United States Nuclear Regulatory Commission - Protecting People and the Environment

Bulletin 74-003: Failure of Structural or Seismic Support Bolts on Class I Components



BL74003 

                              UNITED STATES 
                         ATOMIC ENERGY COMMISSION 
                          WASHINGTON, D.C. 20545 

J. P. O'Reilly, Director of Region I 
N. C. Moseley, Director of Region II 
J. G. Keppler, Director of Region III 
E. M. Howard, Director of Region IV 
R. H. Engelken, Director of Region V 

RO BULLETIN #74-3 "FAILURE OF STRUCTURAL OR SEISMIC SUPPORT BOLTS ON CLASS I 
COMPONENTS" 

The enclosed RO Bulletin #74-3 is forwarded for dispatch to all BWR and PWR 
licensees with Operating Licenses and those licensees having an anticipated 
fuel load date prior to January 31, 1975. 


                                   Harold D. Thornburg, Chief 
                                   Field Support & Enforcement Branch 
                                   Directorate of Regulatory Operations 

Enclosure: 
As stated 

cc:  D. F. Knuth, RO 
     J. G. Davis, RO 
     B. H. Grier, RO 
     A. Giambusso, L 
     J. Hendrie, L 
     D. J. Skovholt, L 
.

 To All BWR and PWR with current Operating             Date: 3/22/74 
     Licenses and those having an anticipated          DRO Bulletin #74-3 
     fuel load date prior to January 31, 1975. 

Gentlemen: 

The enclosed DRO Bulletin No. 3 "Failure of Structural or Seismic Support 
Bolts on Class I Components" is sent to provide you with information 
reported by Connecticut Yankee as an abnormal occurrence at The Haddon Neck 
reactor facility. 

This information may have applicability at your facility(ies). Action 
requested on your part is identified in Section B of the enclosed Bulletin. 

                                   Sincerely, 


                                   Director 

Enclosure: 
DRO Bulletin #74-3 
.

FAILURE OF STRUCTURAL OR SEISMIC SUPPORT BOLTS ON CLASS I COMPONENTS 

We recently received information from the Connecticut Yankee Atomic Power 
Company describing bolt failures found during routine in-service inspections
at the Haddam Neck pressurized water reactor which may relate to the 
installation and serviceability of seismic support bolts or other seismic 
support structures at your facility. 

Description of Circumstances 

A.   During a visual inspection, several steam generator seismic support 
     holddown bolts were observed to be loose. Subsequent inspections by 
     ultrasonic and impact testing of all 256 bolts identified a total of 24
     which had failed and were unable to perform their design function. A 
     preliminary evaluation indicated the bolts had failed in tension 
     apparently from over-torque during the original installation. It was 
     later ascertained by metallurgical and electron microscopic techniques 
     that the failures were the result of stress corrosion, associated with 
     the high pre-stressing, stress risers at the root of the bolt threads 
     and the presence of moisture originally from the concrete and 
     continuing from miscellaneous spills, leakages or high humidity 
     commonly found within containment areas. 

     In reviewing this problem, it was noted that Section XI of the Boiler 
     and Pressure Vessel Code does not specifically address the inspection 
     of-support structures for vessels except for the "support attachment 
     (vessel support skirts) which includes the welds to the vessel and the 
     base metal beneath the weld zone and along the support attachment 
.

                                 - 2 - 

     for a distance of two base metal thicknesses."1/ Support members and 
     structures for piping, valves, and pumps within the system boundary 
     "whose structural integrity is relied upon to withstand the design 
     loads and seismic induced displacements" are subject to examination.2/ 

B.   Action Requested 

     1.   Since the various support structures for vessels within 
          containment are subject to the same environment as other support 
          structures described above, but are not subject to the same 
          examination, it is requested that,during your next scheduled 
          outage, you selectively examine a representative portion of the 
          vessel support members and structures, including the bolting 
          material for two Seismic Category I vessels (as defined in 
          Regulatory Guide 1.29) whose structural integrity is relied upon 
          to withstand design and seismic displacements. This examination 
          should include sufficient coverage of each support structure to 
          provide confidence of serviceability. 
          
     2.   It is requested that you notify the RO Regional Office in writing 
          within 20 days of your proposed schedule for this inspection 
          including the date that your detailed written programs and 
          procedures will be available for RO inspection. Your program shall
          include, but not be limited to: structures to be examined; 
          inspection methods to be used; and number of bolts per structure 
          examined. 

1/ASME Boiler and Pressure Vessel Code; Section XI; Table IS-251: Para. H.
2/ Ibid; Para. K-2. 
.

                                  - 3 -

     3.   If failures are revealed during your inspections, you are 
          instructed to promptly report these as "Abnormal Occurrences" in 
          accordance with the requirements of your license. 
.

                               ENCLOSURE  

                                        Directorate of Regulatory Operations
                                        Information Request No. 74-2  

PWR MAIN STEAM LINE ISOLATION VALVES  

In the past several months Regulatory Operations has received notification 
of a number of abnormal occurrences involving main steam isolation valves at
various PWR facilities. Licensees' investigations following preoperational 
tests and spurious valve closures have indicated that these valves, 
particularly those of the check-valve type, may be subject to generic 
failures.   

To permit evaluation of the extent of the problem, the suitability of 
specific valves for isolation purposes under postulated steam line rupture 
accident conditions, and the appropriateness of possible remedial measures, 
you are requested to provide the following specific information:  

1.   Facility name and unit number.   

2.   A line diagram or sketch showing the locations of the isolation 
     valve(s) with some identification of the type of each valve (check; 
     globe, gate; etc.).   

3.   Assembly or sectional drawings of each valve type with dimensions and 
     identification of material.   

4.   Name of the manufacturer of each valve type.  

5.   Steam line pressure at full power and at hot stand-by conditions, 
     compared with valve design pressure.   

6.   Functional design requirements contained in the original design 
     specifications for each valve. 

7.   An assessment of the adequacy of each valve type to perform the 
     isolation function under postulated steam line rupture accident 
     conditions.   

8.   Discussion of any operational malfunctions of each valve.  

In your response to this request, please also include information relating 
to modifications or other methods of resolution which are planned or which 
may have been made to valves of this type installed, or scheduled to be 
installed, in your facility(ies). 

8011060831 
Page Last Reviewed/Updated Tuesday, July 23, 2013