05000400/LER-2018-002, Reactor Pressure Vessel Closure Head Penetration Nozzle Indications Attributed to Primary Water Stress Corrosion Cracking
| ML18158A292 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 06/07/2018 |
| From: | Hamilton T Duke Energy Progress |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| HNP-18-052 LER 2018-002-00 | |
| Download: ML18158A292 (4) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 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)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i) |
| 4002018002R00 - NRC Website | |
text
£ -, DUKE
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JUK O 7 2018 Serial: HNP-18-052 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Shearon Harris Nuclear Power Plant, Unit 1 Docket No. 50-400/Renewed License No. NPF-63 Subject: Licensee Event Report 2018-002-00 Ladies and Gentlemen:
Tanya M. Hamilton Vice President Harris Nuclear Plant 5413 Shearon Harris Road New Hill, NC 27562-9300 919.362.2502 10 CFR 50.73 Duke Energy Progress, LLC, submits the enclosed Licensee Event Report 2018-002-00 in accordance with 10 CFR 50. 73 for Shearon Harris Nuclear Power Plant, Unit 1. This report details two rejected indications associated with a control rod drive mechanism nozzle penetration, identified during the reactor vessel closure head inspection completed during the last refueling outage. All rejected indications have been restored to code compliance.
This document contains no regulatory commitments. Please refer any questions regarding this submittal to Jeffrey Robertson, Manager - Regulatory Affairs, at (919) 362-3137.
Sincerely, Tanya M. Hamilton Enclosure: Licensee Event Report 2018-002-00 cc:
J. Zeiler, NRC Senior Resident Inspector, HNP M. Barillas, NRC Project Manager, HNP C. Haney, NRC Regional Administrator, Region II
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3.Page Shearon Harris Nuclear Power Plant - Unit 1 05000 400 1
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- 4. Title Reactor Pressure Vessel Closure Head Penetration Nozzle Indications Attributed to Primary Water Stress Corrosion Cracking
- 5. Event Date
- 6. LER Number
- 7. Report Date
- 8. Other Facilities Involved Year I
Sequential I Rev Month Day Year Facility Name Docket Number Month Day Year Number No.
05000 Facility Name Docket Number 04 11 2018 2018 -
002 -
00 06 07 2018 05000
- 9. Operating Mode
- 11. This Report is Submitted Pursuant to the Requirements of 1 O CFR §: (Check all that apply) 20.2201(b) 20.2203(a)(3)(i) 0 50.73(a)(2)(ii)(A) 50.73(a)(2)(viii)(A) 20.2201(d) 20.2203(a)(3)(ii) 50.73(a)(2)(ii)(B) 50.73(a)(2)(viii)(B) 6 20.2203(a)(1) 20.2203(a)(4) 50.73(a)(2)(iii) 50.73(a)(2)(ix)(A) 20.2203(a)(2)(i) 50.36(c)(1)(i)(A) 50.73(a)(2)(iv)(A) 50.73(a)(2)(x) 1 O. Power Level
- 20.2203(a)(2)(ii) 50.36( c)(1 )(ii)(A) 50.73(a)(2)(v)(A) 73.71(a)(4) 20.2203(a)(2)(iii) 50.36(c)(2) 50.73(a)(2)(v)(B)
- 73. 71 (a)(5) 20.2203(a)(2)(iv) 50.46(a)(3)(ii) 50.73(a)(2)(v)(C) 73.77(a)(1) 000 20.2203(a)(2)(v)
- 50. 73(a)(2)(i)(A) 50.73(a)(2)(v)(D) 73.77(a)(2)(i) 20.2203(a)(2)(vi) 50.73(a)(2)(i)(B) 50.73(a)(2)(vii) 73.77(a)(2)(ii) 50.73(a)(2)(i)(C)
Other (Specify in Abstract below or in C. Causal Factors The cause of the indications in nozzle 33 was attributed to PWSCC, which occurs under conditions of high tensile stresses (operating and/or residual), conducive environment (temperature and chemistry), and susceptible material. The CROM nozzles in the SHNPP RVCH were originally constructed from Alloy 600 tubing and Alloy 82/182 weld metal. There is widespread industry operating experience that documents PWSCC of Alloy 600 dissimilar metal weld configurations.
D. Corrective Actions
The two PWSCC indications were repaired using the inside diameter temper bead weld method. All RVCH CROM nozzles were inspected, as required by ASME Code Case N-729-4, due to previously identified PWSCC indications.
E. Safety Analysis
After PWSCC was identified in RFO-17, inspections of the RVCH are required every refueling outage in accordance with ASME Code Case N-729-4, as conditioned by 10 CFR 50.55a. These inspections include NOE for all RVCH penetrations to identify indications, and are supplemented by bare metal visual examinations of the RVCH. If rejectable indications are found, repairs are required in accordance with both ASME Code and with relief requests submitted to the NRC on a case-by-case basis. This ensures indications are identified and repaired before there are any significant impacts on the integrity of the RVCH.
The In-Service Inspection examinations performed on the RVCH did not reveal any through-wall leakage. An ultrasonic leak path assessment and a bare metal visual examination of the reactor vessel head were completed, with no leakage identified. There was not a breach in the fission product barrier, and the structural integrity of the reactor vessel was not significantly compromised. Therefore, there was no significant impact to the health and safety of the public.
F. Additional Information
PWSCC has previously been detected in nozzles 5, 17, 38, 49, and 63 (RFO-17), 37 (RFO-18), 14, 18, and 23 (RFO-19),
and 30, 40, and 51 (RFO-20). LERs 2013-001-00, 2013-003-00, 2015-003-00, and 2016-006-00 all document previous experience with indications in the RVCH CROM penetration nozzles. Page 3
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