05000313/LER-2008-001, Regarding Two Manual Reactor Trips from Power in Response to Abnormal Control Rod Movement Caused by Control Rod Drive Control System Component Degradation
| ML090410662 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 02/10/2009 |
| From: | David Bice Entergy Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 1CAN020904 LER 08-001-00 | |
| Download: ML090410662 (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 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2) |
| 3132008001R00 - NRC Website | |
text
1CAN020904 February 10, 2009 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001
Subject:
Licensee Event Report 50-313/2008-001-00 Arkansas Nuclear One - Unit 1 Docket Nos. 50-313 License Nos. DPR-51
Dear Sir or Madam:
In accordance with 10CFR50.73(a)(2)(iv)(A), enclosed is the subject report concerning two manual reactor trips from power.
There are no new commitments contained in this submittal.
Sincerely, DBB/dce Enclosure Entergy Operations, Inc.
1448 S.R. 333 Russellville, AR 72802 Tel 479-858-4710 David B. Bice Acting Manager, Licensing Arkansas Nuclear One
1CAN020904 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 Estimated burden per response to comply with this mandatory information 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 Records Management Branch (T-6 E6), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by internet e-mail to bjs1@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202 (3150-0104), Office of Management and Budget, Washington, DC 20503. If a means used to impose information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 3. PAGE Arkansas Nuclear One 05000313 1
OF 4
- 4. TITLE Two Manual Reactor Trips From Power in Response to Abnormal Control Rod Movement Caused by Control Rod Drive Control System Component Degradation.
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR SEQUENTIAL NUMBER REV NO MONTH DAY YEAR 05000 FACILITY NAME DOCKET NUMBER 12 12 2008 2008 - 001 - 00 02 10 2009 05000
- 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) 32/100 20.2203(a)(2)(vi) 50.73(a)(2)(i)(B) 50.73(a)(2)(v)(D)
OTHER-Specify in Abstract below or in C. Root Cause The eight control rods of group 7 are typically positioned under automatic control based on demand signals from the Integrated Control System (ICS) to the CRDCS to maintain reactor power at the required level. The ICS signal is provided to the programmer via relays to move rods depending on the command initiated. Firing of silicone controlled rectifiers (SCR) is supervised by a programmer dedicated to group 7. The programmer is a microcontroller based component which responds to commands from ICS (or alternately, manual control) by sequencing the firing of the six phases of power as needed to provide rod motion. If no movement requests are initiated the microcontroller will maintain two phases continuously energized to hold the rods in a fixed position. If the programmer does not provide any SCR firing demand outputs, the rods will be released.
The programmer control power is derived from two redundant AC power sources. The primary source is provided from B631, the secondary source is provided from A501. A control power Automatic Bus Transfer (ABT-3) scheme provides power to the programmer from the designated primary AC source and rapidly switches to the secondary source if a loss of primary voltage is sensed. This is accomplished via the switching of relays K1 and K2 located in the ABT-3.
The most probable root cause of the dropping of the group 7 rods is the intermittent failure of the K1 and/or K2 ABT relays. The K1 and K2 relays were original equipment, and were found to be degraded.
A possible root cause is a failure associated with the programmer assembly. The programmer assembly presents a single point failure vulnerability. The failure of the 15 V power supply (internal to the programmer assembly) would directly cause the failure of the Programmer Micro Controller Unit.
D. Corrective Actions
The degraded K1 and K2 relays and the programmer assembly were replaced and tested.
Additional online monitoring requirements were implemented to obtain additional diagnostic information.
Additional inspection requirements were implemented for the CRDCS programmers.
Preventive Maintenance Program strategy improvement was initiated for the CRDCS. U.S. NUCLEAR REGULATORY COMMISSION (9-2007) LICENSEE EVENT REPORT (LER)
CONTINUATION SHEET
- 1. FACILITY NAME
- 2. DOCKET NUMBER (2)
- 6. LER NUMBER
- 3. PAGE YEAR SEQUENTIAL NUMBER REVISION NUMBER Arkansas Nuclear One - Unit 1 05000313 2008 001 00 4
OF 4
E. Safety Significance
The safety significance associated with these occurrences is considered to be minimal.
Occurrence 1: Although Group 7 Control Rods re-latched and began to withdraw, prompt operator action to perform a manual reactor trip prevented any automatic protective setpoints from being approached. At the time of the reactor trip, the nuclear overpower trip setpoint was set lower than normal, 40% versus 104.9%, while performing physics testing following startup from 1R21 refueling outage. Post trip plant response was normal with no complications and the plant was stabilized in Mode 3, Hot Standby, without incident. All safety systems performed as designed with no automatic safety features actuating.
Occurrence 2: No outward rod motion occurred as a result of this event due to the fact that Group 7 rods re-latched at different heights. This caused an out inhibit associated with the CRD system to prevent outward rod motion. Post trip transient response for this event was normal with no complications and the plant was stabilized in Mode 3, Hot Standby, without incident. All safety systems performed as designed with no automatic safety features actuating. Main Steam safeties lifted for approximately 1 minute as a result of this reactor trip. This is considered to be a normal response due to plant power at 100% at the onset of the event. All main steam safeties reseated properly as designed.
Nuclear safety, Industrial Safety, Radiological Safety and General safety of the public were not challenged at any time during these events.
F. Basis for Reportability The first occurrence of the event was reported on December 12, 2008, at 1048 CST, as a 4-hour Non-Emergency 10CFR50.72(b)(2)(iv)(B) RPS Actuation (scram).
The second occurrence of the event was reported on December 20, 2008, at 1323 CST, as a 4-hour Non-Emergency 10CFR50.72(b)(2)(iv)(B) RPS Actuation (scram).
A manual reactor trip in response to actual plant conditions is reportable pursuant to 10CFR50.73(a)(2)(IV)(A).
G. Additional Information
There are no previous similar events reported by ANO.
Energy Industry Identification System (EIIS) codes are identified in the test as [XX].