ML20107E164

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Proposed Tech Spec Changes Re HPSI Flow
ML20107E164
Person / Time
Site: Calvert Cliffs Constellation icon.png
Issue date: 02/22/1985
From:
BALTIMORE GAS & ELECTRIC CO.
To:
Shared Package
ML20107E151 List:
References
NUDOCS 8502250403
Download: ML20107E164 (39)
v • d • e


Text

1 9.0 TECHNICAL SPECIFICATIONS The Technical Specification changes which are being requested in order to make the Calvert~ Cliffs Unit 1 Technical Specifications consistent with the analyses contained herein are presented in this section. Table 9-1 presents a summary of the Technical Specification changes, excluding those for HPSI flew reduction, 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 change is being requested.

In addition to the Technical Specification changes summarized in Table 9-1 other changes are being requested to provide increased flexibility for acceptable HPSI flow balance test results. This flexibility has been improved by a reduction in analytic conservatisms which are not controlled by Tech. Specs. and by proposed Tech. Spec. revisions crediting charging pump flow delivery on SIAS and a reduced DNB LCO ASI band. The requested Tech. Spec changes for HPSI flow balance testing are presented in Table 9-2 in the form of an action statement and a reason for each change. Following Table 9-2 the existing Technical Specification page with the intended modification is provided fer each Technical Specification for which a change is being requested.

The overall set of HPSI flow changes are supported by the Excess Load (Section 7.1.4) and Steam Line Rupture (Section 7.3.3) analyses of Chapter 7 and by the Samil Break LOCA analysis of Section 8.2. The Small Break LOCA analysis credits the flow of one charging pump on SIAS. An evaluation of the availability of the assumed charging flow has been made. It was determined that the failure of a single diesel generator (DG) would bound-the effects of any single failure impacting availability of both charging and HPSI flow. A review of the effects of a single DG failure on charging flow determined that with the Technical Specification changes of Table 9.2 in effect it is acceptable to credit the flow of one charging pump.

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Table 9-1 Calvert Cliffs 1 Cycle 8 Technical Specification Changes

, Excluding IIPSI Flow Changes Tcch. Spec.

No. rnd Page' Action Explanation Support 3/4.1.1.1 Change shutdown margin, The shutdown margin is being The following safety analyses page 3/4 1-1 a lowered to accomodate the presented in Chapter 7 support T}$>200/F,from

4. Ak/k to 3.5% Ak/k effects of extended burnup, this reduction:

a) Boron Dilution (Sec. 7.1.1) b) Excess Load (Sec. 7.1.4) c) Steam Line Rupture (Sec. 7.3.2) 3.1.1.4 Change HTC positive limit 4 The HTC is being raised to The follwing safety analyses page 3/4 1-5 Power < 705, from +0 5x10- accomodate the effects of presented in Chapter 7 support this i A k/k/ UF to +0.7x10-g long cycles and to simplify increase ho a k/k/ F. startup procedures, ab a) Loss of Load (Section 7.1.3) b) CEA Ejection (Sec. 7.3.1)

ChangeH{Cnegativelimitfrom The HTC is being lowered to The fo?t sing safety analyses

-2.5x10- Ak/k/ F to to accomodate the effects of preser.ted in Chapter 7 support this

-2.7x10-" Ak/k/ F. of extended burnup. decrease:

a) Excess Load (Sec. 7.1.4) b) Asym. Steam Gen. (Sec. 7.2.4) c) Steam Line Rupture (Sec. 7.3.2) 1 4.1.1.4.2 Hodify HTC surveillance Section "b" is being changed Subsection "b" - clarification

page 3/4 1-6 requirement as indicated to make it identical to the Subsection "c" - provides operational i

same Unit 2 subsection; Section "c" flexibility while preserving the is being changed to permit HTC intent of the subsection.

testing up to 7 EFPDs before reaching 300 PPH.

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Table 9-1 (continued)

Tech. Spec.

No. and Page Action Explanation Support 4.2.1.4 Remove flux peaking Augmentation factors are being 1) Detailed discussion and data was paga 3/4 2-2 augmentation factors removed in recognition of the presented in Reference I 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 pins 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.

Reduce the measurement- This uncertainty is being reduced 1) The new value is supported in calculational uncertainty to conform to the approved value Reference 2.

from 7.0% to 6.2% 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 to a level consistent with existing of the fuel pins presented in expansion factor from calculations and to increase Section 4.3 supports this change.

1.0% to 0.2% operating margin.

Figure 4.2-1 Delete Figure 4.2-1 See change for Tech. Spec. See change for Tech. Spec.

pagn 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.

Figure 3.2-4 Hodify Figure 3.2-4 as The allowable negative ASI is being The ECCS performance analysis for page 3/4 2-11 indicated to restrict the reduced to improve the results of the the small break spectrum presented allowable negative.ASI to ECCS performance analysis for the in Section 8.2 takes credit for this

.10 at full power small break spectrum. improvement in demonstrating accept-able results for reduced IIPSI flow.

3/4.7.1.7 Add the indicated This change will allow entry into An evaluation has been performed whis pago 3/4 7-1 Subsection "~da to the Mode 3 with a minimum number of demonstrates that sufficient relievil Action Statement valves operable to facilitate post- capacity exists in Mode 3 with a ning s overhaul and operability testing of mum of 2 valves per steam generator the remaining valves, operable.

n.__. _ - _ - _ _ _ _ _ _ _ _ - _-- _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ - _ _ _

Table 9-1 (continued)

Tcch. Spec. -

No, and Page Action Explanation Support Table 4.7-1 Change lift setting values The lift setting . values and format 1) The following safety analyses paga 3/4 7-4 and format as indicated for the Hain Steam Safety Valves presented in Chapter 7 support (HSSVs) are being changed to increase this change:

operating margin and to eliminate / a) Loss of Load (Sec. 7.1.3) reduce violations of this Tech. Spec. b) Asymmetric Steam Generator (Sec. 7.2.4)

2) The ECCS performance analysis fo' thesmallbreakspectrumpresenh in Section 8.2 supports this change.
3) The setpoint analysis which verified continued acceptability.

of the present thermal margin Tec

'o Speca supports this change.

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B 3/4.1.1.1 Change EOC shutdown margin, See change for Tech. Spec. 3/4.1.1.1 See change for Tech. Spec. 3/4.1.1.

cud Tav B 3/4.1.1.2 to$>.200F,from4.3%Ak/k 5% Ak/k and change DOC shutdown margin, T >2000F, from4.3%ok/kto3y55 Ak/k B 3/4.2.1 Hemove flux peaking augmenta- See change for Tech. Spec. 4.2.1.4. See change for Tech. Spec. 4.2.1.4.

pago B 3/4 tion factors, change measurement-2-1 ,

calculational uncertainty from 7.05 to 6.25 and change axial fuel densification and thermal expansion factor from 1.01 to 0.2% ,

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Table 9-1 (continued) s Tcch. Spec.

Na, and Page Action Explanation Support D 3/4.2.5 Insert the additional text The BASES section for the DNB LCO The text is merely updating the pag 2 B 3/4 concerning limiting criteria is being expanded to more clearly BASES to describe what has been 2-2 on the DNB LCO, as indicated define all of the criteria which standard practice.

are used to establish the Tech.

Spec. values.

H 3/4.7.1.1 Modify the text concerning See change for Tech. Spec. 3/4.7.1. See change for Tech. Spec. 3/4.7.1 pags B 3/4 Hode 3 operation, as 7-1 indicated

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3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BOS ff W CONTROL t '

SHUTDOWN MARGIN - T,yg > 2004

! LIMITING CONDITION FOR OPERATION 3.5 3.1.1.1 The SHUTDOWN MARGIN shall be > ak/k. l APPLICABILITY: MODES 1, 2**, 3 and 4.

ACTION: 3,g With the SHUTDOWN MARGIN < ak/k, immediately initiate and continue i boration at > 40 gpm of 2300 ppm boric acid solution or equivalent until the required SHUTDOWN MARGIN is restored.

SURVEILLANCEREdUIREMENTS

1L S 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be > 4. %* ak/k: _ l
a. Within one hour after detection of an inoperable CEA(s) and at 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 inoperable. ,

. If the inoperable CEA is innovable or untrippable,. the above

. required SHUTDOWN MARGIN shall be increased by an amount at
least equal to the withdrawn worth of the imovable or untrippable CEA(s).
b. 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 that CEA group withdrawal is within the Transient Insertion Limits of Specification 3.1.3.6. ,

c.

N When in MODE 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 criticality by verifying that the predicted critical CEA position is within the limits of Specification 3.1.3.6.

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d. Prior to initia'l operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors of a below, with the CEA groups at the Transient Insertion Limits of Specification 3.1.3.6. - -

.

analyses.

    • See Special Test Exception 3.10.1.
  1. With X,ff > 1.0.

H With r.,ff < 1.,0.

CALVERT CLIFFS - UNIT 1 3/4 l-1 Amendment No. 2J N J/, f/t : -

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.__- - _-6-9

REACTIVITY CONTROL SYSTEMS MODERATORTEN5i56Id$' COEFFICIENT LIMITING CONDITION FOR OPERATION 3.1.1.4 The moderator temperature coefficient (MTC) shall be:

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a. Less positive than x 10-4 ok/k/*F whenever THERMAL POWER is <_ 70% of RA ED THERMAL POWER,
b. Less positive than 0.2 x 10-4 ok/k/*F whenever THERMAL '

j POWER is > 70% of RATED, THERMA!. POWER, and

c. Less negative than . x 10-4 Ak/k/*F at RATED THERMAL POWER.

93.7 APPLICABILITY: MODES 1 and 2*f i ACTION: 3,l l With the moderator temperature coefficient outside any one of the above limits, be in at least HOT STANDBY within f hours.

i SURVEILLANCE REOUIREMENTS

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4.1.1. 4.1 The MTC shall be determined to be within its limits by confirmatory measurements. MTC measured values shall be extrapolated

- and/or compensated to permit direct comparison with the above limits.

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i fSee Special Test Exception 3.10.2.

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CALVERT CLIFFS - UNIT 1 3/4 l'-5 Amendment No, M. 8 3 c ..

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I REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

'4.1.1.4.2 The MTC shall be determined at the following frequencies and THERMAL POWER conditions during,each fuel cycle:

a. Prior to initial operation above 5% of RATED THERMAL POWER, after each fuel loading.

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b. At any THERMAL POWER above 90% of RATED THERMAL POWER, within 7 EFPD after initially reaching an equilibrium conditio at or '

above 90% of RATED THERMAL POWERg a/&.4*cA M .

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. c. At any THERMAL POWER, within 7 EFPD after reaching a RATED THERMAL POWER equilibrium boron concentration of 300 ppm.

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e CALVERT CLIFFS - UNIT 1 3/4 1-6 Amendment No. 8 3 9-3

POWER DISTRI30 TION-LIMITS SURVEILLANCE REGUIREMENTS (Continuedl

. c. V.erifying at least once per 31 days that the AXIAL SHAPE INDEX is

. maintained within the limits of Ft.;ure 3.2-t, where 160 percent ,

of the allowable power represents the maximum THERMAL POWER allowed by the following expression:

MxN

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. where:

1. M is the maximum allowable THERMAL POWER level for the extsting Reactor Coolant Pump com51 nation.
2. N is the maximum allowa5le fraction cf RATED THERMAL POWER as determined 5y the F[y curve of Figure 3.2-3b. l 4.2.1.4 Incore Detector Monitorina System - The incere detector moni- .

toring system may be useo for monitoring tne core power distribution by verifyirg that the incere detector Local Power Density alarms: _

a. Are adjusted to satisfy the requ.frements of the core power distribution map which shall Be updated at least once per 31 days of accumulated operation in MODE 1. ,

3

b. Have their alarm setpoint adjusted to less than or equal 'to the limits shown on Figure 3.2-1 when the following factors are appropriately included in the setting of these alams: .
1. "h: - :" ; " :~ -t:t?;; f::t;.; m .:.;.; '
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A measurement-calculational uncertainty factor of

[ An engineering uncertainty factor of 1.03,

' /.003 l [ A linear heat rate uncertainty factor of 'due to axial l fuel densification and themal expansion, and A THERMAL POWER measurement uncertainty factor of 1.02~ .

l CALVERT CLIFFS - UNIT 1 3/4 2-2 Amendment No. 27, 24, 32, 33 , # , 7 1 1 9-9 .-

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! l CALVERT CLIFFS 'Ji!IT 1 3/4 2-11 tcendnent No. 43 71 i-l 9-11

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3/4.7 PLANT SYSTEMS 3.4.7.1 TURBINE CYCLE SAFETY VALVES LIMITING CONDITION FOR OPERATION 3.7.1.1 All main steam line code safety valves shall be OPERABLE.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

a. With both reactor coolant loops and associated steam generators in operation and with one or more main steam line code safety valves inoperable, operation in MODES 1, 2 and 3 may proceed provided, that within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, either the inoperable valve is restored to OPERABLE status or the Power Level-High trip set-point is reduced per Table 3.7-1; otherwise, be in.at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within f, ~ ; '. -

the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. .

b. With one reactor coolant loop and associated steam generator in operation and with one or more main steam line code safety-valves associated with the operating steam generator inoperable, operation in MODES 1, 2 and 3 may proceed provided:
1. That at least 2 main steam line code safety valves on l

the non-operating steam generator are OPERABLE, and i

2. That within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, either the incperable valve-is restored to OPERABLE status or the Power Level-High trip I setpoint is reduced per Table 3.7-2; otherwise, be in at least HOT STANOBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

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c. The provisions of Specification 3.0.4 are not applicable.

I J.  % Q .iZras&24eM M006 3 2W &MAS hM-M O f l SURVE1LLANCE REOUIREMENTS 4.7.1.1 No additional Surveillance Requireft.ents other than those required i

g by Specification 4.0.5 are applicable for the main steam line code safety l- valves-of Table 4.7-1.

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! CALVERT CLIFFS-UtiIT 1 3/4 7-4 Amendrent .*;o. 21 '

i 9-13

. 1 3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 BORATION CONTROL 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 subcritical 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 suberitical to preclude inadvertent criticality in the shutdown condition.

3 .

SHUTDOWN MARGIN requirgments vary throughout core life as a function of fuel depletion, RCS baron c5ncentration and RCS T The minimum available 3.f j SHUTDOWN MARGIN for no load eratingconditions$59beginning of life is .

Ak/k and at end of life is 4. 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 rupture ent at beginning of cycle conditions, a minimum SHUTDOWN MARGIN of less than 4. ak/k is required to control the ,

reactivity transient, and end of cycle onditions require 4.'% ak/k. Accordinglyd the SHUTDOWN MARGIN requirement is be ed upon this limiting co on and is 7 consistent with FSAR safety analysis ssumptions. With T 200 F, m -3,5,l reactivity transients resulting from any postulated accid $E <are minimal and a ,

k.' 3% Ak/k shutdown margin provides ade uate protection. , With the pressurizer

- l level less than 90 inches, the sourc s of non-borated water are restricted to '

a boren dilution event.

increase the time to criticality duri 3/4.1.1.3 BORON DILUTION.

A minimum flow rate of at least 3000 GPM provides adequate mixing, i prevents stratification and ensures that reactivity changes will be ,

gradual during boron concentration reductions in the Reactor Coolant '

System. A flow rate of at least 3000 GPM will circulate an equivalent ,

Reactor Coolant System volume of 9,601 cubic feet in approximately '

24 minutes. The reactivity change rate associated with baron concen-tration reductions will therefore be within the capability of operator l recognition and control.

3/4.1.1.4 MODERATOR TEMPERATURE COEFFICIENT (MTC)

The limitations on MTC are provided to ensure that the assumptions '

used in the accident and transient analyses remain valid through each l fuel cycle. The surveillance requirements for measurement of the MTC ,

during each fuel cycle are adequate to confirm the MTC value since this coefficient changes slowly due principally to the reduction in RCS boren concentration associated with fuel burnup. The confimation that the 1

measured MTC value is within its limit provides assurances that the coefficient will be maintained within acceptable values throughout each fuel cycle.

l CALVERT CLIFFS - UNIT 1 8 3/4 1-1 ,

Amendment No. 27, 32.42.// d

3/4.2 POWEiFUE TRIBUTION LIMITS BASES 3/4.2.1 LINEAR HEAT RATE The limitation on linear heat rate ensures 'that in the event of a LOCA, the peak temperature of the fuel cladding will not exceed 2200*F.

Either of the two core power distribution monitoring systems, the

- Excore Detector Monit.oring System and the Incore Detector Monitoring .

' System, provide adequate monitoring 6f the core power distribution and are .,

capable of verifying that the linear fieat ra_te does not. exceed its limits.

The Excore Detector Monitoring System perfoms this function by continu-ously monitoring the AXIAL SHAPE INDEX with the OPERABLE quadrant symetric excore neutron flux detectors and verifying that the AXIAL SHAPE INDEX is maintained within the allowable limits of Figure 3.2-2. In conjunction with the use of the excore monitoring system and in establishing the AXIAL SHAPE INDEX limits, the following assumptions are made: 1) the CEA .

insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are satisfied '

the 'L gnin; rrent: tie 'r'-~ re r : 5- '- Fi; . m e. '

ne A IMUTHAL POWER TILT restrictions of Spectftcation 3.2.4 are satisfied, and

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the TOTAL PLANAR RADIAL PEAKING FACTOR does not exceed the limits of pecification 3.2.2.

4 The Incore Detector Monitoring System continuously provides a direct measure of the peaking factors and the alarms which have been established for the individual incere detector segments ensure that the peak linear heat rates will be maintained within the allowable limits of Figure 3.2-1.

The setpoints for these alarms include allowances, set in the conservativ .04 directi ns, for Tl _ n:P' ; r;=ntti;..f::tn a 2 x ' -' n

. i ? a measurement-calculational uncertainty factor of Gr5,(Fran 4 l' engi eering uncertainty factor of 1.03, an allowance of 1. 1 for axial fu densification and themal expans , and un inty factor of 1.02.

a THERMAL uWER measurement g

3/4.2.2. 3/4.2.3 and 3/d.2.4 TOTAL PLANAR AND INTEGRATED RADIAL PE FACTORS-F[y AND Fj AND AZIMUTHAL POWER TILT - T, T

' The limitations on F and T are provided to ensure that the assump-

!' tions used i.n the analysisYfor es!ablishing the Linear Heat Rate and Local Power Density - High LCOs and LSSS setpoints remain valid during coeration a the various allowable CEA group insertien limits. The limitations on y and T qare provided to ensure that the assumptions used in

ALVERT CLIFFS - UNIT 1 33/42-1 Amendment No. 33, 39 l lALVERT CLIFFS - L'::IT 2 . ..:c.W.cnt ..a . 13, 2c 9-15 '~

. . i 1

POWER DISTRIBUTION LIMITS BASES the analysis establishing the DNS Margin LCO, and Thermal Margin / Low Pressure LSSS setpoints remain valid during7 operation at the various o allowable CEA grouc insertion limits. If F

~ F' or T exceed their basic limitations, operation may coni;inue uEer Ehe Td0itioital'restric- #

tiens imposed by the ACTION statements since these additional restric-tions provide adequate provisions to assure that the assumptions used in establishing the Linear Heat Rate, Thermal Margin / Low Pressure and Local Power Density - High LCOs and LSSS setpoints remain valid. An AZIMUTHAL POWER TILT. > 0.10 .i.s_ not e,xpected and if it should occur, sub-sequent operation would be restricted. to only those operations required to identify the cause of this unexpected ti.lt. .

The value of T T

and.F 7

=F 7 (1+T q

) is the measured tilt.that must be used in the9equation The surveillance requirements for verifying that FT 77r and T ge within their limits provide assurance that the actual vMu,es of FT and T do not exceed the assumed values. Verifying F' andF'af8rS r#

each fuel loading prior to exceeding 75% of RATED THEbl POWER provides additional assurance that the core was properly loaded. .

C4 3/4.2.5 DNB PARAMETERS The limits on the DNS related parameters assure that each of the -

parameters are maintained within the nomal steady state envelope of operation assumed in the transient and accident analyses. The limits are ,.

l consistent with the safety analyses assumpticris and have been analytically i demonstrated adequate to maintain a minimum DNBR of 1.23 throughout each I analyzed transip lNSEAT NEW ?)1AA& RAP)f @ //ag g g a m- n' 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 through instru-ment readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation. The 18 month periodic measurement of the RCS total flow rate is adequate to detect flow degradation and ensure correlation of the -

flow indication channels with measured flow such that the indicated percent flow will provide sufficient verification of flow rate on a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> basis.

CALVERT CLIFFS - UNIT 1 B 3/4 2-2 Amendment No. q, gg, {3, 71 9-16 -

l

l In addition to the DNB criteria, there are two other criteria which set the 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 cr'.teria for this condition because the linear heat l l

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 A00s (e.g., CEA withdrawal),

tend to become more limiting than that for LOCA.

! (

't I

i l

l r

f e

9

-~ _ _ _ . _

g.17

1 The as-left lift settings will be no less

! ~

{

than 985 psig to insure that the lift setpoints will remain within specification during the cycle.

3/4.7 PLANT SYSTEMS L -

{

4 BASES

\ ,

! 3/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAFETY VALVES U* 8 The_ OPERA 8ILITY of the main steam line code safety valvds ensures l 1

!, that the secondary system pressure will be limite to within*its design l pressure of 1000 psig during the most severe anti ipated system opera-1

. tional transient.4 The maximum relieving capacit .is associated with a

( turbine trip from 100% RATED THERMAL POWER coi loss of condenser heat i k (i.e., no st, gay)jf ident with an assumed ass W .to the condenser).%

a 4 44sm h 4s(s waved m.a@ -

j

{

\

P: perf'*ed m ; .ift ee++f-~' ...d car r;';; u n ftf- -

accordance with the requiremenJs-t#SeetiondIJof the ASME Boiler and in g PressureKode,1971 Edition.TfThe total relieving capacity for all valves

(on ati or une sr.eam lines is 12.18 x 108 lbs/hr hich i: 102 ;;. . ..; wi q the t t;l rrrrtr; :trr 'M :f !!.2^.i 4 iGC lb./:.. at 100% RATED
j. ' .

DIERMAL POWEV7 -"' r :? I :.";.uc;.; ..i.s., ,J .ea p.c ;:nr. ;r r:ter

ear'2rer t.ht =f'ici=0 rei:.,:..., ;;;;;it; i: r;;il;hl: f;r rc.-c.ia; l
j_ * -InssAT N6W FAMGMPH @ HEMf@ w oN NEXT PAGE_

STARTUP and/or POWER OPERATION is allowable with safety valves

! k ',. incperable within the limitations of the ACTION requirements on the basis of the reduction in secondary system steam flow and THERMAL POWER required

by the reduced reactor trip settings of the Power Level-High channels.

I The reactor trip setpoint reductions are derived on the following bases:

For two loop operation Sp , (X) - (Y)/V) x '106.5 X -

i For single loop operation (two reactor coolant pumps operating in the same loop)

Sp , (X) - (Y)(U) x 46.8

! X 1 . where:

SP. = -reduced reactor trip setpoint in percent of RATED THERMAL POWER '

=

] V maximum number of inoperable safety valves per steam p .

line y

k ' -

} CALVERT CLIFFS - UNIT 1 B 3/4 7 1 4

9-13 .

>>--m,- ,,--..._._-,na,,..n. . _ _ . _ e--, Le, ,_ ,.,,n.,.m.,.-,-_m,,.w,,,,, ,-,,,.,_,,,,4__,.e --

,.n-,-,,.o,-,

i In tiode 3, two main steam safety valves are required operable per steam generator.

These valves will provide adequate relieving capacity for removal of both decay heat and reactor coolant pump heat from the reactor coolant system via either of -

the two steam generators. This requirement is provided to facilitate the post-overhaul setting and operability testing of the safety valves which can only be conducted when the RCS is at or above 500 F. It allows entry into tiode 3 with a minimum number of main steam safety valves operable so that the set pressure for the remaining valves can be adjusted in the plant. This is the most accurate means for adjusting safety valve set pressures since the valves will be in ther-mal equilibrium with the operating environment.

l l

l-l r* -

l l

'^

9-19 .

l

Tchle 0-2 Calvart Cliffs 1 Cycle 8 Technical Specification Changes Concerning IIPSI Flow Tech. Spec.

No. and Page Action Explanation 3/4.1.2.2 a) Add statement clarifying requirements for a) Simply a clarifying statement.

page 3/41-9 operable flow paths from the boric acid storage tanks.

b) Delete the requirements that testing b) This restriction is unnecessary since be performed during shutdown. this test can be performed at any time.

c) Add requirement for demonstrating that c) The boric acid pump can be part of a each boric acid pump starts on SIAS. flow-path which must be operable for the charging pump, in turn, to be operable following SI AS. This addition is necessary to support the crediting of charging pump flow in the Small Break LOCA analysis.

'? 4.1.2.4 Add requirement that charging pumps start This addition is necessary to support the El page 3/41-11 on a SIAS test signal. crediting of charging pump flow in the Small Break LOCA analysis.

4.1.2.6 Add requirement that Tech. Spec. Surveillance A(kts a cross-reference.

page 3/41-13 Requirement 4.1.2.2 be observed.

3.1.2.8 a) Change combinations of water sources, as These changes are being made to assure that a page 3/41-16 indicated, borated water source will be available to the b) Delete applicability to Modes 1 below 80% charging pump following SIAS while the reactor power,2, 3, and 4. is in Mode 1 above 80% power. Such changes are c) Change action statement to recognize necessary to support the crediting of charging pump different combinations of water sources and flow in the Small Break I,0CA analysis.

the sufficiency of reducing power to 80%.

3 /4.1.2.9 a) Change Tech. Spec. 3/4.1.2.8 to 3/4.1.2.9. The changes for Tech. Spec. 3.1.2.8 are not new page b) Change applicability to Mode 1 below 80% necessary for operation in Mode 1 below 80% power 3/4 1-16u power. and Modes 2,3 and 4. Consequently, the old Tech.

Spee. 3/4.1.2.8 is simply being renumbered to 3/4.1.2.9, its applicability to Mode 1 is being altered and its page number is being changed to 3/41-16u.

l .

Table 9-2 (continued)

Tech. Spec. Explanntion No. and Page Action a) ' Clarify maintenance actions which require a) Some maintenance actions will not affect flow 4.5.2 (h & i) characteristics (such as motor resistance checks).

page 3/4 5-Sa verification of valve stem travel. b) This change will provide improved flexibility.in the 7

- b) Change the method of verification of surveillance testing of IIPSI flow without affecting  ?

sufficient ilPSI flow from certifying the integrity of the safety analyses which credit a minimum flow for each injection leg to certifying that the sum of the flow IIPSI flow. '

from the three legs with the lowest flows meets a minimum value, c) This change provides additional margin for verify g'7 c) Lower the total required IIPSI flow. ing that sufficient flow exists; it is supported ~9

~

by the safety analyses presented herein. 3 d) This addition formalizes a surveillance requirement

  • d) Add Subsection "i" which adds a require- for equipment whose proper function is necessary ment that the llPSis deliver a minimum to preserve the assumptions used in the safety pressure head when tested. analyses.

c) This limit prevents the pump from exceeding runout c) Delete the upper limit on IIPSI pump flow. -conditions. Detailed analyses have been performed

,o 43 in order to arrive at the maximum expected IIPSI

~

pump flow with all four (4) flow control valves wide open and the RCS at atmospheric conditions. These ; '

analyses show that the pump will not runout; therefore, the upper limit has been removed.

Testing will be conducted at the beginning of the Unit 1 outage to confirm the mechanical and electrical characteristics of the pump.

The Small Dreak LOCA analysis of Section 8.2 Revise the BASES to reflect the crediting credits charging pump flow to improve the results -

B 3/4.1.2 page B 3/4 of charging pump flow in the Small Dreak of the analysis and permit the lowering of the 1-2 LOCA analysis. required llPSI flow (see change for Tech. Spec.

4.5.2 (h & i) b).

a) Simply an editorial change separating all a) Move the discussion concerning TSP B 3/4.5.2 from the end of the second paragraph discussion of TSP in the BASES from discussions /

page B 3/4 concerning ECCS subsystem performance. .

l to the end of the first paragraph. b) See the change for Tech. Spec. B 3/4.1.2.

5-2 b) Revise the Bases to reflect the ,

crediting of charging pump flow in the Small Break LOCA analysis.

.. e REACTIVITY CONTROL'S b eds.

FLO'A PATHS - OPERATING 1 y gpmjic1 .

m g.gfpc4. 3.I. 3.t y , m- [

{', _

. . LIMITING CONDITION FOR.0PERATION. . . . . . . - .

- \

3.1.2.2 At least two of the folicwing three boren infection ficw paths- .

and one , associated heat tracing. circuit shal.1 be OPERAELE: . .

Ca. Two flow paths from the horic acid stcrage tanks ia either a

' boric acid pu=p or a gravity feed connectien, and a charging pump to the Reacter Ccolant Systam, and -

~

b. The ficw path f cm the refueling watar tank via a charging pums to the Reacter Ccolant System. -

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION: .

I Nith only one of the above required boren injection ficw paths to the Reactor Coolant System OPERABLEi restere at least two boren injection ficw paths to the Reacter Coolant System to OPERABLE status within 72

( hcurs or be in at least HOT STAND 3Y and berated to a SHUTCO'AN PARGIN '

I equivalen tc~ at least .3 Ak/k at 200*F within the next 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />; restore at least two flow paths to OPERABLE stat =s within the nex: 7 days cr be in COLD 5,HUTD0'4N within the next 30 hcurs. -

{}

~

l SURVEILLANCE REOUIREMENTS . . .. . .

l -

1 l 4.1.2.2 At least tko of the above required flew paths shall be demonstrated OPERAELE: -

a. At least once per 7 days by verifying that the tamperature of the heat traced portien of the ficw path from the concentratad boric acid tanks is above the. temperature limit line shcwn an Figure. 3.1-1.
b. At least once per 31- days by verifying that each valve (manual, power operated or. automatic) in the ficw p'a th that is not 1ccked, sealed, or otherwise secured in pcsitien, is in its cor ect pesition. ,
-- ~

. c.: At :1 east once per 18 months l_. ' m- ' _ ^ = by verifying that :

! ,c)* tach automatic valve in the ficw path actua:as s its cor sc:

position on a SIAS test signalj * .d. -

2)M Jeth Mbl f'yS-9 CALVERT CLIFF 5 - UNIT 1 3/4,1-g- Amenc=en: Ne .18

- - . . - -._h.--.. . _ - _ _ _ .

l * .

l

  • l REACTIVITY CONTROL SYs e!S .

(y CHARGING PUMPS - OPERATING ,

LIMITING CONDITION.FOR:0PERATION.

l .

l .. .

3.1.2.4 At least two charging pumps shall be OPERAELE.

l APPLICABILITY: MODES 1, 2, 3 and 4. -

l l ACTION.

l -

l With only one charging pump OPERABLE, restore at least two charging

pumps to- OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY i and borated to a . SHUTDOWN MARGIN equivalent to at least3 % Ak/k at l l 200*F within the 6 ext 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore at least two charging pumps to
OPERABLE status within the next 7 days or be in CCLD SHUTDOWN within the .

next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. -

x~- -

. SURVEILLANCE REOUIREMENTS gd;t.katH dkep*yp j""^4*M.34 h==TM CGMBLE: .

4.1.2.fNo additional Surveillance Requirements other than those requirep Gy Soecification 4.0.5.

' .NY .

N a.'42y.bu;t mo yytrA&u-&Q t y, .446 7 % 'deh%

N -

s. .

L -

CALVERT CLIFFS - UNIT 1 3/4 1-11 Amenv. ment No. 48 9 -2,1

I REACTIVITY CONTROL SYSTEMS l

(, BORIC ACID PUMPS - OPERATING -

1 LIMITING CONDITION FOR OPERATION .

' ~

i 3.1.2.5 ' At least the beric acid pumo(s) irt the baron injection flow path (s) required OPERABLE pursuant to Specification 3.1.2.2a shall be OPERABLE and capable of being powered frem an OPERABLE emergency bus if

~

i the flow path through the boric acid pump (s) in Specification 3.1.2.2a

is OPERABLE. -

APPLICABILITY: MODES 1, 2, 3~ and 4. -

ACTION: .- . .

With one horic acid pump required for the boron injection flo v path (s) '

pursuant to Specification 31.2.2a inoperable, restore the boric acid pump to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY

' within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and borated to a SHUTDOWN MARGIN equivalent to at least'.W Ak/k at 200*F; restore the above required. boric acid pump (s) l

to OPERABLE status within the next 7 days or be in COLD SHUTE 0WN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. ,

((:

~

SURVEILLANCE REGUIREMENTS.

4.1.2.6 No additional Surveillance Requirements other than those required by Specification,4.0.Sg M +,1,g,g, L

-J CALVERT CLIFFS - UNIT I 3/4 1-13 Amenc=an- No . 18

' ~

. 9-24

u.u

....... . ..... o .

BORATED WATER SCURCES - GPERATING . ..

LIMITING CONCITICN 20R OPERATICM

~~~

w H M-:%cf 3.l'2.8 At least two of the following three berated water scurcas shall be OPERABLE:

a. Two beric acid st: rage tank (s). and one ass: cia:ad heat tracing cir:uit cer tank with the contents of :he tanks in ac::rdanca et. wi-h Figure 3.1-1 and the beren c ncentration lici:ad :: < 35, -

.,asede. dad.& W I2. TCMb"b f u'AWS 3.1.2.E a. &

b. /The refueling water tank with:
1. A minimum centained berated wa:ar volume f 400,C00 gallens,
2. A beren c:ncentration of betwaan .2200 and 2700 P;m.. l
3. A minimum solution temperature of 40'F, and 4 A maximum-selH ~' 'a- a-=-'ra cT100*F in MOCE l. -

(> 90%4 Chad TH&aMM s'W.w APPLICASILITY: M00@lpT::6--8::~$p f. S g g cAf. N <

c

. w,-3 4^ Mw ~r w'lN .- W w ACTTCN:

/* & M k With (cniv er3 tera:ad wa ar source 40PERA3LE, restore at leas:":.>. .:.-. :-d s

k f

~2.

.::: :c.r:::- :: GPERA3LE sta:us within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or ': " :: ;;;; :~7 '

'-':: :'l.,,wi thin the nex: E hours.a-d t:r:::d :: : ::.C 7 ','.... . a P." ;.6 % . . _c. :

___ ... . .. .. nee.

u. ...a ...- .
.:. . a.

, . . m_ _ ~ar

......--...~..~.......~..~....:-

.a .

! ,-': ~~~~ ** '.':.

\l --- & p'MGAt1Al.

RATfD 1 POWSBm frj u4 % Po % c) phu I kaa eHLD Q Ne :~JMJud or M ae An

  • Lefes* go% A fur yygg,uge. )w6A aa co sty wWe

\ 9p,:n is. s. t. z 9 _

@ With oniy cne :cra:ec wa:er scurta CPEV ELE, .es::re at leas: twc) l

.t-:. . - :: :..) crated .. -

I water scur:as t: GPERA5LE statu s ' 7" 5;. :.:. '-

i

~- . _ . .. . .

.eu. r :". r. . r_ LmN C.. s' .. '" I n...'m. ."u .*.C - .- ---- _. . _ . . _ . _ _

".l.2.8 A

least -We beratac wa:ar seur:as shali :e demens:ra:ac 0; ERA 8LE:

l -

a. A: least ence :er 7 days by: ,

1.

3

'/erifying :he bcr:n c:ncan:ra-icn in each wa:3r gege':3, ,

l 2. 'lerifying :na ::n ained berated -a a. volume in sacn na:6r scur:3, and '

3.

'Ierifying :he beri: acid s:Ortga :ank sciu:icn :am: era:ure.

. A- leas: Once :er 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> :y veri fyine. :h.a RWT

..=.....o.. 3 .-..r.

when :ne OU:sica air tam: era *ure is < 40',0 .

l '.el.,

t

. L-. a C.-. . r . s - w h. o 3

3,. . .. 1-s-.-a - 'n  ? ..

1 1 -

'1 l

l 1

REACTIVITY IONTRCL SYSTEMS SORATED WATER SOURCES - OPERATING i I

i LIMITING CONOITION FOR OPERATION i T

3.1.S At least two of the follcwing three berated water sources shall i be OPERABLE:

a. Two beric acid storage tank (s) and one asscciated heat tracing cir:ui: per tank witn the c:n:en:s of tne tanks in ac::rcance with F.icure 3.1-1 and tne boron c:ncentration limited c < 6%,

and

b. The refueling water tank with: .
1. A minimum contained borated water volume of 400,000 gallons, ' .
2. A beren c:ncentration of betwaen 2.300 and 2700 ppm, .

l

3. A minimum solutten temperature of 40*F, and A.' A maximum solution temperature of 100'F in MODE 1.

1 AFPLICA5ILITY: MODES. 2, 3 and 4 C

ACTION:

8

- With only one berated water source OPERABLE, restere at least two berated water scur es to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or ce in at isast HOT 4

STANOBY within the next.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and borated to a SHUT 00WN MARGIN equivalent

~

o at least 35 sk/k at 200*F; restore at least two borated. water sources to CPERASLE' status within the nex: 7 days or be in COLD SHUT 00WN within ,

the nex: 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

. SUEVEILL.4NCE REQUIREMENTS t.1.2(3 At leas: two berated water scurces shall be demonstrated OPERABLE:

1 a. At leas: once per 7 days by:

1. ' Verifying the beren concentra:icn in each water scur:e,
2. Verifying the contained berated water volume in each water. scurce , and ' . ..

Verifying thes beric acid st: rage ank solu:icn temperature.

3.

b. A: leas: ence per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying :he RWT :empera:ure y_. when the outside air tem:erature is < dO'F.

CALVERT CLIFFS - UNIT 1 ,

A**".

... *;*_*_ "_ I N

  • 4 0 ' UMI" mm. . ; ,; . .....n..

3 / *4 i -16.P -_ . . . .

f

, q3) t t A+ 4 sate e4 94TGO; 7NGRMM MA .

- 9-26

._ i

l 1

EMERGENCY CORE COOLING SYSTEMS , *

(

't inn $ M 4 9 fts e :vysfew SURVEILLANCE RECUIREMENTS (Continued) -

    • "* '* */* // <5 -

l

/ .

2. Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> followi g ccmoletion of maintenance Yn the valve or its ccerato my measurement of stem travel when the ECCS subsystems are required'te be OPE?A3LE. -

-HRSI SYSTEM Valve Nu=her Valve Number MOV-616 MOV-617 MOV-626 -

MOV-627

.- MOV-636 MOV-637 MOV-646 MOV-647 I

. I <

h. By perfeming a flow balance test during shutdewn following ecmpletion of HPSI system mcdifications that alter system flew characteristics and verifying the fellowing ficw rates @

. W a..&~$f.s KPSI M 4p4A% : -

nr : '"< am Sinole Pu=o

{

,,.... so each injection . .

170 + -

t..cw 7 n+rrf A y . A .c M W J'PNAM p'2AA4s. M h M 7 -

g e++2= -

. .,a p' '

N;du flPSZ f" M P~y4 R

& #~$ *

, ,u. .

e.,m. "" . o. r.

i ppm. .

o y

( .

l .

l 1

q

.CALVERT CLIFF 5 - UNIT 1 . 3/4 5-5a Amendment No.M. 75 [

j 9-27 -

, , - , - - - , - - --,m.--,-. ,....._.-,,,---,__.r,., -

'A & &** W W W u s u s G.S.,& J~B44A LaMhe

$4v$ae~A h/clV @ Y *k% * &

REACTIVITY CONTROL SYSTEMS 4.a d A M AG pft BASES w 3

" *~

- r 3/4.1.1.5 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be.made crijical with the Reactor Coolant System average temperature less than 515 F. This limitation is required to ensure 1) the moderator temperature coefficient is within its analyzed temperature range, 2) the protective instrumentation is within its normal. operating range, 3) the pressuri er is capable of being in an OPERABLE status. with a steam bubble, and 4) the reactor pressure vessel is above its minimum RT emperature.

NDT 3 /4.1.2 BORATION SYSTEMS The baron injection system ensures that negative raactivity control is available during each made of facility operation. (The components required to perform this function include 1) borated water sources, 2) charging pumps, 3) separate flow paths, 4) boric acid pumps, 5). associated heat tracing systems, and 6) an emergency power supply from OPERABLE diesel generators. ,

With the RCS average temperature above 200 F, a minimum of two

( . separate and redundant baron injection systems are provided to ensure single functional capability in the event an assumed failure renders one of the systems inoperable. Allowable out-of-service periods ensure that minor component repair or corrective action may be completed without undue risk to overall facility safety from injection system failures

,during the repair period.

~

The boration capability of either system is sufficient to , provide a

  • I.

~ SHUTDOWN MARGIN from all gperating conditions of 3.0". ak/k after xenon decay and cooldown to 200 F. The maximum boration capability requirement occurs at EOL from full power equilibrium xenon conditions and requires -

5500 gallons of 7.25% boric acid solution from the boric acid tanks or 55,627 gallons of 2300 pp'm borated water from the refueling water tank. However, to be consistent with the ECCS requirements, the RWT is

. required to have a minimum contained volume of 400,000 gallons during MODES 1, 2, 3 and 4 The maximum baron concentration of the refueling water tank shall be limited to 2700 ppm and the maximum baron concentra -

tion of the boric acid storage tanks shall be limited to 8% to preclude the possibility of boron precipitation in the core during long term -

ECCS cooling. -

With the RCS temperature below 200 F, one injection system is acceptable without single failure consideration on the basis of the stable reactivity conditi6n of the reactor and the additional restric-tions prohibiting CORE ALTERATIONS and positive reactivity change in the

{ event the single injection system becomes inoperable.

CALVERT CLIFFS - UNIT 1 -

Amendment No. 27. /B, bf CALVERT CLIFFS - UNIT 2 3 3/4 1-2 Amendment no. 31 .

9-2S

D w- = MF3T p- r m--- ..

  • d sh %'t. N t}%Gk *g-

-->~ h-~ n - nnwp syg, Wqm maps M. 3a. c!CA Aalt.lTY ds An M M

.jffe w p d ' k S

&h&

yS7g ,y, EMERGENCY CO L BASES l

The trisodium phosphate dodecanydrate (TSP) stored in dissolving i baskets located in the centainment basesment is provided to minimi:e the possibility of corresien cracking of certain metal ccmcenents during operation of the ECC5 following a LOCA. The TSF provides this protec:icn by dissciving in the sump water and causing its fjnal pH t'o be raised to

> 7. 0. .

,k The Sur/eillance. ' Requirements provided to ensure OPERASILITY of

~. y each compenent ensures that at a minimu=, the assumstions used in the j safety analyses are met and that subsystem OPERASILITY is maintained.

f Surveillance requirMents for throttle valve position steps and ficw balance testing prc/.de assurance that prcper ECCS flows will be main-l tained in the event cf a LOCA. Maintenance of prcper ficw resistance and pressure drcp in the piping system to each injecticn point is necessary j to: (1) prevent total pump flow frca exceeding runaut ccnditiens when f the system is in its =inimum resistance configuration, (2) provide the t

proper flow split between injection pointe in acccrdance Qith the assumptions used in the ECCS-LOCA analyses, and (3) crevide an accectabla f A level of total ECCS ~ficw to all ~njection coints ecuzi to or aceve tna:

I assumed in the ECCS-LOCA analyses. fThe requirement to c1sscive a repre-tsentatu e sampie c or in a sample of RWT water provides assurance that i

the stored TSP will dissolve in borated water at the postulated pcst LOCA te.moeratures. _

3/4.5.4 REFUELING WATER TANK (RUT) i The OPERASILITY of the RWT as part of the ECCS ensures that a sufficient supply of borated water is available for injection by the ~

ECCS in the event of a LCCA. The limits on RUT minimum volume and boren concentration ensure that 1) sufficient water is available within contain-ment to 'per .it recirculatien cooling flew to the core, and 2) the reac:cr l

will remain subcritical in the cold condition following mixing of the RWT the mosanc the RCS uater volumes with all centrcl rces inserted exceo: fer reactive centrol assemoly. These assumptiens are consistad:

l with the LOCA analyses.

The centained water volume lici: includes an allcwance for water not usable because of, tank discharge line 1ccatien er c:her pnysical characteristics.

i l

o CALVERT C'.*FFI - UNIT i R 2/* ~-2 l- e - " - " " '-

10.0 STARTUP TESTING The startup testing program proposed for Cycle 8 is identical to the program proposed for the reference cycle in Reference 1, except that CEA 5-1, due to its small worth, will not be used for reactivity control and maintaining power. All CEAs in Group 5 or the 4 peripheral Group 5 CEAs will be used instead.

e t '. :- 1

11.0 REFERENCES

References - Chacters 1 Through 3 1 Letter, A. E. Lundvall, Jr. (BG&E) to J. R. Miller (NRC), "Calvert Cliffs Unit 1 supplement 1 to Seventh Cycle License Application,"

September 1, 1983.

2. Letter, A. E. Lundvall, Jr. (BG&E) to R. A. Clark (NRC), Docket Nos.

50-317 and 50-318 " Topical Report for Extended Burnup Operation of C-E Fuel," June 7, 1982; Enclosure CENPD-269-P, " Extended Burnup Opertion of Combustion Engineering PWR Fuel," April 1982

3. CENPD-139-P-A, "C-E Fuel Evaluation Model Topical Report," July 1974 4 CEN-161(B)-P, " Improvement to Fuel Evaluation Model," July 1981.
5. Letter, R. A. Clark (NRC) to A. E. Lundvall, Jr. (BG&E), " Safety Evaluation of CEN-161 (FATES 3)," March 31, 1983.
6. Letter, A. E. Lundvall, Jr. (BG&E) to T. E. Murley (NRC) "Calvert Cliffs Nuclear Power Plant Unit No. 1, Docket No. 50-317 Report of Startup Testing for Cycle 7," February 17, 1984
7. BG&E Calvert Cliffs 1 Slides Depicting SCCUT-1 High Burnup Demonstration Program, presented at BG&E/C-E/NRC meeting in Bethesda, Maryland on December 20, 1978.
8. Letter, A. E. Lundvall, Jr. (BG&E) to R. A. Clark (NRC), Docket No.

50-317, " Sixth Cycle License Application," -February 17, 1982.

9. Letter, A. E. Lundvall, Jr. (BG4E) to R. A. Clark (NRC). Docket No.

50-317 "Fifth Cycle License Application," September 22, 1980.

O 80 S

4 11-1

References - Chapter 4

1. Letter, A. !. Lundvall, Jr. (BG&E) to J. R. Miller (NRC), "Calvert

' Cliffs Unit 1 Supplement 1 to Seventh Cycle License Application,"

September 1, 1983.

2. Letter, A. E. Lundvall, Jr. (BG&E) to R. W. Reid (NRC) . Docket 50-317, " Fourth Cycle License Application," February 23, 1979.
3. Letter, A. E. Lundvall, Jr. (BG&E) to R. A. Clark (NRC). Docket No.

50-317, "Fifth Cycle License Application," September 22, 1980.

4 Letter, A. E. Lundvall, Jr. (BG&E) to R. A. Clark (NRC), Docket No. 50-317 " Sixth cycle License Application," February 17, 1982.

5. Letter, A. E. Lundvall, Jr. (BG%E) to R. W. Reid (NRC), Docket No. 50-317 " Proposed Finding of No Unreviewed Safety Question on Unit 2, Cycle 3 Reload Core Design," July 11, 1979.
6. Letter, A. E. Lundvall, Jr. (BG&E) to J. R. Miller (NRC), Docket Nos.

50-317 & 50-318 " Request for Amendment," (Clad Collapse / Augmentation Factors), December 31, 1984

7. CEN-183(B)-P, " Application of CENPD-198 to Zircaloy Component Dimensional Changes." September 1981
8. Letter, D. H. Jaffe (NRC) to A. E. Lundvall, Jr. (BG&E), "Regarding Unit 1 Cycle 6 License Approval (Amendment #71 to DPR-53 and SER),"

June 24, 1982. '

9. Letter, A. E. Lundvall, Jr. (BG&E) to J. R. Miller (NRC). Docket No.

50-317, " Seventh Cycle License Application Answers to Question Set 2," November 4, 1983.

10. CEN-105(B)-P, "Reconstitutable BgC Type CEA Design for Use in the
  • BG&E Reactor," February 1979.

11 CEN-83(B)-P, "Calvert Cliffs Unit 1 Reactor Operation with Modified CEA Guide Tubes," February 8, 1978, and Letter, A. E. Lundvall, Jr.

~ (SG&E) to V. Stello, Jr. (NRC), " Reactor Operation with Modified CEA Guide Tubes," February 17, 1973.

12. Letter, A. E. Scherer (C-E) to C. O. Thomas (NRC), CEA Guide Tube Wear Sleeve Modification," LD-84-043, August 3, 1984 -
13. Letter, A. E. Lundvall, Jr. (BC&E) to R. A. Clark (NRC), Docket No.59-317. " Report on Fretting Wear Inspection Performed at the End of Cycle 5 on Unit 1," CEN-216(B)-P, September 22, 1982.

la. CENPD-139-P-A, "C-E Fuel Evaluation !fodel Topical Report," July 1974

15. CEN-161(B)-P, " Improvement to Fuel Evaluation Model," July 1981.
15. Letter, R. A. Clark (!!RC) to A. E. Lundvall, Jr. (BG&E), " Safety Evaluation of CEN-151 (FATES 3) " March 31,1983.

e

  • fim*

References - Chapter 5

1. Letter, A. E. Lundvall, Jr. (BE&D) to J. R. Miller (NRC), Docket Hos.

50-317 & 50-318, " Request for Amendment," (Clad Collapse / Augmentation Factors), December 31, 1984

2. CENPD-266-P-A, "The ROCS and DIT Computer Codes for Nuclear Design,"

April 1983.

3 CENPD-153-P, Revision 1, " Evaluation of Uncertainties in the Nuclear Power Peaking Measured by the Self-Powered Fixed In-Core Detector System," May 1980.

4 Letter, A. E. Lundvall, Jr. (BG&E) to J. R. Miller (NRC), "Calvert Cliffs Unit 1 Supplement 1 to Seventh Cycle License Application,"

September 1, 1983.

4 e

11-3

r References - Chapter 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 1977.

4 Letter, P. W. Kruse to W. J. Lippold, " Responses to First Round Quations on the SCU Program: CETOP-D Code Structure and Modeling Methods, (CEN-124(3)-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 #71 to DPR-53 and SER),"

June 24, 1982.

6. CEN-124(B)-P, " Statistical Combination of Uncertainties, Part 2,"

January 1980.

7. CENPD-225-P-A, " Fuel and Poison Rod Bowing," June 1983.
8. Letter, C. O. Thomas (NRC) to A. E. Scherer (CE), " Acceptance for Referencing of Topical Report CENPD-225(P)," February 15, 1983.
9. CEN-124(B)-P, " Statistical Combination of Uncertainties, Part 1,"

January 1980

10. CEN-124(B)-P, " Statistical Combination of Uncertainties, Part 3 " March 1980.
11. Letter, A. E. Lundvall, Jr. (BG&E) to J. R. Miller (NRC), "Calvert Cliffs Unit 1 Supplement 1 to Seventh Cycle License Application,"

September 1, 1983.

11-4 ,

i References - Chaptar 7 1 Letter, A. E. Lundvall, Jr., (BG&E) to J. R. Miller (NRC), "Calvert Cliffs Unit 1 Supplement 1 to Seventh Cycle License Application," September 1, 1983.

2. " 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(B)-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., (BG&E), Regarding Unit 1 Cycle 6 License Approval (Amendments #71 to DPR-053 and SER), June 24, 1982.
6. CENPD-190A, "CEA Ejection, C-E Method for Control Element Assembly Ejection," July, 1976.
7. Letter, A. E. Lundvall, Jr. (BG&E) to R. A. Clark (NRC), Docket No. 50-317 "Fifth Cycle License Application," September 22, 1980.
8. Letter, A. E. Lundvall, Jr. (BG&E) to R. A. Clark (NRC), Docket No. 50-317

" Sixth Cycle License Application," February 17, 1982.

9. Letter, A. E. Lundvall, Jr. (BGE&) to R. A. Clark (NRC), " Amendment to Operating License DPR-69, Fifth Cycle License Application," Docket No. 50-318, October 15, 1982.
10. CENPD-188-A, "HERMITE Space-Time Kinetics," July, 1975.
11. CENPD-161-P, " TORC Code, A Computer Code for Determining the Thermal Margin for a Reactor Core," July, 1975.
12. R. V. MacBeth, "An Appraisal of Forced Convection Burn-Out Data," Proc.

Instn. Mech. Engrs., 1965-66, Vol. 180, Pt. 3C, pp. 37-50. ,

13. O. M. Lee, "An Experimental Investigation of Forced Convection Burnout in High Pressure Water: Part IV, Large Diameter Tubes at About 1600 Psia,"

.AEEW-R, November, 1966. -

9 8

_ap

Rnfarrncna - Chapter 8

1. Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors, Federal Register, Vol. 39, No. 3, y

Friday, January 4, 1974

2. CENPD-132, " Calculative Methods for the CE Large Break LOCA Evaluation Model," August 197'4 (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 LCCA Evaluation Model," July 1975 (Proprietary).

3. Letter, A. E. Lundvall, Jr. (BG&E) to R. A. Clark (NRC), Docket No. 50-317, " Sixth Cycle License Application," February 17, 1982.

4 CENPD-134, "CCMPERC-II, A Program for Emergency Refill-Reflood of the Core," April 1974 (Proprietary).

CENPD- 134, Supplement 1," CCMPERC-II, A Program for Emergency Refill-Reflood of the Core (Modification)," December 1974 (Proprietary).

5. CEMPD-135-P, "STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program," April 1974 CENPD-135-P, Supplement 2-P, "STRIKIN-II, 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. (BG&E) to B. C. Rusche (NRC), "Second Cycle License Application," October 1, 1976
8. CEMPD-138, " PARCH, A FORTRAN-IV Digital Program to Evaluate Pool Boiling, Axial Rod and Coolant Heatup," August 1974 (Proprietary).

CENPD-138, Supplement 1, " PARCH, A FORTRAN-IV Digital Program to Evaluate Pool Boiling, Axial Rod and Coolant Heatup," February 19J5 (Proprietary).

CENPD-138, Supplement 2, " PARCH, A FORTRAN-IV Digital Program to Evaluate Pool Boiling, Axial Rod and Coolant Heatup," January 1977 (Proprietary).

9. CENPD-137, Su pplement . 1," Calculative Methods for the C-E Small Break LCCA Evaluation Model," January 1977 (Proprietary).

11-6

10. L;ttcr, A. E. Lundvall, Jr. (BG&E) to R. W. Raid (NRC), Dcckste 50-317 cnd 50-318, "ECCS Small Brsak LOCA Analysis," Mrrch 13, 1979.
11. CENPD-137, " Calculative Methods for the C-E Small Break LOCA Evaluation Model," August 1974 (Proprietary).
12. CENPD-133, supplement 1, "CEFLASH-4AS, A Computer Program for Reactor Blowdown Analysis of the Small Break Loss-of-Coolant Accident," August 1974 (Proprietary).

CENPD-133. Supplement 3, "CEFLASH-4AS, A Computer Program for Reactor Blowdown Analysis of the Small Break Loss-of-Coolant Accident," January 1977 (Proprietary).

O F

h s

e 11-7 -

I I

f R,fertncea - Chcoter 9 l

r

1. Letter, A. E. Lundvall, Jr. (EG&E) to J. R. Miller (NRC), Docket Nos.

50-317 & 50-318, " Request for Amendment " (Clad Collapse / Augmentation Factors), December 31, 1984

2. CENPD-153-P, Revision 1, " Evaluation of Uncertainties in the Nuclear Power Peaking Measured by the Self-Powered Fixed In-Core Detector System," May 1980.

l 4

=

a s

i i

11-3

E- ]

! i Referrnc*s - Chepter 10

{

1. Letter, A. E. Lundvall, Jr. (BG&E) to J. R. Miller (NRC), "Calvert Cliffs Unit 1 Supplement 1 to Seventh Cycle License Application,a Septer.ber 1, 198 l i

I l

y s

N-..

I 9

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I 11-9

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