05000266/LER-2010-003

LER-2010-003, POINT BEACH
November 11, 2010 NRC 2010-0162
10 CFR 50.73
U. S. Nuclear Regulatory Commission
ATTN: Document Control Desk
Washington, DC 20555
Point Beach Nuclear Plant, Units 1 and 2
Dockets 50-266 and 50-301
Renewed License Nos. DPR-24 and DPR-27
Licensee Event Report 266/2010-003-00
Potential for Residual Heat Removal Trains to be Inoperable During Mode Change
Enclosed is Licensee Event Report (LER) 266/2010-003-00 for Point Beach Nuclear Plant (PBNP),
Units 1 and 2. This LER documents the condition in which both units, when in MODE 4, have
exceeded the fluid temperature in the residual heat removal (RHR) pump suction header which,
under loss of coolant accident conditions, could have resulted in steam voiding and potentially
damaging condensation-induced water hammer. The enclosed LER is submitted in accordance with
10 CFR 50.73(a)(2)(ii)(B) for being in an unanalyzed condition, 50.73(a)(2)(v)(B) as a condition that
could potentially have prevented the fulfillment of a safety function, and 50.73(a)(2)(vii) where a
single cause or condition caused two independent RHR trains to become inoperable. In accordance
with 10 CFR 50.73(a) requirements, this LER is submitted within 60 days following discovery.
This submittal contains no new or revised regulatory commitments.
If you have questions or require additional information, please contact Mr. James Costedio at
920/755-7427.
Very truly yours,
NextEr Energy Point Beach, LLC
arry Meyer
Site Vice President
Enclosure
cc:AAdministrator, Region III, USNRC
Project Manager, Point Beach Nuclear Plant, USNRC
Resident Inspector, Point Beach Nuclear Plant, USNRC
PSCW
NextEra Energy Point Beach, LLC, 6610 Nuclear Road, Two Rivers, WI 54241
NRC FORM 3666 U.S. NUCLEAR REGULATORY COMMISSION
(10-2010)
LICENSEE EVENT REPORT (LER)
(See reverse for required number of
digits/characters for each block)
1. FACILITY NAME
Point Beach Nuclear Plant (PBNP) - Unit 1
4. TITLE
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2. DOCKET NUMBER 3. PAGE
05000266 1 of 4
Potential for Residual Heat Removal Trains to be Inoperable During Mode Change

Point Beach Nuclear Plant (Pbnp) - Unit 1
Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition
2662010003R00 - NRC Website

Event Description:

During a review of NRC Information Notice 2010-0011, it was determined that, while in MODE 4, the temperature of the fluid in the Residual Heat Removal (RHR) pump suction header exceeded the temperature which, under loss of coolant accident (LOCA) conditions, could have resulted in steam voiding and potentially damaging condensation induced water hammer.

Previous analyses of this issue had not considered the potential for failure of RHR pumps in operation at the time of the postulated LOCA, nor had the water hammer effects of re-fill of voided piping, as directed by the station shutdown LOCA procedure, been considered.

A past operability evaluation was performed. The evaluation concluded that there were several periods within the past three years during which operability of the low head safety injection (LHSI) subsystem could not be assured in the event of a shutdown LOCA. The results of the past operability evaluation are discussed below.

This 60-day licensee event report is being submitted in accordance with the requirements of 10 CFR 50.73(a)(2)(ii)(B), 50.73(a)(2)(v)(B), and 50.73(a)(2)(vii).

Event Analysis:

The past operability evaluation considered the RHR/LHSI system to be susceptible for temperatures above 200 degrees F. A review of plant procedures confirmed that existing RHR system cooldown guidance ensures that MODE 3 will not be entered with excessive low head safety injection (LHSI) temperatures. While in MODE 4, Unit 1 was susceptible to steam voiding for a total of 32 hrs 20 minutes over the last three years. The longest single period of susceptibility was 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> 27 minutes. Unit 2 was likewise susceptible for a total of 38 hrs 10 minutes, with the longest single period being 12 hrs 35 minutes. A breakdown of these periods is provided below.

1R32 (Spring 2010) MODE 3 MODE 5 2034 on 3/1/10 to 0450 on 3/2/10 Duration: 8 hrs 16 min MODE 5 MODE 3 1547 on 3/31/10 to 0314 on 4/1/10 Duration: 11 hrs 27 min 2R30 (Fall 2009) MODE 3 MODE 5 1930 on 10/15/09 to 0543 on 10/16/09 Duration: 10 hrs 15 min MODE 5 MODE 3 2355 on 12/2/09 to 1223 on 12/3/09 Duration: 12 hrs 35 min 1R31 (Fall 2008) MODE 3 MODE 5 0154 on 10/6/08 to 0409 on 10/6/08 Duration: 2 hrs 15 min MODE 5 MODE 3 2103 on 11/9/08 to 0725 on 11/10/08 Duration: 10 hrs 22 min 2R29 (Spring 2008) MODE 3 MODE 5 0657 on 4/6/08 to 1006 on 4/6/08 Duration: 3 hrs 9 min MODE 5 MODE 3 1930 on 5/5/08 to 0741 on 5/6/08 Duration: 12 hrs 11 min The site review of previously provided vendor information had not considered the potential for failure of RHR pumps in operation at the time of a postulated LOCA. Station procedures had permitted operation of both RHR pumps with both trains cross-connected during MODE 4. Previous analyses had also not considered the water hammer effects from the refill of voided piping. Operating procedures directed refill by isolating the RHR / LHSI subsystem from the voided Reactor Coolant System (RCS), and then remotely opening the suction from the refueling water storage tank (RWST) motor-operated valve.

Safety Significance:

The potential existed for the ECCS system to not have been available for LHSI and containment sump recirculation had a LOCA occurred during MODE 4. At the reduced temperatures and pressures in MODE 4, the likelihood of occurrence for a design basis accident is lower than at normal operating temperature and pressure.

If the postulated event occurred, emergency procedures direct the operator to maintain injection using charging pumps and/or high head safety injection (HHSI), and to then realign RHR and restore LHSI from the RWST.

Due to the low temperatures, low decay heat load, and low RCS pressure, a postulated LOCA occurring in MODE 4 would not be expected to cause fuel failure. The resultant low radiation and contamination levels would permit continued access to components damaged by the postulated water hammer event, and to facilitate necessary repairs to restore LHSI / sump recirculation functionality.

During the repair / restoration period, multiple sources of continuing makeup to the RWST would support continued injection.

No actual event occurred during the brief period of time when the RHR subsystem was inoperable during MODE 4. Accordingly, the safety significance of the identified condition is low and there were no ramifications that could impact the health and safety of the public.

Cause:

Site procedures did not adequately support all functions of ECCS operation in MODE 4 because technical reviews of previously received operating experience information had been less than adequate.

Corrective Action:

Operations procedure changes are in progress and are being tracked in the site's corrective action program to ensure that administrative controls are in place to maintain one train of RHR isolated and cool while in MODE 4 so that it can be used for LOCA mitigation.

Previous Occurrences: There have been no similar LERs submitted within the last three (3) years.

Failed Components Identified: None