ML19308E255
| ML19308E255 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 04/03/1978 |
| From: | Reid R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19308E253 | List: |
| References | |
| TAC-07763, TAC-7763, NUDOCS 8003240802 | |
| Download: ML19308E255 (10) | |
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NUCLEAR REGULATORY COMMISSION o
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,a WASHINGTON. D. C. 20665
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FLORIDA POWER CORPORATION
_ CITY OF ALACHUA CITY OF BUSHNELL CITY OF GAINESVILLE CITY OF KISSIMMEE CITY OF LEESBURG CITY OF NEW SMYRNA BEACH AND UTILITIES COMMISSION, CITY OF NEW S CITY OF OCALA ORLANDO UTILITIES COMMISSION AND CITY OF ORLANDO SEBRING UTILITIES C0fwISSION SEMIN0LE ELECTRIC COOPERATIVE, INC.
CITY OF TALLAHASSEE DOCKET NO. 50-302 CRYSTAL RIVER UNIT 3 NUCLEAR GENERATING PLANT AMENDMENT TO FACILITY OPERATING LICENSE i
Amendment No.14 License No. OPR-72 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A-The applications for amendment by Florida Power Corporation, et al (the licensees) dated November 8 and December 14, 1977, comply 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; i
B.
The facility will operate in confomity 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 amencment 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 amencment will not be inimical to the comon defense and secu:ity or to the health and safety of the public; and E.
The issuance of this a:nendment is in accordance with 10 CFR Part 51 of the comiss.cn's regulations and all applicable requirements have been satisfied.
8003240 80 2
1 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.14, are hereby incorporated in the license.
Florida Power i
Corporation shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance.
l FOR THE NUCLEAR REGULAT COMMISSION
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4 Robert W. Reid, Chief Operating Reactors Branch #4 Division of Operating Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance: April 3,1978 1
l 1
ATTACHMENT TO LICENSE AMENDMENT NO.14-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 vertical lines indicating the area of change.
The-corresponding overleaf pages are also provided to maintain document completeness.
pages 3/4 4-32 B 3/4 4-13 6-3 6-5
s REACTOR COOLANT SYSTEM 3.4.10 STRUCTURAL INTEGRITY ASME CODE CLASS 1, 2 and 3 COMPONENTS LIMITING CONDITION FOR OPERATION 3.4.10.1 The structural integrity of ASME Code Class 1, 2 and 3 components shall be maintained in accordance with Specification 4.4.10.1.
APPLICABILITY: All MODES.
ACTION:
With the structural integrity of any ASME Code Class 1 component (s) a.
not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature more than 50*F above the minimum temperature required by NDT considerations.
b.
With the structural integrity of any ASME Code Class 2 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature above 200*F.
With the structural integrity of any ASME Code Class 3 component (s) c.
not conforming to the above requirements, restore the structural integrity of the component (s) to within its limit or isolate the affected component (s) from service.
d.
The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.4.10.1 In addition to the requirements of Specification 4.0.5:
a.
The reactor coolant pump flywheels shall be inspected per the recommendations of Regulatory Position C.4.b. of Regulatory Guide 1.14, Revision 1, August 1975.
CRYSTAL RIVER - UNIT 3 3/4 4-31
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REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) b.
Each internals vent valve shall be demonstrated OPERABLE at least once per 18 months
- during shutdown, by:
1 1.
Verifying through visual inspection that the valve body and valve disc exhibit no abnormal degradation, 2.
Verifying the valve is not stuck in an open position, and 3.
Verifying through manual actuation that the valve is fully open with a force of < 425 lbs (applied vertically upward).
N
- TF.a first periodic surveillance of the internals vent valves may be per-formed up to 24 months following the original surveillance (plus 25%
per 4.0.2) but no later than the first refueling outage.
CRYSTAL RIVER - UNIT 3 3/4 4-32 knendment No. 14
h REACTOR COOLANT SYSTEM (Continued)
BASES 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.
The pressure and temperature limits shown on Figures 3.4-2 and 3.4-3 for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR 50.
The number of reactor vessel irradiation survefilance specimens and the frequencies for removing and testing these specimens are provided in Table 4.4-3 to assure compliance with the requirements of Appendix H to 10 CFR Part 50.
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 i
Vessel Code.
The internals vent valves are provided to relieve the pressure j
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 OPERASILITY, 2) ensure that the valves are not stuck open during normal operation, and 3) demonstrate that the valves are fully open at the forces assumed in the safety analysis.
CRYSTAL RIVER - UNIT 3 B 3/4 4-13 Amendment No. 14
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CRYSTAL RIVER - UNIT 3 6-3 Amendment No. E, %. 14 l
p TABLE 6.2-1 MINIMUM SHIFT CREW COMPOSITION #
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APPLICABLE MODES CATEGORY 1, 2, 3 & 4 5&6 SOL 1
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OL 2
1 Non-Licensed 3
1
- Does not include the licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling Individual supervising CORE ALTERATIONS after the initial fuel loading.
- Shift crew composition may be less that the minumum require-ment for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accomodate unexpected absence of op duty shift crew members
)
provided immediate action is taken to restore the shift crew composition to within the minimum requirements of Table 6.2-1.
i 1
CRYSTAL RIVER'- UNIT 3 6-4 V
- k ADMINISTRATIVE CONTROLS 6.3 FACILITY STAFF QUALIFICATIONS 6.3.1 Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except for the Chemistry and Radiation Protection Engineer who shall meet or exceed l
the qualifications of Regulatory Guide 1.8, September 1975.
6.4 TRAINING 6.4.1 A retraining and replacement training program for the facility staff shall be maintained under the direction of the Nuclear Plant Manager i
and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and Appendix "A" of 10 CFR Part 55.
6.4.2 A training program for the Fire Brigade shall be maintained under the direction of the Nuclear Plant Manager and shall meet or exceed the requirements of Section 27 of the NFPA Code-1976, except for Fire Brigade training sessions which shall be held at least quarterly.
6.5 REVIEW AND AUDIT 6.5.1 PLANT REVIEW COMMITTEE (PRC)
FUNCTION 6.5.1.1 The Plant Review Committee shall function to advise the Nuclear Plant Manager on all matters related to nuclear safety.
COMPOSITION 6.5.1.2 The Plant Review Committee shall be composed of the:
Chairman:
Technical Services Superintendent Member:
Operations Superintendent Member:
Technical Support Engineer Member:
Maintenance Superintendent Member:
Chemistry and Radiation Protection Engineer Member:
At large (Designated by Chairman)
ALTERNATES 6.5.1.3 All alternate members shall be appointed in writing by the PRC Chairman to serve on a temporary basis; no more than two alternates shall participate as voting members in PRC activities at any one time.
MEETING FREQUENCY 6.5.1.4 The PRC shall meet at least once per calendar month and as convened by the PRC Chariman or his designated alternate.
CRYSTAL RIVER - UNIT 3 6-5 Amendment No. 5, }6,14
b ADMINISTRATIVE CONTROLS QUORUM
. 5.1. 5 A quorum of the PRC shall consist of the Chariman or his 6esignated alternate and four members including alternates.
d RESPONSIBILITIES 6.5.1.6 The Plant Review Committee shall be responsbile for:
Review of 1) all procedures required by Specification 6.8 and a.
changes thereto, 2) any other proposed procedures or changes thereto as determined by the Nuclear Plant Manager to affect l
nuclear safety.
Review of all proposed tests and experiments that affect b.
nuclear Safety, Review of all proposed changes to the Appendix "A" Technical c.
Specifications.
Review of all proposed changes or modifications to plant d.
systems or equipment that affect nuclear safety.
Investigation of all violations of the Technical Specifications including the preparation and forwarding of reports covering e.
evaluation and recommendations to prevent recurrence to the Director-Power Production and to the Chairman of the Nuclear l
General Review Committee.
Review of events requiring 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> written notif$ cation to the f.
Commission.
Review of facility operations to detect potential nuclear g.
safety hazards.
Performance of special reviews, investigations or analyses and h.
reports thereon as requested by the Chairman of the Nuclear General Review Committee.
Review of the Plant Security Plan and implementing procedures and shall submit recommended changes to the Chairman of the i.
i Nuclear General Review Committee.
Review of the Emergency Plan and implementing procedures and shall submit recommended changes to the Chairman of the Nuclear j.
1 General Review Committee.
l Amendment No. 5 CRYSTAL RIVER - UNIT 3 6-6
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