ML20045F474
| ML20045F474 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 07/01/1993 |
| From: | Mcmeekin T DUKE POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9307070329 | |
| Download: ML20045F474 (5) | |
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' IMe 1%urt Company (704)875 4000 McGuire Nuclear Station 12700 flagers ferry Road fluntersville, NC280781985
. DUKEPOWER i
July 1, 1993 Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Subject:
McGuire Nuclear Station Docket Nos:
50-369 and 370 Supplement to Technical Specification Amendment-Containment Bypass Leakage Rate i
Dear Sir:
By letter dated June 23, 1993, Duke Power submitted a proposed amendment to the Technical Specifications for the McGuire Nuclear
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Station which revised the allowable containment bypass leakage rate.
Review of this change request by the NRC Staff identified l
clarifications that were necessary in portions of the submittal.
Accordingly, please find attached a revised Attachment 3 which i
contains the Safety Basis for the Request, the No Significant Hazards Analysis, and the Environmental Impact Analysis.
Please contact Robert Sharpe at ~(704) 875-4447 if there are any questions regarding this Technical Specification amendment request.
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.i Very truly yours,-
T. C. McMeekin i
060108
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9307070329 930701
'I PDR ADOCK 05000369 E8 -
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U. S., Nuclear _ Regulatory Commission July 1, 1993
'Page 2 xc:
Mr.
S.D.
Ebneter Administrator, Region II'
-U.S.
Nuclear Regulatory Commission j
Mr. Victor Nerses I
U.S.
Nuclear Regulatory Commission Office of Nuclear Reactor Regulation
.l Mr. P.K. VanDoorn
_.i NRC Resident Inspector McGuire Nuclear Station i
American Nuclear Insurers M & M Nuclear Consultants INPO Records Center l
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Attachment'3 Safety Basis for the Request No Significant Hazards Consideration Evaluation Environmental Impact Consideration i
Safety Basis For The Request Recent testing of the bellows assembly M-441 (lD main steam line i
containment penetration) identified increased secondary bypass leakage through the penetration.
This leakage, in addition to other small leaks, results in a total containment bypass' leakage approaching the limit of.07 La as stated in 3.6.1.2 e of the l
McGuire Technical Specifications.
The
.07 La leakage limit j
corresponds to a leakage of 9,427 sccm for McGuire.
The Offsite and Control Room Dose calculations were reviewed to determine the dose significance of increased bypass leakage. These calculations showed that McGuire was currently well within the 10 l
CFR 100 limits for all analyzed accidents which have containment bypass leakage as an input.
The dose most sensitive to a change in j
bypass leakage was the dose to control room personnel in the event of a large break LOCA with substantial fuel melting. This accident also considers a passive failure of the ECCS system which releases 50 gpm of ECCS water for 30 minutes prior to isolation.
General
'e Design Criteria 19 (GDC 19) establishes a dose limit of 5 Rem whole v
body or equivalent for these personnel while responding to an accident.
(The McGuire licensing limit is 5 Rem whole body.and 30 Rem thyroid for control room personnel. ) The accident scenario was reevaluated considering bypass leakage of 14000 sccm or.104 La
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I during the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the accident.
The leakage is assumed to drop to one half this value after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in correspondence to the reduced containment pressure assumed after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
The results of this reanalysis showed that the dose to control room personnel was still in compliance with GDC 19. Specifically, the whole body dose increased.from 1.16 Rem to 1.17 Rem and the thyroid l
dose increased from 24.8 Rem to 28.4 Rem.
Based on this review, it is determined that the increase in bypass j
leakage to.104 La is not a significant impact on.the health and safety of the public.
Sufficient leakage margin above the leakage as measured through tests is.available to ensure that leakage will not increase above the revised limit during the period when the Technical Specification change is in effect.
Since-the bypass l
leakage is primarily. through one penetration, a
program-of increased testing of this penetration is being implemented along with this change in allowable leakage.
This will further. ensure the bypass leakage remains well within the revised limits during
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unit operation.
The main steam penetration assembly acts as a part of the primary l
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containment boundary and is thus a fission product barrier.
The increase in leakage through this barrier results in an increase in consequence for accidents which require containment integrity for accident mitigation.
Analysis of these accidents show that all dose consequences are within acceptable bounds considering increased containment bypass leakage.
There is no increase in the probability of an accident since no accident initiators are i
involved with this change.
The function of all equipment remains the sanc except tha the outer bellows is now considered a part of the containment pressure boundary. The space between the inner and outer bellows is vented to the annulus where any leakage is filtered by the Annulus Ventilation system.
This leakage is considered within the total containment leakage allowable of.3%
l containment air mass per day (134,667 sccm). The amount of leakage following this path is small compared to the total allowable which is considered in dose analysis.
The.6 La limit for total measured j
leakage from penetrations and valves subject to type B and C testing continues to be met considering this increased leakage.
The increase in allowable bypass leakage will not increase dose to the public or personnel required for accident mitigation to an unacceptable level.
Increased monitoring of the penetration will ensure that no unacceptable degradation of containment integrity will occur.
Therefore, it is concluded that the Technical i
Specification change is acceptable with regard to the health and i
safety of the public.
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No Significant Hazards Consideration Evaluation Duke Power Company (Duke) has made the determination that this
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amendment request involves a No Significant Hazards Consideration by applying the standards established by NRC regulations in 10 CFR 50.92. This ensures that operation of the facility in accordance with the revised containment bypass leakage rate for McGuire Unit 1 Cycle 9 would not.
(1)
Involve a
significant increase in the probability or consequences of an accident previously evaluated:
The increase in leakage through the main steam penetration bellows results in an increase in the consequence for j
accidents which require containment integrity for accident j
mitigation.
Analysis of these accidents show that all dose i
consequences are within the McGuire licensing limits l
considering increased containment bypass leakage. There is no j
increase in the probability of an accident since no accident initiators are involved with this change.
(2)
- h (2)
Create the possibility of a new or different kind of accident from any accident previously evaluated Operation of McGuire Unit 1 in accordance with the revised containment bypass leakage rate will not create any failure modes not bounded by previously evaluated accidents.
Consequently, this change will not create the possibility of a new or different kind of accident from any accident previously evaluated.
(3)
Involve a significant reduction in a margin of safety While the conservatively measured leakage through one mechanical penetration bellows increased this outage, this leakage represents a
small fraction of the allowable containment leakage.
The proposed Technical Specification change increases the allowable containment bypass leakage rate.
This still assumes that the containment remains operable and performs its safety function.
The proposed changes to the Technical Specifications will not impact the overall performance of the containment and will not prevent it from performing its safety function. Even with the Technical Specification change, the containment will continue to preven?.
uncontrolled releases to the environment. All other fission product barriers remain in place and function to. limit accident consequences.
In the event of a postulated design basis accident (DBA), the proposed Technical Specification change would not result in doses in excess of NRC acceptance criteria. Analysis results indicated a very slight increase in the radiation dose to control room personnel.
Accordingly, the proposed Technical Specification change would not result in a significant reduction in the margin of safety.
Environmental Impact Analysis The proposed amendment has been reviewed against the criteria of 10 CFR 51.22 for environmental considerations. The proposed amendment does not involve a significant increase in the amounts, and no significant change in the types, of any _ ef fluent that may be-released offsite and that there is no significant increase in individual or cumulative occupational exposure.
Therefore, the proposed amendment meets the criteria given in 10 CFR 51.22(c)(9) for a
categorical exclusion from the requirement for an environmental impact statement.
(3)