05000255/LER-2018-003-01, Indications Identified in Reactor Pressure Vessel Head Nozzle Penetrations
| ML19210D347 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 07/29/2019 |
| From: | Hardy J Entergy Nuclear Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| PNP 2019-032 LER 2018-003-01 | |
| Download: ML19210D347 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 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) |
| 2552018003R01 - NRC Website | |
text
-===-Entergy PNP 2019-032 July 29, 2019 U. S. Nuclear Regulatory Commission A TIN: Document Control Desk Washington, DC 20555-0001 Entergy Nuclear Operations, Inc.
Palisades Nuclear Plant 27780 Blue Star Memorial Highway Covert, MI 49043 Tel 269 7642000 Jeffery A. Hardy Regulatory Assurance Manager 10 CFR 50.73
SUBJECT:
LER 2018-003 Indications Identified in Reactor Pressure Vessel Head Nozzle Penetrations Palisades Nuclear Plant Docket 50-255 License No. DPR-20
Dear Sir or Madam:
Entergy Nuclear Operations, Inc., submits the enclosed Licensee Event Report (LER),
2018-003-01, for the Palisades Nuclear Plant. The event is reportable in accordance with 10 CFR 50.73(a)(2)(ii)(A) as a degraded condition and 10 CFR 50.73(a)(2)(i)(B) as operation in a c~:>ndition prohibited by Technical Specifications. The LER describes a condition in which through-wall and axial flaw indications were identified in reactor vessel head penetrations by inspections performed during a refueling outage.
This letter contains no new commitments and no revisions to existing commitments.
Should you have any questions concerning this report, please contact Mr. Jeffery Hardy, Regulatory Assurance Manager, at (269) 764-2011.
Sincerely, JAH/bed Attachment: LER 2018-003-01, Indications Identified in Reactor Pressure Vessel Head Nozzle Penetrations CC Administrator, Region III, USNRC Project Manager, Palisades, USNRC Resident Inspector, Palisades, USNRC
ATTACHMENT LER 2018-003-01 INDICATIONS IDENTIFIED IN REACTOR PRESSURE VESSEL HEAD NOZZLE PENETRATIONS 3 Pages Follow
NRC FORM 366 U.S. NUCLEAR REGULATORY APPROVED BY OMB: NO. 3150-0104 EXPIRES: 03131/2020 (04*2018) httl2://www.nrc.gov/reading-rm/doc-coliections/nuregs/staff/sr1 0221r3D the NRC may not conduct or sponsor, and a person is not required to respond to, the inlonnation collection.
- 1. FACILITY NAME
~. DOCKET NUMBER
_PAGE PALISADES NUCLEAR PLANT 05000255 10F3
- 4. TITLE Indications Identified in Reactor Pressure Vessel Head Nozzle Penetrations
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED SEQUENTIAL FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR REV MONTH DAY YEAR NUMBER NO.
05000 11 10 2018 2018 003 01 07 29 2019 FACILITY NAME DOCKET NUMBER P5000
- 9. OPERAllNG MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check aI/ that apply) 6 o 20.2201 (b) o 20.2203(a)(3)(i) 181 50.73(a)(2)(ii)(A) o 50.73(a)(2)(viii)(A) 020.2201 (d) o 20.2203(a)(3)(ii) o 50.73(a)(2)(ii)(B) o 50.73(a)(2)(viii)(B) o 20.2203(a)(1) o 20.2203(a)(4) o 50.73(a)(2)(iii) o 50.73(a)(2)(ix)(A) o 20.2203(a)(2)(i) o 50.36(c)(1 )(i)(A) o 50.73(a)(2)(iv)(A) o 50.73(a)(2)(x)
- 10. POWER LEVEL o 20.2203(a)(2)(ii) o 50.36(c)(1 )(ii)(A) o 50.73(a)(2)(v)(A) o 73.71 (a)(4) o 20.2203(a)(2)(iii) o 50.36(c)(2) o 50.73(a)(2)(v)(B) 073.71 (a)(5) 000 o 20.2203(a)(2)(iv) o 50.46(a)(3)(ii) o 50.73(a)(2)(v)(C) o 73.77(a)(1) o 20.2203(a)(2)(v) o 50.73(a)(2)(i)(A) o 50.73(a)(2)(v)(D) o 73.77(a)(2)(i) o 20.2203(a)(2)(vi) 1810 50.73(a)(2)(i)(B) o 50.73(a)(2)(vii) o 73.77(a)(2)(ii) o 50.73(a)(2)(i)(C) o OTHER Specify in Abstract below or in NRC Form 366A
- 12. LICENSEE CONTACT FOR THIS LER ICENSEE CONTACT r lLEPHONE NUMBER (Include Area Code)
~effery Hardy, Regulatory Assurance Manager
~69-764-2011
- 13. COMPLETE ONE UNE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT
CAUSE
SYSTEM COMPONENT MANU-REPORTABLE
CAUSE
SYSTEM COMPONENT MANU-REPORTABLE FACTURER TO EPIX FACTURER TO EPIX B
- 14. SUPPLEMENTAL REPORT EXPECTED
- 15. EXPECTED MONTH DAY YEAR D YES (If yes, complete 15. EXPECTED SUBMISSION DATE)
[8J NO SUBMISSION DATE
~BSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)
On November 10, 2018, with the plant in Mode 6, during bare metal visual inspections of the reactor pressure vessel head (RPVH),
dried boric acid was identified in the area of reactor head nozzle 25, indicative of a through-wall flaw. The flaw had not been identified during review of the original ultrasonic test (UT) data. During re-evaluation of the ultrasonic test (UT) data, analysts identified a leak-path indication and an axially oriented flaw characteristic of primary water stress corrosion cracking (PWSCC). As a result of the discovery in the UT data re-evaluation for reactor head nozzle 25, Framatome extended the UT characteristics to a re-evaluation of the data for the other relevant RPVH nozzles. This extent-of-condition review identified an additional four nozzles, 33, 34, 35, and 36, that required further review. UT analysis determined that reactor head nozzle 33 contained an indication consistent with PWSCC and that reactor head penetration 35 was acceptable. Supplemental eddy current testing (ECT) was performed on reactor head nozzles 34 and 36. ECT inspection of nozzle 34 did not reveal any PWSCC indications and was determined to be acceptable. ECT inspection of nozzle 36 revealed surface breaking PWSCC-type indications.
The plant was in cold shutdown at 0% power and Mode 6 for a refueling outage at the time of discovery. Reactor head nozzles 25, 33 and 36 were repaired and the RPVH was retumed to service. The safety significance of this event was minimal. This event is reportable in accordance with 10 CFR 50.73(a)(2)(ii)(A) as a condition that resulted in a principle safety barrier being seriously I
degraded and 10 CFR 50.73(a)(2)(i)(B) for operation in a condition prohibited by Technical Specifications.
NRC FORM 366 (04-2018)
EVENT DESCRIPTION
SEQUENTIAL NUMBER
- - 003 REV NO.
- - 01 On November 10, 2018, with the plant in Mode 6, at 0% power, during bare metal visual inspections of the reactor pressure vessel [RPV;AB] head (RPVH), dried boric acid was identified in the area of reactor head nozzle [NZL;AB]
25, indicative of a through-wall flaw. The flaw had not been identified during review of the original ultrasonic test (UT) data. Dried boric acid was not observed during the previous inspection in 2017_
During re-evaluation of the UT data, analysts identified a leak-path indication and an axially oriented flaw characteristic of primary water stress corrosion cracking (PWSCC). The Palisades RPVH nozzles are Inconel Alloy 600 material which is known to be susceptible to PWSCC.
As a result of the discovery in the UT data re-evaluation for reactor head nozzle 25, Framatome extended the UT characteristics to a re-evaluation of the data for the other 52 RPVH nozzles. The reactor head vent was not re-performed as it had already been examined using eddy current testing (ECT)_ This extent-of-condition review identified an additional four reactor head nozzles, 33, 34, 35, and 36, that required further analysis. Of the four additional, reactor head nozzle 33 was determined to contain an indication with characteristics consistent with PWSCC.
In addition to the Framatome extent-of-condition review, the data for reactor head nozzles 25, 33, 34, 35, and 36 was sent to the Electric Power Research Institute (EPRI) for an independent third party review. EPRI provided concurrence with the Framatome conclusions for reactor head nozzles 25 (through-wall flaw), 33 (flaw), and 35 (no flaw). EPRI also concluded that the UT data alone was insufficient to make a definitive determination on reactor head nozzles 34 and 36_ As a result, Entergy, Framatome, and EPRI determined that a supplemental inside diameter (10) surface examination, in the form of ECT, was required to adequately evaluate the condition of these two reactor head nozzles.
On November 21, 2018, Framatome completed ECT on reactor head nozzles 36 and 34_ The ECT confirmed that reactor head nozzle 36 contained surface breaking PWSCC-type indications. Reactor head nozzle 34 was determined to be satisfactory.
Framatome provided an in-depth summary of the prior data reviews on reactor head nozzles 25, 33, and 36. The result of this review shows that the ID-initiated axial flaws were present and detectible with the demonstrated inspection method in 2007, 2009, 2010, and 2012_ However, 10 initiation has not been seen by Framatome since the 2001-2002 timeframe, and outside diameter (00) initiation is the primary industry focus of inspection. As a result, these flaws were previously mischaracterized_
Technical Specification (TS) 3.4.13 allows no pressure boundary leakage and requires entry to Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to Mode 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Because it is assumed that the through-wall flaw was present for greater than 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, this event is reportable in accordance with 10 CFR 50(a)(2)(i)(B) for operation in a condition prohibited by TS as well as 10 CFR 50(a)(2)(ii)(A) as a condition that resulted in a principle safety barrier being seriously degraded.
Page 2 of3 (04*2018)
/~
COMMISSION I~~
\\~I LICENSEE EVENT REPORT (LER)
CONTINUATION SHEET (See NUREG-1022, R.3 for instruction and guidance for completing this form http://www.nrc~ov/readino*rm/doc-collections/nureos/staff/sr1 022/r3/\\
APPROVED BY OMB: NO. 3150-0104 EXPIRES: 3131/2020
, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 3. LER NUMBER YEAR PALISADES NUCLEAR PLANT 05000-255 2018
CAUSE OF THE EVENT
SEQUENTIAL NUMBER
- - 003 REV NO.
- - 01 Based on previous intemal and industry operating experience (DE), coupled with the analysis of the data, the cause is PWSCC. In addition, Framatome notes a bias for 00 initiation due to almost exclusive 00 surface flaws over the last 10-15 years in the industry. As a result, 00 initiation is a primary focus during RPVH examinations. Framatome's cause evaluation confirmed the cause to be PWSCC.
ASSESSMENT OF SAFETY CONSEQUENCES
The safety significance of the flaw's presence during operation was minimal. There was no appreciable reactor head wastage due to the boric acid found. The Palisades' RPVH inspection program is in accordance with the requirements of ASME Code Case 729-4, as modified by the additional limitations set forth in 10 CFR 50.55a(g)(6)(ii)(O). This provides assurance against any credible PWSCC degradation event that would challenge nuclear safety. There were no consequences to the general safety of the public, nuclear safety, industrial safety, or radiological safety for this event.
CORRECTIVE ACTIONS
Just-in-Time Training was conducted on the flaw characteristics observed in nozzles 25 and 33. The training was applied to re-inspection of the remainder of the reactor head nozzle population to ensure all flawed nozzles were identified.
Framatome executed half-nozzle replacements using the inside diameter temper bead welding process to repair the flawed reactor head nozzles, and the RPVH was retumed to service.
Future inspections will require prior training on the DE from this event, along with the requirement to identity the reactor head "nozzles of interesf' population similar to that completed during the Extent of Condition review. Once nozzles of interest are identified, techniques up to and including ECT, will be used to resolve indications in that population to ensure all flawed reactor head nozzles are addressed appropriately.
The reactor head inspection contract will be revised to require the use of the improved procedures and training for detection of UT leak path indications.
PREVIOUS SIMILAR EVENTS
LER 2004-002 - Leak Path Indications Identified in Reactor Pressure Vessel Head Nozzle Penetrations, Palisades Nuclear Plant, dated December 9, 2004.