05000338/LER-2012-001, Regarding Degraded Reactor Coolant System Piping Due to Primary Water Stress Corrosion Cracking
| ML12151A441 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 05/18/2012 |
| From: | Gerald Bichof Virginia Electric & Power Co (VEPCO) |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 12-264 LER 12-001-00 | |
| Download: ML12151A441 (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 |
| 3382012001R00 - NRC Website | |
text
1 OCFR50.73 Virginia Electric and Power Company North Anna Power Station P. 0. Box 402 Mineral, Virginia 23117 May 18, 2012 Attention: Document Control Desk Serial No.:
12-264 U. S. Nuclear Regulatory Commission NAPS:
MPW Washington, DC 20555-0001 Docket No.:
50-338 License No.: NPF-4
Dear Sirs:
Pursuant to 10CFR50.73, Virginia Electric and Power Company hereby submit the following Licensee Event Report applicable to North Anna Power Station Unit 1.
Report No. 50-338/2012-001-00 This report has been reviewed by the Facility Safety Review Committee and will be forwarded to the Management Safety Review Committee for its review.
Sincerely, Gerald T. Bischof Site Vice President North Anna Power Station Enclosure
.Commitments contained in this letter: None cc:
United States Nuclear Regulatory Commission Region II Marquis One Tower 245 Peachtree Center Ave., NE, Suite 1200 Atlanta, Georgia 30303-1257 NRC Senior Resident Inspector North Anna Power Station
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 EXPIRES: 10/31/2013 (10-2010)
, the NRC may digits/characters for each block) not conduct or sponsor, and a person is not required to respond to, the information collection.
- 3. PAGE North Anna Power Station, Unit 1 05000 338 1 OF 4
- 4. TITLE Degraded Reactor Coolant System Piping Due To Primary Water Stress Corrosion Cracking
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED MONTH DAY YEAR YEAR SEQUENTIAL REV MONTH DAY YEAR FACILITY NAME DOCUMENTNUMBER I
NUMBER NO.
05000 03 24 2012 2012
--001 --
00 05 18 2012 FACILITY NAME DOCUMENTNUMBER 05000
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)
El 20.2201(b)
Ej 20.2203(a)(3)(i) 50.73(a)(2)(i)(C)
Ej 50.73(a)(2)(vii)
El 20.2201(d)
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20.2203(a)(3)(ii) 50.73(a)(2)(ii)(A)
E] 50.73(a)(2)(viii)(A) 6
[-
20.2203(a)(1)
Ej 20.2203(a)(4) 50.73(a)(2)(ii)(B) 50.73(a)(2)(viii)(B)
E] 20.2203(a)(2)(i)
Ej 50.36(c)(1)(i)(A) 50.73(a)(2)(iii) 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL
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20.2203(a)(2)(ii)
Ej 50.36(c)(1)(ii)(A)
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50.73(a)(2)(iv)(A)
El 50.73(a)(2)(x)
E] 20.2203(a)(2)(iii)
[F]
50.36(c)(2)
El 50.73(a)(2)(v)(A)
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73.71(a)(4)
El 20.2203(a)(2)(iv)
[] 50.46(a)(3)(ii)
El 50.73(a)(2)(v)(B)
E] 73.71(a)(5)
El 20.2203(a)(2)(v)
LI 50.73(a)(2)(i)(A)
El 50.73(a)(2)(v)(C)
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OTHER Refueling j]
20.2203(a)(2)(vi)
Fl 50.73(a)(2)(i)(B)
[
50.73(a)(2)(v)(D)
Specifyin Abstract below or in
2.0 SIGNIFICANT SAFETY CONSEQUENCES AND IMPLICATIONS
A comparison is made between a bounding axial flaw length and the allowable flaw length limit given in Appendix C to ASME Section Xl to demonstrate stability of a bounding through-wall axial flaw. Assuming PWSCC as the only crack growth mechanism, this evaluation demonstrates that the existence of a bounding axial through-wall flaw spanning the entire length of the PWSCC susceptible material (-2.0 inches) could not have reached an unstable axial length during previous operations.
The applied hoop stress is less than the projected allowable stress in ASME Section X1 requirements considering a ligament of 0.57 inches (machined depth of 4.1 inches).
However, a postulated flaw depth of 4.1" (depth-to-thickness ratio a/t=0.88) exceeds the limits of applicability in the code of a/t=0.75. And since the true depths of the uncovered axial flaws prior to machining are unknown, it is reasonable to assume they extended to a higher a/t ratio. Therefore, the as-found flaws, while potentially within acceptable stress limits, did not meet the ASME Section XI code requirements.
Since pressure loading is the only stressor considered for stability of a component containing axial flaws, the only postulated events that would have resulted in any additional challenges to the structural integrity of the nozzle, other than the normal operation experienced, would have been pressure transients. For the Reactor Coolant System (RCS), pressure transients are not significant relative to the normal operating pressure.
Therefore, there would have been no significant challenges to the structural integrity of the nozzle connection related to the discovered axial flaws during previous operation.
The assessments provide reasonable assurance that gross structural failure of the "B" hot leg inlet nozzle would not have occurred during previous operations for any postulated design basis events including Design Basis Earthquake. This is supported by using ASME Section Xl, Appendix C methods for evaluating hoop stress and stability of a bounding axial through-wall flaw. Had PWSCC continued to the point of an axial through-wall leak, the consequences of that outcome would be minimized given that the overall nozzle connection would not have been significantly challenged. As such, the event posed no significant safety implications and the health and safety of the public were not affected by the event.
3.0
CAUSE
The Direct Cause determined the leakage developed due to PWSCC in a susceptible material (i.e. Alloy 82/182). After review of the fabrication records, it is apparent that the primary water stress corrosion cracks were caused by extensive ID weld repairs performed only on the 'B' Hot Leg nozzle. There were no ID weld repairs performed on the 'A' or 'C' Hot Leg nozzles. Repairs performed during SG fabrication to the 'B' Hot Leg DM weld were made with a susceptible material (i.e. Alloy 182) in a sequence which would result in the formation of adverse residual tensile stresses. A combination of tensile stresses and low
chromium weld material on the ID of the DM weld in a high temperature, primary water environment directly led to the formation of stress corrosion cracks. Because NDE has characterized the cracks as being branched and fully contained within the DM weld; the identified cracks are the result of PWSCC and not some other mechanical or thermal loading mechanisms.
4.0 IMMEDIATE CORRECTIVE ACTION(S)
After confirmation that the water which was slowly accumulating along the length of the indication was reactor coolant, the 'B' SG was drained and the through-wall cracks were seal welded under the provisions of Relief Request N1-14-CMP-001.
5.0 ADDITIONAL CORRECTIVE ACTIONS
Extent of Condition volumetric examinations were performed on the Unit 1 SG Cold Leg nozzles during the Spring 2012 refueling outage per the Section XI program. No flaws were identified. The FSWOLs were completed on all three Unit 1 SG Hot Leg nozzles with satisfactory tests results.
Unit 2 SG hot and cold leg nozzles will have UT examinations performed during the 2013 refueling outage. Note that Unit 2 SG nozzles contain an alloy 52/152 weld inlay that significantly reduces susceptibility to PWSCC.
6.0 ACTIONS TO PREVENT RECURRENCE The potential that PWSCC could be present in other systems/components is limited to whether Alloy 600 material was used. Continued inspections and testing performed by the ISI program will identify any future cases of defects propagating in similar components.
7.0
SIMILAR EVENTS
LER 50-339/02-001-00 documents the Unit 2 nozzle through-wall leakage of three reactor vessel head penetrations.
LER 50-339/01-003-00 and Supplemental LER 50-339/01-003-01 documents the Unit 2 nozzle through-wall leakage of three reactor vessel head penetrations.
8.0
ADDITIONAL INFORMATION
Unit 2 was operating in Mode 1, 100 percent power on March 24, 2012.