ML20137Y592

Amend 90 to License DPR-69,changing Tech Specs to Reflect Changes in Analysis to Accommodate Cycle 7 Operation

ML20137Y592
Person / Time
Site:	Calvert Cliffs
Issue date:	11/21/1985
From:	Butcher E Office of Nuclear Reactor Regulation
To:
Shared Package
ML20137Y598	List: ML20137Y592 ML20137Y601
References
NUDOCS 8512110082
Download: ML20137Y592 (28)

Text

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NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 p

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I-BALTIM0RE GAS AND ELECTRIC COMPANY DOCKET NO. 50-318 CALVERT CLIFFS NUCLEAR POWER PLANT, UNIT N0. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.90 License No. DPR-69 1.

The Nuclear Regulatory Comission (the Comission) has found that:

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A.

The applications for amendment by Baltimore Gas & Electric Company (the licensee) dated August 29 and August 30, 1985, comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the applications, the provisions of the Act, and the rules and regulations of the Comission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and

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safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.

The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and The issuance of this amendment is in accordance with 10 CFR Part 51 E.

of the Comission's regulations and all applicable requirements have been satisfied.

r 8512110082 851121 PDR ADOCK 05 COO 318 PDR p

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2.

Accordingly, the license is amended by changes to the Technidhl Specifications as indicated in the attachment to this licensi amendment, and paragraph 2.C.2 of Facility Operating License No. DPR-69 is hereby amended to read as follows:

2.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 90, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Edward J. Butcher, Acting Chief Operating Reactors Branch #3 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: November 21, 1985 9

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ATTACHMENT TO LICENSE AMENDMENT N0. 90 FACILTIY OPERATING LICENSE NO. DPR-69

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DOCKET N0. 50-318 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change. The corresponding overleaf pages are provided to maintain document completeness.

Remove Pages Insert Pages B 2-1 B 2-1 B 2-3 8 2-3 B 2-5 B 2-5 B 2-6 B 2-6 3/4 1-1 3/4 1-1 3/4 1-5 3/4 1-5 3/4 2-2 3/4 2-2 3/4 2-4a 3/4 2-4a 3/4 2-5 3/4 2-5 3/4 7-1 3/4 7-1 3/4 7-4 3/4 7-4 8 3/4 1-1 B 3/4 1-1 B 3/4 2-1 B 3/4 2-1 B 3/4 2-2 B 3/4 2-2 B 3/4 7-1 B 3/4 7-1 e

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2.1 SAFETY LIMITS BASES 2.1.1 PEACTOR CORE The restrictions.of this safety limit prevent overheating of the fuel cladding and possible cladding perforation which'would result in the release of fission products to the reactor coolant. Overheating of the fuel is prevented by maintaining the steady state peak linear heat rate at or less than 22.0 kw/ft. Centerline fuel melting will not occur for this peak linear heat rate. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temper-ature and Pressure have been related to DNB through the CE-1 correlation.

The CE-1 DNB correlation has been developed to predict the DNB flux and the location of DNB for axially uniform and non-unifom heat flux distri-butions. The local DNB heat flux ratio, DNBR, defined as the ration 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.

K The minimum value of the DNBR during steady state operation, normal operational transients, and anticipated transients is limited to 1.21 l

This value corresponds to a 95 percent probability at a 95 percent con-fidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions.

The curves of Figures 2.1-1, 2.1-2, 2.1-3 and 2.1-4 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and maximum cold leg temperature of various pump combinations for which the minimum DNBR is no less than 1.21 for the family of axial shapes and

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corresponding radial peaks shown in Figure B2.1-1.

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

Reactor operation at THERMAL POWER levels higher than 110% of RAIJD THERMAL POWER is prohibited by the high power level trip setpoint specified in CALVERT CLIFFS - UNIT 2 B 2-1 AmendmentNo.J8,H,gg90

t SAFETY LIMITS BASES l ;-

Table 2.1-1.

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

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

4 The reactor protective system in combination with the Limiting Conditions for Operation, is designed to prevent any anticipated combina-tion of transient conditions for reactor coolant system temperature, pressure, and THERMAL POWER level that would result in a DNBR of less than 1.21 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 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 Com Er which permits a maximum transient pressure of 110% (2750 psia) ponents 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

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I' 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 i

demonstrate integrity prior to initial operation.

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CALVERT CLIFFS - UNIT 2 B 2-3 Amendment No.18,7), J[g' 90 e

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2.1 SAFETY-LIMITS BASES e

2.1.1 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel cladding and possible cladding perforation which would result in the i

release of fission products to the reactor coolant. Overheating of the i-fuel is prevented by maintaining the steady state peak linear heat rate at or less than 22.0 kw/ft. Centerline fuel melting will not occur for this peak linear heat rate. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficieri; is large and the cladding surface temperature is slightly above the ccolant 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 durir.g operation and therefore THERMAL POWER and Reactor Coolant Temper-ature and Pressure have been related to DNB through the CE-1 correlation.

The CE-1 DNB correlation has been developed to predict the DNB flux and the location of DNB for axially unifom and non-unifom heat flux distri-butions. - The -local DNB heat flux ratio, DNBR, defined as the ration 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 1.21 l

This value corresponds to a 95 percent probability at a 95 percent con-

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fidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions.

.c The curves of Figures 2.1-1, 2.1-2, 2.1-3 and 2.1-4 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and maximum cold leg temperature of various pump combinations for which the minimum DNBR is no less than 1.21 for the family of axial shapes and j

4 corresponding radial peaks shown in Figure B2.1-1.

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

Reactor operation at THERMAL POWER levels higher than 110% of RAIJD THERMAL POWER is prohibited by the high power level trip setpoint speci.fied in 1

CALVERT CLIFFS - UNIT 2 B 2-1 Amendment No.18, gy, gg 90

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a nonneiumo unoa wixv CALVERT CLIFFS-UNIT 2 B 2-2 Amendment No. 9

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l l SAFETY LIMITS BASES Table 2.1-1.

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

The conditions for the Thermal Margin Safety Limit curves in 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 Conditicns for Operation, is designed to prevent any anticipated combina-tion of trarsient conditions for reactor coolant system temperature, pressure, and THERMAL POWER level that would result in a DNBR of less than 1.21 and preclude the existence of flow instabilities.

l l2.1.2 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

,r 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 r

transient pressure of 110% (2750 psia) of component design pressura.

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

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

.7 CALVERT CLIFFS - UNIT 2 B 2-3 Amendment No. 18, D, f',( 90

E.

t 2.2 LIMITING SAFETY SYSTEM SETTINGS

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BASES 2.2.1 ' REACTOR TRIP SETPOINTS The Reactor Trip Setpoints specified in Table 2.2-1 are the values at which the Reactor Trips are set for each parameter. The Trip Setpoints

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have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their safety limits. Operation with a trip set less conservative than its Trip Setpoint but within its speci-fied Allowable Value is acceptable on the basis that each Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.

Manual Reactor Trip The Manual Reactor Trip is a redundant channel to the automatic protective instrumentation channels and provides manual reactor trip capability, i

s Power Level-High The Power Level-High trip provides reactor core protection against reactivity excursions which are too rapid to be protected by a Pressurizer t

O Pressure-High or Thermal Margin / Low Pressure trip.

1 The Power Level-High trip setpoint is operator adjustable and can be set no tigher than 10% above the indicated THERMAL POWER level. Operator o

action is required to increase the trip setpoint as THERMAL POWER is increased. The trip setpoint is automatically decreased as THERMAL power decreases. The trip setpoint has a maximum value of 107.0% of RATED E

THERMAL POWER and a minimum setpoint of 30% of RATED THERMAL POWER.

Adding to this maximum value the possible variation in trip point due to calibration and instrument errors, the maximum actual steady-state l

THERMAL POWER level at which a trip would be actuated is 110% of RATED THERMAL POWER, which is the value used in the safety analyses.

Reactor Coolant Flow-Low The Reactor Coolant Flow-Low trip provides core protection to prevent DNB in the e.ent of a sudden significant decrease in reactor coo 3 ant flow. Provisions have been made in the reactor protective systeth to permit CALVERT CLIFFS-UNIT 2 B 2-4 Amendment No. M,[ ;

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 1.21 under normal operation I

and expected transients.

For reactor operation with only two or three reactor coolant pumps operating, the Reactor Coolant Flow-Low trip set-points, the Power Level-High trip setpoints, and the Themal 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 1.21 during l

normal operational transients and anticipated transients when only two 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 operation of the pressurizer code safety valves.

Containment Pressure-High The Containment Pressure-High trip provides assurance that a reactor trip is initiated concurrently with a safety injection.

The setpoint for this trip is identical to the safety injection setpoint.

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 coo N wn of the reactor coolant. The setting of 685 psia i

is sufficiently below the full-load operating point of 850 psia'so as not to interfere with nomal operation, but still high enough to provide the required protection in the event of excessively higi steam flow. This setting was used with an uncertainty factor of + 85 psi in the accident analyses which was based on the Main Steam line. Break event.

CALVERT CLIFFS - UNIT 2 B 2-5 Amendment No. 4, M ff 90 -

LIMITING SAFETY SYSTEM SETTINGS BASES 1

a-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 se,tpoint 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 1.21 nor a peak linear heat rate l

which corresponds to the temperature for fuel centerline melting will exist as a consequence of axial power maldistributions. These trip set-points 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.

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Thermal Margin / Low Pressure

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The Thermal Margin / Low Pressure trip is provided to prevent operation when the DNBR is less than 1.21.

1 The trip is initiated whenever the reactor coolant system pressure signal u ops below either 1875 psia or a computed value as described below, wh1 hever is higher. The computed value is a function of the higher of A power or neutron power, reactor inlet temperature, and the number of ret tor coolant pumps operating. The minimum value of reactor coolant flow lite, the maximum AZIMUTHAL POWER TILT and the maximum CEA deviation pern tted for continuous operation are assumed in the genera-tion of this t ip function.

In addition, CEA group sequencing in accor-dance with Spet ifications 3.1.3.5 and 3.1.3.6 is assumed. Finally, the maximum inserti, n of CEA banks which can occur during any anticipated operational occurrence prior to a Power Level-High trip is assumed.

T CALVERT CLIFFS - UNIT 2 B 2-6 Amendment No.78, jf.J;t'90 j

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I 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 B0 RATION CONTROL SHUTDOWN MARGIN - T

> 200*F

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9 LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be > 3.51* ak/k.

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

ACTION:

With the SHUTDOWN MARGIN < 3.51* ak/k, immediately initiate and continue l

boration at > 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 > 3.5%* ak/k:

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

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

I b.

When in MODES 1 or 2, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying that CEA group withdrawal is within the Transient Insertion Limits of Specification 3.1.3.6.

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 c.

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

d.

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

Adherence to Technical Specification 3.1.3.6 as specified in Surveillance Requirements 4.1.1.1.1 assures that there is sufficient available shut-down margin to match the shutdown margin requirements of the safety analyses.

• • See Special Test Exception 3.10.1.

1. With Kg f >_ l.0.

H With Kg f < l.0.

CALVERT CLIFFS - UNIT 2 3/4 1-1 Amendment ho.7. 18, U,$/I(

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6 REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

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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 con-sideration 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% ak/k at least once per 31 Effective Full Power Days (EFPD). ThTs comparison shall consider at least those factors stated in Specification 4.1.1.1.1.e.

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

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CALVERT CLIFFS-UNIT 2 3/4 1-2 s

1

l REACTIVITY CONTROL SYSTEMS MODERATORTEMPERATURECOEFFICIEE LIMITING CONDITION FOR OPERATION 3.1.1.4 The moderator temperature coefficient (MTC) shall be:

Less r,ositive than 0.7 x 10-4 Ak/k/*F whenever THERMAL l

a.

POWER is 1 70! of RATED THERMAL POWER.

b.

Less positive than 0.2 x 10~4 ak/k/*F whenever THERMAL PO4ER is > 70'; of RATED THERMAL POWER, and Less negative than -2.7 x 10-4 ok/k/*F at RATED THERMAL l

c.

POWER.

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 />.

O S_URVEILLANCE REQUIREMENTS 2

4.1.1.4.1 The MTC shall be determined to be within its limits by confirmatory measurements. MTC measured values shall be extrapolated and/or compensated to permit direct comparison with the above limits.

AWith K,ff >_ l.0.

1. See Special Test Exception 3.10.2.

Y CALVERT CLIFFS - UNIT 2 3/4 1-5 Amendment No. 18. II' I.90 J

REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) f-4.1.1.4.2 The MTC shall ne determined at the following frequencies and THERMAL POWER conditions during each fuel cycle:

a.

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

b.

At any THERMAL POWER above 90% of RATED THERMAL POWER within 7 EFPD after initially reaching an equilibrium condition at or above 90% of RATED THERMAL POWER after each fuel loading.

c.

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

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CALVERT CLIFFS-UNIT 2 3/4 1-6 Amendment No.g j

3/4.2 POWER DISTRIBUTION LIMITS LINEAR HEAT RATE

.s' LIMITING CONDITION FOR OPERATION 3.2.1 The linear heat rate shall not exceed the limits shown on Figure 3.2-1.

t APPLICABILITY: MODE 1.

r ACTION:

With the linear heat rate exceeding its limits, as indicated by four or more coincident incore channels or by the AXIAL SHAPE INDEX outside of I

the power dependent control limits of Figure 3.2-2, within 15 minutes initiate corrective action to reduce the linear heat rate to within the limits and either:

.a.

Restore the linear heat rate to within its limits within one i

hour, or b.

Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

i SURVEILLANCE REQUIREMENTS 2

4.2.1.1 The provisions of Specification 4.0.4 are not applicable.

i 4.2.1.2 The linear heat rate shall be determined to be within its limits by continuously monitoring the core power distribution with either the excore detector monitoring system or with the incore detector monitoring i

system.

l 4

Excore Detector Monitoring System - The excore detector moni-4.2.1.3 toring system may be used for monitoring the core power distribution by:

1 a.

Verifying at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> that the full length CEAs are withdrawn to and maintained at or beyond the Long Term Steady State Insertion Limit of Specification 3.1.3.6.

b.

Verifying at least once per 31 days that the AXIAL SHAPE INDEX l

alarm setpoints are adjusted to within the limits shown.on Figure 3.2-2.

.f CALVERT CLIFFS-UNIT 2 3/42-1 Amendment No. 9, 18

a POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) f' c.

Verifying at least once per 31 days that the AXIAL SHAPE INDEX is maintained within the limits of Figure 3.2-2, where 100 percent of the allowable power represents the maximum THERMAL POWER allowed by the follcwing expression:

MxN where:

1.

M is the maximum allowable THERMAL POWER level for the existing Reactor Coolant Pump combination.

2.

N is the maximum allow 9ble fraction of RATED THERMAL POWER as determined by the F curve of Figure 3.2-3b.

xy 4.2.1.4 Incore Detector Monitoring System - The incore detector moni-toring system may be used for monitoring the core power distribution by verifying that the incore detector Local Power Density alarms:

a.

Are adjusted to satisfy the requirements of the core power distribution map which shall be updated at least once per 31 E

days of accumulated operation in MODE 1.

b.

Have their alarm setpoint adjusted to less than or equal to the limits shown on Figure 3.2-1 when the following factors are appropriately included in the setting of these alarms:

1.

A measurement-calculational uncertainty factor of 1.062.

2.

An engineering uncertainty factor of 1.03.

3.

A linear heat rate uncertainty factor of 1.002 due to axial fuel densification and thermal expansion, and 4.

A THERMAL POWER measurement uncertainty factor of 1.02.

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CALVERT CLIFFS-UNIT 2 3/4 2-2 Amendment No. 5, p, 16, JS, 4. ;f S0

t i 19 9 1

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11.54, 1.01 3,o 5

P 3

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0.9 E

M ro (1.785. 0.8) 0.8 N

ACCEPTABLE VALUE g

0.7 7

1 e

0.6 E

0.5 I

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!.45 1.50 1.55 1.00 1.65 1,70 1.75 1.80 F

,o xy 0

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Figure 3.2-3b TOTAL PLANAR RADIAL PEAKING FACTOR vs N o

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CALVERT CLIFFS - UNIT 2 3/4 2-5 Amendment'No. P.75,37,90

POWER DISTRIBUTION LIMITS TOTAL PLANAR RADIAL PEAKING FACTOR Fh w.-

1.IMITING CONDITION FOR OPERATION T

T 3.2.2.1The' calculated value of F*Y, defined as F

=FXY(1+T ), shall be XY 9

limited to 1 1. 70.

APPLICABILITY: MODE 1*.

ACTION:

With F,T > 1.70 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> either:

y Reduc 9THERMALPOWERtobringthecombinationofTHERMALPOWER a.

and F to within the limits of Figure 3.2-3a and withdraw the full Hngth CEAs to or beyond the Long Tem Steady State Insertion Limits of Specification 3.1.3.6; or b.

Be in at least HOT STANDBY.

SURVEILLANCE REQUIREMENTS g

4.2.2.1.1The provisions of Specification 4.0.4 are not applicable.

T T

xy(1+T)andFfy l

4.2.2.1.2F shall be calculated by the expression F

=F y

y q

shall be detemined to be within its limit at the following intervals:

a.

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

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

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

q

• See Special Test Exception 3.10.2.

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CALVERT CLIFFS - UNIT 2 3/4 2-6 AmendmentNo.p.18.M.8I

I, 3/4.7 PLANT SYSTEMS 3.4.7.1 TURBINE CYCLE

[f SAFETY VALVES LIMITING CONDITION FOR OPERATION 3.7.1.1 All main steam line code safety valves shall be OPERABLE.*

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

a.

With both reactor coolant loops and associated steam generators in operation and with one or more main steam line code safety valves inoperable, operation in MODES 1, 2 and 3 may proceed provided, that within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, either the inoperable valve is restored to OPERABLE status or the Power Level-High trip setpoint is reduced per Table 3.7-1; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b.

With one reactor coolant loop and associated steam generator in operation and with one or more main steam line code safety valves associated with the operating steam generator inoperable, opera-tion in MODES 1, 2 and 3 may proceed provided:

l.

That at least 2 main steam line code safety valves on the f

non-operating steam generator are OPERABLE, and 2.

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

The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.1.1 No additional Surveillance Requirements other than those required by Specification 4.0.5 are applicable for the main steam line code safety valves of Table 4.7-1.

h

• Entry into MODE 3 is permitted to determine operability of main steam line code safety valves. During this time, at least two main steam line co'de safety valves per steam generator shall be OPERABLE.

CALVERT CLIFFS - UNIT 2 3/4 7-1 Amendment No.90

's

$9 TABLE 3.7-1 n>

h MAXIMUM ALLOWABLE POWER LEVEL-HIGH TRIP SETPOINT WITH INOPERABLE g

STEAM LINE SAFETY VALVES DURING OPERATION WITH BOTH STEAM GENERATORS P

5 Maximum Allowable Power Maximum Number of Inoperable Safety Level-High Trip Setpoint m

Valves on Any Operating Steam Generator (Percent of RATED THERMAL POWER)

Eq 1

93 m

2 79 3

66 w

?

~

t I

B O

~

Os W

TABLE 3.7-2 g

r-

5i MAXIMUM ALLOWABLE POWER LEVEL-HIGH TRIP SETPOINT WITH INOPERABLE 5

STEAM LINE SAFETY VALVES DURING OPERATION WITH ONE STEAM GENERATOR P

4 Maximum Allowable Power T

Maximum Number of Inoperable Safety Level-High Trip Setpoint E

Valves on The Operating Steam Generator (Percent of RATED THERMAL POWER) w N

1 40 2

35 3

29 R.

w

. o.

9a:

E,

..,.-e

d 10 TABLE 4.7-1 n5 STEAM LINE SAFETY VALVES PER LOOP 5

8 VALVE NUMBER

' LIFT SETTINGS

• ALLOWABLE ORIFICkSIZE l.

I Q

m s,

3 y

a.} RV-3992/4000 s

935-995 psig R

s

[

b.(

RV-3933/4001,

935-995 psig R

c.

RV-3994/4002,

935-1035 psig R'

d.

RV-3995/4003

~ 935-1035 psig R

a e.

RV-3996/4004 935-1065 psig R

y f.

'RV-3997/4005 935-1065 psig R

)

y 9

'RV-3999/,400G 935-1065 psig R

., e h.

RV-3999/4007 935-1065 psig R

)

'g

[

j n

).

g s

e 6

I 'y,,, \\

n

.~

g

.s,

=

• Lift settings for a given steam line are also acceptable if any 2 valves _ lift between 935 and f

y 995 psig, any 2 cther valves lift'between 935 and 1035 psig, and the 4 remaining valvesilft

-s g

between 935 and 1065 psig.

l

)

v.s a

s y

's

l

~;

s s

x

.s 4

m

3/4.1 REACTIVITY-CONTROL SYSTEMS BASES

{

3/4.1.1 80 RATION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN

~~

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

I SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration and RCS T The minimum available SHUTDOWN MARGIN for no load operating conditions $l9beginning of life is 3.5%

1 Ak/k and at end of life is 3.5% Ak/k.

The SHUTDOWN MARGIN is based on the safety analyses performed for a steam line rupture event initiated at no load conditions. The most restrictive steam line rupture event occurs at EOC conditions.

For the steam line rupture event at beginning of cycle conditions,

(~

a minimum SHUTDOWN MARGIN of less than 3.5% Ak/k is required to control the reactivity transient, and end of cycle conditions require 3.5% Ak/k. Accordingly, theSHUTDOWNMARGINrequirementisbaseduponthislimitingconditgonandis consistent with FSAR safety analysis assumptions. With T 200 F, the reactivity transients resulting from any postulated accid l0 <re minimal and a a

3% Ak/k shutdown margin provides adequate protection. With the pressurizer level less than 90 inches, the sources of non-borated water are restricted to increase the time to criticality during a boron dilution event.

3/4.1.1.3 BORON DILUTION E

A minimum flow rate of at least 3000 GPM provides adequate mixing, prevents stratification and ensures that reactivity changes will be gradual during boron concentration reductions in the Reactor Coolant System. A flow rate of at least 3000 GPM will circulate an equivalent Reactor Coolant System volume of 9,601 cubic feet in approximately 24 minutes. The reactivity change rate associated with boron concen-tration reductions will therefore be within the capability of operator recognition and control.

3/4.1.1.4 MODERATOR TEMPERATURE C0 EFFICIENT (MTC)

The limitations on MTC are provided to ensure.t6at the assumptions used in the accident and transient analyses remain valid through each fuel cycle. The surveillance requirements for measurement of the MTC during each fuel cycle are adequate to confirm the MTC value since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup. The confinnation that the imeasured MTC value is within its limit provides assurances that the coefficient will be maintained within acceptable values throughout each fuel cycle.

CALVERT CLIFFS - UNIT 2 B 3/4 1-1 Amendment No. IS,90 U, H, 72,

n l

1 s

REACTIVITY CONTROL SYSTEMS-4 0

BASES J

,o l

s 3/4.1.1.5 MINIMUM TEMPERATURE FOR CRITICALITY F-This specification ensures thati'the reactor will not be made critical g

with the Reactor Coolant System averag'e temperature less than 515 F.

This limitation is required to ensure 1)'the moderator temperature coefficient is within its analyzed temperature rahge, 2) the protective instrumentation is within its normal operating range, 3) the pressurizer is capable of being 1

in an OPERABLE status with a steam bubble, and 4) the reactor pressure vessel is above its minimum RTNDT temperature.

3/4.1.2 BORATIM SYSTEMS The boron injection syst ensures that negative reactivity control is available during each mode of facility operation. The system also provides-coolant flow following a SIAS (e.g., during a Small Break LOCA) to supplement flow from the Safety Injection System. The Small Break LOCA analyses assume flow from a single charging pump, accounting for measwement uncertainties and flow maldistribution effects-in calculating a conservative value of charging flow actually delivered to the RCS. The components required to perform this function include 1) borated water sources, 2) charging pumps, 3) separate flow paths, 4) boric acid pumps, 5) associated heat tracing systems, and 6) an emergency power supply from OPERABLE diesel generators.

0 With the RCS average temperature above 200 F, a minimum of two separate and redundant boron. injection systems are provided to ensure single functional capability in the event an assumed failure renders one of the systems inoper-E able. Allowable.out-of-service periods enssre that minor component repair or

. corrective action may be completed without undue risk to overall facility safety from injection system failures during the repair period.

The boration capability of either syste is sufficient to provide a SHUT-DOWN MARGIN frog all operating conditions of 3.0% ak/k after xenon decay and cooldown to 200 F.

The maximum boration capability requirement occurs at E0L from full power equilibrium xenon conditions and requires 6500 gallons of 7.25% boric acid solution from the boric acid tanks or 55,627 gallons of 2300 ppm borated water from the refueling water tank. However, to be consistent with the ECCS requirements, the RWT is required to have a minimum contained volume of 400,000 gallons during MODES 1, 2, 3 and 4.

The maximum boron concentration of the refueling water tank shall be limited to 2700 ppm and the maximum boron concentration of the boric acid storage tanks shall be limited to 8% to preclude the possibility of boron precipitation in the core during long term ECCS cooling.

0 With the RCS temperature below 200 F, one injection system 3s acceptable without single failure consideration on the basis of the stabl~e reactivity condition of the reactor and the additional restrictions prohibi. ting CORE ALTERATIONS and positive reacti' ity change in the event the single injection v

system becomes inoperable.

CALVERT CLIFFS - UNIT 2 B 3/4 1-2 Amendment No. 31, 89

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

Either of the two core power distribution monitoring systems, the Excore Detector Monitoring System and the Incore Detector Monitoring System, provide adequate monitoring of the core power distribution and are capable of verifying I

that the linear heat rate does not exceed its limits. The Excore Detector Monitoring System performs this function by continuously monitoring the AXIAL SHAPE INDEX with the OPERABLE quadrant symmetric excore neutron flux detectors and verifying that the AXIAL SHAPE INDEX is maintained within the allowable limits of Figure 3.2-2.

In conjunction with the use of the excore monitoring system and in establishing the AXIAL SHAP INDEX limits, the following assump-tions are made:

1) the CEA insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are satisfied, 2) the AZIMUTHAL POWER TILT restrictions of Specification 3.2.4 are satisfied, and 3) the TOTAL PLANAR RADIAL PEAKING FACTOR does not exceed the limits of Specification 3.2.2.

The Incore Detector Monitoring System continuously provides a direct L

measure of the peaking factors and the alarms which have been established for

~

the individual incore detector segments ensure that the peak linear heat rates will be maintained within the allowable limits of Figurc. 3.2-1.

The setpoints for these alarms include allowances, set in the conservative directions, for l

1) a measurement-calculational uncertainty factor of 1.062, 2) an engineering uncertainty factor of 1.03, 3) an allowance of 1.002 for axial fuel densifica-

,- E tion and thermal expansion, and 4) a THERMAL POWER measurement uncertainty factor of 1.02.

3/4.2.2, 3/4.2.3 and 3/4.2.4 TOTAL PLANAR AND INTEGRATED RADIAL PEAKING T

FACTORS - F AND F AND AZIMUTHAL POWER TILT - T 7

q ThelimitationsonFfy and T are provided to ensure that the assumptions q

used in the analysis for establishing the Linear Heat Rate and Local Power Density - High LCOs and LSSS setpoints remain valid during operatio at the various allc,wable CEA group insertion limits. The limitations on F and T areprovidedtoensurethattheassumptionsusedintheanalysisesablishing

{-

the DNB Margin LCO, and Thermal Margin / Low Pressure LSSS setpoints remain vali durin operation at the various allowable CEA group insertion limits.

If F or F or T exceed their basic limitations, operation may continue unde theaditio0alrestrictionsimposedbytheACTIONstatementssincethese

~

additional restrictions provide adequate provisions to assure that the assump-tions used in establishing the Linear Heat Rate, Thermal Margin / Low Pressure I

CALVERT CLIFFS - UNIT 2 B 3/4 2-1 Amendment No. JB,3J,38,67, 90

)s b

POWER DISTRIBUTION LIMITS BASES andLocalPowerDensity-HighLCOsandLSSSsetpointsremainval[id. An AZIMUTHAL POWER TILT > 0.10 is not expected and if it should occ4r,fsubsequent operation would be restricted to only those operations required to identify the cause of this unexpected tilt.

thatmustbeusedintheequationFfy=F,y(1+T)

The value of Tq q

and FT+F7 (1 + T ) is the measured tilt.

7 q

T The surveillance requirements for verifying that F

,F and T are within 7

q T

their limits provide assurance that the actual values of Fxy, F and T do not r

q i

exceed the assumed values.

Verifying F and Ff after each fuel loading prior

]

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

3/4.2.5 DNB PARAMETERS The limits on the DNB related parameters assure that each of the parameters are maintained within the normal steady state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the safety analyses assumptions and have been analytically demonstrated adequate to main-tain a minimum DNBR of 1.21 throughout each analyzed transient.

l~

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

The second criterion is to ensure that the

,l existing core power distribution at full power is less severe than the power distribution factored into the small-break LOCA analysis. This results in a limitation on the allowed negative AXIAL SHAPE INDEX value at full power. The

~

third criterion is to maintain limitations on peak linear heat rate at low power levels resulting from Anticipated Operational Occurrences (A00s).

Figure 3.2-4 is used to assure the LHR criteria for this condition because the linear heat rate LCO, for both ex-core and in-core monitoring, is set to main-tain only the LOCA kw/ft requirements which are limiting at high pcwer levels.

i At reduced power levels, the kw/ft requirements of certain A00s (e.g., CEA withdrawal), tend to become more limiting than that for LOCA.

4 l

The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodic surveillance of these parameters through instrument 4

readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation. The 18 month periodic measurement of the RCS total flow rate is adequate to detect flow degradation and ensure correlation of the flow indication channels l

with measured flow such that the indicated percent flow will provide sufficient

~

j ~

verification of flow rate on a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> basis.

1' l

CALVERT CLIFFS - UNIT 2 B 3/4 2-2 Amendment No.18,3J,38,67, 90

/

1

-. ~ -.., ~ -

l 3/4.7 PLANT SYSTEMS BASES 3/4.7.1 TURBINE CYCLE

[-

3/4.7.1.1 SAFETY VALVES The OPERABILITY of the main steam line code safety valves ensures that the secondary system pressure will be limited to within 110% of its design l

pressure of 1000 psig during the most severe anticipated system operational transient. The total relieving capacity for all valves on all of the steam lines is 12.18 x 106 lbs/hr at 100% RATED THERMAL POWER. The maximum relieving capacity is associated with a turbine trip from 100% RATED THERMAL POWER coincident with an assumed loss of condenser heat sink (i.e., no steam bypass to the condenser). The main steam line code safety valves are tested and maintained in accordance with the requirements of Section XI of the ASME Boiler and Pressure Code,1971 Edition. The as-left lift settings will be no less than 985 psig to ensure that the lift setpoints will remain within specification during the cycle.

In MODE 3, two main steam safety valves are required OPERABLE per steam generator. These valves will provide adequate relieving capacity for removal of both decay heat and reactor coolant pump heat from the reactor coolant system via either of the two steam generators. This requirement is provided to facilitate the post-overhaul setting and operability testing of the safety valves which can only be conducted when the RCS is at or above 500*F.

It allows entry into MODE 3 with a minimum number of main steam safety valves OPERABLE so that the set pressure for the remaining valves can be adjusted in the plant.

This is the most accurate means for adjusting safety valve set pressures since the valves will be in thermal equilibrium with the operating

~

environment.

STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis of the reduction in secondary system steam flow and THERMAL POWER required by the reduced reactor trip settings of the Power Level-High channels. The reactor trip setpoint reductions are derived on the following bases:

For two loop operation SP = ( } "XI }( ) x 106.5 For single loop operation (two reactor coolant pumps operating in the same loop)

SP = (X) -

)(U) x 46.8 where:

SP = reduced reactor trip setpoint in percent of RATED THERMAL POWER V = maximum number of inoperable safety valves per steam line CALVERT CLIFFS - UNIT 2 B 3/4 7-1 Amendment No.90

PLANT SYSTEMS BASES maximumnumberofinoperablesafetyvalvespproperating U

=

steam line 106.5 Power Level-High Trip Setpoint for two loop operation

=

46.8

. Power Level-High Trip Setpoint for single loop operation

=

with two reactor coolant pumps operating in the same loop Total relieving capacity of all safety valves per steam X

=

line in lbs/ hour Maximum relieving capacity of any one safety valve in Y

=

lbs/ hour 3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM The OPERABILITY of the auxiliary feedwater. system ensures that the Reactor i

Coolant System can be cooled down to less than 300 F from normal operating 0

conditions in the event of a total loss of offsite power. A capacity of 400 gpm is sufficient to ensure that adequate feedwater flow is available to remove decay heat and reduce the Reactor Coolant System temperature to less than 3000F when the shutdown cooling system may be placed into operation.

Flow control valves, installed in each leg supplying the steam generators, are set to maintain a nominal flow setpoint of 200 gpm plus or minus 10 gpm for j

operator setting band. The nominal flow setpoint of 200 gpm incorporates a total instrument loop error, band of plus 25 gpm and minus 26 gpm for the motor-driven pump train.

The corresponding values for the steam-driven pump train are plus 37 gpm and minus 40 gpm.

l The operator setting band, when combined with the instrument loop error, results in a total flow' band of 164 gpm (minimum) and 235 gpm (maximum) for the motor-driven pump train. The corresponding values for the steam-driven pump train are 150 gpm (minimum) and 247 gpm (maximum). Safety analyses show that more flow during an over;.ooling transient and less flow during an undercooling transient could be tolerated; i.e., flow fluctuations outside this flow band but within the assumptions used in the analyses listed below, are allowable.

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

(1) Loss of Feedwater 0 gpm Auxiliary Feedwatar Flow (2) Feedline Break 0 gpm Auxiliary Feedwater Flow CALVERT CLIFFS - UNIT 2 B 3/4 7-2 Amendment No. (J 49, $2 78 i

i i

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

--,,.a-,,,

.---,,,,,,,,--,-.,,,.a-_w.-,,,,,., - - -