ML19323F611
| ML19323F611 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 05/07/1980 |
| From: | Parker W DUKE POWER CO. |
| To: | James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| Shared Package | |
| ML19323F613 | List: |
| References | |
| IEB-80-04, IEB-80-4, NUDOCS 8005290231 | |
| Download: ML19323F611 (2) | |
Text
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37 3-4 0 e 3 Mr. James P. O'Reilly, Director U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, Suite 3100 Atlanta, Georgia 30303 Re: RiI:JPO McGuire Nuclear Station Docket Nos. 50-369 and 50-370 IE Bulletin 80-04
Dear Mr. O'Reilly:
As requested by your letter of February 8, 1980, which transmitted IE Bulletin 80-04, the steam line break analysis for McGuire Naclear Station has been reviewed. The review of the containment pressure response analysis was previously conducted as requested by Mr. R. L. Baer, NRC/0NRR, in his letter of September 21, 1979. Duke Power Company's response to Mr. Baer's request for additional information was transmitted by my letter of October 25, 1979.
This response addresses Item 1 of the bulletin and is attached. Also attached is Duke's response to Items 2 and 3 of the bulletin.
f Very truly yours, M
w William O. Parker, Jr.
THH:scs Attachments cc: NRC Office of Inspection and Enforcement Reactor Operations Inspection W.shington, D. C.
20555 4
i OFFICIAL COPY 8005290 NN.
o DUKE POWER COMPANY MCGUIRE NUCLEAR STATION Response to IE Bulletin 80-04 Item 1 See attat ked letter from Mr. W. O. Parker to Mr. H. R. Denton.
Item 2 A review of the analysis of the reactivity increase due to a steam line break has been performed. The assumptions used in the analysis are provided in the attached October 25, 1980 letter. This analysis shows that the core transient results are very insensitive to auxiliary feedwater flow. The first minute of the transient is dominated entirely by the steam flow contribution to primary-secondary heat transfer, which is the forcing function for both the reactivity and thermal-hydraulic transients in the core. The assumptions mentioned above are appropriate and conservative for the short-term aspect of the steam line break transient.
The auxiliary feedwater flow becomes a dominant factor in determining the duration and magnitude of the steam flow transient during later stages in the transient. However, the limiting portion of the transient occurs during the first minute both due to higher steam flows inherently present early in the transient and due to the introduction of boron to the core via the safety injection system.
. Item 3 The review of the steam line break analysis revealed no shortcomings in the analysis from either containment overpressure or reactivity considerations.
Therefore, no modifications are warranted.
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