ML19269F491
| ML19269F491 | |
| Person / Time | |
|---|---|
| Site: | San Onofre, Rancho Seco  |
| Issue date: | 09/14/1979 |
| From: | Mattimoe J SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | Engelken R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| Shared Package | |
| ML13330A826 | List: |
| References | |
| IEB-79-21, NUDOCS 7912210155 | |
| Download: ML19269F491 (5) | |
Text
SMUD SACRAMENTO MUNICIPAL UTILITY DISTRICT C 6201 s street, Box 15830, Sacramento, Cahfornia 95813; (916) 452-3211 September 14, 1979 Nuclear Regulatory Commission Attention:
Mr. R. H. Engelken, Director Region V Office of Inspection &
Enforcement 1990 North California Boulevard Walnut Creek Plaza, Suite 202 Walnut Creek, California 94596 Docket No. 50-312 Rancho Seco Nuclear Generating Station, Unit No.1 IE Bulletin 79-21
Dear Mr. Engelken:
The Sacramento Municipal Utility District has reviewed IE Bulletin 79-21 concerning the effect of temperature on level measurements.
The following information is provided in response to the items in this Bulletin.
Action 1. Review the liquid level measuring systems within containment to determine 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 dest _ription of the type of reference leg shall be included, i.e., open column or sealed reference leg.
RESPONSE
The liquid level measuring systems for the steam generator, pressurizer, core flood tank, containment sump, and pressurizer relief tank have been examined. The measuring systems for the steam generator and pressurizer are of the delta pressure, open column, uninsulated reference leg type.
The core flood tank has a dry reference leg that is not affected by the containment environment. The containment sump and pressurizer relief tank use floats for level measurement which are also not affected by the containment environment.
None of the signals from these measuring systems are used to initiate safety actions.
However, all could provide some information of use following an accident.
Action 2. On those systems described in Item 1 above, evaluate the effect of post-accidert ambient temperatures on the indicated water level to determine any change in indicated level relative 2167 072 79-210 T ri l C S Y 3 ; r..
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R. H. Engelken September 14, 1979 to actual water level. This evaluation must include other
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sources of error including the effects of varying fluid pressure and flashing of reference leg to steam on the water level measurements. The results of this evaluation should be presented in a tabular form similar to Tables 1 and 2 of.
RESPONSE
The effect of the reference leg temperature on the level measure-ment systems which are affected by temperature identified above is listed in Tables 1 and 2.
The errors listed have been maximized to acccunt for varying fluid pressures.
Consideration has been given to boiling in the reference leg and the ejection of water from the reference leg due to the effer-vescence of soluble gases.
These effects will be discussed for each component employing a water reference leg.
Steam generator level measurements are not significantly affected by the effervescence of soluble gases because there is insuffici'.nt soluble gas in the secondary system.
For boiling to occur in the steam generator reference leg, the reference leg must experience high temperatures and almost complete depressurization. The repressurization of the steam generator will refill the reference leg and the errors would be no greater than those listed in the tables.
The pressurizer level could be affected by the effervescence of soluble gases. The ejection of water from a reference leg has been documented in BW-4689 and previously discussed with the NRC. A depressurization from 2000 to 1000 psi will cause an error of approximately 1%.
Larger errors can exist for rapid depressurization to less than 600 psi, but under these conditions, pressurizer level is unimportant.
However, supplementary instruc-tions will be provided to make the operator aware of the possibility of pressurizer level indication errors following a rapid de-pressurization to pressures less than 600 psi.
For boiling to occur in the reference leg, the system presSJre must be below 300 psi and therefore need not be considered as discussed above.
Action 3. Review all safety and control setpoints derived from level signals to verify that the setpoints will initiate the action required by the plant safety analyses throughout the range of ambient tempera-tures encountered by the instrumentation, incluaing accident temperatures.
Provide a listing of these setpoints.
2167 073
R. H. Engelken September 14, 1979
RESPONSE
No reactor protection or safety feature system actions are initiated by these instruments; therefore, the error induced by the increase in the reference leg temperature need only be considered for post-accident monitoring.
During post-accident monitoring, level indication alone is not relied upon but rather system temperature and pressures are used to assure adequate core cooling and to conf.irm the adequacy of the level indications.
Although not related to reactor protection or safety features system actions, the pressurizer level instrumentation is used to deenergize the pressurizer heaters and therefore this action may have to be taken manually in the event of elevated containment temperatures.
Action 4. Review and revise, as necessary, emergency procedures to include specific information obtained from the review and evaluation of Items 1, 2 and 3 to ensure that the operators are instructed on the potential for and magnitude 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 completion date for the revisions and a completion date for operator training on the revisions.
RESPONSE
The District does not feel that revisions to procedures are required as a result of this evaluation, however, ph,t operators will be informed of the possible level indication errors.
Please advise if we can provide any additional informa*. ion, however, we cor. sider this response to complete the requirements of the subject bulletin and will take no further action unless so advised.
Sincerely yours,
} ) );t $ A ohnJIMattimoe Assistant General Manager and Chief Engineer cc: Office of Inspection and Enforcement Division of Reactor Operations Inspection 2167 074
TABLE 1 Correction to indicated water level for post-accident temperature effects of the steam generator operate level, steam generator full range level, and pressurizer level.
Correction to Reference leg temperature indicated level (%)
( F) of full span 100 2.0 150 3.0 200 5.0 250 7.0 300 9.0 350 12,0 400 15.0 Note:
The increase in reference leg temperature causes the measured level to indicate higher than actual level.
2167 075
TABLE 2 Correction to indicated water level for post-accident temperature effects on the steam generator start-t.p level.
Correction to Reference leg temperature indicated level (%)
( F) of full span 100 2.0 150 3.0 200 5.0 250 8.C 300 12.0 350 16.5 400 21.0 Note: The increase in reference leg temperature causes the measured level to indicate higher than actual level.
2167 076