05000456/LER-2013-002, Re Liquid Penetrant Indications in Embedded Flaw Repair of Control Rod Drive Mechanism Penetration 69 During Refueling Outage

Re Liquid Penetrant Indications in Embedded Flaw Repair of Control Rod Drive Mechanism Penetration 69 During Refueling Outage

ML13318A096
Person / Time
Site:	Braidwood
Issue date:	11/13/2013
From:	Kanavos M Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
BW130101 LER 13-002-00
Download: ML13318A096 (5)

LER-2013-002, Re Liquid Penetrant Indications in Embedded Flaw Repair of Control Rod Drive Mechanism Penetration 69 During Refueling Outage

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), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
4562013002R00 - NRC Website

text

November 13, 2013 BW130101 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Braidwood Station, Unit 1 Facility Operating License No. NPF-72 NRC Docket No. STN 50-456 10 CFR 50.73

Subject:

Licensee Event Report 2013-002 Indications in Control Rod Drive Mechanism Nozzle Weld due to Imbedded Flaws Opening Up from Thermal and Pressure Stresses During Operation The enclosed Licensee Event Report (LER) is being submitted in accordance with 10 CFR 50.73, "Licensee Event Report System."

There are no regulatory commitments contained in this letter. Should you have any questions concerning this submittal, please contact Mr. Jamison Rappeport, Regulatory Assurance Manager, at (815) 417-2800.

pL--

Mark E. Kanavos Site Vice President Braidwood Station

Enclosure:

LER 2013-002-00 cc: NRC Project Manager, NRR - Braidwood Station Illinois Emergency Management Agency - Division of Nuclear Safety US NRC Regional Administrator, Region III US NRC Senior Resident Inspector (Braidwood Station)

Illinois Emergency Management Agency Braidwood Representative

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 10/31/2013 (10-2010)

LICENSEE EVENT REPORT (LER)

, the NRC may not conduct or sponsor, and a person is not required to respond to, the jnfnrrn"tion

,Il~ti~

13. PAGE Braidwood Station, Unit 1 05000456 1 of 4

4. TITLE Liquid Penetrant Indications in Embedded Flaw Repair of Control Rod Drive Mechanism Penetration 69 During Refueling Outage

5. EVENT DATE 6, LER NUMBER

7. REPORT DATE

8. OTHER FACILITIES INVOLVED FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR I SEQUENTIAL I REV MONTH DAY YEAR N/A N/A NUMBER NO, FACILITY NAME DOCKET NUMBER 09 14 2013 2013 - 002 -

00 11 13 2013 N/A N/A

9. OPERATING MODE

11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check al/ that apply) 6 o 20.2201(b) o 20.2203(a)(3)(i) o 50.73(a)(2)(i)(C) o 50.73(a)(2)(vii) o 20.2201(d) o 20.2203(a)(3)(ii)

IZ1 50.73(a)(2)(ii)(A) o 50.73(a)(2)(viii)(A) o 20.2203(a)(1) o 20.2203(a)(4) o 50.73(a)(2)(ii)(B) o 50.73(a)(2)(viii)(B) o 20.2203(a)(2)(i) o 50.36(c)(1 )(i)(A) o 50.73(a)(2)(iii) o 50.73(a)(2)(ix)(A) o 20.2203(a)(2)(ii) o 50.36(c)(1 )(ii)(A) o 50.73(a)(2)(iv)(A) o 50.73(a)(2)(x)

10. POWER LEVEL o 20.2203(a)(2)(iii) o 50,36(c)(2) o 50.73(a)(2)(v)(A) o 73,71(a)(4) 000 o 20.2203(a)(2)(iv) o 50,46(a)(3)(ii) o 50,73(a)(2)(v)(B) o 73,71(a)(5) o 20,2203(a)(2)(v) o 50,73(a)(2)(i)(A) o 50.73(a)(2)(v)(C) o OTHER o 20,2203(a)(2)(vi) o 50.73(a)(2)(i)(B) o 50,73(a)(2)(v)(D)

Specify in Abstract below or in NRC Form 366A

12. LICENSEE CONTACT FOR THIS LER FACILITY NAME rELEPHONE NUMBER (Include Area Code)

Jamison Rappeport, Regulatory Assurance Manager (815) 417-2800 CAUSE SYSTEM COMPONENT MANU-TOE~~_

CAUSE

SYSTEM COMPONENT MANU-REPORTABLE FACTURER FACTURER TO EPIX N/A N/A I

N/A N/A N/A c~c N/A N/A N/A N/A N/A

14. SUPPLEMENTAL REPORT EXPECTED

15. EXPECTED MONTH I DAY I YEAR o YES (If yes, complete 15, EXPECTED SUBMISSION DATE)

IZ1 NO SUBMISSION N/A N/A I N/A DATE ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)

On September 14, 2013 at 0100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />, indications exceeding ASME Section III acceptance criteria were discovered in the embedded flaw repair weld of Control Rod Drive Mechanism (CRDM) Penetration 69. A total of 22 recordable indications located within the embedded flaw repair weld were documented, with 13 rounded indications exceeding the acceptance criteria (ASME Section III 1971 Edition through the Summer 1973 Addenda) of dimensions greater than 3/16". The indications in Penetration 69 were removed/reduced to acceptable size or reduced and weld repaired to restore the embedded flaw repair as approved by the NRC in Braidwood Relief Request 13R-09. This LER is being submitted in follow-up to ENS 49343 that was made at 0447 hours0.00517 days <br />0.124 hours <br />7.390873e-4 weeks <br />1.700835e-4 months <br /> on September 14, 2013, Based on industry experience, the cause of this event was determined to be mechanical discontinuities/minor subsurface voids opening up to the weld surface due to thermal and/or pressure stresses during plant operation.

Corrective actions to prevent recurrence include consideration of a long term Alloy 600 mitigation plan and other potential alternatives for the remaining acceptable but unrepaired penetrations on the Braidwood Units 1 and 2 RPV closure heads. This corrective action plan is under development and will also evaluate peening of the remaining acceptable but unrepaired CRDM penetrations.

NRC FORM 366 (10-2010) (10-2010)

U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)

1. FACIUTY NAME

2. DOCKET

6. LER NUMBER

3. PAGE YEAR Braidwood Station, Unit 1 05000456 2

OF 2013 002 00

A. Plant Operating Conditions Before the Event

Event Date:

September 14, 2013 Unit: 1 MODE: 6 Reactor Power: 000 percent Unit 1 Reactor Coolant System [AB]:

Shutdown for refueling outage, no fuel movement in progress No structures, systems or components were inoperable at the start of this event that contributed to the event.

B. Description of Event

On September 14, 2013 at 0100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />, indications exceeding ASME Section III acceptance criteria were discovered in the embedded flaw repair weld of Control Rod Drive Mechanism (CRDM) Penetration 69. The Ai Ri7 examination of the embedded flaw repair in Penetration 69 was performed in accordance with Braidwood Third Interval Relief Request 13R-09 which requires liquid penetrant (PT) examination of embedded flaw weld repairs every refuel outage.

4 This was the first inservice examination of the repair weld since it was applied in A1R16 (April 2012). A total of 22 recordable indications located within the embedded flaw repair weld were documented, with 13 exceeding the acceptance criteria of 3/16". All of the documented indications were characterized as rounded. Per the original Construction Code (ASME Section III 1971 Edition through the Summer 1973 Addenda), unacceptable indications include "Rounded indications with dimensions greater than 3/16 inch."

In addition to the PT examination of the embedded flaw weld repair on Penetration 69, all penetrations were examined by ultrasonic and eddy current methods using procedures and personnel qualified in accordance with the EPRI Performance Demonstration Program. The EPRI program is implemented by 10 CFR 50.55a, "Codes and standards", which includes the use of ASME Section XI Code Case N-729-1, "Alternative Examination Requirements for PWR Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial-Penetration WeldsSection XI, Division 1". No indications of Primary Water Stress Corrosion Cracking (PWSCC) or through wall leakage were observed on any of the remaining penetrations. A bare metal visual inspection of the exterior surfaces of the reactor head and penetrations was also performed during A 1 R17 in accordance with ASME Section XI Code Case N-729-1. There was no indication of through wall leakage observed during the bare metal visual examination.

A 1 R17 was the first outage that an inservice PT examination was performed on the embedded flaw of Braidwood Unit 1 RPV Head Penetration 69. As soon as the unacceptable PT indications in the embedded flaw repair weld at Penetration 69 were identified, contingency plans to remove or reduce these indications were implemented. Once it was determined that additional welding would be required to restore the embedded flaw repair back to the original approved design criteria, repair crews were mobilized and completed the weld repairs on Penetration 69. No other CRDM penetration repairs were required in A 1 Ri?

This event is reportable in accordance with 10 CFR 50.73(a)(2)(ii)(A), any event or condition that resulted in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded.

• NRC FORM 366A (10-2010) (10-2010)

1. FACILITY NAME Braidwood Station, Unit 1

C. Cause of Event

LICENSEE EVENT REPORT (LER)

2. DOCKET YEAR 05000456 2013 U.S. NUCLEAR REGULATORY COMMISSION 002 REV NO.

00

3. PAGE 3

OF 4

Based on industry experience, the cause of these flaws is attributed to existing mechanical discontinuities/minor subsurface voids opening up to the weld surface due to thermal and/or pressure stresses during plant operation.

These flaws were not detectable during the original weld examination using the final acceptance non-destructive examination (NDE) methods (Le., visual and PT examinations). These examinations only detect indications open to the surface. Indications in initial embedded flaw repair welds may not be revealed until after the embedded flaw weld repairs are subjected to plant operating conditions.

Prior to the outage, contingency repair plans were established and were effective in responding to the emergent penetration repairs; however, currently there is no way to effectively predict the likelihood of indications opening to the surface resulting in the need to perform flaw reduction or subsequent weld repairs. Long term corrective actions are under development to address this issue.

D. Safety Consequences

This condition had no actual safety consequences impacting plant or public safety.

The flaw was identified in a timely manner and repaired prior to through wall leakage occurring. The flaw was identified as part of a required periodic inspection. Potentially, if the flaw remained undetected, it could have over time, propagated through the Alloy 600 weld material to form a leak path through the reactor coolant pressure boundary.

Based on the A 1 R 17 documented characteristics and dimensions of the observed PT indications, there was no Safety Significant functional failure (Le., loss of safety function) as a result of these indications. The primary coolant pressure boundary was maintained and capable of preventing the release of radioactive material.

The Rod Drive system remained functional.

E. Corrective Actions

Corrective actions to prevent recurrence:

Develop and implement a corrective action plan that includes consideration of a long term Alloy 600 mitigation and other acceptable alternatives for the remaining acceptable but unrepaired penetrations on the Braidwood Units 1 and 2 RPV closure heads. This corrective action plan will also evaluate peening of the remaining acceptable but unrepaired CRDM penetrations.

Weld repairs of CRDM penetration 69 were completed.

F. Previous Occurrences

Similar inservice indications during PT examinations of embedded flaw repairs have also been encountered at San Onofre Nuclear Generating Station (SONGS) Unit 3 and Beaver Valley Unit 2. The indication observed at SONGS Unit 3 was the first indication associated with an Inconel 52 embedded and raised concerns (10-2010) (10-2010)

1. FACILITY NAME Braidwood Station, Unit 1 U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)

2. DOCKET YEAR 05000456 2013

6. LER NUMBER I

S!:UU!:N IIAL I ~;v NUMBER 002 00

3. PAGE 4

OF 4

The analysis of the boat sample concluded that the apparent cause of the rejectable PT indication was due to original weld fabrication flaws and not associated with any service induced degradation method (e.g., PWSCC).

The final analysis showed a very thin layer of Alloy 52 had opened up allowing the dye penetrant to enter the void beneath it and was found during the PT.

A previous Licensee Event Report was made in April 2012 at Braidwood Station for indications on CRDM penetration 69 (LER 2012-002-00).

G. Component Failure Data

Manufacturer N/A I NRC l (1,

10)

Nomenclature N/A Model N/A Mfg. Part Number N/A