ML20236X899
| ML20236X899 | |
| Person / Time | |
|---|---|
| Site: | Mcguire, Catawba, McGuire, 05000000 |
| Issue date: | 12/02/1987 |
| From: | Tucker H DUKE POWER CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| TAC-61680, TAC-61681, NUDOCS 8712100346 | |
| Download: ML20236X899 (4) | |
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DUKE POWER GOMPANY P.O. HOx 33180 CliARLOTTE, N.C. 28242
, ILAL B. TUCKER TELEPif0NE mvoYa$n"((Ent$m December 2, 1987 U.S. Nuclear Regulatory Commission Document Control Desk Washington, D.C.
20555
Subject:
McGuire Nuclear Station Docket Nos. 50-369 and 50-370 Catawba Nuclear Station Docket Nos. 50-413 and 50-414 Proposed Technical Specification to Revise Allowable Leakage Gentlemen:
Duke Power Company submitted, by letters dated February 18 and March 23, 1987, proposed changes to Technical Specifications to create a graduated response to Reactor Coolant System (RCS) leaks in excess of one gpm.
The Staff responded by.
letter dated July 28, 1987 which identified several items of concern and requested i
that Duke respond with either supporting analysis or the intent to retain the present Technical Specification (TS).
Prior to providing responses to the staff items of concern, Duke. believes it is useful to briefly describe the proposed change.
The purpose of the proposed change is to increase the amount of time available to locate leaks before a plant shutdown is required. The proposed Limiting Condition
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for Operation (LCO) remains the same as in the current Specification. Actions for
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exceeding the LCO leakage limits still require that within 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> the unit be in Hot Standby if the leak is not identified.
Further, the actions in the event the leakage is identified as pressure boundary leakage remain unchanged.
Under the current TS, the unit would be required to be in Cold Shutdown (Mode 5) in the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
The proposed Technical Specification would allow the unit to remain in Hct Standby (operating temperature and pressure, but the reactor is suberitical) for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> before going to Cold Shutdown, rather than proceeding i
directly to Mode 5.
At the end of the 72-hour Allowable Outage Time, or in the event the unidentified leakage exceeds 3 gpm, a shutdown to Mode 5 is required to be initiated.
It is important to emphasize that neither the effect or the purpose of the pro-posed specification is to raise the limit of allowable unidentified leakage, but rather to provide an action time which is appropriate and to facilitate identifi-cation of the leak.
Duke provides the following response to each of the Staff's concerns as presented in the Staff's July 28, 1987 letter.
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8712100346 871202
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PDR ADOCK 05000369
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Document Control Desk December 2, 1987
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Item 1:
The overriding Staff concern is to limit Reactor Coolant System leak by taking appropriate steps before reaching the TS limit of 1 gpm.
Response: Duke fully recognizes that, ideally, leaks should be identified, located and isolated or repaired before becoming an operational concern.
This Technical Specification change will have no effect on Duke's ability or intention to identify, locate, isolate, or repair leaks less than 1 gpm.
What it does provide is additional time at stable operating temperature and pressure, with the reactor shut down to accomplish these actions.
The Specification establishes a threshold at which an unidentified leak requires action to be taken.
This proposed specification improves the plant capability to address the Staff concern.
Item 2:
Additional concerns are identified in IE Information Notice 86-108 and supplement.
Response: The concerns identified in Information Notice 86-108 and supplement have j
been reviewed for applicability to McGuire and Ca tawb a.
The Notice discussed the effects of borated water on carbon steel piping. McGuire i
and Catawba use Stainless Steel piping throughout the Reactor Coolant I
system.
The reactor vessels, steam generators, and pressurizer are fabricated from carbon steel; however, the interiors of these components are clad with stainless steel.
Therefore, it is considered that the concerns addressed in IEN 86-108 and supplement are not a problem in the issue at hand.
Item 3:
The submittal has not demonstrated the stability of a flaw for a leak at 3 gpm.
Response: Westinghouse Report WCAP-10546 provides a stability analysis for a 10 inch crack in the RCS piping, which produces a leak rate in excess of 10 gpm.
Since the margin of stability increases for decreasing crack sizes, it is concluded that WCAP-10546 inherently demonstrates the stability of a 3 gpm leak, which would occur for a much shorter crack size than 10 inches.
Item 4:
Leak-before-break (LBB) analyses based on a 3 gpm leak rate, utilizing the acceptance criteria defined in NUREG-1061, Volume 3,
should be submitted for staff review regarding this matter. The margin applied in NUREG-1061, Volume 3, is 10 on leak rate.
Therefore, in order to demon-strate crack stability based on a 3 gpm actual leak rate, the stability of a 30 gpm leak-size through-wall crack needs be demonstrated.
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Document Control Desk l
December 2, 1987 Page 3 Response: This criterion depends upon the leakage detection capability of the i
plant with no reference to technical specification requirements or actions whatsoever.
The leakage detection capability at McGuire and 1
Catawba vill remain at 1 gpm regardless of technical specification changes. Therefore, according to NRC criteria, the leakage flaw should remain at 10 gpm. A 30 gpm leakage flaw would fit the NRC criteria only if 3 gpm were the minimum leak detection capability.
The minimum leak i
detection capability is not 3 gpm.
A 1 gpm flaw can still be detected and will be acted upon.
The present and proposed Technical Specifica-tions indicate very specific actions to be taken if the leakage is above 1 gpm.
Therefore, a 30 gpm leakage flaw is inappropriate for the leak detection capability of 1 gpm at these plants.
Item 5:
Fracture mechanics analyses should show that a through-wall flaw twice the size of the 30 gpm leak size flaw will be stable under combined normal (pressure, deadweight, and thermal) and safe shutdown earthquake (SSE) loads.
Furthermore, the fracture mechanics analyses need to demonstrate that the 30 gpm leak-size will be stable if the loads are increased to the square root of 2 times the combination of normal and SSE loads.
Response: As noted above in the response to Item 4, the 30 gpm leak size flaw is inappropriate for this application.
The 10 gpm leak size, which is considered appropriate, has been demonstrated to be stable under com-bined normal and safe shutdown loads, as shown in WCAP-10546. A summary of margins available relative to the critical flaw (which gives a leak of 10 gpm) size, with respect to loads, flaw size, and leak rate is provided in WCAP-10546, Section 7.
Item 6:
In reviewing your submittals, it appears that an inadequate application of fracture mechanics in LBB evaluations was made.
Attachments II to your June 9, 1986, and February 18, 1987, submittals state that the critical crack size is 32.5 inches.
This critical flaw size was ob-tained from Westinghouse Report WCAP-10546 which was submitted as a technical basis for the application of LBB at Catawba.
However, this critical flaw size was determined from a limit load analysis without accounting for the Catawba material fracture toughness limitations. The staff considers it necessary to account for the facility material fracture toughness properties in fracture mechanics stability analyses.
Response: Please refer to paragraphs 4.3 and 4.4 of WCAP-10546 where material fracture toughness properties are discussed.
There it can be seen that I
material fracture toughness properties have been accounted for in the analysis.
In determining the critical flew size, it is acceptable practice to use limit load analysis for austenitic stainless steels.
Fracture toughness limitations are then evaluated for the leakage flaw size, not the critical flaw size determined by limit load analysis.
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WCAP-10546 fully complies with these LBB concepts and is adequate in this regard.
Therefore, it is concluded that all appropriate analyses have been properly and successfully conducted.
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Document Control Desk December 2, 1987 Page 4 Conclusion As noted, the purpose of the proposed change is not to raise the limit of allowed unidentified leakage, but to provide additional time to perform the necessary actions in response to a small (1-3 gpm) unidentified leak.
Placing the planc in Mode 3 shuts down the reactor core while maintaining a stable temperature and pressure condition.
This permits searches for leaks to proceed effectively and efficiently and allows leakage calculations to continue due to the maintenance of a constant RCS temperature.
Very truly yours,
.as.k.
Hal B. Tucker SAG /103/jgc xc:
Dr. J. Nelson Grace Regional Administrator U.S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW - Suite 2900 Atlanta, GA 30323 Mr. Darl Hood, Project Manager Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Dr. K.N. Jabbour, Project Manager Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Mr. W.T. Orders NRC Resident Inspector McGuire Nuclear Station Mr. P.K. Van Doorn NRC Resident Inspector Catawba Nuclear Station I
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