05000331/LER-2007-003, Re Linear Indications Found During UT Examination of Safe-End to Nozzle Welds
| ML071150319 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 04/20/2007 |
| From: | Vanmiddlesworth G Duane Arnold |
| To: | Document Control Desk, Plant Licensing Branch III-2 |
| References | |
| NG-07-0364 LER 07-003-00 | |
| Download: ML071150319 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(ix)(A) 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 3312007003R00 - NRC Website | |
text
FPL Energyo Duane Arnold Energy Center FPL Energy Duane Arnold, LLC 3277 DAEC Road Palo, Iowa 52324 April 20, 2007 NG-07-0364 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555-0001 Duane Arnold Energy Center Docket 50-331 License No. DPR-49 Licensee Event Report #2007-003-00 Please find attached the subject Licensee Event Report (LER) submitted in accordance with 10 CFR 50.73. This letter contains no new commitments.
Gary Van Middlesworth Site Vice President, Duane Arnold Energy Center FPL Energy Duane Arnold, LLC cc:
Administrator, Region IIl, USNRC Project Manager, DAEC, USNRC Resident Inspector, DAEC, USNRC
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 06/30/2007 (6-2004)
, the NRC (See reverse for re uired number of may not conduct or sponsor, and a person is not required to respond to, the diqits/characters for each block) information collection.
- 3. PAGE Duane Arnold Energy Center 05000 331 1
OF 4
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED MONTH DAY YEAR YEAR SEQUENTIAL REV MONTH DAY YEAR FACILITY NAME DOCKET NUMBER NUMBER NO.
05000 FACILITY NAME DOCKET NUMBER 02 18 2007 2007 3
0 04 20 2007 05000
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply) 5 20.2201(b)
El 20.2203(a)(3)(i)
E] 50.73(a)(2)(i)(C)
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E] 50.73(a)(2)(viii)(B)
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- 10. POWER LEVEL [j 20.2203(a)(2)(ii)
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Specify in Abstract below or in NRC Form 366A
- 12. LICENSEE CONTACT FOR THIS LER FACILITY NAME ITELEPHONE NUMBER (Include Area Code)
Robert J. Murrell, Regulatory Affairs Engineering Analyst
] (319) 851-7900CAUSE SYSTEM COMPONENT MANU-REPORTABLE CYO MANU-REPORTABLE FACTURER TO EPIX
CAUSE
SYSTEM COMPONENT FACTURER TO EPIX
- 14. SUPPLEMENTAL REPORT EXPECTED
- 15. EXPECTED MONTH DAY YEAR Y
NOSUBMISSIONATE 0]YES (If yes, complete 15. EXPECTED SUBMISSION DATE)
]NODATE ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)
Safe-end to nozzle weld RRF-F002 was scheduled in RFO20 as an examination under BWRVIP-75 and as a successive examination under ASME Section Xl. On 2/18/07, the examination revealed a linear indication at the bottom of the weld that did not meet the acceptance criteria (IWB-3500) of ASME Section XI. Based on this unacceptable linear indication a scope expansion was required for both programs (BWRVIP-75 and ASME Section Xl). The scope expansion included safe-end to nozzle weld RRC-F002. On 2/21/07, this weld also was determined to not meet the acceptance standards (IWB-3500) of ASME Section X1. Both welds were subsequently overlaid under an NRC verbally approved relief request.
The root cause of the indications in both welds (RRF-F002 and RRC-F002) was determined to be Intergranular Stress Corrosion Cracking (IGSCC). This is based on the signal discrimination of the Ultrasonic Test examination, a review of the operating experience (OE) and the independent review of the event by outside organizations.
There were no actual safety consequences and no effect on public health and safety as a result of this event.
NRC FORM 366 16-2004)
PRINTED ON RECYCLED PAPER NRC FORM 366 (6-2004)
PRINTED ON RECYCLED PAPERU.S. NUCLEAR REGULATORY COMMISSION (1-2001)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
DOCKET NUMBER (2)
LER NUMBER 6)
PAGE 3 YEAR SEQUENTIAL REVISION Duane Arnold Energy Center 05000331 NUMBER NUMBER 2007
-- 003 00 2of4 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)
I. Description of Event
During the Duane Arnold refueling outage 20 (RFO20), 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" (1 50mm) 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.
I1. Assessment of Safety Consequences:
This report is being submitted pursuant to 1 OCFR50.73(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.U.S. NUCLEAR REGULATORY COMMISSION (1-2001)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
DOCKET NUMBER (2)
LER NUMBER (6)
PAGE (3) 0YEAR SEQUENTIAL REVISION Duane Arnold Energy Center 05000331 NUMBER NUMBER 2007
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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 IIl. 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 (110-23-78 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 RF022. 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 Xl committee has revised Code Case N-740 to address many of the NRC concerns).U.S. NUCLEAR REGULATORY COMMISSION (1-2001)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
DOCKET NUMBER (2)
LER NUMBER (6)
PAGE (3)
E SEQUENTIAL REVISION Duane Arnold Energy Center 05000331 NUMBER NUMBER 2007
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V. Additional Information
Previous Similar Occurrences:
LER 1999-006, "Indications in Recirculation Riser Nozzle-to-Safe End Welds."
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.
EIIS System and Component Codes:
AD Reactor Recirculation System
Reporting Requirements
This report is being submitted under 10 CFR 50.73(a)(2)(ii)(A)