ML20214J034

Application for Amends to Licenses NPF-35 & NPF-52,deleting Requirement 4.6.1.9.4 for Containment Air Release & Addition Sys Valves & Adding Four Containment Penetration Conductor Overcurrent Protector Devices to Table 3.8-1.Fee Paid

ML20214J034
Person / Time
Site:	Catawba
Issue date:	11/17/1986
From:	Tucker H DUKE POWER CO.
To:	Harold Denton, Youngblood B Office of Nuclear Reactor Regulation
Shared Package
ML20214J037	List: ML20214J034 ML20214J087
References
NUDOCS 8612010126
Download: ML20214J034 (7)

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e DUKE POWER GOMPANY P.O. BOX 33189 CHARLOTTE, N.C. 28242 HALU. TUCKER TELEPHONE WK5 PBretDENT (704) 373 4531 attLeam PRODE?CTION November 17, 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 Prcject Directorate No. 4 RE: Catawba Nuclear Station Docket Nos. 50-413 and 50-414 Technical Specification Amendment

Dear Mr. Denton:

This letter contains proposed amendments to the Technical Specifications for Facility Operating License Nos. NPF-35 and NPF-52 for Catawba Units 1 and 2. The attachments contain the proposed changes 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 amendments do not involve significant hazards considerations.

1) Deletion of Surveillance Requirement 4.6.1.9.4 for the VQ valves;

2) Addition of 4 Containment Penetration Conductor Overcurrent Protective Devices to Table 3.8-1 as a result of two recently completed NSMs;

3) Changes to Table 4.4-5 for the Surveillance capsule withdrawal schedule.

This proposal constitutes an 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, ak Hal B. Tuckar RWO/32/sn 9 8612010126 861117 0 Ol Attachment PDR p

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. Mr. $arold R. Denton, Director November 17, 1986 Page Two xc: Dr. J. Nelson Grace, Regional Administrator U. S. Nuclear Regulatory Commission Region II 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 c/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, New York 10020 INPO Records Center Suite 1500 1100 Circle 75 Parkway Atlanta, Georgia 30339 Mr. P. H. Skinner NRC Resident Inspector Catawba Nuclear Station

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, Mr. harold R. Denton, Director November 17, 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 caid Company to sign and file with the Nuclear Regulatory Commission 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.

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'Ihl B. Tucker, Vice President Subscribed and sworn to before me this 17th day of November, 1986.

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E Attachment 1 Containment Air Release And Addition System Valves

Discussion of Amendment Request It is requested that Technical Specification 4.6.1.9.4 and references to this i specification, be deleted from the Catawba Technical Specifications. This surveillance calls for verifying measured leakage through Containment Air Release and Addition System (VQ) valves with resilient material seals to be less than 0.01 La at least once per 3 months. This surveillance is applicable to two valves per unit (VQ2A and VQ16A). The acceptance criteria for these valves is a total leakage of less than 931 SCCM or 465 SCCM per valve.

Valves VQ2A and VQ16A are 4" diaphragm valves which have an excellent history of i leak tightness. This characteristic is primarily due to the resilience of the rubber diaphragm. Table 1 (attached) lists the leakage history of the VQ valves and two similar valves. Valves VI-77B and VB-38B are the only other similar valves at Catawba that have required type C leak tests performed on them. No maintenance or repair has been performed on either IVQ2A or IVQ16A during this time period.

The measurement of the leakage rate of the VQ Valves is accomplished by pressurizing the test volume auch that pressure is applied between the isolation vaive and a block valve and measuring the flow rate of dry air required to maintain

- test pressure. The leakage measured will be the valve's leak rate. The observed leakage can be attributed mainly to leakage through the boundary (block) valves and l not through the VQ valves. The latest leak test on IVQ2A showed 0.0 SCCM leakage.

This is due to the boundary valve having been modified during the refueling outage.

No maintenance work had been performed on IVQ2A between the last two leak tests.

Valves VQ2A snd VQ16A are containment isolation valves which are required to close autematically upon receipt of a containment isolation signal in response to controls which are intended to effect containment isolation. 10 CFR 50 Appendix J Section II.H. describes Type C Tests as, " tests intended to measure containment isolation valve leakage rate". Additionally, 10 CFR 50 Appendix J Section III.C.3 states that, " Type C tests shall be performed during each reactor shutdown for refueling but in no case at intervals greater than 2 years". Technical Specification Table 3.6-1 identifies the Containment Air Release Penetration (M386) and the ' Containment Air Addition Penetration (M204) as Type "C". These valves are also identified in FSAR Table 6.2.4-2. Nowhere does 10 CFR 50 Appendix J state i that Type "C" leak rate tests on containment isolation valves with resilient trterial seals must be conducted at an increased frequency.

In order to adequately pressurize the test volume to perform the leak rate test, a containment entry must be made. Valve VQ2A must be tested from upper containment

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and valve VQ16A must be tested from lower containment. The testing of these valves at 100% power results in undue hardship due to the radiation exposure (approximately 3000 man-arem/ test) incurred by the testing personnel, Health Physics technician and Operations personnel, and also hardship due to the containment air temperature, whi:h is normally in excess of 100 degrees F.

Based on the fact that these valves exhibit a high degree of reliability and the ALARA concerns involved, extending the Surveillance from every three months to a surveillance that corresponds to Appendix J requirements is justified.

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ANALYSIS OF SIGNIFICANT HAZARDS CONSIDERATION 10 CFR 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 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) Int olve a significant reduction in a margin of safety". t This amendment request would not significantly increase the probability or consequences of an accident previously evaluated. The probability of previously evaluated accidents is not affected since the proposed changes will not affect the normal operation of the plant. The consequences of a previously evaluated accident will not be significantly increased since the valves have an excellent history of leak tightness (Table 1) .

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 affected.

The requested change would not involve a significant reduction in a margin of safety. Since these valves have exhibited a high degree of reliability coupled with the fact that the valves will be tested in accordance with 10 CFR 50, Appendix J, any reduction in a margin of safety would not be significant.

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TABLE 1 LEAKAGE HISTORY IVQ2A IVQ16A Test Date Leakage Test Date Leakage 11/01/86 0.0 SCCM 10/30/86 27 SCCM 08/24/86 23.8 SCCM 08/29/86 44.0 SCCM 06/13/86 6.4 SCCM 06/06/86 34.3 SCCM 03/07/86 225 SCCM* 03/12/86 54.8 SCCM 02/23/85 3.0 SCCM 02/24/85 31.5 SCCM IVB-83B IVI-77B, IVI-312A**

Test Date Leakage Test Date Leakage 09/29/86 1.76 SCCM 09/16/86 74.9 SCCM 07/01/86' 4.60 SCCM 11/11/84 60.0 SCCM 08/19/84 0.16 SCCM 1

• Higher measured leakage than expected was due to a leaking test volume block valve since no repairs were made to IVQ-2A prior to the test performed on 06/13/86.

• • Valves IVI-77B and IVI-312A are leak tested together. IVI-77B is a 2" diaphragm valve and IVI-312A is a 2" globe valve.

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