Information Notice 2001-05, Through-Wall Circumferential Cracking of Reactor Pressure Vessel Head Control Rod Drive Mechanism Penetration Nozzles at Oconee Nuclear Station Unit 3
| ML011160588 | |
| Person / Time | |
|---|---|
| Issue date: | 04/30/2001 |
| From: | Marsh L Operational Experience and Non-Power Reactors Branch |
| To: | |
| References | |
| IN-01-005 | |
| Download: ML011160588 (8) | |
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, D. C. 20555-0001 April 30, 2001 NRC INFORMATION NOTICE 2001-05: THROUGH-WALL CIRCUMFERENTIAL CRACKING
OF REACTOR PRESSURE VESSEL HEAD
CONTROL ROD DRIVE MECHANISM PENETRATION
NOZZLES AT OCONEE NUCLEAR STATION, UNIT 3
Addressees
All holders of operating licenses for pressurized water nuclear power reactors except those who
have 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 to alert
addressees to the recent detection of through-wall circumferential cracks in two of the control
rod drive mechanism (CRDM) penetration nozzles and weldments at the Oconee Nuclear
Station, Unit 3 (ONS3).
It is expected that recipients will review the information for applicability to their facilities and
consider actions, as appropriate. However, suggestions contained in this information notice are
not NRC requirements; therefore, no specific actions or written response is required.
Description of Circumstances
On February 18, 2001, with ONS3 in Mode 5, Duke Energy Corporation (the licensee)
performed a visual examination (VT-2) of the outer surface of the units reactor pressure vessel
(RPV) head to inspect for indications of borated water leakage. This RPV head inspection was
performed as part of a normal surveillance during a planned maintenance outage. The VT-2 revealed the presence of small amounts of boric acid residue in the vicinity of nine of the
69 CRDM penetration nozzles (Figures 1 and 2). Subsequent nondestructive examinations
(NDEs) identified 47 recordable crack indications in these nine degraded CRDM penetration
nozzles. The licensee initially characterized these flaws as either axial or below-the-weld
circumferential indications, and initiated repairs of the degraded areas. NDEs of nine additional
CRDM penetration nozzles from the same heat of material were conducted for extent of
condition purposes, but did not detect recordable indications.(1)
(1)
Axial flaws are flaws that propagate along the inside or outside diameter length of the
CRDM nozzle. Below-the-weld circumferential indications are apparent flaws oriented
around the circumference of the nozzle, beneath the RPV head and below the area where
the nozzle is welded to the RPV head. A recordable indication is one that exceeds the NDE
acceptance criteria. Subsequent dye-penetrant testing (PT) revealed additional indications in two of the nine
degraded penetration nozzles. While affecting further repairs of these indications, the licensee
identified that each nozzle had significant circumferential cracks in the nozzle above the weld.
Further investigations and metallurgical examinations revealed that these cracks had initiated
from the outside diameter (OD) of the CRDM penetration nozzles. The circumferential crack in
the #56 CRDM nozzle was through-wall, and the #50 nozzle had pin hole through-wall
indications. These cracks followed the weld profile contour, and were nearly 165° in length.
The licensee stated that pre-repair ultrasonic testing (UT) examinations had identified
indications in these areas during the initial inspections, but these indications had been
misinterpreted as craze cracking with unusual characteristics. The characterization for these
two nozzle indications was revised after the initial post-repair PT examinations. The licensee
concluded that the root cause for the CRDM penetration nozzle cracking was primary water
stress corrosion cracking (PWSCC). This conclusion was based on metallurgical
examinations, crack location and orientation, and finite element analyses.
Discussion
The 69 CRDM nozzles at ONS3 are approximately 5 feet long and are J-groove welded to the
inner radius of the RPV head, with the lower end of each nozzle extending about 6 inches
below the inside of the RPV head (see Figure 2). The nozzles are constructed from 4-inch OD
Alloy 600 Inconel procured in accordance with the requirements of Specification SB-167, Section II to the 1965 Edition (including Addenda through Summer 1967) of the American
Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code. During initial
construction, each nozzle was machined to final dimensions to assure a match between the
RPV head bore and the OD of the nozzle. The nozzles were shrink-fit by cooling to at least
minus 140 degrees F, inserted into the closure head penetration, and then allowed to warm to
room temperature (70 degrees F minimum). The CRDM nozzles were tack-welded and then
permanently welded to the closure head using 182-weld metal (see Figure 2). The shielded
manual metal arc welding process was used for both the tack weld and the J-groove weld.
During weld buildup, the weld was ground and PT inspected at each 9/32 inch of the weld.
The final weld surface was ground and PT inspected. The weld prep for installation of each
nozzle in the RPV head was accomplished by machining and buttering the J-groove with 182- weld metal.
Axial cracking in pressurized water reactor (PWR) CRDM nozzles has been previously
identified, evaluated, and repaired. Numerous small-bore Alloy 600 nozzles and pressurizer
heater sleeves have experienced leaks attributed to PWSCC. Generally, these components
are exposed to temperatures of 600 degrees F or higher and to primary water, as are the
ONS3 CRDM nozzles. However, circumferential cracks above the weld from the OD to the
inside diameter (ID) have not been previously identified in the U.S.
An action plan was implemented by the NRC staff in 1991 to address PWSCC of Alloy 600
vessel head penetrations (VHPs) at all U.S. PWRs. This action plan included a review of the
safety assessments by the PWR owners groups (Westinghouse Owners Group, Combustion Engineering Owners Group, and Babcock & Wilcox Owners Group) submitted for staff review
on June 16, 1993, by the Nuclear Management and Resource Council (NUMARC, now the
Nuclear Energy Institute [NEI]).
After reviewing the industry's safety assessments and examining the overseas inspection
findings, the NRC staff concluded, in a safety evaluation (SE) dated November 19, 1993, that
PWR CRDM nozzle and weld cracking was not an immediate safety concern. The bases for
this conclusion were that if PWSCC occurred (1) the cracks would be predominately axial in
orientation, (2) the axial cracks would result in detectable leakage before catastrophic failure, and (3) the leakage would be detected during visual examinations performed as part of
surveillance walkdown inspections before significant damage to the RPV head would occur.
However, the NRC staff noted concerns about potential circumferential cracking (which would
need to be addressed on a plant-specific basis), high residual stresses from initial manufacture
and from tube straightening sometimes done after welding, and the need for enhanced
leakage monitoring.
By letter dated March 5, 1996, NEI submitted a white paper entitled "Alloy 600 RPV Head
Penetration Primary Stress Corrosion Cracking," which reviewed the significance of PWSCC in
PWR VHPs, described how the PWR licensees were managing the issue. NEI assumed that
the issue was primarily an economic issue rather than a safety issue, and described an
economic decision tool to be used by PWR licensees to evaluate the probability of a VHP
developing a crack or a through-wall leak during a plant's lifetime. This information would then
be used by a PWR licensee to evaluate the need to conduct a VHP inspection at their plant.
To verify the conclusions in the industrys safety assessments, sampling inspections were
performed at three PWR units in 1994. The results of these domestic inspections were
consistent with the February 1993 analyses by the PWR owners groups, the staffs November
19, 1993, SE, and the PWSCC found in European reactors. On the basis of the results of the
first five inspections of U.S. PWRs, the PWR owners groups' analyses, and the European
experience, the NRC staff determined that it was probable that CRDM penetrations at other
plants contained similar axial cracks, but that such cracking did not pose an immediate- or
near-term safety concern. Further, the NRC staff recognized that the scope and timing of
inspections may vary for different plants, depending on their individual susceptibility to this
form of degradation. In the long term, however, the staff determined that degradation of the
CRDM and other RPV head penetrations is an important safety consideration because of the
possibility of (1) exceeding the ASME Code safety margins if the cracks are sufficiently deep
and continue to propagate during subsequent operating cycles and (2) eliminating a layer of
defense in depth for plant safety.
On April 1, 1997, NRC issued Generic Letter (GL) 97-01, Degradation of Control Rod Drive
Mechanism Nozzle and Other Vessel Closure Head Penetrations, which requested
addressees to inform the staff of their inspection activities related to VHPs. Based on the
industrys GL 97-01 response, which took credit for periodic inspections of the RPV head, the
staff agreed that the conclusions in its November 19, 1993, SE remained valid. The recent identification of significant circumferential cracking of two CRDM nozzles at ONS3 raises concerns about a potentially risk-significant generic condition affecting all domestic
PWRs. RPV head penetrations, including CRDM nozzles, provide the function of maintaining
the reactor coolant system (RCS) pressure boundary. Cracking of CRDM nozzles and welds
is a degradation of the primary RCS boundary. Industry experience has shown that Alloy 600
is susceptible to stress corrosion cracking (SCC). Further, the environment in the CRDM
housing annulus will likely be far more aggressive after any through-wall leakage, because
potentially highly concentrated borated primary water will become oxygenated, increasing
crack growth rates.
The repair activities at ONS3 were extensive. The licensee stated that all flaws would be
removed entirely from both weld material and nozzle base metal and repaired prior to plant
restart. The licensee plans to perform a thorough visual inspection of the Unit 2 RPV head
penetrations during the next outage and is investigating the eventual replacement of the RPV
heads on all three units to prevent recurrence of this event. Foreign PWRs in France and
Japan have already replaced a number of their RPV heads.
The NRC held a public meeting with the Electric Power Research Institute (EPRI) Materials
Reliability Project (MRP) personnel on April 12, 2001, to discuss CRDM nozzle circumferential
cracking issues. During the meeting, the industry representatives said that they were
developing a generic safety assessment, recommendations for revisions of near-term
inspections, and long-term inspection and flaw evaluation guidelines.
The ONS3 cracking reinforces the importance of examining the upper PWR RPV head area
(e.g., visual under-the-insulation examinations of the penetrations for evidence of borated
water leakage or volumetric examinations of the CRDM nozzles) and of using appropriate NDE
methods (e.g., UT, ET, PT, etc.) to adequately characterize cracks. Presently, licensees are
not required to remove RPV head insulation to visually inspect the head penetrations;
however, some licensees have recently performed expanded VT-2 examinations by using
cameras to inspect between the CRDM nozzles and the insulation.
The NRC has recently developed a Web page to keep the public informed of generic activities
on PWR Alloy 600 weld cracking (http://www.nrc.gov/NRC/REACTOR/MRP/index.html). The
NRC will update this Web page and assess the need for further generic action as new
information becomes available.
This information notice requires no specific action or written response. If you have any
questions about the information in this notice, please contact one of the technical contacts
listed below or the appropriate Office of Nuclear Reactor Regulation (NRR) project manager. Related Generic Communications
+ Generic Letter 88-05, Boric Acid Corrosion of Carbon Steel Reactor Pressure
Boundary Components in PWR Plants, March 17, 1988
+ Generic Letter 97-01, Degradation of Control Rod Drive Mechanism Nozzle and Other
Vessel Closure Head Penetrations, April 1, 1997
+ Information Notice 90-10, Primary Water Stress Corrosion Cracking of INCONEL 600,
February 23, 1990
+ Information Notice 96-11, Ingress of Demineralizer Resins Increases Potential for
Stress Corrosion Cracking of Control Rod Drive Mechanism Penetrations, February
14, 1996
+ NUREG/CR-6245, Assessment of Pressurized Water Reactor Control Rod Drive
Mechanism Nozzle Cracking, October 1994
/RA/
Ledyard B. Marsh, Chief
Events Assessment, Generic Communications
and Non-Power Reactors Branch
Division of Regulatory Improvement Programs
Office of Nuclear Reactor Regulation
Technical contacts: Ian Jung, NRR James Medoff, NRR
301-415-1837 301-415-2715 E-mail: ixj@nrc.gov E-mail: jxm@nrc.gov
C. E. (Gene) Carpenter, NRR
301-415-2169 E-mail: cec@nrc.gov
Attachments:
1. Figure 1: Oconee Reactor Pressure Vessel Head Map
2. Figure 2: Oconee CRDM Nozzle Penetration (Typical)
3. List of Recently Issued NRC Information Notices
ML011160588 Temple #=NRR-052
ÿ Publicly Available ÿ Non-Publicly Available ÿ Sensitive ÿ Non-Sensitive
OFFICE REXB Tech Editor EMCB EMCB REXB C:REXB
NAME IJung* PKleene* JMedoff* CCarpenter* JTappert/txk*1. LMarsh
DATE 04/25/2001 04/25/2001 04/26/2001 04/25/2001 04/26/2001 04/30/2001
Figure 1: Oconee Reactor Pressure Vessel Head Map
Figure 2:Oconee CRDM Nozzle Penetration (Typical)
Attachment 3 LIST OF RECENTLY ISSUED
NRC INFORMATION NOTICES
_____________________________________________________________________________________
Information Date of
Notice No. Subject Issuance Issued to
______________________________________________________________________________________
2001-04 Neglected Fire Extinguisher 04/11/01 All holders of licenses for nuclear
Maintenance Causes Fatality power, research, and test
reactors and fuel cycle facilities
2001-03 Incident Reporting 04/06/01 All industrial radiography
Requirements for Radiography licensees
Licensees
2001-02 Summary of Fitness-for-Duty 03/28/01 All holders of operating licenses
Program Performance Reports for nuclear power reactors, and
for Calendar Years 1998 and licensees authorized to possess
1999 or use formula quantities of
strategic special nuclear material
(SSNM) or to transport formula
quantities of SSNM
2001-01 The Importance of Accurate 03/26/01 All material licensees
Inventory Controls to Prevent
the Unauthorized Possession
of Radioactive Material
2000-17, Crack in Weld Area of Reactor 02/28/01 All holders of operating licenses
Supp. 2 Coolant System Hot Leg Piping for nuclear power reactors except
at V.C. Summer those who has ceased operations
and have certified that fuel has
permanently removed from
reactor vessel
2000-22 Medical Misadministrations 12/18/00 All medical use licensees
Caused by Human Errors authorized to conduct gamma
Involving Gamma Stereotactic stereotactic radiosurgery
Radiosurgery (GAMMA KNIFE) treatments
2000-21 Detached Check Valve Disc 12/15/00 All holders of OLs for nuclear
not Detected by Use of power reactors except those who
Acoustic and Magnetic have ceased operations and have
Nonintrusive Test Techniques certified that fuel has been
permanently removed from the
reactor
______________________________________________________________________________________
OL = Operating License
CP = Construction Permit