05000247/LER-2010-007, Regarding Automatic Reactor Trip Due to a Turbine Trip as a Result of a High Steam Generator Level Trip After Transition to Single Feedwater Pump Operation
| ML103130040 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 11/02/2010 |
| From: | Joseph E Pollock Entergy Nuclear Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-10-110 LER 10-007-00 | |
| Download: ML103130040 (7) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(1), Submit an LER, Invalid Actuation 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| 2472010007R00 - NRC Website | |
text
~En teigy Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 Buchanan, N.Y. 10511-0249 Tel (914) 734-6700 J. E. Pollock Site Vice President NL-10-110 November 2, 2010 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Stop O-P1-17 Washington, D.C. 20555-0001
SUBJECT:
Licensee Event Report # 2010-007-00, "Automatic Reactor Trip Due to a Turbine Trip as a Result of a High Steam Generator Level Trip After Transition to Single Feedwater Pump Operation" Indian Point Unit No. 2 Docket No. 50-247 DPR-26
Dear Sir or Madam:
Pursuant to 10 CFR 50.73(a)(1), Entergy Nuclear Operations Inc. (ENO) hereby provides Licensee Event Report (LER) 2010-007-00. The attached LER identifies an event where the reactor was automatically tripped, which is reportable under 10 CFR 50.73(a)(2)(iv)(A). As a result of the reactor trip, the Auxiliary Feedwater System was actuated, which is also reportable under 10 CFR 50.73(a)(2)(iv)(A). This condition was recorded in the Entergy Corrective Action Program as Condition Report CR-IP2-2010-05484.
There are no new commitments identified in this letter. Should you have any questions regarding this submittal, please contact Mr. Robert Walpole, Manager, Licensing at (914) 734-6710.
Sincerely, JEP/cbr cc:
Mr. William Dean, Regional Administrator, NRC Region I NRC Resident Inspector's Office, Indian Point 2 Mr. Paul Eddy, New York State Public Service Commission LER Events@ inpo.org
Abstract
On September 3, 2010, during a scheduled plant shutdown, an automatic reactor trip (RT) was initiated as a result of a turbine trip due to a high steam generator (SG) water level.
All control rods did not indicate fully inserted as the Individual Rod Position Indicator (IRPI) for rod H-8 indicated 38 steps withdrawn and its rod bottom light failed to light.
Rod H-8 was verified to be fully inserted by alternate means.
All primary systems functioned properly.
The plant was stabilized in hot standby with decay heat being removed by the main condenser.
The Auxiliary Feedwater System automatically started as designed due to automatic trip of the MBFPs as a result of closure of the MBFP discharge valves from the SG high level trip.
The direct cause of the RT was a turbine trip on a high SG level.
The root cause was inadequate design control of the proportional band and reset tuning settings for critical plant controllers.
There was less than optimum controller settings on the MBFP speed controller, Feed Regulatory Valve (FRV) flow controllers, and the SG level controllers for low power operation.
Significant corrective actions include: I&C procedures developed from I&C Preventive Maintenance documents have been reviewed to ensure that the instrument calibration requirements have been transferred into the I&C procedures, a list of critical controllers was generated and the Equipment Data Base (EDB) updated with known existing settings, I&C procedures will be reviewed to identify changes to ensure controller calibrations maintain required settings and procedures will be revised to incorporate testing of critical parameters, an engineering evaluation will be issued and the EDB updated in the work control program with findings on controller settings.
The event had no effect on public health and safety.
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This event was bounded by the analyzed event described in FSAR Section 14.1.10, "Excessive Heat Removal Due to Feedwater System Malfunctions."
Excessive FW additions is an analyzed event postulated to occur from a malfunction of the FW control system or an operator error which results in the opening of a FW control valve.
The analysis assumes one FW valve opens fully resulting in the excessive FW flow to one SG.
For the FW system malfunction at full power, the FW flow resulting from a fully open control valve is terminated by the SG high level signal that closes all FW control valves and trips the MBFPs.
Trip of the MBFPs automatically actuates the AFWS.
The SG high water level signal also produces a signal to trip the main turbine.
A TT initiates a RT.
The analysis for all cases of the excessive FW addition initiated at full power conditions with and without automatic rod control, show that the minimum DNBR remains above the applicable safety analysis DNBR limit, the primary and secondary side maximum pressures are less than 110% of the design values, and all applicable Condition II acceptance criteria are met.
For this event, rod control was in automatic.
All control rods did not indicate fully inserted as the Individual Rod Position Indicator (IRPI) for rod H-8 indicated 38 steps withdrawn and its rod bottom light failed to light.
Rod H-8 was verified to be fully inserted by alternate means.
Troubleshooting determined that two electrolytic capacitors of IRPI required replacement.
The AFWS actuated and provided required FW flow to the SGs.
RCS pressure remained below the set point for pressurizer PORV or code safety valve operation and above the set point for automatic safety injection actuation.
Following the RT, the plant was stabilized in hot standby.