05000315/LER-2008-004

LER-2008-004, Power Company
Nuclear Generation GroupINDIANA One Cook PlaceMICIGAN Bridgman, MI 49106
POWER aep.com
June 23, 2008 AEP-NRC-2008-2
10 CFR 50.73
Docket No. 50-315
U. S. Nuclear Regulatory Commission
Attn: Document Control Desk
Mail Stop O-P1-17
Washington, DC 20555-0001
Donald C. Cook Nuclear Plant Unit 1
LICENSEE EVENT REPORT 315/2008-004-00
NON-ISOLABLE REACTOR COOLANT SYSTEM
PRESSURE BOUNDARY LEAK
In accordance with the criteria established by 10 CFR 50.73, Licensee Event Report System, the
following report is being submitted:
LER 315/2008-004-00: "Non-Isolable Reactor Coolant System Pressure Boundary Leak."
There are no commitments contained in this submittal.
Should you have any questions, please contact Mr. John A. Zwolinski, Regulatory Affairs Manager, at
(269) 466-2428.
Sincerely,
Lawrence J. Weber
Site Vice President
JEN/rdw
Attachment
U. S. Nuclear Regulatory CommissionJ AEP-NRC-2008-2
Page 2
c:JJ. L. Caldwell — NRC Region III
K. D. Curry — AEP Ft. Wayne, w/o attachment
INPO Records Center
J. T. King — MPSC, w/o attachment
MDEQ — WHMD/RPS, w/o attachment
NRC Resident Inspector
P. S. Tam — NRC Washington DC
.04
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1. FACILITY NAME 2. DOCKET NUMBER 3. PAGE
Donald C. Cook Nuclear Plant, Unit 1 05000315 1 of 3
4. TITLE
Non-Isolable Reactor Coolant System Pressure Boundary Leak
11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)

Donald C. Cook Nuclear Plant
Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded
Initial Reporting
ENS 44171	10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded
3152008004R00 - NRC Website

Conditions Prior to Event Unit 1 was in Mode 4 with Reactor Coolant System (RCS) [AB] pressure near 1000 psig.

Description of Event

identified a non-isolable RCS pressure boundary leak on a 3/4-inch instrument line connected to the RCS. Technical Specification Limiting Condition for Operation '(LCO) 3.4.13 limits the RCS to no pressure boundary leakage and is applicable in Modes 1, 2, 3, and 4.

Unit 1 RCS temperature and pressure was being increased during a refueling outage.

During the 1000 psig leak inspection walkdown of the RCS pressure boundary, a through-wall leak was identified on a 3/4-inch instrument line upstream of 1-NFP-222-V2, Reactor Coolant Loop 2 Channel 3 Differential Flow Instrument 1-NFP­ 222 Low Pressure Side Root Valve. Steam was noted coming from an instrument line weld between the RCS Loop 2 piping and an instrument isolation valve. In that location the leak was non-isolable.

At 2210 on April 25, 2008, CNP Unit 1 entered LCO 3.4.13 Condition D which, with the observed Pressure Boundary Leakage, requires the plant to be placed into Mode 3 in six hours and Mode 5 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Action was initiated to cool down and approximately seven hours later, CNP Unit 1 entered Mode 5. Upon reaching Mode 5, Unit 1 was no longer in the Mode of Applicability, and LCO 3.4.13 Condition D for Pressure Boundary Leakage was exited.

This event is reportable in accordance with 10 CFR 50.73(a)(2)(ii)(A), Any event or condition that resulted in the condition of the nuclear power plant, including its, principal safety barriers, being seriously degraded.

This event was also reportable within eight hours in accordance with 10 CFR 50.72(b)(3)(ii), Any event or condition that results in: (A) The condition of the nuclear power plant, including its principal safety barriers, being seriously degraded. This report (Event Number 44171) was submitted Within the required eight hours on April 26, 2008, at 0511.

Cause of Event

The apparent cause of the leak was determined to be vibratory fatigue based on the configuration making it susceptible to high cycle fatigue and the source of the energy being the Reactor Coolant Pump. This high cycle fatigue caused flaw propagation and the subsequent leak.

Analysis of Event

The leak was small (well within the capability of normal charging), found while the unit was shut down as part of start-up preparation during a refueling outage, and posed no significant nuclear safety threat. The leak did not directly affect any safety/risk significant equipment, and therefore posed no significant risk. From a deterministic standpoint, all plant safety function equipment was available at the time this leak developed. This event was not a threat to public safety and health in that: Unit 1 core decay heat load was extremely low following the, refueling outage; all required safety function equipment was available; the leak was small; and the other similar welds that were examined were acceptable.

Corrective Actions

Repair was performed using a weld overlay procedure per ASME Code Case N-666, Weld Overlay of Class 1, 2, and 3 Socket Welded Connections, Section XI, Division 1.

Use of Code Case N-666 for the repair as an alternative to a code repair was verbally approved by the Nuclear Regulatory Commission (NRC) on April 26, 2008.

Written documentation of the NRC approval is forthcoming.

The repair consisted of a seal weld of the leak followed by a VT-1 Examination prior to the overlay procedure. Additionally, a VT-1 Examination and a Dye-Penetrant Examination were completed following the overlay to verify the repair was performed as described in the Code Case, 'the Work Order Task, and the Proposed Alternative to the ASME Section XI Repair Requirements.

Following the weld repair, vibration data was taken. This was performed to meet the requirements of Code Case N-666. The results were found to fall within the acceptance criteria, making the piping acceptable without further analysis.

An extent of condition review was performed that included examination of the other welds of similar construction and design for size and quality. The examination met the standards for VT-1. The examined welds met the quality and size specification requirements. Additionally, actions for similar examinations on Unit 2 have been entered into the Corrective Action System.

Previous Similar Events

A search of the past ten years found no previous similar events of critical weld failures causing Non-Isolable Reactor Coolant System Pressure Boundary Leakage.