On May 22, 2010 at 1645 with the Millstone Power Station Unit 2 at 100% power in Mode 1, the reactor was manually tripped by operators due to a high water level condition in steam generator (S/G) number 2. Investigation determined that level oscillations occurred because of S/G level control problems associated with number 2 feedwater ( FW) regulating valve.
Safety systems functioned as expected based upon the signals received. The operators took actions as trained and in accordance with established procedures. No equipment was damaged as a result of the event. The unit was brought to a-stable condition in hot-standby. The-cause of the 'reactor trip was vibration induced wear of the number 2 FW regulating valve positioner beam screw. The degraded valve positioner was replaced and the frequency of on-line preventive maintenance was changed.- Design changes are being evaluated to reduce -FW regulating valve positioner beam screw wear.
This event is being reported pursuant to 10 CFR 50.73(a)(2)(ivXA) as-an-event that resulted in manual or ' automatic actuation of systems listed in 10CFR50.73(a)(2)(iv)(B). |
1. !Event Description On May 22, 2010 at 1645 with the Millstone Power Station Unit 2 (MP-32) at 100% power in Mode 1, the reactor [AA] [RCT] was manually tripped by operators due to a high water level condition in the number 2 steam generator oscillations occurred because of feedwater level [SJ] control problems associated with the number 2 feedwater regulating valve [V] (2-FW-518). When the number 2 S/G water level rose to 84.4% on narrow range instrumentation, the operators manually tripped the reactor in accordance with established operating procedures.
In accordance with their training and also established procedures, the operators closed the number 2 steam generator feedwater regulating valve block valve-(2-FW-42B) and tripped both main feedwater pumps.
The auxiliary feedwater [BA] system automatically actuated,as expected on low level in number 1 steam generator at the setpoint of 18.1% water level. The steam dump valves continued to control main steam [SB] pressure.
Safety systems functioned as expected based upon the signals received. The operators took actions as trained and in accordance with established procedures. No equipment was damaged as a result of the-event. The unit was brought to a stable condition in hot-standby (Mode 3) and the degraded valve positioner was replaced.
This event is being reported pursuant to 10 CFR 50.73(a)(2XivXA) as an event that resulted in manual or automatic actuation of the reactor protection system and the auxiliary feedwater system.
This-event is being reported pursuant to 10CFR50.73(a)(2Xiv)(A) as an event that resulted in manual or automatic actuation of the reactor protection system and the auxiliary feedwater system.
2. Cause The cause of this event was determined to be vibration induced worn threads on the number 2 feedwater regulating valve (2-FW-51B) positioner beam screw.
3. Assessment of Safety Consequences The safety consequences associated with this event were very low. The reactor was manually tripped by operators when the number 2 S/G water level rose to 84.4% and the feedwater control system was not responding as expected. All control rods inserted into the reactor following the trip.
There was no safety injection system actuation. There was no significant increase in core power prior to the manual reactor trip. Neither the departure from nucleate boiling or fuel centerline melt fuel design limits were challenged. As such, there were no challenges to the fuel, reactor coolant system or containment fission product barriers.
NRC 'FORM 366A (9-2007)
- 4. Corrective Action The following compensatory actions have been completed:
- The degraded feedwater regulating valve successfully retested and returned to service on May 23, 2010.
- The on-line preventative maintenance frequency has been increased for installed feedwater regulating valve positioners to improve reliability.
Long term corrective actions:
- Design changes are being evaluated to reduce FRV positioner beam screw wear.
Additional corrective actions are being taken in accordance with the station's corrective action program.
5. Previous Occurrences No previous similar events/conditions were identified.
Energy Industry Identification System (ElIS) codes are identified in the-text as [XX]
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| | | Reporting criterion |
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| 05000346/LER-2010-001 | Steam and Feedwater Rupture Control System Re-Energizes in a Blocked Condition On Loss of Offsite Power | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition
10 CFR 50.73(a)(2)(v), Loss of Safety Function
10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | | 05000334/LER-2010-001 | Beaver Valley Power Station Unit Number 1 05000334 1 of 5 | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition
10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor
10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident
10 CFR 50.73(a)(2)(vii), Common Cause Inoperability
10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | | 05000346/LER-2010-002 | 5501 North State Route 2
FirstEnergy Nuclear Operating Company Oak Harbor, Ohio 43449
Barry S. Allen 419-321-7676
Vice President - Fax: 419-321-7582
•
September 30, 2010
L-10-258 10 CFR 50.73
ATTN: Document Control Desk
United States Nuclear Regulatory Commission
Washington, D.C. 20555-0001
SUBJECT:
Davis-Besse Nuclear Power Station
Docket Number 50-346, License Number NPF-3
Licensee Event Report 2010-002 Revision 01
Enclosed is Revision 01 to Licensee Event Report (LER) 2010-002, "Control Rod Drive
Nozzle Primary Water Stress Corrosion Cracking and Pressure Boundary Leakage."
This LER is being reported in accordance with 10 CFR 50.73(a)(2)(i)(B), operation in a
condition prohibited by the Technical Specifications, and 10 CFR 50.73(a)(2)(ii)(A),
condition of the plant, including its principal safety barriers, being seriously degraded.
This LER is being revised to provide results of the completed Root Cause evaluation.
There are no new regulatory commitments contained in this letter or its enclosure. The
actions described represent intended or planned actions and are described for
information only. If there are any questions or if additional information is required,
please contact Mr. Patrick J. McCloskey, Manager, Site Regulatory Compliance, at
(419) 321-7274.
Sincerely,
Barry S. Allen
GMW
•
Enclosure: LER 2010-002-01
.CC: NRC Region III Administrator
NRC Resident Inspector
NRR Project Manager
Utility Radiological Safety Board
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 EXPIRES 8/31/2010
(6-2004)
Estimated burden per response to comply with this mandatory collection
request: 80 hrs. Reported lessons learned are incorporated into the
licensing process and fed back to industry. Send comments regarding burden
estimate to the Records and FOIA/Privacy Service Branch (T-5 F52), U.S.
Nuclear Regulatory Commission, Washington, DC 20555L0001, or by intemetLICENSEE EVENT REPORT(LER) e-mail to infocoltects@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 an information
collection does not display a currently valid OMB control number, the NRC may(See reverse for required number of not conduct or sponsor, and a person is not required to respond to, the
digits/characters.for each block) information collection.
1. FACILITY NAME 2. DOCKET NUMBER 3. PAGE ,
Davis-Besse Nuclear Power Station 05000346 1 OF 4
4. TITLE '
Control Rod Drive Nozzle Primary Water Stress Corrosion Cracking and Pressure Boundary Leakage
., | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded | | 05000336/LER-2010-002 | Manual Reactor Trip on High Steam-Generator Water Level | 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
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