of contents is to be altered to delete sections which no longer appear in the technical specifications for units and 2. Page 9 of the unit technical 1 1 specifications is to be changed to correct a typographical error. Table 4.2.J is to be changed for all three units (page 108 units and 2, page 105 1
unit 3) to add surveillance requirements which would make the table "consistent with table 3.2.J . Page 120 of the unit 3 technical specification is to be changed to complete a sentence which, as it stands now, makes no sense. Section 4.5.G.2 is to be changed for all three units (page 157 units 1 and 2, page 161 unit 3) to make it read consistent with the associated, limiting condition for operation, 3.5.G.2. Sections 3.6.B .6 and 4.6.B.b.d are being changed for units 1 and 2 (page 179) to correct typographical errors in those sections. For unit 3, section 4.6.B.3.d is being changed to reference the correct limiting condi,tion for operation (page 188). Section 4.6.H is to be changed for all units (page 185 units 1 and 2, page 198 unit 3) to correctly identify surveillance instructions which have been reissued. The section 3.7 .A.1 is to be revised to delete an incorrect reference to 3 .7 .A . 2 for all units . Table 3 .7 .A is to be changed for units 2 and 3 to correctly identify the normal position of valves in each table, which had been incorrectly specified. Table 3.7.D is to be changed for unit 2 to correctly identify the function of two valves in the radiation monitoring system. 'able 3.7.E is to be changed for all units to correctly identify two RHR valves. Finally, for unit 3, section 3.10.B.1.b.2 is to be changed to correct a misspelled word.
A C
~ .~
lg

ML18029A905

Text

Navigation menu

Search