05000255/LER-2018-003, Indications Identified in Reactor Pressure Vessel Head Nozzle Penetrations
| ML19003A239 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 01/03/2019 |
| From: | Hardy J Entergy Nuclear Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| PNP 2018-056 LER 2018-003-00 | |
| Download: ML19003A239 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 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)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i) |
| 2552018003R00 - NRC Website | |
text
- ~ Entergy PNP 2018-056 January 3, 2019 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Entergy Nucl~ar Operations, Inc.
Palisades Nuclear Plant 27780 Blue Star Memorial Highway Covert, MI 49043 Tel 269 764 2000 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),
2018-003-00, for the Palisades Nuclear Plant. The event is reportable in accordance with 10 CFR 50.73(a)(2)(ii)(A) as a degraded condition. 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, 9i\\k\\~
JAH/bed Attachment: LER 2018-003-00, Indications Identified in Reactor Pressure Vessel Head Nozzle Penetrations CC Administrator, Region III, USNRC Project Manager, Palisades, USNRC Resident Inspector, Palisades, USNRC
ATTACHMENT LER 2018-003-00 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: 03/3112020 (04*2018) httQ:/Iwww.nrc.gov/reading*rm/doc-collections/nuregs/stafflsr1 022/r3!')
the NRC may not conduct or sponsor, and a person is not required to respond to, the information 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 00 01 03 2019 FACILITY NAME DOCKET NUMBER 05000
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply) 6 020.2201 (b) o 20.2203(a)(3)(i) 18150.73(a)(2)(ii)(A) o 50.73(a)(2)(viii)(A) o 20.2201 (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) o 73.71 (a)(5) 0 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) o 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 LICENSEE CONTACT r rLEPHONE 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 S
- 14. SUPPLEMENTAL REPORT EXPECTED
- 15. EXPECTED MONTH DAY YEAR IZI YES (If yes, complete 15. EXPECTED SUBMISSION DATE) o NO SUBMISSION 07 01 2019 DATE ABSTRACT (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 iden@ed 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 returned 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 degraded.
NRC FORM 366 (04*2018)
EVENT DESCRIPTION
SEQUENTIAL NUMBER
- - 003 REV NO.
- - 00 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.
Page 2 of3 (04-2018)
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COMMISSION
......... 1 LICENSEE EVENT REPORT (LER)
CONTINUATION SHEET (See NUREG-1 022, R.3 for instruction and guidance for completing this form http://www.nrc.aov/readina-rm/doc-collections/nureas/staff/sr1 022/r3/\\
APPROVED BY OMB: NO. 3150-0104 EXPIRES: 313112020
, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 3. LEA NUMBER YEAR PALISADES NUCLEAR PLANT 05000-255 2018
CAUSE OF THE EVENT
SEQUENTIAL NUMBER
- - 003 REV NO.
- - 00 Based on previous internal and industry operating experience (OE), coupled with the analysis of the data, the cause is PWSCC. In addition, Framatome notes a bias for OD initiation due to almost exclusive OD surface flaws over the last 10-15 years in the industry. As a result, OD initiation is a primary focus during RPVH examinations. Framatome's cause evaluation of this event is not yet complete.
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)(D). 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 returned to service, Future inspections will require prior training on the OE from this event, along with the requirement to identify 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 cause evaluation being completed by Framatome will be reviewed. If warranted, this LER will be supplemented to incorporate appropriate conclusions from that evaluation.
PREVIOUS SIMILAR EVENTS
LER 2004-002 - Leak Path Indications Identified in Reactor Pressure Vessel Head Nozzle Penetrations, Palisades Nuclear Plant, dated December 9, 2004.