ML20134N141
| ML20134N141 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 08/30/1985 |
| From: | BALTIMORE GAS & ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20134N133 | List: |
| References | |
| NUDOCS 8509050016 | |
| Download: ML20134N141 (26) | |
Text
..
O.0 TECHNICAL SPECIFICATIONS The Technical Specification changes which are being requested in order to make the Calvert Cliffs Unit 2 Technical Specifications consistent with either the reference cycle (Reference 1) analyses which have been verified j
for Unit 2. Cycle 7 or the analyses contained herein are presented in this section.
All changes except two which are being requested herein for Unit 2 were approved for Unit 1 in Reference 2.
The first new change makes the affected Unit 2 Technical Specification identical to the corresponding Unit 1 Technical Specification.
The second new change lowers the minimum DNBR Technical Specification limit to make it consistent with the final DNBR limit approved in Reference 3.
Table 9-1 presents a sumary of the Technical Specification changes, in the form of: 1) an action statement for each change;. 2) the reason for each change and 3) a reference to the supporting analyses which demonstrate acceptable safety analyses results for each change.
Following Table 9-1 the existing Technical Specification page with the intended modification is provided for each Technical Specification for which a ch.nge is being requeste<i.
The Technical Speci fication changes requested herein for Unit 2 are identical to those that were approved for Unit 1 in Reference 2 with the following exceptions:
- 1) All HPSI flow related changes for Unit 2 (including ASI limitations) have already been submitted in Reference 4 for apolication to Cycle 6.
- 2) The change in the surveillance interval for CEA insertability in
~
Technical Specification 3/4.10.1 has already been submitted in Reference 5 for application to Cycle 6.
- 3) Changes in the lift setting values and format of the Main Steam Safety Valve (MSSV) Technical Specification, and changes to permit entry into Mode 3 with 2 MSSVs per steam generator operable are being submitted separately for application to Cycle 6.
- 4) The minimum DNBR SCU based limit (see Section 6) is being lowered to make it consistent with the final minimum non-SCU based DNRR limit approved by the NRC in Reference 3.
- 5) The radial peaking factor at which full power operation may proceed when operating on the excore monitoring system is being raised to make it identical to the Unit i value and to increase coerating margin.
1 8509050016 850830 ADOCK 0 % g 8 DR 9-1
Table 9-1 Calvert Cliffs 2 Cycle 7 Technical Specification Changes
' Tech. Spec.
No. and Page Action Explanation Support H.2.1.1, H.2.2.1 Change minimum DNAR The minimum SCll based DNRR limit The discussion in Chapter 6, concern-pages 82-1, limit fran 1.23 to is being lowered to make it ing the derivation of the new SCU 02-3, 82-5, 1.21 consistent with the final mini-based DNRR limit using the NRC approv-H2-6 mum non-SCll based DNRR limit ed non-SCll hased final DNBR limit and approved by the NRC.
previously approved SCU and rod how penalty methodologies, supports this reduction.
3/4.1.1.1 Change shutdown margin, The shutdown margin is being g
T,N>200"/F,fran page 3/4 l-1 lowered to accomodate the 4.
ak/k to 3.51,Ak/k effects of extended burnup.
3.1.1.4 Change HTC positive limit.4 The HTC is being raised to The safety analyses presented page 3/4 1-5 Power < 70%, from +0 5x10-accomodate the effects of in Chapter 7 of Reference 1 A k/k/*F to *0.7x10-4 long cycles and to simpilfy and verification in Chapter 7
.,o a k/k/ F.
startup procedures.
of this document that these su analyses are applicable to Unit 2 Cycle 7 support these ChangeH{Cnegagivelimitfrom The HTC is being lowered to changes.
-2.5x10-ak/k/ F to tn accomodate the effects of
-2.7x10-4 ak/k/ F.
of extended burnup.
U 4.2.1.4 Remove flux peaking Augmentation factors are being
- 1) Detailed discussion and data was page 3/4 2-2 augmentation factors removed in recognition of the presented in Reference.6 to demonstrated lack of gap for-support this change, mation in pre-pressurized non-
- 2) The thermal design analysis of densifying fuel and to increase the fuel pirs presented in operating margin, Section 4.3 supports the change.
- 3) The ECCS performance analysis for the large break spectrum presented in Section 8.1 supports this change.
'l Table 9-1 (cnotinued)
Tech. Spec.
No. and Page Action Explanation Support 4.2.1.4 Reduce the measurement-This uncertainty is heing reduced
- 1) The new value is supported in (cont.)
calculational uncertainty to conform to the approved value Reference 7 fran 7.01 to 6.21 and to increase operating margin.
- 2) The thermal design analysis of the fuel pins presented in Section 4.3 supports this change.
Reduce the axial fuel This uncertainty is being reduced The thermal design analysis densification and thermal I
to a level consistent with existing of the fuel pins presented in expansion factor from calculations and ho increase Section 4.3 supports this change.
1.01 to 0.2%
operating margin.
i figure 3.2-3h Hodi fy Figure 3.2-3h as This radial peaking factor, in the The setpoint analysis for tinit 2 l
page 3/4 2-4a indicated to increase the form of the variable
'N', is being Cycle 7 supports this change.
the radial peaking factor at increased to make the ilnit 2 Figure which full power operation 3.2-3h identical to the correspond-
,,j, may proceed when operating ing linit I figure and to increase on the excore monitoring operating margin.
t system fran 1.50 to 1.54 Figure 4.2-1 Delete Figure 4.2-1 See change for Tech. Spec.
See change for Tech. Spec.
page 3/4 2-5 4.2.1.4 which covers removal of 4.2.1.4 which covers removal of flux peaking augmentation factor.
flux peaking augmentation factors.
i f
l 4
1 d
Table 9-1 (continued)
Tech. Spec.
No. and Page Action Explanation Support R 3/4.1.1.1 Change ESC shutdown margin, See change for Tech. Spec. 3/4.1.1.1 See change for Tech. Spec. 3/4.1.1.1 t8y3>.200"f, f ran 4.3% ak/k and T
B 3/4.1.1.2 St a k/k and change ROC 8
p:ge B 3/4 shutdown margin, T
>200 f, 1-1 fran 4.31a k/k to $50% a k/k R 3/4.2.1 Remove flux peaking augmenta-See change for Tech Spec. 4.2.1.4.
See change for Tech. Spec. 4.2.1.4.
page B 3/4 tion factors, change measurement-2-1 calculational uncertainty fran 7.01 to 6.2% and change axial fuel densification and thermal expansion factor from 1.01 to
.o 0.21 L
B 3/4.2.5 Change minimum DNBR limit from See change for Tech. Specs. R.2.1.1 See change for Tech. Specs. R.2.1.1 ptge B 3/4 1.23 to 1.21 and 8.2.2.1 and 8.2.2.1 2-2 Insert the additional text The BASES section for the DNR LCO The text is merely updating the concerning limiting criteria is being expanded to more clearly BASES to descrlhe what has been on the DNB LCO, as indicated define all of the criteria which standard practice.
are used to establish the Tech.
Spec. values.
8
2.1 SAFETY LIMITS i
BASES 2.1.1 REACTOR CORE.
1 The restrictions of this safety limit prevent overheating of the fuel cladding and possible cladding perforation which would result in the l
release of fission products to the reactor coolant.
Overheating of the fuel is prevented by maintaining the steady sta:a peak linear heat rate l
at or less than 22.0 kw/ft.
Centerline fuel melting will not occur l
for this peak linear heat rate.
Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime wnere the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.
Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNS) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reac:cr Coolant Temper-a:ure and Pressure have been related to DNS through the CE-1 correlation.
T4p CE-1 DNB correlation has oeen caveloced to predic the DNS flux and
- ne location of DNB for axially uniform and non-uniform neat flux cistri-but.ans. The local DNS heat flux ratio, DNBR, defined as the ration of the heat flux tnat would cause ON8 at a carticular care location to :ne iccal hea flux, is indicative of :ne margin :c 045.
The minimum value of the ONBR during steady stata aceration, -
..al Operational transients, ano anticicated transients is limitad to 1.2' l.11 l
This value corresponos to a 95 per ant pr:bability at a 9E :ercent con-fidence level that DNS will not oc:ur and is chosen as an acpropriata margin to DNB for all operating conditions.
{,21 The curves of Figures 2.1-1, 2.
. 2.1-3 and 2.1-4 snew the loci of points of THERMAL POWER, -
ctor Coolant System pressure and maximum cold leg temperature f ricus pumo c:mbinations f:r *nica :ne i
minimum DNBR is no less than
.c" for the family of axial shaces ano l
correspanoing radial peaks shown in Figure 32.1-1.
The limits in Figures 2.1-1, 2.1-2, 2.1-3 and 2.1-4 were calcula:ec f:r reac::r c: clan:
inlet temceratures less than or ecual to 580*F.
The dashed line a: 580*F coolant inlet :amcerature is not a safety limit; however, cceraticn acave 580*F is not possible because of the actuation of ne main.:aam line safety valves *nich limit the maximum value of reac:cr inier :am era ure.
Reac cr oceraticn at THERMAL POWER levels nigner than 110% Of RATED THER. MAL 90WER is pronibi ad cy :t.e nign power level ria set:cin: s:ecifiac in i
\\
s CALVERT CLIFF 5 - UNIT 2 5 2-1 Atendrent No. ?3, 37, 31 9-5 m
..-,--..,.r-
---c
,y
SAFETY LIMITS
(
BASES Table 2.1-1.
The area of safe operation is below and to the left of these lines.
The conditions for the Thermal Margin Safety Limit curves in Fi 2.1-1, 2.1-2, 2.1-3 and 2.1-4 to,be valid are shown on the figures. gures The reactar protective system in ecmbination with the Limiting Conditions for Operation, is designed to prevent any anticipated combina-tion of transient conditions for reactor coolant system temperature, pressure, and THERMAL POWER level that would result in a ONBR of less e
than(}]{)and preclude the existence of flow instabilities.
l 2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reac:ce Coolant System fecm over:res:uri:ation anc :nerecy prevents -he 0
release of radionuclides containec in :ne reac:ce c: alan:
(
the containment atmos:nere.
fr:m reacning The react:r pressure vessel and pressuri:er are cesicned to Secti:n I:!,1967 Edi-icn, of :ne ASME Code for Nuclear Fewer Plant Cc=conents wnien :ennits a maximum transient pressure of 110% (2750 :sia) of casign The Reac:cr Coolant System pioing, valves and fi :ings, are cressure.
designed to ANSI 3 31.7, Class I,1969 Edition, whien permits a maximum transient pressure of 110% (2750 psia) of c:moonent design pressure.
The Safety Limit of 2750 psia is therefore c:nsistent with the cesign criteria ano associatec code recuirements.
The entire Reac:cr Coolan: 5y' stem is hydro:asted a: 3125 :sia to demonstrate integrity prior to initial operation.
L CALVERT CL:F?S - UNIT 2 3 2-3 kren c=en: No. 73,2), 31 9-6
1 LIMITING SAFETY SYSTEM SETTINGS
\\
BASES l
- 1. Ll
\\
operation of the reactor at reduced power if one or l
wo reactor coolant pumps are taken out of service.
The low-flew trip etpoints and Allowable Values for the various reactor coolant pumo combi tions have been derived in consideration of instrument errors nd response times of equipment involved to maintain the DNBR above 1.2 under normal oceration I
and expected transients.
For reac:or operation with only two or three reactor coolant pumps operating, the Reactor Coolant Flcw-Lew trip set-points, the Power Level-High trip setpoints, and the Thermal Margin / Low Pressure trip setpoints are automatically changed when the pumo ccnditicn selector switch is manually set to the desired.two-or three-pump position.
Changing these trip set oints during two and three pumo
/ lLl operation prevents the minimum value of ONBR fr0m going below 1.23 during l
normal ocerational transients and anticipated transients wnen anny two or three reactor coolant pumcs are operating.
P"essurizer Pressure-Hich The Pressuri:er Pressure-Hign tric, backec up q' y One pressuri:er c:ce safety valves and main steam line safety valves, Orovices reac::r :: alan.:
i system protaction against overaressuri:ation in :ne event of loss of loac.
without reac:cr tric.
This trio's se :oint is 100 psi tel:w ne n:=inal lif setting (2500 :sia) of :ne cressuri:er code safe:y valves anc its c:ncurren: Ocera: ion wi:n the :cwer-ocera:ac relief valves avoics ne uncesiracle aceration of the pressuri:er code safety valves.
Containment Pressure-Hicn The Containment Pressure-Hign trip provides assurance Ona: a react:r tric is initiatec c:ncurrently wita' a safety injection.
The se::cint for this trio is icentical to :Me safety injection se: point.
Steam Generet:r 3esssure-Low The Steam Generator P'ressure 8.:w trio :r:vides cratacticn agains an excessive rate of heat ex raction fr:m :ne steam genera: Ors ano subsecuent c:olcewn of :ne reac:Or c:alant.
The setting of 555 sia l
is sufficiently below the full-Teac ocerating :cin: of 350 :sia se as nct to in:arfere wi:n normal cceration, but still ni;n enougn :
- revide :ne recuirec protection in :ne event of excassivelv ni:n s: sam ficw.
This se::ing was usec tita an uncartainty fac::r of'- 35 05 i
in tne ac:icent analyses which was basec cn the Main 5taam Tine 3reak
{.
tvent.
I CALVERT CLIFFS - UNIT 2 3 2-5 Amenc=en: No. 73,27, 31 9-7
I i
1 I
LIMITING SAFETY SYSTEM SETTIi;G5 I
BASES Steam Generator Water Level The Steam Generator Water Level-Low trip provides core protection by preventing cperation with the steam generator water level below tne minimum volume required for adequate heat removal cacacity and assures that tne pressure of the reac:cr coolant system will not exceed its Safety Limit.
The specified setpoint in combination with the auxiliary feedwater actuation system ensures that sufficient water inventory exists in both steam generators to remove decay heat following a loss.of main feedwater ficw event.
Axial Flux Offset f, gf ine axial flux offset trip is provide to ensure na; excassive l
axial ceaking will not cause fuel damage.
The axial flux offset is detemined frem the axially soli execre datec:ces.
The trip se:coints i
ensure tnat neitner a DNSR of less than 1.23 nor a peak linear heat rate wnien correspencs to the temcerature for ruel centerline reiting will l
exist as a consecuence of axial gewer malcistributions. These tric set-points were derived from an analysis of-many axial cower snaces wi n Q.)
allowances for instrumentation inaccuracies and the uncertainty asscciated with tne execre :c incere axial flux offsat relationsnic.
Themal Marcin/Lew Pressure The Thernal Margin / Low P-assure trio is previced to crevent coeration when the DNBR is less tnanhf,gl l
The trip is initiated wnenever the reac:cr coolant syste.9 pressure signal droos below ei ater 1875 psia or a ccmcuted value as described l
below, whichever is higner.
The comouted value is a function of the higher of AT power or neutren ;cwer, reactor inlet tamcerature, anc the numoer of reactor coolant pumps ocerating.
The minimum value of reac:cr coolant ficw rate, the maximum AZIMUTHAL POWER TILT anc :ne maximum CIA deviation cermitted for continucus aceration are assumec in :ne genera-tion of this trip function.
In addition, CIA group secuencing in accer-dance witn Soecifica:icns 3.1.3.5 and 3.1.3.5 is assumed. Finally, :ne maximum' insertion of CEA banks wnich can ccour during any anticicated operational occurrence price to a Pcwer Level-Hign tric is assumec.
.,)
CALVERT CLIFFS - UNIT 2 3 2-6 Amencment No.;a, L 31 9-8
i 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUTDOWN MARGIN - T,y, > 200*F LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be d.3P.* kk/.
l APPLICABILITY: MODES 1, 2", 3 anc 4 ACTION:
3.5 With the SHUTDOWN MARGIN <
3f.* ak/k, immediately initiate and continue
[
boration at 1 0 gpm of 2300 ppm boric acid solution or equivalent until 4
the required SHUTDOWN MARGIN is restored.
SURVEILLANCE RE0'UIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be detennined to be 1 k/k:
l Within one hour after detection of an inoceraole CEA(s) and at a.
least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter while the CEA(s) is inoceracle.
If the inocerable CEA is immovable or untripoable, tne aoove required SHUTDOWN MARGIN shall be increased by an amount at least equal to the withdrawn worth of the ic=cvable or untripoable CEA(s).
When in MODES 1 or 2*, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying b.
that CEA group wi:ndrawal is within tne Transient Insertion Limits of Specification 3.1.3.6.
When in tiODE 2", within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to achieving reactor c.
criticality by verifying that the predicted critical CEA position is within the limits of Soecification 3.1.3.5.
d.
Prior to initial operation above 5: RATED THERMAL POWER after eacn fuel loading, by consideration of the factors of e celow, with the CEA groucs at the Transient Insertion Limits of Specification 3.1.3.6.
Adherence to Tecnnical Specification 3.1.3.5 as specified in Surveillance Requirenents 4.1.1.1.1 assures that there is suf ficient availacie snu:-
I down margin to match the shutdown margin requirements of the safet/
i analyses.
See Special Test Excection 3.10.1.
With K,ff 1 1.0.
- With X,ff : 1.0.
CALVERT CLIFFS - UNIT 2 3/4 1-1 Amencment No. 3. 73, U, Jg, 72 9-9
i REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION 3.1.1.4 The moderator temperature,oefficient (MTC) shall be:
a.
Less positive than.9 x 10~ ak/k/*F whenever THERMAL POWER is ;, 70" of RAi:D THERMAL POWER, b.
Less positive than 0.2 x 10'# ak/k/*F wnenever THERMAL POWER is > 70" of RATED THERMAL POWER, and c.
Less negative than 2.
x 10'4 ak/k/*F at RATED THERMAL l
POWER.
2.7 APPLICABILITY
M0'CES-1 and 2*f ACTION:
With the moderator temoerature coefficient outside any one of the above limits, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
o SURVEILLANCE RECUIREMENTS 4.1.1.4.1 The MTC shall be determined to be witnin its limits by confirmatory measurements. MTC measured values snall be extracolated and/or compensated to permit direct comparison with the above lie.its.
Twitn K,f,>,1.0.
- See Special Test Exception 3.10.2.
CALVERT CLIFFS - UNIT 2 2/4 1-5 Amendment No. 73, 3!, 72
'9-10
e P0HER DISTRIBUTION LIMITS SURVEILLANCE REOUIREMENTS (Continued) c.
Verifying at least once per 31 days that the AXIAL SHAPE INDEX is maintained within the limits of Figure 3.2-2, where 100 percent of the allowable power represents the maximum THERMAL POWER allowed by the following excression:
MxN where:
1.
M is the maximum allowable THERMAL POWER level for the existing Reactor Coolant Pump comoination.
2.
N is the maximum allowable fraction of RATED THEFF.AL PONER as detemined by the Fj curve of Figure 3.2-3b.
l y
4.2.1.4 Incore Detector Monitorinc Svstem - The incore detector moni-taring system may se usec for monitoring tne core power distributien by verifying that the incere detector Local Power Oensity alarms:
a.
Art adjusted to satisfy the recuirements of the core ::ower distribution man wnich shall be uodated at least once ser 31 days of accumulated operation in MODE 1.
b.
Have their alam setpoint ad3usted to less than or ecual to the limits shown on Ficure 3.2-1 when the following factors are appropriately included in the setting of these alarms:
. p y.,in; __,........ 10 #::::-- : the
'- c' p :
1.062 A measurement-calculational uncertainty factor of 1.0s,
3.
An engineering uncertainty factor of 1.03, 3
1.002 A linear heat rate uncertainty factor of e to
- 4. xial fuel densification and themal exoanslen, and a
intRMAi. POWER measurement uncertainty factor of 1.02.
a.
a CALVERT CLIFF 3-UNIT 2 3/4 2-2 Amendment No. 5, 3, 76, 73, Lgy 9-11 4
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20 40 60 El 100 120 140 CCRE HEIGHT. INCHES FIGURE 4.2-1 Augmentation t sciar vs Distance from Bottom of Core CALVEFT CLIFF 3 - UNn' 2 3/a 2-5 Amendmen: No. 3, 73, 31 J
9-13
3/4.16 REACTIVITY CONTROL SYSTEMS BASES i
3/4.1.1 80 RATION CONTROL l
3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made suberitical from all operating conditions, 2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and 3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.
, 3. 5 SHUTDOWN MARGIN requirement ary throughout core life as a function of 3.5 fuel depletion, RCS boron conc tration and RCS T The minimum availa SHUTDOWN MARGIN for no load erating conditions N9beginning of life is Ak/k and at end of life is.
Ak/k. The SHUTDOWN MARGIN is based on t e safety analyses performed for a steam line rupture event initiated at no load conditions. The most restrictive steam line rupture event occurs at EOC
~
conditions. For the steam line ruptur ent at beginning of cycle conditions, a minimum SHUTDOWN MARGIN of less than
% Ak/k is requir to control the reactivity transient, and end of cycle co ditions require. 5 Ak/k. Accordingly,
the SHUTDOWN MARGIN requirement is based upon this limiti ondition and is consistent with FSAR safety analysis assumptions. With i N 200"F, the reactivity transients resulting from any postulated accidE are minimal and a -
3% Ak/k shutdown margin provides adequate prote'ction. With the\\ pressurizer level less than 90 inches. the sources of non-borated water are restricted to increase the time to' criticality during a baron dibtion event. \\
\\
L 3/4.1.1.3 BORON DILUTION f.5 3.5 A minimum flow rate of at least 3000 GPM provides adecuate mixing, prevents stratification and ensures that reactivity changes will be gradual during baron concentration reductions in the Reactor Ccolant System. A flow rate of at least 3000 GPM will circulate an equivalent Reactor Ooolant System volume of 9.601 cubic feet in approximately 24 minutes. The reactivity change rate associated with baron concen-tration reducticns will therefore be within the capability of operator recognition and control.
3/4.1.1.a MODERATOR TEMPERATURE COEFFICIENT (MTC)
The limitations on MTC are provided to ensure.that the assumotions used in the accident and transient analyses remain valid through each fuel cycle. The surveillance requirements for measurement of the MTC jduring each fuel cycle are adequate to confirm the MTC value since this
- coefficient changes slowly due principally to the reduction in RCS baron
! concentration associated with fuel burnuo. The confinnation that the 4
' measured MTC value is within its limit provides assurances that the coefficient will be maintained within acceptable values througneut each fuel cycle.
CALVERT CLIFFS - UNIT 2 S 3/4 l-1 kendment No. I2. 3I,8J/, 7 9-14
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- ne ana!ysg asublisning te ONE Ma gin LCO, an: D er=al Ma,in/'j:w Pressure L.uS se=c1nu reain valid curing':; era-icn a: 2e va.-icus allowa:ie CIA gr:us insar. ion limits.
If.:
F' cr 7 ex: sed ceir basic li.iuti:ns,. egeration may =n-inue uIda,r Ine ad$iti:nai tstrie-tiens im::sec ey =e ACT*0N 3.1 aments sinca =ese a:di-icnal restrie-ti::s : : vide adecua:a pr: visions := assu t na: na assu==:i:ns usac in asu:lisning =e Linear Hea: Rau, ne-.a1 Margin /L:w ?ressurs anc L:ca'. Fewer Censity - Hign L.Os and LIII se ;:in.
sc:ain valid An m.h. v.W..,CW.s i m > 0.10 is not ex:ecud and if n. sh:u!d ec=r, sub-secuen ::erati:n wcuid be restric ad :: :nly 2:se :: ara-i:ns escuirse
- = identify :ne cause of =is unex;ac.ad til:.
S e value of 7 24: r.:s: be used in ne ecuati:n.:'
=? *Y (1 - 7 )
and F' =.:r (1-7 ) Ys =e measurte -ili.
27 9
De su-vefilanca recuirmenu f:r verifying =a
.:'.,, :* anc.7, ar.e wiu n meir li=in :r:vice assunnca =a: =a ac ual va ues':".:;. - F.
and 7,., = ne: ex:anc =e assumee values.
Vehfying .. an:. '. after eacn *uel icacing :rier = ex:ascing 75:
f,.:.A..e n...'aAL ;c'WF.:. ;=vicas acciti:nal assunn=a =a: ce =rt.45 pr:;erl.y 1:acac.
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ne limi.s en =t CNE -slanc ;a-tesurs assu t =a: eaen :' te I
- artme ars art maintained wirin me ner=ai suscy s.2
- a anvei :e :1
- ert:1:n assu=ed in =e : scsient anc acci:an; analyses.
The l i=i.s a t
=nsistant win ce safety analyses asst==:i:ns anc n ve :een acaiy ially cam =. s: 2:ac acecuau.: =ein.ain'/.a mini =:= CN:R
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9 9-16 l
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A In addition to the DNB criteria, there are two other critbria which set the i
specification in Figure 3 2-4 The second criteria is to ensure that the existing core power distribution at full power is less severe than the power distribution factored into the small-break LOCA analysis.
This results in a limitation on the allowed negative AXIAL SHAPE INDEX value at full power. The third criteria is to maintain limitations on peak linear heat rate at low power levels resulting from Anticipated Operational Occurrences (A00s).
Figure 3.2-4 is used to assure the LHR criteria for this condition because the linear heat rate LCO, for both ex-core and in-core monitoring, is set to maintain only the LOCA kw/ft requirements which are limiting at high power levels.
At reduced power levels, the kw/ft requirements of certain ACOs (e.g.,
CEA withdrawal),
tend to become more limiting than that for LOCA.
e O
9-17
I i
10.0 STARTUP TESTING The startup testing program proposed for Cycle 7 is identical to the program proposed for the reference cycle in Reference!1.
0 9
2 9
11.0 REFERENCES
References - Chapters 1 Through 3 1.
- Letter, A. E. Lundvall, Jr. (BG&E) to J. R. Miller (NRC),
Docket No. 50-317, "Calvert Cliffs Unit 1 Eighth Cycle License Application,"
February 22, 1985.
l t
2.
Calvert Cliffs Nuclear Power Plant Units 1 and 2 Updated Final Safety Analysis Report Chapter 3.
3.
Letter,. O. H. Jaffe (NRC) to A. E. Lundvall, Jr. (BGAE), Occket No.
50-317, " Safety Evaluation of Calvert Cliffs Unit 1 Cycle 8," May 20, 19RS.
4
- Letter, A. E. Lundvall, Jr. (BG&E) to R. A. Clark (NRC), Occket Nos.
50-317 and 50-318, " Topical Report for Extended Burnuo Operation of C-E Fuel," June 7,
1987; Enclosure CENPD-760-P, " Extended Burnup Operation of Combustion Engineering PWR Fuel," April 1982.
5.
- Letter, A. E. Lundvall, Jr. (BG&E) to T. E. Murley (NRC), "Calvert Cliffs Nuclear Power Plant Unit No. 2, Docket No. 50-318 Report of Startup Testing for Cycle 6,"
October 8, 1984 6.
CEN-105(R)-P, "Reconstitutable 8C Type CEA Design for Use in the 3
BGAE Reactor," February,1979.
7.
CE-NPSD-225 P, "Results of Calvert Cliffs Unit 2 EOC 4 Poolside Inspection *of Contro1' Element Assemblies," April, 1983; 8.
CE-NPSD-266-P, "Results of Calvert Cliffs-2 EOC-5 Poolside Inspection of Control Element Assemblies," May, 1985.
11-1
. ~, -
I References - Chapter.4 1.
- Letter, A. E. Lundvall, Jr. (BG&E) to J. R. Miller (NRC), Docket No.
50-317, "Calvert Cliffs Unit 1' Eighth Cycle License Application,"
February 22, 1985.
Calvert Cliffs Nucleah Power Plant Units 1 and 2 Updated Final Safety 2.
Analysis Report Chapteh 3.
3.
- Letter, A. E. Lundvall, Jr. (BG&E) to R. A. Clark (NRC), Docket No.
50-318, "Calvert Cliffs Unit 2 Fifth Cycle License Application,"
October 15, 1982.
4
- Letter, A. E. Lundvall, Jr. (BG&E) to R. W.
Reid (NRC), Docket No.
50-317, " proposed Finding of No Unreviewed Safety Ouestion on Unit 2, Cycle 3 Reload Core Design," July 11, 1979.
5.
- Letter, A. E. Lundvall, Jr. (BGAE) to R. W. Reid (NRC), Docket No.
50-318, " Unit 2 Cycle 2 License Application," July 26, 1978.
6 Letter, A. E. Lundvall, Jr. (RGAE) to J. R. Miller (NRC), Docket Nos.
50-317 4 50-318, " Request for Amendment," (Clad Collapse / Augmentation Factors), December 31, 1984 7.
EPRI Report NP-3966-CCM, Volume 5,
" Evaluation of Interpellet Gap Formation and Ciad Collapse in Modern PWR Fuel Rods," April 1985.
8.
I,etter, O. H. Jaffe (NRC) to A. E. Lundvall, Jr. (BG&E), Occket No. 50 '
317, "Safet'y Evaluation of Calvert Cliffs Unit 1 Cycle 8," May 20,'1985.
9.
CEN-183(B) 0,
" Application of CENPD-198 to Zircaloy Component Dimensional Changes," September 1981, 10 Letter, D. H. Jaffe (NRC) to A. E. Lundvall, Jr. (BGAE), "Regarding Unit 1 Cycle 6 License Approval (Amendment 471 to OPR-53 and SER),"
June 24, 19R2.
- 11. Letter, A. E. Lundvall, Jr. (BGAE) to J. R. Miller (NRC), "Calvert Cliffs Unit 1 Supplement I to Seventh Cycle License Application,"
September 1, 1983.
- 12. Letter, A. E. Lundvall, Jr. (BGAE) to J. R. Miller (NRC), Docket No.
50-317
" Seventh Cycle License Application Answers to Question Set 2,"
November 4, 1983.
- 13. CEN-83(B)-P, "Calvert Cliffs Unit 1 Reactor Operation with Modified CEA Guide Tubes," February 8, 1978, and Letter, A. E. Lundvall, Jr.
(BGAE) to V. Stello, Jr. (NRC), " Reactor Operation with Modified CEA Guide Tubes," February 17, 1978 la. Letter, A. E. Scherer (C-El to C. O. Thomas (NRC),
"CEA Guide Tube Wear Sleeve Modification," LO-84-043, August 3, 1084
- 15. CENPD-139 P-A, "C-E Fuel Evaluation Model Topical 4ecort," July 1974
- 16. CEN-161(B) 0, "Imorovement to Fuel Evaluation Model," July 1981.
- 17. Letter, R.
A.
Clark (NRC) to A. E. Lundvall, Jr. (BG&E),
" Safety Evaluation of CEN-161 (FATES 3)," March 31,1983.
11 9
l References - Chapter 5 1.
- Letter, A. E. Lundvall., Jr. (BG&E) to J. R. Miller (NRC), Docket No.
50-317, "Calvert Cliffs Unit 1 Eighth Cycle License Application,"
February 22,,1985.
1 2.
Letter, A. E. Lundvall, Jr. (BE&D) to J. R. Miller (NRC), Docket Nos.
50-317 A 50-318, " Request for Amendment," (Clad Collapse / Augmentation Factors), December 31, 19R4 3.
Letter, O. H. Jaffe (NRC) to A. E. Lundvall, Jr. (BGAE), Docket No.
50-317, " Safety Evaluation of Calvert Cliffs Unit 1 Cycle 8,"
May 20, 1985.
e 11-3
i heferences - Chacter 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 Modeling Methods,"
January 1477.
4
- Letter, P.
W.
Kruse to W.
J.
Lippold, " Responses to First Round Ouestions on the SCU Program:
CETOP-0 Code Structure and Modeling Methods, (CEN-124(B)-P, Part 2)," May 1981 and letter, P. W. Kruse to W. J. Lippold (above document), BGE-9676-576, May 1, 1981.
5.
- Letter, D.
H. Jaffe (NRC) to A. E. Lundvall, Jr. (BG&E), "Regarding
~
Unit 1 Cycle 6 License Approval (Amendment 471 to DPR-53 and SER),"
June 24, 1982.
6.
CEN-124(B)-P,
" Statistical Combination of Uncertainties, Part 2,"
January 1980.
7.
Letter, C. O.
Thomas (NRC) to A.
E.
Scherer (C-E), " Acceptance for Referencing of Topical Report CENPD-207," November 2, 1984 l
8.
CENPD-207-A, " Critical Heat Flux Correlation for C-E Fuel Assemblies with Standard Spacer
- Grids, Part 2:
Nonuni form Axial Power Distributions," December, 1984 1
9.
- Letter, A. E. Lundvall, Jr. (BGAE) to J. R. Miller (NRC),
Dockct No.
50-317 "Calvert Cliffs Unit 1 Eighth Cycle License Apolication,"
February 22, 1985.
- 10. CENPD-225-P-A, " Fuel and Poison Rod Bowing," June 1983.
- 11. LQter, C. O. Thomas (NRC) to A.
E.
Scherer (C-E), " Acceptance for Referencing of Topical Report CENPD-225(P)," February 15, 1983.
l
- 12. CEN-124(R) D,
" Statistical Combination of Uncertainties, Part 1,"
January 1980, t
- 13. CEN-124(B)-P, " Statistical Combination of Uncertainties, Dart 3," March l
1980.
I 1
f 11-4
p References - Chapter 7 1.
- Letter, A.
E.
Lundvall, Jr. (BG&E) to J.
R. Miller (NRC), Docket No.
50-317, "Calvert Cliffs Unit 1 Eighth Cycle License Application,"
February 22, 1985.
{." Statistical Combination of Uncertainties Methodology; Part 1;
C-E Calculated Local Power Density and Thermal Margin / Low Pressure LSSS for Calvert Cliffs Units I and II," CEN-124(B)-P, December,1979.
3.
" Statistical Combination of Uncertainties Methodology; Part 2; Combination of System Parameter Uncertainties in Thermal Margin Analyses for Calvert Cliffs Units I and II," CEN-124(8)-P, January 1980.
4
" Statistical Combination of Uncertainties Methodology; Part 3;
C-E Calculated Local Power Density and Departure from Nucleate Boiling Limiting Conditions for Operation for 'Calvert Cliffs Units I and II," CEN-124(B)-P, March 1980 5.
Letter, D. H. Jaffe (NRC) to A. E. Lundvall, Jr. (BGAE), Regarding Unit 1 Cycle 6 License Approval (Amendments 471 to OPR-53 and SER), June 24, 1982.
=
11-5
I r
d References - Chapter 8 1.
Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors, Federal Register, Vol. 39, No.
3, Friday, January 4,1974 2.
Letter, A. E. Lundvall, Jr. (BGAE) to J. R. Miller (NRC), Docket No.
50-317, "Calvert Cliffs Unit 1 Eighth Cycle License Appifcation,"
February 22, 1985.
3.
Letter, D. H. Jaffe (NRC) to A. E. Lundvall, Jr. (BGAE), Docket No.
50-317, " Safety Evaluation of Calvert Cliffs Unit 1 Cycle 8," May 20, 19RS.
4 CENPD-132,
" Calculative Methods for the CE Large Break LOCA Evaluation Model," August 1974 (Proprietary).
CENPD-132, Supplement 1, " Updated Calculative Methods for the CE Large Break LOCA Evaluation Model," December 1974 (Proprietary).
CENPD-132, Supplement 2, " Calculational Methods for the CE Large Break LOCA Evaluation Model," July 1975 (Proprietary).
5.
CENPD-135-P, "STRIKIN-II, A
Cylindrical Geometry Fuel Rod Heat Transfer Program," April 1974
-d CENPO-135-P, Supplement 2-P, "STRIKIN-IT, A Cylindrical Geometry Fuel
- Rod Heat Transfer Program (Modifications)," February 1975.
CENPD-135-P, Supplement 4-P, "STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program," August 1976.
CENPD-135-P, Supplement 5-P, "STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program," April 1977 6.
CEN-161(B)-P, " Improvements to Fuel Evaluation Model," July 1981.
7 Letter, A. E. Lundvall, Jr. (BGAE) to B. C. Rusche (NRC), "Second Cycle License Application," October 1, 1076.
8.
Letter, J. A. Mihalcik (BGAE) to D. H. Jaffe (NRC), "Large Break LOCA ECCS Performance Evaluation," July 25, 1985.
11-6
Refe'rences - Chaoter 9 1.
- Letter, A. E. Lundvall, Jr. (BGAE) to J. R. Miller (NRC),
Docket No.
50-317, "Calvert Cliffs Unit 1 Eighth Cycle License Application,"
February 22, 1935.
2.
Letter, D. H. Jaffe (NRC) to A.
E. Lundvall, Jr. (BGAE)
Docket No.
50-317, " Safety Evaluation of Calvert Cliffs Unit 1 Cycle 8,"
May 20, 19R5.
3.
- Letter, C. O. Thomas (NRC) to A.
E.
Scherer (C-E), " Acceptance for Referencing of Topical Report CENPO-207," November 2, 19R4 4
Letter A. E. Lundvall, Jr. (BG&E) to J. R. Miller (NRC), Docket No.
50-318, " Request for Amendment," (HPSI Flow), April 10, 1985.
5.
- Letter, A. E. Lundvall, Jr. (BGAE) to J. R. Miller (NRC), Docket Nos.
50-317 and 50-318 " Request for Amendment," (Change in the Surveillance Interval for CEA Insartability in Technical Specification 3/4.10.1),
February 26, 1985.
6.
- Letter, A. E. Lundvall, Jr. (BGAE) to J. R. Miller (NRC), Occket Nos.
50-317 A 50-318, " Request for Amendment," (Clad Collapse / Augmentation Factors), December 31, 1984 7.
CENPD-153 P, Revision 1,
" Evaluation of Uncertainties in the Nw: lear Power Peaking Measured by the Self-Powered Fixed In-Core Detector System," May 1980.
9 11-7
f References. Chapter 10 4
.O 1.
- Letter, A.
E.
Lundvall, J r.
(BG8E) to J. R. Miller (NRC), Docket No.
50-317 "Calvert Cliffs Unit 1 Eighth Cycle License Application,"
February 22, 1985.
l l
6 O
l l
11-8 l
_.. _ _ _... _ _ _ _ _ _ _ _. _ _ _ _ _ _ _ _ _ _.. _ _.,, _.. _ _. _ _. _ _. _..