ML20045C542

LER 93-010-00:on 930522,SG Level Oscillations Occurred in One SG Resulting in SG Being Overfed & Causing high-high SG Water Level.Caused by Personnel Error.Power Reduced to Approx 2.5% & FW to SGs restored.W/930618 Ltr

ML20045C542
Person / Time
Site:	Seabrook
Issue date:	06/18/1993
From:	Feigenbaum T, Peschel J NORTH ATLANTIC ENERGY SERVICE CORP. (NAESCO)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-93-010, LER-93-10, NYN-93087, NUDOCS 9306230270
Download: ML20045C542 (7)

Text

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P.O. Box 300 '

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seedrook. "H 03874 Telephone (603)474-9521

[C Facsimile (603)474-2987

' Energy Service Corporation Ted c. Feigenbaum -

Senior Vice President and Chief NuclearOfficer NYN-93087 -

June 18,1993 United States Nuclear Regulatory Commission Washinct on, D.C. 20555 Attention: Document Control Desk

Reference:

Facility Operating License No. NPF-86, Docket No. 50-443

Subject:

Licensee Event Report (LER) 93-10 00: " Engineered Safety Features Actuation- Feedwater Isolation" Gentlemen:

Enclosed is Licensee Event Report (LER) No. 93-10-00 for Seabrook Station. This submittal documents an event which was discovered on May 22,1993. This event is being reported pursuant to 10CFR50.73(a)(2)(iv).

Should you require further information regarding this matter, please contact Mr.

James M. Peschel, Regulatory Compliance Manager at (603) 474 9521 extension 3772. ,

Very truly yours,

[fC Ted C. Feigenba m TCF:EWM/cwm ,

Enclosures:

NRC Forms 366/366A i

2200C7 9306230270 930618 E a member of the Northeast Utilities system

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9 United States Nuclear Regulatory Commission June-18,1993 -

Atterition: Document Control Desk Page two ,

.I cc: Mr. Thomas T. Martin )

Regional Administrator ,

United States Nuclear Regulatory Commission l Region 1 _.'

475 Allen' dale Road King of Prussia, PA 19406-Mr. Albert W. De Agazio, Sr. Project Manager Project Directorate I-4 -1 Division of Reactor Projects j U.S. Nuclear Regulatory Commission Washington, DC 20555 Mr. Noel Dudley .

NRC Senior Resident Inspector P.O. Ilox 1149 Seabrook,' NH 03874 ,

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INPO Records Center 700 Galleria Parkway Atlanta, GA 30339-5957 i F

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i OtC IORM 366 U.S. NUCtLAR REGUI AIORY COPHISSION APPROULD B7 OMD CD. 3150-0104 i (5-92) EXPlRES 5/31/95 i 1

ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS.

LICENSEE EVEN'I, REPORT (LER) FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCd (MNBB 7714), U.S. NUCLEAR REGULATORY COMMISSION.

(See reverse for required number of digits / characters for each block) WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OfflCE OF ,

MANAGEMFNT AND BUDGET, WASHINGTON, DC 20503.  !

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I ACILilY N/ME (1) DOCKL1 NUMBLR (2) PAGE (3)

Seabrook Station 05000443 1 OF 5 Till t' (4)

Engineered Safety Features Actuation- Feedwater Isolation (Vf MI DAll (5) tfR NUMBfR (6) RfPORI DAlf (7) OlHIR f ACit ITIES INV0t Vf D (8)

MONTH DAY YEAR YEAR NUMBER NUMPER MONTH DAY YEAR 05000 FA M I M NAML NET ""

05 22 93 93 10 00 06 18 93 O 0O OPERATING y IHIS RFPORT IS SUBMIlllD PURSUANI 10 THE REQUIRIMINIS Of 10 CFR %: (Check one or more) (11)

MODE (9) 20.402(b) 20.405(c) X 50.73(a)(2)(iv) 73.71(b)

POWER 20.405(a)(1)(1) 50.36(c)(1) 50.73(a)(2)(v) 73.71(c)

I4 LEVEL (10) 20.405(a)(1)(ii) 50.36(c)(2) 50.73(a)(2)(vii) OTHER 20.405( a)(1)( t il) 50.73(a)(2)(i) 50.73(a)(2 )(v iii)( A) (Specify in low 20.40$(a)(1)(iv) 50.73(a)(2)(li) 50.73(a)(2)(viii)(B) /[jt ct 20.405(a)(1)(v) 150.73(a)(2)(iii) 50.73(a)(2)(x) NRC Form 366A)

TIC [N5FI ONIACI FOR THIS 1IR (12)

NAME lELEPHONE NUMBER (include Area Code)

Mr. James M. Peschel, Regulatorf# Compliance Mgr. (603) 474-9521 Ext. 3772 COMPIElf ONI IINI F OR F ACH COMPONI NT F Alt VRf Df SOtlBf D IN THIS RFPORI (13)

CAUSL SYSTLM COMPONfNT MANUFACTURER CAUSE SYSTEM COMPONENT MANUFACTURER 0

MONTH l SUPPlf MINIAl RfIVtT IXPf CIID (14) EXPECTED DAY YF AR YLS SUDMISSION (If yes, complete EXPECTED SUBMISSION DATE). X NO DATE (15) ,

"5BSIRACI (Limit to 1400 spaces, i.e. approximately 15 single-spaced typewritten lines) (16)

During a plant startup (approximately 14 percent power) on May 22,1993, at approximately 1000 EDT, Steam Generator (SG)  !

level oscillations occurred in one SG resulting in the SG being overfed and causing a high-high SG watt r level. This caused a I feedwater isolation to all four SG's and additionally caused a turbine trip signal. Operator response to the event prevented a l reactor trip and reactor power was reduced to approximately 2.5% At 1154 EDT, North Atlantic made a four-hour  !

notiGeation to the NRC pursuant to 10 CFR 50.72(b)(2)(ii) since this event constituted an ESF actuation.  ;

I There were no adverse safety consequences as a result of this event. A'l equipment functioned as designed and all operator j actions v ere determined to be appropriate to ensure the safety of the plant and the public.

l The root cause for this esent was determined to be personnel error on the part of the feed station operator. Specifically, the l feed station operator knew the steps involved to successfully transfer to the main feedwater regulating valves (FRV's) but did l not properly anticipate and interpret the SG level conditions with respect to existing plant conditions. The extreme sensitivity of the FRV's was not fully realized, as shown by the 'A' FRV being opened approximately 6 percent initially.

Immediate corrective actions, as a result of the Feedwater Isolation, were to reduce power to approximately 2.5% and restore feedwater to the SGs which prevented a reactor trip. Long term corrective actions include: procedure enhancements; low power operator training; an evaluation of simulator enhancements with regard to low power and feedwater operations; and pre-shift bricGngs for low power evolutions.

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Y FCRM 366A U.S. NUCLEAR REGULATORY C0pellSSIC~J AIPROVED BY OMB NO. 3150-0104 EXPIRES 5/31/95 l (5-92) l ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH I

, THIS INFORMATION COLLECTION REQUEST: 50.0 HRS.

FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO I LICENSEE EVENT REPORT (LER) THE INFORMATION AND RECORDS MANAGEMENT BRANCH (MNBB 7714), U.S. NUCLEAR REGULATORY COMMISSION, TEXT CONTINUATION WASHINGTON, REDUCTION PROJECTDC 20555-0001, (3150-0104), OFFICEANDOf TO THE P MANAGEMENT AND BUDCET, WASHINGTON, DC 20503.

F AClll1Y NAME (1) DOCKET NUMBER (2) LER NUMBER (6? PAGE (3)

SEQUENilAL REVISION YEAR Scabrook Station NUMBER NUMBER 05000443 93 00 2 OF 5 10 TEXT (if more space is required, ese t'-itional copfes of NRC form 366A) (t7)

Description of Event During a plant startup (approximately 14 percent power) on May 22,1993, at approximately 1000 EDT, Steam Generator (SG) level oscillations occurred in one SG, resulting in the SG being overfed and causing a high-high SG water level. This caused a fcedwater isolation to all four SG's and additionally caused a turbine trip signal. Operator response to the event prevented a reactor trip and reactor power was reduced to approximately 2.5%.

The plant was in the process of starting up following the unit trip that occurred on May 20,1993 (See LER 93-09). The turbine was off-line on the turning gear at the time of the event. The main turbine was reset and was in the process of shell and chest warming. The shift assumed the watch during the evening of May 22, 1993 and were continuing the steps of main plant evolution, (MPE), procedure OS1000.02, Plant Startup from Hot Standby to Minimum Load. An additional control room operator was available to augment the crew as feed station operator during planned low power feedwater operations.

The USS directed the primary control room operator to increase power using control rods. This action was taken primarily to increase feed flow and move away from the low power conditions to a point where SG 1evel control sensitivity to feed flow changes would be less severe. This order was acknowledged by the primary operator and rods were withdrawn a couple of steps.

The feed station operator recognized that the Feedwater Regulating Bypass Valves were 80% open and were near the upper end of their control range. The feed station operator asked the USS for permission to start transferring feed flow from the Feedwater Regulating Bypass Valves to the Feedwater Regulating Valves (FRV's). The USS then directed the feed station operator to commence transferring feedflow from the Feedwate r Regulating Bypass Valves to the FRV's by starting with the ' A' SG. This direction was acknowledged by the feed station operator.

The feed station operator, in attempting to open the 'A' FRV slightly, held the valve open pushbutton too long, which resulted in opening the FRV approximately 6 percent. The extreme sensitivity of the FRV's was not realized, as shown by the 'A' FRV being opened approximately 6 percent initially. Shortly afterwards, the feed station operator took manual control of feedwater flow to the other 3 SGs and attempted to open their respective FRV's slightly to compensate for the inventory changes that would result from the earlier power increase. The feed station operator was under the impression that, in the process of transferring the 'A' SG from the bypass to the FRV, it would also be necessary to take manual control of the feedwater flow to the other 3 SGs to compensate for inventory changes due to the power increase and changes in feed flow distribution when swapping from the bypass valve to the FRV on one SG. The feed station operator did not inform the USS of his actions on the 'B', 'C', and 'D' SGs.

hRC f DiiH 366A T5-92)

INRC' FORM 366A U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 (5-92) EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST:

50.0 HRS. FORWARD COMMENTS REGARDING LICENSEE EVENT REPORT (LER) BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (MNBB 7714), U.S.

TEXT CONTINUATION NUCLEAR REGULATORY COMMISSION WASHINGTON, DC 20555-0001, AND TO TIIE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.

FACILITY MAME (1) DOCKET NUMBER LER NUMBLR (6) PAGE (3)

YEAR SEQUENTIAL REVIS10 Seabrook Station 05000443 93 00 3 OF 5 0

TEXT (if more space is required, use additional copies of NRC form 366A) (t?)

(Continued)

The MPE procedure, OS1000.02, provides a method for plant startup through 22 percent power. The procedure specifically addresses feedwater temperature, limitations on the Feedwater Regulating Bypass Valves, power stabilization for enhanced feedwater level control, and turbine generator operations. The transfer of feedwater flow control from the Feed Regulating Bypass Valves to the Main Feed Regulating Valves was attempted with the turbine generator on the turning gear, therefore there was effectively no feedwater heating. The USS proceeded with the unit startup despite not having feedwater preheating due to the concern of feedwater instability at low power levels. This lack of feedwater preheating, was not discussed with the shift or feedwater station operator. The sequence of bringing all four main feedwater regulating valves off of their closed seats, coupled with the earlier power increase, induced oscillations in all four SG's.

The oscillations were made worse by cold feedwater temperatures and rapidly progressed to the point at which the feed station operator requested assistance.

The USS assigned a second operator to the feed station. Before the oscillations could be brought under control, the 'B' SG narrow range indicated level reached the F-14 setpoint (86% narrow range level) and a Feedwater Isolation occurred. Although the Feedwater Isolation was not avoided, the combined effort of two experienced feed station operators, (both had been through the power ascension low power feed control training and had successfully been involved with several startups as feed station operators), did prevent a reactor trip due to low SG level.

At 1154 EDT, North Atlantic made a four-hour notification to the NRC pursuant to 10 CFR 50.72(b)(2)(ii) since this event constituted an ESF actuation.

Safety Conseauences There were no adverse safety consequences as a result of this event. All equipment functioned as designed and the operator actions in response to the event were determined to be appropriate to ensure the safety of the plant and the public. At no time during this event was there any impact on the health and safety of plant employees or the public.

Root Cause it is recognized that SG level control and feedwater regulating valve transfer at low power is sensitive and difficult evolution. Ilowever, the use of a dedicated feed station operator and training have been proven to i

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E FORM 366A U.S. MUCLEAR REGULATORY COMMISSION APPROVLD BY OMB NO. 3150-0104 (5-92) EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY

. WITH THIS INFORMATION COLLECTION REQUEST:

50.0 HRS. FORWARD COMMENTS REGARDING LICENSEE EVENT REPORT (LER) BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (MNBB 7714), U.S.

TEXT CONTINUATION NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT , OFFICE OF MANAGEMENT AND (3150-0104)HINGTON, BUDGET, WAS DC 20503.

IACitIIY NAME (1) DOCKET NUMBER tfR NUMBER (6) PAGE (3)

YEAR SEQUENilAL REVIS10 Seabrook Station 05000443 93 00 4 OF 5 10 LEX 1 (if more space is required, use additional copies of MRC f orm 366A) (11)

(Continued) be extremely successful in the past. The root cause for this event was determined to be personnel error on the part of the feed statian operator. Specifically, the feed station operator knew the steps involved to successfully transfer to the FRV's but did not properly anticipate and interpret the SG level conditions with respect to existing plant conditions. The extreme sensitivity of the FRV's was not realized as shown by the

'A' FRV being opened approximately 6 percent initially.

Several contributing factors were identified that if corrected or improved may have supported the feed station operator during this evolution. These factors are as follows:

1. The time span between the first valve opening and the power increase was not entirely consistent with the power / steam generator level stability guidance in the main plant evolution procedure.

2. The MPE and operating procedure did not include or reference appropriate portions of the Westinghouse power ascension feedwater control training or lessons learned from previous successful performance of this evolution.

3. Incomplete communication existed as the feed station 'erator did not announce to the crew that he was taking manual control of the other three FRV's.

Corrective Actions Immediate corrective actions, in response to the Feedwater Isolation, were to reduce plant power to approximately 2.5% and restore feedwater flow to the steam generators. Long term corrective actions to prevent reoccurrence include the following:

1. The normal operating procedure used for transferring the Feedwater Regulating Valves Dypass to the Main Feedwater Regulating Valves will be revised to provide additional guidance on plant parameters to monitor when initiating and completing the transfer.

2. Operator training will be provided in the simulator on the revised procedure and low power operations.

3. The simulator design will be reviewed for enhancements that could improve simulator response during low power operations aad feedwater station operations.

4. The guidelines for pre-shift crew briefing will be reviewed to ensure that low power feedwater system l

! operations are adequately addressed.

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l MRC'f0RM 366A U.S. NUCLEAR RECHLAIORY COMMISSION APPROVED BY OMB NO. 3150-0104 EXPIRES 5/31/95 (5 92)

ESTIMATED DURDEN PER RFSPONSE TO COMPLY

. WITH THIS INFORMATION COLLECTION REQUEST.

50.0 HRS. FORWARD COMMENTS REGARDl!(G LICENSEE EVENT REPORT (LER) BURDEN ESTIMATE TO THE INFORMATION aND b.S.

TEXT CONTINUATION RECORDS MANAGEMENT NUCLEAR REGULATORY BRANCH (MNBB COMMISSION WASHlh 7714),GTON, DC 20555-000), AND TO IIIE PAPERWORK REDUC 110N PROJECT (3150-0104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, OC 70503.

FACIllIY NAME (1) DOCKET NUMHFR t[R NUMBER (6) PAGE (3)

YEAR SEQUENTIAL REVIS10 Seabrook Station 05000443 93 00 5 OF 5 0

1LK1 (11 more space is required, use additional copies of NRC form 366A) (11)

Plant Conditions At the time of this event the plant was in MODE 1, at 14 percent power, with a Reactor Coolant System temperature of 558 degrees Fahrenheit and pressure of 2235 psig.

Similar Events This is the second occurrence where a high-high level in a SG resulted in a Feedwater Isolation that involved operator action. LER 92-017 documents an event which occurred on September 7,1992 with the plant in MODE 1. During a routine shutdown to begin a refueling outage while at approximately 12 percent power SG level oscillations led to one SG being overfed, initiating a Feedwater Isolation. The root cause for this event was determined to be incomp.ete communication, in that the operations crew did not effectively communicate and coordinate actions that were taking place on rod control, feedwater control, and turbine control. The Feedwater Isolation described in this LER differed in that it occurred during a plant startup and was induced by operator action while the Feedwater Isolation of LER 92-017 was induced by the plant.

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