05000306/LER-2007-001, Re Unit 2 Reactor Trip
| ML071550268 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 06/04/2007 |
| From: | Wadley M Nuclear Management Co |
| To: | Document Control Desk, Plant Licensing Branch III-2 |
| References | |
| L-PI-07-038 LER 07-001-00 | |
| Download: ML071550268 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(v), Loss of Safety Function |
| 3062007001R00 - NRC Website | |
text
Prairie Island Nuclear Generating Plant Operated by Nuclear Management Company, LLC June 4,2007 L-PI-07-038 10 CFR 50.73 U S Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Prairie Island Nuclear Generating Plant Unit 2 Docket 50-306 License No. DPR-60 LER 2-07-01, Unit 2 Reactor Trip The Licensee Event Report (LER) for this occurrence is attached. Nuclear Management Company, LLC (NMC) notified the NRC of this event, as required by 10 CFR 50.72(b)(2)(iv)(A) and 10 CFR 50.72(b)(2)(iv)(B) on April 5, 2007. Please contact us if you require additional information related to this event.
Summarv of Commitments This letter contains one new commitment and no revisions to existing commitments:
NMC will submit a supplement to this LER after the associated Root Cause Evaluation has been completed.
Michael D. Wadley Site Vice president, Prairie Island Nuclear Generating Plant Nuclear Management Company, LLC Enclosure cc:
Administrator, Region Ill, USNRC Project Manager, Prairie Island, USNRC Resident Inspector, Prairie Island, USNRC Glenn Wilson, State of Minnesota 171 7 Wakonade Drive East Welch, Minnesota 55089-9642 Telephone: 651.388.1 121
ENCLOSURE LICENSEE EVENT REPORT 2-07-01 3 pages follow
Estimated burden per response to comply with this mandatory collection request: 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-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 and FOIAIPrivacy LICENSEE EVENT REPORT (LER)
Service Branch (T-5 F52), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by internet e-mail to infocollects@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (31 50-0066),
Office of Management and Budget. Washington, DC 20503. If a means used to impose an information collection does A Safety Injection (SI) actuation occurred resulting in Reactor Protection System (RPS) actuation. Train A SI was in "Test" at the time and should not have caused the RPS trip. At approximately 091 3 the operating crew manually actuated Train B SI as required by emergency operating procedures. Reactor Coolant System (RCS) pressure momentarily decreased below the shutoff head of the high head Emergency Core Cooling System (ECCS) pumps during the transient, resulting in ECCS discharge to the RCS. At approximately 0920 safety injection was terminated per emergency operating procedures. All systems operated as expected and operator response and recovery actions were as expected. U.S. NUCLEAR REGULATORY COMMISSION (1-2001)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION TEXT (If more space is required, use additional copies of NRC Form 366A) (17)
EVENT DESCRIPTION
On April 5, 2007, Prairie Island Nuclear Generating Plant (PINGP) Unit 2 was operating at 100%
power. At approximately 0908 CDT, durin surveillance testing of Unit 2 Train A safeguards logic at 8
power, a spurious Train A Safety Injection (SI) actuation occurred resulting in Reactor Protection system2 (RPS) actuation. Train A SI was in "Test" at the time and should not have caused the RPS trip. At approximately 0913 the operating crew manually actuated Train B SI as required by emergency operating procedures. All automatic actions for a reactor trip and safety injection occurred as required. Reactor Coolant system3 (RCS) pressure momentarily decreased below the shutoff head of the high head Emergency Core Cooling System (ECCS) pumps during the transient, resulting in ECCS discharge to the RCS. At approximately 0920 safety injection was terminated per emergency operating procedures. All systems operated as expected and operator response and recovery actions were as expected.
EVENT ANALYSIS
The trip of the Unit 2 reactor and the actuation of the emergency core cooling system are required to be reported per 10 CFR 50.73(a)(2)(iv)(A).
Impact on Safetv Svstem Functional Failure Performance Indicator This event did not result in a loss of the safety injection system since the Unit 2 Train B safety injection was manually actuated and performed as expected. Therefore, this event does not represent a loss of safety function. Consequently, this event is not reportable per 10CFR 50.73(a)(2)(v).
SAFETY SIGNIFICANCE
The plant was stabilized in Mode 3 after the trip and all systems performed as expected in response to the reactor trip. Therefore, this event did not affect the health and safety of the public.
' Ells System Code: BQ Ells System Code: JC 3 Ells System Code: AB U.S. NUCLEAR REGULATORY COMMISSION (1-2001)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION I/
FACILITY NAME (I) 11 DOCKET NUMBER (2) (1 LER NUMBER (6)
U TEXT (If more space is required, use additional copies of NRC Form 366A) ( I 7 )
1 Prairie Island Nuclear Generating Plant Unit 2 11 05000306
CAUSE
Initial investigation of the cause of the automatic reactor trip was determined to be a deficiency with the safety injection relay equipment. Instrument and Control technicians discovered a safety injection relay4 with high contact resistance. The high resistance contact caused the relay to not reset when exiting the safeguards logic test. With the relay not reset, and safeguards logic not in test, a spurious SI actuation occurred. The defective relay was replaced and tested to verify proper operation prior to the Unit 2 reactor startup. NMC is conducting a Root Cause Evaluation (RCE) to determine the cause of the event; after the RCE is completed, a supplement to this LER will be submitted.
YEAR
CORRECTIVE ACTION
SEQUENTIAL Immediate:
- 1. The defective relay was replaced and tested.
- 1.
NUMBER 07 01 0
Subsequent:
- 2. Safeguards logic test procedures have been revised to preclude a similar event.
Planned:
- 3. The Root Cause Evaluation may recommend additional corrective actions when it is complete.
The recommended corrective actions will be described in the supplement to this LER.
PREVIOUS SIMILAR EVENTS
Both Unit 1 and Unit 2 have experienced unplanned reactor trips in the past. The only other unplanned reactor trip in the last three years occurred in 2006. The 2006 reactor trip was not caused by the safety injection system. There were no other reportable events in the past three years that were related to relay failures or inadvertent SI actuation.
4 Ells Component Code: RLY