ML19340A377
| ML19340A377 | |
| Person / Time | |
|---|---|
| Site: | Oconee, Crystal River  |
| Issue date: | 07/24/1978 |
| From: | Reid R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19340A378 | List: |
| References | |
| NUDOCS 8003250687 | |
| Download: ML19340A377 (8) | |
Text
_
47 o
UNITE D STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555
'****/
FLORIDA POWER CORPORATION
_ CITY OF ALACHUA CITY OF BUSHNELL CITY OF GAINESVILLE CITY OF KISSItHEE CITY OF LEESBURG CITY OF NEW SMYRNA BEACH AND UTILITIES COMMISSION, CITY OF NEW SMYRNA BEACH CITY OF OCALA ORLANDO UTILITIES COMMISSION AND CITY OF ORLANDO SEBRING Ui!LITIES C0ftISSION SEMIN0LE ELECTRIC COOPERATIVE, INC.
CITY OF TALLAHASSEE
^
DOCKET NO. 50-302 i
CRYSTAL RIVER UNIT 3 NUCLEAR GENERATING PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.15 License No. DPR-72 y
1.
The Nuclear Regulatory Comission (the Comission) has found that-A. The applications for amendment by Florida Power Corporation, et al (the licensees) dated May 30 and June 28, 1978, comply with i
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 confonnity with the application, I
the provisions of the Act, and the rules and regulations of the Comission; 5
C.
There is reasonable assurance (1) that the activities authorized
}
by this amendment can be conducted without endangering the health i
and safety of the public, and (ii) that such activities will be j
conducted in compliance with the Comission's regulations; 1
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;
?4 and i
E.
The issuance of this arendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements i
have been satisfied.
- h1 i
SOO3250 5
i 0.
2.
Accordingly, Facility Operating License No. DPR-72 is hereby amended as indicated below and by changes to the Technical Specifications as indicated in the attachrent to this license amendment:
~ Add a new' paragraph 2.B.(7) to read as follows:
2.B.(7)
Florida Power Company, pursuant to the Act and 10 CFR Parts 30 and 70, to receive and possess, but not separate, that by-product and special nuclear materials associated with four (4) fuel assemblies (B&W Identifi-cation Numbers 1 A-01, 04, 05 and 36 which were previously irradiated in the Oconee Nuclear Station, Unit No.1) acquired by Florida Power Corporation from Duke Power Company for use as reactor fuel in the facility.
Revise Paragraph 2.C.(2) to read as follows:
2.C.(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.15, 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 k
Robert W. Reid, Chief Operating Reactors Branch #4 Division of Operating Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance: July 24, 1978
=
9 e
i ATTACHMENT TO LICENSE AMENDMENT NO.15 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 number and contain vertiell lines indicating the area of change. The corresponding overleaf pages are also provided to maintain document completeness.
I Pages 3/4 4-29 B 3/4 4-12 B 3/4 4-13 O
'l
!il Fl
~
i BASES I
recalculated when the ART detemined from the surveillance capsule is different from the calculbd ART for the equivalent capsule radiation NDT exposure.
The closure head region is significantly stressed at relatively low temperatures (due to mechanical loads resulting from bolt pre-lead). This region largely controls the pressure-temperature limitations r' the first several service periods. The outlet nozzles of the reactor wssel also affect the pressure-temperature limit curves of the first several service periods. This is due to the high local stresses at the inside corner of the nozzle which can be two to three times the membrane stresses of the shell. After the first several years of neutron radiation exposure, the RT temperature of the beltline region materials will be high i
enough N that the beltline region of the reactor vessel will start to N
control the pressure-temperature limitations of the reactor coolant 4
pressure boundary. For the service period for which the limit curves are established, the maximum allowable pressure as a function of fluid temperature is obtained through a point-by-point comparison of the limits imposed by the closure head region, outlet nozzles, and beltline region.
The maximum allowable pressure is taken to be the lower pressure of the three calculated pressures. The calculated pressure temperature limit t
curves are then adjusted by 25 psi and 10*F for possible errors in the j
pressure and temperature sensing instruments. The pressure limit is also adjusted for the pressure differential between the point of system pressure measurement and the limiting component for all operating reactor coolant pump combinations. The limit curves were prepared based upon j
the most limiting adjusted reference temperature of all the beltline region materials at the end of the fifth effective full power year. The fifth effective full power year was selected because the second surveil-3 lance capsule will be withdrawn at the end of the fifth cycle. The time-difference between the fifth cycle and fifth effective full power year provides adequate time for establishing the operating pressure and temperature limitations for the period of operation after the fifth effective full power year.
. The actual shift in RT of the bel',line region material will be l
established periodically dubg operation by removing and evaluating, in f
accordance with Appendix H to 10 CFR 50, reactor vessel material irradia-J tion surveillance specimens installed near the inside wall of the reactor f
vessel in the core area.
Since the neutron spectra at the irradiation samples and vessel inside the radius are essentially identical, the measured transition shift for a sample can be applied j
I i
CRYSTAL RIVER - UNIT 3 B 3/4 4-11 j
e l
I
1 REACTOR COOLANT SYSTEM (Continued)
BASES ll l
All pressure-temperature limit curves are applicable up to the fifth effective full power year. The protection against non-ductile failure is assured by maintaining the coolant pressure below the upper limits of Figures 3.4-2, and 3.4-3, and 3.4-4.
l The pressure and temperature limits shown on Figures 3.4-2 and 3.4-4 i
for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements 3
of Appendix G to 10 CFR 50.
The number of reactor vessel irradiation surveillance specimens and l
the frequencies for removing and testing these specimens are provided in Table 4.4-5 to assure compliance with the requirements of Appendix H to 10 CFR Part 50.
j The limitations imposed on pressurizer heatup and cooldown 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 ASME Code requirements.
3/4.4.10 STRUCTURAL INTEGRITY The inspection programs for ASME Code Class 1, 2 and 3 components, except steam generator tubes, ensure that the structural integrity of these components will be maintained at an acceptable level throughout the life of the plant. To the extent applicable, the inspection program for these components is in compliance with Section XI of the ASME Boiler and Pressure
" Vessel Code.
The internals vent valves are provided to relieve the pressure generated by steaming in the core following a LOCA so that the core remains sufficiently covered.
Inspection and manual actuation of the internals vent valves 1) ensure OPERABILITY, 2) ensure that the valves are not stuck open during nornal operation, and 3) demonstrate that the valves are fully open at the forces assumed in the safety analysis.
d
[
5 9
l C'
1 3
i CRYSTAL RIVER - UNIT 3 8 3/4 4-13 Amendment No. )(,15
)
1
_1
3 TABLE 4.4-5
- o REACTOR VESSEL MATERIAL IRRADIATION SURVEILLANCE SCHEDULE P
Capsule Installation Removal 9
h A
At 270110 EFPD of First Cycle Standby z
]
C At 270110 EFPD of First Cycle End of Ninth Cycle E
At 270110 EFPD of First Cycle Standby 8
Initial Fuel Load At 270110 EFPD of First Cycle D
Initial Fuel Load End of Fifth Cycle R
F At 270110 EFPD of First Cycle End of Fifth Cycle l
e
- g j
a i
n E.
i
,