United States Nuclear Regulatory Commission - Protecting People and the Environment

Information Notice No. 90-10: Primary Water Stress Corrosion Cracking (PWSCC) of Inconel 600

                                UNITED STATES
                        NUCLEAR REGULATORY COMMISSION
                    OFFICE OF NUCLEAR REACTOR REGULATION
                           WASHINGTON, D.C.  20555

                              February 23, 1990


Information Notice No. 90-10:  PRIMARY WATER STRESS CORROSION CRACKING
                                   (PWSCC) OF INCONEL 600


Addressees:

All holders of operating licenses or construction permits for pressurized 
water reactors (PWRs).

Purpose:

This information notice is intended to alert addressees to potential 
problems related to primary water stress corrosion cracking (PWSCC) of 
Inconel 600 that has occurred in pressurizer heater thermal sleeves and 
instrument nozzles at several domestic and foreign PWR plants.  It is 
expected that recipients will review the information 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 Calvert Cliffs Unit 2 (CC-2), visual 
examination detected leakage in 20 pressurizer heater penetrations and 1 
upper-level/pressure tap instrument nozzle.  Leakage was indicated by the 
presence of boric acid crystals at the penetrations and on the nozzle.  Non-
destructive examinations (liquid penetrant and eddy current examinations) 
were performed on 28 thermal sleeves and 3 instrument nozzles.  Crack indi-
cations were reported in 24 thermal sleeves, including the 20 originally 
identified to be leaking as well as the leaking nozzle.  No crack 
indications were found in the two lower instrument nozzles.  The 
examinations showed that all cracks in the sleeves and the nozzle were 
axially oriented with a maximum length not greater than 10.5 inches.  The 
mode of failure for the thermal sleeves was identified as PWSCC.

The heater sleeves and the instrumentation nozzles were made of Inconel 600 
tubing and bar materials, respectively, supplied by INCO.  All thermal 
sleeves were made in a high strength heat (NX8878) with a reported yield 
strength of 63.5 ksi.  No chemical contaminants were found on the sleeve 
fracture surfaces.  A review of the fabrication records showed that all 120 
thermal sleeves in CC-2 were reamed 3.5 inches from the top before welding 
and all but two were also 





9002160093 
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                                                       IN 90-10
                                                       February 23, 1990
                                                       Page 2 of 4


reamed after welding to facilitate the insertion of the heater rods.  All
cracks in the sleeves were reported to be located inside the reamed region 
either above or below the J-groove weld.

All instrument nozzles were made from heat no. NX8297 and its yield strength 
was reported to be 36 ksi.  The licensee indicated that four upper 
instrument nozzles, including the leaking one, had been extensively reworked 
when the faulty J-groove welds were repaired.  Based on the results of the 
investigations, the licensee, Baltimore Gas and Electric (BG&E), postulated 
that the leaking in the pressurizer thermal sleeves and the instrumentation 
nozzle was due to PWSCC.  The licensee is in the process of removing a 
metallurgical sample from the leaking instrument nozzle for failure analysis 
to identify the mode of failure.

On February 27, 1986, a small leak (about 0.15 gpm) was observed on a 
3/4-inch-diameter upper pressurizer level instrument nozzle at San Onofre 
Nuclear Generating Station (SONGS) Unit 3 while the plant was in hot 
standby.  An axial flaw about 5/8 inch in length was identified on the 
inside diameter surface of the nozzle.  The flaw appeared to originate from 
the end of the nozzle inside the pressurizer and extended beyond the 
attachment weld (1/2 inch in depth) approximately 1/8 inch into the annulus 
area of the nozzle assembly.  The flawed nozzle was cut out, including the 
entire attachment weld.  The results of the metallurgical examination 
performed on the flawed nozzle assembly indicated that the cracking was 
PWSCC.

In spring 1989, leakage from instrument nozzles was observed in two foreign 
PWRs (one from each 1300-MW plant) when the hydrostatic pressure testing of 
the primary system was performed during the first refueling outage.  The 
instrument nozzles were made of Inconel 600 material.  The installation of 
the nozzles included mechanically rolling a portion of the nozzle into the 
pressurizer shell.  Nondestructive examinations (NDEs) were performed on the 
leaking nozzles and found the cracks to be principally axial in orientation; 
however, some circumferential cracking was observed.  Destructive 
examination of these two leaking nozzles to identify the root causes has not 
been completed.  Additional NDEs were performed on all the instrument 
nozzles of five 1300-MW PWRs.  Crack indications were found in 12 instrument 
nozzles.

Discussion:

Extensive laboratory testing has shown that intergranular stress corrosion 
cracking (IGSCC) requires the presence of the following three key elements:  
an aggressive environment, susceptible material, and sufficient tensile 
stresses for crack initiation and propagation.  PWSCC refers to IGSCC in the 
primary water environment of PWRs.  The laboratory demonstration of PWSCC in 
Inconel 600 was first reported by Coriou almost 30 years ago.  The studies 
of PWSCC in Inconel 600 have been documented in numerous published reports.  
However, the mechanism for PWSCC in Inconel 600 is still not well 
understood.  In PWRs, PWSCC of Inconel 600 was first reported in steam 
generator tubing.

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                                                       IN 90-10
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The cracking to date in the thermal sleeves and the instrument nozzles of 
the domestic PWRs has been reported as being only axially oriented.  The 
safety implication of an axial crack is not considered a significant threat 
to the structural integrity of the components and most likely will result in 
a small leak.  However, limited circumferential cracking was reported in the 
instrument nozzles of several foreign PWRs.  The difference in the cracking 
morphology has been attributed to the different type of mechanical working 
(rolling vs. reaming) being performed on these nozzles and thermal sleeves.  
The appearance of the crack in the rolled instrument nozzle is consistent 
with that observed in the roll-expanded region of steam generator tubing.  
Circumferential cracking poses a more serious safety concern because if it 
were to go undetected it could lead to a structural failure of a component 
rather than to a limited leak.

The reported cracking of the pressurizer thermal sleeves and the instrument 
nozzle at Calvert Cliffs Unit 2 and the instrument nozzle at SONGS Unit 3 
are the most current PWSCC events for Inconel 600 in domestic PWRs besides 
the cracking problem associated with the steam generator tubing and plugs.  
The pressurizer thermal sleeves in Calvert Cliffs Unit 1 (CC-1) were also 
made of the same heat of susceptible material, but the recent inspection of 
the CC-1 pressurizer did not reveal any leaking or cracking of the thermal 
sleeves.  The licensee indicated that the major difference in the 
fabrication of thermal sleeves between CC-1 and CC-2 is that the 
pre-installation reaming operation was not performed on CC-1 sleeves.  The 
Combustion Engineering Owners Group (CEOG) performed an evaluation of 
pressurizer heater sleeve susceptibility to PWSCC for plants designed by 
Combustion Engineering.  The CEOG recommended a visual inspection program 
for the thermal sleeves.  The inspection program for the thermal sleeves 
varied, depending on the degree of susceptibility of the sleeve materials.  
The sleeve susceptibility was rated by the elements described above.  The 
staff notes that at CC-2 the yield strength of the thermal sleeve material 
is higher than that of the instrument nozzle material.  However, PWSCC 
occurred in both heats of materials.  This circumstance may indicate that 
the yield strength of the material is not necessarily a reliable screening 
criterion for PWSCC susceptibility.  The CEOG is performing additional work 
to address PWSCC of Inconel 600.  The CEOG programs include the following 
activities:  evaluations to gain better understanding of the cracking 
mechanism in pressurizer thermal heater sleeves and instrument nozzles; an 
analytical determination of a temperature profile for the heater sleeves; 
review of the fabrication history of all Inconel 600 penetrations in the 
primary system components; conduct of a test that is primarily a system 
leakage test on a mock-up of the flawed components; and improvement of NDE 
methods for cracking evaluation.

PWSCC of Inconel 600 is not a new phenomenon.  However, very little special 
attention has been given to the inspection for PWSCC in Inconel 600 applica-
tions other than that associated with the steam generator tubing.  As a 
result of the recently reported instances of PWSCC in the pressurizer heater 
thermal sleeves and instrument nozzles in several domestic and foreign PWRs, 
it may be prudent for licensees of all PWRs to review their Inconel 600 
applications in the primary coolant pressure boundary, and when necessary, 
to implement an augmented inspection program.

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                                                       IN 90-10
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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:  William H. Koo, NRR
                     (301) 492-0928

                     Robert A. Hermann, NRR
                     (301) 492-0911


Attachment:  List of Recently Issued NRC Information Notices
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