05000219/LER-2006-004, Re Operation Exceeding Maximum Power Level
| ML070380317 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 02/02/2007 |
| From: | Rausch T AmerGen Energy Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 2130-07-20450 LER 06-004-00 | |
| Download: ML070380317 (5) | |
| 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 |
| 2192006004R00 - NRC Website | |
text
AmerGen.
AmerGen Energy Company www.exeloncoTp.com An Exelon Company Oyster Creek US Route 9 South, P.O. Box 388 Forked River, NJ 08731-0388 10 CFR 50.73 February 2, 2007 2130-07-20450 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 - 0001 Oyster Creek Generating Station Facility Operating License No. DPR-16 NRC Docket No. 50-219
Subject:
Licensee Event Report 2006-004-00, Operation Exceeding Maximum Power Level Enclosed is Licensee Event Report 2006-004-00, Operation Exceeding Maximum Power Level. This event did not affect the health and safety of the public or plant personnel.
This event did not result in a safety system functional failure. There are no new regulatory commitments made in this LER submittal.
If any further information or assistance is needed, please contact Richard Milos, Regulatory Assurance at 609-971-4973 or Sylvain Schwartz, Engineering, at 609-971-4558.
Sincerely, o
ch, Vice President Oyster Creek Generating Station
Enclosure:
NRC Form 366, LER 2006-004-00 cc:
Administrator, USNRC Region I USNRC Project Manager, Oyster Creek USNRC Senior Resident Inspector, Oyster Creek File No. 07050
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 06/3012007 (6-2004)
, the NRC may not conduct or digits/characters for each block) sponsor, and a person is not required to respond to, the information collection.
- 3. PAGE
,Oyster Creek, Unit 1 05000 ;2119'3/4 1
OF 4
- 4. TITLE
'3/4>>>Operation Exceedin~g Maximium Power. Level
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED 11111UENTIAL REV FjAACIII T
YN A M E DOCKET NUMBER MONTH DAY YEAR YEAR NUMBER NO MONTH DAY YEAR 05000 12 08 2006 2006 004 00 02 02 2007' FACILITY NAME DOCKET NUMBER 12_____08_____
__200611._:
2006__70005000
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)
N 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) 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) 100 20.2203(a)(2)(iii) 50.36(c)(2) 50.73(a)(2)(v)(A) 73.71 (a)(4)
'0 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 20.2203(a)(2)(vi)
X 50.73(a)(2)(i)(B) 50.73(a)(2)(v)(D)
Specify in Abstract below I
or in (If more space is required, use additional copies of (If more space is required, use additional copies of NRC Form 366A)
Analysis of the Event
There were no actual safety consequences associated with this event. The potential safety consequences of this event were also minimal. Operators took prompt action to decrease reactor power, which terminated the overpower condition in approximately 40 seconds.
The peak reactor power of 102.46% was of very short duration and would have a negligible impact on the decay heat for a Loss of Coolant Accident (LOCA). The LOCA analysis assumes that the initial conditions of the event are at 102% power at steady state, due to uncertainties in measurement and detection of core power. These analysis assumptions bound this event.
Opening of an EMRV at power is an evaluated event that is significantly bounded by the limiting events evaluated each reload and the associated operating limits documented in the Core Operating Limits Report. As such, no reactor vessel (pressure boundary) or fuel safety limits or other Specified Acceptable Fuel Design Limits (SAFDLs) were challenged A Prompt Investigation has been performed in accordance with current plant procedure "Event Response Guidelines". The investigation team determined that there was no evidence that this event was influenced by human performance. The investigation determined that the most likely cause of this event was a malfunction of the IA83D Pressure Sensor. A review of the as left data for EMRV Pressure Sensor test and calibration surveillance indicates that during the last surveillance test prior to this event, the set point was within specification. However, set point drift of the pressure switch since last surveillance occurred, lowering the set points by approximately 20-25 psig. The as found set point for IA83D after the event was still above the operating pressure at the time of the event. Component history review indicates that in July and August 2005, the IA83D pressure switch had excessive drift identified during the EMRV Pressure Sensor Test and Calibration. This component had a history of drifting in the past year. Furthermore, Operating Experience from Oyster Creek in 1993 indicates an EMRV pressure switch failed due to dirty contacts from natural aging. Based on the above, the degraded pressure switch was considered to be the most probable cause of the spurious opening of the D EMRV.
Cause of the Event
The most probable cause of the event was malfunction of the pressure switch connected to the D EMRV.U.S. NUCLEAR REGULATORY COMMISSION (1-2001)
LICENSEE EVENT REPORT (LER)
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE Oyster Creek, Unit 1 05000219 YEAR SEQUENTIAL REVISION O1 04NUMBER NUMBER 2006 004 00 1 4OF 4
- 17. NARRA I IVE (If more space is required, use additional copies of NR; i-orm Jw366)
Corrective Action Completed.
The IA83D pressure switch and its associated relay were replaced.
Corrective Action Planned.
The new pressure switch performance will be monitored for a year to determine if periodic replacement of the pressure switches is warranted.
Previous Similar Occurrences There were no previous occurrences of a spurious EMRV opening that caused reactor power to exceed the licensed maximum power level.
Component Data.
Component:
Cause
System:
Component:
Manufacturer:
Model number:
IA83D, Reactor High Pressure Pressure switch degradation SB (Main/Reheat Steam System)
PS (Pressure Switch)
Barksdale B2S-M12SS -
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