United States Nuclear Regulatory Commission - Protecting People and the Environment

Bulletin 88-02: Rapidly Propagating Fatigue Cracks in Steam Generator Tubes

                                                       OMB No.:  31500011
                                                       NRCB 88-02

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

                                February 5, 1988


NRC BULLETIN NO. 88-02:  RAPIDLY PROPAGATING FATIGUE CRACKS IN STEAM
                         GENERATOR TUBES


Addressees: 

     For Action - All holders of operating licenses or construction permits 
     for Westinghouse (W)-designed nuclear power reactors with steam 
     generators having carbon steel support plates.  Steam generators in this 
     category include Westinghouse models 13, 27, 44, 51, D1, D2, D3 and D4, 
     and the Westinghouse model E steam generators at South Texas Unit 1.

     For Information - All other holders of operating licenses or construction 
     permits for Westinghouse (W) and Combustion Engineering (CE) designed 
     nuclear power reactors.  

Purpose: 

The purpose of this bulletin is to request that holders of operating licenses 
or construction permits for Westinghouse (W)-designed nuclear power reactors 
with steam generators having carbon steel support plates implement actions 
specified herein to minimize the potential for a steam generator tube rupture 
event caused by a rapidly propagating fatigue crack such as occurred at North 
Anna Unit 1 on July 15, 1987.  

Description of Circumstances: 

On July 15, 1987, a steam generator tube rupture event occurred at North Anna 
Unit 1 shortly after the unit reached 100% power.  For several days prior to 
the event, operators had observed erratic air ejector radiation monitor read-
ings.  Grab samples were taken prior to the tube rupture for purposes of 
performing environmental release calculations.  Subsequent analysis of this 
data indicated that increasing primary to secondary leakage had occurred over 
a 24- to 36-hour period before the tube rupture event.  This leakage had been 
below the limit given in the Technical Specifications.  The ruptured tube was 
located in Row 9 Column 51 in steam generator "C."  The rupture location in 
this model 51 steam generator was at the top support plate on the cold leg 
side of the tube.  The rupture extended circumferentially 360ø around the 
tube. 

The cause of the tube rupture has been determined to be high cycle fatigue.  
The source of the loads is believed to be a combination of a mean stress level 
in the tube and a superimposed alternating stress.  (The mean stress is 
produced by denting of the tube at the uppermost tube support plate, and the 
alternating stress is the result of out-of-plane deflection of the U-bend 
portion of the 

8802020035 
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                                                            February 5, 1988
                                                            Page 2 of 5


tube above the uppermost support plate, caused by flow-induced vibration.)  
Denting also shifts the maximum tube bending stress to the vicinity of the 
uppermost tube support plate.  These loads are sufficient to produce fatigue 
in an all volatile treatment (AVT) water chemistry environment.  

The specific mechanism for the flow-induced vibration has been determined to 
be a fluid-elastic instability.  The fluid-elastic mechanism has a significant 
effect on tube response in cases where the fluid-elastic stability ratio 
equals or exceeds 1.0.  The stability ratio, SR, is defined as: 

     SR =   V eff / V c

where V eff is the effective crossflow velocity and V c is the critical 
velocity beyond which the displacement response to the tube increases rapidly. 

The most significant contributors to the occurrence of a high fluid-elastic 
stability ratio are believed to have been (1) a reduction in damping at the 
tube-to-tube support plate intersection caused by denting and (2) locally high
flow velocities caused by non-uniform antivibration bar (AVB) penetrations 
into the tube bundle.  The presence of an AVB support will restrict tube 
motion and thus preclude the deflection amplitude required for fatigue.  The 
original design configuration required AVBs to be inserted to Row 11.  
However, inspections have shown that some AVBs in the North Anna Unit 1 steam 
generators penetrate to Row 8, exceeding the minimum AVB design depth.  
However, no AVB support was present for the Row 9 Column 51 tube that 
ruptured. 

Discussion: 

Based on available information, the staff concludes that the presence of all 
the following conditions could lead to a rapidly propagating fatigue failure 
such as occurred at North Anna: 

     (1)  denting at the upper support plate

     (2)  a fluid-elastic stability ratio approaching that for the tube that 
          ruptured at North Anna

     (3)  absence of effective AVB support

Actions Requested: 

Within 45 days following receipt of this bulletin, addressees having 
Westinghouse steam generators with carbon steel support plates shall submit a 
written report detailing the status of their compliance with the actions 
specified below for purposes of minimizing the potential for rapidly 
propagating fatigue failure such as occurred at North Anna 1.  The report 
shall include an appropriate schedule for completion of the analyses described 
under item C below, if applicable. 

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                                                            February 5, 1988
                                                            Page 3 of 5


A.   The most recent steam generator inspection data should be reviewed for 
     evidence of denting at the uppermost tube support plate.  Inspection 
     records may be considered adequate for this purpose if at least 3% of the 
     total steam generator tube population was inspected at the uppermost 
     support plate elevation during the last 40 calendar months.  "Denting" 
     should be considered to include evidence of upper support plate corrosion
     and the presence of magnetite in the tube-to-support plate crevices, 
     regardless of whether there is detectable distortion of the tubes.  The 
     results of this review shall be included as part of the 45-day report.  
     Where inspection records are not adequate for this purpose, inspections 
     of at least 3% of the total steam generator tube population at the 
     uppermost support plate elevation should be performed at the next 
     refueling outage.  The schedule for these inspections shall be included 
     as part of the 45-day report and the results of the inspections shall be 
     submitted within 45 days of their completion.  Pending completion of 
     these inspections, an enhanced primary-to-secondary leak rate monitoring 
     program should be implemented in accordance with paragraph C.1. below.

B.   For plants where no denting is found at the uppermost support plate, the 
     results of future steam generator tube inspections should be reviewed for 
     evidence of denting at the uppermost support plate.  If denting is found 
     in the future, the provisions of item C below should be implemented.  
     Commitments to implement these actions shall be submitted when the 
     results of A above are submitted.

C.   For plants where denting is found, the NRC staff requests that the fol-
     lowing actions be taken: 

     1.   Pending completion of the NRC staff review and approval of the 
          program described in C.2 below or completion of inspections 
          specified in item A above to confirm that denting does not exist, an 
          enhanced primary-to-secondary leak rate monitoring program should be 
          implemented as an interim compensatory measure within 45 days of the 
          date of receipt of this bulletin.*  Implementation of this program 
          shall be documented as part of the 45-day report.  The enhanced 
          monitoring program is intended to ensure that if a rapidly 
          propagating fatigue crack occurs under flow-induced vibration, the 
          plant power level would be reduced to 50% power or less at least 5 
          hours before a tube rupture was predicted to occur.  The 
          effectiveness of this program should be evaluated against the 
          assumed time-dependent leakage curve given in Figure 1.

____________________

*While this bulletin was being prepared, licensees for a few plants committed 
to an enhanced primary-to-secondary leak rate monitoring program at the 
staff's request.  These plants had been identified on a preliminary basis by 
Westinghouse as being potentially susceptible to rapidly propagating fatigue 
cracks.  These enhanced programs should be upgraded as necessary to comply 
with this paragraph.  However, no relaxation from current commitments should 
be made without prior approval by the NRC staff.  
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                                                            February 5, 1988
                                                            Page 4 of 5


          This program should consider and provide the necessary leakage 
          measurement and trending methods, time intervals between measure-
          ments, alarms and alarm setpoints, intermediate actions based on 
          leak rates or receipt of alarms, administrative limits for 
          commencing plant shutdown, and time limitations for (1) reducing 
          power to less than 50% and (2) shutting down to cold shutdown.  
          Appropriate allowances for instrument errors should be considered.  
          Finally, the program should make provision for out of service 
          radiation monitors, including action statements and compensatory 
          measures.

     2.   A program should be implemented to minimize the probability of a 
          rapidly propagating fatigue failure such as occurred at North Anna 
          Unit 1.  The need for long-term corrective actions (e.g., preventive 
          plugging and stabilization of potentially susceptible tubes, 
          hardware, and/or operational changes to reduce stability ratios) 
          and/or long-term compensatory measures (e.g., enhanced leak rate 
          monitoring program) should be assessed and implemented as necessary.
          An appropriate program would include detailed analyses, as described 
          in subparagraphs (a) and (b) below, to assess the potential for such 
          a failure.  Alternative approaches and/or compensatory measures 
          implemented in lieu of the actions in subparagraphs (a) or (b) below 
          should be justified.  

          Although the 45-day report shall provide a clear indication of 
          actions proposed by licensees, including their status and schedule, 
          a detailed description of this program and the results of analyses 
          shall be submitted subsequently, but early enough to permit NRC 
          staff review and approval prior to the next scheduled restart from a 
          refueling outage.  Where the next such restart is scheduled to take 
          place within 90 days, staff review and approval will not be 
          necessary prior to restart from the current refueling outage.  An 
          acceptable schedule for submittal of the above information should be 
          arranged with the NRC plant project manager by all licensees to 
          ensure that the staff will have adequate time and resources to 
          complete its review without adverse impact on the licensee's 
          schedule for restart.  

          (a)  The analysis would include an assessment of stability ratios 
               (including flow peaking effects) for the most limiting tube 
               locations to assess the potential for rapidly propagating 
               fatigue cracks.  This assessment would be conducted such that 
               the stability ratios are directly comparable to that for the 
               tube which ruptured at North Anna.

          (b)  The analysis would include an assessment of the depth of pene-
               tration of each AVB.  The purpose of this assessment is 
               twofold: (1) to establish which tubes are not effectively 
               supported by AVBs and (2) to permit an assessment of flow 
               peaking factors.

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                                                            February 5, 1988
                                                            Page 5 of 5


               (Note: Most steam generators have at least two sets of AVBs.  
               This applies only to the set that penetrates most deeply into 
               the tube bundle.)  The methodology used to determine the depth 
               of penetration of each individual AVB shall be described in 
               detail in the written report.  The criteria for determining 
               whether a tube is effectively supported by an AVB shall also be 
               identified.  (Note: An AVB that penetrates far enough to 
               produce an eddy current signal in a given tube may not 
               penetrate far enough to provide a fully effective lateral 
               support to that tube.)

If addressees cannot perform this suggested approach or meet this suggested 
schedule, they should justify to the NRC their alternative approaches and 
schedules.  

The written reports shall be submitted to the appropriate Regional 
Administrator under oath or affirmation under provisions of Section 182a, 
Atomic Energy Act of 1954, as amended.  In addition, the original of the cover 
letter and a copy of the reports shall be transmitted to the U.S. Nuclear 
Regulatory Commission, Document Control Desk, Washington, D.C. 20555 for 
reproduction and distribution. 

This request for information was approved by the Office of Management and 
Budget under blanket clearance number 31500011.  Comments on burden and 
duplication may be directed to the Office of Management and Budget, Reports 
Management, Room 3208, New Executive Office Building, Washington, D.C. 20503. 

The NRC intends to review the information collected under this bulletin and 
determine the adequacy of specific actions proposed by each licensee.  The 
information will be analyzed and placed in the NRC Public Document Rooms.  

If you have any questions about this matter, please contact one of the 
technical contacts listed below or the Regional Administrator of the 
appropriate regional office. 




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

Technical Contacts:  Emmett Murphy, NRR 
                     (30l) 492-0945

                     Keith Wichman, NRR 
                     (30l) 492-0908

Attachments:  
1.   Figure 1 Leak Rate Versus Time Chart 
2.   List of Recently Issued NRC Bulletins 
Page Last Reviewed/Updated Tuesday, July 23, 2013