Information Notice 1990-04, Cracking of the Upper Shell-to-Transition Cone Girth Welds in Steam Generators,
| ML031470418 | |
| Person / Time | |
|---|---|
| Issue date: | 01/26/1990 |
| From: | Rossi C Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| IN-90-004 | |
| Download: ML031470418 (3) | |
Information Notice No. 90-04: Index Site Map I FAQ I Help I Glossary Contact Us l-IuISeaCh A
U.S. Nuclear Regulatory Commission E
HomeWhatNuclear
WhoWeRAre Nuclear Radioactive ll Public
I 0 fl Materials
11 ~ ~~~~Reactors Waste Involvement
Home > Electronic Reading Room > Document Collections > General Communications > Information Notices > 1990 > IN 9 UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, D.C. 20555 January 26, 1990
Information Notice No. 90-04: CRACKING OF THE UPPER SHELL-TO-TRANSITION
CONE GIRTH WELDS IN STEAM GENERATORS
Addressees
All holders of operating licenses or construction permits for Westinghouse- designed and Combustion Engineering-designed nuclear power reactors.
Purpose
This information notice is intended to alert addressees to continuing
problems related to cracking of the upper shell-to-transition cone girth
welds in the steam generators (SGs) originally described in Information
Notices 82-37, Cracking in the Upper Shell to Transition Cone Girth Weld of
a Steam Generator at an Operating Pressurized Water Reactor, and 85-65,
'Crack Growth in Steam Generator Girth Welds., It is expected that
recipients will review the in-formation for applicability to their facilities
and consider actions, as appropriate, to avoid similar problems. However, suggestions contained in this information notice do not constitute NRC
requirements; therefore, no specific action or written response is required.
Description of Circumstances
During the 1989 refueling outage at Zion Unit 1, a scheduled inservice
inspection (ISI) was performed on the SG D- upper shell-to-transition cone
girth weld. The ultrasonic testing (UT) detected flaw indications that
exceeded the allowable standard of Section XI of the ASME Code, Article
IWC-3000 (Table IWB-3511-1). Based upon these results, the extent of UT was
initially expanded to include the girth weld in SG C' and further expanded
to include SGs "A' and B.' All surface indications were removed by
grinding, contoured to established profiles, and accepted by magnetic
particle testing (MT) methods. The deepest repair excavation was
approximately 0.50 inch in depth by 6.45 inches in length. Boat samples
were removed for metallography. The results of the metallography are still
under investigation by the licensee.
During the 1987 refueling outage at Indian Point Unit 2, flaw indications
were detected during a scheduled ISI of the same upper shell-to-transition
cone girth weld. Visual examination of the inside circumference revealed
essentially horizontal intermittent linear indications around the entire
weld length of SG #22. Subsequently, UT and MT were extended to essentially
100 percent of this girth weld in all SGs. A total of 291 surface
indications were reported in the four SGs, with the most severe cracking
occurring in SG #22. The linear indications were predominantly in the
vicinity of the weld heat-affected zones.
9001220165 IN 90-04 January 26, 1990 http://www.nrc.gov/reading-rm/doc-collections/gen-comm/info-notices/1 990/in90004.html 03/13/2003
tnformation Notice No. 90-04: A repair program was completed that included progressive grinding to
established profiles and nondestructive examination. All observed cracks
detected by MT were removed; however, the corrosion pits outside the repair
areas were not removed before the plant started up after the refueling
outage. The repair resulted in a series of grooves that extended around
essentially the entire circumference of SG #22 with the maximum depth of
excavation approximately 1.07 inch, whereas the wall thickness is typically
3.5 inches. Eight boat samples were removed for metallurgical analysis. On
the basis of this analysis, the licensee concluded that the cracking was
most likely caused by corrosion fatigue.
During the 1989 refueling outage at Indian Point Unit 2, an MT was initially
conducted on one third of the inside circumference of the SG #22 girth weld.
Linear indications were detected during this examination. Subsequently, 100
percent of the inside circumferences of the girth welds in all SGs were
inspected. Linear indications were also detected in these additional
examinations. All observed cracks were ground out again; the maximum depth
of grinding to remove the new flaw indications was 0.95 inch. A weld repair
of localized areas and a post-weld heat treatment PWHT) were accomplished
on SG #22. An MT performed after the PWHT detected additional surface
indications, which were later removed. The licensee concluded that the
probable cause of the cracking was corrosion fatigue resulting from the
combined action of thermal cycling, oxygen in the auxiliary feedwater, and
copper alloys from the feedwater system. The licensee removed the downcomer
flow resistance plate to minimize the thermal cycling mechanism. The
licensee also committed to shutdown for an MT inspection during a mid-cycle
outage to evaluate the effectiveness of corrective actions.
Discussion:
Cracks and linear indications on the inner circumference have been detected
in the upper shell-to-transition cone girth weld in 18 SGs in the United
States. In addition, linear indications have been found at one foreign
plant. The degree of cracking ranges from severe in the case of Indian
Point Unit 2 to isolated and dispersed at Zion Unit 1. At the domestic
plants flaws have been observed only in Westinghouse Model 44 and Model 51 vertical recirculating U-tube SGs with the feedwater ring design.
The manufacturer, the affected licensees, and the NRC staff are still
evaluating the available information to establish the root cause of the
cracking problem and its generic implication. A common factor was the
general corrosion pitting on the inside surface of the SGs. Metallography
found that the surface pits served as crack initiation sites. The current
information indicates that the degradation probably results from
corrosion-assisted thermal fatigue. Thermal cycling results from relatively
cold water that impinges upon the weld region during reactor trips from full
power and certain transient operations. At Indian Point Unit 2, copper
alloys from the feedwater system and the downcomer flow resistance plate
probably were contributing factors.
IN 90-04 January 26, 1990 The flaw indications can be detected with enhanced UT procedures that are
performed by experienced nondestructive examination personnel. The upper
shell-to-transition cone weld is located at a gross structural
discontinuity. The weld is relatively wide and typically has an irregular
crown. These inherent geometric features commonly result in innocuous
reflectors. In addition, subsurface flaw indications are known to exist
near the inside diameter surface of SGs at several plant sites. In order to
distinguish innocuous reflectors from cracks, the following processes may be
necessary: scanning at a high gain, the use of multiple transducers with
optimum angles, careful plotting of reflector locations, and examination by
experienced personnel.
The rules of Section XI of the ASME Code require a volumetric examination of
one upper shell-to-transition cone weld during each 10-year inspection
interval. The required examinations may be limited to one SG or may be
distributed among all the SGs. However, if general corrosion pitting of the
http://www.nrc.gov/reading-rmldoc-collections/gen-commlinfo-notices/1990/in90004.html 03/13/2003
Information Notice No. 90-04: SG shell is known to exist, the requirements of Section XI of the ASME Code
may not be sufficient to differentiate isolated cracks from inherent
geometric conditions. In lieu of volumetric examinations, visual and MT
examinations of the interior circumference of the girth weld were used by
the licensee of Indian Point Unit 2 to detect the surface-connected flaws.
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 NRR project
manager.
Charles E. Rossi, Director
Division of Operational Events Assessment
Office of Nuclear Reactor Regulation
Technical Contacts: Martin R. Hum, NRR
(301) 492-0932 Robert A. Hermann, NRR
(301) 492-0911 Attachment: List of Recently Issued NRC Information Notices
.ENDEND
http://www.nrc.govlreading-rmldoc-collections/gen-commlinfo-notices/1990/in90004.html 03/13/2003