05000455/LER-2007-001, From Byron Station Unit 2, Re Reactor Pressure Vessel Head Control Rod Drive Mechanism Penetration Nozzle Weld Indication Due to an Initial Constructive Weld Defect Allowing the Initiation of Primary Water Stress Corrosion Crac
| ML071590211 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 06/08/2007 |
| From: | Hoots D Exelon Generation Co, Exelon Nuclear |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| BYRON 2007-0072 LER 07-001-00 | |
| Download: ML071590211 (4) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii) 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) 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 |
| 4552007001R00 - NRC Website | |
text
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Exe~onGeneration Company, LLC www.exeloncorp coin N
1 Byron Btation 4430 North German Church Road Byron, FL 61010-9794 June 8, 2007 LTR:
BYRON 2007-0072 File:
1.10.0101 United States Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Byron Station, Units 2 Facility Operating License No. NPF-66 NRC Docket No. STN 50-455
Subject:
Licensee Event Report (LER) 455-2007-001 -00, Reactor Pressure Vessel Head Control Rod Drive Mechanism Penetration Nozzle Weld Indication Due to an Initial Construction Weld Defect Allowing the Initiation of Primary Water Stress Corrosion Cracking Enclosed is an LER for the issue involving the April 9, 2007 discovery of a flaw in the reactor pressure vessel head control rod drive mechanism penetration nozzle weld. This condition is reportable to the NRC in accordance with 10 CFR 50.73 (a)(2)(ii).
Should you have any questions concerning this matter, please contact Mr. William Grundmann, Regulatory Assurance Manager, at (815) 406-2800 Respectfully, David M. Hoots Site Vice President Byron Nuclear Generating Station DMH/JELrah Attachment LER 455-2007-001-00
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION (6-2004)
LICENSEE EVENT REPORT (LER)
(See reverse for required number of
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APPROVED BY 0MB: NO. 3150-0104 EXPIRES: 06/30/2007
, the NRC may not conduct or sponsor, and a person is not required to respond to, the infnrmntinn ~l~tion
- 1. FACILITY NAME Byron Station Unit 2
- 2. DOCKET NUMBER
- 3. PAGE 05000455 1 of 3
- 4. TITLE Reactor Pressure Vessel Head Control Rod Drive Mechanism Penetration Nozzle Weld Indication Due to an Initial Construction Weld Defect Allowing the Initiation of Primary Water Stress Corrosion Cracking
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED MONTH DAY YEAR YEAR SEQUE TIA NUMBER L NO.
MONTH DAY YEAR FACILITY NAME DOCKET NUMBER N/A 04 09 2007 2007 001 00 06 08 2007 FACILITY NAME DOCKET NUMBER
- 9. OPERATING MODE 6
- 11. THIS REPORT IS SUBMITTED PURSUANTTO THE REQUIREMENTS OF 10 CFR~:(Check all that apply) o 20.2201(b) 0 20.2203(a)(3)(i) 0 50.73(a)(2)(i)(C) 0 50.73(a)(2)(vii) 0 20.2201(d) 0 20.2203(a)(3)(ii)
Z 50.73(a)(2)(ii)(A) 0 50.73(a)(2)(viii)(A) o 20.2203(a)(1) 0 20.2203(a)(4) 0 50.73(a)(2)(ii)(B) 0 50.73(a)(2)(viiD(B) o 20.2203(a)(2)(i) 0 50.36(c)(1)(i)(A) 0 50.73(a)(2)(iii) 0 50.73(a)(2)(ix)(A) 0 20.2203(a)(2)(ii) 0 50.36(c)(1)(ii)(A) 0 50.73(a)(2)(iv)(A) 0 50.73(a)(2)(x) o 20.2203(a)(2)(iii) 0 50.36(c)(2) 0 50.73(a)(2)(v)(A) 0 73.71 (a)(4) o 20.2203(a)(2)çv) 0 50.46(a)(3)çi) 0 50.73(a)(2)(v)(B) 0 73.71 (a)(5) 0 20.2203(a)(2)(v) 0 50.73(a)(2)(i)(A) 0 50.73(a)(2)(v)(C) 0 OTHER o 20.2203(a)(2)(vi) 0 50.73(a)(2)(i)(B) 0 50.73(a)(2)(v)(D)
Specify in Abstract below or in NRC Form 366A
- 10. POWER LEVEL 000
- 12. LICENSEE CONTACT FOR THIS LER FACILITY NAME TELEPHONE NUMBER (Include Area Code)
Byron Station, William Grundmann, Regulatory Assurance Manager (815) 406-2800CAUSE SYSTEM COMPONENT FACTURER REPORTABLE
CAUSE
SYSTEM COMPONENT FACTURER REPORTABLE N/A N/A N/A N/A N/A
- 14. SUPPLEMENTAL REPORT EXPECTED o YES (If yes, complete 15. EXPECTED SUBMISSION DATE)
NO
- 15. EXPECTED SUBMISSION DATE MONTH DAY YEAR BSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)
In accordance with NRC order EA-03-009, Issuance Of Order Establishing Interim Inspection Requirements For Reactor Pressure Vessel Heads At Pressurized Water Reactors, Byrons Unit 2 reactor head was ranked in the low Primary Water Stress Corrosion Cracking (PWSCC) susceptibility category. This low ranking required the initial Control Rod Drive Mechanism (CRDM) penetrations volumetric examinations to be completed in the recent spring 2007 refuel outage (B2R13). During the B2R13 volumetric examination of the Unit 2 CRDM penetration nozzle 68, an ultrasonic test (UT) indication was discovered. Subsequently, bare-metal visual examination was completed, and no evidence of boric acid was identified. A boat sample from CRDM penetration 68 was obtained and analyzed. The root cause was determined to be a lack of fusion during initial construction fabrication that resulted in a weld defect. The weld defect allowed initiation of PWSCC. The indication was repaired and the reactor vessel head returned to service. Future inspections will occur in accordance with NRC Order EA-03-009. This condition had minimal safety consequences. The CRDM penetration indications were identified prior to any through-wall leakage. This condition is reportable to the NRC in accordance with 10 CFR 50.73 (a)(2)(ii), as a condition that resulted in a principal safety barrier being seriously degraded.
Page 2 of 3NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (7-2001)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
DOCKET NUMBER (2)
LER NUMBER (6 PAGE (3)
YEAR SEQUENTIAL REVISION NUMBER NUMBER Byron Station Unit 2 05000455 2007 001 00 age 2 o 3 (If more space is required, use additional copies of NRC Form 366A)(17)
A. Plant Condition Prior to Event:
Event Date: April 9, 2007 Unit 2 - Mode 6 Refueling Reactor Coolant System (AB) was at ambient temperature and depressurized.
B. Description of Event
In accordance with the requirements of NRC order EA-03-009, Issuance Of Order Establishing Interim Inspection Requirements For Reactor Pressure Vessel Heads At Pressurized Water Reactors, Byrons Unit 2 reactor head was ranked in the low Primary Water Stress Corrosion Cracking (PWSCC) susceptibility category. This low ranking required the initial Control Rod Drive Mechanism (CRDM) penetrations nozzle volumetric examinations to be completed in the recent spring 2007 refuel outage (B2R13).
During the B2R13 volumetric examination of the Unit 2 CRDM penetration 68, an ultrasonic test (UT) indication was discovered from 13.5 degrees to 22.5 degrees. The 0-degrees reference is located on the downhill side of the penetration where the nozzle extension is shortest. The indication is located 1.95 inches from the end of the penetration to 2.47 inches towards the J-Groove weld. The bottom toe of the J-Groove weld is located 1.99 inches from the bottom of the tube and the top of the J-Groove weld is located 3.71 inches from the bottom of the penetration. The indication was not surface connected below the J-groove and/or fillet weld.
Following the discovery of the UT indication, a liquid penetrant test (PT) examination was performed to determine if the indication was open to the surface. The PT examination revealed two surface indications: one rounded indication 0.05 inches in diameter and one linear axial indication 0.15 inches long. The indications aligned with the general location and orientation of the UT indication.
Since the indication was consistent with PWSCC, a bare-metal visual (BMV) examination was subsequently completed on 100 percent of the reactor head penetrations, including 360 degrees around each penetration in accordance with NRC order EA-03-009. No evidence of boric acid was identified during the BMV.
Prior to the weld repair, a boat sample from CRDM penetration 68 was obtained for metallurgical evaluation.
This condition is reportable to the NRC in accordance with 10 CFR 50.73 (a)(2)(ii), as a condition that resulted in a principal safety barrier being seriously degraded.
Page 3 of 3NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (7-2001)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
Byron Station Unit 2 DOCKET NUMBER (2)
LER NUMBER (61 I
PAGE (3)
YEAR I
SEQUENTIAL I
REVISION I
NUMBER 05000455 2007 001 00 Page 3 of 3 (If more space is required, use additional copies of NRC Form 366A)(17)
C. Cause of the Event
A boat sample was removed from the penetration to determine the cause of the indications. A PT exam was performed on the boat sample excavation site, and a subsurface linear defect which intersected the axial indication was revealed. A portion of the subsurface defect was captured in the boat sample and was identified as a lack of fusion between the next-to-last weld pass and the tube surface. Additionally, hot cracking and lack of fusion between weld passes were observed in the boat sample. The metallurgical evaluations indicated the axial indication was the result of PWSCC which initiated at a subsurface location and propagated toward the wetted surface of the weld. The initiation of PWSCC is attributed to wetting of the subsurface lack-of-fusion defects which created an atypical crevice corrosion environment in the high stress J-groove weld region.
D. Safety Analysis
This condition had minimal actual safety consequences. The CRDM penetration 68 indications were identified and repaired prior to any through wall leakage occurring.
The CRDM penetration nozzle weld indications were identified as part of a required inspection to look for flaws of this type. Left undetected, the flaw could have propagated through the alloy 600 weld material, and potentially the penetration material, to form a leak path through the RCS pressure boundary. The consequences of this failure would have been high. Although the leak would have been very small, industry experience has shown that boric acid wastage of the carbon steel reactor head is possible.
The flaw was detected as a small indication and no through wall leakage was noted. Verification of no through wall leakage was accomplished through a bare metal head inspection near penetration 68 as well as the remainder of the reactor head penetrations.
Byron Unit 2 will undergo more frequent inspections consistent with the requirements of NRC order EA-03-009.
E. Corrective Actions
The CRDM penetration nozzle number 68 was repaired prior to return to service of the Unit 2 reactor head.
Consistent with NRC Order EA-03-009, Byron Unit 2 reactor vessel head has been reclassified into the high susceptibility category. Consequently, Byron Unit 2 will comply with the Orders requirements for inspections for a high susceptibility reactor vessel head.
F. Previous Occurrences
There have been no previous occurrences of this nature at Byron.(7-2001)