05000382/LER-2019-003, Relevant Indications Identified in Reactor Coolant Pump 1A and 2A Suction Drain Nozzle Dissimilar Metal Welds Resulting in the Condition of the Nuclear Power Plant, Including Its Principal Safety Barriers, Being
| ML19078A172 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 03/15/2019 |
| From: | Signorelli J Entergy Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| W3F1-2019-0021 LER 2019-003-00 | |
| Download: ML19078A172 (6) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded |
| 3822019003R00 - NRC Website | |
text
Entergy Operations, Inc.
17265 River Road Killona, LA 70057-3093 Tel 504-739-6032 John V. Signorelli Manager, Regulatory Assurance (Acting) 10 CFR 50.73 W3F1-2019-0021 March 15, 2019 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 The enclosed report is being sent pursuant to 10 CFR 50.73.
There are no regulatory commitments contained in this correspondence.
If you have any questions or require additional information, please contact the Acting Regulatory Assurance Manager, John V. Signorelli, at (504) 739-6032.
Respectfully, John V. Signorelli JVS/mmz cc:
NRC Region IV Regional Administrator NRC Senior Resident Inspector - Waterford Steam Electric Station, Unit 3 NRR Project Manager
Subject:
Licensee Event Report (LER) 2019-003-00 Relevant Indications Identified in Reactor Coolant Pump 1A and 2A Suction Drain Nozzle Dissimilar Metal Welds Resulting in the Condition of the Nuclear Power Plant, Including its Principal Safety Barriers, Being Seriously Degraded Waterford Steam Electric Station, Unit 3 (Waterford 3)
NRC Docket No. 50-382 Renewed Facility Operating License No. NPF-38
Enclosure:
Waterford 3 Licensee Event Report 2019-003-00
ENCLOSURE W3F1-2019-0021 Entergy Operations, Inc.
Waterford 3 Licensee Event Report 2019-003-00
NRC FORM 366 (04-2018)
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION (04-2018)
LICENSEE EVENT REPORT (LER)
(See Page 2 for required number of digits/characters for each block)
(See NUREG-1022, R.3 for instruction and guidance for completing this form http://www.nrc/gov/reading-rm/doc-collections/nuregs/staff/sr1022/r3/)
APPROVED BY OMB: NO. 3150-0104 EXPIRES: 03/31/2020
, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 1. Facility Name Waterford Steam Electric Station, Unit 3
- 2. Docket Number 05000382
- 3. Page 1 OF 4
- 4. Title Relevant Indications Identified in Reactor Coolant Pump 1A and 2A Suction Drain Nozzle Dissimilar Metal Welds Resulting in the Condition of the Nuclear Power Plant, Including its Principal Safety Barriers, Being Seriously Degraded
- 5. Event Date
- 6. LER Number
- 7. Report Date
- 8. Other Facilities Involved MONTH DAY YEAR YEAR SEQUENTIAL NUMBER REV NO.
MONTH DAY YEAR FACILITY NAME DOCKET NUMBER 01 17 2019 2019 -
003
- - 00 03 15 2019 FACILITY NAME DOCKET NUMBER
- 9. Operating Mode
)
No Abstract (Limit to 1400 spaces, i.e., approximately 14 single-spaced typewritten lines)
On January 17 and 18, 2019, relevant indications were detected on the Reactor Coolant System (RCS) Loop Reactor Coolant Pump (RCP) suction drain nozzle to safe-end butt welds07-009 (RCP 1A) and 11-007 (RCP 2A) during the performance of Phased Array Ultrasonic Examination of Alloy 600 dissimilar metal piping welds during planned inspections which were being conducted during Refueling Outage 22. This resulted in an event or condition that resulted in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded because they did not meet acceptance criteria referenced in American Society of Mechanical EngineersSection XI, IWB-3514-2 and Code Case N-770-2. The RCS was declared inoperable per Technical Requirements Manual (TRM) requirement 3.4.9, "Structural Integrity."
The cause of this event was Primary Water Stress Corrosion Cracking. Action was taken to restore full component integrity and assure structural integrity of the affected welds by applying dissimilar metal weld full structural weld overlays over the affected components. The RCS was subsequently declared operable on February 18, 2019 and TRM 3.4.9 was exited.
EVENT DESCRIPTION
A. Plant Status When the indications were identified, Waterford 3 was in Mode 6 (Refueling). There were no other structures, systems or components out of service that contributed to this event.
B. Event Chronology and Description of Relevant Indications Detected On January 17 and 18, 2019, relevant indications were detected on the Reactor Coolant System (RCS) [AB] Loop Reactor Coolant Pump (RCP) [P] suction drain nozzle [NZL] to safe-end butt welds07-009 (RCP 1A) and 11-007 (RCP 2A) during the performance of Phased Array Ultrasonic Examination of Alloy 600 dissimilar metal piping welds during planned inspections which were being conducted during Refueling Outage 22. The welds contribute to maintaining the RCS pressure boundary. The RCP suction drain nozzles are categorized as American Society of Mechanical Engineers (ASME) Code Class 1 components. The indications are axially oriented and located within the butt weld and weld butter. The axial indications are inside surface connected and are exposed to the reactor coolant.
The indication on weld 07-009 is in the axial direction and measures 0.59 which is 55% through-wall. The width of the indication is 0.40. The total thickness of the weld is 1.08. The remaining ligament 0.49 is between the flaw and the outer diameter surface.
The indication on weld 11-007 is in the axial direction and measures 0.58 which is 50% through-wall. The width of the indication is 0.60. The total thickness of the weld is 1.17. The remaining ligament 0.59 is between the flaw and the outer diameter surface.
On January 17, 2019, at 1858 CST, following identification of the indication on weld 07-009, the RCS was declared inoperable in accordance with Technical Requirements Manual requirement 3.4.9, "Structural Integrity," due to the weld did not meet acceptance criteria referenced in ASME Section XI, IWB-3514-2 and ASME Code Case N-770-2. Operations entered TRM 3.4.9 action a, which requires that with the structural integrity of any ASME Code Class 1 component(s) not conforming to the ASME requirements, restore the structural integrity of the affected component(s) prior to increasing the RCS temperature more than 70 degrees Fahrenheit above the minimum temperature required by nil ductility temperature (NDT) considerations.
This event is reportable pursuant to 10 CFR 50.73(a)(2)(ii)(A), "Any event or condition that resulted in: (A) The condition of the nuclear power plant, including its principal safety barriers, being seriously degraded," due to welding defects in the RCS that cannot be found acceptable under ASME Section XI. Page 3
of 4 COMMISSION (04-2018)
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/r3/)
APPROVED BY OMB: NO. 3150-0104 EXPIRES: 03/31/2020
, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 3. LER NUMBER Waterford Steam Electric Station, Unit 3 05000382 YEAR SEQUENTIAL NUMBER REV NO.
2019 003 00 C. Event Causes A failure mode analysis and an equipment failure evaluation were performed. This identified that the most likely cause of the indications is Primary Water Stress Corrosion Cracking (PWSCC).
CORRECTIVE ACTIONS
(1) Review for extent of condition per ASME Code Case N-770-2 paragraph 2430(a), which requires that if an examination reveals an unacceptable flaw, then the scope of Inservice Inspections (ISI) must be extended to include additional weld inspections during the current outage. The result of this review was that all welds which would have been included in the code required extent of condition were scoped into the refueling outage and examined using Phased Array Ultrasonic Examination as part of the regularly scheduled outage activities. (complete)
(2) Restore full component integrity and assure structural integrity of the affected welds by obtaining NRC approval of request for relief for and application of dissimilar metal weld full structural weld overlays over the affected components. (complete)
(3) Update the Alloy 600 Project technical report and issue additional actions to revise the Alloy 600 Aging Management program contingency or mitigation plans for any identified deficiencies.
(planned)
(4) Assess the likelihood of other cold leg weld failures occurring and, as appropriate, generate action to revise the Alloy 600 Aging Management program contingency or mitigation plans for any identified deficiencies. (planned)
(5) Verify similar susceptible welds do not have unacceptable indications. (complete)
(6) Review Entergy ISI plans to identify welds that may require mitigation. (complete)
(7) Ensure Entergy ISI plans require encoded phased array inspection of all applicable dissimilar metal welds in the future. (planned)
(8) Ensure current ASME Section XI inspection intervals are sufficient to identify indications prior to their becoming unacceptable.
SAFETY EVALUATION The actual consequence of this event was that the indications were not acceptable under ASME Section XI. There were no other actual consequences to safety of the general public, nuclear safety, industrial safety, and radiological safety for this event. Page 4
of 4 COMMISSION (04-2018)
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/r3/)
APPROVED BY OMB: NO. 3150-0104 EXPIRES: 03/31/2020
, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 3. LER NUMBER Waterford Steam Electric Station, Unit 3 05000382 YEAR SEQUENTIAL NUMBER REV NO.
2019 003 00 The potential consequence to safety of the general public, nuclear safety, industrial safety and radiological safety of this event if the indication was through-wall is a loss of the RCS pressure boundary. This would have resulted in a small, non-isolable reactor coolant leak that would have been detected by local radiation detectors and through RCS leakrate surveillance by plant personnel.
PREVIOUS OCCURRENCES
A review of the Waterford 3 corrective action program and previous Licensee Event Reports for the previous 12 years revealed no similar events as described in NUREG-1022 guidance.
Energy Industry Identification System (EIIS) codes and component codes are identified in the text as [XX].