05000321/LER-2016-003, Regarding Reactor Coolant System Piping Has Unacceptable Weld Indication Discovered During Refueling Outage

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Regarding Reactor Coolant System Piping Has Unacceptable Weld Indication Discovered During Refueling Outage
ML16105A219
Person / Time
Site: Hatch 
Issue date: 04/14/2016
From: Vineyard D
Southern Co, Southern Nuclear Operating Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-16-0536 LER 16-003-00
Download: ML16105A219 (6)
v • d • e


LER-2016-003, Regarding Reactor Coolant System Piping Has Unacceptable Weld Indication Discovered During Refueling Outage
Event date:
Report date:
Reporting criterion: 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded
10 CFR 50.73(a)(2)(i)
10 CFR 50.73(a)(2)
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
3212016003R00 - NRC Website
v • d • e

text

David A. Vineyard Vrce President* Hatch April 14,2016 Docket Nos.: 50-321 Southern Nuclear Operating Company, Inc.

Edwm I Hatch Nuclear Plant 11 028 Hatch Parkway North Baxley. Georgra 31513 Tel 912 537 5859 Fax 912 366 2077 U. S. Nuclear Regulatory Commission ATIN: Document Control Desk Washington, D. C. 20555-0001 Edwin I. Hatch Nuclear Plant Licensee Event Report 2016-003-00 SOUTHERN '*

NUCLEAR A SOUTHERN COMPANY NL-16-0536 Reactor Coolant System Piping Has Unacceptable Weld Indication Discovered During Refueling Outage Ladies and Gentlemen:

In accordance with the requirements 10 CFR 50.73(a)(2)(ii)(A), Southern Nuclear Operating Company hereby submits the enclosed Licensee Event Report.

This letter contains no NRC commitments. If you have any questions, please contact Greg Johnson at (912) 537-5874.

Respectfully submitted, D. A. Vineyard Vice President - Hatch DRV/jcb Enclosures: LEA 2016-003-00

U.S. Nuclear Regulatory Commission NL-16-0536 Page2 cc:

Southern Nuclear Operating Company Mr. S. E. Kuczynski, Chairman, President & CEO Mr. D. G. Bost, Executive Vice President & Chief Nuclear Officer Mr. D. R. Vineyard, Vice President - Hatch Mr. M. D. Meier, Vice President-Regulatory Affairs Mr. D. R. Madison, Vice President-Fleet Operations Mr. B. J. Adams, Vice President-Engineering Mr. G. L. Johnson, Regulatory Affairs Manager-Hatch RTYPE: CHA02.004 U. S. Nuclear Regulatory Commission Ms. C. Haney, Regional Administrator Mr. M. D. Orenak, NRR Project Manager-Hatch Mr. D. H. Hardage, Senior Resident Inspector-Hatch

Edwin I. Hatch Nuclear Plant Unit 1 Licensee Event Report 2016-003-00 Reactor Coolant System Piping Has Unacceptable Weld Indication Discovered During Refueling Outage

X NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMS: NO. 3150..(]104 EXPIRES: 01131/2017 (02*2014)

Est;mated burden per response to comply with th1s mandatory collgct1on request eo hours.

i:

~~...

Reported lessons learned are incorporated Into the licensing process and fed back to industty

-~~

Send commen:s regarding burden e;timate to the FOIA. Privac-1 and lnfurmation Coll<;ehons

  • ~~~!/ LICENSEE EVENT REPORT (LER)

Branch (T*S F53). US Nuclear Regulatory CommiSSIOn, Washrngton, DC 20555*000t. or by internet e-mail to lnfocol!ecls.AesciJice@nrc.gov, and to the Desk Officer OHice of Information and Regulatory Aliairs NEOEl-10202, (3150.010') Office ol Management and 8L'<lge1, Washington, DC

~0503. If a means used to impose an mlorma110n colled1cn does not display a currenLy valrd OMS ccn1rcl number, the NRC may not conduct cr sponsor, and a eerson IS not required to respond to, the information colleclion.

3. PAGE Edwin I. Hatch Nuclear Plant Unit 1 05000 321 1 OF 3
4. TITLE Reactor Coolant System Piping Has Unacceptable Weld Indication Discovered During Refueling Outage
5. EVENT DATE
6. LER NUMBER
7. REPORT DATE B. OTHER FACILITIES INVOLVED YEA-I SEQUENTIAL _I REV FACILITY NAME DOCKET NUMBER MONTH DAY YE:AR NUMBER NO MONTH DAY YEAR 02 16 2016 2016 - 003 - 00 4

14 FACILITY NAI.l£ DOCKET NU:.:a:R 2016

9. OPERATING MODE
11. THIS REPORT IS SUBMITIED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)

D 2o.22o1 (bl D 20.2203(a)(3)(i)

D 50.73(a)(2)(i)(C)

D 50.73(a)(2)(vli) 5 D 2o.22o1 (dl D 20.2203(a)(3)(ii) 181 50.73(a)(2)(ii)(A)

D 50.73(a)(2)(viii)(A)

D 20.2203(a)(1) 0 20.2203(a)(4)

D 50.73(a)(2)(ii)(B)

D 50.73(a)(2)(viii)(B)

D 20.2203(a)(2)(i)

D 50.36(c)(1 )(i)(A)

D 50.73(a)(2)(iii)

D 50.73(a)(2)(ix)(A)

10. POWER LEVEL D 20.2203(a)(2)(ii)

D 50.36(c)(1 )(ii)(A)

D 50.73(a)(2)(iv)(A)

D 50.73(a)(2)(x)

D 20.2203(a)(2)(iii)

D 50.36(c)(2)

D 50.73(a)(2)(v)(A)

D 73.71(a)(4)

D 20.2203(a)(2)(iv)

D 50.46(a)(3)(ii)

D 50.73(a)(2)(v)(B)

D 73.71(a)(5) 0 D 20.2203(a)(2)(v)

D 50.73(a)(2)(i)(A)

D 50.73(a)(2)(v)(C)

D OTHER D 20.2203(a)(2)(vi)

D 50.73(a)(2)(i)(B)

D 50.73(a)(2)(v)(D)

Spec1fy '" Absltact below or 1n NRC Fcrm 36SA

12. LICENSEE CONTACT FOR THIS LER UCENSEE CONTACT

_rELEPHONE NUMBER (lncludo Area Code)

Edwin I. Hatch I Carl James Collins - Licensing Supervisor 912-537-5900 ext 2342 CAUSE SYSTEM COMPONENT MANU-REPORTABLE

CAUSE

SYSTEM COMPONENT MANU*

REPORTABLE:

FACTURER TO EFIX FACTURER TO EPIX E

AD PSP y

14. SUPPLEMENTAL REPORT EXPECTED
15. EXPECTED MONTH DAY YEAR DYES {If yes, complete 15. EXPECTED SUBMISSION DATE) 181 NO SUBMISSION DATE ABSTRACT (Umit to 1400 spaces, i.e.. approximately 15 smgle*speced typewritten lines)

During the 2016 Unit 1 R27 refueling outage, plans were put in place to upgrade the 1831-1 RC-12BR-E-5 (1 B31-E5) design weld overlay (WOL) to a full structural weld overlay (FSWOL) in order to allow for code qualified examinations. On February 16, 2016 at 0631 EST, during surface preparation work, axial indications were found on the WOL. Evaluation of he indications found in the weld overlay suggests that the non-satisfactory PT examination was a result of the propagation of the original flaw that was found on the 1 E Recirculation Loop Piping. The original indication had propagated into the lnco,lel Alloy 82 WOL material installed in 1988. It was determined that the as-found condition of the flaw did not meet ASME Section XI acceptance criteria. The indications were removed from the WOL and 1 B31-E5 was upgraded to a full structural weld overlay using intergranular stress corrosion cracking (IGCSS) resistant Alloy 52 weld material.

NRC FORM 265 (02*2014)

NRC FORM 388A (1-2001)

PLANT AND SYSTEM IDENTIFICATION

General Electric-Boiling Water Reactor (BWR)

Energy Industry Identification System codes appear in the text as (EllS Code XX)

DESCRIPTION OF EVENT

On February 16,2016 at 0631 EST, with Unit 1 at 0 percent rated thermal power due to a scheduled refueling outage, it was discovered that an axial flaw found on the recirculation Inlet nozzle 1 831-1 RC-12BR-E-5 (1 B31-E5) to safe-end weldment had propagated Into the lnconel Alloy 82 Weld Overlay (WOL) material installed In 1988. Evaluation of the as-found condition of the flaw did not meet ASME Section XI acceptance criteria.

As part of normal pre-outage scope activities, plans were put in place to upgrade the 1 831-ES partial WOL to a FSWOL in order to allow for code qualified examinations. During surface preparation work, axial indications were found on the partial WOL. It was then discovered that the original flaw had propagated into the lnconel Alloy 82 WOL material installed in 1988.

The flaw was repaired in accordance with an approved code alternative and 1 B31-E5 was upgraded to a full structural weld overlay (FSWOL) using intergranular stress corrosion cracking resistant Alloy 52 weld material.

CAUSE OF EVENT

The unacceptable as-found condition of the defect found in the WOL for 1 831-ES was due to intergranular stress corrosion cracking (IGSCC). The 304 stainless steel piping and lnconel Alloy 1 82 weld materials are susceptible to this failure mode in a BWR environment.

The station also lost track of the different design attributes of 1 B31-E5 weld repair. This led to mis-characterization of the weld as a FSWOL and extending the exam frequency, thus causing the station to not adequately monitor the growth of the flaw. Reliefs that reduced the frequency of performance of lSI code exams were approved by the NRC. Approval was based upon the ASME Class 1 WOL's meeting the requirements of being a full structural WOL type. Additionally, during the implementation of BWRVIP*075, a risk classification was required to prioritize the welds to be examined. However, the assumption that all Unit 1 WOL repairs were full structural welds affected the priority placed upon this weld. The flaw was repaired in accordance with an approved code alternative and 1831-ES was upgraded to a full structural weld overlay (FSWOL) using intergranular stress corrosion cracking resistant Alloy 52 weld material.

NRC FORM 388A (02*2014)

NRC FORM 38eA (1-2001)

NRC FOAM 366A (02-2014)

LICENSEE EVENT REPORT (LER)

CONTINUATION SHEET U.S. NUCLEAR REGULATORY COMMISSION

1. FACILITY NAME
2. DOCKET Edwin I. Hatch Nuclear Plant Unit 1 05000 321 NARRATIVE REPORT ABILITY AND SAFETY ASSESSMENT
6. LER NUMBER YEAR I SEQUENTIAL l NUMBER 2016-003- 00 REV.

NO.

3. PAGE 30F3 This event is reportable per 10 CFR 50.73(a)(2)(ii)(A) due to a defect in the primary coolant system that could not be found acceptable under ASME Section XI. Upon performance of the subsequent liquid penetrant testing (PT) examination, it was discovered that the as-found condition of the flaw did not meet ASME Section XI acceptance criteria.

Evaluation of the flaw found in the weld overlay suggests that the non-satisfactory PT examination is a result of the propagation of the original flaw that was found on the 1 E Recirculation Loop Piping.

Though the weld flaw exceeded the acceptance criteria of ASME Section XI, no leakage from this flaw was identified at any time during operation or shutdown. There is reasonable assurance that there was not a breach in the credited RCS boundary due to the axial flaw not having grown through the weld overlay. Additionally, engineering evaluation of the structural integrity of the weld shows that the flawed component had adequate margin for all design basis events. The evaluation has shown that the axial flaw identified in Weld 1 B31-1 RC-12BR-E-5 located on the recirculation inlet system, meets the ASME Code,Section XI structural margin, considering a Service Level D structural factor in the evaluation, as required by the NRC Inspection Manual. Even though the flaw has a depth of 100% of original wall thickness, which exceeds the ASME Code,Section XI allowable flaw depth of 75% of wall thickness, the safety of the reactor pressure boundary was not compromised. It is, therefore, concluded that the flawed component had adequate margin for all Design Basis Loading Events when the flaw was identified and this condition had a very low safety significance.

CORRECTIVE ACTIONS

As part of corrective actions, a weld repair on 1B31-1 RC-12BR*E-5 was completed to restore piping back to original code. The design type weld overlay was upgraded to a full structural weld overlay using lnconel Alloy 52 weld material in accordance with an NRC approved alternative (HNP-ISI-ALT-15-01).

As part of an extent of condition review, a similar weld repair was performed on 1 B31-1 RC-12BR-C-5 to upgrade its weld from a design type weld overlay to a full structural weld overlay. Also, Unit 1 and 2 weld overlays containing Alloy 82 weld material will be re-examined in the upcoming refueling outages.

ADDITIONAL INFORMATION

Other Systems Affected: No systems other than those mentioned in this report were affected by this event.

Failed Components Information

None.

Commitment Information: This report does not created any new licensing commitments.

Previous Similar Events

None.

NRC FORM 3eSA (02*2014)

David A. Vineyard Vrce President* Hatch April 14,2016 Docket Nos.: 50-321 Southern Nuclear Operating Company, Inc.

Edwm I Hatch Nuclear Plant 11 028 Hatch Parkway North Baxley. Georgra 31513 Tel 912 537 5859 Fax 912 366 2077 U. S. Nuclear Regulatory Commission ATIN: Document Control Desk Washington, D. C. 20555-0001 Edwin I. Hatch Nuclear Plant Licensee Event Report 2016-003-00 SOUTHERN '*

NUCLEAR A SOUTHERN COMPANY NL-16-0536 Reactor Coolant System Piping Has Unacceptable Weld Indication Discovered During Refueling Outage Ladies and Gentlemen:

In accordance with the requirements 10 CFR 50.73(a)(2)(ii)(A), Southern Nuclear Operating Company hereby submits the enclosed Licensee Event Report.

This letter contains no NRC commitments. If you have any questions, please contact Greg Johnson at (912) 537-5874.

Respectfully submitted, D. A. Vineyard Vice President - Hatch DRV/jcb Enclosures: LEA 2016-003-00

U.S. Nuclear Regulatory Commission NL-16-0536 Page2 cc:

Southern Nuclear Operating Company Mr. S. E. Kuczynski, Chairman, President & CEO Mr. D. G. Bost, Executive Vice President & Chief Nuclear Officer Mr. D. R. Vineyard, Vice President - Hatch Mr. M. D. Meier, Vice President-Regulatory Affairs Mr. D. R. Madison, Vice President-Fleet Operations Mr. B. J. Adams, Vice President-Engineering Mr. G. L. Johnson, Regulatory Affairs Manager-Hatch RTYPE: CHA02.004 U. S. Nuclear Regulatory Commission Ms. C. Haney, Regional Administrator Mr. M. D. Orenak, NRR Project Manager-Hatch Mr. D. H. Hardage, Senior Resident Inspector-Hatch

Edwin I. Hatch Nuclear Plant Unit 1 Licensee Event Report 2016-003-00 Reactor Coolant System Piping Has Unacceptable Weld Indication Discovered During Refueling Outage

X NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMS: NO. 3150..(]104 EXPIRES: 01131/2017 (02*2014)

Est;mated burden per response to comply with th1s mandatory collgct1on request eo hours.

i:

~~...

Reported lessons learned are incorporated Into the licensing process and fed back to industty

-~~

Send commen:s regarding burden e;timate to the FOIA. Privac-1 and lnfurmation Coll<;ehons

  • ~~~!/ LICENSEE EVENT REPORT (LER)

Branch (T*S F53). US Nuclear Regulatory CommiSSIOn, Washrngton, DC 20555*000t. or by internet e-mail to lnfocol!ecls.AesciJice@nrc.gov, and to the Desk Officer OHice of Information and Regulatory Aliairs NEOEl-10202, (3150.010') Office ol Management and 8L'<lge1, Washington, DC

~0503. If a means used to impose an mlorma110n colled1cn does not display a currenLy valrd OMS ccn1rcl number, the NRC may not conduct cr sponsor, and a eerson IS not required to respond to, the information colleclion.

3. PAGE Edwin I. Hatch Nuclear Plant Unit 1 05000 321 1 OF 3
4. TITLE Reactor Coolant System Piping Has Unacceptable Weld Indication Discovered During Refueling Outage
5. EVENT DATE
6. LER NUMBER
7. REPORT DATE B. OTHER FACILITIES INVOLVED YEA-I SEQUENTIAL _I REV FACILITY NAME DOCKET NUMBER MONTH DAY YE:AR NUMBER NO MONTH DAY YEAR 02 16 2016 2016 - 003 - 00 4

14 FACILITY NAI.l£ DOCKET NU:.:a:R 2016

9. OPERATING MODE
11. THIS REPORT IS SUBMITIED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)

D 2o.22o1 (bl D 20.2203(a)(3)(i)

D 50.73(a)(2)(i)(C)

D 50.73(a)(2)(vli) 5 D 2o.22o1 (dl D 20.2203(a)(3)(ii) 181 50.73(a)(2)(ii)(A)

D 50.73(a)(2)(viii)(A)

D 20.2203(a)(1) 0 20.2203(a)(4)

D 50.73(a)(2)(ii)(B)

D 50.73(a)(2)(viii)(B)

D 20.2203(a)(2)(i)

D 50.36(c)(1 )(i)(A)

D 50.73(a)(2)(iii)

D 50.73(a)(2)(ix)(A)

10. POWER LEVEL D 20.2203(a)(2)(ii)

D 50.36(c)(1 )(ii)(A)

D 50.73(a)(2)(iv)(A)

D 50.73(a)(2)(x)

D 20.2203(a)(2)(iii)

D 50.36(c)(2)

D 50.73(a)(2)(v)(A)

D 73.71(a)(4)

D 20.2203(a)(2)(iv)

D 50.46(a)(3)(ii)

D 50.73(a)(2)(v)(B)

D 73.71(a)(5) 0 D 20.2203(a)(2)(v)

D 50.73(a)(2)(i)(A)

D 50.73(a)(2)(v)(C)

D OTHER D 20.2203(a)(2)(vi)

D 50.73(a)(2)(i)(B)

D 50.73(a)(2)(v)(D)

Spec1fy '" Absltact below or 1n NRC Fcrm 36SA

12. LICENSEE CONTACT FOR THIS LER UCENSEE CONTACT

_rELEPHONE NUMBER (lncludo Area Code)

Edwin I. Hatch I Carl James Collins - Licensing Supervisor 912-537-5900 ext 2342 CAUSE SYSTEM COMPONENT MANU-REPORTABLE

CAUSE

SYSTEM COMPONENT MANU*

REPORTABLE:

FACTURER TO EFIX FACTURER TO EPIX E

AD PSP y

14. SUPPLEMENTAL REPORT EXPECTED
15. EXPECTED MONTH DAY YEAR DYES {If yes, complete 15. EXPECTED SUBMISSION DATE) 181 NO SUBMISSION DATE ABSTRACT (Umit to 1400 spaces, i.e.. approximately 15 smgle*speced typewritten lines)

During the 2016 Unit 1 R27 refueling outage, plans were put in place to upgrade the 1831-1 RC-12BR-E-5 (1 B31-E5) design weld overlay (WOL) to a full structural weld overlay (FSWOL) in order to allow for code qualified examinations. On February 16, 2016 at 0631 EST, during surface preparation work, axial indications were found on the WOL. Evaluation of he indications found in the weld overlay suggests that the non-satisfactory PT examination was a result of the propagation of the original flaw that was found on the 1 E Recirculation Loop Piping. The original indication had propagated into the lnco,lel Alloy 82 WOL material installed in 1988. It was determined that the as-found condition of the flaw did not meet ASME Section XI acceptance criteria. The indications were removed from the WOL and 1 B31-E5 was upgraded to a full structural weld overlay using intergranular stress corrosion cracking (IGCSS) resistant Alloy 52 weld material.

NRC FORM 265 (02*2014)

NRC FORM 388A (1-2001)

PLANT AND SYSTEM IDENTIFICATION

General Electric-Boiling Water Reactor (BWR)

Energy Industry Identification System codes appear in the text as (EllS Code XX)

DESCRIPTION OF EVENT

On February 16,2016 at 0631 EST, with Unit 1 at 0 percent rated thermal power due to a scheduled refueling outage, it was discovered that an axial flaw found on the recirculation Inlet nozzle 1 831-1 RC-12BR-E-5 (1 B31-E5) to safe-end weldment had propagated Into the lnconel Alloy 82 Weld Overlay (WOL) material installed In 1988. Evaluation of the as-found condition of the flaw did not meet ASME Section XI acceptance criteria.

As part of normal pre-outage scope activities, plans were put in place to upgrade the 1 831-ES partial WOL to a FSWOL in order to allow for code qualified examinations. During surface preparation work, axial indications were found on the partial WOL. It was then discovered that the original flaw had propagated into the lnconel Alloy 82 WOL material installed in 1988.

The flaw was repaired in accordance with an approved code alternative and 1 B31-E5 was upgraded to a full structural weld overlay (FSWOL) using intergranular stress corrosion cracking resistant Alloy 52 weld material.

CAUSE OF EVENT

The unacceptable as-found condition of the defect found in the WOL for 1 831-ES was due to intergranular stress corrosion cracking (IGSCC). The 304 stainless steel piping and lnconel Alloy 1 82 weld materials are susceptible to this failure mode in a BWR environment.

The station also lost track of the different design attributes of 1 B31-E5 weld repair. This led to mis-characterization of the weld as a FSWOL and extending the exam frequency, thus causing the station to not adequately monitor the growth of the flaw. Reliefs that reduced the frequency of performance of lSI code exams were approved by the NRC. Approval was based upon the ASME Class 1 WOL's meeting the requirements of being a full structural WOL type. Additionally, during the implementation of BWRVIP*075, a risk classification was required to prioritize the welds to be examined. However, the assumption that all Unit 1 WOL repairs were full structural welds affected the priority placed upon this weld. The flaw was repaired in accordance with an approved code alternative and 1831-ES was upgraded to a full structural weld overlay (FSWOL) using intergranular stress corrosion cracking resistant Alloy 52 weld material.

NRC FORM 388A (02*2014)

NRC FORM 38eA (1-2001)

NRC FOAM 366A (02-2014)

LICENSEE EVENT REPORT (LER)

CONTINUATION SHEET U.S. NUCLEAR REGULATORY COMMISSION

1. FACILITY NAME
2. DOCKET Edwin I. Hatch Nuclear Plant Unit 1 05000 321 NARRATIVE REPORT ABILITY AND SAFETY ASSESSMENT
6. LER NUMBER YEAR I SEQUENTIAL l NUMBER 2016-003- 00 REV.

NO.

3. PAGE 30F3 This event is reportable per 10 CFR 50.73(a)(2)(ii)(A) due to a defect in the primary coolant system that could not be found acceptable under ASME Section XI. Upon performance of the subsequent liquid penetrant testing (PT) examination, it was discovered that the as-found condition of the flaw did not meet ASME Section XI acceptance criteria.

Evaluation of the flaw found in the weld overlay suggests that the non-satisfactory PT examination is a result of the propagation of the original flaw that was found on the 1 E Recirculation Loop Piping.

Though the weld flaw exceeded the acceptance criteria of ASME Section XI, no leakage from this flaw was identified at any time during operation or shutdown. There is reasonable assurance that there was not a breach in the credited RCS boundary due to the axial flaw not having grown through the weld overlay. Additionally, engineering evaluation of the structural integrity of the weld shows that the flawed component had adequate margin for all design basis events. The evaluation has shown that the axial flaw identified in Weld 1 B31-1 RC-12BR-E-5 located on the recirculation inlet system, meets the ASME Code,Section XI structural margin, considering a Service Level D structural factor in the evaluation, as required by the NRC Inspection Manual. Even though the flaw has a depth of 100% of original wall thickness, which exceeds the ASME Code,Section XI allowable flaw depth of 75% of wall thickness, the safety of the reactor pressure boundary was not compromised. It is, therefore, concluded that the flawed component had adequate margin for all Design Basis Loading Events when the flaw was identified and this condition had a very low safety significance.

CORRECTIVE ACTIONS

As part of corrective actions, a weld repair on 1B31-1 RC-12BR*E-5 was completed to restore piping back to original code. The design type weld overlay was upgraded to a full structural weld overlay using lnconel Alloy 52 weld material in accordance with an NRC approved alternative (HNP-ISI-ALT-15-01).

As part of an extent of condition review, a similar weld repair was performed on 1 B31-1 RC-12BR-C-5 to upgrade its weld from a design type weld overlay to a full structural weld overlay. Also, Unit 1 and 2 weld overlays containing Alloy 82 weld material will be re-examined in the upcoming refueling outages.

ADDITIONAL INFORMATION

Other Systems Affected: No systems other than those mentioned in this report were affected by this event.

Failed Components Information

None.

Commitment Information: This report does not created any new licensing commitments.

Previous Similar Events

None.

NRC FORM 3eSA (02*2014)

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