ML20236J009
| ML20236J009 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 07/27/1987 |
| From: | Tucker H DUKE POWER CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| Shared Package | |
| ML20236J011 | List: |
| References | |
| TAC-65861, TAC-65862, NUDOCS 8708060018 | |
| Download: ML20236J009 (5) | |
Text
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Duxe POWER GOMPANY P.O. nox 33189 CHARLOFTE, N.C. 28242 HALH. TUCKER res.zenown (704) 373-4831
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wuotnas rococonow July 27, 1987 i
U. S. Nuclear Regulatory Commission At.tention:- Document Control Desk Washington, D. C. 20555 Re: Catawba Nuclear Station, Units 1 and 2 Docket Nos. 50-413 and 50-414 Technical Specification Amendment Turbine Trip On Stop Valve EH Pressure - Low
Dear Sir:
This letter contains a proposed amendment to the Technical Specifications for Facility Operatinr, License Nos. NPF-35 and NPF-52 for Catawba Units 1 and 2.
The request involves changes to the Unit 1 Technical Specifications to reflect a modification which will be performed during the upcoming Unit 1 refueling outage.
This modification will move the Turbine trip generation signal from the main control valves to the main stop valves. The attachment contains the proposed change and a discussion of the justification and safety analysis. The analysis is included pursuant to 10 CFR 50.91 and it has been concluded that the proposed amendment does l
not involve significant hazards considerations.
This request involves one amendment' request to Catawba's Technical Specifications.
Accordingly, pursuant to 10 CFR 170.21 a check for $150.00 is enclosed.
Pursuant to 10 CFR 50.91 (b) (1) the appropriate South Carolina State Official is being provided a copy of this amendment request.
1 Very truly yours, j
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Fal B. Tucker
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Attachment 00 6,3 I
8708060018 870727 PDR ADOCK 05000413 PDR p
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- If. S._.Nuclocr Ragulatory Commission-
. July 27,J1987
- Page Two xc:
Dr. J. Nelson Grace, Regional Administrator U. S. Nuclear Regulatory Commission
'RegfonlII 101 Marietta Street, NW, Suite 2900-Atlanta,' Georgia 30323
-Mr. Heyward Shealy, Chief Bureau of Radiological Health South Carolina Department of Health &
Environmental Control 2600 Bull Street Columbia, South Carolina _29201 American Nuclear Insurers e/o Dottie Sherman, ANI Library The Exchange,-Suite 245
.270 Farmington Avenue Farmington, CT 06032 M&M Nuclear Consultants 1221 Avenue of the Americas New York, NewLYork 10020 INPO Records Center Suite 1500 1100 circle 75 Parkway Atlanta, Georgia 30339 Mr. P. K. Van Doorn 3
I NRC Resident Inspector Catawba Nuclear Station l
i 1
i 1
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-Mr. Harold R. Denton, Director s.
July 27, 1987 Page Three HAL B. TUCKER, being duly sworn, states that he is Vice President of Duke Power Company; that he is authorized on the part of said Company to sign and file with the Nuclear Regulatory Consnission this revision to the Catawba Nuclear Station Technical Specifications, Appendix A to License Nos. NPF-35 and NPF-52; and that all
. statements and matters set forth therein are true and correct to the best of his knowledge.
(^
Pal B. Tucker, Vice ' resident f
Subscribed and sworn to before me this 27th day of July, 1987.
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1 DISCUSSION AND ANALYSIS OF NO SIGNIFICANT HAZARDS CONSIDERATIONS a
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The proposed changes to Technical Specification 2.2.1-Table 2.2-1, 3/4.3.1-Tables I
3.3-l' 3.3-2 and 4.3-1 reflect a modification to the Unit 1 Turbine Trip circuitry which is to be implemented during the upcoming Unit 1 refueling outage.
The anticipatory reactor trip on turbine trip which is currently installed is actuated by a low pressure signal from two-out-of-four control valve electro-hydraulic fluid pressure switches, or by valve closed signals from four-out-of-four turbine steam stop valve limit switches. A turbine trip initiates a j
direct reactor trip only when reactor power is above the P-9 setpoint. This trip provides conservatism and protection beyond that required to assure the health and safety of the public, and is included as good engineering practice and prudent design. No credit is taken for this trip in.any of the safety analyses of Chapter 15.
On April 19, 1985, with Catawba. Unit 1 at 100% percent power, a full load rejection was initiated.
During ?.his transient, the General Electric Mark II Turbine Electrohydraulic Control (EWC) System sensed a 40% steam power versus generator output power mismatch.
This mismatch resulted in the actuation of the power load unbalance circuit. The actuation of this circuit results in the fast closure of the turbine control and intercept valves through actuation of their respective fast acting solenoid valves.
l The Westinghouse NSSS initiates a reactor trip as a result of the turbine trip j
signals. The present General Electric design utilizes the turbine stop valve
" CLOSED" limit switches (four-out-of-four logic) or the decay of the turbine centrol valve Emergency Trip System (ETS) pressure as sensed at the control valves (two-out-of-four logic).
That is, the pressure switches that provide the reactor i
i trip inputs sense ETS pressure, and ETS pressure is decreased on both a Turbine Trip and a Power-Load Unbalance. With actuation of the power load unbalance circuit an unnecessary reactor trip occurred as a result of an invalid turbine trip indication.
]
The modification to be implemented will move the four pressure switches and associated control packs from the main control valves to the main stop valves. With this modification in place, a main Turbine Power-Load Unbalance signal will still result in the closure of the control and intercept valves but an unnecessary reactor trip signal will not be generated since the main stop valves will not produce the trip signal.
This amendment is for a Unit 1 m7dification, Unit 2 is affected only administrative 1y L
since the Technical Specifications are combined.
This modification has already been installed on Unit 2, as indicated by the current Technical Specifications.
The Unit I refueling ia scheduled to begin on October 2, 1987 and end by December 1, 1987. Since this specification is applicable in Mode 1 only, this amendment is needed no later than December 1, 1987.
10 CFR 50.92 states that a proposed amendment involves no significant hazards considerations if operation in accordance with the proposed amendment would not:
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DISCUSSION AND ANALYSIS OF NO SIGNIFICANT HAZARDS CONSIDERATIONS (CONT'D)
(1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.
The proposed amendment does not involve an increase in the probability or consequences of any previously evaluated accident. The probability of an accident is not increased because the anticipatory reactor trip on turbine trip signal will continue to be functional. The proposed modification will eliminate unnecessary reactor trips and challenges to the Reactor Protection System.
This modification and Technical Specification amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated since these changes do not affect the design or operation of the plant.
The Reactor Trip on Turbine trip signal will still be generated when it is needed.
These changes do not involve a significant reduction in a margin of scfety since the Reactor Trip on Turbine Trip signal will still be generated. The overall affect of this change will be an increase in the margin of safety since unnecessary Reactor trips and challenges to the Reactor Protection System will be eliminated.
For the above reasons, Duke Power concludes that the proposed amendment does not involve significant hazards considerations.
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