ML19318C848
| ML19318C848 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 06/16/1980 |
| From: | Mills L TENNESSEE VALLEY AUTHORITY |
| To: | James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| References | |
| IEB-80-04, IEB-80-4, NUDOCS 8007020347 | |
| Download: ML19318C848 (3) | |
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TENNESSEE VALLEY AUTHORITY
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...,,p 400 Chestnut Street Tower II June 16, 1980 ;0 JUN 19 A 9: 3I Mr. James P. O'Reilly, Director Office of Inspection and Enforcement U.S. Nuclear Regulatory Commission Region II - Suite 3100 101 Marietta Street Atlanta, Georgia 30303
Dear Mr. O'Reilly:
OFFICE OF INSPECTION AND ENFORCEMENT BULLETIN 80 RII:JP0 50-327 -
SEQUOYAH NUCLEAR PLAKT UNIT 1 The subject letter dated February 8,1980, transmitted OIE Bulletin 80-04.on Analysis of a PWR Main Steam Line Break with Continued Feed-water Addition. An interim report was submitted on May 8, 1980.
Enclosed is our response for Sequoyah N'iclear Plant unit 1.
If you have any questions concerning this inatter, please get in touch with D. L. Lambert at FTS 857-2581.
Very truly yours, 1
TENNESSEE VALLEY AUTHORITY 1
f L. M. Mills, Ma age /
Nuclear Regulation and Safety
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Faclosure Director of Reactor Operations Inspection (Enclosure) cc:
Office of Inspection and Enforcement U.S. Nuclear Regulatory Commission Washington, DC 20555 7
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ENCLOSURE SEQUOYAH NUCLEAR PLANT V
ANALYSIS OF A PWR MAIN STEAM LINE BREAK WITH CONTINUED FEEDWATER ADDITION RESPONSE TO IE BULLETIN 80-04 The TVA response to the subject IE Bulletin is as follows:
Response to Item 1-The response to this item has been previously addressed by TVA's response to Sequoyah FSAR question Q6.56B.
Response to Item 2 The assumptions made for main and auxiliary feedwater flow as they
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apply to licensing basis steamline break transients have been reviewed.
Several of the relevant assumptions used in all core transient analyses follow, and are further explained in the Sequoyah FSAR sections 6.2.*.3.11 and 15.5.4.
1.
The reactor is assumed initially to be at hot shutdown conditions, at the minimum allowable shutdown margin.
2.
For the Condition IV breaks, i.e., double-ended rupture of a main steam pipe, full main feedwater is assumed from the beginning of the transient at a very conservative cold temperature.
3.
All auxiliary feedwater pumps are initially assumed to be operating, in addition to the main feedwater. The flow is equivalent to the rated flow of all pumps at the steam generator design pressure.
4.
Feedwater is assumed to continue at its initial flow rate until feedwater isolation is complete, approximately 10 seconds after the break occurs, while auxiliary feedwater is assumed to continue at its initial flow rate.
5.
Mairi feedwater is completely tenninated following feedwater isolation.
Based on the manner in which the analysis is performed for Westinghouse plants, 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 effect of auxiliary feedwater runout (or failure of runout protection where applicable) is minimal.
Greater feedwater flows during the large steamline breaks serves to reduce secondary pressures, accelerating the automatic safe-guards actions, i.e., steamline isolation, feedwater isolation and safety injection.
The assumptions described above are therefore appropriate and conservative for the short-term aspect of the steamline break transient.
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y The auxiliary feedwater flow becomes a dominant factor in determining the (i
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.
In conclusion, the affect of runout aux ~iliary feedwater flows in the core transient for steamline break has been evaluated; and based on this evaluation, it has been determined that the assumptions presently made are appropriate for use as a licensing basis.
The concerns outlined in the introduction to IE Bulletin 80-04 relative to, (1) limiting core conditions occurring during portions of the transient where auxiliary feedwater flow is a relevant contributo to plant cooldown; and (2) incomplete isolation sf main feedwater flow, are not representative of the Westinghouse NSSS designs and associated Balance of Plant requirements.
Response to Item 3-Based on the response to items 1 and 2, no corrective action is necessary, 1
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