ML13330A827
| ML13330A827 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 09/14/1979 |
| From: | Head J SOUTHERN CALIFORNIA EDISON CO. |
| To: | Engelken R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| Shared Package | |
| ML13330A826 | List: |
| References | |
| IEB-79-21, NUDOCS 7912210142 | |
| Download: ML13330A827 (5) | |
Text
Southern California Edison Company P. 0. BOX 800 2244 WALNUT GROVE AVENUE J. T. HEAD, JR.
ROSEMEAD, CALIFORNIA 91770 TELEPHONE cet PQesioC~r 213-572-1472 September 14, 1979 U. S. Nuclear Regulatory Commission Office of Inspection and Enforcement Region V Suite 202, Walnut Creek Plaza 1990 North California Boulevard walnut Creek, California 94596 Attention:
Mr. R. H. Engelken, Director Docket No. 50-206 San Onofre - Unit 1
Dear Sir:
IE Bulletin 79-21 Temperature Effects on Level Instrumentation Reference is made to your correspondence of August 13, 1979, forwarding the subject IE Bulletin.
This Bulletin identified potential errors in level measurements due to the effects of increased containment tenperature on reference leg water columns.
This letter presents the results of our review of this matter.
Responses to individual items specified in the Bulletin are listed below.
Item 1 "Review the liquid level measuring systems within containment to de termine if the signals are used to initiate safety actions or are used to provide post-accident monitoring information.
Provide a description of systems that are so employed, a description of the type of reference leg shall be included, i.e., open column or sealed reference leg."
Response
A review of the San Onofre Unit I liquid measurement systems inside containment indicates that there are six such level instruments that needed to be considered in this evaluation.
7912210 7-
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U. S. Nuclear Regulatory Commission Page 2 Three of these instrunents, LT-430, LT-431 and LT-432, are located on the pressurizer and provide input to the pressurizer level control system and the high pressurizer level reactor trip circuit.
These instrunents also provide post-accident monitoring of pressurizer level.
LT-430, LT-431 and LT-432 consist of differential pressure transmitters with ex ternal sealed colum reference legs.
The remaining three level instrunents, LT-450, LT-451, and LT-452, are provided for wide range level indication on steam generators A, B and C, respectively.
They consist of differential pressure transmitters with external sealed column reference legs.
These instruments do not initiate safety actions but are powered from the station vital buses and could be available for post-accident monitoring of steam generator level following an accident.
Additionally, there are three level instrunents, LT-453, LT-454, and LT-455, provided for narrow range level indication on steam generators A, B, and C, respectively.
Cutput f ran these level instruments is fed to the steam generator level control system and to the turbine generator high level trip logic.
There are no signals fran these instruments which are used to ini tiate safety actions nor are they designed to provide post-accident moni toring information.
Therefore, no further discussions will be provided for the narrow range steam generator level indicators.
Item 2 "On those systems described in Item 1 above, evaluate the affect of post accident ambient temperatures on the indicated water level to determine any change in indicated level relative to actual water level. This eval uation must include other sources of error including the affects of varying fluid pressure and flashing of reference leg to steam on the water level measurements.
The results of this evaluation should be pre sented in a tabular form similar to Tables 1 and 2 of Enclosure 1."
Reconse:
An increase in the tererature of the reference leg water column results in an increase in indicated level.
Table 1 provides a suwary of the effects of post-accident contairment temperature on indicated pressurizer and steam generator level.
Pressurizer level instrunentation is available for post-accident monitoring, hcwver, its accuracy would be suspect and operators have been instructed and procedures will be revised accordingly.
Steam generator level instrumentation, although not presently required, could be used for post-accident monitoring.
However, as in the case of the
U. S. Nuclear Regulatory Commission Page 3 pressurizer, its accuracy would be suspect and operators have been in structed and procedures revised accordingly.
Boiling could conceivably occur in the reference leg following depressuri zation of either the steam generators or pressurizer with high containment temperature.
This combination of conditions could occur only following an accident which results in an alrmst total depressurization of the pres surizer or steam generators, respectively.
Recent generic analyses by our NSSS supplier indicate that such reference leg boiling would not occur.
If such boiling were to occur, it could cause a major bias in the indicated level for a short period of time, in the the extreme case indicating 100%
level when the vessel is actually empty.
However, since the only safety trip associated with any of the level instruments is the pressurizer high level trip (see discussion under Item 3),
any error introduced by boiling wculd be in the conservative direction.
The post-accident monitoring function of the instrunentation would ccrmence subsequent to major system transients, when conditions are relatively stable.
Therefore, the effects of large pressure variations and reference leg boiling are not considered below.
TABLE 1 Deviation of indicated pressurizer level for reference leg heatup effects due to post-accident containment teperature.
Reference Leg Deviation of Pressurizer Temperature ( 0F)
Level Indication (% of Span) 100.
0 150 3
200 6
250 10 300 14 Deviation of indicated steam generator level for reference leg heatup effects due to post-accident containment temperature.
Reference Leg Deviation of Steam Generator Tenerature ('F)
Level Indication (% of Span) 100 0
150 2
200 5
250 7
300 11
U. S. Nuclear Regulatory Commission Page 4 Iten 3 Review all safety and control setpoints derived fran level signals to verify that the setpoints will initiate the action required by the plant safety analyses throughout the range of ambient temperatures en countered by the instrunentation, including accident terrperatures.
Pro vide a listing of these setpoints.
If the above reviews and evaluations require a revision of setpoints to ensure safe operation, provide a description of the corrective action and the date the action was completed.
If any corrective action is temporary, submit a description of the proposed final correction action and a timetable for implementation."
Response
A review of safety and control setpoints derived fran level signals in dicates that the pressurizer high level reactor trip is the only such setpoint which initiates action required by the Safety Analysis.
This setpoint, the maximtrn safety systen trip setting for pressurizer high level, is 68% of the pressurizer level span (i.e. 27.3 ft. above the bottan of the pressurizer).
Heatup of the reference leg causes an in crease in indicated level and therefore would result in a reactor trip at a pressurizer level below the normal setpoint.
Since this would result in a conservative action by the reactor protection system, no change in the setpoint will be made.
Item 4 "Review and revise, as necessary, emergency procedures to include specific information obtained fran the review and evaluation of Items 1, 2, and 3 to ensure that the operators are instructed on the potential for and mag nitude of erroneous level signals.
All tables, curves, or correction factors that would be applied to post-accident monitors should be readily available to the operator.
If revisions to procedures are required, provide a comple tion date for the revisions and a capletion date for operator training on the revisions."
Response
All station emergency procedures have been reviewed relative to this matter.
Emergency Procedure S-3-5.20, Steam Generator High Energy Pipe Break, spe cifies minimum pressurizer level as a condition for resetting safety injec tion.
This procedure will be revised to reflect the results of our review of the pressurizer level instrunentation per Items 1 through 3 above.
This revision and any required operator training will be completed by September 30, 1979.
U. S. Nuclear Regulatory Commission Page 5 Previously, cautionary statements were added to Emergency Procedure S-3-5.20 alerting the operator of the potential for erroneous steam generator level signals due to reference leg heatup and boiling.
Ad ditionally, Station Order S-O-104, Reactor Standard for Cperation, has been revised to instruct operators not to base decisions solely on a single plant parameter when more than one confirmatory indication is available.
Should you have any further questions concerning this matter, please contact me.
Sincerely, cc:
Director, Office of Inspection and Enforcement, Division of Reactor Cperations Inspection