ML20041A648
| ML20041A648 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 02/17/1982 |
| From: | BALTIMORE GAS & ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20041A639 | List: |
| References | |
| NUDOCS 8202220377 | |
| Download: ML20041A648 (27) | |
Text
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l 2.1 SAFETY LIMITS .
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{
EASE 3 2.1.1 REACTOR' CORE . .
The restricticns . this safety limit prevent overheating of the ssible cladding perfora:icn which wculd result in the fuel cladding and pdrocucts to the reactoi coolant. Overneating of the ..
releaseoffissien$ymaintainingthesteadystatepeaklinearheatrata fuel is prevente j at or less than 21 kw/ft. Centerline fuel =el:ing will not ec:ur :
for this peak linear heat rate. Overheating of the fuel cladding is.
prevented by restricting fuel cperatien to within the nucleate boiling !
regime whers :ne hea: transfer coefficient is large and :ne cladding surface temperature,is slightly above the c olant saturatien tampgrature.
Operation abcve the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures :ecause of the enset of departure frcm nuclea:e boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. CNB is not a directly measurable parameter during operatien and therefore THERSAL POWER and Reac er Coelan: Temper- ;
ature and Pressure h(ve been related :s Dii3 thr: ugh the CE-1 c:rrelatien.
The CE-1 DNS correlatien has been developed t: predic the DN5 flux and the locatica cf DN5 fer axially unifem and ncn-unifem heat flux cistri-4 i butiens. The 1ccal DN5 heat flux ratio, CNER, defined as the ration of 1
the heat flux that would cause CNB at a particular core lccation to the local he2: flux, is indicative cf the margin to DNS. .
The minimum value of the CNER during steady state coeration e
- operatienal transients, and antici:ated transients is limited l.2 3 ltoh, This value c:rres: ends :: a g5 percent pr::a:ility a: a 95 percent cen-fidence level that CH3 will nc: oc:ur and is chosen as an appropriate .
=argin to DN3 for all cperating conditions. /23 The curves of Figures 2.1-1, 2.1-2 .1-3 and 2.1-4 shew the loci of points of TH5??AL 70WER, Rea r Ccolant System pressure and maxi:mn ccid lac tem:erature of.H: va-for cus cum:
One c:mhinations family fer whien of axial sha:es andthe l
minimum GN5R is'no less : nan correspending radial peaks shewn in Figure 52.1-1. The limits in Figures 2.1-1 , 2.1-2, 2.1-3 and 2.1 4 were calculated fer reac::r c:elant inlet :a :eratures "ess than er ecual :c 550*F. The dasned line a 520*F c clan: inle: tamperature is no a safety limit; hewever, c;eratien a:cve 520*F is no: ;cssible because of the actuatien of the main steam line safety valves wnich limit the maxi =um value of react:r inlet tam:erature.
React:r c: era icn at THE:?al FCWER levels higher than@cf tvu .s snR"AL PCWER is p(chibitec by the high pcwer level trip se poln specified in i10 %
! Amend =ent No. M, 45 CALVERT CLIFF 5 - UNIT 1 B 2-1 8202220377 820217 -
i 1711
- 1
. SAFETY LIMITS -
I BASES
. I Table 2.1-1. The area of safe operation is beicw and to the left of )
these lines. .
1 The conditions fer the Thermal Margin Safety Limit curves in Figures 2.1-1, 2.1-2, 2.1-3 and 2.1-4 to be valid are shewn en the figures.
The reactor protective systam in c:=binatien with the Limitina -
Conditions for Operatien, is desigt.ed to prevent any antici;a:ed ce=bina-tion of transient cord'.tions fer reactor ceciant syste= ta=:erature, pressure, and THERMAL FCWER level that would result in a DNBR of less than 1.19s and praclude the existance of flew instabilitias.
/. 2 3 2.1.2 REACTOR CCOLANT SYSTEP. PRESSURE The restriction tf this Safety Limit protects the integrity of the Reacter Coolant System fr:m over;ressuri:ation and thereby prevents the release of radienuclides c:ntained in the reacter coolant fr:m reaching the containment a:=cspnere.
The reactor pressure vessel and pressuri:er are designed to Section III,1967 Ecition, of the ASME C:de f:r Nuclear Fewer Plant Cc=penents which pent.1:s a maximum transien pressure of 110 (2750 psia) of design pressure. The React:r Ccoiant Systen piping, valves and fittings, are desicned to ANSI 5 31.7, Class I,195g Editi:n, wnien per=its a =axi===
transient pressure of 11Ct (2750 psia) of c:=ponen: design pressure.
The Safety Limit cf 2750 psia is :nerefere consistent with :ne cesign criteria and asscciated coda requirements.
The entire Reactor Ccoian: System is hydrotested at 3125 psia to demenstrate integrity prict to initial cperation.
I CALVERT CLIFFS - UNIT 1 B 2-3 Amend =en: Nc. 73, ?*, 40
ut 2.2 LIMITING SAFETY SYSTEM SETTINGS .
( -
I BASES 2.2.1 RUCTCRTRI?SETFO!kTS The React:r Trip Setpoints s:ecified in Table 2.2-1 are the values at which the React:r Trips are se f:r each parameter. The Trip Set:cints have been selecte: to ensure that the react:r c:re and react:r :: lant systa are prevented frc= exceeding their safety limits. 0:eratien with a tric set less c:nservative than its Trio Se:pcint but within its speci-fied Allcwable Value is ac:eptable en :ne basis tha the difference between the trip se: point and the All:wable Value is equal t: er less' than the drift allcwance assumed for each trip in the safety analyses.
Manual Reacter Tri The Manual React:r Trip is a redundant channel t: the aut: atic protective instrumentation channels and provides manual reac::r trip-capability. ,,
Pcwer Level-Hich The Power Level-High trip provides react:r c re pr:te::icn against reactivity.excursiens wnich are t:: rapid :: be pr:: acted by a Pressuri:er Pressure-Hign er Thermal Margin /Lew Pressure trip.
The Pewer Level-High tric set:: int is c: erat:r adjustable andC;can be era:Or set no hicner than 10% ab ve :ne incicated THERMAL FCWER level.
action is'recuirec :c increase :ne trip set:cin: as THERMAL FGWER is increased. The trip se:; in is aut::atically de reased as THERMAL pcwer l decreases. The trip se::: int has a v.axi=u.a value cf 107.0% cf RATED THERMAL F0WER anc a =inimu= se : int cf 3C: of RATED THERMAL FCWER.
Addinc :: this maximum value the p ssible varia:icn in trip point due :=
errers, the maximu= actual steadv-state calibrati:n and instru=en cf RATED
. THERMAL PCWER level at which a trip would be actua:ec is d2:
THERMAL F0WER, wnich is the value used in the safety analyses. llo 7a peacter C clant Flew-l w
'The Reacter C:olant Flew-L:w trip provides c:re pr::ecti:n to prevent DNS in the event of a sudcen significant decrease in rea:::r :::ian: ,
flew. Provisiens have been =ade in the react:r pr: ective systa= :: per=1;
(
CALVERT CL:FFS - UNIT 1 B 2-4 Amend =ent No. 33
173l LIMITING SAFET( SYSTEM Su iIN35 ..
I BASES ,
i
- I.23 operation of the reacter at reduced power if one or we reacter coolant pumps are taken out of service. The icw-ficw trip etpoints and Allcwable Values for the varicus reacter c:olant pu=p c:=bi tiens have been derived in censideration of instrument errors an response times of equipment involved to maintain the DNER above .19' under normal oceration l
. and expected transients. For reacter operation with only two or three reactor c:clant pu=:s operating, the Reactor Ccolant Flew-Lew trip set-points, the Pcwer Level-High trip set;oints, and the Thermal Margin /Lew Pressure trip se:;oints are aut==atically changed when the pump condition selector switch is manually se to the desired two- or three-pu=p g./.3 position. Changing these trip se:coints during two and three pum operatien prevents the minimum value of DNER frem going belcw 1.195 during l normal operational transients and anticipated transients when only two er three react:r coolant pumps are operating. -
I pressurizer pressure-Hich .
. The Pressuri:e'r Pressure-High trip, backed uo by the pressuri:er cede safety valves anc main steam line safety valves, previces reactor coolant system protacticn against over:ressuri:ation in the event of loss of load withcut reac:cr trip. This trip's ser;cint is 100 psi below the ncminal lift setting (2500 psia) of the pressuri:er code safety valves and its concurrent operation with the pcwer-operated relief valves avoids the undesirable operation of the pressuri:er c de safety valves.
Centainment pressure-Hieh The Containment Pressure-High trip provides assurance that a reactor trip is initiated c ncurrently witn a safety injectien. The setpoint for this trip is identical to the safety injecti:n se: point.
Steam Generater pressure-Low The Steam Generat:r Pressure-Lew trip provides protec:icn against an excessive rate of heat extracticn fr:m tne steam generators anc subsecuen: c:oldewn of the reac:ce ecolan:. The setting of 570 psia l is sufficiently belcw the full-leac c:erating poin of 550 psia so as not to interfere with ner. al c: era:icn, but still high enough to provide the required pr:tec:icn in the even: of excessively high steam flew. This setting was used with an uncertainty fact:r of 1 22 psi in the ac:ident analyses.
, r l CALVERT CLIFFS - UNIT 1 B 2-5 Amend:ent No. 37.48
- a. ,\
i',
< LIMITIl:0 SAFETY SYSTEM S . .I?:GS 1
i
~
BASES t
Steam Generat:r Water Level .
The Steam Generater Water Level-Lew trip provides c:re ;retection by preventing c;eratien with the steam genera::r water level beicw the '
minimum volu=e required fer adequate heat re=cval capacity and assures
- that the pressure of the reac cr :: lan: system will not ex:end its '
Safety Limit. The specified se::: int previces allewance tha: there wf11 be sufficient water invent:ry in the steam genera: rs at the tice of-trip t: pr: vide a margin of mere than 13 minutes before auxiliary feedwater i.s requ.i red. ..
Axial Flux Offset l.13 The axial flux effset trip is pr vided : ensure that excessive axial peaking will net cause fuel damage. T e axialThe fluxtrip offse: is set;cints determined frem the axiaily split exc:re d .sc :rs. :
ensure that neither athe DNER of less than({}fuei tec;erature f:r centerline melting will{)ncr a :tak; which c:rrespencs t: exist as a c:nsecuen:e of axial pcwer =aldistributicas. These trip set-points were derived fr:m an analysis of =any axial p:wer sha';es with allewances f r instrumenta:ien inac:uracies and the uncertainty asse:izted with the excern t: incere axial flux cffse relatienship. Thermal Marcin/Lew pressure The Thermal Margin /L:w Fra sure trip is provided :: prevsat perati:n when the DNER is less than 1.19- w,g3 The trip is initiated whenever the react:r c: clan: systam pressure signal dr::s belew either 1750 csia er a c:::utad value as ces:ribed below, whichever is hicher. The c =puted value is a fun::icn of :ne
- hicher of ai ;:wer er neutr:n p:wer, react:r inlet tem:erature, and the nuh.ber cf react:r c:Olan: ;u=:s c:erating. The minimum value of rea ::r c: clan: ficw rate, :ne maxi =um AI;y.UTHAL pCWER TILT and the maximum CEA deviatica permitted f:r ::ntinu:us ::eration are assumed in the genera-tien of this tri: fun::icn. In additien, CEA gr:u: sacuencing in ac::r-Finally, One dance with 5 ecifica:icns 3.1.3.5 and 3.1.3.5 is assu=ec.
maxi =um insertien of CEA banks wnich can c::ur during any antici;ated I cperational oc:urrence prior := a F:wer Level-nign trip is assumed. l 5 2-6 Amendment Nc. 23, ?? ,#3 CALVERT CLIFF 5 - UNIT 1 h
. . 175 ,
. l I
LIMITING SAFETY SYSTEM SETTINGS l l l ( BASES
+o INCLMOS The Thermal Margin /Lew Pressure trip etpoints 1 : d: d '-- the 4:r: ::'ety : ?"--"f arp-:*': f equipmentres:ensetime,measurementunce/.;;r:~'=-=alicwancesfortain
' ::f:t> ::- ' ': prr/id:d 'i:h in:hd :: er s' r in:: ;f 5% :,f
- ATC 7"~.T J.; '~;;C :: :: :n:::: #:r ;;f:nti:1 ;; :r :::u- -2 :r :r; On : :u n n f 2* te n n n:::: #:r ; 2-M
'5- e- r-2 -" nrr:nt F ert:' y; and a further allcwance of @ psia to ce=pensate for T//S l
;-e -- e n:r;. :nt ~r , tri; :yt t: y ?: '1 :~:r, rd time delay associated with providing effective termination of the occurrence that exhibits the most rapid decrease in margin to the safety limit. 5: C l
-p:n =11-="-= ': d: ; :# : 22 ;:i: ;rn:m . . n;.c;.;;nt ii n n::
erM i " 7F: W: d::;:. hwan. l Asymmetric Steam Generatre Transient Drotection Tric Function (ASGTpTF) The ASGTPTF utilize;. steam generator pressure inputs to the TM/LP calculator, which causes a reactor trip when the difference in pressure between the two steam generators exceeds the trip se: point. The ASGTPTF is designed to provide a reactor trip for those Anticipated Goerational Occurrences associated with secondary system malfunctions which resul: in asy= etric primary icop ccclant tem:eratures. The most limiting event is the loss of load to one steam generator caused by a single Main Steam Isolation Valve closure. The equipment trip setpoint and allcwable values' are calculated to account for instrument uncertainties, and will ensure a trip at or before reaching the analysis setpoint. _ Loss of Turbine A Loss of Turbine trip causes a direct reacter trip when cperating above 15'; of RATED THERMAL POWER. This trip provides turbine pre:ection, reduces the severity of the ensuing transient anc helps avoid the lifting of the main steam line safety valves during the ensuing transient, thus extending the service life of these valves. NoItscredit was taken in the functicnal cacabili:y accident analyses for cperation of this trip. at the specified trip setting is reccired to enhance the overall reliability of the Reac er Prc ection System. Rate of Chan e cf p:wer-Mich The Rate cf Change of Power-High trip is provided to prc:ect the cere during star v; c erations and its use serves :ne acministra- i tively enf:rced 3:artup ra:e limit. Its tripas a backuo se:::in ccest:not c:rres:end ! to a Safety Limit anc no credit was taken in the acciden; analysas f:r i opera:icn cf this trip. Its functional capability a. -he 3:ecified trip setting is recuired :: ennance tne overall reliability of the Reac::r Protection System. S 2-7 Amendmen No. 27, 32, J?,'E CALVERT CLIFFS - UNIT 1 . I
3/4.1 REACTIVITY CCilTRCL SYSTEMS . 3/4.1.1 50RATICtl CONTROL SHUTDOWN MARGIt - T,ye > 200*F. LIMITING C0tlDITI0tl FOR 00Ep.aTION - p374 ' 3.1.1.1 The SHUTDOWN MARGIN shall be't ak/k. APPLICABILITY: MODES 1, 2**. 3 and 4. ACTION: gM ~ ak/k, immediately initiate and centinue With the SHUTDOWN MARGIN < boration at > 40 gpm of 2300 ppm boric acid solutien or equivalent until the requiredSHUTOCWN MARGIN is restored. SURVEILLANCE REOUIREMENTS
, $' 3 ?a 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be 114.3$ Ak/k: l
- a. Within one"hcur after detection of an inoperable CEA(s) and at least once per 12 hours thereafter while the CEA(s) is inoperable.
If the inoperable CEA is immevable er untric;able, the above v required SHUTOOWN MARGIll shall be increased by an amcunt at least equal to the withdrawn worth of the immevable or un:rippable CEA(s) .
- b. When in MODES 1 or 2 , at least once per 12 hours by verifying that CEA group withdrawal is within the Transient Insertien Limits of Specificatica 3.1.3.6.
a
- c. When in MODE 2", within 4 hcurs prior to achieving reacter criticality by verifying that the predictec critical CEA position is within the limits of Specificatien 3.1.3.6.
- d. Prior to initial operatien abcve 5% RATED TriE?fGL FOUER after each fuel leading, by censideratien of the fac:;rs of a belew, with the CIA grcups at the Transient Insertien Limits of Specification 3.1.3.5.
* - Adherence to Technical Specification 3.1.3.5 as specified in Surveillance Requirements 4.1.1.1.1 assures that there is sufficient available shut-
. down margin to ca:ch the shutdown margin requirements of the safety analyses.
** See Special Tes: Exce::tien 3.10.1. -
- # With Xerr .. >-1.0.
i! With Ker f < t.0. v Amendment No. 2.7, 37,'O l CALVERT CLIFFS - UNIT 1 3/4 i-1
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W3mO41vnw3H131sVM011Y 40 NO113YWd 5o g a. O CALVEP.T CLIFFS . UNIT 1 3/4 1-27 Amendment No. 2J, 2/,33
U O POWER DISTRIBUTION LIMITS SURVEILLANCE REGUIREMENTS (Centinued) - c.' Verifying at least once per 31 days that the A%IAL SHAPE INDEX is maintained within the limits of Figure 3.2-2, where 100 percent of the allowable power represents the maximum THETWAL POWER allowed by the following expressien: MxN where: , , i 1. M is the maximum allcwable THERMAL POWER level for the - existing Reactor Coolant Pump ccmSination. ,
- 2. N is the maximum a11 cwa 51e fraction of RATED THEN*d.AL POWER as detemined by the F'Y X curve of Figure 3.2-3b
- 0. - . -
,. 4.2.1.4 Incore Detector Monitorino System - The incere detector meni , toring system may be used fcr menitoring tne core power distributien by verifying that the incore detector Lccal Pcwer Density alarms:
- a. Are adjusted to satisfy the requirements of the core power distribution map which shall be updated at least once per 31 days of accumulated operatien in MODE 1.
- b. Have their alarm setpoint adjusted to less than or equal to the limits shewn on Figure 3.2-1 when the following factors are
- appropriately included in the setting of these alarms:
' ' 1. Flux p,eaking augmentation factors as shown in Figure 4.2-1,
- 2. A measurement-calculational uncertainty fact,r of 1.070, l
- 3. An enginescing uncertainty factor of 1.03,
- 4. A linear heat rate uncertainty factor of 1.01 due to axial fuel densificatien and thennal expansien, and
- 5. A THE?c'AL POWER measurement uncertainty factor of 1.02.
CALVERT CLIFFS - UNIT l 3/42-2 A=endment No. 77, Ef , 72, 73,3 9
179
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180 ' POWER DISTRIEUTIO!! LIMITS l , TOTAL PLA!iAR RADIAL PEAXIliG FACTOR - Fh -
- i LIMITI?:G CC!iDITIO!! FCS OPEDATIO!i T T
.e es atedvalueofF,y,definedasF,y=F,y(1+T),sna11be q 14T APPLICABILITY: M00E 1*.
ACTIO?l: ,g . WithFfy>1A,.within6hourseither: 3 . 2. -3 0- l
- a. Reduca THERMAL POWER to bring the c::chinati of THERMAL POWER and F' to within the limits of Figure 2.: 2-and withdraw the full Ungth CEAs to or beyond the Long Term Steady State Insertion Limits of Specification 3.1.3.6; or
- b. Se in at least HOT STA!iDBY.
( . SURVEILL2.!!CE REOUIRE"E?tTS
- k+
- .2.2.Y gk. provisiens of Specification 4.0.4 are not applicable.
4.2.2. Fjy shall be calculated by the expression Ffy = Fxy,(1+T)andFjy q sha e detennined to be within its limit at the following intervals:
- a. Prior to operation above 70 percent of RATED THEF/.AL PCWER after each fuel leading.
l b. At least once per 31 days of accu =ulated cperation in MODE 1, and (-
- c. Within four hcurs if the AIIMUTHAL POWIR TILT (T q ) is > 0.030.
i "See Special Test Exceptien 3.10.2. , I ! CALVERT CLIFF 5 - U:i!T 1 3/4 2-6 Amendment !io. 27, 2A, 32,
??, 48 t- .
ISI
~
PC'4ER DISTRIBUTION LIMITS SURVEILLANCE RE'0UIREMENTS (Continued) - 13
- 4. 2 . 2 .*
shall be d'eterm'ned each time a calculation of Ffyis required xy by using the incere detectors to obtain a power distribution map with all full length CEAs at or above the Long Term Steady State Insertien Limit for the existing Reactor Coolant Pump combination. This determination shall be limitti to core planes between 15% and 855 of full core height inc ive ang aall exclude regions influenced by grid effects. T 4.2.2. shall be determined each time a calculation of F*7 is required " . 9 T
. an - value of Tqused to determine Fxy 'shall be the measured v'alue of Tq . .
sl , t l s . 4 l 4 Anendment No. 27, 32 CALVERT CLIFFS - UNIT 1 3/4 2-7
[ *
. N2.
s . PC'JE? O!57.IEUTIC?! LIMITS : w,..e. -
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APptICAEILITY: MODE 14 When o?cto.\vnS m. a.ccoch.ance.b dk' spec 6scMton 4.g. t.3
,.,C.10 ?! ..
T , , ,.,n With F k
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- b. Be in at least HOT STA!!C3Y. 'R,A uce 4.k e udue d N u2d m spec 6cefeson 4 tt3 to wi h
. .t.he \{mW s a Fsc.u.v e ' M - W o v-SURVEILLUICE RECUIRE"E?iTS i @
~
4.2.2.'W ..e crevisiens of Specifica icn 4.0.4 are nct applicable. s1.%
. fl T : F I4.2.2.1s x'y shall he calculated by the expressien Fxy xy (1+Tq ) and F xy sha.. e deter::ined tc he within its limitt.gy at the folicwing intervay:
.mondcWS 'Fxy u
a r
,a . Prior to cperation a.ove 70 percent or. , .vsir - .D te. .,.v.nl r.ac.x after each fuel 1cadi ,
- b. At least ence per J days / of ace =ulated cperatien in .".CCE 1 I
- and v.
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( . CALVEFsi C:.IFF5 - U!iIT I 4/? --!
, 4 2-7a.
~*
~
o
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POWER DISTRIBUTION LIMITS S URVEILLANCE 0 RE'UIREMENTS (Continued) r3 T is required ( 4.2.2.Y xy shall be d'etermined each time a calculation of F y b : g the incore detectors to obtain a power distribution map with all full length CEAs at or above the Long Term Steady State Insertion Limit for the existing Reactor Coolant Pump combination. This determination shall be limited to core planes between 15% and 855 of full core height inclusive and shall exclude regions influenced by grid effects. tk . T ( 4.2.2. shall be determined each time a calculation of F*Y is required ' . q T
- a. . value of Tqused to determine F xy shall be the measured v'alue of Tq . .
[ .
~* '
l l l - l . 1 . ( w. 3/; 2 Amendment No. 27, 32 CALVERT CLIFFS - UNIT 1 3/42-? b
184
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C 8 . . . POWER DISTRIIUTICi! LIMITS G - TOTAL IriTIGRATED FADIAL PEAKIttG FACTOR - Ff LIMITI?!G C0!!DITIO!! FOR 07E7TI0:1 3.2.3 The calculated value of F , defined as FI = F (1+T ), shall ha r' b .* lied'LCd T HE?tri L ~PCW ER t:.s b-mg l
- 1. leio -
Se combmed:: ton o f T"HERM A L payen an& yy s, % w m w h msh or APPLIC,. .:: ILITY: P.00 : 1*.
.24ajwin A W M kn@
ACTIO!{.* (.E R s + o or w =yona. -t-be Lon 'rei-m
> , ,.,g .g g o sMe Lswub of See hcabot dith FT r
. within 5 hours eithe.r: Wady%
a li.3 al WM m Wu.s o f t=J m B Asssj or
- a. Be in at least HOT STAtt0BY, er n M 3.2 -3 a.
C f. of THEFPAL 70WER - Reduca and THEFFAL F to within POWER
.the 11r1ts of Figure to bring the ccmbina[: and withdraw lengt.. CEAs to er beyond the Long Term Steady state Insertien
- Limits of Specif'.catien 3.1.3.5. The THER'AL PCWER limit detemined frem Figure W shall then be used to establish a revised upper .Th, .-FAL POWER level limit en Ficure 3.2 4 (truncat'a
'v Figure 3.2-4 a the all pabl fractfen of RAT!D THEast pcwgg determined by r,igure c a o subsecuent ccera-icn shall be maintained within th . educed cceptable cperation regien of Figure 3.2-.,. 32-3 A 3 2-3'A
. _ , . SURVEILLA!1CE REOUIREME?TS 0 .2.3 .1 The provisiens of Scecificatien 4.0.4 are not applicable.
4.2.3.2 Ff shall be calculated by the expressicn Ff = FrQ+Tq) and Ff shall be detamined to Be within its limit at the folicwing ir.:ervals:
- a. Prior to cperatien above 70 percent of RATO THEFPAL PCKER
! after each fuel leading,
- b. At least once per 31 days cf accu =ulated cperatien in MOCE 1, and
- c. Within (cur hcurs if the AZIPUTHAL PCWER TILT (T q ) is > 0.C20.
5ee special iese Excestien 3.10.2.
v
'3/4 2-9 Amenc=ent :tc. 27,2.4,32, jALVERTCLIFFS-UttIT1 23,M,a
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PERIPHERA1. AXI AL SHAPC INDCX (Y ) . FIGURE 3.2-4 DNQ Axial Flux Ofhet Control Umits
) '
/
l hend.v.cnt. tio. f I8 CALVERT CLIUS-UtilT l 3/4 2 11
..wo
- )
? .
r
-s .
l
~
POUERb1STRIBUTIONLIMITS T.' .
- I i
Dfi8 PARAfiETERS ' i
. . i i
LIllITItG CONDITION FOR OPERATIOil .
- 3.2.6 The following DNB related parameters shall be maintained within ( ,
the limits shown on Table 3.2-1: i
- a. Cold Leg Temperature
- b. Pressurizer Pressure . ,
- c. Reactor Coolant System Total Flow Rate
- d. AXIAL SHAPE ItiDEX, Core her APPLICABILITY: MODE 1. T ..-
l ACTIO!!: J With any of the above parameters exu2eding its limit, restore the parameter ,
" to within its limit within 2 hours or reduce THERMAL POWER ' to less than 5%
of R4TED THERMAL POUER within the next 4 hours. . SURVEILLANCE REQUIREMENTS 4.2.6.1. Each of the parameters of Table 3.2-1 shall be verified to be - 1 l within their limits at least once per 12 hours. I
'4.2.6.2 The Reactor Coolant System total flow rate shall be determined to be within its limit by measurement a+] east once per 18 months..
[
- l. -
l 1 I 3/4 2-14 Amendment ::o. 21 JALVERT CLIFFS - UtlIT 1
. .. ~
.TA5LE 3.2-1 . .
n . -
. <N -
..- DNB PARAMETERS _' * .
~.
LIMITS' * , n , , r- . Three Reactor Two Reactor Two Reactor Four Reactor
~
OI - ' Coolant Pumps Coolant Pumps
- T
~
sCoolant Pumps Coolant Pumps .
.0pera ting O_pera ting _ Operating-Same Looii Oper: ting-Opposite Lcop_
E-Parameter - Cold Leg Temperature < 548"F Pressurizer Pressure ,_2225
> psla*
Reactor Coolant System . .. ' Total flow Rate > 370,000 gpm .'**. AXIAL SHAPE INDEX . , M"
~
7 ' Limit r.ot applicable during either a THER!ML POWER ramp increase in excess of 5% of RATED TilEPJiAL POWD Q per minute or a TIIERMAL POWER step increast of greater than 10% of RATED TilERMAL POWER.
**These values lef t blank pending NRC approval of ECCS analyses for operation with less than
- four reactor coolant pumps operating. . .
*** The AXIAL SHAPE INDEX, Core Power shall be maintained within the limits established by the .
Better Axial Shape Selection System (BAS'iS) for CEA insertions of the lead bank of <55% uhen BASSS is operable, or within the limits of FIGURE 3.2-4 for CEA insertions speciffeif by FIGURE 3.1-2. g . s
. *e.
. - . w
. 8 4
( (
'~
( . n TABLE 3.3r2
?*
,1 REACTOR PROTECTIVE IllSTRtJMENTATIO!1 RESPONSE T1HES G .
n - FullCTI0flAL llNIT RESP 0ftSE TIME
- 1. Manual Reactor Trip tiot Appitcable 12 0 -
- 2. Power Level - liigh .
< 0.40 seconds *f and < secondsif a 3. Reactor Coolant Flow - Low ., 1 0.50 seconds
- 4. Pressurizer Pressure - liigh 1 0.90 seconds S. Containment Pressure - liigh .- -
1 0.90 seconds
- 6. Steam Generator Pressure - Low < 0.90 seconds
- 7. Steam Generator Water Level - Low < 0.90 seconds gy.0 8 Axial Flux Offset -
< 0.40 sec'onds*# and.-<
- seconds ## . .
.o. . . . . _
9.a. Thermal llargin/ Low Pressure <0.90secends*#and<Bjdse,conds'##'
- b. Steam Generator Pressure Di f ference '- liigh . 1 0.90 seconds
- 10. Loss of Turbine--Ilydraulic y Fluid Pressure - Low ilot Applicable h .
- 11. Wide Range Logaritlnic ticutron Flux Monitor tiot Appilcable
[f.
,+
I - 2 *lleutron detectors are exempt from response time testing. Response time of the neutron' flux signal portion of the channel shall be measured from detector output or input of first electronic component in channel.
# Response time does not include contribution of RTDs. .
i.. F ##itTO response time only. This value is equivalent to the time interval required for the RTDs output
, ,, to achieve 63.2% of its total change when subjected to a step , change in RTD temperature. .
y> 8
. . 191 m
("' v ' TABLE 3.3-5 (Continued) ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSETIMEINShCON05
- 6. Steam Generator Pressure-Low
- a. Main Steam Isolation 1 j2 9
- b. Feedwater Isolation 1 80
- 7. Refueline Water Tank-Low
- a. Containment Sump Recirculation 1 80
- 8. Reactor Trio a, ,Feedwater F. low Reduction to 5% 1 20
- 9. Loss of Pcwer (F . a. 4.16 kv Emergency Bus Undervoltage (Loss of
~
< 2.2 Voltage) see
- b. 4.16 kv Emergency Bus -
< 8.4 Undervoltage (Degraded Voltage)
TABLE NOTATION Diesel generator starting and sequence loading delays included. i ** Diesel generator starting and sequence loading' delays not included. Offsita pcwer available. Response time measured frem the incidence of the undervoltage conditen to the diesel generator start signal. CALVERT CLIFFS - UNIT 1 3/4 3-21 h endment No. 4 0 l l . .
~
y PLANT SYSTEMS - MAIN STEAM LINE ISOLATION VALVES . LIMITING CONDITION rg ,0PERATION 3.7.1.5 Each main st<!am line isolation valve shall be OPERABLE. APPLICABILITY: MODES 1, 2 and 3. ACTION: f MODE 1 - With one main steam line isolation valve inoperable, POWER OPERATION may continue provided the inoperable
. valve is either restored to OPERABLE status or closed
- within 4 hours; otherwise, be in HOT SHUTDOWN within the next 12 hours.
. MODES 2 - With Ene main steam line isolation valve inoperable, sub-and 3 sequent operation in MODES 1, 2 or 3 may proceed provided:
- a. The isolation valve is maintained closed.
[ b. The provisions of Specification 3.0.4 are not applicable. Otherwise, be in HOT SHUTDOWN with the next 12 hcurs. SURVEILLANCE REOUIREMENTS 4.7.1.5 Each main steam line isolation valve shall be demonstrated OPERABLE by verifying full closure within seconds when tested pursuant to Specification 4.0.5. i l b.v CALVERT CLIFFS-UNIT 1 3/4 7-9
- 193 3/4.1 REACTIVITY CITROL SYSTEMS BASES 3/4.1.1 BORATION C0flTROL ,
3/4.1.1.1 and 3/4.1.1.2 SHUT 00WN MARGIN A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made suberitical frem all operating conditiens, 2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and 3) the reacter will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition, li 3fe 4 5 7. SHUTDOWN MARGIll requirements v y throughout core life as a function of fuel depletion, RCS boren concent . tion and RCS T The minimum available SHUTDOWN MARGIN for no load ra tingconditions$l9beginning of life is 61% ak/k and at end of life is 4.3%Jak/k. The SHUTDOWN MARGIN is based on the safety analyses performed for a ' steam line rupture event initiated at no lcadE 3fe conditions. The most restrictive steam line rupture event occurs at EOC conditions. For the steam line rupture a ant at beginning of cycle c . ions, a minimum SHUTCOWN MARGIN of less than a.1% Ak/k is requirad *o a rol the Accordingly, reactivity transient, and end of cycle con tions retuire! 4.3% ak/k. the SHUTCCWN MRGIN requirement is based up n this limitir, c.,nditfonandis consistent with FSAR sifety analysis assumoJons. With T < 200 F, the reactivity transient's resulting frem any pas ulated accid! M are minimal and a 8 3% Ak/k shutdown margin provides adecuata pr action. With the pressuri:er level less than 90 inches, the sources of ncn bora:ed water are restricted to increase the time to criticality during a core dilution event. 3/4.1.1.3 SORON DILUTION. A minimum flew rate of at least 3000 GpM 3revides adecuate mixing, prevents stratificatien and ensures that reactivity changes will be gradual during boren concentration reductions in the Reactor Coolant System. A flow rate of at least 3000 GFM will circulate an equivalent Reactor Ccolant System volume of 9,501 cubic feet in acproximately 24 minutes. The reactivity change rate associated with boren concen-tration reductions will therefore be within the capability of operator reccgnition and control. , 3/4.1.1.4 MCDERATOR TEMPERATUP.E COEFFICIENT (MTC) The limitatiens on MTC are previded to ensure that the assumotiens used in the accident and transient analyses remain valid thrcugh each fuel cycle. The surveillance requirements for measurament of the MTC during each fuel cycle are adequate to ecnfirm the MTC value since this icoefficient changes slcwly due crincipally to the recucticn in RCS beren
*!concentrationasscciatedwithfuelburnup. The ccnfinaticn that the measured MTC value is within its limit prcvides assurances that the coefficient will be maintained within acceptable values threucheut each r fuel cycle.
8 3/4 1-1 Amendment t;c. 27, 37,4r, CALVERT CLIFFS - UNIT 1
MSl \
! PO'<.'ER DISTRIBUTION LIMITS -
t BASES I t he analysis establishing the DN3 Margin LCO, and Themal Margin /Lew Pressure L555 se:;:in s remain valid during' ;crstien at the varicus l allowable CEA group inser:icn limits. If F F' or T exceed their basic limitations, operation may c:ntinue uEEe,r Ehe ad81ticnal restrie-tions imposed by the ACTION statements since these additicnal restrie- ' tions provide adecua:e provisions : assure tha: the assu==:icns used in establishing the Linear Heat Rate. Themal Margin /Lew Pressure and An Local 7:wer Density - Hign LCOs and L555 se::: ints remain valid. AZIMUTHAL FCUER TILT > 0.10 is not expected anc if it sh uld occur, sub-sequent coeration wculd be restric ed to only these cperations required
- to identify the cause of this unexpected tilt.
' T
'he value of T that must he used in the equation F*Y , 7*Y () , 79) -
and rFI=Fr (1+T q ) Ts the measured tilt. T 77 and,.T are The surveillance recuirements fer verifyine that F within their limits provide assurance tha: :ne actual.vEu,es"cf F',) F' VerifyingF.'..,andFiafter and i de not exceed.the assumed values. G uel leading prior to exceeding 75: of RATED THE:fAL FC',iER provides each additional assurance tha: the ::re was pr:perly leaded i 3/4.2.4 PJEL RESIDENCE T!"E . This specificati n deleted. . . 3/4.2.5_ DNS PA J METERS The li=its en the DM3 related part=eters assure that each cf the parameters are caintained within the nomal steacy stata The envele::e limits are Of opera:icn assumed in the transien: and ac:ident analyses. l demcastrated adecuate to maintain a minimum DN3R of 1.195 thr:ughout each
/ 2,3 analy:ed transian:.
The 12 hour periedic surveillance ofthe these parameters parameters are rest:thrredugh instru-ment read:ut is sufficient to ensure tha:within .neir limits f:llowing l The is men:h ;iericcie measure en: cf =e RCS :::al flew Oe ra:e c:eration. is adecuate to detec: 11:w degraca:i:n and ensure c:rrelati:n c7 he indicated ficw indica:icn enannels with easured ficw su:n tha: percent ficw will previce suffician: verifica:icn of ficw rate on a 12 hcur basis. . knend:ent No. 27, 32, CALVERT ctIyps . gnI7 1 3 3/4 2-2 ,, c ,, ves ors *w
, . 195 .
(,, DESIGN FEATURES DESIGN PRESSURE AND TEMPERATURE 5.2.2 The reactor containment building is designed and shall be main-tained for a maximum internal pressure of 50 psig and.a temperature of 276*F. 5.3 REACTOR CORE FUEL ASSEFSLIES
- 5.3.1 The reactor core shall contain 217 fuel assemblies with each fuel assembly containing a maximum of 176 fuel rods clad with Zircaloy-4.
Each fuel rod shall have a nominal active fuel length of 136.7 inches and contain a maximum total weight of 3000 grams uranium. The initial core loading shall have a maximum enrichment of 2.99 weight percent U-235. Reload fuel shall be similar in physical design to the initial core loading and ch:l' h2>c : maximum :nrich.:nt Of 2.7 . !;ht ;;ccc..
, a . <.
CONTROL ELEMENT ASSEMBLIES v . 5.3.2 The reactor core shall contain 77 full length and no part length l control element assemblies.
~ '
_5 . 4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE r 5.4.1 The reactor coolant system is designed and shall be maintained:
- a. In accordance with the code requirements specified in Section 4.2 of the FSAR with allowance for normal degradation pursuant of the applicable Surveillance Requirements, t b. For a pressure of 2500 psia, and
- c. For a temperature of 650*F, except for the pressurizer which l is 700*F.
' ( 5- , , CALVERT CLIFFS UNIT 1 5-4 Amendment No. 32 i
939
.)
10.0 STARTUP TESTING i i l l
. The startup testing program proposed for Cycle - 6 is -identical to - the program proposed for the reference cycle ~ in Reference 1. i l
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