Text

LER 97-004-01:on 970327,plant Operators Manually Scrammed Reactor as Required by TS Due to Indication of Entry Into Region a of power-to-flow Map.Caused by Inadequate Attention to Detail.Established Event Evaluation teams.W/970924 Ltr
	ML17292B064
Person / Time
Site:	Columbia
Issue date:	09/24/1997
From:	Arbuckle J, Bemis P; WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
	GO2-97-175, LER-97-004, LER-97-4, NUDOCS 9710010084
	Download: ML17292B064 (9)
	v • d • e

CATEGORY 3y

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REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS) j~ v ACCESSION NBR:9710010084 DOC.DATE: 97/09/24 NOTARIZED: NO DOCKET t FACIL:50-397 WPPSS Nuclear Project, Unit 2, Washington Public Powe 05000397 AUTH.NORE . AUTHOR AFFILIATION ARBUCKLE,J.D. Washington Public Power Supply System BEMIS,P.R. Washington Public Power Supply System RECIP.NAME RECIPIENT AFFILIATION

SUBJECT:

LER 97-004-01:on 970327,plant operators manually scrammed reactor as required TS due to indication of entry into Region A of power-to-flow may.Caused by inadequate attention to detail. Established event evaluation teams.W/970924 ltr.

DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR ENCL SIZE:

TITLE: 50.73/50.9 Licensee Event Report (LER), Incident Rpt, etc.

NOTES:

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N NOTE TO ALL "RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE. TO HAVE YOUR NAME OR ORGANIZATION REMOVED FROM DISTRIBUTION LISTS OR REDUCE THE NUMBER OF COPIES RECEIVED BY YOU OR YOUR ORGANIZATION, CONTACT THE DOCUMENT CONTROL DESK (DCD) ON EXTENSION 415-2083 FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 25 ENCL 25

WASHINGTON PUBLIC POWER SUPPI.Y SYSTEM PO. Box 968 ~ Richhtttil, 1Vttshitttltoll 99352-0968 September 24, 1997 G02-97-175 Docket No. 50-397 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555 Gentlemen:

Subject:

NUCLEAR PLANT WNP-2, OPERATING LICENSE NPF-21 LICENSEE EVENT REPORT NO. 97-004-01 Transmitted herewith is revised Licensee Event Report No. 97-004-01 for WNP-2. This report is submitted pursuant to 10 CFR 50.73 and discusses the items of reportability, corrective action taken, and action to preclude recurrence.

Should you have any questions or desire additional information regarding this matter, please call me or P.J, Inserra at (509) 377-4147.

~estectfully, P.

Vice President, Nuclear Operations Mail Drop PE23 Enclosure CC: EW Merschoff - NRC RIV KE Perkins, Jr. - NRC RIV, Walnut Creek Field Office TG Colburn - NRR NRC Senior Resident Inspector - 927N DL Williams - BPA/399 PD Robinson - Winston k, Strawn 97i00i0084 970924 050003'tt7 lllllllllilllilllil'IIIII)llllilllllll PDR ADOCK S PDR

LICENSEE EVENT REPORT (LER)

EAOLITY NAIIK(11 tACK (Sl Washington Nuclear Plant Unit 2 0 5 0 0 0 3 9 7 1 of 6 TITIK tai TECHNICAL SPECIFICATION REOUIRED MANUALSCRAM DUE TO INDICATION OF ENTRY INTO REGION A OF THE POWER-TO.FLOW MAP IKR IASIIRR tel RttORT OATS 1st OTHER EACAEERS etvOIYKO Itn SKOVKNTIAL EACKSTY RAISES oocxKT IARISKRISI IRRISKR N/A 0 5 0 0 0 03 27 97 97 0 0 4 0 1 09 24 97 0 5 0 0 0 OtOIATSIO TISS RttORT IS SVSMITTKO tURSUANT TO TI% RKOUtRKMKNTS OF to CER: 111)

MODE tnt I 20.402(b) I 20.405c I X I 50.73(s)(2)(iv) I 73.71(b)

I 20.405(s)('1)(i) I 60.38(cl(1) I 50.73(e)(2)(v) I 73.71(c) tOWOI LKYKL I 20.405(s)(1)(ii) I 50.38(c)(2) I 60.73(s)(2)(vii) I OTHER ISpecify bI Abstract bdow snd tta in Text NRC Form 388A) 0 4 6 I 20.405(s)(1)(i'n ) I 50.73(s)(2)(i) I 50.73(s)(2)(viii)A 20.405(s) (1)(iv) I 60.73(s)(2)(ii) I 50.73(s)(2)(vtsi)B I 20.405(s)(1)(v) 50.73(s)(2) (iii) I 50.73(s)(2)(xl IICKNKEK CONTACT TOII KISS LKR IIKI IttttttONK NUMSKR J.D. Arbuckle, Licensing Technical Specialist AREA COOK 509 377-4601 COMtLKTK OIC LSC EOR KACN MFWCIRTIT FALImK OKSCRRKTI et Tice RttoRT tlat RttORTASIK RttORTAIKK To Ievlos To Ievtos Sl LKIAMKIETALRKFORT KXFKCTTss I sat EXPECTED SUBMISSION YES (if yes, complete EXPECTED SUBMISSION DATE) IXI NO DATE (15)

ABSTRACr (l6)

On March 27, 1997 at 0907 hours0.0105 days 0.252 hours 0.0015 weeks 3.451135e-4 months with the plant in Mode 1 at 46 percent power, plant operators manually scrammed the reactor as required by the Technical Specifications due to indication of entry into Region A of the power-to-flow map during Reactor Recirculation (RRC) and Reactor Feedwater (RF%) System testing.

gt was determined by followup engineering analyses that Region A of the power-to-flow map had not been entered.) The post modif-ication testing was heing performed to demonstrate that the feedwater level control system and recirculation flow runback feature would avoid a scram following the trip of a single reactor feedwater pump during power operation.

During the test and after the planned trip of pump RFW-P-1B, an expected reactor recirculation pump speed runback to 27 Hz was observed. Following this planned runback, a second unexpected runback to 15 Hz occurred which, based upon evaluation of plant heat balance data, appeared to have placed the reactor into Region A of the power-to-flow map. The second runback was caused by a reactor recirculation pump differential temperature cavitation interlock condition.

PIant operators responded conservatively by immediately scramming the reactor and maneuvering the plant to )

a safe shutdown condition in accordance with procedures.

The root cause of this event was inadequate attention to detail and a lack of a questioning attitude during the design phase of the modifications for the adjustable speed drive and digital feedwater level control systems.

Corrective actions included the establishment of internal and external event evaluation teams to investigate the event, and successfully repeating the reactor feedwater pump trip test.

There were no indications of power oscillations during the event period. This event posed no threat to the health and safety of the public or plant personnel.

LICENS VENT REPORT (LER) TEXT COhijIUATION ACILlTY MME (1) DOCKFT MUMBER (2)

VfAR sfotrOrRAL NVMIOI Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 97 -004 0 1 OF TEXT (17)

EvntD ri i n On March 27, I997 at 0907 hours0.0105 days 0.252 hours 0.0015 weeks 3.451135e-4 months with the plant in Mode I at 46 percent power, plant operators manually scrammed the reactor as required by the Technical Specifications due to indication of entry into Region A of the power-to-flow map during Reactor Recirculation (RRC) System [AD]

and Reactor Feedwater (RFW) System [JB] testing. (It was determined by followup engineering analyses that Region A of the power-to-flow map had not been entered.)

At the time of the event, post-modification testing of the digital feedwater level control and RRC adjustable speed drive systems was being conducted to demonstrate that the level control system and recirculation fiow runback feature would avoid a scram following the trip of a single feedwater pump during power operation.

During the test and after the trip of pump RFW-P-1B [P], a reactor recirculation pump speed runback to 27 Hz was observed as expected. Following this runback, a second unexpected runback to iS Hz occurred which, based upon evaluation of plant heat balance data, appeared to [

have placed the reactor into Region A of the power-to-flow map. The second runback to 15 Hz was caused by a reactor recirculation pump differential temperature cavitation interlock condition existing for greater than 15 seconds.

As expected in response to the manual scram, reactor vessel water level decreased to just below

+13 inches. In response to the low level condition, reactor feedwater flow increased and level was recovered to above +13 inches. During this evolution and at approximately +18 inches, pump RFW-P-1A P'] speed failed to decrease in response to the increasing level control signal.

As a result, reactor vessel water level reached +54 inches (Reactor Vessel Water Level - High:

Level 8) and the pump tripped on high level.

Plant operators successfully returned both reactor feedwater pumps to operation and restored level to within normal limits.

Immediate Corrective Action Plant operators maneuvered the plant to a safe shutdown condition in accordance with procedures.

Further Evaluation

1. Plant operators made a conservative decision to manually scram the reactor when plant heat balance data indicated that Region A of the power-to-flow map had been entered.

However, based upon a followup review of plant data it was determined by Reactor Engineering personnel that, although close, Region A of the power-to-flow map had not been entered during the event. A review of Average Power Range Monitor [IG] and heat flux data confirmed that conclusion.

LICENSE ENT REPORT (LERI TEXT CONOIUATION FACILITY NANE {1) DOCKET NURSER {2)

Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 97 -004 0 1 OF TEXT {17)

2. Based upon the followup conclusion that Region A of the power-to-flow map had not been entered, this event is being reported pursuant to 10 CFR 50.73(a)(2)(iv) as, "Any event or condition that resulted in a manual or automatic actuation of any Engineered Safety Feature, including the Reactor Protection System."

This event was originally determined to be reportable as a four-hour Reactor Protection System actuation in accordance with 10 CFR 50.72(b)(2)(ii). The initial notification to the NRC Operations Center was made on March 27, 1997 at 1013 hours0.0117 days 0.281 hours 0.00167 weeks 3.854465e-4 months (66 minutes after the event). Upon further review, it was determined that the event should have also been reported within one hour as an initiation of a plant shutdown required by the Technical Specifications in accordance with 10 CFR 50.72(b)(1)(i)(A).

Technical Specification 3.4.1, "Recirculation Loops Operating," requires that the reactor mode switch be placed in the shutdown position within 15 minutes of operation in Region A of the power-to-flow map. On March 27, 1997 at 1308 hours0.0151 days 0.363 hours 0.00216 weeks 4.97694e-4 months , the NRC Operations Center was notified of the determination that the event was also reportable within one hour.

However, since Region A of the power-to-flow map was not actually entered, this event is reportable pursuant to 10 CFR 50.72(b)(2)(ii).

3. Plant design includes an interlock to reduce RRC pump drive flow during conditions which could cause cavitation in the RRC system. A low temperature differential between the RRC pump suction and the vessel steam dome indicates a reduction in subcooling of reactor coolant in the reactor vessel annulus region. A reduction in RRC drive flow under these conditions prevents system cavitation and avoids equipment damage from prolonged vibration.

Plant design at the time of the event was such that a reduction in drive flow would be initiated if the minimum differential temperature of 9.9 degrees Fahrenheit was exceeded for more than 15 seconds. If this occurred, the control logic would automatically reduce drive speed to 15 Hz, and RRC pump flow would be reduced to 25 percent.

4. Following the planned RRC runback to 27 Hz, evaluation of plant data validated that a differential temperature condition of greater than 9.9 degrees Fahrenheit was present for longer than 15 seconds, and this condition caused the RRC runback to 15 Hz as designed.

5. The differential temperature cavitation interlock was not expected to exist for greater than 15 seconds. The digital feedwater level control and adjustable speed drive systems were previously modeled in the plant simulator and in simulation programs used by General Electric and the Supply System. There was no previous simulator indication that the cavitation interlock would be initiated due to a single reactor feedwater pump trip from 100 percent reactor power.

LICEN EVENT REPORT (LER) TEXT CO INUATION FACILlTY kAHE (1) DOCKET kUHBER (2)

SEOUENNL IIVMOOI Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 97 -004 0 1 4 oF TEXT (17)

6. The digital feedwater level control and adjustable speed drive systems performed as designed to the planned trip of the reactor feedwater pump up to the point of the second runback to 15 Hz, Following the manual scram, the digital feedwater level control system did not respond as designed to control vessel level below the Reactor Vessel Water Level-High (Level 8) setpoint of +54 inches.

Root au es

+Runback The root cause of this event was inadequate attention to detail and lack of a questioning attitude that led to acceptance of previously known design information without in-depth challenge during the design phase of the adjustable speed drive and digital feedwater level control system modifications. The integrated effect of these modifications, in conjunction with implementation of power uprate one year earlier, was not considered in relation to the cavitation interlock protection setpoint. This resulted in failure to identify the potential for activation of the differential temperature cavitation interlock.

Level-8 The root cause of the failure of the digital feedwater level control system to maintain reactor water less than +54.5 inches (Level-8) was attributed to a design deficiency with the governor controller. This deficiency has existed since the original plant design and it was anticipated that installation of the digital system would correct the problem. Following extensive analyses, testing and intrusive investigations, it was concluded that the only effective means of avoiding vessel overfill was improvement in the responsiveness of the governor controller following a scram. The required changes are being evaluated.

Fu her orrective Ac ion An event evaluation team, composed primarily of Supply System personnel, was established to investigate the event and present the results and conclusions to plant management. Specific areas evaluated by the team included analytical results of testing performed, performance of the on-shift Operations personnel during this event, performance of the digital feedwater level control and adjustable speed drive systems, the adequacy of the test procedure as it relates to the conditions surrounding the scram, and the adequacy of the design related to the installation of the adjustable speed drive and digital feedwater level control systems. As a result of the evaluation, several recommendations were developed. These recommendations have been reviewed and are being implemented as part of the Problem Evaluation Request process.

A second independent evaluation team, consisting of primarily non-Supply System personnel, evaluated the event and performed a critical review of the investigation conducted by the Supply System event evaluation team. The independent team validated the findings of the event evaluation team.

LICENS VENT REPORT (LER) TEXT CO UATION FACILiTY NAME (1) DOCKET HUNBER (2) KOVWAAL Washington Nuclear Plant - Unit 2 05 0 0 0 3 9 7 97 004 0>5 oP 6 TEXT (17)

The cavitation interlock setpoint and methodology were reviewed to determine if any runback value or cavitation logic changes were necessary. As a result of this review, the differential temperature cavitation interlock setpoint was changed from 9.9 degrees Fahrenheit to 10,7 degrees Fahrenheit. In addition, the differential time delay to runback was increased from 15 seconds to ten minutes.

To demonstrate that the feedwater level control system and recirculation flow runback feature would prevent a scram following the trip of a single feedwater pump during power operation, the was successfully repeated on July 23, 1997 when the plant returned to stable, full power

'est operation following the R-12 Maintenance and Refueling Outage. The systems responded as designed.

An assessment of the Level-8 design function was performed. Based on this assessment, it was concluded from the evaluation that a post-scram, Level-8 trip of this type does not represent a safety concern.

As men of fe on uences The probability of thermal-hydraulic oscillations is greatly increased if the plant is operating in Region A of the power-to-flow map. Plant operators made a conservative decision to manually scram the reactor when plant heat balance data indicated that Region A of the power-to-flow map had been entered. This action was taken well within the 15 minute time-frame allowed by the Technical Specifications. Subsequent operator actions were prompt and correct to maneuver the plant to a safe shutdown condition.

Upon further evaluation it was determined that, although close, the plant had not actually entered Region A of the power-to-flow map. There were no indications of power oscillations during the event period. Accordingly, the event did not represent a safety concern.

An evaluation was performed to determine the impact of continued operation of the plant without a final design resolution of this problem. It was concluded from the evaluation that a post-scram, Level-8 trip of this type does not represent a safety concern. The post scram response is not dissimilar to what would have been seen with the previous analog system. Since the Level-8 trip was reached post-scram, there was no adverse impact on the fuel thermal limits. For long-term cooling and inventory make-up, the High Pressure Core Spray [BG] and Reactor Core Isolation Cooling [BN] Systems would be available once the water level lowered to their initiation setpoints. Therefore, the transients in the FSAR are still bounding and, as a result, the consequences of an accident as analyzed in the FSAR were not increased.

Although additional operator actions may be required to preclude or recover from a Level-8 trip, plant operators are trained that a Level-8 trip following a scram may occur. This training includes actions that can be taken to either avoid the trip or respond accordingly should a trip occur.

LlCENSE VENT REPORT (LER) TEXT CO UATION

)ACILITY HN(E (1) OOCKET NlNBER (2)

YEAR 'SEOVENEIAl, ' RQAQSN NVMSOI, . NVMSOI Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 97 -004 0 1 TEXT (l7) imil Even Licensee Event Report 96-004 reported a problem involving a manual reactor scram due to a reactor water level transient during testing of the digital feedwater system. After performing a step of the test procedure which reduced reactor water level by six inches, level continued to decrease past the intended value. The feedwater level control system responded to control vessel level below the Level-8 setpoint during this event (a Level-8 trip signal was not received).

The cause of this event was determined to be a manufacturer's programming error in the digital feedwater system testing software which caused mismatches to occur in the feed/steam signal, resulting in unwanted level step changes. The software was modified by the vendor to delete the possibility of unwanted step changes. The software change was verified and validated and functionally tested on the plant simulator prior to installation in the plant.

Corrective action taken in response to the previous event would not have been expected to preclude this event. The action was designed to correct a specific software problem that was causing mismatches in the feed/steam signal.

ML17292B064

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