05000461/LER-2004-004

LER-2004-004, An Delon Company
Clinton Power Station
R. R. 3, Box 228
Clinton, IL 61727
10 CFR 50.73
U-603690
September 10, 2004
U. S. Nuclear Regulatory Commission
ATTN: Document Control Desk
Washington, D. C. 20555-0001
Clinton Power Station, Unit 1
Facility Operating License No. NPF-62
NRC Docket No. 50-461
Subject:LLicensee Event Report 2004-004-00
Enclosed is Licensee Event Report (LER) No. 2004-004-00: Reactor Scram While Placing
Residual Heat Removal B into Shutdown Cooling. This report is being submitted in
accordance with the requirements of 10CFR50.73.
Should you have any questions concerning this report, please contact Mr. William Iliff,
Regulatory Assurance Manager, at (217)-937-2800.
Respectfully,
/401e
R. . Bement
Si e Vice President
Clinton Power Station
JLP/blf
Enclosure:LLicensee Event Report 2004-004-00
cc:LRegional Administrator — NRC Region III
NRC Senior Resident Inspector — Clinton Power Station
Office of Nucleai- Facility Safety — IEMA Division of.Nuclear Safety

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, 1.FACILITY NAME I 2. DOCKET NUMBER I 3. PAGE
1 Clinton Power Station I 05000461 I 1 OF 3 •
IT I4. T LE
Reactor Scram While Placing Residual Heat Removal S' Into Shutdown Cooling

Docket Number
Event date:
Report date:
4612004004R00 - NRC Website

A. PLANT OPERATING CONDITIONS PRIOR TO THE EVENT:

Unit: 1 � Event Date: 7/14/2004 Mode: 3 (HOT SHUTDOWN) Event Time: 0044 Central Daylight Time Reactor Power: 0 percent

B. DESCRIPTION OF THE EVENT:

On July 14, 2004, operators were warming the Residual Heat Removal (RHR) [BO] B loop while making preparations for placing the Shutdown Cooling (SDC) mode of RHR into service. At 0044 hours5.092593e-4 days <br />0.0122 hours <br />7.275132e-5 weeks <br />1.6742e-5 months <br /> (Central Daylight Time), reactor pressure vessel (RPV) water level decreased from +32 inches to +9 inches, resulting in an automatic RPV low water level (Level 3) scram actuation and a containment isolation (Groups 2, 3, and 20). At the time of the event, the plant was in Mode 3 (Hot Shutdown) with all control rods fully inserted from the reactor scram that occurred on July 13, 2004 at 1608 hours0.0186 days <br />0.447 hours <br />0.00266 weeks <br />6.11844e-4 months <br /> (Reference LER 2004-003), and operators were preparing to proceed to Mode 4 (Cold Shutdown).

After completing the flush of the RHR B system on July 13, 2004 at 2128 hours0.0246 days <br />0.591 hours <br />0.00352 weeks <br />8.09704e-4 months <br />, RHR B system warmup commenced at 2255 hours0.0261 days <br />0.626 hours <br />0.00373 weeks <br />8.580275e-4 months <br />. At 0017 hours1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br /> on July 14, 2004, warming was stopped and a fill and vent of the RHR B heat exchanger was started. A sharp drop in temperature and pressure was observed and the fill and vent evolution was terminated. At 0030, operators commenced re- warming of the RHR B loop and opened valve 1E12F040 (RHR B flush valve to radwaste) for about 8 to 9 seconds, and the 1E1 2F003B (RHR B heat exchanger outlet valve) was throttled open for about 1 to 2 seconds. This allowed water from the RHR B heat exchanger and the RHR B discharge header to be discharged to radwaste. At 0043 hours4.976852e-4 days <br />0.0119 hours <br />7.109788e-5 weeks <br />1.63615e-5 months <br />, operators opened the 1E1 2F040 valve for an additional 3 to 4 seconds to continue the warming process. Approximately one minute after throttling the valve open, RPV water level dropped approximately 24 inches in about 20 seconds. At 0044 hours5.092593e-4 days <br />0.0122 hours <br />7.275132e-5 weeks <br />1.6742e-5 months <br />, a Reactor Protection System (RPS) actuation (i.e., scram) due to RPV water level reaching Level 3 occurred. The operators entered procedure CPS No. 4401.01, "Reactor Scram," and Emergency Operating Procedure (EOP) EOP-1, "RPV Level Control." At 0047 hours5.439815e-4 days <br />0.0131 hours <br />7.771164e-5 weeks <br />1.78835e-5 months <br />, all isolations occurred as expected for reaching RPV Level 3. At 0051 hours5.902778e-4 days <br />0.0142 hours <br />8.43254e-5 weeks <br />1.94055e-5 months <br />, the scram was reset, and at 0111 hours0.00128 days <br />0.0308 hours <br />1.835317e-4 weeks <br />4.22355e-5 months <br />, EOP-1 was exited. At 0129 hours0.00149 days <br />0.0358 hours <br />2.132936e-4 weeks <br />4.90845e-5 months <br />, CPS 4401.01 was exited.

The opening of 1E1 2F040, while using valve 1E1 2F003B to throttle and control the flow of reactor coolant to warm the system, allowed the phenomenon of "column separation" to occur in the system.

This phenomenon occurs when a line is drained without a vent path, creating a vacuum in the pipe.

Subsequent analysis of the event showed that due to the low differential pressure between the RPV, which was at 18 psig, and the RHR B heat exchanger, which was at 10 psig, there was not enough differential pressure between the RPV and the RHR B system to develop flow through the RHR B pump discharge check valve (1E1 2F031B) to make up for the flow to radwaste. This allowed additional column separation to occur in the discharge piping. When the 1E12F040 valve was throttled an additional 3 to 4 seconds open for additional warming, the RHR B heat exchanger pressure decreased. At that point, the differential pressure was sufficient to open the 1E1 2F031B check valve and the RHR B system was refilled with water from the RPV, causing the RPV level to drop, equalizing level between the RPV and RHR B heat exchanger.

Condition Report 235832 was initiated to investigate this event.

C. CAUSE OF THE EVENT:

A root cause of this event was that the RHR SDC procedure did not adequately prohibit or provide adequate guidance for re-warming of the system. The procedure introduced a latent error through valve sequencing that allowed the column separation effect to occur while warming the system.

D. SAFETY ANALYSIS:

This event did not constitute a risk to the public, the plant, or station personnel. Sufficient engineered safety features existed to prevent more serious consequences. There were no actual safety consequences associated with his event. The event was reviewed for analyzed transients discussed in Chapter 15 of the Clinton Power Station Updated Safety Analysis Report. The analysis determined that this event was within the design basis of the plant.

No safety system functional failures occurred during this event.

E. CORRECTIVE ACTIONS:

The warming sequence of the RHR SDC procedure will be revised to preclude column separation.

F.�PREVIOUS OCCURRENCES:

There have been no previous similar events at Clinton Power Station or in the industry due to the phenomenon of column separation.

G. COMPONENT FAILURE DATA:

None.