NRC Generic Letter 80-105, Implementation of Guidance for USI A-12, Potential for Low Fracture Toughness and Lamellar Tearing on Component Supports
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GL80105
UNITED STATES NUCLEAR REGULATORY COMMISSION 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"
Gentlemen:
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.
.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
Enclosure:
Summary of November 12, 1980 Meeting
cc: Service List, w/o encl.
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UNITED STATES NUCLEAR REGULATORY COMMISSION 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
SUBJECT: SUMMARY OF NOVEMBER 12, 1980 MEETING REGARDING INCLUSION OF LINEAR ELASTIC FRACTURE MECHANICS IN THE RESOLUTION OF USI A-12 (POTENTIAL FOR LOW FRACTURE TOUGHNESS AND LAMELLAR TEARING ON COMPONENT SUPPORTS)
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.
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K. KnielThe 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
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K. KnielAttachments:
- 1. List of Attendees
- 2. EPRI/APTECH Slides
cc: All Attendees H. Levin R. Vollmer S. Norris