Information Notice 2015-04, Fatigue in Branch Connection Welds
| ML15023A054 | |
| Person / Time | |
|---|---|
| Issue date: | 04/23/2015 |
| From: | Michael Cheok, Kokajko L Office of New Reactors, Division of Policy and Rulemaking |
| To: | |
| Popova A | |
| References | |
| IN-15-004 | |
| Download: ML15023A054 (6) | |
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
OFFICE OF NEW REACTORS
WASHINGTON, DC 20555-0001 April 24, 2015 NRC INFORMATION NOTICE 2015-04: FATIGUE IN BRANCH CONNECTION WELDS
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 who 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 certification, 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 about recent operating experience related to the structural integrity of recirculation
system piping in boiling-water reactors (BWR) and to raise industry awareness regarding the
possibility of emerging fatigue cracking in branch connections in all light-water reactors. The
NRC expects that recipients will review the information contained in this IN for applicability to
their facilities and consider actions, as appropriate, to avoid similar issues at their facilities.
However, suggestions contained in this IN are not NRC requirements; therefore, no specific
action or written response is required.
DESCRIPTION OF CIRCUMSTANCES
The licensee event reports (LERs) detailed below represent three examples of fatigue failures in
recirculation system piping, specifically in full-penetration groove welds in branch connections.
Similar incidences occurred in recirculation system instrument lines in BWR units due to
vibrations and were addressed in IN 95-16, Vibration Caused by Increased Recirculation Flow
in a Boiling Water Reactor, dated March 9, 19951.
1 Available in the Agencywide Documents Access and Management System (ADAMS) at Accession
No. ML031060310.
ML15023A054 Hope Creek Nuclear Generating Station (2001)
On October 10, 2001, while performing a primary containment-walk down at the beginning of a
refueling outage at Hope Creek Nuclear Generating Station, the licensee (PSEG Nuclear, LLC)
observed a leak on the elbow tap of the A recirculation pump suction piping. The leak was
producing a 3-4 foot long spray with the reactor vessel pressure at approximately 300-400 psig.
Further investigation revealed that the leak was coming from the weld area of a 1-inch
pipe-to-branch connection weld and was, therefore, a through-wall breach of the reactor coolant
system pressure boundary. The licensee identified that a vibration-induced weld failure caused
the leak. According to the LER, the vibration-induced failure was most likely caused by the
second natural frequency of the piping becoming resonant with the 5X vane-passing frequency
of the "A" recirculation pump. Additionally, the licensee found that accelerometers, which had
been installed in 1991, added weight to the piping and altered its original frequency.
As a result of this event, the licensee: (1) removed the accelerometers, (2) performed
walk-downs of all recirculation lines for indication of other failures and inspected equipment
around the area of the leak to ensure no damage from leak impingement, (3) performed
radiographic examinations on other lines and penetrant tests on all other susceptible welds (to
determine extent of condition), and (4) removed the cracked weld and affected section of pipe.
Additional information is available in LER-50-354/2001-006-00, dated December 7, 2001 and
can be found on the NRCs public website in ADAMS under Accession No. ML020220237.
Hope Creek Nuclear Generating Station (2005)
On March 27, 2005, PSEG Nuclear, LLC identified a steam leak at the Hope Creek Nuclear
Generating Station. The steam leak was coming from a 4-inch crack in the insulated
decontamination port, which is connected to the B reactor recirculation loop between the
suction of the pump and the suction isolation valve. The decontamination port is 4-inch SA-376 Type 304 stainless steel piping. The fatigue crack started near the outside diameter of the
pipe-to-branch connection weld.
The through-wall crack in the decontamination port was caused by fatigue initiation and
propagation. The licensee determined that an original subsurface weld defect propagated due
to high cycle vibrations. Analysis revealed that, based on the geometry of the decontamination
port, the natural frequencies of the B decontamination port coincided with the 5X vane-passing
frequency of the reactor recirculation pump at normal operating speeds. The natural frequency
of the decontamination port ranged from 122-135 Hertz (Hz), and the 5X vane-passing
frequency of the reactor recirculation pump at normal operating speeds was 120-125 Hz.
Because of this event, the licensee modified the decontamination ports for both recirculation
loops. The length of 4-inch decontamination port was reduced to increase its natural frequency
in order to not coincide with the normal recirculation pump 5X vane-passing frequency.
Additionally, the licensees review consisted of: finite element analysis, vibration analysis, modal analysis, isometric review, and review of previous in-service inspections (ISIs). Based on
these reviews and as part of its extent of condition determination, the licensee inspected 23 welds using nondestructive examination techniques, and the inspections were satisfactory.
Additional information is available in LER-50-354/2005-002-00, dated May 26, 2005, and can be
found on the NRCs public website in ADAMS under Accession No. ML051540027. Susquehanna Steam Electric Station, Unit 1 (2012)
On June 19, 2012, the licensee (PPL Susquehanna, LLC) for the Susquehanna Steam Electric
Station Unit 1 identified the source of an increasing trend in drywell leakage to be an
approximately 3-inch long crack. The crack was in the weld of the branch connection to the
4-inch diameter chemical decontamination pipe to the 1A" reactor recirculation pump suction
line.
Metallurgical examinations revealed fatigue caused the crack. The root cause was that the
calculated stresses were underestimated by use of an incorrect stress intensification factor (1.1 vs. 1.8) in the vibration analysis. The calculated vibration stress for the 4-inch decontamination
line connection, considering the maximum extended power uprate vibration and stress data, exceeded the endurance limit by approximately 26 percent. Considering the primary frequency
of vibration was 128.5 Hz, the fatigue life was approximately 4.9 years when exposed to the 5X
recirculation pump vane-passing frequency, although the original vibration analysis yielded an
infinite life. The normal operating speed range for the reactor recirculation pumps correlated to
a 5X vane-passing frequency of 122-135 Hz. The configuration of the reactor recirculation
pump suction line decontamination flange was such that its natural frequency was 135.8 Hz
based on ambient temperature conditions. However, at operating temperature the natural
frequency was approximately 129 Hz. Because the 5X vane-passing frequency of the reactor
recirculation pumps was similar to the natural frequency of the decontamination flange, the
vibrational accelerations were magnified. This led to relatively large bending moments and
stresses at the branch connection (exceeding the endurance limit of the material) and resulted
in fatigue failure. The crack was not caused by intergranular stress corrosion cracking (IGSCC);
however, indications of IGSCC were found unrelated to the fatigue crack.
Because of this event, the licensee changed the natural frequency of the assembly by
redesigning the 4-inch diameter chemical decontamination flange connection on both the "1A"
and "1B" reactor recirculation pump suction lines. The length of the 4-inch diameter pipe was
reduced from 6-inch to approximately 3.5-inch to produce a new configuration that was not
susceptible to the cyclic fatigue caused by reactor recirculation pump 5X vane-passing
frequency similarity to the natural frequency of the assembly. This modification also removed
the sections with IGSCC, which were identified during metallurgical examinations. The modified
decontamination flange connections were pressure tested (1035 psig) to ensure pressure
boundary integrity. In addition, the licensee performed extent of condition inspections of similar
reactor recirculation and reactor water cleanup system piping. Vibration related issues, additional fatigue flaws, or IGSCC were not identified. At the time of the event, the licensee
determined that the Unit 2 recirculation system piping was capable of performing its required
design function. The licensee planned for similar actions for Unit 2 during the next refueling
outage in order to prevent recurrence.
Additional information is available in LER-50-387/2012-007-01, dated November 20, 2012, and
can be found on the NRCs public website in ADAMS under Accession No. ML123250703.
BACKGROUND
Related NRC Generic Communications
NRC IN 2005-08, Monitoring Vibration to Detect Circumferential Cracking of Reactor Coolant
Pump and Reactor Recirculation Pump Shafts, dated April 5, 2005. The NRC issued this IN to
alert addressees to the importance of timely detection of circumferential cracking of reactor coolant pump and reactor recirculation pump shafts to minimize the likelihood of consequential
shaft failures.
NRC IN 2005-23, Vibration-Induced Degradation of Butterfly Valves, dated August 1, 2005.
The NRC issued this IN to alert addressees to the vibration-induced degradation (loss of taper
pins) of butterfly valves.
NRC IN 2006-15, Vibration-Induced Degradation and Failure of Safety-Related Valves, dated
July 27, 2006. The NRC issued this IN to alert addressees of vibration-induced degradation and
failure of valves.
NRC IN 2007-21, Pipe Wear Due to Interaction of Flow-Induced Vibration and Reflective Metal
Insulation, dated June 11, 2007. The NRC issued this IN to alert addressees that a licensee
identified significant wear marks on the outside wall of the chemical volume control system
piping, which was subject to flow-induced vibration conditions.
DISCUSSION
The above operating experiences discuss that unexpected fatigue failures could occur in branch
connection welds during normal operating conditions. These failures were caused by the
affected piping becoming resonant with the reactor recirculation pump 5X vane-passing
frequency. In addition, since fatigue failure is progressive, the affected plants ISI programs
failed to identify the problems before actual failure took place.
Similar weld fatigue failures could be minimized by designing piping with natural frequencies
that avoid pump vane-passing frequencies and by selecting welds that are vulnerable to fatigue
failure for examination under the ISI program. In addition to NRC requirements, some licensees
have chosen to implement programs for monitoring fatigue in branch connection welds.
CONTACT
S
This information notice does not require any specific action or written response. Please direct
any questions about this matter to the technical contacts listed below.
/RA/ /RA/
Michael C. Cheok, Director Lawrence E. Kokajko, Director
Division of Construction Inspection Division of Policy and Rulemaking
and Operational Programs Office of Nuclear Reactor Regulation
Office of New Reactors
Technical Contacts: Ganesh Cheruvenki, NRR/DE/EVIB
301-415-2501 E-mail: Ganesh.Cheruvenki@nrc.gov
Varoujan Kalikian, NRR/DLR/RARB
301-415-5590
E-mail: Roger.Kalikian@nrc.gov
Alfred Issa, NRO/DCIP/IGCB
301-415-5342 E-mail: Alfred.Issa@nrc.gov
ML15023A054 *via e-mail TAC MF5566 OFFICE NRR/DE/EVIB* NRR/DLR/RARB* NRO/DCIP/IGCB* QTE* NRR/DE/EPNB*
NAME GCheruvenki RKalikian AIssa DAlley
DATE 03/02/2015 02/27/2015 03/11/2015 02/18/2015 03/09/2015 OFFICE NRR/DLR/RARB* NRO/DE/MCB* NRO/DCIP/IGCB* RES/DE/CIB* NRR/DE/EVIB*
NAME DMorey MMitchell BAnderson DRudland SRosenberg
DATE 03/03/2015 03/06/2015 03/11/2015 03/03/2015 03/09/2015 OFFICE NRR/DE* NRR/DPR/PGCB* NRR/DPR/PGCB* NRR/DPR/PGCB* NRO/DCIP
NAME MJRoss-Lee APopova ELee SStuchell MCheok
DATE 03/31/2015 04/01/2015 04/02/2015 04/07/2015 04/09/2015 NRR/DPR NRR/DPR
OFFICE
NAME AMohseni LKokajko
DATE 4/22/2015 4/23/2015