05000458/LER-2014-002, Regarding Reactor Scram Due to Average Power Range Monitor High-flux Signal Following a Malfunction of the Main Turbine Electro Hydraulic System
| ML15006A029 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 12/16/2014 |
| From: | Olson E Entergy Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| RBF1-14-0181, RBG-47522 LER 14-002-00 | |
| Download: ML15006A029 (6) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(ix)(A) 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(iv), System Actuation |
| 4582014002R00 - NRC Website | |
text
S' Entergy Entergy Operations, Inc.
River Bend Station 5485 U. S. Highway 61 N St. Francisville, LA 70775 Tel 225 381 4374 Fax 225 381 4872 eolson@entergy.com Eric W. Olson Site Vice President RBG-47522 December 16, 2014 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555
Subject:
Licensee Event Report 50-458 / 2014-002-00 River Bend Station - Unit 1 Docket No. 50-458 License No. NPF-47 RBF1-14-0181
Dear Sir or Madam:
In accordance with 10 CFR 50.73, enclosed is the subject Licensee Event Report.
This document contains no commitments. Please note that this report number has been reassigned from a previous report submitted on March 12, 2014, which was subsequently cancelled.
If you have any questions, please contact Mr. Joseph Clark at 225-381-4177.
Sincerely, EWO/dhw Enclosure
'-Z
Licensee Event Report 50-458 / 2014-002-00 December 16, 2014 RBG-47522 Page 2 of 2 cc:
U. S. Nuclear Regulatory Commission Region IV 1600 East Lamar Blvd.
Arlington, TX 76011-4511 NRC Sr. Resident Inspector P. 0. Box 1050 St. Francisville, LA 70775 INPO (via ICES reporting)
Central Records Clerk Public Utility Commission of Texas 1701 N. Congress Ave.
Austin, TX 78711-3326 Department of Environmental Quality Office of Environmental Compliance Radiological Emergency Planning and Response Section Ji Young Wiley P.O. Box 4312 Baton Rouge, LA 70821-4312
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 01131/2017 (02-2014)
Estimated burden per response to comply with this mandatory collection request 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />.
Reported lessons learned are incorporated into the licensing process and fed back to industry.
Send comments regarding burden estimate to the FOIA, Privacy and Information Collections LBranch (T.5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by LICENSEE EVENT REPORT (LER) internet e-mail to Infocollects.Resource@nrc.gov, and to the Desk Officer, Office of Information and (See Page 2 for required number of Regulatory Affairs, NEOB-10202 (3150-0104), Office of Management and Budget Washington, DC 20503. If a means used to impose an information collection does not display a currently valid OMB digits/characters for each block) control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 3. PAGE River Bend Station - Unit 1 05000 458 1 OF 4
- 4. TITLE Reactor Scram Due to Average Power Range Monitor High-flux Signal Following a Malfunction of the Main Turbine Electro-hydraulic System
- 5. EVENT DATE
- 6. L.R NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED MONTH DAY YEAR YEAR SEQUENTIAL REV M
FACIUTY NAME DOCKET NUMBER NUMBER MONTH DAY YEAR 05000 FACIUTY NAME DOCKET NUMBER 10 17 2014 2014 -
002 00 12 16 2014 05000
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR§: (Check all that apply)
[]
20.2201(b)
E] 20.2203(a)(3)(i)
[
50.73(a)(2)(i)(C)
[
50.73(a)(2)(vii)
[] 20.2201(d)
[] 20.2203(a)(3)(ii)
El 50.73(a)(2)(ii)(A)
E] 50.73(a)(2)(viii)(A) 1:1 20.2203(a)(1)
[]20.2203(a)(4)
[j 50.73(a)(2)(ii)(B)
Ej 50.73(a)(2)(viii)(B)
E] 20.2203(a)(2)(i)
[
50.36(c)(1)(i)(A)
[
50.73(a)(2)(iii)
[
50.73(a)(2)(ix)(A)
- 10. POWER LEVEL
-- 20.2203(a)(2)(ii)
[]
50.36(c)(1)(ii)(A)
[] 50.73(a)(2)(iv)(A)
L] 50.73(a)(2)(x)
Fl 20.2203(a)(2)(iii)
[
50.36(c)(2)
[I 50.73(a)(2)(v)(A)
[
73.71(a)(4) 100 [l 20.2203(a)(2)(iv)
[
50.46(a)(3)(ii)
[] 50.73(a)(2)(v)(B)
[] 73.71(a)(5)
Fl 20.2203(a)(2)(v)
[
50.73(a)(2)(i)(A)
[
50.73(a)(2)(v)(C)
OTHER 20.2203(a)(2)(vi)
[
50.73(a)(2)(i)(B) 50.73(a)(2)(v)(D)
Specify in Abstract below or in
REPORTED CONDITION On October 17, 2014, at approximately 3:03 a.m. CDT, a reactor scram occurred in response to a high neutron flux signal from the average power range monitors (APRMs). The plant was operating at 100 percent power at the time. Immediately prior to that signal, an apparent malfunction in the main turbine electro-hydraulic control (EHC) (**JI**) system caused both the main turbine steam bypass valves to fully open, and also commanded all four main turbine control valves to close. The resulting increase in reactor steam pressure caused reactor power to immediately rise to the trip setpoint of the APRMs, at which point the actuation of the reactor protection system (RPS) (**JC**) occurred.
All reactor control rods inserted as designed in response to the RPS actuation, and operators implemented the appropriate response procedures. After the scram occurred, an operator in the auxiliary control room erroneously removed all the main condensate system demineralizers (**SF**) from service, isolating condensate flow to the suction header of the main feedwater pumps (**SJ**). The "C" feedwater pump tripped on low suction pressure; the "A"' and "B" feedwater pumps had been previously secured by the operators.
The mis-operation of the demineralizer system was promptly recognized and corrected, and the main feedwater system was restored to service within approximately six minutes. In the interim, reactor water level decreased to Level 3, causing a second actuation of the RPS system, but remained well above the point at which actuation of the emergency core cooling systems is required. The Level 3 signal also caused an automatic actuation of the containment isolation valves in the suppression pool cooling system, as designed.
Reactor recirculation pump "B" failed to downshift to slow speed, and instead, tripped off. A controlled plant cooldown was commenced, and proceeded into cold shutdown conditions.
This event is being reported in accordance with 10CFR50.73(a)(2)(iv) as an automatic actuation of the RPS system.
INVESTIGATION and IMMEDIATE ACTIONS
- 1. EHC Malfunction An investigation team was formed to analyze the operating parameters of the EHC system and determine the source of the malfunction.
Earlier in that same shift, there were two actuations of the main control room alarm indicating that the turbine steam bypass valves were open. At 7:26 p.m., the alarm actuated and immediately cleared. Plant computer data indicated that both bypass valves had opened slightly, while the number 4 turbine control valve closed slightly., No accompanying changes were noted in reactor pressure or in any EHC parameters. At 11:14 p.m., the alarm actuated a second time. Computer data then showed that turbine bypass valve no. I had fully opened, and bypass valve no. 2 had opened approximately 20 percent. Turbine control valve no. 4 closed significantly, and there was a slight change in the turbine steam flow reference signal in the EHC system.
Analysis of EHC parameters immediately prior to the scram found that the steam flow reference signal went from 92 percent to negative 41 percent with no corresponding change in total steam flow demand. The error in the steam flow reference signal cleared in approximately five seconds, but the resulting pressure transient had caused the reactor scram by that point. The movements of the bypass valves and turbine control valves were appropriate for the loss of the steam flow reference signal.
A detailed circuit analysis was performed to identify which components within the system could cause the loss of this signal. Efforts to recreate the loss of the signal were unsuccessful. Based on the system response earlier in the shift, and the lack of any apparent fault, it was concluded that an intermittent failure with the steam flow reference signal was occurring.
The investigation initially postulated that failure in any of these five EHC circuit cards and their subcomponents could have caused this event:
" pressure load gate amplifier
" control valve amplifier input standby transfer
" load limit set runback analog / logic
" control valve flow reference signal
" load limit and load set nnback All these circuit cards, except one, were replaced and sent to the vendor for analysis. Discussion with the vendor concerning the load limit and load set runback card concluded that its potential role as a source of the failure was extremely low.
If a definite cause of the EHC malfunction is found, that information will be provided in a supplement to this report.
- 2. Mis-operation of the main feedwater system When the operators in the auxiliary control room heard the plant announcement of a reactor scram, they began removing main condensate demineralizers from service to maintain proper system parameters for the expected decrease in feedwater flow. However, they inappropriately isolated all the demineralizers, isolating all flow to the main feedwater pump suction header. Main control room operators had already secured two of the three pumps; the last pump tripped on low suction pressure. The error was promptly recognized, and demineralizers were restored to service, allowing the re-start of a feedwater pump.
A human performance error review was conducted, which found the following:
- - The auxiliary control room operators did not correctly follow the system operating procedure guidance on maintaining flow though individual demineralizers within specifications as overall system flow decreased. The operators imposed unnecessary haste by focusing on demineralizer channeling effects, and did not demonstrate knowledge of integrated system operation.
- Procedural guidance for the removal of condensate demineralizers from service following a scram was less than adequate, in that it promoted the practice that operators in the field have the authority to determine when they can perform actions without specific direction from the main control room.
- Operations department management had not clearly defined standards and expectations to preclude auxiliary operators from taking actions without direction from main control room operators, or a supervisor, during certain plant conditions. In this event, the condensate demineralizers were removed from service and then returned to service without direction from main control room, with the intent of not distracting those operators from the scram recovery. The system operating procedure promotes this long-standing practice.
CORRECTIVE ACTION TO PREVENT RECURRENCE
- 1. EHC malfunction During the upcoming refueling outage in February 2015, an evaluation of a potential replacement of the obsolete card will b2 completed. A complete replacement with a digital EHC system is planned for the refueling outage to occur in early 2017. These actions are being tracked in the corrective action program.
- 2. Mis-operation of the main condensate demineralizer system The system operating procedure for the main condensate demineralizers has been revised to clarify the precautions regarding how many demineralizers are to remain in service. The scram response procedure has been revised to add a step for the main control room operators to communicate with the auxiliary control room operator regarding system operations.
PREVIOUS OCCURRENCE EVALUATION No event reported by RBS within the last three years was caused by unexplained, erratic signals originating within the main turbine EHC control circuits.
SAFETY SIGNIFICANCE
At the time of the reactor scram, the reactor core isolation cooling system was out of service for planned maintenance. No plant parameters that would have required its actuation were exceeded. No other engineered safety features were out of service at the time.
The plant response to this transient was as designed. Thus, this event was of minimal significance to the health and safety of the public.
(NOTE: Energy Industry System Identification codes are annotated as (**XX**).)