05000259/LER-2008-002
| Docket Numbersequential Revmonth Day Year Year Month Day Yearnumber No. None N/A | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded |
| 2592008002R01 - NRC Website | |
I. PLANT CONDITION(S)
Unit 1 was in Mode 4, approximately 1055 psig. Units 21 and 3 were at 100 percent power (3458 Megawatts thermal) and unaffected by the event.
II. DESCRIPTION OF EVENT
A. Event:
On November 23, 2008, at approximately 1200 hours0.0139 days <br />0.333 hours <br />0.00198 weeks <br />4.566e-4 months <br /> Central Standard Time (CST), during the performance of the Unit 1 vessel hydrostatic test, ASME Section XI System Leakage Test of the Reactor Pressure Vessel and Associated Piping, 1-SI-3.3.1.A, a leak was discovered on an unisolatable instrument line connected to the reactor vessel. This instrument line is an ASME Code Class 1 equivalent component, 2-inch pipe, near pressure vessel nozzle N11B. BFN entered Technical Requirements Manual (TRM) Section 3.4.3 - Structural Integrity, which requires the integrity of ASME Code Class 1, 2, and 3 equivalent components shall be maintained through the life of the plant.
Unit 1 was in Mode 4 at the time of discovery; it remained in Mode 4 until the repairs were completed.
TVA is submitting this report in accordance with 10 CFR 50.73(a)(2)(ii)(A) as any event or condition that resulted in the nuclear power plant, including principal safety barriers, being seriously degraded.
B. Inoperable Structures. Components. or Systems that Contributed to the Event:
None.
C. Dates and Approximate Times of Maior Occurrences:
November 23, 2008 1200 hours0.0139 days <br />0.333 hours <br />0.00198 weeks <br />4.566e-4 months <br /> CST� Identified Code Class 1 pressure boundary leak.
November 23, 2008 1600 hours0.0185 days <br />0.444 hours <br />0.00265 weeks <br />6.088e-4 months <br /> CST� Unit 1 is depressurized.
November 23, 2008 1813 hours0.021 days <br />0.504 hours <br />0.003 weeks <br />6.898465e-4 months <br /> CST TVA makes an eight-hour non-emergency notification to NRC in accordance with 10 CFR 50.72 (b)(3)(ii)(A).
D. Other Systems or Secondary Functions Affected None.
E. Method of Discovery The leak was identified by visual examination (VT-2) during a scheduled performance of 1-SI-3.3.1.A.
F. Operator Actions None.
G. Safety System Responses None.
III. CAUSE OF THE EVENT
A. Immediate Cause
The immediate cause of the event was a through wall leak in an 2 inch line nozzle near the safe end/instrument line interface weld very close to the heat affected zone (HAZ).
B. Root Cause
The root cause of the event was residual stress introduced to the safe end inside diameter during initial fabrication. The vessel manufacturer fabricated the safe end from a stainless steel forging. The forging was butt welded to the instrument nozzle. Then the inside diameter of the safe end was machined in place. This fabrication method resulted in high residual stresses on the inside diameter surface of the safe end. The butt welding operation typically results in higher heat input than socket or fillet weld. The higher heat input associated with the weld results in a larger HAZ and a sensitized microstructure that is conducive to IGSCC. In addition, the water in the area of the instrument penetrations contains oxidants, which can create an aggressive environment for the growth of IGSCC. The combination of the water chemistry, sensitized microstructure, and fabrication methodology promoted the growth of IGSCC and the eventual through wall leak.
C. Contributing Factors
None.
IV. ANALYSIS OF THE EVENT
There are three key contributors required to promote IGSCC in 304 type stainless steel pipes:
weld sensitized microstructure, an oxygenated environment, and tensile stress. A sensitized microstructure results from heating and cooling the material at various time and temperature combinations that form chromium carbides at the grain boundaries. These carbides deplete the surrounding area of chromium, providing a continuous path of lower corrosion resistance along the grain boundaries for the propagation of cracks. Welding can create this condition. The HAZ of the weld is susceptible to this carbide depletion. The through wall crack identified by the inspection was close to the HAZ.
A BWR environment contains dissolved oxygen that in the correct amounts promotes IGSCC growth. The water in the location of the safe end is high in oxidants, making the location of the N11B safe end susceptible to IGSCC growth. Residual stresses from the welding, grinding, and machining process, all contribute to the overall tensile stress in the safe end. The manufacturing and installation process for the N11B safe end created stresses that, when combined with the weld residual stress, the oxygenated environment and the sensitized weld HAZ, exceeded the yield strength of material. Once cracking initiated, it propagated through IGSCC susceptible metal and became a through wall leak.
V. ASSESSMENT OF SAFETY CONSEQUENCES
The safety consequences of this event were not significant. Plant Technical Specifications (TSs) require monitoring of reactor coolant leakage. When leakage limits are met, the TSs requires the reactor be placed in Mode 4. During the previous operating cycle, reactor coolant leakage was less than the TS limits.
The visual inspection during the performance of 1-SI-3.3.1.A identified the through wall leak at the instrument line connection to the reactor vessel. BFN entered Technical Requirements Manual (TRM) 3.4.3, Structural Integrity, Condition A, which requires immediate restoration of the structural integrity of the affected component or maintain the reactor in Mode 4 or 5 or the reactor coolant system less than 50 degrees F above the minimum temperature required for nondestructive testing considerations, until each indication has been investigated and evaluated. Until repairs were completed, BFN maintained the reactor in accordance with these requirements. Therefore, the event did not adversely affect the safety of the public and plant personnel.
VI. CORRECTIVE ACTIONS
A. Immediate Corrective Actions
Once TVA determined there was an unisolatable leak in the ASME Class 1 reactor pressure boundary, the reactor was depressurized to the pre-test pressure and Mode 4 was maintained.
B. Corrective Actions to Prevent Recurrence (1) To determine the extent of the through wall crack, the N11B safe end was examined ultrasonically (UT). The through wall leak was repaired by weld overlay. TVA added inservice examination requirements for N11B weld overlay to the Unit 1 inservice inspection program.
BFN ultrasonically examined the remaining Unit 1 small-bore (less than 4 inches in diameter) instrument nozzle safe ends and the core delta pressure - Standby Liquid Control [BR] line nozzle safe end. The examinations did not identify any other recordable indications.
VII. ADDITIONAL INFORMATION
A. Failed Components
None.
B. Previous LERs on Similar Events None.
C. Additional Information
Corrective action document for this report is PER 157918.
D. Safety System Functional Failure Consideration:
This event is not a safety system functional failure according to NEI 99-02.
E. Scram With Complications Consideration:
This event was not a complicated scram according to NEI 99-02.
VIII. COMMITMENTS
None.
TVA does not consider the corrective actions regulatory requirements. TVA tracks completion of the actions the Corrective Action Program.