Implementation of Guidance for USI A-12, Potential for Low Fracture Toughness and Lamellar Tearing on Component Supports (Generic Letter 80-105)


                               UNITED STATES 
                          WASHINGTON, D.C.  20555 

Generic Task No. A-12 

All Power Reactor Licensees and Applicants 

Subject:  Implementation of Guidance for Unresolved Safety Issue (USI) A-12,
          "Potential For Low Fracture Toughness and Lamellar Tearing on 
          Component Supports" 


Attached for your review and guidance is the summary of the November 12, 
1980 meeting of the NRC staff, EPRI, and interested licensees, license 
applicants, and consultant organizations.  The meeting was held to discuss 
the ongoing EPRI program for the inclusion of fracture-mechanics - based 
analyses as part of the implementation program for resolution of USI A-12.  
A final meeting is to be held on Wednesday, December 17, 1980, in room P-118 
of the Phillips Building, 7920 Norfolk Avenue, Bethesda, Maryland.  The 
meeting will commence at 9:00 A.M. and your attendance, or that of your 
representatives, is encouraged. 

Subsequent to the November 12 meeting, some participants expressed confusion
regarding the failure-consequence analysis. Although the impression may have
been given that a decision could be made to perform such an analysis without
attempting to classify the material by its properties (NOT, CVN and possibly
fracture mechanics parameters), this is neither what the NRC intends or will
allow. Failure-consequence analysis must follow an attempt to determine the 
properties of the material by use of the methods provided to you in the May 
19, 1980 letter to licensees and May 20, 1980 letter to license applicants. 
As you know, the guidance provided in those letters will be modified, before
presentation in the final NUREG document, by consideration of the comments 
provided by the industry. 

In addition, and as a result of questions raised during the meeting, the NRC
staff is considering a major modification to the implementation program. As 
you are aware, the Sandia study (Appendix C to NUREG-0577) recommended the 
classification of plants and materials into three groups. Group I contained 
plants and materials on which significant questions remained; Group II was 
for plants and materials for which no conclusion could be drawn as to their 
being satisfactory enough for immediate approval or bad enough to require 
further attention.  Group III contained the plants on which the conclusion 
could be drawn that the design and materials were satisfactory. 

                                  - 2 -

It was the staff's intention at the time that the Group II plants, and 
possibly the materials , would be given further examination, if necessary, 
after the conclusion of the review of the Group I plants.  This would have 
required substantial staff involvement. 

The lack of staff time for such involvement has already necessitated 
significant reassessment of the A-12 implementation program and in fact was 
the impetus behind the May 19 and May 20, 1980 letters.  Further review has 
revealed that the Group II classification will leave certain plants and 
materials essentially in limbo and will not provide a "clean" resolution of 
the program.  Therefore, we are considering removing the Group II 
classification for plants and materials, and reclassifying such plants and 
materials as Group I for the purposes of the implementation review.  Please 
recognize that, although this will add to your review effort now, it will 
decrease the amount of interface effort with the staff necessary later and 
will help assure that you can meet the intended deadline.  This deadline, 
incidentally, is again being reconsidered for further extension.  Both 
subjects will be discussed during the December 17, 1980 meeting. 

As a final point of clarification, licensees of the PWR plants listed below 
are reminded that the ongoing review of their  steam generator and reactor 
coolant pumps by the Franklin Research Center does not relieve them of the 
requirement to review other applicable supports by the criteria determined 
during the present NRC/industry review effort and to be published in the 
final version of NUREG-0577.  This also applies to additional support 
components (e.g. snubber arms) not previously reviewed and presently being 
considered for inclusion in the review.  These plants are: 

     Arkansas Nuclear One, Unit 2 
     San Onofre 1 
     Turkey Point 3 and 4 
     Indian Point 2 and 3 
     D. C. Cook 1 and 2 
     Salem 1 and 2 
     Zion 1 and 2 

Please refer any questions to Richard Snaider at 301-492-7876. 

                                        Darrell G. Eisenhut, Director 
                                        Division of Licensing 

Summary of November 12, 1980 Meeting

cc:  Service List, w/o encl.

                              UNITED STATES 
                          WASHINGTON, D.C.  20555 
                                NOV 25 1980 

Generic Task No. A-12 

MEMORANDUM FOR:     K. Kniel, Chief 
                    Generic Issues Branch 
                    Division of Safety Technology 

FROM:               R. Snaider, Task Manager 
                    Unresolved Safety Issue (USI) A-12 


On Wednesday, November 12, 1980 meeting was held to discuss industry efforts
to develop a fracture mechanics-based program that meets NRC criteria and 
could therefore be included in the staff's requirements for the resolution 
of USI A-12. Attendees include representatives of the Electric Power 
Research Institute, licensees, license applicants, architect/engineering 
firms, and consultants. A list of attendees is attached. 

The option of a fracture mechanics analysis had been included in the draft 
NUREG-0577 issued in November 1979 but had been removed in the May 19, 1980 
and May 20, 1980 letters to licensees and applicants, respectively. Related 
information regarding the proposed use of a fracture mechanics analysis can 
be found in the September 10, 1980 meeting summary of the August 27, 1980 
meeting and in the October 6, 1980 generic letter to all power reactor 
licensees and applicants. The September 10, 1980 letter also includes the 
NRC-established criteria which the fracture mechanics program must meet. 

As part of its introductory statements, the NRC staff clarified its position
with regard to the use of the proposed fracture mechanics program if it is 
approved when presented in its final form to the staff. The generic letter 
of October 6, 1980 states that licensees and applicants must make a choice 
and commit their organizations to either the NRC program delineated in the 
May 19th and 20th, 1980 letters (to be amended by industry comments) or the 
EPRI-proposed alternative program if approved. The staff has since modified 
its position such that both programs may be used as necessary during the 
analysis. For example, if a fracture mechanics analysis of a specific 
material is estimated to be too expensive or time-consuming, the NDT or CVN 
approach of the May 19th and 20th letters (as amended) may be used. 

K. Kniel                          - 2 -

The remainder of the meeting was devoted to a presentation by EPRI and 
APTECH (EPRI contractor) personnel regarding the status of the A-12 
resolution program.  Applicable slides are attached.  The programs for 
stress-corrosion cracking (SCC) and lamellar tearing are in their infancy, 
with a two-pronged approach (experience data base plus relevant materials 
properties review) being planned for SCC and a program for lamellar tearing 
being negotiated with a potential contractor.  Although no further 
information was discussed with regard to these aspects of the program, the 
NRC staff did note that the December 17, 1980 final resolution meeting was 
the deadline for presentation of the proposed SCC program and that, absent 
such a proposed program, the NRC will impose the program of its May 19th and 
20th letters (as amended). 

The EPRI representatives noted that their review of the NRC (non-fracture 
mechanics) procedures of the May 19th and 20th letters was complete. A "flow
diagram" was presented which demonstrated how a material would be proven 
either satisfactory or unsatisfactory. A "simplified" procedure was also 
presented. It was noted that the plant-specific portion of this procedure 
would be the most difficult aspect of the evaluation. It is in the 
plant-specific evaluation that decisions must be made regarding materials 
testing versus consequence analyses and that plant management must determine
what cost and inconvenience would be involved in operational control of 
support temperature by ancillary heating. 

The potential impact of the program on existing plants was reviewed and was 
followed by a discussion of what benefit fracture mechanics will have with 
respect to minimizing the impact on existing plants. It was noted that 
APTECH has already classified 22 materials as part of their materials data 
base development, and that emphasis is being applied to pin-column supports 
because they would be least likely to satisfy a consequence analysis. 

In summary, EPRI stated that they view the September 10, 1980 NRC summary of
the August 27, 1980 meeting as a statement of criteria that their program 
must meet and that they are working to meet this objective. EPRI intends to 
have a meeting with licensees and applicants prior to the final resolution 
meeting with the NRC. This final meeting with the NRC is scheduled for 
Wednesday, December 17, 1960, in Room P-118 of the Phillips Building, 7920 
Norfolk Avenue, Bethesda, Maryland. The meeting begins at 9:00 A.M. 

                                        R. P. Snaider, Task Manager 
                                        Unresolved Safety Issue A-12 

Attachments:  See Next Page 

K. Kniel                          - 3 -

1.   List of Attendees
2.   EPRI/APTECH Slides 

cc:  All Attendees
     H. Levin
     R. Vollmer
     S. Norris 


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