ML20205T665
| ML20205T665 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 06/10/1986 |
| From: | Allen C COMMONWEALTH EDISON CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.F.2, TASK-TM 1718K, GL-84-23, NUDOCS 8606160133 | |
| Download: ML20205T665 (4) | |
Text
f~x Commonwealth Edison
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') 72 West Adams Street Chicego Illinois
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- 7 Address Reply to: Post Offics Box 767
%,/ Chicago, Ilknois 60690- 0767 June 10, 1986 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555 i
Subject:
LaSalle County Station Units 1 and 2 I
Generic Letter 84-23 and NUREG-0737 Item II.F.2 NRC Docket Nos. 50-373 qd 50-374 References (a): December 4, 1984 letter from G. L.
Alexander to H. R. Denton (b): March 26, 1985 letter from A. Schwencer to D. L. Farrar (c): May 3, 1985 letter from G. L. Alexander to H. R. Denton
Dear Mr. Denton:
The purpose of this letter is to inform you of the status of the modification of the Reactor Vessel Water Level Instrumentation System and to revise our schedule for its implementation. Based on additional review by engineering and station personnel, the design to be implemented has been changed from that identified in reference (a). The changes reflect information gathered by detailed walkdown of the drywell and are discussed in the attachment.
Reference (a) contained our initial response to Generic Letter 84-23
" Reactor Vessel Water Level Instrumentation in BWRs".
In that response, we stated that we expected to complete the required modifications on both units by the end of their respective first refueling outage. During a March, 1985, telephone conference between Dr. A. Bournia of the NRC and Mr. J. Marshall of Commonwealth Edison (CECO), we were requested to review our commitments and provide a revised schedule if necessary. Based on our projection, we committed to the second refueling outage for Unit 1 and the first refueling outage for Unit 2.
Reference (b) forwarded a Safety Evaluation of our original submittal, but did not include the revised schedule provided during the aforementioned telephone conversation. Reference (c) reiterated the revised CECO commitment for installation of the modification and requested that the NRC amend reference (b) to incorporate the revised schedule.
8606160133 860610 d g 7, PDR ADOCK 0500 3
P ll, l
Mr. H. R. Denton June 10, 1986 Operation of LaSalle Unit 2 has not permitted drywell access in time to complete the detailed engineering design for installation of the rcodification during the Unit 2 first refueling outage. Therefore, we have rescheduled implementation of this work until the Spring, 1988, refueling outage. Design work will continue to the extent practical, and an effort will be made to install the modification during the first refueling outage.
The modification for LaSalle Unit 1 is still expected to be completed during the second refueling outage. This is based upon completion of drywell walkdowns during the current refueling outage to complete the design and allow installation.
Please address any questions that you or your staff may have concerning this response to this office. One signed original and fifteen copies of this letter with attachments are being provided for your use.
Sincerely, C. M. Allen Nuclear Licensing Administrator 1m Attachment cc:
Dr. A. Bournia LaSalle Resident Inspector M. C. Parker - IDNS l'118K
ATTACHMENT The primary objective of the reactor vessel water level instrumenta-tion system (RVWLIS) improvement is to provide a reliable and accurate water J
1evel display to the operator in the control room following a postulated post-accident high drywell temperature environment and subsequent depressuri-zation of the reactor vessel which may result in the flashing and boil-off of the reference legs. This objective will be accomplished by installing two new reference legs in the drywell for that level instrumentation that will be used by the operator to monitor the reactor vessel level in the post-accident environment.
Design Basis 1.
Two new reference legs with condensing chambers will be designed to feed four wide range level transmitters for providing reactor vessel level indication in the control room as shown in Table 1.
These instruments will be used by the operator to monitor the reactor vessel level in the event of reference leg boil-off or flashing following postulated high drywell temperature.
2.
The maximum piping drop of the new reference leg system in the drywell shall be limited to allow an indicated level at the bottom of the wide range instruments when the actual level is just above reactor vessel variable leg tap of the wide range instruments.
3.
All components of the new reference leg system will be seismic Category
)
1, Quality Group B and safety-related (Class lE).
4.
The new system will be designed to accomodate a reference leg break and a simultaneous failure of a level transmitter without loss of level indication.
5.
The equipment and components of the reference leg system will be environmentally qualified for their installed locations. The environmental parameters used for the qualification of the equipment and components are included in the FSAR.
6.
A single failure of the new reference leg piping and components shall not degrade the safety status of the existing systems and components.
System Description
The two new reference legs will be installed as shown in Figure 1.
(
One of the reference legs with new condensing chamber LX-1B21-D367 will be connected upstream of the condensing chamber on the existing reference leg l
connected to the reactor vessel nozzle N14D. The piping of this reference leg will be routed outside containment through containment penetration M-52.
l This piping will then be connected to the two existing level transmitters l
.' B21-N026BA and B21-N026AA on panels H22-p027 and H22-P004 respectively. One of these transmitters will feed post-accident level recorder LR-B21-R884B and the other will feed a new post-accident level indicator LI-B21-R887 in the control room.
The second reference leg with a new condensing chamber LX-B21-D368 will be connected upstream of the condensing chamber on the reference leg connected to the reactor vessel nozzle N14A. The piping from this reference leg will be routed outside the containment using existing penetration M-21.
This piping outside the containment will be connected to the two existing level transmitters B21-N026DA and B21-N026CA. The transmitter B21-N026CA located on panel H22-P005 will feed the level indicator B21-R604 in the control room. The transmitter B21-N026DA located on a local panel will feed post-accident level recorder LR-B21-R884A in the control room.
System Operation The four wide range level indicators (two recorders and two indicators) will be calibrated based on normal power operating pressure and temperature conditions. During normal plant operating conditions these level instruments will provide reactor vessel level indication in the control room over the wide range (-160" to +60").
Since the vertical piping drop in the containment for the reference legs of these instruments will be limited to less than 4 feet, the boil-off or flashing of these reference legs in the event of a high drywell temperature and low reactor pressure will introduce insignificant error in the level indication by these instruments in the control room. A reliable level indication will thus be available in the control room following a postulated high temperature condition in the drywell.
This design would result in the following benefits:
l 1.
Comparison of level indications being fed from "old" and "new" reference legs would demonstrate the existence of flashing in the "old" leg.
2.
No recalibration of level trip units would be required. Only minimal revision of maintenance procedures and training would be required.
3.
Radiation exposure during installation would be substantially l
reduced compared to the four leg reroute.
4.
Divisional separation would be maintained as originally designed inside containment.
1718K
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