Information Notice 2007-21, Pipe Wear Due to Interaction of Flow-Induced Vibration and Reflective Metal Insulation
| ML20225A204 | |
| Person / Time | |
|---|---|
| Issue date: | 12/11/2020 |
| From: | Chris Miller NRC/NRR/DRO/IOEB |
| To: | |
| Mark Litz NRR/DRO/IOEB, 415-4051 | |
| References | |
| IN 2007-21, Supp 1 | |
| Download: ML20225A204 (6) | |
IN 2007-21, Supp. 1 ML20225A204
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, DC 20555-0001
December 11, 2020
INFORMATION NOTICE 2007-21, Supplement 1: PIPE WEAR DUE TO INTERACTION OF
FLOW-INDUCED VIBRATION AND
REFLECTIVE METAL INSULATION
Addressees
All holders of operating licenses, construction permits, or combined licenses for nuclear power
reactors, except those that have permanently ceased operations and have certified that fuel has
been permanently removed from the reactor vessel.
Purpose
The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice (IN) to alert
addressees to recent information related to wear of nuclear power plant piping caused by
flow-induced vibration (FIV) conditions that has not been captured in industry operating
experience reports. In this supplement to IN 2007-21, Pipe Wear Due to Interaction of Flow- Induced Vibration and Reflective Metal Insulation, dated June 11, 2007 (Agencywide
Documents Access and Management System (ADAMS) Accession No. ML071150051), the
NRC staff discusses additional recent instances of piping wear due to FIV conditions. The NRC
expects that addressees will review the information for applicability to their facilities and
consider actions, as appropriate, to identify and address similar problems. However, suggestions contained in this IN are not NRC requirements; therefore, the NRC requires no
specific action or written response to this IN.
Description of Circumstances
In IN 2007-21, the NRC staff described multiple wear marks on the chemical and volume control
(CVCS) system stainless steel piping downstream of the CVCS system letdown orifices at
Catawba Nuclear Station (Catawba), Unit 1, identified during the refueling outage in the fall of
2006. The licensee determined that these marks were the result of abrasive wear between the
stainless steel reflective metal insulation (RMI) end caps and the American Society of
Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPV Code) Class 2 piping.
The cause of this abrasive wear was most likely FIV causing an interaction between the end cap
and the piping. The licensee assembled RMI by clipping together short segments of insulation
with the end caps located at the intersection of each insulation segment. The Catawba, Unit 1, licensee discovered a total of 84 wear marks from its initial and extent-of-condition inspections
and repaired the piping by light grinding. In addition, the licensee used ultrasonic examination
of the repaired areas to confirm acceptable pipe thickness and liquid penetrant testing to
confirm the absence of surface cracks. The licensee installed temporary stainless steel cuffs
directly on the piping at some RMI end cap locations, and placed fiberglass insulation pads
IN 2007-21, Supp. 1 where it could not install the cuffs. As follow-up actions, the licensee planned to consider
installing an improved RMI design and to inspect Catawba, Unit 2, for similar piping wear.
Since the issuance of IN 2007-21, additional licensees have identified instances of RMI fretting
(also referred to as abrasive wear) of nuclear power plant piping in other systems. Examples
include Beaver Valley Power Station Units 1 and 2 Integrated Inspection Report 05000334/2019004 and 05000412/2019004, page 4 (ADAMS Accession No. ML20044D313)
and Arkansas Nuclear One Integrated Inspection Report 05000313/2020002 and
05000368/2020002, page 26 (ADAMS Accession No. 20218A442). In the spring of 2020, the
licensee of Arkansas Nuclear One (ANO), Unit 2, identified multiple wear marks on piping in the
ASME BPV Code Class 1 pressurizer spray line. In some instances, the wear marks were
circumferentially oriented and only detectable by visual examination after RMI removal. The
depth of wear marks ranged from a simple surface scratch to a 25-percent through-wall deep
groove that stretched up to 360 degrees around the circumference of the pipe. See Figures 1 and 2 in the enclosure to this IN. The ANO, Unit 2, licensee determined that fretting wear from
FIV between the ASME BPV Code Class 1 piping and the RMI end caps caused the damage.
At ANO, Unit 2, the licensee confirmed that the RMI end caps had been installed in accordance
with plant procedures and vendor instructions. The licensee repaired the deeply worn areas of
the ASME BPV Code Class 1 piping that did not meet the required minimum wall thickness by
welding to restore the piping to the original construction code. The licensee repaired the other
worn areas of the piping by surface conditioning to meet the original construction code. The
licensee then added stainless steel cuffs, or banding, to protect the Class 1 piping at the ANO,
Unit 2, locations that experienced abrasive wear.
Discussion
The NRC regulations in General Design Criterion 1, Quality standards and records, of
Appendix A, General Design Criteria for Nuclear Power Plants, to Title 10 of the Code of
Federal Regulations (10 CFR) Part 50, Domestic licensing of production and utilization
facilities, require the following:
Structures, systems, and components important to safety shall be designed, fabricated, erected, and tested to quality standards commensurate with the
importance of the safety functions to be performed. Where generally recognized
codes and standards are used, they shall be identified and evaluated to
determine their applicability, adequacy, and sufficiency and shall be
supplemented or modified as necessary to assure a quality product in keeping
with the required safety function.
The ASME BPV Code contains no specific requirements for licensees to remove insulation
periodically to visually inspect piping for RMI wear. For example, the ASME BPV Code
requirements for visual examinations (VT-2) do not require removal of piping insulation.
However, the NRC regulations in 10 CFR Part 50, Appendix A, General Design Criterion 1, include general requirements for the design, fabrication, erection, and testing of nuclear power
plant piping with provisions to supplement codes and standards as necessary to assure a
quality product in keeping with the required safety function.
The NRC staff notes that RMI wear of ASME BPV Code Class 1 piping could result in an
unisolable leak from the reactor coolant pressure boundary in nuclear power plants. For
IN 2007-21, Supp. 1 example, one of the wear marks on the pressurizer spray line at ANO, Unit 2, reached a depth
of 25 percent of the through-wall thickness of the piping. The ANO licensee discovered the
wear in this section of the pressurizer spray line because of its piping inspections in response to
vibration-related failures of snubber connections to the pressurizer spray line. Unchecked, this
type of wear could cause a small break loss of coolant accident and challenge the plant
emergency core cooling systems.
The NRC staff is issuing this supplement to IN 2007-21 to alert licensees to the potential for FIV
to cause RMI wear that could impact the integrity of nuclear power plant piping. Without specific
ASME BPV Code requirements to remove insulation periodically to inspect piping, licensees
might not be aware of ongoing piping wear. Thus, the abrasive wear can continue without
detection and impact the leak tightness and structural integrity of the piping.
To mitigate the wear of affected piping, some licensees have installed temporary stainless steel
cuffs or banding (or, alternatively, fiberglass insulation if cuffs could not be added) directly on
the outer surface of the piping at the RMI end cap locations. This mitigation method provides a
physical barrier that protects the piping from abrasive wear until the insulation can be replaced
with an improved design.
Contacts
This IN requires no specific action or written response. Please direct any questions about this
matter to the technical contacts listed below or to the appropriate Office of Nuclear Reactor
Regulation project manager.
/RA/
Christopher G. Miller, Director
Division of Reactor Oversight
Office of Nuclear Reactor Regulation
Technical Contacts:
Nick Taylor, Region IV
301-415-1328
817-200-1137
Ali.Rezai@nrc.gov
Nick.Taylor@nrc.gov
301-415-2702
301-415-2794 John.Tsao@nrc.gov
Thomas.Scarbrough@nrc.gov
Enclosure:
Figures 1 and 2
Note: NRC generic communications may be found on the NRC public Web site, http://www.nrc.gov, under Electronic Reading Room/Document Collections.
Figure 1. First example of ANO, Unit 2, Pressurizer Spray Line Wear caused by Flow-Induced
Vibration between Piping and RMI End Cap
Figure 2. Second example of ANO, Unit 2, Pressurizer Spray Line Wear caused by
Flow-Induced Vibration between Piping and RMI End Cap
IN 2007-21, Supp. 1 INFORMATION NOTICE 2007-21, SUPPLEMENT PIPE WEAR DUE TO INTERACTION OF
FLOW-INDUCED VIBRATION AND REFLECTIVE METAL INSULATION,
DATE: December 11, 2020
ADAMS Accession Number: ML20225A204 OFFICE
NRR/DEX/EMIB
NRR/DNRL/NPHP
NRR/DNRL/NVIB
QTE
RIV
NAME
ABuford
MMitchell
HGonzalez
JDougherty
NTaylor
DATE
10/01/2020
10/01/2020
10/01/2020
08/13/2020
10/02/2020
OFFICE
NRR/DRO/IOEB
NRR/DRO/IOEB
NRR/DRO/IOEB
DRO:D
NAME
IBetts
MLintz
LRegner
GSuber for CMiller
DATE
11/02/2020
10/02/2020
11/17/2020
12/11/2020
OFFICIAL RECORD COPY