05000286/LER-2005-002, Re Automatic Reactor Trip Due to 32 Steam Generator Steam Flow/Feedwater Flow Mismatch Caused by Low Feedwater Flow Due to Inadvertent Condensate Polisher Post Filter Bypass Valve Closure
| ML052500486 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 07/05/2005 |
| From: | Dacimo F Entergy Nuclear Northeast |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-05-078 LER 05-002-00 | |
| Download: ML052500486 (8) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| 2862005002R00 - NRC Website | |
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Entergy Nuclear Northeast Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 Buchanan. NY 10511-0249 Tel 914 734 6700 Fred Dacimo Site Voce Presdent Admnzistration July 5, 2005 Indian Point Unit No. 3 Docket Nos. 50-286 NL-05-078 Document Control Desk U.S. Nuclear Regulatory Commission Mail Stop O-Pl-17 Washington, DC 20555-0001
Subject:
Licensee Event Report # 2005-002-00, "Automatic Reactor Trip Due to 32 Steam Generator Steam Flow/Feedwater Flow Mismatch Caused by Low Feedwater Flow Due to Inadvertent Condensate Polisher Post Filter Bypass Valve Closure."
Dear Sir:
The attached Licensee Event Report (LER) 2005-002-00 is the follow-up written report submitted in accordance with 10 CFR 50.73. This event is of the type defined in 10 CFR 50.73(a)(2)(iv)(A) for an event recorded in the Entergy corrective action process as Condition Report CR-IP3-2005-02478.
There are no commitments contained in this letter. Should you or your staff have any questions regarding this matter, please contact Mr. Patric W. Conroy, Manager, Licensing, Indian Point Energy Center at (914) 734-6668.
Sincerely, Vice President Indian Point Energy Center
Docket No. 50-286 NL-05-078 Page 2 of 2 Attachment: LER-2005-002-00 cc:
Mr. Samuel J. Collins Regional Administrator - Region I U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Resident Inspector's Office Resident Inspector Indian Point Unit 3 Mr. Paul Eddy State of New York Public Service Commission INPO Record Center
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Abstract
On May 6, 2005, at approximately 10:32 hours, an automatic reactor trip was initiated due to a 32 steam generator (SG) steam flow/feedwater flow mismatch with low SG level trip during Instrumentation & Control (I&C) troubleshooting of the condensate system (CS).
All control rods fully inserted and all required safety systems functioned properly.
The plant was stabilized in hot standby with decay heat being removed by the main condenser. There was no radiation release.
The Emergency Diesel Generators did not start as offsite power remained available. The Auxiliary FW system automatically started as expected due to a SG low level. The cause of the event was personnel error due to poor work practices during troubleshooting of the CS.
The control switch for the condensate polisher (CP) post filter bypass valve was in auto when a wire was disconnected from a relay of a seal-in circuit causing the post filter bypass valve to close.
The post filter bypass valve closure caused inadequate feedwater flow that resulted in the mismatch and low SG level.
Operations and I&C personnel failed to verify and recognize the required switch position for the condition due to overconfidence.
Corrective actions include counseling operations and I&C personnel and applicable plant staff on management expectations for a questioning attitude, pre-job briefs, peer checking, and use of the formal work process.
The CP system operating procedure was revised for proper CP valve control switch position and seal-in circuit clearance. The work control procedure will be revised to define when work becomes troubleshooting and decide the level of risk assessment to be performed. The event had no effect on public health and safety.
(if more space Is required, use additional copies of (If more space Is required, use addidional copies of (if more space Is required, use additional copies of NRC Forn 3664) (17)
The root causes were; 1) Operations and I&C personnel failure to verify and recognize the required switch position for the condition due to overconfidence, and 2) Operations misunderstanding whether the activity constituted operational maintenance or troubleshooting allowing I&C to work outside the normal work process. Significant contributing causes (CC) were as follows: (CC-1) Poor communication; Pre-job briefs between the Control Room Supervisor (CRS) and I&C Supervisor and between the CRS and Reactor Operators (ROs) were inadequate and not formal.
Technicians only received verbal guidance. There was no written plan or work package.
(CC-2) Inadequate procedures; The CP System Operating Procedure (3-SOP-C-002) only required the valve to open.
The SOP did not specify the bypass valve's control switch position as a function of plant condition.
The work control process procedure IP-SMM-WM-100 and the troubleshooting procedure IP-SMM-MA-103 do not have a clear interface. Troubleshooting appears to be a subset of the different types of work described in the work control process.
The troubleshooting procedure was not utilized for this event because I&C believed they were performing operational maintenance. (CC-3) Poor use of Human Performance tools, lack of a questioning attitude, (CC-4) Knowledge/Training; Operator's knowledge of how the bypass valve worked in AUTO. The CRS authorizing the work misunderstood what work control procedure governed the work being performed.
There are no Licensed Operator Requalification tasks associated with the CP.
CORRECTIVE ACTIONS
The following corrective actions have been or will be performed under the CAP to address the causes of this event.
- Operations, I&C, Planning Scheduling & Outage (PS&O), Maintenance, and Engineering were briefed on management's expectation concerning a questioning attitude, pre-job briefs, peer checking, and the use of the IPEC formal work process. These expectations for I&C and Operations will be strengthened through training and/or alignment meetings.
- Operations and I&C individuals involved in this event were counseled.
- Procedure 3-SOP-C-002, ~Condensate System Operation,' was revised to ensure proper CP configuration for system post filter bypass and seal-in circuit clearance.
- Procedure IP-SMM-WM-100,0Work Control Process, will be revised to define when work becomes troubleshooting and decide the level of risk assessment to be performed for operational maintenance, investigative work and emergency work.
Procedure revision is scheduled to be completed July 29, 2005.
- Formally review this event with all operating crews.
Completion is scheduled for July 31, 2005.
- The operator training program will be evaluated and determined if it needs to be revised for condensate polisher operations.
Completion is scheduled for July 31, 2005.
(If more space Is required, use addiffonal copies of (If more space Is required, use addifionalcopies ofNRCFom, 3664) (17)
This event was bounded by the analyzed event described in FSAR Section 14.1.9, Loss of Normal Feedwater. A Low-Low water level in any one SG initiates actuation of two motor-driven AFW pumps and one steam driven AFW pump.
The AFW System has adequate redundancy to provide the minimum required flow assuming a single failure.
The analysis of a loss of normal FW shows that following a loss of normal FW, the AFWS is capable of removing the stored and residual heat plus reactor coolant pump waste heat thereby preventing either over pressurization of the RCS or loss of water from the reactor.
For this event, rod control was in automatic and the reactor scrammed immediately upon receiving the reactor trip signal.
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.
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