05000366/LER-2007-008, Regarding Reactor Scram on Low Reactor Water Level Due to Partial Loss of Condensate System
| ML072780182 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 10/05/2007 |
| From: | Madison D Southern Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-07-1820 LER 07-008-00 | |
| Download: ML072780182 (4) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| 3662007008R00 - NRC Website | |
text
D. R. Madison (Dennis)
Southern Nuclear Vice President - Hatch Operating Company, Inc.
Plant Edwin I. Hatch 11028 Hatch Parkway, North Baxley, Georgia 3151 3 Tel 912.537.5859 Fax 91 2.366.2077 Energy to Serve Your World" October 5, 2007 Docket No.:
50-366 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Edwin I. Hatch Nuclear Plant - Unit 2 Licensee Event Report Reactor Scram on Low Water Level due to Partial Loss of Condensate Svstem 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 Licensee Event Report (LER) concerning a partial loss of the condensate system which resulted in an automatic reactor scram and primary containment isolation actuation.
This letter contains no NRC commitments. If you have any questions, please advise.
Sincerely, D. R. Madison Vice President - Hatch Enclosure: LER 2-2007-008 cc:
Southern Nuclear Operatina Companv Mr. J. T. Gasser, Executive Vice President Mr. D. R. Madison., Vice President - Hatch Mr. D. H. Jones, Vice President - Engineering RTYPE: CHA02.004 U. S. Nuclear Reaulatorv Commission Dr. W. D. Travers, Regional Administrator Mr. R. E. Martin, NRR Project Manager - Hatch Mr. J. A. Hickey, Senior Resident Inspector - Hatch
ponse to comply with this mandatory collection request:
ns learned are incorporated into the licensing process Send comments regarding burden estimate to the LICENSEE EVENT REPORT (LER) cy Service Branch (T-5 F52), U.S. Nuclear Regulatory n, DC 20555-0001, or by internet e-mail to (RPS) actuation on Low Reactor Water Level and a Group 2 Primary Containment Isolation System (PCIS) isolation. The investigation determined that the loss of condensate system was caused by the loss of the 2D 4160V switchgear. The 2D bus provides power to the 2A and 2B Condensate and Condensate Booster Pumps. The loss of the 2D 4160V bus was the result of the inadvertent actuation of the over-current relay protecting one of the three phases of the normal supply breaker. This relay had been removed from service as part of a routine I&C periodic relay calibration. During reinstallation of the relay cover following completion of the calibration, the relay actuated tripping the normal supply breaker for the 2D 4160V switchgear and de-energizing the bus providing power to the pumps. All systems functioned per their design given the water level transient.
The root cause of this event was determined to be ineffective execution of a screening procedure written to determine scramltransient potential of I&C activities. The screening procedure was executed for the calibration of the overcurrent relay and errantly determined that there was no reactor trip potential when NRC FORM 366 (9-2007)
U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)
Edwin I. Hatch Nuclear Plant - Unit 2
NARRATIVE
PLANT AND SYSTEM IDENTIFICATION
General Electric - Boiling Water Reactor Energy Industry Identification System codes appear in the text as (EIIS Code XX).
DESCRIPTION OF EVENT
This event is reportable, per 10 CFR 50.73 (a)(2)(iv)(A), because an event occurred which resulted in an automatic scram and automatic closure of primary containment isolation system valves.
On August 7,2007 at 1506 EDT, Unit 2 was in Mode 1 at 2804 CMWT, 100 percent power. An inadvertent actuation of contacts on an over-current relay protecting one of the three phases for the normal supply breaker of the 2D 4160V switchgear occurred. This resulted in a partial loss of the Condensate System (EIIS Code SD) causing a reduction in Feedwater (EIIS Code SJ) flow. The reduction in Feedwater flow caused a decrease in reactor water level. The Recirculation System (PCIS, EIIS Code AD) runback setpoint was reached and the runback initiated. With the loss of two of the three condensate pumps the Recirculation System runback was not able to prevent the continued decrease in reactor water level. As water level continued to decrease, the Reactor Protection System (RPS, EIIS Code JC) reached the reactor low water setpoint and initiated an automatic scram signal. The Group 2 Primary Containment Isolation System (PCIS, EIIS Code JM) actuation setpoint was reached and the isolation signal initiated. Following the reactor scram, water level was recovered automatically with the normal condensate and feedwater system. No automatic initiation setpoints were reached for the Emergency Core Cooling Systems and the operators had no need to manually actuate those systems.
CAUSE OF EVENT
The root cause of this event was determined to be ineffective execution of a screening procedure written to determine scramltransient potential of I&C activities. The screening procedure was executed for the calibration of the overcurrent relay and errantly determined that there was no reactor trip potential when performing the procedure on-line and did not include a precaution for installation of the relay cover.
11
SAFETY ASSESSMENT
Following the automatic scram on low reactor water level, reactor vessel water level continued to decrease due to void collapse. Level reached a minimum of about thirty two inches below instrument zero (about 126 inches above the top of the active fuel). The decrease in water level resulted in a Group 2 PCIS isolation on low water level and thus closure of the Group 2 Primary Containment Isolation Valves per design. The RPS and PCIS are Engineered Safety Feature systems.
The operating Reactor Feedwater Pumps automatically restored water to its designed setpoint. Operations personnel verified correct system response and restored the isolation valves and the RPS to their normal configuration.
1 1
- 2. DOCKET 05000366 All I&C procedures previously screened will be re-evaluated per the revised guidance. Implementation of this corrective action will be tracked under the corrective action program.
ADDITIONAL INFORMATION
- 6. LER NUMBER YEAR SEQUENTIAL REVISION NUMBER NUMBER 2007 1-008 1- 00 Other Systems Affected: None
- 3. PAGE OF
Failed Components Information
None Commitment Information: This report does not create any permanent licensing commitments.
Previous Similar Events
LER 2-2006-002 identified an instance where performance of a calibration procedure resulted in an automatic reactor scram. The root cause of that event identified an error in the calibration procedure which allowed work to be performed on-line that would directly cause an automatic scram if performed on-line.
Additional corrective actions for that event required a review of calibration procedures to determine if an automatic scram would be a potential effect of performing the procedure on-line. This review failed to achieve the desired result of identifying procedures which could result in a unit scram if performed on-line.
Therefore the corrective action for the event reported in LER 2-2006-002 was not effective in preventing the current event.