05000265/LER-2004-002
| Quad Cities Nuclear Power Station Unit 2 | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded |
| 2652004002R00 - NRC Website | |
FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6 PAGE (3) Quad Cities Nuclear Power Station Unit 2 05000265 (If more space is required, use additional copies of NRC Form 366A)(17)
PLANT AND SYSTEM IDENTIFICATION
General Electric - Boiling Water Reactor, 2957 Megawatts Thermal Rated Core Power Energy Industry Identification System (EIIS) codes are identified in the text as [XX].
EVENT IDENTIFICATION
Axial Flaws Detected in Recirculation Piping During Inservice Inspection
A. CONDITION PRIOR TO EVENT
Unit: 2 � Event Date: March 9, 2004 � Event Time: 1200 hours0.0139 days <br />0.333 hours <br />0.00198 weeks <br />4.566e-4 months <br /> Reactor Mode: 5 � Mode Name: Refueling � Power Level: 000% Refueling (5) - Mode switch in the. Shutdown or Refueling position with fuel in the reactor vessel and one or more vessel head closure bolts less than fully tensioned or with the head removed.
B. DESCRIPTION OF EVENT
On March 9, 2004, during a refueling outage on Unit 2 (Q2R17), two inside diameter connected indications that were axially oriented (i.e., axial flaws) were detected during a scheduled Inservice Inspection (ISI) of a Recirculation System [AD] weld (02B-S7). The length and depth of the flaws were indeterminate due to being orientated axially on the pipe side (downstream) of the weld. Currently there is no qualified Performance Demonstration Initiative (PDI) Ultrasonic Testing (UT) procedure to perform length and depth sizing of axial flaws in piping welds.
Although the axial indications could not be fully characterized using qualified PDI inspection techniques, the flaw sizes were estimated to be approximately 0.35 and 0.25 inches long and approximately 0.39 inches deep. Axial flaws are not a structural concern (as circumferential flaws are) but can be a leakage concern. In this case the flaws were not through-wall.
Weld 02B-S7 was part of the original ISI inspection scope for Q2R17. The configuration is a cross to pipe weld on the 22" "B" Recirculation System Piping Ring Header. The piping is austenitic stainless steel and is susceptible to intergranular stress corrosion cracking (IGSCC) and hence is part of the GL 88-01, "NRC Position on IGSCC in BWR Austenitic Piping," population. A review performed by Corned (now Exelon) in 1999 titled "Unit 2 GL 88-01 Remediation Plan" concluded that the weld received ineffective Induction Heat Stress Improvement (IHSI) in 1983 because the minimum temperature gradient was not met. The temperature gradient is a recommended IHSI process control parameter as described in BWRVIP-61, "Induction Heating Stress Improvement Effectiveness on Crack Growth on Operating Plants.
Subsequently, in 1999 the weld was reclassified from Category C (non-resistant materials stress improved after two years of operation) to Category D (non-resistant materials, no stress improvement) in accordance with the GL 88-01 program. During the Spring 2002 refueling outage (Q2R16), Mechanical Stress Improvement Process (MSIP) was applied on weld 02B-S7 to provide additional mitigation to IGSCC. Pre- and post-MSIP UT inspections that met PDI requirements were performed and no recordable indications were noted. Additionally, previous exams were conducted in FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6 PAGE (3) Quad Cities Nuclear Power Station Unit 2 05000265 (If more space is required, use additional copies of NRC Form 366A)(17) 1997, 1993, 1990, 1988, 1986, 1983, and 1975 with acceptable results.
Based on the review of the MSIP records for 02B-S7 during Q2R16, it was concluded that MSIP was successfully applied. Specifically, the minimum required change in pipe circumference was specified as 0.35" with a maximum change of 1". The actual change on weld 02B-S7 subsequent to MSIP application was measured at 0.36".
ASME Section XI, 1995 Edition through 1996 Addenda, Article IWB-3000, "Acceptance Standards," provides methodologies to evaluate acceptability of flaws for continued service. The flaws found in 02B-S7 could not meet the requirements of Articles IWB- 3410-1 "Acceptance Standards" or IWB-3600 "Analytical Evaluation of Flaws" and as a result were determined to be reportable in accordance with 10 CFR 50.73 (a)(2)(ii)(A) and NUREG 1022, Revision 2, section 3.2.4(A)(2). Specifically, Article IWB-3640 "Evaluation Procedures and Acceptance Criteria for Austenitic Piping" provides methodology, evaluation and acceptance criteria for continued service of a detected flaw. However, as previously stated, the flaws noted were axial and there currently is no PDI qualified procedure to size these types of flaws and therefore they could not be evaluated using criteria provided in Section XI.
Since the axial flaws could not be evaluated for continued service, a repair consisting of a two-layer overlay was applied in accordance with Code Case N-504-2, "Alternative Rules for Repair of Classes 1, 2, and 3 Austenitic Stainless Steel Piping Section XI, Division 1," paragraph (f)(3). Code Case N-504-2 has been endorsed by the NRC as documented in Regulatory Guide 1.147, Rev. 13, "Inservice Inspection Code Case Acceptability Section XI, Division 1," and the repair is therefore a Code-accepted repair.
The post-overlay UT inspection was successfully performed on March 14, 2004. In accordance with the GL 88-01 Program and BWRVIP-75, "BWR Vessel and Internals Project Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedules," (as approved by NRC SER dated May 14, 2002), the Quad Cities Nuclear Power Plant ISI program requires a re-inspection of weld 02B-S7 within three outages (no later than Q2R20). If no additional cracking is observed during the inservice examination, the weld will be reclassified as a Category E (cracked, reinforced by weld overlay) weld.
Quad Cities Nuclear Power Station has effectively utilized both Noble Metal Chemical Addition (NMCA) and Hydrogen Water Chemistry (HWC) to mitigate IGSCC. Unit 2 has been operating under HWC since 1990 and NMCA since February 2000. Additionally, the water chemistry meets the Electric Power Research Institute (EPRI) guidelines. Unit 2 HWC system availability has been in excess of 96% for the previous cycle.
C. C CAUSE OF EVENT Weld 02B-S7 was examined following the Reactor Recirculation System decontamination to minimize inspection dose. It is concluded that more than likely these flaws were present and undetected by previous inspections. The observance of the axial flaws during this inspection is the result of the Low Oxidation State Metal Ion (LOMI) decontamination process, performed during Q2R17, prior to the inspection. An EPRI study ("Influence of Induction Heating Stress Improvement Treatment on Ultrasonic Reflectivity of IGSCC", EPRI NP-3815-SR, dated December 1984) has shown that decontamination efforts have been known to remove corrosion product from the crack tip and from the crack mouth of pre-existing flaws, thereby producing a better UT reflector. Specifically, corrosion products and oxide layers that filled the cracks DOCKET NUMBER (2) LER NUMBER (6 PAGE (3) FACILITY NAME (1) Quad Cities Nuclear Power Station Unit 2 05000265 � (If more space is required, use additional copies of NRC Form 366A)(17) were removed during the LOMI (decontamination) process. Hence, once the oxides were removed from the cracks the UT detectability was enhanced.
EPRI has acknowledged that the detection rate of axially oriented IGSCC flaws is significantly lower than for circumferentially oriented IGSCC flaws. HWC/NMCA have been effectively utilized to mitigate IGSCC. Coupled with the additional benefits of various stress improvement mitigation techniques (e.g. IHSI and MSIP), the probability that cracking occurred following Q2R16 is low. The most probable cause of the axial flaw detection was the increased reflectivity of the existing cracks as a result of the piping decontamination process.
D. SAFETY ANALYSIS
The safety significance of this event was minimal. The flaw indication was not through-wall, so there was no leakage. The flaw indication was axial and did not present a structural concern. An overlay of the weld was performed in accordance with industry guidance. At the time that the flaw indication was identified, weld 02B-S7 was classified as a Category D weld, and there were five Category D welds (including 02B-S7) on Unit 2. The required sample expansion resulted in all of the Category D welds on Unit 2 being inspected. No flaws were found on the remaining four welds, three of which were also in the LOMI-treated zone. The sample expansion performed was in accordance with the requirements of the GL-88-01 program and BWRVIP-75 "BWR Vessel and Internals Project Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedules" (as approved by NRC SER dated May 14, 2002).
E. CORRECTIVE ACTIONS
Immediate Actions
An overlay was applied to weld 02B-S7 in accordance with industry guidance.
The other four Category D welds were inspected, and no additional flaws were identified.
F. PREVIOUS OCCURRENCES
A root cause evaluation performed in November 1998, titled "Quad Cities Reactor Recirculation System IGSCC Root Cause Report (Problem Identification Form Q1998- Remediation Project. The purpose of the Remediation Project was to determine the effectiveness of previous IGSCC mitigation efforts. Weld 02B-S7 was identified as having an ineffective application of IHSI because the minimum temperature gradient was not met. As a result, weld 02B-S7 was reclassified and the subsequent MSIP (during Q2R16) and inspection (during Q2R17) were corrective actions to address the ineffective IHSI. There have been no previous events regarding detectability of axial flaws at Quad Cities Station. Also, no flaw indications have been identified previously on weld 02B-S7.
FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6 PAGE (3) Quad Cities Nuclear Power Station Unit 2 05000265 (If more space is required, use additional copies of NRC Form 366A)(17) G. � COMPONENT FAILURE DATA There were no component failures associated with this event.
NRC FOR/4 366A (7-2001)