ML17223B197
| ML17223B197 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 05/23/1991 |
| From: | Berkow H Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17223B198 | List: |
| References | |
| NUDOCS 9106040124 | |
| Download: ML17223B197 (8) | |
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UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 FLORIDA POWER
& LIGHT COMPANY ORLANDO UTILITIES COMMISSION OF THE CITY OF ORLANDO FLORIDA AND FLORIDA MUNICIPAL POWER AGENCY DOCKET NO. 50-389 ST.
LUCIE PLANT UNIT NQ.
2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.
License No. NPF-16 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Florida Power 8 Light Company, et al. (the licensee),
dated October 24,
- 1988, as supplemented June 1, 1989, October 19, 1989, March 27,
- 1990, November 8, 1990 and modified December 18, 1990, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; ~
C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
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2.
Accordingly, Facility Operating License No.
NPF-16 is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and by amending paragraph 2.C.2 to read as follows:
2.
Technical S ecifications The Technical Specifications contained in Appendices A and B,
as revised through Amendment No.
5O
, 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 COMt1ISSION
Attachment:
Changes to the Technical Specifications He ert N. Berkow, Director Project Directorate II-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation Date of Issuance:
t1ay 23, 1991
ATTACHMENT TO LICENSE AMENOMENT NO.
5O TO FACILITY OPERATING LICENSE NO.
NPF-16 DOCKET NO. 50-389 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 also provided to maintain document completeness.
Remove Pa es 8 2-4 3/4 3-7 Insert Pa es B 2-4 3/4 3-7
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SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES
- 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 Coolant System components are designed to Section III, 1971 Edition including Addenda to the
- Summer, 1973, of the ASME Code for Nuclear Power Plant Components which permits a maximum transient pressure of 110K (2750 psia) of design pressure.
The Safety Limit of 2750 psia is therefore consistent with the design criteria and associated code requirements.
The entire Reactor Coolant System was hydrotested at 3125 psia to demonstrate integrity prior to initial operation.
2,2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 AC R
RIP S
PO N
S The Reactor Trip Setpoints specified in Table 2.2-1 are the values at which the Reactor Trips are set for each functional unit.
The Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their Safety Limits during normal operation and design basis anticipated operational occurrences and to assist the Engineered Safety Features Actuation System in mitigating the consequences of accidents.
Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.
Manual Reactor Tri The Manual Reactor 'Trip is a redundant channel to the automatic protective instrument'ation channels and provides manual reactor trip capability.
ST.. LUCIE - UNIT 2 8 2-3
SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES Variable Power Level-Hi h A Reactor trip on Variable Overpower is provided to protect the reactor core during rapid positive reactivity addition excursions which are too rapid to be protected by a Pressurizer Pressure-High or Thermal Margin/Low Pressure Trip.
The Variable Power Level High trip setpoint is operator adjustable and can be set no higher than 9.6'bove the indicated THERMAL POWER level.
Operator 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 107.0X of RATED THERMAL POWER and a minimum setpoint of 15.0X 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 THERMAL POWER level at which a trip would be actuated is 112K of RATED THERMAL POWER, which is the value used in the safety analyses.
Pressurizer Pressure-Hi h
The Thermal Margin/Low Pressure trip is provided to prevent operation when the DNBR is less than 1.23.
The trip is initiated whenever the Reactor Coolant System pressure signal drops below either 1900 psia or a computed value as described below, whichever is higher.
The computed value is a function of the higher of hT power or neutron power, reactor inlet temperature, the number of reactor coolant pumps operating and the AXIAL SHAPE INDEX.
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 Specifica-tions 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.
The Thermal Margin/Low Pressure trip setpoints are derived from the core safety limits through application of appropriate allowances for equipment response time measurement uncertainties and processing error.
A safety margin is provided which includes:
an allowance of 2.0X of RATED THERMAL POWER to compensate for potential power measurement error; an allowance of 3.0 F to compensate for potential temperature measurement uncertainty; and a further allowance of 125 psia to compensate for pressure measurement error and time delay associated with providing effective termination of the occurrence that exhibits the most rapid decrease in margin to the safety limit.
The l25 psia allowance is made up of 'a 55'sia pressure measurement allowance and a
70 psia time delay allowance.
ST.
LUCIE - UNIT 2 Amendment No.8', 50, B 2-4 The Pressurizer Pressure-High trip, in conjunction with the pressurizer safety valves and main steam safety valves, provides Reactor Coolant System protection against overpressurization in the event of loss of load without reactor t'rip.
This trip's setpoint is at less than or equal to 2375 psia which is below the nominal lift setting 2500 psia of the pressurizer safety valves and its operation minimizes the undesirable operation of the pressurizer safety valves.
Thermal Mar in/Low Pressure
SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES Variable Power Level-Hi h
A Reactor trip on Variable Overpower is provided to protect the reactor core during rapid positive reactivity addition excursions which are too rapid to be protected by a Pressurizer Pressure-High or Thermal Margin/Low Pressure Trip.
The Variable Power Level High trip setpoint is operator adjustable and can be set no higher than 9.61K above the indicated THERMAL POWER level.
Operator 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 107.0X of RATED THERMAL POWER and a minimum setpoint of 15.0X 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 THERMAL POWER level at which a trip would be actuated is 112K of RATED THERMAL POWER, which is the value used in the safety analyses.
Pressurizer Pressure-Hi h
The Thermal Margin/Low Pressure trip is provided to prevent operation when the DNBR is less than 1.28.
The trip is initiated whenever the Reactor Coolant System pressure signal drops below either 1900 psia or a computed value as described below, whichever is higher.
The computed value is a function of the higher of hT power or neutron power, reactor inlet temperature, the number of reactor coolant pumps operating and the AXIAL SHAPE INDEX.
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 Specifica-tions 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.
The Thermal Margin/Low Pressure trip setpoints are derived from the core safety limits through application of appropriate allowances for equipment response time measurement uncertainties and processing error.
A safety margin is provided which includes:
an allowance of 2.0X of RATED THERMAL POWER to compensate for potential power measurement error; an allowance of 3.0'F to compensate for potential temperature measurement uncertainty; and a further allowance of 125 psia to compensate for pressure measurement error and time delay associated with providing effective termination of the occurrence that exhibits the most rapid decrease in margin to the safety limit.
The 125 psia allowance is made up of a 55 psia pressure measurement allowance and a
70 psia time delay allowance.
ST, LUCIE - UNIT 2 Amendment No.f, 50, B 2"4 The Pressurizer Pressure-High trip, in conjunction with the pressurizer safety valves and main steam safety valves, provides Reactor Coolant System protection against overpressurization in the event of loss of load without reactor t'rip.
This trip's setpoint is at less than or equal to 2375 psia which is below the nominal lift setting 2500 psia of the pressurizer safety valves and its operation minimizes the undesirable operation of the pressurizer safety valves.
Thermal Mar in/Low Pressure
InM fll FUNCTIONAL UNIT TABLE 3.3-2 (Continued)
REACTOR PROTECTIVE INSTRUMENTATION RESPONSE TIMES
RESPONSE
TIME 1Q.
Loss of Component Cooling Water to Reactor Coolant Pumps ll.
Reactor Protection System Logic 12.
Reactor Trip Breakers 13.
Wide Range Logarithmic Neutron Flux Monitor 14.
Reactor Coolant Flow - Low t
15.
Loss of Load (Turbine Hydraulic Fluid Pressure
- Low)
Not Appl icabl e Not Applicable Not Applicable Not Applicable 0.65 second Not Applicable Neutron detectors are exempt from response time testing.
Response
time of the neutron flux signal portion of the channel shall be measured from detector output or input of first electronic component in channel.
Based on a resistance temperature detector (RTD) response time of less than or equal to 14.0 seconds where the RTD response time is equivalent to the time interval required for the RTD output to achieve 63.2X of its total change when subjected to a step change in RTD temperature.