ML20002A278
| ML20002A278 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 11/04/1980 |
| From: | BALTIMORE GAS & ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20002A276 | List: |
| References | |
| NUDOCS 8011050515 | |
| Download: ML20002A278 (27) | |
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TABLE 1 Supplement to Calvert Cliffs I Cycle 5 Technical Specification Changes Change # Tech Spec # Action 1 Figure 2.1-1 page 2-2 Replace Figure 2.1-1 wito enclosed Figure 2.1-1 2 Table 2.2-1 page 2-9 fio change from original request. 3 Table 2.2-1 page 2-10 fio change from original request. 4 Table 2.2-1 page 2-10 fio change from original request. 5 Figure 2.2-1 page 2-11 Replace Figure 2.2-1 with enciesed Figure 2.2-1 6 Figure 2.2-2 gage 2-12 fio change from original request. 7 Figure 2.2-3 pagt 2-13 fio change from original request. 8 B.2.1.1 page B2-1 flo change from original request. 9 B.2.1.1 page B2-1 Change hign power level from 110% to 1121. (flo change from reference cycle.) 10 B.2.1, B.2.2 Change minimum DNBR value from 1.23 to 1.195 pages B2-1, B2-3, B2-5, as indicated on noted pages. B2-6 11 B.2.2.1 Change maximum high power level trip actuation from 110% of rated thermal power to 112%. (flo change fron reference cycle.1 12 B.2.2.1 page B2-5 fio change from original request. 13 B.2.2.1 page B2-7 fio change from original request. 14 3.1.1.1 page 3/4 1-1 flo change from original request. 15 3.1.1.2 page 3/4 1-3 fio change from original request. 16 3.1.1.4 page 3/4 1-5 flo change from original . equest. 17 3.1.2.2 page 3/4 1-9 ik change from original request. 18 3.1.2.4 page 3/4 1-11 flo change from original request. 19 3.1.2.6 page 3/4 1-13 flo change from original request. 20 3.1.2.7 page 3/4 1-14 flo change from orininal request. 21 3.1.2.7 page 3/4 1-14 Ito change from original request. 22 Figure 3.1-1 Page 3/4 1-15 fio change from original request. . 8 01105 0 5lS
TABLE 1 (cont'd) Change # Tech Spec # Action 23 3.1.2.8 page 3/4 1-16 fio change from original request. 24 Figure 3.2.1 page 3/4 2-3 110 change from original request. 25 Figure 3.2-2 page 3/4 2-4 Replace Figure 3.2-2 with enclosed Figure 3.2-2. 26 Figure 4.2-1, - page 3/4 2-5 No change from original request. 27 3.2.2 page 3/4 2-6 Changed calculated value of Fxy from
<l.700 to <l.620 and FxyT >1.700 to Fxy >l.620.
28 Figure 3.2-3 Replace Figure 3.2-3 with enclosed page 3/4 2-8 Figure 3.2-3 29 3.2.3 page 3/4 2-9 Change calculated vs.lue of Fr from
<l.700 to <l.620 and change TrT >1.700 to FrT > 1.620.
30 Figure 3.2-4 Replace Figure 3.2-4 with enclosed Figure 3.2-4. page 3/4 2-11 31 Table 3.3-1, No change from original request. page 3/4 3-2 32 Table 3.3-1, No change from original request. page 3/4 3-4 33 Table 3.3-2 No change from original request. page 3/4 3-6 34 Table 4.3-1 Page 3/4 3-7 No change from original request. 35 Table 3.3-3 No change from original request, page 3/4 3-15 36 Table 3.3-4 No change from original request, page 3/4 3-17 37 Table 3.3-5 No change from original request. page 3/4 3-20 38 3.5.1 page 3/4 5-1 No change from original request except change safety injection tank minimum boron concentration from ninimum of 2300 ppm to between 2300 ppm and 2800 ppm.
TABLE 1 (cont'd) Change # Tech Spec.# Action 39 3.5.4 page 3/4 5-7 !!o change from original request. 40 3.9.1 page 3/4 9-1 fio change from original request. 41 3.10.1 page 3/4 10-1 flo change from original request. 42 B 3/4.1.1.1 and B 3/4.1.1.2, flo change from original request. Page B 3/4 1-1 43 B 3/4.1.2, pages B 3/4 1-2, B 3/4 1-3 fio change from original request. 44 8 3/4.1.2, page B 3/4 1-3 fio change from original request. 45 B 3/4.2.5, page Change minimum DflBR from 1.23 to minimum B 3/4 2-2 Df1BR of 1.195. 46 B 3/4.9.1, page B 3/4 9-1 flo change from original request. 47 3.4.1 page 3/4 4-2 flo change from original request. The following Tech Spec change has been added to the list of changes from the original request: 48 3.1.1.2, page 3/4 1-3, Replace pages 3/4 1-3 and B 3/4 1-1 with and B 3/4 1.1.1, enclosed pages 3/4 1-3 and B 3/4 1-1. B 3/4 1.1.2 page B 3/4 1-1 49 4.5.2, e.3 and e.4 Change minimum volume of TSP from 75 cubic pg. 3/4 5-5 feet to 100 cubic feet and change sample volume to 4.01 0.1 gms in 3.51 1 liters of RWT water. l L
TABLE 2 Explanations for Cycle 5 Tech Spec Changes Supplement Change # Tech Spec # Explanation 1 Figure 2.1-1 Thermal limit lines have been changed to reflect different radial peaking factors and the exclusion of margin recovery programs. 5 Figure 2.2-1 The LHR LSSS has been changed to reflect different radial peaking factors and the exclusion of margin recovery programs. 9 B.2.1.1 Removal of Statistical Combination of Uncer-tai.nties. Pro, gram requires determi.nistic treatment of power uncertainty. 10 B.2.1, B.2.2 The minimum DNBR has been changed .. . . to 1.195 to be consistent with the treatment of fuel rod bowing in Section 6.2 and the exclu- ~
'sion of margin recovery programs.
11 B.2.2.1 Removal of Statistical Combination of Uncer-tainties program requires deterministic treatment of power uncertainty. 25 Figure 3.2-2 The LHR LC0 has been changed to reflect different radial peaking factors and the exclusion of margin recovery programs. 27 3.2.2 Radial peaking factors, both FxyT and FrT, have been revised for Cycle 5 due to the exclusion of margin recovery programs. 28 Figure 3.2-3 Radial peaking factors, both FxyT and FrT , have. been revised for Cycle 5 due to the exclusion of margin recovery programs. f 29 3.2.3 Radial peaking factors, both FxyT and FrT, have been revised for Cycle 5 due to the exclusion of margin recovery programs.
Change # Tech Spec # Explanation 30 Figure 3.2-4 The DNB LC0 has been changed to reflect different radial peaking factors and the exclusion of margin recovery programs. 38 3.5.1 The refueling safety injection tank boron 3/4 5-1 concentration has been changed to reinsert an upper limit of 2800 ppm on the allowable soluble baron. 45 B 3/4.2.5 The minimum DNBR has been changed to 1.195 to be consistent with the treatment of fuel rod bowing in Section 6.2 and the exclusion of margin recovery programs. 48 3.1.1.2 and Additional requirements to the pressurizer B 3/4 1.1.1 level have been included t'o increase the time to criticality during a boron dilution event. 49 4.5.2 The minimum volume of TSP needed to raise e.3 and e.4 the PH of the borated water of the ECCS to 7.0 is 100 cubic feet. In order to test the ability of the TSP to raise the PH of the borated water of the ECCS, the ratio of the volume of TSP to the volume of ECCS borated water must be the same in containment as it is in the laboratory.
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-0.6 -0.4 -0.2 0 0.2 0.4 0. 6 PERIPHERAL AXIAL SHAPE INDEX, Y 1
l Figure 2.2-1 PERIPHERAL AXIAL SHAPE INDEX, Y 1 vs FRACTION OF RATED THERMAL POWER - Calvert Cliffs - Unit 1 2-11
a . 2.1 SAFETY Ll!11TS S C
.W _ BASES u.
I 2.1.1 REACTOR CORE
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<t gl fuel cladding cnd possible claJding perforatica 0 bhich t:The V. 6 restrictio l
the_ release of fission peeriutts to tM roacter cols nt. M r mit._i.n__, - I lthe fuel is prevented by maintaining the staaay state peak lir.ai.e heatOverimnin ( g . (rate below the leval at s.hich centarlina fuel c.altirm will ccar
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OverheaTiiig if the tuct cle.ddir.g is prevented oy rest.ricting 6.el cperitica ._ to within the nucleata boiling regime where ti.a heat transfer coef ficient " is large saturation c,oolant and the cladding surface temperature is slightly above the temperature. s, ., Operation above the upper boundary of the nucleate boiling regime could result in excessive claddina temoeratures because of the onset of , ,' departure from nucleate boiling (6ftS) and the resultant sharp reduction j h in heat transfer coefficient. Of;B is not a directly measurable parameter during operation and therefore THE0JGL POWER and Reactor Coolant Temper-ature and Pressarc have baen related to OfiS through the CE-1 correlation. , m The CE-1 DNB correlation has been developed to predict the OrtB flux and .* g the location of Ofi3 for axially unifom and non-unifom heat flux distri-butions. The local DHE heat flux ratio, DfiBR, defined as tne ratio of l w/ the heat flux that would cause Of;S at a particular core location to the local heat flux, is indicative of the margin to Dfi3. I s The minimum value of the DNSR during stead < state operation, normal I*l95 .. operatio'ial transients, and anticipated transie'nts is limited to W l This value :orresponds to a 95 percent probability at a 95 percent con-fidence level that Of;B will not occur and is chosen as an appropriate (*
- margin to OfiB for all operating conditions.
The' curves of Figures 2.1-1, 2.1-2, 2.1-3 and 2.1-4 show the loci of points of THERIML POWER, Reactor Coolant Systen pressure and ' maxinum cold leg temperature of various pumo combinations for which the - l minimum DNBR is no less than 1-47 for the f amily of axial shapes and corresponding radial peaks shown in Figure B2.1-1. The limits in Figures M} 2.1-1, 2.1-2, 2.1-3 and 2.1-4 were calculated for reactor coolant inlet temperatures less than or equal to SSO:F. The dashed line at SE0'F coolant inlet temperature is not a safety limit- however, operation above 580'F is not possible because of-the actuation of the main steam line . t I safety valves which limit the maximum value of reactor inlet temperature. , Reactor operation at THER*1AL POWER levels higher than(U290f RATED THERMAL g POWER is prohibited by the high power level-trip setpo 'nt specified in
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"', I g fu II CALVERT CLIFF 5 - UNIT I B 2-1 E Amendment No. 31 i j.
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[s M SAFETY LIMITS BASES
, Table 2.1 l. The arca of safe operation is below and to the lef t of 2 these lines.
The conditions for 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 she.:n on the figures. The reactor protective system in ccmbiratien with the Limiting i g Conditicns for Operation, is designed to prevent any anticipated ccnbina-tion of transient conditiens for reactor ccolant system te. perature, pressure, and THER!'AL F0WER level that would result in a CriBR of less than g and preclude the existence of ficw instabilities. 12 .I9s- I ] 2.1.2 REACTOR C00LAtlT SYSTEM PRESSURE 4 b- The restricticn of this Safety Limit protects the integrity of the Reacter Cociant System frcm overpressurization and thereby prevents the
' V release of rad b uclides contained in the reactor coolant from reaching the containment. atmosphere.
The reactor pressure vessel and pressuri:cr are designed to Section III, 1967 Editicn, of the ASMF Code for !;uclear Power Plant Ccmponents which permits a maximum transient pressure of 1105 (2750 psia) of design
, . pressure. The Reactor Ccolant System piping, valves and fi.ttings, are designed to AtiSI B 31.7, Class I,1959 Edition, which permits a maxict.m transieit pressure of 110" (2750 psia) of ccmponent design pressure.
The Safety Limit of 2750 psia is therefore consistent with the design criteria and associated code requirements. . The entire Reactor Coolant System is hydrotested at 3125 ps_ia to demonstrate integrity prior to initial operation. i i b 3 CALVERT CLIFFS - UtilT 1 B 2-3 Amendment lio. 33, 3 9
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- u 2.2 LIMITitiG SAFETY SYSTEM SETTitiGS ASES 2.2.1 REACTOR TRIP SETP0!flTS The Reactor Trip Setpoints specified in Table 2.2-1 are the values at which the Reactor Trips are set for each parameter. The Trip Setpoints have been selected to ensure that the reactor core and reactor ccolant system are prevented frca exceeding their safety limits. Operaticn with
- a trip set less conservative than its Trip Setpoint but within its secci-fied Alicuable Value is acceptable on the tiasis that the difference a between the trip setpoint and the Alicuable Value is equal to or less than the drif t allowance assumed for each trip in the safety analyses.
Manual Reactor Trio
' The Manual Reactor Trip is a redundant channel to the autcmatic protective instrumentaticn channels and prcvides manual reactor trip capability.
V Power Level-High The Power Level-High trip provides reactor core protection against reactivity excursiens which are tco rcpid to be protected by a Pressurizer Pressure-Hign or Thermal Margin / Low Pressure trip. The Power Level-High trip setpoint is operator adjustable and can be set no higher than 105 above the indicated THEFy.AL PO'nER level . Operator action is required to increase the trip setpoint as THEP"AL TOWER is increased. The trip setpoint is autcmatically decreased as THEMtAL pcwer l decreases. The trip setpoint has a maximum value of 107.0% of EATED THEP3.AL F0WER and a minimum setpoint of 30", of P.ATED THEM!AL PO'iER. Adding to this maximum value the possible variation in trip point due to calibra tion and ins tru .ent errors , the maximum ac tual steady.-s tate THEPJtAL F0WER level at which a trip would be actuated is {2].(of FATED THEF#.AL F0WER, which is the value used in the safety analyses. g%.
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Reactor Coolant Flow-Low The Reactor Coolant Flcw-Low trip provides core protection to prevent Di(B in the event of a sudden significant decrease in reactor coolant flow. Provisions have been made in the reacter protective system to permit j' CALVEPT CLIFFS - UtilT 1 B 2-4 l.mendmen t flo. 3 '3 'l l
l
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w c .m LIMITitiG SAFETY SYSTEli SETTINGS , BASES l 4 operation of the reactor at reduced power if one or two reactor coolant pumps are taken out of service. The icw-flow trip setpoints and Allowable ~ Values for the various reactor ccolant pur:p combinations have been derived in consideration of instrument errors ard response tings _nf-pBf'/,jT equipment involved to maintain the D':5R above L44 n er ncrmal operation I and expected transients. For reactor cperaticn with cnly two 'r three reactor coolant pumos operating, the Peactor Coolant Flow-Lew trip set-g points, the Power level-High trip setpoints, and the Thermal !1argin/ Low
' Pressure trip setpoints are autcmatically changed when the pump condition selector switch is manually set to the desired two- or three-pump position. Changing these trip setpoints during two and three puto K $ I' M -_
operation prevents the minimum value of D:!3R from going below%M curing l , normal operational transients and anticipated transients when only two or three reactor coolant pumps are cperating. Pressurizer Pressure-Hioh Q V The Pressurizer Pressure-High trip, backed up by the pressurizer code safety valves and main steam line safety valves, provides reactor coolant
- system protection against overpressurizatica in the event of loss of load without reactor trip. This trip's setpoint is 100 psi below the ncminal lift setting (2500 psia) of the pressurizer code safety valves and its concurrent operation with the power-cperated relief valves avoids the undesirable operation of the pressuri er code safety valves.
1 Containment Pressure-Hich The Containment Pressure-High trip provides assurance that a reactor trip is initiated concurrently with a safety injection. The setpoint
, for this trip is identical to the safety injection setpoint.
Steam Generator Pressure-Lew The Steam Generator Pressure-Lew trip provides protection against an excessive rate of heat extraction frca the steam generators and p10 subsequent cooldown of the reactor coolant. Thesettingof@5VpI is sufficiently below the full-load operating point of 650 psia so as not to interfere with nomal cperation, but still high enough to O prcvide the required protection in the event of excessively high steam O ficw. This setting was used with an uncertainty factor of + 22 psi in the~ accident analyses.
. cal. VERT CLIFFS - UlllT 1 B 2-5 Amendment flo. 33
0 .- oo e .- .m d,a LIMITillG SAFETY SYSTE!1 SETTIfiGS , i . BASES Steam Generator Water Level
~,
The Steam Generator Water Level-Low trip pcevides core protection by preventing operation with the steam generator water level below the minimun volume required for adequate 1. cat rem val capacity and assures that the pressure of the reactor ccolant system will not exceed its Safety Limit. The specified setpoint provides allowance that trere will be sufficient water inventory in the steam generators at the time of
\ trip to provide a margin of more than 13 minutes before auxiliary feedwater is required.
Axial Flux Of fset [* The axial flux offset trip is provided to cn ure that excessive axial peaking will not cause fuel danage. TheAxial flux of fset is determined frca the axially split excore de dctors. The trip setpoints p ensure that neither a OfiBR of less than Q3 nor a peak linear heat rate V which corresponds to the temperature for fuel centerline melting will exist as a consequence of axial power maldistributions. These trip set-points were derived f rem an analysis of many axial power shapes with allcwances for instrumentation inaccuracies and the uncertainty asscciated with the excore to incere axial flux of fset relationship.
~
. Therrial Marcin/ Low Pressure The Thermal Margin / Low pressure trip is provided to prevent operatien when the CitBR is less than({ fA % pg The trip is :nitiated whenever the reactor coolant system pressure signal drops below either 1750 psia or a computed value as described below, whichever is higher. The temputed value is a function of the higher of aT power or neutron power, reactor inlet temperature, and the number of reactor coolant pumps operating. The minimum value of reactor
' coolant flow rate, the maximum AZIMUTHAL P0'JER TILT and the maximum CEA deviation permitted for ccntinuous operation'are assumed in the genera-tion of this trip function. In addition, CEA group secuencing in accor-dance with Specifications 3.1.3.5 and 3.1.3.6 is assumed. Finally, the maximum insertion of CEA banks which can occur during any anticipated operational occurrence prior to a Power Level-High trip is assumed.
O . V-
~
B 2-6 Amendment flo. 33, 3 9
. CALVERT CLIFFS - Uf(IT 1
1.0 __, + , ..- , ,= = = . r
== = .7 . ~ -- + -3
" UNACCEPTA BLE:: =xr7 = = ; 5 5 f 5 ; 5 5 : 5 UNACCEPTA BLE =i-i F OPERATION EEEEEIEtEE E"~ ele OPERATION --
si REGION !EEtEiElgiEt=iEE EEg=!EE REGION =:
+l t. =:irr =:==;==:=
- =~ = ' = ~ '= *==* 2= =="= =:=={==n==:=j==":==}"==~;=~~'t=:=
='="= = "t'T =
0*9 -
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=. :=t= C _ .4 -- J4 - ! -
c f ._= _. _:_ __=:= .cin._
.: =_ ._::===l==_1=r
= :=rL= * = f ==l = t: =!r= = U $- :' ~2 .~ 1~ l Z-.
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. - --4 I
< 0. 8 i=iassaiM'Efsk/lhissZlE
- = ,==; = =;=
- EflEZIE ~ =4 - !
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s w : .:t==2 ==== ==/:=q==u=;==== =y:= q:
= e-
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; = tf /. . _ _=-.-.
I v= t_ f.=__ets e s = Et
- w=_::s... . ._ t __i: s._s_.. f._r=r=a..__ 1 ; m a ts a._-!, = := =i=
l-- o. .- - -I
- - - - - - . - o o -l: a= pas a:ftsisjasEEa,n smE i ! 0.7 2
z.:: .z c:1.r:r._:n p:i si: 1 art:==::: ~ l z.:11::1J
$:sud: EEEE/: EsrisisiEEsiEE. =E , ,i.Z: =k=1 - ~~-l i- : 2. b =: = = =1 ==: =; == = j=-"==. . _ . . . . . = a= 2 ,
w 1- ._ i t :-: x:- rf::= :=:= t:rn =: =f rr t.1: .}1 a . 2 .: 7. :. . : o 1- 1 z - : n(-0.3, . 63) 7"~~~ ^ ~ -1 -1 h: (0.3 .63) !E = o . _ 3 = =v=_ _!Es"=is saj~a s[== _u.=,I= 4 u b6 -0 " - - - - - - "-- " " ~ ~ - ~ ~ ' gl; _ja=} sis sesaissteig5glE=ajgh =i.;=EjaEgjg;: < = ei =.=r=:= ===== ,== = : = i = =: = i = a .. .=u=i= = % : t =:=- - ='= = := =t:=<=n= t =rt = t =:= :l_r_ x t r _ = ;= 1 :r
; - n ic tr 7:= :r-- ~ h== -~2:=1=!rnj=;=in.I' r r r{ r !-~11. r :!r '-
Tr ~ - J'.-- : 11 .- c~ : { '- r;-~
~*
't 1
**._ f5 = : == =:n - i - - ~ - t ~ ~ ' -- +r . : = L--
rL L
. t _:
0.5 -
=_r+
mis 155 ?= ACCEPTABLE 55
- +2 seas OPERATION
. . _ . . ~
T *- EE@tEEE
~E~ =l * +=
1M
='= =
ml=:= _ ===. . =u: _ REGION =* = 't - = = = - ."=":==.F~ =-
==
. = :=ifil=EE.
____4.__ . E!E-EiEEME EEElfiECis =MEigsstiiE 0.4 e ~=]'*1='==^=~=~=*==*= = = ~ === == * ==_=.
-0.6 -0.4 -0.2 0 0.2 0.4 0. 6 PERIPHERAL AXIAL SHAPE INDEX, Y I
Figure 3.2-2 LINEAR HEAT RATE , AXIAL FLUX OFFSET CONTROL LIMITS ! Calvert Cliffs - Unit 1 3/4 2-4
I C v , POWER DI3TRIBUT10ft LIMITS - TOTAL PLAf!AR RADIAL PEAK!!!G FACTOR - Ffy LIMITit:G C0':DITIO!! FCR OPERATIC 1 T T 3.2.2 The calco sted value of F *Y, defined as F *Y = F*Y(1+T 9 ), shall be limited to 1 .660 M l APPLICABILITY: MCDE 1*. O
\ ACT10:1: l,go i .6 r, within 6 hcurs either:
WithFfy> l l a. Reduty THER'tAL PCUER to bring the conbinaticn of TiiERMAL POWER and F to within the limits of Figure 3.2-3 and withdraw the full Mngth CEAs to cr ceyend the Long Tena Steady State
; losertion Limits of Specificaticn 3.1.3.6; or
- b. Be in at least HOT STA!!DBY.
J SURVEILLA!:CE REOUIREMEllTS 4.2.2.1 The provisions of Specification 4.0.4 are not applic ble. Ffy shall be calculated by the expression F 4.2.2.2 =F xy (1+Tq) and F shall be detennined to be within its limit at the followir.g intervals:
- a. Prior to operation above 70 parcent of RATED THERMAL POWER after each fuel loading,
- b. At least once per 31 days of accur.ulated operation in MODE 1, and I
- c. Within four hours if the AZIMUTHAL POWER TILT 9(T ) is > 0.030.
1
*See Special Test Exception 3.10.2.
D . v. v CALVERT CLIFFS - UtilT 1 3/4 2-6 T+.c ndne n t tic . 71, 7 4 i
F ' l - 1 "iIllIIIIlllllllllllll ' j 1.00 T jl.621.00) 7, UNACCEPTA BLE _ _ _ OPERATION g m i
%,J qt;g s i
i ll REGION '1 g q 'N : 3_ .4 .t _ . 7 _ . 9Hmumnummu .- + 5 a. 2 g 0.90 bk .
- >,g Fhy LIMIT CURVE r ,
(
-1
@ 3 'Ff LIMIT CURVE ; I ~
%q y "
g
~
i
'ii n hd, ' '
F-(1.70,.85) c- o ' j' [ T
- a. W 0.80 = .
~
T t
' ~ ~
~
< I l %[11ih l
j I .... _ _ _ _ l (1.695, .775)j _ ACCEPTABLE l' W a $ p 0.70 - OPERATION REGION o !
< .l..
l 7 g - _ _ a _ _ 1L . _ _ q .... _ m u3 p .(( ; l 0.60 .~.
-1 - --
.Il +- -
p- . - j )-- - I 9 !I i i i, i i !
/c I l ! li l O! l . I -llii ll! l l ! ; !! .I g
j r gt )L1
-1 y ,
7 y - 7 7-0.50 1 .I I il l l I 2 L .l .l _ 1 1. t lI.lll .I j l 1 .._ _ a 1.56 1.58 1.60 1.62 1.64 1.66 1.68 1.70 1.72 T T FXY ' p r Figure 3.2-3 TOTAL RADIAL PEAKING FACTORS vs ALLOWABif FRACTION OF RATED THERMAL POWER
1 I I D""}D o o Ju
]D h }S oJuuSu. .m
- p .
P0'.!ER O!STRIPitT!0:1 LIMITS t TOTAL IftiEG' RATED RADIAL PEAKII;G FACTOR - F - - r i L1M1TIf!G CONDIT10tl FOR OPER ATIGN - T T 3.2.3 The cair' ,ed value of F r, defined as F r = Fr (1+Tq ), shall be lipited to <~ .571 _l APPLICABILITY: t'0 0E 1* .
.60 ACTION:
O j
\ With FT> .57 within 6 hcurs either:
r g a
- a. Be in at least HOT STAtiDBY, or
- b. Reduc and F[p THER:'AL to within F0WER the limits of Ficure 3.2-3 to bring the ccnbination and '.:ithdraw lengt i CEAs to or t eyend the Ler.g'Tcrr. Steady State insertic1 the full o
{ Limi ts of Spec i fication 3.1.3.6. Tt.e THEFl%L FCUER limit O determined frca Figure 3.2-3 shall then be used to establish a
~
revised upper THEF"al FCWER le eci lic:it en Figurc 3.2-4 (truncate s Figure 3.2-4 at the alicuable fractica of RATED THEF 9L FC'..ER determir.ed by figure 2.2-3) and subsecuent operation shall te
, maintained within the reduced acceptable operction region of Figure 3.2-4.
l SURVEILL A'; E RECUIRE?'E!;TS I 4 i 1 1.2.3.1 The provisions of Specification 4.0.4 are not applicable. T ] 1.2.3.2 F r shall be calculated by the expression FTr = F rW T q) and FTr shal1 ~ be detemined to be within its limit at the following intervals :
- a. Prior to cperetion above 70 percent of RATED THERi%L PCWER af ter each fuel loading,
- b. At leas t once per 31 days of accumulated operatien in M00E '1, and -
- c. Within four hours if the AZIMUTHAL POWER TILT (T ) is > O'.030, q
p -
'$ce special Test Exception 3.10.2.
+-
.O jALVERT CLIFFS - UNIT 1 3/4 2 9 Amendment No . ;'1, ?.t . ,U 33 .
z O l-= I m 1.1 . m = , z i: u=t == =r + - ===.= =2= :=i.=:= v o = = .
@ . j UNACCEPTA BLE MM$[5fC llMj=lC UNACCEPTA BLE"=
o i OPERATION I==== == = h== = === ! =~ OP ERATION c_ u :H REGION fa=HEEEEsMEMslEE4!?M REGION j 3 - __ _ .. .. ._ 4 _ _ . - - -r - . - "e 1.07 _'=_._.;-k. ...-. m _ _ . _ ,r. M_ (-10, uU(.15,1.0)9_..M..~*.,
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ut. L H t.
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x W
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- .!! 1. US 5.55[5 ff:}E 55531Idi3:_h!5-351.-Ii: NI.'
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' ' --~ ~ ~ ~ ~ - - - - ~ ' - ' ~~ -
g -'. T. .l aZ ====. t = ==/= ==1== = = ._= j rrr-Ifl! / ..=+C=22-*
= =; _=1= ~ ~ "'Z: ==
=; v = r_
17._'p{.Z2 ' .i. t:11' : ~1} T_ y l1
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,_.....p.,___ _ . _ _ .
g._ ._ _ , _ - . . ....._ s , t.7_
- o. :( .30, .75) CTJ 1 5!5525 -- 5=P=5: ==
w m. Isi== ===ts = (.30, .75) ZE __2 4_ =i= =. _s =. ..=.=. =4= t==.=_y:==_. r= - < 0.7 i. : 0 !=E sas m _ .._ 2 ._ s s=E= .! asis ~5=uE_ _ _ _ - . _ . . =:t ._ is_a.. . .ssist.= ws. = $ l' ' 3:iF 'i=E5tE'555 .T:E~
% ; : h_ _~ ~ O : -
o Y . 5 5"55 : :
; u .-t=z
! . ::11 : z .=:x t u =:=
- ==: == =trx = =[~=;r; > ==== =.nz' = : =:t :n -_. __ ri.r . ~m1.a _. ~t.- #
< L += =.. : :.
- s. _u._s :=. a_ __E__s_._m.._sj+.s_m._e_ .s__s__ sic..1_+
_r=_=
.. = r_. _ a. m..t. _ .
=-
$ h'
..r:= =:n r n:: nr-..:= .c.- ,
=c=I_.==
Ei- = ."z ACCEPTA 2 ;nnt=1= BLE =i5 5'd
~
D0 h' .. ; .555 sl= 55Si!5 ft 36 F. 4 i= x ss se OPERATION nElE== EH=4EE!E"~ H. . : . - g. =_i._E_ E_.i5 REGION E. ._i._E_ l_ _~E._~ E_Ei_ET_ ...'.E_E.. EIE. _E. E.iE_ =. =..> g
. u a= cnn. =:= r -+ ==:nntu--- :_.._t_------ un:= r---~ x * = u . _
t- .. - .
=:t-rtz: = 1.- - + - - -
= :=t=1--=-+ ---t:--tr ==tr:=fx2 :: -'
u- ir z= =i=t= == - - -
==== =i=;=.
0 '5 r 1= = i === =;= i=r- -._ -w== . == _ =t=,= r -
. - i 8 -0.6 -0.4 -0.2 0 0. 2 0.4 0.6 C PERIPHERAL AXIAL SHAPEINDEX Y I a
x w Figure 3.2-4 DNB AXIAL FLUX OFFSET CONTROL LIMITS Calvert Cliffs - Unit 1 3/4 2-11
ID j v 3/4.5 EIERGEllCY CORE C00LillC SYSTE!!5 (ECCSJ
%/
SAFETY lilJECT10ft TA!!KS l ^ LIMITitlG C0';DIT10:1 FOR OPERAT10!! 3.5.1 Each reactor coolant system safety injection tank shall be . OPERABLE with:
- a. The isolation valve open,
- b. A contained borated water volume of between 1113 and 1179 cubic feet of borated water (eauivalent to tank levels of between 187 and 199 inches, respectively),
#,7; , f %wt* %^ p '3" -
- c. A(oron concentration of W::"---172C c.nd :20; w..., ad-- l btNesw 2300 AwA 2too p
- d. A nitrogen cover-pressure of between 200 and 250 psig.p,and, APPLICABillTY: MODES 1, 2 and 3.*
( ACT10ti: O"
- a. With one safety injection tank inoperable, except as a result :
V of a closed isolation valve, restore the inoperable tank to , OPERABLE status within one hour or be in HOT SHUTDOWil within the next 12 hours.
- b. With one safety injection tank inoperable due to the isolation valve being closed, either in:nediately open the isolation valve or be in HOT STA::DBY within one bour and be in HOT
. SHUTDOWil within the next 12 hours.
1 SURVEILLA:tCE RE0VIRE!4Ef:TS 4.5.1 Each safety injection tank shall be demonstrated OPERACLE:
- a. At least once per 12 hours by:
- 1. Verif.,ing the contained borated water volume and nitrogen cover-pressure in the tanks, and
- 2. Verifying that each safety injection tank isolation' valve
,(- is open. . r *With pressurizer pressure > 1750 psia.
i CALVERT' Cliffs - UtliT 1 3/4 5-1 Amendment lio.? ' r
. _j .
i
%m/
POWER DISTRIBUTI0il LIMI_TS BASES the analysis establishing the Df;B Margin LCO, and Thermal f targin/Lc.-: Pressure LSSS setpoints remain valid during7 c peration at tha various . allowable CEA grcup inserticn limits. If F .,, F'. or i exceed their basic limitations, operaticn may c:ntinue uf. der tne ad3itional restric-tions imposed by the ACTICS state. rents since these additional restric-tions provide adequate previsicn3 to assure that the assumptions used in establishing the Linear rica Rate, ir.cr al 'targir./Lc.; Pressure and Local Po.4er Density - High LCOs anc LSSS setpcints remain valid. An AZIMUTHAL FO'.iER TILT > 0.10 is not expected and if it sneuld occur, sub-sequent operation would be restricted to cnly those cperatiens requircJ to identify the cause of this unexpected tilt. T that must be used in the equation F *7 =F *# (1 + T9) ' andF}hevalueofTr r
=F q (leT ) is the measured T T tilt.
The surveillance recuiremen.ts for verifying that F ,, Fr and7 T are within their limits provide assurance tnat Verifying the actualF 7vuues F., gr-F af .,, ehr and T do not exceed the assumed values. y each ?uel leading prior to exceeding 75', of RATEC . L PO'n'ER THEd,
- p. ovides, and additienal assurance that the core was prcpcrly loaded.
3/4.2.4 FUEL RESIDE. ICE TIME , This specification celeted. 3/4.2.5 D!iB FAPAMETERS The limits en the 0l;S related parameters assure that cach of the parameters are maintained witnin the ncmal steady state cnvelope of operation assumed in tne transient and accident analyses. The limits arc consistent with the safety analyses assu.T ptions and hwe been analytically demonstrated adequate to maintain a minimu.:1 C.';3R of .19 .hrca-hcut eacn analyzed transient. . jg I The 12 hour periodic surveillance of these parameters through instru-ment readout is suf ficient to ensure that the parameters .are restcrid within their limits following lead changes and other excected transient operation. The 18 mentn pcriodic ce2surement cf the RCS total fica rate is adequate to detect floo degradation and ensure correlation of the flow indication channels with measured ficw sucn that the indicated percent flow will provide suf ficient verification of ficw rate on a 12 hour basis. Amendmen t tio . 7 7, J/ . 3 3. g .... CALVERT' CLIFFS - UNIT 1 B 3/4 2-2
-g -
0 , , . .
^ 1:f ACT IVilY CO'ilR01 SYSTf!15 V S1101D0'.!!! ItARGiff - T;ivo -e 200"F d .
D a 3,Q : Ll!11TI:tG C0:10lT10l1 IOR OPEP.A'110:;
/ ---
3.1.1.2 The S!!UT00U:1 ItAP.Glit shall be . 0. ok/k. cs . Pr ... ..mrie, ie.,.ci 2 cie ;n,ms Gc. wite.- A w p s.4%.er APPLIC ACII. l TY : !;0CE 5. - b Ee/t *a.r N <l ( cio in<.n .. 3 r: .,, ornn,,, ef g pre,31,,. ,y,. ACTIO,, nrA cll scuws e f nc.s bonaed va -( " ef ,,3 a i. 2300 a.With the SUUTCC'.:'t l'ARCl!! 1 . . I., irmediately initiate and ccntinue ] = boration at > 40 9;n of
~
, r;.m t oric acid solution or equivalent until the required SWJ1DO.!!! Ik61:i is restored.
b b.n & pefutrser drnincATo 6 90 .m.e., and rail Mirw cf re..- tcatteA sdc.- > '37 3 ,%_ 6minerlordely Sysed (dl #f(.Ltici.n InNelvit pc *ativ e_ r(qctivd y Che,yp u.n,\*. The. O ng(,.cW) 1%kbiu is inuivist.A To c.crapc;n:.<tc Gr we. neic(cim ,id ,s en,c, e,y n:,g. be rta ed v.r.ier c e redw< % d<w. c.f ni..s.bc.dcA vtter To 6 % c pra SUR'.'E ILI./yrE PEQUl.JjQ'igT S d 4.1.1.2 The SHUT 00'.::1 itARGl:1 shall be determined to be 1 1.F 23./ L J
- a. Within ene bour af ter detection of an ino; erable CEA(s) and at least onse per 12 hours thereaf ter t.hile the CEA(s) is increrr;le.
If the inoperable CD is inrrvable er untrippable, t.he abe;c required SHUTCO'..:t :'ARGli; shall be ir. creased by an a . cunt at least equal to the withdra.,n s.octi. of the ir.r.ovable or untri;:paDli t CEA(s). , 1 b. At least once per 24 hours by consideration of the follo iirg; factors:
- 1. Reactor coolant system baron concentration, l
- 2. CEA positien, 3.. Reacter ccolant systen average temperature,
- 4. fuel burnup based en gross thernal energy generation,
- 5. Xenon ccnce.it.rc:L ion , and
- 6. Sai.urium concentration.
]# 4. p. 2. 2.
Lhh W. prouviw donned, to 4 'tc iunes Mteremne :
<. k dhh ec.w hcw cd twq it 'nou s 1%c.,5tew %r % '!gsu *,n tw.,
3,4 r%ch.< cccsa,a system is 9bc.it % btTtc.w ef iht het kg neub mM O b Mhm c... hcu <tnet tv, y 12 hws Thundkc ter.; W .earr.% d ne.)-be.n.tcA wxx m i 75 gpro W thc ahd r\ct.o iw .p h ,s cca p.. n s. t.A E.- A toirl tic .s d b. sd . i CAL.VI.HT CLiiTS - ll'ill 1 3/4 1-3 .
'~ -~ ~ - .... ..... .... ,.... .. .... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , . . . . . . . . . .
g 4 , - e- - - - - -
J d
~
(O 3/4.1 RfACTIVITY CutlTROL SYS1Etts 4p j BASES 3/4.1.1 BORAT100 CD:iTROL i 3/4.1 1 1 and 3/4.1.1.2 SHUTDOWN tlARGit; _. A sufficient SHUTDOWN l'ARGIN ensurce that 1) the reactor can be
! ' subcritical from all c;:erating conditions, 2) the reactivity tecnsients i .
associated with postulated cccident conciticns are controllab1c within j cccept:ble limits, and 3) the reactor will be maint:ined sufficiently subcritical to preclude inadvertent critic:litv.in the shutdc.m cor.dition. i \
! [ J2, vi
-5MTMW:L.40.7,42L fuci c.cic'.ia , ". S l'. recu i :~;c t; v&Ae@et
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- r. a
, _ . s t.1 '. t ; Mi+-Ste.-i.L 3 cc.14e,+t-and_x e ni tU~ w r; *?'.*" =M - ? %~
- W c ; . a i ,, ., a ginh;..y ewr v y p ^ ate m (i s f -t M s--t e e M +re.-r tha re?:'4'4*, U:r- * +, 4 j c, ; n 3 ., g 3 , ,gpjpgg g g ,y g ]
Accordingly, the SHUTOOUM !!AT. GIN rcauire .ent is based upon analysis this limiting concition and is censistent with FSAR safety assumptions. With T < 200*F, the reactivity tecnsients result-s (f ing frem any postulated accidS,E Ere minimal and c apshutdown k/k ruargin provides adequate protection. O.ihe- 3 '7. i 3/4.1.1.3 BOR0t OlLUTICN D'"4 e re,wr e l<,1 le,.thr.n 'ie in< hc N., nee bo n ut.w y r.;w,tt a w, 3 yg* ' A ninimum ficu ra te of a t leas t 3039 0;:g.t,,u. m n.3,,,9,7 4, .,,, , ._ m y , , n prdvides cdeque.te mixing, prevents stratificatien and cnsures that reactivity changes 1.ill be '
- grcdual Syste. durinc bcrcn cor. centration re1,c ticns in the Peacter Ceolant u A flin< ra te of a t ,
Reictor C'olant Syster vciure of 9,601 cubic fec:1 cast 3000 Gin will ir, cpproximatcly Y@-py circulat 24 minates. The reactivity change rate !stcciated with borcn ccncen-tratien reductions recognition will therciore t,c within the cai: ability of operator and control. 3/4.1.1.4 MODEP.ATOR TEMPERATURE COEFTICIENT (MTC) The limitatiens on MTC are provided to ensure that the assu: ptions used in the accident and transient analyses ren:3in valid thrcugh each fuel cycle. l The surveillance requirewnts for r.:cisurement of the itTC during each fuel cycle are adequate to confirm the tiiC value since this coefficicnt changes slowly due principally to the reduction in RCS boren concentration anscciated with fuel burnup. The confirnation that the ! measuicd itTC value is within its limit prevides asturances that the coef ficient will be maintained within acceptat,lc values thruughout each fuel cycle.
. r..
CALVERT CLIFFS - UtilT 1 B 3/4 1-1 Amendment tio, n , 32
~. __ _ _
I (I!lh EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
- e. At least once per 18 months by: -
- 1. Verifying automatic isolation and interlock action of the shutdown cooling system from the Reactor Coolant System when the Reactor Coolant System pressure is'above 3.
psia.
- 2. A visual inspection of the containment sump and verifying that the subsystem suction inlets are not restricted by debris and that the sump components (trash rac.ks, screens, etc.) show no evidence of structural distress or corrosiun.
l00
- 3. Verifying that a minimum total oWcubic feet of solid granular trisodium phosphate dodecahydrate (ISP) is contained within the tsp storace baskets.
3.G t o.I N<lers 4.0 t o.i go,
~7
- 4. Verifying that whe representative sample of-O R 4 lbs-of TSP from a/ TSP storage basket is submerged, without agitation, in -80 ;+ gallons of 77 + 10'F borated water from the RWT, the pH of the mixed solution is raised to
> 6 within 4 hours.
- f. At least once per 18 months, during shutdown, by:
1 Verifying that each automatic valve in the flow path actuates to its correct position on a Safety Injection Actuation test signal.
- 2. Verifying that each of the following pumps start auto-matically upon receipt of a Safety Injection Actuation Test Signal:
- a. High-Pressure Safety Injection pump,
- b. Low-Pressure Safety Injection pump.
- g. By verifying the correct position of each electrical position i stop for the following Emergency Core Cooling System throttle valves:
- 1. During each performance of valve cycling required by Specification 4.0.5 by observation of valve position
-M k::: on the control boards.
CALVERT CLIFFS - UNIT 2 3/4 5-5 Amendment No. 26
I i
- 10. STARTUP TESTIflG l
fio change from original request. O
- m. _ a _ _
l l l l e l i 1 4 i SECTION
11.0 REFERENCES
4
REFERENCES for Sec, tion 6
- 1. CENPD-161-P, " TORC code, A Computer Code for Determining the Thermal .
Margin of a Reactor Core", July 1975.
- 2. CENPD-162-P-A (Proprietary) and CENPD-162-A (Nonproprietary), " Critical Heat Flux Correlation for C-E Fuel Assemblies with Standard Spacer Grids Part 1 Uniform Axial Power Distribution", April 1975.
- 3. CENPD-206-P, " TORC Code, Verification and Simplified Modelin9 Methods",
Janua ry , 1977.
- 4. CEN-83-(B)-P, "Calvert Cliffs Unit 1 Reactor Operation with Modified CEA Guide Tubes", dated February 8,1978 and letter, A. E. Lundvall, Jr.
to V. Stello, Jr. , " Reactor Operation with Modified CEA Guide Tubes", dated February 17, 1977.
- 5. Supplement 3-P (Proprietary) to CENPD 225P, " Fuel and Poison Rod Bowing",
June, 1979.
- 6. Letter from D. B. Vassallo (NRC) to A. E. Scherer (C-E), dated June 12, 1978 k
i Ie4GM - :==-
REFERENCES FOR SECTION 7 (NON-LOCA TRANSIENT ANALYSES)
- 1. CENPD-199-P, "C-E Setpoint Methodology", April,1976.
- 2. Letter from A. E. Lundall to R. W. Reid, Dated February 23, 1979.
- 3. Letter from A. E. Lundall Jr. to B. C. Rusche, "Second Cycle License Application," October 1, 1976.
- 4. Letter from J. W. Gore Jr. to E. G. Case, " Third Cycle Licensee Application", December 1,1977, As modified by Letter from A. E. Lundall to R. W. Reid, " Request for Amendment to Operating License",
May 8, 1978. l
APPENDIX A ASYMMETRIC STEAM GENERATOR TRANSIENT PROTECTION TRIP FUNCTION . l ! No change from original request. APPENDIX B > ! METHOD FOR CALCULATING SPACE-TIME SCRAM REACTIVITIES Not used in this supplement. I APPENDIX C i DESCRIPTION OF MODEL USED TO SIMULATE NSSS BEHAVIOR DURING STEAM LINE RUPTURE EVENT. No change from original request. APPENDIX 0
~
PROTOTYPE ASSEMBLY DESCRIPTION No change from original request. APPENDIX E DESCRIPTION OF MODIFIED ASSEMBLIES No change from original request. I d j i
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