ML20247B818

Proposed Tech Specs Supporting Cycle 9 Reload Application

ML20247B818
Person / Time
Site:	Calvert Cliffs
Issue date:	05/08/1989
From:	BALTIMORE GAS & ELECTRIC CO.
To:
Shared Package
ML20247B812	List: ML20247B806 ML20247B818
References
NUDOCS 8905240249
Download: ML20247B818 (30)

Text

_

~

5

)

0

'0 0

1 1

9 4,

0 0

4 0

(

'0

+

I S

A 5

0 3

0

=

3

)

A 0

0 0

2 1

y 0

2 0

(

1 8

0 i

I i

S 0

I A

g 2

8

)

0 6

2 0

1 1

1

' 0 i

2 N

0 I

0 T

1 1

x

(

+

6 1

2 S

7 0

I A

1 5

+

0

)

3 R

0~

3

=

Q

(.

1-1

)

A S

j A

(

2 9

1 8

4

^

2 8

0 g

=

P B

I R

N R A D T V.

0-P_

5 0

g 5

4 3

2 1

O 9

1 1

1 1

1 0,

g A

D$c0D D40E> OS<

C t

>CO> O $O0JD t

t TC2 7Z m

o>qQ othw z4 w

' m y3eg Beg

~

=

'i 1

.)

1

)

(1.2, 1.2) '

1.2 1.1 1.0 (1.0,1.0) 0.9 (0.6, 0.85) 0.8 0.7 OR g

~

0.6 0.5 0.4 ODNB = Ag x ORg 0.3 <

(0.0, 0.3)

AND 0.2 DNB + 17.16 x TIN - 10682 VAR 0.1 0.0 O.0 0.1; 0.2.

0.3 0.4 0.5, 0.6 0.7 0.8, 0.9 1.0 1.1 1.2 FRACTION OF RATED THERMAL POWER FIGURE 2.2-3 THERMAL MARGIN / LOW PRESSURE TRIP SETPOINT PART 2 (FRACTION OF RATED THERMAL POWER VERSUS OR )

j O#%Ars 7tWM hn NA aisam mA A

d

^

l 2.1 SAFETY LIMITS BASES l

2.1.1 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel cladding and possible cladding perforation which could result in the release of fission products to the reactor coolant. Overheating of the fuel is prevented by maintaining the steady state peak linear heat rate at or less than 22.0 kw/ft.

Centerline fuel melting will not occur for this peak linear heat rate. 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 surface temperature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (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 Temperature 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 distributions.

The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local 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 the DNB SAFDL of 1.15 in conjunction with the Extended Statistical Combination of Uncertainties (ESCU). This DNB SAFDL assures with at least a 95 percent probability at a 95 percent confidence level that DNB will not occur.

The curves of Figures 2.1-1, 2.1-2, 2.1-3 and 2.1-4 show conservative loci of points of THERMAL POWER, Reactor Coolant System pressure and maximum cold leg temperature of various pump combinations for which the DNB SAFDL is

_ not violated for the family of axial shapes and 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 limit the maximum value of reactor inlet temperature.

Reactor operation at THERMAL POWER levels higher than 110%

of RATED THERMAL POWER is prohibited by the high power level trip setpoint specified in CALVERT CLIFFS - UNIT 2 B 2-1 Amendment No. JE,77,EJ,99,

SAFETY LIMITS BASES Table 2.1-1.

The area of safe operation is below and to the left of these l

lines.

The conditions for the Thermal Margin Safety Limit curves in Figures 2.1-1, 2.1-2, 2.1-3 and 2.1-4 to be valid are shown on the figures.

The reactor protective system, in combination with the Limiting Conditions for Operation, is designed to prevent any anticipated combination of transient conditions for reactor coolant system temperature, pressure, and THERMAL POWER level that would result in a DNBR of less than 1.15, in conjunction with the ESCU methodology, and preclude the existence of flow instabilities.

2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the

~

' Reactor Coolant System from over pressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.

The reactor pressure vessel and pressurizer are designed to Section III, 1967 Edition, of the ASME Code for Nuclear Power Plant Components which,

permits'a maximum transient pressure of 110% (2750 psia) of design pressure.

The Reactor Coolant System piping, valves, and fittings are designed to ANSI B 31.7, Class I, 1969 Edition, which permits a maximum transient pressure of 110% (2750 psia) of component design pressure.

The Safety Limit of 2750 psia is therefore consistent with the design criteria and associated code requirements.

The entire Reactor Coolant System is hydrotested at 3125 psia to demonstrate integrity prior to initial operation.

CALVERT CLIFFS - UNIT 2 B 2-3 Amendment No. JE,77,EJ,9,

4 LIMITING SAFETY SYSTEM SETTINGS BASES

)

operation of the reactor at reduced power if one or two reactor coolant pumps are taken out of service. The low-flow trip setpoints and Allowable Values for the various reactor coolant pump combinations have been derived in consideration of instrument errors and response times of equipment involved to maintain the DNBR above the DNB SAFDL of 1.15, in conjunction with ESCU methodology, under normal operation and expected transients.

For reactor operation with only two or three reactor coolant pumps operating, the Reactor Coolant Flow-Low trip setpoints, the Power Level-High trip setpoints, and the Thermal Margin / Low Pressure trip setpoints are automatically 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 the DNB SAFDL of 1.15, in conjunction with ESCU methodology, during normal operational transients and anticipated transients when only two or three reactor coolant pumps are operating.

Pressurizer Pressure-Hich The Pressurizer Pressure-High trip, backed up by the pressurizer code safety valves and main steam line safety valves, provides reactor coolant system protection against over pressurization in the event of loss of load without reactor trip. This trip's setpoint is 100 psi below the nominal lift setting (2500 psia) of the pressurizer code safety valves and its concurrent operation with the power-operated relief valves avoids the undesirable operation of the pressurizer code safety valves.

Containment Pressure-Hiah The Containment Pressure-High trip provides assurance that a reactor trip is initiated prior to, or at least concurrently with, a safety injection.

Steam Generator Pressure-Low The Steam Generator Pressure-Low trip provides protection against an excessive rate of heat extraction from the steam generators and subsequent cooldown of the reactor coolant. The setting of 685 psia is sufficiently 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 protection in the event of excessively high steam flow. This setting was used with an uncertainty factor of 85 psi in the accident analyses which was based on the Main Steam Line Break event.

l

\\

CALVERT CLIFFS - UNIT 2 B 2-5 Amendment No. JE,77,EJ,99, SS, j

)

4.

LIMITING SAFETY SYSTEM SETTINGS BASES Steam Generator Water Level The Steam Generator Water Level-Low trip provides core protection 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. The specified setpoint in combination with the auxiliary feedwater actuation system ensures that sufficient water inventory exists in both steam generators to remove decay heat following a loss of main feedwater flow event.

Axial Flux Offset The axial flux offset trip is 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 the DNB SAFDL of 1.15, in conjunction with ESCU methodology, nor a peak linear heat rate which corresponds to the temperature for fuel

-centerline melting will exist as a consequence of axial power maldistributions.

These trip setpoints were derived from an analysis of many axial power shapes with allowances for instrumentation inaccuracies and the uncertainty associated with the excore-to-incore axial flux offset relationship.

~

\\

Thermal Marain/ Low Pressure The Thermal Margin / Low Pressure trip is provided to prevent operation when the DNBR is less than the DNB SAFDL of 1.15, in conjunction with ESCU methodology.

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

In addition, CEA group sequencing in accordance 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 2 B 2-6 Amendment No. JE,77,EJ,9p,

3/4.1 REACTIVITY CONTROL SYSTEMS

'3/4.1.1 BORATION CONTROL SHUTDOWN MARGIN - Tava

> 200*F LIMITING CONDITION FOR OPERATION 3.1.1.1 "The SHUTDOWN MARGIN shall be equal to or greater than the limit line of Figure 3.1-lb.

APPLICABILITY:

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

ACTION:

With the SHUTDOWN MARGIN less than the limit line of Figure 3.1-lb, immediately initiate and continue boration at 2 40 gpm of 2300 ppm boric acid solution or equivalent until the required SHUTDOWN MARGIN is restored.

SURVEILLANCE REQUIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be equal to or greater than the limit line of Figure 3.1-lb:

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 immovable or untrippable, the above required

~

SHUTDOWN MARGIN shall be increased by an amount at least equal to the withdrawn worth of the immovable 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.

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 criti-cality by verifying that the predicted critical CEA 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 l

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.

With Keff 2 1.0.

With Keff <

1.0.

CALVERT CLIFFS - UNIT 2 3/4 1-1 Amendment No.99,JpE,

11

)t l

,1I I

s n

C O

)

E 0

5

,CO E

(

N IG RA M

NW OD TUHS E

b L

M C

1 U

Y 1

M C

3 IN N

E I

I R

M E

U M

G I

I T

F E

E L

L N BN BO N AO N I

ATO TI TO TAI P AI P R G E RG 1

E E E C EE C P R CPR CO AO A

NU i

)

5 3

,CO B

C

(

O 0

0 0

0 0

0

0. _ B 6

5 4

3 2

1 0

Q ye$ z3oOlS*

rj*

cg m

{.h=

D3$o3oE zO.

o>Em"

2 0 7 i

8

, 8 2

4 06 4 8

5 1

6 0

0 i

1

, 00 8

4 1

8 8

2 0

08 8 7 i

i 20 2

1 0

4 0

I

, 06 6 4 3

5 1

2 0

0 6

0 6

00 0

i 6

1 4

8 1

3 3.

8 N

2 08 W

P P

0 i

G G

, 8 7

, 20 2

0, 2

1 N A O R I

D 0 4 0

TH 0

0 06 RT 64 3

EI i

5 SW 1

3 n

N A I

6 0

8 1

, 00 AE U

0 0

4 EC G

5 8

1 IF C S E

8 2

4 H

08 0

i

% P 20 8 7 C

8 i

5 G

2 N

1 I

8 T

EIM' 0

4 6

5 0

06 TI 0

3 i 64

%5 0

MAL 8

i 5

0 RT 1

P ESN2 8

6

%@G T

O TDT @

YI 0

i 0 6 5

00 41 5

8 0 56 RA R4 5

1 8

E P 5

OE P 0 S R blS HT

@p 8

S N G 5

2 i 2 0 0

0 8_

I 5

5 g, J

U '-

2 1_

5 P~

8 7 Il l 8

I 3

3 PO G

I R

0 T

O EM 5

TI %

G 0 4 5.

5 O6 AL5 64 3

1 i

P P 7

MT 2lIl l I I i1 5

G_

G 0

RS N

+

1 5

E O@

,0 9 T YI 6

D T SN 0

5 1

0 GA R I

1 4

NE EP SO 8

OT SR SI A G NG LS I L

B E 8

l l l l I l I g g R

08 I

E 20 LG 1

BN I

A T 0

WA S 06 OR P 3

0. _

9 8

7, 6

5 4

3 2

1

0. _ L E U 1

L P O 1

0 0

o 0

0 0

0 0

0 O AO RG g30n. a.[E% ldgoj< g bPOm"-

u nn f

QEr:H @@7 cE[

w*4 y3go3$ i;. =

l I

l

E 1

l0

{

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 1

A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made suberitical from all operating conditions, 2) the reactivity transients associated with postulated accident conditions are controllable within ac-ceptable limits, and 3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

The most limiting SHUTDOWN MARGIN requirement at.beginning of cycle is determined by the requirements of several transients, including Boron Dilution and Steam Line Rupture. The SHUTDOWN MARGIN requirements for these transients are relatively small and nearly the same. However, the most limiting SHUTDOWN MARGIN requirement at end of cycle comes from just one transient, the Steam Line Rupture event.

The requirement for this transient at end of cycle is significantly larger than that for any other event at that time in cycle and, also, considerably larger than the most limiting requirement at beginning of cycle.

The variation in the most limiting requirement with time in cycle has been incorporated into Technical Specification 3.1.1.1, in the form of a specified SHUTDOWN MARGIN value which varies linearly from beginning to end of cycle.

This variation in specified SHUTDOWN MARGIN is conservative relative to the actual variation in the most limiting requirement. Consequently, adherence to Technical Specification 3.1.1.1 provides assurance that the available SHUTDOWN MARGIN at any time in cycle wi.ll exceed the most limiting SHUTDOWN MARGIN requirement at that time in cycle.

In MODE 5, the reactivity transients resulting from any event are minimal and do not vary significantly during the cycle. Therefore, the specified SHUTDOWN MARGIN in MODE 5 via Technical Specification 3.1.1.2 has been set equal to a constant value which is determined by the requirement of the most limiting event at any time during the cycle, i.e., Boron Dilution with the pressurizer level less than 90 inches and the sources of non-borated water restricted. Consequently, adherence to Technical Specification 3.1.1.2 provides assurance that the available SHUTDOWN MARGIN will exceed the most limiting SHUTDOWN MARGIN requirement at any time in cycle.

CALVERT CLIFFS - UNIT 1 B 3/4 1-1 Amendment No. J////p//77, yg//195,

o O

e 3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 BORATION CONTROL 3/4.1.1.3 BORON DILUTION A minimum flow rate of at least 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 24 minutes. The reactivity change rate associated with boron concentration reductions will therefore be within the capability of operator 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 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-concentration associated with fuel burnup. The confirmation 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-la Amendment No. J///fE//77, 95//195,

1.3

~

~

~~

(0.0,1.17)

UNACCEPTABLE UNACCEPTABLE OPERATION OPERATION REGION REGION 1.1

(-0.18,1.0)

(0.2,1.0) 1.0 6

m0 0.9 s

l ACCEPTABLE OPERATION ul 0.8 N

PEGION 8

~

~

0.7 8

2Op 0.6 k

E 0.5

(-0.6, 0.4)

(0.6, 0.4) 0.4 l

l l

-0.3 1

0.2 i

I I

I I

I 0.15

-0.8

-0.6

-0.4

-0.2 0.0 0.2 0.4 0.6 0.0 PERIPHERAL AXtAL SHAPE INDEX, Y g FIGURE 2.2-1 PERIPHERAL AXIAL SHAPE INDEX, Y g vs. FRACTION OF RATED THERMAL POWER CALVERT CLIFFS - UNIT 2 2-11 Amendment No. 61,108 L___-___-________-__

REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION e.

With one or more CEAs misaligned from any other CEAs in its group by more than 7.5 inches but less than 15 inches, operation in MODES 1 and 2 may continue, provided that within one hour the misaligned CEA(s) is either:

1.

Restored to OPERABLE status within its above specified alignment requirements, or 2.

Declared inoperable. After declaring the CEA inoperable, operation in MODES 1 and 2 may continue for up to 7 days per occurrence with a total accumulated time of s 14 days per calendar year provided all of the following conditions are met:

a.

The THERMAL POWER level shall be reduced to s /0% of the maximum allowable THERMAL POWER level for the existing Reactor Coolant Pump combination within one hour; if negative reactivity insertion is required to reduce THERMAL POWER, boration shall be used.

b.

Within one hour after reducing the THERMAL P'0WER as required by (a) above, the remainder of the CEAs in the group with the inoperable CEA shall be aligned to within 7.5 inches of the inoperable CEA while maintaining the allowable CEA sequence and insertion limits shown on Figure 3.1.2; the THERMAL POWER

~

level shall be restricted pursuant to Specification

~

3.1.3.6 during subsequent operation.

f.

With one CEA misaligned from any other CEA in its group by 15 inches or more, operation in MODES 1 and 2 may continue, provided that the misaligned CEA is positioned within 7.5 inches of the other CEAs in its group in accordance with the time allowance determined by the Better Axial Shape Selection System (BASSS) or, if the BASSS time allowance is unavailable, the time allowance shown in Figure 3.1-3.

If Figure 3.1-3 is used, the pre-misaligned FJ value used to determine the allowable time to realign the CEA from Figure 3.1-3 shall be the latest measurement taken within 5 days prior to the CEA misalignment.

If no measurements were taken within 5 days prior to the misalignment, a pre-misaligned FJ of 1.70 shall be assumed.

g.

With one CEA misaligned from any other CEA in its group by 15 inches or more at the conclusion of the permitted time allowance, immedi-ately start to implement the following actions:

1.

If the THERMAL POWER level prior to the misalignment was greater than 50% of RATED THERMAL POWER, THERMAL POWER shall be reduced to less than the greater of:

CALVERT CLIFFS - UNIT 2 3/4 1-18 Amendment No. R9,

REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION a) 50% of RATED THERMAL POWER b) 75% of the THERMAL POWER level prior to the misalignment within one hour after exceeding the permitted time allowance.

2.

If the THERMAL POWER level prior to the misalignment was s 50% of RATED THERMAL POWER, maintain THERMAL POWER no higher than the value prior to the misalignment.

If negative reactivity insertion is required to reduce THERMAL POWER, boration shall be used. Within one hour after establishing the appropriate THERMAL POWER as required above, either:

1.

Restore the CEA to within the above specified alignment requirements, or 2.

Declare the CEA inoperable. After declaring the CEA inoperable, POWER OPERATION may continue for up to 7 days per occurrence with a total accumulated time of s 14 days per calendar year provided the remainder of the CEAs in the group with the inoperable CEA are aligned to within 7.5 inches of the inoperable CEA while maintaining the allowable CEA sequence and insertion limits shown on Figure 3.1-2; the THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation.

h.

With more than one CEA inoperable or misaligned from any other CEA in its group by 15 inches (indicated position) or more, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

i.

For the purposes of performing the CEA operability test of TS 4.1.3.1.2, if the CEA has an inoperable position indication channel, the alternate indication system (pulse counter or voltage dividing network) will be used to monitor position.

If a direct position indication (full out reed switch or voltage dividing network) cannot be restored within ten minutes

- from the commencement of CEA motion, or CEA withdrawal exceeds the surveillance testing insertion by > 7.5 inches, the position of the CEA shall be assumed to have been > 15 inches from its group at the commencement of CEA motion.

CALVERT CLIFFS - UNIT 2 3/4 1-19 Amendment No. J7,E7,Jpf, l

t

0 7

~

1 R

i O

T C

)

0 A

F 7

e G

6 1

N I

(

KA i

EP L

R 5

A l

F 1 6 D

I 0

1 A

0 6

R D

i 2

E 0

T 0

A 1

RG i

ET N

I LATO T

i DEN G

0 IL 6

A i

g 1

S N

I O

M I

i G

ER E

P R

D D

E e

E R

W-U~

)-

~

0 O

S 6

L A

L i

7 A

EM 5

1

(

5 5

1 0

0 0

0 0

0 0

7 6

5 3

2 1

AMw>-

l

> wet E <u[9Yw J

O am $ OCmm* Cz4 N g aae 3$a!a z0, i

l'

!I it

.J I'i l,l l'

m 00

~

~

5 00 P

I 4

UNRUB

.sv N

E O

N S

T O

Y A

IT A

R I

A T

R A

D T

E R

0 1

A 0

R P

E 3

E 2 E i

O P

H W

3 O

E O

E R L

E t

B L

t P

R A A

B L

U EN L

GI T

A P

T U

IF L E

P F

K C

E E

A C

C V

E A

C T

P I

N A

0 C

E U

.' 0 i

2 E

L F

B F

A E

WO LLA 00 i

1 0

6 2

5 4

3 1

1 1

1 5

1 EC

$Ol=$aOj$w~

E gA'3M. WEE kE E<WE' M<E wam43OaJ<

Otm m ' Cz4 N Ea Yu

>9I ya3 zo * *.

O>Qm:t D-i r

l lllljIlll

,l' l

.l<

l

i1

]l l'

t1;

~

0 8

1

)

8 0

5 8

7 5

7 1

8 I

(

1 0

7 I

1 4

N

.sv RO TC 5

A F

6 I

G 1

N I

b K

3-A 2 E E

Y U

3 P L

TXF E L A

R A V

0 U D I

6 G A E

I I

L 1

F R BA R

T A

P N

E A

C L

C P

A L

5 A

5 T

I O

1 T

)0 1

05 8

I 0

1 5

1(

5 4

0..

9 8

7 6

5 1

1 0

0 0

0 0

N O>Er:y Or9- () hah N P=MA=

>ge3o3e?. zo,s',o nD nf lllll

+

POWER DISTRIBUTION LIMITS TOTALPLANARRADIALPEAKINGFACTOR-F}y LIMITING CONDITION FOR OPERATION The ' calculated value of F[y shall be limited to 11.70.

3.2.2.1 APPLICABILITY:

MODE 1*.

ACTION:

WithF}y > 1.70, within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> either:

a.

Withdraw and maintain full length CEAs at or beyond the Long Term Steady State Insertion Limits of Specification 3.1.3.6 and reduce THERMAL POWER as follows:

1.

Reduc 9THERMALPOWERtobringthecombinationofTHERMALPOWER and F within the limits of Figure 3.2-3a, or xy 2.

Reduce THERMAL POWER to less than or equal to the limit estab-lished by the Better Axial Shap2 Selection System (BASSS) as a function of FT ; or xy b.

Be in at least HOT STANDBY.

SURVEILLANCE REQUIREMENTS 4.2.2.1.1 The provisions of Specification 4.0.4 are not applicable.

is determinId with a non-full core power distribution }5ppiNh s(y+st0m) w T

4.2.2.1.2 F shall be calculated by the expression F

=F 1T m

be calculated as FT-F when determined with a full core power distribution I# sN511 be determined to be within its limit at the mapping system.

F followingintervalI5 a.

Prior to operation above 70 percent of RATED THERMAL POWER af'er each fuel loading, b.

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

Within four hours if the AZIMUTHAL POWER TILT (T ) is > 0.030.

q

• See Special Test Exception 3.10.2.

~

CALVERT CLIFFS - UNIT 2 3/4 2-6 Amendment No. 9,JE,7J,JJ, 1

E ____-

I POWER DISTRIBUTION LIMITS 3

SURVEILLANCE REQUIREMENTS (Continued) 4.2.2.1.3 FT shall be determined each time a calculation is required by usine 67 incore detectors to obtain a power distribution map with all full length Ofis at or above the Long Term Steady State Insertion Limit for the existing Reactor Coolant Pump combination. This determination shall be limited to core planes between 15% and 85% of full core height inclusive and shall exclude regions influenced by grid effects.

T is made 4.2.2.1.4 T shall be determined each time a calculation of F usinganon-fu11copepowerdistributionmappingsystemandthevaYdeofT 9

q used to determine F shall be the measured value of T.

xy q

1 l

l CALVERT CLIFFS - UNIT 2 3/4 2-7 Amendment No. f,JJ,gJ, 1

POWER DISTRIBUTION LIMITS TOTALPLANARRADIALPEAKINGFACTOR-F[y LIMITING CONDITION FOR OPERATION 3.2.2.2 The value of N presently used in Specification 4.2.1.3 shall be in accordance with Figure 3.2-3b.

APPLICABILITY: H0DE 1 when operating in accordance with Specification 4.2.1.3.

ACTION:

With the value of N presently used in Specification 4.2.1.3 exceeding the limit shown in Figure 3.2-3b, within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> either:

a.

Reduce the value of N used in Specification 4.2.1.3 to.within the limits of Figure 3.2-3b; or l

b.

Be in at least HOT STANDBY.

SURVEILLAt'CE REQUIREMENTS 4.2.2.2.1 The provisions of Specification 4.0.4 are not applicable.

T T

4.2.2.2.2 F shall be calculated by the expression F

=F 1T when F l

is determinN5 with a non-full core power distribution In5ppinl#sy(s+te0)and shaI5 T

-F l

distribution mappink systN5.when determined with a full core power be calculated as F N shall be determined to be within its limit by T

monitoring F at the following intervals:

xy a.

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

At least once per 3 days of accumulated operation in H0DE 1.

l CALVERT CLIFFS - UNIT 2 3/4 2-8 Amendment No. 9,JE,JJ,EJ, L _ _-_ _ _____-_____ __________-_____

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) using the incore[iletectors to obtain a power distribution map with all full 4.2.2.2.3 F

shall be determined each time a calculation is required by length CEAs at or above the Long Term Steady State Insertion Limit for the existing Reactor Coolant Pump combination. This determination shall be limited to core planes between 15% and 85% of full core height inclusive and shall exclude regions influenced by grid effects.

T is made 4.2.2.2.4 T shall be determined each time a calculation of F usinganon-ful1cogepowerdistributionmappingsystemandthevaNeofT 9

4 used to determine F chall be the measured value of T.

xy q

CALVERT CLIFFS - UNIT 2 3/4 2-8 (cont.)

Amendment No. S,JE,EJ,

o j

1 POWER DISTRIBUTION LIMITS TOTALINTEGRATEDRADIALPEAKINGFACTOR-Ff LIMITING CONDITION FOR OPERATION 3.2.3

' The calculated value of Ff shall be limited to s 1.70.

)

APPLICABILITY: MODE 1*,

ACTION:

utu WithFf>1.70,within6hourseither:

a.

Be in at least HOT STANDBY, or b.

Withdraw and maintain the full length CEAs at or beyond the Long Term Steady State Insertion Limits of Spec'fication 3.1.3.6 and reduce THERMAL POWER as follows:

1.

Reduc THERMAL POWER to bring the combination of THERMAL POWER and F within the limits of Figure 3.2-3c, or 2.

Reduce THERMAL POWER to less than or equal to the limit estab-lished by the Better Axial Shape Selection System (BASSS) as a T

function of F.

r When the THERMAL POWER is determined from Figure 3.2-3c, it shall be used to establish a revised upper THERMAL POWER LEVEL limit on Figure 3.2-4 (i.e., Figure 3.2-4 shall be truncated at the allowable fraction of RATED THERMAL POWER determined by Figure 3.2-3c).

Subsequent operation shall be maintained within the reduced accept-able operation region of Figure 3.2-4.

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 - Fr (1+T is determinedwithanon-fullcorepowerdistributionmappingsyst0m)whenF and shall be mappingsystem.fFfshallbedeterminedtobewithinitslimitatthefollow-calculated as F F when determined with a full core power distribution ing 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 AZIMUTHAL POWER TILT (T ) is > 0.030.

q

• See Special Test Exception 3.10.2.

1 CALVERT CLIFFS - UNIT 2 3/4 2-9 Amendment No. 9,Jg,JE,JJ, pf,

SURVEILLANCE REQUIREMENTS (Continued) 4.2.3.3 Ff shall be determined each time a calculation is required by using the incore detectors to obtain a power distribution map with all full length CEAs at or above the Long Term Steady State Insertion Limit for the existing Reactor Coolant Pump combination.

shall be determined each time a calculation of Ff is made using a 4.2.3.4 T

non-full cofe power distribution mapping system and the value of T used to q

shall be the measured value of T.

determine Fr q

CALVERT CLIFFS - UNIT 2 3/4 2-10 Amendment N'o. 9,JE,J3,JJ, EJ,

__-____-.u_m____

0

)

N 8

8 I

I O

0 1

I G

E 5

E V

R R

8 U

7 N

C 1

(

O I

T T

I A

M I

R L

EP r

O F

EL BATPECC 5

A 7

I I

N 1

U N

i' O

IG E

r R

F NO ITA REP 0

O

)

1 E

0 L

7 I

l B

0 1

7 AT 1

P

(

ECCA i

5 6

0 9

8 7

6 51 1

0 0

0 0

0 ew3on.s.<E Ez> owt m Eo $b4aw Wan<hs4 e

p9 M=Ci u

Sa" O gga9E. zo O>EQ n

E4

{

l POWER DISTRIBUTION LIMITS DNB PARAMETERS LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB related parameters shall be maintained within the limits shown on Table 3.2-1:

a.

Cold Leg Temperature b.

Pressurizer Pressure c.

Reactor Coolant System Total Flow Rate d.

AXIAL SHAPE INDEX, THERMAL POWER APPLICABILITY: MODE 1.

ACTION:

With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

~

SURVEILLANCE RE0VIREMENTS 4.2.5.1 Each of the parameters of. Table 3.2-1 shall be verified to be within their limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

472.5.2 The Reactor Coolant System total flow rate shall be determined to be within its limit by measurement at least once per 18 months.

4.2.5.3 The Better Axial Shape Selection System (BASSS) may b'e used for monitoring THERMAL POWER as a function of AXIAL SHAPE INDEX.

BASSS monitoring shall be limited to CEA insertions of the lead bank 5 55%.

CALVERT CLIFFS - UNIT 2 3/4 2-13 Amendment No. 9,JE,EJ,

---

m_

CPO A

P ETP RNO A -

%y OLG 5b WON n

5 TOI D

a y

d CT E

h b5e A

T t

i R

AD dff E

RE s

eoi P

T s

h c

O fA e

ske oR l

inp l as

%f h

bb 5o t

a s

P i

tdn O

f%

w sao O

o0 eei L

1 n

lt S

s o

s r

RPE sn i

t ee OMM ea t

ih s TUA ch a

mt n CPS xt r

i i

A e

e lf ET -

r p

oA RN ne o

e E

AG it h sC OLN a

r t n WOI ee o

or TOT sr f

nio CA ag itf R

e s

hr E

rf e

t e4 P

co s

is -

O n

y wn2 e

l i

i s

a d

3 pa n

eA RS me a

NEE OPj ar iCR TMG rc S

a U

S CUN n

C t rG 1

R API Ri C

noI E

E' T

E E

ifF 2

T RTA Wp a

E NR O e f

m)f 3

M EAE Pt o

S o A

ELP s

eS E

R ROO L

l bSs L

A HO AR a

At B

P TC ME v

l Bi A

RW o

l ( m T

B EO r

a i

N HP p

h ml D

T p

se L

a.

t e RP p

aA g

Rsh S

m M

C n E yt OMG a

g rR Ri WS TUN F

i eE Nt O

n CPI s

0 hH a

P ni A

T p

0 tT gr oh ETA 8

0, i

ne L' i t i p Ati RNR 4

0 ea AE 5

0 0

d o Mcw RLP 2

7 gr n

Re UOO 1

2 3

no es El r i

pp Heo OO FC 2

2 re m

TS ut ku du np

,eE n

a X pL ei l t EaB l m bn DhA b

a NSR ar tl I

E m

ce f o l P e

ip eo EaO e

e t

l l c Pi r

r s

pRR Axs PEE sr HAi2 u

u ye t

s St X

aWW eo S

a s

a ER OO ut rS1 r

e tR DE tPP l c LeS e

r n

NW o

aa AtS3 p

P aw IO nLL ve ItA m

l o P

AA r

XeBE e

r ol E

tMM e

AB R

T e

oF PL iRR sr nU R

z C

AA mEE eu eeeG E

g i

l HM iHH h o hhhI T

e r

ra SR LTT Tf Tt wF E

l u

ot E

M s

t o LH A

d s

cT AT R

l e

a I

A o

r e

X P

C P

R A

>G9.PMG h b ?%

>j a.Po A 2 ? ? M i

e 9

REACTIVITY CONTROL SYSTEMS BASES Overpower margin is provided to protect the core in the event of a large misalignment (2 15 inches) of a CEA.

However, this misalignment would cause distortion of the core power distribution. The reactor protective system would not detect the degradation in radial peaking factors and since varia-tions in other system parameters (e.g., pressure and coolant temperature) may not be sufficient to cause trips, it is possible that the reactor could be operating with process variables less conservative than those assumed in generating LC0 and LSSS setpoints. The ACTION statement associated with a large CEA misalignment requires prompt action to realign the CEA to avoid excessive margin degradation.

If the CEA is not realigned within the given time constraints, action is specified which will preserve margin, including reductions in THERMAL POWER.

For a single CEA misalignment, the time allowance to realign the CEA (Figure 3.1-3 or as determined by BASSS) is permitted for the following reasons:

1.

The margin calculations which support the power distribution LCOs forDNBRarebasedonasteady-stateF)asspecifiedinTechnical Specification 3.2.3.

2.

When the actual FT is less than the Technical Specification value, additional margin exists.

3.

This additional margin can be credited to offset the increase in F) with time that will occur following a CEA misalignment due to xenon redistribution.,

The requirement to reduce power level after the time limit of Figure 3.1-3 or the ime limit determined by BASSS is reached offsets the continuing increase in F that can occur due to xenon redistribution. A power reduction is not requir d below 50% power.

Below 50% power there is sufficient conservatism in the DNB power distribution LCOs to completely offset any, or any additional, xenon redistribution effects.

The ACTION statements applicable to misaligned or inoperable CEAs include requirements to align the OPERABLE CEAs in a given group with the inoperable CEA. Conformance with these alignment requirements bring the core, within a short period of time, to a configuration consistent with that assumed in generating LC0 and LSSS setpoints. However, extended operation with CEAs significantly inserted in the core may lead to perturbations in 1) local burnup, 2) peaking factors, and 3) available shutdown margin which are more adverse than the conditions assumed to exist in the safety analyses and LC0 and LSSS setpoints determination. Therefore, time' limits have been imposed on operation with inoperable CEAs to preclude such adverse conditions from developing.

CALVERT CLIFFS - UNIT 2 B 3/4 1-4 Amendment No. E7,7J, 19),

POWER DISTRIBUTION LIMITS -

BASES and Local Power Density - High LCOs and LSSS setpoints remain valid. An AZIMUTHAL POWER TILT > 0.10 is not expected ~and if it sho'.1d occur, subsequent operation would be restricted to only those operations required to identify the cause of this unexpected tilt.

thatmustbeusedintheequationF[y-Fxy (1 + T ) and Ff The value of Tr (1 + T ) is the measured tilt.

q q

F q

The surveillance requirements for verifying that Ffy, F[ and T are q

within their limits provide assurance that the actual values of F[y, Ff and Verifying F[y and Ff after each fuel T do not exceed t' assumed values.

q loading prior to exceeding 75% of RATED THERMAL POWER provides additional assurance that the core was properly loaded.

3/4.2.5 DNB PARAMETERS The limits on the DNB related parameters assure that each of the parame-ters are maintained within the normal steady state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the safety analyses assumptions and have been analytically demonstrated adequate to maintain a minimum DNB SAFDL of 1.15 throughout each analyzed transient.

In addition to the DNB criterion, there are~two other criteria which set the specification in Figure 3.2-4.

The second criterion is t6 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 criterion 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 criterion for this condition because the linear heat rate LCO, for both ex-core and in-core monitoring, is set to maintain only the LOCA kw/ft requirements which are limiting at high power levels. At reduced power levels, the kw/ft requirements of certain A00s (e.g., CEA withdrawal), tend to become more limiting than that for LOCA.

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 instrument readout is sufficient to ensu_re 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' hour basis.

l CALVERT CLIFFS - UNIT 2 8 3/4 2-2 Amendment No. JE,7J,JE, 57,99,

5 1

PLANT SYSTEMS BASES 3/4.7.1.2 (Continued)

(3)

Main Steam Line Break 1550 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 requirad by the existing plant condition.

CALVERT CLIFFS - UNIT 2 -

B 3/4 7-2a Amendment No.7E,

L a

RESIGN' FEATURES DESIGN PRESSURE AND TEMPERATURE 5.2.2 The reactor containment building is designed and shall be maintained for a maximum internal pressure of 50 psig and a temperature 6~f 276*F.

5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 217 fuel assemblies with each fuel assembly containing a maximum of 176 fuel rods clad with Zircaloy-4.

Each fuel rod shall have a nominal active fuel length of 136.7 inches and contain a maximum total weight of 3000 grams uranium. The initial core loading shall have a maximum enrichment of 2.99 weight percent U-235.

Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum enrichment of 4.35 weight percent U-235.

5.3.2 Except for special test as authorized by the NRC, all fuel assemblies under control element assemblies shall be sleeved with a sleeve design previously approved by the NRC.

CONTROL ELEMENT ASSEMBLIES 5.3.3 The reactor core shall contain 77 full length and no part length control element assemblies.

5.4 REACTOR COOLANT SYSTEM-DESIGN PRESSURE AND TEMPERATURE 5.4.1 The reactor coolant system is designed and shall be maintained:

a.

In accordance with the code requirements specified in Section 4.2 of the FSAR with allowance for normal degradation pursuant of the applicable Surveillance Requirements, b.

For a pressure of 2500 psia, and c.

For a temperature of 650*F, except for the pressurizer, which is 700*F.

l l

CALVERT CLIFFS - UNIT 2 5-4 Amendment No. JE,JJ,EJ, I

____:-____.__________.______._.