ML20134M745
| ML20134M745 | |
| Person / Time | |
|---|---|
| Issue date: | 02/13/1997 |
| From: | Lainas G NRC (Affiliation Not Assigned) |
| To: | Reedy R AFFILIATION NOT ASSIGNED |
| References | |
| NUDOCS 9702200391 | |
| Download: ML20134M745 (6) | |
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February 13, 1997 Mr. Roger Reedy Reedy Engineering, Inc.
3425 S. Bascom Avenue, Suite 210 i
Campbell, California 95008 j
SUBJECT:
RESPONSE TO LETTER OF JANUARY 14, 1997, REGARDING NRC MEETING WITH 1
THE NUCLEAR ENERGY INSTITUTE
Dear Mr. Reedy:
4 f
I am responding to your letter of January 14, 1997, to Mr. Stewart L. Magruder l
of the U. S. Nuclear Regulatory Commission (NRC) staff.
In your letter, you comment on the NRC staff's summary of a staff meeting on December 19, 1996, with industry representatives, which was issued on December 26, 1996. The purpose of the meeting was to discuss the industry's response to Generic Letter (GL) 96-06, " Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions," which was issued on September 30, 1996.
In your letter, you assert that certain statements in the staff's meeting sun ary are misleading. These statements concern whether strain criteria are prescribed in the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) and whether overpressure due to the heating of trapped fluid should be considered a primary stress. The NRC staff's review of your letter finds no merit to your assertions, as discussed below.
With regard to strain criteria, the Nuclear Energy Institute (NEI) proposed that licensees use a strain-based acceptance criteria to evaluate the potential for overpressure of isolated sections of piping due to thermal expansion of trapped fluid. The staff's response, documented in the meeting summary, contains the following statement:
...there is no regulatory precedent or technical basis for approving the proposed strain criteria."
In the second paragraph of your letter, you claim that the staff's statement regarding strain-based acceptance criteria is not quite true because paragraph F-1322.5 of Appendix F to Section III of the ASME Code permits the owner to establish limits for permissible permanent strain.
Paragraph F-1322.5 of Appendix F contains the following statement:
"In addition to the limits given in this Appendix, the strain or deformation limits (if any) provided in the Design Specification shall be satisfied." The paragraph merely indicates that if strain or deformation limits are specified in the Design Specification, the I
limits should also be satisfied.
It does not provide an alternative to a
meeting the allowable stress criteria in Appendix F.
The staff does not believe that your argument regarding paragraph F-1322.5 of Appendix F is relevant to the issue of overpressure of isolated sections of piping due to Q
M thermal expansion of trapped fluid.
In the last page of your letter, you also infer that the meeting summary is inaccurate in its implication that pressure due to heating trapped water causes primary stresses.
The ASME Code categorizes stresses in components as primary or secondary, depending on the nature of the applied load. The issue
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R. Reedy 2
of whether the pressure stresses caused by the heating of trapped fluid should be considered primary or secondary was discussed at the ASME Code Section III Working Group on Piping and the Subgroup on Design meetings at Colorado Springs during the ASME Code week of December 9-13, 1996. The NRC staff representative was present during the discussions. A proposed ASME Code interpretation that would have classified these stresses as secondary was withdrawn at the Subgroup on Design meeting. The staff sees no basis for l
changing its position that the pressure caused by the heating of trapped fluid should be evaluated using the ASME Code criteria for primary stresses.
l Sincerely, 4
hSIGNEDRv Gus C. Lainas, Acting Director Division of Engineering l
Office of Nuclear Reactor Regulation cc:
G. Eisenberg, ASME D. Modeen, NEI 0FFICIAL DOCUMENT NAME: G:\\ FAIR \\ REEDY.RS3
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of whether the pressure stresses caused by the heating of trapped fluid should be considered primary or secondary was discussed at the ASME Code Section III Working Group on Piping and the Subgroup on Design meetings at Colorado Springs during the ASME Code week of December 9-13, 1996. The NRC staff representative was present during the discussions. A proposed ASME Code j
interpretation that would have classified these stresses as secondary was withdrawn at the Subgroup on Design meeting. The staff sees no basis for changing its position that the pressure caused by the heating of trapped fluid should be evaluated using the ASME Code criteria for primary stresses.
Sincerely, hSIGNEDBY Gus C. L inas, Acting Director Division of Engineering Office of Nuclear Reactor Regulation cc:
G. Eisenberg, ASME D. Modeen, NEI 0FFICIAL DOCUMENT NAME: G:\\ FAIR \\ REEDY.RS3
- See previous concurrences Distribution:
Central Files /PDR EMEB RF/CHRON GHolahan BSheron BWetzel FMiraglia AThadani GHubbard RZimmerman SMagruder TMartin WTravers To receive a copy of this document, indicate in the box C= Copy w/o attachment / enclosure E= Copy with attachment / enclosure N = No copy 0FFICE EMEB:DE C
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in the meeting summary regarding this issue.
Those meeting attendees who believe that the staff's meeting summary contains inaccurate statements hould submit their concerns directly to the NRC.
The ASME Code categorizes stresses in components as primary or secor}he
- ary, depending on the nature of the applied load. The issue of whether pressure stresses caused by the heating of trapped fluid should by considered l
primary or secondary was discussed at the ASME Code Section III ) forking Group on Piping and the Subgroup on Design meetings at Colorado Springs during the ASME Code week of December 9-13, 1996. The NRC staff represeptative was present during the discussions. A proposed ASME Code interpyetation that would have classified these stresses as secondary was eventdally withdrawn at the Subgroup on Design meeting. Therefore, if the stressps are not classified f
as secondary, then the only alternative is to consider t,he stresses primary.
We continue to believe that the pressure caused by the/or primary stresses.
eating of trapped fluid should be evalua ted using the ASME Code criteri;t f As discussed in previ s correspondence with you, t NRC is not necessarily bound by an ASME Code nterpretation. However, w are perplexed that you j
continue to assert tha these stresses should be reated as secondary stresses inspiteoftheactiontakenbytheASMECodeSdtionIIISubgrouponDesign.
Finally, the staff note that you continue t offer your own interpretations of the ASME Code requir ments regarding the issues addressed in GL 96-06. As documented in our previo s letter regardi g GL 96-06, the staff does not agree with certain of the stat ments you have - de in your letters and considers it unproductive to continual y rehash the ame issues with you,
- incerely, Gus C. Lainas, Acting Director Division of Engineering Office of Nuclear Reactor Regulation j
cc:
G. Eisenberg, ASME I IAL DOCUMENT NAME:
G:\\ FAIR \\ REEDY.RS3 D. Modeen, NEI
- See previous concurrences Distribution:
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i in the meeting summary regarding this issue Those,m/
eeting attendees who j
believe that the staff's meeting summary ntains inaccurate statements should submit their concerns directly to the N
.l The ASME Code categorizes stresses i components as' primary or secondary, depending on the nature of the appl d load. The issue of whether the stresses due to the heating of tra ed fluid should be considered primary or secondary was debated at the ASME Aode Working Group on Piping and the t
Subgroup on Design meetings at C9 orado Springs during the ASME Code week of 1
December 9-13, 1996. The NRC s 'aff representative was present during the debates. 'A proposed ASME Code nterpretation that would have classified these stresses as secondary was evey ually withdrawn at the Subgroup on Design' meeting. Therefore, if the tresses are not classified as secondary, the only alternative is to consider e stresses primary. As discussed in previous correspondenhewithyou,t NRC is not necessarily bound by an ASME Code s
interpretation.\\However, e are perplexed that you continue to assert that these stress 9s are secon ry in spite of the action taken by the ASME Code Subgroup on Design \\
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Finally, the staff non s that you continue to offer your own interpretations of the ASME, Code req rements regarding the issues addressed in GL 96-06. As documented in our p vious letter regarding GL 96-06, the staff does not agree with many of the s ateme'ats you have made in your letters and considers it unproductive to e ntinual rehash the same issues with you. Although the staff has had nu rous in actions with licensees regarding these issues, not one has cited y ur argument as a basis for resolving the issues in GL 96-06.
Ij Sincerely,
/
Gus C. Lainas, Acting Director
/
Division of Engineering
/
Office of Nuclear Reactor Regulation
/
cc: / G Eisenberg, ASME DOCUMENT NAM G:\\ FAIR \\ REEDY.RS3 f
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R. Reedy 2
believe the staff's meeting summary contains inaccurate statements should submit their concerns directly to the NRC.
The ASME Code categorizes stresses in components as primary or secondary depending on the nature of the applied load. The issue of whether the stresses due to the heating of trapped fluid should be considered primary or s'econdary was debated at the ASME Code Working Group on Piping and S Design meetings at Colorado Springs during the Code week of Decepber$ group on 9-13, 1996. The NRC staff repr A
proposedCodeinterpretat\\esentativewaspresentduringthede)ates.
on'that would have classified thes'e stresses as secondary was eventually w\\thdrawn at'the-Subgroup _on-Dest'gn. Therefore, the only alternative is to consider the stresses primary.f s discussed in A
previous correspondence with\\you, the NRC is not necessarily bound by a Code interpretation.
However, we Age perplexed that you' continue to assert that these stresses are secondary in spite of the act 6n taken by the ASME Code Subgroup on Design.
Finally, the staff notes that you ontinue offer your own interpretations of the ASME Code requirements regard ng t issues addressed in GL 96-06. As documented in our previous letter reg (grdt g GL 96-06, the staff does not ag with many of your statements in your le)ters and considers it unproductive to continually rehash the same issues wit)i\\you. Alttough the staff has had numerous interactions with licensees regarding these issues, not one has come forth citing your arguments as a ba s to esolve the issues in GL 96-06.
Sincere y, GusC.Lajnas,ActingDirector Division of Engineering Office of Nuclear Reactor Regulation cc:
G. Eisenberg, ASME D. Modeen, NEI D CUMENT NAME: G:\\ FAIR \\RE DY.RS3 Distribution: Central iles/PDR EMEB RF/CHRON G lahan BSheron BWetzel FMiraglia AThadani GH bard RZimmerman SMagrude',
TMa tin WTravers r
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