05000247/LER-2018-001, Re Penetration Indications Discovered During Reactor Pressure Vessel Head Inspection

Re Penetration Indications Discovered During Reactor Pressure Vessel Head Inspection

ML18149A126
Person / Time
Site:	Indian Point
Issue date:	05/21/2018
From:	Vitale A Entergy Nuclear Northeast
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-18-039 LER 2018-001-00
Download: ML18149A126 (6)

LER-2018-001, Re Penetration Indications Discovered During Reactor Pressure Vessel Head Inspection

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(1), Submit an LER, Invalid Actuation 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)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
2472018001R00 - NRC Website

text

• ==~ Entergx NL-18-039 May 21, 2018 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Stop O-P1-17 Washington, D.C. 20555-0001 Entergy Nuclear Northeast Indian Point Energy Center 450 Broadway, GSB P.O. Box249 Buchanan, N.Y. 10511-0249 Tel (914) 254-6700 Anthony J Vitale Site Vice President SUBJECT: Licensee Event Report# 2018-001-00, "Penetration Indications Discovered During Reactor Pressure Vessel Head Inspection" Indian Point Unit No. 2 Docket No. 50-247 DPR-26

Dear Sir or Madam:

Pursuant to 10 CFR 50.73(a)(1), Entergy Nuclear Operations Inc. (ENO) hereby provides Licensee Event Report (LER) 2018-001-00. The enclosed LER identifies a degraded condition discovered during the reactor pressure vessel (RPV) head inspection conducted during the recently completed Indian Point Unit 2 Spring 2018 refueling outage (2R23).

The degraded condition resulted from RPV head penetration indications that could not be dispositioned as acceptable per the requirements of 1 O CFR 50.55a. This condition is reportable under 10 CFR 50. 73(a)(2)(ii)(A) as a condition that resulted in the nuclear power plant, including its principal safety barriers, being degraded. In addition, one of the indications was found to have a through-weld component, with evidence of reactor coolant leakage. Because Technical Specification 3.4.13 (Reactor Coolant System (RCS) Operational Leakage) does not allow pressure boundary leakage, the condition is also reportable under 10 CFR 50. 73(a)(2)(i)(B) as a condition prohibited by the plant's Technical Specifications. The described condition was recorded in the Entergy Corrective Action Program as Condition Reports CR-IP2-2018-01944, 01985, 02071, and 02122.

NL-18-039 Page 2 of 2 There are no new commitments identified in this letter. Should you have any questions regarding this submittal, please contact Mr. Robert Walpole, Manager, Regulatory Assurance at (914) 254-6710.

Sincerely,

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AJV/cdm cc:

Mr. David Lew, Acting Regional Administrator, NRC Region I NRC Resident Inspector's Office Ms. Bridget Frymire, New York State Public Service Commission

NRCFORM366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY 0MB: NO. 3150-0104 EXPIRES: 03/31/2020 (02-2018)

, the NRG may not conduct or sponsor, and a oerson is not reauired to resoond to the information collection.

,3. Page Indian Point 2 05000-247 1 OF4

4. Title Penetration Indications Discovered During Reactor Pressure Vessel Head Inspection

5. Event Date

6. LER Number

7. Report Date

8. Other Facilities Involved Sequential Rev Facility Name Docket Number Month Day Year Year Number No.

Month Day Year 05000 03 31 2018 2018

- 001

- 00 05 21 2018 Facility Name Docket Number 05000

9. Operating Mode

11. This Reoort is Submitted Pursuant to the Reciuirements of 10 CFR §: (Check all that ann/vl D 20.2201(b)

D 20.2203(a)(3)(i)

~ 50.73(a)(2)(ii)(A)

D 50.73(a)(2)(viii)(A) 6 D 20.2201(ct)

D 20.2203(a)(3)(ii)

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

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

D 20.2203(a)(1)

D 20.2203(a)(4)

D 50.73(a)(2)(iii)

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

D 20.2203(a)(2)(i)

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

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

D 50.73(a)(2)(x)

10. Power Level D 20.2203(a)(2)(ii)

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

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

D 73.71 (a)(4)

D 20.2203(a)(2)(iii)

D 50.36(c)(2)

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

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

D 50.46(a)(3)(ii)

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

D 73.77(a)(1)

D 20.2203(a)(2)(v)

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

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

D 73.77(a)(2)(ii)

D 20.2203(a)(2)(vi)

[8] 50. 73(a)(2)(i)(B)

D 50.73(a)(2)(vii)

D 73.77(a)(2)(iii)

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

D Other (Specify in Abstract below or in NRC Form 366A

12. Licensee Contact for this LER Licensee Contact relephone Number (Include Area Code)

Nelson Azevedo, Supervisor, Engineering I (914) 254-6775 Cause System Component Manufacturer Reportable To ICES

Cause

System Component Manufacturer Reportable To ICES B

AB RPV W120 y

14. Supplemental Report Expected Month Day Year D Yes (If yes, complete 15. Expected Submission Date) J::gj No

15. Expected Submission Date Abstract (Limit to 1400 spaces, i.e., approximately 14 single-spaced typewritten lines)

On March 28, 2018, during the Indian Point Unit 2 Spring 2018 refueling outage (2R23}, a white substance, indicative of reactor coolant leakage, was observed on the annulus region of reactor pressure vessel (RPV) head Penetration No. 3 while performing the visual testing (VT)-2 bare metal examination of the RPV head as required by American Society of Mechanical Engineering (ASME) Code Case N-729-4, as amended by the Nuclear Regulatory Commission (NRC) in 10 CFR 50.55a(g)(6)(ii)(D). No base material wastage was identified.

On March 31, 2018, liquid penetrant testing and eddy current testing of the partial penetration J-groove weld of RPV head Penetration No. 3 confirmed the presence of two relevant indications on the inside wetted surface of the J-groove weld.

Both identified indications were determined to have lengths of approximately 0.085 inches. One of the two indications was characterized as being axially oriented with a measurable through-weld component, and it was concluded that this was the source of the reactor coolant leakage. The cause of the Penetration No. 3 indications was attributed to Primary Water Stress Corrosion Cracking (PWSCC) of the original Alloy 600 (Alloy 82/182) partial penetration J-groove weld.

Penetration No. 3 was repaired in accordance with the NRC approved Westinghouse embedded flaw weld overlay method (WCAP-15987-P, Revision 2-P-A) and relief request IP2-ISI-RR-06.

This event had no effect on the public health and safety. The event was reported to the NRC on March 31, 2018 under 10 CFR 50. 72(b )(3)(ii)(A) as a condition of the nuclear power plant, including its principal safety barriers, being degraded.

NRG FORM 366 (02-2018)

Note: The Energy Industry Identification System Codes are identified within the brackets { }.

DESCRIPTION OF EVENT

SEQUENTIAL NUMBER

- 001 REV NO.

- 00 On March 28, 2018, during the Indian Point Unit 2 (IP2) Spring 2018 refueling outage (2R23), a white substance was observed on the annulus region of reactor pressure vessel (RPV) {AB, RPV} head Penetration {PEN} No. 3 while performing the visual testing (VT)-2 bare metal examination of the RPV head as required by American Society of Mechanical Engineering (ASME) Code Case N-729-4, as amended by the Nuclear Regulatory Commission (NRC) in 10 CFR 50.55a(g)(6)(ii)(D). The VT-2 visual examination did not identify any other relevant indications or base material wastage. Subsequent gamma spectroscopy testing of a sample of the white substance detected the presence of Cesium (Cs)-137, Cobalt (Co)-58, and Co-60, which is indicative of reactor coolant {AB} leakage, and that the leakage had commenced within the past 90 to 180 days.

On March 29, 2018, ultrasonic testing (UT) and eddy current testing (ET) surface examinations of the control rod drive mechanism (CROM) {AA} nozzle {NZL} of RPV head Penetration No. 3 were performed. UT is a volumetric technique used for examination of the CROM tube, while ET is a technique designed to identify indications open to the wetted surface of the component. The UT and ET examinations were performed from the inside surface of the Alloy 600 penetration tube. Although no indications in the tube were identified, the UT leak path assessment did provide evidence of a change from previous outage examinations.

On March 31, 2018, follow up liquid penetrant testing (PT) and ET surface examinations of the partial penetration J-groove weld of RPV head Penetration No. 3 confirmed the presence of two relevant indications on the inside wetted surface of the J-groove weld. Both identified indications were determined to have linear components with estimated lengths of approximately 0.085 inches. One of the two indications was characterized as being axially oriented with a measurable through-weld component, and it was concluded that this was the source of the reactor coolant leakage previously identified in the annulus region of Penetration No. 3.

The two indications described above were determined to not meet the requirements of ASME Code N-729-4 due to evidence of reactor coolant leakage identified coincident with surface-connected flaws. ASME Code Case N-729-4 and ASME Boiler and Pressure Vessel Code Section XI do not allow through-wall flaws in pressure boundary material in Class 1 (i.e., reactor coolant pressure boundary) components. Indications that cannot be dispositioned as acceptable per ASME Code Section XI in a Reactor Coolant System (RCS) pressure boundary are reportable under 10 CFR 50.72(b)(3)(ii)(A) and 10 CFR 50.73(a)(2)(ii)(A) as a condition of the nuclear power plant, including its principal safety barriers, being degraded. The Penetration No. 3 examination that could not be dispositioned as acceptable per the requirements of 10 CFR 50.55(a) was reported to the NRC under 10 CFR 50. 72(b)(3)(ii)(A) as required on March 31, 2018 (Event Number: 53305).

Primary Water Stress Corrosion Cracking (PWSCC) occurs in pressurized water reactors when a susceptible material is exposed to the borated primary water environment, and when in the presence of residual tensile stresses caused by welding. The IP2 RPV head Penetration No. 3 Alloy 600 (Alloy 82/182) J-groove weld in which the indications were discovered meets these criteria. The cause of the Penetration No. 3 indications was attributed to PWSCC of the original Alloy 600 partial penetration J-groove weld. An extent of condition analysis was performed, which identified the IP2 and Indian Point Unit 3 (IP3) locations with Alloy 600 materials susceptible to PWSCC. Since compensatory actions in the form of increased inspection frequencies as specified in 10 CFR 50.55(a) have already been implemented, no additional actions were deemed necessary.

The RPV heads for both IP2 and IP3 were fabricated by Combustion Engineering {C490} under a Westinghouse

{W120} Nuclear Steam Supply System (NSSS) design specification. Because the indications in the IP2 RPV head Page 2 of 4 l(04-2017)

U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY 0MB: NO. 3150-0104 EXPIRES: 3/31/2020 tw\\

~i~ll LICENSEE EVENT REPORT (LER)

CONTINUATION SHEET (See NUREG-1022, R.3 for instruction and guidance for completing this form http://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr1022/r3D

, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.

3. LERNUMBER YEAR Indian Point 2 05000-247 2018 SEQUENTIAL NUMBER

- 001 REV NO.

- 00 Penetration No. 3 Alloy 600 J-groove weld could not be accepted by analytical evaluation, a weld repair of the J-groove weld was necessary. All of the RPV head penetrations were VT-2 and UT inspected during the IP2 Spring 2018 refueling outage (2R23) in accordance with ASME Code Case N-729-4, as amended by 10 CFR 50.55a(g)(6)(ii)(D), and only Penetration No. 3 was found to have indications requiring repair.

Pursuant to 10 CFR 50.55a(z)(1), Entergy Nuclear Operations, Inc. (Entergy) submitted relief request IP2-RR-ISl-06, which proposed to repair the J-groove weld of RPV head Penetration No. 3 using the Westinghouse embedded flaw repair process as an alternative to the defect removal and weld repair provisions of ASME Section XI, IWA-4000 and ASME Section Ill, NB-4450. The proposed embedded flaw repair process is described in WCAP-15987-P, Revision 2-P-A, and has been approved by the NRC. The embedded flaw repair methodology has been successfully employed multiple times for repairs of indications on the J-groove weld or the outside surface of penetration nozzles.

Entergy received NRC verbal authorization on April 9, 2018 (ML18099A373) for relief request IP2-RR-ISl-06, allowing use of the embedded flaw repair method for the period of the relief request (i.e., IP2 Cycle 24 that will end in Spring 2020). The Penetration No. 3 weld repair utilizing the embedded flaw repair process involved sealing the entire Alloy 600 J-groove weld and a portion of the Alloy 600 penetration tube with an Alloy 690 (Alloy 52/52M) weld metal overlay. The weld repair was completed prior to IP2 entering Mode 5 (Cold Shutdown).

CAUSE OF EVENT

The cause of the identified RPV head Penetration No. 3 indications was attributed to PWSCC of the original Alloy 600 (Alloy 82/182) partial penetration J-groove weld. This is a known failure mechanism that is addressed by the requirements of 1 O CFR 50.55a(g)(6)(ii)(D). The repair of Penetration No. 3 utilized the Westinghouse WCAP-15987-P, Revision 2-P-A, embedded flaw weld overlay process to encapsulate the indication with an Alloy 690 (Alloy 52/52M) PWSCC resistant material. A weld overlay of PWSCC resistant material to isolate the indications from the borated primary water environment arrests any further PWSCC on the penetration. This was the first weld overlay repair performed at either IP2 or IP3 using the alternative embedded flaw weld overlay methodology.

CORRECTIVE ACTIONS

RPV head Penetration No. 3 has been repaired in accordance with the NRC approved Westinghouse embedded flaw weld overlay method (WCAP-15987-P, Revision 2-P-A) and relief request I P2-ISI-RR-06.

EVENT ANALYSIS

The event is reportable under 10 CFR 50. 73(a)(2)(ii)(A). The licensee shall report any event or condition that resulted in the condition of the nuclear power plant, including its principal safety barriers, being degraded. This condition meets the reporting criteria because the indications identified in the J-groove weld of RPV head Penetration No. 3, which is considered part of the RPS pressure boundary, could not be dispositioned as acceptable under ASME Code Section XI. In addition, one of the indications was found to have a through-weld component, with evidence of reactor coolant leakage. Because Technical Specification 3.4.13 (Reactor Coolant System (RCS) {AB}

Operational Leakage) does not allow pressure boundary leakage, the condition is also reportable under 10 CFR 50.73(a)(2)(i)(B) as a condition prohibited by the plant's Technical Specifications.

PAST SIMILAR EVENTS No previous similar Licensee Event Reports (LERs) were identified for either IP2 or IP3. Page 3 of 4 (04-2017)

U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY 0MB: NO. 3150-0104 EXPIRES: 3/31/2020 LICENSEE EVENT REPORT (LER)

CONTINUATION SHEET (See NUREG-1022, R.3 for instruction and guidance for completing this form http://www.nrc.gov/reading-rm/doc-collections/n u regs/staff/sr1 022/r3D

, the NRG may not conduct or sponsor, and a person is not required to respond to, the information collection.

3. LERNUMBER YEAR Indian Point 2 05000-247 2018

SAFETY SIGNIFICANCE

SEQUENTIAL NUMBER

- 001 REV NO.

- 00 This event had no effect on the health and safety of the public. There were no actual safety consequences for the event because the through-weld reactor coolant leakage in the annulus region of RPV head Penetration No. 3 was minor and the timely RPV head inspection and repair of the identified indications prevented significant propagation of the indications. All of the RPV head penetrations were VT-2 and UT inspected during the I P2 Spring 2018 refueling outage (2R23), and only Penetration No. 3 was found to have indications requiring repair. During the period of plant operation with the through-weld reactor coolant leakage, the primary containment function to act as an additional fission product barrier was not affected, the Control Rod Drive System {AA} remained structurally capable of performing its intended safety functions, and IP2 Technical Specification 3.4.13 would have required a plant shutdown in the unlikely event that the reactor coolant leakage had exceeded the operational limit. With the implementation of the NRC approved Westinghouse embedded flaw repair process on Penetration No. 3, the PWSCC was mitigated and no longer a credible crack growth mechanism. The embedded flaw repair is considered bounded by previous analyses and the structural integrity of the RPV closure head and Penetration No. 3 are maintained for one fuel cycle of operation in accordance with the NRC approved relief request (IP2-ISI-RR-06).

In addition, an extensive industry safety assessment of PWSCC in RPV head penetrations concluded that an inspection program of periodic nonvisual, non-destructive examinations at appropriate intervals supplemented by periodic bare metal visual examinations provides adequate protection against safety-significant failures. The safety assessment also demonstrates that the bare metal visual examination of the head top surface performed at appropriate intervals provides assurance against significant wastage of the low-alloy steel head material, even given a leaking penetration nozzle. Based on the industry safety assessment and the results of the IP2 examinations, it is reasonable to conclude that the Indian Point Energy Center inspection program, which is in accordance with ASME Code Case N-729-4, as modified by the additional limitations set forth in 10 CFR 50.55a(g)(6)(ii)(D), 'provides assurance against any credible PWSCC degradation event that would challenge nuclear safety. Page 4 of 4