ML17227A304
| ML17227A304 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 02/25/1992 |
| From: | Berkow H Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17227A305 | List: |
| References | |
| NUDOCS 9202280055 | |
| Download: ML17227A304 (19) | |
Text
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+y*~k UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON. D.C, 20555 FLORIDA POWER 5 LIGHT COMPANY DOCKET NO. 50-335 ST.
LUCIE PLANT UNIT NO.
1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.
License No.
DPR-67 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Florida Power 5 Light Company, et al. (the licensee),
dated December 17,
- 1991, 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.
9202280055 920225,"T PDR ADOCK 05000335 P
2.
Accordingly, Facility Operating License No.
DPR-67 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.
- 113, 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 its date of issuance and shall be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION
Attachment:
Changes to the Technical Specifications Herbert N. Berkow, Director Project Directorate II-2 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation Date of Issuance:
February 25, 1992
ATTACHMENT TO LICENSE AMENDMENT NO.
113 TO FACILITY OPERATING LICENSE NO.
DPR-67 DOCKET NO. 50-335 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 area of change.
The corresponding overleaf pages are also provided to maintain document completeness.
Remove Pa es 3/4 4-22 3/4 4-24 83/4 4-12 Insert Pa es 3/4 4-22 3/4 4-24 B3/4 4-12
3/4.4. 9 PRESSURE/TEMPERATURE LIMITS REACTOR COOLANT SYSTEM lIMITING CONDITION FOR OPERATION 3.4.9.1 The Reactor Coolant System (except the pressurizer).
temperature and pressure shall be limited in accordance with the limit lines shown on Figures 3.4-2a, 3.4-2b and 3.4-3 during heatup, cooldown, criticality, and inservice:
leak and hydrostatic testing.
APPLICABILITY: At all times.*8 ACTION:
With any of the above limits exceeded, restore the temperature and/or pressure to within the limits within 30 minutes; perform an analysis to determine the effects of the out-of-limit condition on the fracture toughness properties of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operations or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce the RCS T" to less than 200'F within the'olldwing 30'ours in accordance with'igures 3.4-2b and 3.4-3.
- When the flow path from the RWT to the RCS via a single HPSI pump is established per 3.1.2.3, the heatup and cooldown rates shall be established in accordance wi;th Fig. 3.1-lh.
8During hydrostatic testing operations above system design pressure, a
maximum temperature change in any one hour period shall be limited to 5'F.
ST.
LUGIE - UNIT 1 3/4 4-21 Amendment No. b, pg, io4
REACTOR COOLANT SYSTL
~
'URVEILLANCE REQUIREMENTS 4.4.9.1 a
~
b.
C.
The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup,
- cooldown, and inservice leak and hydrostatic testing operations.
The Reactor Coolant System temperature and pressure conditions shall be determined to be to the right of the criticality limit line within 15 minutes prior to achieving reactor criticality.
The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties as required by 10 CFR 50 Appendix H.
The results of these examinations shall be used to update Figures 3.4-2a, 3.4-2b and 3.4-3.
ST.
LUCIE
UNIT 1
3/4 4-22 Amendment No. K-,
- 113,
DELETED ST.
LUGI E UNIT 1
3/4 4-24 Amendment No. 460, 113,
DELETED ST.
LUCIE - UNIT 1 B 3/4 4-11 Amendment No. 8l
R ACTOR COOLANT SYST BASES for piping, pumps and valves.
Below this temperature, the system pressure must be limited to a maximum of 20X of the system's hydrostatic test pressure of 3125 psia.
The limiations imposed on the pressurizer heatup and cooldown rates and spray water temperature differential are provided to assure that the pressurizer is operated within the design criteria assumed for the fati-gue analysis performed in accordance with the ASHE Code requirements.
3 4.4.10 STRUCTURAL INTEGRITY The inservice inspection program for ASHE Code Class 1,
2 and 3
components ensure that the structural integrity of these components will be maintained at an acceptable level throughout the life of the plant.
This program is in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR
'Pa} t 50.55a(g) except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50.55a(g)(6)(i).
Components of the reactor coolant system were designed to provide access to permit inservice inspections in accordance with Section XI of the ASHE Boiler and Pressure Vessel Code 1971 Edition and Addenda through Winter 1972.
ST.
LUCIE UNIT 1 B 3/4 4-12 Amendment No. 90,
- 113,
.~p,R REGS(1p0 UNITED STATES NUCLEAR REGULATORY COMIVllSSION WASHINGTON, D.C. 20555 FLORIDA POWER 5 LIGHT COMPANY ORLANDO UTILITIES COMMISSION OF THE CITY OF ORLANDO FLORIDA AND FLORIDA MUNICIPAL POWER AGENCY DOCKET NO. 50-389 ST.
LUCIE PLANT UNIT NO.
2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 54 License No. NPF-16 1.
The Nuclear Regulatory Commission (the Commi'ssion) has found that:
A.
The application for amendment by Florida Power 5 Light Company, et al. (the licensee),
dated December 17,
- 1991, 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 regul'ations; 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.
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. 54
, 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 its date of issuance and shall be implemented within 30 days.
FOR THE NUCLEAR GULATORY COMMISSION
) 'r-,-)eM ~
Herbert N. Berkow, Director Project Directorate II-2 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
February 25, 1992
ATTACHMENT TO LICENSE AMENDMENT NO.
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 area of change.
The corresponding overleaf pages are also provided to maintain document completeness.
Remove Pa es XXIV 3/4 4-30 3/4 4-33 B3/4 4-11 Insert Pa es XXIV 3/4 4-30 3/4 4-33 B3/4 4-11
LIST OF TABLES INDEX TABLE 1.2
- 2. 2-1 FREQUENCY NOTATION.
OPERATIONAL MODES REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS PAGE 1-8 1-9 2-4
- 3. 1-1
- 3. 2-1 3.2-2 3.3-1 3.3-2 MONITORING FREQUENCIES FOR BACKUP BORON DILUTION DETECTION FOR ST. LUCIE-2.....
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D DNB MARGIN LIMITS............................
3/4 1-17 3/4 2-11 I
3/4 2-15 REACTOR PROTECTIVE INSTRUMENTATION........................ 3/4 3-2 REACTOR PROTECTIVE INSTRUMENTATION RESPONSE TIMES......... 3/4 3-6
- 4. 3-1 3.3-3 REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE REQUIREMENTS..
ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION.
.... 3/4 3-8 3/4 3-12 3.3-4 3.3-5 4.3-2
- 3. 3-6
- 4. 3-3 3 ~ 3 7
- 4. 3-,4 3.3-8 4.3-5 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS.....
3/4 3-22 RADIATION MONITORING INSTRUMENTATION...................... 3/4 3-25 RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.
~........
3/4 3-28 SEISMIC MONITORING INSTRUMENTATION........................ 3/4 3-33 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.............
3/4 3-34 METEOROLOGICAL MONITORING INSTRUMENTATION.
3/4 3-36 METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.
3/4 3-37 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES...............................
3/4 3-17 ENGINEERED SAFETY FEATURES RESPONSE TIMES.................
3/4 3-19 ST.
LUCIE-UNIT 2 XXIII Amendment No. 8
LIST OF TABLES (Continued)
TABLE INDEX PAGE 3.3-9 4.3-6 3.3-10 4.3-7 3.3-11 3.3-12 4.3-8 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION 3/4 RADIOACTIVE LIQUID fFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE Rf(UIREMENTS............................
3/4 3-49 3-51 REMOTE SHUTDOWN SYSTEM INSTRUMENTATION...............
3/4 3-39 REMOTE SHUTDOWN SYSTEM INSTRUHfNTATION SURVEILLANCE REQUIREMENTS...................
3/4 3-40 ACCIDENT MONITORING INSTRUMENTATION.......
3/4 3-42 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS..
3/4 3-43 FIRE DETECTION INSTRUHENTS...................
3/4 3-45 3.3-13 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION..
. 3/4 3-54 4.3-9 4.4-1 4.4-2 3.4-1 3.4-2 4.4-3 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE Rf(UIREHENTS....................
3/4 3-57 MINIMUM NUMBER OF STEAM GfNERATORS TO BE INSPfCTED DURING INSERVICE INSPECTION...
...... 3/4 4-16 STEAM GENERATOR TUBE INSPECTION 3/4 4-17 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES...... 3/4 4-21 REACTOR COOLANT SYSTEM CHEMISTRY 3/4 4-23 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE REgUIREHfNTS.............
3/4 4-24 4,4-4 k
PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS ROGRAH ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~
~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~
~
~ 3/4 P
4-27 4.4-5 3.4-3 1
3.4-4 3.6-1 3.6-2 DELETED LOW TEMPERATURE RCS OVERPRESSURE PROTECTION RANGE..... 3/4 4-37a MINIMUM COLD LEG TEMPERATURE FOR PORV USE FOR LTOP.... 3/4 4-37a CONTAINMENT LEAKAGE PATHS....
3/4 6-5 CONTAINMENT ISOLATION VALVES,.......................
3/4 6-21 ST.
LUCIE - UNIT 2 XXIV Amendment No. 8, 6a; 54,
REACTOR COOLANT SYSTEM 3/4.4. 9 PRESSURE/TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and 'pressure shall be limited in accordance with the limit lines shown on Figures 3.4-2, 3.4-3 and '3.4-4 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing.
APPL ICABILITY:
At al 1 times.
ACTION:
With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operations or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce the RCS T
to less than 200'F within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> in accordance with Figurel 3.4-3 and 3.4~4.
SURVEILLANCE REQUIREMENTS
- 4. 4.9. l. 1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup,
- cooldown, and inservice leak and hydrostatic testing operations.
ST, LUGI E UNIT 2 3/4 4-29 Amendment No.
PP, p)L, 46,
REACTOR COOLANT SYST SURVEILLANCE REQUIREMENTS (Continued) 4.4.9. 1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, as required by 10 CFR 50 Appendix H.
The results of these examinations shall be used to update Figures 3.4-2, 3.4-3 and 3.4-4.
ST.
LUCI E
UNIT 2 3/4 4-30 Amendment No. 46-,34, 54,
'DELETED ST.
LUCIE UNIT 2 3/4 4-33 Amendment No. 46-,-8k, 54,
REACTOR COOLANT SYSTEM PRESSURIZER HEATUP/COOLDOWN LIMITS LIMITING CONDITION FOR OPERATION 3.4.9.2 The pressurizer temperature shall be limited to:
a.
A maximum heatup of 100~F in any 1-hour period, and b.
A maximum cooldown of 200 F in any 1-hour period.
APPLICABILITY: At al 1 times.
ACTION:
With the pressurizer temperature limits in excess of any of the above limits, restore the temperature to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the "-out-of-limit condition on the structural integrity of the pressurizer; determine that the pressurizer remains acceptable for continued operation or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce the pressurizer pressure to less than 500 psig within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE RE UIREMENTS 4.4.9.2 The pressurizer temperatures shall be determined to be within the limits at least once per 30 minutes during system heatup or cooldown.
ST.
LUCIE - UNIT 2 3/4 4 Amendment No.
16
C OR C 0 ANT SYS BASES asaaaaaaaaaaam-=aaa==a a aaaaaamaaaaaaaasaaaaaaasaaaaem-a-aaeaaaaaaaaa-aaaaa The actual shift in RTM>> of the vessel material will be established periodically during operation by removing and evaluating, in accordance with ASTH E185 and 10 CFR 50 Appendix H, reactor vessel material irradiation surveil-lance specimens installed near the inside wall of the reactor vessel in the core area.
Since the neutron spectra at the irradiation samples and vessel inside radius are essentially identical, the measured transition shift for a sample can be applied with confidence to the adjacent section of the reactor vessel.
The heatup and cooldown curves must be recalculated when the delta RT >>
determined from the surveillance capsule is different from the calculated delFa RT>> for the equivalent capsule radiation exposure.
The pressure-temperature limit lines shown on Figures 3.4-2, 3.4-3 and 3.4-4 for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature require-ments of Appendix G to 10 CFR 50.
The maximum RTg>> for al 1 Reactor Cool ant System pressure-retaining mate-rials, with the exception of the reactor pressure
- vessel, has been determined to be 60'F.
The Lowest Service Temperature limit line shown on Figures 3.4-2, 3.4-3 and 3.4-4 is based upon this RT>> since Article NB-2332 (Summer Addenda of 1972) of Section III of the ASME Baler and Pressure Vessel Code requires the Lowest Service Temperature to be RTg>> + 100 F for piping,
- pumps, and valves.
Below this temperature, the system pressure must be limited to a maxi-mum of 20X of the system's hydrostatic test pressure of 3125 psia.
The limitations imposed on the pressurizer heatup and cooldown rates and spray water temperature differential are provided to assure that the pressurizer is operated within the design criteria assumed for the fatigue analysis performed in accordance with the ASHE Code requirements.
The OPERABILITY of two PORVs, two SDCRVs or an RCS vent opening of greater than 3.58 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold leg temperatures are less than or equal to the LTOP temperatures.
The Low Temperature Overpressure Protection System has adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either (1) a safety injection actuation in a water-solid RCS with the pressurizer heaters energized or (2) the start of an idle RCP with the secondary water temperature of the steam generator less than or equal to 40 F above the RCS cold leg temperatures with the pressurizer water-solid.
ST.
LUGIE - UNIT 2 B 3/4 4-11 Amendment No. 46-,84-,46, 54,
REACTOR COOLANT SYSTEM BASES 3/4.4. 10 REACTOR COOLANT SYSTEM VENTS Reactor Coolant System vents are provided to exhaust noncondensible gases and/or steam from the primary system that could inhibit natural circulation core cooling.
The OPERABILITY of at least one reactor Coolant System vent path from the reactor vessel head and the pressurizer steam space ensures the capa-bility exists to perform this function.
The redundancy design of the Reactor Coolant System vent systems. serves to minimize the probability of inadvertent or irreversible actuation while ensuring that a single failure of a vent valve, power supply, or control system does not prevent isolation of the vent path.
The function, capabilities, and testing requirements of the Reactor Coolant System vent system are consistent with the requirements of Item II.b. 1 of NUREG-0737, "Clarification of TMI Action Plan Requiremegts,"
November 1980.
3/4. 4. 11 STRUCTURAL INTEGRITY The inservice inspection and testing programs for ASME Code Class 1, 2, and 3 components ensure that the structural integrity and operational readiness of these components will be maintained at an acceptable level throughout the life of the plant.
These programs are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR Part 50.55a(g) except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50.55a (g) (6) (i).
Components of the reactor coolant system were designed to provide access to permit inservice inspections in accordance with Section XI of the ASME Boiler and Pressure Vessel Code 1971 Edition and Addenda through Summer 1973.
ST.
LUCIE - UNIT 2 B 3/4 4-12