Text

Tech Spec Replacement.* Forwards Tech Specs Requested at 900323 Prehearing Conference Re Certain Proposed Contentions to Enable Board to Determine Nature of Difference Between Two Sets of Sections
	ML20042D851
Person / Time
Site:	Turkey Point
Issue date:	04/04/1990
From:	Reis H; FLORIDA POWER & LIGHT CO., NEWMAN & HOLTZINGER
To:	Anderson G, Bloch P, Johnson E; Atomic Safety and Licensing Board Panel
References
	CON-#290-10196 OLA-5, NUDOCS 9004110015
	Download: ML20042D851 (90)
	v • d • e

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,.,..r i..c,r 6 1 I .mn. u. &cDTT SL.96.as,4.. I W I ( 1 t.rsh.LD, 6i6W t.w as top ..o u..,6,+.. L4 64(-44 0w Peter B. Bloch, Chairman Atomic Safety and Licensing Board r i U.S. Nuclear Regulatory Commission i j Washington, D.C. 20555 i l Dr. George C. Anderson Elisabeth B. Johnson j 7719 Ridge Drive, N.E. Oak Ridge National Laboratory i Seattle, WA 98115 P.O. Box 2088 Bethel Valley Road, Bldg. 3500 l Mail Stop 6010 l i Oak Ridge, TN 37831 c Ret Florida Power f. Licht Co. (Turkey Point Plant, Unit I Nos. 3 and 4), Docket Hos. 50-250-OLA-5 and 50-251-OLA-5 (Technical Specifications Replacement) l t

Dear Licenairig Board Members:

During the prehearing conference held on March 23, i 1990, the Licensing Board requested the Applicant to submit sections of the current and revised Turkey Point Technical i specifications related to certain proposed contentions to enable i the Board to determine the nature of the difforence, if any, i between the two sets of sections. In response to that request, please find enclosed the following: A table that identifies the relevant sections of the Technical Specifications. j Copies of the relevant sections. Si cerely, l 9004110035 900604 r PDR ADOCK 05000250 /1 0 FDfi Harold F. Reis ( l cc (w/ encl): Service List l 1 s y

t Dest 1 Sectieses of the Correset ased Davlend Ttschey poiset Techselcel specificetless pelated to cozinia Prescoed e-_ M _*_ I Lw__.f meleveset Sectless of Curreset Deleweest Smetles of Davised Dneeripties of Ayylicable fas_tgestics

prh Ieat sameari m aamm 3gsclue& sal._EpeG111cetjene C _

4 Table 4.1-1 Sheet 5 (Attachment 1) Table 1.1 (Attachment 2) The correset Techalcel syncificatieses include a hoy for the fam..c y i cedme ese ehest 5 of the table. The reviend i Techr.leel speelficatione include a boy for thm ) f w codes en a _eie table (Table 1.1) in the Definitiesse sectlen. l .i 5 Table 4.1-1 (Attachment 1) Table 4.3-1 (Attachment 2) The revloed table includes additiesnel reetractieses esed relasetiosas. Tlso relasetlesse are 41eewooed en pagee App. A 3/4 3-1 to 3-5 of the Jesse 5, 1999 applicotten i (Attacw 3). i i e t t .m_, m e~..~ v -.m-. ,.w.-- w+--ww -. ~==. =r, - < yw --e-- w- -e ,-.m-ww-e--n'- e+-w-o--gwmv -w-ww=-e - - --wwwww w w.ew-.+- w-ew-'

PeeE D Sections of thus Current and Dewlead Tethemy Pelet Tactenical specificetlene marInted to Certala P W f*==8==* toes Lw.2 melevent section of current helevant sectiese of ghuwjemd Speerj tion g( Agg1& cable p Costantion h Era 1 Emwo-4 r me inam h E.ca1 herars,meaa== Cheese i 7 Section 2.1 and page Section 2.1.1 Section 2.1 ef tDee (Notes See 3.1-7 (Attachseent 4) and 3/4.4.1 {Attoclument 3) current Thical eleo pi end specificatiouse providne Conterition 30 for pe==r below) operations oath three loope, two loops, esse t leep or sisterol circulation esed provides limite for each. Nome, r, the flow requir===nte for Dus

parametere et g-og= 3.1-7 in the curreset Technical specifications r= quire 3 Icep eperation at pe==r.

The reviend Techselcel Specificetter.= claer up thie Isecosteletessey by requiring 3 loops in poden 1 esed 21/. t 7 Figure 2.1-1 (P.ttachment 4) Tigure 4.1-1 (Attoclummet 5) Eks changa to corvos or setpolate. A grid le added to sehe cerves esey to reed by pleet f i OPeretore. l J/ ohe Modes are defined in revised Technical Specification Table 1.2 (Attachemet 23). t. I t 1 i .-.,e_-s- .... _ -. - -.. - - - - ~,, - - _ - - -m.- ,,. -_--.-._ y

4 Man 3 Sections of the Curreset asul stevised Terboy Feint i Technical Specificationa steleted t.o c giele Fa.. _hsnes e t F, %. f melevant section of Currwest televant section of meessed peersytlen of Appliemble CentsutLisms .1mochnicet smartrtratinne ph= t,-= 1 h=cIrsc= eta== _e_. __ i 7 section 3.1.1.c sactione 3.4.2.1 and 3.4.2.2 The corrent T=chnteel and B2.2 (Attachment 17) (Attachment 24) Specifications reqpsire preeserizer oefety volves i to be operable, but do not eracify tt velve set-preseere toler cas. 2/ The review Technical Specificetlene require preeeurteer esfety velves to be operable with a lift setting of 2405 poig +/- 14. i i 4 2/ section 82.2 of the seees for the current Technical specificatione state thet the settings of the oefety volves are 2485 pelg, and that *velve set pressere tolerances should be those stated in applicable codme.* Wseder Section 4.1.7 and Table 4.1-3 of the Updated Final Safety Analysis 5t= port for Turkey Point, the presevelser oefety velves are ASME Code velves; the ASNE Code, Section III specif1*e +/- 14 tolerance. Further, the Stenderd Technical specificatione for meetinghomme plants hos required +/- It tolerancee on tneee valves since 1976. ,,--.m -...ws e,. ,4-.__,-*w- --%ve--_,-_r- .vv.,_,,e,,----e_,.,,~_.,,.m-_y, w .,. ~.., m,_..-.- + - - sw.._ __w,,

I mO i Sections of the Corrent and Itsmrleed W Point i wchalcat Sp=cificee.sene noneted to C W =t= L-,- __^_Ceef48-Jsse 71,_._ l Delevent SectIm of Current pel-went Sectjese of sesmrlead Damer$pt.len of applicable Estatention w hmical Sascifications W cal Sessificatines C _ 8 Section 2.2 end Section 1.1 Section 2.1.2 ( Attachment 5) T9= corrent T=chnical (Attachment 6) and 5=ction 6.7.1.d Specificatione r m ire potes Petitionero eleo reference (Attachment 8)

ehetdoen* If a oefety Section 2.1.2 and Section 6.3 11elt le encended (no

( Attoclasent 7). There to no Section time limit opaciflod)- 2.1.2 in the corrent Technical The revloed w hnical Specification and Section 6.3 Specificatione sculd appears to be irrelevant. Include en ACTION statement regelring the reactec to be in het etendby in I hour if a

Safety Lielt le encoeded.

enceeded. i 9 Section 2.3 (Attachment 9) Section 2.2 (Attachment 18) In the revised m leelcel Specificatione, a prowleton for A11emoble Veloce le added to the table of trip omtyeinte, but the Alleueble volese are left blank. Page 2-19 of the regioed Technical Specificatione states that if no Alloweble hien le epocified, the trip oot t point le meteetly the Alleueble Velse. g ? 1 l

9 Pass O I Sectione of the Current and newtood W Point Techalcol Specificatione heleted to Car *= h Fr __ 2 W a ^1._ 1 2 6,c nelevant section of Current mel=went sectIon of n=wlead n=ectlytles of appliemble E9mLERL19B W GSI Saec Hicstless feckedc_Nir.stiges Change 11 4.0 (Attachmant 25) 4.0.1 (Attachment 26) The rewlead h hnical specifications would add a n=w section 4.9.1, which states thet serve 111once requiremonte only apply in the medes in which the swaa a g ing egyulpment le reggelred to be eparable (unless otherwise specified).

Useder the corrent Technical specificatione, tInere are eene instances in which a sorwe111nece

~ requirement epy!!se shon the -.__ - Ting ognipment le *wt required to be operable. l -i j = e-_,, e -n -e.-.e

4 t enos O Sections of the Currwat and newined Ttartey Potet. TactueIcel Specifications meisted to h _ -.-_" ' not amt Efine Aw;; helevant section of Current selevent. Secties of newtood smacription of Aprilcable Egetsuntips 23g;.hmiral W Irlemt.Enem 2prJudse2_SymrI rIceLiups C _ i 21 3.2.3 (Attachment 11) 3.1.3.4 (Attachment 12) The current Technical Specificettene state thet th= control red drop time shell be no greeter then 2.4 seconde from the

beginning of red motion to deshpet entry.* The revloed Technical

[ Speelficatione state that the centrol red deer time j ohell be no geweter then 2.4 encende frem the

beginning of decoy of l

stetlenery gripper cell l voltage to f: A A entry.* l 25 4.20.1 (Attachment 13) 4.4.9.1.2 (Attachemet 14) We sobetontive chenT*= 1 t O .P s ~g .--.. ~. -. y --w-3 - - ---,*ww etm - ---er-v - +-- e -+-~-+-w,, e+ +==-v-o-, ,-c --w m-- - - --3--w-~me- -*e--e-..-.-w+---e-e* =ay

t pass 7 4 sections of the corrwest ased a=vle=d Tischey point 1echsencal specificatione meisted to W =n= & r % = l* - L_ melevant sectlen of corrent selevent section of movissed sneeriptles of appI1 cable Cententine 2HELBumisal Spec 111rJtt19ee 2pGhaical See_giticagiens C 30 3.1 and 3.4.1.c (Attachment 15) 3.4.1.1 (Attaciument 16) In the currwat Techselcal Specificatione, the reoctor moy be eparated j with less these all roector coolant pumpe, i etmee genereters, etc. operable; houseme, all three r=ector coolent Iceps meet be operable with rooynet to the Safety In)octiese and meeldeel most memovel Systems er the reacter most be in mot sesetdeses in one hour. In the revloed Tactueelcel Specificetlene, all l roector coolant loope j meet he in operetten in Isodne 1 and 2 or the reacter emot be la mot Stessey in ela heers. i 1 I ( _... ~<m -e<w ,--.---~=w ..ee- ,----~-r -3,-.--wr =.--w---w+-e.v* e.w-- --we- ~-.we--%e.#-, --+%

L pass O Sectione of the 0%rrent and aswisest Tterboy Peint Technical Specifications Deleted to W M = 7 _ __ f M --? "__ _ + Le2 melevant Sectles of 0%rrent selevent smetlen of newleed Descriptlee et appliceMe fas4 6 Technical 6

== chm & cal Samriftrattame F- ] 35 3.1.1.e (Attachment 17) 3.4.4 (Attachment IS) The current TechnIcel 4 Specificatione resguire both the power operated rollet velves (PURT) and i 4 the eseociated block velves to be operable i when the roectcr coeleet le greater then 359*F. i Tha revised Technical Specificatione require the PORT block velves to ' be operable during modse 1, 2, and 3 (greater then er equel to 350*F). l 35 3.15.3 (Attachment 19) 3.4.9.3 (Attachment 20) so sabotentive chenye. i 51 3.7 and 4.3.1 (Attachment 21) 3.8.1.1 and 4.8.1.1.4 The revloed Tochalcel (Attac h t 22) Specifiestione incledo 2 ACTION Statemente and Serve 111once pegysiremente related to the crenkinT diesel. - -_nore that l ar= not in the current i hical Specifications. [ f a e I

n-

+ -.. - - -. -,., _., -r

.-,.__-~,--,-------,..n.,-,..

Attach ent pe. 1 ~ m l vasaz 4.i.-l Mininum reequencies son cercas cas.smaarness ame Tast or insteentwr imamatas canneet asscarrrios caect cas.namata Test messess t s(l) o(2) se(3)

1) 1.eed we. fles caree, or aT vs. reacter poser G 3.

. neeseer Fe==r sense (check, Colthrete end Test only se*(4) Eg*(4)

2) Theemet yemer ceteetetsee applicable ebees I M of

3) staael to 8t', bestable settee rated pomer.)

(peentestes, sed step, tripel l

4) Opper & leser detectose for speestrSe of fset (45 to -31).

d E

b. Famer Statettetten IEsp H(I) l) Felle 4eq thttlet leedtag and prter to (2) opereteen above FSE pomer.

0)

2) esee per esseettee rett pesar emeth.

l

3) centsee lyt ensenet tecemee entban steate.

4 l 2.- m-n.e, sateemedtete aseen s(1)+

u. A.

F(2) I) eseelehaft up to SSE R.F. l

2) tes teost; hastehne setsen (reseaseten, ved eemp, trap) 3.

Eb-laa,Seurte asses s(l) II. A. F(2) l) Snee/ shaft eben as servlee.

2) Diesble, settee (elese, trep)

4. -

Roseter costeet Temperature s+ R S/W(B)* I) overtempeestese a T (2)+

2) oveeposer a i

S. Se eter cessant ytes s+ a u* l 4 6. Pressurseer taster too.1 s+ a m* i 7. Pro rs e ere s+ a n+ l S. 4 kw Weltage & Freyseecy N. A. 'R** a Reactor preteettee eteestte enty i l 9. Asenes and esenesee s+ a n+ unge ete,es ster,. Amendment. Res. M & 74 i h 'n n---,-,,--,----- -.. - - - --, -,--- - - _ n -,_

Attac?nsent No. 1 TABt.E 4.1-1 SHEET 2 Channel Description Check Calitrate Test Rc --_ b t 10. Rod Position Bank Counters St N.A. N.A. With analog Rod Position i1. Steam Generator Level St R Mt t 2. Charging Flow N. A. R N.A. i-13. Residual Heat Removal Pump Flow N.A. R N.A. 14. Boric' Acid Tank Level T R N. A. i 13. Refueling Water Storage Tank Level Wt R N.A. 16. Volume Control Tank Level N.A. R N.A. 17A. Containment Pressure - Narrow Range i#t R N. A. 178. Containment Pressure - Wide Range - l#t R N.A. 13 A. Process Radiation * *

D N.A.

M t I88. Area Radiation D A M 19. Boric Acid Control N.A. N.A. R 20. Containment Sump Level N.A. R N.A. 21. Accumulator Level and Pressure T R N.A. 22. Steam Line Pressure i R W i ..TaJ..c;; Nos.110 and 104 ^ _ =- 1

._w, ~ - - - - - 8 NG 9* N. t eorem f_.s Attachunent No. 1-Clieck Calibrane Tgg stennerks Channel r'- -= - i KA. K A. Mi i 21. Logic Channels l R A. A M i 2e. Emer. Persable Survey baseremments, l 1 RA. N.A. Q Make trace. Tees bessery (csenege m ' _ - l l 23. Sessanegraph 1 Mt R RA. i 26. Anaminary Feedesser FIsar Rees j Mt R KA. i RCS Subceeheg Margin tennieer .j 27. t Oseck cameises of smenteering l 28. PORV Reeleien Indcasar ~ Mt R A. R (Ptemary Deesceer) i Mt

R A.

R indecated @ and verifyleg i PORV Stock Valve Resisten bedecator 29. Mt R KA. by aboarveelen of reteoed \\ i. 30. Safety Valve Resialen bedicacer J f e MA. N.A. R For APW acessaien as pauser j

a. t. ass of Vetenge M thw knessess) enly i

e 31. l l

b. underwetenge theek 4KV busses and 5t R

Mt l l 4se vele Lead ceneerslaa R A. KA. R For APW acoussion es peeper l l 12. Trip of heeh Main Foodweser ~ enty l l Pissup threakers ) 3). Turbine Trip N.A. R. g.A. (Aeste-Seep 011 Freesere Seeiteties) 1 l I Tha 6eem does noe apply ese these D une64 af ter iniplemeneassen of PC/44 79-186 and en the4e 4 esteil after "" d of f PC/M se-44. Amendment Mos.110 and 104 j Tl8:e 3 1 ..,..n

Attochment No.1 ~ 4 t TAmt.E 4.1-4 5HEET 4 'W c _. ";"__ Check Cashrase Test memesens ] M. Centainment Tater I.evel(Narrow 8tangel Mit R M.A. i I 31. Centainment Water Invel(Fede Rangel Mtt R tiL6 l w. c-seggi Raase Area Madiation 5tt R(Note in M j

37. Centainment Stydrogen RAnnieers 5t Q(Il

'A t (t) Onannel calibration ensing i

a. One vetrue percent sample gascentainings

. N belance nieregen.

b. Four vehsee percent belance regin Range eshte Gas Efftsent AAnnitors m

18. I I

a. Plant Vent Eshassa 5

R M

b. tanta 1 Seent Insel Fit Eshessa 5

R u

c. ch Air Epoctors 5t R

4t l

d. MehnSesem Unos 5t R

wt

19. Decese W Mit R

N.AA5ee Steee Tl l -(Cese Esta ^___ -- ; l l i 40. Reacter Vessel Invel Monleering System Mtt R as.A. l i = ? l i s i i f h Amendment Mes.I25_ am8 II9 l 4

Attachment No. 1 TABLE 4.1-1 SHEET 5 i

- Using moveable in-core detector system.
- - Frequency only l
* * - Applies to containment particulate (RI l) and gaseous (R12) monitors only. For etfluent monitors, refer to Tables 4.1-3 and 4.1-4.

PR - Prior to each release 5 - Each Shift D - Daily W - Weekly 4/M - At least 4 per month at intervals of no greater than 9 days and a minimum of 48 per year B/W - Every Two Weeks M - Monthly Q - Quarterly P - Prior to each startup if not done previous week j R - Each Refueling Shutdown A - Annually N.A. - Not applicable l t - K A. during cold or refueling shutdowns. The specified tests, however, shall be performed within one surveillance i interval prior to starttp. tt - K A. during cold or refueling shutdowns. The specified tests, however, shall be performed within one surveillance intervai prior to heatup above 200F. NOTES: 1) Acceptable criteria for calibration is provided in Table ILF.1-3 of NUREG 0737.

2) Compliance will depend on instrumentation operability.

b AMnent Nm.HO and #

AttEchment No. 2 ^ TABLE 4.3-1 5lli! REACTOR TRIP SYSTEN INSTRUNENTATION SURVEILLANCE REQUIREE NTS Q TRIP i mS ANALOG ACTUATING M00E5 FOR I CHANNEL DEVICE WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED c-z d 1. Manual Reactor Trip N.A. N.A. N.A. R(11) N.A. 1, 2, 3*, 4*, 5* 2. Power Range, Neutron Flux a. High Setpoint 5 D(2,4), M N.A. N.A. 1, 2 i M(3,4), Q(4,6), R(4) b. Low Setpoint S R(4) M N.A. N.A. 1***, 2 w1 3. Intermediate Range, S R(4) S/U(1),M N.A. N.A. 1***, 2 w 4 Neutron Flux i 4. Source Range, Neutron Flux S R(4) S/U(1),M(9) N.A. N. A. 2**, 3, 4, 5 f 5. Overtemperature AT S R(12) M N.A. M.A. 1, 2 i 6. Overpower AT S R M N.A. N.A. 1, 2 k 7. Pressurizer Pressure--Low S R M N.A. N.A. I 5 M 8. Pressurizar Pressure--High S R M N.A. N.A. 1, 2 4 9. Pressurizer Water Level--High S R M N.A. N.A. 1 z8

10. Reactor Coolant Flow--Low S

R M N.A. N.A. 1 I R

11. Steam Generator Water Level-- S R

M N.A. N.A. 1, 2 o low-Low I 4 I o l en a) l

Attachement No. 2 m .BLE 4.3-1 -4 C g N REACTORTRIPSYSTEAINSTRUENTATIONSURVEILLANCEREqulREENTS 0 TRIP S ANALOG ACTUATING MODES FOR CHANNEL DEVICE WNICH e CHANNEL CHAleIEL OPERATIONAL OPERAil0NAL ACTUATION SURVEILLANCE c FUNCTIONAL UNIT CHECK CAllBRATION TEST TEST LOGIC TEST IS REQUIRE 0 z U

12. Steam Generator Water 5

R M N.A. N.A. 1, 2 Level--Law Coincident with Steam /Feedwater Flow [ Mismatch

13. Undervoltage - 4.16 kV N.A.

R N.A. N.A. N.A. 1 l i Busses A and B 3 ~

14. Underfrequency - Trip of N.A.

R N.A. N.A. W.A. 1 Reactor Coolant Pump g Breakers (s) Open 4 i

15. Turbine Trip i

T a. Autostop 011 Pressure N.A. R N.A. S/U(1,10) N.A. 1. b. Turbine Stop Valve j Closure N.A. R N.A. S/U(1,10) N.A. I 16. Safety Injection Input from ESF N.A. N.A. .N.A. R N.A. 1, 2 i 17. Reactor Trip System Interlocks g i m a. Intermediate Range i j Neutron Flux,-P-6 N.A. R(4) M N.A. N.A. 2** E b. Low Power Reactor Trips Block, P-7 N.A. R(4) M(8) N.A. N. A. - 1 i 5 (includes P-10 input i P and Turbine First l Stage Pressure) E c. Power Range Neutron Flux, P-8 N.A. R(4) N(8) N.A. N.A. 1 i b ~ i oo os (LP i

Attcchmerit No. 2 ^ JLE 4.3-1 i e jlll REACTOR TRIP SYSTEN INSTRUNENTATION SURVEILLANCE REWIREMENTS '2 TRIP ,o ANALOG ACTUATING MDOES FOR 5 CHANNEL DEVICE WHICN f CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST 15 RE@lRE0 e i5 i d

17. Reactor Trip System Interlocks (Continued) w d.

Pouer Range Neutron Flux, P-10 N.A. R(4) M(8) N.A. N.A. 1, 2 i 18. Reactor Coolant Pimp N.A. N.A. .N.A. R N.A. 1 Breaker Position Trip 19. Reactor Trip Breaker ~ N.A. N.A. N.A. M(7,11) N.A. 1, 2, 3*, 4*, 5*: w1

20. Automatic Trip and Inter-i wa lock logic N.A..

N.A. N.A. N.A. M(7,14) 1,.2, 3*, 4*, 5*- o 21. Reactor Trip Bypass Breaker M.A. N.A. M.A. M(13),R(15) N.A. 1, 2, 3*, 4*, 5* '1 ~ i 5 1 E E i G ~ l z l ? i e ( l 0,:

TABLE 4.3-1 (Continued) Attachment No. 2 i TABLE NOTATIONS l CWhen the Reactor Trip System breakers are closed and the Control Rod Drive i ( System is capable of rod withdrawal. neg ),w p.6 (Intermediate Range Neutron Flux Interlock) Setpoint. anag,jow p.10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint. (1) If not performed in previous 7 days. (2) Comparison of calorimetric to excore power indi. cation above 15% of RATED THERMAL POWER. Adjust excore channel gains consistent with calorimetric power if absolute difference is greater than 2%. The provisions of Specification 4.0.4 are not applicable to entry into MODE 2 or 1. (3) Single point comparison of incore to excore AXIAL FLUX DIFFERENCE above 15% of RATED THERMAL POWER. Recalibrate if the absolute difference is greater than or equal to 3%. The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1. (4) Neutron detectors may be excluded from CHANNEL CALIBRATION. Thi' table Notation number is not used. (5) s (6) Incore-Excore Calibration, above 75% of RATED THERMAL POWER (RTP). If between 30% y surveillance requirement coincides with sustained operation the quarterl and 75% of RTP, calibration shall be performed at this lower ( power level. The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1. l (7) Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS. l (8) With power greater than or equal to the Interlock Setpoint the required ANALOG CHANNEL OPERATIONAL TEST shall consist of verifying that the i interlock is in the required state by observing the permissive annunciator window. (9) Monthly surveillance in MODES 3*, 4*, and 5* shall also include verifica-tion that permissive P-6 and P-10 are in their required state for exist-ing plant conditions by observation of the permissive annunciator window, Monthly surveillance shall include verification of the High Flux at Shut-down Alarm Setpoint of 1/2 decade above the existing count rate. (10) Setpoint verification is not applicable. (11) The TRIP ACTUATING DEVICE OPERATIONAL TEST shall include independent verification of the OPERABILITY of the undervoltage and shunt trip attachment of the Reactor Trip Breakers. I TURKEY POINT - UNITS 3 & 4 3/4 3-11 AMENDMENT N05. AND FEB to n.a

Attach m t No. 2 TABLE 4.3-1 (Continued) TABLEkOTATIONS l

12) CHANNEL CALIBRATION shall include the RTD bypass loops flow rate.

l (13) Remote manual andervoltage trip when breaker placed in service. (14) Interlock togic Test shall consist of verifying that the interlock is in its required state by observing the permissive annunciator window. (15) Automatic undervoltage trip. em e 0 0 t b TURKEY POINT - UNITS 3 & 4 3/4 3-12 AMENDMENT NOS. AND FEB 2 81559

Attcch:Ont No. 2 i TABLE 1.1 FREQUENCY NOTATION i N,0TATION FREQUENCY s At least once per 12 hours. O At least once per 24 hours. W At least once per 7 days. M At least once per 31 days. Q At least once per 92 days. I SA At least once per 184 days. I R At least once per 18 months. 1 S/U Prior to each reactor startup. N.A. Not applicable. P Completed prior to each release. l I 1 i 4 l l 1 TURKEY POINT - UNITS 3 & 4 1-7 AMENDMENT N05. AND

..]

Attach:2nt No. 3 N0 $1GNIFICANT HAZARDS EVALUATION ( PROPOSED TECHNICAL SPECIFICATION TITLE: REACTOR TRIP SYSTEM INSTRUMENTATION NO: 3/4.3.1 A. DESCRIPTION OF CHANGES 1) Present Condition of License: As described in the current Turkey Point Units 3 and 4 Technical Specification in Specification 3.5.1 and Table 4.1 1. 2) Proposed Condition of License: a. The amendment consolidates the current requirements into this specification and explicitly states the LCO, APPLICABLE MODES, ACTION Limits and SURVEILLANCE REQUIREMENTS. b. The revision is more complete than the current Technical Specification as follows: 1. A complete list of trip channel and trip actuation devices (reactor trip breakers) is included in Table 3.3-1. ( 2. Trip Channel OPERABILITY requirements are included for MODES 3, 4, and 5. c. The revision relaxes the following current requirements: 1. The current bi-weekly surveillance interval for the OTd-T and OPd T analog channel operational test is relaxed to monthly, r 2. The 480 V Load Centers, Reactor Trip Breaker and Automatic Trip and Interlock Logic Surveillance interval is relaxed from monthly to bi-monthly on a staggered test basis. j 3. The CTS Table 3.5-1 implies applicability for Modes 1 and 2 for; 1 Item 5 Pressurizer Low Pressure Item 7 Pressurizer Hi Water Level Item 8 Low Loop Flow Item 9A 4 KV Bus Undervoltage Item 98 4 KV Bus Underfrequency item 9C RCP Breakers and the RTS Table 3.3-1 indicates only Mode 1 applicability for those channels. 4. The proposed change relaxes the channel out of service requirements by permitting an inoperable channel in a 2 out of I 4 logic to be bypassed for up to 2 hours to provide for surveillance testing of the other three channels. l App. A 3/4 3 1

I Attcch ent No. 3 Proposed Technical Specification No. 3/4.3.1 i 5. The applicability of the Calorimetric Power to Nuclear ( Instrumentation power indication comparison surveillance is relaxed from 10% to 15% of RATED THERMAL POWER. j B. BASIS FOR NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION' ) The standards used to arrive at a proposed determination that the changts i described above involve no significant hazards consideration are included in 10 CFR 50.92. The regulations state that if operation of the facility in accordance with the proposed amendment would not: (1) involve a significant increase in the probability or consequences of an accident previously evaluated, or (2) create the possibility of a new or different kind of accident from any accident previously evaluated, or (3) involve a significant reduction in a j margin of safety, then a no significant hazards determination can be made. The Commission has provided guidance concerning the application of the standards for determining whether a significant hazards consideration exists by providing certain examples (48 FR 14870) of amendments that are considered not likely to I involve a significant hazards consideration. Example (i) relates to a purely administrative change to Technical Specifications: for example, a change.to achieve consistency throughout the Technical Specifications, correction of an error, or a change in nomenclature. Example (ii) relates to a change that constitutes an additional limitation, restriction, or control not presently included in the Technical Specifications for example, a more stringent surveillance requirement. 1) The proposed change as described in Item 2.a is similar to example (i)tes of f 48 FR 14870 in that it is an administrative change which consolida \\ current requirements into a technical specification format consistent with the Standard Technical Specifications and does not involve technical or plant modifications. 2) The proposed changes as described in items 2.b.1 and 2.b.2 are similar to example (ii) of 48 FR 14870 in that additional controls in the form of a more complete. list of trip channels is included in the list of trip channels required to be OPERABLE and trip channel OPERABILITY requirements for MODES 3, 4, and 5 are included. A surveillance requirement on the reactor trip breaker is also included. 3) The proposed change as described in Items 2.c.1 and 2.c.2 to relax the OTd-T and OPd T trip channel surveillance intervals from bi weekly to monthly. and the trip channel actuation logic surveillance interval from monthly to bi-monthly on a staggered test basis, does not involve a significant hazards consideration because this change would not: a. Involve a significant increase in the probability of or consequences of an accident previously evaluated. Thc analog channel operational test verifies that the trip channels are able to perform their trip function. Past experience at Turkey Point over a typical 12 month interval consisting of OT T and OP T bi-weekly trip functional surveillance tests have shown that trip channels failed the surveillance procedure acceptance criteria on App. A 3/4 3-2

Attachment Ns 3 Proposed Technical Specification No; 3/4.3.1 j only 5 of 150 tests. Furthermore, the surveillance procedure I ( acceptance criteria is more restrictive than the Technical Specification acceptance criteria for setpoints and includes both high and low side setpoint drift. Based on this observed trip channel reliability the proposed relaxed .turveillance interval will not degrade the trip system reliability. Therefore, this change will not significantly increase the probability of or consequences of any previously evaluated accident. In addition, the Westinghouse owner's group has recently completed a reliability and risk analysis of the reactor trip system which is documented in the WCAP 10271 series of documents. This analysis shows that the analog channel surveillance test intervals in the reactor trip system can be relaxed from monthly to quarterly with no increase in risk as estimated by the core melt frequency prediction. This analysis also contains some calculations on the sensitivity of the trip system reliability to changes in the actuation logic test interval. These calculations show that the system reliability is insensitive to relaxation of the actuation logic test interva,1. Therefore, these changes will not significantly increase the probability of or consequences of any previously evaluated accident. l Also, the monthly surveillance of OTd T and OPd-T trip channels and the bi-monthly staggered actuation logic test interval is consistent 1 with industry practice in that it is the surveillance interval in the Standard Technical Specifications. i ( i b. Create the possibility of a new or different kind of accident from any previously analyzed because the proposed change introduces no new mode of plant operation nor involves a physical modification to the

plant, c.

Involve a significant reduction in a margin of safety because of the high reliability of the OTd-T and OPd T trip channels and actuation logic as demonstrated by the current surveillance test results and owner's group programs which quantify the reactor trip system reliability and contribution to total plant risk. t i 4) The proposed change to relax the applicability for the reactor trip channels identified in 2.c.3 above, does not involve a significant hazards consideration because the change would not: r a. Involve a significant increase in probability of or consequences of an accident previously evaluated. These trip channels are blocked for operation below the P-7 setpoint of 10% power. As this block functions under 10% power, it effectively makes the trip inoperable in modes other than Mode 1 (as defined in theRTS). The change to show Mode 1 applicability is consistent with and does not change the actual function of these protective trips and consequently does not increase the probability of or consequences of an accident previously evaluated. App. A 3/4 3-3 _...,y

Attcchment No. 3 i Proposed Technical Specification No. 3/4.3.1 l I b. Create the possibility of a new or different kind of accident from any previously analyzed because the proposed change does not introduce ( e new mode of operation or involve a physical modification to the plant. c. Involve a significant reduction in a margin of safety because the i protective function of the reactor trips has not been changed. 5) The proposed changa as described in 2.c.4 to provide for a bypass of the inoperable channel of a 2 out of 4 logic for 2 hours does not involve a significant hazards consideration because the change does not: a. Involve a significant increase in the probability of or consequences of an accident previously evaluated. The channel bypass relaxation is justified, as a 2 out of 4 sigr.a1 logic provides sufficient redundancy. When the inoperable channel i is bypassed, the logic becomes 2 out of 3 and with the channel under surveillance being put into the tripped mode, any one of the remaining 2 channels is sufficient to make up the trip signal. As the makeup of the trip signal is not impaired, there is no increase in the probability of or consequences of an accident previously evaluated. b. Create the possibility of a new or different kind of accident from any previously evaluated because the proposed change does not introduce a new mode of plant operation or involve a physical modification to the plant. I Involve a significant reduction in the margin of safety as the change c. does not alter the generation of a trip signal from an actuation j signal of two channels. 6) The proposed change as described in Item 2.c.5 to relax the applicability of the Calorimetric Power to Nuclear Instrumentation power indication comparison does not involve a significant hazards consideration because the change does not: a. Involve a significant incr6ase in the probability of or consequences j of an accident previously evaluated. Applicability of this surveillance at 15% is in accordance with industry practice and is the value used in the Standard Technical Specifications. Due to accuracy considerations, Calorimetric measurements below 15% RTP have no practical value, b. Create the possibility of a new or different kind of accident from any previously analyzed because the proposed change introduces no new mode of plant operation outside the requirements of the LCO nor involves a physical modification to the plant. App. A 3/4 3-4 t

Attcchment N). 3 i Proposed Technical Specification No. 3/4.3.1 l Involve a significant reduction in a margin of safety because the ( c. proposed change in applicable power level does not involve changes in plant design, mode of operation or affect any safety analysis assumption. Based on the above considerations the changes included in the development of proposed Technical Specification 3/4.3.1 are considered not to involve a significant hazards consideration as defined in 10 CFR 50.92. Further, there is reasonable assurance that the health and safety of the public will not be endangered by the proposed changes. i App. A 3/4 3-5 MM/3 4-3.] dp

~ Attachment N3. 4 0 ( / i ...[ \\ 1 1 ) SATEthM,c TITS.AND LIMITING SATETY SYSTEM SETTINGS 2.0,,Y. 2.1,,,*j y).~ SAFETY LIMIT, REACTOR CORE v }'f.Y. ~, ,g. Applicability: Applies to the limiting combinations of thermal power, Reactor Coolant System pressure, average coolant temperat.gr,e r I and flow *during power operation'. Obiective: To maintain fuel cladding integrity. Specification: 1. THR'IE LOOP OPERATION n. :..: - ..c... s - The combination of thermal power level, coolant , pressure and average coolant temperature shall not exceed [ I the limits shown in Figure 2.1-1 for full flow ~ from three reactor coolant pumps. 2. TWO LOOP OPERATION ( i s The combination of thermal power, level, coolant pressure and average coolant temperature shall not exceed l the limits shown in Figure 2.1-2 for full flow from two reactor coolant pumps. 3. ONE LOOP OPEPATION The thermal power level shall not exceed 20%, coolant pressure shall be maintained in the 1820 '2400 psig* range, and the average coolant temperature shall not exceed 590 F for full flow from one reactor coolant pump. ~ 4. NATURAL CIRCULATION The thermal power level shall not ex'ceed 12%, coolant pressure shall be maintained in.the 2135 - 2400 psig range and the average coolant temperature shall not exceed 602 F, when no reactor coolant pumps are in operation. 2.1-1 Amendment Nos. 73 & 67 g.7 ,e-

Attachment N3 4 l b t l IP + tuttCY FOINT UNITS ! 6 4 gg, 655 .,200 650<

5 fe 625<

648 ".

- fe 615<

E50-

G25<

200g Ds te 620

615<

~ 610< IEES Dste (_ EC5 " ECD. e 5.0 < 555< 550 575 O. 1 .2 .5 4 .5 .6 .7 .8 .l 1. 1.1 1.2 PCVER (fraction of nominell l Figure 2.1-1 Reactor Core Ther:.al and Hydraulic Safety Limits. Three Loop Operation i This amendment effective as of date of issuance for Unit 3 and date of startup. Cycle 10 for Unit 4. Amendment Nos. 99and 93

l Attcchme t Np. 4 l H C i. .\\ . 1; l 1 e

)

l l i i l i l + b I 6 ( This Figure it:tentionally. deletec. i i .I 4 I Figure 2.1.la This amendment effective as of date of issuance for Unit 3 and date of startup. Cycle 10, for Unit 4. Amendment Nos. 99 and 93 l9

Attachment No. 4 i i i k a f ~ ( This Figure intentional'1y deletec. i. l l l l i This amendment effective as of date of issuance for Unit 3 and date of startup, Cycle 10. for Unit 4. Figure 2.1 1b Amendment Nos. 99 and 93 20

Attachment No. 4 q 6% l' 2600 P$la 630 2260 PSIA ~ 620 _w ~, g 610 s + isoo rsia d600 1700 Psla 8_:f. 590 f x.I g580 aW l $_, 570 i W .C 560 sso L 17 I l' I I m o 20 e so ao too RATEDPOWER(PERCENT) l Figure 2.1-2. Reactor Core Thermal and Hydraulic lr, Safety Umth, Two Loop Operation 1 I g fo/24/74 O ,..r... ,._.~..---% ,m, . -. _ _.. -. -... ~..,, -.. -.r..

i ) Attt:chnert flo. 4 i 0 DNS PARAMETER 5 1 The following DNB related parameter !!mits shall be maintained during power operation. \\

a. Reactor Coolant System Tavg5578.2 F

]

b. Pressuriser Pressure 32220 pala'

c. Reactor Coolant Flow 3268,500 gpm i

With any of the above parameters exceeding its limit, restore t,he parameter to within its limit within 2 hours or reduce thermal power to less than 5% of rated thermal power using normal shutdown procedures. Compilance with a. and b. Is demonstrated by verifying that each of the parameters is within its limits at least once each 12 hours. ] Comp!!ance with c. Is demonstrated by verifying that the parameter is within its !!mits after each refueling cycle. l i l Ii i

l 1

i

Limit not apolicable during either a THERMAL POWER ramp increase in excess of (5%)

RATED THERMAL POWER per' minute or a THERMAL POWER stap increase in excess of (10%) RATED THERMAL POWER.. This amendment effective as of date of issuance for Unit 3 and date of Start-up, Cycle 10, for Unit is. 2E 3.1-7 Amendment Nos.110 and%

Attechment No. 5 j

2. 0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS

( 2.3 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest l operating loop coolant temperature (T,yg) shall not exceed the limits shown in Figure 2.1-1, for 3 loop operation. APPLICABILITY: MODES I and 2. ACTION: Whenever the point defined by the combination of the highest operating loop average temperature and THERMAL POWER has exceeded the appropriate pressurizer pressure line, be in HOT STANDBY within 1 hour, and comply with the require-ments of Specification 6.7.1. REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig. APPLICABILITY: H0 DES 1, 2, 3, 4, and 5. ACTION: MODES 1 and 2: Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within I hour, and comply with the requirements of Specification 6.7.1. MODES 3, 4 and 5: ( Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within its limit within r l 5 minutes, and comply with the requirements of Specification 6.7.1. i k i l l l TURKEY POINT - UNITS 3 & 4 2-1 AMENDMENT N05. AND TE0 2 E ic:a

AttO?hment No. 5 L7b 6Lb 2400 PStA N N UNACCEPTABLE m '55 ^ OPERATION N 2250 PSIA g\\ i N as s \\ N "5 2000 PSIA N ^Q l q F N N p as N 1825 PSIA N NI N \\ N N \\< ,,3 N N \\ l I N N T N \\ \\ .0, N x x N \\g \\ N 595 \\ \\ i \\ e, 575 0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 1.1 1.2 POWER (FRACTION Of NOMINAL) FIGURE 2.1-1 REACTOR CORE SAFETY UMIT l TURKEY POINT - UNITS 3 & 4

~ .I 3/4.4 REACTOR C0(LANT SYSTEM '/4.4.1 REACTOR C30LANT LOOPS AND COOLANT CIRCULATION STARTUP AND POWER 0.' ERAT!0N LIMITING CONDITION FOR OPERATION 3.4.1.1 All reactor coolant loops shall be in operation. APPLICABILITY: MODES 1 and 2. i-TCTION: With less than the above required reactor coolant loops in operation, be in at least HOT STAN08Y within 6 hours. i SUREVEILLANCE REQUIREMENTS I 4.4.1.1 The above required reactor coolant loops shall be verified in opera-tion and circulating reactor coolant at least once per 12 hours. 1 4 TURKEY POINT - UNITS 3 & 4 3/4 4-1 AMENDMENT N05. AND i muam

Attcchment No. 6 l l 2.2 SATETY LIMIT, REACTOR COOLANT SYS7Di PRESSURE j l l Applicability: Applies to the maximum limit on. Reactor Coolant System pressure. Objective: TomaEntaintheintegrityoftheReactorCoolantSystem. ) Specification, The Reactor Coolant System pressure shall not exceed 2735 psig with fuel assemblies installed.in the reactor vessel. 4 e 1 e 1 m 2.2-1 26

Attechnent N3. 6 CON;7f01.lfp* O'%.;,0k 1.0 CEFIH!T!0NS n, l t, g The definitions used for these specifications follow. l , 1.1 SAFETY LIMITS Safety limits are the necessary quantitative restrictions placed upon those process variables that must be controlled in order to reasonably i protect the integrity of certain of the physical barriers which guard against the uncontrolled release of radioactivity. If ar1y safety limit is exceeded, the associated reactor shall be shut down until the AEC authorizes resumption of operation. 1.2 LIMITING SAFETY SYSTEM SETTINGS Limiting safety system settings are set points for automatic protective l devices responsive to the variables on which safety limits have been' pl aced. These set points are so chosen that automatic protective actions will correct the most severe, anticipated abnormal situation so that a safety limit is not exceeded. 1.3 LIMITING CONDITIONS FOR OPERATION Limiting conditions for operation are those restrictions on reactor operation, resulting from equipment performance capability, that must be enforced to ensure safe operation of the facility. 1.4 OPERABLE - OPERABILITY A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY uhen it is capable of performing its specified function (s). Implicit in this definition shall be the assumption that all necessary attendant instrumentation, controls, normal and emergency electrical power sources, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, f component or device to perfom its function (s) are also capt.ble of performing their relcted support fun: tion (s). krend.nents i g7 v

Attachm nt No. 7 ] g. The Operations Supervisor shall hold a Senior Reactor j Operator License. ( h. The Operations Superintendent shall either hold or have held a Senior Reactor Operator Lleense on the Turkey Point Plant, or have held a Senior Rosetor Operator License on a similar plant (l.a. another pressurized water reactor). 6.3 FACILITY MAFF Q11ALIFICATIONS i 4.3.1 Each member of the faellity staff shall meet or exceed the minimum quellfloations of ANSI N18.1-1971 for comparable positions except for the Shift Technical Advisor who shall have a bachelor's degree or equivalent in a seientifle or engineering discipline with speelfle training la plant design and in the response and analysis of the plant for transients and aceldents, and the Operations Superintendent whose requirement for a Senior Reactor Operator Lleense is as stated in Speelfloation 0.1.2.h. 6.3.2 plealth Physles Sumervlaar Quallflentless 6.3.2.1 The Hestth Physles Supervisor at the time of appointment to the position, shall, except as indiented below, meet.the i following 1. He shall have a bachelor's degree or equivalent in a selence or engineering subject, including some formal training in radiation protection. ( 2. He shall have five years of professional experience in applied radiation protections where a master's degree in a related field is equivalent to one year experience and a doctor's degree in a related fleid is equavelent to two years of experience. 3. Of his five years of esperience, three years shall be in applied radiation protection work in a nuclear faellity dealing with radiologloal problems similar to those encountered at Turkey Point Plant. S.S.S.S When the Health Physles Supervisor does not meet the above requirements, compensatory estion shall be' taken l whleh the Plant Nuclear Safety Committee determines and the NRC Offlee of Nuelear Reactor Regulation concurs that the setion meets the intent of Speelfloation 6.3.2.1. L r l l l l gh 6-5 Amendment Nos. 135 and 129 4 -_,y,, .y. ,,s.,-...c-+ ---w m.,,-*we3

-e=

+-- - - = * * ---"' " - * ' ' * - - " * * - ' - " * * - ' * * ' * * - ' ' ' " * ' " ' ' ~ - - - ~ " * * * ' - " ~ - "

ADMINISTRATIVE CONTROLS F ACTIVITIES (Continued) t b. Individuals responsible for reviews performed in accordance with Specification 6.5.3.1 (a) shall be members of the plant staff previously designated by the Plant Manager-Nuclear and meet or exceed the minimum qualifications of ANSI N18.1-1971, Sections 4.2, 4.3.1, 4.4 and 4.6.1. c. Each review shall include a determination of whether or not addi-tional, cross-disciplinary review is necessary, if deemed necessar such review shall be performed by qualified personnel of the appro y, priate discipline, d. Each review will include a determination of whether or not an unreviewed safety question is involved. 65.3.2 Records of the above activities shall be provided to the' Plant Manager, PNSC, and/or the CNRB as necessary for required reviews. 6.6 REPORTABLE EVENT ACTION 6.6.1 The following actions shall be taken for REPORTABLE EVENTS: The Commission shall be notified and a report submitted pursuant to t a. the requirements of Se: tion 50'.73 to 10 CFR Part 50, and t b. Each REPORTABLE EVENT shall be reviewed by the PNSC, and the results of this review shall be submitted to the CNR8 and the Senior Vice President-Nuclear. 6.7 SAFETY LINIT VIOLATION l 6.7.1 The following actions shall be taken in the event a Safety Limit is l violated: a. In accordance with 10 CFR 50.72 be notified by telephone as soon, the NRC Operations Center, shall as practical and in all cases l within one hour after the violation has been determined. The Senior l Vice President-Nuclear, and the CNR8 shall be notified within 24 hours. l j TURKEY POINT - UNITS 3 & 4 '6-12 AMENDMENT NOS. AND FF A P A 1000

y Attachment NG. 8 1 ADMIN!$7AATIvt CONTROL $ AFITY LIMIT v!0LATION (Continued) b. A Licensee Event Report shall be prepared in accordance with 10 CFR 50.73. The License Event Aeport shall be submitted to the Commission in c. accordance with IC CFR 60.73, and to the CNtB, and the Senior Vice President Nuclear within 30 days af ter discovery of the event, d. Critical operation of the unit shall not be resumed until authorized l by the Nuclear Regulatory Commission, j g.8 PE0CEDURES AND PROGRAMS I 6.8.1 Written procedures sh611 be established, implemented, and maintained covering the activities referenced below The applicable procedures recommended in Appendix A of Regulatory $1 ~~ a. Cuide 1.33, Revision 2. February 1978, Sections 5.1 and 5.3 of AN N18.7 1972; b. The emergency operating procedures required to implement the requirements of NUREG 0737 and Supplement 1 to NUREG 0737 as stated ( in Generic Letter No. 82-33; c. Security Plan implementation; d. Emergency Plan implementation; e. PROCESS CONTROL PROGRAM implementation; l f. OFFSITE DOSE CALCULATION MANUAL implementation; I g. Quality Control Program for effluent monitoring using the guidance in Regulatory Guide 1.21, Revision 1, June 1974; h. Facility Fire Protection Program; and 1. Quality Control Program for environmental monitoring using the guidance in Regulatory Guide 4.1, Revision 1, April 1975. 6.8.2 Each procedure of Specification 6.8.1 (a through h), and changes thereto, shall be reviewed and approved prior to implementation and reviewed periodically as set forth in Specification 6.5.3 and administrative procedures. 6.8.3 Temporary changes to procedures of $pecification 6.4.1 (a through 1) Cay be made provided: a. The intent of the original procedure is not altered; i TURKEY POINT - UNITS 3 & 4 6 13 AMEN 0 MENT N05. AND ] FEB 2 8 Itan

AttO;hment No. 9 i 2.3 t!MITINC SATETT $YSTEM SETTIN05. FkOTICTIVE INSTRUMEh*TATION Arrlicability: Applies to trip settings for instruments monitoring l reacter povert reacter coolant pressure, temperature and fluvi pressuriser levelt and steas generator level. Objective: To prevent the princita) process variables from exceeding a safety limit. e Specificatient Reacter trip settings shall be as follovst l Nuclear Taux 7 ~ Fover range (low set point) equal to er less than 251 of rate,d power, u.ay be bypassed when power is greater l than 101 of rated power. j Power range (high set point) equal to or less than l 1091 of rated power. i l ( f i l 2.3 -1 III l 5

Att:chment Co. 9 l I f J,, TACTOR C00LwT Trot.O n'ti l ~ C,varte.rperature 4 1,1 47, tKg - 0. 010 7 (7 3 74 ) + 0 000 & S 3 (? -2135 ) - ! (,, g)) Ittisated 4; at rated ptver. T 7-Average to=7erature. T F-Pressuriser pressure, psig l t !(.ig)= a futetics cf the 1 dt'tated di!!trette betvets tsp 4:4. bettec I detesters of the pover-rasga sudear tot chasters; vith gair.a to be sa.lected hased es measured is.strutant respor.se tuttag startw l ' tests svth that: i r I Tor (qt - 9b) vitkia + 10 partest and -14 partent store tt a tt t, i are the percent power la the top and bettcu halves of the tera l respectively, and tt + 9b is totti sete power in percett ci rated power, f (A g) = 0. ) t Tor each pertest that the segnitu.de cf (q - gg) e.xceed + 10 t h e De.h t s-7 trip s e t p o it t s h a.!L b,e a v t cu t ic a.11y r e t s t e d per:est, by 3.5 percent of its value at irtaram power. Tor eat.h percent that the magnitutt et (gg - (b) a.xtends 4 4 { p e r : e r.t. the telta-7 trip se tpoist.Dh:11 te astvr.atica.11y redvt e! by 2 perter,t c! its valve at ir.tt:t: pcVer. ( K (**r.r e e I.c ep Op e r a tio n) = 1. 095 t l (;vo keep Ciaration)

0.48 i

l j l 1 1 i i i f i i i i i This amendment effective as of date of issuance for Unit 3 and date of startup. Cycle 10 for Unit 4. 2 3-2 Mendment Nos. 99 and 93

Attachment No. 9 Ov er; c wc r 4 ; igt, 1+09 *K1 -K2 (T

I' ) "'I (O'4) 4 7,.

Ir. cit a t ed T at ratti ;:ver. T Ave r ag e t t.g e r a tur a, T = T* !stitsta! averags ts:; eratura at ac-tr.:.1 staditle:s ar.d = rated power. T E 0 f or detransiss r.*erage tetyerature; 0 2 sec.h* ter = g attransing,sverage staperatura L = 0.00068 f o r ; e gu.a.1 t o o r as t e tha e 7 * ; O t e r T les s tha.m M=

p. ate c! chatst c! t ag a rature T/ sac it

! (a g) =A s (e.!i:st ab eve. Jressurt:er L: V 7:essvrtter ;rassure - eta.at tr er greater that 1435 psig X;gh Pr e s t writ e r p res s wr e - e t, s:.1.t.t cr less :P. r 2333 pr:g. ( i Kj g r. 7 t ts s a ur is t r va t t r i t'. a:, - e wat to o r. n s s tr.ar. S d' s c a.. e.

! t ual

+ 1 J.es:ter1:etattT{ev L:V rea ter ccc14:t fiev - et,va' tr.t1:a t a.t !! v. . t o e r g r e a : t r t r.ar. I C T' ! :: : a ' l L v reatttr teolant pump ::t:r f ragver. y that 361 Es. es.41 tr or 3: ester C:darveltage en reatter ecolar.: greater than 60: et sortal v ltage.;u=7 star h s - t;ual to er Jt ems Camerato rs tev-lev steam generate r vatar level - egual to or greater than 13% c! narrev range instruze t scale. l This amendment effective as of date of issuance for Unit 3 and date Cycle 10. for Unit 4 1.5-3 knendment Nos. 9% id o'4

AttC:hne;t No. 9 .,e Feactor Trit Interlocks ] ] / protective instrucentation settings for reactor trip interlocks i shall be as follows: 1. Above 10% of rited power, the lov pressuriser pressure trip, high pressuriser level trip and the low reactor coolant flow trip (for two or more loops) are made functional, t i 2. Above 45% of rated power tt.e single loop loss of flow trip is made functional. It need not be made functional o belov 60% of rated pover for two loop operation if the l I overtemperature 47 trip setpoint has been adjusted using i K3 = 0.66. I e I 9 L 6 e I i 2.3-4 3l4" i i

Attschae:t N3. 10 l l $AFFTY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS f i i I I j.! LIMITING SAFETY 5YSTEM $[TTING,5 i l ,RJ; ACTOR 1 RIP $Y$T[M INSTRUMENTATION $(TP0fMT$ j 2.2.1 The Reactor Trip System Instrumentation and Interlock Setpoints shall I' be set consistent with the Trip Setpoint values shown in Table 2.2+1. l l APPLI,CASILITY: As shown for each channel in Table 3.3 1. ACTION: a. With a Reactor Trip System Instrumentation or Interlock Setpoint less conservative than the value shown in the Trip 5etpoint column but more conservative than the value shown in the Allowable Value I column of Table 2.2 1, adjust the setpoint consistent with the Trip setpoint value. b. With the Reactor Trip System Instrumentation or Interlock Setpoint i less conservative than the value shown in the Allowable Values l column of Table 2.2-1, declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1 until l the channel is restored to OPERA 8LE status with its setpoint adjusted l consistent with the Trip Setpoint value. ( I i 1 TURKEY POINT - UNITS 3 & 4 2-3 AMEN 0 MENT N05. AND t-*n ' b I.)

t i ^ Attach =Get No. 13 ^ i TABLE 2.2-1 t a E REACTOR TRIP SYSTEM INSTEINENTATION TRIf SETPOINTS l A 1 o FUNCTIONAL ISIIT TRIP SETPOINT ALLOlmSLE VALUE # 5 1. Manual Reacter Trip N.A. N.A. c 2. Power Range. Heutron Flux l l a. Nigh Setpoint 1109E ef RTP** <[ ]E of RTP** J i \\ w b. Low setpoint 125% of RTP** 1[ ]E of RTP** w e-3. Intermediate Range, $2sz of RTP** $[ ]E of Bir** l Neutron Flest l l 4. Searce Range. Heutron Flest 118 cys $[ ] x 10 cys l 5 5 S. Overtemperature AT See IInte 1 l 7 6. L.'c. aT See Ilote 3 I 7. Pressurizer Pressure-Law >1835 psig >[ ] psig t l;

s. Pressurizer Pressure-mish

~ 123e5 psis $[ ] psig 9. Pressurizer water Level-nigh 1925 et lastrument span $[ ]E er instrument span j

10. Reacter Coolant Flene-Lew

>98K of leap >[ ]E of leap l k 3esign flow 3esign flow

l a

1 5 'l R

11. Steam Generator Water

>1SK of narrow range >[ ]E of narrow range lastrument 5 Level Low-Lew Testrument span span 8 i i 4 y y5 t m U " Leap design flow = 89,500 gum

- RTP = RATED TIENI4L POWER O

M, 36 l l .-__..__---m. , _ ~ _ _ _ _ _ _ _ _ _ _ _ _ -

Attachment No. 1] m TABLE 2.2-1 (Continued) k REACTOR TRIP SYSTEM INSTRUNENTATION TRIP SETPOINTS l A ] FUNCTIONAL UNIT TRIP SETPOINT Att0WABLE VALUE #

12. Steam /Feedwater Flow feed Flow Feed Flow Mismatch

<0.64 x 108 lb/hr <[ ] x IOS ib/hr i Coincident with Selow steam flow Selow steam flow c-5g Steam Generator Water >15% of narrow 1[ JE of narrow range instrument Level-tow range instrument span span w b

13. Undervoltage - 4.16 kV

>24% volts- >[ ] volts-Busses A and B Each bus each bus

14. Underfrequency - Trip of Reactor

->56.1 Hz ->[ ] Hz Coolant Pump Breaker (s) Open 7

15. Turbine Trip w

a. Auto Stop Oil Pressure >45 psig >[ ] psig b. Turbine Stop Valve Fully Closed *** Fully Closed *** Closure

16. Safety Injection laput N. A.

N.A. from ESF g 5 17. Reactor Trip System 5 Interlocks 5 2 a. Intermedfete Range >l x 10 88 amp >[ ] amp Neutron Flux, P-6 o E O ac4 """ Limit switch is set when Turbine Stop Valves are fully closed. o 37

I l 1 ,1 ( l0 i 2 ~ \\. l1, J. l L t i L i I E E E -s!s s N RR R R l e s s ss i l 'l 5 W vi Vi Vi Yi d

== I 1 l >= t i l .g i -n i . E m l 5 E i l E e _g [ [ I ( t f i $ j "s l s a v- ~ d b i g g' g g .N i l I E E m m er m vi vi vi Al E E E l [ >g t E i i l 5 ~ a h b i = y l 1! 1 1-i i ? 'W Ic 1 si}4 Je t! l," L &A W i S j$ L . bb 5 l 2 & s' wC u 2 E g* 5 5 s g5-{3j3 5 s. ~ j 4 u 4 b& ~ ] y E R ma TURKEY POINT - UNITS 3 & 4 2-6 AMENDMENT N05. AND IEB 3 81;h 1 ,.__.,.,..,_.,-m.- ... _., _... _ _... ~,. _., _.. _,..,, _.. - _ _. _ _. _. ,.m.-

Attachment We. 13 TASTE 2... (Continued) f { TABLE NBIATIONS l c i !ll2 NOTE 1: OVERTEfrERATURE AT i b I II * '2SI [T ( I AT ( ) < AT ) - T']

K (P - P') - f (&I)l

~ l [K -K s 2 (1 + ta ) 1+T5 8 1+t5 e s 4 c i 5 lesere: AT Measured ai by RTS Instrimentation; i = j i l 1 tag compensatw en measome 4T: = 1 ,,3 e-j Tlee constants utilized in the las compensater for ATJg i 1 = %+Ribhp oc (e c 2.5 sf)g AT, Indicated AT at RATED TIEment peter; = 1.995; l K = jwj 2 e.s187/*F; j y K = u I * '*f = dynamic compensation; ~ l The funct.len generated by the lead-lag compensater for T g '8 8'8 i i ts. is Time constants attilzed in the leed-lag compensator for T ,1 = 25s, = 13 = 3 s; i Average temperature. *F; j T = y I G tag c ':xde en esasured Taug; = l I + t,5 o i A i x [ t, Tlee constant. utilized in the meesered T 1as c,1 :dw.) = 0 a = i o" T' < 574.2*F (Neelmal T at RATES TIEment POWER); g,tn p % =2.5s-J 0.000453/psig; K = i 5 P i rressurizer press re, psig; r = w r i to's i i l a 39 2 - -- .-- a

Attachment Je. 16 - m TABLE 2.2-1 (Continued) -e y, TABLE NOTATIONS (Continuedy A NOTE 1: (Continued) 2235 psig (Nominal RCS operating pressure); g P' = Laplace transfers operator, s 8; 5 = E U and f (AI) is a function of the indicated difference between top and betten detectors of the power-range neutron len chambers; with gains to be selected based en messered instrument response during plant startup tests such that: e-(1) Fer'g g between - 14% and + 1 5 f (AI) = 0, where g and g are percent RATES TIE WWE j reWER in the top and bottaa halves of the core respectively, and g + g is total THEWWE PauEn in percent of antes THEmot reWER; l (2) For each percent that the magnitude of g g exceeds - 16, the AT Trip Setpoint shall Y he automatically reduced by 2.gE of its value at RATED TIEWWE POWER; and 4 as For each percent that the magnitude of g g enceeds + 15, the AT Trip 5etseint stEll j (3) he automatically reduced by 3.SE of its value at RATED TIEmmt peter. NOTE 2: (This note W is not used.) ~ i T E i E i ~ 5 .= g E! l C 1 ~ l h I ._ -_-.,. _-.~_,-_ -,_-_ -.._-_-_._.._..

2 Attachusent No. 13 j TAetE 2.z-1 (Continued) i E TABLE NOTATIONS (Continued) l M i NOTE 3: OVERPeiER AT x 7 AT ( 1 ) < 37 gg _g ( tss ) ( 1 )T-K [T I - T"] - f (AI)! I I i g (1 + t:5) - e 5 (1 + 1 5) (1 + Y,5) (1 + t 5) 5 U. v i As defined in Note 1 w namere: AT = i e-I i 1+1 5 As defined in Note,1, = As defined in Note 1 1 = As defined in Note 1 AT, = ] 3 K. = 1.09 o.e2 M for lacreasing average temperature and e for decreasing average Es = temperature, The function generated by the rate-lag compensator for T dynamic = 3 5 compensation, i Tlee constants utlitred in Line rate-lag compensator for Tag, is = le s, =

g ts h

^* #

I" "I' I* i I + 1,5 m i "i l z 14 As defined in Ilote 1, = l 8 i i e i E o 4 m D tt? t 4% ii 9 41

Attachment No.13 - m TAttE 2.2-1 (Continued) --e ]m TABLE NOTATIONS (Continued)' Q y NOTE 3: (Continued) z { 0.00068/*F for T > T" and K. = 0 for T < T*, K. = E T As detined in Note 1 = 4 i } T" Indicated T at RATED TEWIAL POER (Calibration temperature for AT l = w avs e. Instrumentation, < 574.2*F), I i r S = As defined in Mete 1, and l f (AI) As defined in Note 1. = ~ NOTE 4: (This note ma=her is not used.) b t i i I i R i 3 E E5 I E f If no alleuable valve is specified as indicated by [ ], the trip set point shall aise be the alleuable value. l 5 l o 4 m to [ w ] on i

,7 4!

O ..~.,,...,,.----.---.--..---.-,-.,-,..,..-,.~,-------__-,_~__-..,_.--____-.b

. _ _ ~ 3.2 'CONff.01. ROD AND FCrJER DISTR!iVTION L1 HITS Attcchoe:t No. 11 I 1 Applicability Applies to the operation of the contrv1 rods and power-distribution limits. ) j r i j Objeettve To enture (1) core subcriticality siter a reactor trip, f . (2) a liedt on potential reactivity insertions from a hypo l j rhetical control red ejection, an'd (3) an accept,able enn i povar distributf.an during power operation. 4 J Specification: 1. CONTROL ROD IN$tRTION LIMIT $ 1

a. *Whenever the reacter is critical, except for physics i

tests and control red exercises, the shutdodn control. l rods shall be fully withdraws. b l .b. For Unit 4, whenever the reactor is critical, l l except for physics tests and control rod exercises,- l 1 the control group rods shall be no further inserted [ J than the lisiits shown on Figure 3.2-1 for three y loop operation and on Tigure' 3.2-1(a) for two loop I ] operation. 1 l l l c. For Unit 3, whenever,the reactor is critical, ex- [ cept f or physics tests and control rod exercises. [ the control group rods shall be no further in. sorted than the 2.imits shown on Figure 3.2-1(b)

i for three loop operation and on Figure 3.2-1(c) l for two loop operation.

d. The Unit 4 control rod instreion limits shown on Figure 3.2-1 and the Unit 3 control rod insertion l limits shown, on Figure 3.2-1(b) may be revised on the basis of physics calculations and physics data obtained during startup and subsequent operation.

part length rods shall not be permitted in the core e.

except for low power physics tests and.for axial of fset l calibration tests performed below 75% of rated power. . Any reference to part length was no longer applies after the part length rods are remove.d from the reactor. 43 .s*2 1 9/21/78 i

Attechme2t W3. Il a j i 1 l Except for low pewse physics setts, the t f. shutdewn serpin with allevante for a stuck contre ed shall esteed the i applicable valve shown en Figure 3.2-1 i 1 I under all steady-stste operating codi-tiens free sere to fv11 power, hecluding l effects of asial power distribution. l i The shutdown margin as used here is } 1 l defined as the asevnt by which the reat-j ter core wvid be soberitical at het i s,huttern conditions (540'F) if all een-t tre) rHs were tripped, assumieg that } the highest wrth centrol red remained fv11y withdrawn, and assuming ne changes in senon, beren concentration er part-l j 1ength red position. i During physics tests 'and teatrol red g. eserc'ses, the insertion limits ned not be set, but the retviced shutdown ur-gin, Figure 3.1-2 must be maintained er esteeded. l M!& ALIGNED CONTROL R00 2. l If a part lensth' or foil length control red i { (- is more than 12 steps out of alignment with l i it's bank, and is not corrected wtthin 8 i* hours, power shall lie reduced se as not to. l qsceed 715 of interie power for 3 leep er 45% or interte power for two leep operation, unless the het thannel facters are shown to l be no greater than allowed by Section 64 of Seecification 3.2 l 400 Otop TTME 3. Tse drop t<en of each control red shall be t i no greater than 2.4 seconds at full flew and l l operating temperature free the beginning of. red motion to dashpet entry. 4. INDpfRABLE CONTROL 200$ l No more than one inoperable control red 4. Shall be permitted during sustained l l power operation, except St shall not be permitted if the red has a potential l j l I ' Any reference to part-length rods no longer applies after the part 1ength rods are removed from the reactor. l This amendment effective as of date of issuance for Unit 3 and date of l startup. Cycle 10. Unit 4 Amendment Nos. 98 and 92 3.1-1 i .,. _ _ _ _. _.. -.... ~...

AttaehmeQt No. 12 REActivtrY CONTROL SYSTEMS r j R00 DROP TIME ( j LIMIT!NG CONDITION FOR OPERATION ) ~ l l 3.1.3.4 The individual full length (shuteown and control) rod drop time from l l the fully withdrawn position sha:) be less than er equal to 2.4 secon6s from ~ j beginning of decay of itationary gripper coil voltage to dashpot entry with: l l T,yg greater than or equal to 541'F, and a. l b. All reactor coolant pumps operating, j l l APPLICABILITY: MODES 1 and 2. l ACTION: !~ Rth the drop time of any full-length rod determined to exceed the above limit, restore the rod drop time to within the above limit prior to proceeding l to MODE 1 or 2. { i i SURVE!LLANCE REQUIREMENTS l 4.1.3.4 The rod drop time of full-length rods shall be demonstrated through measurement prior to reactor criticality: a. For all rods following each removal of the reactor vessel head, j b. For specifically affected individual rods following any maintenance on or modification to the Control Rod Drive System which could affect the drop time of those specific rods, and c. At least once per 14 months. 4 I I E P l TURKEY POINT - UNITS 3 & 4 3/4 1 24 AMENDMENT N05. AND 2

Attscheest N>. 13 i i nE ACTOR IMTERIAL SURVfJLLANCE PROGRAM 4l 6,30.1 The following lerediation Specimer Schedule shall be fellowed: l ) CAPSULE RDADVAL SCHEDULA t M M M i j Y ) 12 years i t V* 4 24 years X 3 )) years j ( X 4 Standby i 1 Capsules U, v. Y, and I for Units 3 and 4 are held in standby. 4.10.1 The above surveillance obsti be conducted using the Tenslie [nd Charey V Notch Test. r l t i i i i l I t i i l ( I i I I e l } i b a 1 i ) i i l i 1 i t i I f I I l l e 112 106 j 4.20 1 Amendment Nos. and l + l I

ittCthee:t N>. 14 RlACTOR COOLANT $YSTEM j i ( 3/4.4.9 PRES $URE/T[MPERATURE LIMITS RfACTOR COOLANT SYSTEM i LIMITING CONDITION FOR OPERATION i b 3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and l pressure shall be limited in accordance with the limit lines shown on Figures 3.4-2 and 3.4 3 and 3.4 4 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with: ? a. A maximum heatup of 100'F in any 1-hour period, l b. A maximum cooldown of 100'F in any 1 hour period, and c. A maximum temperature change of less than or equal to b'F in any 1 hour period during inservice hydrostatic and leak testing opera-tions above the heatup and cooldown limit curves. APPLICABILITY: At all times. ACTION: ( With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes perform an engineering evaluation to determine the effects of the out-of-1}mit condition on the structural integrity i of the Reactor Coolant System; determine that the Reactor Coolant System remains i acceptable for continued operation or be in at least HOT STAND 8Y within the next and pressure to less than 200'F and 500 psig, 6hoursandreducetheRCST,yhing30 hours respectively, within the follo SURVEILLANCE REQUIREMENTS 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be I 1 determined to be within the limits at least once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operations. 4.4.9.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, as i required by 10 CFR Part 50. Appendix H in accordance with the schedule in i Table 4.4-5. The results of these examinations shall be used to update Figures 3.4-2, 3.4 3 and 3.4-4. i i l TURKEY POINT - UNITS 3 & 4 3/4 4-30 AMENDMENT NOS. AND l MAY 0 $ 1983

Attachment C3. 14 fI I l TAgtE 4 435 f( REACTOR VESSEL MATERIAL. SVRVilLLANCE PROGR$ - WITHDRAWAL SCHEDUQ ) I UNIT 3 4 CAPSULE VE5SEL LEAD NUMBER LOCATION FACTOR WITHDRAWAL TIME l U 30' O.49 Steney $pecimen withdrawn f Y at 12 years j 1 W 40' O.34 5tandby l 1 X 50' O.34 33 years Y 150' O.49 Staney { Z 230' O.34 Staney i i 1 ) UNIT 4 CAPSVLE VESSEL LEAD l NUMBER LOCATION FACTOR WITH0RAWAL TIME ( V 30' O.49 $taney J i i V 290' O.79 24 years W 40' O.34 Standby X 50' O.34 Standby Y 150' O.49 Standby Z 230' O.34 Standby i i t l 1 r TURKEY POINT - UNITS 3 & 4 3/4 4 34 AMENDMENT N05. AND I wu o s na i

Attcchment Wa. 15 3.1 2E AC70R COOL ANT 5Y$ TEM i i Aoolleability: Applies to the operating status of the Reactor Coolant System. i Objective: To speelfy those limiting conditions for operation of the Acactor Coolant System which must be met to assure safe reactor operatlen, j i Soeelfication: 1. OPERATION A1, COMPONE NTS l i

a. Reactor Coolant Pumas j

!. A minimum of ONE pump shall be in operation when the reactor is in power operation, except during low power physics tests. [

7. A minimum of ONE pump, er ONE Residual Heat Removal Pump, shall be in operation during reactor coolant boron f

concentration reduction. I

3. Reactor power shall not exceed 10% of rated power unless at

) least TWO reactor coolant pumps are in operation. I r 4 Reactor power shall not exceed 43% of rated power with only two pumps in operation unless the overtemperature AT trip j setpoint, Kg, f or two loop operation, has been set at 0.88.

3. A reactor coolant pump shall not be started when cold leg i

i temperature is $27)*F unless steam generator secondary j water temperatura is less than 50cF above the RCS temperature (including instrument error).

b. Steam Generators

( l

1. A minimum of TWO steam generators shall be operable when t

( the average coolant temperature is above 350*F. l l l 1 3.1 -1 Amendment Nos.110 and 104 49 l

Attcchment No. 15 l l

c. Pressurlace Safety Valves,

i

1. ONE valve sha!! be operable whenever the head is on the reactor vessel except during hydrostatic test.

l i

2. THREE valves shall be operable when the reactor coolant l

average temperature is above 3500F or tne reactor is critical, j i

d. Pressuriner

{ The pressurizer shall be operable with a steam bubble, and with at f least 123 KT of pressurlser heaters capable of being supplied by l emergency power, when the reactor coolant is heated above 3300F. I

e. Relief Valves I

1. A power operated re!!ef valve (PORW and its associated block i

valve shall be operable when the reactor coolant is heated l t above 3300F.

2. If the average coolant temperature is greater than 3500F and

{ the conditions of 3.1.1.e.1 cannot be met because one or more PORY(s) is inoperable, within 1 hour elther restore the l PORY(s) to operable status or close the associated block valve (s) and remove power from the block valve (s) otherwise, be in a condition with Keff < 0.99 within the next 6 hours and In COLD $HUTDOWN within the following 30 hours.

3. If the average coolant temperature is greater than 3500F anJ the conditions of 3.1.1.e.1 cannot be met because one or more

( block valve (s) is inoperable, within I hour either restore the block valve (s) to operable status or close the block valve (s) and [ remove power from the block valve (s): otherwise, be in a condition with Keff < 0.99 within the next 6 hours and in COLD 5HUTDOWN within the following 30 hours. I i l I l 3.1 -l a Amendment Nos.110 and 104 50 l

Attechme:t W). 15

f. Reactor Coolant $ystem Vents

l. At least one reactor coolant system vent path consisting of at j

least two valves in series powered from emergency busses j shall be OPF.RABLE and closed at each of the following i i locations when Tavg s greater than 2000F: i i s. Reactor Yessel Head b. Pressurlser steam space j i

2. With one of the above reactor coolant system vent paths inoperable, STARTUP and/or POWER OPERATION may 1

continue provided the inoperable vent path is maintained l 1 closed with power removed from the valve actuator of all the i valves in the Inoperable vent paths restore the inoperable vent l path to OPERABLE status within 30 days, or, be in HOT SHUTDOWN within 6 hours and in COLD $HUTDOWN within the following 30 hours. t I

3. At power Operation. With both reactor coolant system vent paths inoperable, maintain the inoperable vent paths closed j

with power removed from the valve actuators of all the valves In the Inoperable vent paths, and restore at least one of the vent paths to OPERABLE status within 72 hours or be in HOT SHUTDOWN within 6 hours and in COLD SHUTDOWN within j the f ollowing 30 hours. j l

4. At Suberltleal Conditions and Tava > 2000F. With both reactor i

coolant system vent paths inoperable, maintain the Inoperable l vent paths ' closed with power removed from the valve actuators of aLI the valves in the Inoperable vent paths, and i restore at least one of the vent paths to OPERABLE status l within 72 hours or be in COLD SHUTDOWN within the. following 30 hours. i i I g 3.1-lb Amendment Nos. and

Attcchme;t N). 15 s. i 3.4 GCINEtkED SATETY TT.ATL*RES j i deptfeebilityi Applies to the operating status of the En51neered safety l restures. @ ective To define those limiting :enditions (or operation that i are necessary: (1) to ramove decay heat (rom the core l in emergency or normal shutdown situations. (2) to re-move heat from containeent in normal operating and [ emergency situations, and (3) to remove airborne iodine f rom the containeent atmosphere in the event of a Maximum Hypothetical Accident. Seeeif!caeioni 1. SATETY INJECT 10N AND F.E$! DUAL NEAT RDf0 VAL SYSTDf5 ,.r,;8.M a. The reactor shall n Iref tical, except for low power h unless the following condition mets i ( 3 l 1. The refueling water tank shall contain not less i than 320,000 gal. of water with a boron con-centration of at least 1950 ppe. 2. The boron injection tank shall contain not less than 900 gal, of a 20,000 to 22,500 ppm boron i solution. The* solution in the tank, and in isolated portions of the inlet and outlet piping, shall be maintained at a temperat'ure of'at lesst 145F. TWO channels of heat tracing ahall be operable for the flow path.* l l I 3. Each accumulator shall be pressurised to et 3 least 600 pois and contain 475-891 f t of water with a boren concentration of at least 1950 rpm, and shall not be isolated. i ) 4 TOUR safety injection pumps shall be operable. i i

See reference (11) on Page 33.4-2 l

i 3.4-1 Amendments 78 1 72 I i

_, [ Attechtent N3. 1$ i I J, I

3. TWO residual hast removal pumps shall be oporable.

6. TWO residual heat eschangers shall be oporable.

i

7. All valves, interlocks and elping asseclated with the above j

components and required for post aceldent operation, shall be I operable escept valves that are posittened and locked. Yalves j 462 A'and S 86bA and 81864-A and S 843 A. 8, and Ci and l 866-A and 8 shall have power removed from their meter operaters by locking open the circuit breakers at the Motor Control Centers. Two air supply te valve 738 shall be shut of f to the valve operater. I

b. Duriry; d to. allow one of the following components to bep modif,e i

lneperable (includleg associated valves and pipt at any one time j If t system is not except for the cases stated in A4.l.b.2. restored to meet the requirements of Lt.la within the, time i I , period specified, the reacter shall be placed in the het shutdtwn j condition. If the requirements of A4.la are not satistled within j an additional 48 hours, the reacter shall be placed in the cold shutdown condition. Specification A0.1 app!!as to 1.4.1.b. 4 1 h i

1. ONE accum61 ster may be out of service for a period of up to 4 hours.

L ONE of FOUR safety infection pumps may be out of service I ( for 30 days. A second safety injection pump may be out of service, provided the pump is restored to operable status l within 24 hours. TWO of the POUR safety injectlen pumps l shall be tested to demonstrate operab!!!ty before irtitiating i l maintenance of the Inoperable pumps.

3. ONE channel of heat tracing on the flow path may be out of f

service for 24 hours.' j

4. ONE residust heat removal pump may be out of. service, j

provided the pump is restored to operable status within 24 wurs. In add, tion the other residual h6at re neval pump shall i be tested to demonstrate operability prior to initiating j maintenance of the inoperable pump. \\ i i j i i i i i

See reference (11) on page B.3.4 2 i

i I .53 ) 'l 1.a-1 Amendment Nos.101 and 91

j Attachment No. 15 'I i

3. oNE residual heat enchanger may be out of service for a period of 24 hours.

6. Any valve in the system may be' inoperable provided repairs are completed witMn 24 hours.

Prior to initiating i maintenance, all valves that provide the duplicate function shall be tested to demonstrate operability. i

7. To permit temporary operation of the valve, e.a.

for i surveillance of valve operability, for the purpose of, valve maintenance, etc., the valves specified in 3.4.l.a.7 may be i . unlocked and may have supp!!ed air or electric power restored I for a period not to escoed 24 hours. J l i

c. During power. operation three Reacter Coolant Loops shall bo in i

operatten. f y l

1. With less than three Reactor Coolant Loops in operation; the reacter must be in hot shutdown witMn one hour.

i

d. In het shutdown at least two Reactor Coolant Loops shall be operable and at least one Reactor Coolant Loop shall be in operation.'

j

1. With less than two Reactor Coolant Loops operable, restore the required Coolant Loops to operable status witMn 72 hours r

( or reduce Tavs to less than or equal to 3M F witMn the next 12 hours. l

2. With no Reactor Coolant Loop in operation, suspend all i

operations involving a reduction in boren concentration of the l Reactor Coolant System and immediately initiate corrective j j action to return the required Coolant Loop to operation,

e. With average coolant temperature less than 3M P, at least two I

Coolant Loops shall be operable or immediate corrective action l must be taken to return two Coolant Loops to operable is soon as possible. One of these Coolant Loops shall be in operation.' j

1. With no Coolant Loop in operation, suspend all operations Involving a reduction in boron concentration of the Reactor i

j Coolant System and immediately initiate corrective action to i return the required Coolant Loop to operation. l I l l

All reactor coolant pumps and residual heat removal pumps may be de-energised for up to i

I hour provided 1) no operations are permitted that would cause dilution of the reactor l coolant system bdron concentration, and 2) core outlet temperature is maintained as last 10 F below saturation temperature. i l t 5+ ... s... u.. i n i..a = = l

ff Attcchmer.t H). 16 3/4.4 REACTOR COOLANT $Y$7EM ) ( '/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION j i STARTUP AND POWER OPERATION LIMITING CONDITION FOR OPERATION i l 3.4.1.1 All reactor coolant loops shall be in operation. i APPLICABILITY: MODES I and 2. \\CTION With less than the above required reactor coolant loops i'n operation, be in f at least HOT STANDBY within 6 hours. [ i i f i l SUREVElt. LANCE REQUIREMENTS 4.4.1.1 The above required reactor coolant loops shall be verified in opera-l tion and circulating reactor coolant at least once per 12 hours. l i 4 1 1 P 55 j TURKEY POINT - UNITS 3 & 4 3/4 4-1 AMENDMENT N05. AND FEB 181989 l

Attcchsent N>. M 3.1 RE ACTOR COOL ANT SYSTEM I Apolleability: App!!es to the operating status of the Reactor Coolant System. Objective: To speelfy those limiting conditions for operation of the Reactor Coolant System which must be met to assure safe reactor operation. Soeelfleetion: 1. OPERATIONAL COMPONENT 5

a. Reactor Coolant Pumas t

1. A minimum of ONE pump shall be in operation when the f

reactor is in power operation, except during low power physics tests.

7. A minimum of ONE pump, or ONE Residual Heat Removal Pump, shall be in operation during reactor coolant boron concentration reduction.

(

3. Reactor power shall not exceed 10% of rated power unless at least TWO reactor coolant pumps are in operation.

4. Reactor power shall not exceed 43% of rated power with only two pumps in operation unless the overtemperature AT trip i setpoint, K[, for two loop operation, has been set at 0.88. S. A reactor coolant pump shall not be started when cold leg temperature is $2730F unless steam generator secondary water temperature is less than 300F above the RCS I temperature (including instrument error). i

b. Steam Generators i

l

l. A minimum of TWO steam generators shall be operable when the average coolant temperature is above 3500F.

j t 3.11 Amendment Nos.110 and 104 4

Attcchme t N3. 17 l

c. Prestwirer $afety Valves,

(.

1. ONE valve shall be operable whenever the head is on the I

reactor vessel escept during hydrostatic test.

2. THREE valves shall be operable when the reactor coolant l

averag'e te'nperature is above 3300F or the reactor is critical.

d. Presswirer a

l The pressurtzer shall be operable with a steam bubble, and with at least 12) KW of pressuriser heaters capable of being supplied by f emergency power, when the reactor coolant !: heated above 330cF. I

e. Reflef Valves

l. A power operated relief valve (PORY) and its associated bloc.k valve shall be operable when the reactor coolant is heated I

above 3500F.

2. If the average coolant temperature is greater than 3300F and the conditions of 3.1.1.e.1 cannot be met because one or more

} PORY(s) is Inoperable, within I hour either restore the l PORY(s) to operable status or close the associated block l valve (s) and remove power from the block valve (s): otherwise, ) be in a condition with Keff < 0.99 within the next 6 hours and f In COLD 5HIJTDOWN within the following 30 hours. l I j

3. If the average coolant temperature, is greater than 3300F and j

I the conditions of 3.1.1.e.1 cannot be met because one or more block valve (s) is inoperable, within I hour either restore the [ block valve (s) to operable status or close the block valve (s) and [ i remove power from the block valve (sis otherwise, be in a d condition with Keff < 0.99 within the next 6 hours and in COLD f l SHUTDOWN within the following 30 hours. I I i I 3.1-l a Amendment Nos.110 and 104 G7 i _-,-,-----,nne .-w ,-w.e

/, t Attachment No. 17 82.2 SASES FOR SATETY 1.IM17. REACTOR C001AVT SYSTEM PRkSSURE The maximum transient pressure allowable in the reactor vessel under the ASMI Code. Section III is 110% of design pressure. The maximum transient pressure allowable in the Reacter Coolant System piping' valves and fittings under USAS Section 831.1 is 120% of design pressure. System design pressure is 2485 pois.

The settings of the power-operated relief valves (2335 psig),

the reactor high pressure trip (2385 peig) and the safety valves (2485 psig) have been established to prevent overpressure. 4 Valve set pressure tolerances shall be those stated in applicable codes. \\ L l i l I B2.2 1 i

Attechtsnt No. 18 REACTOR COOLANT SYSTEM 3/4.4.4 RELIEF VALVES I LIMITING CONDITION FOR OPERATION ~ ( Each power-operated relief valve (PfRVf) b ock valve shall'be OPERABLE.

3.4.4 APPLICABILITY

MODES 1, 2 and 3. ACTION: a. With one or more block valve (s) inoperable, within I hour either restore the block valve (s) to OPERABLE status, or close the block valve (s) and remove power from the block valve (s); otherwise, be in at least H0T STANOBY within the next 6 hours and HOT SHUTDOWN within the following 6 hours. b. The provisions of Specification 3.0.4 are not applicable. ( SURVEILLANCE REQUIREMENTS 4.4.4 Each block valve shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one cemplete cycle of full travel unless the block valve is closed with power removed in order to meet the requirements of Specification 3.4.4 or is closed to provide an isolation i function. 1 e si -l 9 TURKEY POINT - UNITS 3 & 4 3/4 4-10 AMENDMENT NOS. AND 59 FEB 2 81989 ...-.w- ,.-,n-

J Attcchment No. 19 l

3,15 CVERPRESSURE MIT! GATING $YSTct l

Applicability Establishes operating limitations to assure that the limits of 10 CrR 50, Appendix G, are not exceeded. ( ( Ob!ectives: To sininise the possiblity of an overpressure transient which l could exceed the limits of 10 CTR 50 Appendix C. l I specification: 1. At RCS temperature less than er equal to 380'T and with RCS pressure boundary integrity, valves HOV-*- 443A, MOV *-8435 and MOV *-869 shall be closed and their breakers racked out. 2. If any of the valves listed in 3.15.1 are found to be open when required to be closed by 3.15.1, perform at least one of the following within d.e next 8 houras l l

a. block the corresponding flow path to the reactor vessel, or c

b close the valwow -

c. depressuriae and ven,t' t'he*RCS,through an opening with an area of at least 2.20 square inches, or.

d. verify at least one pressuriser power operated ralief valve is saintained open.

3. At RCS temperature less than or equal to 275'T with RCS pressure boundary integrity established, two pro 6suriser power operated relief valves shall be operable with a setpoint of '415 psig i 15 ps1. ( ( ' ('

a. If one power operated relief valve required by i-3.35.3 is inoperable, perform at least one of the

'I following within 7 days: (1) restore operability of the power operated relief valve, or (2) depressurise and vent the RCS through an opening with an area of at least 2.20 souare i inches, or (3) verify at least one presssuriser power operated relief valve is maintained open. i

b. If both power operated relief valves required by

{ ,3 3.15.3 are inoperable, perform at least one of the following within the next 24 hours ~ (1) restore operability of at least one power operated rellaf. valve, or l! (2) depressurise and vent the RCS through an i opening with an t.rea of at least 2.20 square lI d inches, or l ] l f. l (3) verify at least one pressuriser power operated relief valve is maintained open. 3.15 1 Amend:.unt Hos. 79173 l 60 i

Attechnent No. 20 i REACTOR COOLANT SYSTEM OVERPRESSURE MIT! GATING $YSTEMS l LIMITING CONDITION FOR OPERATION t 3.4.9.3 The high pressure safety injection flow paths to the Reactor Coolant System (RCS) shall be isolated, and below an RCS average coolant' temperature of 275'F at least one of the following Overpressure Mitigating Systems shall be OPERABLE: 4 4 a) Two power operated relief valves (PORVs) with a lift setting of 415 1 15 psig, or b) The RCS depressurized with a RCS vent of greater than or equal to 2.20 square inches. I APPLICABILITY: MODES 4, 5 and 6 with the reactor vessel head on. l.. " AGTION: l a. WiththehighpressuresafetyinjectionflowpathstotheRCS unisolated, restore isolation of these flow paths within 4 hours, and b. In MODE 4 with RCS average coolant temperature less than or equal to 275'F, and in MODE 5 or in Mode 6 with the reactor vessel head on: ( 1. With one PORV inoperable, perfom at least one of the following I within the next 7 days: i a) Restore the inoperable PORV to OPERABLE status, or b) Depressurize and vent the RCS through at least a 2.20 square inc l c) Depressurize and maintain a RCS vent through at least one open PORV and open associated block valve. 2. With both PORVs inoperable, depressurize and vent the RCS through at least a 2.20 square inch vent within 24 hours, 3. In the event either the PORVs or a 2.,20 square inch vent is'used to mitigate an RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specifica-tion 6.9.2 within 30 days. The report shall describe the circumstances initiating the transient, the effect of the PORVs l or RCS vent (s) on the transient, and any corrective action j necessary to prevent recurrence. TURKEY POINT - UNITS 3 & 4 3/4 4-36 AMENDMENT NOS. AND FEB 2 81989

v Attachment No. 20 REACTOR COOLANT SYSTEM 0VERPRESSURE MITIGATING SYSTEMS SURVEILLANCE REQUIREMENTS 4.4.9.3.1 Each PORV shall be demonstrated OPERABLE by: l i Performance of an ANALOG CHANNEL OPERATIONAL TEST on the PORV actua-a. tion channel, but excluding valve operation, within 31 days prior to entering a condition in which the PORY is required OPERABLE and at least once per 31 days thereafter when the PORV is required i OPERABLE. b. Perfomance of a CHANNEL CALIBRATION on the PORV actuation channel at least once per 18 months; and v Verifying the PORY block valve is open at least once per 72 hours c. when the PORV is being used for overpressure protection. d. While the PORVs are required to be OPERABLE, the backup air supply ~~ shall be verified OPERABLE at least once per 24 hours. 4.4.9.3.2 The 2.20 square inch vent shall be verified to be open at least once per 12 hours

when the vent (s) is being used for overpressure protection.

4.4.9.3.3 Verify the high ressure in,jection flow path to the RCS is isolated at least once per 24 hours by closed valves with power removed or by locked closed manual valves. T P i

Except when the went pathway is provided with a valve which is locked, sealed.

or otherwise sequred in the open position, then verify these valves open at least once per 31 days. d TURKEY POINT - UNITS 3 & 4 3/4 4-37 AMENDMENT N05. AND FEB 8 1968

c Attachment N3. 21 CCHP.v 3.7 ELECTRICAL SYSTEMS 1 ( Applicability: Applies to the availability of electrical power ] for the operation of auxiliaries. Ob.iective: To define those conditions of electrical power i availablity necessary (1) to provide for safe reactor operation, and (2) to provide for the continuing availability of engineered safety features. Specification: 1. Either reactor shall not be started from a l cold shutdown without* i a. The associated 239 KY-4160 volt start-up transformer in service. b. 4160-volt busses A and 8 of the associated unit, and either bus A or B of the second uni,t, energized. c. THREE out of FOUR 480-volt load centers and 480-volt motor control centers A, B or C, and D of the associated unit energized. d. TWO diesel generators operable with on site supply of 40,000 gallons of fuel ( available. e. Four batteries and associated DC systems are operable with FOUR out of $1X battery chargers operable. 2. During power operation or restarting from hot shutdown the following components may be inoperable: l a. ONE start-up transformer may be out of service provided both diesel generators I are operable. The NRC shall be notified l within 24 hours and be advised of plans to restore the transfomer to service. 4 L endments f/'& 3* b1 $3

Attechnent No. 21 i b. Power operation may continue if. ONE diesel generator is out of service provided (1) ) the renaining diesel generator is tested j daily and its associated engineered safety features are operable, and (2) either l start-up transfonner is operable. If the diesel outage is to be seven (7) days or more the NRC shall be notified. l c. ONE battery may be out of service for a period of twenty four hours. d. Specification 3.0.1 applies to 3.7.2. l r ( 'l Amendments S 3.7-2

Attachment No. 21 4.8 EMERGENCY POWER SYSTEM PERIODIC TESTS 1 Applicability: Applies to periodic testing and surveillance requirements for the emergency power system. Ob.iective: To verify that the emergency power system will respond promptly and properly. Specification: The following tests and surveillance shall be performed as stated: j 1. Diesel Generator Each diesel generator shall be demonstrated OPERA 8LE: a. On a staggered test basis (nonconcurrently) at the frequency specified by Table 4.8-1 by: 1. Verifying fuel level in the day tank and in the engine-mounted fuel tank. 2. Verifying fuel level in the fuel storage tank. 3. Verifying that a fuel transfer pump can be started and transfers fuel from the Diesel 011 Storage Tank to the Day Tank. 4. Verifying that the diesel generator starts from ambient conditions and accelerated to provide ( 6011.2 Hz frequency and 41601624 volts in < 15 seconds. 5. Verifying that the generator is synchronized, loaded to >2500 kw within 10 minutes and operates for >60 miliutes. l 6. Verifying that the diesel generator cooling system functions within design limits during the 1-hour full load test required by Specification 4.8.1.a.5. b. At least once per g2 days by verifying that a sample of diesel fuel from the fuel storage tank'is within acceptable limits when checked for viscosity, water, and sediment. c. During each Unit 3 refueling outage by: 1. Subjecting the diesel to an inspection in conjunction with its manufacturer's reconnendations for this class of standby service. ' d. At least once per 18 months by: l 1. Verifying the diesel generator's capability to: 1 4.8-1 Amendment Nos.120 & 114

Attechnent No. 21 l (a) Reject a lead of 200 kw without exceeding 4160+624 volts and 60+1.2 Hz. l (b) Reject complete load without exceeding 4160+624 volts, and without exceeding t overspeed limits. l 2. Verifying that diesel generator trips which are l operable during the test mode of diesel operation i are inoperable when the diesel is not in the test t mode of operation. 3. Alternately initiating one of the following two l l diesel startup tests: I (a) Simitate a safety injection signal, and allow the diesel generator to achieve nominal rated voltage and speed. Then 9 simulate a loss of offsite power, and 3 allow the diesel generator to load and stabilize. i (b) Simulate a loss of offsite power, and allow the diesel generator to load and stabilize. Then simulate a safety injection signal, and allow the diesel pnerator to sequence i safety loads and stab'lize. ( 4 Monitoring the tests specified in 4.8.1.c.4 to: I 5 (a) Verify proper doenergitation and load shedding from the 4160 volt busses. i (b) Verify that the diesel generator starts from ambient conditions and accelerates to l provide 60+1.2 Hz frequency and 416+624 volts in iTB seconds. ~ 5. Verifying that the diesel generator operates for { 3 at least 8 hours by performing the following tests: (a) Load the diesel generator to >2750 kw during the first 2 hours of the 8 hour test. i (b) Load the diesel generator to >2500 kw during L the last 6 hours of the 8 hour test. l (c) Verify that voltage, frequency, and cooling system functions are within design limits during the 8 hour full-load test. i i 6. Demonstrating the ability to sequentially: I 1 4.8-2 Amendment Nos. 120 & 114 C.

AttCchment N3. 21 l l 1 (a) Synchronize the diesel generator with offsite power while the generator is supplying emergency loads: i (b) Transfer the emergency load to offsite powers . i (c) Isolate the diesel generators and (d) Return the diesel generator to standby status. 7. Verifying that auto-connected loads to each diesel generator do not exceed 2730 kw. At least once per 10 years or after any modification that could affect e. k diesel generator independence, start both diesel generators simultaneously I at a time when both reactors are shutdown and verify that both diesel generators provide 60 g 1.2 Hz frequency and 4160 1 624 volts in less than 15 seconds. l l ( c i 1 l l l + c I i e a f 4.8-3 Amendment Nos.127 and 121 y ,.,w _, _,,, -. -.,, .,_-~.,...,n,.,n ,,--.,,,,-,,,y..y -,,..,.n,_..,.-.-,ne,,.,, --,.,,..,, n

Attach:Ont No. 22 3/4.8 ELECTRICAL POWER SYSTEMS I _3/4.8.1 AC SOURCES OPERATING LIMITING CONDITION FOR OPERATION 3.8.1.1 As a minimum, the following AC electrical power sources shall be OPERABLE: a. TWO 239 KV-4160 volt startup transformers with associated circuits, b. TWO diesel generators each with a day and skid-mounted fuel tank containing a minimum volume of 2,000 gallons of fuel, and c. A fuel storage system containing a minimum volume of 38,000 gallons of fuel and capable of transferring fuel to day tanks via a fuel transfer pump. APPLICABILITY: MODES 1, 2, 3, 4

ACTION:

a. With either startup transformer inoperable, ( 1) Demonstrate the OPERABILITY of both diesel generators by performing surveillance requirement 4.8.1.1.2.a.4 separately, for each diesel generator within 24 hours, if the diesel generator has not been suc-cessfully tested within the past 24 hours, and at least once per 24 hours while the startup transformer is inoperable, 2) Notify the NRC within 24 hours of declaring a startup transformer inoperable, l 3) Demonstrate the OPERABILITY of at least two cranking diesel generators by performing surveillance requirement 4.8.1.1.4 within 12 hours. The requirements of specification 3.0.3 do not apply to this ACTION statement, i 4) Demonstrate the OPERABILITY of the other startup transformer and its ~ associated circuits by performing surveillance requirement 4.8.1.1.1 l within I hour and at least once per 24 hours thereafter, and 5) a) For the unit with its startup transformer inoperable in MODE 1, i restore the inoperable startup transformer to OPERA 8LE status within 24 hours or reduce THERMAL POWER to < 30% RATED POWER within the next 6 hours. Restore the inoper@le sTartup transformer to OPERABLE status within 30 days e place both units in at leist HOT STANDBY within the next 12 hours and in COLD SHUTDOWN within the following 30 hours. l 4 TURKEY POINT - UNITS 3 & 4 3/4 8-1 AMENDMENT N05. AND OG8 MAY 0 51999

i E b) With the unit in MODES 2, 3 or 4. restore the inoperable startup transformer to OPERABLE status within 24 hours or place the unit in at least HOT STANDBY within the next 6 hours and in COLD l SHUTOOWN within the following 30 hours, b. With either diesel generator inoperable, for reasons other than the performance of surveillance requirement 4.8.1.1.2.c. l 1) Demonstrate the OPERABILITY of the remaining diesel generator by performance of surveillance requirement 4.8.?.1.2.a.4 within 24 hours i and once per 24 hours thereafter while the diesel generator is inoperable, j 1 2) Within 2 hours verify that the engineered safety features that depend on the remaining. diesel generator are OPERA 0LE, and verify compliance with specification 3.8.2.1, 1 3) Demonstrate the OPERABILITY of the startup transformers and their associated circuits by performing surveillance requirement 4.8.1.1.1 within I hour, and at least once per 24 hours thereafter, and 4) Restore the inoperable diesel generator to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 12 hours and in COLD SHUTDOWN within the following 30 hours. This ACTION applies to both units simultaneously. I ( c. With either diesel generator inoperable, for the performance of surveillance requirement 4.8.1.1.2.c, 1) Demonstrate the OPERABILITY of the remaining diesel generator by performance of surveillance requirement 4.8.1.1.2.a.4 within 24 hours and once per 24 hours thereafter while the diesel generator is inoperable, 2) Within 2 hours verify that the engineered safety features that depend on the remaining diesel generator are OPERABLE, and verify compliance with specification 3.8.2.1, 3) Demonstrate the OPERABILITY of the startup transformers and their associated circuits by performing surveillance requirement 4.8.1.1.1 within 1 hour, and at least once per 24 hours thereafter, and 4) Restore the inoperable diesel generator to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within t.he following 30 hours. d. With one startup transformer and one diesel generator inoperable, 1) Demonst4tetheOPERABILITYoftheremainingdieselgeneratorby performance of surveillance requirement 4.8.1.1.2.a.4 within 8 hours and once per 24 hours thereafter while the diesel generator is inoperable, I TURKEY POINT - UNITS 3 & 4 3/4 8-2 AMENDMENT N05. AND

Attcch:2nt ND. 22 4 r 2) Within 2 hours verify that the engineered safety features that I depend on the remainin diesel generator are OPERABLE, and verify l compliance with specif cation 3.8.2.1, 3) Demonstrate the OPERABILITY of the remaining startup transformer and its associated circuits by performing surveillance requirement 4.8.1.1.1 5 within 1 hour, and at least once per 24 hours thereafter, 4) Demonstrate the OPERABILITY of at least 2 cranking diesel generators ' by performing surveillance requirement 4.8.1.1.4 within 12 hours. The requirements of specification 3.0.3 do not apply to this ACTION, statement, 5) Comply with the requirements of specification 3.8.1.1 ACTION a.5', and ACTION b.4 or ACTION c.4 whichever is applicable, and 6) Notify the NRC within 4 hours of declaring both a startup transformer and a diesel generator inoperable. With two diesel generators inoperable, demonstrate the OPERABILITY of e. both startup transformers and their associated circuits by performing l surveillance requirement 4.8.1.1.1 within I hour and, at least once per j 24 hours thereafter. Restore at least one of the inoperable diesel generators to OPERABLE status within 2 hours or be in at least HOT STANDBY within the next 12 hours and in COLD SHUTDOWN within the following 30 hours. This ACTION applies to both units simultaneously, f. With two startup transformers inoperable, 1) Demonstrate the OPERABILITY of both diesel generators by performance of surveillance requirements 4.8.1.1.2.a.4 within 8 hours and once per 24 hours thereafter while the startup transformer (s) are inoperable, unless the diesel generators are already operating, 2) Restore one of the inoperable startup transformers to OPERABLE status within 24 hours or place one unit in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. Then place the other unit in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours, and 1 3) Notify the NRC within 4 hours of declaring both startup transformers inoperable. SURVEILLANCE REQUIREMENTS 4.8.1.1.1 Each required startup transformer and its associated circuits shall be determined OPERABLE at least once per 7 days by' verifying correct breaker alignment and indicated power availability. l 4.8.1.1.2 EachdieselgeneratorshallbedemonstratedOPERABLE: l l a. In accordance with the frequency specified in Table. 4.8-1, with diesel generator surveillances performed nonconcurrently by: TURKEY POINT - UNITS 3 & 4 3/4 8-3 AMENDMENT NOS. AND MAY 0 5195 i

Attcchment No. 22 L l 1) Verifying the day and skid-mounted fuel tanks contain a minimum volume of 2,000 gallons of fuel. I 1 2) Verifying the minimum fuel volume of 38,000 gallons in the Diesel Oil Storage Tank. 3) Verifying that a fuel transfer pump can be started and transfers fuel from the Diesel Oil Storage Tank to the Day Tank. j 4) Verifying that the diesel generator starts from normal conditions and accelerates to provide 60 + 1.2 Hz frequency and 4160 + 624 ~ ~ volts in < 15 seconds". within 10 minutes

generator is synchronized, loaded to > 2500 kw Verifying that the 5) and operates for > 60 minutes, and the cooling system operates within design limits!

6) Verifying that the diesel is aligned to provide standby power to the l associated emergency buses, b. At least once per 92 days by verifying that a sample of diesel fuel from i the Diesel Oil Storage Tank is within acceptable limits when checked for viscosity, water, and sediment. c. During each Unit 4 refueling outage by: ( 1) Subjecting the diesel to an inspection in conjunction with its manufacturer's recommendations for this class of standby service. d. At least once per 18 months by: I 1) Verifying the diesel generator's capability to: a) Reject a load of greater than or equal to 380 kw without l exceeding 4160 t 624 volts and 60 1 1.2 Hz. I b) Reject a load of greater than or equal to 2500 kw without tripping. The generator voltage shall retdn to less than or equal to 4784 volts within 2 seconds following the load rejection. P

The diesel generator start (15 sec) from normal conditions shall be performed at least once per 184 days in these surveillance tests.

All other engine starts for the purpose of this surveillance testing may be preceded by an engine prelube period and/or other warmup procedures recommended by the manu-facturer so that mechanical stress and wear on the' diesel engine is minimized. e s.' TURKEY POINT - UNITS 3 & 4 3/4 8-4 AMENDMENT N05. AND 71 uw o s me

ph { Attcchment N3. 22 2) Verifying that diesel generator trips that are made operable during the test mode of diesel operation are inoperable when the diesel is j not in the test mode of operation. 3) Alternately initiating one of the following two diesel startup tests, a) Simulate a safety injection signal, and allow the diesel generator to achieve nominal rated voltage and speed. Then i t simulate a loss of offsite power, and allow the diesel generator to load and stabilize. b) Simulate a loss of offsite power, and allow the diesel generator to load and stabilize. Then simulate a safety injection signal, and allow the diesel generator to sequence safety loads and stabilize. 4) Monitoring the tests specified in 4.8.1.1.2.d.3 to: a) Verify proper deenergization and load shedding from the 4160 l volt bussas. b) Verify that the diesel generator starts from ambient conditions and accelerates to provide 60 i 1.2 Hz frequency and 4160 1 624 volts in 5,15 seconds. 5) Verifying that the diesel generator operates for at least 8 hours by ( performing the following tests: a) Load the diesel generator to > 2750 kw during the first 2 hours of the 8 hour test. During tEis 2 hour period, increase the load to > 2850 kw until the generator electrical load is stabilized and then decrease back to > 2750 kw. b) Load the diesel generator to 1 2500 kw during the last 6 hours of the 8 hour test. c) Verify that voltage, frequency, and cooling system functions are within design limits during the 8 hour full-load test. 6) Demonstrating the ability to sequentially: t l l a) Synchronize the diesel generator with offsite power while the generator is supplying emergency loads: b) Transfer the emergency load to offsite power; c) Isolate the diesel generator; and d) Re}urnthedieselgeneratortostandbystatus. l 7) Verifying the auto-connected loads to each diesel generator do not i. exceed 2750 kw. ( TURKEY POINT - UNITS 3 & 4 3/4 8-5 AMENDMENT N05. AND g h1AY 0 5 nes

- _ = - Attcchment NO. 22 y At least once per 10 years or after any modification that could affect e. ( diesel generator independence, start both diesel generators simultaneously at a time when both reactors are shutdown and verify that both diesel generators provide 60 1 1.2 Hz frequency and 41601 624 volts in less than or equal to 15 seconds.

4. 8.1.1. 3 Reports - All valid diesel generator failures shall be reported to the Commission in a Special Report pursuant to Specification 6.9.2 within 30 days.

Reports of diesel generator failures shall include the information recommended in Regulatory Position C.3.b of Regulatory Guide 1.108, Revision 1. August 1977. If the number of failures in the last 100 valid tests is greater than or equal to 7, the report shall be supplemented to include the additional information recommended in Regulatory Position C.3.b of Regulatory Guide 1.108. Revision 1, 1 August 1977.

4. 8.1.1. 4 At least two cranking diesel generators shall be demonstrated OPERABLE as required by specification 3.8.1.1 ACTIONS a.3 and d.4 by verifying that the cranking diesel generators manually start from normal conditions and acceleratetoprovide60t1.2QGifrequencyand4160t624voltsandarecapable of being aligned to either 4160M olt safety bus.

3 ( T l l l r b i l 4 6' TURKEY POINT - UNITS 3 & 4 3/4 8-6 AMENDMENT N05. AND MAY 0 51989

p5) Attachment No. 22 TABLE 4.8-1 i DIESEL GENERATOR TEST SCHEDULE 3 i NUMBER OF FAILURES IN t LAST 20 VALID TESTS

TEST FRE0VENCY il Once per 31 days

>2** Once per 7 days 'l i h

Criteria for determining number of failures and number of valid tests shall be in accordance with Regulatory Position C.2.e of Regulatory Guide 1.108, Revision 1, August 1977, but determined on a per diesel generator basis.

( For the purpose of determining the required valid test frequency, the previous valid test failure count may be reduced to zero if a complete diesel overhaul to like-new condition is completed, provided that the overhaul, including appropriate post-maintenance operation and testing, is specifically approved by the manufacturer and if acceptable reliability has been comonstrated. The i reliability criterion shall be the successful completion of 14 consecutive valid tests in a single series. Ten of these valid tests shall be in accord-l ance with the routine Surveillance Requirements 4.8.1.1.2.a.4 ar.d 4.8.1.1.2.a.5; and four valid tests in accordance with the 184 day testing requirement of l Surveillance Requirements 4.8.1.1.2.a.4 and 4.8.1.1.2.a.5. If this criterion is not satisfied during the first series of valid tests, any alternate cri-terion to be used to transvalue the failure count to zero requires prior NRC approval.

- The associated valid test frequency shall be maintained until seven consecu-tiva failure free demands have been performed and the number of failures in the last 20 valid demands has been reduced to one.

l 4 se l TURKEY POINT - UNITS 3 & 4 3/4 8-7 AMENDMENT N05. AND 74 tw 0 s m

V l Attcchesnt N3. 23 TABLE 1.2 l OPERATIONAL MODES e ( .} ' REACTIVITY % RATED AVERAGE COOLANT MODE, CONDITION. S THERMAL POWER

TEMPERATURE l

gf 1. POWER OPERATION > 0.99 > 5% > 150'F f 2. STARTUP 1 0.99 3 5% 1 350'F 3. HOT STANDBY < 0.99 0 > 350'F 4. HOT SHUTDOWN < 0.99 0 350'F > T > 200'F avg 5. COLD SHUTDOWN < 0.99 0 5 200'F 6. REFUELING ** $ 0.95 0 1 140'F 7 " Excluding decay heat. ^*Futi in the reactor vessel with the vessel head closure bolts less than fully i tensioned or with the head removed. f. l '. ' '.i P 4 9 d e s TURKEY POINT - UNITS 3 & 4 1-8 AMENDMENT N05. AND ) " ;"3 75 \\

f Attach =nt No. 24 REACTOR COOLANT SYSTEM 3/4.4.2 SAFETY VALVES O.m0m LIMITING CONDITION FOR OPERATION AminimumofonepressurizerCodesa[etyvalveshallbedPERA8LE*with 3.4.2,1 l a lif t setting of 2485 psig

IL**

APPLICABILITY: MODES 4 and 5. ACTION: I With no pressurizer Code safety valve OPERABLE, immediately suspend all opera-tions involving positive reactivity changes and place an OPERABLE RHR loop into operation in the shutdown cooling mode. i OuRvEIttANCE RE0u1REMENTs 4.4.2.1 No additional requirements other than those required by Specification 4.0.5. While in MODE 5, an equivalent size vent pathway may be used provided that the vent pathway is rpt isolated or sealed.

- The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

) O TURKEY POINT - UNITS 3 & 4 3/4 4-7 AMEN 0 MENT NOS. AND 7(p FEB 2 81989 1 ~J.

'Z""

1 ? .~ f.1 J

Att chment Na. 24 j REACTOR COOLANT SYSTEM OPERATING IMITING CONDITION FOR OPERATION f i 3.4.2.2 All pressurizer Code safety valves shall be OPERABLE with a lift setting of 2485 psig 1 1%.* APPLICABILITY: MODES 1, 2 and 3. ACTION: fith ene pressurizer Code safety v41ve inoperable, either restore the inoper-cble valve to OPERABLE status within 15 minutes or be in at least NOT STANDBY within 6 hours and in at least HOT SHUTDOWN within the following 6 hours. f SURVEILLANCE REQUIREMENTS No additional, requirements other than those required by IJp.4.2.2 ecification 4.0.5. l 1 l 0

The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperatura and pressure.

TURKEY POINT - UNITS 3 & 4 3/4 4-8 AMENDMENT N05. AND FEB : s 1989 ~ -,--,--4-,--.- .-m----

Attachunt No. 25 [ 4.0 $URVEILLANCE REoutREENTS 4.0.1 5peelflod intervals may be adjusted plus or minus 25% to accommodate i normal test schedules. [ i When the reactor is in a shutdown condition, some of the surveillance 4.0.2 requirements discussed in this section are not required to be satisfied provided that the safety !!mits or umiting conditions for operation for the shutdown status are satistled. When a surve!Uance activity is not completed because.the reactor is shutdown and the surveluence is not j required, the surveluance requirement shall be met prior to the time indicated in the applica.'? footnote. Surveulance Requirements for inservice tempaction of ASME Code Class 1, 1 6.0.3 2, and 1 components shau be applicable as feuows: l I Inservlee insoection of ASME Code Class 1,2, and 3 components s%a!! a) be periormed in accordance with Section 10 et the ASME Soller and i i Pressure Vessel Code and appucable Addenda as required by 10 CPR 50, Section 50.S$a(g), except where specific written reHet tas been the Commission pursuant to 10 CPR S0, Section granted by(0. 50.55a(gX6) b) Surveulance intervals speelflod in Section XI of the ASME.Soller and Pressure Vessel Code and appucable Addenda for the inservice inspection activities required by the ASME Soller and Pressure Yessel Code and app!! cable Addenda shau be applicable as fouows in these t Technical Specifications: I A5ME Soller and Pressure Vessel Code and appucable Addenda Required frequencies for l terminology ior Inservice performing inservice inanection activities letian actMtles I Weekly At least once per 7 days Monthly At least once per $1 days Quarterly or every 3 months At least once per 92 days Semlannunuy or every 6 months At least once per !$4 days Every 9 months At least once per 716 days Yearly or annusuy At least once per 366 days c) The provisions of Spectf! cation 4.0.1 are applicable to the above required frequencies for performing inservice inspection netivities. / d) Performance of the above inservice inspection activities shall be in addition to other specified Surveluence Wrements, e)

Nothing in the ASME Bouer and Pressure Vessel Code shan be construed to

_.: t the requirements of any Technical Speelfication.

i Amendment Nos. n2 and,U1 l e.o.1 y

Attachment No. 26 APPLICABILITY ( SURVEILLANCE REQUIREMENTS i 4.0.1 Surveillance Requirements shall be met during the OPERATIONAL MODES or other conditions specified for individual Limiting Conditions for Operation unless otherwise stated in an individual Surveillance Requirement. 4.0.2 Each Surveillance Requirement shall be performed within the specified time interval with a maximum allowable extension not to exceed 25% of the surveillance interval. l 4.0.3 Failure to perform a Surveillance Requirement within the allowed surveillance interval, defined by Specification 4.0.2, shall constitute noncompliance with the OPERABILITY requirements for a Limiting condition for Operation. The time limits of the ACTION requirements are applicable at the time it is identified that a Surveillance Requirement has not been performed. The ACTION requirements may be delayed for up to 24 hours to permit the completion of the surveillance when the allowable outage time limits of the ACTION requirements are less than 24 hours. Surveillance Requirements do not have to be perfomed on inoperable equipment. 4.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made unless the Surveillance Requirement (s) associated with a Limiting Condition for Operation has been performed within the stated surveillance g interval or as otherwise specified. This provision shall not prevent passage through or to OPERATIONAL MODES as required to comply with ACTION requirements. 4.0.5 Surveillance Requirements for inservice inspection and testing of ASME Code Class 1, 2, and 3 components shall be applicable as follows: a. Inservice inspection of ASME Code Class 1, 2, and 3 components and inservice testing of ASME Code Class 1, 2, and 3 pumps and valves ) shall be performed in accordance with Section XI of the ASME Boiler l and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g), except where specific written relief 1 has been granted by the Commission pursuant to 10 CFR 50, Section 50.55a(g)(6)(i). l 4 TURKEY POINT - UNITS 3 & 4 3/4 0-3 AMENDMENT N05. AND. 79 FEB 28 m 1

i l 1 i UNITED STATES OF AMERICA t NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC EAFETY AND LICmMSIMG BohmD + l l ) l l In the Matter of ) i ) FLORIDA POWER & LIGHT COMPANY ) Docket Nos. 50-250 OLA-5 ) 50-251 OLA-5 i ) (Turkey Point Plant, Units 3 ) (Technical Specifications and 4) ) Replacement) ) camTIFIcaTE or aEnvIca I hereby certify that copies of a letter dated April 4, 1990, together with the enclosures, from Harold F. Reis to the Licensing Board Members in the above captioned proceeding were served on the following by deposit in the United States mail, first-class postage paid on the date shown below. 3 Peter B. Bloch, Chairman Atomic Safety and Licensing Board U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Dr. George C. Anderson 7719 Ridge Drive, N.E. Seattle, WA 98115 t Elizabeth B. Johnson Oak Ridge National Laboratory P.O. Box 2008 Bethel Valley Road, Bldg. 3500 Mail Stop 6010 Oak Ridge, TN 37831 Atomic Safety and Licensing Board Panel Adjudicatory File U.S. Nuclear Regulatory Commission Washington, D.C. 20555 (two copies) Atomic Safety and Licensing Appeal Board Panel Adjudicatory File U.S. Nuclear Regulatory Commission washington, D.C. 20555 (three copies) 1 e,., n -n.-- m

b i I 9 Office of the Secretary U.S. Nuclear Regulatory Commission 1 Washington, D.C. 20555 Attention: Chief Docketing and Service Section (Original plus two copies) Thomas J. Saporito, Jr. - Executive Director Nuclear Energy Accountability Project P.O. Box 129 Jupiter, Florida 33468-0129 Janice E. Moore, Esq. Patricia A. Jehle, Esq. Office of General Counsel ) U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Richard Goddard, Esq. Regional Counsel U.S. Nuclear Regulatory Commission, Region II 101 Marietta Street, N.W.,

2900 Atlanta, GA 30323 l

John T. Butler, Esq. Steel, Hector & Davis 4000 Southeast Financial Center Miami, Florida 33131 April 4, 1990 Harold F. Reis Newman & Holtzinger, P.C. 1615 L Street, N.W. Suite 1000 Washington, D.C. 20036 _.}}

ML20042D851

Text

Dear Licenairig Board Members:

3.4.4 APPLICABILITY

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