ML19209B791

Responds to IE Bulletin 79-21 Re Temp Effects on Level Measurements.Liquid Level Measuring Sys for Steam Generator, Pressurizer & Core Flood Tank Have Been Examined.Tables Re post-accident Generator Startup & Liquid Level Encl

ML19209B791
Person / Time
Site:	Davis Besse
Issue date:	09/20/1979
From:	Crouse R TOLEDO EDISON CO.
To:	James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
References
NUDOCS 7910100476
Download: ML19209B791 (7)

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6 TOLEDO Docket No. 50-346 License No. NPF-3 Serial 1-90 September 20, 1979 Mr. James G. Keppler Regional Director, Region III Office of Inspection and Enforcement U. S. Nuclear Regulatory Commission 799 koosevelt Road Glen Ellyn, Illinois 60137

Dear Mr. Keppler:

Attachment A is Toledo Edison's response to IE Bulletin 79-21 for the Davis-Besse Nuclear Power Station, Unit 1, coverir.g temperature effects on level measurements.

Very truly yours, ffh= = '-

Richard P. Crouse Vice President - Energy Supply RPC/TJM cc:

Director Division of Reactor Operations Inspection Office of Inspection and Enforcement U. S. Nuclear Regulatory Commission Washington, D. C. 20555 SEP 251979 1910100 h 76 THE TOLEDO EDISON COMPANY EDISON PLAZA 300 MADISON AVENUE TOLEDO. OHIO 43652 1124 021

Docket No. 50-346 License No. NPF-3 Serial No. 1-90 September 20, 1979 Page 1 Response to IE Bulletin 79-21 Davis-Besse Nuclear Power Station Temperature Effects on Level Measurement 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 description of the type of reference shall be included, i.e. , open column or sealed reference leg.

RESPONSE

The liquid level measuring systems frr the steam generator, pressurizer, and core flood tank have been examined. These measuring systems are of the dalta pressure, open column, uninsulated reference leg type.

The pressurizer water level is monitored to provide an indication of reactor coolant inventory, to control the makeup, and is used for post-accident monitoring. The Core Flood Tank water level is monitored to provide an ima :atien during normal operation that an adequate supply of water is available for .a acci-dent as required by Tech Specs. However, Core Flood tank level indication is 20t considered post-accident monitoring instrumentation. The Steam Generator water level is monitored to provide an indication and a control of the steam generator water inventory below 15% power. The level is monitored in three ranges:

" Start-up", " Operating" and " Full Range". The " Start-up Range" is also used for the following safety functions:

1. Start Auxiliary Feedwater on a partial or total loss of Main Feedwater.

2. Control the water level in the unaffected Steam Generator af ter a steam line or feedwater line break inside the containment.

3. Control the w.!ter level in the Steam Generctors after a small loss of coolant accident.

4. Post-Accident Monitoring.

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Docket No. 50-346 License No. NPF-3 Serial No. 1-90 September 20, 1979 Page 2 Action 2. On those systems described in Item 1 above, evaluate the ef fect of post-accident ambient temperatures on the indicated water level to determine any change in indicated level relative to actual water level. This evaluation must include other sources of error including the effects of varying fluid pressure and flashing of reference leg to steam on the .ater level measurements. The results of this evaluation should be presented in a tabular form similar to Tables 1 and 2 of Enclosure 1.

RESPONSE

The effect of the reference leg temperatures on the level measurement instrumentation systems identified above in lisced in Tables 1 and 2. The errors listed have been maximized to account 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 effervescence of soluble gases.

For the core flood tanks, the level indication would be af fected by elevated temperature, but the safety function of the tank is not affected by level indication, and tank discharge can be readily confirmed by pressure indication alone.

Steam generator level measurements are not significantly affected by the effervescence of soluble gases because there is insufficient 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.

This depressurization will not occur on both steam generators at Davis-Besse af ter a steam line break as the steam generators are both isolated by the Steam and Feedwater Rupture Control System and only one steam generator will be dep essurizec by the break. The repressurization of the steam generator will refill the reference teg and the errors would be no greater than those listed in the Tables.

The pressurizer level could be affected by the ef fervescence of soluble gases.

The ejection of water from a reference leg has been documented in BW-4689 and pre-viously discussed with the NRC. A depressurization from 2000 to 1000 psi will cause an error of approximately 1% cr about 3.2 inches. Larger errors can exist for rapid depressurization to less than 600 psi, but under these conditions, pressurizer level is unimportant. However, supplementary instructions will be provided to make the operator aware of the possibility of pressurizer level indication errors following a rapid depressurization to pressures less than 600 psi. For boiling to occur in the reference leg, the system pressure must be below 300 psi and therefore necd not be considered as discussed above.

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Docket No. 50-346 License No. NPF-3 Serial No. 1-90 September 20, 1979 Page 3 Action 3. Review all safety and control setpoints derived from level setpoints to verify that the setpoints will initiate the action required by the plant safety analyses throughout the range of ambient temperatures.

Provide a listing of these setpoints.

RESPONSE

The steam generator start-up range level is the only safety ralsted level instrumentation inside the containment vessel which initiates safety actions at Davis-Besse. The initiating action will not be affected by partial or total loss of Main Feedwater flow as there are no temperature transients on the instrumentation resulting f rom this event.

These low steam generator level instruments do not provide any initiating function for main feedwater or main steam line breaks inside of containment. Low steam generator pressure witches mounted outside of containment are used to protect against these breaks inside of containment.

No reac*or protection system (RPS) reactor trips are initiated by steam generator start-up r;nge level instruments; therefore, the error induced by the increase in the reference leg temperatute need only be considered for post-accident monitoring and steam generator level control. During post-accident r7nitoring, level indication alone is not relied upon but rather system temperature and pressures are used to assure adequate core cooling and to confirm the adequacy of the level indications. The steam generator dual level control system is used to control the speed of the auxiliary feedpumps which in turn controls the water level in the steam generators.

The highest temperature whleh will affect the steam generator level instru-mentation will occur during a steam line break inside of containment. The unaffected steam generator would be used to remove decay heat from the Reactor Coolant system, and its start-up level instrumentation will be affected.

The water level control associated with the start-up level ir.atrumentation is as follows:

With a " 41600 psig Reactor Coolant Pressure" Safety Features Actuation System (SFAS) trip and nct forced Reactor Coolant System circulation, the system controls the steam generator level at 120 inches of 5500F water.

Without a "dc1600 psig Reactor Coolant Pressure" SFAS trip and no forced Reactor Coolant circulation, the s'mtem controls the steam generator level at 35 inches of 5500F water.

If there is forced circulation in the Reactor Coolant System, the system controls the steam generator level at 28 inches.

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Docket No. 50-346 License No. NPF-3 Serial No. 1-90 September 20, 1979 Page 4 Action 3 - Response (continued)

The pressurizer level instrumentation is used as Post-Accident Monitoring instrumentation and to de-energize the pressurizer heaters on low pressurizer level and therefore this level may have to be controlled manually in the event or elevated containment temperatures. The low level pressurizer heater cutout la 40 inches.

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 Emergency Procedures which involve the steam generators start-up level, and pressurizer level will be revised as required to ensure that the operators have the instructions to be able to compensate for potentially erroneous level signals so that they can manually control these levels properly.

The necessary revisions to the procedures and the operatt training on those revisions will be completed by October 31, 1979.

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Docket No. 50-346 License No. NPF-3 Serial No. 1-90 September 19, 1979 TABLE 1 Correction to indicated water level for post-accident temperature effect ' of the steam generator operate levcl, steam generator full range level, pr+3sutizer level, and core flood tank level (for tanks with water reference legs).

Correction tc Reference leg temperature indicated level N)

(OF) 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.

L 1124 026

Docket No. 50-346 License No. NPF-3 Serial No. 1-90 September 19, 1979 TABLE 2 Correction to indicated water level for post-accident temperature effects on the steam generator start-up level.

Correction to Reference leg temperature indicated level (%)

(cF) of full span 100 2.0 150 3.0 200 5.0 250 8.5 300 12.0 350 16.5 400 21.0 NOTE: The increase in reference leg tenperature causes the measured level to indicate higher than actual level.

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