ML20076J860

Proposed Tech Spec Changes Decreasing Required Shutdown Margin & Increasing Allowed Moderator Temp Coefficient

ML20076J860
Person / Time
Site:	Calvert Cliffs
Issue date:	09/01/1983
From:	BALTIMORE GAS & ELECTRIC CO.
To:
Shared Package
ML20076J850	List: ML20076J846 ML20076J860
References
NUDOCS 8309070350
Download: ML20076J860 (40)

Text

77 TABLE 9-1 CALVERT CLIFFS 1 CYCLE 7 TECHNICAL SPECIFICATION CHANGES Change Tech. Spec. # Action 1 Table 2.2-1 Change Steam Generator Low Pressure Trip page 2-9 setting from 635 psia to 685 psia.

2 Table 2.2-1 Change Steam Generator Low Pressure Trip page 2-10 Bypass limit from 710 psia to 785 psia.

'3 B 2.1.1 Change LHGR to centerline melt ~ limit page B 2-1 from 21.3 kw/ft to 22.0 kw/ft.

4 B 2.2.1 Change Steam Generator Low Pressure Trip page B 2-5 setting from 635 psia to 685 psia and change uncertainty from 87 psia to 85 psia.

5 B 2.2.1 Revise description of the Steam page B 2-6 Generator Low Water Level Trip.

6 3/4.1.1.1 Change shutdown margin, Tavg> '

page 3/4 1-1 from 5.3% ak/k to 4.3% a k/k.

7 3 1.1.4 Change moderator temperature coefficient 0

page 3/4 1-5 negative ymit from -2.2x10- ak/k/ F to -2.5x10 ak/k/ F.

8 4.1.1.4.2 Change MTC surveillance Item (b).

page 3/4 1-6 9 Figure 3.2-3b Replace Figure 3 2-3b with enclosed page 3/4 2-4a Figure 3.2-3b.

10 3.2.2.1 Change calculated value of F xy T from page 3/4 2-6 1.65 to 1.70.

11 Figure 3.2-3a Replace Figure 3.2-3a with enclosed new page 3/4 2-7a Figure 3.2-3a 12 3.2.3 Change Figure 3.2-3a to Figure 3.2-3c.

page 3/4 2-9 13 Figure 3.2-3c Insert enclosed new Figure 3.2-3c after page 3/4 2-10a (new) page 3/4 2-10.

14 Table 3.2-1 Change minimum pressurizer pressure from page 3/4 2-14 2225 psia to 2200 psia.

8309070350 830901 PDR ADOCK 05000317 F PDR

78 TABLE 9-1 (continued)

Change Tech. Spec. # Action 15 Table 3 3-1 Change Steam Generator Low Pressure Trip page 3/4 3-4 Bypass limit from 710 psia to 785 psia.

16 Table 3 3-3 Revise and augment the Auxiliary page 3/4 3-14 Feedwater Section as indicated on the sample page.

17 Table 3 3-3 Change Main Steam Line Isolation (SGIS) page 3/4 3-15 Steam Generator Low Pressure Trip Bypass limit from 710 psia to 785 psia.

18 Table 3 3-4 Change Main Steam Line Isolation (SGIS) page 3/4 3-17 steam Generator Low Pressure Trip setting from 635 psia to 685 psia.

19 Table 3.3-4 Revise and augment the Auxiliary page 3/4 3-19 Feedwater Section as indicated on the sample page.

20 Table 3 3-5 Add the line for AFAS.

page 3/4 3-20 21 Table 3.3-5 Revise and augment the Steam Generator page 3/4 3-21 Level-Low Section and aca the Steam Generator AP-High Section, 'us indicated on the sample page.

22 Table 4 3-2 Revise and augment the Auxiliary page 3/4 3-23 Feedwater Section as indicated on the sample page.

23 3.7.1.2 Replace Pages 3/4 7-5 and 3/4 7-Sa with pages 3/4 7-5 and enclosed Pages 3/4 7-5 and 3/4 7-Sa.

3/4 7-Sa 24 Figure 3.7-1 Delete Figure 3.7-1.

page 3/4 7-Sb 25 3.10.4 Change 4.10.5.2 to 4.10.4.2.

, page 3/4 10-4 26 B 3/4.1.1.1 and Change shutdown margin, T 200 0F, B 3/4.1.1.2 from 5 3% a k/k at EOC and 4.5fy$>k/k at page B 3/4 1-1 BOC to 4.3% a k/k at both EOC and BOC.

27 B 3/4.7.1.2 Change bases for Tech. Spec. 3/4.7.1.2 page B 3/4 7-2 as indicated on sample page.

79 TABLE 9-2 E7.PLANATIONS FOR CYCLE 7 TECH. SPEC. CHANGES Changes Tech. Spec. # Explanation 1 Table 2.2-1 The Steam Generator Low Pressure Trip setting is being raised to accommodate the Safety Grade Auxiliary Feedwater Actuation System 2 Table 2.2-1 The Steam Generator Low Pressure Trip Bypass limit is being raised to reflect the change in the trip setting (see Change No. 1).

3 B 2.1.1 LHGR to centerline melt is being raised to increase operating margins and flexibility.

4 B 2.2.1 See Change No. 1.

5 B 2.2.1 The basis for the Steam Generator Low Water Level Trip is being adjusted to be consistent with the Safety Grade Auxiliary Feedwater Actuation System.

6 3/4 1.1.1 The shutdown margin is being lowered to reduce operating requirements with regard to shutdown boron levels, consistent with the generic SLB analysis presented herein (see Table 7-2) and existing safety analyses (see note to Table 5-2).

7 3 1.1.4 The moderator temperature coefficient negative limit is being increased to accommodate the effects of extended burnup.

8 4.1.1.4.2 The surveillance requirements on MTC are being modified to allow the use of MTC determinations made during power ascension startup measurements for the purpose of satisfying surveillance requirements. This change is consistent with the objective of assuring that the most positive MTC at power conditions, which occurs at the highest boron concen-tration, meets Tech. Spec. 3.1.1.4.b.

9 Figure 3 2-3b Figure 3.2-3b is being revised due to the segaration of the allowable Fxy and F curves into separate figures (seeCbangeNos.11, 12 and 13).

80-TABLE 9-2 (continued)

Changes . Tech. Spec. # Explanation 10 3.2.2.1 The planar r adial . peaking factor, T

F is being raised for Cycle 7 to indre,ase operating margins and flexibility.

11 Figure 3.2-3a The radial peaking factor ,

F [T, planar i!rease .

is being raised for Cycle 7 to operating and mygins flexibility and the F cyve is being separated from tYe curve T

to accommodate different F[xy and F

r values.

12 3.2.3 Figure 3.2-3c is being added to separation of the i

fac{litiate tge F and F curves to dale diffe[ent FxyT andaccommoT F

r values.

13 Figure 3.2-3c See Change No. 12.

14 Table 3 2-1 The minimum steady state pressurizer pressure has been lowered to increase operating flexibility.

15 Table 3 3-1 See Change No.'2.

16 Table 3 3-3 The description of the Engineered Safety Feature Actuation System Instrumentation for Auxiliary Feedwater is being revised and augmented to reflect the implementation of the Safety Grade

Auxilary Feedwater Actuation System.

17 Table 3 3-3 The Main Steam Line Isolation (SGIS)

Steam Generator Low Pressure Trip Bypass limit is being raised to reflect the change in the trip setting (see Change No. 18).

18 Table 3 3-4 The Main Steam Line Isolation (SGIS)

Steam Generator Low Pressure Trip setting is being raised to accommodate the Safety Grade Auxiliary Feedwater Actuation System.

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81 TABLE 9-2 (continued)

Changes Tech. Spec. # Explanation 19 Table 3 3-4 The Engineered Safety Feature Actuation System Instrumentation Trip values for Auxiliary Feedwater are being revised and augmented to reflect the implementation of the Safety Grade Auxiliary Feedwater Actuation System.

20 Table 3 3-5 The line for AFAS is being added due to the implementation of the Sa fety Grade Auxiliary Feedwater Actuation System.

21 Table 3.3-5 The Engineered Sa fety Feature Response Times for Steam Generator Level-Low are being revised and augmented and those for Steam Generator AP-High are being added to reflect the implementation of the Safety Grade Auxiliary Feedwater Actuation System.

22 Table 4.3-2 The Engineered Safety Feature Actuation System Instrumentation Surveillance Requirements for Auxiliary Feedwater are being revised and augmented to reflect the implementation of the Safety Grade Auxiliary Feedwater Actuation System.

23 3 7.1.2 The Limiting Condition for Operation (LCO) of the Auxiliary Feedwater System is b eing revised and augmented to reflect the implementation of the Safety Grade Auxiliary Feedwater Actuation j System.

24 Figure 3.7-1 Figure 3 7-1 is b eing removed to be consistent with revised and augmented

Tech. Spec. 3.7.1.2 (see Change No. 23).

25 3.10.4 A " typo" is being corrected.

26 B 3/4.1.1.1 and The shutdown margin is being decreased B 3/4.1.1.2 to make it consistent with Tech. Spec.

3/4.1.1.1.

27 B 3/4.7.1.2 The bases for the Auxiliary Feedwater LCO are being modified to reflect the implementation of the Safety Grade Auxiliary Feedwater Actuation System.

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n TABLE 2.2-1 (Cent'd) .

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y REACTOR PROTECTWE INSTRUHplTA_T_Igl TRIP SETPO,IN_T__ LIMITS a '

i n L .C FUNCTIONAL UNIT -TRI P._S.E.T.. POI NT ALLOWABLE VALUES y 4. Pressurizer Pressure - liigh 1 2400 psia < 2400 psia E 5. Containment Pressure - High < 4 psig < 4 psig

6. Steam Generator Pressure - Low (2) - la .

7. Steam Generator Water Level - Low > 10 inches below top > 10 inches below top of feed ring. of feed ring.

8. Axial flux offset (3) Trip setpoint adjusted to Trip setpoint adjusted to not exceed the limit lines not exceed the limit lines of Figure 2.2-1. of , Figure 2.2-1.

7* -

9. Thermal Margin / Low Pressure (1) k

a. Four Reactor Coolant Pumps Trip setpoint adjusted to Trin setpoint adjusted to Operating not exceed the limit lines be not less than the larger of Figures 2.2-2 and 2.2-3. of (1) the value calculated from Figures 2.2-2 and 2.2-3 and (2) 1875 psig.

b. Steam Generator Pressure -< 135 psid < 135 psid Difference - High (1) -

10. Loss of Load N.A. fl . A.

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g 11. Rate of Change of Power - High (4) 1 2.6 decades per minute < 2.6 decades per minute

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l TABLE NOTATION

-4 (1) Trip may be bypassed be}ow 10 % of RATED TilERMAL POWER; bypass shall be automatically removed when TilERMAL POWER is > 10- % of RATED TilERNAL POWER.

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G TABLE 2.2-1 (Cont'd)

E TABLE NOTATIONS (Cont'J) 7tg vi (2) Trip may be manually bypassed belowhst bypass shall be automatically removed at or above psia.

i (3) Trip may be bypassed below 15%.of RATED THERMAL POWER; bypass shall be automaticall'y removed when c- THERHAL POWER is > 15% of RATED THERMAL POWER.

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(4) Trip may be bypassed below 10-41 and above 12% of RATED THERMAL POWER.

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84 f 2.1 SAFETY LIMITS

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BASES' 2.1.1 REACTOR CORE 23'O The res'rictiot of this safety limit prevent dVerheating'of t'he .

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fuel cladding and ssible cladding perforation which would result in the release of fissio products to the reactor coolant. Overheating of the fuel is prevent maintaining the steady state peak linear heat rate at or less than 1. kw/ft. Centerline fuel melting.will not occur l for this peak linear. heat ra.tt. Overheating of the fuel cladding is .

( prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient.is large and the cladding i surface temperature is slightly above the, coolant saturation temperature. '

Operation above the upper boundary 6f the nucleate boiling regime could result in' excessive cladding temperatures because of the onset of  :

departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temper-ature and Pressure have been related to DNB through.the CE-1 correlation.

The CE-1 DNB correlation has been developed to predict the DNB flux and

(-). the. location of DNB for axially uniform and non-uniform heat flux distri-butions. The local DNB heat flux ratio. DNBR, def.ined as the ration of the heat flux that would cause DNB at a particular core location to the jocal. heat flux, is indicative of the margin to DNB.

The minimum value of the DNBR during steady state operation, normal operational transients, and anticipated transients is limited to 1.23. l This value corresponds to a 95 percent probability at a 95 percent con-fidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions.

The curves of. Figures 2.1-1, 2.1-2, 2.1-3 and 2.1-4 show the loci of points' of THERMAL POWER, Reactor Coolant System pressure and maximum cold leg temperature of various ptsnp combinations for which the minimum DNBR is no less than 1.23 for the family of axial shapes and l corresponding radial peaks shown in Figure B2.1-1. The limits in Figures  !

2.1-1, 2.1-2, 2.1-3 and 2.1-4 were calculated for reactor coolant inlet temperatures less than or equal to 580*F. The dashed line at 580*F coolant inlet temperature is not a safety limit; however, operation above, 580'F is not possible because of the actuation of the main steam line safety valves which limi,t the maximum value of reactor inlet temperature.

Reactor operation at THERMAL POWER. levels higher than 110". of RATED THERMAL l POWER is prohibited by the high power level trip setpoint specified in CALVERT CLIFFS - UNIT 1 B 2-1 Amendment No. M , Q , y

_ . _ . _ . _ _ _ _ - _ _ ._ _ _ _ _ _ ~ - _ - - - - - - - -

7 LIMITING SAFETY' SYSTEM SETTINGS BASES -

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operation of the reactor at reduced power if one or two reactor coolant pumps l are taken out of service. The low-flow trip setpoints and Allowable Values i for the various reactor coolant pump combinations have been derived in l consideration of instrument errors and response times of equipment involved  !

to maintain the DNBR above 1.23 under normal operation and expected transients. l l For reactor operation with only two or three reactor coolant pumps operating. -

the Reactor Coolant Flow-Low trip setpoints, the Power Level-High trip set '

points, and the Thermal Margin / Low Pressure trip setpoints are automatically l

. changed when the. pump condition selector switch is manually set to the desired two- or three-pump position. Changing these trip setpoints during two and ,

'three pump operation prevents the minimum value of DNBR from going below 1.23 l l during normal operational transients and anticipated transients when only' two -

f or three reactor coolant pumps are operating. . .

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Pressurizer Pressure-High .

l The' Pressurizer Pressure-High trip, backed up by the pressurizer cod'e i safety. valves and main steam line safety valves, provides reactor coolant  !

system protection against overpressurization in the event of loss of load without reactor trip. This trip's setpoint is 100 psi below the nominal lift 1 setting (2500 psia) of the pressurizer code safety valves and its concurrent l operation with the power-operated relief valves avoids the undesirable opera- ,

tion of the pressurizer code safety valves.

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Containment Pressure-High .  !

'* The Containment Pressure-High. trip provides assurance that a reactor trip is initiated concurrently with a safety., injection. The setpoint for

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this trip is 1.dentical to the safety injection setpoint. '

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Steam Generator Pressure-Low -

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L The Steam Generator ' iPressure-Low trip . provide protection against an l excessive rate of heat extraction from the steam nerators and subsequent

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cooldown of the reactor coolant. The setting of 63 psia is sufficiently l

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below the full-load operating point of 850 psia so as not to interfere  ;

! with normal operation, but still high enough to provide the required protec- l tion in the event of excessively igh steam flow. This setting was used l with an uncertainty factor of t 87 psi which was based on the main steam line )

break event. g .

l CALVERT CLIFFS - UNIT 1 B 2-5 Amendment; No. 33, 4,71 4 l

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LIMITING SAFETY SYSTEM SETTINGS BASES

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Steam Generator Water Level The ' Steam Generator Water Level-Lrv trip provides cori prote.ction ,

by preventing operation with the steam generator water level below the minimum volume required for adequate heat removal capacity and assures that the pressure of the reactor coolant system will not exceed its Safety Limit.4t:r*:':;::f'i:d  ::t;;f-t pr:vid:: :11::::-

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Axial Flux Offset g& A g M &M +

.The axial flux offset trip i provided to ensure that excessive axial peaking will not cause fuel damage. The axial flux offset is .

determined from the axially split excore detectors. The trip setpoints ensure that neither a DNBR of less than 1.23 nor a peak linear heat rate

• which corresponds to the temperature for fuel centerline melting will exist as a consequence of axial power ma1 distributions. These trip set- gs points were derived from an analysis of many axial power shapes with G allowances for instrumentation inaccuracies and the uncertainty associated with the excore to incore axial flux offset relationship. -

l Thermal Margin / Low Pressure The Thennal Margin / Low Pressure trip is provided to prevent operation when the DNBR is less than 1.23. . , .

The trip is initiated whenever the reactor coolant system pressure signal drops below either 1875 psia or a computed value as described l below, whichever is higher. The computed value is a function of the

. . higher of AT power or. neutron power, reactor inlet temperature, and the number of reactor coolant pumps operating. The minimum value of reactor coolant flow rate, the maximum AZIMUTHAL POWER TILT and the maximum CEA deviation permitted for continuous operation are assumed in the genera-tion of this trip function. In addition, CEA group sequencing in accor-dance with Specifications 3.1.3.5 and 3.1.3.6 is assumed. Finally, the

• maximum insertion of CEA banks which can occur during any anticipated operational occurrence ~ prior to a Power Level-High trip is assumed.

CALVERT CLIFFS - UNIT 1 B'2-6 Amendment No. 33, W .48, 71 l

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I 87 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL

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SHUTDOWN MARGIN - T,,,> 200*F -

LIMITING CONDITION FOR ODERATION 43% '

3.1.1.1 -The SHUTDOWN MARGIN shall be > 5.3 Ak/k. ~ '.

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, APPLICABILITY: MODES 1, 2**, 3 and 4. .

ACTION: g r/, .

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With the SHUTDOWN MARGIN < Ak/k, imediately initiate and continue , I boration at > 40 gpm of 230 ppm boric acid. solution or equivalent until -

the required-' SHUTDOWN MARGIN is restored., .

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SURVEILLANCE REQUIREMENTS 43%

4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be > 5.3 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 imovable or untrippable, the above M( required SHUTDOWN MARGIN shall be increased by an amount at i least e

• CEA(s) qual to the withdrawn worth of the imovable or untrippable I

b. When in MODES 1 orI 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 withdrawaJ 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 r

criticality by verifying that the predicted critical CEA 1

position is within the limits of Specification 3.1.3.6.

. d. ~ Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors of e below, with the CEA groups at the Transient Insertion Limits of Specification 3.1.3.6.

Adherence to Technical Specification 3.1.3.6 as specified in Surveillance

. Requirements 4.1.1.1.1 assures that there is sufficient available shut-down margin to match the shutdown margin requirements of the safety

, analyses.

• • See Special Test Exception 3.10.1.

1. With K,ff > 1.0.

1. 1. With K,ff < 1.0.

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CALVERT CLIFFS - UNIT 1 3/4 1-1 Amendment No. 27. YJ.AE,71

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REACTIVITY CONTROL SVaiu6 e i 2

l MODERATOR ipFERATURE COEFFICIENT .: l 6

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LIMITING CONDITION FOR OPERATION T- - -

3.1.1.4 Themoderatortemperaturecoefficient(MTC)shall,be:

a. .Less positive than 0.5 x 10 Ak/k/*F whenever THERMAL -

POWER is <_ 705 of RATED THERMAL POWER,

b. Less positive than 0.2 x 10 Ak/k/*F whenever THERMAL -

POWER is > 705 of RATED THERMAL POWER, and -

c. Less negative than -2.

x 10i Ak/k/*F at RATED THERMAL POWER.

.2.5 APPLICABILITY: MODES 1 and 24 4c ACTION: .!

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With the moderator tempe'rature coefficient outside any one of the above ~

limits, be in at least HOT STAND 8Y within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. .

' 4 .

l SURVEILLANCE REQUIREMENTS 4.1.1.4.1 The MTC shall be determined to be within its lidiits by confirmatory measurements. MTC measured values shall be extraoolated and/or compensated to permit direct comparison with the above 1imits. ,

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1 h ith K,ff > 1.0. -

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1. See Special Test Exception 3.10.2.

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CALVERT CLIFFS - UNIT 1 3/4 1-5 Amendment No. 48

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REACTIVITY CONTROL SYSTEMS .

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SURVEILLANCE REQUIREMENTS.(Continued). .

4J.T.4.~2 The 'MTC shall be determined at the following frequencies and

THERMAL POWER conditions during each fuel cycle: -

a. Prior to iniistal operation above 55 of RATED THERMAL POWER s..

after each fuel loading. PATgo THfSMAL Po,gJ g *-

yy ge y,4 -

b. Wa 4AfiiD At any THERMA.

O ;; iittrPOWERLwithin h ter;.. xc.:: t 7 EFPD aftegA :" E ;p.

• r i "C" .

c. At any THERMAL POWER, within.7 EFPD af r reaching a RATED s THERMAL, POWER equilibrium boron concen tion of 300 ppm.

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91 POWER DISTRIBUTION LIMITS -

TOTAL PLANAR RADIAL PEAKING FACTOR Fh LIMITING CONDITION FOR OPERATION T T 3.2.2.1 The ulated value of F*Y,

.lisited.to f_ defined as F*Y,.= F*?(*1 4 +T ), shall be gyg '

. APPLICABILITY: ' MODE l'*. -

ACTION.

(.70 .

With F > 1. 6 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> eittrer: - -

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a. Reduca THERMAL POWER to bring the tombination of THERMAL. POWER and1 to within the limits of Figure 3.2-3a and withdraw the l full length CEAs .to or beyond the Long Tern Steady State

. -Insertion Limits of Specification 3.1.3.6; or -

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b.' 'Be in at least HOT STANDBY. .

' SURVEILLANCE REQUIREMENTS 9

4. 2. 2.1.1 The provisions of Specification 4.0.4 are not applic'able.- l 42.2.1.2 Ffy shall be calculated by the' expression F =Fxy(1+T ) and l q,

'Fxy shall be d.etermined to be within its limit at the following intervals:

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a. Prior to operation above 70 percent of RATED THERMAL POWER after ea'ch fyel loading,

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b. At least' once per 31. days of accumulated operation'in MODE 1 and ,

c. Within four hours'if the AZIMUTHAL POWER. TILT q(T ) is > 0.030.

• See Special Test Exception.3.10.2.

l CALVERT. CLIFFS - UNIT 1 3/4 2-6 Amendment No. 32, 33, (8,71

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_yM Mys _i h -

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.o.z = 'A OC 1

!_ W n .L'

__s ._

}_._- _T i . N

• "~"

e ,

a - - -

E I I I I, I

__,-_._a, _ _ .  ; y_a. m o4

• a y w g xx.L.n~ o R. L n, -

t_ _._ _ . . .M,e** .

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+-- h- e -d 1

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4. 2.- 7 4.

93' , ,

~~ '

POWER DISTRIBUTION LIMITS

~ '

f TOTAL INTEGRATED RADIAL PEAKING FACTOR'-- Ff , -;. .

LIMITING CONDITION FOR OPERATION ,_ -

3.2.3 to < 1.650. The-calculated value of1Ffo ldefined ~as Ff)=iF IIU ')* s~ha r

,,- y.%...

J APPLICABILITY: MODE 1*. ,

ACTION: ,

WithFf>1.650within6 hours,either: '

. ), ,

a. Be in at least HOT STANDBY, 'or 31 - -

b. Reduce THERMAL POWER ~to bring the, combination of THERMAL POWER and Ff to within the limits of Figureh withdraw the' full length CEAs to or beyond the Long Term Steady State Limits of Specification 3.1.3.6, and insert new value of Ff in BASSS; or

c. 32.-36 Reduce THERMAL POWER to bring the combination of THERMAL POWER and T

F to within the lic.sts of Figure 3.2-3a and withdraw the full length I h A'2 ~ N CEAs to or beyond the Long Term Steady State Insertion Limits of.

The THERMAL POWER limit determined from

( Specification 5.1.3.6.Figur#.2-33shall then be used to establish Ia revi PUWER level limit on Figure 3.2-4 (truncate Figure 3.2-4 at the wable fraction of RATED THERMAL POWER determined by Figure 3.2-3 and subsequent operation snall be maintained within the reduced acceptable operation region of Figure 3.2-4.

3 2-34 *-

SURVEILLANCE REQUIREMENTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicable.

4.2.3.2 Ff shall be calculated by the expression Ff = F (1+T ) and Ff r q shall be-determined to be within its limit.at the following intervals:

a. Prior to operation above 70 percent of RATED THERMAL POWER after each fuel loading, ,,

b. At least once per 31 days of accumulated operation in MODE 1.,

and

c. Within four hours if the AZ;IMUTHAL POWER TILTq(T ) is > 0.030.

C

• See Special Test Exception 3.10.2. .

CALVERT CLIFFS - UNIT 1 3/4 2-9 , Amendment No. 33, 33, 39. M ,'/;

94 w i

m, H

w -4

-t-m

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(

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j u- y

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(,m.--. ) y _2 mg _

(/ 4-t~m _ _ 2

'a V_ a' "

NTD b '

'a g _; q Rb-n, ~ =

s p

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i . m r my _

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s , .. y I -' / 5 e c, 1

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=

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___ 1 6-- =:

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?4 z - io a.

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j

~

TABLE 3.2-1 .

n

$ DNB PARAMETERS

!3

--s

p LIMITS i

5 . Four Reactor Three Reactor Two Reactor Two Reactor Coolant Pumps Coolant Pumps Coolant Pumps Coolant Pumps

[ ,

Operating Operating Operating-Same Loop Operating-Opposite Loop 2 Parameter i

j Cold Leg Temperature < 548*F 2Z00 ,

,' Pressurizer Pressure > 2225 psia

• 3

Reactor Coolant System Total Flow Rate > 370,000 gpm i

l AXIAL SHAPE INDEX w ..

m 2

• Limit not applicable during e'ther a THERMAL POWER ramp increase in excess of 5% of RATED THERMAL POWER

% per minute or a THERMAL POWER s,tep increase of greater than 10% of RATED THERMAL POWER.

**These values left blank pending NRC approval of ECCS analyses for operation with less than four reactor coolant pumps operating.

***The AXIAL SHAPE INDEX, Core Power shall be maintained within the limits established by the Better l Axial Shape Selection System (BASSS) for CEA insertions of the lead bank of '< 55% when SASSS is l ko OPERABLE, or within the limits of FIGURE 3.2-4 for CEA insertions specified by FIGURE 3.1-2,

a n

O j g u

~

• j- n r

c .

TABLE 3.3-1 (Continued)

-TAB'LE NOTATION

')

• With the protective system trip breakers in the closed position and the CEA drive system capable of CEA withdrawal.

The provisions of Specification 3.0.4~ are not applicable.

..  : s - .

(a) o of RATED THERMAL POW Trip may be bypassed bel' w.10-4ba automatically of RATED removed whe THERMAL POWER. ,

g (b) Trip may be manually bypassed belo 71 psia; byriass shall be ~ '

automatically removed at'ol above,1 psia.

(. 715 '- - .

(c) Trip may be bypassed below 15% of RATED ~ THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is > 15% of ~

. RATED THERMAL POWER. .

(d) Trip may be bypass' d below e 10-4% and above 12% of RATED. THERMAL'. -

l POWER. l (e) Trip may be bypassed during. testing pursuant to Special Test.Excep-tion 3.10.3. '

(f) There shall be at least two decades of overlap between the Wide Range Logarithmic Neutron Flux Monitoring Channels and the Power h l Range Neutron Flux Monitoring Channels.

~

ACTION STATEMENTS ,

ACTION 1 -

With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within ... .

1. P, . hours or be in NOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> '

ar.d/or open the protective system trip breakers.

ACTION 2 -

With the number of OPERABLE channels one less than the Total Number of Channels,-STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a. The inoperable channel is placed in either the bypassed or tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. For the purposes of testing and maintenance, the inoperable channel may be bypassed for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> from time of initial loss of OPERABILITY; however, the inoperable channel shall then be either restored to OPERABLE status or placed in the tripped condition. ,

CALVERT CLIFFS - UNIT 1 3/4 3-4 Amendment No. AS, 71 1 I

- - - , . , _ _ _m.- - _ . _ _ - - . - - ,. . , . . , .- . - . , _ . _7-., ____,, _ . . _ _

TABLE 3.3-3 (Continued)

ENGINEEllEO SAFETY FEATultE ACTUATION SYSTEM INSTItUMENTATION MINIMUM TOTAL NO. CilANNELS CilANNELS APPLICABLE .

FUNCTIONAL UNIT OF CilANNELS TO TitiP OPEltABLE MODES ACTION

8. CVCS ISOLATION

a. Manual (CVCS isolation Valve Control Switches) 1/ Valve 1/ Valve . 1/ Valve 1, 2, 3, 4 6

b. West Penetration lloom/ Letdown lleat Exchanger Itoom Pressure - liigh 4 ,

2 3 1, 2, 3, 4 7' g 9. AUXILIAltY FEEDWATElt . .

N Actuation System

a. Manual (Trip Buttons) 2 sets of 2 I set of 2 2 sets of 2 1, 2, 3 6 per S/G per S/G per S/G

b. Steam Generator 1,2,3 7 p

Level - Low 4/SG 2/SG 3/SG

c. Steam Generator A Pliigh 4/SG 2/SG 3/SG 1,2,3 7 k

4 9

98 ,

TABLE 3.'3-3 (Continued)

TABLE NOTATION

.. < s .

(a) Trip function may be bypassed'in:this MODE when pressurizer pressure' is < 1800 psia; bypass shall be automatically removed when pressurizer pressure is 2.1800 psia. 7g (c) Trip f' unction may be bypassed in thi.s MODE lo h psha;' bypass shall be automatically. removed at or above 710 psia.

785

• The provisions of Specification ~3.0.4 are not applicable.

, . ACTION STATEMENTS ,

[

ACTION 6 -

With the number of OPER'ABLE channels one less than the - -

Total Number of Channels, restore the ino~perable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-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 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 />.

ACTION 7 -

With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed provided ~

the following conditions are satisfied:

O '

a. The inoperable channel is placed in either the bypassed or tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. For the purposes of testing and maintenance, the inoperable channel may be bypassed for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> from time of initial loss of OPERABILITY;~ however, the inoperable channel shall then be either restored to OPERABLE status or, placed in the tripped condition.

. .b. Within one hour, all functional units receiving an input from the inoperable channel are also placed in the same condition (either bypassed or tripped, as applicable) as that required by a. above for the inoperable channel.

c. The Minimum Channels OPERABLE requirement is met; however, one additi_onal channel may be bypassed for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> while performing tes'.s and maintenance on that channel provided the other '1 operable channel is placed in the tripped condition.

CALVERT CLIFFS - UNIT 1 3/4 3-15 Amendment No.AB, 71 '

1 .,

l TABLE 3.3-4 9

G ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES i g -

r"-

ALLOWABLE j y _ FUNCTIONAL UNIT TRIP SETPOINT

, ' VAL _UES l 1.

i SAFETY INJECTION (SIAS) e a. Manual (Trip Buttons) Not Applicable Not Applicable 2

~

b. Containment Pressure - High 1 4.75 psig 1'4.75 ps.ig

]

c. Pressurizer Pressure - Low > 1725' psia' > 1725 psia 'l .

2. CONTAINMENT SPRAY.(CSAS)

a. Manual (Trip Buttons) Not Applicable J ' Not Applicable

b. Containmen,t Pressure -- High 1 4.75 psig 1 4.75 psig

3. CONTAINMENT ISOLATION (CIS) #

y a. Manual CIS (Trip Buttons)

Not Applicable Not" Applicable

b. Containment Pressure - High 1 4.75 psig 1 4.75,psig

4. MAIN STEAM LINE ISOLATION a.

Manual (MSIV Hand Swit'ches and Feed Head Isolation

• y Hand Switches) Not Applicable . Not Applicable S

y b. Steam Generator Pressure - Low > 635 psia > 635 psia '

l i 6US , 4IS

= _

?

~

1. Containment isolation of non-essential penetrations is also initiated by SIAS (functional

• p units 1.a and 1.c).

• 8 ,

s

• e

r-O

- D .

, TAllLE 3.3-4 (Continued)

O ENGINEEllED SAFETY FEATullE ACTt1ATION SYSTEM INSTitt1 MENTATION TitIP VALUES E. '.

I i ALLOWAllLE ,

q FilNCTIONAL UNIT Tittl* VALUE VAltltiS

.Il

(/,

8. CVCS ISOLATION 8

West l'enetration Roosn/ Letdown lleat (0.5 psig {0.5 psig i9 Exchanger Room Pressure - liigh tl j,13 9. AUXILIAltY FEEDWATElt ACTUATION SYSTEM 3

a. Manual (trip buttons) Not Applicable Not Applicable

_. -m . - _ __

)

b. Steam Generator (A or 15) Level - Low -lig =cun re -sgia.Aes -194 im.hes To -M haber ,

). _ _

Wiesive

)

c. Steam GeneratorAP-liigh -

.f 135.0 psi .6135.0 psi P jig'i..' (SG-A 2 SG-il) &

O' lso '

l d. Steam Generator AP-liigh 6135.0 psi 6135.0 psi *j l (SG-li x SG-A)

(1) % of the distance between steam generator upper and lower level instrument nozzles.

t l

4 I!I I

>l.ilti i i>i q.

l l .

101 TABLE 3:3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES

, ')

INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONOS

1. Manual

a. SIAS Safety Injectio'n(ECCS) Not Applicable

b. CSAS Containment Spray Not App'11 cable -

c. CIS Containment Isolation Not Applicable *

d. RA$

Containment Sump Recirculation Not Applicable .

AFAS Auxiliary Feedwater Initiation Not Applicable

2. Pressurizer Pressure-Low

a. Safety Injection (ECCS) 1 30*/30** g i

3 .' Containment Pressure-High .

a. Safety Injection (ECCS) 1 30*/30**'

b. Containment Isolation 103

c. Containment Fan Coolers 1 35*/10**

4. Containment Pressure--High

a. Containment Spray ~ 1 60*/60** (l)

5. Containment Radiation-High ,,

a. Containment Purge Valves Isolation 15

• _1:

.J Amendment No. AE, 54 CALVERT CLIFFS - UNIT 1 3/4 3-20 CALVENT C:.:FFS - U;!IT 2 Amendment i:c. 31,37

• 102 TABLE 3.3-5 (Continued)

ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS

6. Steam Generator Pressure-Low -

a. Main Steam Isolation 16.9

b. Feedwater Isolation 5 30

7. Refuelinit Water Tank-Low

a. Containment Sump Recirculation $30

3. Reactor Trio ,

a. Feedwater Flow Reduction to 5% -

f20

9. Loss of Power (1 a. 4.16 kv Emergency Bus Undervoltage $ 2.2*+

(Lpss of Voltage)

b. 4.16 kv Emergency Bus Undervoltage f 3.4+*

(Degraded Voltage)

Steam Generator Level- Low 10.

n

\

a. Steam Driven AFW Pump $54.5

b. Motor Driven AFW Pump ,

$_54.5* / 14.5++ g L1. Steam Generator AP-High

a. Auxiliary Feedwater Isolation f20.0 TABLE NOTATION

+ Diesel generator starting and sequence loading delays included.

++ Diesel generator starting and sequence loading delays not included.

Offsite power available.

"+ Response time measured from the incidence of the undervoltage condition to the diesel generator start signal (1) Header fill time not included.

CALVERT CLIFFS - UNIT 2 3/4 3-21 Amendment No.

~. --- - -

i O .

../

TAllLE 4.3-2 (Continued) .'

ENGINEEllED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS f

l o

! [*: CilANNEL MODES IN WillCil l' CilANNEL CilANNEL FUNCTIONAL SURVEILLANCE l': FUNCTIONAL IJNIT CllECK CAlll1 RATION TEST REQUIRED _

9

. I, 5. CONTAINMENT SUMP l ", ItEClitCULATION (RAS)

! , a. Manual RAS (Trip 11uttons) N.A. N.A. R N.A.

lg b. Itef ueling Water i n Tank - Low N.A. R M 1, 2, 3

'~l c. Autornatic Actuation Logic N.A. N.A. M(1) I, 2, 3 j

6. CONTAINMENT PURGE VALVES ISOLATION ##

a. Manual (Purge Valve Control j Switches N.A. -

N.A. R N.A.

b. Containsnent Radiation - liigh

,,, Area Monitor S R M 6 D ' '5 w

, ,, 7. LOSS 01: POWER A, a. 4.16 kv Emergency llus

]" Undervoltage (Loss of j Voltage N.A. R M 1, 2, 3

b. 4.16 kv Emergency llus j tlndervoltage (Degraded j Voltage N.A. R M 1, 2, 3 4

8. CVCS ISOL ATION N.A. R M 1, 2, 3, 4 i West Penetration lloom/

i E,

Letdown lleat Exchange l !j

! 1 11oom Pressure - liigh i f 9. AUXILIAltY Fl:1:hW ATE:lt Manua6 Trip liuttons))g

} il a.

b. Steam Generator t.evel-Low N.A.

S N.A.

R M R N.A.

l. 2, 3 ,4

.g 9 . 5 eam Generator AP-liigh S R M 1, 2, 3

d. Automatic Actuation Logic N.A. N.A. M(1) I, 2, 3 j j ## Contaisunent purge valve isolation is also initiated by SIAS (functional units 1.a, l.h and 1.c).

1 s

104-PLANT SYSTEMS AUXILIARY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION ,

3.7.1.2 Two auxiliary feedwater trains consisting of one steam-driven and one motor-driven pump and associated flow paths capable of automatically initiating flow shall be OPERABLE.* (An OPERABLE steam-driven train shall consist of one pump aligned for automatic flow initiation and one pump aligned in standby.)**

PPLICABILITY: MODES 1,2 AND 3 ACTION:

a. With 13(23) motor-driven pump inoperable

1. Align the standby steam-driven pump to automatic initiating 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 HOT shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and

2. Restore 13(23) motor-driven pump to OPERABLE status within the next 14 days or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b. With one steam-driven pump inoperable:

1. Align the OPERABLE steam-driven pump to automatic initiating 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 HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and

2. Restore the inoperable steam-driven pump to standby status (or automatic initiating status if the other steam-driven pump is to be placed in standby) within the next 30 days or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

c. Whenever a subsystem /s (a subsystem consisting of one pump, piping, valves and controls in the direct flow path) required for operability is inoperable for the performance of periodic testing (e.g. manual discharge valve closed for pump Total Dynamic Head Test or Logic Testing) a dedicated operator /s will be stationed at the local station /s with direct communication to the Control Room. Upon completion of any testing, the subsystem /s required for operability will be returned to its proper status and verified in its proper status by an independent operator check.

d. The requirements of Specification 3.0.4 are not applicable whenever one motor t.'d one steam-driven pump (or two steam-driven pumps) are aligned for automatic flow initation.

• For a period of up to 30 days following the entering into Mode 3 (up through and including MODE I operation) from the Cycle 7 Unit I startup the automatic actuation features of the auxiliary feedwater system may be inoperable.

• • A standby pump shall be available for operation but aligned so that automatic flow initiation is defeated upon AFAS actuation.

CALVERT CLIFFS UNIT 1 & 2 3/47-5

-' , 105 PLANT SYSTEMS AUXILIARY FEEDWATER SYSTEM SURVEILLANCE REOUIREMENTS 4.7.1.2 Each auxiliary feedwater flowpath shall be demonstrated OPERABLE:

a. At least once per 31 days by:

Verifying that each steam <friven pump develops a Total Dynamic Head of

~

1.

12300 ft. on recirculation flow (if verification must be demonstrated during startup, surveillance testing shall be performed upon achieving an RCS temperature 1300 F and prior to entering MODE 1).

2. Verifying that the motor-driven pump develops a Total Dynamic Head of > 3100 ft. on recirculation flow.

3. Cycling each testable, remote-operated valve that is not in its operating position through at least one complete cycle.

4 Verifying that each valve (manual, power operated or automatici in the direct flow path is in its correct position.

b. Before entering MODE 3 after a COLD SHUTDOWN of at least 14 days by completing a flow test that verifies the flow path from the condensate storage tank to the steam generators.

c. At least once per 18 months by verifying that each automatic valve in the flow path actuates to its correct position and each auxiliary feedwater pump automatically starts and delivers flow to each flow leg upon receipt of each auxiliary feedwater actuation system (AFAS) test signal.

CALVERT CLIFFS UNIT 1 & 2 3/4 7-Sa

I 106 SPECIAL TEST EXCEPTIONS CENTER CEA MISALIGP9 TENT ,

LIMITING CONDITION FOR OPERATION 3.10.4 The requirements of Specifications 3.1.3.1 and 3.1.3.6 may be  !

suspended during the performance of PHYSICS TESTS to detemine the isothermal temperature coefficient and power coefficient provided:

a. Only the center CEA (CEA #1) is misaligned, and

, b. The limits of Specification 3.2.1 are maintdnad and determined as specified in Specification @ 0.5.)below.

APPLICABILITY: MODES 1 and 2. 4,/O.f.?_

ACTION:

With any of the limits of Specification 3.2.1 being exceeded while the requirements of Specifications 3.1.3.1 and 3.1.3.6 are suspended, either: @

a. Reduce THERMAL POWER sufficiently to satisfy the requirements of Specification 3.2.1, or

b. Se in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.10.4.1 The THERMAL POWER shall be determined at least once per hour during PHYSICS TESTS in which the requirements of Specifications 3.1.3.1 and/or 3.1.3.6 are suspended and shall be verified to be within the test power plateau.

4.10.4.2 The linear heat rate shall be determined to be within the -

limits of Specification 3.2.1 by monitoring it continuously with the Incore Detector Monitoring System pursuant to the requirements of Specifications 4.2.1.3 and 3.3.3.2 during PHYSICS TESTS above 5% of RATED THERMAL POWER in which the requirements of Specifications 3.1.3.1 and/or 3.1.3.6 are suspended.

F v

CALVERT CLIFFS - UNIT 1 3/4 10-4

. 107 3/4.1 REACTIVITY CONTROL SYSTEMS- '~ -

(. BASES -

3/4.1' .1 BORATION CONTROL - . .

e 3/4.1.1.1 and '3/4.1.1.2 SHUTDOWN ~ MARGIN

.

. ~ ~

A's~ufficient SHUTDOWN MARGIN ensures that:1) 1ihe reactor'can be made f subcritical from all operating conditions ~,-2) the 'reactivit.f trinsients

• L associated with postulated accident condit. ions are controllable w'ithin i

• acceptable limits, and 3) the reactor will be maint'ained suffic.iently suberitical to preclude inadvertent criticality in the shutdown condit' ion.'

~ 4.3*A SHUTDOWN MARGIN require. ts vary throughout cor'e life as a function of

~

fuel depletion, RCS b'oron c' centration 'and 'RCS T The minimum avafia

+N, SHUTDOWN MARGIN for no load rating'conditionsll9beginning of life is 4.5 .

Ak/k and at end of life is .3 Ak/k; The SHUTDOWN MARDIN is based on the safety analyses performed for a steam line rupture event . initiated at no load

~

a conditions. The most restrictive steam line rupture event occurs'at .EOC conditions. For the steam line rupture nt at beginning of cycle conditions, a minimum SHUTDOWN MARGIN of less1than 4.5% Ak/k'is requir to control the reactivity transient, and end of cycle on itions req'uire .3 Akik. Accordingly, the SHUTDOWN MARGIN requirement is bas d upon this; limiting conditgon and 1s X consistent with FSAR safety analysis sumptions. With T 200 F, the a reactivity transients res'ulting from' ny postulated accid 8N <re a minimal and a 4 3 9e

,Q. 3% Ak/k shutdown margin provides adeq ate protection. With the pressurizer level less than 90 inches, the source ; of non-borated water are restricted to increasethetimetocriticalityduriga,borondilutionevent. . . .

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3/4.1.1.3 BORON DILUTIOR 'i.'5 9e ~

A minimum flow rate of'at leas.t 3000 GPM.provides adequate mixing, 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 2.4 minutes. The reactivity change rate associated with boron concen

. tration reducti-ons will therefore be within the capability of operator recognition and control.

3/4.1.1.4 MODERATOR TEMPERATURE COEFFICIENT (MTC) l The limitations on MTC are provided to ensure that the ass'umptions used in the accident and transient analyses remain valid through each 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 boron i -concentration associated with fuel burnup. The confinnation that the

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

{

CALVERT CLIFFS - UNIT 1 B 3/4 1-1 Amendment No. 27, 22.4d. 71

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PLANT SYSTEMS BASES U = maximum nun.ber of inoperable safety valves per operating steam line q 106.5 ~= Power Level-High Trip Setpoint for two loop operation 46.8 = Power Level-High Trip Setpoint for single loop operation with two reactor coolant pumps operating in the same loop X = Total relieving capacity of all safety valves per steam line in Ibs/ hour Y. = Maximum relieving capacity of any one safety valve in Ibs/ hour -

3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM The OPERABILITY of the auxiliary feedwater system ensures that the Reactor Coolant System can be cooled down to less than 3000 F from normal operating conditions in the event of a total loss of offsite power. A capacity of 400 gpm is sufficient to ensure that adequate feedwater flow is available to remove decay heat and reduce the Reactor Coolant System temperature to less than 300 F when the shutdown cooling system may be placed into operation.

Flow control valves, installed in each leg supplying the steam generators, are set to maintain a nominal flow setpoint of 160 gpm plus or minus 10 gpm for operator set' ting band. The nominal flow setpoint of 160 gpm incorporates a total instrument loop error band of plus 47 gpm and minus 60 gpm. The operator setting band, when combined

• with the instrument loop error, results in a total flow band limits of 90 gpm (minimum and 217 gpm (maximum). Safety analyses show that more flow during an overcooling transient and less flow during an undercooling transient could be tolerated; i.e., f!ow fluctuations outside this flow band but within the assumptions used in the analyses listed below, are allowable. .

In the spectrum of events analyzed in which automatic initiation of auxiliary feedwater occurs, the following flow conditions are allowed with an operator action time of 10 minutes.

(1) Loss of Feedwater: 0 gpm Auxiliary Feedwater Flow (2) Feedline Break: 0 gpm Auxiliary Feedwater Flow (3) Main Steam Line Break: 1300 gpm Auxiliary Feedwater Flow (This being the maximum flow through the AFW suction line, with one unit requiring flow, prior to pump cavitation due to low NPSH.)

At 10 minutes after an Auxiliary Feedwater Actuation Signal the operator is assumed to be available to increase or decrease auxiliary feedwater flow to that required by existing plant conditions.

CALVERT CLIFFS UNIT I & 2 B 3/4 7-2

-109 10.0 STARTUP TESTING The startup testing progran proposed for Cycle 7 is identical to the program proposed for the reference cycle in Reference 1.

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11.0 REFERENCES

Reforences - Chapters 1 Through 5

1. IAtter, A. E. Lundvall, Jr. (BG&E) to R. A. Clark _(NRC), " Amendment to Operating License DPR-69, Fifth Cycle License Application," Docket No. 50-318, October 15, 1982.

2. Letter, A. E. Lundvall, Jr. (BG&E) to R. A. Clark (NRC), Docket No. 50-318, " Supplement 1 to Fifth Cycle License Application," November 17, 1982.

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. Letter, A. E. Lundvall, Jr. (BG&E) to R. A. Clark (NRC), Docket No. 50-317, " Supplement 1 to sixth Cycle License Application," April 29, 1982.

5. Letter, A. E. Lundvall, Jr. (EG&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.

6. Letter, A. E. Lundvall, Jr. (BG&E) to R. A. Clark (NRC), Docket No. 50-317, " Report of Startup Testing for Cycle Six," October 15, 1982.

7. Letter, A. E. Lundvall, Jr. (BG&E) to R. A. Clark (PRC), Docket No. 50-317. "Fifth Cycle License Application," September 22, 1980.

8. BG&E Calvert Cliffs 1 Slides Depicting SCOUT-1 High Burnup Demonstration Program, presented at BG&E/C-E/NRC meeting in Bethesda, Maryland on December 20, 1978.

9. Letter, J. W. Gore, Jr. (BG&E) to E. G. Case (NRC), Docket No. 50-317.

" Third Cycle License Application," Decmeber 1,1977.

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

11. " Updated Final Safety Analysis Report, Calvert Cliffs Units 1 and 2,"

Docket Nos. 50-317/318.

12. Letter, A. E. Lundvall, Jr. (BG&E) to R. W. Reid (NRC), Docket No. 50-318, " Unit 2 Cycle 2 License Application," July 26, 1978.

13 CENPD-187, "CEPAN Method of Analyzing Creep Collapse of Oval Cladding," June 1975.

14. CEN-182(B)-P, " Statistical Approach to Analyzing Creep Collapse of Oval Fuel Rod Cladding Using CEPAN," September 1981.

15. Letter, D. H. Jaffe (NRC) to A. E. Lundvall, Jr. (BG&E), "Regarding I Unit 1 Cycle 6 License Approval (Amendment #71 to DPR-53 and SER),"

June 24, 1982.

111

16. CEN-183(B)-P, " Application of CENPD-198 to Zircaloy Component Dimensional Changes," September 1981.

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

(BG&E) to V. Stello, Jr. - (NRC) , " Reactor Operation with Modified CEA Guide. Tubes," February 17, 1978.

18. Letter, A. E. Lundvall, Jr. (BG&E) to R. W. Reid (NRC), Docket No. 50-318, " Supplement to Request for Amendment to Operating License,"

September 7, 1978.

19 Letter, A. E. Lundvall, Jr. (BG&E) to R. A. Clark (NRC), Docket No. 50-317, " Report on Fretting Wear Inspection Performed at the End of Cycle 5 on Unit 1," CEN-216(B)-P, September 22, 1982.

20. Letter, R. C. L. Olson (BG&E) to R. W. Reid (NRC), Docket No. 50-318, "ECT Results from Calvert Cliffs Unit 2 Inspection Program," CEN-118(B)-P, November 12, 1979.

21. CENPD-139-P-A, "C-E Fuel Evaluation Model Topical Report," July 1974.

22. CEN-161(B)-P, " Improvement to Fuel Evaluation Model," July 1981.

23 Le tter , R. A. Clark (NRC) to A. E. Lundvall, Jr. (BG&E), "Sa fety Evaluation of CEN-161 (FATES 3)," March 31, 1983.

24. 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.

25. Letter, J. A. Mihalcik (BG&E) to R. A. Clark (NRC), Docket No. 50-317.

" Application of HERMITE Methodology to the Steam Line Break Analysis for Cycle 7," March 28, 1983

26. Letter, 9. A. Clark (NRC) to A. E. Lundvall , Jr. (BG&E), " Positive Conclusion Regarding Acceptability of HERMITE/MacBeth Methodology for Steam Line Break Analysis and SER Covering MacBeth," July 15, 1983

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

April 1983.

28. Le tter , C. O. Thomas (NRC) to A. E. Scherer (C-E), " Acceptance for

! Referencing of Licensing Topical Report CENPD-266-P, CENPD-266-NP 'The ROCS and DIT Computer Codes for Nuclear Design'," April 4, 1983.

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-112 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(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 #71 to DPR-53 and SER),"

June 24, 1982.

6. CEN-124(B)-P, "Statiscal Combination of Uncertainties, Part 2," January 1980.

7. CEN-83(B)-P, "Calvert Clifs Unit 1 Reactor Operation with Modified CEA Guide Tubes," February 8,1978, and letter, A. E. Lundvall, Jr. to V.

Stello, Jr., " Reactor Operation with Modified CEA Guide Tubes,"

February 17, 1978.

8. CENPD-225-P-A, " Fuel and Poison Rod Bowing," June 1983

9. Letter, C. O. Thomas (NRC) to A. E. Scherer (CE), " Acceptance for Referencing of Topical Report CENPD-225(P)," February 15, 1983

10. CEN-124(B)-P, " Statistical Combination of Uncertainties, Part 1,"

January 1980.

11. CEN-124(B)-P, " Statistical Combination of Uncertainties, Part 3," March 1980.

12. Letter, A. E. Lundvall, Jr. (BG&E) to R. A. Clark (NRC), " Amendment to Operating License DPR-69, Fifth Cycle License Application," Docket No.

50-318, October 15, 1982.

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x 3 113 References -' Chapter 7 1a.' Letter, A. E. Lundvall, Jr. (BG&E) to R. A. Clark (NRC), " Amendment to Operating License DPR-69, Fifth.. Cycle License Application," Docket No. 318, October 15, 1982.

1b. Letter, A. E. Lundvall, Jr. (BG&E) to R. A. Clark (NRC), Docket No. 50-318,

" Supplement 1 to Fifth Cycle License Application," November 17, 1982.

2a. " 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 2b. " 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.

2c. " 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.

3 Letter, D. H. Jaffe (NRC) to A. E. Lundvall, Jr. (BG&E), Regarding Unit 1 Cycle 6 License Approval ( Amendments #71 to DPR-53 and SER), June 24, 1982.

4. CENPD-188-A, "HERMITE Space-Time Kinetics," July, 1976.

5. CENPD-161-P, " TORC Code, A Computer Code for Determining the Thermal Margin for a Reactor Core," July, 1975.

6. R. V. MacBeth, "An Appraisal of Forced Convection Burn-Out Data," Proc.

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

7. D. 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.

8. CENPD-206-P, " TORC Code Verification and Simplified Modeling Methods,"

January, 1977.

9. R. E. Henry, H. K. Fauske, "The Two Phase Critical Flow of One-Component Mixtures in Nozzles, Orifices, and Short Tubes," Journal of Heat Transfer, Transactions of the ASME, May, 1971.

,. n 114 References - Chapter 8

1. Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear "TVer Reactors, Fed eral Register, Vol. 39, No. 3, Friday, January 4, 1974.

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

3 Letter, D. H. Jaffe (NRC) to A. E. Lundvall, Jr. (BG&E), Docket No. 50-317. " Amendment to Facility Operating License," June 24, 1982.

4. CEN-161(B)-P, " Improvements to Fuel Evaluation Model," July 1981.

5. Letter, R. A. Clark (NRC) to A. E. Lundvall, Jr. (BG&E), "Sa fety Evaluation of CEN-161 (FATES-3)," March 31, 1983

6. CENPD-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.

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115 Refarrners - Ch ptnr 9

-1. Letter, A. E. Lur.dvall, Jr. (BG&E) to R. A. Clark (NRC), Docket No. 50-317.

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

2. Letter, A. E. Lundvall, Jr. (BG&E) to R. A. Clark (NRC), Docket No. 50-317.

" Supplement 1 to Sixth Cycle License Application," April 29, 1982.

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116

.Refarcne's - Chapttr 10

1. Letter, A. E. Lundvall, Jr. (BG&E) to R. A.~ Clark (NRC), '" Amendment to Operating License DPR-69, Fifth Cycle License Application," Docket No. 50-318. October 15, 1982.

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