United States Nuclear Regulatory Commission - Protecting People and the Environment

Information Notice No. 90-49: Stress Corrosion Cracking in PWR Steam Generator Tubes

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

                               August 6, 1990


Information Notice No. 90-49:  STRESS CORROSION CRACKING IN PWR STEAM 
                                   GENERATOR TUBES


Addressees:

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

Purpose:

This information notice is intended to inform licensees of recent problems 
involving stress corrosion cracking (SCC) in PWR steam generator (SG) tubes.  
In particular, this information notice is intended to alert licensees to 
recent findings at Millstone Unit 2 and to recent problems in SCC detection 
during inservice inspections.  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:

1.  Circumferential Cracking at Millstone Unit 2 

In October 1989, the licensee for Millstone Unit 2 conducted a mid-cycle 
inspection of the SG tubing using eddy current testing (ECT).  
Circumferential SCC had been observed in previous inspections to be 
affecting the outer diameter (OD) surface (that is, the secondary side) of 
the tubes at the expansion transition at the top of the tubesheet.  The 
mid-cycle inspection followed a previous inspection during the February 1989 
refueling outage and was performed, in part, out of concern for the 
relatively high rate of SCC growth observed during the previous inspection 
and to ensure that SCC did not excessively degrade the integrity of the 
tubes.  Just before the mid-cycle outage, leakage from the primary side to 
the secondary side was less than 5 gallons per day (gpd).  The plant's 
Technical Specifications limit for such leakage is 144 gpd.

The ECT inspections at Millstone 2 were conducted with a rotating pancake 
coil (RPC) probe to ensure optimal sensitivity to circumferential cracks.  
Tubes found with circumferential crack indications were also inspected by 
ultrasonic testing (UT) to obtain additional information regarding the 
length and depth of the cracks.  In addition, two tubes with circumferential 
crack indications were removed and examined.


9007310195 
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The ECT/RPC inspections revealed 104 tubes with circumferential cracks at 
the expansion transition.  The macrocracks, as defined by ECT/RPC, consisted 
of several discontinuous microcracks that were separated by small ligaments 
of sound material.  The discontinuous nature of the array of microcracks was 
confirmed by the UT and examination of the removed tube specimens.  As 
measured by UT, the macrocracks ranged in circumference from 84 degrees to 
329 degrees and ranged in depth up to 100-percent throughwall. 

All tubes with crack indications were staked and plugged.  In addition, the 
licensee evaluated the residual strength of the cracked tubes to assess 
their capability to sustain normal operating and postulated accident 
loadings before their removal from service.  This structural evaluation 
considered the profiles for each crack obtained from the UT examination.  
This evaluation revealed one cracked tube which failed to meet the ASME 
Code, Section III, NB-3225 and Appendix F stress limits for postulated 
accident conditions.  (Regulatory Guide 1.121, "Bases for Plugging Degraded 
PWR Steam Generator Tubes," states that margins should be consistent with 
the stress limits in Section III of the code.)  Based on these findings, the 
staff concludes that the integrity of the subject tube was not ensured under 
postulated accident conditions.

The staff has recently identified service induced, circumferential SCC, such 
as at Millstone Unit 2, to be a source of significant degradation to tubes 
in PWR steam generators.  Such cracking is particularly noteworthy because 
it is generally not detectable with conventional bobbin probes used 
routinely for inservice inspection.  Such cracking is generally only 
detectable through the use of specialized probes, such as the RPC probe.

Most circumferential cracking has been observed at tube expansion 
transitions at or near the top of the tubesheet.  In addition to Millstone 
Unit 2, circumferential cracking at the expansion transition has recently 
been identified at one other Combustion Engineering (CE) plant (Maine 
Yankee), at three plants with Westinghouse Model 51 steam generators (North 
Anna Unit 1, Trojan Unit 1, and Sequoyah Unit 1), and at one plant with 
Westinghouse Model D steam generators (McGuire Unit 1).  Tubes in the 
affected CE and Westinghouse Model 51 steam generators were explosively 
expanded against the tubesheet.  Tubes in the McGuire Model D steam 
generators were expanded against the tubesheet by mechanical rolling.
In addition to being found at the expansion transition location, widespread 
circumferential SCC has been observed at drilled-hole support plate 
locations at Palisades (CE steam generators).  Isolated instances of 
circumferential SCC have been reported at the uppermost support plate of a 
pre-replacement Westinghouse Model 44 steam generator of Indian Point Unit 3 
and at a row 1 U-bend of a Model 51 steam generator at Zion Unit 1.  The 
circumferential SCC at Palisades and Indian Point Unit 3 appears to be 
associated with significant denting at the support plates.

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2.  Axial Cracking at Support Plates

Licensees have reported secondary side-initiated, axial SCC at several 
plants with Westinghouse Model 51 and Model D steam generators at support 
plate intersections exhibiting little or no denting.  Recent difficulties 
experienced in the detection of such cracks were the subject of Westinghouse 
Customer Information Letter GEN-LTR-90-006, "Steam Generator Tube Outer 
Diameter Stress Corrosion Cracking at Tube Support Elevation-Eddy Current 
Detection Issue," which was issued on or about February 8, 1990, to all 
utilities with Westinghouse steam generators.  Westinghouse reported that 
metallographic examinations of tubes removed from the field have revealed 
the presence of OD-initiated SCC at tube support plate (TSP) intersections 
that were not reported by personnel using a bobbin probe to perform field 
eddy current tests.  For example, these examinations revealed one tube 
containing axial cracks within two 30-degree-wide bands on opposite sides of 
the tube, with the deepest crack penetrating to 62-percent throughwall.  The 
field eddy current interpretation of the signal for this TSP location was 
"no detectable degradation" (NDD) using the plant voltage threshold 
criteria.  

The EPRI "Steam Generator Examination Guidelines, Revision 2" contains 
Figure C-55 showing the qualitative relationship between bobbin probe signal 
amplitude and crack depth determined metallographically.  The staff believes 
that the amplitude threshold criteria used at the plant in the 
above-mentioned example were taken from the actual data used to develop 
Figure C-55.  The recent evidence cited by Westinghouse suggests that this 
data is not conservative for all plants.

Industry meetings attended by representatives from a number of vendors 
providing SG inspection services, EPRI, and the Westinghouse Owners Group 
have been conducted to examine various proposals to detect and measure SCC 
at TSP locations.  The minutes of the EPRI Guidelines Revision Committee 
meeting on February 13, 1990, note that general principles still apply, 
pending development of updated guidance, for the detection and measurement 
of SCC at support plates.  Section 4.6.1 of the EPRI guidelines states that 
as a general rule, an "extremely conservative position" should be adopted 
for the resolution of distorted indications or undefined signals not covered 
by existing analysis guidelines.  Specifically, tubes with these types of 
indications should be recommended for plugging unless other supporting data 
exists (tube pulls or NDE diagnostic data) that justifies their retention as 
active tubes. 

Discussion:

The reliable detection and sizing of SCC during inservice inspections pose a 
significant challenge to current ECT technology and practice in view of the 
low signal-to-noise ratios frequently exhibited by such cracks.  Experience 
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indicates that SCC is frequently not detected until it has penetrated beyond 
40-percent throughwall.  Fortunately, the vast majority of SCC flaws consist 
of short axial or circumferential crack segments.  The staff believes that 
such flaws can be detected before they grow sufficiently large to degrade 
the structural margins of the tube to below the Regulatory Guide 1.121 
criteria.

Tubes are generally inspected once per refueling cycle.  Depending on the 
rate of crack growth and the number of tubes involved, this frequency may or 
may not be sufficient to ensure that all cracks are detected before they 
become sufficiently large to degrade structural margins to less than the 
Regulatory Guide 1.121 criteria.  A structural assessment of the crack 
geometries found during an inspection, such as performed at Millstone Unit 
2, provides a means for assessing whether the inspection frequency is 
sufficient to ensure adequate structural margins for all tubes between 
inspections.

The staff believes that the effectiveness of eddy current testing for 
detecting and sizing SCC can be enhanced through improved criteria for the 
qualification and performance demonstration of the eddy current data 
acquisition equipment (including probes), data analysis procedures, and data 
analysts.  The staff and the industry, including the EPRI Steam Generator 
Reliability Project, are evaluating this issue.  In the meantime, field 
experience indicates that careful attention to the potential for SCC at the 
tubesheet region, at tube expansion transitions, at support plates, and at 
tube U-bends is important.  The recommendations in Section 4.6.1 of the EPRI 
steam generator examination guidelines indicate the importance of being 
alert to distorted and undefined signals at these locations and employing 
diagnostic measures as appropriate to establish the cause of these signals 
and to validate the plant data analysis procedures and criteria.

Finally, the findings at Millstone Unit 2 illustrate that cracks will not 
necessarily cause leakage approaching the Technical Specifications limit 
before the structural margin in the affected tube drops below the Regulatory 
Guide 1.121 criteria for postulated accident conditions.  This point is 
further illustrated by the steam generator tube rupture (SGTR) event that 
occurred at McGuire Unit 1 on March 7, 1989, as a result of axially oriented 
SCC.  Leakage before the SGTR event was about 15 gpd, which was small 
compared to the plant's Technical Specifications limit of 500 gpd.  The 
McGuire event occurred under normal operating conditions.  Thus, the McGuire 
event was preceded by a period during which the subject tube was vulnerable 
to rupture if challenged by a postulated accident.

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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 
the technical contact listed below or the appropriate NRR project manager.




                            Charles E. Rossi, Director
                            Division of Operational Events Assessment 
                            Office of Nuclear Reactor Regulation

Technical Contact:  E. Murphy, NRR
                    (301) 492-0710

Attachment:  List of Recently Issued NRC Information Notices
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