05000321/LER-2008-004, Power Supply Card Failure Causes Loss of Feedwater Flow Resulting in Manual Reactor Scram
| ML090140150 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 01/13/2009 |
| From: | Madison D Southern Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-09-0028 LER 08-004-00 | |
| Download: ML090140150 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2) 10 CFR 50.73(a)(2)(viii)(B) |
| 3212008004R00 - NRC Website | |
text
D. R......ison (Dennis)
Soudllnt NlClHr Vice President - Hatch Ope_III CoaIpI"', IIIC.
Plant Edwin I. Hatch 11028 Hatch Parkway, North Baxley, Georgia 31513 Tel 912.537.5859 Fax 912.366.2077 SOUTHERN.\\
January 13, 2009 COMPANY Docket No.:
50-321 NL-09-oo28 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Edwin I. Hatch Nuclear Plant - Unit 1 Ucensee Event Report Power Supply Card Failure Causes Loss of Feedwater Flow Resulting in Manual Reactor Scram Ladies and Gentlemen:
In accordance with the requirements of 10 CFR 50.73 (a)(2)(iv)(A), Southern Nuclear Operating Company is submitting the enclosed report for a condition that occurred on November 22, 2008.
This letter contains no NRC commitments. If you have any questions, please advise.
Sincerely,
~~~
D. R. Madison Vice President - Hatch DRMlMJKldaj Enclosure: LER 1-2008-004 cc: Southern Nuclear Operating Company Mr. J. T. Gasser, Executive Vice President Mr. D. H. Jones, Vice President - Engineering RTYPE: CHA02.004 U. S. Nuclear Regulatorv Commission Mr. L. A. Reyes, Regional Administrator Mr. R. E. Martin, NRR Project Manager - Hatch Mr. J. A. Hickey, Senior Resident Inspector - Hatch
"'RC FORM 316 U.S. NUCLEAR REGUlATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 0813112010 9-2(07)
Estimated buldln 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 />. Repotled lessons learned are incorporated into the licensing process and fed back to Industry. send convnents ~ buldln UCENSEE EVENT REPORT (LER) estimate to the Records and FOIAlPr1va~ Service Branch
- - 5 F 2). U.S.
Nuclear ~tory Corrvnlsslon. Wu~n, DC 20555-0001, or::rc Internet e-mail to OColleclsOnlt~, and to the Desk OlIIcer, OIIIee 01 I ormation and Regulatory Affairs. HE 10202, (3150-0104), 0IIlce aI Management and Budget. Washington, DC 20503. If a means used to Impose an Inlormation COIlec:t1on does not display a cunently valid OMB control number, the NRC may not conduc:~~, ancl a person is not required to respond to, the
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- 13. PAGE Edwin 1. Hatch Nuclear Plant Unit 1 05000321 1 OF 4 14.11TLE Power Supply Card Failure Causes Loss of Feedwater Flow Resulting in Manual Reactor Scram
- 5. EVENT DATE
- 5. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILmES INVOLVED SEQUENTIAL REV FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR NUMBER NO.
MONTH DAY YEAR 05000 FACILIlY NAME DOCKET NUMBER 11 22 2008 2008 004 0
01 13 2009 05000 II. OPERAllNG MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFRI: (Check aU that apply) o 20.2201 (b) o 2O.2203(a)(3)(ij o SO.73(a)(2)(1)(C) o 50.73(a)(2)(vll) 1 o 20.2201 (d) o 2O.2203(a)(3)(li) o SO.73(a)(2)(ii)(A) o SO.73(a)(2)(viil)(A) o 2O.2203(a)(1) o 2O.2203(a)(4) o SO.73(a)(2)(ii)(B) o 50.73(a)(2)(viii)(B) o 2O.2203(a)(2)(I) o SO.36(c)(1)(i)(A) o SO.73(a)(2)(III) o SO.73(a)(2)(ix)(A)
- 10. POWER LEVEL o 2O.2203(a)(2)(11) o SO.36(c)(1 )(ii)(A) 181 5O.73(a)(2)(lv)(A) o SO.73(a)(2)(x) o 2O.2203(a)(2)(iii) o 5O.36(c)(2) o 5O.73(a)(2)(v)(A) o 73.71(a)(4) 99.8 o 2O.2203(a)(2)(Iv) o 5O.46(a)(3)(11) o 5O.73(a)(2)(v)(B) o 73.71 (a)(5) o 2O.2203(a)(2)(v) o 5O.73(a)(2)(i)(A) o 5O.73(a)(2) (v)(C) o OTHER o 2O.2203(a)(2)(vl) o 5O.73(a)(2)(i)(B) o SO.73(a)(2)(v)(D)
Specify In Abstlllet below or in NRC Form 366A
- 12. UCENSEE CONTACT FOR THIS LER ACILIlY NAME I~ELEPHONE NUMBER (Indude Area Code)
Edwin 1. Hatch I Kathy Underwood, Performance Improvement Supervisor 912-537-5931
- 13. COMPLETE ONE UNE FOA EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT
CAUSE
SYSTEM COMPONENT MANU*
REPORTABLE
CAuse
SYSTEM COMPONENT MANU*
REPORTABLE FACTURER TO EPIX FACTURER TO EPIX X
SD JX YOO6 Y
- 14. SUPPLEMENTAL REPORT EXPECTED
- 15. EXPECTED SUBMISSION MONTH DAY YEAR o YES (If yes, complete 15. EXPECTED SUBMISSION DATE) 181 NO DATE ABSTRACf (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)
On November 22,2008 at approximately 1019 EST, Unit 1 was in the Run mode at a power level of approximately 2800 CMWf, 99.8 percent rated thermal power. A manual scram was inserted due to Reactor Water Level (RWL) decreasing to 10 inches above instrument zero and continuing to decrease. High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) automatically started on low RWL, Level 2. RWL decreased to approximately negative 68 inches (about 90 inches above the top of active fuel) prior to it being recovered by HPCI and RCIC operation. Due to the RWL reaching the Anticipated Transient Without Scram - Recirculation Pump Trip (ATWS-RPT) low level, the recirculation pumps tripped as designed. As RWL was recovering, HPCI was manually secured and RCIC flow was decreased. The lA Reactor Feed Pump (RFP) was subsequently restarted and RWL control was then transitioned to the lA RFP.
Investigation determined that the direct cause of the event was failure of DC power supply IN21-K088 which provides power to the differential pressure (DP) controller for the (Steam Jet Air Ejector) SJAB Intercondenser Cooling water control valve IN21-F211.
The DC power supply IN21-K088 was replaced and a repetitive task has been created to replace the component at a prescribed interval.
PRINTED ON RECYCLED PAPER NRC FORM 388 (9-2007)
4 NACFOAM38IA LICENSEE EVENT REPORT (LER) u.s. NUCLEAR REGULATORY COMMISSION
(~2007)
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Edwin I. Hatch Nuclear Plant Unit 1 NARRATNE (If more space is required, use additional copies of NRC Form 366A)
PLANT AND SYSTEM IDENTIFICATION
General Electric - Boiling Water Reactor Energy Industry Identification System codes appear in the text as (EllS Code XX).
DESCRIPTION OF EVENT
On November 22,2008 at approximately 1019 EST, Unit 1 was in the Run mode at a power level of approximately 2800 CMWT, 99.8 percent rated thennal power. A manual scram was inserted due to Reactor Water Level (RWL) decreasing to 10 inches above instrument zero and continuing to decrease.
Prior to this, the Condensate Booster Pumps (CBP) (EnS Code SO) and the Reactor Feed Pumps (RFP)
(EllS Code Sl) low suction pressure alarms were received. A Recirculation Pump (EnS Code AD) runback to 61 % speed was initiated by design due to the low suction pressure condition. Reactor operators responded to the transient by manually reducing recirculation flow further. A System Operator was dispatched to the Condensate Demineralizer (EllS Code SD) panel. The System Operator observed demineralizer flows oscillating between 0 and 1200 gpm and the demineralizer system OP at approximately 17 psid.
CBP discharge pressure increased momentarily with the initial reduction in recirculation flow and then began decreasing again. The lA CBP tripped and was followed by the tripping of the IA and IB RFP's. At that time, a manual scram was inserted. Reactor Water Level continued to decrease with High Pressure Coolant Injection (HPCI) (EllS Code BJ) and Reactor Core Isolation Cooling (RCIC) (EllS Code BN) automatically starting on low RWL, Level 2. RWL decreased to approximately negative 68 inches (68 inches below instrument zero or about 90 inches above the top of active fuel) prior to it being recovered by HPCI and RCIC operation. The peak reactor pressme reached was approximately 1053 psig, which is below the setpoint of 1150 psig for the actuation of the Safety Relief Valves (EllS Code SB). Oue to the RWL reaching the Anticipated Transient Without Scram - Recirculation Pump Trip (ATWS-RPT) low level, the recirculation pumps tripped as designed.
As RWL was recovering, HPCI was manually secured and RCIC flow was decreased to 270 gpm. RWL continued to increase and the RWL high level trip, Level 8, was then received due to level swell and RCIC operation. The high level trip resulted in a trip of RCIC. As RWL decreased due to steaming from decay heat, RCIC was manually initiated for RWL control. The lA RFP was subsequently restarted and RWL control was then transitioned to the lA RFP.
CAUSE OF EVENT
Investigations determined that the direct cause of the event was failure of DC power supply IN21-K088.
This power supply provides control power for OP control for the SJAE Intercondenser Cooling water control valve IN21-F211. This valve is on the primary condensate 30-inch line and controls Op by being throttled closed. With the failure of the power supply, the valve failed closed isolating the main condensate 30-inch line to the Condensate Demineralizer, thereby creating a backpressure and forcing cooling water through a 12-inch line to the SJAE. The 12-inch SlAE cooler supply line did not have adequate capacity for the 3-2-2 alignment of the Condensate, CBP. and RFP's. resulting in low suction pressure trips for the lA CBP and the lA and IB RFP's.
PRINTED ON RECYClED PAPER NRC FORM aeeA ("'2007)
NRC FORM 388A LICENSEE EVENT REPORT (LER) u.s. NUCLEAR REGULATORY COMMISSION
(~2007)
COtfrlNUATION SHEET
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05000321 2008 004 0
Edwin 1. Hatch Nuclear Plant Unit 1 REPORTABll..ITY ANALYSIS AND SAFETY ASSESSMENT This report is required by 10 CFR 50.73 (a)(2)(iv)(A), actuation of the Reactor Protection System (RPS)
(EllS Code JC) including: reactor scram or reactor trip. Specifically, the manual insertion of a reactor scram based on RWL decreasing to to inches above instrument zero and continuing to decrease.
Prior to this event, the Condensate Booster Pumps (CBP) and the Reactor Feed Pumps (RFP) low suction pressure alarms were received. The 'A' Recirculation Pump runback to 61% speed was initiated due to the low suction pressure condition. CBP discharge pressure increased momentarily with the reduction in recirculation flow and then began decreasing again. The lA CBP tripped fol.lowed by the IA and IB RFP's tripping. At that time, a manual scram was inserted. Reactor Water Level continued to decrease with High Pressure COOlant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) automatically starting on low RWL, Level 2. RWL decreased to approximately negative 68 inches (68 inches below instrument zero or about 90 inches above the top of active fuel) prior to it being recovered by HPCI and RCIC operation. The RFP's were available immediately following the manual scram to maintain level but were not immediately used. Reactor pressure reached a pressure of approximately 1053 psig which is below the setpoint of 1150 psig for the actuation of the Safety Relief Valves. Due to the RWL reaching the Anticipated Transient Without Scram - Recirculation Pump Trip (ATWS-RPT) low level, the recirculation pumps tripped as designed.
As RWL was recovering, HPCI was manually secured and RCIC flow was decreased to 270 gpm. RWL continued to increase and the RWL high level trip, Level 8, was then received due to level swell and RCIC operation. The high level trip resulted in a trip of RCIC. As RWL decreased due to steaming from decay heat, RCIC was manually initiated for RWL control. The lA RFP was subsequently restarted and RWL control was then transitioned to the lA RFP.
All systems functioned as expected and per their design given the water level transient. Water level was maintained well above the top of the active fuel throughout the transient and was restored to its desired value. Therefore, it is concluded the event had no adverse impact on nuclear safety. This analysis is applicable to all power levels.
CORRECTIVE ACTIONS
DC power supply IN21-K088 was replaced.
A repetitive task for replacement of the IN21-K088 DC power supply card has been created.
Any additional corrective actions that are determined to be appropriate as a result of the cause investigation will be tracked in the plant's corrective action program.
ADDITIONAL INFORMAnON Other Systems Affected: No systems other than those already mentioned in this report were affected by this event.
PRINTED ON RECYCLED P.....ER NRC FOAM 3lIllA. (9-2007)
4 NRC FOAM 381A LICENSEE EVENT REPORT (LER) u.s. NUCLEAR REGULATORY COMMISSION
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Edwin I. Hatch Nuclear Plant Unit 1 05000321
Failed Components Information
Master Parts List Number: IN21-K088 EllS System Code: SD Manufacturer: Yokogawa (YOO6)
Reportable to EPIX: Yes Model Number: WAIV Root Cause Code: X Type: Circuit Board EllS Component Code: JX Commitment Information: This report does not create any permanent licensing commitments.
A previous similar event in the last two years in which the reactor scrammed due to low feedwater flow due to an equipment failure was reported in the following Licensee Event Report:
LER 2-2007-008 identified an instance where a panialloss of the Condensate System caused low feedwater flow resulting in a Reactor Protection System (RPS) actuation on Low Reactor Water Level. The root cause of that event was determined to be ineffective execution of a screening procedure written to determine scram/transient potential of I&C activities. The procedures revised to correct this event were related to I&C activities and were not required to be used during this event Therefore the corrective actions taken for that event would not prevent the occurrence of this event PRINTED ON RECYCLED PAPER NRC FORM 3e8A (.2007)