ML20203C462
| ML20203C462 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 04/02/1986 |
| From: | Beckham J GEORGIA POWER CO. |
| To: | Grace J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| References | |
| 2465N, SL-517, NUDOCS 8604210166 | |
| Download: ML20203C462 (6) | |
Text
Georgia Power Company h
333 Piedmont Avenue Atlanta. Geergia 30308 Te!ephone 404 526-7020 Mait;ng Accregg.
Wst Off,ce 903 4343 Attanta, Georg,a 39302 I
P]2,-
UIS U POLn'r
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J. T. Beckham Jr.
M": d ' 't' ewrtr m. tem Vice Pres >0cnt and Ger'eral Manacer
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SL-517 2465N April 2, 1986 U. S. Nuclear Regulatory Comission
REFERENCE:
Office of Inspection and Enforcement RII: RDW Region II - Suite 2900 50-321 101 Marietta Street, NW Inspection Report Atlanta, Georgia 30323 85-32 ATTENTION: Dr. J. Nelson Grace Gentlemen:
The following additional information is submitted in response to Inspection Report 85-32, which concerns the inspection conducted by Messrs.
P.
Holmes-Ray and G.
M.
Nej felt of your office from October 12 to November 9, 1985 Georgia Power Company respectfully requests that you consider the following request for reconsideration of violation 321/85-32-01 provided you in our letter of January 13, 1986 Additionally, information responsive to the request in your letter of March 5,1986, is provided as an enclosure to this letter.
VIOLATION:
"10 CFR 50.72(b)(2)(1) requires that a four hour report be made of any event, found while the reactor is shut down, that, had it been found while the reactor was in operation, would have resulted in the nuclear power plant, including its principal safety barriers, being seriously degraded or being in an unanalyzed condition that significantly compromises plant safety.
Contrary to the above, on December 15, 1984, in service inspection (ISI) was being performed in Unit 1 on selected pipe welds using the magnetic particle inspection method.
During testing, a linear through wall crack approximately 2 3/4-inches long was discovered in weld IT48-2 CPI-18-PID-6 This weld is located in the 18-inch nitrogen inerting and purge line between penetration X-25 and valve IT48-F307 This nonisolable crack was an unanalyzed degradation of a safety barrier (containment) required to be reported under 10 CFR 50.72(b)(2)(i).
No four hour report was made.
This is a Severity Level V violation (Supplement I)."
8604210166 860402 PDR ADOCK 05000321 G
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r GeorgiaPower A U. S. Nuclear Regulatory Commission Office of Inspection and Enforcement Region II - Suite 2900 April 2,1986 Page Two RESPONSE TO VIOLATION:
Admission or denial of alleged violation:
Georgia Power Company finds that this alleged violation must be denied.
We agree that there was an unisolable crack in the purge line.
We also agree that the event is reportable as a Licensee Event Report.
However, it is not reportable as defined by 10 CFR 50.72(b)(2)(1).
Reason for the denial:
The basis of the denial of the violation in question hinges on the words "unanal compromises plant safety" rather than { zed condition that significantly that may significantly (emphasis added) as presented in your letter of March 5, 1986.
Our denial is based on the interpretation of 10 CFR 50.72 published at the time of its issuance (48 Federal R3 ster 39042, August 29, 1983) in that the 1
rationale provided in your letter conflicts with the Statement of Consideration.
Specifically, under the
" Paragraph by Paragraph Explanation of the Rule" the. regulation provides: "Since the types of events intended to be capturea'by this reporting requirement are similar to 50.72(b)(1)(ti) except that the reactor is shutdown, the reader should refer to the explanation of 50.72(b)(1)(fi) for more details on intent."
Under paragraph 50.72(b)(1)(ii), "The Commission recognizes that the licensee may use engineering judgment and experience to determine whether an unanalyzed condition existed," at 39042 Note that the word is " judgment," not " analysis."
The Shift Supervisor at the time of discovery examined the information available to him and judged the crack to be of a significance which did not require a four hour report.
The regulation does not require, nor does it intend to require, the performance of a full engineering evaluation within four hours of discovery.
If that were the case, the NRC would be inundated by four hour reports since virtually every off-normal condition found during an outage is "unanalyzed" and cannot be analyzed, in the sense of an in-depth engineering analysis, within four hours to prove that it does not "significantly compromise plant safety."
Further, an underlying goal of the final regulation was to decrease the number of reports
- which, in fact, did not constitute a safety concern.
Additional engineering
- analysis, which was applied when determining the reportability of the Licensee Event Report, substantiated the Shift Supervisor's judgment.
Additionally, in the same paragraph of the Federal Register: " Finally this paragraph also includes material problems that cause abnormal 700775
k Georgialbwer h U. S. Nuclear Regulatory Commission Office of Inspection and Enforcement Region II - Suite 2900 April 2, 1986 Page Three t
RESPONSE TO VIOLATION:
(Continued) degradation of the principal safety barriers... (f) loss of containment function or integrity including: (1) containment leakage rates exceeding authorized limits..."
This Section cites, as intent, the actual failure, not the mere potential failure.
No leakage existed at the time of discovery.
Furthermore, it is questionable whether a DBA LOCA would have resulted in a leakage rate which would have exceeded authorized i
limits.
The size of the crack was such that there would be minimal material release, if any, through it under any condition short of a DBA LOCA.
The location of the crack was such that any radioactivity escaping through the crack would have been released through a monitored, filtered release point.
If the release through the crack occurred during normal operations, even though we do not believe this to be plausible, it would be released through the Reactor Building Vent.
If the release through the crack occurred during a DBA LOCA, it would be released through the Standby Gas Treatment System.
For reasons noted in your meeting summary on this event, dated November 29, 1985, we do not consider catastrophic failure of the subject weld possible.
It must be noted that it is only because of Georgia Power Company's conservative policy to report items which are potentially reportable that the NRC discovered this alleged violation in the first place.
As stated in the inspection report, the violation was based on a revision to the Licensee Event Report, not in-field investigation at the time of discovery.
Therefore, we found and repaired the crack and had the engineering evaluation complete before the end of the reporting period in which it was cited as a violation.
This is not to imply that our actions were limited to evaluation.
As delineated in the Licensee Event Report and reiterated in the enclosure to this letter, Georgia Power Company took sufficient action to prevent recurrence of the problem before the citation for failure to appropriately report the event was issued.
We did not believe at the time of our initial
- response, January 13, 1986, that the cited violation was justified.
We still do not believe that the violation is justified.
However, the information included in the enclosure does identify the cause of the problem which we found and our corrective actions taken.
b GeorgiaPower d I
U. S. Nuclear Regulatory Commission Office of Inspection and Enforcement Region II - Suite 2900 April 2,1986 Page Four In summation, we agree with you that events of this nature should be reported to the NRC.
Our disagreement is on the appropriate vehicle for said reporting.
For the reasons delineated above, i.e., that the citation is in conflict with the bases as stated in the published Statement of Consideration, we do not believe that any violation should have been issued for this event.
Therefore, we respectfully request that you withdraw the citation for 321/85-32-01.
If-you feel a meeting would be beneficial to discuss this issue, we are ready to join you at our mutual convenience.
Please contact this office if you have any questions.
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Very truly yours,
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. T. Beckham, Jr.
MJBlackwood/lc i-enclosure i
c: Mr. J. P. O'Reilly Mr.' L. T. Gucwa Mr. H. C. Nix, Jr.
Senior Resident Inspector GO-NORMS 5
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Georgialbwerb ENCLOSURE Response to NRC l
Request for Additional Information The following information is provided pursuant to the request contained in the letter dated March 5,
- 1986, from D r.
J.
Nel son Grace, Regional Administrator, United States Nuclear Regulatory Consnission, Region II, to Mr. R. J. Kelly, Executive Vice President, Georgia Power Company.
Corrective steps which have been taken and the results achieved:
The l
crack was ground out.
The weld was repaired and satisfactorily l
inspected.
Five additional welds in the same line were magnetic l
particle inspected, and both the subject weld and the next weld in the l
line were radiographed.
No additional failure indications were detected.
The corresponding welds in the Unit 2 nitrogen inerting line l
were checked using the magnetic particle inspection method.
No failure indications were detected.
l An engineering evaluation was conducted to determine the cause of i
failure.
Visual and radiographic analysis of the crack, examination of i
the pipe geometry in the area of the failure, and the system operating conditions indicate that the crack was caused by thermally induced stress or growth of a pre-existing weld defect due to applied thermal stress.
A two-inch nitrogen makeup line enters the eighteen-inch purge line on the opposite side and approximately two feet upstream from the crack.
Cold gas, entering the purge line from the makeup line would make contact with the purge line in the area of the weld, the contour of which would make it the most likely spot for a thermally induced failure to occur.
During a makeup evolution, makeup flow is controlled and drywell pressure is monitored.
However, makeup gas temperature was not monitored.
A " Nitrogen Inert System Make-up Low Temperature" alarm existed, but there was no response indicated in any annunciator response procedure.
To prevent recurrence of this event, the following actions have been taken for both units:
1)
The setpoint for the "Ni trogen Inert System Make-up Low Temperature" annunciator has been raised from 00F to 100F; 2)
The "Drywell and Torus Inerting System Trouble" annunciator response procedure has been revised to include the " Nitrogen Inert System Make-up Low Temperature" annunciator and the direction to terminate the makeup flow when the annunciator occurs;
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Georgia Rnver h Response to NRC Request for Additional Information (Continued) l 3)
The " Primary Containment Atmospheric Control System" operating procedure has been revised tn require the stopping of nitrogen makeup flow if the "Ni tr agen Inert System Make-up Low Temperature" annunciator is actuated.
Corrective steps which will be taken to avoid further violations:
The above actions are sufficient to prevent recurrence of the problem.
Date when full compliance was achieved:
Full compliance was maintained j
throughout the event.
The crack was repaired on December 19, 1984 l
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