05000321/LER-2016-003

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

Edwin I. Hatch Nuclear Plant Unit 1
Event date:	02-16-2016
Report date:	4-14-2016
Reporting criterion:	10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded
3212016003R00 - NRC Website

LER 16-003-00 for E.I. Hatch, Unit 1, 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 R 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)

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PLANT AND SYSTEM IDENTIFICATION

General Electric- Boiling Water Reactor (BWR) Energy Industry Identification System codes appear in the text as (EIIS 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 1B31-1RC-12BR-E-5 (1631-E5) to safe-end weldment had propagated into the Inconel 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 11331-E5 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 Inconel Alloy 82 WOL material installed in 1988.

The flaw was repaired in accordance with an approved code alternative and 1631-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 1631-E5 was due to intergranular stress corrosion cracking (IGSCC). The 304 stainless steel piping and Inconel Alloy 182 weld materials are susceptible to this failure mode in a BWR environment.

The station also lost track of the different design attributes of 1631-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 ISI 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 1B31-E5 was upgraded to a full structural weld overlay (FSWOL) using intergranular stress corrosion cracking resistant Alloy 52 weld material.

REPORT ABILITY AND SAFETY ASSESSMENT

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 1B31-1RC-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 11331-1RC-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 Inconel 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 1631-1RC-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.