05000321/LER-2004-001, Water Intrusion Into Relay Panel 2H21-P232 Results in the Start of the 2C Emergency Diesel Generator
| ML041070241 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 04/13/2004 |
| From: | Sumner H Southern Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-04-0615 LER 04-001-00 | |
| Download: ML041070241 (6) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(iv), System Actuation |
| 3212004001R00 - NRC Website | |
text
H. L Sumner, Jr.
Southern Nuclear Vice President Operating Company, Inc.
Hatch Project Post Office Box 1295 Birmingham, Alabama 35201 Tel 205.992.7279 SOUTHERNta April 13, 2004 COMPANY Energy to Serve Your World Docket Nos.:
50-321 NL-04-0615 50-366 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Edwin 1. Hatch Nuclear Plant Licensee Event Report Water Intrusion into Relay Panel 2H21-P232 Results in the Start of "2C" Emergency Diesel Generator 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 the water intrusion into relay panel 2H21-P232 resulting in the start of the "2C" emergency diesel generator.
This letter contains no NRC commitments. If you have any questions, please advise.
Sincerely, H. L. Sumner, Jr.
HLSIIL/daj Enclosures: LER 50-366/2004-001 cc:
Southern Nuclear Operating Company Mr. J. B. Beasley, Jr., Executive Vice President Mr. G. R. Frederick, General Manager - Plant Hatch RTYPE: CHAO2.004 U. S. Nuclear Regulatory Commission Mr. L. A. Reyes, Regional Administrator Mr. C. Gratton, NRR Project Manager - Hatch Mr. D. S. Simpkins, Senior Resident Inspector - Hatch
Abstract
On 2/18/2004 at 1547 ET, Unit 2 was in the Run mode at a power level of approximately 2772 CMWT (98.8% rated thermal power). At that time, an 8 inch Plant Service Water (PSW) pipe was being cut in support of implementation of a planned design change. Despite precautions taken some water drained onto the Unit 2 '2C' Emergency Diesel Generator (DG) relay panel 2H21-P232 which caused the '2C' DG to start from an apparent invalid signal. The DG did not tie to the bus, and the bus remained energized throughout the event. Licensed Operations personnel shut down the DG by 1604 ET on 2/18/2004, and the DG's mode switch was left in the test position until the completion of trouble shooting activities.
This event resulted from personnel failing to anticipate the volume of water that could be released from the pipe cut due to the combination of the pipe configuration and sediment build-up in the pipe. A contributing cause was failure to protect relay panel 2H21-P232 from any potential water damage.
Troubleshooting and repair activities were completed; and the Operability Surveillance (34SV-R43-006-2S, DIESEL GENERATOR 2C SEMI-ANNUAL TEST) was completed on 2/20/2004 at 0505 ET.
Procedural guidance is being incorporated into the appropriate plant procedures to ensure that electrical equipment is protected when breaching nearby systems.
'NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (1-2001)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
DOCKET LER NUMBER (6)
PAGE (3)
YEAR I SEQUENTLAL l REVISION YEAR NUMBER Edwin I. Hatch Nuclear Plant - Unit 2 05000-366 2004 001
-* 0 2 OF 5 TEXT [if more space Is required, use additional copies of NRC Form 366A) (17)
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
PLANT AND SYSTEM IDENTIFICATION General Electric - Boiling Water Reactor Energy Industry Identification System codes are identified in the text as (EIIS Code XX).
DESCRIPTION OF EVENT
On 2/18/2004 at 1547 ET, Unit 2 was in the Run mode at a power level of approximately 2772 CMWT (98.8% rated thermal power). An 8-inch pipe cut on the Plant Service Water (PSW, EIIS Code BI) system was made during the implementation of Design Change Request (DCR) 03-026T. Piping downstream of the cut area was degraded. The DCR required removal of a section of pipe, capping the upstream side, and abandoning the degraded downstream piping. The cut area was located approximately 12 ft. above the floor and was directly over a cable tray in the 2G switchgear room.
In preparation for the pipe cut, a clearance boundary was established and protective measures were taken to prevent water damage to surrounding equipment. The cut was made by an air-driven cutting tool. As the cut was nearly completed, the pipe shifted. The volume of water release was more than anticipated and traveled beyond the area prepared to handle water.
Investigation revealed that the trapped water was not fully drained from the piping. Personnel failed to realize that a combination of sediment build-up and a piping section without a drain could trap a considerable volume of water in the pipe. As a precaution, pipefitters had drilled a hole in the top of the 8-inch pipe close to the cut location; however, because the hole was located in the portion of the pipe being discarded, the area available for inspection was downstream of the cut. Upstream of the inspection hole there was a sediment build-up resulting in trapped water. The pipe appeared dry when it was inspected through the drilled hole. When the cut was made, the pipe moved and the trapped water was released. Although the electrical equipment directly beneath the cut was covered and adequately protected from water, the adjacent electrical equipment was not protected because the amount of water anticipated was underestimated.
Water from the cut traveled along cable trays and scaffold boards and drained onto surrounding panels and junction boxes. Panel 2H21-P232 (the '2C' Emergency Diesel Generator Relay Panel), received enough water to cause an auto-start of the '2C' DG (EDG, EIIS Code EK). Additionally, a Generator NRC FonT S64A (1.2001)
- - NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (1-2001)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
DOCKET LER NUMBER (6)
PAGE (3) r YEAR I SEQUENTML REVSION YEAR NUMBER Edwin I. Hatch Nuclear Plant - Unit 2 05000-366 2004 001 0
3 OF 5 TEXT (If more space Is required, use additional copies of NRC Form 366A) (17)
Field Ground and a Battery Malfunction alarm (caused by a DC distribution ground) were both received. The water flow stopped after an estimated 30 to 50 gallons had drained from the pipe cut.
The 2H21-P232 panel contains '2C' DG control logic. There are six contacts located in this panel that if shorted by water would start the DG. During this event the normal supply breaker to the bus did not trip and the bus remained energized. The '2C' DG did not tie to the bus. Licensed Operations personnel shut down the DG by 1604 ET, and the DG's mode switch was left in the test position.
Troubleshooting and repair activities were completed; and the Operability Surveillance (34SV-R43-006-2S, DIESEL GENERATOR 2C SEMI-ANNUAL TEST) was completed on 2/20/2004 at 0505 ET.
It was concluded from the event that at least one of the start contacts for '2C' DG were shorted by the water intrusion causing an invalid start signal to be generated resulting in the '2C' DG starting.
Additionally, since the DG's control switch could be taken to and maintained in the test position, the LOCA relays were not initiating a start signal to the DG and no LOCA signal was received.
Recovery from the PSW intrusion into the '2C' DG control panel (2H21-P232) involved drying out the panel, inspecting the HGA and HFA relays in the panel associated with the DG start logic and load shed logic. Relays were dried, calibrated, repaired, or replaced as required. No water was found during the inspection of the 6 HGA and 14 HFA relays. When relay 2R43-K770 was removed from its housing and the housing dried out the DC distribution ground and Field ground cleared.
After all recovery actions were completed, the '2C' DG was functionally tested by performing the DG quick start test. This test most closely approximates an automatic start of the DG.
CAUSE OF EVENT
Personnel failed to anticipate the water that could be released from the pipe cut because the combination of pipe configuration and sediment build-up was not considered. A contributing cause of the event was the failure to protect all electrical equipment from water damage. At the time the pipe was cut, water drained onto 2H21-P232 ('2C' Relay Panel) from the 8-inch PSW pipe being cut shorting an undervoltage relay causing an invalid signal to start the '2C' DG.
NRC FomI 356A (I*Zgg01
lNRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (1-2001)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
DOCKET LER NUMBER (6)
PAGE (3)
YEAR SEQUENTIAL I REVISION I
YEAR NUMBER Edwin I. Hatch Nuclear Plant - Unit 2 05000-366 2004 001 0
4 OF 5 TEXT (If more space Is required, use additional copies of NRC Form 366A (17)
REPORTABILITY ANALYSIS AND SAFETY ASSESSMENT This event is reportable per 10 CFR 50.73 (a)(2)(iv) because an automatic actuation of a system listed in paragraph (a)(2)(iv)(B)(8) occurred. Specifically, the '2C' DG automatically started from an invalid signal when water drained onto the 2H21-P232 relay panel.
The Unit 2 Class IE AC Electrical Power Distribution System AC sources consist of the offsite power sources (preferred power sources, normal and alternate), and the onsite standby power sources (DGs
'2A', '2C', and 'I B'). As required by 10 CFR 50, Appendix A, GDC 17, the design of the AC electrical power system provides independence and redundancy to ensure an available source of power to the Engineered Safety Feature (ESF) systems. The Class IE AC distribution system is divided into redundant load groups, so loss of any one group does not prevent the minimum safety functions from being performed. Each load group has connections to two preferred offsite power supplies and a single DG. Offsite power is supplied to the 230 kV and 500 kV switchyards from the transmission network by eight transmission lines. From the 230 kV switchyards, two electrically and physically separated circuits provide AC power, through startup auxiliary transformers 2C and 2D, to 4.16 kV ESF buses 2E, 2F, and 2G.
Startup auxiliary transformer (SAT) 2D provides the normal source of power to the ESF buses 2E, 2F, and 2G. If any 4.16 kV ESF bus loses power, an automatic transfer from SAT 2D to SAT 2C occurs.
The onsite standby power source for 4.16 kV ESF buses 2E, 2F, and 2G consists of three DGs. DGs
'2A' and '2C' are dedicated to ESF buses 2E and 2G, respectively. DG 'IB' (the swing DG) is a shared power source and can supply either Unit 1 ESF bus IF or Unit 2 ESF bus 2F. A DG starts automatically on a loss of coolant accident (LOCA) signal (i.e., low reactor water level signal or high drywell pressure signal) or on an ESF bus degraded voltage or undervoltage signal. After the DG has started, it automatically ties to its respective bus after offsite power is tripped as a consequence of ESF bus undervoltage or degraded voltage, independent of or coincident with a LOCA signal.
In this event, it was concluded that at least one of the start contacts for '2C' DG was shorted by the water intrusion causing an invalid start signal to be generated resulting in the '2C' DG starting. The DG did not tie to the bus and was not required to. The bus remained energized throughout the event.
There were no other equipment actuations associated with this event. The DG performed as designed given the invalid start signal introduced by the water intrusion into the relay panel.
During this event the '2A' and 'IB' DGs were available for Unit 2. With the '2C' DG inoperable, the plant entered Technical Specifications LCO 3.8.1, Condition B. Required Action B.4 requires, in part, the DG be restored to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The TS Bases for this Required Action states that Regulatory Guide 1.93 provides guidance that operation in Condition B. may continue for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />..
Based on this analysis, it is concluded that this event did not adversely affect nuclear safety.
NRC Fonn 366A (1.2001)
II
- - NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (1-2001)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
DOCKET LER NUMBER (6)
PAGE (3)
YEAR SEQUENTIAL REVISION I
YEAR NUMBER Edwin I. Hatch Nuclear Plant - Unit 2 05000-366 2004 001 0
5 OF 5 TEXT (Il more space Is required, use additional copis of NRC Form 366AI (17)
CORRECTIVE ACTIONS
Troubleshooting and repair activities were completed; and the Operability Surveillance (34SV-R43-006-2S, DIESEL GENERATOR 2C SEMI-ANNUAL TEST) was completed on 2/20/2004 at 0505 ET.
The personnel involved in determining that the PSW line was isolated and free from water were made aware of their error and the consequences of it.
Additional procedural guidance is being incorporated into the appropriate plant procedures to ensure that electrical equipment is protected when breaching nearby systems. These procedure revisions will be completed by June 2004.
ADDITIONAL INFORMATION
No systems other than those previously described in this report were affected by this event.
This LER does not contain any permanent licensing commitments.
There were no previous similar events reported in the past two years in which an automatic actuation of a system listed in paragraph (a)(2)(iv)(B)(8) occurred because of inadequate equipment protection when breaching a system.
Cause Code: A