05000341/LER-2008-005, Regarding Loss of High Pressure Coolant Injection System Safety Function Due to Closure of Steam Supply Valve
| ML090420065 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 01/26/2009 |
| From: | Plona J DTE Energy |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NRC-09-0003 LER 08-005-00 | |
| Download: ML090420065 (4) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(v), Loss of Safety Function 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)(i)(B), Prohibited by Technical Specifications |
| 3412008005R00 - NRC Website | |
text
Joseph H. Plona Site Vice President 6400 N. Dixie Highway, Newport, MI 48166 Tel: 734.586.5910 Fax: 734.586.4172 DTE Energy-10 CFR 50.-73 January 26, 2009 NRC-09-0003 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington D C 20555-0001
Reference:
Fermi 2 NRC Docket No. 50-341 NRC License No. NPF-43
Subject:
Licensee Event Report (LER) No. 2008-005 Pursuant to 10 CFR 50.73(a)(2)(v)(D), Detroit Edison is submitting the enclosed LER No. 2008-005. This LER documents the December 4, 2008 loss of the High Pressure Coolant Injection System safety function due to closure of a steam supply valve in accordance with Technical Specification requirements.
No commitments are being made in this LER.
Should you have any questions or require additional information, please contact Mr. Rodney W. Johnson of my staff at (734) 586-5076.
Sincerely, cc: NRC Project Manager NRC Resident Office Reactor Projects Chief, Branch 4, Region III Regional Administrator, Region III Supervisor, Electric Operators, Michigan Public Service Commission
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: No. 3150-0104 Expires 8/31/2010 (9/2007)
, 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.
- 3. PAGE Fermi 2 05000341 1 OF 3
- 4. TITLE Loss of High Pressure Coolant Injection System Safety Function Due to Closure of Steam Supply Valve
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED MONTH DAY YEAR YEAR I SEQUENTIAL REV MONTH DAY YEAR FACILITY NAME DOCKET NUMBER NUMBER NO 05000 12 04 2008 08 005 00 01 26 2009 FACILITY NAME DOCKET NUMBER 05000
- 9. OPERATING MODE
- 11. THIS REPORT SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR§: (Check all that apply)
EJ 20.2201 (b) 020.2203(a)(3)(i) 050.73(a)(2)(i)(C) 0 50.73(a)(2)(vii) 1l 20.2201 (d) 020.2203(a)(3)(ii) 0 50.73(a)(2)(ii)(A) 0 50.73(a)(2)(viii)(A)
EJ 20.2203(a)(1) 020.2203(a)(4) 50.73(a)(2)(ii)(B) 0 50.73(a)(2)(viii)(B) 20.2203(a)(2)(i) 050.36(c)(1)(i)(A) 050.73(a)(2)(iii) 0 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL 020.2203(a)(2)(ii) 0 50.36(c)(1)(ii)(A) 0 50.73(a)(2)(iv)(A) 0 50.73(a)(2)(x)
EJ 20.2203(a)(2)(iii) 0 50.36(c)(2) 050.73(a)(2)(v)(A) 73.71 (a)(4) 100%
EJ 20.2203(a)(2)(iv) 050.46(a)(3)(ii) 050.73(a)(2)(v)(B) 073.71 (a)(5) 20.2203(a)(2)(v) 0 50.73(a)(2)(i)(A) 050.73(a)(2)(v)(C)
[: OTHER 020.2203(a)(2)(vi) 0 50.73(a)(2)(i)(B) 50.73(a)(2)(v)(D)
Specify in abstract below or in (If more space is required, use additional copies of NRC Form 366A)
Initial Plant Conditions
Mode 1
Reactor Power 100 percent Description of the Event On December 04, 2008 at 17:35 EST during High Pressure Coolant Injection (HPCI) [BJ] system steam line warming performed as part of a planned system pump and valve surveillance test, the HPC1 Steam Supply Outboard Isolation Valve Bypass Valve (E4150F600) position indication lights did not function as expected.
During a partial opening stroke, the valve's Open light did not come on as expected and its Close light flickered while pushing the Open pushbutton for one to two seconds. The Close light continued to flicker for about five seconds after the Open pushbutton was released, at which time all indication was lost and a HPCI Motor Operated Valve Motor Overload/Loss of power alarm was received in the Control Room.
Upon discovery of the problem and performance of minor troubleshooting, Operations determined that the ability of the E4150F600 valve to close on containment isolation signal was unreliable, the valve was declared inoperable. Containment Isolation Limiting Condition for Operation (LCO) 3.6.1.3 Condition A was entered for the E4150F600 valve with the required action to isolate the penetration within four hours.
On December 04, 2008 at 21:18 EST an unplanned HPCI inoperability occurred when the HPCI Steam Supply Inboard Isolation Valve (E4150F002), was closed to isolate the HPCI steam line and satisfy LCO 3.6.1.3 Condition A Required Action to address the E4150F600 inoperability. A 14 day LCO was entered at that time for an inoperable HPCI system per LCO 3.5.1.
System troubleshooting, diagnostic testing, and analysis were performed, and control relay CR1 was replaced.
Post maintenance testing was successfully completed on December 05, 2008 at 23:45 EST. The HPCI Steam Supply Inboard Isolation Valve was opened, the system returned to service, and the LCOs cleared by December 06, 2008 at 00:12 EST.
Significant Safety Consequences and Implications
The HPCI system, a single train safety system, was rendered inoperable when the HPCI steam isolation valve was closed in response to meet the requirements of the containment isolation Technical Specification. The purpose of the HPCI system is to provide emergency core cooling in the event of an accident involving loss of coolant from a small break. Reactor steam is used to drive the HPCI turbine, which in turn drives the main and booster pumps to provide a source of high pressure water to the reactor. The Reactor Core Isolation Cooling [BN] and Standby Feedwater [SJ] systems remained available for high pressure injection in the event of an emergency.
Additionally, the Automatic Depressurization System [JE] was available to reduce reactor pressure to within the capabilities of the low pressure Emergency Core Cooling Systems.
1U.S. NUCLEAR REGULATORY COMMISSION (9-2007)
LICENSEE EVENT REPORT (LER)
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE YEAR SEQUENTIAL REVISION Fermi 2 05000341 2
NUMBER NUMBER 3 OF 3 2008 005 0
This event resulted in approximately 27-hours where HPCI was inoperable. Technical Specification 3.5.1 allows HPCI to be taken out of service for planned outages for up to 14-days. The risk increase associated with HPCI being out of service for approximately 27-hours has been evaluated by the Probabilistic Safety Analysis (PSA)
Model and determined to be very low.
This report is made in accordance with 10 CFR 50.73(a)(2)(v)(D), for any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident. An eight-hour non-emergency notification was made pursuant to 10 CFR 50.72(b)(3)(v)(D) as a condition that at the time of discovery could have prevented the fulfillment of a safety function to mitigate the consequences of an accident based on loss of a single train safety system (EN 44698).
Cause of the Event
The cause for HPCI being inoperable was the isolation of the main steam line penetration to meet Primary Containment Technical Specification requirements. The cause for the main steam line being closed is the loss of indication on the Outboard Isolation Valve Bypass Valve (E4150F600) attributed to a faulty CR1 relay with degraded electrical contacts. This relay connects the control power supply to the MOV operational controls and the valve indications.
Corrective Actions
Valve power fuses, control power fuses, and control power relay CR1 were replaced for the E4150F600 valve, and testing was performed to ensure proper valve operation and indication. This event has been added into the Corrective Action Program and additional actions may be taken as determined by the program.
Additional Information
A.
Failed Components:
Component:
240 VDC relay (CRI Relay)
Function: Connects the control power supply to the MOV operational controls and the valve indications Manufacturer: Siemens 3TH30 Model Number: 3TH3031-OBQ4 Failure Cause: Electrical contact degradation B.
Previous LERs on Similar Problems:
In July 2003 the High Pressure Coolant Injection System was declared inoperable due to closing of the main steam supply valve. The component and cause of the failure were unrelated to the E4150F600 valve control circuitry. Therefore, the corrective actions for that event could not have precluded this event. No other similar problems were noted within the past five years.
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