ML20212F274
| ML20212F274 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 10/31/1997 |
| From: | FLORIDA POWER CORP. |
| To: | |
| Shared Package | |
| ML20212F271 | List: |
| References | |
| NUDOCS 9711040238 | |
| Download: ML20212F274 (119) | |
Text
{{#Wiki_filter:_ ._ _ - - _ _ _ . . _ _ _ _ _ _ . -- l l FLORIDA POWER CORPORATION l l CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/ LICENSE NUMBER DPR-72 ATTACIIMENT C LICENSE AMENDMENT REQUEST 214 REVISION 0 STRIKEOUT /SIIADOW PAGES Technical Specifications & Bases Each change is ludicted by a shadow box. Deletions are indicated by strikeout. Additional and replacement text are ludicated by shading. l 1 !Ta' P 188 n FJ8 & 2 l PDR
EFIC Manual Initiation i 3.3.12 3.3 TNSTRUMENTATION ; t 3.3.12 Emergency Feedwater Initiation and Control (EFIC) Manual Initiation LCO 3.3.12 Two manual initiation switches per actuation channel for i each of the following EFIC Functions shall be OPERABLE:
- a. Steam generator (OTSG) A Main Feedwater (MFW) Isolation;
- b. OTSG B MFW Isolation;
- c. OfSC A Main Steam Line Isolation; ;
- d. OTSG B Main Steam Line Isolation; and
- e. Emergency Feedwater (EFW) Actuation.
p.. =.... = ....;.:.;...Un;.n ... C Z NOTE -- E A----- G C W- l- Z - r- G 2-? Only:Et he manuallinitiation switches! associated with.EFW actuation ^are required r -
............ ..............t....
e ;
,tobeOPERASLEjiniMODE(4.- ,
4............. 33...................... APPLICABILITY: MODES-1T-27-and-3r ACTIONS
.((Insert)M06E5E273.'andf.4-
..................................... NOTE-~~--------------~~~~----------------
Separate Condition entry is allowed for each Function. CONDITION REQUIRED ACTION COMPLETION TIME A. One or more EFIC A.1 Place trip module for 72 hours l Function (s) with one the associated EFIC manual initiation Function (s) in trip, switch inoperable in one actuation channel. B. One or more EFIC B.1 Restore one manual 72 hours Function (s) with both initiation switch to manual initiation OPERABLE status, switches inoperable in-one actuation channel, l l (continued) Crystal River Unit 3 3.3-30 Amendment No. 149 l.
EFIC M nual Initiation 3.3.12 l ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME C. One of more EFIC C.1 Place trip modules for 1 hour ' functions with one the associated EFIC manual initiation Function (s) in trip. Switch inoperable in both actuation channels. D. One or more EFIC D.1 Restore one actuation 1 hour Function (s) with both channel for the manual initiation associated EFIC < switches inoperaole in Function (s) to OPERABLE both actuation status, channels. E. Required Action and E.1 Be in MODE 3. 6 hours . associated Completion Time not met AND F " for: manual initiationi F switchel=Det j E.2 Be in MODE 4. 12 hours i h associated with EFW 'i k actuationx x 2 '(gi; m
}
Ff? Required Action?and? 7/ F;1Tlei inL. MODE 13;l ': . E'M 6: hours?JT Z Si Lassociated Completion a bl / Time.not met'for? ANDi"il ".21 6 ~1. :'"a (1 manual? initiation: [ switches' associated' F;22 Belin? MODE 151 l "' m l30 hours;: in M rr zwith EFW actuation m l I l SURVEILLANCE REQU' EMENTS SURVEILLANCE FREQUENCY SR 3.3.12.1 Perform CHANNEL P,NCTIONAL TEST. 31 days p m ,- I l. i , Crystal River Unit 3 3.3-31 Amendment No. 149
I PAM Instrumentation i 3.3.17 l i 3.3 INSTRUMENTATION 3.3.17 Post Accident Monitoring (PAM) Instrumentation ; i LCO 3.3.17 The PAM instrumentation for each Function in Table 3.3.17-1 shall be OPERABLE. APPLICABILITY: MODES-17-2 and-3r ! linseH i "Nides;10 2 d ii3 T " l ACTIONS - !
..................................... NOTE----------------------------------- !
- 1. -LCO-3.0.4 is not applicable. l
- 2. Separate Condition entry is allowed for each Function. .
CONDITION REQUIRED ACTION COMPLETION TIME A. One or more Functions A.1 Restore required 30 days : with one required channel to OPERABLE channel inoperable, status. j B. Required Action and- B.1 Initiate action in Immediately associated Completion accordance with , Time of Condition A Specification 5.7.2a. , not met. t C. One or more Functions C.1 Restore one channel to 7 days with two required OPERABLE status channels inoperable. ' (continued) ; CrystalLRiver Unit 3 3.3-38 . Amendment No. 149 p
PAM Instrumentation 3.3.17 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME , D. Required Action and 0.1 Enter the Condition Immediately associated Completion referenced in Time of Condition C Table 3.3.17-1 for not met. the Function. E. As required by E.1 De in MODE 3. 6 hours Required Action D 1 and referenced in AND Table 3.3.17-1. E.2 3e in E-4 . hours
- ugiwnean maano F. As required by F.1 Initiate action in Immediately Required Action D.1 accorriance with and referenced in Specification Table 3.3.17-1. %.7.2.a.
l Crystal River Unit 3 3.3-39 Amendment No. 149 l
RCS Loops-MODE 4 3.4.5 4 3.4 REACTOR COOLANT SYSTEM (RCS) IiNsidfii65IWit5 loops er.one AC5iloop 3.4.5 RCS Loops-MODE 4 and one decay; heat; ; r'emovali(DN gR) l g _ , t LCO 3.4.5 Two loops consisting of any-combination-of-RCS-loops 1nd decay-heat-removal-(DHR)-loops shall be OPERABLE and at l least one loop shall be in operation. : i
.........................-NOTE-------------------------- l All reactor coolant pumps (RCPs) may be de-energized for j s 8 hours per 24 hour period for the transition to or from '
the DHR System, and all RCPs ared DHR pumps may be i de-energized for s I hour per 8 hour period for any other . reason, provided:
- a. No operations are permitted that would cause reduction of the RCS boron concentration; and
- b. Core outlet temperature is maintained so as to assure subcooling throughout the RCS.
F ADPLICABILITY: MODE 4. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One RCS loop A.1 Initiate action to Immediately inoperable. restore a second loop to OPERABLE status. IND Two-OHR-1 s inoperable, w I(15iiditj[RequiredI (continued) b3P_._.L__ l-4 l ! Crystal River Unit 3 3.4-8 Amendment No. 149 t
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i i RCS Loops-MODE 4 3.4.5 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME : B. One@R-loop B.1 Initiate action to Immediately i - m- restore a-second loop i rabier [hl***.J. to OPERABL status. b- ((5drtIan[Rd5 QR wwa Two RCS loops inoperable. B . 2_ Be in MODE 5. 24 hours C. All RCS and DHR loops C.1 Suspend all operations- Immediately inoperable.- involving a reduction in RCS boron 08 concentration. No RCS or DHR loop in ANQ ; operation. ' C.2 Initiate action to Immediately restore one loop to OPERABLE status and operation. SURVEILLANCE REQUIREMENTS __ SURVEILLANCE FREQUENCY SR 3.4.5.1 Verify one DHR or RCS loop is in operation. 12 hours SR 3.4.5.2 Verify correct breaker alignment and 7 days indicated power available to the required pump that is not in operation. ..; Crystal River Unit 3 3.4-9 Amendment No. 149-i r.e,-,..,,--.,,,-i..,-.,-,,.n.n.-,,,,--,--~,..,,..,,,v..-n,,.n.,,,,-,,.w,--,,,.-,v,
.,-...,.,.Lv.,~, -- n, , -.,,.r ..-,,,.r,-.- , .- r
ECCS-Shutdown i 3.5.3 ! I SURVEILLANCE RLQUIREMENTS SURVEILLANCE FREQUENCY l t SR 3.5.3.1 ------------------NOTE------------------ . 4 An-LPI-subsysts my be considered-OPERABLE :
~
/ during-ali gnment-and-oper a tion-foe-decay i[Mifef]'+>heatremoval-if-capable-of-being-manually L ----- re-aligned-to-the-ECC w.ede of-operation For all . equipment required to be OPERABLE 1 In accordance '
the following SRs are applicable, with applicable SRs SR 3.5.2.1 SR 3.5.2.6
- SR 3.$.2.2 SR 3.5.2.1 ;
SR 3.5.2.5 l i Crystal River Unit 3 - 3.5-8 Amendment No. 149-1 1
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l
.i
.TBVs 3.7.4 l 3;7 PLANT SYSTEMS 3.7.4 Turbine Bypass Valves (TBVs) i LCO 3.7.4 Each TBV shall be OPERABLE.
----------------------------NOTE-------------------------
N N00E I4,Pik19"tETSVs~isioci atWivi thTWeTOPERABLT RCSIl JMshare%uMMMEwa_-ww x camaad f[hisert]; L_ w ...-...... -............-...-_...-----...-.....------- --- APPLICABILITY: MODES-1, 2, ar.4-3r (( Insert]'ii0lf'1,2,L3nandE47] ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One of more TBVs A.1 Restore-TBV(s) to 7 days inoperable. OPERABLE status. 08-A.2 Verify by 7 days administrative means OPERABILITY of associated steam generator atmospheric dump valve (ADV). ' B. Required Action and B.1 Be in MODE 3. 6 hours associated Completion Time not met. AND B.2 Be.in i ottes (( [.- Insert]7>;p] l[ Insert](30] 6 g.+_J yurse C n, t SURVEILLANCE REQUIREMENTS . SURVEILLANCE. FREQUENCY SR 3.7.18.1 Perform one complete cycle of each TBV. 24 months I Crystal- River Onit 3 3.7-8 Amendment No. 149
3.7.5 (( Ins (tDN System-Operating] [T4 S) 3.7 PLANT SYSTEMS gg, gy 3.7.5 Emergency Feedwater (EFW) Sys*em Operating LCO 3.7.5 Two EFW trains shall be OPERABLE. !
----------------------------NOTE------------------------- ,
Only one EFW train, which includes a motor driven pump, is i required to be OPERABLE in MODE 3 with steam generator ! pressure-< 200 psig. J APPLICABILITY: MODES 1, 2, and 3. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One-steam supply to A.1 Restore steam supply 7 days the turbine driven EFW to OPERABLE status, pump inoperable. AND 10 days from discovery of failure to meet the LCO B. One EFW train B.1 Restore EFW train to 72 hours inoperable for reasons OPERABLE status, other'than AND Condition A. 10 days from discovery of failure to meet the LCO (continued) Crystal River Unit 3 3.7-9 Amendment No. 149
/FW-System
$ h5j[E,W5yj[E{@[@hC ACTIONS (continued) _
CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action ar.d C.1 Be in MODE 3. 6 hours associated Completion Time of Conditior, A or 6HD 0 not met. C.2 Be in MODE 4. 12 hours D. Two EFW trains D.1 Initiate action to Immediately inoperable. restore one EFW train to OPERABLE status.
'L Crystal diver Unit 3 3.7-10 Amendment No. 149
. _ _ ~ . - ,
l FW-Syttem l 3.7.5
}!asert][5Ni$ystWQtd]
SURVEILLANCE REQUIREMENTS { SURVEILLANCE FREQUENCY SR 3.7.5.1 Verify each EFW manual, power operated, and 45 days , automatic valve in each water flow path and in both steam supply flow paths to the turbine driven pump, that is not locked, sealed, or otherwise secured in position, is in the correct position. SR 3.7.5.2 -------------------NOTE-------------------- Not required to be performed for the. turbine driven EFW pump, until 24 hours after reaching 200 psig in the steam generators. Verify the developed head of each EFW pump at 45 days on a , the flow test point is greater than or & qual STACCERED TEST to the required developed head. BASIS SR 3.7.5.3 -------------------NOTE-------------------- Not required to be performed until 24 hours after reaching 200 psig in the steam generators. Verify each_EFW automatic valve that is not 24 months locked, senyd, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal. ! (continued) ; i l s Crystal River Unit 3 3.7-11 Amendment No. 149 l
_ - . . __ _ _ _ . __ . . _ . _ _ . _ _ _ _ . _ _ . _ _ . - _ . . ~ . . . _ - . _ . . . . _ _ _ _ _ _ E FW-System _ _ _ - , . _ _ .3.7.5 '
. L_In_se.
T. ., rtl,1 EPWisystem-Operitl. ngN _.]. . SURVEllLANCE REQUIREMENTS (continued)-
~ SURVEILLANCE FREQUENCY
-SR 3.7.5.4 --------- ---------NOTE--------------------
Not required to.be performed until 24 hours after reathing 200 psig in the steam generators.
' Verify each [FW pump starts automatically on 24 months ,
an actual or simulated actuation signal. SR 3.7.5.5 Verify proper alignment of the EFW flow paths Prior to by verifying flow from the EFW tank to each entering MODE 2 - steam generator. whenever plant has been in MODE 5 or 6 for -
> 30 days 1
. Crystal' River Unit 3' 3.7-12 Amendment No. 149
- . . + ,- , . . . .. . . - , , , , - , - . , . , ..m...-.. . . . , - . _ , . . . . _ ~ . . , -. , . . . . - , , . , . .-..,-.,my,
Emergency Feedwater Tank 3.7.6 3.7 PLANT SYSTEMS , 3.7.6 Emergency Feedwater (EFW) Tank i LCO 3.7.6 EFW tank water volume shall be a 150,000 gal. > [linsert[M00E5f1', 2L,[3, Lan{y APPLICABILITY: MODES-1,-- 2.-- and-3r . ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. EFW tank water volume A.1 Verify by 4 hours not within limit, administrative means OPERABILITY of backup AMQ water supply. Once per 12 hours thereafter AND A.2 Restore EFW tank 7 days water tank volume to within limit. 2 1 B. Required Action and 8.1 Be in MODE 3. 6 hours associated Completion
- Time not met. ARD B.2 Be in MODE-4r hours
/ /
t(~.-a[., =Iins M M..T5+ 7 ,f[ Insert);30; hoursf i b T Crystal-River Unit 3 3.7-13 Amendment No. 149
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.it.023,.7.,18 w,2Motot dr,1venz EFE.,. tra_in~, s . hall! be m0PERABLE.,f_A:1J;,w.R.d., n 3 m m.w smm, mm - yr. y .m MODE,A,,,,.,w _ ~ o. y.mau.
, APPLICABILITY. 7
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A.T Mot 6rldriverFEFWEtFaini X.' 1" Inhiate sctionitd7El Immediate19 C h w inoperable., L , j i r ktore(motor driveni, ir *I EFittr'ain - to 2 OPERABLE! L __ ; status.; . a J I ,y,,
. - - -- .g% j3gye , .- ,.-- v- 4 gwg y j m m2s<...s4.mg- 7 B.15: d, ed* nu.m M00Em 3_Aah a ~24: hours;.
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nn . , . . cJ- . L a m N SURVE.ILLANCE,u . m e /a n.r,,w i.; o J REQUENCYu -
.s - .. _. .- .- - _.
SR ,s 3,. 7;18,,1 For_; alls ,." equip.- . ment t required'to b.me OPERABLE,,.,- . - .In accorda.._ _,nce Q j
- x. i ;the;following1SRsiare applicable.7.ma .3 withlapplicablei
- SRs ~> ~~1 ' "
~ -N m.
- m. ,,_ ..
SR317;5,:1,; . _ '.- SRms 3'.7l5;,3 _- SRJ3.7,5.2M a SR .13;7j5t4[ au . m %i { t i
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j .r, .,, ,y y 9 gy. 3e .g. q . ,.,,g- .. __.9, rystal.g . d Ri ._< $.verJnit.s3m i h n A 13u .37u M:n . a : K A Amendment No.- 2 _ C. .. - s;L
FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/ LICENSE NUMBER OPR-72
, ATTACHMENT C (cont.)
LICENSE AMENDMENT REQUEST 214 REVISION 0 STRIKEOUT / SHADOW PAGES Base.i . Each change is indicted by a shadow box, Deletions are indicated by strikeout. Additional and replacement text are indicated by shading. L M t._____ _____im._ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - - - - - _ - - - - --- - - - - - - - - - - - - - - - - - - - - '
J l EFIC Instrumentation' l B 3.3.11 BASES' APPLICABILITY- during the MODES and specified conditions listed in Table 3.3.11-1. Each Function has its own requirements based on the specific accidents and conditions.for_ which it-is designed to provide protection. The initiation of EFW on the Loss of MFW Pumps is applicable. in MODE 1 and in MODES 2 and 3 when not in shutdown bypass. Below these plant conditions, EFW initiation on low OTSC
-level occurs fast enough to prevent' primary system-overheating.
EFW-Initiation on low OTSG 1evel shall be OPERABLE at all times the OTSG is required for heat removal.- These conditions include MODES 1.-2, and 3. To avoid automatic actuation of the EFW pumps during heatup and cooldown, the low OTSG pressure Function can be-bypassed at or below a . secondary pressure of _750 psig. This-secondary-side pressure occurs during MODE 3 operation. EFW initiation on loss of all RCPs is required to be-
- 0PERABLE at a 10% RTP. This power level coincides with the bypass permissive signal provided t'y RPS.
The MFW, Main Steam Line Isolation, and EFW Vector Valve Control Functions shall Le OPERABLE in . MODES 1, 2, and :, with OTSG pressure a 750 psig because OTSG inventory can be high enough to contribute significantly to the peak pressure following a secondary side break. Both the normal feedwater and the EFW must be isolatable on each OTSG to limit overcooling of the _ primary and mass and energy
-releases to the RB. Once OTSG pressures decrease below 750 psig, the Main Steam Line and MFW Isolation Functions-can be bypassed to prevent actuation during cooldown. . _ The EFW Vector Valve Control logic will not perform any function L -when both 'OTSG pressures are low; thus, the logic is also
[13ert]3hel bypassed at the same time the OTSG pressure low Functions is automatict 3 %, bypassed. In MODES 4, 5, and 6, primary and secondary side
'actuationi o ; energy levels are reduced and the feedwater flow rate is low features?of?EFIC4 or nonexistent. Because of.thi .-EFIC instra.entation is instrumentat ionp :not required to be OP n these' MODES.
arei sen < n
~
. ThsTFsisWesisnif66the~'essiNEisatidsfea^tuFM7Aff EFICM ii n :M00E141aroldi scussed u nr Speci fi cation 13112 k&dCgd kI Mb (continued) l l
Crystal River Unit 3 8 3.3-94 Amendment No. 7 I l l
EFIC' Minual Initiation B 3.3.12 8 3.3- INSTRUMENTATION- ., B 3.3.12 Emergency Feedwater Initiation and Control (EFIC) Manual Initiation
- j.-
l-BASES BACKCROUND :The EFIC manual initiation capability provides the operator with the capability to actuate certain EFIC Functions in' the absence of an automatic initiation condition. Functions with the capability to be manually actuated include Main -
-Feedwater (MFW) Isolation for Once Through. Steam Generator (OTSG) A MFW Isolation for OTSG B, Main Steam Line Isolation for OTSG A Main Steam Line Isolation for OTSG B, and Emergency Feedwater (EFW) Actuation.
The EFIC manual initiation circuitry satisfies the manual initiation and single-failure criterion requirements of IEEE-279-1971 (Ref. 1). Although not part of this LCO, the EFIC Functions listed ' above can also be remotely manually initiated from the EFIC cabinets. APPLICABLE EFIC Functions credited in the safety analysis are SAFETY ANALYSES automatic. However, EFIC manual initiation Functions are
- required by design as backups to the automatic trip Functions. This allows the operator to actuate EFW,-Main Steam Line Isolation, or MFW Isolation whenever conditions dictate and one has not already automatically occurred. As such, they are backup Functions to those performed automatically by EFIC.
- In" MODE T4 EWTH maF9~ aridFecondaFyTsi def^ese rgyIl Del sTaiil
'reducediandifeedwater flowiratelis- lown $ Sufficientz time?is%
a
- ,__ j"available,td(the(operator 7td manu' lly eventithatlrequiresna}.safetyigrade initiate EFICli_ns.ail ;U sourcelof;feedwatterAbd
{nse L l l , (continued) l-Crystal River Unit 3 8 3.3-100 Amendment No. 149 L 1
i EFIC Minual Initiation B 3.3.12 > BASES LCO Two manual initiation switches per actuation channel (A and B) of each Function-(OTSG A.and B MFW Isolation, OTSG A and B Main Steam Line Isolation, and EFW Actaation) are required to be OPERABLE whenever the OTSGs are relied on to remove heat from the primary. Each Function (MFW Isolation, , Main Steam Line Isolation, and EFW. Initiation) has two actuation or " trip" channels, channels A and B. Within each channel A actuation logic there are two manual trip switches. When one manual switch is depressed, a half trip - l (continued) L Crystal. River Unit 3 8 3.3-100a Amendment No. 149 i I
EFIC Minual Initiation B-3.3.12 BASES i 1 LCO occurs. When both manual switches are depressed, a full (continued) trip of channel A actuation occurs for that particular Function. Similarly, channel B actuation logic for each Function has two manual trip switches. Both_ switches per actuation channel _must be OPERABLE and must be depressed to get a full manual trip of that channel. The use of two manual trip switches for each channel of actuation logic allows for testing without actuating the end devices and also reduces the possibility of accidental manual actuation. APPLICABILITY The MFW and Main Steam Line Isola 4on manual initiation Functions shall be OPERABLE in MODES 1, 2, and 3 because OTSG inventory can be at a sufficiently high energy level to contribute significantly to the peak containment pressure during a secondary side break. In MODES 4, 5, and 6, the lInsert]'M09ES~ primary and secondary side energy levels are reduced and l'i2,I3.fand_4L
- faadwater flow rate is low or nonexistent; and the Function is ot required to be OPERABLE.
The EF manual initiation Function shall be OPERABLE in MODES 1, 2, End-3 because the OTSGs are relied on as a heat
-- . ---- sink for the Reactor Rolant System and the core itself. In n[ Insert]jMODES: MODES-4r+-andHi, heat removal requirements are reduced and 5'and 6,2 : ca M rovided by the Decay Heat Removal System.
ACTIONS A Note has been added to the ACTIONS indicating that separate Condition entry is allowed for each EFIC manual initiation Function. Ail With one manual initiation switch of one or more EFIC Function (s) inoperable in one actuation channel, the trip module for the associated EFIC Function (s) must be placed in the tripped condition within 72 hours. With the channel in the tripped condition, the single-failure criterion is met
- and the operator can still initiate one actuation channel
! given a single failure in the other channel. Failure to i L (continued) Crystal River Unit 3 8 3.3-101 Amendment No. 149
EFIC Manual Initiation B 3.3.12-
. BASES ggyym &+3py g gyny ym %w;7, a;r mn n~w:y yxwyq;y; (continuAd h o w n m M <
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':Iff thei Requi red :Act" ions ; cannot: be ' met M this thA[ia
,~ .
lationiTimes?for manualkinitiation switches!forjEFW M 9-
~
ctuationythe plantimustlbe7placedfin alMODEf in which the?$ LCO:doesinotcappip; fTo ' achieve 1this status @the? plant mustij
'[Inser.t]2xm
.be placed)inTatLleast! MODE l31within: 6)hoursLand"in1 MODE:52 ?!
inithin 30; hoursATheiailowed Comp 1'etioneTimesfare;1 de TreasonableibasedfonToperating"sxperiente,toireach?the0lgj
~
T tj 3
;mannerz requiredandiwi l MODES 1fromengi thoutichall fullipowericonditionsLin an1orderlpi y;;d ng i pl antj systems '.musAE SURVEILLANCE SR. 3.3.12.1 REQUIREMENTS This SR requires the performance of.a CHANNEL FUNCTIONAL TEST to ensure that the channels can perform their intended functions.-This SR, like all other CHANNEL FUNCTIONAL TESTS, does not include actuation of the end device. This is-due to the risk of a plant transient caused by the closure of valves associated with MFW and Main Steam Line Isolation or actuating EFW with the reactor at power. The Frequency of 31 days is based on operating experience and industry-accepted practice.
( : REFERENCES 1. IEEE-279-1971, April 1972. l l l i f ! (continued)
' Crystal River Unit-3 8 3.3-104 Amendment No. 149-
l 1 EFIC-EFW-Vzctor Valva Logic B 3.3.14 : BASES ACTIONS- dul (continued) With one channel inoperable, the system cannot meet the single-failure criterion and still satisfy the dual functional criteria described above. Therefore, when 'one vector valve -logic channel is inoperable, the channel must be restored to OPERABLE status within 72 hours. This Condition is analogous to having one EFW train inoperable; wherein a - 72 hour Completion Time :is provided byLthe Required Actions of ifC 3.7.4, " Erd System." ~As such, the-Completion Time' of. hours is based on engineering j udgment. -
- fOnsert)1.LC0]3;7.!5,;,.
[E K System-Operating % B.1 and B.2 If Required- Action A.1 cannot be met within the associated Completion Time, the plant must be placed in a MODE in which the LCO does not apply. To achieve this status, the plant must be placed in at least . MODE 3 within 6 hours and in MODE 4 within 12 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3.3.14.J REQUIREMENTS SR 3.3.14.1 is the performance of a CHANNEL FUNCTIONAL TEST every 31 days, This test demonstrates that the EFIC-EFW vector valve logic is capable of performing its intended function. Tne Frequency is based on op.; rating experience that demonstrates failure of more than one channel within the same 31 day interval is unlikely. REFERENCES None. (continued) , ' Crystal River Unit 3 'B 3.3-113 Amendment No. 149
PAM Instrumentation B 3.3.17 BASES APPLICABILITY The PAM instrumentation requirements are applicable in ((fseR]P MODES-17-27-and-3. These variables are related to the diagnosis and pre-planned actions required to mitigate
$ES'1.J2, !3 band _4 d DB/.s . The applicable DBAs are assumed to occur in MODES-ly h-and-3. In-M00E5r-4, 5, and-6, plant operating conditions - gggg are su that the likelihood of an event occurring that
- amt require PAM instrumentation is low; therefore, PAM
.N._ ODE.S_5
~J _and 6 instrumentation is not required to be OPERABLE in these MODES.
ACTIONS The ACTIONS are modified by two Notes. Note 1 was added to indicate the restrictions of LCO 3.0.4 are not applicable. This exception allows ents y into an applicable MODE while relying on the ACTIONS even though the ACTIONS would eventually require a shutdown. This exception is acceptable due to the passive function of the instruments, the operator's ability to respond to an accident utilizing alternate instruments and methods, and the low probability of an event requiring these instruments. Note Two was addeo to clarify the application of Completion Time rules to this Specification. The *~onditions of this Specification are entered indepe~ for each Function listed in Table 3.3.17-1. The Cony .cion Time (s) of the inoperable channels of a Function will be tracked separately for each Function starting from the time the Condition was entered for that Function, eul Wnen one or more Functions have one required channel fnoperable, the inoperable channel must be restored to OPERABLE status within 30 days. The 30 day Completion Time is based on engineering judgment and a variety of considerations. These considerations include availability of the remaining OPERABLE channel, the passive nature of the instrument (no critical automatic action is assumed to occur from these instruments), and the low probability of an event requiring PAM instrumentation during this i n tervr.l . (continued) Crystal River Unit:3 8 3.3-139 Amendment No. 149
PAM Instrumentati8n - B 3.3.17 BASES
- ACTIONSi -
RJ (continued)- . Required Action D.1 directs entry into the appropriate-Condition referenced in_ Table 3.3.17-1. The applicable Condition referenced in the Table is Function dependent. Each time an inoperable channel has'not met any Required Action'and associated Completion Time of Condition C, , Condition D is entered for that Function and the operator is directed to the appropriate subsequent Condition.
~
E11 If the Required Action and associated Completion Time of Conditions C is not met and Table 3.3.17-1 directs entry into Condition E, the plant must be placed in a MODE in which the requirements of this LCO do not apply. To achieve this status, the plant must be placed in at least MODE 3 within-6 hours and in C 4 withiri 12 hoves. The , allowed Completion Times are rea onable, based on operating experience, to reach the require plant conditions from full power conditions ir an orde ly manner and without challenging. plant systems.. ~ (( Insert) MODE 45fwithinf p0 hours? lN > Alternative means of monitoring containment area radiation and reactor vessel level are available and may be relied upon if the r:ormal PN4 channels cannot be restored to OPERABLE status within the associated Completion Time. Based upon this capability, it is inappropriate to require plant shutdown -in this condition. Rather, in conjunction with the alternate monitoring means, the Required Action specifies action be immediately initiated in accordance with Specification 5.7.2.a. "Special Reports," in the 4 Administrative Controls section of the Technical
' Specifications. The report provided to the NRC should discuss the alternate means of monitoring, describe the -
degree to which the alternative means are equivalent to the installed PAM channels, justify the areas in which they are not equivalent, and provide a schedule for restoring the normal PAM channels to OPERABLE status. l (continued)
-'CrystalR[verUnit3' B 3.3-141 Amendment No. 149
RCS-Loops-Mode 4 8 3.4.5-
\
~
B-3.4- REACTOR; COOLANT SYSTEM (RCS)L
~B.3.4.5' RCS. Loops-MODE 4 BASES BACKGROUND In MODE 4', the primary function of the reactor coolant is ,
the removal of decay heat and transfer of this heat to the steam generators (OTSGs) or decay heat removal (DHR) heat exchangers. The secondary-function of the reactor coolant is to act as a transport medium for soluble neutron poison,- boric acil I!. MODE 4, either. reactor coolant pumps (RCPs) or DHR pumps can be used for coolant circulation. The number of pumps in operation can vary to suit the operational needs. The intent of this LCO is to provide forced flow from at:least one RCP or one DHR pump for decay heat removal and transport. The flow provided by one RCP or one DHR pump is adequate for heat removal. The other intent of this LCO is to require that two paths (loops) be available to provide redundancy for heat removal. APPLICABLE No safety analyses are performed with initial conditions in 4 SAFETY ANALYSES MODE 4. The flow provided by one reactor coolant or one
$ decay heat removal pump is adequate to-prevent boron stratification in the vessel core region during a reduction of boron concentration.
RCS loops MODE 4 satisfies the requirements of NRC Policy Statement.= While none of the three criteria directly apply, this Specification assures that reactivity control is maintained, thus Criterion 2 is the appropriate criterion, because boron dilution and reactivity control in natural circulation are unanalyzed. Potential reactivity increases would be outside the bounds of the safety analysis. RCS loops- MODE 4 was identified in. the NRC Policy Statement as an important contributor to risk reduction. LCO .The purpose of.this LCO is..to require that two loops, RCS cr
.. _ DHR,' be OPERABLE in MODE' 4 and one of these loops be _in
.. Inserti b _ peration. The-tEO allews the twe loops that are required
- itheeftwoM to be Oi'E m mm 4 floops Systea; ih"0LC s. Anyto censist one loopsfinany ce.abination operation providesmenough
_ mm + ,j,eRC$ilol n ort ops ___, ..-- _ [Insertf0nlytone DHR?loopdsf available in MODE y"Mone?0HRg p" 4 J N +- _ [% since: theiotherJOHR' trainYist required to psafdedicated!LPI/ECCS? function (infaccordance6
;.4th ; Speci fi cation (3 ( $ i3 P ECCShShutdown -
^M Crystal. River Unit 3 ~B 3.4-22 Amendment No. 149
^ '
RCS Loops-MODE 4 8 3.4.5 BASES LCO flow to remove -the decay heat from the core with forced (continued) circulation. The second loop that is required to be OPERABLE provides a redundant path fc,r heat removal. An OPERABLE RCS loop consists of at least one OPERABLE RCP and a flow path for circulating reactor coolant around the loop. RCPf, are OPERABLE if they are capable of being powered anj are able to provide flow if required. Similarly for the DHR System, an OPERABLE DHR loop is comprised of the OPERABLE DHR pump (s) capable of providing forced flow to the DHR heat exchanger (s). DHR pumps are OPERABLE if they are capable of being povered and are able to provide flow if required. The Note permits a limited period of operation during which all RCPs may be de-energized for s 8 hours per 24 hour period for the transition to or from the DHR System. This allows the RCPs to be secured prior to reducing RCS pressure below that needed to place DHR in service. In this pressure range, accelerated RCP seal degradation can potentially occur due is inadequate NPSH. The Note also permits all DHR and RC pi': ops to be stopped for s 1 hour per 8 hour period for any reason. During this period, natural circulation will provide core decay heat removal. The Note prohibits the reduction of RCS boron concentration when forced flow is stopped because an even concentration distribution cannot be ensured. Core outlet temperature is to be maintained so as to assure subcooling throughout the RCS so that no vapor bubble may form and possibly cause a natural circulation flow obstruction. APPLICABILITY In MODE 4, the heat generated is lower than at power; therefore, on RCS loop in operation is adequate for transport a eat removal. A second RCS loop is required to be OP LE in order redundant heat removal capabil gelete],g
- : not have to be in operation.
This LCO allows use of either DHR or RCS loops because it is possible to remove core decay heat and to provide proper boron mixing with either system. (continued) Crystal River Unit 3 8 3.4-23 Amendment No. 149
i
>RCS Loops-MODE 4 B 3.4.5 BASES'
- APPLICABILITY ; Forced circulation is required in all MODES and is addressed - t (continued) _by the following Specifications:-
LCO 3.4.4, "RCS Loops MODE 3"; ~ LCO - 3.4.6, "RCS Loops MODE 5, - Loops Filled"; - LCO-3.4.7, _ "RCS LA>ps MODE 5, Loops Not Filled"; LCO 3.9.4, " Decay Heat Removal (DHR) and Coolant-Circulation High Water Level" (MODE 6);and LCO 3.9.5, " Decay Heat Removal. (DHR) and Coolant Circulation Low Water: Level" (MODE-6). Forced circulation is implicitly required in MODES 1 and 2 in order to prevent _a Reactor Protection System actuation (Ref. LCO 3.3.1).
'ACTIOWS M
- If one RCS loop and -h DHR loops'are inoperable redundant
-[InsertfloopM
! forced flow heat removal capability is lost. Action must be
- eitherian RC5 immediately initiated to restore a seco oop to OPERABLE Ioop orkthe' status. The imniediate Complet e reflects the
;equired10HR? _importance of maintai e-availability of two paths for BOO #e + $ ' heat removal . In t s Condition the OPERABLE RCS loop must be in operation, except as allcwed by the Note in the LCO, .
otherwise Condition C is also w licable. B.1 and B.2
-((InsettL antRCS3 Q If sne DHR-4ccp and both RCS loops are inoperable, a second loop must be restored to OPERABLE status to satisfy single failure considerations. Action to restore the loop must:be initiated immediately reflecting the urgency of restoring redundant heat removal capability. One DHR loop is still available for cooldown given the redu d heat loads of-this operating MODE. -((InsedHThe reqsired]
((_ Insert)j IsI hWCohditichHi:he"Fegsifedl0PERABLE?DHRIlcsgi%if tiasid t l:
'operationM excepti as allowed byfthei Note iin/theiLC00 ' ; @i ll
'otherwi se / Condi tion i Cii sl ai solappli cablehrmai& wad If restoration cannot be reasonably accomplished,
_ ._ consideration should be given = to' placing the plant in [ i((Deltte)? j N MODE-5. Placing the plant in "00C 5 is a conservative laetion with regard te decay heat reic. eval, bst-does necess4 tete-RCS heatsp scctt-tc stilize the OTSC5 fer decay (continued) Crystal River Unit'3 B 3.'4-24 Amendment No.-149
._u_ . ._. . _ _ -.. __ _ _ x _ _ - . _ _ .
l 4 a RCS Loops-MODE ~4 B 3.4.5-BASES-1
~
ACTIONS B.1 and B.2 (continued) ((Tnseff E he lequiredl Jattemoval.- With only - -DHR loop _0PERABLE, redundancy ((Elete[3 for decay heat removal is lost and, . in .the event of a_ loss f inTe~rt]p of the, remaining DHR loop, there is core- temperature margin
,47,gc^ ,_to g ling in MODE 5 ~ (s 200 F) than MODE 4 (200 F to 280 F). The Completion Time of 24 hours is reasonable, based on ,
operating experience, to reach MODE 5 in' an orderly manner and without challenging plant systems. C11:and C.2 a This Condition is not. entered when using the allowance in the Note to'the LCO to de-energize all RC and DHR pumps. If no RCS or DHR loops are OPERABLE or in operation, all operations involving a reduction of RCS boron concentration must be suspended and_ action to restore one RCS or DHR loop to OPERABLE status and operation must be initiated. Boron dilution requires forced circulation _for proper mixing, and the margin to criticality must not be reduced without forced flow. The immediate Completion Times reflect the importance of maintaining decay heat removal capability and forced circulation. The action to restore a loop to OPERABLE status must. continue until one loop is restored to operation. SURVEILLANCE SR 3.4.5.1 REQUIREMENTS This Surveillance requires verification every 12 hours that one DHR or RCS loop, with a minimum of one pump, is in operation to ensure forced flow is providing decay heat removal. Verification includes flow rate, temperature, or pump status monitoring. The 12 hour Frequency has been shown by operating practice to be sufficient to regularly-assess operating loop status. In addition, control room indication and alarms indicate loop status and will typically alert operations personnel to anomalons flow / loss ;' of flow, should this occur. (continued) E Crystal River Unit 3 B 3.4-25 Amendment No. 149 _ - --- , u v. - , --- - . , - , - . , -
ECCS-Shutdown B 3.5.3 8 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) B 3.5.3 ECCS Shutdown BASES BACKGROUND The Background section for Bases B 3.5.2 is applicable to these Bases, with the following modification. The ECCS flow paths consist of piping, valves, heat exchangers, and pumps, such that water from the borated water storage tank (BWST) can be injected into the Reactor Coolant System (RCS) following the accidents described in Bases 3.5.2. APPLICABLE The Applicable Safety Analyses section of Bases 3.5.2 is SAFETY ANALYSES applicable to these Bases. Due to the stable conditions associated with operation in MODE 4 and the reduced probability of occurrence of a-Design Basis-Accident-(DBA), the ECCS operational requirements are reduced. Included in these reductions is that certain automatic Engineered Safeguards Actuation System (ESAS) actuation is not available. In this MCOE sufficient time ggg 7, exists for manual actuation of the required ECCS to mitigate the consequences o DBA. !accidentrequirihgy i
- ECCSLinjection 4-
- i!L:
1 L a -- >
~._
(continued) Crystal River Unit 3 B 3.5-20 Amendment No. 149
ECCS-Shutdown B 3.5.3 BASES
'APPLICABLET ~ ~ . 7An?LPITsubsist'emidedicst'e'd'toitakingissction;frbef the BWSTU SAFETY ANALYSISL rand:not;used forLdecay:heatiremoval, minimizesithe?timesforj l(continued)? l operator /l actions ~;to;_ initiate LPI,and reduces operator. burden
- since re-alignmentfof the DHR-; system to:the ECCS mode of> ;
M l operation: wouldi not: be 'necessaryh ! Additionally,^ LPI ? b . l t . capability is' protected since:a:breaktof the.reactorecoolant~ h .outletcline could:possibly_damageLthe DHP.=pumpLwithout: 2 j mitigating ~operatoriactions:.The time available-for operator' r j actionsJare.improvedibased on~the; ability;to detect RCS R - _ La _ _leakagelasia.LOCALprecursorT.L T _ i tzo m- o m a.=a Only one ECCS train is required for MODE 4. This {EsErif requirement dictates that single failures are not considered during this MODE LCO In MODE 4, one ECCS train is required to ensure sufficient ECCS flow is available to the rare following a-OBA. In MODE 4, an ECCS train consists sf an HPI subs stem and an LPI subsystem. Each train includes the piping, i struments, and controls to ensure un OPERABLE flow path capab e of r taking suction from the BWST upon an ESAS signal an ---+ l nsert] e an -:- t manually transferring suction to the reactor building emergency sump. QC , ECCS- [i_njection'. 1 (continued) Crystal River Unit 3 B 3.5-20a Amendment No. 149 i
. .. L
ECCS-Shutdown.
.B 3.5.3
-BASES
.LCO: .
~ During an event requiring ECCS actuation, a-flow path is (continued) required to provide an abundant supply of water from the e BWST to the RCS, via the ECCS pumps and their. respective:
discharge flow paths, to each of the four cold leg injection nozzles and the reactor vessel. . In the long term, this flow path may be switched to take its supply -from the~ reactor building- emergency sump and to supply its flo.v to the RCS . hot and cold legs. This LCO is modified by a Note which states that HPI may be deactivated,in accordance with Low Temperature Overpressure UEFtg; hCCident'--/ Protection (LTOP) administrative controls. Operator action is then required to initiate HPI. In the event of a-oss-ef coelent-accident-{LOCA) requiring HPI actuation, the t:me required for operator action has been shown by analysis to be acceptable. APPLICABILITY In MODES 1, 2, and 3, the OPERABILITY requirements for the
.ECCS are covered by-LCO 3.5.2 ."ECCS Operating."
In MODC 4 with the RCS temperature below 280 F, one OPERABLL' ECCS train is acceptable without single failure consideration, on the basis of the stable reactivity conditirn of the reactor and 'the limited core cooling require nents.
. In MODIS S and 6, plant conditions are such that the probability of an event requiring ECCS injection is extremely low. Core cooling requirements in MODE 5 are addressed by LCO 3.4.6 "RCS Loops - MODE 5, Loops Filled,"
and LCO 3.4.7, "RCS Loops - MODE 5, Loops Not Filled." MODE 6 core cooling requirements are addressed by LCO 3.9.4,
" Decay Heat Removal and Coolant Circulation High Water Level," 'and LCO 3.9.5, " Decay Heat Removal and Coolant Circulation Low Water Level."
(continued) Crystal' River Unit 3 8 3.5 Amendment No. 149.
ECCS-Shutdown B 3.5.3 BASES ACTIONS ful If-no-LPI-subsystem-is-OPERABLEr-the-unit--is net-prepared-to -
- respond-to-a-LOCA-or to- continue-cooldown-using-the-OHR/LPI pumps-and-d ecay-hea t-hea t-e xch angers-The-i mmed i a te Completion-Time-ensures-that prem+t- action-is-initiated-to res to re-t he-requi red-cooHng-eapacity---Normallyv-in-MODE-47 reactor-deesy-hest-must be removed by a DHR/tPI-train
[ dei _ete] ._] + ope ra ti n g-wi th-s ueti ofr-from-t he-RC S , If no DHR/LPI-train-4s OPERABLE-for-this-ftinction,- reactor-decay-heat-must-be remove d-by-s ome-a l t ern a te-me t hodT-s uch-a s-tt s em f-th e-ste am generator (t)-{OTSG) --The-alternate-means-of-heat-remeval must-continue-until-the-inoperable-EECS-t.fI-subsystem-ean-be re sto red -to-ope ra ti on-s o-th a t-con ti n u a ti on-of-de edy-hea t removal-(DHR}-is-providedr
- If noTECCS' LPIis'utssystem3s 0PERABLE,ldue0 t o thei _ '
{inoperability ofithe LPI'pumpc or flow ' path from1 the BWST,1 :i ithe plant zis;notfprepared to. provide flow- pressureiresponse : f ;tofaccidents' requiring ECCS responsed The Completion Timers rfj - nse g -j to initiate the restoration 'of tat least one ECCS LPI:
'l subsystem to OPERABLE! status ensureshthat prompt; action;is m
.. X taken:.to provide the = roquired; cooling capacity'or to. .D l initiate, actions c to : place! the: plantiin MODE : 5, L where Lan ; ECCS ltraintisLnot! required m zi_ b ma. -.1 A . 1 maa IL1 If no ECCS HPI subsystem is OPERABLE, due to the inoperability of the HPI pump or flow path from the BWST, the plant is not prepared to provide high pressure response rjI -
to Design-Basis-Events requiring ECCS response. The 1 hour 3 lc .. th Cop pfetion Time to restore at least one ECCS HPI subsystem
. o ;m to OPERABLE status ensures that prompt action is taken to
~ ~ ~
provide the required cooling capacity or to initiate actions to place the plant in MODE 5, where an ECCS train is not required. (continued) Crystal River Unit 3 B 3.5-22 Amendment No. 149
- ~. - . - ~ . - - .. - - - - - .. . . . _ . . , .- _. ~ -_. . . -- .
ECCS-Shutdown-B 3.5.3 BASES-3 ACTIONS.. L1
-If the Required Actions and associated Completion Times are not met. . the plant must be placed in a MODE in which the 1 Specification does not apply. . When the Required Actions of Condition B cannot be completed within the associated :
Completion Time, a controlled shutdown should' be initiated.. provided adequate decey heat-ren. oval-capability-exists. The , allowed Completion Time of 24 hours is reasonable, based on :' operating experience, to reach MODE 5.from MODE 4 conditions r( @ ~it3 ] L -in an orderly manner and without challenging plant systems. ' Should-adequate decey heet-removal-capability net existm Required-Action A.1 net-be ceapleted-within-its- associated , Complet4on-Time, censideration-should-be givea te pursuing Di1reretionary-Enforcement free-the-NRC en--the requirement-to proceed to "0DE-57 SURVEILLANCE 51 3.5.3.1 REQUIREMENTS The applicable Surveillance descriptions from Bases 3.5.2 apply. This-SR-is-modified-by-a-Note-which slie 3 a Olin train to be considered-OPERABt:E-during-alignaent end cperation-for decay : heat-removal, if capable of being-manually realigned Creeete < [5Elete)3 *for-local)--te the CCG mede inoperabler--This-allows of speration-and-cperstien in the Oi R medenot during ether ise MODE 4, if necessaryr REFERENCES The applicable references from Bases 3.5.2 apply, f i p 1 l- (continued)
+
Crystal- River Unit 3 8 3.5-23 Amendment No. 149
- - -, , - . - - , . -..ev ..w ~ n , n.. . - . . _
. - . . . - . . . , , - - + . ~ , .,
TBVs B 3.7.4 BASES LCO Each TBV (two per OTSG) is required to be OPERABLE for this LCO. Failure to meet the LCO can result in the inability to cooldown to DHR System entry conditions-following a SGTR event while maintaining offsite doses to a minimum. A TBV is considered OPERABLE when it is capeble of providing a controlled relief of the main steam flow, and is capable-of fully opening and closing when manually commanded to do so by the operator. APPLICABILITY In MODES 1, 2, and 3, the pressures and temperatures in the RCS are high enough to initiate a SGTR and require secondary side depressurization. Therefore, the TBVs are required to be OPERABLE in these MODES.
~
The'APPCICABILITYfisfmodified!by(a' note'whichfonlp~fequires3 the!TBVs associated with sthe RCS loop (s) required ' operable ~ !
* '# ** * ~ " ' "
(Insertl d
~InTMODEI4E EFW ma9'befusedi to 'fremoVe; coredesapIheat} int 'l j laccordanceiwith thefBases Bs3,7.18.- -Theilicensing-basist
- l does not require Lthe assumption of alloss'of offsite. powep resulting;from;an; accident in MODETand,:therefore,e the d iTBVs s are avail abl e ,fori secondaryiside; heat . removal mica In MODES 4, 5, and 6, a SGTR is not a credible event due to the reduced stresses in the generator tubes and low driving head for release to the environment.
ACTIONS A.1 and A1 2 With one or more Tbv(s) inoperable, action must be taken to restore all TBVs to OPERABLE status. The 7 day Completion Time is asonable to repair inoperable TBVs, based on the availabiHty of other means of depressurizing the RCS following a SGTR, and the low probability of this event occurring during the 7 day period. As an alternative to restoring the TSV(s) to OPERABLE status, the associated OTSG ADV must be verified to be OPERABLE within 7 days. This entails verifying that SR 3.7.4.1 is " current" for the ADV, or performing the Surveillance. Reliance on the ADV to satisfy the ACTIONS of this Specification is considered acceptable based on the early analysis. (continued) Crystal River Unit 3 8 3.7-21 Amendment No. 149
-TBVs l B 3.7.4 ;
-BASES'-
ACTIONS ( . 321 and B.2 (continued) -
-If the TBVs cannot be restored to OPERABLE status within the associated-Completion Time, the plant must-- be placed in a
_ _. l MODE in which the LCO does not apply. To achieve this [in_s_e_r_t)_r W1 - status, the plant must be placed .in at least MODE 3 within-WWE liin?._ : 6 hoursgM00C 4 within-liHvours. The allowed-30thours M Cvmpicuon Times are reasonable, based on operating experience,- to reach the required plant conditions from full power- conditions in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3.7.4.1 1 REQUIREMENTS To perform a controlled cooldown of the RCS. the TBVs must be able to be opened remotely and throttled through their full range. This SR ensures that the TBVs are tested through a full control- cycle at least once per fuel cycle. , Performance of inservice testing or use of a TBV during a'
-plant cooldown satisfies--this requirement. Operating experience has shown that these components usually pass the Surveillance when performed at the 24 month Frequency.
Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. REFERENCES 1. FSAR, Section 10.2.1.4. 7 FSAR. Section 14.2.2.2.
, ...m__m.. m l
l i
~
(continued) LCrystalLRiver Unit 3 B 3.7-22 Amendment No. 149 1 i _.
- - . . - - ~ ,
I EFW-Syotem ! 8 3.7 PLANT SYSTEMS : ((irThilSystem B 3.7.5 Emergency feedwater (EFW) System [ 'perating O ~ ~ " ~ ~
~. ~~
BASES BACKGROUND The Emergency feedwater (EFW) System is designed to provide adequate flow to one or both steam generators (OTSGs) for decay heat removal with the generators at the maximum operating pressure of 1050 psig plus suitable margin for post-accident pressure increase (Ref. 1, 2). The principal function of the EFW system is to remove decay heat from the Reactor Coolant System upon the unavailability of normal feedwater supply. This is accomplished by supplying water from the emergency feedwater tank (EFT-2) to the OTSG secondary side via the high nozzles. Steam produced in the OTSGs is condensed in the main condenser via the turbine bypass valves or, if the atmospheric dump valves (ADVs) or main steam safety valves (M55Vs) have actuated, discharged directly to the atmosphere. The EFW System consists of one motor driven EFW pump and one steam turbine driven EFW pump, each having a nominal 100% capacity (Ref. 3). The motor driven EFW pump is powered from the 4160 volt ES bus 3A. The turbine driven EFW pump receives steam from one main steam line per OTSG via connections upstream of the associated main steam isolation valve. An alternative source of steam is available from the fossil units, Crystal River Unit 1 and 2 (Ref. 1), but cannot be relied upon to consider the EFW train OPERABLE. The diverse motive power of the two trains enhances both system availability and reliability. The preferred water source for both EFW pump trains is the Seismic Class I, missile protected ot.dicated EFW tank. Backup supplies of l e argency feedwater are provided by the condensate storage tank and the main condenser hotwell. The pumps tie into common discharge headers providing the capability to feed either or both of the OTSGs. DC powered block and control valves are actuated to feed the appropriate steam generator by the Emerger. y Feedwater Initiation and Control (EFIC) System. The capacity of either EFW pump is sufficient to remove decay heat and cool the plant until the Reactor Coolant System (RCS) pressure and temperature are low enough to place the Decay Heat Removal (DHR) System in service. (continued) Crystal River Unit 3 8 3.7-23 Amendment No. 149 l l
hseFijiEFW;5ystE-OperatI5i) B
. BASES : - BACKGROUND' Automatic actuation of the EFW System occurs on the (continued) following:
- 1. Trip of both main feedwater pumps with reactor po..er greater than 20% or the NI/RPS not in shutdown bypass;
- 2. Low level-in either OTSG; ,
- 3. Low pressure in either OTSG;
- 4. Trip of all four reactor coolant pumps;
- 5. High pressure injection (HPI) actuation on both Channel A and B Engineered Safeguards Actuation System (ESAS) channels; and -
- 6. AMSAC actuation.
The EFIC is a " smart" system which will feed either or both OTSGS with indications of low levels,-but will isolate EFW to ' faulted steam generator having a significantly lower steam pressure than the other. The EFW System is designed to ANSI B 31.1 ES Seismic Class I and in accordance with General Design Criteria 2, 4, 5,19, 44, 45, and 46 (Ref. 3, 4). APPLICABLE The EFW System is sized to provide sufficient decay heat SAFETY ANALYSIS removal capability to cooldown the RCS to the temperature and pressure at which the DHR System can be placed in . service for any of the following events: J
. loss of main feedwater (LMFW);
. LMFW with loss of offsite power;
-. main feedwater line break;
. main steam line break; and
. small break loss of coolant accident (LOCA).
The EFW System is designed to remain functional following the maximum hypothetical earthquake. It will also remain (continued) Crystal River Unit 3 B 3.7-24 Amendnwnt No.149
~
"I'# Systent
}i~nseft] 5MSystem-OperaW9] 8 3.7.5 BASES APPLICABLE functional following a single failure in addition to any of SAFETY ANALYSES the above events with the exception of the loss of all AC (continued) power (Ref. 3) No single failure prevents EFW from being supplied to the intact OTSG nor allows EFW to be supplied to tne faulted OTSC. Note that in most cases of a main feedsater break or a steam line break, the depressurization of the affect < OTSG would cause the automatic initiation of EFW. However, there will be some small break sizes for which automatic detection will not be possible. For these small breaks, the operator will have sufficient time in which to take appropriate action to terminate the event (Ref. 1).
The EFW System satisfies Criterion 3 of the NRC Policy Statement. LCO Two independent emergency feedwater pumps and their associated flow paths are required to be OPERABLE. The OPERABILITY of the EFW pumps requires that each be capable of developing its required discharoe pressure and flow. Additionally, the OPERABILITY of the turbine driven pump requires that it be capable of being powered from an OPERABLE steam supply through ASV-5. ASV-204 was installed to improve EFW reliability and is not required for OPERABILITY, The motive power for the turbine driven pump is supplied from either OTSG from a main steam header upstream of the main steam isolation valves so that their closure does not isolate the steam supply to the turbine. Both steam supply flow paths to the turbine driven pump are required to be OPERABLE. The OPERABILIT: of the associated EFW flow paths requires all valves be in their correct positions or be capable of actuating to their correct positions on a valid actuation signal. Inoperability of the EFW System may result in inadequate decay heat removal following a transient or accident during whicn main feedwater is not available. The resulting RCS heatup and pressure increase can potentially result in significant loss of coolant through the pressurizer code safety valves or the PORV. The LCO is modified by a Note indicating that only one EFW train, which includes a motor driven EFW pump, is required in MODE 3 with OTSG secondary side pressure < 200 psig. At (continued) Crystal River Unit 3 B 3.7-25 Amendment No.149 l l
{_nsert] EFW System-Operatingl jFW-System B 3.7.5 BASES LCO less than 200 psig, the turbine driven EFW pump is not (continued) capable of producing the accident analysis flow rate. Under these conditions, the flow rate produced by the pump may well be adequate to perform the heat removal function. However, this has not been analytically demonstrated. Thus, the allowance to not have this pump OPERABLE is acceptable based upon the reauced heat removal requirement and the short duration of this condition. Similar to the concept employed in LCO 3.5.2 "ECCS Systems - Operating," cross-connecting portions of both trains to produce a single OPERABLE train, is acceptable in this plant condition. APPLICABILITY In MODES 1, 2, and 3, the EFW System is required to be OPERABLE and to function in the event that main feedwater is lost. In addition, the EFW System is required to supply enough makeup water to replace the setondary side inventory lost as the plant cools to MODE 4 conditions. [Dn'si}[] N\The" req'uirements for-EFW irs' MODE'4 are addressed by gggg Technical Specification 3.7,18.3 .. s . .
. MODES 5; andj AMODES-47-57-and-6, the OTSG need not be used to cooldown E 1 -- _J the RCS. Therefore, the EFW System is not required to be OPERABLE in thes2 MODES.
ACTIONS Ad With one of the two steam supplies to the turbine driven EFW pump inoperable, action must be taken to restore the steam supply to OPERABLE status witnin 7 days. Allowing 7 days in this Condition is reasonable, based on the redundant OPERABLE steam supply to the pump and the low probability of an event occurring that would require the inoperable steam supply to the turbine driven EFW pumps. The second Completion Time for Required Action A.1 establishes a limit on the maximum time allowed for any combination of Conditions to be entered during any continuous failure to meet this LCO. The 10 day Completion Time provides a limitation time allowed in this specified Condition after discovery of failure to meet the LCO. This limit is considered reasonable for situations in which Conditions A and B are entered concurrently. The 'AN_D' connector between 7 days and 10 days dictates that both Completion Times apply simultaneously, and the more restrictive must be met. (continued) Crystal River Unit 3 8 3.7-26 Amendment No. 149
"P# Sy5bern
){erQEFW Syst'em-prat {p9] B 3.7.5 l BASES ACTIONS - R11 '
(continued) If one of the EFW trains is inoperable, action must be taken to restore the train to OPERABLE status within 72 hours. The 72 hour Completion Time is reasonable, based on the redundant capabilities afforded by the EFW System, time needed for' repairs, and the low probability of a DBA occurring during this time period. This Condition includes the loss of two steam supply lines to the turbine driven EFW pump. The second Completion Time for Required Action 8.1 establishes a limit on the maximum time allowed for any combination of Conditions to be entered during any continuous failure to mcet this LCO. The 10 day Completion Time provides a limitation time allowed in this specified Condition after discovery of failure to meet the LCO. This limit is considered reasonable for situations in which Conditions A and B are entered concurrently. The 'aNQ' connector between 72 hours and 10 days dictates that both Completion Times apply simultaneously, and the more-restrictive mest be met. C.1 and Ct 2 If Required Action A.1 or Required Action B.1 cannot be completed within the associated Completion Time, the plant must be placed in a MODE in which the LCO does not apply. To achieve this status, the plant must be placed in at least MODE 3 within 6 hours and.in MODE 4 within 12 hours. The allowed Completion Times are reasonable,- based on operating experience,-to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. D_d With both EFW trains inoperable, the plant is in a seriously ' degraded condition with no safety related means for conducting a cooldown. In such a condition, plant operation should not be perturbed by a forced action, including a power change, that might result in a trip. For this reason, l t (continued) Crystal River Unit 3 8 3.7 Amendment No. 149
TA' System [DEerWEFSystem?Ofe3f9] B 3.7.5 BASES ACTIONS Q11 (continued) the Technical Specifications do not mandate a plant shutdown. Rather the ACTIONS allow the plant to dictate the most prudent course of action (including plant shutdown) for the situation. The seriousness of this condition requires that action be initiated immediately to restore at least one EFW train to OPERABLE status. SURVEILLANCE SR 3.7.5.1 REQUIREMENTS Verifying the correct alignment for manual, power operated, and automatic valves in the EFW water and steam supply flow paths providos assurance that the proper flow paths exist for EFW operation. The valves verified by this SR include valves in the main flow paths and the first normally closed valve in a branch line. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since those valves are verified to be in the correct
~
position prior to locking, sealing, or securing. There are several other exceptions for valve position verification due to the low potential for these types of valves to be mispositioned. The valve types vehich are not verified as part of this SR include vent or drain v-ive outside the Rd, relief va'ves outside the RB, and instre.u ation valves (both inside and outside the RB). This SR also does not apply to valves that cannot be inadvertently misaligned, such as check valves. This Surveillance does not require any testing or valve manipulation; rather, it involves verification that those valves capable of potentially being mispositioned are in the correct position. The 45 day Frequency is based on engineering judgment and is consistent with the Frequency established for SR 3.7.5.2. SR 3 7.5.2 requires extensive EFW valve manipulation in order to perform the pump flow rate verification, such that a flow path verification is
- .ecessary following each performance.
SR 3 . 7 . 5 . 2, This SR verifies that the EPf pumps develop sufficient discharge pressure to deliver the required flow at the full open pressure of the MSSVs. Because it is undesirable to (continued) Crystal River Unit 3 B 3.7-28 Amendment No. 149
)
- . -- . ~ . . - - .- -- - - - - .
.m .-_,_
~kI N .*2 1 D I N E. Q. +jfW: System B 3.7.5 BASES SURVEILLANCE SR 3.7.5.2 (continued)
- 1 REQUIREMENTS introduce cold EFW into the OTSGs while they are operating, this test is performed on recirculation at a reduced flow-rate.
Periodically comparing the reference differential-pressure developed at this reduced flow detects trends that might be indicative of degrading pump performance. Performance of inservice testing discussed in the ASME Code,-Section XI (Ref. 5), at 3-month intervals, is satisfied by this SR. The 45 day Frequency on a STAGGERED TEST BASIS results in testing-each pump once every 3 months, as required by Reference 5. This SR is modified by a Note indicating that the SR may be
. deferred until suitable test conditions are established.
This SR 3.0,4 type exception may be necessary during any plant start-up because there is insufficient steam pressure in the secondary side of the OTSCs to perform this SR on the turbine-driven pump. SR 3.7 123 This SR verifies that EFW can be delivered to the appropriate OTSG in the evant of any accident or transient that generates an EFIC signal by demonstratino that each automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, actuates to its correct position on an actual or simulated actuation signal. Valves secured in the correct position need not demonstrate the capability to achieve this configuration. ADVs also need not demonstrate the capability to satisfy this SR since their operation is not credited as part of any DBA. The SR also verifies the EFW control and block valves attuate to the isolation position on a simulated or actual vector valve control signal. _This SR is a test of the integrated system response to an _ actuation signal and as such, it is not necessary to verify-the EFW System actuates on each EFIC signal. Any of the initiation signals- described in the Background Section of these Bases is. adequate, given that the various EFIC (continued) Crystal River Unit-3 8 3.7-29. Amendment No. 149 1
,N_"_.5f_t[
I Q Systey p eya h gj_ ._B 3.7.5 _.___ jfW-System BASES SURVEILLANCE L L 1,l M (continued) REQUIREMENTS instrumentation is surveilled separately in Section 3.3
" Instrumentation" of the ITS.
The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during an outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. The 24 month Frequency is also acceptable based on operating experience and design reliability of the equipment. This SR is modified by a Note indicating that the SR may be deferred until suitable test conditions are established. This SR 3.0.4 type exception is necessary because there is insufficient steam pressure to perform the test while in MODE 4. SIL_3.LM This SR verifies that the EFW pumps start in the event of any accident or transient that generates an EFIC initiation signal by demonstrating that nach pump starts automatically on an actual or simulated actuation signal. This SR is a test of the integrated system response to an actuation signal and as such, it is not necessary to verify the EFW system actuates on each EFIC signal.. Any of the initiation signals described in the Background Section of these Bases is adequate, given that the various EFIC instrumentation is surveilled separately in Section 3.3 " Instrumentation" of the ITS. The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a unit outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. This SR is modified by a Note indicating that the SR may be deferred until suitable test conditions are established. This SR 3.0.4 type exception is necessary because there is insufficient steam pressure to perform the test while in MODE 4. SR 3.7.5.5 This SR ensures that the EFW System is properly aligned by verifying the flow paths to each OTSG prior to entering MODE 2 after more than 30 days in MODE 5 or 6. OPERABILITY (continued) Crystal River Unit 3 B 3.7-30 Amendment No. 149
_ . . . _ _ _ - ~ _ _ _ _ _ . . . _ _.. _. _ . _ _ ~ . _ _ _ . _ _ _ . _ _ . _ _ _ _ _ . - . - . i r ew - . ~ .- > [i y e'iifIs { 5 d i W W p E % ] B 33.5 i I BASES i SURVEILLANCE $R 3.7.5.5 (continue]) REQUIREMENTS i of EFW flow paths must be demonstrated before sufficient ! core heat is generated that would require the operation of the EFW System during a subsequent shutdown. The Frequency ' is reasonable, based on engineering judgment, in view of other administrative controls to ensure that the flow paths . are OPERABLE. To further ensure EFW System alignment, flow ! path OPERABILITY is verified, following extended outages to : determine no misalignment of valves has occurred. This SR ensures that the flow path from the EFW tank to the OTSGs is properly aligned. This requirement is Sased upon the . ; recommendation of NUREG 0737. The Freqsency was. modified ; slightly during ITS dr.slopment (prior to entering MODE 2)
- to provide an SR 3.0.4 type exceptian. As written, the SR '
allows the plant to achieve and maintain MODE 3 conditions in order to perform the verification. REFERENCES 1. Enhanced Design Basis Document for the Emergency Feedwater and Emergency Feedwater Initiation and
- Control System, Revision 1, dated September 27, 1991 with Temporary Changes 156, 230, 247, and 249.
- 2. BAW-10043, " Overpressure Protection for B&W Reactors",
dated May-1972.
- 3. FSAR, Section 10.5.
- 4. 10 CFR 50, Appendix A.
- 5. ASME, Boiler and Pressure Vessel Code, Section XI.
Inservice Inspection, Subsection 1WP.
/
(continued) i Crystal River Unit 3 8 3.7-31 Amendeent No. 149 9
?
. , - . . . , - . - - . -.-m.- . .. . . , , - _ , . , _ . . - -. _, . - . . . _ . . . _ _ . . _ . , , , , _ . , . _ _ ,
Emerg:ncy Fccdmater Tcnk l B 3.7.6 i BASES APPLICABLE was not used as an input to these safety analyses. 1 SAFETY ANALYSES OPERABILITY of the EFW System, and therefore the (continued) EFW tank. is essential to the mitigation of the following svents (Ref. 3): i Loss of main feedwater (LMFW) i LMFW with a loss of offsite power , Main feedwater line break : Main steam line break Small break loss of coolant accident (LOCA) The required minimum volume of usable condensate in the EFW tank is 150,000 gallons. This amount is sufficient to remove decay heat for a period of 18 hours at MODE 3 conditions (Ref. 4). This time-period is considered adequate to allow plant conditions to be stabilized and another source of-water to be made available for natural circulation cooldown until MODE 4 is achieved. In MODE 4, the RCS pressure will be decreased to the point that allows - the alignment of the Decay Heat Removal (DHR) System to the
- RCS.
Fo6eventsToccs r H ng fi n? MODE f 4 E theT EFW itaink"woul d " provi de 7 the;safetylgradetsourceiof_wateritoatheiEFWl System.i .u s d Although the single failure criteria is applicable to the EFW System in the evaluation of the previously mentioned events, the EFW tank performs its safety function in a passive manner and is thereby excluded from application of the single failure criterion. The EFW tank satisfier. Criterion 3 of the NRC Policy Statement. LCO In the event of a loss of offsite power, or other conditions resulting in a complete loss of main feedwater, a means of removing heat from the RCS must be immediately available. i(fise7t[ The EFW tank minimum usable water volume limit of 150.000
'Forievent]s? 9allons is necessary to provide assurcnce that the EFW
'or.curringK System can supply the volume of secondary coolant needed to Kn- N00ESil,' remove decay heat in MODE 3 conditions for 18 hours (Ref. 4) while "~ sources of water are made available for
{and3l.o.1
~
""" subsequent cooldown to below 280 degrees, if required.
(continued) Crystal' River Unit 3 B 3.7-33 f.mendment No. 149
. . . - .. - ~ . . _ . . - -.. - _ . .
- = - _ . -- . . .- .. - - - _ . - - - - _ _ - - - . - - . - - . . ._.
Emerg2ncy F02daater Tank , B 3.7.6 l l BASES (continued) _
-y
- m. ~ , , - -. - - .- - -
3 g
- (continued)'< Forl such events occurring i(MODE 4, the Eip tank would! ,
i
. s eJ iam m-, Apryide:a;safetygradesourceofifeedwater.m
/Cc,vitance with the LCO is verified by maintaining tank ,
.[ Insert) + - ~ level at or above the minimum requirs.d level. j l
}
APPLICABLITY In-M00ES-1, 2, and-3 the OTSGs are the operating heat sink (Insert);In for RCS h t removal. The EFW system, and thus, the EFW MODES I.-2, 3,t tank, m t be OPERABLE during these MODES, to assure the pnd.4L Evallauility of a safety grade means of RCS heat removal following any event which results in a loss of main feedwater. In--M00ES-4---5, and-6, the EFW Tank water volume is not required to be ' thin limits consistent with the requirements for EFW Sys em OPERABILITY. r _ t[ Insert] eIri M00ES 5 and 6,JL-ACTIONS L Land A Z As an alternative to restoring the EFW Tank volume to within limit, the OPERABILITY of the backup water supply can be verified within 4 hours and once every 12 hours thereafter. The OPLRABILITY of the backup feedwater supply must include verification, by administrative means, of the OPERABILITY of flow paths from the backup supply to the EFW pumps and availability of the required volume of water in the backup supply. Typically, the condensate storage tank is the preferred back-up since the volume in this tank is available without the time delay associated with having to ' break' condenser vacuum in the hotwells. The EFW Tank volume must be restored to within limit within - 7 days because the backup supply is not designed to the same , criteria as the EFW tank and may be fulfilling the requirements of this Specification in addition to its normai operational functions. The 4 hour Completion Time is reasonable, based on operating experience, to verify the OPERABILITY of the backup water supply. The 7 day Completion Time is reasonable, based on an OPERABLE backup water supply being available, and the low probability of an event occurring during this time period, requiring the use of the water from the EFW Tank. (continued)
. Crystal River Unit 3 8 3.7-34 Amendment No. 149
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Emerg:ncy FC dnater Tank B 3.7.6 BASES (continued) , l ACUONS !!d aad_L2 (continued} , If the EFW Tank volume cannot be restored to within limit in the associated Completion Time, the plant must be placed in a MODE in which the LCO does not apply. To achieve this y[ insert).___.__-,MODE 54 6status, hours, and the plant in MODE-+-with4a-12-hours. must be placed in at least MODE 5 within l withinj30. hours ' The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. l i (continued) Crystal River Unit 3 8 3.7-34a Amendment No. 149
ggnmmpwyyy r ~ m y y n m m e ~ n m gy;gg,,,,y gg,,g,,g L u . A h s wd d b m h d a _ m e m M & E 'i l t5L3.7;18 I % Insert; new specification)) : y,J ,.7+,PLANL 5YSTENs w w w n,vwnenm w a.wnrw . a u a u Lm n- e. e n m n ,. x wu u . wua .m 0 % 738! Embreenc9? FeedwaterI(EPW)? $9stinM $hutdownWCH2X?2CAU MStianwau, w e m m waar , - cn m w aa m mn a:w.w y- x-y m m - ~ c - - .mwnm y r.emwy m u .a i l l N"" #ThitaderWnd3dEtiWfoF8asesit~l3]7;5Eappliciabis;to'3 ! it d a d a a d h se Bases.n b & xm . . , m:ca.uAhna .3 ! m - m- .
=
APPLICABLE,N. ..,,M. ne,>,,Ap p .licable' Safe..ty: -. ._,. Anal,y, sis- :_s. 3ct. . oi~n L of,Ba~ses'. 3.y 7.c5.[.is,p aq SAFEl% ANALYSISLjasplicable tolthese10aseshim h a anec e n ' s onditio. -ns. asso,da teAw .. . ith_:operatio_n,: b_ue. Lto~th_e stab,le_c. . i,n ,a, i MODEi4 Land:the reduced' probability bf!occurrencelof an event
' requiring EFW !the EPW operational? requirements 3areireduced? -i
! Included'in thesesreductions is thatiautomatic4mergency: d ,
.FeedwateriInitiation, and Contro1*$ystem (EFIC)1actuationSis j
~notJavailable.1 In this MODE,< sufficient timeLexists1for 9 d manual; actuation ^iof EFW. to mitigate 3heionsequences .of and !
' event ; requiring EFWh:.JMGReca um& &M .
~ . . . ,
only,r_onet train _o.f,EFW capable._f ,o <;b_ein_g su_pplie~d J from e theE y l meter!dr,1ven pump'isfreguiredjin MODE:4. eThis; requirement 9 - ' dictates?that singleffailures areinoticonsidered.during(this' NODE.z a us u u idM n.a ~ ; w a w n .:
}
LC0 id1C hlhe1LC0 Mecti onsofs Basei; 3 lT.' 5 ".Li s Espol i cablifto ithe'selta'ses? In )DDEE4 ?an EFWitraiF capible~of(being lupplied^byXthi M I
'moter' driven' pump?is4 required l tol ensure; sufficient' decay! , ; '
_ heat = removal)capabilityNollowinman eventirequiring EFWe a employed JniLCOT3.5.2,j"ECCS Systeery - [pisilarttothe-concepts 0peratingF'scross3onnecting por.tions ' of b prodscota(s hgle: operable train:1s?acceptabletinithistplanti 4 condition.w M Q c A E A A d n u s s u m ~i d y-.,--m n ~_j yr 4s -._.py. # o . y .r,g., .c. -
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stor, pump' s: cons er - - .
; etuotor. .v MODE: 4 'wnh ii t'/ hand i swi tck put ki n pul l kto llockil n order; tol
'preventlinadvektent EFW actuationidhamm mesamnd pgmygwy:5p9emm ~mgpr wmmmm eym- Pyr(dondssd7 CrystahR%erluniCl?EGFGEEs?):7isGE M B E FZ Amendment K 2i m
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.ApplICASILITYWRInEMODEE4pthe!EFWisystem may1Wused7for1heatiremosalivia?? l N w _ L A ..l6 utheisteam; generators.L i xs i & n s.u.d a i m.m ;a.m. - ,
1
^ ^~
lIn M00ESIS"and'6Tt W steam gene'ratorsTarFnot"uded'foN " i
~ decay;. heat; removal andithe EFW System:isinot;requiredu.u d 1 I
i i gg;t;. Sp ) jy i s 'a g q* q*y v f 9 - p Sis' em g4.p y. ds. w ww 4 .g tgg OM.y' q g y jt ,W-79 -}_r-fpf sqwepgg
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With,, the,. 4:.smotor. driv., i,.en EFW . rtrain. Jg. (p, ;u.gmp or flowop
; inoperable,caction'eust be taken 11mmediately:to:restoreL thed=
' required 1trainito OPERA 8LELstatusMThe;immediate Completiori i Timeiensures5that Tprompt action d slinitiatoditoireste,re L the)
' required : cooling icapacity J J In the Dinteria,(alternativei 4 i
'means of: heat removal;must:be availablefsuch:asethes .
-3 auxi11arylfeedwater train-(FWp-7) or the; main;feedwateri j systenf capable.Lofl supplying 1 flow to the- emergency feedwaterM i
'nonles.T 4If restoration 'cannot b(ecreasonably; accomplished.1 lthe plant:must;be placed,in. Mode $41rnaccordance1with2 A1 Actionm8.1. h !/ & m w a a d L w , A m A m i ,, M T Y i f .6%%,f
;&. a'wa,, YWT wm'"en[oz C qr%
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I."l'a $:a"do a m a. , If Required Xttioh"Nf14asot bFcompleSidlinithin the"V?l
~
'associsted Completion,TimeL theiplant;must:be'placed:in LaJ M M
~
MODE in whi_.ch;the:LCO does not. apply.-4Tojachieve thisK r,tatusb the plant must.be!placed;incatnisast MODET5 within A
,241 hours. 3Theiallowed: CompletionlTimei1sL reasonableh based j on operating experience,6tojreach;the required. plant-E ,
conditions nf an orderlyimanner!and'without challenghgi d]
'plantxsystems.i>. m A w u e u _ u u w s u a .s e l f i
_ ym mnynwnmmyuymmm ~ vnnyvwmmqp gy,4,,,g Ecriss15M yedhii t!3%AJEE.ESI3 s7s47EE32322MAmendments Noi3HQ . t 1 e s -yw-a +.=-,...s,. v--- * ,--es-ted 1,w.- ,---e m+ .e v--*w-rr--+ww---*-+-m-,,w---r-w-w---enw,--+ wee-e ev.,-+v,--w. r w w-wr-- w- r+ vre -*a-'r' '
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- 5. . , ya " ~
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. - - . - - - - /
(( _-Insert _i__ new specification. ] ]: : SURVGILLANCE "" JSR *1J.'is.16 p' ' ' "i AEQUISEMENisia 1 ._ - . . . .
. .. .. . , - . f . a. < '2 The' applicable Surveillance ' descriptions from Bases 3.7;5: ;!
apply.. a m . , a . a m.m _. ~ .ma# vfLs,. ga,J
- - . m, 4 m - -. - - - - - - s- -e ~ . n . w, ,,
1,,.,a.Th,e!appliv. ca ble.~, referen .~ fro. m, Bases,3 ' 5. y .-- -C apply.,y a. ,J REFERENCES.-,D ; 0 ( . cesi . . W 4 1 I gn.7
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1 i FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 , DOCKET NUMBER 50-302/ LICENSE NUMBER DPR-72 ATTACIIMENT D LICENSE AMENDMENT REQUEST 214 REVISION 0 REVISION BAR PAGES Technical Specifications & Bases l
l ATTACHMENT TO LICENSE AMENDMENT NO. IAcillTY OPERATING LICENSE NO. DPR-72 ; 1 DbCKET NO. 50-302 , (Page-1 of 2) Replace the following sages of the Appendix "A" Technical Specifications with
- the attached pages. 11e revised pages are identified by amendment number and ;
contain vertical lines indicating the area of change. The corresponding ,
*spillover pages are also provided to maintsin document completeness.
Remove Rt01 ace ; iii 111 vil vii 3.3-30 3.3-30 3.3-31 3.3-31 3.3-38 3.3-38 3.3 39 3.3-39 3.4 8 3.4-8 3.4-9 3.4-9 , 3.5 8 3.5-8 3.7 8 3.7-8 3.7-9 3.7-9 3.7-10 3.7-10 ; 3.7-11 3.7-11 3.7-12 3.7-12 L 3.7-13 3.7-13 3.7-37 l B 3.3-94 8 3.3-94 8 3.3 100 B 3.3-100 B 3.3-101 B 3.3-101 B 3.3-102* 8 3.3-102 B 3.3-103 B 3.3-103* , B 3.3-104 8 3.3-104*
'B 3.3 113 8 3.3-113 8 3.3-139 B 3.3-139 '
!- B 3.3-141 B 3.3-141 B 3.4-22 B 3.4-22
- 8 3.4-23 8 3.4-23 B 3.4-24 8 3.4-24 8 3.4 B 3.4-25.
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l ATTACHMENT TO LICENSE AMENDMENT NO. 1 EACILITY OPERATING LICENSE NO. DPR 72 !
, , DOCKET NO. 50-302 )!
(Page2of2) i' Replace the following sages of the Appendix "A" Technical. Specifications with the attached pages. Tie revised pages are identified by amendment number and contain vertica linesjndicatingtFe-areaofchange. The corresponding-
*spillover pages are also provided to maintain document completeness. ;
Pamove Renlace . 1 B 3.5 20 B 3.5-20 -; B 3.5 21 B 3.5-21 8 3.5-22 8 3.5-22
- B 3.5 23 8-3.5-23 1 i
B 3.7 21 B 3.7 21 B 3.7-72 B 3.7-22 - B 3.7 23- B 3.7 23 - B 3.7-24 B 3.7 24 l B 3.7-25 8 3.7-25 , B 3.7-26 B 3.7-26 l 8 3.7-27 8 3.7-27 8 3.7-28 B 3.7-28 B 3.7-29 B 3.7-29 l B 3.7 30 8 3.7-30 B 3.7-31 B 3.7 B 3.7 33 8 3.7-33 B 3.7-34 8 3.7-34 B 3.7-35 8 3.7-35
....-. B 3.7 85
------ B 3.7-86 ,
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- _ - - - . - . - . - _ . ~ _ - - - - -..- __- -. - _-. TABLE OF CONTENTS 3.6 CONTAINMENT SYSTEMS (continued) 3.6.6 Reactor Building Spray and Containment Cooling Systems . . . . . . . . . . . . . . . . . 3.6 17 3.6.7 Containment Emergency Sump pH Control System (CPCS) . . . . . . . . . . . . . . . . . . 3.6 21 3.7 PLANT SYSTEMS .......... .......... 3.7-1 3.7.1 Main Steam Safety Valves (MSSVs .......... 3.7-1 3.7.2 Main Steam Isolation Valves (MS Vs) ........ 3.7-4 3.7.3 Main feedwater Isolation Valves (MFIVs) ...... 3.7 6 ' 3.7.4 Turbine Bypass Valves (TBVs) . . . . . . . . . . . . 3.7 8 3.7.5 Emergency feedwater (EFW) System-Operating .... 3.7-9 l 3.7.6 Emergency feedwater (EfW) Tank. .......... 3.7 13 3.7.7 Nuclear Services Closed Cycle Cooling Water (SW) System . . . . . . . . . . . . . . . . 3.7 15 3.7.8 Deacy Heat Closed Cycle Cooling Water (DC) System . . . . . . . . . . . . . . . . . . . 3.7 17 3.7.9 Nuclear Services Seawater System . . . . . . . . . . 3.7-19 3.7.10 Decay Heat Seawater System . . . . . . . . . . . . . 3.7-21 ' 3.7.11 Ultimate Heat Sink (VHS) . . . . . . . . . . . . . . 3.7-23 3.7.12 Control Room Emergency Ventilation Syster(CREVS) ................. 3.7-24 3.7.13 fuel Storage Pool Water Level ........... 3.7-27 3.7.14 Spent fuel Pool Boron Concentration ........ 3.7-28 3.7.15 Spent fuel Assembly Storage ............ 3 . 7 .M 3.7.16 Secondary Specific Activity ............ 3.7 34 3.7.17 Steam Generator Level ............... 3.7-35 3.7.18 Emergency feedwater (EFW) System-Shutdown . . . . . 3.7 37 l 3.8 ELECTRICAL POWER SYSTEMS . . . . . . . . . . . . . . . . 3.8-1 3.8.1 AC Sources-0)erating ............... 3.8-1 3.8.2 AC Sources--51utdown . . . . . . . . . . . . . . . . 3.8-11 3.8.3 Diesel fuel Oil, Lube Oil, and Starting Air .... 3.8-14 3.8.4 DC Sources-0)erating ............... 3.8-17 3.8.5 DC Sources-S iutdown . . . . . . . . . . . . . . . . 3.8-21 3.8.6 Battery Cell Parameters .............. 3.8-23 3.8.7 Inverters-0)erating . . . . . . . . . . . . . . . . 3.8-27 3.8.8 Inverters-S1utdown ................ 3.8 29 3.8.9 Distribution Systems-0)erating .......... 3.8-31 3.8.10 Distribution Systems-51utdown . . . . . . . . . . . 3.8-33 3.9 _ REFUELING OPERATIONS . . . . . . . . . . . . . . . . . . 3.9-1 3.9.1 Boron Concentration ................ 3.9-1 3.9.2 Nuclear Instrumentation .............. 3.9-2 3.9.3 Containment Penetrations . . . . . . . . . . , , . . 3.9-4 3.9.4 Decay Heat Removal (DHR) and Coolant Circulation-High Water Level . . . . . . . . . . 3.9-6 (continued) Crystal River Unit '3 iii Amendment No. _ .__ _ . , _ _ ~ . _ . - _ _ , _.
r TABLE OF CONTENTS l r B 3.6 CONTAINMENT SYSTEMS (continued) f B 3.6.7 Containment Emergency Sump pH Control (CMS) . . . . B 3.6 47 B 3.7 PLANT SYSTEMS .....................B3.71 8 3.7.1 Main Steam Safety Valves (MSSVs) . . . . . . . . . . B 3.71 ! B 3.7.2 . . . . . . . . B 3.7-7 i B 3.7.3 MainSteamisolationValves(MSIVs)Vs) MainfeedwaterIsolationValves(MFI . . . . . . B 3.7 13 ; B 3.7.4 Turbine Bypass Valves (TBVs) . . . . . . . . . . . . B 3.7 19 8 3.7.5 ....B3.723l B 3.7.6 Emergency Emergency feedwater feedwater Tank (EFW)(System-OpratingEFT-2) . . . . . . . . . , B 3.7.7 Nuclear Services Closed Cycle Cooling Water System (SW) . . . . . . . . . . . . . . . . B 3.7 36 B 3.7.8 Decay Heat Closed Cycle Cooling Water System . . . . B 3.7 41 B 3.7.9 Nuclear Services Seawater System . . . . . . . . . . B 3.7 46 B 3.7.10 Decay Heat Seawater System . . . . . . . . . . . . B 3.7 51 B 3.7.11 Ultimate Heat Sink (VHS) . . . . . . . . . . . . . . B 3.7 56
-B 3.7.12 Control Room Emergency Ventilation System (CREVS) . . . . . . . . . . . . . . . . . B 3.7 60 B 3.7.13 fuel Storage Pool Water Level . . . . . . . . . . . B 3.7 66 8 3.7.14 Spent fuel Pool Boron Concentration . . . . . . . . B 3.7 69 8 3.7.15 Spent Fuel Assembly Storage . . . . . . . . . . . . B 3.7 72 ,
B 3.7.16 Secondary Specific Activity . . . . . . . . . . . . B 3.7-77 ; B 3.7.17 Steam Generator Level . . . . . . . . . . . . . . . B 3.7 81 B 3.7.18 Emergency Feedwater (EFW) System-Shutdown . . . . . B 3.7-85 l 8 3.8 ELECTRICAL POWER SYSTEMS . . . . . . . . . . . . . . . . B 3.8 1 8 3.8.1 AC Sources-Operating ...............B3.81 , B 3.8.2 AC Sources-S 1utdown . . . . . . . . . . . . . . . . B 3.8-24 8 3.8.3 Diesel fuel Oil, Lube Oil, and Starting Air . . . . B 3.8 30 8 3.8.4 DC Sources-Operating . . . . . . . . . . . . . . . B 3.8 39 B 3.8.5 DC Sources-Shutdown . . . . . . . . . . . . . . . . B 3.8 49 B 3.8.6 Battery Cell Parameters . . . . . . . . . . . . . . B 3.8 52 : B 3.8.7 inverters-0'a erating . . . . . . . . . . . . . . . . B 3.8 59 l l B 3.8.8 inverters-51utdown . . . . . . . . . .-. . . . . . B 3.8 64 8 3.8.9 Distribution Systems-0)erating . . . . . . . . . . B 3.8 67 0 3.8.10 Distribution Systems-S1utdown . . . . . . . . . . . B 3.8 77 B 3.9 REFUELING OPERATIONS . . . . . . . . . . . . . . . . . . B 3.9 1 B 3.9.1 Boron Concentration . . . . . . . . . . . . . . . . B 3.9 1 B 3.9.2 Nuclear Instrumentation . . . . . . . . . . . . . . B 3.9-5 B 3.9.3 Containment Penetrations . . . . . . . . . . . . . . B 3.9 9 B 3.9.4 Decay Heat Removal (DHR) and Coolant Circulation-High Water Level . . . . . . . . . . B 3.914 8 3.9.5 . Decay Heat Removal (DHR) and Coolant Circulation-Low Water Level . . . . . . . . . . B 3.9-18
- B 3.9.6 Refueling Canal Water level . . . . . . . . . . . . B 3.9-23 t
i Crystal River Unit-3 vli Amendment No.
EFIC Manual Initiation f 3.3.12 ; i 3.3 INSTRUMENTATION- l 3.3.12 Emergency feedwater Initiation and Control (EFIC) Manual Initiation l LCO 3.3.12 Two manual initiation switches per actuation channel for i each of the following EFIC Functions shall be OPFRABLE: l
- a. Steamgenerator(OTS3)AMainfeedwater(MFW) Isolation; l OTSG B MFW ! solation;
- b. )
- c. OTSG A Main Steam 1.ine Isolation; j
- d. OTSG B Main Steam Line Isolation; and j
, e. Emergencyfeedwater(EFW) Actuation. !
... ...............................-NOTE - ----------- - --~~~ - -- - - -
i On1 the manual initiation switches associated with EFW actuation are required : to e OPERABLE in MODE 4. j I APPLICABillTY: MODES 1, 2, 3 and 4. ACTIONS-
..................................... NOTE--- ---- - -- - ---------------------
Separate Condition entry is allowed for each function. ! CONDITION REQUIRED ACTION COMPLETION TIME i t A. One or more EFIC A.1 Place tr;p module for 72 hours i Function with one the associated EFIC me.nualitin(s)iation Function (s) in trip. 1 switch inoperable in # . one actuation channel. B. One or more EFIC B.1 Restore one manual 72 hours function (s) with both initiation switch to manual initiation . ~ OPERABLE status, switches inoperable in one actuation channel.._ (continued) Crystal River Unit 3 3.3 30 Amendment No. ym._,._.m._ ,y .,_y._ ,, . . . .,,,, . ,,._,,...,y_,,,._. , , _ . , , , , , , , , _ , . _ , _ . , , ,
EFIC Manual Initiation 3.3.12 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME i C. One of more EFIC C.1 Place trip modules I hour functions with one for the associated manual initiation EFIC function (s) in ' switch inoperable in trip. both actuation channels. D. One or more EFIC D.1 Restore one actuation I hour function channel for the manual itiation in(s) with both associated LFIC switches inoperable in Function s) to both actuation OPERABLE (status, channels. E. Required Action and E.1 Be in MODE 3, 6 hours associated Completion Time not met for 6MQ manual initiation switches Ini E.2 Be in MODE 4. 12 hours associatec with EFW actuation. F. Required Action and F.1 Be in MODE 3. 6 hours associated Completion Time not met for AND manual initiation switches associated F.2 Be in MODE 5. 30 hours with EFW actuation. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.12.1 Perform CHANNEL FUNCTIONAL TEST. 31 days Crystal Rist- itt 3 3.3-31 Amendment No.
PAM Instrumentation 3.3.17 3.3 INSTRUMENTATION 3.3.17 Post Accident Monitoring (PAM) Instrumentation LCO 3.3.17 The PAM instrumentation for each function in Table 3.3.17-1 shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3 and 4. I ACTIONS
..................................... NOTES --- --------~~--------------------
- 1. LCO 3.0.4 is not applicable.
- 2. Separate Condition entry is allowed for each function.
CONDITION REQUIRED ACTION COMPLETION TIME A. One or more Functions A.1 Restore required 30 days with one required channel to OPERABLE channel inoperable. status. B. Required Action and B.1 initiate action in Immediately associsted Completion accordance with Time of Condition A Specification not met. 5.7.2.a. C. One or more functions C.1 Restore one channel 7 days with two retuired to OPERABLE status, channels inoperable. (continued) Crystal River Unit 3 3.3-38 Amendment No.
PAM instrumentation 3.3.17 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME
- 0. Required Action and 0.1 Enter the Condition immediately associated Completion referenced in Time of Condition C Table 3.3.17-1 for not met, the function.
E. As required by E.1 Be in MODE 3. 6 hours Recuired Action 0.1 and referenced in AHQ Table 3.3.17-1. E.2 Be in MODE 5. 30 hours I
- f. As required by F.1 initiate action in immediately Recuired Action D.1 accordance with anc referenced iii Specification Table 3.3.17-1. 5.7.2.a.
l l l ! Crystal River Unit 3 3.3-39 Amendment No. l l
= - - - _ _ _ _ ,
RCS Loops-MODE 4 3.4.5 3.4 REACTOR COOLANT SYSTIM (RCS) 3.4.5 RCS Loops-MODE 4 LCO 3.4.5 Two loops consisting of either two RCS loops or one RCS loop and one decay heat removal (OHR) 1 cop shall be OPERABLE and at least one loop shall be in operation. l
............................N01E- ------ -- -- ----- - -- -
RCPs) may be de-energized for , All s 8 reactor hours percoolant 24 hourpumps perio (d for the transition to or from the DHR System, and all RCPs and DHR pumps may be de-energized for s I hour per 8 hour period for any other ' reason, provided:
- a. No o)erations-are permitted that would cause reduction of tie RCS baron concentration; and
- b. Core outlet temperature is maintained so as to assure subcooling throughout the RCS.
?
APPLICABILITY: MODE 4. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME f A. One RCS loop A.1 Initiate action to Immediately inoperable, restore-a second loop to OPERABl.E status. AND Required DHR loop i inoperable. (continued) Crystal-River Unit 3' 3.4-8 Amendment No.
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RCS Loops-MODE 4 3.4.5 ACTIONS (continued) ,, CONDITION REQUIRED ACTION COMPLETION TIME i B. Two RCS loops B.1 Initiate acilon to immediately l >' inoperable. restore an RCS loop to OPERABLE status, i 03 , B.2 Be in MODE 5. 24 hours C. All RCS-and DHR loops C.1 Suspend all Imediatzly inoperable, operations involving a reduction in RCS 0E boron concentration. No RCS or DHR loop'in MQ operation. C.2 Initiate action to Immediately restore one loop to OPERABLE status and operation. SUR'JEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.5.1 Verify one DHR or RCS loop is in operation. 12 hours SR 3.4.5.2 Verify correct breaker alignment and 7 days indicated power available to the required pump that is not in operation. Crystal- River Unit 3 3.4-9 Amendment No.
1 ECCS-Shutdown 3.5.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY
$R 3.5.3,1 for all equipment required to be OPERABLE In accordance I the following SRs are a3plicable, with applicable SRs SR 3.5.2.1 SR 3.5.2.6 SR 3.5.2.2 SR 3.5.2.7 SR 3.5.2.5 Crystal River Unit 3 3.5-8 Amendment No.
TBVs 3.7.4 3.7 PLANT SYSTEMS 3.7.4 Turbine Bypass Valves (TBVs) LCO 3.7.4 fach TBV shall be OPERABLE.
...........................N0TE-..-------
In MODE 4, only the TBVs associated with the OPERABLE RCS loop (s) are required OPERABLE. APPLICABILITY: H0 DES 1, 2, 3, and 4. I ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more TBVs A.1 RestoreTBV(s)to 7 days inopersble. OPERiviLE status. QS A.2 Verify by 7 days administrative means OPERABILITY of associated steam generator atmospheric dump valve (ADV). B. Required Action and B.1 Be in MODE 3. 6 hours associated Completion Time not met. AND B.2 Be in MODE 5. 30 hours I SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.4.1 Perform one complete cycle of each TBV. 24 months .c Crystal River Unit 3 3.7-8 Amendment No.
EfW Syste -.0perating l 3.7.5 3.7 PLANT SYSTEMS 3.7.5 Emergency feedwater (EfW) System . Operating l , LCO 3.7.5 Two EfW trains r. hall be OPERABLE.
............................G0TE-----
Only one EfW train which includes a motor driven pump, is requiredtobeOPEb8LEinMODE3withsteamgenerator pressure < 200 psig. APPLICABILITY: MODES 1, 2, and 3. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One steam supply to A.1 Restore steam supply to 7 days the turbine driven OPERABLE status. EfW pump inoperable. 6HD 10 days from discovery of failure to meet the LCO B. One EfW train B.1 Restore EfW train to 72 hours inoperable for OPERABLE status. reasons other than aNQ Condition A. 10 days from discovery of failure to meet the LC0 (continued) Crystal River Unit ? 3.7-9 Amendeent No.
EfW Systea-Operating l l 3.7.5 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME j C. Required Action and C.1 Be in MODE 3.- 6 hours associated Completion ! Time of Condition A MD or B not met. C.2 Be in MODE 4. 12 hours E D. Two EfW trains D.1 Initiate action to- Immediately l inoperable. restore one EfW train to OPERABLE status. h r t I i Crystal-River Unit 3 3.7-10 Amendment No.. t
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i
'I EFW SysteQ-Operating l i 3.7.5 i SURVEILLAliCE REQUIREMENT 5 ,
f SURVEILLANCE FREQUENCY SR 3.7.5.1 Verify each EFW manual, power operated, ar.1 45 days automatic valve in each water flow path and in both steam supply flow paths to the turbino driven pump, that is not locked, sealed, or otherwise secured in position, is in the correct position. - SR 3.7.5.2 -- ---- ----- ---- NOTE -- ------------ --- Not required to be performed for the until 24 hours ' turbine driven EFW after reaching 200 pump,in psig the steam - generators. i Verify the developed head of each EFW pump 45 days on a at the flow test point is greater than.or. STAGGERED TEST equal to the required developed head. BASIS r SR 3.7.5.3 -------------------NOTC------------ ------- Not required to be performed until 24 hours after reaching 200 psig in the steam generators. Verify each EFW automatic valve that is not 24 months locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal. (continued) i Crystal River Unit 3 3.7-11 Amendment No. s
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EFW Systeo--Operating l 3.7.5 SURVEILLANCE-REQUIP.EMENTS (continued) I l SURVEILLANCE FREQUENCY I i SR 3.7.5.4 ------------- -----NOTE.--------- --------- j Not required to be performed until 24 hours after reaching 200 psig in the steam i generators. Verify each EFW pump starts automatically 24 months on an actual or simulated actuation signal. - SR 3./.5.5 Vertfy proper alignment of the EFW flow Prior to paths by verifying flow from the EFW tank entering MODE 2 to each steam generator, whenever plant has been in MODE 5 or 6 for
> 30 days I
i Crystal' River Unit-3 3.7-12 Amendment No. - _ _ _ _ _ . . _ . z_ _ __ _ __.___ . _ _ ._;_.__.._._.__._ __ _ ,_. . _ , , _ _ ,
EFW Tank 3.7.6 3.7 PLANT SYSTEMS 3.7.6 EmergencyFeodwater(EFW) Tank LCO 3.7.6 EFW tank water volume shall be 2150,000 gal. APPLICABILITY: MODES 1, 2, 3, and 4. I ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. EFW tank water volume A.1 Verify by 4 hours not within limit, administrative means OPERABILITY of backup AND water supply. Once per 12 hours thereafter 88D A.2 Restore EFW tank 7 days water volume to within limit. B. Required Action and B.1 Be in MODE 3. 6 hourt associated Completion Time not met. Ah9 B.2 Be in MODE 5. 30 hours I Crystal River Unit 3 3.7-13 Amendment No.
EFW Systea--Shutdown 3.7.18 8
'3.7 PLANT SYSTEMS 3.7.18 Emergency feedwater (EFW) System - Shutdown LCO 3.7.18 Notor driven EFW train shall be OPERABLE.
APPLICABILITY: MODE 4. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Motor driven EFW train A.1 Initiate action to Immediately inoperable restore motor driven EFW train to OPERABLE status. B. Required Action and 8.1 Be in MODE 5. 24 hours associated Completion Time not met.
-SURVEILLANCE REQUIREHENTS SURVEILLANCE FREQUENCY SR 3.7.18.1 for all equipment required to be OPERABLE, In accordance the following SRs are applicable, with applicable SRs SR 3.7.5.1 SR 3.7.5.3 SR 3.7.5.2 SR 3.7.5.4 Crystal River Unit 3 3.7-37 Amendment No.
EFIC Instrumentatien B 3.3.11 BASES (continued) l I APPLICABILITY EFIC instrumentation OPERABILITY requirements are applicable during the MODES and specified conditions listed in l Table 3.3.11-1. Each function has its own requirements based on the specific accidents and conditions for which it . is designed to provide protection. 4 The initiation of EFW on the loss of MFW Pumps is ap)11 cable in MODE I and in MODES 2 and 3 wh(r not in shutdown )ypass. Below these plant conditions, EFW initiation on low OTSG 1evel occurs fast enough to prevent primary system overheating. EFW Initiation on low OTSG 1evel shall be OPERABLE at all times the OTSG is required for heat removal. These conditions include MODES 1, 2, and 1 To avoid automatic ' actuation of the EFW pumps during L. tup and cooldown, the low OTSG pressure function can be bypassed at or below a secondary pressure of 750 psig This secondary-side pressure occurs during MODE 3 operation. EFW initiation on loss of all RCPs is required to be OPERABLE at 2 10% RTP. This power level coincides with the bypass permissive signal provided by RPS. The MFW, Main Steam Line Isolation, and EFW Vector Valve Control Functions shall be OPERABLE in MODES 1, 2, and 3 with OTSG pressure 2 750 psig because OTSG inventory can be high enough to contribute t 4nificantly to the peak pressure following a secondary side treak. Both the normal feedwater and the EFW must be isolatable on each OTSG to limit overcooling of the primary and mass and energy releases to the RB, Once OTSG pressures decrease below 750 asig, the Main Steam Line and MFW Isolation functions can se bypassed to-prevent actuation durint; cooldown. The EFW Vector Valve Control logic will not perform any function when both OTSG pressores are low; thus, the logic is also bypassed at the same time the OTSG pressure low functions is bypassed. In MODES 4, 5, and 6, priinary and secondary side energy levels are reduced and the feedwater flow rate is low or nonexistent. Because of this, the automatic actuation features of EFIC instrumantation are not required to be
-OPERABLE in these MODES.
The requirements for the manual actuation features of EFIC l in MODE 4 are discussed in Specification 5.3.12. (continued) Crystal ' River Unit 3- B 3.3-94 ' Amendment No,
- i. . .. . . . __ -
EFIC Manual Initiation B 3.3.12 B 3.3 INSTRUMENTATION B 3.3.12 Doergency Feedwater Initiation and Control (EFIC) Hanual Initiation BASES
= ,
BACKGROUND The EFIC mu,ual initiation capability provides the operator with the capability to actuate certain EFIC Functions in the absence of an automatic initiation condition. Functions with the capability to be manually actuated include Main Feedwater (MFW) Isolation for Once Through Steam Generator (OTSG)- A, MFW Isolation for OTSG B, Main Steam Line Isolation for OTSG A, Main Steam Line Isolation for OTSG B, and Emergency feedwater (EFW) Actuation. The EFIC manual initiation circuitry satisfies the manual initiation and single-failure criterion requirements of IEEE-279-1971(Ref.1). Although not part of this LCO, the EFIC functions listed above can also be remotely manually initiated from the EFIC cabinets. APPLICABLE EFIC Functions credited in the safety analysis are SAFETY ANALYSES automatic. However, EFIC manual initiation Functions are required by design as backups to the automatic trip Functions. This allows the operator to actuate EFW, Main Steam Line Isolation. or MFW Isolation whenever conditions dictate and one has not already automatically occurred. As such, they are backup Functions to those performed automatically by EFIC. In MODE 4, the primary and secondary side energy levels are reduced and feedwater flow rate is low. Sufficient time is available to the operator to manually initiate EFIC in an evert that requires a safety grade source of feedwater. LCO Two manual initiation switches per actuation channel (A and B) of each function (OTSG A and B MFW Isolation, OTSG A and B Main Steam Line Isolation, and EFW Actuation) are required to be OPGABLE whenever the OTSGs are relied on to remove taat from the primary. Each Function (MFW Isolation, Main Steam Line Isolation, and EFW Initiation l nas two (continued) Crystal River Unit 3 8 3.3-100 Amendment No.
EFIC Hanual Initiation B 3.3.12 d BASES LCO actuation or " trip" channels, channels A and B. Within each (continued) channel A actuation logic there are two manual trip switches. When one manual switch is depressed, a half trip occurs. When both manual switches are depressed, a full trip of channel A actuation occurs for that particular ' Function. Similarly, channel B actuation logic for each Functkn has two nanual trip switches. Both switches per , actuation channel must be OPERABLE and must be depressed to get a full manual trip of that channel. The use of two-manual trip switches for each channel of actuation logic @ allows for testing without actuating the end devices and also reduces the possibility of accidental manual actuation. APPLICABILITY The MFW and Main Steam Line Isolation manual initiation Functions shall be OPERABLE in MODES 1. 2, and 3 because OTSG inventory can be at a sufficiently high energy level to contribute significantly to the peak containment pressure during a secondary side break, in MODES 4, 5, and 6, the pn nary and secondary side energy levels are reduced and feedwater flow rete is low or nonexistent, and the Function is not required to be OPERABLE. The EFW manual initiation Function shall be OPERABLE in MODES 1, 2, 3, and 4 because the OTSGs are relied on as a l heat sink for the Reactor Coolant System and the core itself. In MODES 5 and 6, heat removal requirements are I reduced tnd can be provided by the Decay Heat Removal System. ACTIONS A Note has been added to the ACil0NS indicating that separate Condition entr3 is allowed for each EFIC manual initiation Function. Ad With one manual initiation switch of one or mere EFIC Function (s) incperable in one actuation channel, the trip module for the associated EFIC Function (s) must be placed in the tripped condition within 72 hours. With the channel in the tripped condition, the single-failure criterion is met and the operator can still initiate one actuation channel given a single failure in the other channel. Failure to (continued) Crystal River Unit 3 B 3.3-101 Amendment No.
EFIC Manual Initiation B 3.3.12 BASES ACTIONS 621 (continued) perform Required Action A.1 could allow a single failure of a switch in the other manual initiation channel to prevent manual actuation of the Function from the MCB. The Completion Time allotted to trip the trip module allows the operacor to take all the appro)riate actions for the failed switch and still ensure that tie risk involved in operating with the failed stitch is acceptable. 2 H21 With both manual initiation switches of one or more EFIC Function (s) inoperable in one actuation channel, one manual initiation switch must be restored to OPERABLE status within 72 hours. In this Condition, the operator has the capability to manually initiate the affected EFIC Function from the MCB utilizing the manual initiation switches in the i other actuation channel. Mowever, the systems single failure provisions are no longer provided and must be restored within 72 hours. ~he associated Completion Time of 72 hours is acceptable ba:td upon engineering judgment and is consistent with similar EFW-related Required Actions addressing a loss of redundancy. From a functional perspective, this Condition is equivalent to Condition A since each actuation channel is a two-out-of-two logic (i.e., the actuation channel will not function with one or both switches inoperable). The difference between the two Conc '. ions lies in the specified Required Actions. The trip n~Jules associated with both inoperable manual initiation switches in a given actuation channel cannot be simultaneously placed in trip without receiving an EFIC actuation. Lil With one manual initiation switch of one or more EFIC Function (s) inoperable in both actuation channels, the EFIC trip modules associated with the inoperable manual initiation switch must be placed in trip within I hour. In this Condition, the operator does not have the capability to manually initiate the affected EFIC Function from the MCB. The Function can still be manually actuated from the EFIC Cabinets (124 ft elevation Control Complex). The associated Completion Time of I hour reflects the immediacy of (cor.tinued) Crystal River Unit 3 B 3.3-102 Amendment No.
l EFIC Manual Initiaticn B 3.3.12 BASES ACTIONS Gd -(continued) restoring manual initiation capability at the HCB and is reasonable based on operating experience. With one actuation channel for the associated EFIC Function restored to OPERABLE status, operation may continue in accordance with Condition A or Condition B. From a functional perspective, this Condition is equivalent to Condition D since each actuation channel is a two-out-of-two logic (i.e., neither actuation channel will not function with one switch inoperable). The difference between the two Conditions lies in the specified Required Actions. The trip modules associated with both inoperable manual initiation switches in a given actuation channel cannot be simultaneously placed in trip without receiving an EFIC actuation. DJ With one or both manual initiation switches of one or more EFIC Function (s) inoperable in both actuation channels, one actuation channel for each function must be restored to OPERABLE status within 1 hour. With the channel restored, operation may continue in accordance with Condition A or B. The Completion Time allotted to restore the channel allows the operator to take all the appropriate actions for the failed channel and still ensures that the risk involved in operating with the failed channel is acceptable. [ 1 and E.2 If the Required Actions cannot be met within the associated Completion Times, the plant must be placed in a MODE in whicF the LC0 does not apply. To achieve this status, the plant must be placed in at least MODE 3 within 6 hours and in MODE 4 within 12 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required MODES from full power conditions in an orderly manner and without challenging plant systems. (continued) Crystal River Unit 3 B 3.3-103 Amendment No.
EFIC Manual Initiatien B 3.3.12 BASES ACTIONS'- F.1 cnd i 2 j (continued) If the Required Actions cannot be met within the associated Completion Times for manual initiation switches for EFW actuation, the plant must-be placed in a MODE in which ti;e LCO does not apply. To achieve this status, the plant must be placed in at least MODE 3 within 6 hours.and in MODE 5 within 30 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required MODES from full power conditions in an orderly manter and without challenging plant systems. SURVEILLANCE SR '3.3.12.1
- REQUIREMENTS This SR requires the 3erformance of a CHANNEL FUNCTIONAL TEST to ensure that tie channel; can perform their intended functions. This SR, like all other CHANNEL FUNCTIONAL TESTS, does not include actuation of the end device. This is due to the risk of a plant transient caused by the closure of valves associated with MFW and Main Steam Line Isolation or actuating EFW vith the reactor at power. The Frequency of 31 days is based on cperating experience and industry-accepted practice.
REFERENCES 1. IEEE-279-1971, April 1972.
- Crystal River Unit 3 B 3.3-104 . Amendment No.
1 EFIC-EFW--Vector Valve Logic B 3.3.14 BASES ACTIONS 621 (continued) With one channel inoperable, the system cannot meet the single-failure criterion and still satisfy the dual functional criteria described above. Therefore, when one vector valve logic channel is inoperable, the channel must be restored to OPERABLE status within 72 hours. This Condition is analogous to having one EFW train inoyerable; wherein a 72 hour Completion Time is provided by tie ' Required Actions of LC0 3.7.5, "EFW System--Operating." As l such, the Completion Time of 72 hours is based on engineering judgment. B.1 and B.2 If Required Action A.1 cannot be met within the associated Completion Time, the plant must be placed in a MODE in which the LCO does not apply. To achieve this status, the plant must be placed in at least MODE 3 within 6 hours and in H0DE 4 within 12 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCC SR 3t L11d REQUIREMENTS SR 3.3.14.1 is the performance of a CHANNEL FUNCTIONAL TEST every 31 days. This test demonstrates that the EFIC-EFW--vector valve logic is capable of performing its intended function. The Frequency is based on operating experience that demonstrates failure of more than one channel within the same 31 day interval is unlikely. REFERENCES None. I Crystal River Unit 3 B 3.3 113 Amendment No. 1
l l PAM Instrumentation B 3.3.17 BASES (continued) APPLICABILITY The PAM instrumentation requirements are applicable in MODES 1, 2, 3, and 4. These variables are related to the l diagnosis and pre-)lanned actions required to mitigate DBAs. The applica)le DBAs are assumed to occur in MODES 1, 2, 3 and 4. In MODES 5 and 6, plant operating conditions l are such that the likelihood of an event occurring that would require PAM instrumentation is low; therefore, PAM instrumentation is not required to be OPERABLE in these MODES. ACTIONS The ACTIONS are modified by two Notes. Note I was added to indicate the restrictions of LCO 3.0.4 are not applicable. This exception allows entry into an applicable MODE while relying on the ACTIONS even though the ACTIONS would eventually require a !,hutdown. This exception is acceptable due to the passive function of the instruments, the operator's ability to respond to an accident utilizing alternate instruments and methods, and the low probability of an event requiring these instruments. Note Two was added to clarify the application of Completion Time rules to this Specification. The Conditions of this Specification are entered independently for each Function listed in Table 3.3.17-1. The Completion Time (s) of the inoperable channels of a Function will be tracked separately for each Function starting from the time the Condition was entered for that function. Ad When one or more Functions have one required channel inoperable, the inoperable channel must be restored to OPERABLE status within 30 days. The 30 day Completion Time is based on engineering judgment and a variety of considerations. These considerations include availability of the remaining OPERABLE channel, the passive nature of the instrument (no critical automatic action is assumed to occur from these instruments), and the low probability of an event requiring PAM instrumentation during this interval. (continued) Crystal River Unit 3 B 3.3-139 Amendment No. l
PAM Instrumentation B 3.3.17 BASES ACTIONS D.d
-(continued)
Required Action 0.1 directs entry _into the. appropriate Condition referenced in Table 3.3.17-1. The applicable-Condition referenced in the Table is function dependent.
- Each time an inoperable channel has not met any Required Action and associated Completion Time of Condition C, Condition 0 is entered for that Function-and the operator is directed to the appropriate subsequent Condition, f.d If the Required Action and associated Completion Time of Conditions C is not met and Table 3.3.17-1 directs entry into Condition E, the plant must be placed in a MODE in which the requirements of this LC0 do not apply. To-achieve this status, the plant must be placed in at least MODE 3 within 6 hours and in MODE 5 within 30 hours. The I allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems, f.d Alternative means of monitoring containment area radiation and reactor vessel level are available and may be relied upon if the normal PAM channels cannot be restored to OPERABLE status within the associated Completion Time.
Based upon this capability, it is inappropriate to require plant shutdown in this condition. Rather, in conjunction with the alternate monitoring means, the Required Action speciftet actisa be immediately initiated in accordance with Specification 5.7.2.a. "Special Reports," in the Administrative Controls section of the Technical Specifications. The report provided to the NRC should discuss the alternate means of monitoring, .1escribe the degree to which the alternative means are equivalent to the installed'PAM channels, justify the areas in which they.are not equivalent, and provide a schedule for restoring the normal PAM channels to OPERABLE status. (continued) Crystal River Unit 3 B 3.3-141 Amendment No.
l RCS Loops-MODE 4 B 3.4.5 B 3.4 REACTOR COOLANT OYSTEM (RCS) B 3.4.5 RCS Loops-MODE- 4 BASES - BACKGROUND In MODE 4, the primary function of the reactor coolant is the removal of decay heat and transfer of this heat to the steam generators (OTSGs) or= decay heat removal (DHR) heat exchangers. The secondary function of the reactor coolant is to act as a transport medium for soluble neutron poison, boric acid. In MODE 4, either reactor coolant pumps (RCPs) or DHR pumps can be used for coolant circulation. The number of ) umps in operation can vary to suit the operational needs. T ie intent of this LCO is to provide forced flow frem at least one RCP or one DHR pump for decay heat removal and transport. The flow provided by one RCP or one DH9 pump is adequate for heat removal. The other intent of this LCO is to recuire that two paths (loops) be available to provide redunc ancy for heat removal. APPLICABLE No safety analyses are performed with initial conditions in SAFETY ANALYSE! MODE 4. The flow provided by one reactor coolant or one decay heat removal ) ump is adequate to prevent boron stratification in t1e vessel core region during a reduction of boron concentration. RCS loops-MODE 4 satisfies the requirements of NRC Policy Statement. While none of the three criteria directly apply, this Specification assures that reactivity control is maintained, thus Criterion 2 is the appropriate criterion, because boron dilution and reactivity control in natural circulation are unanalyzed. Potential reactivity increases would be outside the counds of the safety analysis. RCS loops-MODE 4 was identified in the NRC Policy Statement as an important contributor to risk reduction. LCO The purpose of this LCO is to recuire that two loops, either two RCS loops or one RCS loop anc one DHR loop, be OPERABLE in MODE 4 and one of these loops be in operation. Only one DHR loop is available in MODE 4 since the other DHR train is required to sarve as a dedicated LPI/ECCS function in accordance with Specification 3.5.3, ECCS-Shutdown. Any one (continued) Crystal River Unit 3 B 3.4-22 Amendmert No.
'CS Loops-MODE 4 B 3.4.5 BASES LCO loop in operation provides enough flow to remove the decay (continued) haat from the core with forced circulation. The second loop that is required to be OPERABLE provides a redundant path for heat removal.
An OPERABLE RCS loop consists of at least one OPERABLE RCP and a flow path for circulating reactor coolant around the loop. RCPs are OPERABLE if they are capable of being powered and are able to provide flow if required. Similarly for the DHR System, an OPERABLE DHR loop is comprised of the OPERABLE DHR pump (s) capable of providing forced flow to the DHR heat exchanger (s). DHR pumps are OPERABLE if thsy are capable of being powered and em able to provide flow if required. The Note permits a limited period of operation during which all RCPs may be de-energized for s 8 hours per 24 hour period for the transition to or from the DHR System. This allows the RCPs to be secured prior to reducing RCS pressure below that needed to place DHR in service. In this pressure range, accelerated RCP seal degradation can potentially occur due to inadequate NPSH. The Note also permits all DHR and RC pumps to be stopped for s I hour per 8 hour period for any reason. During this period, natural circulation will provide core decay heat removal. The Note prohibits the reduction of RCS boron concentration when forced flow is stopped because an even concentration distribution cannot be ensured. Core outlet temperature is to be maintained so as to assure subcooling throughout the RCS so that no vapor bubble may form and possibly caus3 a natural circulation flow obstruction. APPLICABILITY In MODE 4, the heat generated is lower than at power; therefore, one loop in operation is adequate for transport and heat removal. A second loop is required to be OPERABLE in order to provide redundant heat removal capability, but does not have to be in operation. This LC0 allows use of either DHR or RCS loops because it is possible to remove core decay heat and to provide proper boron mixing with either system. (continued) Crystal River Unit 3 B 3.4-23 Amendment No. i I
' i : RCS Loops-MODE 4 h B 3.4.5 i BASES APPLICABillTY forced circulation is required in all MODES and is addressed (continued) by the following Specifications: LCO 3.4.4, "RCS Loops-MODE 3"; LC0 3.4.6, "RCS Loops-MODE 5, Loops Filled"; LCO 3.4.7, "RCS Loops-MODE 5, Loops Not Filled"; LC0 3.9.4, " Decay Heat Removal (DHR) and Coolant Circulation-High Water Level" (MODE 6); and LCO 3.9.5, " Decay Heat Removal (DHR) and Coolant Circulation-Low Water Level" (MODE 6). Forced circulation is implicitly required in MODES 1 and 2 in order to prevent a Reactor Protection System actuation (Ref. LCO 3.3.1). ACTIONS 16 1 If one RCS loop and the required DHR loops are inoperable, l redundant forced flow heat removal capability is lost. Action must be ir.imediately initiated to restore a second loo), either an RCS loop or the required DHR loop, to OPE 1ABLE status. The immediate Completion Time reflects the importance of maintaining the availability of two paths for heat removal. In this Condition the OPERABLE RCS loop must be in operation, except as aliowed by the Note in the LCO, otherwise Condition C is also applicable. B.1 and B.2 If both RCS loops are inoperable, an RCS loop must be I restored to OPERABLE status to satisfy single failure considerations. Action to reatore the loop must be initiated immediately reflecting the urgency of restoring redundant heat removal capability. The required DHR loop is still l available for cooldown given the reduced heat loads of this operating MODE. In this Condition, the required OPERABLE DHR loo) must be in operation, except as allowed by the Note in the .C0, otherwise Condition C is also applicable, if restoration cannot be reasonably accomplished, consideration should be given to placing the plant in MODE 5. (continued) Crystal River Unit 3 8 3.4-24 Amendment No.
RCS Loops-MODE 4 B 3.4.5 BASES ACTIONS B.1 and B.2 (continued) % With only the required DHR loop OPERABLE, redundancy for l decay heat removal is lost and, in the event of a loss of the required DHR loop, there is more temperature margin to I boiling in MODE 5 (s 200*F) than MODE 4 (200'F to 280'F). The Completion Time of 24 hours is reasonable, based on operating experience, to reach MODE 5 in an orderly manner and without challenging plant systems. C.1 and C.2 This Condition is not entered when using the allewance in the Note to the LCO to de-energize all RC and DHR pumps. If no RCS or DHR loops are OPERABLE or in operation, all operations involving a reduction of RCS boron concentration must t,a suspended and action to restore one RCS or DHR loop to OPERABLE status and operation must be initiated. Boron dilution requires forced circulation for proper mixing, and the margin to criticality must not be reduced without forced flow. The immediatt Completbn Times reflect the importance of maintaining decay beat rc.noval capability and forced circulation. The action to restore a loop to OPERABLE status must continue until one loop is restored to operation. SURVEILLANCE SR 3.4.5.1 REQUIREMENTS This Surveillance requires verification every 12 hours that one DHR or RCS loop, with a minimum of one pump, is in operation to ensure forced flow is providing decay heat removal. Verification includes flow rate, temperature, or pump status monitoring. The 12 hour Frequency has been shown by operating practice to be sufficient to regularly assess operating loon Sti. us. In addition, control room indication and ala ms indicate loop status and will typically alert operations personnel to anomalous flow / loss of flow, should tais occur. (continued) Crystal River Unit 3 8 3.4-25 Amendment No. l
1 L ECCS-Shutdown B 3.5.3 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) B 3.5.3' ECCS-Shutdown BASES BACKGROUND The Background section for Bases B 3.5.2 is applicable to
-these Bases, with the following modification The ECCS flow paths consist of piping, valves, heat exchangers, and pumps, such that water from the borati:4 water storage tank (BWST) can be injected into the Reactor Coolant System (RCS) following the accidents described in Bases 3.5.2.
APPLICABLE The Applicable Safety Analyses section of Bases 3.5.2 is SAFETY ANALYSES applicable to these Bases. Duc to the stable conditions associated with operation in MODE 4 and the reduced probability of occurrence of an accident requiring ECCS injection, the ECCS operational requirements are reduced. Included in these reductions is that certain automatic Engineered Safeguards Actuation System (ESAS) actuation is not available. In this MODE sufficient time exists for manual actuation of the required ECCS to mitigate the consequences of an accident requiring ECCS injection. An LPI subsystem dedicated to taking suction from the BWST, and not used for decay heat removal, minimizes the time for operator actions to initiate LPI and reduces operator burden ,
~
since re-alignment of the DHR system to the ECCS mode of operation would not be necessary. Additionally, LPI capability is protected since a break of the reactor coolant outlet line could possibly damage the DHR pump without mitigating operator action. The time available for operator actions are improved based on the ability to detect RCS leakage as a LOCA precursor. Only one ECCS train is required for MODE 4. This requirement dictates that single failures are not considered during this MODE. (continued) Crystal River Unit 3 B 3.5-20 Amendment No.
ECCS-Shutdown B 3.5.3 BASES LC0 In MODE 4, one ECCS train is required to ensure sufficient ECCS flow is available to the core following an accident requiring ECCS injection. In MODE 4, an ECCS train consists of an HPI subsystem and an LPI subsystem. Each train includes the piping, instruments, and controls to ensure an OPERABLE flow path capable of taking suction from the BWST upon an ESAS signal and manually transferring suction to the reactor building emergency sump. During an event requiring ECCS actuation, a flow path is required to provide an abundant supply of water from the BWST to the RCS, via the ECCS pumps and their respective discharge flow paths, to each of the four cold leg injection nozzles and the reactor vessel. In the long term, this flow path may be switched to take its supply from the reactor building emergency sump and to supply its flow to the RCS hot and cold legs. This LCO is modified by a Note which states that HPl may be deactivated in accordance with Low Temperature Overpressure Protection (LTOP) administrative controls. Operator action is then required to initiate HPl. In the event of an accident requiring HPI actuation, the time required for operator action has been shown by analysis to be acceptable. APPLICABILITY In MODES 1, 2, and 3, the OPERABILITY requirements for the ECCS are covered by LC0 3.5.2, "ECCS-Operating." In MODE 4 with the RCS temperature below 280*F, one OPERABLE ECCS train is acceptable without single failure consideration, on the basis of the stable reactivity condition of the reactor and the limited core cooling requirements. In MODES 5 and 6, plant conditions are such that the probability of an event requiring ECCS injection is extremely low. Core cooling requirements in MODE 5 are addressed by LC0 3.4.6, "RCS Loops-MODE 5, Loops Filled," and LCO 3.4.7, "RCS Loops-MODE 5, Loops Not Filled." MODE 6 core cocling requirements are addressed by LC0 3.9.4,
" Decay Heat Removal and Coolant Circulation-High Water Level," and LC0 3.9.5, " Decay Heat Removal and Coolant Circulation-Low Water Level."
(continued) Crystal River Unit 3 B 3.5-21 Amendment No.
ECCS-Shutdown . B 3.5.3 BASES (continued) ACTIONS AJ If no ECCS LPI subsystem is OPERABLE, due to the inoperability of the LPI pump or flow path from the BWST, the plant is not prepared to provide low pressure response to accidents requiring ECCS response. The Completion Time-to initiate the restoration of at least one ECCS LPI subsystem to OPERABLE status ensures that prompt action is taken to provide the required cooling capacity or to initiate actions to place the plant in MODE 5, where an ECCS train is not required. 161 If no ECCS HPI subsystem is OPERABLE, due to the inoperability of the HPI pump or flow path from the BWST, the plant is not prepared to provide high pressure response to accidents requiring ECCS response. The I hour Completion l Time to restore at least one ECCS HPI subsystem to OPERABLE status ensures that prompt action is taken to provide the required cooling capacity or to initiate actions to place the plant in MODE 5, where an ECCS train is not required. This condition does not apply to HPI subsystem components which are deactivated for the purposes of complying with Low Temperature Overpressure Protection (LTOP) administrative ; control commitments. With these components deactivated, the HPI subsystem is still considered OPERABLE based upon guidance in NRC Generic Letter 91-18. This guidance allows substitution of manual operator action for otherwise automatic functions for the purposes of determining OPERABILITY. The substitutions are limited and must be evaluated against the assumptions in the accident analysic,. In the case of deactivating HPI in MODE 4, the components are available for injection following manual operator action to restore the system to OPERABLE status and this action can be accomplished within the time frame required to respond to the transient / accident. 2 (continued) Crystal River Unit 3 B 3.5-22 Amendment No.
- . . .- ~ ..- - - . - - . . . . - - . - . . .... ~ . -.. ~ . . . .
-C ,);
f:
;g ,
ECCS-Shutdown .
/- B 3.5.3 9
BASES 1 ;
, ACTIONS' .
. LJ Y !
(' "" " 'qg If thel Required Actions and associated Completion Times:are
> 1
--not'. met, the plant must be placed in a MODE:in which the .
' Spacification.does-not apply. When-the Required Actions of i Condition 8 cannot be c apleted within the associated
- Completion'The(> a controlled shutdown should be initiated. '
The allowed Comphtion -Time of 24 hours is . reasonable,' based l. , M , 'on operating experience', to reach MODE 5 from MODE 4 conditions-in an orderly 4 manner and without challenging plant > systems.
.; } l I
SURVEILLANCE- =SR- 3.5.3.1 -!
'The applicable Survet1' lance descriptions from Bases 3.5.2 apply, i
REFERENCES- LThe applicable references from Bases 3.5.2. apply. t f P i e u Crystal: River Unit 3; B 3.5-23 Amendment No. L s O _ :_ , _ , . . . , . v. . _ _ _ __ . . _ _ . - - -.
TDVs B 3.7.4 BASES (continued) . LCO Each TBV (two per OTSG) is required to be OPERABLE for this LCO. Failure to meet the LCO can result in the inability to cooldown to DHR System entry conditions following a SGTR event while maintaining offsite doses to a minimum. A TBV is considered OPERABLE when it is capable of providing a controlled relief of the main steam flow, and is capable of fully opening and closing reo manually commanded to do so by the operator. APPLICABILITY In MODES 1, 2, and 3, the pressures and temperatures in the RCS are high enough to initiate a SGTR and require secondary side depressurization. Therefore, the TBVs are required to be OPERABLE in these MODES. The APPLICABILITY is modified by a note which only requires the TBVs associated with the RCS loop (s) required operable for decay heat removal. Ih MODE 4, EFW may be used to remove core decay heat in accordance with the Bases B 3.7.18. The licensing basis does not s equire the assumption of a loss of offsite power resulting from an accident in MODE 4 and, therefore, the TBVs are available for secondary side heat removal. In MODES 4, 5, and 6, a SGTR is not a credible event due to the reduced stresses in the generator tubes and low driving head for release to the environment. ACTIONS A.1 and A.2 With one or more 1BV(s) inoperable, action must be taken to restore all TBVs to OPERABLE status. The 7 day Completion Time is reasonable to repair inoperable TBVs, based on the availability of other means of depressurizing the RCS following a SGTR, and the low probability of this event occurring during the 7 day period. As an alternative to restoring tha TBV(s) to OPERABLE status, the associated 0TSG ADV must be verified to be OPERABLE within 7 days. This entails verifying that SR 3.7.4.1 is " current" for the ADV, or performing the Surveillance. Reliance on the ADV satisfy the ACTIONS of this Specification is considereo acceptable based on the early analysis. (continued) Crystal River Unit 3 B 3.7-21 Amendment No.
TBVs B 3.7.4 BASES (continued) ACTIONS B.1 and B.?, (continued). If the TBVs cannot be restored to OPERABLE status within the associated Completion Time, the plant must be placed in a MODE in which the LC0 does not apply. To achieve this status, the plant must be placed in at least MODE 3 within. 36 hours, and in MODE 5 in 30 hours. The allowed . Completion l Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
- SURVEILLANCE SR 3.7.4.1 REQUIREMENTS To perform a controlled cooldown of the RCS, the TBVs must be able to be opened remotely and throttled through-their full range. This SR ensures that the TBVs are tested through a full control cycle at least once per fuel cycle.
Performance of inservice testing or use of a TBV during a plant cooldown satisfies this requirement. Operating experience has shown that these components usually pass the Surveillance when performed at the 24 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. REFERENCES 1. FSAR, Sectic;1 10.2.1.4.
- 2. FSAR, Section 14.2.2.2.
Y (cortinued) Crystal River Unit 3 8 3.7-22 Amendment No.
EFW Syste -Operating I B 3.7.5 B 3.7 PLANT SYSTEMS B 3.7.5 Emergency Feedwater (EFW) System - Operating l BASES _ _ . _ _ _ _ , . , i BACKGROUND The Emergency Feedwater (EFW) System is designed to provide adequate flow to one or both steam generators (OTSGs) for decay heat removal with the generators at the maximum operating pressure of 1050 psig plus suitable margin for post-accident pressure increase (Ref. 1, 2). The principal function of the EFW system is to remove decay heat from the Reactor Coolant System upon the unavailability of normal feedwater supply. This is accomplished by supplying water from the emergency feedwater tank (EFT-2) to the OTSG secondary side via the high nozzles. Steam produced in the OTSGs is condensed in the main conden:,er via the turbine bypass valves or, if the atmospheric dump valves (ADVs) or main steam safety valves (HSSVs) have actuated, discharged directly to the atmosphere. The EFW System consists of one motor driven EFW pump and one steam turbine driven EFW pump, each having a nominal 100% capacity (Ref. 3). The motor driven EFW pump is powered from the 4160 volt ES bus 3A. The turbine driven EFW pump receives steam from one main steam line per OTSG via connections upstream of the associated main steam isolation valve. An alternative source of steam is available from the fossil units, Crystal River Unit I and 2 (Ref.1), but cannot be relied upon to consider the EFW train OPERABLE. The diverse motive power of the two trains enhances both system availability and reliability. The preferred water source for both EFW pump trains is the Seismic Class I, missile protected dedicated EFW tank. Backup supplies of emergency feedwater are provided by the condensate storage tank and the main condenser hotwell. The pumps tie into common discharge headers providing the capability to feed either or both of the OTSGs. DC powered block and control valves are actuated to feed the appropriate steam generator by the Emergency Feedwater Initiation and Control (EFIC) System. The capacity of either EFW pump is sufficient to remove decay heat and cool the plant until the Reactor Coolant System (RCS) pressure and temperature are low enough to place the Decay Heat Removal (DHR) System in service. (continued) Crystal River Unit 3 8 3.7-23 Amendment No.
l l EFW Systea-Operating l B 3.7.5 BASES u, BACKGROUND Automatic actuation of the EFW System occurs on the (continued) following: 4 1. Trip of both main feedwater pumps with reactor )ower greater than 20% or the NI/RPS not in shutdown aypass;
- 2. Low level in either OTSG; .
- 3. Low pressure in either OTSG;
- 4. Trip of all four _ reactor coolant pumps;
- 5. High pressure injection (HPI) actuation on both Channel A and B Engineered Safeguards Actuation System (ESAS) channels; and-
- 6. AMSAC actuation.
The EFIC is a " smart" system which will feed either or both 0TSGs with indications of low levels, but will isolate EFW to a faulted steam generator having a significantly lower steam pressure than the other. The EFW System is designed to ANSI B 31.1 ES Seismic Class I and in.accordance with General Design Criteria 2, 4, 5,19, 44, 45, and 46 (Ref. 3, 4). APPLICABLE The EFW System is sized to provide sufficient decay heat SAFETY ANALYSIS removal-capability to cooldown the RCS to the temperature and pressi;re at which the DHR System can be placed in service for any of the following events:
. lots of main feedwater (LMFW);
. LMFW with loss of offsite power;
. main feedwater line break;
. main steam line break; and small' break-loss of coolant accident (LOCA).
The EFW System is designed to remain functional following the maximum hypothetical earthquake. It will also remain (continued) Crystal River Unit 3- B 3.7-24 Amendment No.
a , EFW System Operating i B 3.7.5 BASES APPLICABLE functional following a single failure in addition to any of SAFET)lANALYSES the above events with the exception of the loss of all AC (continued) power (Ref. 3). No single failure prevents EFW from being supplied to the intact OTSG nor allows EFW to be supplied to the faulted OTeG. Note that in mostr cases of a main feedwater break or a steam line break, the depressurization of the affected OTSG would cause.the automatic initiation of EFW. However, there will be some'small break sizes for which automatic detection will not be possible. For these small breaks, the operator will have sufficient time in which to take appropriate action to terminate the event (Ref. 1). The EFW Systea satisfies Criterion 3 of the NRC Policy Statement.
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LCO Two independent emergency feedwater pumps and their associated flow paths are required to be OPERABLE. The OPERABILITY of the EFW pumps requires that each be capable e of developing its required discharge pressure and flow. Additionally, the OPERABILITY of the turbine driven pump
< requires that it be capable of being powered from an OPERABLE steam supply through ASV-5. ASV-204 was installed to improve EFW reliability and is not required for OPERABILITY. The motive power for the turbine driven pump , is rteam supplied from either OTSG from a main steam header s upstream of the main steam isolation valves so that their clesure does not isolate the steam supply to the turbine.
Bot h steam supply flow paths to the turbine driven pump are required to be OPERABLE. The OPERABILIM of the associated EFW flow paths requires all valves be in their correct posit!ons or be capable'of actuating to their correct positions on a valid actuatien signal. Inoperaatlity of the EFW System may result in inadequate decay heat removal following a transient or accident during which main feedwater is not available. The resulting RCS heaf.up and pressure increase can potentially result in significant loss of c6clant through the pressurizer code o safety valves or the PORV. The LC0 is modified by a Note indicating that only one EFW
-train, which includes a motor driven EFW pump, is required in MODE 3 with 0TSG secondary side pressure < 200 psig. At (continued)
Crystal River -Unit 3 B 3.7-25 Amendment No.
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EFW Systen-Operating i B 3.7.5 BASES LCO - less than 200 psig, the turbine driven EFW pump is not (continued) capable of producing the accident analysis flow rate. Under these conditions, the flow rate )roduced by the pump may well be adequate to perform the Teat removal function. Ilowever, this has not been analytically demonstrated. Thus, the allowance to not have this pump OPERABLE is accept,ble based upon the reduced heat removal requirement and the short duration of this condition. Similar to the concept employed in LC0 3.5.2 "ECCS Systems - Operating," cross-connecting portions of both trains to produce a single OPERABLE train, is acceptable in this plant condition. APPLICABILITY In MODES 1, 2, and 3, the EFW System is required to be OPERABLE and to function in the event that main feedwater is lost. In addition, the EFW System is required to supply enough makeup water to replace the secondary side inventory lost as the plant cools to MODE 4 conditions. The requirements for EFW in MODE 4 are addressed by Technical Specification 3.7.18. In MODES 5 and 6, the OTSG need not be used to cooldown the l RCS. Therefore, the EFW System is not required to be OPERABLE in these MODES. ACTIONS A_d With one of the two steam sup>1ies to the turbine driven EFW pump inoperable, action must )e taken to restore the steam supply to OPERABLE status within 7 days. Allowing 7 days in this Condition is reasonable, based on the redundant OPERABLE steam supply to the pump and the low probability of an event occurring that would require the inoperable steam supply to the turbine driven EFW pumps. The second Completion Time for Required Action A.1 establishes a limit on the maximum time allowed for any combination of Conditions to be entered during any continuous failure to meet this LCO. The 10 day Completion Time provides a limitation time allowed in this s)ecified Condition after discovery of failure to meet the .00. This limit is considered reasonable for situations in which Conditions A and B are entered concurrently. The 'AND' connector between 7 days and 10 days dictates that both Completion Times apply simultaneously, and the more restrictive must be met. (continued) Crystal River Unit 3 8 3.7-26 Amendment No.
EFW Syste:2-Operating i B 3.7.5 + BASES ACTIONS Bil (continued) If one of the EFW trains is inoperable, action must be taken to restore the train to OPERABLE status within 72 hours. The 72 hour Completion Time is reasonable, based on the redundant capabilities afforded by the EFW System, time needed for repairs, and the low probability of a DBA occurring during this time period. This Condition includes the lass of two steam supply lines to the turbine driven EFW pump. The second Completion Time for Required Action B.1 establishes a limit on the maximum time allowed for any combination of Conditions to be entered during any continuous failure to meet this LCO. The 10 day Completion Time provides a limitation time allowed in this specified Condition after discovery of failure to meet the LCO. This limit is considered reasonable for situations in which Conditions A snd B are entered concurrently. The '680' connector between 72 hours and 10 days dictates that both Completion Times apply simultaneously, and the more restrictive must be met. C.1 and C.2 If Required Action A.1 or Required Action b.1 cannot be completed within the associated Completion Time, the plant must be placed in a MODE in which the LC0 does not apply. To achieve '.his status, the plant must be placed in at least MODE 3 within 6 hours and in MODE 4 within 12 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. D21 With both EFW trains inoperable, the plant is in a seriously degraded condition with no safety related means for conducting a cooldown. In such a condition, plant operation should not be perturbed by a forced action, including a power change, that might result in a trip. For this reason, (continued) Crystal River Unit 3 B 3.7-27 Amendment No.
EFW Syste2-Operating i B 3.7.5 BASES ACTIONS DJ (continued) the Technical Specifications do not mandate a plant shutdown. Rather the ACTIONS allow the plant to dictate the most prudent course of action (including plant shutdown) for the situation. The seriousness of this condition requires that action be initiated immediately to restore at least one EFW train to OPERABLE status. SURVEILLANCE SR 3. 7. 5J REQUIREMENTS Verifying the correct alignment for manual, power operated, and automatic valves in the EFW water and steam supply flow paths provides assurance that the proper flow paths exist for EFW operation. The valves verified by this SR include valves in the main flow paths and the first normally closed valve in a branco line. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since those valves are verified to be in the correct position prior to locking, sealing, or securing. There are several other exceptions for valve position verification due to the low potential for these types of valves to be mispositioned. The valve types which are not verified as part of this SR include vent or drain valves outside the RB, relief valves outside the RB, and instrumentation valves (both inside and outside the RB). This SR also does not apply to valves that cannot be inadvertently misaligned, such as check valves. This Surveillance does not require any testing or valve manipulation; rather, it involves verification that those valves capable of potentially being mispositioned are in the correct position. The 45 day Frequency is based on engineering judgment and is consistent with the Frequency established for SR 3.7.5.2. SR 3.7.5.2 requires extensive EFW valve manipulation in order to perform the pump flow rate verification, such that a flow path verification is necessary following each performance. SR 3.7.5.2 This SR verifies that the EFW pumps develop sufficient discharge pressure to deliver the required flow at the full open pressure of the MSSVs. Because it is undesirable to (continued) Crystal River Unit 3 B 3.7-28 Amendment No.
EFW System-Operating i B 3.7.5 BASES
-u SVRVEILLANCE SR 3.7.5.2 (continued)
REQUIREMENTS introduce cold EFW into the OTSGs while they are operating, this test is performed on recirculation at a reduced flow rate. Periodically comparing the reference differential pressure developed at this reduced flow detects trends that might be indicative of degrading pump performance. Performance of inservics testing discussed in the ASME Code, Section XI (Ref. 5), at 3 month intervals, is satisfied by this SR. The 45 day Frequency on a STAGGERED TEST BASIS results in testing each pump once every 3 months, as required by Reference 5.
; This SR is modified by a Note indicating that the SR may be deferred until suitable test conditions are established.
This SR 3.0.4 type exception may be necessary during any plant start-up because there is insufficient steam pressure in the secondary side of the OTSGs to perform this SR on the turbine-driven pump. - SR 3.7.5.3 This SR verifies that EFW can be delivered to the appropriate OTSG in the event of any accident or transient
, that generates an EFIC signal by demonstrating that each automatic valve in the flow path, that is not locked,
' sealed, or otherwise secured in position, actuates to its correct position on an actual or simulated actuation signal.
Valves secured in the correct position need not demonstrate the capability to achieve this configuration. ADVs also need not demonstrate the capability to satisfy this SR since their operation is not credited as part of any DBA. The SR also verifies the EFW control and block valves actuate to the isolation position on a simulated or actual vector valve control signal. ' This SR is a test of the integrated system response to an actuation signal and as such, it is not necessary to verify the EFW System actuates en each EFIC signal. Any of the
. initiation signals described in the Background Section of l these Bases is adequate, given that the various EFIC (continued)
Crystal. River Unit 3 8 3.7-29 Amendment No.
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. SURVEILLANCE ~ SR 3.7.5.3 (continued)-
- REQUIREMENTS instrumentation is surveilled separately in Section 3.3
" Instrumentation" of-thelITS.
The-24 month Frequency is based-on the need to perform this , Surveillance under the conditions that apply during.an 1
= nutage and the potential for an unplanned transient if the Surveillance were-performed with the reactor.at power.. The q 24 month Frequency is also acceptable based on: operating-- :i o
experience and design reliability of the equipment. This SR is modified by a. Note indicating that the SR may be deferred until suitable test conditions'are established. This-SR 3.0.4. type exception-is necessary because.there is insufficient steam pressure to perform the test while in , MODE 4. 1 1 SR 3.7.5.4 1 This SR verifies that the EFW pumps start-in the event of . any accident or transient that generates an EFIC initiation ~l signal by demonstrating that each pump starts automatically on an actual or simulated actuation signal.- This SR is a
- test of the integrated system response to an actuation-signal and as such, it is not necessary to verify the EFW i system actuates on each EFIC signal. Any of the initiation !
signals described in the Background Section of these Bases is-adequate,-given that the various EFIC-instrumentation is surveilled~ separately in Section 3.3 " Instrumentation" of ! the ITS. The 24 month Frequency is based on the need to ; perform this Surveillance under the conditions that apply _ during a unit outage and the potential for an unplanned
. transient if the Surveillance were performed with-the reactor-at' power. This SR is modified.by a Note indicating L__
that:the SR may be deferred until-suitable test conditions are e:tablished. This SR 3.0.4 type exception is necessary-because-there is insufficient steam pressure to perform _the test while in MODE-4. SR 3.7.5.5 This SR ensures that the EFW System is properly aligned- by - verifying' the flow paths to each OTSG prior to. entering 1: MODE 2.after more than_30 days in MODE 5 or 6. OPERABILITY-n -(continued) L i .
, Crystal River Unit.3 8 3.7-30 ._ Amendment No.
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a EFW System-Operating l B 3.7.5
- BASESL
~ SR 3.7.5.5=(continued)L SURVEILLANCE-REQUIREMENTS-of EFW flow paths.must be demonstrated before sufficient core heat is: generated that would require the_ operation of the:EFW System during a subsequent shutdown. The Frequency is reasonable, based on engineering judgment, in view of other administrative controls:to ensure that the flow paths are OPERABLE. To further ensure EFW System alignment, flow path OPERABILITY is verified, following extended outages to determine no misalignment of-valves has occurred. This SR ensures-that the flow path from the EFW tank to the OTSGs is properly aligned. This requirement is based upon the recommendation of NUREG 0737. The Frequency was_ modified slightly during ITS development (prior to entering MODE 2) to provide an SR 3.0.4 type exception. As written, the-SR allows the plant to achieve and maintain MODE 3 conditions in order to perform the verification.
REFERENCES 1. Enhanced Design Basis Document for the Emergency Feedwater and Emergency Feedwater Initiation and Control System, Revision 1, dated September 27, 1991 with Temporary Changes 156, 230, 247, and 249,
- 2. BAW 10043, " Overpressure Protection for B&W Reactors",
- i- dated May 1972.
- 3. FSAR, Section 10.5.
- 4. 10 CFR 50, Appendix A.
- 5. ASME, Boiler and Pressure Vessel Code, Section XI, Inservice Inspection, Subsection IWP. ,
4 4 .5 a Crystal ' River _ Unit _3L B 3.7-31 Amendment No.
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Emergency Feedwater Tank B 3.7.6 BASES APPLICABLE was not used as an input to.these safety analyses, SAFETY ANALYSES OPERABILITY of the EFW System, and therefore the - (continued) EFW tank, is essential to the mitigation of the following events (Ref. 3): Loss of main feedwater (LMFW)-
. LMFW with a loss of offsite power
. Main feedwater line break
. Main steam line break
. Small break loss of coolant accident (LOCA)
The required minimum volume of usable condensate in the EFW tank is 150,000 gallons. This amount is sufficient to remove decay heat for a period of 18 hours at MODE 3
. This time period is considered conditions (Ref.p 4)lant conditions to be stabilized and adequate to allow another source of water to be made available for natural circulation cooldown until MODE 4 is achieved. In MODE 4, the RCS pressure will be decreased to the point that allows the alignment of the Decay Heat Removal (DHR) System to the RCS.
For events occurring in MODE 4, the EFW tank would provide the safety grade source of water to the EFW System. Although the single failure criteria is applicable to the EFW System in the evaluation of the previously mentioned i events, the EFW tank performs its safety function in a passive manner and is thereby excluded from application of the single failure criterion. The EFW tank satisfies Criterion 3 of the NRC Policy Statement. LCO In the event of a loss of offsite power, or other condition resulting in a complete loss of main feedwater, a means of removing heat from the RCS must be immediately available. For events occurri.1g in MODES 1, 2, and 3, EFW tank minimum l usable water volume limit of 150,000 gallons is necessary to provide assurance that the EFW System can su) ply the volume of secondary coolant needed to remove decay 1 eat in MODE 3 conditions for 18 hours (Ref. 4) while other sources of water are made available for subsequent cooldown to below 280 degrees, if required. (continued) Crystal River Unit 3 B 3.7-33 Amendment No. o
Emergency feedwater Tank B 3.7.6 BASES LCO for such events occurring in MODE 4, the EFW tank would (continued) provido a safety grade source of feedwater. Compliance with the LCO is verified by maintaining tank level at or above the minimum required level. APPLICABILITY In MODES 1, 2, 3, and 4 the OTSGs are the operating heat l sink for RCS heat removal. The EFW system, and thus, the EFW tank must be OPERABLE during these MODES, to assure the availability of a safety grade means of RCS heat removal following any event which results in a lo n of main faedwater, in MODES 5 and 6, the EFW TanA water volume is l not required to be within limits consistent with the requirements for EfW System OPERABILITY. s AC'il0NS A.1 and A.2 As an alternative to restoring the EFW Tank volume to within limit the OPERABILITY of the backup water supply can be verifledwithin4hoursandonceevery12hoursthereafter. The OPERABILITY of the backup feedwater sup)1y must include verification, by administrative means, of tie OPERABILITY of flow paths from the backup supply to the EFW pumps and availability of the required volume of water in the backup supply. Typically, the condensate storage tank is the preferred back-up since the volume in this tank is available without the time delay associated with having to ' break' condenser vacuum in the hotwells. The EFW Tank volume must be restored to within limit within 7 days because the backup supply is not designed to the same criteria as the EFW tank and may be fulfilling the requirements of this Specification in addition to its normal operational functions. The 4 hour Completion Time is reasonable, based on oaerating experience, to verify the OPERABILITY of the bac(up water supply. The 7 day Completion Time is reasonable, based on an OPERABLE backup water supply being available, and the low probability of an event occurring during this time period, requiring the use of the water from the EFW Tenk. (continued) Crystal River Unit 3 B 3.7-34 Amendment No. ,
i Emergency feedrater Tank , B 3.7.6 ; BASES , AC10 tis s.1 and 8 1 (continued) If the EFW Tank volume cannot be restored to within limit in the associated Completion Time, the plant must be placed in a MODE in which the LCO dr-es nat apply. To achieve this status, the plant must br :. laced in at least MODE 3 within 6 hours and in MODE 5 w h.iln 30 hours. The allowed l . Completion Times are reasonable, based on operating i experience to reach the required plant conditions from full > power condItiors in an orderly manner and without challenging plant systems. - SVRVftLLANCE SR 3.7.6.1 REQUldEMENTS This SR verifies that the EFW Tank contains the required , volume of cooling water. The 7 day frequency is bared on operating experience which indicates the water volur.e in the tank is normally static and the assumption the operator will be aware of plant operations that may affect the EFW tank inventory between performances of the SR. The frequency is considered adequate in view of other indications in the control room, including alarms, to alert the operator to reductions in EFW tank level. REFERENCES 1. FSAR, Table 10-2.
- 2. MAR 82 09-19-01, dated December 23, 1985.
- 3. FSAR, Chapter 14.
- 4. FSAR, Section 10.5.3.
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l Crystal River Unit 3 B 3.7-35 Amendment No. L
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EfW Syst0L-Shutdown B 3.7.18 8 3.7 PLANT SYSTEMS B 3.7.18 Emergency feedwater (EfW) System - Shutdown BASES - BACKGROUND The Background section for Bases B 3.7.5 1s applicable to these Bases. APPLICABLE 1he Applicable Safety Analysis section of Bases 3.7.5 is SAFETY ANALYSIS applicable to these Bases. 000 to the stable conditions associated with operation in MODE 4 and the reducea probability of occurrence of an event requiring EfW, the EfW operational requirements are reduced. Included in these reductions is that automatic Emergency feedwater Initiation and Control System (EFIC) actuation is not available. In this MODE, sufficient time exists for manual actuation of EfW to mitigate the consequences of an event requiring EfW. Only one train of EfW capable of being supplied from the motor driven pump is required in MODE 4. This requirement dictates that single failures are not considered during this MODE. LC0 The LCO section of Bases 3.7.5 is applicable to these Bases, in MODE 4, an EFW train capable of being supplied by the motor driven pump is required to ensure sufficient decay heat removal capability following an event requiring EfW. Similar to the concept employed in LC0 3.5.2, "ECCS Systems-Operating," cross connecting portions of both trains to produce a single operable train is acceptable in this plant condition. The motor driven feedwater pump is considered OPERABLE in MODE 4 with its hand switch put in pull to lock in order to prevent inadvertent EfW actuation. (continued) Ctystal River Unit 3 8 3.7-85 Amendment No.
EFW System - Shutdown B 3.7.18 BASES APPLICABILITY In MODE 4, the EFW system may be used for heat removal via I the steam generators. l l In h0 DES 5 and 6, the steam generators are not used for decay heat removal and the EFW System is not required. ; ACTIONS M , With the motor driver. EFW train (pump or flow path) inoperable, action must be taken immediately to restore the required train to OPERABLE status. The immediate Completion Time ensures that prompt action is initinted to restore the required cooling capacity. In the interim, alternative means of heat removal must be available such as the auxiliary feedwater train (FWP-7) or the main feedwater i system ca nozzles. pable of supplying If restoration cannotflow to the emergency be reasonably feedwater accomplished, the plant must be placed in Mode 5 in accordance with Action ! B.l. , M If Required Action A.1 cannot be completed within the associated Completion Time, the plant must be placed in a MODE in which the LC0 does not apply. To achieve this status, the plant must be placed in at least MODE 5 within 24 hours. The allowed Completim The is reasonable, based on operating experience, to reach the required plant conditions in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3.7.18.1 REQUIREMENTS The applicable Surveillance descriptio.1s from Bases 3.7.5 apply. REFERENCES 1. The applicable references from Bases 3.5.2 apply. I' Crystal River Unit 3 8 3.7 86 Amendment No.
1 FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUM11ER 50-302/ LICENSE NUM11ER DPR-72 ATTACIIMENT E LICENSE AMENI) MENT REQUEST 214 REVISION 0 OPERATOR ACTIONS l 1 l l l l l l l 1
1 ATTACllMENT E OPERATOR ACTIONS As discussed in the cover letter, Florida Power Corporation (FPC) is addressing the issues related to decay heat removal requirements in Mode 4 through a combination of Technical Specification changes and operator actions. This integrated approach ensures that the necessary accident mitigation systems are available. Tables I A and lit list the operator actions necessary to mitigate a Mode 4 LOCA. Table 1 A identifies those operator actions used to address the earlier stages of a RCS pipe break when the RCS leakage is less than the capability of the makeup system and is not classified as a LOCA. These actions are taken prior to emering the Eraergency Operating Procedures (ECP). Table til lists those operator actions taken to address a Mode 4 LOCA using the EOPs. To support the NRC's review of LAR 214, the results of simulator exercises of two scenarios are provided. These scenarios involve the tmlikely occurrence of an instantaneous reactor coolant system piping break greater than the decay heat drop line and a progressive break of the cold leg pump discharge pipe, beginning as a 30 gpm crack and advancing to a full rupture. Those operator actions taken after the RCS leakage has exceeded the capability of a aakeup pump and is classified as a LOCA are based on the EOPs. The EOPs affected by this LAR are in various stages of revision, and therefore, the information provided in this table is draft, llowever, FPC does not expect the final EOPs to affect these tables or the conclusions of this submittal. FPC requests that the NRC review these operator actions as an integral part of LAR 214 in order to achieve a comprehensive review of the Mode 4 LOCA strategy. l l l l l
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U.S. Nuclear Regulatory Commission Attachment E 3F1097-18 Page 2 TABLEIA OPERATOR ACTIONS PRIOR TO LOSS OF SUBCOOLING MARGIN OA Operator Action Time Basis Prior NRC Revieur Reference 1 Enter AP-520 upon mdcaton of RCS leakage N/A This procedure is entered upon an No AP-520 (Draft G). "1.oss of RCS and control Pressurizer Level using MUV-31, increase m RCS leakage or uncontroned Coolant or Pressure." Step 311 pressurizer level control decrease in RCS pressure. 2 Determne leak locaton N/A in the gwen scensno, the leak is in the No AP-520. Step 3.17 > RCS cold se g at 30 gpm and increases to 200 gpm over a 10 rrwnute per od indcatons of leakage would become evident to the operator upon recept of RB l sump level alarm, decreaseg makeup tank level, or makeup tank low level alarm.
^
3 hP Vents N/A isolated un:y00md are restored n a No AP-520. Steps 320 thru 322 DHV-3. Drop Line og step. PZR Spray Valves PORV and Block Valve Sample Lines Letdown Cooler inlet Valves 4 Isolate the RB sump by clossng RB sump N/A isolating the RB sump penetraten will No AP-520. Step 324 pump discharge valves, maintarn inventory in the containment for possible ECCS pump suden for long term reorculation 5 Enter EOP-8. "LOCA Cooldown.* if RCS N/A RCS leakage > 100 gpm is an entry No AP-520. Step 328 leakage is > 100 gpm u.n 66n. for EOP-8 EOP-8 (Draft O)."LOCA Cooldown"
U.S. Nuclear Regulatory Commission Attachment E 3F1097-18 Page 3 TABLE I A OPERATOR ACTIONS PRIOR TO LOSS OF SUBCOOLING MARGIN OA Operator Action Time Basis Prior NRC Review Reference 6 Establish RCS Cooldown usmg TBVs or N/A truttates RCS cooldown to acheve end Yes Use of TBVs and ADVs are ADVs pomt of event (start of decay heat) discussed in CR-3 Tectncal Specr5caton 3.7.4.and assocrated Bases. EOP-8. Step 3.18 7 Transfer ECCS section to the RB Sump N/A Sucbon transfer is requeed when the Yes Transfemng from the BWST to the BWST levelis < 15 ft RB sump is drseussed in Bases for CR-3 Techncal Specnicaton 3.5.2. EOP-8. Step 321 requres imibire w of EOP-14. Endosare 19."ECCS Sucbon Transfer
- 8 A!gn Decay Heat Pump (DHP) for decay heat N/A Assume other makeup M were tagged No EOP-8. Steps 3.78 thru 3 89 shgn removal and start pump out per An operatog procedure pnor and start DHP-1B to entenng Mode 4. Adequate subcoolog margm exists the HP1 pump is alegned to the RB sump and buildog spray pumps are not runnog 1
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U.S. Nuclear Regulatory Commission Attaciuneet E 3F1097-18 Page 4 TABLEIB OPERATOR ACTIONS AFTER LOSS OF SUBCOOLING MARGLN OA Operator Act>on Time Basis Prior NRC Revi*.er Re8erence 1 Tnp all runn:ng RCPs <2 Requred for loss of subcocing marge Yes NRC letter to FPC dated 5/29,1Mi rrunutes based on vodng condcon of reactor (Genenc Letter 86-05) refers to coolant B&W Owners Group (BWOG) stud es wNdh concluded that , w.v;.w.c& vnth 10 CFR 50 46 is
&,r e.ad if operator acbon to tnp RCPs is taken witNn 2 rrunutes.
EOP-3 (Draft Q) Step 2.1 and EOP-13 Rule 1." Loss of SCM" 2 Actuate LPI < 7.5 Depress the manual avbatron buttons on No EOP-3. Step 2.1 and EOF .13
- starts standby LPI pump minutes both *A' and H trams. Rule 1.* Loss of SCM"
- opens valves to BWST for both LPI pumps i 3 ff Subcoohng Marge (SCM) is lost, irutiate 10 Required for loss of subcoolmg marg n Part al EOP-3. Step 2.1 and EOP-13 Ru8e manual HPl and Reactor Budding isolation minutes (precedes automabe rubabon) 1.
- Loss of SCM" and Coohng (RBIC) Note: In this wedw. with LTOPS brrutations. HP1 vanes would not open B has been NRC letter dated 7EIT9 (SER for
-isolates letdown (MW-49/MUV-567)
-intbates HPl (see note) when HPIis iruhated. # ' "' O " 2 I
-isolates normal makeup (MUV-27) rw E. -> p p
- Wu RCP ml W M oW vab Note: EFIC wdl not feed due to RCP Procedure EP-3. currently EOP-3.
- actuates EFIC BypasWeset b Bypass. as wu'so-..g u - W to We
- irutiates Erimigw.cf RB coolog HM and m Ne HR h
U.S. Nuclear Regulatory CommissBan Attachment E 3FIO97-18 Page5 TABLEIB OPERATOR ACTIONS AFTER LOSS OF SUBCOOLING MARGIN OA Operator Action Time Basis Prior NRC Review Reference 4 Ensure a!I four HPI injecton vatwes are open 10 Eru.ures HPI injechon valves are Yes NRC letter to FPC dated S/29i79 minutes repc, c.M Per;wsit Solubon to SBLOCA
- switch power supply for affected injechon , .
to valves by manipulatog swrtches in control
'
- M b coen a'Yected HPI vatves by 10 maiutes.
See EOP-3. Step 3.3 5 Close DHV-3 decay heat drop kne isolabon N/A Protects rurnng DHP durrig loss of SCLt No EOP 3. Step 3.7 Prowdes for a second LPI trac 6 1solete RCP seat injecbon 20 Requeed to maximize HPl flow to reador Partal The need to isotate RCP Seal nwrutes irsectron was discovered in 1995 to be necessary due to descowry that operators reted on nort Reg Gude 1.97 instrumerfatson to measure thrs flow when determang HPt pump runout flow Emrts (see LER 95-026). Sealingechon isolation was a!so determined resary donng Refuel 10 in 1996 upon discovery that worst case instrumant error may result in inadeTzate HP1 flow (see LER 96-006) FPC letter dated July 7.1997 (NOV 96-07} discusses the add. tonal need for closure of RCP l seal contro' fed bleed-off (CBO) valves a*ter 90 seconds if seal ryecbon has not been restored. EOP-3. Step 3 8
1U.S. Nuclear Regulatory Commission - j Attachment E
- . 3F1097-18 Page'6 ,
1 TABLEIB. t-OPERATOR ACTIONS AFTER LOSS OF SUBCOOLING MARGIN f , + l OperatorAcenon OA Tsnee Beess Prior fGIC lloweeur Esfesence . , .
. 7, - Ensure adequate HPI Sow (isolate a broken 20' Requred only for tweak in HPI Ene. Partial FPC issued LER 9tK)D7 on 3/15/96 :
migocteiline) mwumes isolaban entens is > 50 gpm behoeen two to report a deogn beess condeon knes snwohnng HPI flow instrumentsten. The Dow denoenoes descobed 1- thereen were addressed by revised ! SBLOCA anefyses provided by , Framstome Technosospes in Apre j i 1996 which requeed isolehen of me l1 affected HPI kne venus 1 -4 i- Most recent FTIanefyses have ; , provided new isolabon enterm i t See EOP-3. Step 3 6 1
.8 Ensure adeauste EFV1 flow 20 Rasses OT5G Ievels to ISCM setpowit Partial NRC lenerto FPC dated 8/30!85 mnutes (90%) prowdes a SER for NUREG 0737 '
i- * " ** I N '. " W teem tlK3.30."SBLOCA Methods *
!q OM M Section I!!.5.a of the SER states h 1 _. trning of operator acton to raise the ! !~
seenndary system waterlevelto } 95% was found not to be cnhcal" ) EOP - 3. Step 3.9 l 9 Ensure Control Comosex Venhisbon is 30 Prowdes control complex cochng. No EOP-3. Step 3.12 ,
" N EOP-14 (Draft T). Eidoeure 17 p'
7 " Control Complex Ecc-rn~y i j Venhishon Systent* ! i -. 10 Transeon to EOP-8. N/A Transson when LPI flowis > 1400 gpm No EOP-3. Step 3.14 J a t f i 4 i I
U.S. Nuclear Regulclori; Commission Attachment E 3F1097-18 Page 7 SCENARIO 1 RCS BREAK GREATER TIIAN DECAY llEAT DROP LINE INITIAL CONDITIONS: f 0% Rated Thermal Power Reactor Tripped (all rods in) Governing Operating Procedure is OP-209 k0 Effective Full Power Days RCS Temp of 238'F (Mode 4) RCS Pressure of 217 psig Turbine driven feedwater pump (EFP 2)out of service MUP-1 A and MUP-lc tagged out PORV selected to low range llPI valves de-energized at power source switches on MCB and administrativelycontrolled RPS in shutdown bypass MUP H running , EFIC Channels A/B/C/D selected to *RCP Bypass" Motor driven feedwater pump (EFP-1) block and control valves energized EFP-1 in normal after stopposition EFIC Cabinets energized OTSG levels at 350" for wet lay up. RCS is on Decay lleat using DilP-1 A 2 e RCS leak develops in cold leg (1 ft ) ENPECTED PROCEDURE US.&GE1 E0P 13 "EOP Rules" EOP-3 " Inadequate Subcooling Margin" E0P-8 "LOCA Cooldown" E0P-14 "EOP Enclosures" CIIALLENGl?Si Upon indications of RCS leakage (RB sump level alarm, or decreasing pressurizer level), Operators will be alerted to the LOCA in progress. Subcooling margin will be lost as indicated by the Safety Parameter Dkplay System (SPDS). Operators will perform Rule 1 of EOP-13, and enter EOP-3, "Inadecuate Subcooling Margin." RCPs are verified tripped, and manual llPI and manual RBIC are actuated per Rule 1. LPI will be manually actuated ; by the operator after confirming that RCS pressure is .<;.150 psig (per new step in Rule 1). EOP 3 ensures all IIPI valves are opened by swapping pov/er supplies (Step 3.3).
U.S. Nuclear Regulatory Commission Attachment E ; 3F1097-18 Page 8 SCENARIO 1 ! EOP-3 step 3.7 (new step) will close DlIV-3, drop line flow to DilP-1 A. EOP-3 step 3.9 requires EFW to feed OTSGs to the ISCM setpoint. EFIC must be controlled in manual per EOP 13, Rule 3 because EFIC is in "RCP bypass." ; EOP 8 is entered due to LPI flow > 1400 ppm. Scenario ends when EOP 8 is completed and ECCS has been transferred to the RB Sump. ; SjGNIFICANT RESULTS: Purpose of scenario is to demonstrate the operator's response to an instantaneous , rupture of the RCS the size of the decay heat drop line. Operator response was acceptable in that:
. Loss of subcooling was detected and appropriate EOPs were entered, e LPI was initiated within 2 minutes of break,
- _ llPI was initiated and power selector switches were used to re-power the valves (in
. LTOPS mode),
o EFW was initiated, and .
- Cooldown was initiated using appropriate EOPs.
I L umye- y -- W e- -n.9 ,Tt - W4-" = -t' e - --- - + -*--'M'>eW' T *T* eer
U.S. Nucleer Regulatory Commission Attachment E 3F1097-18 Page 9 SCENARIO 2 PROGRESSIVE HREAK OF THE COLD LEG PUMP DISCHARGE INITIAL CONDITIONS: 0% Rated Thermal Power Reactor Tripped (all rods in) Plant Cooldown in progress per OP 209 300 effective full power days RCS temperature of 268'F RCS pressure of 442 psig Turbine drive emergency feedwater pump out of service Makeup pumps I A and IC out of service Makeup pump til running PORV selected to low range llPI valves de-energized at power source switches on MCil and administrative controlled RPS in shutdown bypass EFIC channels A/B/C/D selected to "RCP Ilypass" Motor driven emergency feedwater pump (EFP-1) block and control valves energized EFP 1 in normal after stop position EFIC cabinets energized e RCS leak develops in cold leg at 30 gpm.
- Leak ramps up to 200 gpm over a 10 minute period, e When RCS is on Decay lleat using DilP-lil, cold leg breaks open (5000 gpm break)
EXPECTED PROCEDURE USIMG]h AP-520 " Loss of RCS Coolant OR Pressure" EOP-8 "LOCA Cooldown" EOP-3 " Inadequate Subcooling Margin" EOP 13 "EOP Rules" EOP-14 "EOP Enclosures" CHALLENGES: Upon indications of RCS leakage (RB sump level alarm, or decreasing MUT level), AP-520 will be entered. Pressurizer level will be maintained with normal makeup via MUV-31 pressurizer level control. Upon completion of AP-520, EOP-8 will be entered due to RCS leakage > 100 gpm. EOP-8 will direct a conldown and establish DilR using DilP-1B because the A-train MUP is running. After DilR is established and RCPs are secured, the cold leg break increases to approxinutely I square foot.
U.S. Nuclear Regulatory Conunission Attachment E 3F109718 Page 10 SCENARIO 2 (cont.) Subcooling margin is lost as indicated on SPDS. EOP-3 is entered, and Rule 1 Loss of SCM is performed. RCPs are verified tripped, and manual llPI and manual RillC are actuated. EOP-3 ensures all llPI valves are opened by swapping power supplies (Step 3.3). EOP-3 step 3.8 requires EFW to feed OTSGs to the ISCM setpoint. EFIC must be controlled in manual per EOP-13 Rule 3. EOP-3, Step 3.9 will require manual LPI > 1400 ppm. Scenario ends when EOP-8 is completed prior to ECCS transfer to Ril sump. SIGNIFICANT RESUI,TS: Purpose of scenario is to demonstrate the operator's response to the RCS leakage and entry into the required EOPs. Operator response was acceptable in that:
- RCS leakage was detected.
- Attempts were made to isolate the source of the leak, e RCS inventory was maintained, and
. Cooldown was initiated using appropriate EOPs.
FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUh1BER 50-302/ LICENSE NUMBER DPR-72 ATTACHMENT F SAFETY ASSESSMENT
Mmic 41,0CA Safety Assessmmt introduction and Hackground The purpose of this safety assess:nent is to document Florida Power Corporation's approach to addressing hiode 4 Loss of Coolant Accidents (LOCAs). This assessment includes background information on the impetus behind this effort, a description of analyses demonstrating that sufficient time is available for operators to mitigate this unlikely event, and simulator exercises (Attachment 4) demonstrating that the current draft abnormal and emergency cperating procedures at Crystal River 3 are adequate for mitigation. This effort addresses Technical Speci6 cation Change Request Notice (TSCRN) 210, which included a commitment that "An engineering evaluation is being performed on issues concerning decay heat removal in biode 4." This evaluation is required to be performed prior to restart. An analysis' performed to support TSCRN 210 determined that Once-Through Steam Generator (OTSG) cooling is required for certain small break less-of-Coolant Accidents (SBLOCA). There are several variables that define OTSG cooling mission time: break size and location, decay heat load, and high pressure injection (IIPI) flow which is a function of the number of IIPI lines and pumps assumed in operation. If only one llPI oump is oparational, then OTSG cooling is required for 35 hours. With two IIPI pumps operating, OTSG cooling is required until the Emergency Core Cooling System (ECCS) is configured for
'* piggyback " operation to support taking suction from the reactor building emergency sump (RDES).
In addition, TSCRN 213 will inco ,, orate a new Limiting Condition for Operation (LCO) into the improved Technical Specifications (ITS) for implen enting Low Temperature Overpressure Protection System (LTOPS) requirements. There are operational limitations identified in TSCRN 213 that are designed to minimize an overpressure event during low temperature conditions. Several limitations are associated with the ECCS which include " ..a maximum of one hiakeup pump capable of injecting into the RCS, Iligh Pressure Injection (IIPI) deactivated, the Core Flood Tanks (CFT) isolated..." These limitations potentially affect the plant con 0guration during the mitigation of a hiode 4 LOCA. Further, certain Action Statements require shutdown to hiode 4 within 12 hours. ITS currently does not require emergency feedwater (EFW) OPERABLE in hiode 4. In addition, LCO 3.5.3 requires only one train of ECCS OPERABLE and allov.s an LPI subsystem to be considered OPERABLE when aligned for decay heat removal as stated in Surveillance Requirement (SR) 3.5.3.1. The Plant Cooldown operating procedure (OP-209), would deactivate llPI as soon as possible after entering hiode 4 by closing and de-energizing the IIPI valves and racking out the makeup pumps (hiUP) not in service. In addition, EFW would be
' Fil Document 51 1268138 01, submitted to the NRC in TSCRN 210 8
HPl pump (s) aligned to take suction from the discharge of Low Pressure injecthn pump (s) i
- Page 1 -
U.S. Nuclear Regulatory Commission Attachment F 3F109718 Page2 removed from service (the next procedural step in OP-209) by closing and de-energizing the EFW block valves, racking out the breaker for the motor driven EFW pump (EFP-1), and placing the control handle for EFP-1 in pull-to-lock. For these reasons. FPC is committed to addressing the concerns associated with Mode 4 LOCAs by providing comprehensive ITS changes which facilitate the mitigation of such events. Discussion CR3 is pursuing an opera'.or action based strategy, supported by analyses and Emergency Operating Procedure (EOP) validation, to mitigate Mode 4 LOCAs. Framatome Technologies Incorporated (FTI) was requested to provide an engineering evaluation to address mitigation of a LOCA postulated to occur in Mode 4. Fil documented their preliminary evaluation in letters' to Florida Power Corporation (FPC). This safety assessment provides a summary of those FTl letters and FPC's review of Mode 4 LOCAs. This evaluation, which will be finalized by the end of November 1997, considered breaks in the RCS piping, as well as breaks in the piping attached to the RCS hot and cold legs. The simulator exercises of CR 3's EOPs are described below. Or.rator Actions The key assumption for the analysis is operator action. Since the ECCS automatic actuation function would be bypassed in Mode 4, manual operator action would be required to make the necessary ECCS con 0guration changes to initiate injection now through additional lines to provide greater How rates than what is available through the normal makeup line. As described in the Introduction and Background section, the llPI system end devices would be currently deactivated by removing power at circuit breakers which are not located within the control room. This method of administrative control, used to implement LTOPS, prevents
- timely restoration of IIPI components needed to cope with LOCAs.
As an alternative, the current control strategy signincantly improves accident response by enabling restoration of IIPI valves from within the control room. FPC intends to implement the de-activation of IIPI when in LTOPS conditions by de powering the four IIPI injection
, valves using the power source selector switches on the Control Board. By applying this LTOPS protection, a minimum of two operator actions would be required to initiate IIPI.
These actions are to actuate IIPI and then restore power to the llPI valves. With the implementation of the IIPI valve administrative control method, now through all four IIPI lines can be initiated from the control room within 10 minutes, in addition, LPI can be initiated from the control room within 10 minutes either by manual ESAS actuation, or by manually starting the LPI pumps and aligning the LPI How path. These assumptions are 8 INS-97 3R74. Task 4101026. dated 10/1/97 I'rt Division Personnel Support Activities for htode 4 LOCA Evaluation INS 97-3939. Task 4101026 dated 10/3/97 - 171 Division Personnel Support Activities for hiode 4 LOCA Evaluation
U.S. Nuclear Regulatory Conunission Attachment F 3F109718 Page 3 reasonable and consistent with the Nuclear Regulatory Commission's position on simple operator actions (taken within the control room)'. The purpose of TSCRN 213 is to ensure that inadvertent actuation of IIPI in certain low temperature conditions does not overpressurize the reactor vessel. Break Site Analyses With credit for operator action within 10 minutes, the analysis demonstrated successful mitigation of hiode 4 LOCAs on RCS and attached piping of various break sizes. Breaks of up to 0.25 ft' in the RCS hot and cold legs have been evaluated. The results indicate that initiating flow from one llPI or LPI pump within 10 minutes will prevent core uncovery and ensure that no significant fuel heatup occurs. Although the licensing basis postulates a hiode 4 LOCA, there are minimal driving forces to cause a large break of the reactor primary piping at the lower temperature and pressures of hiode 4. Therefore, it is reasonable to infer that a decreased rate of reactor primary piping degradation would occur and operators would have additional time to be alerted and respond to such an impending LOCA. These operator actions would include attempting to identify and isolate the source of the leak and ensuring RCS inventory is maintained through normal makeup. Further, CR 3 ITS 3.4.12 requires the plant to be placed in hfode 5 if pressure boundary leaks exist or if identified or unidentified leakage exceeds certain limits. These actions are a part of cristing CR-3 procedures. Also, emergency operating procedures require initiation of LPI to maintain the core covered with reactor coolant if such a leak did propagate into a LOCA in hiode 4. Consideration of the decreased rate of reactor primary piping degradation in hiode 4 provides additional margin for the time available for the operator to respond to a postulated large break of reactor coolant primary piping. Simulator exercises of the operators' response to both small and large break LOCAs confirm operator capability to implement the abnormal and emergency operating procedures required and maintain the core covered with reactor coolant. 2 Several breaks of attached piping were also examined. A break of the 0.44 ft Core Flood Tank (CFT) line was also analyzed. The results showed that this break did not uncover the core within 10 minutes, permitting the operators to diagnose the event and take mitigative action. On the other hand, a break of the pressurizer surge line is large enough that it may result in core uncovery in less than 10 minutes, similar to the Decay lleat (Dil) drop line discussed below, llowever, the consequerces of a surge line break are bounded by the Dli drop line analysis, since the surge line has a smaller diameter than the Dll drop line (resulting in a longer time to core uncovery), and because the surge line connects to the hot leg at a higher elevation than the Dil drop line.
- Ixiter 3N097847 USNRC to FPC, long-Term Modifications on liCCS SmaH ikeak Analpis I
l
U.S. Nucle:r Regul: tory Commission Attachment F 3F109718 Page 4 The analysis shows that a break of the Dil drop line (0.737 ft') is the most severe break of the attached piping postulated in blode 4. Fcilowing a break in the Dil drop line, the operator may not have adequate time to protect the decay heat pump aligned to the Dil drop line due to vortexing and the subsequent loss of NPSil that will occur as the RCS liquid level becomes too low to sustain pump operation. With the requirements imposed by LAR 214, one train of ECCS would be available to replace the reactor coolant lost through the break. For conservatism, no credit will be given to the LPI pump aligned for decay heat removal to satisfy the LCO for hiode 4 ECCS. , Following a break of the Dil drop line, core uncovery is predicted in approximately 7.5 minutes. This is the bounding scenario for attached RCS piping because of its break size and location. Consistent with the analyses supporting TSCRN 210, llPI line pinch breaks require either two llPI pumps or one IIPI pump and OTSG cooling to mitigate the LOCA. IIPI hne breaks will also require injection How to be established through four IIPI lines, with isclation of a detectable llPI line break (i.e., flow difference between broken llPI line and next highest IIPI line flow rate of greater than 50 gpm). As such, LAR 214 will continue to require one train of ECCS OPERAHLE in hiode 4 without allowing an LPI subsystem to be considered OPERAHLE while in the decay heat removal mode. EFW will also be required to be OPERAHLE in hiode 4, using EFP-1. Abnormal and Emergency Operating Procedure To demonstrate that the draft abnormal and emergency operating procedures are adequate to assist in the diagnosis and mitigation of a hiode 4 LOCA, a simulator exercise was run for a 2 1 ft cold leg break. This exercise showed that with the use of these procedures, the operators could act to initiate LPI within approximately 2 minutes, and IIPI within 10 minutes, in addition, to assure that the symptoms associated with a Leak Before Break scenario can be diagnosed and appropriate actions taken, a simulator exercise was performed for a cold leg break. The exercise demonstrated that the symptoms of pipe degradation could be correctly diagnosed, and that the draft Emergency Operating Procedures and Abnormal Procedures provided the necessary steps to mitigate the event. Summary and Conclusions As a result of analytical work performed by FTI to support TSCRN 210, a better understanding o.' the role OTSG cooling plays in SHLOCA management was realized. Accordingly, LAR 214 provides for additional ECCS requirements in hiode 4. It is reasonable to postulate that a decreased rate of reactor primary piping degradation in hiode 4 would occut r,nd operators would have additional time to be alerted and respond to 2 such an impending LOCA. For Core Flood line breaks and breaks less than 0.25 ft , initiating How from one llPI or LPI pump within 10 minutes will prevent core uncovery and ensure that
U.S. Nuclear Regulatory Commission Attachment F 3F1097-18 , Page5 no significant fuel heatup occurs. IIPI line breaks will also require injection flow to be , established through four llPI lines, with isolation of a detectable llPI line break (i.e., flow ; difference between broken llPI line and next highest ilPI line flow rate of greater than 50 gpm). Mode 4 LOCAs are unlikely given the lower temperatures arx! pressures, stable operating conditions, and decreased rate of pipe degradation. Additionally, Mode 4 is a tran:itory mode in which CR 3 in recent history has been la for less than 1% of time. Ilowever, should such an event occur. It can be mitigated with operator actic .s. Provided that the llPI valves are . activated from within the control room, llPI 116w ca* e initiated within 10 minutes. LPI will ! also be available within 7.5 minutes upon implen" .,n of LAR 214. In addition EFW via EFP-1 will also be available to manage decay h. for those SBLOCAs that require OTSG cooling.
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