05000461/LER-2007-003

LER-2007-003, IGSCC Causes Pressure Boundary Leak and Reactor Shutdown

Docket Number Sequential Revmonth Day Year Year Month Day Year Nonenumber No. 05000
Event date:	06-18-2007
Report date:	08-16-2007
4612007003R00 - NRC Website

PLANT OPERATING CONDITIONS PRIOR TO THE EVENT

Unit: 1� Event Date: 6/18/07 � Event Time: 0433 Central Daylight Time Mode: 1 (Power Operation)� Reactor Power: 97 percent

DESCRIPTION OF EVENT

On June 18, 2007, the plant was operating in Mode 1 at 97 percent power. At 0433 hours0.00501 days <br />0.12 hours <br />7.159392e-4 weeks <br />1.647565e-4 months <br />, the Main Control Room (MCR) received several alarms [ALM], including Transient Test system trouble, fission product particulate high radiation, fission product high iodine, and Drywell (DW) continuous air monitor iodine channel and low noble gas channel alarms. The Transient Test system showed the "C" Main Steam [SB] Line (MSL) elbow tap differential pressure spiked from 9.8 to 10.1 pounds per square inch differential.

Operators checked the MSL guard pipe temperatures and Reactor Recirculation [AD] seal [SEAL] parameters and found them unchanged. The DW head temperature increased from 200 degrees Fahrenheit (F) to 211 degrees F. The DW pressure rate of change also increased. At 0520 hours0.00602 days <br />0.144 hours <br />8.597884e-4 weeks <br />1.9786e-4 months <br />, the DW floor drain in­ leakage was 0.3 gallons per minute (gpm) and slowly trending up. These conditions were indicative of a possible high energy leak in the drywell.

Operators entered the Abnormal Release of Airborne Radioactivity and the Reactor Coolant Leakage emergency operating procedures.

At 0800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br />, the unidentified leakage was less than the Technical Specification (TS) limits of 5 gpm (0.7 gpm) for unidentified leakage, of 30 gpm (2.7 gpm) for total leakage averaged over the previous 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, and of 2 gpm increase in identified leakage in the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (went from 0.09 to 0.6 gpm) in Mode 1.

Operators continued to monitor the steam leak indications, and at 1241 hours0.0144 days <br />0.345 hours <br />0.00205 weeks <br />4.722005e-4 months <br />, Operators made a decision to perform a plant shut down starting at 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> in order to access the DW to identify the source of the steam leak and make repairs. The shutdown process started at 2011 hours0.0233 days <br />0.559 hours <br />0.00333 weeks <br />7.651855e-4 months <br />.

On June 19, at about 0635 hours0.00735 days <br />0.176 hours <br />0.00105 weeks <br />2.416175e-4 months <br />, with the plant in Mode 2 (Startup/Hot Standby) and reactor power at about 1 percent, Maintenance personnel entered the DW and identified the source of the steam leak was a one­ inch diameter ASME Section III Class II stainless steel braided, flexible hose assembly [PSX] on the "C" Main Steam Line flow elbow low-pressure instrumentation tap. Leakage from this source is classified as pressure boundary leakage; therefore, operators entered the actions of Technical Specification 3.4.5 that require the plant be in Mode 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and Mode 4 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Operators fully inserted all control rods by 1103 hours0.0128 days <br />0.306 hours <br />0.00182 weeks <br />4.196915e-4 months <br /> on June 19.

The braided, flexible hose assembly had a steam leak mid-way through the side of the outer bellows, on the outside bend of the component. During disassembly and examination of the removed braided flexible hose assembly, the leak was found to be a circumferential crack on the inner bellows, approximately 180° and 1/16" wide on an outer convolution, located 2-1/2" from the steam end of the flexible hose assembly. Critical portions of the flexible hose assembly were sectioned and shipped to an off-site laboratory for failure analysis. The analysis by the off-site laboratory identified that Intergranular Stress Corrosion Cracking (IGSCC) was the cause of the braided flexible hose failure.

Specifically, IGSCC caused the inner bellows failure. The crack initiated from the inner diameter surface.

After the inner bellows pressure boundary failed, the outer bellows was exposed to the instrument line environment and also failed due to IGSCC. The bellows failures do not appear to be associated with improper installation, external mechanical damage, or fabrication seam weld defects.

No other inoperable equipment or components directly affected this event.

The root cause investigation and corrective actions for this event are tracked under Issue Report 641375.

CAUSE OF EVENT

The cause of this event was IGSCC. IGSCC will eventually occur when a susceptible material is installed in susceptible environment. IGSCC cannot form in a location where any one of three elements, (susceptible material, tensile stress and corrosive environment) are not present. Contributing causes were: (1) The cold forming process required to manufacture bellows from straight tube produced a susceptible material by changing the material microstructure. The new microstructure is susceptible to IGSCC if the other elements for IGSCC are present. (2) The bellows has increased residual stresses due to cold forming process. High Tensile stress is a component required for IGSCC initiation. The high residual stress in the bellows allows IGSCC initiation with smaller contributions from the other two elements for IGSCC initiation. (3) The Main Steam line provides the corrosive environment necessary for IGSCC initiation. The low pressure side of the flow elbow contains a mostly steam environment with some condensate. The steam environment with a presence of condensate appears to provide an IGSCC initiation condition. Thus, the three elements necessary for IGSCC were present.

SAFETY ANALYSIS

This event is reportable under the provisions of 10 CFR 50.73 (a) (2) (i) (A) for the plant shutdown required by TS 3.4.5, 10 CFR 50.73 (a) (2) (ii) (A) for the condition of the nuclear power plant including its principal safety barriers being seriously degraded, and 10 CFR 50.73 (a) (2) (i) (B) for operation prohibited by Technical Specifications, in that TS 3.4.5 allows no pressure boundary leakage.

This event had minimal safety significance. During this event, TS limits for RCS leakage were not exceeded.

The initial leakage was considered to be unidentified leakage and was less than the TS limits of: 5 gpm for unidentified leakage; 30 gpm for total leakage averaged over the previous 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period; and 2 gpm increase in identified leakage in the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in Mode 1.

Although the steam leak added heat load to the ventilation system and condensed water volume to the floor drain system, the leakage was within the system capabilities of the reactor coolant system inventory makeup, the drywell floor drain system, and the drywell ventilation system. There was no release of radioactive material to the environment. Operators continually monitored the leakage during this event.

This event report does not identify any safety system functional failures.

CORRECTIVE ACTION

The flexible hose assembly that failed has been replaced. Other in-service flexible hose assemblies installed in IGSCC susceptible locations are scheduled to be replaced. Susceptible flexible hose assemblies not currently in service are scheduled to be cut out and the lines capped. Additionally, preventive maintenance activities will be established to replace the susceptible hose assemblies at a frequency of 16 years. A modification will be required for the replacement hose assemblies since the original manufacturer is no longer in business. Engineering will evaluate replacement materials for the new hoses as part of the modification process.

PREVIOUS OCCURRENCES

None

COMPONENT FAILURE DATA

Manufacturer � Nomenclature � Manufacturer Model Number Metal Bellows � 1-inch ASME III, � N/A Corporation � Class II, Flexible � Part Number 78664 Braided High Pressure Stainless Steel Hose