ML20038A860
| ML20038A860 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 11/19/1981 |
| From: | BALTIMORE GAS & ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20038A849 | List: |
| References | |
| NUDOCS 8111240302 | |
| Download: ML20038A860 (15) | |
Text
.
9.0 TECHNICAL SPECIFICATIONS l
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8111240302 811119 DR ADOCK 05000317 P DR_-
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9.0 Technical Specifications Some of the Technical Specification changes which are being I,
requested for Cycle 6 operation are contained herein. With i
the exception of Change Nc. 3, the changes are very similar l{
to those requested in the Cycle 5 license submittal which used the Statistical Combination of Uncertainties (SCU) methodology and the CETOP code Table 9-1 presents a summary of the lechnical Specification changes requested. Table 9-2 presents the explanations for the changes summarized in Table 9-1.
Each page from the Technical Specifications for which a change is requested is contained herein with the modification included.
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Table 1 Calvert Cliffs I Cycle 6 Technical Specification Changes il t
Change #
Tech Spec #_
Action 1
Figure 2.1-1,page 2-2 Replace Figure 2.1-1 with enclosed Figure 2.1-1 2
Figure 2.2-1 Replace Figure 2.2-'I with enclosed Page 2-l'.
Figure 2.2-1 Change LHGR to centerline melt limit 3
B.2.1.1 page B2-1 from 21 kw/ft to 21.3 kw/ft Change minimum DNBR value from 1.195
'4 B.2.1.1, B.2.2.1 pages B2-1, B2-3 to 1.23 as indicated on noted pages B2-5, B2-6 5
B.2.1.1, 8.2.2.1 "hange high power level trip and maximum pages B2-1, B2-4 high power level trip actuation from 112%
of rated thermal power to 110%
6 Figure 3.2-2 Replace Figure 3,2-2 with enclosed page 3/4 2-4 Figure 3.2-2 7
3.2.2 Change calculated valte of FxvT from Page 3/4 2-6 s.l.62 to 11.700 and FxyT >1.620 to F
>1.700 Change calculated value of FrT from I
8 3.2.3 T
l page 3/4 2-9 s.l.620 to sj.700 and change Fr
>l.620 to FrT >l.700 I
Change minimum DNBR of 1.195 to minimum l.
9 8 3 4.2.5, page B 3 4 2-2 DNBR of 1.23 l
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Table 9-2 Explanations for Cycle 6 Tech Spec Changes Change,_#
Tech Spec #
Explanation
[,'
1 Figure 2.1-1 Thermal Limit Lines have been changed to i'
reflect higher radial peaking factors and implementation of margin recovery programs.
2 Figure 2.2-1 The LHK LSSS has been changed to reflect higher radial peaking factors and the implementation of margin recovery programs 3
B. 2.1.1 LHGR to certerline melt is being raised based upon Cycle 6 analyses 4
B. 2.1.1, The minimum DNBR has been increased to 1.23 to B.2.2.1 be consistant with Statistical Combination of Uncertainties 5
B. 2.1.1, Statistical Combination of Uncertainties has B.2.2.1 removed the 2% power uncertainty from the transient analyses 6
Figure 3.2-2 The LHR LC0 is being changed as a result of higher radial peaks and the imple'entation of margin recovery programs.
7 3.2.2 Radial peaking factors, both FxyT and FrT, cre beina raised for Cycle 6.
T and FrT, are 8
3.2.3 Radial peaking factors, both Fr,y being raised for Cycle 6.
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9 B 3/4.2.5 The minimum DNBR has been increased to be l
consistent with Tech Spec 5.2.1.1 o
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.O.6 PERIPHERAL AXIAL SHAPE INDEX.Y, FIGURE 2.2-1 Peripheral Axial Shape Index. Y, Versus Fraction of RATED THERMAL POWER Amendment No. U, 24 2-11 CALVERT CLIFFS - UNIT 1
m
' ' 2.1 SAFETY Lift!TS O
II, BASES
,i v 2 I. 3 6k
~
2.1.1 REACTOR CORE
. this safety limit prevent overheating of the fuel cladding and pa sible cladding perforation which would result in th The restrictions Overheating of the i
recucts to the reactor coolant. state peak linear heat rate release of fission d
ey maintaining the stea yCenterline fuel melting will no occur d
fuel is prevente at or less than 21 kw/ft.
Overheating of the fuel cladding is.
prevented by restricting fuel creration to within the nucleate boiling for this peak linear heat rate.
regime where :ne heat transfer coefficient is large a Operation above the upeer boundary of the nucleate boiling regime could result in excessive cladding temperatures :ecause of the ons departure frem nuclea:e boiling (0:G) and the resultant sharp red during operation and therefore THERMAL POWER and Reactor Coo in heat transfer coefficient.
l tion.
ature anc Pressure have been related to dig thrcugh the CE-1 ccrre a The CE-1 CNE correlation has been ceveloped te predict the OfG flux cistri-the loca icn Of D:3 fer axially unifom and ncn-uniform heatT h
the heat flux that would cause DN5 at a particular core location to t e butiens.
(
flux, is indicative of the margin to DNS..
local hes rmal The minimum value of the DNER during steady state doeration, 1g5 113 l operaticnal transients, and anticipated transients is limited to This value ::rres::nds := a 95 percen: prc:a:ility at a 95 percent con-fidence level that OtG will no: oc:ur and is chosen as an appropriate margin to DIG for.all cperating conditions'.
/.2 3 -
The curves of Figures 2.1-1, 2.1-2
.1-3 and 2.1-4 show the d
sr Coolant System pressure an loci of points of THE??AL POWER, Reaous numa ccmbinations for whien the l
f vae maximun coid leg temoerature
.ir for tne family of axial shapes and The limits in Figures mini =vm CNER is no less : nan corresponding radial peaks shewn in Figure 52.1-1.2.1 inlet temperatures less than er ecual to SE0*F.
tamcerature is not a safety limit; hewever, operation above,
580*F is nc: pessible because of the actuation coolan: inle Reactor creratien at THE yAL PCWER levels higher th in (10 To Amendment No.M. 48
/
8 2-1
, CALVERT CLIFFS - UNIT 1 8
I
,----,,,,,,-n
-,-.-,m--
_e_,,,,,,
,c
SAFETY LIMITS I
I t I
BASES l1 1
I The area of safe operation is below and to the left of Table 2.1-1.
these lines.
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 shown on the figures.
The reactor protective system in ccmbinatien with the Limiting -
Conditions for Coeration, is cesigr.ed to prevent any anticipated ccmtina-tion of transient conditions for reactor cociant system temoerature,a pressure and preclude the existence of ficw instabilitias.
than 1.192
/. 2 3 2.1. 2 REACTOR CCO'. ANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reacter Coolant System frem overpressurization and thereby prevents the
~
release of radienuclides contained in :ne reacter coolant frem reaching
^
the containment acmespnere.
The reactor pressure vessel and pressurizer are designed to Se-tion III,1967 Ecition, of the ASME C:de for Nuclear Power F13nt Cecconents
) of design which permits a maxicum transient pressure of 110t (2750 psia desiered to ANSI B 21.7, Class I,1959 Editi:n, wnien permits a maximum pressure.
transient pressure of 110t (2750 psia) of c:mponen-design pressure.
The Safety Limit of 2750 psia is therefore consistent with tne cesign criteria and associated cede requireTents.
The entire Reactor Coolant System is hydrotested at 3125 psia to demenstrate intagrity prict to initial cperatien.
/
Amendment No. 73, #,40 t
B 2-3 CALVERT CLIFFS - UNIT 1 i;
9 y
_-,____,_w..
n
LIMITING SAFETY SYSTEM SETTIN35
- r 2.2
?,
A Z
f, BASES i
li
_2. 2.1 REACtCR TRIP SETPOINTS The Reactor Trip Setpoints see:ified in Table 2.2-1 are the values The Trip Setpoints t which the Reactor Trips are set for each parameter.to ensure that th a
ystem are prevented from exceeding their safety limits.
ave been sele e:
h i
ied Allowable Value is ac:eptable on the basis th s
less,
a etween the trip se: point and the A11cwable Value is equal to or f
f analyses.
than the drift allowance assumed ter each trip in the sa ety b
Panuel Reacter Trio The Manual React:r Trip is a redundant channel to the automati l reactor trip protective instrumentation enannels and provides manua capability.
Power Level-Hich i t The Power Level-High trip provides reactor core prete: fon aga d by a Pressurizer reactivity excursions wnich are too rapid to be protecte Pressure-Hign or Thermal Margin /Lew Pressure trip.
be The Pcwer Level-High trip set:cint is c:erator adjustable and level. Cpera:Or set no higher than 10% above the incicated THER"AL FCWER AL F0WER is action is recuirec to increase the trip setsoint as THERM The trip se:pcint is automatically de:reased as THERMAL l
The trip se::: int has a maximum value of 107.0% of RATE increased.
POWER.
THERMAL ?OWER anc a minimum se:;oint c decreases.
i due to l ste* v-state calibratien and instrumen errors, the maximum actua Il0 7o 12 of RATED HERMAL POWER, wnich is the value use Reactor Coolant Flow-low i
to prevent
'The Reactor Coolant Flew-Low trip provides core protect on l
Or ccolant DNS in the event of a succen significant decre,ase in reac:
permit Provisions have been made in the reactor protective sys flow.
/
Amendment No. 39
(
B 2-4 CALVERT CLIFFS - UNIT 1 k
,.3--
v-
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--t---
ese
ew-->
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'-a Ti
6 s
Lin' TING SAFETY SYSTEM SETTIN35 l
BASES I,
~
/. 2 3 l{ i operation of the reacter at reduced power if one or.wo reactor coolante: points an The low-flow trip pumps are taken out of service.
Values for the varicus reactor ecolant pump combi.tiens have been response times of d
derived in consideration of instrument errors an.19' under normal coeration I
equipment involved to maintain the DNER aboveFor reactor operation w1:n only two or and expected transients.pumes operating, the Reactor Coolant Ficw-Lew trip set-reactor coolan points, the power Level-High trip setecints, and the Therm selector switch is mar.ually set to the desired two-or three-pumo Changing these trip setooints during two anc three pump i
operation prevents the minimum value of DNSR frcm going belcw 1.195 during position.
normal operational transients and anticipated transients when only two or three reactor coolant pu=ps are operating.
Pressurizer Pressure-Hich The Pressurize'r Pressure-High trip, backed up by the pressurizer cede safety valves anc main steam line safety valves, previces reactor coolant system protection against overpressuri:ation in the event of loss of load This trip's se:;oint is 1C0 psi below the nominal
~
without reactor trip.
lift setting (2500 psia) of the pressarizer coce safety valves and its concurrent operation with the pcwer-operated relief valves avoids the undesirable operation of the pressurizer code safety valves.
Containrent Pressure Wieh The Contair.nent Pressure-High trip provides assurance that a reactor The setpoint trip is initiated cencurrently witn a safety injection.
for this trip is identical to the safety injection setpoint.
. Steam Generater pressure-Low The Steam Generater pressure-Low trip provides protection against an excessive rate of heat extraction from the steam generators and The setting of 570 psia l
subsecuent ecoldcwn of the reactor coolant.
is sufficiently below the full-1cac cperating point-of 850 psia so as not to interfere with normal coeration, but still high enough to provide the required protection in the event of excessively high steam 22 psi This setting was used with an uncertainty factor of A flow.
in the accident analyses.
Amendment No. 37,4L
(
,CALVERT CLIFFS - UNIT 1 B 2-5 l
..e MW W ehW e*
--w
V' o
o LIMITit:0 SWETY SYSTEM SETTI?:GS i
g b
I, I
BASES
'l I t
lj i
Steam Generater Water Level The Steam Generater Water Level-Low trip provides core protection by preventing c;eration with the steam generator water level below the minimum volume required for adequate neat removal capacity and assures its that the pressure of the reactor c olant system will not exce be sufficient wa er inventory in the steam generators at the time of' Safety Limit.
i trip to provide a margin of more than 13 minutes before auxil ary feedwater is required.
1.1'3 Axia1 Flux Offset ensure that ex:essive The axial flux offset trip is provided tT.e axial flux offset is axial peaking will not cause fuel damage.
The trip setpoints l
detemined from the axially split excere d.ectors.
ensure that neither a D!GR of less than,.1 - nor a Deak linear heat rat i
ill which corresp:ncs to the temperature for fuel centerline melt ng w These trip set-exist as a c:nse:cen:e of axial power =aldistributiens.fr:m an analy i ted points were derive: allowances for instrumentatien inaccuracies and
~
with the excore to intore axial flux offset relationship.
Thermal Marcin/Lew Pressure i
The Thermal Margin /Lew Pr ssure trip is provided to prevent o f,g 3 when the DNER is less than 1.19:
The trip is initiated whenever the reactor coolant system pressure ibed signal dr:;s below eitner 1750 psia cr a e m:vted value as des:rT below, whichever is higher. higher of ai power or neutron pcwer, r i m CEA number of reactor coolant pum:s coerating.
coolant flow rate, the maximum AZIMUTHAL POWER TILT and t deviation pemitted for continu:us coeration are assumed Finally, the tion of this tri; function.
dance with Specifications 3.1.3.5 and 3.1.3.6 is assumed.
icipated maximum insertion of CEA banks which can d
J Amendment No. 32, M',48 I
B 2-6 f
CALVERT CLIFFS - UNIT 1 I
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FIGURE 3.2-2 Linear Heat Rate Axial Flux Of fset Control Limits D
e
/
CALVET<i CLIFFS - UNIT I 3/4 2-4 Amendment No. 9,18 l
' POWER DISTRIEUTION LIMITS ff 7 TOTAL PLA.GR RADIAL PEAKING FACTOR - F Ili LIMITING CCNDITION FOR OPEF.' TION
~
T T
XY(1+T ), sna11 be
[
F 3.2.2 The cal ' lated value of F*Y, defined as F
=
9 XY limited to <
.62s.
7g,,
APPLICAEILITY_: MODE 1*.
ACTI0ti:
f,7oo I
With F
> 1.62,within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> either:
T Reduce THERF.AL POWER to bring the combination of THERMAL POWE xy to within the limits of Figure 3.2-3 and withdraw the a.
and F' full Ength CEAs to or beyond the Long Term Steady State Insertion Limits of Specification 3.1.3.6; or b.
Be in at least HOT STAND 3Y.
SURVEILL ANCE RIOUIREME'iTS The provisions of Specification 4.0.4 are not applicable.
4.2.2.1 shall be calculated by the expression F
= F,y (1+T ) and F q
4.2.2.2 F shall be detemined to be within its limit at the following intervals:
Prior to operation above 70 percent of RATED THERMAL POWER a.
after each fuel loading, At least once per 31 days of accumulated operation in MODE 1 b.
and Within four hours if the AZIMUTHAL POWER TILT (T ) is q
c.
'See Special Test Exception 3.10.2.
3/4 2-6 Amendment No. 27,24,22.
M. 48
~
CALVERT CLIFFS - UNIT 1
~
e
a
(
, POWER DISTRIEUTION LIMITS
,t T
s TAL INTESRATED RADI AL PEAKING FACTCR - F r
~ t TO ll !i i
LIMITING CONDITICM FOR ODER*. TION T = F (1+T ), shall be l
T The cal lated value of F, defined as Fr r
q r
3 2.3 limited to <
.e2.
/.700.
~
APPLICA51LITY: MODE 1*.
ACTIOfi:
,/,700 t
~
within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> eithe'r:
I With F >
.520 r
Se in at least HOT STANDBY, or a.
Reduce THERMAL POWER to bring the ccm b.
length CEAs to or beycnd the Long Term Steidy 5:ete Insertion The THERMAL PCWER limit Limits of Specificatien 3.1.3.5.
determined from Figure 3.2-3 shall then be tise the allcwa:1e fracticn cf RATED THEIiAL F0WER maintained within the reduced accepteble ope Ficure 3.2 4 a:
Figure 3.2-4.
SUR"EILLc"CE RE:UIEriENTS The provisions of Soecification 4.0.4 are not applicable.
f
- .2.3.1 Ff shall be calculated by the expressien Ff = F (1+T l a 7
q l
- .2.3.2 shall be detennined to 5e within its limit at the fcilowing interva s
Prior to operation above 70 percent of RATED THERMAL POWE i
a.
after each fuel loading, At least once per 31 days of accumulated operatica in MODE i
b.
and Within four hours if the AZIKUTHAL POWER TILT q
i I
c.
I 43es Speciai iest Exception 3.10.2.
~
'3/4 2-9 Amendment No. 27, 24, 32,
{ALVERT CLIFFS - UNIT 1 33, 79,43 1
(
f l
r_
y-
,iPOVER DISTRIBUTION LIMITS g
3, BASES I,t i /Lew I
he analysis establishing the DNS Margin LCO, and Th t
If F' F' or T exceed their P
llowable CEA group insertion limits.asic limitatier.s. operation ma i
a l restric-ions imposed by the ACTION s:atements since these additienathe ass,m;; io b
ions provide adecuate provisiens to assure tha:
n establishing the Linear Heat Rate, Thermal Margin / Low Pre t
t id. An ocal Power Density - High LCOs and LSSS setecints remain val i
b ZIMUTHAL FC"ER TILT > 0.10 is not expected and if it should occu quired L
equert coeration wculd be restricted to only these operations re A
to identify the cause of this unexpected tilt.
s that must be used in the equation F*Y = F*7 (1 + T )
T 9
he value of T and F{r = Fr(1+T)Tsthemeasuredtilt.T T
q F and,.T are The surveillance recuirements for verifying that F1vEu,es#cfF'$F' r
within their limits provice assurance tha, tne actua T VerifyineF,,,,andFiafEr do not exceed the assumed values.
E2f.AL POWER provides each fuel leading prior to exceeding 75 of RATED and T I
FUEL RESIDENCE T!".E_
3/a._2.4 This specification deleted.
DN3 PARAMETERS f the 3/4.2.5 The limits en the DN3 related parameters assure that each o e envelcpe of parameters are maintained within the normal steacy statand at:ice The limits are operaticn assumed in :ne transien:
consistent witn the safety analyses ass.:mp:icns and have 95 throughout each l
demonstrated adecuate to maintain a minimum DN3
>/.2.3 analy:ec transien:.
h instru-The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodic surveillance of these parameters thro ment readcut is sufficient to ensure that the para xpected transient h
within their limits following 1 cad chances and ot er eTh flow rate l
ion of the is adequate to detect flow degradation and ensure corre at indicated operation.
flow indication enannels with measured flew such f flow rate on a percent flow will provide sufficient verification o 5
12 hcur basis.
/
Amendment No. 27, 72, B 3/4 2-2 33, M,48 CALVERT CLIFFS - UNIT 1 u
,