05000313/LER-2010-004, Regarding Automatic Reactor Protection System Actuation That Resulted in a Reactor Trip Due to Inadequate Procedure Use and Adherence and Workers Acting Independently
| ML101750581 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 06/22/2010 |
| From: | David Bice Entergy Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 1CAN061002 LER 10-004-00 | |
| Download: ML101750581 (6) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 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)(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 |
| LER closed by | |
| IR 05000313/2012003 (9 August 2012) | |
| 3132010004R00 - NRC Website | |
text
1CAN061002 June 22, 2010 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001
Subject:
Licensee Event Report 50-313/2010-004-00 Arkansas Nuclear One - Unit 1 Docket No. 50-313 License No. DPR-51
Dear Sir or Madam:
In accordance with 50.73(a)(2)(iv)(A), enclosed is the subject report concerning an automatic Reactor Protection System (RPS) actuation that resulted in a reactor trip.
There are no commitments contained in this submittal.
Sincerely, DBB/slp Enclosure - LER 50-313/2010-004-00 Entergy Operations, Inc.
1448 S.R. 333 Russellville, AR 72802 Tel 479-858-4710 David B. Bice Acting Manager, Licensing Arkansas Nuclear One
1CAN061002 Page 2 cc:
Mr. Elmo Collins Regional Administrator U. S. Nuclear Regulatory Commission Region IV 612 E. Lamar Blvd., Suite 400 Arlington, TX 76011-4125 NRC Senior Resident Inspector Arkansas Nuclear One P.O. Box 310 London, AR 72847 Institute of Nuclear Power Operations 700 Galleria Parkway Atlanta, GA 30339-5957 LEREvents@inpo.org
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION (9-2007)
LICENSEE EVENT REPORT (LER)
(See reverse for required number of digits/characters for each block)
APPROVED BY OMB NO. 3150-0104 EXPIRES 8/31/2010 the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 3. PAGE Arkansas Nuclear One, Unit 1 05000313 1
OF 4
- 4. TITLE Automatic Reactor Protection System Actuation that Resulted in a Reactor Trip Due to Inadequate Procedure Use and Adherence and Workers Acting Independently
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED MONTH DAY YEAR YEAR SEQUENTIAL NUMBER REV NO MONTH DAY YEAR FACILITY NAME DOCKET NUMBER FACILITY NAME 04 25 2010 2010 - 004 - 00 06 22 2010 DOCKET NUMBER
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply) 20.2201(b) 20.2203(a)(3)(i) 50.73(a)(2)(i)(C) 50.73(a)(2)(vii) 20.2201(d) 20.2203(a)(3)(ii) 50.73(a)(2)(ii)(A) 50.73(a)(2)(viii)(A) 20.2203(a)(1) 20.2203(a)(4) 50.73(a)(2)(ii)(B) 50.73(a)(2)(viii)(B) 1 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) 50.73(a)(2)(x) 20.2203(a)(2)(iii) 50.36(c)(2) 50.73(a)(2)(v)(A) 73.71(a)(4) 20.2203(a)(2)(iv) 50.46(a)(3)(ii) 50.73(a)(2)(v)(B) 73.71(a)(5) 20.2203(a)(2)(v) 50.73(a)(2)(i)(A) 50.73(a)(2)(v)(C)
OTHER-Specify in Abstract below or in B. Event Description - continued Indicated excore NI power had rose from 30% to 49.55% resulting in a high neutron flux trip on RPS Channel C. The amount of energy added to the Reactor Coolant System (RCS) [AB] by the control rod withdrawal caused RCS temperature and pressure to rise resulting in a high RCS pressure trip on "A" Channel RPS. The trip of these two channels caused an automatic reactor trip at ~2126 CDT.
Post reactor trip system response was normal as expected. All control rods fully inserted into the core and no safety systems, other than RPS actuated. Emergency Feedwater (EFW) [BA] did not actuate and was not needed. No primary safety valves lifted. Seven secondary safety valves lifted and subsequently reseated. The plant stabilized in Hot Standby (Mode 3) conditions.
C. Root Cause The failure to appropriately use and adhere to procedures was determined to be the major factor in this event. The procedure step was not performed as written and not read verbatim to the operator as required by procedure guidance. The I&C procedure specifically required Operations to place the Reactor Demand H/A and Diamond Rod Control Station into manual using OP-1105.004, "Integrated Control System Operating Procedure." The lead technician should have communicated the procedure step exactly as written. This procedure step clearly stated that the Diamond Rod Control was to be placed in manual.
Additionally, the CBOT made the decision to act independently without guidance from supervision by allowing the I&C technician to proceed by communicating an interpretable, non-standard message. The CBOT also failed to properly verify and have his action peer checked by an independent control room operator.
D. Corrective Action
As a result of this event, an immediate Operations and Maintenance supervisor meeting was conducted. This meeting was followed by a site stand-down. Additional control room oversight was established which included focus on Operations and Maintenance interface.
Interim guidance was established for work activities conducted in the control room to ensure plant conditions are appropriately established prior to commencing maintenance.
The lessons learned from this event, supervisory skills, and human performance training are being incorporated into Operations and Maintenance requalification training.
E. Safety Significance
A risk evaluation for the ANO-1 automatic reactor trip during NI calibration was performed to evaluate the condition with respect to safety risk for the general public, nuclear safety, radiological safety, and industrial safety. The change in core damage risk for ANO-1 from the event was calculated and compared to the criteria for significance, as defined in MD 8.3, RG 1.174, and EPRI TR-105396. The risk for the transient event is deemed insignificant. Post reactor trip system response was normal as expected. All control rods fully inserted into the core and no safety systems, other than RPS actuated. EFW did not actuate and was not needed and no primary safety valves lifted. Seven secondary safety valves lifted and subsequently reseated.
The plant stabilized in Hot Standby (Mode 3) conditions.
The increased risk of having a Pellet Cladding Interaction (PCI) fuel failure at ANO-1 due to the power increase is judged to be insignificant. This is based on review of the available data from ANO-1, mechanisms that can result in PCI fuel failures, engineering judgment, and discussions with the vendor. RPS actuated as designed to shutdown the reactor. Post reactor trip system response was normal for the plant and no industrial safety issues occurred. There were no radiological safety issues created by this event and at no time during the course of the event was the general safety of the public compromised.
F. Basis for Reportability Actuation of the RPS is reportable under 10 CFR 50.73(a)(2)(iv)(A). Additionally, actuation of the RPS when the reactor is critical is reportable under 10 CFR 50.72(b)(2)(iv)(B). An immediate notification was made to the NRC Operations Center on April 25, 2010.
G. Additional Information
During the last five years, there was one other previous similar event reported as LER-2005-003-00 for ANO-1 concerning actuation of the RPS; however, the root cause was not similar. This condition was due to a Main Turbine trip caused by low turbine bearing lube oil pressure.
Energy Industry Identification System (EIIS) codes are identified in the text as [XX].