ML18009A962

LER 91-010-00:on 910603,automatic Reactor Trip Occurred Due to Pressure Spike in Flow Transmitter Ref Leg for Reactor Coolant Loop Low Flow.Caused by Perturbation in Sensing Lines.Rcs Flow Calibr Procedures revised.W/910703 Ltr

ML18009A962
Person / Time
Site:	Harris
Issue date:	07/03/1991
From:	Hamby M, Richey R CAROLINA POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
HL-910096-(O), LER-91-010-01, LER-91-10-1, NUDOCS 9107100289
Download: ML18009A962 (7)

Text

ACCELERATED DISTRIBUTION DEMONS7PWTION SYSTEM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:9107100289 DOC.DATE: 91/07/03 NOTARIZED: NO DOCKET FACIL:50-400 Shearon Harris Nuclear Power Plant, Unit 1, Carolina 05000400 AUTH. NAME AUTHOR AFFILIATION HAMBY,M.R. Carolina Power & Light Co.

RICHEY,R.B. Carolina Power & Light Co.

RECIP.NAME RECIPIENT AFFILIATION

SUBJECT:

LER 91-010-00:on 910603,automatic reactor trip occurred due to pressure spike in flow transmitter ref leg for reactor D coolant loop low flow. Caused by perturbation in sensing lines.RCS flow calibr procedures revised.W/910703 ltr.

DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR ENCL I SIZE:

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

NOTES:Application for permit, renewal filed. 05000400 A D

RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL D PD2-1 LA 1 1 PD2-1 PD 1 1 MOZAFARI,B. 1 1 INTERNAL: ACNW 2 2 ACRS 2 2 AEOD/DOA 1 1 AEOD/DSP/TPAB 1 1 AEOD/ROAB/DSP 2 2 NRR/DET/ECMB 9H 1 1 NRR/DET/EMEB 7E 1 1 NRR/DLPQ/LHFB10 1 1 NRR/DLPQ/LPEB10 1 1 NRR/DOEA/OEAB 1 1 NRR/DREP/PRPB11 2 ' NRR/DST/SELB 8D 1 1 NRR/DST/SICB8H3 1 1 RRPD~/QPLB8 Dl 1 1 NRR/DST/SRXB 8E 1 1 REGALE~ 02 1 1 RES/DSIR/EIB 1 1 RGN2 FILE 01 1 1 EXTERNAL: EG&G BRYCE,J.H 3 3 L ST LOBBY WARD 1 1 NRC PDR 1 1 NSIC MURPHY,G.A 1 1 R NSIC POORE,W. 1 1 NUDOCS FULL TXT 1 1 D

D D

NOTE TO ALL "RIDS" RECIPIENTS:

PLEASE HELP US TO REDI.'CE WASTF! CONTACT THE DOCUMENT CONTROL DESK, ROOM Pl-37 (EXT. 20079) TO ELIMINA1E YOUR NAiWIE FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!

FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 33 ENCL 33

Carolina Power & Light Company P.O. Box 165 ~ New Hill N.C. 27562

~

R. B. RICHEY Vice President Harris Nuclear Project Letter Number: HO-910096 (0)

U.S. Nuclear Regulatory Commission ATTN: NRC Document Control Desk Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT UNIT 1 DOCKET NO. 50-400 LICENSE NO. NPF-63 LICENSEE EVENT REPORT 91-010-00 Gentlemen:

In accordance with Title 10 to the Code of Federal Regulations, the enclosed Licensee Event Report is submitted. This report fulfills the requirement for a written report within thirty (30) days of a reportable occurrence and is in accordance with the format set forth in NUREG-1022, September 1983.

Very truly yours R. B. Richey Vice President Harris Nuclear Project RBR:dmw Enclosure cc: Mr. S. D. Ebneter (NRC - RII)

Ms. B. L. Mozafari (NRC - RII)

Mr. J. E. Tedrow (NRC - SHNPP)

MEM/LER91-010/1/Osl 9$ Ct7 J.t.'tClg89 5 iO7O3 PDi( Ajt.t'C,'i, O5(.tOO4Ct(y pDI>>

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSI APPROVED OMB NO. 31504)104 (64)9)

EXPIRES: 4130I92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REQUESTr 50.0 HRS. FORWARD LlCENSEE EVENT REPORT {LER) COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS AND REPORTS MANAGEMENT BRANCH (P630), V,S, NUCLEAR REGULATORY COMMISSION, WASHINGTON, OC 20555, AND TO THE PAPERWORK REDUCTION PROJECT (31500104). OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON. DC 20503.

DOCKET NUMBER (2) PA E 3 FACILITY NAME (II N 1 ar Power Plant 0 5 0 0 0 4 P Q 1 OF.O 5 Reactor Tiip During Surveillance Testing and One Reactor Trip Breaker Failed To 0 en EVENT DATE (5) LER NUMBER (6) REPORT DATE (7) OTHER FACILITIES INVOLVED (SI MONTH DAY YFAR YEAR ~yc.'SEOIIENTIAL RE"0~ MONTH DAY YEAR FACILITYNAMES DOCKET NVMBERISI NVMSER gas NVMBEII N/A 0 5 0 0 0 0 6 03 9 191 0 1 0 007 03 9 1 0 5 0 0 0 THIS REPORT IS SUBMITTED PURSUANT T 0 THE RtOUIREMENTS OF 10 CFR (): (Chrch onr or morr of rhr followrnpl (1$

OPERATING MODE (9) 20A02(is) 20AOS(c) 50.73(e l(2)(ir) 73.7$ (t)

POWER 20AOS( ~ IIII(i) SOW(cl(1) 50.73(e) (2Hr) 73.7$ (c)

LEVEL GTHER (specify In Aosrrect

00) 20.405(sl(1)(EI 50.36(cl(2) 50.73(el(2( (vii)

(resow enrf In Test, HIIC Form c8 v 9'I' 60.73(e) (2)(viii) I A) 366AI ry 20AOS( ~ l(1)(ii)l 50.73( ~ I(2)(I) x 20,405 (s l(1 I I Iv) 50.73(e l(2)(ii) 60.73(e) l2)(viii)(B) 20.405(el (1 )(v) 50.73(el(2) (iii) 50.73(el(2)(x)

LICENSEE CONTACT FOR THIS LER (12)

NAME TELEPHONE NUMBER AREA CODE M.R..Hamb Pro'ect S ecialist Re ulator Com liance 193 62- 22 COMPLETE ONE I.INE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

CAUSE SYSTEM COMPONENT MANUFAC TVRER REPORTABLE TO NPRDS CAUSE SYSTEM COMPONENT MANVFAC.

TVRER EPORTABLE TO NPRDS RAN)

Ãi45~~ 4%3)IM SUPPLEMENTAL REPORT EXPECTED (14) MONTH OAY YtAR EXPECTED SUBMISSION DATE (161 YES (II yes, completr EXPECTED SVShrISSIDII DATEI NO ABSTRACT (Llmlt to f400 specrs, I e., rpproxlmrrrly fifrrrn slnplrepecr typewrirrrn lines) 116)

The plant received an automatic Reactor Trip on RCS Low Flow during the performance of a maintenance calibration procedure on an RCS flow transmitter. All rods fully inserted on the trip signal. The "A" Reactor Trip Breaker did not open on the automatic reactor trip signal but did open on a subsequent manual trip signal. The failure of the "A" Reactor Trip Breaker to open.was due to a failed undervoltage output driver card in the "A" Solid State Protection System. The failure of the undervoltage output driver card was the same as described in IE Notice 85-13 and had apparently occurred during maintenance on May 18, 1991. The failed undervoltage output driver card was replaced with a modified card which prevents this failure mechanism.

This event is being reported as a Technical Specification violation and Engineered Safety Features actuation per 10CFR50.73 (a)(2)(i)(B) and 10CFR50.73 (a)(2)(iv).

NRC Form 366 (64)9)

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSIO APPROVED 0MB NO. 31500104 (64)9)

EXPIRES: 4/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EVENT REPORT (LER) INFORMATION COLLECTION REGUEST: SOil HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT BRANCH (P4)30). U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO 1'HE PAPERWORK REDUCTION PROJECT (31500104), OFFICE OF MANAGEMENTAND BUDGET. WASHINGTON, DC 20503.

FACILITY NAME ('Il DOCKET NUMBER (2) LER NUMBER (6) PAGE IS)

YEAR ~o: SEQUENTIAL ~I": REVISION

WN NUMBER ~Rr9 NUMBER Shearon Harris Nuclear Power Plant 0 5 0 0 0 400 9 0 1 0 0 0 02OFO 5 TEXT /// mare 4/Mce lI rer/u/red, uee addle'aae/ NRC Farm 36549/ (IT)

EVENT DESCRIPTION:

At 1533, on June 3, 1991, an automatic Reactor Trip occurred due to a pressure spike in the flow transmitter common reference leg for "A" reactor coolant loop low flow.

All rods fully inserted on the reactor trip signal. The "A" reactor trip breaker failed to open on the automatic reactor signal but did open on a subsequent manual reactor trip signal. An Auxiliary Feedwater (AFW) actuation signal was initiated on low steam generator water level. All AFW pumps started as required and were later secured.

The low loop flow, signal occurred during the performance of surveillance test MST-I0056, Reactor Coolant Flow Instrument (F-0415) Calibration, on one of the three "A" reactor coolant loop flow transmitters. Manipulation of a transmitter isolation valve caused a pressure perturbation in the sensing lines of the two inservice transmitters, resulting in the generation of the low flow signal. MST-I0056 was recently revised because the RCS flow transmitters were changed to a new type during the previous refueling outage. As required for the new type of flow transmitter, steps to correct for instrument zero shift at pressure were added to the procedure.

It appears 'the reactor trip occurred when the high side (common to all three flow transmitters) isolation valve was opened to perform this zero shift check.

Procedures have since been revised to use the low side (non-common) isolation valve to perform this zero shift check.

The failure of the "A" reactor trip breaker to open on the automatic reactor trip signal was caused by a failed undervoltage output driver circuit board in the Solid State Protection System (SSPS). The failure of the undervoltage output driver card was initially determined to be a random failure. The undervoltage output driver card was replaced, "A" train SSPS was tested and the plant was =restarted. During subsequent investigation it was determined that random failure was probably not the it cause of the undervoltage output, driver card failure and that apparently failed during maintenance performed to correct a breaker closing problem on the "A" reactor trip breaker on May 18, 1991. The failure that prevented the "A" reactor trip breaker from opening was the same as described in IEN 85-13 and Westinghouse Technical Bulletin NSID-TB-85-16. This failure mechanism results in the output voltage from SSPS being maintained even if an automatic trip signal is present.

and automatic shunt trips. Manual This prevents both the automatic undervoltage trips remained available during this time period.

CAUSE:

The reactor trip was caused by a perturbation in the sensing lines during isolation valve manipulation while performing MST-I0056, Reactor Coolant Flow Instrument (F-0415) Calibration. This perturbation was caused either by opening the isolation valve too quickly or because the high side (common reference leg) isolation valve was used. Engineer review determined that using the high side isolation valve should be acceptable but that using the low side isolation causes less perturbations and is acceptable for the zero shift check. The failure of "AU Reactor Trip Breaker NRC Form 366A (64)9)

I NRC FORM366A U.S. NUCLEAR REGULATORY COMMISSIC APPROVED OMB NO. 31500104 (6J)9)

E XPIR ES: 1/30/92 ESTIMATED BURDEN PER AESPONSE TO COMPLY WTH THIS LICENSEE EVENT REPORT HLER) INFORMATION COLLECTION REOUESTI 500 HRS. FOAWARD COMMENTS REGARDING BUADEN ESTIMATE TO THE RECORDS I)

TEXT CONTINUATION AND REPORTS MANAGEMENT BRANCH (F430). U.S. NUCLEAR .

REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO 1HE PAPERWORK REDUCTION PROJECT (31500104), OFFICE I OF MANAGEMENTAND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)

SEQUENTIAL REV S ON YEAR N~o r 2 NUMSER gg (r5 NUMSER Shearon Harris Nuclear Power Plant 400 91 Ol 0 03 OF 0 TEXT ///mors T/roco /I ror)rr/rod, rrJo odd/I/orM/HRC Farm 366A3/ (IT)

CAUSE: (continued) to open was a result of a failed undervoltage output driver card in the "A" Solid State Protection System. This failure apparently occurred during maintenance on the "A" reactor trip breaker closing circuitry on May 18, 1991. The post-maintenance testing performed after this maintenance verified the close circuit problem was corrected but failed to detect the problem in the SSPS undervoltage output driver card.

There has been one previous event, reported in LER 91-009, where manipulation of instrument valves resulted in an Engineered Safety Features actuation when a common reference leg was affected. In that event, Auxiliary Feedwater actuated on a low steam generator level signal. There have been no previous events that were similar to the RAR Reactor Trip Breaker failure.

SAFETY SIGNIFICANCE:

During this event all systems functioned properly except for the failure of the "A" Reactor Trip Breaker to open. The safety significance of the "A" Reactor Trip Breaker to open on an automatic signal is mitigated by the fact that the "B" Reactor Trip Breaker was always available for automatic actuation and that manual actuation of both Reactor Trip Breakers was always available. Additionally, emergency procedures require that the operator immediately verify all rods fully inserted on a reactor trip signal, if not the operator immediately inserts a manual reactor trip if the "B" Reactor Trip Breaker had not signal. This would have been performed opened automatically.

The reactor trip and AFW actuation are reported as an Engineered Safety Feature actuation per 10CFR50.73 (a)(2)(iv).

The failure of the "A" Reactor Trip Breaker to open is being reported as a Technical Specification violation per 10CFR50.73 (a)(2)(i)(B).

CORRECTIVE ACTIONS:

Revise the Reactor Coolant System flow calibration procedures to perform the zero shift check using the low side (non-common) pressure and include precautions to perform valve manipulation very slowly to prevent perturbations in the sensing lines.

2. Replaced the failed undervoltage output driver card with a new fused card as recommended by Westinghouse Technical Bulletin NSID-TB-85-16.

Placed non-fused undervoltage output driver cards on administrative hold and"ordered additional fused undervoltage output driver cards.

NRC Form 366A (689)

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSI APPROVED OMB NO. 31500104 (64)9)

E XP I R ES; 4/30/92 ESTIMATED BURDEN PER RESPONSE TO COMP(.Y WTH THIS LICENSEE EVENT REPORT (LER) INFORMATION COLLECTION REGUEST: 508) HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT BRANCH (P4)30), U,S. NUCLEAR i

REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO

')HE PAPERWORK REDUCTION PROJECT (31504104). OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)

YEAR SEOVENTIAL REVISION NUMSER NUMBER Shearon Harris Nuclear Power Plant 0 5 0 0 0 4 p p'9 0 1 0 00. 04 OF 0 5 TEXT ///mare 4/Mce /4 rer)II(red, Iree edd/dorM///RC Farrrr 36543/ l)T)

CORRECTIVE ACTIONS: (continued)

4. The SSPS Logic test will only be deleted as a post-maintenance testing requirement for reactor trip breaker work with the Manager - Operations approval.

5. Additional troubleshooting guidance is being developed and will be incorporated into appropriate procedures.

An umbilical cord will be manufactured to supply 125V DC for reactor trip breaker maintenance and troubleshooting. This umbilical cord will not connect to the 48V SSPS output. A temporary 48V power supply will be used to supply a simulated SSPS input signal for any future reactor trip breaker work.

Appropriate site personnel involved in maintenance and modification work onsite will be sensitized to this event and the requirement that post-maintenance/post-modification testing be extensive enough to ensure no additional problems are created by the maintenance or modification activities.

NRC Form 366A (669)

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