ML19268C312
| ML19268C312 | |
| Person / Time | |
|---|---|
| Issue date: | 04/07/1978 |
| From: | Snaider R Office of Nuclear Reactor Regulation |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19268C313 | List: |
| References | |
| REF-GTECI-A-12, REF-GTECI-EQ, TASK-A-12, TASK-OR NUDOCS 8001110462 | |
| Download: ML19268C312 (3) | |
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NUCLEAR REGULATORY COMMissiCN UNITE D ST ATES
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April 7, 1978 MEMORANDUM FOR:
Darrell G. Eisenhut, Assi: tant Director for Systems and Projects, 00R THRU:
Jon K. Davis, Chief Systematic Evaluation Progr,.
Branch, 00R FROM:
Richard 9. Snaider, Systematic Evaluation Pragram Branch, DOR
SUBJECT:
SUMt%RY OF MEETING WITH SANDIA l BORATORY TO DIEUSS GENERIC TECHNICAL ACTIVITY A-12 (FRACTURE TOUGHNESS OF STEAM GENERATOR AND REACTOR C000 PUMP SUPPORTS)
On Thursday, March 30, 1978, representatives of Sandia Laboratory met with the NRC to discuss re-direction of Generic Technical Activity A-12.
A list of attendees is attached. The Sandia slides used in the presentation are also attached.
Santfia noted that it had become apparent, early in their investigation of lice.isee responses to our ie;,temcer 1977 questionnaire, that deter-mination of those plants with poteraial problems was not as easy a task as both we and Sandia had earlier envisioned it.
Sandia therefore p"oposed, for the examination of lamellar tearing potential and its results, that a series of calculations be performed using a comnuter model for a single joint to determine the amount of tearing and residual stress due to welding and then the remaining load carrying capabilities of the torn joint.
The NRC staff objected to this because of the many uncertain assumptions which would be necessary as inputs to the programs.
Instead, the staff suggested a number of alternatives, including assuming that a realistic size lamellar tear exists in the joint.
The size of the postulated tear shoulo be determined from past experi-ence for similar joints and correlated with the computer model results, including residual stresses. The output of this initial analysis will then be utilized to determine remaining load-carrying capacity of the 95001253 8 001110 %d
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-2 April 7,1978 joint in question.
Sandia agreed to utilize this approach, with staff concurrence that the fall-back position with more detailed analyses would still remain a viable option, if it appeared that the initial results adequately represent past experience.
The detemination of fracture toughness of the materials in PWR reactor coolant pump and steam generator supports also presents a problem because of the variability of the material properties as assessed from available literature.
With the staff's concurrence, Sandia has agreed to perform a parametric assessment of critical joints in selected struc-tures.
A set of material-independent parametric ourves, linking values _
of assumed flaw hi:e, stress, and crack-tip stress state, will be the result of such study.
These curves will then be used in determining what minimum fracture toughness values must exist for the material in question and the particular geometry (joint).
NRC and Sandia agreed to confer during the week of April 3,1978, in order to arrive at agreed upon wording for a revision of the Task Action Plan.
Sucn revision will also take into account schedule enanges due to slowness of response frcm a few licensees.
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. Sn der Systematic Evaluation Program Branen Division of Operating Reactors Atta chments :
1.
Attendance list 2.
Slides 95001254
AT'ACHMENT 1 ATTENDANCE LIST R. P. Snaider, NRC R. D. Krieg, Sandia G. A. *norovsky, Sandia G. C.
' cn, Jr, Sandia H. Levin, NRC J. Hodge, NRC H. Conrad, NRC R. Gamble, NRC K. Wichman, NRC D. Sellers, NRC U. Potapovs. NRC V. Noonan, NRC
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H. R. W. Cobb, NRC G. Georgiev, NRC e
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