05000458/LER-2003-001, Unplanned Reactor Scram Due to Fluid Leak in Main Turbine Electrohydraulic Control System
| ML031180571 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 04/22/2003 |
| From: | King R Entergy Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| RBF1-03-0068, RBG-46105 LER 03-001-00 | |
| Download: ML031180571 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(v), Loss of Safety Function |
| 4582003001R00 - NRC Website | |
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- Entergy Entergy Operations, Inc.
River Bend Station 5485 U S Highway 61 P 0 Box 220 St Francisville, LA 70775 Tel 225 336 6225 Fax 225 635 5068 Rick J. King Director Nuclear Safety Assurance April 22, 2003 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555
Subject:
River Bend Station Docket No. 50-458 License No. NPF-47 Licensee Event Report 50-458 / 03-001-00 File Nos.
G9.5, G9.25.1.3 RBG-46105 RBF1 0068 Ladies and Gentlemen:
In accordance with 1 OCFR50.73, enclosed is the subject Licensee Event Report.
There are no commitments in this document.
Sincerely, RJK/dhw enclosure I4 ZC
Licensee Event Report 50-458 / 03-001-00 April 22, 2003 RBG-46105 Page 2 of 2 cc:
U. S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011 NRC Sr. Resident Inspector P. O. Box 1050 St. Francisville, LA 70775 INPO Records Center E-Mail Mr. Jim Calloway Public Utility Commission of Texas 1701 N. Congress Ave.
Austin, TX 78711-3326 Mr. Prosanta Chowdhury Program Manager-Surveillance Division Louisiana DEQ Office of Radiological Emergency Planning and Response P. O. Box 82215 Baton Rouge, LA 70884-2215
NRC FORM 366 U.S. NUCLEAR REGULATORY APPROVED BY OMB NO. 3150-0104 EXPIRES 6-30-2001 (1-2001)
COMMISSION
, the NRC may not conduct or sponsor, and a nerson is not required to respond to, the information collection FACILITY NAME (1)
DOCKET NUMBER (2)
PAGE (3)
River Bend Station 050- 458 1 OF 3
TITLE (4)
Unplanned Reactor Scram Due to Fluid Leak in Main Turbine Electrohydraulic Control System EVENT DATE (5)
LER NUMBER (6)
REPORT DATE (7)
OTHER FACILITIES INVOLVED (8)
FACILITY NAME DOCKET NUMBER MO DAY YEAR YEAR NUMBER NO MO DAY YEAR 05000 FACILITY NAME DOCKET NUMBER 02 22 2003 2003 001 00 04 22 2003 05000 OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THF REQUIREMENTS OF 10 CrR 6, (Check all that aoolv) Ill)
MODE (9) 1
_ 20.2201 (b)
_ 20 2203(a)(3)(1i) 50.73(a)(2)(ii)(B)
_ 50 73(a)(2)(x)(A)
POWER
_ 20.2201 (d)
_ 20.2203(a)(4) 50.73(a)(2)(iii) 50 73(a)(2)(x)
LEVEL (10) 90% _
20.2203(a)(1) 50 36(c)(1)(i)(A)
X 50.73(a)(2)(iv)(A) 73 71(a)(4) 20.2203(a)(2)(i)
_ 50.36(c)(1)(ii)(A)
.50-73(a)(2)(v)(A)
_ 73.71(a)(5) 20.2203(a)(2)(i) 50 36(c)(2) 50.73(a)(2)(v)(B)
OTHER 20 2203(a)(2)(i) 50 46(a)(3)(in)
=
50.73(a)(2)(v)(C) ify in AbRC bo (If more space is required, use additional copies of NRC Forn 366A) (17)
REPORTED CONDITION At approximately 1:23 a.m. CST on February 22, 2003, with the plant operating at 90 percent power in end-of-cycle coastdown, an unplanned manual reactor scram was initiated. During the scram recovery, the reactor core isolation cooling (RCIC) system was manually actuated to assist in reactor water level control. This event is being reported in accordance with 10CFR50.73(A)(2)(iv)(a) as a valid actuation of the reactor protection system (RPS) and RCIC.
INVESTIGATION Approximately ten minutes prior to the scram, an alarm actuated in the main control room indicating a low level in the main turbine (**TRB**) electrohydraulic control (EHC) oil reservoir. An operator was dispatched to investigate, and reported that the EHC pump (**P**) was cavitating and that there appeared to be smoke in the area of the turbine front standard. The scram was initiated in anticipation of losing EHC system function. The main turbine unloaded normally as steam pressure decreased, and the main generator output breaker tripped on reverse power as designed. The EHC system was then shut down.
Following the scram, reactor water level lowered to the low alarm setpoint (Level 3).
The feedwater level setpoint setdown, reactor recirculation pumps downshift, and the suppression pool cooling system trip all occurred as expected. Approximately 35 seconds later, reactor water level rose to the high alarm setpoint (Level 8). Due to a pre-existing deficiency, the "C" feedwater regulating valve (**FCV**) was not able to close to less than 80 percent open. A high reactor water alarm (Level 8) signal was received, tripping all three main feedwater pumps as designed. RCIC was initiated to provide water level control. The "A" loop of the residual heat removal system was placed into the suppression pool cooling mode as required for RCIC operation.
Subsequent to the RCIC initiation, main feedwater pump "A" was restarted to provide level control. Three minutes after the "A" main feedwater pump was restarted, reactor water level reached Level 8 for a second time. This automatically closed the RCIC system injection valve and tripped the feedwater pump. Five minutes later, operators restarted main feedwater pump "B" and reactor water level increased to Level 8 for the third time approximately 4 minutes after the pump start. Reactor water level control using RCIC was established at 2:01 a.m. The RCIC system was shut down at 4:00 p.m.
on the day of the event after the main feedwater system was put back into service.
During the refueling outage subsequent to the reported event, internal inspections of the "C" feedwater regulating valve, the "C" main feed pump discharge check valve (If more space is required, use additional copies of NRC Form 366A) (17)
(**V**), and the downstream isolation valve (**ISV**) were performed. Part of a washer missing from the discharge check valve was found inside the feedwater regulating valve that was preventing it from closing past the 80 percent open position.
Additionally, the disc nut from the check valve was missing, and was found lodged in the seat of the isolation valve. This was preventing the isolation valve from closing completely. These conditions were determined to be significant contributing factors to the difficulty in maintaining reactor water level less than Level 8 during this event.
CAUSAL ANALYSIS AND CORRECTIVE ACTIONS The investigation determined that the leak occurred when a section of hydraulic tubing near the main turbine control valves developed a through-wall crack. The analysis of this event is continuing. A complete root cause analysis and corrective action plan will be provided in a supplement to this LER.
SAFETY SIGNIFICANCE
The plant was shutdown following the discovery of a leak in the main turbine EHC system. A manual reactor scram was initiated and all safety systems operated per design. Prior to the event, reactor power was at 90 percent in end-of-cycle coastdown.
The core was operating with approximately 15 percent margin to thermal power margins. No power excursion or pressure excursion were seen in post-trip data.
Therefore, the safety limits specified in Technical Specification were not challenged.
Reactor water level was maintained above the Level 2 setpoint at which emergency core cooling systems are actuated. Water level was controlled and maintained by the RCIC system. Reactor feedwater pumps remained available following their automatic shutdown upon reaching Level 8, although the decision was made to utilize RCIC for an expeditious recovery.
Reactor water level was returned to a normal post-shutdown range in a timely manner.
There was no release of steam to the containment, drywell, or suppression pool; therefore, there was no challenge to containment integrity. The reactor vessel pressure did not exceed the first safety relief valve set point of 1133 psig, or the high pressure scram setpoint (nominally 1094.7 psig); therefore, there was no challenge to the reactor coolant pressure boundary.
As none of the barriers to fission product release were challenged, there was no adverse effect on nuclear safety. This event had minimal effect on the health and safety of the public.
(NOTE: Energy Industry Component Identification codes are annotated as (**XX**).)