L-89-329, Forwards Addl Info Re 890605 Proposed License Amend to Implement Revised Tech Specs,Including Proposed Changes to Fsar.Requests Mod of Requirements & Bases for Tech Specs 4.6.2.2.a

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Forwards Addl Info Re 890605 Proposed License Amend to Implement Revised Tech Specs,Including Proposed Changes to Fsar.Requests Mod of Requirements & Bases for Tech Specs 4.6.2.2.a
ML17347B402
Person / Time
Site: Turkey Point  
Issue date: 11/03/1989
From: Harris K
FLORIDA POWER & LIGHT CO.
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
L-89-329, NUDOCS 8911130158
Download: ML17347B402 (131)
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{{#Wiki_filter:ACCELERATED ISTRIBUTION DEMONSTRATION SYSTEM ~'J) ~~I REGULATORY INFORMATION DISTRXBUTION SYSTEM (RXDS) ACCESSION NBR: 8911130158 DOC. DATE: 89/11/03 NOTARIZED: NO FACIL:50-250 Turkey Point Plant, Unit 3, Florida Power and Light C 50-251 Turkey Point Plant, Unit 4, Florida Power and Light C AUTH.NAME AUTHOR AFFILIATION HARRIS,K.N. Florida Power & Light Co. RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk) DOCKET 05000250 05000251

SUBJECT:

Submits addi info re proposed license amend revised Tech Specs. DISTRIBUTION CODE: A001D COPIES RECEIVED:LTR I ENCL Q SIZE: 6~ TITLE: OR Submittal: General Distribution I NOTES: RECIPIENT ID CODE/NAME PD2-2 LA EDISON,G INTERNAL: NRR/DEST/ADS 7E NRR/DEST/ICSB NRR/DEST/RSB 8E NUDOCS-ABSTRACT OGC/HDS2 RES/DSIR/EIB EXTERNAL: LPDR NSXC COPIES LTTR ENCL 1 1 5 5 1 1 1 1 1 1 1 1 1 0 1 1 1 1 1 1 RECIPIENT ID CODE/NAME PD2-2 PD NRR/DEST/ESB 8D NRR/DEST/MTB 9H NRR/DOEA/TSB 11 OC ENB 01 NRC PDR COPIES LTTR ENCL 1 1 1 1 1 1 1 1 1 0 1 1 1 1 NOTE TO ALL"RIDS" RECIPIENTS: 4k~ > PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM Pl-37 (EXT. 20079) TO, ELIMINATEYOUR NAMEFROM DISTRIBUTION LISIS FOR DOCUMENTS YOU DON'T NEED! TOTAL NUMBER OF COPIES REQUIRED: LTTR 21 ENCL 19

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P.O. Box14000, Juno Beach, FL 33408-0420 NOVEM8ER 0 g 1989 L-89-329 10 CFR 50.90 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555 Gentlemen: Re: Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 Proposed License Amendment Revised Technical Specifications Additional Information The purpose of this letter is to provide additional information to support the NRC review of FPL's proposed license amendment to implement the Revised Technical Specifications (RTS) for Turkey Point Units 3 and 4. This license amendment was submitted by an FPL letter (L-89-201) dated June 5, 1989 and a package of expected Final Safety Analysis Report (FSAR) revisions was provided by FPL letter (L-89-235) dated July 12, 1989. The following information is provided to further assist in the review process: Attachment 1 contains proposed changes to the FSAR to add Reactor Protection System and Engineered Safety Feature Actuation setpoints to Chapter 7 and to delete reference to operational values of these setpoints found in various logic diagrams. Also included is a clarification of the description of control room ventilation isolation initiation. 2. Attachment 2 consists of proposed FSAR changes for Chapter 6 to add the list of existing automatic containment isolation valves which are required by RTS 3.6.4. New FSAR Table 6.6-3 lists automatic valves receiving Containment Phase A, Phase B, and Containment Ventilation Isolation signals in FSAR Table 6.6-1 which are required to meet Turkey Point General Design Criteria 53. This list does not represent. a change from our current Technical Specification (T.S. 3.3.3) which specifies isolation valves receiving Phase A, Phase B and Containment Ventilation Isolation signals. The isolation times listed with the valves, reference the In Service Test (IST) program requirements with the exception of the 85'lil30158 89ii03 i DR ADOCK OSOOO2SO gyO] an FPL Group company

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U. S. Nuclear Regulatory Commission L-89-329 Page two containment purge valves which are based on accident analysis assumptions. Minor text changes are also included which provide clarification and continuity. 3 ~ 4 ~ Attachment 3 consists of proposed FSAR changes to clarify the auxiliary feedwater actuation signals. To clarify one of the actuating signals, a minor revision is also provided in the listing of Loss of Power setpoints. The coincident open diesel generator breaker, which is needed to complete the loss of power logic, is added to the engineered safety feature actuation system instrumentation setpoint table. Attachment 4 provides an explanation of the basis for selected RTS parameters and setpoints. The majority of these parameters are not found in the Current Technical Specifications or in the FSAR. As provided previously in our July 12, 1989 letter, the revisions proposed in Attachments 1, 2 and 3 are subject to subsequent modification for editorial purposes or to incorporate other non-related changes as part of the FSAR update activities performed under 10 CFR 50.71. In addition, in order to clarify the testing requirements for the emergency containment coolers (ECCs) in RTS 4.6.2.2.a, we request that the Surveillance Requirements and the Bases for this specification be modified. The proposed RTS changes are indicated in Attachment 5. These changes were discussed with the NRC Staff on September 28, 1989. This modification clarifies the Surveillance Requirement to ensure that the cooling water flow obtained in the surveillance test correlates to the 2,000 gpm flow required for the design basis condition. The surveillance frequency shown in the RTS is also changed from monthly to once per 18 months. We understand the monthly frequency is more appropriate for plants using service water systems for cooling which could result in tube fouling if not checked on a frequent basis. Because Turkey Point uses component cooling water for ECC cooling, rapid fouling is not expected, and an 18-month interval is appropriate. This interval is consistent with the current technical specification surveillance requirement. Also included in this attachment are marked up pages to clarify the containment isolation initiation under safety injection and to clarify the manual alignment of containment spray. We understand these changes are acceptable to the NRC Staff. The RTS represents a large increase in the level of detail compared to the current Technical Specifications. As the staff is aware, the Turkey Point FSAR does not contain a comparable level of

U. S. Nuclear Regulatory Commission L-89-329 Page three detail. Further, the FSAR often describes equipment capabilities, which do not represent design basis requirements. The NRC staff's independent auditor had comments on the comparison of sections of the RTS and FSAR. Resolution of these comments required no technical changes to the RTS. Proposed FSAR changes resulting from the auditor's comments are reflective of the Turkey Point FSAR level of detail and editorial style. Increasing the level of detail or descriptive style of the FSAR to match the RTS was not included in the scope of the upgrade project as previously agreed upon by FPL and NRC management. Should there be any questions on this information, please contact us ~ Very truly yours, h;H a~aws K. N. Harris Vice President Turkey Point Plant Nuclear KNH/PLP/gp Attachments cc: Mr. Stewart D. Ebneter, Regional Administrator, Region II, USNRC Senior Resident Inspector, USNRC, Turkey Point Plant

ATTACHMENT 1 REACTOR PROTECTION SYSTEM AND ENGINEERED SAFETY FEATURES SETPOINTS FSAR CHANGES

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During power operation, a sufficient amount of rapid shutdown capability in the form of control rods is administratively maintained by means of the control rod insertion limit monitors. Administrative control requires that all shutdown group rods be in the fully withdrawn position during power operation. ,prS cr A list of reactor trips> means of actuati q and 'tIIe coincident circuit requirements is given in Table 7.2-1. The interlock circuits, referred to in Table 7.2-1, are listed in Table 7.2-2. Manual Tri The manual actuating devices are independent of the automatic trip circuitry, and are not subject to failures which make the automatic circuitry inoperable. Either of two manual trip devices located in the control room can initiate a reactor trip. Hi h Nuclear Flux (Power Ran e) Tri This circuit trips the reactor when two of the four power range channels read above the trip set-point. There are two independent trip settings, a high and a low setting. The high trip setting provides protection during normal power operation. The low setting which provides protection during startup can be manually bypassed when two out of the four power range channels read above approximately 10K power (P10). Three out of the four channels below 101 automatically reinstates the trip function. The high setting is always active. Hi h Nuclear Flux (Intermediate Ran e) Tri This circuit trips the reactor when one out of the two intermediate range channels reads above the trip set-point. This trip which provides protection during reactor startup can be manually bypassed if two out of four power 7.2-19

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TABLE 7.2-1 4 Sheet 1 of~ LIST OF REACTOR TRIPS & CAUSES OF ACTUATION OF: ENGINEERED SAFETY FEATURES CONTAINMENT. AN NE ISOLATION & AUXILIARYFEEDWATER 2. REACTOR TRIP Manual High neutron flux ~ <O't 4 fat COINCIDENCE CIRCUITRY AND INTERLOCKS 1/2, no interlocks 2/4, no interlocks COMMENTS High and lov settings; manual block and automatic reset of low setting by P-10, Table 7.2-2 3. 4. 5. 6. 7. 8. 9. Overtemperature AT ploTF I Overpower AT Low pressurizer pressur P )gag ps/ (fixed set point) High pressurizer pressur C~ W Phj (fixed set point) oem'f High pressurizer vater leve 5~ Low reactor coolant flow Monitored electrical supply to reactor coolant pumps: 2/3, no interlocks /3, no interlocks /3, interlocked with P-7 2/3, no interlocks /3, interlocked vith P-7 2/3 per loop, interlocked vith P-7, and P-8 how flow in 2 loops permitted below P-7. Lov flow in 1 loop permitted below P-8. 9A. 9B. Undervoltage Underfrequency p gpss Valb >S'l,l Ng Loss o'f power on 1 out of 2 on each of 2 buses Under frequency on 1 out of 2 on each of 2 buses ..)Under frequency on 2 'out of 2 buses villtrip'll reactor coolant pumps and consequently cause reactor trip; interlocked vith P-7 and P-8.

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TABLE 7.2-1 Sheet 2 of% REACTOR TRIP COINCIDENCE CIRCUITRY AND INTERLOCKS COMMENTS 9C Reactor coolant pump breakers 10. Safety injection signal (Actuation) interlocked with P-7 and P-8 Low pressurizer pressure (2/3), or 2/3 high containment pressure; or 2/3 high differential pressure between any steam line header and steam line; or 2/3 high steam flow in coincidence with 2/3 low T or 2/3 steam line ave

pressure, or man5al 1/2 (See 7.2 System Description-Protective Action for Interlocks).

ll. Turbine-generator trip h /3, low auto stop oil pressure interlocked with P-7, or 2/2 stop valve closure indication (interlocked with P-7) 12. Steam/Peedwater flow mismatch, coincident with low steam generator level 1/2, (steam/feedwater flow mismatch) n coincidence with 1/2 low steam enerator water level per loop 13. Low-low steam generator water level 14. Intermediate range neutron flux 2/3, per loop 1/2, manual block permitted by P-10 Manual block and automatic reset 15. Source range neutron flux 1/2, manual block permitted by P-6, interlocked with P-10 Manual block and automatic reset

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e INSERT I TRIP SETPOINT gc. NA 10. NA ll. > 45 psig FULLY CLOSED 12. FEED FLOW < 0.64 X 10 1b/hr BELOW STEAM FLOW 15% of narrow range instrument span 13. > 15% of narrow range instrument span 14. 25% of RTP 15. 1 x 10 CPS

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TABLE 7.2-1 7@it Sheet 3 of~ P CONTAINHENT ISOLATION hCTVATION C~KTPCI OINCIDENCE CIRCUITRE RND INIE~ INTERLOCKS 16. Phase h - Safety In)ection Signa 17. Phase B - Containment pressure 18. High contaicment activity See Item 10; 2 oe"tery push butt uttona ~ pressing of either push button (1]2) will actuate. Coincidence of two 2/3 containment Hi pressure and 2/3 Hi-Hi pressure, (same signal which actuates contai~ent spray), or aanual 2/2 f ~f High activity signal, from air articulate detector or radiogas detector. (1/2) ~Sf 5 COOQggZS AcCuaCea all eaaanti l service conCaisaenC isolation Crip valvaa ia a44i Cional ~iinal cloaca conCainaent puttee ~upply and exhauat valves ~ Actuate n aaaanc) ~l ic+ conCai~C trip valva ENGINEERED SAFETY FEATVRES hCTUhTION 19. 20. Safety in)ection signal (s) Containment spray signal (P) 21. Emergency containment cooling and filtering ASST'r 0 See Item 1O XHSCsT3 2 out of 3 high containment pressure in coincidence with 2/3 High-High contairment pressure~ or manual 2 out of 2. Safety in)ection signal initiates NA starting of all fans in accordance with the Safety In)ection Starting Sequence. 4DI caVl

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0 TABLE 3.3-3 (Continued) ENGINEERED SAFETY FEATURE ACTUATIOH SYSTEM UM N FUNCTIONAL UNIT 3. Containment Isolation a. Phase "A" Isolation TRIP 5ETPOINT ALLOWABLE VALUEP 1) Manual Initiation N.A. 2) Automatic Actuation Logic H.A. and Actuation Relays H.A. H.A. 3) Safety Injection b. Phase "B" Isolation See Item 1 above for all Safety Injection Trip Setpoints and Allowable Values. 1) Manual Initiation N.A. H.A. 2) Automatic Actuation Logic H.A. and Actuation Relays <30.0 psig c. Containment Ventilation Isolation 1) Containment Isolation N.A. Manual Phase A or Phase B 2) Automatic Actuation Logic N.A. and Actuation Relays 3) Containment Pressure-High-High Coincident with: Containment Pressure-High <6.0 psig N.A. <[ ] psig <[ ) psig H.A. N.A. 3) Safety Injection 4) Containment Radio-activity-High (1) See Item 1. above for all Safety Injection Trip Setpoints and Al1 owab le Values. Particulate (R-11) [ 3 <6.1 x 10s CPM Cas ous (R-12) ~>spy ee TURKEY POINT - UNITS 3 4 4 3/4 3-24 AMENDMENT NOS. AND ~88 ss ~+>

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e FUNCTIONAL UNIT TABLE 3. 3-3 (Continued) ENGINEEREO SAFETY FEATURE ACTUATION SYSTEM N U 'TRIP SETPOINT ALLOWABLE VALUES 9. Control Room Isolation (Continued) c. Containment Radioactivity-High (1) Particulate (R-11) [ 3 <S.l x 10s CPM t'aseous (R-12) See (2) d. Containment Isolation Manual Phase.A or Phase B e. Air Intake Radiation Level N.A. < 2 IR/hr N.A. Z.93 ~R/4 t-TABLE NOTATIONS (1) Either the particulate or gaseous channel in the OPERABLE status will satisfy this LCO. 3 2xlOi (2) Containment Gaseous Monitor Setpoint =

CPM,

( F ) ~MSCR.7+ Where F Actual Pur e Flow Oesign Purge Flow (35,000 CFM) Setpoint may vary according to current plant conditions provided that the release rate does not exceed allowable limits provided in Specification 3.11.2.1. (3) A fli e f d 1 t ttl tt i f 1 d d h 5p Nfl'.7.1.2. SIf no allowable value is specified so indicated by t', 3, the trip setpoint shall also be the allowable value. URKEY POINT - UNITS 3 4 4 3/4 3-28 AMENOMENT NOS. AND 8 ).-.-. FEB ti

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TABLE 3.3-3 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM FUNCTIONAL UNIT 1. Safety Injection (Reactor Trip, Turbine Trip, Feedwater Isolation, Control Room Isolation, Start Diesel Generators, Containment Cooling Fans, Containment Filter Fans, Start Sequencer, Component Cooling Water, Start Auxiliary Feedwater and Intake Cooling Mater) a. Manual Initiation TRIP SETPOINT N.A. ALLOWABLE VALUEt N.A. b Automatic Actuation Logic N.A. N.A. c. Containment Pressure-High d. Pressurizer Pressure-Low e. High Differential Pressure Between the Steam Line Header and any Steam Line. f. Steam Line Flow-High Coincident with: Steam Generator Pressure-Low or T -Low Contas pr ay <6 psig >1715 psig. <150 psi <A function defined as follows: A hp corresponding to 0.64 x 10e lbs/hr at OX load increa-ing linearly to a hp corresponding to 3.84 x 10~ lbs/hr at full load. >600 psig >531OF <t: 3 psig ] psig ] psi >t,'] psig >E l'F a. Automatic Actuation Logic and Actuation Relays b. Containment Pressure-High-High Coincident with: Containment Pressure-High TURKEY POINT - UNITS 3 4 4 '<30.0 psi <6.0 psig 3/4 3-23 N.A. <k 1 psig <t: 3 psig AMENDMENT NOS. AND FEB 2 S lgag

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0: TABLE 7.2-1 filI f'TEAN LINES ISOLATION ACTUATION I COINCIOENCE CIRCUITRY AND INTERLOCKS Sheet 4 ofA" CURRENTS 22. Steam Flow ~WgFA'7 High steam line flow in 2 out of 3 loops coincident with either low T in 2 out of 3 loops or low steam lfn5 pressure in 2 out of 3 loops. 23. Containment pre sure hs)h h:Sl "hsing 24. Ha ua e C. 4.o t'sf)

c. 3qo fslg 2/3 high containment pressure signal in coincidence with 2/3 high-high containment pressure 1/1 per steam line AUXILIARY.FEEINhTER-ACTUATION 25.

Turbine driven pumps Coincidence of 2/3 low-low level in any steam generator; or loss of voltage on both 4160 volt buses', or auto trip of main feed water

pumps, or safety injection signal; or manual 1/2 HAIN FEEINATER ISOLATION 26A. Close main feedwater control valves (fast closure)

Actuated by: 1. Safety injection (see ¹10) 2. 2/3 high-high feedwater level (SOZ) in steam generator 3. Reactor trip coincident with low T (slow closure) avg 26B. Close bypass feedwater control valves 1. Safety xngectxon or hxgh-hzgh level zn steam generators 27. Trip steam generator feed pumps ~ ~ rt P I L. l>i.i ~ triP Safety injection signal Coincidence of 2/3 high-high level (80Z) in any steam generator Rev. 3-7/85

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TABLE 3.3-3 (Continued) ENGINEERED SAFETY FEATURE ACTUATION SYSTEM FUNCTIONAL UNIT 4. Steam Line Isolation a. Manual Initiation b. Automatic Actuation Logic and Actuation Relays c. Containment Pressure--High-High Coincident with: Containment Pressure-High 'TRIP SETPOINT N.A. N.A. <30.0 psig <6.0 psig ALLOWABLE VALUES N.A N.A. l psig <{ l psig f. Stean Line Flow-High <A function define [ ] as follows: A hp corresponding to 0.64 x 10e lbs/hl at OX load increa-ing linearly to a hp corresponding to 3.84 x 10'bs/hr at full load. Coincident with: Steaa Line Pressure-Low or Tv-Low 5. Feedwa solat on a. Automatic Actuation Logic and Actuation Relays >600 psig 531oF N.A. >{ ] psig N.A. b. Safety In)ection See Itta I. above for all Safety Infection Trip Setpoints and Allowable Values. 6. Auxiliary Feedwater (3) a. Autoeatic Actuation Logic '.A. and Actuation Relays N.A. b. Steaja Generator Water Level-Low-Low >15K of narrow range instrument span. >{ Q of narrow range instrument span c. Safety Injection See Itea 1. above for all Safety Injection Trip Setpoints and Allowable Values. 3/4 3-25 AMENDMENT NOS. AND

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~ I, ~ r I \\ K.r. 1 . ~ '! I I '. 1 4 ~- TABLE.2-/ Continued TABLE NOTATIONS NOTE 1: OVERTEHPERATURE dT ST ( ) < ST (K - K ~ [T ( ) - T'] + K (P - P') - f (S))) 1+ x S o ~ r (1+ TIS) 1+ rrS O. 5o4 & Mhere: hT 1 ~+x, Heasured AT by RTO Instruaentation; Lag coapensator on aeasured hT; Tiae constants utilized in the lag coapensator for hT,"~~~ h h h ~ ~ '"r .'r r.,' ~ hT Kq Kg ~1+ r S Xgp t3 Indicated hT at RATEO THERNAl. POMER~ 1.095; 0.0107/ FI g+R7bey~se 4i < => ~S~. The function generated by the lead-lag coapensator for T dynaaic coapensation; Tiae constants utilized in the lead-lag coapensator for T, xq = 25s, f3 = 3 sl Average teaperature, 4F; 'ag coapensator on aeasured Tav Tiae constant utilized in the aeasured T lag coapensator, ~~ avg 57I.2 F (Noainal T at RATEO THERNL POMER); ~+gQ ~~k~e=~.SS> 0.000453/psig; Pressurizer

pressure, psig;

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) wf I TABLE TA-l Continued TABLE KOTATIOKS Continue HOTE OVERPOWER hT yW e gf ~ ~ Where: dT As defined in Hote 1, ~e = As defined in Note 1, .1 < AT (K - K ~ T K (T T"] - f (Ai)) (1 + ssS) o ~ s (1 + ssS) (1 + teS) e (1 + seS) r 5 ~+so Ts 1 + ty As defined in Hote 1, As defined in Hote 1, 1.09, 0.02/ F for increasing average ter))perature and 0 for decreasing average teaperature, The function generated by the rate-lag coapensator for.T dynaIic coapensation, Tiae constants utilized 'in the rate-lag col)pensator for Tav ts 10 s, As defined in Hote 1, As defined in Kote 1,

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ts N IS Nl ls IlVl I -{@ NIS ~ aa IINIgIQCC INININP N NN ISA V NlINCAN N Iasaa Ial. 'NNP NN NNN+I t N NN NIIae. KICK L I I VS II aw aas Na Kll KSCN ~I ISN c5 lk 5 8 aaacee aeas ea 1 N alaaaaaae p v we aa lail INaa NI INK ~4 eaN NIICI ~NN I NNIN4 PMIMC ~WICN IC NN S Nl W Caela NIICIQ e. ~ I I PICK Kt I 141'N ~41 N ~ ~I CSIN -I '- aw h w El Nl Iat ~ Cs IWK aal PN Nsa NMeMC ttNI144 Na aa aassMIN NN'4 alaaaa tl cheat ~ 44 a'NN aa ICC ICV4 ~SCASIC ISSOR SSSIIN PRCSSSNC JPSCSSQCNI SN ~L5!LRI ASNCCICICICR CSCS INCCSKCCfl1 ACICSNR ZItJaLSCLEJILRL SIAIVS tSSV CCIS rRIINRS ACICNCO XSILLRILLCEELILRL le ea la Nhl N NCI NIm sv sl al h K N:eak NN IS Ia skat, saam ~C eal ~Nt ~ ~NINAI aas ea tNC ~Naala IMetah NKC w 444 N KNP w aslp cd laa SNICC P IP ~lANN INIca cs PN )S Nl WP Cheka NNI ~I'eaa ISV 4 KNN w Ncpackl NNcs W IPI VIKP W ~ We Sa aa N aas ~ N ~p sh aap P IKC N NP ea N ep IN ~PC IIPC IP IPI INC N saaN ~ NI WI~ KSNN VNNW ~4 WC NINa NIP ~Nlaa ~P 'IN NNIP NNIIW N ~Is WQI N~ ICMP NNCI ~ ItlWwat a INN SNI tsa NI ~el tIIC ~Catleks PINS VS I Icc Iael la avl wa NM Ia eca Naa~ ) Nt IN NS ea IN Iee K CVN Stkal NNINSP Ksats IMIIIIWIa Naa WNCIIP aaal NCVNlaal lelta NC NI ICNINNI&tN h SN ~ slal 444 IN aaek ~4 Plate Ate%I 4 44 4 N taael MNNSNCNNIP Waa leal 14 N ~salsa QWNNL NlCkt IIC Nehsekea ~ Ie SSlV.S ASCII Rtt ONO CCICS1 CSOOICA SII I IRIV SS fLORIDAPOWER 6 LIGHTCOMPANY TURKEYPOINT PLANT UNITS 3 4 4 PRESSURIZER PAESSUAE PROTECTION 6 OVERPRESSURE MITIGATIONSYSTEM CONTROL SYST'M DIAGRAM FIGURE 7.2 1 la

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~NCNSC ~NW % SNSIIl %W D NlD ~Nt IDINS +E~ a~ <<1~ I "SVEP+I-N N~ -~sw~ N aCIVINNISN @Zany& spl CS IINIVI DN IW 551 tOIN15 CCZ t5555CNI t514 IN PCN SCP C IC-EE I ~OE Nl PW DP 0 ~C ED lhl I SN DN S SN NN D W PI N Pt IDION + I CPN IDION Sl SESSN PIECED Ac hdst 5 - ~ca rvst ~EN INC PCI SS ECICN WS PEN VEI N PCI Daa EIENW Ps DNE POSE WEt PEDI Pl CEDE P. WS WNt EESN D Sl IEDN<<W EIE N PD IStl

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~NC PCC N ONNN ~fhlE I NIDNCWNNNN15lWOOIDINICVNNNCI WW ra ID SE NII+IIW ~ PD NE I IWWICCPDE SONIC ISI NN W PD QEID CC N EEI ICIID P NWa ~E ~ EW WD N IE DCDIDESNE ~ wl VEINS DECI II.PEN 5 IWWIWICD EID CWWIDW EW Sl SD INN N IVIIVI WNK PSPESSJC EIWIISVIIDEN EIEDISt5 Nl I+ N IEO E NE ~ ~S SNIWC EIDCIIS Sa VVS IDEPIIS D AlCTCS: 5+I AC,gn,l~/ ~P< I ifiCt I I yn 0 4 D v VIV.~ ISNSI +~~LCC. fLOSIDAPOWER 5 LIGHTCOMPANY TURKEYPOINT PLANTUNITS S a 4 PAESSVRI 2ER PRESSV RE CONTROL CONTROL SYSTEM 0lAGRAM FIGURE 72-lib Est ONCCI ICIDTCSCNICS IIIV 11

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I '~/ '~l / AIrkrlI~ j RWIQ4~ rl k ~ I rl k kl1 ~I kec IlrI ~ I ~ I 10 0klrlIkrrl r wrlrr '4 krlrk ~klk Irl ~0 IM ~RBLJ!ERS!K 011~ ~%I~ krLSlc1k w 0 0 ~4 01 IV 01 IH klflak MI SI 4 0110 rI ~I k kI 01 ~r CJ 01 irk ATV.~ IIIIII FLORIDA POWER 5 LIGHT COMPANY TURKEYPOINT PLANTUNITS 3 El 4 000 OW4l 0010 T.LINL 20 (IICV.0I PRIMARYCOOLING SYSTEM REACTOR TRIP 4 TAVGINTERLOCK LOGIC DIAGRAM FIGURE 7280

k 1 t ~ ~ ~ ' k f

as~ r olas~ r soac c Iscos a vva Iss ~ v Qs ~so r OUS~ 1 aal C r aas I I soac~ ralg ral~ras~ ~ AUes le ~ I Ol OS ~ I ~ I O ~ %S ~ I ~I 4I Ol ~I ~ 0 ~I ~ Sk ~ I ~I ~l ~ I IC ~S O ~ IC Ol IC ~I Ufo~ Sl V NS ASIII AN UN SSU a Otal Ue ~ O so I5 Ue so U as U~ gs ICO IO ~V~11 ae ] I~ICOSI ASS W ~I SU US.AI ~IA OlUe IASN UASU IW IIISSOIN NOUN Us U IW I SOOSI~ rNial~ s Nial~ VVIIO~I VVV'~I VVSO~ IO ICO~ g OS%i ASUa ~ I ~ I O'I OI Ol Ol ~I ~ I Ol 4a Jas Lse Ul 4s 4s 4a 4e os Sl S 'S UIC4 %1 NSI Ieee se ~ SI ~ ANIISAI ~ ANIN~ Il Iee ~Ie Ie AN a I+AN AEV.~ Ishsl QR. %PEfornlA vW v->-> IIE~ ISSSOI 0010T LS SSL 'l0 ISKV.SI FLORIDA POWER 4 LIGHTCOMPANY TVRKEYPOINt PLANTVNITS 5 44 STEAM GENERATOR CAUSED REACTOR TRIP ANDSAFETY INJECTION SIGNALS LOGIC DIAGRAM F!QVRE 744b

0 il'

CONTROL ROOM VENTILATION ISOLATION Recirculation Phase In the unlikely event of an MHA the automatically be placed in a recirculation fans and dampers to the cable spreading areas will automatically close. control room air conditioning will phase. The fresh air dampers, exhaust

room, locker and toilet, and shower zA($EA'7 g The following provisions are made to reduce unfiltered infiltration and pressurize the control room.'.

Automatic start, opening or cloying signal, as required for emergency for SP-1 and dampers D-l, D-2, D-3 and D-ll. 2. Temporary motor operated isolation dampers at fans EP-8 and EF-9, including automatic closure signal to the damper D-14 to EF-9 and damper D-22 to EF-20 ~ 3. Two volume control dampers, one in the emergency makeup supply D-20 and one in the recirculation duct D-21 to SP-l. 4. Supply and return ducts added to the mechanical equipment room. A controlled air leakage type door is installed. 5. Sealed opening in mechanical equipment room doors and walls, in Control Cabinet Rooms 212 and 213, in duct seams on the suction and supply side of Exhaust Fan EF-7, and in roof, floor and wall penetrations.

9. 9-2 Rev. 3-7/85

e

Insert A "The recirculation phase is automatically actuated in the event of a safety injection signal, or high containment Radiation Monitor signal (R-ll, R-12), or Air Intake Radiation level signal."

I y

ATTACHMENT 2 CONTAINMENT ISOLATION VALVE CLOSURE TIMES

C 0 t'

TABLE 6.6-3 CONTAINMENT ISOLATION VALVE CLOSURE TIMES Phase "A" Isolation Valves VALVE NUMBER PENETR ~Nl ~IIUM ~ETE N ll CLOSURE TIME SECONDS CV-*-200A CV-*-200B CV-*-200C CV-*-204 HOV-*-381 CV-*-516 CV-*-519A CV-*-739 CV-*-855 CV-*-956A CV-*-956B CV-*-956D HOV-*-1417 HOV-*-1418 CV-*-2821 CV-~-2822 SV-*-2911 SV-*-2912 SV-*-2913 CV-*-4658A CV-*-4658B CV-*-4659A CV-*-4659B CV-*-4668A CV-*-4668B CV-*-6165 SV-*-6385 MOV-*-6386 SV-*-6428 14 14 14 14 25 5 7 13 42 8 9 55 21 22 23 23 33 32, 33 10 10 31 31 52 52 30 5 25 20 CVCS Normal Letdown CVCS Normal Letdown CVCS Normal Letdown CVCS Normal Letdown CVCS Excess Letdown/RCP Seal Water Return PRT Gas Analyzer Sample PRT Makeup Primary Water Supply CCW Return from Excess Letdown HX Nitrogen Supply to Accumulators Pressurizer Steam Space Sample Pressurizer Liquid Space Sample Accumulator Sample CCW Supply to Normal CTHT Coolers CCW Return from Normal CTHT Coolers Containment Sump Discharge Containment Sump Discharge Containment Air Sample Containment Air Sample Return Containment Air Sample RC Drain Tank Vent RC Drain Tank Vent RC Drain Tank Gas Analyzer Sample RC Drain Tank Gas Analyzer Sample RC Drain Tank Pump Discharge RC Drain Tank Pump Discharge Breathing Air Supply PRT Gas Analyzer CVCS Excess Letdown/RCP Seal Water Return RCS Hot Leg Sample Note 2 Note 2 Note 2 Note 2 Note 2 Note 2 Note 2 Notes 3, 5 Note 2 Note 2 Note 2 Note 2 Notes 2, 5 Notes 2, 5 Note 2 Note 2 Note 2 Note 2 Note 2 Note 2 Note 2 Note 2 Note 2 Note 2 Note 2 Note 4 Note 2 Note 2 Note 2

r V J

TABLE 6.6-3 (continued) CONTAINMENT ISOLATION VALVE CLOSURE TINES B Phase "B" Isolation Valves VALVE NUMBER Note 1 PENETR N I ~ITEN UNCT CLOSURE TIME SECONDS MOY-*-626 MOV-*-716A MOV-*-716B MOV-*-730 43 CCW Return from RCP Thermal Barrier CLRS CCW Supply to RCPs CCW Supply to RCPs CCW Return from RCP Oil Coolers Notes 2, 5 Notes 2, 5 Notes 2, 5 Notes 2, 5 C Containment Ventilation Isolation Valves VALVE NUMBER Note 1 PENETR NIINNEII ~SYETEM EUNCTI N CLOSURE TIME SECONDS POV-*-2600 POV-*-2601 POV-*-2602 POV-*-2603 CV-*-2819 CV-*-2826 35 35 36 36 63 63 Containment Purge Supply Containment Purge Supply Containment Purge Exhaust Containment Purge Exhaust Instrument Air Bleed Instrument Air Bleed 5 5 5 5 Note 2 Note 2 NOTES: 2. 3. 4. 5.

  • = Unit number (3 or 4)

The isolation times of each automatic valve shall be within the limits established for testing in accordance with Section XI of ASME Boiler and Pressure Vessel code and applicable Addenda as required by 10 CFR 50.55 a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i). Testing requirements are per letter L-89-358 from J. H. Goldberg to U. S. Nuclear Regulatory Commission dated October 3, 1989. CV-*-6165 is locked closed, and is not required to be stroke time tested. Valve is not subject to Type C local leak rate testing of 10CFR50, Appendix J.

T ,h7 t ~

P eveqt of a loss-of-coolant accident. Isolation of a line inside the containment prevents flow from the reactor coolant system or any other large source of radioactive fluid in the event th'at a piping rupture outside the containment occurs. A piping rupture outside the containment, at the same time as a loss-of-coolant accident occurs, is not considered credible. yacc Pung% spective isolation signal, the requirements of integrity at peak containment pressure and temperature for the double-ended coolant line break are met. KAscR7 6 Containment isolation becomes mandatory under the same conditions that require operation of the other engineeged safeguards. The containment isolation signal is derived from the same signals which automatically acti-vate safety in5ection. the containment have been designed to assure that they are capable of with-standing the maximum hypothetical earthquake. To assure their adequacy in this respect: (a) Valves are located in a manner to reduce the accelerations on the valves. Piping spans have been designed for adequacy of the loads to which the span would be subjected. Valves are mounted in the position recommended by the manufacturer. (b) Earthquake forces on the operating parts of the valve are calculated to be small compared to the other forces present in the piping system. 6.6-2

1 ~

  • O t S.

0 ~bi

4 ~ ~ (c) Control cables to the valve operators are designed and installed to assure that the flexure of the line does not endanger the control system. Appendages to the valve, such as position indicators and operators, are designed for structual adequacy. Containment Isolation Valve Criterion Vcnc~gr ) 4 rc ufr~ closer~ 4r Wc. c<n&ln~c<k u c0:orle, ore. pmkr~4<A p'v44e at least two barriers for redundancy against leakage of radioactive fluids to the environment in the event of a loss-of-coolant accident. These

barriers, in the form of isolation valves or closed
systems, are defined on an individual line basis.

rtISeg7 With respect to numbers and locations of isolation valves, the criteria applied are generally those outlined by the five categories described herein. 6.6.2 SYSTEM DESIGN The five general categories listed below, describe the methods by which lines penetrating containment may be classified. Also described are the basic isolation valve arrangements used to provide two barriers between the Reactor fho5C 'pCfjedre 0 s'o AS Coolant System or containment atmosphere and the environment ~ Soystem design rc uric.d 4~ ~ j, caAhai~~enf 4'unct;o~, AC ~is such that failure of one valve to close does not prevent isolation, and no manual operation is required for immediate isolation. 6.6-3

t A 91 1 ~ V; I t" li 2",g

The end of the fuel transfer tube inside containment is closed by a e double-gasketed blind flange, to prevent leakage of spent fuel pit water into the containment during operation. This flange also serves as protection against leakage from the containment following a loss of coolant accident. The space between these gaskets can also be pressurized by the penetration test system.

Also, as a special
case, because of size, the containment purge supply and exhaust ducts are each equipped with two tight-sealing butterfly valves, one inside containment and one outside.

As discussed in Subsection 9.8.2, the opening angle of the valves has been limited. These 'valves can be closed either manually or automatically upon a signal of high radioactivity level in the containment or by the containment isolation signal. These valves are normally closed during reactor power operation. See Table 6.6-2 for a single failure analysis of the containment purge valves. Debris screens are present upstream {with respect to Post LOCA flow) of the supply and exhaust purge valves. These debris screens will prptect the containment isolation valves from debris that might follow as a result of a LOCA. 6.6.3 ISOLATION VALVES AND Figure 6.6-1 shows all valve or to closed systems on both defines the nomenclature and INSTRUMENTATION DIAGRAMS arrangements in lines leading to the atmosphere sides of the containment barrier. Figure 6.6-2 symbols used. Valve Parameters Tabulation A summary of the fluid systems lines penetrating containment is presented in Table 6.6-1. Each valve is described as to type, operator, position indication and open or closed status during normal operation, shutdown and accident conditions. Information is also precented on valve preferential failure mode and automatic trip by the containment isolation signal. 6.6-6 Rev 4 7/86

P 'gvw g ld IN '

Containment isolation valves are provided with actuation and control equip-ment appropriate to the valve type. For example, air operated globe and diaphragm (Saunders Patent) valves are generally equipped with air diaphragm operators, with fail-safe operation ensured by redundant control devices in the instrument air supply to the valves. Motor operated gate valves are capable of being supplied from reliable on-site emergency power as well as their normal power source. Manual and check valves, of course, do not require actuation by control systems. 1 Containment isolation "trip" valves are actuated to the closed position by the containment isolation signal, derived automatically from the signals as 4.c -3 listed in Table ~h. Non-automatic isolation valves, ie., remote stop valves and manual valves, are used in lines which must remain in service, at least for a time, following a MHA. These are closed manually if and when the lines are taken out of service. The large butterfly valves used to i,solate the containment purge ventilation ducts are equipped with air-cylinder operators, with spring returns. These valves fail to the closed position on loss of control signal or instrument air. 6.6-7 Rev 4 7/86 I

roc 11 t)

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e'" VALVE OPERABILITY dhrhjctlsolation valves located inside containment are subject to the high

pressure, high temperature, steam laden atmosphere resulting from an acci-dent.

Operability of these valves in the accident environment is ensured by proper design, construction and installation, as reflected by the following considerations: (a) All components in the valve instal&tion, including valve bodies, trim and moving parts, actuators, instrument air lines and control and power wiring, are constructed of materials sufficiently temperature resistant to be unaffected by the accident environment. Special attention is given to electrical insulation, air operator diaphragms and stem packing material. (b) In addition to normal pressures, the valves are designed to withstand maximum pressure differentials in the reverse direction imposed by the accident conditions. This criterion is particularly applicable to the butterfly type isolation valves used in the containment purge lines.

6. 6-8

>A

Insert A "Closure times for other automatic isolation valves in lines that provide a direct pathway between the containment atmosphere and the environs are selected to minimize offsite radiological consequences and to assure that ECCS effectiveness is not degraded by a reduction in the containment backpressure following a LOCA." INSERT B "Table 6.6-3 provides a listing of required isolation valves that actuate on a Phase A, Phase B, or Containment Ventilation Isolation Signal. Automatic closure of the valves listed in Table 6.6-3 will achieve containment isolation in accordance with the requirements of GDC 53 and the plant Technical Specification. Additional automatic valves that receive an ESFA's signal for other than containment isolation purposes are not included in Table 6.6-3." Valves that provide a containment integrity function, but which are not isolated automatically (e.g., manual

valves, check valves, remote manual valves) are also excluded from Table: 6.6-3.

INSERT C Phase A Isolation is initiated by a Safety Injection signal. Phase B Isolation is initiated by Containment Pressure High-High coincident with Containment Pressure High. Containment Ventilation Isolation is initiated by Safety Injection or high Containment Radioactivity (R-11 or R12) signals. INSERT D "Table 6.6-3 lists closure times for containment isolation valves that receive an automatic isolation signal (Phase A, Phase B or Containment Ventilation Isolation). Valve capability to meet the required closure times is verified periodically during performance of the In-Service Testing (IST) Program. Operability requirements for these automatic isolation valves are as provided in the Technical Specifications."

I 0

ATTACHMENT 3 AUXILIARYFEEDWATER ACTUATION PARAMETERS

II 1F l 0 gt,i E'

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9'11.2 AUXILQNY FEEDWATER PUMPS t Three, quick starting steam turbine driven, auxiliary feedwater pumps are provided for Turkey Point Units 3 and 4. Each pump is capable of delivering 600 gpm to the steam generators between 1085 psig at 5900 rpm and 120 psig at 3200 rpm. The three pumps are installed such that each supplies auxiliary feedwater to either Unit 3 or 4, with any single pump supplying the total feedwater requirement of either unit. Two pumps (B6C) are normally aligned to AFW Train 2 and the third (A) is normally aligned to AFW Train l. The turbine driven pumps are supplied with steam from the unit which has lost its normal feedwater supply. RPM indicators are provided locally and in the control room to provide indication that the AFW pump/turbine is running. The turbines have an atmospheric exhaust. Steam can also be supplied from the unit having normal feedwater supply or from an auxiliary steam connection to Units 1 and 2. The supply valves will automatically open by any one of the following four signals. 1. Safety Injection 2. Low"Low Level in any of the three steam generators. Loss of both feedwater pumps under normal operating conditions-K'AlscAy' The turb ne casing is provided with a sentinel type relief valve for warning purposes only. Impulse type steam traps are provided upstream of the MOVs and drain to the condenser. The turbine casing drains, the exhaust pipe drain, the gland seal

drain, the governor valve and the HP and LP steam leakoffs in the throttle trip valve drain to a drain trough.

The pump recirculation is controlled by an 'rifice in the recirculation piping.'he pumps continue to supply reduced amounts of water to the steam generators until steam pressure is reduced to 120 psig. The pump output in pounds per hour is greater than the steam consumption until the 120 psig point is reached. At this point the Residual Heat Removal System is started and the auxiliary feedwater pumps are shutdown. 117F/ 9.11-2 Rev 7 7/89

I ll I' e 0

JPN-PTN-SENJ-89-091 Revision 1 Insert A 4. Bus Stripping a ~ b. c ~ Loss of voltage on either the A or B 4. 16 KV bus. Degraded voltage on one 480V load center (instantaneous) coincident with safety injection and the diesel generator breaker open. Degraded voltage on one 480V load center (delayed) coincident with the diesel generator breaker open.

E 'l 0

ABLE 7.2-1 Sheet 4 o STEAN LINES ISOLATION ACTUATION COlNCIDENCE CIRCUITRY AND INTERLOCKS COHHENTS 22. Steam Flow 23. Containment pressure High steam line flow in 2 out of 3 loops coincident with either low T in 2 out of 3 loops or low steam land pressure in 2 out of 3 loops. 2/3 high containment pressure signal in coincidence with 2/3 high-high containment pressure 24; Hanual per steam loop AUXILIARYPEEDQhTER hCTUATION 1/I per steam line 25. Turbine driven pumps Coincidence of 2/3 low-low level in any steam generator; or ~us St c sf pin vcR~eees; or auto trip of main feed water pumps, or safety injection signal; or manual 1/2 HAIN FEEDMATER ISOLATION 26A. Close main feedwater control valves (fast closure) Actuated by: 1 Safety injection (see f10) 2. 2/3 high-high feedwater level (80X) in steam generator 3. Reactor trip coincident with low T (slow closure) avg 26B Close bypass feedwater control valves 27. Trip steam generator feed pumps / 28. Turbine and main feedwater pumps trip 1 Safety injection or high-high level in steam generators Safety ingectxon signal Coincidence of 2(3 high-high level (80X) in any steam generator Rev. 3-7/85

t ~ ~

ATTACHMENT 4 SELECTED RTS SETPOINTS AND PARAMETERS

0

Page 1 ADDITIONAL INFORMATION ON REVISED TECHNICAL SPECIFICATIONS PARAMETERS RTS Table 2.2-1 PARAMETER: Intermediate Range Neutron Flux DOCUMENT/SECTION/VALUE: RTS: Table 2.2-1 < 254 RTP CTS N A DISCUSSION: This trip provides core protection during reactor startup to mitigate the consequences of an uncontrolled rod cluster control assembly bank withdrawal from a subcritical condition. This value was provided by Westinghouse, and is the STS Iigeneric" value for this function. This value is applicable to Turkey Point. RTS Table 2.2-1 PARAMETER: Source Range Neutron Flux DOCUMENT/SECTION/VALUE: RTS: Table 2.2-1 10 CPS CTS N A DISCUSSION: This trip provides core protection during reactor startup to mitigate the consequences of an uncontrolled rod cluster control assembly bank withdrawal from a subcritical condition. This value was provided by Westinghouse, and is the STS "generic" value for this function. This value is applicable to Turkey Point.

RTS Table 2.2-1 Page 2 PARAMETER: Reactor Coolant Flow-Low DOCUMENT/SECTION/VALUE: RTS: Table 2.2-1 CTS: 2.3 90~ of loop design flow 90% of normal indicated flow DISCUSSION: The RTS value is consistent with CTS 3.1.6.C, which requires Reactor Coolant flow of 268,500 gpm during power operation. This is equivalent to 89,500 gpm/loop, which is the RTS value for loop design flow. RTS Table 2.2-1 PARAMETER: RTS Interlock, Intermediate Range Neutron Flux, P-6 DOCUMENT/SECTION/VALUE: RTS: Table 2.2-1 CTS: N A > 1 x 10 amp DISCUSSION: On increasing power P-6 allows the manual block of the Source Range trip (i.e., prevents premature block of Source Range trip) and de-energizes the high voltage to the detectors. On decreasing

power, Source Range Level trips are automatically reactivated and high voltage restored.

(RTS Bases) The setpoint of 1 x 10 amp provides a transition point between the Source Range and Intermediate Range. It is a STS "generic" value provided by Westinghouse, and is typical of many other Westinghouse plants. This value is applicable to Turkey Point.

t g) W4

...RTS Table 2.2-1 Page 3 PARAMETER: RTS Interlock, Low Power Reactor Trip Block, P-7 DOCUMENT/SECTION/VALUE: RTS: Table 2.2-1 CTS: 2.3 10%'TP (P-10 Input) 10% Turbine Power (Turbine First Stage Pressure) 10< RTP DISCUSSION: RTS/CTS are equivalent for pressurizer low pressure, pressurizer high level and low flow in more than'ne reactor coolant loop. The RTS bases also specifies enables for two or more RCP breakers

open, RCP bus undervoltage, and turbine trip.

The RTS turbine power setpoint is an STS "generic" value supplied by Westinghouse, and is typical of many other Westinghouse plants. The generic interlock setpoint is applicable to Turkey Point because the Turkey Point reactor protection system setpoints are similar to other Westinghouse plants. PARAMETER: Trip delay and block on undervoltage trip (P-7) DOCUMENT/SECTION/VALUE: RTS: B.2.2.1 CTS: N A < 1.3 second delay from 10% power with 1st stage turbine pressure at 104 =DISCUSSION: The RTS 1.3 second delay was verified by Westinghouse to be consistent with the Turkey..Point Safety Analysis. The. FSAR value of 1.6 seconds includes the time for reactor trip breaker to open and the rod to release.

I ~' I

~ ~ RTS 3.1.1.4 Page 4 PARAMETER: RCS Lowest Tavg DOCUMENT/SECTION/VALUE: RTS: 3.1.1.4 ~CTS: N A RCS lowest Tavg > 541'F DISCUSSION: The value of 541'F was provided by Westinghouse as the appropriate value. This limit ensures that (1) the moderator temperature coefficient is within its analyzed temperature rangei (2) the trip instrumentation is within its normal operating range; (3) the pressurizer is capable of being in an OPERABLE status with a steam bubble; and (4) the reactor vessel is above its minimum RTNDT temperature (RTS BASES). RTS 3.1.2.4.a.l PARAMETER: BAST Volume, Modes 5 and 6 DOCUMENT/SECTION/VALUE: ~CTS: N A BAST indicated vol. of 500 gal in Modes 5 and 6 DISCUSSION: This BAST volume for the RTS was provided by Westinghouse. The BAST volume compensates for RCS contraction and temperature induced reactivity due to cooldown from 200'F to ambient.

I ~ ~ i rakM

.,RTS 3.1.2.4. .2 PARAMETER: BAST Boron Concentration, Modes 5 and 6 DOCUMENT/SECTION/VALUE: RTS: 3.1.2.4.a.2 ~CTS: N A BAST boron con. between 20,000 and 22,500 ppm Page 5 DISCUSSION: The BAST Boron Concentration in Modes 5 and 6 is ,identical to the CTS BAST Boron Concentration required by CTS 3.6.b.3 for Modes 1 and 2. The min and max values are used in the safety analysis for determining shutdown margin and to ensure solubility. The tech spec max and min values are consistent with the,.analysis values. RTS 3.1.2.4.b.1 PARAMETER: RWST Volume, Modes 5 and 6 DOCUMENT/SECTION/VALUE: ~ ~ ~ ~ ~ ~ RTS: 3.1.2.4.b.1 > 20,000 gal. CTS N A DISCUSSION: This RWST volume for the RTS was provided by Westinghouse. The RWST volume compensates for RCS contraction-and temperature induced reactivity due. to cooldown from 200'F to ambient. RTS 3.1.2.4.b.2 PARAMETER: RWST Boron Con., Modes 5 and 6 DOCUMENT/SECTION/VALUE: RTS: 3.1.2.4.b.2 CTS: N A > 1950 ppm DISCUSSION: The RWST Boron Concentration in Modes 5 and 6 is identical to the CTS RWST Boron Concentration in CTS 3.6.b.3 for Modes 1 and 2.

4 S sC 1 br l

RTS 3.1.2.5.b 3 & 4 ~ ~ ~ Page 6 PARAMETER: RWST Temp. DOCUMENT/SECTION/VALUE: RTS: 3.1.2.5.b.3 & 4 39 F 100 F CTS N A DISCUSSION: These values are assumed in the containment integrity and large break LOCA analyses. RTS 3.1.3.2 PARAMETER: Group demand position for Control Rods (Operating) DOCUMENT/SECTION/VALUE: RTS: 3.1.3.2 CTS N A + 2 steps DISCUSSION: This is the accuracy of the setpoint for rod position upon an operator command. The basis for this tolerance is from the Westinghouse Vendor Manual, "Rod Position Indication System". The accuracy associated with the control band output voltage is +14 of full scale voltage span. The integer of 1% full scale is +2 steps. RTS 3.1.3.3 PARAMETER: Group demand position for Control Rods (Shutdown) DOCUMENT/SECTION/VALUE: RTS: 3.1.3.3 CTS: N A + 2 steps DISCUSSION: This is the accuracy of the setpoint for rod position upon an operator command. The basis for this tolerance -is. from the Westinghouse Vendor Manual, "Rod Position Indication System". The accuracy associated with the control band output voltage is +1: of full scale voltage span. The integer of 1% full scale is +2 steps.

0

RTS 3.2.4 PARAMETER: Quadrant Power Tilt Ratio (QPTR) DOCUMENT/SECTION/VALUE: Page 7 RTS: 3.2.4 CTS: 3.2.6.h 1.02 1.02 and < 1.09 1.09 < 24 > 24 and < 104 > 104 DISCUSSION: The safety analysis for Turkey Point 3 & 4 assumes a QPTR of 2% or 1.02 at the start of a DBA. The RTS LCO and action statements are more conservative. RTS 3.2.5 PARAMETER: DNB Limits: Tavg Pressurizer

Pressure, RCS flow DOCUMENT/SECTION/VALUE:

RTS: 3.2.5 CTS: 3.1.6 Tavg < 576.6'F Pressurizer Pressure 2209 psig RCS flow 277,900 gpm Tavg < 578.2'F Pressurizer Pressure > 2220 psia RCS flow 268,500 gpm DISCUSSION: These values are more conservative than the CTS values. These values represent the allowable values with greater margin from the respective Safety Analysis limits. In STS Rev. 5, the DNB parameters are defined as indicated values, implying an allowance for instrument error. Westinghouse provided these numbers.

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RTS Table 3.3-3 Page 8 PARAMETER: Loss of Power 480 Load Centers Instantaneous Relays Degraded Voltage DOCUMENT/SECTION/VALUE: 'RTS: Table 3.3-3 CTS: Table 3.5-4 3A 436V + 5v 3B 416V + 5v 3C 417V + 5v 3D 428V + 5v 4A 415V + 5v 4B 414V + 5v 4C 401V + 5v 4D 403V + 5v (10 sec + 1 (10 sec + 1 (10 sec + 1 (10 sec + 1 (10 sec + 1 (10 sec + 1 (10 sec + 1 (10 sec + 1 sec delay) sec delay) sec delay) sec delay) sec delay) sec delay) sec delay) sec delay) 3A 436V (10 sec delay) 3B 416V (10 sec delay) 3C 417V (10 sec delay) 3D 428V (10 sec delay) 4A 415V (10 sec delay) 4B 414V (10 sec delay) 4C 401V (10 sec delay) 4D 403V (10 sec delay) all with tolerance of + 5v DISCUSSION: The CTS calls out a 10 sec. delay. The RTS adds an explicit tolerance limit of + 1 sec. to the 10 sec. delay. The repeat accuracy of the setting is 154, therefore a tolerance of + 1 sec. has been added as representative. The analytical requirements for these relays is a delay time of > 7 seconds and < 20 seconds. RTS Table 3.3-3 PARAMETER: Control Room Air Intake Radiation Level DOCUMENT/SECTION/VALUE: 2 mR/hr (Setpoint) < 2.83 mR/hr (Allowable) DISCUSSION: This parameter is from the Radiation Monitoring Instrumentation Table in the STS. This value applies to Turkey Point and is the specified setpoint per the modification which installed this system." The allowable - value is based on the measured instrument accuracy/drift.

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Page 9 RTS 3.3.3.1 PARAMETER: SFP Unit 3 High Gaseous DOCUMENT/SECTION/VALUE: RTS: Table 3.3-4 CTS N A 5 5 x 10 uCi/cc DISCUSSION: The SFP alarm setpoints were obtained from the ODCM (Off-Site Dose Calculation Manual). RTS: 3.3.3.1 PARAMETER: SFP Unit 4 High Gaseous DOCUMENT/SECTION/VALUE: RTS: Table 3.3-4 CTS N A 2 ' x 10 e uCi/cc or 10 cpm DISCUSSION: The SFP alarm setpoints were obtained from the ODCM-(Off-Site Dose Calculation Manual).

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4 1 Page 10 RTS 3.3.3.2 ~ ~ PARAMETER: g Detector Thimbles DOCUMENT/SECTION/VALUE: RTS: 3.3.3.2 CTS: 3.2.7 16 to calibrate Excore Neutron Flux and Monitor Quadrant Tilt

Ratio, 2 per quadrant; 38 to monitor FN+Hg Fq(Z) f F~ (Z) 16 thimbles, 2

per quadrant DISCUSSION: The RTS requirement (38 operable thimbles) for F< and FNiH surveillance is the generic STS limit (i.e., 754 of total thimbles) and is related to the generic F< and FNiH surveillance uncertainty in Spec 3/4.2.2 and 3/4.2.3 (i.e., 54 on F< and 44 on FNiH). In the

CTS, the same generic F< and FNiH uncertainties apply with no corresponding thimble OPERABILITY requirement.

RTS 3.4.1.3 PARAMETER: Core Outlet Temp. Below Saturation (Mode 4) DOCUMENT/SECTION/VALUE: RTS: 3.4.1.3* CTS: N A 10'F below saturation DISCUSSION: This tech spec ensures that the primary coolant remains in the liquid phase. Any amount of subcooling will meet this requirement. 10'F is an STS

value, which provides an allowance for instrument uncertainty.

This is a generic Westinghouse value that.is applicable to Turkey Point.

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,RTS 3.4.1.4.1* ~ ~ ~ ~ 3.4.1.4.2** Page 11 PARAMETER: Core Outlet Temp Below Saturation (Mode 5) DOCUMENT/SECTION/VALUE: RTS: 3.4.1.4.1* (loops filled) 3.4.1.4.2** loo s not filled CTS: N A > 10'F below saturation DISCUSSION: This tech spec ensures that the primary coolant remains in the liquid phase. Any amount of subcooling will meet 'this requirement. 10'F is an STS

value, which provides an

~allowance for instrument uncertainty. This is a generic Westinghouse value that is applicable to Turkey Point. RTS 3.4.3 PARAMETER: Pressurizer Heater Capacity DOCUMENT/SECTION/VALUE: RTS: 3.4.3 CTS: 3.1.1.d 2 Groups, 1 group > 125 KW > 125 KW each if RCS > 350'F. DISCUSSION: The requirement for 2 groups is generic STS. 2 groups provides redundancy of function to improve system reliability relative to single failures and it is more restrictive than CTS.

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RTS 3.5.1.d Page 12 PARAMETER: Accumulator Nitrogen Pressure DOCUMENT/SECTION/VALUE: RTS: 3.5.1.d CTS: 3.4.1.a.3 3.1.9-2 600-675 psig > 600 psig DISCUSSION: The upper limit is based on the nitrogen cover gas pressure regulator. In an accident

analysis, the minimum accumulator pressure is limiting.

No accident analysis assumption is sensitive to the maximum pressure. The regulator setpoint prevents pressure from exceeding the FSAR design limit of 700 psig. The "normal" pressure listed in the FSAR is a nominal value, and has no safety significance within the RTS range of 600-675 psig. The RTS limit on max. pressure is consistent with the generic STS format which includes minimum and maximum accumulator pressure values. RTS 4.5.2.b.3 PARAMETER: RHR Pump Parameters DOCUMENT/SECTION/VALUE: RTS: 4.5.2.b.3 CTS: 4.5.2.a.2 Must meet Fig. 3.5-1 Pump Curve Pumps shall start and reach required head for normal or recirc. flow, whichever is greater. DISCUSSION: These values are from a safety evaluation performed per 10 CFR 50.59 to determine pump flows. The values for pump heads/flows were provided by Westinghouse and are acceptable with regard to the applicable accident analyses.

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RTS 4.5.2.c.l PARAMETER: SI Pump Parameters DOCUMENT/SECTION/VALUE: Page 13 RTS: 4.5.2.c.1 CTS: 4.5.2.a.2 1126 psid 9 > 300 gpm normal alignment or > 1156 psid 9 > 280 gpm Pumps shall start and reach required head for normal or recirc. flow, whichever is greater. DISCUSSION: These values are from FPL Calculation, "ECCS Pump Minimum Static Heads at Various Flows." The input values for pump head/flows were.provided by Westinghouse and are based on flows (at various RCS pressures) assumed in the applicable accident analyses. RTS 3.5.4.c & d PARAMETER: RWST Temp. DOCUMENT/SECTION/VALUE: RTS: 3.5.4.c & d ~CTS: N A 6.2.11-2 39 F 100 DISCUSSION: These values are assumed in the containment integrity and large break LOCA analyses. RTS 3.6.1.3 PARAMETER: Air Lock Leakage DOCUMENT/SECTION/VALUE: RTS: 3.6.1.3.b CTS: 4.4.1& 4.4.2 < 0.05 La at Pa, 49.9 psig Sum of local leak rate tests is < 0.60 La DISCUSSION: CTS 4.4.1 and 4.4.2 require only that the local leak rate tests is < 0.60 La. The RTS value of 0.05 La is an STS value which will allow detection of gross failure of the airlock'eal before the total containment leakage limit is exceeded.

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,.RTS 3. 6. 1. 5 ~ ~ ~ PARAMETER: Containment Air Temp. DOCUMENT/SECTION/VALUE: Page 14 RTS: 3.6.1.5 < 125'F; limited to 336 equivalent hrs/yr above 120'F CTS N A DISCUSSION: The RTS value of 125'F is based on recent FPL analysis performed under the requirements of 10 CFR 50.59. RTS 4.6.1.6.1a&b PARAMETER: Containment Vessel Structural Integrity DOCUMENT/SECTION/VALUE RTS: 4.6.1.6 CTS: 4.4.5 12 random tendons within lift-off limit; 240,000 psi for wire strand samples; Hoop 5 Vertical 4 Dome 3 9 select tendons Horizontal 3 Vertical 3 Dome 3 DISCUSSION: 240, 000 psi is the minimum ultimate strength requirement specified in ASTM A421-65 which is the standard to which the prestressing wires were purchased. 12 tendons were proposed by FPL to comply with a proposed Reg. Guide. It is more conservative than the CTS requirement of 9 tendons. RTS 4.6.1.7.2 PARAMETER: Purge Valves Leakage DOCUMENT/SECTION/VALUE: RTS: 4.6.1.7.2 ~CTS: N A < 0.05 La at Pa DISCUSSION: This limit is an STS generic limit. CTS 4.4.2 only requires that the sum of all local leak rate tests be < 60% of the total allowable leakage (0.60 La at Pa).

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RTS 4.6.2.1.b ~ ~ PARAMETER: Containment Spray Pump Parameters DOCUMENT/SECTION/VALUE: Page 15 RTS: 4.6.2.1.b 241.6 psid in recirc. CTS N A DISCUSSION: These values are from a safety evaluation performed per 10 CFR 50.59 to determine pump flows. The values for pump heads/flows were provided by Westinghouse and are acceptable with regard to the applicable accident analyses. RTS 4.6.3 PARAMETER: Emergency Containment Filters DOCUMENT/SECTION/VALUE: RTS: 4. 6. 3 CTS: 4.7.1.1 and 4.7.1.2.a 37 g 500 + 10% CFMg pressure drop < 6<<H~O. > 99% DOP removal, > 994 halogenated hydrocarbon > 99.94 elemental iodine 37,500 + 104 CFM pressure drop 6 HpOg > 994 DOP removal, > 994 halogenated hydrocarbon > 99.94 elemental iodine DISCUSSION: RTS/CTS values are identical. FSAR removal values are reflective of design capabilities and not design bases requirements. RTS 3.7.1.4 PARAMETER: Secondary Specific. Activity DOCUMENT/SECTION/VALUE: RTS: 3.7.1.4 CTS: 3.8.2 < 0.10 uCi/gm Dose Equiv. I-131 0.67 uCi/gm Dose Equiv. I-131 DISCUSSION: The RTS value is an STS "generic" value, and is more l restrictive than the CTS.

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RTS 3.7.4 PARAMETER: Ultimate Heat Sink DOCUMENT/SECTION/VALUE: Page 16 RTS: 3.7.4 CTS: N A average supply water temp. to ICWS < 100 F DISCUSSION: 100'F in conjunction with CCW/ICW technical specification requirements for heat exchanger cleanliness, ensures capability to remove accident heat loads, as analyzed in a 10 CFR 50.59 evaluation. RTS 4.7.8.1.3 P2GDMETER: Fire Pump Diesel Starting Battery DOCUMENT/SECTION/VALUE: RTS: 4.7.8.1.3 ~CTS: N A Electrolyte level above plates. Battery voltage 24 volts DISCUSSION: Electrolyte level and voltage are STS criteria which ensure battery operability and are applicable to the Turkey Point equipment. RTS 3.9.3 PARAMETER: Decay Time DOCUMENT/SECTION/VALUE: RTS: 3.9.3 CTS: 3.10.5 Reactor shall be subcritical for at least 100 hrs. Reactor shall be subcritical for at least 100 hrs. DISCUSSION: RTS/CTS values are identical. The FSAR value of 96 hours is the design time that shielding design is based on. The RTS/CTS duration is conservative with respect to this design value.

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Page 17 RTS 3.9.6 PARAMETER: Manipulator Crane Capacity DOCUMENT/SECTION/VALUE: RTS: 3.9.6 CTS N A Capacity 2750 lbs cutoff limit < 2700 lbs. DISCUSSION: These limits are STS values, and were provided by Westinghouse. They are applicable to Turkey Point because Turkey Point fuel assemblies are the same as those on which the STS values ,are based. RTS 3.9.6 PARAMETER: Aux. hoist capacity DOCUMENT/SECTION/VALUE: RTS: 3.9.6 ~ ~ CTS N A Capacity 610 lb. load indicator 600 lb. DISCUSSION: These limits are STS values, and were provided by Westinghouse. They are applicable to Turkey Point because Turkey Point drive rods are the same as those on which the STS values are based.

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RTS 4.9.8.1.1 & 4.9.8.2 Page 18 PARAMETER: RHR flow DOCUMENT/SECTION/VALUE: RTS: 4.9.8.1.1 Ei ~CTS: N A 4.9.8.2 > 3000 gpm DISCUSSION: The 3000 gpm RHR flow is a generic STS limit, used in many Westinghouse STS plant specs. The safety analysis basis for the flow is to ensure uniform boron concentration within the RCS. The STS generic limit is applicable to Turkey Point, based on the Turkey Point RHR system design flow rate. The FSAR value represents the pump design flow, not a system requirement. RTS 3.10.3 PARAMETER: Special Test: Power, Intermediate Range & Power Range Setpoints, RCS Lowest Loop Tavg ~ ~ ~ ~ DOCUMENT/SECTION/VALUE: RTS: 3.10.3 CTS: Several 54 RTP < 254 RTP Trip Setpoint 531 Example 3.2.1.a: "except for physics tests" DISCUSSION: This TS allows suspension 'of various TS during low power physics tests as long as the listed parameters are maintained. The CTS physics test exceptions do not explicitly call out compliance with the listed parameters.

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...RTS 4.10.5 P2DGLMETER: Position Indication System Special Test DOCUMENT/SECTION/VALUE: Page 19 RTS: 4.10.5 CTS N A 12 steps-stationary 24 steps-during motions DISCUSSION: This is a generic STS surveillance to ensure position indication system operability during rod drop time measurements. The Rod Position Indication is calibrated to an accuracy of + 5.04 of full scale rod withdrawal, which corresponds to the limit of + 12 steps. During calibration of the rod position indication, the distance between calibration points is approximately 24 steps. The case of a rod being in motion will correspond to the additional uncertainty in position between statically calibrated points. This additional uncertainty will be at its worst at the midpoint between any two such static points. This will add an additional + 12 steps for a total of + 24 steps for total uncertainty. RTS 5.6.1.1.a & b PARAMETER: d-K/K uncertainty in fuel racks DOCUMENT/SECTION/VALUE: RTS: 5.6.1.1.a & b ~CTS: N A Keff = < 0.95 single region 2.554 d-k/k Region 1: 0.97~ d-k/k Region 2: 1.96> d-k/k DISCUSSION: The Region 1..and Region 2 dk/k uncertainties were provided by Westinghouse. .This is historical design information used in the design of the fuel racks.

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ATTACHMENT 5 MARKED UP SPECIFICATION PAGES

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~ I ONTAINMENT SYSTEMS ERGENCY CONTAINMENT COOLING SYSTEM IMITING CONDITION FOR OPERATION 3.6.2.2 Three emergency containment cooling units shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: 'a I b. -With one of the above required emergency containment cooling units inoperable restore the inoperable cooling unit to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in COLO SHUTDOWN within the following 30 hours. With two or more of the above required emergency containment cooling units inoperable, restore at least two cooling units to OPERABLE status within 1 hour or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. Restore all of the above required cooling units to OPERABLE status within 72 hours of initial loss or be in at least HOT STANDBY within the next 6 hours and in COLO SHUTDOWN'ithin the following 30 hours. RVEILLANCE RE UIREMENTS 4.6.2.2 Each emergency containment cooling unit shall be demonstrated OPERABLE: a. b. At least once per 31 days by)( P starting each cooler unit from the control room and verifying that each unit motor reaches the nominal operating current for the test conditions and operates for at least 15 minutesg a 2) Verifying a cooling water flow rate of greater than or equal to 2000 gpm to each cooler. A I At least once per 18 months by'erifying that each unit starts automatically on a safety injection (SI) test signal a,~ TURKEY POINT - UNITS 3 4 4 3/4 6-14 AMENDMENT NOS. ANO I:gg g8 1989

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CONTAINMENT SYSTEMS BASES CONTAINMENT VENTILATION SYSTEM (Continued) resilient material seal degradation and will allow opportunity for repair before gross leakage failures could develop. The 0.60 L leakage limit of Specifica-a tion 3.6. 1.2b. shall not be exceeded when the leakage rates determined by the leakage integrity tests of these valves are added to the previously determined total for all valves and penetrations subject to Type B and C tests. 3/4. 6. 2 DEPRESSURIZATION AND COOLING SYSTEMS 3/4.6.2. 1 CONTAINMENT SPRAY SYSTEM The OPERABILITY of the Containment Spray System ensures that containment depressurization capability wi 11 be available in the event of a LOCA. The pressure reduction and resultant lower containment leakage rate are consistent with 'the assumptions used in the safety analyses. The allowable out-of-service time requirements for the Containment Spray System have been maintained consistent with that assigned other inoperable ESF equipment and do not reflect the additional redundancy in cooling capability provided by the Emergency Containment Cooling System. Pump performance requirements are obtained from the accidents analysis assumptions. 3/4.6.2.2 EMERGENCY CONTAINMENT COOLING SYSTEM The OPERABILITY of the Emergency Containment Cooling System ensures that adequate heat removal capacity is available during post-LOCA conditions. The emergency containment coolers are a full capacity system and are redundant to the spray system in terms of heat removal function for design basis accident. The allowable out-of-service time requirements for the Containment Cooling System have been maintained consistent with that assigned other inoperable ESF equipment and do not reflect the additional redundancy in cooling capability provided by the Containment Spray System. jn s+rt'/4.6.3 EMERGENCY CONTAINMENT FILTERING SYSTEM The OPERABILITY of the Emergency Containment Filtering System ensures that sufficient iodine removal capability will be available in the event of a LOCA. The reduction in containment iodine inventory reduces the resulting SITE BOUNDARY radiation doses associated with containment leakage. The operation of this system and resultant iodine removal capacity are consistent with the assumptions used in the LOCA analyses. System components are not subject to rapid deterioration. Visual inspection and operating/performance tests after maintenance, prolonged operation, and at the required frequencies provide assurances of system reliability and will prevent system failure. Filter performance tests are conducted in accordance with the methodology and intent of ANSI N510- 1975. TURKEY POINT - UNITS 3 8( 4 8 3/4 6-3 AMENDMENT NOS. AND >,'.Pi'( 0 5 198g

INSERT The surveillance requirement for ECC flow is verified by correlating the test configuration value with the design basis assumptions for system configuration and flow. An 18-month surveillance interval is acceptable based on the use of water from the CCW system, which results in a low risk of heat exchanger tube q fouling.

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3/4. 6. 2 DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT SPRAY SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2. 1 Two independent Containment Spray Systems shall be OPERABLE with each Spray System capable of taking suction from the RWST~ ard ~~null(y Cransf~rrsnp s'~g+gyy tel 4~ ~ cori 0/ame

g89 E -0 ÃI m TABLE 3.3-2 I I ENGINEEREO SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION FUNCTIONAL UNIT TOTAL NO. CKA(NELS OF CHANNELS TO TRIP HINIHUH CHANNELS OPERABLE APPLICABLE NODES ACTION Z,' C/l 1.. Safety Injection (Reactor Trip, Turbine Trip, Feedwater Isolation, Control Room d~~tIla+i~~ Isolation, Start Oiesel - Generators, ontainment Cooling Fans, Contain-ment Filter Fans, Start Sequencer, Component Cooling Water, Start Auxiliary Feedwater and Intake Cooling Water). C <<fai~~c. + PAs ia 6 Isola'tc ~ ('~wr<p+~a ~l zz') fll C3 m a. Hanual Initiation b. Automatic Actuation Logic and Actuation Relays c. Containment - Pressure-High d. Pressurizer Pressure - Low '2 2 2 2 1, 2, 3, 4 1, 2, 3, 4 1, 2, 3 1, 2, 38 17 15 15 e. High Oifferential Pressure Between. the Steam Line Header and any . Steam Line 3/steam line 2/steam line 2/steam 1, 2, 3" in any steam line line 15 I ~ J Ld l II II h I a H e TABLE 3.3-3 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM RUM N S N FUNCTIONAL UNIT y<<>jj~f~t<~~ TRIP SETPOINT ALLOWABLE VALUE¹ 1. Safety Injecti (Reactor Trip, Turbine Trip, eedwater Isolation, Control Room solation, Start Diesel Generators, on a>nment Cooling Fans, Containment Filter Fans, Start Sequencer, Component Cooling Water, Start Auxiliary Feedwater and Intake Cooling Water) &nf<i<m<~7 1 J? ww A gio/a pxo+ f +~ <<FC<pr ~an<>/ Z+g a. Manual Initiation N.A. N.A. b Automatic Actuation Logic N.A. N;A. e. 0, High Differential Pressure Between the Steam Line Header and any Steam Line. Steam Line FlowHigh c. Containment Pressur eHigh d. Pressurizer Pressure Low <6 psig >1715 psig- <150 psi <[ 3 psig >[ ] psig <[ 3 psi <A function defined [ as follows: A hp corresponding to 0.64 x 106 lbs/hr at OX load increa-ing linearly to a dq corresponding to 3.84 x 10 lbs/hr at full load. Coincident with: Steam Generator Pressure--Low or T Low Containment Spray Automatic Actuation Logic and Actuation Relays >600 psig >531 F N. A. >[ ] psig 3'F N.A. . b. Containment Pr essure High-High Coincident wi.th: Containment Pressure High <30.0 psig <6.0 psig ] psig <[ j psig TURKEY POINT - UNITS 3 5 4 3/4 3-23 AMENDMENT NOS. AND FEB g 8 ))g k V g ~ gl J I 1 Q p,y Ql Q II ~ E ~ ~ II . ~ '.3-2 I I ENGINEERED SAFETY FEATURE ACTUATION SYSTEN INSTRUMENTATION C/1 CHANNEL 'FUNCTIONAL UNIT TRIP ANALOG ACTUATING NODES CHANNEL DEVICE FOR WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE CHECK CALIBRATION TEST TEST LOGIC TESTS IS RE UIRED m C3 m b. C. d. Automatic Actuation N.A. Logic and Actuation Relays Containment Pressure N.A. High Pressurizer Pressure S Low e. High Differential Pressure Between the Steam Line Header and any Steam Line Safety Injection (Reactor Trip, Turbine Trip, Feed-water Isolati on Control Room Isolat>on, Start Diesel Generators, on asn-ment.Cooling Fans, Contain-ment Filter Fans, Start Sequencer, Component Cooling Water, Start Auxiliary Feed-water and Intake Cooling Water) a. Hanual Initiation N.A. N.A. N.A. N.A. N.A. N.A. N.A. N.A. N.A. N.A. N.A. l C+cu~ +~1 a. tPB'~w A +o(wf(i~ (aggzft n'e" vW SX) 1, 2; 3 1, 2, 3(P) x 1 2 3 1 2 3(P) ~ x .1) 2, 3(g) xx ll' CM f. Steam Line FlowHigh S Coincident with: Steam Generator Pressure--Low S or T --Low S N.A. N.A. N.A. N.A. N. A. N.A. 1, 2, 3(E) 5 1,2,3(m 1, 2, 3(Q ~/ 4 ~ ri'r 4 ~ h~rg O lh b r 4 b hI}}

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