ML18030A978

Proposed Tech Spec Changes,Supplementing Util 840823 & 850403 Applications for Amend to License DPR-52 to Include Editorial Corrections,New Info & Updated Pages Re Turbine Control Valve Fast Closure or Turbine Trip Scram

ML18030A978
Person / Time
Site:	Browns Ferry
Issue date:	12/30/1985
From:	TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML18030A977	List: ML18030A976 ML18030A978 ML18030A980
References
TVA-BFNP-TS-199, NUDOCS 8601060169
Download: ML18030A978 (38)

Text

ENCLOSURE 1 PROPOSED TECHNICAL SPECIFICATION REVISIONS BROWNS FERRY NUCLEAR PLANT UNIT 2 (TVA BFNP TS 199 SUPPLEMENT 1) 86020602b9 851230

-

ADQCK 050002b0

'DR P PDR J

'iack',+Kg we PROPOSED CHANGES UNXT 2

'PRO~OS' -"' p"CXFrcz"'vz J,o fS BROriNS FERRY NUCLEAR PLANT i8$ XT 2 (Tl'Z 3L"i~fP TS 199 SUPPI ENEliT 3.)

bAS! s ff<<n (uel dasLade ass>>aint a steady state operation as the tip vetelnR over

~ C

~ nc [re scelseulat tan f lou range. The narbin co the Safety Lfolt increases the f lov dec ceases for the spvc 1 f 1cd. t s p sett lnR s ersus 1 1 lou sc latfonshlp; (bees jpre she uorst ease WCPR uhich could occur due 1>>R s eady- ~ tate operation

~

~ s inst of satrd shrrnal rcwer brc ause of she AplVi red bloeb t rip eettlnb The padres d ps( lbus Ini I>> sbw core L'stabl l shed by spec 1 1<<d control sod sequences

~

~ s gv c ~

~ nd li nvnl<ored cont tnuously by the fn-core OIW sysseo,

$ 1 1

/

Reactor uater Lov level sabras and I oint icn (Face c ttaln sit ael ines)

The set point for the lou level scram ls above the bot toss of the separator skirt.

.his level has been used ln transtent analvsci deal ant vieh coolant inventory decrease. The res<<lt s reported ln FSAR subsect ion 1<.S sb<<M that scram and isolation ol all process lines (except naLn stem) at this level adequately pcotccts chc fuel and the pressure barrier, because 11CPR 1s Rreater than 1.07 ln all cases, and

~ ysteo pressute does not reach the safety valve sett lnss. 'The scraLs sett i>>a ls

~ pproafsLacely )1 inches beloM the noraal opetatfnR range and ls thus adequate to

~ void spur1ous acres.

n. n w ~ ~il<< ~ ., s Thi turbinr. stop valve closurr trip anticipates the pressure, neutron flux anl 4< dl, flux increases that would result. from closuse of the stop valves.

'Mith a trip setting of 10" of valve closure fromm full open, the resultant increase in heat flux is such that adequate therma'l margins are n@inta<ncd even during the worst case trarsient that assumes the turbine bypass valves remain closed. (Reference 2)

V.. T>>rhine Control Vnlvn Fnnt Closure or Turbine Tri Scram Turbine control valve fast closure or turbine trip scram anticipates the pr< so>>re, n<<>>tron flux, a>>d bent fl>>x increase that co>>ld result from r>>ntrol valve fn>>t clos>>rc due to load rejection or rnnrrol valve closure d>>e tn t>>rhl>>c trip; each without bypass valve capahiiity. Thc reactor protection "ystcm initiates a scram in less than 30 milliseconds after the start of control valve fast closure due to load re)ection or control valve closure due to turbine trip. This scram is achieved by rapidly reducinR hydra>>lie control oil pressure at the main turbine control valve actuator disc dump valves.

This loss of prcssure ic srnsed by prcssure switches whose contacts form thr one-out-ot'-two-twice ionic input to the reactor protection system.

This trip scttinq, a noniinally 50" nreatrr closure tip'nd a diffe>cnt valve character istic from that of the turbine stop valve, combine tu produce transients very similar to that for the stop valve.

Relevant transient analyses are discussed

.in tlcfcrenccs 1 ai)d 2 . This scram is bypassed when turbine stcam flow is below 30Ã of rated, as m asurcd by L>>rbinc first slate pressure.

23

TABLE irl>A REACTOR PROTECTION SYSTEH (SCRAH) IHFPQ~ATION FUHCTIOHAI; TESTS KWQUH PVHCTIORQ. TEST FREgUKHCIES POR SAPETT IHSTRI> AHD COHTROLJCIRCUpg.

r ~I

.:

~

' QKDVIi~g Puoctfooai".Teat I

l

"

.J,>>Hfnfoui Fte ueocy ())>

o

%de S&tcb fo Place'"Horde Sbitch frl'Shutdoun,. '.-.Each Refuelfog,outiRe-A

'

~

rJ n Haouaf Scran r>

A ).;

J<

Trfp Chiaoel",aad Alai', -"Eac'ty ) Hontha

~

Q >I loa~ ) m 7 r ta a Q a IRH (4 ) ', j,y,

• ~ 1 ~

r Hfgh Plum C j Trfp Chaane Lccod Ala pa Once> Fat Veal'urloR 'R efue ffn D

~u <aod Before Each Stiitaup.

o J>

lnopcratf ao '>> C

~J Ttfp Cbiooel add Ala~ >'I Ja rOacca Pet Veelr, DurfoR Refuelfn oand Before Each, Startup'

>J .

'a 1 o AFRH  ! 'I >J

' r C'/Vector.

r C T<

> /clays (4) I J

'I J iBeofotc hach Stcttup >a'nd Veetl 0 'Vben Rcgufred to bcOpetalJIo Iligh Flux .(FloM Biased) ~,B Trip Output->Relays (4) ', . Once/M'cek QfBh Flea (F)xed Trip) o B Trip Outpu~ Rclcya (4) %I t> <Ooe ~ ' O>a r C r~ ~ \ e r: 1 I laoperctfac " Output R'elaya (4) r'rfp Once/Vega ~ . g ! Do+a 4 ca Ic I>>B Trip Outpuc Relays (4) ce/Veen u I Flou Bfca (l'J) \P J > (g) >J w r. l J N r aa ~r ~~ u>>

o "cet I Ca DIRb Reactor Ereaoute Trfp Channel.,"aod Alan< (7)=" I.month (PIS-3-2?AA, BB, C, 0) '- '.

0 -'. a >1 r) Ia >>>>>>> = >* "IAr-"sNs Y-b'), ~ a>" 'a .==:.' ~ )";- Trfp CbannoL anJL 'lara (7). >! ')month "~\ ~ Ar) ~ >> 1 JI r ~ 1 ~ f, o J ., ~ Reactor Lou Vatet Level a Trip Channel aod Alans " ence/ l-month ~ A-0) 'LIS-3-203 B o" c >>>J >> u J' g1 Vatcr Ecacl fo Scran Tant c r> ~ ~1 HIRh rJ o JJ  ! >J ~ 'A Dfachar'gc'loat lJJ SMitches Trip Channel and Ala rm g Ollhe/month r (LS-85-45 C-F),' r", rr E. >J Level . 'lectronic Switches Trip Cliannel yh6 Ala rm (7) Once/ month (LS-85-45A, B, G, H) >> Hain Steam Line lliph Radiation Trip Channel anJl hlnrm (4) Once/3 months (8) TABLt a 1. i REACTOR PRO'I ECTION SYSTEM (SCRAM) IH.'iTi?:: 4L: ~ a'ATION FUHCT IOflAL TESTS -'1II41HUH PUbC'TIOIIAL TEST PREQUENCIES FOR SAI= ~ I I MS.R. AMD CONTROL CIRCUITS Group (2) Functional Test Minimum Frauen y (3) Main Steam Line Isolatron Valve Closure Trip Channel and Alarm OnCer 3 montgS(8) Turbine control Valve Fast; Closure Trip Channel Chancre ...i Alarm Once/Honth (1) or Turbine Trip Turbine First Staqe Pressure Permissive Trip and Alarm(7) Every 3 Honths )PIS-)-81 QB~ /IS-1-91 ASB) Trip C:.anncl and Alarm Once/Month (1) NOTES FOR TABLE 4 ~ 1.A Initially the minimum frequency for thc kndiratrd tests shall he once pcr month. r deicription of the three groups is included in the H 2.. ',A Bases of this specification. . Il 3.":Functional tests arc not required when thc systems are not required to he opcrablc or are operating, (i.e., already tripped). Tf tests are .,missed,= they shall be performed prior to returning thc system., to an operahli status.

4. This instrumentation is exempted from the instrument channel test definition. This instrument channel functional test will consist of injecting a simulated electrical signal into the measurement channels.

(DELETED) 6.. The Functional test of the flow bias network is performed in accordance with Table 4.2.C.

7. Functional test. consists of the injection of a simulated signal into thc electronic 'trip circuitry in place of the sensor signal to verify oper'ability of, the trip and alarm functions.

8 . The functional test frequency decreas ed to once /3 months to reduce chal t cages to relief valves par NUREC 0737 1 tern iZ.K.J. 16. 39 TABLE 4~ 1~ B REACTOR PROTEC?IOH SIST EN (SCRAH) IHSTRUHZHT CALIBRATIOH HIHIHOH CALIBRATIOH PREQOE1C'IES FOR REACTOR PROTECTIOH IHSTROHEHT CHAHHELS Instrwent Channel Croup (1) Calibration )tiniam Frequency II) IRH High Flux Caeparison to APRH on Control note (4) led etartupa (6) ARRH R)gh Flux Output Signal Heat Balance Once every I days F lov Bf as Signal Cali "rate Flo~ Bias Signal {7) Once/operating cycle LPRH Signal TIp System Traverse (I) Every 1000 Efiective Full PoMer Hours "'I'PS-7-'>O'A'; Standard Pressure Source S, C, 0) Once/Operatinq Cycle (q) Standard Pressure Source Onte/Operating'ycl e (9) '-tran "r NVA",.1 Pressure standard Once/Operating Cycle (9) High Hater Level in scran Discharge voluse Floa t Switches ( LS-85-45 C- F ) A Note (5) Note (5) Electronic Level Switches B Calibrated Mater Column Once/Operating Cycle (9) (LS-85-<5 A, S, 0, H Turbine Condenser Low Vacuum Standard Vacuum Source Every 3 Months Main Steam Line Isolation Valve Closure A Note (5) Note (5) Main Steam Line High Radiation B Standard Current Source (3) Every 3 Months Turbine First Stage Pressure Permissive Standard Pressure Source (PIS-1-81 ASB, PIS-1-91 ASB) B Once/Operatinq Cycle (0) Turbine Stop Valve Closure A Note (5) Note (5) Turbine Cont. Valve Fast Closure Standard Pressure on Turbine Trip Source Once/ODerating Cycle tow. 4< 1. N

• TABLE 3 2 F SURVEILLANCE ZtiSTRUMEHTATIntt

<r< i)(b n'; -c p- >4 J Isa~] Hininulr' of trd;t.rrvt,,) > Type Indicatxon ~ 1 per able Instrument and Range Notes 0 Instrument Instrument Channels 2 LI-3-58A 4 ReaCtOr Materi'LeVdlrI J r') IOL<a . Indicator - 155", ... $ 60" to, -r'g d(1) (2) (3) LI-3-58B PI-3-74A Reactor PresVure .'.= T 'r=': , Incicator 0-)200, pst 9 .-(I (/I g) (3) PI-3-74B 11 'jcvr 'R-64-So Dryvell Pressure "anv< Recorder 0-80~gpia,r)r Indicator 0-80 .'t g(1) l hg (2) . (3) P I-64-678 '< Indicator (3) psia'ecorder, ~ J (7) NI TI-:64-52AB Dr@sell Temperature ta..,rd 0 400oF l XR 69 50 Ai r Recorder 0-400OF (1) (2) (3) XR-69-S2 Suppression.,Chachyr I Temperatur"e,"", ' "' ~>) "r-- '-'-fn~~ 4;c),-" N at:d .:1 'V-Cut;,, co<IT  : 3.',Jnt),-.. )', EO,J) (5) N/A, Control RodaPqoigion, r6V Indicating y, < L'ights I-' ~ y .onths SRM, IRM, LPRH ) (1) (2) (3) (") Neutron Monitor ing tt/A <I tot 4, "') 1',1 r,rc 0 to 100% pover )

PS-64-67B p, Dryvell Pre"oury~ Alarm at 35 psig, )Blat)riq C'",:

TS-64-52A8 DryMell Temperature and Alarm if 281OF and enp. ) ) (1) (2) (3) (4) PIS-64-58AR Prcssure <and Timer IS-64-67A ~ tnt '1 l Pressure > g: psf8.) after 30 minute' delay ) LI-84-2A CAD Tank A Level Indicator 0 to Tank "B" Level to 100'ndicator LI-84-13A CAD 0 100% TABLE 3 2.P SURVEILLANCE It!STBU!!ENTATIOB Hfninua I cf Operable In r'--.cn'hannels ~ r ~ ~cL.dInhfrctfon Rcn-! Rrtes !!8 2 IG 9! Dryvcff end 0 1 20r Torus R H 2 - g6 lo4 .  !!jdroRen .Ccncertraticn Pdi-64-13I Drwclf to Indicator (1) (2) (3) Suppression 0 to 2 paid PdI-64-138 Chsnber Differential Pressure 1/Valve Belief Valve (5) Tailpipe Themocoupfe Tcnpcraturc or Acoustic Ronftor on Relief Valvo Tailpipe BR-90-272CD  !!fgh Range Recorder> (7) (8) Prinary 1 - 10 R/!fr RR-90-ZI3CD Contaf anent Radiation Bccordcrs LI-64-159A Suppression Indicator, / XR-64-159 Chanbcr 'Lister Level-Midc Range Recorder 0-240" (l) (2) (3) PI 160A Dryvcll Pressure Indicator, Recorder) (1,) (2) (3) XB-64-159 Mfde Range 0-300 psig ) TI-64-161 Suppres fcn Pcol Indicator, Bceordcr) (1) (2) (3) (4) N) TB-64-161 Oulk, ) TI>>64-162 Tcoperatutc - r Tn-6'162 3o 23o ') RR-90-322A Wide Ranpe Recorder Gaseou's 'oble Gas) Effluent )0 7 10"5 pCi/cc)(7)(8) Radiation Particulates) Honitor iodf)e 10" and 10+2 )fCi/cc) ()) prom end after tho data that ono o these parameters ia reduced to one indication, continued operaticn ia pemsaiblo dur-ng aha succeeding dirty daya unleaa such inst"um~tac'cn sooner made opcrahla. (") I".om and after the data Jet cno o{ thcaa para=ctcra js not indicated in tho cont"ol roon, continued operation Ls p+rmiaaphlo during the ouccccdwg aaven d ya unlcao auc5 ~". Ls aoonor +ado oporahla. 'natruncntaticn (3) If tho requircaoats of aotas (I) and (2) cscaot ba cat, snd if one of the indications car~at bo restored--in (6) hours, sn orderly" shusdawn shall be ini.tiatad and ho riiactor shall ba in a cold . conditAoa within "4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. ' (a) .heaa au=veillancc inatwznta are conaido od to 5 redundant. co 'oach other. IJ ~ g iA (5) 7rcu and after the date that both the, acoustic monitor and tha ra~7 -"temperature indication on any one valve fails t'o indicare ~m,the if> < ~

conrrol roca, 'cont'nued operation is pemiasible during,the succeeding hirty days, unless one of the two monitoring cl~anela is, sooner~cade

,operablc. I~'both the pricary and secondary indication'on'ny SRV tail , pipe ia inopara51e, the torus temperature will'e -onitored a't least F: once'per shift to observe any unexplained tecpcraturc increa e which 'might bc indicative of an open-,SRV. )iran ~o

• I's wl Ek f,

1t 1 il tl (6) h c Haniicl consistj of,'B:, ensors, *one,;from, each alternating rnrus bay, Seven" sensors must, be ~operable;for the';channe1= to ~be, ', 'ioped;abler;-,gy w ~ ~ .<< .e ~ ~. i ~~ ~ ~ ~

• L) ~

l0 (7) 'lihun ono of 'these instruments is inoperable for more than 7 days, in nf any other report required by specification 6.7.2, prepare and suhmit a Special Report to the Commission pursuant .to'spo~.ificatlon 6.7.3 within the next 7 days outlining the action ~tal'en, the'ause nf inoperabilityand the, plans agd,sd)edule for. .resto>-ing the system to operable, status. 4 vi (8),'.With the plant in the power operation, startup, or hot shutdown condition and with the number of operable channel's"less chan the required operable channels, either restore the inoperable channel(s) to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or initiate the preplanned alternate method of monitoring the appropriate parameter. 80 TAbLE a ~ 3.h SORYEILLLHCE REQOIRKHEHTS IOR RRIHhRT CottlhlttNEttc hHD REACTOR SOILDIHG ISOLhT IOH IIISZRIJKEÃfhTIOH inunction tunct lonal Test Calibration Frequency Instcunent Checir, InatcusIsat Cbannol- II) (27) Once/Operating Cycle (28) once/d y Reactor Lov taster Lovel (Lzs-3-203A-0) Inotrusont Channel once/3 tsonths none Reactor IIIgh preaauco Instrument Channel- II) (27) Once/Operating Cycle (28) once/day Reactor Lov uatec Level ~ V (LrS-3-5eA-D) inatcusont Cbannol II) (27) Once/Operating'Cycle (28) Sigh Dryvell Pressure (Prs-64-SGA-D) Inatcutsant Channel (29) once/day Oigh Radiation Hain Stean LIne Tunne 1 Instcuoent Cbannol- (29) (27) Ance/Aper atinII Cycle (28) Line )PIS-i-V3 /6 82, 86) instr+ant Channol- (29)(27) Once/Operating Cycle (28) once/day Blgh tlov Hain Stean Line Pdzs-l-13A-D, 25A-D, 36A-D. SOA-D) natrment Channol (Z9) once/operating cycle none Hain Stean Line Tunnel Blgh Taoperaturv Inotrusatnt Channel ( I) {Ia) Ill) once/3 eonths once/day Ib) Reactor building Ventilation IIIgh Radiation Reactor tone TABLE CD SURVKTLLAXCE REQUTRPfP. TS FOR TMSTRL~ATlnlf THAT TM?rlATK nl COVFLOL TRK CSCS Function Functional Test Cslfbrstfon Tnstrus4ant Chaclr. laotruaent Channel (1) (27) Ance/Operating Cycle (28} Reactor Lou Mater Level (LTS-3-SGA-D) ance/des'astruaeat 4 ~ Channel (1) (27) Once/Operating Cycl e (28) ance/dsf Reactor Lcv Mater Level (L?S-3-184 R 105) 4 Taatruacat Channel <1) (27) Ance/Operating Cycle, (28) l.. aa<</d41 Roactar Lc4t Mater Level ( (LIS-3-52 8 62) Tnatnrscnt Channel (1) (27) 4 4 f, 4<' 4II 'u. IIP-Once/Operating Cycle (2B } Reactor Lou Mater Level ( LIS-3-56A-D) Zactruuant Channql (1) (27) Once/Operating Cycle (2 ) Reactor High Preasura . I 4 (P IS-3-204A-0) Taatruacat Cheanel (1) (27) Once/Operatinq Cycle Drywall UKRh Praaaura (PIS-64-58E-H) (r Eaat~t Chan>el (1) (27) (@)'nce/Operating Cycle (28) Dr3n~ll HtRh Preaoaro fPIS-64-58A-D) lnat~t Chaaaal Drywall Btgh Preaaura (1) (27) Once/Operating Cycle (28) (PIS-64-57A-D) Zaat~at C~ol. (1) (27) Once/Operating Cycle (28) QaactoL.Lm Paeceara (PIS-3-74A88, PS-3-74AIlB) (PIG-68-95, PS-68-95) (PrS-68-96, PS-68-96) TABLE a ~ 2ec SQRVEZLLA14 E RF4QZRENENTS POR ZNSTRQkENTATZON tHAT INITIATE ROO DLOCKS 0 ~ ~ Function Functional Test Calibration {17) Instrument Check APRk Qpscale {tl~ Dias) {11 . {13) once/3 sonths once/day (8) APRk Dpscale (Startup kode) {1) {131 once/3 sonths once/day {8) APRk Dovtlscale {11 {13) once/3 sonths once/4ay {8) APRk Inoperative {1) {131 once/day {8) REk Qpscale {tie Bias) {1) {13) once/6 nonths once/day {8) RSk Downscale (1) {13) once/6 sonths once/day (8) RSk Inoperative (1) (13) once/day (8) IRk Upscale {1) {2) {13) once/3 scathe ence/4ay (8) IRk Dowlscale (1) (2) {13) once/3 sonths once/day (8) IRk Detector not in startup (2) {once/operating once/operating cycle (12) S/A Position cycle) IRx Znoperatise (1) (2) {13) M ~ SRk Qpscale (1) {21 (13) once/3 aaaths once/day {8) SRk Dovnscale (1) {2) (13) once/3 sonths once/day {8) SRk Detector not in Startup {2) (once/operating once/operating cycle {12) N/A Position cycle) SRk Inoperative {1) {21 {13) Floe Ries cooperator {11 {1>) once/operating cycle {20) tlcw Siss Qpscale {1) {1$ ) once/3 soaths Rod Slock Logic ('16) k/A RSCS Restraint oncel3 sonths IVA West Scram Discharge once/quarter once/operating. cycle N/A Tank Water Level High (LS-85-45L) East Scram Discharge once/quarter once/operating cycle'/A Tank Water Level High (LS-85-45M) 0 TABLEq).,2. Fc HIIIIMUM TEST AHD CA'LIBRATION FtrFQUI:ttCY'OR SUIIVEIIrLAHCE IttSTRUtlt:trTATIOH h C ~ g h Il h, hh Er ~ =.h = i Crhh:a Instrument Channel Instrument Check

1) Reactor Water Level (LI-3-58A88) tr ilail Once/6 months " ~  :.. ~:y I;., Each Shift

2) Reactor Pressure Once/12 months 43'.

Each Shift (PI-'3-74ASB) C

3) Drywell Pressure Once/6 rqonths;,<,, ~

Qyy t>t Each Shift (PI-64-67B) and XR-64-50,l

4) Dr ell Temperature r;3, Once/6 mBHths . ~ Crh <<I Each Sllift TI'-64-52AB) and XR-64-50 trt '

trrt

5) Suppression Chamber Air Tepperature Once/6 morlt)ls '-.Ct Each Shift (XR-64 52) h heal

~:3+ 'res Control Position Shift lh

8) Ror) ttA Each C hh I'1

9) Neutron Honitoring pl 'jh (2) ,-I Each Shi ft h

10) Drylrell Pressure (PS-64-67,B) Once/6 moil'ths

'l) Drywel 1 Pressure (PIS-64-58'A) Once/6 monttie 'IA r g ttc

12) Drywel.l Temperature (TS-64-$ 2$ ) I Once/6 mon ths. HA l h

13) Timer -(ZS-64-67A) Once/6 months tr ~

14) CAD Tank Level Once/6 moll t lls Once/day c'lce/quar<or o ",> ""., retie ~

h)l re

15) Con tiiiiraerrt Atmosphere Horlit:ore Once/6 moll tile Once/day.

hrS-l6) Dryqcll.to Suppreooion Chamber once /6 rrontho Each Shift Differential Freosure mr ir 0 TABLE 4 ' ' llINIHUN TEST ANO CALIBRATION FREQUEllCY FOR SURVEILLANCE IHSTRU! lENTATION Instrument Channel Calibration Fre uenc Instrument Check I7 Relief valve Tailpipe Once/month (24) Thermocouple Temperature lS Acoustic Vnnitor on Once/cycle (25) Once/month (26) Relief Valve Tailpipe 19 High-Range Primary Containment Once/cycle ( 3o) Ance/month Radiation lionitors (RR-90-272CD) (RR-')0-273CD) 20 Suppression Chamber Hater Once/cycle Ance/month Level-Hide Range (LI-64-159A) (XR-64-15q) 21 Drywell Pressure-Hide Range Once/cycle Ance/shi est (PI-64-160A) (XR-64-159) 22 Suppression Pool Bulk Temperature Once/cycle Once/shift (TI-64-161) (TR-64-161) (TI-64-162) (TR-64-162) 23 High Range Gaseous Effluent Once/cycle Once/shift Radiation 1fonitor (RR-90-322A) rrorts fOR TASI CS 4. 2.A TNROUCII 4. 2.ll ConC tnurd

14. Upscale trip te functionally tested during functional teat ttrLe ao raqutred by oect,ton 4.7.b.l.a and 4.7.C.l.c.

15. The ftov bias d'osparator-'vill"tie tested..bv putting. one Clov uoft Ln-Tesc" (producing 1/2 rcrarr) and ad)uottng the teac tnput:Ito obtain coaparacor rod block. Tha Clov bias upscale uttf be verified by observing ~ local upscale trip light duitng opera ion and vrrtfLad that it vL11 produce ~ rod block during 'the operattng cycLo.

16. Parforrred during operating cycle. Portions of the logic io checked rrora frequently durLng Cuncttorral teoto of tha functions that produce

~ rod block.

17. Thts calibration conatsca oC reerovtng tho function Crora oarvtca and perfomfnb an electronfc calfbration of tha channel.

14. PunctLonal tert ts ltrrtted to the condition adhere secondary contatnuont integrtty ts not requtred ao opect fied tn oections ).7.C,2 and ),7.C.).

19 ~,PunctLonot ccac ts Ltrrt ted to the t trrre vhere the SCTS to required to .'rreet the rrqutrerrento of aectton 4.7.C.l.e.

20. Caltbratton of the cooperator requtreo the inputs Crorr both recLrculation Loops to be tncerrupted, thereby rerravtng the flov btaa afgnal to the

.hPRPI and AN a rd scra~tng the reactor. This calfbratfon can only be .pcrforrrcd durtng an outage. 'r I ~ l'Q 21 ~ Logic trot to llrrttad to the,ctrse vhero actual operation oC ths equipoant 'fs I e pervrtrstbte.

22. One channel of etther the reacce one ur r!Cueltna aonr Reactor building

Vent tta(ton Radiation Horrttoring Syotarr rray be adrrtntatrat tvety byparord lor a period not to cacerd 'ours Cor functional testing and calibration.

(Deleted)

24. This instrument check consists of cocrparing the therrrmcouple reodinps for all valves for consistence and for notrfnal expected values (not required during refueling outages).

25. During each refueling outage, all acoustic =~n!torin'4 channels shall ba calibrated. This calfbrotion includea verification of actelerocreter response due to crechanfcal excitation in tho vicinity of tho sensor.

26. This tnstrunent check consists of conpartng the background signal levels Cnr all valves for consistency and for nocrfnal expected values (not required durtng refueling outageo).

110 NOTES FOR TABLES 4.2.A THROUGH 4.2.H Continued 27 ~ Functional test consists of the in)ection of a simulated signal into the electronic trip circuitry in place of the sensor signal to verify operability of the trip and alarm functions. I Calibration consists of the ad)ustment of the primary sensor and associated components so that they correspond within acceptable range and accuracy to hnown values of the parameter which the channel monitors, including ad)ustment of the electronic trip circuitry, so that its output relay changes state at or more conservatively than the analog equivalent of the trip level setting.

29. The fmccicnal cast frequency decreased tn once/3 months to reduce chal}cnges so relief valves per NUREG-0737, Item Zl.K.3.16.

v 3p. Oalibration shall consist of an electronic calibration of the channel, not including the detector, for range dccadog above 10 Ruhr and a one-point oourcc check of the detector bclov 10 R/hr with an installed or portable gamma source. llOa LI!'.ITING CONDITIONS FOR OPERATION SURVEILLANCE R r.@VI RFHE NTS 3.5 CORF. AND CONTAIN"NT 4.5 CORE AND CONTAINYiEHT COOLING SYSTE'iS *--: - COOLING SYSTE11S 8 'h l 3'.Y ""Pin'imum Critical 'Power. 4.5.K. Mininum -C~itical Power Ra'tio (1".CPR) Ratio (HCPR) nxmum" rcritibal.:sawer;rati,o---n- 'he m 1,->>.YCPR,@hall, be determ ned,"rdaily (llCPR) a~ a-Suncttorriof'.CScrarr > ,during .reactor powe~'-oPYrat.5on-tine anrL cor e rf$ ow;, rshall-be. equal ~ 25$ p'at~a'0'hermal power and ~ 't to or preater than shown in followinr-any change in power Figure 3.5.K-1 multiplied by the level or distribution that F;f shor1n in'igure,.3;5,.2, where;..., would cause ope"ation with a "3,imitinp control rod pattern f = 0 or~av A - B, - r-R whichever gr eater is as descrihed in:.the bases for Specification ~.3. Jr~~~D 'll ~ A=0.90 sec (Specificati'on '3~3.C;1 -.ct,".r 2.. cThe rHCPR ) limit .pha3,3.be deter- ~ ,scran time",limit'-to 20$ - hzined for. each fuel type BXB, tion" from'full.'withdrawn) c'nser 8~ "- BXBR, PBXBR, fron'Figure 3.5.K-1 respectively using: ~ B=O.710+1.65 N 'z (0.053)'Ref. 2J n a.'9i 0.0 prior to in'tial scram time measurements for <ave = ai the cycle performed in accordance with Specification 4.3.C. 1 ~ n = number of surveillance rod tests performed to date in cycle (in- , h. Was defined in Specification cluding BOC test). 3.5.K following the conclusion of each scram = scram time to 20< insertion f'rom time surveillance test fully withdrawn of the. ith rod requi. ed by Specification 4.3.C.1 and 4.3.C.2. N = total number of'ctive rods measured in Specification 4.3.C. 1 The determination of the at BOC limit must be completed with 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of'ach scram If at, any time during steady state time surveillance required operation it is determined by normal by Specification 4.3.C. surveillance that the limiting value for HCPR is being exceeded, action shall be ini.tiated within 15 minutes to ".estore operation to within the prescribed limits. If the steady state 11CPP. is not returned to within the prescribed limits within two (2) hours, the reactor shall be brought to the Cold Shutdown conditi.on within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, su. veillance and corresponding ac ion shall continue urrtil ".eactor operation is within the prescr'bed limits. 160 TABLE ).S.I- 1 I'O'LIICR VL'RSV" AVERACE PLA)IAR D(l'OSVRE Fuel T>oes: PQDRB284L, QUAD+ and 8DRB284L @vers<le Pls<<sr Exposure NAPLHCR I)"..Id/c ) (kM/Ec) 200 11. 2 ), o<)0 II 3~ S,O00 I I. <) IO,OOO 12.<) 15,000 12.0 2O,OOO 11. 8 'S,OOO Ln,non LO,R )S.ooo 10. 0 cn.non 9,4 Table 3.5.I-MAPLHCR VERSUS AVERAGE PLANAR EXPOSURE Fuel Typ< <:: P8l)R82GSH Av<!rn);e I'10<<;<r Exposure MAPLHCR (Mvd/t) (IEW/<L) 200 11.5 1,000 11. 6 5,000 11.9 10,000 12.1 15,000 12.1 20,000 12.0 25,000 11 ~ 6 30,000 I l.2 35,000 L0.9 40,000 10. 5 45,000 10. 0 I ~ ~ >>~>> ~ ~ >>>> C)Q'I)I) ' vv ~ vv v vv v'v ~ v ~ ~ i ~ ' ~ ~ ~ v v ~ ~ ~ ~ . v v A v ~ i ~ ~ ~ ~ v' ~ ~ ~ i ~ ~ v ~~ ~ ~ ~ ~ v ~ ~ ~ v ~ ~ ~ i~ ~ ~ i ~ v ~ ~ ~. ~ v v l ~ ~ v ~ ~ ~ ~ I ~ ~ v %i V<< . ~ 0 ~ v' I iv'c ~ ~,' iv ~ li v >Ca. ~ i ~ i<vr ~ ~' PgSCv v4v ~ ~ ~ v ~ v '4, ivjv%'Si3 v i ~ ~ ~ i ~ ~ ~ ~ I~ ~ ~ ~ ~ ~ ~ ~ i I ~ ~ v ~ ~ ~ /~ 0 0.1 0.2 v ~ 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 ) v Figure 3.5.K-l MCPR Limits for P8 x 8R/8 g 8R/ qUAD+ -172- TAMH3 1 A PRIHARZ CONTAIHHENT IQ)LATION VALVES Huuher of Pobder Haxiuua Action on Operated Valves Operating Horual Initiating Croup Valve Ideutif ication zuboard outboard Tiao (sec.) Position Signal Hain steaal inc isolation valves 3<T< 5 (PCV 1 1-15d 27d 14'6t 38 37r 6 51( 6 52) 1 Hain steauline drain isolation 0 GC valves (PCV-1-55 6 1-56) Reactor Mater sauple line isola-tion valves RHRS shutdovn cooling supply isolation valves (PCV-74-48 6 47) 40 RHRS - LPCI to reactor 30 C SC (PCV-74-53 6 61) Iv U< 2 HHRS flush and drain vent to CI suppression chaubcr 20 C SC (PCV-74-102'03d 119' '120) 2 'uppression Chaaber Drain 15 0** GC (PCV 75 51 6 58) 2 Dryuell equipnent drain discharge isolation valves (pcv-77-15A 6 158) 15 Dryvell floor drain discharge isolation valves (pCV-71-2A 6 28) 15 CC

• • Tllcse valves arc normally open when thc prcssure suppress)on head tank is to serve the RIIR and CS <lisc)>argo piping and closed when rhc condensate head aligned is used to serve the RIIR and CS discharge piping (Scc speaification 3.5.1)) tank

~ these valves isolate only on. reactor vessel low low water level (470") and main steam line high radiation of Group 1 isolations. TABLE 3.7.B TESTABLE PENETRATIONS WITH DOUBLE 0-RING SEALS Penetrat.ion No. Identification X-1A Equipment Hatch X-1B Equipment Hatch X-4 Head Access, Drywel x-6 CRD Removal Hatch X-25 Flange on 64-18 X-25 Flange on 64-19 X-25 Flange on 84-8A X-25 Flange on 84-8D X-26 Flange on 64-31 X-26 Flange on 64-34 X-35A TIP Drive X-35B TIP Drive X-35C TIP Drive X-35D TIP Drive X-35E TIP Drive X-35F TIP Indexer Purge X-35<: Spare X-47 Power Operation Tes t X-200A Suppression Chamber Ace'ess Hatch X-200B Suppression Chamber Access. Hatch Drywell Head Shear Lug No. 1 Shear Lug No. 2 Shear Lug No. 3 Shear Lug No. Shear Lug No. 5 Shear Lug No. 6 , Shear Lug No. 7 Shear Lug No. 8 X-205 Flange on 64-20 X<<205 Flange on 64-21 X-205 Flange on 84-8B X-205 Flange on 84-8C X-205 Flange on 76-18 X-205 Flange on 76-19 X-223 Suppression Chamber Access Hatch X-231 Flange on 64-29 X-231 Flange on 64-32 ":256- TABLE 3.7.C TESTABLE PENETRATIONS PITH TESTABLE BELLOWS X-7A P r imary S team 1 inc X-11 Steamline to HPCI Turbine X-7B Primary Steamline X-12 RHR Shutdovn Supply Line X-7C Primary Steamline X-13A RHR Return Line X-7D Primary Steamline X>>13B RHR Return Line X-S . Primary Steamline Drain X-14 Reactor Water Cleanup Line X-9h Feedvater Line X-16h Core Spray Line X-9B Fcedvater Line X-16B Core Spray Line X-10 Gteamline to RCIC Turbine X-17 Blank 257 ENCLOSURE 2 .. QW BFNP TS 199 SUPPLEMENT 2} BFNP UNIT 2 Determination of No Significant Hazards Considerations Description of Amendment Request The amendment would'revise the Technical Sp'ecifications (T.S.) of the operating license to: (1) modify the core, physics,. thermal and,.hydraulic 'Limits to b'e'"consist:ent with the reanalyses associated with replacing about one-third of the core during the cycle 6 core reload outage, and (2) reflect " changes in var'ious specifications as a result of plant modifications performed during the outage.- Specifically.;'the amendment would result in changes to the T.S. in the following areas: Core Reload Changes related to the cycle 6 core reload involve removal of depleted fuel assemblies in about one-third of the nuclear reactor core and replacement with new fuel with attendant T.S. changes in the core protection safety limits. The new fuel will include fuel assemblies of the same type as previously loaded, plus four Westinghouse "QUAD+" demonstration assemblies. The latter assemblies will be located in non-limiting locations. The actual T.S. changes include changes in the Operating Limit Minimum Critical Power Ratio (OLMCPR), deletion of tables on maximum average planar exposure for fuel types no longer used, and changes to the references cited in the bases to reflect that TVA performed the reload analyses.

2. Accident Monitoring Instrumentation Changes to T.S. instrumentation tables to add new instrumentation for high-range gaseous effuent monitors and containment high-range radiation monitors, and replace drywell pressure and suppression chamber water level instruments with new wide-range instruments in response to requirements in NUREG - 0737; items II.F.'l.1, II.F.1.3, II,F.1,4 and II.F. 1.5. A note similar to Standard Technical Specifications will also be added to describe operating limitations with less than the required instrumentation channels operable.

3. Analog Instrumentation Modify the T.S. to apply the new calibration frequency and indicator range for the new reactor pressure instrumentation. In the tables for surveillance requirements and calibration frequency for the instrument replaced, adjust'he instrument range- and change the calibration requirements to incorporate an extended calibr ation interval. The new calibration requirements, together with the new instrumentation, are expected to provide a more reliable instrumentation system.

Basis for No Significant Hazards Consideration Determination

1. Core Reload The proposed reload involves fuel assemblies of the same type (P8X8R and 8X8R) as previously found acceptable by the staff and loaded in the core in previous cycles. The reload also includes four Westinghouse fuel assemblies (QUAD+) in non-limiting locations. These assemblies are analytically similar to the P8X8R fuel such that results of analytical methods used by licensee for the P8X8R fuel bound the QUAD+ assemblies. Therefore, this proposed amendment will not create the possibility of a new or different kind of accident from any accident previously evaluated. The analytical methods used by the licensee to demonstrate conformance to the technical specifications are applicable to P8X8R, 8X8R and QUAD+ fuel and have not been significantly changed from those previously approved by the staff. Since each replacement fuel assembly is of the same type as previously added to all three Browns Ferry units and other BWRs, or is analytically similar to those fuel asssemblies, and since the codes, models, and analytical techniques used to analyze the reload have been approved by the NRC, the changes to the T.S. associated with the reload will not involve a significant increase in the probability or consequences of an accident previously evaluated. Finally, the proposed amendment will not involve a significant reduction in a margin of safety due to the reasons given above since no changes have been made to the acceptance criteria for the technical specification changes involved. Therefore, TVA proposes to determine that the proposed amendment does not involve a significant hazards consideration.

2. Accident Monitoring Instrumentation Item II.F.1 of NUREG>>0737, "Clarification of TMI Action Plan Requirements," requires all licensees to install five new monitoring systems and provide onsite sampling/analysis capability for a specified range of radionuclides. For all six categories, NUREG-0737 states: "Changes to technical specifications will be required." During this refueling outage, the licensee will install: (a) a gaseous effluent high-range radiation monitoring system, (b) a containment high-range radiation monitoring system, (c) a drywell wide-range pressure monitoring system, and (d) a suppression chamber wide<<range water level monitoring system.

These items were required by NUREG<<0737, items II.F.1.1, II.F.1.3, II.F.1.4, and II.F.1.5, respectively. The changes to the T.S., which track the model T.S. provided to the licensee by the staff, are to add operability and surveillance requirements on the new monitoring systems. The revisions also delete the present drywell pressure and suppression chamber water level instruments since they are being replaced by items (c) and (d) above. The changes to the technical specifications are necessary administrative follow-up actions required by the Commission. <<a i. I The proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated; or create the possibility of new or different k1nd of accident, from any accident=previously .evaluated since no modifications are made to.any safety related;equipment and.: procedures;for..plant. operation are not changed..Neither does the proposed;amendment.involve a .significant,'reduction'ina. margin of safety,-sinceftechnical>specification-acceptance, criteria are not, reduced;.;. Therefor e',,;TVA;.praposescto det'eemine;that;.thecproposed, amendment daes notninvolve"a>significant:..-hazards consideration. ~ y1] ",

3. Analog Instrumentation, iiv i" \ C'i ~ ii&1 UI 4 I I << '4 v 'tJiI ' -

The modification,'involves-.removing,'one. devic'e"and substituting APPPll Ak 0 4 .t " -} V4 +-'I c- ~ another device to perform-'the same funct1on-..". Changes. in';-design bases';" protect!ive. function, .redundancy,=- setpoints-and logi'c'-ar e not involved. .However, the new indicator range is.0-1200 psig and the c'alibration-".interval has been increased commensurate with the reduced drift-for'"the new instrument: 'However'~because the" - modification..and T.S. change will. not eliminate or modify any"-i protective. functions nor:.permit any new operational 'conditions, they"do="not create'he. possibility- of a'ew'-kind of.'accident or significantly increase 'the probability or consequences of- an-- accident previo'usly "-evaluated. Because 'of the increased reliability and-stability,='and-r educed>> drift of-=.the analog trip system, the increased cal1bration.intervals would:not-reduce any safetyimargin" 'e  :;ro,:o=".'~end".on;- c.c .-s no- 1:.'.-.; .'"igns f'.c .. '..c~ar . '" considcri "'.'-'c:.. h Therefore, TVA proposes to determine that the proposed amendment does not involve significant hazards considerations. c ~a 4 . i v V>> T'i.~ ~'f pppp,r<

> limitations with less'than the required instrumentation:.channels o'operable for. the instruments zadded by-<the,-orggkpal~~amengnent. in -=responsento requirements "Xn NUREG-0737verÃokq 8~qequfrqs an,~ s-a'lternateimcnitoeing;method'ctocbe used when less than the required operable channels are available. Therefore, the amendment'does.not adversely:,effect safe plant operationo -.,~',- Jl" ~ << 2.. Pages.78 and 105 -.Page 78,.shows the. correct instrument range (0- =-]200psig) < for= the=reactor, pressure indicator and, page,105. shows anthe required, calibration,.frequency of-once. per.,12 months for this indicator at its proposed range. The original amendment request pdiscussed-replacing the old reactor pressure, instrument with a elqew. analog-system .-. This;new~system-.is more..accurate.and.,3.ess , -pr one to drift than the:old system, and the required calibration frequenoy for the new reactor pressure indicatorI.3ias .been .-;evaluated and determined.to.be greater than 12 months. Since a 0- . 1200 psig, range is acceptable for all, required postaccident monitoring functions and the 12-month calibration interval is ,.preferred to,a,6-month intervaland this, combination maintains., ~>the required accuracy, .this changeiwill.not adversely effect ~plantsafety.,. icy ,'as xnac~crtenzd cp't~~~. 3.-;.Pages 171 and 172.- -Revise ~ the~tables.for,iHAPLHGR,and..the..Figure d.3.5>K-1 for -HCPR limits to reflect the updated limits for cycle 6 operations. The Justification and safety analysis for these ..revisions are described in TVA-RLR-002 ~Revision,1. A Jg* <<P +PhfOPi Q

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