ML18016A897

LER 99-004-00:on 990312,unit Trip Was Noted.Caused by Degraded Condition of SG Water Level Flow Control Valve. Replaced Positioners on All Three FW Regulating Valves.With 990408 Ltr

ML18016A897
Person / Time
Site:	Harris
Issue date:	04/08/1999
From:	Brooke Clark, Fleming C CAROLINA POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
HNP-99-0063, HNP-99-63, LER-99-004, LER-99-4, NUDOCS 9904130344
Download: ML18016A897 (9)

Text

CATEGORY 1

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REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

.ACCESSION NBR:9904130344 DOC.DATE: 99/04/08 NOTARIZED: NO DOCKET FACIL:50-400 Shearon Harris Nuclear Power Plant, Unit 1, Carolina 05000400 AUTH.NAME AUTHOR AFFILIATION FLEMING,C.W. Carolina Power &. Light Co.

CLARK,B.H. Carolina Power & Light Co.

RECIP.NAME RECIPIENT AFFILIATION

SUBJECT:

LER 99-004-00:on 990312,unit trip was noted. Caused by degraded condition of SG water level flow control valve.

Replaced positioners on all three FW regulating valves. With 990408 ltr.

DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR ENCL SIZE'-

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

NOTES:Application for permit renewal filed. 0500040d-.

RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD2-1 PD 1 1 FLANDERS,S 1 1 INTERNAL: ACRS'= 1 1 AEOD/SPD/RRAB 1 1 FILE CENTER- 1 1 NRR/DRCH/HOHB 1 1 NRR/DRCH/HQMB 1 1 NRR/DRPM/PECB 1 1 NRR/DSSA/SPLB 1 1 RES/DET/EIB 1 1 RGN2 FILE 01 1 1 EXTERNAL: L ST LOBBY WARD 1 1 LMITCO MARSHALL 1 1 NOAC POORE,W. 1 1 NOAC QUEENER,DS 1 1 NRC PDR 1 1 NUDOCS FULL TXT 1 '

NOTE TO ALL "RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE. TO HAVE YOUR NAME OR ORGANIZATION REMOVED FROM DISTRIBUTION LIST OR REDUCE THE NUMBER OF COPIES RECEIVED BY YOU OR YOUR ORGANIZATION, CONTACT THE DOCUMENT CONTRO DESK (DCDj ON EXTENSION 415-2083 FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 17 ENCL 17

Carolina Power 8 light Company Harris Nuclear Plant P.O. Box 165 New Hill NC 27562 I

Ape o a $ 99 U.S. Nuclear Regulatory Commission Serial: HNP-99-063 ATTN: NRC Document Control Desk 10CFR50.73 Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT UNIT 1 DOCKET NO. 50-400 LICENSE NO. NPF-63 LICENSEE EVENT REPORT 1999-004-00 Sir or Madam:

In accordance with 10CFR50.73, the enclosed Licensee Event Report is submitted. This report describes a condition which resulted in an automatic trip of the reactor and the automatic actuation of various Engineering Safety Features Actuation Systems.

Sincerely, B.H. Clark General Manager Harris Plant CWF/cwf Enclosure c: Mr. J. B. Brady (HNP Senior NRC Resident)

Mr. R. J. Laufer (NRC - NRR Project Manager)

Mr. L. A. Reyes (NRC Regional Administrator, Region II) 9904130344 990408 PDR ADQCK 05000400 8 PDR 5413 Shearon Harris Road New Hill NC

U. S. Nuclear Regulatory Commission

'ocument Control Desk / HNP-99-063 Page 2 of 2 bcc:

Ms. D. B. Alexander Mr. C. S. Hinnant Mr. G. E. Attarian INPO Mr. R. H. Bazemore Mr. W. D. Johnson Mr. T. C. Bell Mr. M. B. Keef Ms. P. P. Burns Mr. G. J. Kline Mr. H: K. Chernoff Ms. W. C. Langston Mr. B. H. Clark Mr. R. D. Martin Mr. W. F. Conway Mr. J. W. McKay Mr. J. M. Curley Mr. R. O. Moore Mr. G. W. Davis Mr. T. C. Morton Mr. W. J. Dorman Mr. P. M. Odom Mr. R. J. Field Mr. W. S. Orser Ms. J. L. Gawron Mr. R. M. Poulk Mr. J. W. Gurganious Mr. J. Scarola Mr. K. N. Harris Mr. F. E. Strehle Ms. L. N. Hartz Mr. J. M. Taylor Mr. J. D. Henderson Licensing File(s)

Mr. W. J. Hindman Nuclear Records

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED OMB NO. 3150.0104 EXPIRES 06/30/2001 Estimated burden per response to comply with this mandatory information (6-1998) collection request: 50 hrs. Reported lessons learned are incorporated into the licensing process and fed back to industry. Forward comments regarding LICENSEE EVENT REPORT (LER) burden estimate to the Information and Records Management Branch (T+

F33), U.S. Nuclear Regulatory Commission, Washington, DC 205554001 ~

and to the Paperwork Reduction Proiect (31504104), Office of Management (See reverse for required number of and Budget, Washington, DC 20503. If an information collection does not display a currently valid OMB control number, the NRC may not conduct or digits/characters for each block) sponsor, and a person is not required to respond to, the information collection.

FACILITYNAME (1( DOCKET NUMBER l2l PAGE (3) 1 OF 3 Harris Nuclear Plant, Unit 1 OS000400 TITLE (4)

Unit trip due to the degraded condition of a steam generator water level flow control valve.

MONTH DAY YEAR YEAR SEOUENzIAL AEvlsloN MONTH DAY YEAR FACILITYNAME DOCKET NUMBER NUMBER NUMBER 12 1999 1999 - 004' 00 1999 FACILITYNAME DOCKET NUMBER 05000 OPERATING MODE (9) 20.2201(b) 20.2203(a)(2)(v) 50.73(a) (2) 5) 50.73(a)(2)(viii)

POWER 100 20.2203(a) (1) 20.2203(a)(3)(i) 50.73(a)(2)(ii) 50.73(a)(2) (x)

LEVEL (10) 20.2203(a)(2)(i) 20 2203(a)(3)(ii) 50.73(a) (2)(iii) 73.71 20.2203(a)(2)(ii) 20.2203(a)(4) X 50.73(a)(2)(iv) OTHER 20.2203(a)(2)(iii) 50.36(c)(1) 50.73(a)(2) (v) Specify in Abstract below 20.2203(a) (2)(iv) 50.36(c) (2) 50.73(a)(2) (vii) or In NRC Form 366A NAME TELEPHONE NUMBER liireiude Area Code)

Carey W. Fleming, Principal Analyst - Licensing (919) 362-2313 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE CAUSE 'YSTEM COMPONENT MANUFACTURER REPORTABLE To EPIX To EPIX JB FCV Bailey EXPECTED MONTH DAY YEAR YES X NO (If yes, complete EXPECTED SUBMISSION DATE).

ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) (16)

At 06)39 on March 12, 1999, the unit experienced an automatic reactor trip from 100% power. The unit tripped on a high water level trip signal on one of the three steam generators. The 'hi-hi'team generator water level signal trip setpoint is 82.4 percent on the narrow range )eve( scale. Just prior to the automatic trip, the 'C'team Generator (SG) water level was observed to be increasing in an uncontrolled manner with its Feedwater Regulating Valve (FRV) in automatic control. The control room operators took manual control of the 'C'RV, in an attempt to restore water level to its normal range. The FRV response, as observed from the controller output signal and the resulting feed flow indications, was slower and not as uniform as expected. After approximately eight minutes of attempting to control the 'C'team generator water level in manual, the performance of the valve positioner degraded to the point where operator control was highly questionable. The operators had set a limit, at which they would initiate a manual reactor trip; however, an automatic trip occurred prior to their taking the planned action.

The most probable failure mechanism for the 'C'RV is a combination of a loose stroke adjustment screw on the air-operated valve positioner and a sticking pilot valve within that same positioner. This degradation in the material condition of the valve's positioner is attributed to an inadequate preventative maintenance program/schedule for the Feedwater Regulating Valves.

Corrective actions that have been taken include: 1) replacing the positioners on all three Feedwater Regulating Valves, including a verification that the stroke adjustment screw on each is not loose, and 2) Determining that similar air operated valve positioners do not have a potential for similar failures which may jeopardize plant safety or reliability. Planned corrective actions include reviewing and revising, as applicable, the preventative maintenance schedules for Bailey air operated valve positioners.

NRC FoRM 366 (6-1998I

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (6 98)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITYNAME (1) DOCKET LER NUMBER (6) PAGE (3)

Harris Nuclear Plant, Unit 1 05000400 YEAR SEQUENTIAL NUMBER REVISION NUMBER 2 OF 3 1999 004 00 TEXT fifmore spece is required. use eddidonsl copies of NRC Form 3GGAJ (17)

I. DESCRIPTION OF EVENT At 06:39 on March 12; 1999, the unit experienced an automatic reactor trip from 100% power. The unit tripped on a high water level trip signal on one of the three steam generators. The 'hi-hi'team generator water level signal trip setpoint is 82.4 percent on the narrow range steam generator water level scale. This trip signal causes the main turbine to trip and a main feedwater isolation signal to be sent to various components for the purpose of limiting any further increase in water level, thereby protecting the turbine and other main steam system components from water damage. When the plant is above approximately 10 percent power, the turbine trip signal will cause an automatic reactor trip.

Just prior to the automatic trip, the 'C'team Generator (SG) water level was observed to be increasing in an uncontrolled manner with its Feedwater Regulating Valve (FRV) (EIIS: JB, FCV) in automatic control. The control room operators took manual control of the 'C'RV, in an attempt to restore water level to its normal range. The FRV response, as observed from the controller output signal and the resulting feed flow indications, was slower and not as uniform as expected. After approximately eight minutes of attempting to control the 'C'team generator water level in manual, the performance of the valve positioner degraded to the point where operator control was ineffective (i.e., some control was available, but the control characteristics were not as expected).'he operators had set a limit, at which they would initiate a manual reactor trip; however, an automatic trip occurred prior to their taking the planned action. The reactor trip recovery proceeded normally with minor equipment deficiencies noted on some non-safety secondary systems. Following the trip, level in the 'C'team generator was reduced to its post-trip level band by a combination of 'shrink'nd steaming for decay heat removal.

The reactor was not manually tripped prior to automatic action due to the Unit Senior Control Operator (USCO) establishing a verbal limit which was only OA percent of narrow range level indication les's than the trip setpoint.

Both the investigation team and Operations management have determined that this guidance provided by the USCO was not optimal. Procedures existing at the time of the event did not specify a manual trip limit for steam generator levels for the given situation.

An investigation team devised a plan and performed testing and inspections on the 'C'RV controller, as well as inspections and comparisons of the 'A'nd 'B'alve controllers. When the 'C'RV was disconnected from its operator, an abnormal (e.g., jerky) stroke of the valve persisted on the operator (i.e., a positioner problem and not a mechanical binding of the valve itself). Disassembly of the valve revealed a loose stroke adjustment screw, which was not the case in the 'A'nd 'B'ositioners. Inspection of a pilot valve assembly in the 'C'ositioner indicated that dirt build-up and wear on the pilot valve stem. Although the pilot valve stems for 'A'nd 'B'RV positioners also experienced some wear, the investigation team still considered the wear issue to be a likely cause for the symptoms observed. The team believes that constant modulation and vibration of the 'O'RV without more frequent preventative maintenance was the cause of the positioner failure.

II. CAUSE OF EVENT The most probable failure mechanism for the 'C'RV failure is a combination of a loose stroke adjustment screw on the air-operated valve positioner and a sticking pilot valve within that same positioner. The sticking pilot valve may have been caused by excessive wear or dirt build-up. This degradation in the material condition of the valve's positioner is attributed to an inadequate preventative maintenance program/schedule for the Feedwater Regulating Valves.

NRC FORM 366 (6-98)

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (6 96)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITYNAME (1) DOCKET LER NUMBER (6) PAGE (3)

Harris Nuclear Plant, Unit 1 05000400 YEAR SEQUENTIAL NUMBER REVISION NUMBER 3 OF 3 1999 004 00 TEXT IIfmore spsce Is required, use eddidonel copies of NRC Form 366AI (17)

III. SAFETY SIGNIFICANCE There were no actual safety consequences as a result of this event. All systems required to limit an overfeeding event of a steam generator, as enumerated in FSAR sections 15.1.2 and 15.0.8, remained operable throughout the event. Additional features available to protect the unit from an over-feeding event are the overtemperature and overpower delta-T reactor trips. These features remained available, and were not challenged on the March 12, 1999 trip. No safety limits were exceeded and the event neither initiated nor exacerbated any radiological releases.

This report is being submitted pursuant to the criteria of 10 CFR 50.73(a)(2)(iv) for an unplanned automatic actuation of the Reactor Protection System (RPS) and the unplanned, automatic Engineered Safety Features (ESF) actuations as follows:

A) The automatic trip of the turbine and feedwater isolation signal from the "hi-hi steam generator level trip signal" (P-14);

B) The automatic reactor trip, which is required following a turbine trip with the plant power above the P-7 setpoint (approximately 10 % power);

C) The automatic start of the two motor driven auxiliary feedwater pumps, indirectly.due to the feedwater isolation signal (i.e., loss of both running main feed pumps on the feedwater isolation signal);

D) The automatic start of the turbine driven auxiliary feedwater pump on the low-low steam generator levels in the 'A'nd 'B'team generators, due to the main feed isolation signal and the expected water level

'shrink'ollowing the trip.

IV. CORRECTIVE ACTIONS Corrective actions that have been taken:

1. Replaced the positioners on all three Feedwater Regulating Valves, including a verification that the stroke adjustment screw on each is not loose.

2. Determined that similar air operated valve positioners do not have a potential for similar failures which may jeopardize plant safety or reliability.

Planned corrective actions:

3. 'Review and revise, as applicable, the preventative maintenance schedules for Bailey air operated valve positioners (by 8/31/99).

V. SIMILAR EVENTS This is the first trip of the Harris Plant involving a 'hi-hi'team generator water level. Harris Plant LER 97-001-00 describes a reactor trip on low steam generator water level, due to degraded equipment on the positioner of a major valve in the main feedwater system. This LER is distinguishable from that event in that the manufacturer and make of the valves are very dissimilar. Corrective actions from that event had no effect on the FRVs for steam generator water level control.

Investigations following the trip did not reveal any industry operating experience which would be directly on point with this occurrence.

NRC FORM 366A (6-96)

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