ML20214C692

Application for Amend to License NPF-35,changing Tech Specs to Extend 18-month Surveillances Associated W/Esf Until First Refueling Outage Scheduled for Aug 1986.Surveillances Currently Due by 860428.Fee Paid

ML20214C692
Person / Time
Site:	Catawba
Issue date:	02/12/1986
From:	Tucker H DUKE POWER CO.
To:	Harold Denton, Youngblood B Office of Nuclear Reactor Regulation
Shared Package
ML20214C695	List: ML20214C692 ML20214C697
References
NUDOCS 8602210235
Download: ML20214C692 (7)

Text

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DUKE POWER GOMPANY P.O. Isox 33189 CHAltLorrE, N.C. 28242 HAL II. TUCKER

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..m.. . .cm. trasU m su February 12, 1986 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Attention

Mr. B. J. Youngblood, Project Director PWR Project Directorate No. 4 Re: Catawba Nuclear Station Docket No. 50-413 l Technical Specification Amendment

Dear Mr. Denton:

This letter contains proposed amendments to the Technical Specification for Facility Operating License No. NPF-35 for Catawba Unit 1. The attachment.contains the proposed changes and a i

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 amendments do 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.

Very truly yours, l 4d /$

Hal B. Tucker fi RWO sib 6 e od W Attachment ' o 160pf[4 8602210235 860212 PDR ADOCK 05000413 P PDR

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Mr. Harold R. Denton, Director February 12, 1986 Page Two cc: Dr. J. Nelson Grace, Regional Administration U.-S. Nucelar Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 Mr. P. H. Skinner NRC Resident Inspector Catawba Nuclear Station Mr. Heyward Shealy, Chief Bureau of Radiological Health South Carolina Department of Health &

Environmental Control 2600 Bull Street Columbia, South Carolina 29201 INPO Records Center Suite 1500 1100 Circle 75 Parkway Atlanta, Georgia 30339 American Nuclear Insurers c/o Dottie Sharman, ANI Library The Exchange, suite.245 270 Farmington Avenue Farmington, CT 06032 M&M Nuclear Consultants 1221. Avenue of the Americas New York, New York 10020 L

l Mr. Harold R. Denton, Director February 12, 1986 Page Four 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 Commission this revision to the Catawba Nuclear Station Techncial Specifications, Appendix A to License No. NPF-35; and that all statements the and.

best of his matters set forth therein are true and correct to knowledge- .

C a4 / .

Hal B. Tucker, Vice President Subscribed and sworn to before me this 12th day of February, 1986.

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Notary Public My Commission Expires:

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DISCUSSION OF AMENDMENT REQUEST It is requested that the 18 month surveillances associated with Engineered Safety Features (ESF) be extended until the Catawba Unit 1 first refueling outage. The surveillances were last conducted on June 13, 1984. Given the grace period allowed by Technical '

Specification 4.0.2, the 18 month surveillances are due to be performed no later than April 28, 1986. Unit 1 is currently scheduled to enter the first refueling outage in the latter part of August, 1986. Therefore, an extension for the time to conduct the surveillance is necessary.

The surveillances requiring an extension are those surveillances which can only be conducted while the Unit is in COLD SHUTDOWN or REFUELING. The majority of the surveillances require initiation of a Safety Injection, Phase "A" Isolation, Phase "B" Isolation or Loss of Offsite Power signal along with verification of proper valve movement and pump performance. Injection into the core is also a part of the surveillance requirements.

The above noted signals and subsequent ESF actuations are conducted as an integrated test to verify overall ESF capability. Other periodic surveillances required by the Technical Specification and the IWV/IWP program will continue to be performed as required.

During on-line operation of the reactor, all of the Engineered Safety Features analog and logic circuitry will be fully tested.

In addition, essentially.all of the Engineered Safety Features final actuators will be fully tested. The remaining few control circuits whose operation is not compatible with continued on-line plant operation will be checked by means of continuity testing through the final actuator or the final safety device before signal isolation.

During normal operation, the operability of testable final actuation devices of the Engineered Safety Features final actuators will be fully tested. The remaining few control circuits whose operation is not compatible with continued on-line plant operation will be checked by means of continuity testing through the final actuator or the final safety device before signal isolation.

During normal operation, the operability of testable final actuation devices of the Engineered Safety Features Systems will be tested by manual initiation from the control room.

During reactor operation the basAs for Engineered Safety Features Actuation Systems acceptability is the successful completion of the overlapping tests performed on the initiating system and the Engineered Safety Features Actuation System (see PSAR Figure 7.3.2-1). Checks of process indications verify operability of the sensors. Analog checks and tests verify the operability of the analog circuitry from the input of these circuits through to and including the logic input relays except for the input relays

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associa'ted with the containment spray function which are tested .

during-the solid state logic testing. Solid state logic testing  !

also checks the digital signal path from and including logic input relay contacts through the logic matrices and master relays add performs continuity tests on the coils of the output slave relays; final actuator testing operates the output slave relays and verifies operability of those devices which require safeguards actuation and which can be tested without causing plant upset. a continuity check is performed on the actuators of the untestable devices. Operation of the final devices is confirmed by control board indication and visual observation that the appropriate pump breakers close and automatic valves have completed their travel.

I The basis for acceptability for the Engineered Safety Features interlocks is control board indication of proper receipt of the signal upon introducing the required input at the appropriate setpoint.

The procedures described provide capability for checking completely from the process signal to the logic cabinets and from there to the individual pump and fan circuit breakers or starters, valve contactors, pilot solenoid valves, etc. including all field cabling actually used in the circuitry called upon to operate for an accident condition. For those devices whose operation could adversely affect plant or equipment operation, the same procedure provides for checking from the process signal to the logic rack. '

To check the final actuation device a continuity test of the individual control circuits is performed.

The procedures require testing at various locations.

(1) Analog testing and verification of bistable setpoint are accomplished at process analog racks. Verification of bistable relay operation is done at the main control room status lights.

(2) Logic testing through operation of the master relays and low voltage application to slave relays is done at the logic rack test panel.

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(3) Testing of pumps, fans, ano va2.cs is done at a test panel located in the vicinity of the logic racks in combination with the control room operator.

(4) Continuity testing for those circuits that can not be operated is done at the same test panel mentioned in 3 above. 3,

' Additional assurance of system and component operability has been demonstrated through several inadvertent ESF actuations. Licensoo Event Reports (LER's) 413/85-51, 85-34, 85-27, 85-07, 84-31 and

~ 84-28 described incidents of diesel generator start signals being generated. In all cases the appropriate diesel auto-started upon

F receipt of the ESP signal. During two of these auto-starts the diesels automatically started, loaded and operated until normal system power was restored. In the other incidents load shedding was not necessary due to the short duration of the transient.

LER's 413/85-12 and 85-09 describe two incidents of Safety Injection initiation. In each case the ESF components actuated as designed.

Therefore, it can be concluded that postponement of the indicated surveillances by approximately four months will not result in an unacceptable risk to plant operation or the health and safety of the public. The conduct of the other required surveillances coupled with the short duration of the exemption request plus the inadvertent actuations which demonstrated that the systems and components function properly ensure that ESF performance will not degrade.

In order to conduct the specified surveillances, Unit I would have to be placed in Mode 5 (COLD SHUTDOWN) and then into Mode 6 (REFUELING). Cooling down, testing and subsequent heatup would require at least two weeks. Granting this amendment request would allow the unit to stay on line and avoid a thermal cycle on the Reactor Coolant System.

ANALYSIS OF SIGNIFICANT HAZARDS CONSIDERATION 10 CPR 50.92 (c) states that "a proposed amendment... involves no significant hazards considerations, if operation of the facility in accordance with the proposed amendment would not:

(1) Involve a significant increase in.the probabil'ity 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."

This amendment request would not significantly increhse the probability or consequences of an accident previously evaluated.

The probability of previously evaluated accidents is not affected since the proposed changes will only affect ESF components thus normal plant operation will not be affected. The consequences of a previously evaluated accident will not he significantly increased since the majority of the affected ESF components and circuitry will be tested as required by other applicable Technical Specification Surveillance Requirements and IWV/IWP requirements.

This amendment request would only affect the ESP actuation of certain components by extending the required surveillance by four or five months. This increase is not viewed as significant especially when coupled with the other surveillances conducted on the individual components and circuitry.

This proposed amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated since the design and operation of the plant will not be a f fec ted.

The proposed amendment would not cause a significant reduction in a margin of safety. The extension of time in which to do the required surveillance would be on the order of four or five additional months beyond that allowed by the Technical Specifications. Coupled with the fact that individual component and circuit tests are conducted on a regular basis as provided in other Technical Specifications and the Pump and Valve Inservice Testing Program, there is no significant reduction in a margin of safety. An increase in safety is gained by the avoidance of an extra cooldown and heatup cycle of the Reactor Coolant System.

For the reasons stated above, it is concluded that the proposed amendment does not involve significant hazards considerations.

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