ML20058B412
| ML20058B412 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 07/14/1982 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Florida Power Corp, City of Alachua, FL, City of Bushnell, FL, City of Gainesville, FL, City of Kissimmee, FL, City of Leesburg, FL, City of New Smyrna Beach, FL, Utilities Commission, City of New Smyrna Beach, FL, City of Ocala, FL, City of Orlando, FL, Orlando Utilities Commission, Sebring Utilities Commission, Seminole Electric Cooperative, City of Tallahassee, FL |
| Shared Package | |
| ML20058B416 | List: |
| References | |
| TAC 11945, DPR-72-A-054 NUDOCS 8207260007 | |
| Download: ML20058B412 (7) | |
Text
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'o UNITED STATES E'*
NUCLEAR REGULATORY COMMISSION e
D E
W ASHINGT ON. D. C. 20665 1.
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FLORIDA POWER CORPORATION CITY OF ALACHUA CITY OF BUSHNELL CITY OF GAINESVILLE CITY OF KISSIMMEE CITY OF LEESBURG CITYa0F NEW SMYRNA BEACH AND UTILITIES COMMISSION, CITY OF NEW SMYRNA BEACH CITY OF OCALA ORLANDO UTILITIES COMMISSION AND CITY OF ORLANDO SEBRING UTILITIES COMMISSION SEMIN0LE ELECTRIC COOPERATIVE, INC.
CITY OF TALLAHASSEE DOCKET NO. 50-302 CRYSTAL RIVER UNIT 3 NUCLEAR GENERATING PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 54 License No. DPR-72 1.
The Nuclear Regulatory Commission (the Comission) has found that:
A.
The application for amendment by Florida Power Corporation, et al (the licensees) dated March 21, 1979, as supplemented November 27, 1979, and February 15, 1980, complies with the standards and require-ments 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 Comission; 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 Comission'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 Comission's regulations and all applicable requirements have been satisfied.
8207260007 820714 PDR ADOCK 05000302 P
. l 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-72 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 54, are hereby incorporated in the license.
Florida Power Corporation 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 L
ohn
. Stolz, Chief Operating Reactors Branch #4 sion of Licensing
Attachment:
Changes to the Technical i
Specifications l
)
Date of issuance: July 14, 1982 l
ATTACHMENT TO LICENSE AMENDMENT NO. 54 FACILITY OPERATING LICENSE NO. DPR-72 DOCKET NO. 50-302 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment i
numb 6r and contain vertical lines indicating the area of change. The corresponding overleaf pages are also provided to maintain document completeness.
3/47-2 B3/4 7-1 a
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3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE 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:
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 Nuclear Overpower 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 />. The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.7.1.1 No additional Surveillance Requirements other than these required by Specification 4.0.5, are applicable for the main steam line code safety valves of Table 4.7-1.
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I CRYSTAL RIVER - UNIT 3 3/4 7-1
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k TABLE 3.7-1 g
MAXIMUM ALLOWABLE NUCLEAR OVERPOWER TRIP SETPOINT WITH IN0PERABLE.
g STEAM LINE SAFETY VALVES U
r-Maximum Allowable Nuclear
=
g Maximum Number of Inoperable Safety Overpower Trip Setpoint Valves on Any Steam Generator (Percent of RATED THERMAL POWER) e C5 1
96.9 m
82.4 2
3 67.9 N
8e
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3 /4.7 PLANT SYSTEMS BASES 3/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAFETY VALVES The OPERABILITY cf the main steam line code safety valves ensures that the secondary system pressure will be limited to within its design pressure of 1050 psig during the most severe anticipated system,operationa] transient.
The maximum relieving capacity is associated with a turbine trip.from 100%.
RATED THERMAL POWER coincident with an assumed loss of condenser heat sink (l.e., no steam bypa,ss to the condenser).
The specified valve lift settings and relieving capacities are in accordance with the requirements of Sect. ion III of the ASME Boiler and Pressure Vessel Code, 1971 Edition. The total relieving capacity for all valves on all of the steam lines is 13,007,774 lbs/hr which is 118.3 percent of the total secondary steam flow of 11.0 x 106 lbs/hr at 100% RATED THERMAL POWER.
STARTUP and/or POWER OPERATION is allowable with safet'y valves inoperable within the limitations of the ACTION requirements on the basis of the reduc-tion in secondary system steam flow and THERMAL POWER required by the re-duced reactor trip settings of the Nuclear Overpower channels. The reactor trip setpoint reductions are derived on the following bases:
SP =
X - AY x 105.5 X 1 --
.here: SP = reduced Nuclear Overpower Trip Setpoint in per.
w cent of Rated Thermal Power X = total actual relieving capacity of each steam gene,rator in lbs/ hour (6,503,887 lbs/ hour)
A = maximum number of inoperable safety valves per steam generator a
Y = maximum relieving capacity of each of the larger capacity safety valves in lbs/ hour (845,759 lbs/ hour)
IX = total required relieving capacity of each steam generator for 112% Rated Thermal Power in lbs/ hour (6,160,000 lbs/ hour) 105.5 = Nuclear Overpower Trip Setpoint specified in Table 2.2.1 CRYSTAL RIVER - UNIT 3 8 3/4 7-1 Amendment No.54 1
PLANT SYSTEMS BASES 3/4.7.1.2 EMERGENCY FEEDWATER SYSTEMS The OPERABILITY of the eme.rgency feedwater systems ensures that the Reactor Coolant System can be cooled down to less 2han 280*F from normal operating conditions in the event of a total loss of offsite ppwer.
The electric driven emergency feedwater pump is capable of delivering a totai feedwater flow of 740 gpm at a pressure of 1144 psig to theEach entrance of the steam generators.
pump is capable of delivering a total feedwater flow of 740 gpm at a This pressure of 1144 psig to the entrance of the steam gener t.o remove decay heat and reduce the Reactor Coolant System temperature to Removal System may be placed into operati.n 280 F where the Decay Heat less tha on.
3/4.7.1.3 CONDENSATE STORAGE TANK The OPERABILITY of the condensate storage tank with the minimum water volume ensures that sufficient water is available for coold the Reactor Coolant System to less than 280*F in the event of a total The minimum loss of offsite power or of the main feedwater system.
water volume is sufficient to maintain the RCS at HOT STANDBY for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> with steam discharge to atmosphere concurrent with loss The contained water volume limit includes' an allowance for water not usable because of tank discharge line location or other of offsite power.
physical characteristics.
3/4.7.1.4 ACTIVITY The limitations on secondary system specific activity ensure that the resultant offsite radiation' dose will be limited to a small frac This of 10 CFR Part 100 limits in the event of a steam line rupt These tube leak in the st'eam generator of the affected steam line.
values are consistent with the assumptions used in the safety analyses.
3/4.7.1.5 MAIN STEAM LINE ISOLATION VALVES
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The OPERABILITY of the main steam line isolation valves ensures that no more than one steam generator will blowdown in the event of a This restriction is required to 1) minimize the steam line rupture.
B 3/4 7-2 CRYSTAL RIVER - UNIT 3