ML19305C402

Revised LER 79-105/01T-2:on 791109,during Review of IE Bulletin 79-21 Response,Low Level Trip Setpoints to Steam & Feedwater Rupture Control Sys Channels 1 & 2 Exceeded Tech Specs.Caused by Setpoint Determination Method Error

ML19305C402
Person / Time
Site:	Davis Besse
Issue date:	11/21/1979
From:	Trautman D TOLEDO EDISON CO.
To:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
Shared Package
ML19305C392	List: ML19305C390 ML19305C402
References
LER-79-105-01T, LER-79-105-1T, NUDOCS 8003280496
Download: ML19305C402 (4)

Text

U.S. NUCLEAR REGULATORY CoMMISSfoN LRC FORM 3G6

• LICENSEE EVENT REPOR1 (PLEASE PRINT OR TYPE ALL REQUIRED INFORMATION)

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@] DATE 1 l 1 ] 3 l M9 l 7 l 9 ]@l1 80l 1 l 2 l 1 l 74 REPORT DATE 7 8 60 61 DOCKET NUM8EH EVENT DESCRIPTION AND PROBABLE CONSEQUENCES h1400 hours during rev lo l 3 l l On 11/9/79 at Steam and Feedwaterg

,o g ,l covered that our steam generator (SG) low level trip setpoints to i

gn,,, { Rupture Control System (SFRCS) Channels 1 and 2 were less conservative t This is being reported in accordance with T.S. 6.9.1.8.h. A Mode 3l l o i s i l of T. S. 3. 3. 2.2.

0 6 l restraint was placed on the unit until new values were determined and adjustments in all cases requiring SFRCS actua-Safety evaluation results have shown that g l made.

@ tion from SG level, none is found to cause a safety concern. (NP-32-79-12) l 80 gl 8 9 S E COMPONENT CODE SUSCbOE C DE CODE SUSC E

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, g o l l The cause of the occurrence was an error in the method used in determining the The approach failed to take into I 1 g ,j l steam generator low level bistable setpoints.

l g,,,,l account the actual pressures and temperatures as seen by the steam generat On 11/12/79, the adjustments were made, and ST 5031.14}

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TOLEDO EDISON COMPANY DAVIS-BESSE NUCLEAR POWER STATIC! UNIT ONE SUPPLEMENTAL INFORMATION FOR LER NP-32-79-12 DATE OF EVENT: November 9, 1979 FACILITY: Davis-Besse Unit 1 IDENTIFICATION OF OCCURRENCE: Steam Generator low level trip setpoints found to be less conservative than assumed.

Conditions Prior to Occurrence: The unit was in Mode 5, with Power (MWT) = 0, and Load (Gross MWE) = 0.

Description of Occurrence: On November 9, 1979 at 1400 hours0.0162 days <br />0.389 hours <br />0.00231 weeks <br />5.327e-4 months <br />, during the review cf NRC ISE Bulletin 79-21 response, it was discovered that our steam generator low level trip setpoints to Steam and Feedwater Rupture Control System (SFRCS)

Channels 1 and 2 were less conservative than the limits of Technical Specification 3.3.2.2. The steam generator low level bistable trip setpoint was 23 inches + 2 inches. However, the re-evaluation of the method used to determine the setpoints based on actual versus indicated level found that the actual level would be less conservative than the minimum 20 inches of water above the lower tube sheet as required by Technical Specifications.

As a result, a Mode 3 restraint was placed on the unit until new values could be determined and adjustments to the bistable trip setpoint made. This occurrence is being reported in accordance with Technical Specification 6.9.1.8.h which re-quires prompt notification with a followup report within two weeks for an error in analysis, which would have permitted reactor operation in a manner less conser-vative than assumed.

4 Designation of Apparent Cause of Occurrence: The apparent cause of the occurrence was found to be an error in the method used in determining the steam generator low level bistable setpoints. The approach failed to take into account the actual pres-cures and temperatures as seen by the steam generator level transmitter during cperation and in transients.

(I Analysis of Occurrence: Results of the safety evaluation performed have shown that  !

) none of the cases requiring SFRC3 actuation from steam generator level is found to I cause a safety concern, See attached copy of Safety Evaluation.

Corrective Action: On November 17, 1979, new steam generator low level bistable trip setpoints were established and adjustments were made. All four channels were

, proven operable by performance of ST 5031.14, "SFRCS Monthly Stirveillance Test". The mode restraint relating to this event was removed as of 1755 hou s on November 19, 1979.

Failure Data: There have been no previous similar occurrences.

' LER #79-105 J

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Safety Evalistion for Licensee Event Report 79-105 on Steam Generator Low level SFRCS Trip Setpoints The reason for this reportable occurrence is that the steam generator (SG) low level (LL) trip setpoints to the steam and feedwater rupture control system (SFRCS) weze found to be less conservative (lower) than the limits of Technical Specification 3.3.2.2.

The following cases are analyzed for safety evaluation:

I. SG Underfed Case A - At low powers (<25%) and with an underfeed cond? tion, the resulting SG LL will trip the SFRCS, which will start the auxiliary feedwater pumps and trip the turbine and reactor. At low powers

(<25%), the ICS using the startup feedwater control valve should control the SG level at the LL limit. If either SG is underfed, the SG level will fall until the LL trip of SFhCS occurs. With forced circulation in the RCS and only decay heat generation, a SG can remove the heat from the RCS until the SG becomes essentially dry. As shown by the B&W analysis, " Evaluation of transient behavior and small reactor coolant system breaks in the 177 EA plant (raised loop), Vol III, Fig 4.5-1, the plant can withstand up to a 30 minute delay in AFW initiation without causing core uncovery.

Therefore, even though the low level trip setpoint was set at a non-conservative low value, any delay in the low level trip (assumed 2

here to be the primary protective function) would have been negligible as compared to 30 minute.

Case B - For higher powers (>25%), the reactor would trip due to high RCS pressure and then only decay heat is left to be removed.

In this case, the low SG 1evel will trip the SFRCS if the ICS and/or the startup feedwater control valve fails to maintain the SG leve3 at the LL limit. Again, time to trip the SFRCS should be much less than 30 minutes.

Case C - If either SG is underfed and a loss of forced circulation occurs at the same time, the loss of four reactor coolant pumps will cause a immediate direct trip of the reactor and SFRCS. In this case, the low SG 1evel trip of SFRCS is not required to start the auxiliary feedwater or trip the reactor.

II. Main Feedwater Line Break Case A - Break Upstream of the Feedwatar Check Valve - Such a break will be located anywhere upstream of the main feedwater check valve. In such 1 case, the reverse AP across the check valve would j trip the SFRCS o start the auxiliary feedwater pumps and trip the i reactor.

f Case B - Break Downstream of the Feedwater Check Valves - This is analyzed in response to NRC Question 15.2.12 in the Davis-Besse Nuclear Power Station Unit No. 1 FSAR.

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In this case, the reverse AP across the fu Jwater check valve feeding into the healthy steam generator will trip the SFRCS one second af ter the break and start the auxiliary feedwater pumps and trip the reactor.

As a backup, even if LL trip was called upon to start the auxiliary feedwater pumps, it takes less than one second to reduce the steam generator level from 5 ft. (60") to 0". This is shown in Figure 15.2.12-2 in the FSAR. Hence, the low level limit anywhere between 0" and 60" would have caused a delay of less than one second to start the auxiliary feedwater pumps. As shown in this diagram, the LL trip occurs within the first 5 seconds of the transient.

III. Steam generator goes completely dry without initiating auxiliary

_feedwater - In case the steam generator goes dry for lack of any feedwater, we will have enough superheated steam at 1050 psig (lowest steam code safety valve setpoint) in the steam generator to start the auxiliary feedwater pumps. This has been shown in a previous letter to the NRC, Serial No. 522, dated June 23, 1979.

IV. Main steam line break - Steam generator low level trip is not used to trip the SFRCS for this kind of a break. Instead, low pressure is the prime protective trip of the SFRCS, which will start the auxiliary feedwater pumps on the unaffected steam generator and 2 trip the reactor.

CONCLUSION As shown in all the above cases, none is found to cause a safety concern.

In all the cases, once the auxiliary feedwater has been f, tarted,.the steam generator level would have been r .nually controlled at >35 inches on the startup range.

In summary:

1. The probability of occurrence or the consequence of an accident or malfunction of equipment important to safety previously evaluated in the FSAR has not been increased.

2. The possibility of an accident or a malfunction of a different type that is not bounded by the previous analysis in the FSAR has not been created.

3. The margin of safety as defined in the basis for any Davis-Besse Unit 1 Technical Specification has not been decreased.

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