05000331/LER-2007-003
| Docket Numbersequential Revmonth Day Year Year Month Day Yearnumber No. 05000 | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded |
| 3312007003R00 - NRC Website | |
FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6 PAGE (3) Duane Arnold Energy Center 05000331 YEAR
I. Description of Event:
During the Duane Arnold refueling outage 20 (RF020), ultrasonic inspections of welds susceptible to IGSCC were performed in accordance with the BWRVIP-75 and ASME Section XI inspection programs.
On 2-18-07, ultrasonic inspection of the N2F reactor recirculation riser safe-end to nozzle weld, RRF-F002, identified a circumferential flaw. The RRF-F002 circumferential weld indication was approximately 5.9" (150mm) long by 0.59" (15mm) deep (55.6% through-wall (TW)) and identified as ID surface connected.
As part of the BWRVIP and ASME Section XI Inspection Programs, the inspection population was expanded to determine the extent of condition (CAP 47960). On 2-21-07 another circumferential indication was identified in the N2C reactor recirculation riser safe-end to nozzle weld, RRC-F002. The flaw indication in the N2C safe end weld, RRC-F002, was 6.3" (160mm) long by 0.787" (20mm) deep (74% TW) and identified as ID surface connected. Both welds were dissimilar metal welds (ASME Category B-F) made with alloy 82/182 weld material. Both flaws exceeded the acceptance criteria of 10.4% TW in ASME Section XI, Table IWB-3410-1 and Table IWB-3514-2. A component that does not meet the acceptance standards of Table IWB-3410-1 and IWB-3514 for Category B-F welds shall be corrected by repair/replacement (IWB-3132.2), or justified for acceptance using analytical evaluation (IWB-3132.3).
Both welds were subsequently overlaid under an NRC verbally approved relief request.
This event was reported to the NRC as an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> event under 10 CFR 50.72(b)(3)(ii)(A), Any event or condition that results in: (A) The condition of the nuclear power plant, including its principal safety barriers being seriously degraded.
II.Assessment of Safety Consequences:
This report is being submitted pursuant to 10CFR50.73(a)(2)(ii)(A).
An evaluation of the known sized flaws was performed to determine whether the flaws met the acceptance criteria established in ASME Section XI, IWB-3600. The report concluded the following:
1. The allowable through-wall flaw depth for the two flawed nozzle-to-safe end welds is very large. They are at least 75% of pipe wall and could be as large as through-wall if the 75% through-wall limit imposed by ASME Section XI is not considered.
2. The N2C and N2F nozzle flaws met the ASME Code Section XI IWB-3640 requirements (75% of wall for the as-found lengths) at the time of the inspection.
3. Under Normal Water Chemistry (NWC) conditions, it will take an initial 10% flaw at least 19 months to reach the allowable flaw depth (75% of wall) and under hydrogen water chemistry conditions, it will take at least 195 months for the initial 10% through-wall flaw to reach the allowable flaw depth (75% of wall).
4. Under Hydrogen Water Conditions (HWC) conditions, the flaw at the N2F nozzle weld would be acceptable for at least one additional operating cycle based on the crack growth analysis results.
Although the N2C flaw was slightly less than the 75% allowable flaw depth limit, the flaw would be predicted to exceed the 75% limit during the next operating cycle.
5. Based on the fact that Duane Arnold Energy Center (DAEC) is on HWC, it is very likely that both flaws have been present for a significant time, e.g. multiple cycles, and were not identified by previous inspections.
FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3) Duane Arnold Energy Center 05000331 YEAR Therefore, this event has no nuclear safety significance because the flaws were within Code allowable when discovered and the affected areas have overlays installed to restore long term structural integrity
III.Cause of Event:
The cause of the circumferential flaws identified in the RRF-F002 (N2F) and RRC-F002 (N2C) reactor recirculation riser safe-end to nozzle welds is IGSCC. The cause is based on the following considerations:
The industry experience that alloy 82/182 butt welds are susceptible to IGSCC, even in HWC. Weld repairs made on the N2F safe end weld (10-23-78 to 10-25-78 in N2F repair history), and weld repairs and potential oxidation of the root of the N2C (10-28-78 N2C repair notation) during the 1978/79 recirculation riser safe end replacement, may have increased susceptibility to stress corrosion cracking (SCC) and contributed to the flaw initiation. The flaw length, depth, and orientation characteristics are comparable to other flaws identified in the BWR industry that have also been attributed to IGSCC. In addition, a review of the automated Non-Destructive Examination (NDE) data by EPRI concluded that the flaw ultrasonic responses have similar characteristics to a crack that was specifically fabricated to simulate the response from SCC in this type of weld configuration.
IV.Corrective Actions:
Corrective Actions to Restore Two Work Orders were planned to perform repairs of the safe-end to nozzle welds (WO 1139064 (RRC- F002) and WO 1139059 (RRF-F002)). The repair was to overlay the flawed welds with material (Alloy 52M) resistant to SCC. The implementation of the overlay required a relief request from the NRC to approve the use of two code cases (N-504-2 and 638-1). The overlays were successfully applied and the ultrasonic examination results were found to be acceptable.
Interim Corrective Actions
None identified.
Corrective Actions to Prevent Recurrence
CATPR 1-1 (CA045700), Mitigation techniques for the remaining safe-end to nozzle welds that have not been overlaid (RRA-F002, RRE-F002, RRG-F002 and RRH-F002) shall be completed prior to completion of RFO22. The mitigation techniques available are Mechanical Stress Improvement (MSIP) or weld overlay.
The overlays are considered acceptable as a mitigation technique for the two welds (RRC-F002 and RRF- F002). For the remaining safe-end to nozzle welds (RRA-F002, RRE-F002, RRG-F002 and RRH-F002) an evaluation of a mitigation technique shall be completed prior to next refueling outage. The mitigation techniques available are Mechanical Stress Improvement (MSIP) or weld overlay. It is known that the Pressurized Water Reactors are addressing their primary water stress corrosion cracking (PWSCC) issue by implementing "mitigative" overlay and using Code Case N-740 (note that the ASME Section XI committee has revised Code Case N-740 to address many of the NRC concerns).
FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3) Duane Arnold Energy Center 05000331 YEAR
V. Additional Information:
Previous Similar Occurrences:
In 1999, stress corrosion cracking was found in two safe-end to nozzle welds (RRB-F002 and RRD-F002).
Both of these welds were overlaid using an approved relief request from the NRC. The specifics are discussed in the above LER.
El IS System and Component Codes:
AD Reactor Recirculation System Reporting Requirements:
This report is being submitted under 10 CFR 50.73(a)(2)(ii)(A)