Information Notice 2019-08, Flow-Accelerated Corrosion Events
| ML19065A123 | |
| Person / Time | |
|---|---|
| Issue date: | 10/08/2019 |
| From: | Anna Bradford, Chris Miller NRC/NRO/DLSE, NRC/NRR/DIRS/IRGB |
| To: | |
| Lintz M, 415-4051, NRR/DIRS | |
| References | |
| IN-19-008 | |
| Download: ML19065A123 (4) | |
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
OFFICE OF NUCLEAR MATERIAL SAFETY AND SAFEGUARDS
WASHINGTON, DC 20555-0001 October 8, 2019 NRC INFORMATION NOTICE 2019-08: FLOW-ACCELERATED CORROSION EVENTS
ADDRESSEES
All holders of an operating license or construction permit for a nuclear power reactor under
Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Domestic Licensing of
Production and Utilization Facilities, except those that have permanently ceased operations
and have certified that fuel has been permanently removed from the reactor vessel.
All holders of and applicants for a power reactor combined license, standard design approval, or
manufacturing license under 10 CFR Part 52, Licenses, Certifications, and Approvals for
Nuclear Power Plants. All applicants for a standard design certification, including such
applicants after initial issuance of a design certification rule.
PURPOSE
The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice (IN) to inform
addressees of recent operating experience in which flow-accelerated corrosion (FAC) events
resulted in reactor trips. The NRC expects that recipients will review the information for
applicability to their facilities and consider actions, as appropriate, to avoid similar problems.
INs may not impose new requirements, and nothing in this IN should be interpreted to require
specific action.
DESCRIPTION OF CIRCUMSTANCES
Indian Point Energy Center, Unit 3
On September 18, 2018, while in Mode 1 at 100 percent reactor power, operators at Indian
Point Unit 3 manually tripped the reactor and closed all main steam isolation valves in response
to a steam leak on a 6-inch elbow located upstream of the 36C feedwater heater. The direct
cause of the steam leak was FAC. The root cause was attributed to the program engineers not
using the replacement history to identify susceptibility to FAC, as components on this line had
been replaced in 2007 because of previous failures. Contributing causes included weaknesses
in the setup of the FAC program software model and inadequate procedure guidance for scope
expansion from the 2007 failure. Corrective actions included replacing the failed component, revising the model to split the reheater drain branches into three separate runs with one run per
heater, and revising procedures on scope expansion and system replacement history.
Additional information appears in Indian Point - Integrated Inspection Report 05000247/2018004 and 05000286/2018004, dated February 7, 2019, on the NRCs public
website in the Agencywide Documents Access and Management System (ADAMS) Accession
No. ML19038A398, Indian Point - Integrated Inspection Report 05000247/2019002 and
ML19065A123 05000286/2019002, dated August 13, 2019 (ADAMS Accession No. ML19225C606), and
Indian Point Licensee Event Report 50-286/2018-003-00, dated November 19, 2018 (ADAMS
Accession No. ML18341A122).
Davis-Besse Nuclear Power Station
On May 9, 2015, while in Mode 1 at 100 percent reactor power, field operators at Davis-Besse
reported a steam leak on a 4-inch pipe in the moisture separator reheater system. After
initiating a rapid shutdown, the operators manually tripped the reactor from approximately
30 percent power. The direct cause of the steam leak was FAC. An incorrect data input
caused the FAC software model to underestimate the predicted wear rate, so inspections had
not been performed to identify the wall thinning before failure. Additionally, corrective action
from a comparable event in 2006 did not include a validation of all critical data inputs.
Corrective actions from the more recent event included improvements in the fidelity of the data
in the FAC software model and improvements in the corrective action program with respect to
root cause evaluations.
Additional information appears in Davis-BesseNRC Integrated Inspection Report 05000346/2015003, dated October 21, 2015 (ADAMS Accession No. ML15295A107), and
Davis-Besse Licensee Event Report 50-346/2015-002, dated July 8, 2015 (ADAMS Accession
No. ML15194A013).
BACKGROUND
Related NRC Generic Communications
NRC Bulletin 87-01, Thinning of Pipe Walls in Nuclear Power Plants, dated July 9, 1987, requested addressees to submit information on their programs for monitoring the thickness of
pipe walls in high-energy single-phase and two-phase carbon steel piping systems.
NRC Generic Letter 89-08, Erosion/Corrosion-Induced Pipe Wall Thinning, dated May 2, 1989, required addressees to provide assurances that they have implemented a program consisting of
systematic measures to ensure that erosion/corrosion does not lead to degradation of
single-phase and two-phase high-energy carbon steel systems.
DISCUSSION
These events demonstrate the importance of adequately implementing a FAC program, as both
events resulted in reactor trips. While neither of these events caused personnel injury, workers
have been seriously injured or killed in previous events because of failures resulting from FAC.
In 1986, four workers died at Surry Power Station after a catastrophic failure of a pipe because
of FAC. This event prompted the NRC to issue Bulletin 87-01, which requested the
implementation of a program for monitoring the wall thickness of piping at each site.
It is important to apply appropriate engineering judgement and not to place overreliance on the
FAC program software model. Correctly inputting data into the model ensures accurate
modeling and consequential accurate wear rate prediction. Different inputs include, but are not
limited to, diameter, geometry, chemistry, thermodynamic properties, and material content. For
example, if an incorrect diameter is used, or if the presence of trace chromium is inputted when
it is not present, then nonconservative wear rates may be predicted. As such, addressees may
consider performing periodic verifications and validations of the model, in accordance with an approved QC/QC program, and of the assumptions made to the initial setup and updates to the
model.
CONTACT
S
Please direct any questions about this matter to the technical contacts listed below.
/RA/ /RA/
Christopher G. Miller, Director Anna H. Bradford, Deputy Director
Division of Inspection and Regional Support Division of Licensing, Siting, Office of Nuclear Reactor Regulation and Environmental Analysis
Office of New Reactors
Technical Contacts: Catherine Nolan, NSIR James Gavula, NRR
301-415-1535 630-829-9755 Catherine.Nolan@nrc.gov James.Gavula@nrc.gov
ML19065A123 *concurred via email
OFFICE TECH EDITOR NSIR/DPR/OB NRR/DNLR/MCCB NRR/DIRS/IRGB/LA NRR/DIRS/IOEB/BC
NAME KAzariah-Kribbs* CNolan* JGavula* IBetts RElliott*
DATE 9/23/19 9/24/19 9/24/19 09/20/19 9/25/19 OFFICE NRR/DIRS/IRGB/PM NRR/DIRS/IRGB/BC NRO/DLSE/DD NRR/DIRS/D
NAME BBenney PMcKenna ABradford CMiller
DATE 9/26/19 9/26/19 10/1/19 10/10/19