ML20149F397

Proposed Tech Specs,Supporting Reload Core for Unit 1 Cycle 10 & Changing Figures to Broaden Range of Core Operability at Lower Power Levels Re Axial Shape Offset Limits for Unit 2

ML20149F397
Person / Time
Site:	Calvert Cliffs
Issue date:	02/12/1988
From:	BALTIMORE GAS & ELECTRIC CO.
To:
Shared Package
ML20149F376	List: ML20149F369 ML20149F397
References
NUDOCS 8802170130
Download: ML20149F397 (73)

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FIGURE 2.2 2 Thermal Margin / Low Pressure Trip Setpoint N

e.

Part 1 (ASI Versus A )

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0.0 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 QR )

g i

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CALVERT CLIFFS UNIT 1 2 13 Amendment No.

9-11

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+ 17.16 x T 10682 lDN3 l

l 7y VAR 0.10 0.00 0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0,8 o,9 1.0 1.1 1.2 FRACTION OF RATED THEV.AL POWER FICW E 2.2 3 Ther:nal Margin /Lov Pressure Trip se:poin:

28r: 2 (Frac:fon ot RATED THE AL PCVER versus QR )

V CALVERT c;;773. tl NIT 1 2,13 Amendment tio.

1

/An DAlb SArDL whiGdA8 &

& jr. d EN-3 M(8)- P 2.1 SAFETY LIMITS BASES 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 coolar,t. 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 j

for this peak linear heat rate. Overheating of the fuel cladding is f

prevented by restricting fuel operation to within the nucleate boiling I

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 reaime 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 THER. L POWER and Reactor Coolant Temper-V ature and Pressure have been related to DNS through the CE-1 correlation.

The CE-1 DNB correlation has been developed to predict the DNB flux and the lo'ce. tion of DNB for axially uniform and non-uniform heat flux distri-The local DNB heat flux ratio, DNBR, defined as the ratio of l

butions.

the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.

,-*CONB SA FDL)

[ The minimum value of the DNBR during steady state operation, nomal I

ope ational transients, and anticipated transients is limited to 1.2.

This va u

--- > " a 95 percent probability at a 95 percent con-che:r : er ppr0prf:tc fidencE'. evel fihat DN,B will not occur. erd ':

~/L._

___o.,__.

7... s.. w _.,.,

_.,,4.

/

The curves of Figures 2.1-1 2.1-2, 2.1-3 and 2.1-4 show the trMC& 5 loci of points of THERMAL POWER,,eactor Coolant System pressure and maximum cold leg temperature of v rious pump combinations for which the I

nie- ?!EP f:

k:: th:r 1.21 for the family of axial shapes and corresponding radial peaks shown in Figure 82.1-1.

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 teaperature.

Reactor operation at THERML POWER levels higher than 110% of RATED THERMAL POWER is prohibited by the high power level trip setpoint specified in s

.w P. 4 A-W N A of**$

S a.

-~

CALVERT CLIFFS - UNIT 1 B 2-1 Amendment No. 33,/8,U,EB,10/30/86 9-12

2.1 SAFETY LIMITS BASES 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 presented 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 whue 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 consistent with the methods described in CEN-348(B)-P.

This DNB SAFDL, when used in conjunction with the appropriate uncertainty allowance, 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 82.1-1.

The limits in Figures 2.1-1, 2.1-2, 2.1-3, ard 2.1-4 were calculated for reactor coolant inlet temperatures less than or equel 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 I B 2-1 Amendment No. JJ///E///J, BB//19/J9/ts,

pG ONB SAFDL gg A een-una)-F SAFETY LIMITS

~

BASES Table 2.1-1.

The area of safe operation is belcw and to the left of these 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, pressu and THEFJtAL POWER 1evel that would result in a MBR of less than and preclude the existence of flow instabilities.

l 2.1. 2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolaat System from overpressurization 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 1 B 2-3 Amendment No. 33,39,48,7J,10 /30/86 9-13

o l

SAFETY LIMITS BASES Table 2.1.-l.

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

The conditions for the Thermal Mcrgin 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 the ONB SAFDL consistent with the methods described in CEN-348(B)-P 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 overpressurization 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 pre;sure 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 3215 psia to demonstrate integrity prior to initial operation.

1 CALVERT CLIFFS - UNIT 1 B 2-3 Amendment No. JJ//JJ///E, 111/19/19/99,

p DNB SAfDL h Y Y y p ;

sen-s w e)-p LIMIT!NG SAFETY SYSTEM SETTINGS BASES j

operation of the reactor at reduced power if one or two reactor coolant pumps are taken out of service.

he low-flow trip setpoints and Allowable Values for the various reactor coo lant pump combinations have been derived in consideration of instrument rors and response times of equipment involved to maintain the DNBR above

,.2 under normal operation and expected transients. I For reactor operation with on y 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 changed when the pump condition selector switch is manually set to the desired two-or three-pump position. Changing these trip setpoints during two an three pump operation prevents the minimum value of DNBR from going below 1.21 l

during normal operational transients and anticipated transients when only wo or three reactor coolant pumps are operating.

Pressurizer pressure-High 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 overpressurization 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 opera-tion of the pressurizer code safety valves.

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

j 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 l

with normal operation, but still high enough to provide the required protec-i tion in the event of excessively high steam flow. This setting was used with en uncertainty factor of + 85 psi which was based on the main steam line break event inside containment!

CALVERT CLIFFS - UNIT 1 B 2-5 Amendment No. 33, AB,77,38, yy7,10/30/86 9-14

o 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 consistent with the methods described in CEN-348(B)-P under normal operation and expected transients.

For reactor operation witi, 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 DNB SAFDL consistent with the methods described in CEN-348(B)-P during normal operational transients and anticipated transients when only two or three reactor coolant pumps are operating.

Pressurizer Pressure-Hiah 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 overpressurization 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 valvet.

_ Containment Pressure-Hiah The Containment Pressure-High trip provides assurance that a reactor trip is initiated prior to, or at least concurrertly 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 which was based on the main steam line break event inside containment.

CALVERT CLIFFS - UNIT 1 B 2-5 Amendment No. JJi//51/7J, 951/1171/19/19/55, I

Ms. DNB SAf 0L LIMITING SAFETY SYSTEf1 SETTINGS M MM O~ 3 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 ensure that excessive axial peaking will not cause fuel damage.

The axial flux offset is determined from the axially split excore ectors.

The trip setpoints ensure that neither a ONBR of less than 1.2 nor a peak linear heat i

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 associate.d with the excore to incore axial flux offset Yelationship.

Thormal Marcin/ Low Pressure The Themal Margin / Low P essure trip is provided to prevent operation when the DNBR is less than

.2.

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

The computed value is a function of the higher of aT power or neutron power, reacto-fnlet 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 pemitted 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, he 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. 33,39,/3,7J,38,10/ 30/86 9-15

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 consistent with the methods described in CEN-348(B)-P 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 consistent with the methods described in CEN-348(B)-P.

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 opercting. 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 1 B 2-6 Amendment No. JJ//19///E, 11//55//19/19/B%,

t 3/a.1 REACTIVITY CONTROL SYSTE!'S 3/a.1.1 50 RATION CONTROL SHUTDOWN MARGIN Tavg " 200 LIMITING CONDITION FOR OPER:T!ON 3.1.1.1 The SHUT 00WN MARGIN shall be.>y d h o f' J ' N M[h AFFLICABILITY:

PCES 2**, 2 and 4.

ACTION:

& :t k ( Q 3.1-Iby With the SHUTDOWN MARGIN.0 AX4sr.cmficriediately initiate and continue beratien at > 40 gem of 2300 ppm beric acid solution or ecuivalent until

ne requirec SHUTDOWN MARGIN is restored.

Si.'RVEILL ANCE REOUIREMENTS m.*~.HUTDOWN MARGIN shall be determined to be.y>"O c' f' b

4.1.1.1.1 The.S 4' Typ at-tb :

rM Within one hour after detection of an inoperable CEA(s) and at least a.

once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereaf ter while the CEA(s) is ineperacle.

If tne inecerable CEA is imovable er untrippable, the ab ve required SHUT-DOWN MARGIN shall be increased by an amount at least equal to the withdrawn worth of the imovable er untrippable CEA(s).

b.

When in MODE 2 &, witnin 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to achieving react:r critical-ity by verifying that the predicted critical CEA position is within the limits of Specification 3.1.3.6.

When in MODES 3 er 4, at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by con-c.

sideratien of the following facters:

1 1.

Reactor coolant syster beron concentration,

)

2.

CEA positien, 1

3.

Reactor coelant system average temoerature, 4

Fuel burnus based on gross thermal energy generation, 5.

Xenen cencentratien, and 6.

Samarium concentration.

4.1.1.1.2 The overall core reactivity balance shall be c:meared to predicted values te demonstrate agreement within + 1.00 t.k/k at least ence per 31 Effective Full Power Days (EFFD).

ThTs ccmparison shall censider at least those factors stated in Specification 4.1.1.1.1.c, above.

The predicted reactivity values shall be adjusted (nor.alized) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 Effective Full Power Days after each fuel loading.

4With X,ff < 1.0.

CALVERT CLIFFS - UNIT 1 3/4 1-1 Amendment No. !B, 77, 22, 104 9-16

e 3/4.1 REACTIVITY C0tlTROL SYSTEMS 3/4.1.1 B0 RATION CONTROL SHUTDOWN MARGIN - Tavg > 2000F LIMITING CONDITION FOR OP: RATION 3.1.1.1 The SHUTDOWN MARGIN shall be equal to or greater than the limit line of Figure 3.1-lb.

APPLICABILITY:

MODES 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 MODE 2##, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to achieving reactor criticality by verifying that the predicted critical CEA position is within the limits of Specification 3.1.3.6.

c.

When in MODES 3 or 4, at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by consideration of the following factors:

1.

Reactor coolant system boron concentration, 2.

CEA position, 3.

Reactor coolant system average temperature, 4.

Fuel burnup based on gross thermal energy generation, 5.

Xenon concentration, and 6.

Samarium concentration.

4.1.1.1.2 The overall core reactivity balance shall be compared to predicted values to demonstrate agreement within 1.0% 4 k/k at least once per 31 Effective Full Power Days (EFPD).

This comparison shall consider at least those factors stated in Specification 4.1.1.1.1.c, above.

The predicted reactivity values shall be adjusted (normalized) to correspond to the actual core conditions to exceeding a fuel buinup of 60 Effective Full Power Days after each fuel loading.

With Keff < l.0 CALVERT CLIFFS - UNIT 1 3/4 1-1 Amendment No./E///I//EE/

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ACCEPTABLE 5

OPERATION 5.0 REGION (E0C,5.0)-"

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REACTIVI_TY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION 3.1.1.4 The moderator temperature coefficient (MTC) shall be:

t n0 10-4 a /k whene e THr L POW LeIt7' h sitiv s

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APPLICABILITY: f.0 DES 1 and 2*f ACTION:

With the moderattr temperature coefficient outside any one of the above limits,'be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE RE0VIREME?RS 4.1.1.4.1 The MTC shall be determined to be within its limits by confir-natory measurements. MTC measured values shall be extrapolated and/or compensated to pemit direct comparison with the above limits.

• With K,ff 1 1.0.

1. See Special Test Exception 3.10.2.

CALVERT CLIFFS - UNIT 1 3/4 1-5 Amendment No. 48, 28. 104 9-18

REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION 3.1.1.4 The moderator temperature coefficient (MTC) shall be:

a.

Less positive than the limit line of Figure 3.1-la, and b.

Less negative than -2.7 x 10-4 A k/k/ F at RATED THERMAL POWER.

0 APPLICABILITY:

MODES 1 and 2*#

ACTION:

With the moderator temperature coefficient outside any one of the above limits, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLA%CE REQUIREMENTS 4.1.1.4.1 The MTC shall be determined to be within its limits by confirmatory measurements. HTC measured values shall be extrapolated and/or compensated to permit direct ccmparison with the above limits.

Lith r,.

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CAli(dT CLIFFS - UNIT 1 3/4 1-5 Amendment No. /E//E$//Jpf,

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'-0.20,Gp 3 W 60%

w 1

0.10 lo

\\1

"-0.00, Gp 3 @ 60%-

o p" 0.00 Allowable BASSS <5--

Grp 5 9 55%

Ocerating Region Groups:

5 3

1 i

e i

i i

a L.

i i

i i

i e

i i

0 20 40 60 80 100 0

20 40 60 80 100 0

20 40 60 80 100 136.0 108.8 81.6 54.4 27.2 0

136.0 108.8 81.6 54.4 27.2 0 136.0 108.8 81.6 54.4 27.2 0 4

2 i

i e

i e

i a

a i

e i

g 0

0 2d 40 60 80 100 0

20 40 60 80 100

{

13G.0 108.S 81.6 54.4 27.2 0

136.0 103.8 81.6 54.4 27.2 0

s

% CEA INSERTION 5

INCHES CEA WITHORAWN Figure 3.1-2 CEA GROUP INSERTION LIMITS VS. FRACTION OF ALLOWABLE THERMAL POWER FOR EXISTING RCP COMBINATION

.._=

N E

1.00

" -1.00,Gp 5 9 35%

g 1

1-a 0.90 40.90,Gp 5 9 35%

p

(

l c.

0.80

.75,Gp 5 9 50%

.70,Gp 5 9 60%

0.70 N

Q65,Gp5985%

N 0.60 l

56,Gp 4 9 50%

~

0.50 jN l

Steady State \\-

IShort Term Long Term 0.40 lSteadyState Insertion

' Insertion 8 Limit

[-

l Limit 0.30

' Grp 5 9 25%

'Grp 4 9 20%

\\

' 0.20,Gp 3 @ 60%-

0.20 u

3 0.10

" 0.00, Gp 3 9 00%

/o 0.00 w

Allowable BASSS <(---

Grp 5 9 55%

Operating Region Groups:

5 3

1 i

i I

I I

I 9

I I

i i

I I

I I

t I

I O

20 40 60 80 100 0

20 40 60 80 100 0

20 40 60 80 100 p

136.0 108.8 81.6 54.4 27.2 0

136.0 108.8 81.6 54.4 27.2 0 136.0 108.8 81.6 54.4 27.2 0 4

2

; conditions.

For the steam lin eure event

. ginning of yele conditions, g

kaminimumSHUTCOWNMARGINoflessthan"'

. /k is required to control the

" reactivity transient, and end of e e condit1 e:vire 3. M ik/k.

/c:crd-3 ingly, :ne SnUTDOWN MARGIN re ~ ement is based u:en..'

limiting c0ndition

, and is consistent with ~~ s safety analysis assumpti:ns, ne reactivity tr>. ents' resulting from any postulated acciden.n '

• h Tavg,1 2000F, e minimal and a 3% l.k outd0wn margin provides adequate protection.

With :n N pressu '.r level less than 90 inches, the scur:es of non-berated water zr

. icted to increase the tire to criticality during a boren dilution event.

r 3/4.1.1.3 BORON DILUTION A minimum flow rate of at least 3000 GFM provides adequate mixing, prevents stratificatien and ensures that reactivity changes will be gradual during boren c:n=entration reductions in the Reactor Ceolant System.

A flow rate of at leas: 3000 GPM will circulate an equivalent Reactor Ceolan-System volume of 9,501 cubic feet in appr:ximately 24 minutes.

The reactivi;y change rate associated with boren concentration reductions will therefore be witnin the capability of operator recognition and c:ntrol.

)

j3/4.1.1.4 M20ERATOR TEM:E:ATURE COE::!C!ENT (MTC)

The limitations on MTC are provided to ensure that the assum:: fens used

'in he accident anc transient analyses remain valid througn ea:n fuel cycle.

. Ine surveillance requiremen:s for measurement of the FC curing each fuel

.. cycle are ade:uate ::

)

confir :ne MTC value since this coeffi: Tent changes jislowly dus principally to the reduction in RCS toren concentration ass:ciated with fuel turnup.

The ::nfirma:i:n tha: the measured MTC value is wi:nin its i

limit provides assurances that the :cefficient will be maintained within

,, a:ceptable values inrougneut each fuel cy:le.

!i:i I

CAL','ERT CLIFFS - UNIT 1 S 3/4 1-1 Amen men: No. 32, /2, 71, 22, 104 9-21 i

24-29a(66II)/cp-41 r

O SHUTDOWN MARGIN requirements for sub-critical conditions vary throughout the cycle as a function of fuel depletion, RCS boron concentration and RCS Tavg.

These requirements are satisfied via adherance to Technical Specifications 3.1.1.1 and 3.1.1.2 in the form of minimum RCS shutdown boron concentrations versus time in cycle (fuel depletion), RCS boron concentration at HFP and RCS Tavg.

The appropriate minimum RCS shutdown boron concentrations are determined on a cycle specific basis.

The most limiting SHUTDOWN MARGIN requirement for Modes 2 (Keff < 1.0), 3 and 4 conditions at beginning of cycle is determined by the requirements of several transients, including Baron 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 for these same medes at end of cycle comes from just one transient, the Steam Line Rupture event initiated at Mode 2 (Keff < 1.0) conditions with RCS Tavg=532*F.

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 1

incorporated in the SHUTDOWN MARGIN Technical Specification for Modes 2 (Keff 1.0), 3 and 4, i.e., 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 has been determined to be conservative relative to the actual variation in the most limiting requirement.

Consequently, adherance to Technical Specification 3.1.1.1 provides assurance that the available SHUTDOWN MARGIN at anytime in cycle will exceed the most limiting SHUTDOWN MARGIN requirement at that tire 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, adherance to Technical Specification 3.1.1.2 provides essurance that the available SHUTDOWN MARGIN will exceed the most limiting SFUTDOWN MARGIN requirement at any time in cycle.

i 1

l j

9-22

o 3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 B0 RATION CONTROL 3/_4 J,.L)_and 3/4.1.1.2 SHUTDOWN MARGIN A sufficient SHUIDOWN 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 accept-able limits, and 3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

Specifications 3/4.1.1.1 and 3/4.1.1.2 ensure that adequate SHUTDOWN P U 3IN is available for the sub-critical modes (MODES 2 {K Adequate SHUTDOWN MARGIN for the critical modes (MODlkf < l.0), 3, 4, and 5).

1 and 2 (KefI 2 1.0)) is assured via specifications in the "Movable Control Assemblies section (3/4.1.3).

SHUTDOWN MARGIN requirements for sub-critical conditions vary throughout the cycle as a function of fuel depletion, RCS boron concentration, and RCS Tavg.

These requirements are satisfied via adherence to Technical Specifications 3.1.1.1 and 3.1.1.2 in the form of minimum RCS shutdown boron concentrations versus time in cycle (fuel depletion), RCS boron concentration at HFP and RCS Tavg.

The appropriate minimum RCS shutdown boron concentrations are determined on a cycle specific basis.

The most limiting SHUTDOWN MARGIN requirement for Modes 2 (Keff < l.0), 3 and 4 conditions at beginning of cycle is determined by the requirements of several transient events, including Boron Dilution and Steam Line Rupture.

The requirements for these transients are relatively small and nearly the same, such that one transient does not dominate with regards to SHUTDOWN MARGIN requirement at beginning of cycle.

However, the most limiting SHUTDOWN MAR'IN requirement for these same modes at end of cycle comes from just one transient, the Steam Line Rupture event initiated in Mode 2 (Keff < l.0) conditions with RCS Tavg - 532 F.

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

This variation in specified SHUTDOWN MARGIN has been determined to be 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 anytime in cycle will exceed the most limiting SHUTDOWN MARGIN requirement at that time in cycle.

CALVERT CLIFFS - UNIT 1 B 3/4 1-1 Amendment No. J////E/////

$$//Jff,

o 3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 B0 RATION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN 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 has been set equal to a constant value which is larger than 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 3.1.12 provides assurance that the available SHUTDOWN MARGIN will exceed the most limiting SHUTDOWN MARGIN requirement at any time in cycle.

3/4.1.1.3 BORON DILUTION A minimum flow rate of at least 3000 GPM ;rovides 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 throu?h each fuel cycle.

The surveillance requirements for measurement of the HTC during each fuel cycle are adequate to confirm the MTC value since this coefficient changes l

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 1-1A Amendment No. 32///E/////

15//Jpf,

h J

s

,a.

.e i

p) & (%/k %d 20Wl 0)

REACTIVITY CONTROL SYSTEMS I

BASES I

The boron capability required below 200'F is based upon providing a 35 Ak/k SHUTDOWN MARGIN after xenon decay and cooldown from 200"F to 140*F. This condition requires either 737 gallons of 7.251 boric acid solution from the boric acid tanks or 9.844 gallons of 2300 pga borated 1

water frza the refueling water tank.

1 The OPERABILITY of one boron injection system during REFUELING j

ensures that this system is available for reactivity control whil in MODE 6.

3/4.1.3 MOVA8LE CONTROL ASSEMBLIES 4

t(1) acceptable f

The specifications of this section enau inimum SHUTD0bN MARGIN +gower l

distribution limits are maintained. (2) the s

maintained, and (3) the potential effects of a CEA ejection accident are limited to acceptable levels.

The ACTION statements which pemit limited variations from the basic 1

requirements are accompanied by additional restrictions which ensure that I

the original criteria are met.

The ACTION statements applicable to a stuck or untrippable CEA and to a large misaligrinent (t 15 inches) of two or more CEAs. require a prompt shutdown of the reactor since either of these conditions may be indicative of a possible loss of mechanical functional capability of the CEAs and in the event of a stuck or untrippable CEA. the loss of SWT-DOWN MARGIN.

small degradation in the peaktiig factors relative to those assumed)in For small misaligrenents (< 15 inches) of the CEAs. there is 1 a generating LCOs and LSSS setpoints for DN8R and linear heat rate. 2) a

=

small effect on the time dependent long tem power distributions rela-tive to those used in generating LCOs and LSSS setpoints for DN8R and i

linear heat rate. 3) a small effect on the available SWTDOWN MARGI.

I arel 4) a small effect on the ejected CIA worth used in the safety analysis. Therefore, the ACTION statement associated with the small misaligrenent of a CEA pemits a one hour time interval during which j

attempts may be made to restore the CEA to within its aligrinent require-

]

ments prior to initiating a reduction in THERMAL POWER. The one hour time limit is sufficient to (1) identify causes of a misaligned CEA (2) take appropriate corrective action to realign the CEAs and (3) minimize the effects of xenon redistribution.

t CALVERT CLIFFS - UNIT 1 8 3/4 1-3 heendment No. B. MIOS i

i 9-23 i

24-29a(8611)/cp-42 i

SHUTDOWN MARGIN requiremerts for critical conditions vary throughout the cycle as a function of fuel depletion, RCS boron concentrations and power level.

These requirements are satisfied via adherance to the Technical Specifications of Section 3/4.1.3 which specify the appropriate operability status and position of the full length CEAs.

Cycle specific calculations are performed to verify that the restrictions of Section 3/4.1.3, in particular, the Transient Insertion Limits of Technical Specification 3.1.3.6, yield appropriate SHUTCOWN MARGINS.

The most limiting SHUTCOWN MARGIN recuirements for Modes 1 and 2 (Keff> 1.0) conditions at beginning of cycle are determined by the requirements of several transients, e.g., Loss of Flow, Seized Rotor, etc. However, the most limiting SHUTDOWN MARGIN requirements for these same modes at end of cyc'e come from just one transient, the Steam Line Pupture.

The requirements of the Steam Line Rupture event at end of cycle for both the full power and no load conditions are significantly larger than those of any other event at that time in cycle and, also, considerably larger than the rest limiting requirements at beginning of cycle.

Although the most limiting SHUT 00WN MARGIN requirements at end of cycle are much larger than those at beginning of cycle, the available SHUTDOWN MAPGINS, obtained via the scramming of the CEAs, are also substantially larger due to the much lower boren concentration at end of cycle.

To verify that adequate SHUTDOWN MARGINS are available throughout the cycle to satisfy the changing requirements, calculations are performed at both beginning and end of cycle.

It has been determined that calculations at these two times in cycle an sufficient since the differences between available SHUTDOWN MARGINS and the most limiting requirements are the smallest at these tires in cycle.

The measurement of CEA bank worth performed as part of the Startup Testing Program demonstrates that the core has the expected shutdown capability.

Consequently, adherance to the Technical Specifications of Section 3/4.1.3 provides assurance that the available SHUTDOWN MARGINS at any time in cycle will exceed the most limiting SHUT 00WN MARGIN requirements at that time in cycle.

1 2

9-24 4

REACTIVITY CONTROL S,JTEMS BASES 0

The boron capability required below 200 F is based upon providing a i

0 0

3% A k/k SHUTDOWN MARGIN after xenon decay and cooldown from 200 F to 140 F, This condition requires either 737 gallons of 7.25% boric acid solution from the boric acid tanks or 9,844 gallons of 2300 ppm borated water from the

)

refueling water tank.

1 The OPERABILITY of one boron injection system during REFUELING ensures that this system is available for reactivity control while in MODF 6.

3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that (1) acceptable power distribution limits are maintained, (2) the required SHUTDOWN MARGIN at critical conditions (MODES 1 and 2 (Keff 21.0)) is maMained, and (3) the potential effects of a CEA ejection accident are limited to acceptable levels, j

SHUTDOWN MARGIN requirements for critical conditions vary throughout the cycle as a function of fuel depletion, RCS boron concentrations, and power level. These requirements are satisfied via adherence to the Technical Specifications of Section 3/4.1.3 which specify the appropriate operability i

status and position of the full length CEAs.

Cycle specific calculations are performed to verify that the restrictions of Section 3/4.1.3, in particular, the Transient Insertion Limits of Technical Specification 3.1.3.6, yield appropriate SHUTDOWN MARGINS.

The most limiting SHUTDOWN MARGIN requirements for Modes 1 and 2 (Keff 2 1.0) conditions at beginning of cycle are determined by the requirements of several transients, including loss of Flow, Feedline Break, Small Break LOCA, etc.

The requirements of these transients are relatively the same, such that one event does not dominate with regard to SHUTDOWN MARGIN requirements at beginning of cycle.

However, the most limiting SHUIDOWN MARGIN requirements for these same modes at end of cycle come from must the Steam Line Rupture events.

The requirements of the Steam Line Rupture event at end of cycle for both the full power and no load conditions are significantly larger than those of any other event at that time in cycle and, also; considerably larger than the most limiting requirements at beginning of cycle.

Although the most limiting SHUTDOWN MARGIN requirements at end of cycle are much larger than those at beginning of cycle, the available SHUTDOWN MARGINS retained via the scramming of the CEAs is also substantially larger at end of cycle due to the much lower boron concentration. To verify that adequate SHUTDOWN MARGINS are available throughout the cycle to satisfy the changing requirements, calculations are performed at both beginning and end of cycle.

It has been determined that calculatiors at these two times in cycle are sufficier.t since the available SHUTDOWN MARGINS increase throughout the cycle at a faster rate than the most limiting SHUTDOWN MARGIN requirements.

CALVERT CLIFFS - UNIT I B 3/4 1-3 Amendment No. JJ// M //J @,

REACTIVITY CONTROL SYSTEMS BASES In addition, the measurement of CEA bank worth performed as part of the Startup ?esting Program reinforces the adequacy of these calculations.

In summary, adherence to the fechnical Specifications of Section 3/4.1.3 and appropriate Startup Testing results provide assurance that the available SHUTDOWN MARGINS at any time in cycle will exceed the most limiting SHUIDOWN MARGIN requirements at that time in cycle.

The ACTION statements which permit limited variations from the basic requirements are accompanied by additional restrictions which ensure that the original criteria are met.

The ACTION statements applicable to a stuck or untrippable CEA and to a large misalignment ( 215 inches ) of two or more CEAs, require a prompt shutdown of the reactor since either of these conditions may be indicative of a possible loss of mechanical functional capability of the CEAs and in the event of a stuck or untrippable CEA, the loss of SHUTDOWN MARGIN.

For small misalignments ( 115 inches ) of the CEAs, there is 1) a small degradation in the peaking factors relative to those assumed in generating LCOs and LSSS setpoints for DNBR and linear heat rate, 2) a small effect on the time dependent long term power distributions relative to those used in generating LCOs and LSSS setpoints for DNBR and linear heat rate, 3) a small effect on the available SHUTDOWN MARGIN at critical conditions (MODES 1 and 2 (K ff 11.0)), and 4) a small effect on the ejected CEA worth used in the safetyanalysis.

Therefore, the ACTION statement associated with the small misalignment of a CEA permitt, a one hour time interval during which attempts may be made to restore the CEA to within its alignment requirements prior to initiating a reduction in THERMAL POWER. The one hour time limit is sufficient to (1) identify causes of a misaligned CEA, (2) take appropriate corrective action to realign the CEAs and (3) minimize the effects of xenon redistribution.

CALVERT CLIFFS - UNIT 1 B 3/4 1-3A Amendmont No. JJ///E//JpE,

t REACTIVITY CONTROL SYSTU's BASES The maximum CEA drop time restriction is consistent with the assumed CEA drop time used in the safety analyses. t;easurement with T

> 515*F and with all reactor coolant pumps operating ensures that the S.EsIred drup times will be representative of insertion times experienced during a reactor trip at operating conditions.

The LSSS setpoints and the power distribution LCOs were generated based upon a core burnup which would be achieved with the core operating in an essentially unrodded configuration.

Therefore, the CEA insertion limit specifications require that during tiODES 1 and 2, the full length CEAs be nearly fully withdrawn. The amount of CEA insertion per.r.itted by the Steady State Insertion Limits of Specification 3.1.3.6 will not have a significant effect upon the unrodded burnup assumption but will still provide sufficient reactivity control. The Transient Insertion Limits of Specification 3.1.3.6 are provided to enhshat (1) accectable power distribution limits are maintained, (2) theCminimun)dnUTDOWN MARGIN 1s maintained, and (3) the potential effects of a CIA ejection 7ccident are limited to acceptable levels; however, long term operation I

at these insertion limits could have adverse effects on core cower distribution during subsequent operation in an unrodded configuration.

~

&. h(W $ & Wk.44L hl'0 l

CALVERT CLIFFS - U:1IT 1 B 3/4 1-5 k:endment t'o. 32 9-25

REACTIVITY CONTROL SYSTEMS BASES The maximum CEA drop time restriction is consistent with the assumed CEA drop time used in the safety analyses.

Measurements of T 1 515 F and with av all reactor coolant pumps operating ensures that the measur0d drop times will be representative of insertion times experienced during a reactor trip at operating conditions.

The LSSS setpoints and the power distribution LCOs were generated based upon a core burnup which would be achieved with the core operating in an essentially unrodded configuration. Therefore, the CEA insertion limit specifications require that during MODES 1 and 2, the full length CEAs be nearly fully withdrawn. The amount of CEA insertion permitted by the Steady State Insertion Limits of Specification 3.1.3.6 will not have a significant effect upon the unrodded burnup assumption but will still provide sufficient reactivity control.

The Transient Insertion Limits of Specification 3.1.3.6 are provided to ensure that (1) acceptable power distribution limits are maintained, (2) the required SHUTDOWN MARGIN at critical conditions (MODES 1 and 2 (Keff 11.0)) is maintained, and (3) the potential effects of a CEA ejection accident are limited to acceptable levels; however, long term operation at these insertion limits could have adverse effects on core power distribution during subsequent operation in an unrodded configuration.

CALVERT CLIFFS - UNIT 1 B 3/4 1-5 Amendment No. J.2,

i 24-29a(8611)/cp-43 TABLE 9-2 TECHNICAL SPECIFICATION CHANGES APPLICABLE TO BOTH CALVERT CLIFFS UNIT 1 CYCLE 10 AND UNIT 2 CYCLE 8 The following Technical Specification changes are proposed for both Unit 1 Cycle 10 and Unit 2 Cycle 8:

1.

Figure 2.2-1 (Axial Power Distribution Trip Limiting Safety System Settings (APD LSSS)).

This figure is associated with Technical 1

Specification 2.2-1 (Table 2.2-1. Item 8).

The new limits are wider in both the positive and negative ASI regions below 70% power.

2.

Figure 3.2-2 (Linear Heat Rate Axial Flux Offset Control Limits).

This figure is associated with Technical Specification 3.2.1 (Linear Heat Rate Limiting Condition for Operation (LHR LCO)). The new limits are wider in the negative ASI region below 50% power.

3.

Figure 3.2-4 (DNB Axial Flux Offset Control Limits).

This figure is associated with Technical Specification 3.2.5 (DNB Limiting Cundition for Operation (DNB LCO)).

The new limits are wider in the negative ASI region below 50% power.

The proposed new limits for these Technical Specifications are shown on the enclosed modified figures.

BASIS FOR TECHNICAL SPECIFICATION CHANGES During plant startup, when power is below the APD trip bypass of 15% power, the measured Axial Shape Index (ASI) can be more negative than the current APD trip limit. This condition is caused by xenon and reactivity feedback induced perturbations in the axial power shape, which are associated with rapid changes in power and temperature. Technical Specification 2.2-1 requires that J

the plant remain below 15% power until the ASI falls within the APD LSSS operating band. Widening of the APD trip at this power level will not affect safety since sufficient conservatism exists in the current APD LSSS to permit relaxation of the ASI operating band without compromising any safety margin.

Such relaxation at these very low power levels will be very helpful to plant operations, as it will allow the induced perturbations in the axial shape to dampen cut before a significant power level is reached (i.e., in the 20-40%

power range).

To be consistent with the changes being made to the APD LSSS and to provide additional operating flexibility, widenings of the LHR LCO and DNB LCO AST i

limits are also being proposed.

Such changes will not affect safety sinca sufficient conservatism exists in these LCO's to permit relaxation of the ASI j

operating bands without compromising any safety margins.

9-26 1

24-29a(86I1)/cp-44 SAFETY EVALUATION The proposed changes were evaluated by Combustion Engineering (C-E) using the approved Setpotnt Methodology described in Reference 5.

The proposed changes were evaluated for potential impact on the following items:

a.

The transient analyses b.

The margin to fuel centerline melt limits c.

The margin to DNB limits d.

The margin to the LOCA peak linear heat rate limit e.

The core power versus planar radial peaking factor LCO (Technical Specification Figure 3.2-3a) f.

The Thermal Margin / Low Pressure LSSS (Technical Specification Figures 2.2-2 and 2.2-3) g.

The core power versus integrated radial peaking factor LCO (Technical Specification Figure 3.2-3c)

Each of these items is addressed below, l

a.

Accident Analyses The APD trip is not credited directly in any transient analysis for Calvert Cliffs Units 1 and 2 (i.e., its performance is not modeled).

Instead, the APD trip is credited indirectly.

By virtue of the Setpoint Pethodology, the APD LSSS ensures that tne Specified Acceptable Fuel Design Limit (SAFDL) on fuel csnterline melt is not violated during Design Basis Events (DBE). This methodology assumes that no Technical Specifications are violated. Consequently, a transient simulation is not required except for those events which result in system changes not considered in the APD LSSS analysis (e.g., CEA drop which results in a violation of the Power Dependent insertion Limits Technical t

Specification Figure 3.1-2).

The transient simulations dete :nine the margin degradation for those events not covered by the APD LSSS aralysis and the corresponding Required Overpower Margin (ROPM) data to be factored into the Limiting Conditions for Operation. This methodology is described in Chapter 1 of C-['s approved setpoint methodology topical report (Reference 5).

Therefore, the proposed change to the APD LSSS will not impact the transient analyses of Chapter 14 At zero power conditions the operating band set by the DNB LC0 and LHR LCO is credited in the following accident analyses:

1.

Co.itrol Element Assembly (CEA) Withd.*awal, l

2.

Excess Load, and 3.

CEA Ejection The CEA Ejection analysis considers axial shapes of positive ASI values.

i Since no changes are proposed in the positive ASI region of the LCOs, the CEA Ejection event will not be affected. Thus, thc proposed changes will only affect the CEA Withdrawal and Excess Lead events.

9-27 i

p

24-29a(86I1)/cp-45 l

l The proposed changes have an insignificant impact on the site boundary doses calculated for the CEA Withdrawal and Excess load events.

Moreover, the radiological consequences for both the CEA Withdrawal and Excess load analyses are less adverse than the Loss of Non-Emergency AC Power event for which the thyroid and whole body doses are 0.04 and 0.0006 REM, compared to the 10 CFR 100 guidelines of 300 and 25 REM respectively.

Consequently, the proposed changes will have no impact on the site boundary doses which are already negligible compared to the 10 CFR 100 guidelines.

With regard to the non-radiological criteria the proposed change affects only the transient minimum DN8R. An evaluation of the CEA Withdrawal and Excess Load Analyses showed that the proposed changes will make the i

transient minimum DNBP, more limiting.

However, the impact is very small and the current minimum DNBR values of 5.44 for Excess Load and 1.86 for i

CEA Withdrawal are high enough that the DNBR design limits will not be violated.

The reduced minimum DNBR values are estimated to be approximately 5.3 for Excess Lead and 1.7 for CEA Withdrawal. Therefore, the proposed changes to the DNB LCO and LHR LCO do not significantly impact the Unit 1 Cycle 10 and Unit 2 Cycle 8 transient analyses.

b.

Margin to Fuel Centerline Melt Limits The proposed change to the APD LSSS was analyzed for both Calvert Cliffs Unit 1 Cycle 10 and Unit 2 Cycle P.

This analysis showed that at power levels less than 70% of rated power there exists sufficient conservatism to accomodate the proposed change, c.

Margin to DNB Limits The proposed change to the DNB LCO was analyzed for both Calvert Cliffs Unit 1 Cycle 10 and Unit 2 Cycle 8.

This analysis showed that at power levels less than 50% of rated power there exists sufficient conservatism to accomodate the proposed change.

d.

Margin to the LOCA Peak Linear Heat Rate Limit (Ex-core Monitoring)

The LHR LCO using ex-core monitoring assures that the LOCA peak linear heat rate (PLHR) limit is not exceeded.

This LCO consists of two parts.

The first part is Technical Specification Figure 3.2-2.

This figure gives the allowed core power as a function of measured ASI.

The second part is Technical Specification Figure 3.2-3b.

This figure, called the N factor curve, applies a scaling factor to the allowed core power as a t

function of the measured planar radial peaking factor.

Because of this tradeoff with planar radial peaking factor, the margin to the LOCA PLHR limit is determined by the N factor curve.

The proposed change to the LHR LCO tent was analyzed for both Calvert j

j Cliffs Unit 1 Cycle 10 and Unit 2 Cycle 8.

This analysis showed that at 1

pewer levels less than 50% of rated power there exists sufficient I

conservatism to accomodata the proposed change.

t P

9-28

J 24-29a(8611)/cp-46

)

It shculd be noted that during normal plant operation the LOCA PLHR limit is monitored with the in-core monitoring system. Technical Specification Figures 3.2-2 and 3.2-3b are used as a backup when the in-core system or plant computer are out of service.

e.

Core Power Versus Planar Radial Peaking Factor LCO The core power versus planar radial peaking factor tradcoff curve (Figure 3.2-3a) ensures that the APD LSSS retains conservative if the measured unrodded planar radial peaking factor becomes greater than the Technical Specification limit of 1.70.

This limit curve was analyzed for both Unit I cycle 10 and Unit 2 Cycle 8 using the proposed change to the APD LSSS.

The results showed that the current Unit 1 Cycle 10 and Unit 2 Cycle 8 Technical Specification limit remains conservative.

i f.

Thermal Margin / Low Pressure LSSS The APD LSSS is credited in the Thermal Margin / Low Pressure (TM/LP) LSSS analysis.

For Calvert Cliffs Unit 1 Cycle 10 the TM/LP LSSS is being modified (see Table 9-1).

The analysis supporting the TM/LP LSSS modification assumed the change to the APD t.SSS proposed herein (Table 9-2).

Th!s analysis shewed that for Unit 1 Cycle 10 there exists sufficient conservatism to accommodate the proposed changes to the APD LSSS and TM/LP LSSS.

The Calvert Cliffs Unit 2 Cycle 8 TM/LP LSSS has not been modified.

Therefore, the TM/LP LSSS was analyzed using the current Unit 2 Cycle 8 TM/LP LSSS and the proposed change to the APD LSSS.

This analysis showed that for Unit 2 Cycle 8 there exists sufficient conservatism to accommodate the proposed change to the APD LSSS.

g.

Core Power Versus Integrated Radial Peaking Factor LCO The core power versus integrated radial peaking factor tradeoff curve (Figure 3.2-3c) ensures that both the TM/LP LSSS and DNB LCO remain conservative if the measured unrodded integrated radial peaking factor becomes greater than the Technical Specification limit of 1.65.

This limit curve was analyzed for both Unit 1 Cycle 10 and Unit 2 Cycle 8 because the APD LSSS TM/LP LSSS, and DNB LC0 which are changing are used in the enalysis of this curve.

The results showed that the current Unit 1 Cycle 10 and Unit 2 Cycle 8 Technical Specification limit remains conservative.

9-29

L 1.30 i.

i 1

(0,1.20) 1.20 UNACCEPTABLE UNACCEPTABLE OPERATION

\\

OPERATION REGION REGION 1.10 1.00

(-0.2,1,00)

(0.2,1,00)'

l 0.90 d

t 0.80 t

l ACCEPTABLE w

OPERATION REGION a

0.70 Y

2 4

h.

O 0.60 5

i a

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[

i 0.40

+ ( 0.6,.40)

(0.6,.40) u 1

1 0.30 3

0.20 i

0.15 0.8

-0.6 0.4 0.2 0.0 0.2 0.4 0.6 0.8 PERIPHERAL AXIAL SHAPE INDEX, Y 7

Figure 2.2 1 Peripheral Axial Shape Index, Y vs Fraction of Rated Thermal Power g

i

)

CALVERT CLIFFS 2 11 Amendment No.

9-30 i

1.30 i

i i

(0,1.20)

UNACCEPTABLE UNACCEPTABLE 1.20 OPERATION OPERATION i

REGION RECION 1.10 i

( 0.2,1.00)

(0.2,1.00) 1.00 0.90 5

a 0.80 l

ACCEPTABLE w

OPERATION PlGION 0.70 g

9 I

w0 0.60 8

6 h

0.50 l

1' ( 0.6,.40)

(0.6. 40) u 0.40 O.30 0.20 f

0.15

-0.8

-0.6 0.4

-0.2 0.0 0.2 0.4 0.6 0.8 PERIPHERAL AXIAL SHAPE INDEX, Y j

7 Figure 2.2-1 Peripheral Axial Shape Index, Y vs Fracdon of Rated nemal Pwer j

7 CALVERT CLIFFS 2-11 Amendment No.

J

t J

i 1.10 i

i i

i L

( 0.06,1,00) *

• (0.12,1,00) 1.00 UNACCPETABLE UNACCEPTABLE 0.90 OPERATION OPERATION REGION REGION 1

P O.80 i

dl 0.70

( 0.3,.70)

• ACCEPTA3LE

(0.3. 70)

OPERATION g

REGION 0.60 g

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( 0.3. 50)

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)

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)

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PERIPHERAL AXIAL SHAPE INDEX, Y 7

i l

Figure 3.2 2 j

Linear Heat Rate Axial Flux Offset Control Lirnits t

(

0 CALVERT CLIFFS 3/4 2 4 Amendment No.

[

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i 9-31

.m e

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0.90 PlGION P.EGION I

i O.80 I

El 0.70 (0.3,.70) p

( 0.3,.70)

• ACCEPTABLE 9

OPERATION g

REGION l

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i Figure 3.2 2 Linear Heat Rate Axial Flux Offset Control Limits 4

l I

CALVERT CLIFFS 3/4 2 4 Amendment No.

j i

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=

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OPERATION OPERATION i

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( 0.3,.80),

ACCEPTABLE

• (0.3. 80)

OPERATION REGION E E f p 0.70 si me 0.60 I

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Figure 3.2 4 J

DNB Axial Flux Offset Control Liteits CALVERT CLIFFS 3/4 2 11 Amendment No.

i i

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9-32 i

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1.10 i

F 1.00

( 0.1,1.00) ;

• (0.15,1.00) j UNACCEPTABLE UNACCEPTABLE OPERATION OPERATION 0.90 REGION REGION i

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DNB Ax h1 Flux Offset Control Limits CALVERT CLIFFS 3/4 2 11 Amendment !!o.

1 1

I

24-29a(86II)/cp-47 10.0 STARTUP TESTING The startup testing program proposed for Unit 1 Cycle 10 ~is identical to the program proposed for the reference cycle (Unit 2 Cycle 8) in Reference 1.

l I

l 10-1

24-29a(86II)/cgh-48

11.0 REFERENCES

References - Section 1 1.a. Letter, J. A. Tiernan (BG&E) to Document Control Desk (NRC),

Docket No. 50-318, "Request for Amendment, Eighth Cyr.le License Applcation," February 6,1987.

b. Letter, J. A. Mihalcik (BG&E) to Document Control Desk (NRC),

"Calvert Cliffs Nuclear Power Plant Unit No. 2; Docket No.

50-318, Changes to Request for Amendment, Eighth Cycle License Application," April 7,198.

c. Letter, J. A. Mihalcik (BG&E) to S. A. McNeil (NRC), "Calvert Cliffs Unit 2 Cycle 8 Reload License Submittal," (Rodded F /F r xY Correction), FCM 87-117, May 26, 1987.

2.

Letter, J. E. Baum (C-E) to J. A. Michalcik (BGAE), "Calvert Cliffs Unit 1 Cycle 9 Reload Design Report," B-86-R-070, December 15, 1986.

3.a. Letter, S.

A.

McNeil (NRC) to J.

A.

Tiernan (BG&E),

Docket No. 50-318 (Safety Evaluation Report of Eighth Cycle License Application), May 4, 1987.

b. Letter, S.

A.

McNeil (NRC) to J.

A.

Tiernan (BG&E)

Docket No. 50-318, "Revised Safety Evaluation Supporting Amendment No. 108 to Facility Operating License No. DPR-69,"

June 30, 1987.

4 CEN-348(B)-P, "Extended Statistical Combination of Uncertainties," January 1987.

5.

Letter, S.

A.

McNeil (NRC) to J.

A.

Tiernan (BG&E),

Docket Nos. 50-317 and 50-318, "Safety Evaluation of Topical Report CEN-348(B)-P,' Extended Statistical Combination of Uncertainties' (TACS 64985 and 64986)," October 21, 1987.

6.

Letter, A. E. Lundvall, Jr (BG&E) to R. A. Clark (NRC), Docket Nos. 50-317 and 50-318, "Topical Report for Extended Burnup Operation of C-E Fuel,"

June 7, 1982; Enclosure CENPD-269-P.

"Extended Burnup Operation of Combustion Engineering PWR Fuel,"

April 1982.

I 7.

Letter, E. J. Butcher (NRC) to A. E. Lundvall, Jr. (BG&E),

j Docket Nos. 50-317 and 50-318. "Safety Evaluation for Topical Report CENPD-369-P, Revision 1-P," (Extended Burnup Operation),

October 10, 1985, i

11-1 l

= _.

I 24-29a(86II)/cgh-49 Reference - Section 2

-1.

Letter, J. A. Tiernan (BG&E) to J. M. Allen (NRC), "Calvert Cliffs Nuclear Power Plant Unit No. 1; Docket No. 50-317 Report of Startup Testing for Cycle 9," April 15, 1987.

P l

4 I

l j

n 6

4 I

I j

i I

l

)

1 11.2 t

i f

=

24-29a(86II)/cgh-50 Reference - Section 3 1.

Letter, J. A. Tiernan (BG8E) to Document Control Desk (NRC),

I

-Docket No. 50-318. "Request for Amendment, Eighth Cycle License i

Application," February 6,1987.

l r

e i

f f

7 i

e 4

i f

f I

4 r

j t

4 I

'f 4

i 5

i i

h 1

11-3 1

J 1

24-29a(85II)/cgh-51 References - Section 4 1.

Letter, J. A. Tiernan (BG&E) to Document Control Desk (NRC).

Docket No. 50-318. "Request for Amendment, Eighth Cycle License Application," February 6,1987 2.

Calvert Cliffs Nuclear Power Plant Units 1 and 2 Updated Final Safety Analysis Report Chapter 3.

3.

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

4.

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.

5.

Letter, A. E. Lundvall, Jr. (BG&E) to J.

R. Miller (NRC),

"Calvert Cliffs Unit 1 Supplement 1 to Seventh Cycle License Application," September 1,1983.

6.

Letter, A.

E.

Lundvall, Jr. (BG&E) to J. R.

Miller (NRC),

Docket No, 50-317, "Seventh Cycle License Application Answers to Question Set 2," November 4,1983.

7.

CENPD-139-P-A, "C-E Fuel Evaluation Model Topical Report,"

July 1974, i

1 8.

CEN-161(B)-P, "Improvements to Fuel Evaluation Model,"

July 1981.

9.

CEN-161(B)-P, Supplement 1-P, "Improvements to Fuel Evaluation Model," April 1986

10. Letter, S. A. McNeil (NRC) to J. A. Tiernan (BG&E) Docket Nos.

l 50-317 and 50-318. "Safety Evaluation of Topical Report CEN-161

)

(B)-P Supplement 1-P, ' Improvements to Fuel Evaluation Model',"

February 4,1987.

11-4

24-29a(56II)/cgh-5?

Peferences - Section 5 1.

Letter, J. A. Tiernan (BGAE) to Document Control Desk (NRC),

Docket No. 50-318, "Request for Amendment, Eighth Cycle License Application," February 6,1987 2.

CENPD-275-P, "Core Designs Containing Gadolinia-Urania Burnable Absorbers," March 1987.

3.

Letter, S. A. McNeil (NRC) to J. A. Tiernan (BG&E), Cocket No.

50-318, "Request for Additional Information - Unit 2 Cycle 8 Reload," March 12, 1987.

4 Letter, J. A. Tiernan (BG&E) to Document Control Desk (NRC),

"Calvert Cliffs Nuclear Power Plant Unit No. 2; Docket No.

Request for Additional 50-318, Unit 2 Cycle 8 Reload Information," Parch 27, 1987.

.I, a

1 11-5 4

24-29a(86I1)/cgh-53

,)

\\

j References - Section 6 1.

CENPD-161-P-A, "TORC Code, A Computer Code for Determining the Thennal Margin of a Reactor Core," April 1986, 2.

CENPD-162-P-A (Proprietary), "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-A, "TORC Code. Verification and Simplified Modeling Methods," June 1981.

4.

CEN-191(B)-P, "CETOP-D Code Structure and Modeling Methods for Calvert Cliffs Units 1 and 2," December 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-348(B)-P, "Extended Statistical Combination of Uncertainties," January 1987.

7.

Letter,S.A.McNeil(NRC)toJ.A.Tiernan(BGSE),00cketNos.

50-317 and 50-318 "Safety Evaluation of Topical Report CEN-348(B)-P,

' Extended Statistical Combination -

of Uncertainties'," October 21, 1987.

8.

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

9.

Letter, J. A. Tiernan (BG&E) to Document Control Desk (NRC),

Docket No. CO-318. "Request for Amendment. Eighth Cycle License Application " February 6,1987.

11-6

24-29a(86II)/cgh-54 References - Section 7 1.

Letter, J. A. Tiernan (BG&E) to Document Control Desk (NRC),

Docket No. 50-318, "Request for Amendment, Eighth Cycle License Application," February 6,1987, 2.

Letter, A.

E.

Lundvall, Jr. (BG&E) to J.

R.

Miller (NRC),

Docket No. 50-317, "Amendment to Operating License DPR-53 Eighth Cycle License Application," February 22, 1985.

3.

CEN-348(B)-P, "Extended Statistical Combination of Uncertainties," January 1987.

4 Letter, S. A. McNeil (NRC) to J. A. Tiernan (BG&E), Docket Nos.

50-317 and 50-318, "Safety Evaluation of Topical Report CEN-348(B)-P,

' Extended Statistical Combination of Uncertainties'," October 21, 1987.

5.

CEN-124(B)-P, "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,"

December 1979.

6.

CEN-124(B)-P, "Statistical Combination of Uncertainties Methodology; Part 2;

Combination of System Parameter Uncertainties in Thermal Margin Analyses for Calvert Cliffs Units I and II," January 1980.

7.

CEN-124(B)-P, "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," March 1980.

8.

Letter, D. H.

Jaffe (NRC) to A. E.

Lundvall, Jr.

(BGLE),

i Regarding Unit 1 Cycle 6 License Approval (Amendments #71 to

^

DPR-53andSER), June 24, 1982.

1 l

i d

i 11-7

o 24-29a(86II)/cgh-55 References - Section 8 1.

Ac:eptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors, Federal Register, Vol. 39, No. 3, Friday, January 4, 1974.

4 2.

Letter, A. E. Scherer (C-E) to J. R. Miller (NRC), LO-81-095, Enclosure 1-P, "C-E ECCS Evaluation Model Flow Blockage Analysis, "(Proprietary), December 15, 1981, 3.

Letter, A. E. Scherer (C-E) to C. O. Thommas (NRC), LD-86-027, "Responses to Questions on C-E's Revised Evaluation Model for Large Break LOCA Analysis," (Proprietary), June 17, 1986.

4.

Letter, A. E. Scherer (C-E) to C. O. Thomas (NRC)

LD-85-032, "Revision to C-E Model for Large Break LOCA Analysis", July 3, 1985.

5.

CENPD-133, Supplement 5-P, "CEFLASH-4A, A FORTRAN 77 Digital Computer Program for Reactor Blowdown Analysis," June 1985.

6.

CENPD-134, Supplement 2-P, "COMPERC-II, A Program for Emergency Refill-Reflood of the Core," June 1985.

7.

CENPD-132-P, Supplement 3-P, "Calculative Methods for the C-E Large Break LOCA Evaluation Mcdel for the Analysis of C-E and W Designed NSSS," June 1985.

8.

Letter, A. E. Scherer (C-E) to C. O. Thomas (NRC). LD-85-050, Enclosure, "Supplemental Material for Inclusion in CENPD-132, Supplement 3-P", (Proprietary), November 5,1985.

9.

Letter D. M. Crutchfield (NRC) to A. E. Scherer (C-E), "Safety Evaluation of Combustion Engineering ECCS Large Break Evaluation Model and Acceptance for Referencing of Related Licensing Topical Reports", July 31, 1986.

10. Letter, J. A. Tiernan (BG&E) to Document Control Desk (NRC).

Docket No. 50-318, "Request for Amendment, Eighth Cycle License Application," February 6,1987.

11. 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.

11-8

24-29a(86II)/cgh-56

12. CENPD-138-P, and Supplement 1-P, "PARCH, A FORTRAN IV Digital Program to Evaluate Pool Boiling, Axial Rod and Coolant Heatup", February 1975.

CENPD-138 Supplement 2-P, "PARCH - A FORTRAN-IV Digital Program to Evaluate Pool Boiling, Axial Red and Coolant Heatup",

January 1977,

13. CEN-161(B)-P, Supplement 1-P, "Improvement to Fuel Evaluation Model," April 1986.

14 CEN-161(B)-P, "Improvements to Fuel Evaluation Model,"

July 1981.

15. Letter, S. A. McNeil (NRC) to J. A. Tiernan (BG&E), Docket Nos.

50-317 and 50-318 "Safety Evaluation of Topical Report CEN-161(B) Supplement 1-P,

' Improvements to Fuel Evaluation Model'," February 4, 1987.

16. Letter, A.

E.

Lundvall, Jr. (BG&E) to J.

R. Miller (NRC),

Docket No. 50-317, "Calvert Cliffs Unit 1 Eighth Cycle License Application," February 22, 1985,

17. Letter, D.

H.

Jaffe (NRC) to A.

E.

Lundvall, Jr. (BG&E),

Subject:

Safety Evaluation Report for Amendment No. 104 to DPR-53, Calvert Cliffs Unit 1 Cycle 8, May 20, 1985.

11-9

j 24-29a(8611)/cgh-57 References - Section 9 1.

Letter, J. A. Tiernan (BG&E) to Document Control Desk (NRC),

Docket No. 50-318. "Request for Amendment, Eighth Cycle License Appleation," February 6,1987.

2.

Letter,

_S.

A.

McNeil (NRC) to J.

A.

Tiernan (BG&E).

Docket No. 50-318 (Safety Evaluation Report of Eighth Cycle License Application), May 4, 1987.

j 3.

CEN-348(B)-P, "Extended Statistical Combination of Uncertainties," January 1987.

4.

Letter, S. A. McNeil (NRC) to J. A. Tiernan (BG&E), Docket Nos.

50-317 and 50-318, "Safety Evaluation of Topical Report CEN-348(B)-P,

' Extended Statistical Combination of Uncertainties'," October 21, 1987.

5.

CENPD-199-P, Rev.1-P-A, "C-E Local Power Density and DNB LSSS-and LCO Setpoint Methodology for Analog Protection Systems,"

March 1982, i

i I

I l

l l

11-10

24-29a(86I1)/cghe58 References - Section 10 1.

Letter, J. A. Tiernan (BG&E) to Document Control Desk (NRC),

Docket No. 50-318. "Recuest for Amendment. Eighth Cycle License Applcation," February 6, 1987.

t I

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i 11-11 I

4 1

APPENDIX TO CALVERT CLIFFS UNIT 1 CYCLE 10 LICENSE SUBMITTAL 1

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ANF 88 019 1 y'j

.. s.3 TABLE OF CONTENTS Section Pacq 1.0, INTRODUCTION AND SUM. MARY I

2.0 MECHANICAL DESIGN DESCRIPTION AND COMPATIBILITY EVALUATION 3

3.0 NEUTRONICS DESIGN AND COMPATIBILITY ANALYSES 6

3.1 Neutronic Design Parameters.....................

6 3.2 Enrichment Level and Distribution.

6 3.3 Neutronics Design Methodology....

6 3.4 Neutronic Compatibility.........

7 4.0 THERMAL-HYDRAULIC COMPATIBILITY ANALYSIS 9

qL e

4.1 Thermal-Hydraulic Design Criteria.

9 4.2 Hydraulic Characterization of ANF and Co-Resident Fuel 10 4.3 MDNBR Evaluation of ANF Lead Assemblies and Co-Resident Fuel 10 4.4 Fuel Centerline Temperature.....................

12 8

4.5 LOCA/5CCS Performance........................

12

\\

5.0 REFERENCES

16 A

Tl 3

t

t l

IG l

G 11 ANF 89 019 i

list of Tables Table Pace er 2.1 Fuel Bundle Assen.oly and Component Comparison............

5 h

3.1 Key Neutronic Design Parameters for Calvert Cliffs lead Assemblies 8

4.1 MDNBR Results for CE and ANf Lead Assemblies 14 4.2 Nominal Reactor and Fuel Design Parameters 15 N

J II C

M R

II 3

a 15

.i

~

l ANF 88 019 JII

1.0 INTRODUCTION

AND

SUMMARY

Evaluations have been performed to show that the four (4) lead fuel assemblies (LFAs) manufactured by Advanced Nuclear Fuels Corporation (ANF) are compatible

g with the existing fuel assemblies in the Calvert Cliffs reactor.

These ANF

., J lead fuel assemblies are scheduled to be inserted into the Calvert Cliffs reactor during the next refueling outage (March 1988).

g The ANF Calvert Cliffs fuel assembly is a 14x14 array containing 176 fuel rods in a cage structure of 5 guide tubes and 9 grid spacers.

Both the guide tubes and the fuel rod cladding are made of Zircaloy-4 for low neutron absorption i r]

and high corrosion resistance.

Two of the four lead assemblies have standard

.W ANF bi-metallic spacers made of a Zircaloy 4 structure with Inconel 718 gq springs; the remaining two assemblies contain ANF high thermal performance i

id (HTP) spacers which have superior coolant mixing and structural characteristics.

The HTP spacers are made of all Zirtaloy 4 The fuel assembly tie plates are stainless steel castings with Inconel holddown springs.

The fuel assembly upper tie plate is mechanically locked to the fl guide tubes and may be easily removed to allow inspection of irradiated fuel rods.

Compatibility evaluations have been performed in the areas of mechanical design, neutronics, and thermal-hydraulics.

Mechanical compatibility of the 14x14 lead fuel assemblies is established with in-reactor components and fuel.

l The fuel has also been designed to be hydraulically and thermally compatible

{j with the co-resident fuel.

The LFAs have been designed to be neutronically compatible with the co resident 14x14 fuel.

These compatibility analyses were l

performed with NRC approved methodologies.

It is assumed that the ANF lead assemblies will be inserted in non-limiting I

reduced power locations in the core.

This assumption is considered appropriate for a demonstration program.

The thermal-hydraulic and LOCA/ECCS assessments therefore assume that the ANF lead assemblies operate at a power u

i

2 Af1F 88 019 I

level Sr. below that of the hot assembly in the core.

MDNBR and fuel J,

centerline temperature requirements are met.

The LFAs are expected to exhibit LOCA/ECCS peak cladding temperatures below those of the limiting co resident i,

fuel due to their reduced power location in the core, f

1 fl I

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.A 3

3 ANF 88 019 h."

[I 2.0 MECHANICAL DESIGN DESCRIPTION AND COMPATIBillTY EVALUATION

o1 Mechanical design analyses were performed for the Calvert Cliffs lead assemblies.

These analyses include the design description, design criteria, and the results of the mechanical design analyses (Reference 1).

The analyses

.r}

l

,r d were performed in accordance with the current ANF methodology consistent with the NRC approved "Qualification of Exxon Nuclear Fuel for Extended Burnup,"

f XN NF 82 06.

Each of the four (4) Calvert Cliffs lead fuel assemblies consists of a 14x14 array with 176 fuel rods and 5 guide tubes.

Each assembly contains 164 enriched fuel rods and 12 rods containing gadolinia, a burnable absorber contained in natural enriched UO.

The fuel rod pitch is maintained by 9 2

spacers.

Two of the assemblies have all bi-metallic spacers made of Zircaloy 4 structure with an Inconel 718 spring.

A third assembly has a bi-metallic spacer at each end and seven HTP spacers in the remaining grid locations.

The fourth assembly has all HTP spacers.

Mechanical design features of the ANF lead fuel assemblies and the co-resident CE fuel assemblies are shown in Table 2.1.

As shown in the table, most of the p'

assembly and core interface dimensions are the same for the two fuel types Y]

with the exception of the upper and lower end fitting height and overall assembly height.

The maximum relative centerline displacement of the spacer i

grids in the two vendor assemblies is 0.29 inch, with a minimum grid overlap 2

of 1.09 inches.

These differences will not affect the performance of either i

fuel assembly.

Per Reference 1,

the ANF asscc.bly has been shown to be compatible with the core plate spacing and meets the requirements for fuel assembly holdcown.

The upper tie plate assembly has been tested for compatibility with the reactor handling tools.

The detailed design description of the ANF 14x14 Calvert Cliffs lead fuel assemblies is reported in Reference 1.

I

a 4

ANF 88 019 The Calvert Cliffs co-resident fuel design is similar to the ANF fuel design inserted in reload quantities at the Maine Yankee, Fort Calhoun, and the St.

Lucie Unit I reactors.

In these plants, the ANF reloads were inserted adjacent to CE 14x14 reload fuel.

No operational difficulties were encountered at those plants and none are expected at Calvert Cliffs.

i l

l 1

l I

l 1

i I

1 I

I f

nm b'

5 ANF-88-019

.y

.1 P

Table 2.1 Fuel Bundle Assembly and Component Comparison

• jd w

0 Calvert Cliffs 4NF

.,,3 hda Overall fuel assembly length 157.241 157.104 Overall fuel assembly width 8.115 max, 8.110

'[

lower end fitting height 2.937 3.145 Upper end fitting height 5.766 6.101 Active fuel length 136.700 136.700

~

Bottom of fuel asse-bly to bottom of active core 3.375 3.825 Grid spacing: Bottom of fuel assembly to first grid 5.234 4.813 1st span 11.325 11.497 2nd span 18.559 15.612 3rd thru 7th spans 13.359 13.359 5th span 17.750 17.550 Number of grids (3) Zircaloy (9) Si-metallic

]

5 (Zirc 1 Inc)

]~J (1) Inconel or (2) Si metallic my (7) Zircaloy d

or (9) Zircaloy

-[

Spacing of lower end fitting legs 4.640 sq.

4.640 sq.

10 of lower end fitting legs 1.270 1.270 Spacing of upper end fitting posts 4.540 sq.

4.640 sq.

Upper end fitting post head 00 (outer 4) 1.559 1.659 c

Fuel / burnable absorber rod pitch 0.530 0.530

, 11

'j Fuel assemoly pitch 8.150 3.180 Number of guide tubes 5

5

"/l Suide tube ID 1.035 1.035 s?

Fuel / burnable absorber rod CD 0.440 0.440

• 0imensions in inches.

l iU

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~

6 ANF 88 019

.I s

,7

. I 3.0 NEUTRONICS DESIGN AND COMPATIBILITY ANALYSES k i I'

'y Thc results of the neutronic design and compatibility analysis for ANF's 14x14 lead fuel assembly design for Calvert Cliffs are presented in this section.

} t 1,,

Key neutronic parameters are compared against pareeters of the CE 14x14 i,,

reference fuel design to demonstrate neutronic compatibility.

q j

3.1 Neutronic Desien Parameters

[

The key neutr:nic design parameters for the ANF 14x14 lead fuel assembly design an0 the CE 14x14 reference fuel assembly design are presented in Table

,y 3.1.

',s}

t 3.2 EnHea ent level and Distribution i

e n:-iral earicn ent level (average fissile centent) of the enriched lattice

f tre 2NF lax!: lea: fuel asse-clies is 3.37 w/o,U-235.

The enrichment of the n:n :urnable atsorber rocs in the ANF lead asse-blies is 4.05 w/o U 235.

i This enrien ent is the same as that use: in the reference fuel cesign supplied

?

f Cc tusticn Engineering.

The burnacle absorter design of the ANF assemblies is consists of 12 rods of 10 w/o Gd 02 3 contained in natural (0.71 w/o U-235) uranium.

This enrichment and burnable absorber design was chosen to provide 6

l asse-bly neutronic characteristics similar to the reference fuel design.

17f i3 3.3 Neutronics Desian Methodoloay

! A 1

The principal co puter odels e ployec in the asse-bly and core design

~':

calculations to show neutrenic co patibility are the MICSURN-2( )/CASM0 2E C3I I) codes.

These NRC approved codes were used in accordance with and Xi3PWR ANF's approvec neutr:nics meth0dology('7'*o'3) to perform asse-bly neutronic

A m

4 %

calculations fer the ANF fuel asse-ely anc reference CE fuel asse-bly in order

{

t

pare neutr:nic para eters.

I i

I

---

,,.,_.-----.n,-

7 ANF SS 019 In addition to comparing fuel assembly characteristics, an evaluation was performed to determine the impact of the lead assemblies on the core axial cower distribution. ANF's reactor simulator model, XTGPWR, was utilized.

3.4 Neutronic Cc~eatibility various infinite assembly calculations were performed for the ANF lead assemolies in order to provide reactivity comparisons and demonstrate neutronic compatibility with the reference assembly design.

A comparison of the results show that the differences between assembly reactivities, control rod worth, Do;pler and moderator temperature coefficients are small.

Alto, because the ANF lead fuel assemblies contain 176 active fuel rods compared to the 164 active fuel rods of the reference design, the assemoly average LHGR is less in the ANF assembly by 7% for equal assembly power (Nt).

In addition to these infinite assembly calculations, three-di~ensional core calculations usirg cross sections representative of the ANF lead and reference assembly designs were performed.

These calculations were performed in order to assess the impact the lead assemblies would have on the core axial poaer distribution.

The core axial power distribution forms the bases for the axial shape index which is an irportant parameter in the determination of limiting Safety System Settings and Limiting Conditions of Operation.

The resulting dif ference in the core average axial power shapes was found to be negligible.

As a result of these comparisons, it is concluded that the ANF leads and reference assemblies are neutronically compatible.

Reference 1

shews additional calculation detail.

r 4:

t-

a A

~

8 ANF 88 019 I,{5 1

I 1 Table 3.1 Key Neutronic Design Parameters for 3,?

Calvert Cliffs lead Assemtliec l

Calvert ANF Cliffs Enrichment, w/o U-235 4.08 4.08 Number of Fuel Pins 176 164 Fuel Pin Array 14x14 14X14 Fuel Assembly Pitch, in.

8.18 3.18 i

Fuel Red Pitch, in.

0.ESO 0.580 a tive Fuel Length, in.

135.7 136.7 jJ c

Fuel Pellet 00, in.

0.3700 0.3755

,]

Clad CD, in.

0.440

. 440

'ad 10, in.

0.375 0.354

)

^ ice Tube OC. in.

1.115 1.115 Suice Tute ID, in.

1.d35 1.035 i

uel Red. Mass, kg UO / Rod

2 Standard Rod 2.45 2,52 Eurnable absorber Rod 2.17 Burnable absorber Red Data

,'-]

j Nu-ber of Absor:er Rods 12 12 Gacolinia Concentratdan, w/o Gd 023 10.0 Active Absorber gms B10/in.

0.036 ll Active Absorber, Length, in.

122.7 122.7 Natural UO2 Length Rod Top / Bottom, in.

7.0/7.0 J

'bsorber Pellet 00, in.

0.3700 0.3620 i%a j

j Abscrter Clad CD, in.

0.440 0.440 I

'Morrer Clad ID, in.

0.373 0.3S4

.1! ',

4l'I

l

!7 t c

'l 9

ANF 88 019 r,

- t1 fi 4.0 THEU'AL HvCRAUllC COP?ATIBILITY ANALYSIS 5.1 The co patibility analyses demonstrate that applicable thermal-hydraulic l

design criteria are met, including MONSR and fuel centerline temperature s

q, requirements.

LOCA/ECCS performance of the ANF lead assemblies is also

,ti addressed.

These' analyses are reported in greater detail in Reference 1.

k, It is assumed that the ANF lead as',tmblies will be inserted in non-limiting, oga reduced coaer locations in the core.

This assumption is considered

,}'

a:p opriate fo a demonstration program.

The calculations therefore simulated the INF lead asse.clies at a power level 5'; belcw that of the het assembly in i

the core.

'he results

-dicate that insertion of the ANF lead asse.blies into the

'3 ert Clis c:re will net significantly affect the limiting asse oly MONER.

e l'. F 'ea: asse.: lies are ex:ected to exhibit LCCA/ECCS peak cladding te :erstares re':w those of t e li.itirg :c resident fuel due to their recuced a

cae" 10catier 'n the core.

I1]

.1 Thr al N r nlic Desic "riteria l

i i}

The primary ther al-hydraulic design criteria for ANF reload fuel assure that tuel rod integrity is maintaines during normal operation and Anticipated C:eratienal Cc:urrences (200s).

Specific criteria are:

l

}

1)

Avoicance of CNS for the limiting red in the core with 95*. pr:bability at a 95', confidence level.

4 j

2)

Fuel centerline temperatures re ain telow the melting point of the fuel j

l~

pellets.

1 1

i 11 i

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

- - = - -

i 10 ANF.es.019 I

i Observance of these criteria is considered conservative relative to the requirement that ACOs not result in fuel rod f ailures or loss of functional capability.

Hydraulic Characterization of ANF and Co Resident Fuel 4.2 Co.bustion 1.oss coef ficients for the ANF standard fuel and the co resident Engineering (CE) fuel were derived fron pressure drop tests performed in ANF's facility (10)

The loss coefficient for the HP portable loop hydraulic test spacer was estimated from pressure drop test data for a si-ilar design.

Results indicate that the overall assembly loss coefficient fy the ANF The fl:. d'.ersion effect star.dard fual exceeds that of the CE fuel by 3.7%.

of these differences in loss coefficient is assessed in the "F ',' 2 R calculations.

The four ANF lead assemblies will not siEnificantly affect total core flow.

A3 MENE: h?l u a t i e n o '.' N 3.ead asse-blies and Cn et de

• I':e1 i

a'3 toad "C'iER calculations were perfor ea to assess the perfor.ance

' t$e l

The calculatices e ? c. sed tne asse-olies and their impact on the core MONER.

III) and ANf's approved thermal nydraulic etnedology for

-t re!

XCCERA-Ill code II2)

The results of these calculations are given in hole 4.1.

cores The tNE ONE correlatien MONERs are calculated using the XNS DNS correlation.

is de-onstrated to be applicable to the co resident CE fuel and the a'J respec* vely.

The A'J h7 standarc fuel asser.blies in References 13 ano 14, spacer is specifically designed to yield i provec ;NE ;er'cr ance rela!'<e *:

l

7x!7 Hi; soace-Flow.ixing data for the s1-tiar the ANF standard spacer.

design demonstrate significantly teproved mixine relative to !"e ANF stand No cMcM aa; spacer, supporting the expectation of improved CNB ;er'or ance.

th's analysts.

this indication of i'; roved CNER perfor-ance

'9 taken for

's concluded that the XNE correlation ay te ccnservative' so; lied to t*e

J MIP spacer leads in this analysis-i i

I i

W ul l

,3

._ o 11 ANF-88 019 r]

s i

Four cases were considered in the MONER analysis.

A MONER of 1.46 was calculated for the liniting assembly in an all Combustion Engineering (CE) core.

This case provides a value to which the subsequent mixed core case results may be comparec.

A MDNER of 1.52 was calculated for the ANF standard fuel assembly, containing nine bi metallic spacers.

A MDNBR of 1.70 was calculated for the ANF HTP fuel, containing nine HTP spacers.

The MONER for y

the limiting CE asse-bly in a core containing four ANF HTP assemblics was calculated to be 1.45, essentially the same as the all CE core reference case.

Nominal reactor and fuei design parameters employed in the calculations are given in Table 4.2.

The ethodology e ployed accounts explicitly for the ef fects of flow diversion between fuel assemblics having different hydraulic character.

The ANF lead asser: lies exhibit greater MDNERs than tne reference all CE 3:se-Oly he t0 t*e 5',

'cwer asse 01.

wer at anich t"e A'J leads were

mulated.

The UJ LT? leal asse cij has a 1cwer resistaare spacer than the UF standard assem0'y cr tae co resident CE fuel and thus benefits frcn civersior, crossflcw.

This ircretsed flow in the HTP asse-Oly is responsible f:r its improved MDNER performance relative to the c'6 standard fuel.

No credit has been taken for the HTP spacers' irproved c:clant mixing capability, i

j The MCNBRs calculated for the standard ar.d all HTP assemblies are expected to tound that of the single hybrid asse-oly.

The hytrid asse-bly contains 7

_]

interior HTP spacers with bi-metallic spacers in the top and bottom positions.

]

Ine MCNER calculations indicate that the ANF lead assemblies will have MDNBRs larger than that of the hot CE asse-bly.

Insertion of the lead assemblies

[1 dces not significantly affect the MDNER of the hot CE asse-bly, which asse~bly establishes the core MCNER.

Thus, the core MDNBR is expected to be unchanged I~,

"alvert Cliffs core has been shown to meet the 95/95 DNER criterion in previous submittals prepared y the censee.

Ce this basis, it is concluded

12 ANF.SE.019 t

i that the design criterion on DNER is satisfied by the mixed core containing ANF leads.

i 4.4 Fuel Centerline Te-cerature i

i The power level required to produce centerline celt in zircaloy clad urania I

fuel rods is about 21 kW/f t.

Loss-of coolant accident considerations for Calvert Cliffs limit the steady state peak linear heat generation rate (LHGR)

]

to 15.5 kW/ft.

The 35". margin between the 21 kW/f t design limit and the 15.5 1

kW/ft operating limit is large enough that fuel centerline melt is not a limiting factor for anticipated operr.tional transients.

4 1

)

4.5 LOCI /ECCS Performance i

to eet the i

The limiting co resident fuel at Calvert Clif fs has been shca a

1 l

a:pr:priate LCC A/ECCS acceptance criteria.

The fuel red :esi;n Of the ;'if l

lead asse?0 lies is similar to that of the co resident fue!

T ere':ro, at

(

a ';u i '. a i e n t p:wer l ev el s, the ANF lea 0 asse Dlies are ex;ecte to ein1 bit I

(

LOCA/ECCS performance characteristics similar to that Of 19e 14-iting co-resident fuel.

J l

The ANF lead assemblies will be operated at a power level at least 5'i belod that of the limiting co resident assembly in the core.

Inis reduction in poaer level would be expected to yield s igni fi c an'.ly i prove: LOCAJECCS f

results relative to that of the limiting co resident fuel.

j l

3 Ine lead asse-bly fuel rod design is essentially the sa e as

• hat Of other 1NF 14x14 CE type reload fuel.

ANF has de onstrated that this design eets applicatie acceptance criteria in other CE reactors sie.ilar to Calvert Cliffs.

l Sased on the similarity of the lea: asse-bly fuel ro: :esign to :thers for l

which acceptable LOC A/ E C".5 perfor ance has been ce onstratec, and on the l

re:uced power at which the lead asse-blies will be 0:e"a'e:. 't 's ::n:lVded h

4

{ l 13 ANF SS 019 J

I that the lead asse-blies will exhibit acceptable LOCA/ECCS performance in the Calvert Cliffs reactor.

t J.

6

?

a 1,

e.

4 J

l 1;e i

I l-I i

4 i,

i 1

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)

i j

1

! L. 4 1

I.,

\\

~.

- - - - - - - - - - - -. ~

l l

14 ANF-88-019 s

Table 4.1 MONBR Results for CE and ANF Lead Assemblies MDNBR for MDNBR for limitina CE Assembly ANF lead Assembly All CE Core 1.46 CE Core with Lead ANF Standard

>l.46 1.52 Assembly CE Core with Lead ANF HTP 1.46 1.70 Spacer Assembly A

n

+

~

15 ANF-88-019

,g a

i Table 4.2 Nominal Reactor and Fuel Design Parameters

1. l r

Coeratina Conditions Value Core Rated Power 2700 MWt Fraction of Heat Generated in Fuel 0.975 Minimum Pressurizer Pressura 2200 psia

.l._

Core Inlet Temperature 548'F Total Reactor Coolant Flow 370,000 gpm Bypass Flow Fraction 0.0385 Pa_akirc ictors limitinc CE Lead ANF Axial (Fz) 1.37 1.37 Engineering 1.03 1.03

~

Radial Assembly (F )

1.65 1.57 R

-]

Total Nuclear Peaking 2.33 2.22 ruel Para eters Cl AN.F Fuel Rod CD 0.440 inches 0.440 inches Guice Tube 00 1.115 inches 1.115 inches Rod Array 14x14 14x14 Rod Pitch 0.580 inches 0.580 inches

' ~1 j

Number of Fuel Rod Positions /

176 176 Assembly Numter of Guide Tubes 5

5 j

I Active Fuel Length 136.7 inches 136.7 inches Number of Spacers 9

9 I

w

w-g L

i 16 ANF-88-019 Ii 5.0 REF ERENCEji Il l.

ANF-87-176(P),

Rev.

O, "Design Report for Calvert Cliffs lead Assemblies," Advanced Nuclear Fuels Corp., Richland, WA, January 1988.

I-2.

S tud svi k/NR-82/153 (restricted),

"MICSURN-2:

Microscopic Burnup in Burnable Absorber Rods," Studsvik Energiteknik AB.

3.

Studsvik/NR-81/3 (restricted),

"CASM0-2E:

A Fuel Assembly Burnup Program," Studsvik Energiteknik AB.

4 XN-CC-28(A),

Rev.

3, "XTG - A Two Group Three-Dimensional Reactor Simulator Utilizing Coarse Mesh Spacing,"

Exxon Nuclear Company, Richland, WA, January 1975.

I.

I 5.

XN-75-27(A), "Exxon Nuclear Neutronics Design Methods for Pressurized Water Reactors," Exxon Nuclear Company, Richland, WA, June 1975.

5.

XN-75-27(A), Supplement 1, September 1976.

7 XN-75-27(A), Su.rplement 2, December 1977.

I:

3.

XN-75 27(A), Supplement 3, November 1980.

9.

XN-75-27(A), Supplement a, Cecemcer 1985.

(

10.

XN-NF-78-21, "Single Phase Hydraulic Performance of Exxon Nuclear and g 1 Combustion Engineering Maine Yankee Fuel As sembl i e s, " Exxon Nuclear 3j Company, Richland, WA, February 1979.

11.

XN-NF-75-21(P)(A), Rev. 2, "XCOBRA-IllC: A Computer Code to Determine the f.

i Distribution of Coolant During Steady-State and Transient Core Operation," Exxon Nuclear Company, Richland, WA, January 1986.

{'

12.

XN-NF 82-21(P)(A), Rev.

1, "Application of Exxon Nuclear Company PWR I

Thermal Margin Methodology to Mixed Core Configurations," Exxon Nuclear Company, Richland, WA, September 1983.

13.

XN-NF-621(P)(A), Rev.

1, "Exxon Nuclear CNB Correlation for PWR Fuel Designs," Exxon Nuclear Company, Richland, WA, September 1983.

14 XN-NF-83-08(P), "Jus t i fica t ion of XNB Departure from Nucleate Roiling Correlation for St. Lucic Unit 1,"

Exxon Nuclear Company, Richland, WA, February 1983.

1

,