Bulletin 89-01: Supplement 2, Failure of Westinghouse Steam Generator Tube Mechanical Plugs

                                                OMB No:  3150-0011 
                                                NRCB 89-01, Supplement 2 


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

                                June 28, 1991


NRC BULLETIN NO. 89-01, SUPPLEMENT 2:  FAILURE OF WESTINGHOUSE STEAM 
                                       GENERATOR TUBE MECHANICAL PLUGS


Addressees:

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

Purpose:

This bulletin supplement requests that actions similar to those requested in 
NRC Bulletin 89-01, "Failure of Westinghouse Steam Generator Tube Mechanical 
Plugs," be extended to include all Westinghouse mechanical plugs fabricated 
from thermally treated Inconel 600.  These actions are to ensure that these 
plugs will continue to provide adequate assurance of reactor coolant 
pressure boundary (RCPB) integrity under normal operating, transient, and 
postulated accident conditions.

Description of Circumstances:

Bulletin 89-01 requested that licensees determine whether certain mechanical 
plugs supplied by Westinghouse are installed in their steam generators and, 
if so, that an action plan (including plug repairs and/or replacement) be 
implemented to ensure that these plugs will continue to provide adequate 
assurance of RCPB integrity.  This request applied only to plugs fabricated 
from Inconel 600 heats NX-3279, NX-3513, NX-3962, and NX-4523 (hereafter 
referred to as group 1 heats) on the basis of field experience and 
laboratory studies indicating that plugs from group 1 heats are highly 
susceptible to primary water stress corrosion cracking (PWSCC).  Such 
cracking led to a gross plug failure at the North Anna Power Station, Unit 
1, resulting in a 75-gallon-per-minute primary-to-secondary leak.  The 
bulletin requested that the subject repairs and/or replacements be 
accomplished according to a schedule consistent with an algorithm developed 
by Westinghouse (Reference 1, Revision 1), using as a benchmark the most 
conservative corrosion rate data from the field (observed at Millstone, Unit 
2, for a plug fabricated from heat NX-3513).

After the NRC issued Bulletin 89-01, Westinghouse issued Revision 3 of 
Reference 1 providing complete listings of plug lifetimes categorized by 
plant, date of installation, and heat number.  Lifetimes were listed for all 
Westinghouse mechanical plugs fabricated from thermally treated Inconel 600, 

9106250162 
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including plugs fabricated from group 1 heats and plugs fabricated from all 
other heats of thermally treated Inconel 600, hereafter referred to as group 
2 heats.  Group 2 heats include heat numbers NX-1989, NX-2386, NX-2387, 
NX-5222, NX-6135, NX-6323 (HW), and NX-6323 (CR).  The lifetimes listed for 
plugs fabricated from group 1 heats were consistent with the approach 
requested in the bulletin.  The lifetimes listed for plugs fabricated from 
group 2 heats were estimated using the same approach as for group 1, except 
that the estimates were adjusted upward by heat-specific "microstructural 
factors" to reflect the comparative time-to-cracking performance of each 
group 2 heat with heat NX-3513 during Westinghouse corrosion tests.

During the summer and autumn of 1990, the staff received reports of 
instances of PWSCC at two plants affecting group 2 heats of thermally 
treated Inconel 600.  These instances (at Sequoyah Nuclear Power Plant, Unit 
1, and North Anna Power Station, Unit 2) were described in Supplement 1 to 
Bulletin 89-01.  At the Sequoyah Nuclear Power Plant, Unit 1, five plugs 
fabricated from heat NX-5222 were removed from the field to examine their 
condition.  One of these five plugs was found to exhibit a circumferential 
crack above the expander.  This crack consisted of two small cracks with a 
total length of approximately 15 degrees around the plug circumference and a 
maximum depth of penetration of 0.009 inch.  Another plug was found to 
exhibit axial cracks below the expander.  These five plugs had accumulated 
only 21 percent of the calculated plug lifetime, as given in Revision 3 of 
Reference 1.

At North Anna, Unit 2, 15 plugs fabricated from heat NX-6323(HW) were 
inspected using a Westinghouse eddy current test technique.  Nine of these 
plugs were located on the hot-leg side, five of which exhibited evidence of 
minor leakage.  Eight of the nine hot-leg plugs exhibited indications of 
axial and/or circumferential cracking above the expander.  One of the plugs, 
which was removed from the field and examined, was found to contain a crack 
that extended 360 degrees around the plug circumference.  The crack varied 
in depth between 74 percent and 99 percent of the plug wall thickness.  No 
indications were found in the six cold-leg plugs that were inspected.  The 
accumulated service time on these plugs was less than 20 percent of the 
calculated lifetime, as given in Revision 3 of Reference 1.

As documented in Reference 2, Westinghouse revised its algorithm for 
estimating plug PWSCC lifetimes in light of the accumulated plug experience 
to date.  The revised algorithm included revised heat-specific 
"microstructural factors" for plugs fabricated from group 2 heats based on a 
more conservative treatment of Westinghouse corrosion test data.  The 
revised algorithm also included a revised "time to PWSCC failure" versus 
plug temperature relationship, based on operating experience trends.   

Table 2 of Reference 2 provided lifetime estimates, based on the revised 
algorithm, for all Westinghouse mechanical plugs fabricated from both group 
1 and group 2 heats.  These lifetime estimates were based on the heat 
specific microstructural factors given in Table 1 of Reference 2 and with 
the exception of heat NX-5222, were benchmarked against the aforementioned 
field data from Millstone, Unit 2.  The lifetime estimates in Reference 2 
for plugs fabricated 

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                                                June 28, 1991
                                                Page 3 of 6 


from heat NX-5222 were benchmarked against the most conservative field data 
from Sequoyah, Unit 1 (for heat NX-5222).

The staff believes that the field data for heat NX-5222 from Sequoyah Unit 1 
are insufficient to justify a less conservative approach than Westinghouse 
is employing for other heats of thermally treated Inconel 600 (Reference 3).  
At the request of the NRC staff, Westinghouse has issued new lifetime 
estimates for plugs fabricated from heat NX-5222 (Reference 4).  These 
revised estimates reflect the heat-specific microstructural factors given in 
Table 1 of Reference 2 and are benchmarked against the aforementioned field 
data from Millstone, Unit 2, consistent with the approach taken by 
Westinghouse for plugs from all other heats of thermally treated 
Inconel 600.  The most up-to-date listing of plug lifetime estimates for all 
heats of thermally treated Inconel 600 is given in Table 2 of Reference 4.  

Discussion:

Based on the recent field experience at Sequoyah, Unit 1, and North Anna, 
Unit 2, and the results of corrosion tests performed by Westinghouse, the 
staff has concluded that a systematic remedial action program, similar to 
that requested in the original bulletin for plugs fabricated for group 1 
heats, is needed for all Westinghouse mechanical plugs fabricated from 
thermally treated Inconel 600 to ensure the continued integrity of the RCPB 
over both the short-term and long-term.  For this reason, the actions 
requested in this bulletin supplement apply to all Westinghouse mechanical 
plugs fabricated from thermally treated Inconel 600.  The requested schedule 
for implementing the necessary remedial actions is consistent with the 
Westinghouse plug lifetime estimates in Reference 4. 

It is possible that future refinements to the plug lifetime estimates will 
become appropriate as additional field experience is accumulated.  In 
addition, technical industry organizations such as Westinghouse and the 
Electric Power Research Institute/Steam Generator Reliability Project are 
continuing to examine this issue.  The NRC staff is also monitoring 
developments on the issue and will issue further supplements to this 
bulletin if found to be warranted.  Although not required by this bulletin 
supplement, addressees are encouraged to monitor the condition of the plugs 
through sample removal and examination of plugs as was done at Sequoyah, 
Unit 1, or by eddy current testing as was done at North Anna, Unit 2, 
thereby adding to the operational experience database. 

Actions Requested:

1.   Addressees are requested to verify that information contained in 
     Table 2 of Reference 4 for their plants is correct for plugs fabricated 
     from group 2 heats.  (Addressees have previously verified similar 
     information for group 1 plugs in response to the original bulletin.)  
     The specific information to be verified is the number of Westinghouse 
     mechanical plugs installed in the hot-leg and cold-leg side of each 
     steam generator, categorized by heat number and date of installation.  
     The plug operating temperatures for each plant given in this Table 
     should also be verified.  If information from this 

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                                                June 28, 1991 
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     Table is incorrect, addressees should provide correct information.  
     Addressees are requested to so state if their plants have not installed 
     Westinghouse mechanical plugs from group 2 heats.

2.   Addressees are requested to take the following actions, to be 
     implemented initially during any refueling outage or extended outage 
     (greater than four weeks) which ends 60 days or more following receipt 
     of this bulletin and during all future refueling outages.  For the 
     period of time between receipt of the bulletin and 60 days, the actions 
     requested in the original version of this bulletin continue to be 
     applicable for plugs fabricated from group 1 heats.

     a)   Addressees should implement appropriate remedial actions (i.e., 
          repair and/or replacement) for all plugs whose estimated lifetime 
          in item 2b, below does not extend to the next refueling outage.  

     b)   Remaining lifetime estimates (in effective full power days (EFPD)) 
          are given in Table 2 of Reference 4 in the column entitled "Remain 
          EFPD to MIN."  These remaining lifetime estimates are relative to 
          reference dates given in the column entitled "Reference CALC 
          Dates." These remaining lifetime estimates may be used directly.  
          These estimates should be adjusted to reflect any corrections 
          noted in Actions Requested, item 1.  
          
     c)   For refueling outages or extended outages ending prior to 
          November 30, 1991, remedial actions for plugs fabricated from 
          NX-5222 may be deferred until the next scheduled refueling outage.

     d)   Installation of Westinghouse mechanical plugs fabricated from 
          Inconel 600 should be discontinued.

     e)   If for any refueling outage, the addressee does not plan to 
          satisfy items 2a to 2d above, an alternative plan for insuring 
          plug integrity, with appropriate technical justification, should 
          be submitted to the NRC at least 30 days before the end of the 
          refueling outage.

     f)   Prior to any plug repairs or replacement, addressees are reminded 
          that their responsibilities under ALARA require analysis of the 
          various plug repair or replacement methods.  In choosing a plug 
          repair or replacement method, the licensee should consider the 
          accessibility of the plugs and the dose reduction benefit of using 
          robotic manipulators.  Prior to plug repair or replacement, the 
          licensee should consider steam generator decontamination and/or 
          local shielding to reduce working area dose rates.  

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Reporting Requirements:

Pursuant to Section 182a of the Atomic Energy Act and 10 CFR 50.54(f), 
addressees shall submit a letter within 30 days of receipt of this bulletin 
containing:

1.   The information under Actions Requested, item 1.

2.   A statement indicating whether or not the actions in Actions Requested, 
     item 2 have been or will be taken.  If alternative actions are 
     proposed, supporting justification shall be provided.

The written reports required above shall be addressed to the U.S. Nuclear 
Regulatory Commission, ATTN:  Document Control Desk, Washington, D.C. 20555, 
under oath or affirmation.  In addition, a copy shall be submitted to the 
appropriate regional administrator.

Backfit Discussion:

This supplement expresses new staff positions which are considered to be 
backfits justified under the criteria of 10 CFR 50.109(a)(4)(i) [compliance 
exception backfit]; in addition, the requirement for a response is 
considered to be justified under the criteria of 10 CFR 50.54(f) 
[information request]. 

Plant Technical Specifications require plugging or sleeving of steam 
generator tubes which are found to be defective during inservice inspection.  
Such plugs or sleeves are necessary to ensure the continued integrity of the 
RCPB under normal operating, transient, and postulated accident conditions.  
However, mechanical plugs supplied by Westinghouse are subject to PWSCC 
attack, which adversely affects the RCPB.  The actions requested in this 
bulletin supplement are necessary to ensure that these plugs will continue 
to provide adequate assurance of RCPB integrity under the above mentioned 
conditions.  Thus, these actions are to ensure that defective tubes continue 
to be effectively plugged as required by the Technical Specifications, to 
ensure compliance with General Design Criteria 14 and 31 of 10 CFR Part 50, 
Appendix A, and to ensure compliance with the quality assurance requirements 
of Criterion XVI of 10 CFR Part 50, Appendix B.  

A documented evaluation of the type described in 10 CFR 50.109(a)(6) was 
prepared to state the objectives of and reasons for the modification, and 
the basis for invoking the compliance exception.  Because the bulletin 
supplement also requires submittal of written reports under 10 CFR 50.54(f), 
the document also contains the reasons for the information request in view 
of the potential safety significance of the problem.  This document, which 
is a revised review package submitted to the Committee to Review Generic 
Requirements (CRGR), will be made available in the Public Document Room in 
association with the minutes of the 203rd meeting of the CRGR.  

This request is covered by Office of Management and Budget Clearance 
Number 3150-0011 which expires on June 30, 1991.  The estimated average 
burden hours are 160 man-hours per licensee response, including assessing of 
the new 
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                                                June 28, 1991
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recommendations, searching data sources, gathering and analyzing the data, 
and preparing the required letters.  These estimated average burden hours 
pertain only to these identified response-related matters and do not include 
the time for actual implementation of the requested actions.  Send comments 
regarding this burden estimate or any other aspect of this collection of 
information, including suggestions for reducing this burden, to the Records 
and Reports Management Branch, Division of Information Support Services, 
Office of Information Resources Management, US. Nuclear Regulatory 
Commission, Washington, D.C. 20555; and to the Paperwork Reduction Project 
(3150-0011), Office of Management and Budget, Washington, D.C. 20503.

If you have any questions about this matter, 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:  Emmett Murphy, NRR
                    (301) 492-0710


References:

1.   Westinghouse reports WCAP-12244 (proprietary version) and WCAP-12245  
     (non-proprietary version), "Steam Generator Tube Plug Integrity Summary 
     Report," Revision 1, April 1989; Revision 3, November 1989.

2.   Addendum to Revision 3 of Westinghouse reports WCAP-12244 (proprietary 
     version) and WCAP-12245 (non-proprietary version), "Steam Generator 
     Tube Plug Integrity Summary Report," December 1990.

3.   Memorandum, C. Y. Cheng to J. E. Richardson, "Summary of Meeting with 
     Westinghouse on October 11, 1990, Concerning Steam Generator Tube Plug 
     Issue," May 22, 1991.  (Available in the Public Document Room).

4.   Addendum 2 to Revision 3 of Westinghouse reports WCAP-12244 
     (proprietary version) and WCAP-12245 (non-proprietary version), "Steam 
     Generator Tube Plug Integrity Summary Report," May 1991.


Attachment:  List of Recently Issued NRC Bulletins

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