Information Notice No. 90-04: Cracking of the Upper Shell-To-Transition Cone Girth Welds in Steam Generators

                                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 
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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.  

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                                                            IN 90-04 
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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 
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 
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