ML19256E395
| ML19256E395 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 10/09/1979 |
| From: | Moody D YANKEE ATOMIC ELECTRIC CO. |
| To: | Grier B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| References | |
| WYR-79-117, NUDOCS 7911050130 | |
| Download: ML19256E395 (3) | |
Text
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7n Telephone 617 366-90\\l rwx 710 390-0739 YANKEE ATOMIC ELECTRIC COMPANY
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f 20 Turnpoke Road Westborough, Massachusetts 01581
, ANK$E B.4.1.1 WYR 79-117 October 9,1979 United States Nuclear Regulatory Commission Office of Inspection and Enforcement Region I 631 Park Avenue King of Prussia, PA 19406 Attention:
Mr. Boyce H. Grier, Director
References:
(a) License No. DPR-3 (Docket No. 50-29)
(b) USNRC Letter to YAEC dated August 13, 1979; I&E Bulletin No. 79-21
Dear Sir:
Subject:
Response to I&E Bulletin No. 79-21 The following information is being provided in response to I&E Bulletin No. 79-21.
The responses address in sequence the concerns raised in the bulletin.
There are four liquid level measuring systems which are located within
.ontainment and used for safety actions or post-accident monitoring information.
a.
The Steam Generator Narrow Range and Wide Range Level Systems use vented reference legs open to the top of the steam generator and differential pressure transmitters. The narrow range system is used to provide a low level trip input to the Reactor Protection System.
The wide range system is used for post-accident monitoring.
b.
The Pressurizer Narrow Range and Wide Range Level System use vented reference legs open to the top of the pressurizer and differential pressure transmitters.
The wide range system is used to provide a high level trip input to the Reactor Protection System and can be used for post-accident monitoring. The narrow range system is used for post-accident monitoring and level control.
4 g' 776 7911050
United States Nuclear Regulatory Commission October 9, 1979 Attention:
Mr. Boyce H. Grier, Director Page 2 An evaluation of the effects of post-accident ambient heatup on these level systems has been completed.
Plant protection for high energy line breaks inside containment is assured by a number of reactor protection system trip functions, including:
1.
high neutron flux 2.
high pressurizer water level 3
low pressurizer pressure 4.
low steam generator water level, and 5.
high containment pressure The primary reactor protection system function is dependent on a number of parameters, including location of break (primary, steamline, feedline) and size of break.
1.
Primary loss of Coolant - For a primary loss of coolant event, the primary reactor trip function is low pressurizer pressure.
Backup is via a high containment pressure trip.
2.
Steamline Rupture Inside Containment - For a steamline rupture in containment, the primary reactor trip function is a high containment pressure signal. Backup trips are provided by high neutron flux and low pressurizer pressure signals.
3 Feedwater Line Rupture Inside Containment - The worst case of feedwater line rupture inside containment is limited to a single steam generator blowdown due to Lystem design and piping configurations.
The primary plant protection for this event is dependent on the rupture size and blowdown rates.
For small ruptures the primary reactor protection is provided by the low steam generator wate? level trip on the intact steam generator, with backup vi-
..igh containment pressure and high pressurizer water level trips.
For large feedwater line ruptures, the primary plant protection is the high containment pressure trip with low steam generator water level and high pressurizer water level trips as backup.
For feedwater line ruptures which would have the low steam generator water level trip as the primary trip function, containment conditions will be such that the impact on the measured steam generator level will be insignificant.
For feedwater line ruptures which may yield containment conditions severe enough to affect steam generator level monitoring accuracy the plant will be tripped via high containment pressure.
Even for these severe cases, it is not expected that steam generator water level measurement will be significantly different than actual level.
In addition, contrary to the standard Westinghouse design, there is no automatic initiation of auxiliary feedwater at Yankee Rowe.
Thus it becomes obvious that the concerns for reactor protecticn during the events mentioned are not applicable to Yankee Rowe design.
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United States Nuclear Regulatory Commission October 9, 1979 Attention:
Mr. Boyce H. Grier, Director Page 3 A concern was also expressed that high containment temperatures following a secondary line break could affect level indications during post-accident monitoring. While it is true that the containment temperatures at Yankee Rowe could attain high temperatures (>3000F) during a main steamline rupture event the duration is short and near normal condition will be attained within a few hours.
Thus it can be shown that even though steam generator and pressurizer level indications may briefly contain a slight error this is corrected in a relatively short period of time and post-accident monitsring is not significantly affected.
We trust that this information is satisfactory; however, should you require any additional information please call us.
Very truly yours, YANKEE ATOMIC ELECTRIC COMPANY Mw D
E. Moody Manager of Operations JKT/dep 1?;
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