ML19323F455
| ML19323F455 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 05/14/1980 |
| From: | Wagner P Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19323F456 | List: |
| References | |
| NUDOCS 8005290042 | |
| Download: ML19323F455 (3) | |
Text
4 s
i MEETING SUWARY DISTRIBUTION Licensee:
Mr. W. C. Jones Division Manager, Production Operations Omaha Public Power District 1623 Harney Street Omaha, Nebraska 68102 H. Denton E. G. Case Docket File NRC POR L POR WAA NSIC
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ORB Project Manager ORB Licensing Assistant OELD AE0D IE-3 RFraley, ACRS-16 Program Support Branch Meeting Summary File NRC Participants Noce: NRC Participants only to get enclosures.
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May 14, 1980 Docket No. 50-285 LICENSEE:
0MAHA PU3LIC POWER DISTRICT (0 PPD)
FACILITY:
FORT CALHOUN STATION, UNIT NO. 1
SUBJECT:
SUMMARY
OF MEETING HELD ON MAY 7, 1980 A meeting was held on May 7,1980 with OPPD to discuss Asymetric Loads methods and calculations requirements.
Representatives from Combustion Engineering (CE) and Ebasco assisted OPPD in the presentation. A list of attendees is attached.
The meeting was called at the request of OPPD to inform the staff of -
the status of the analyses being performed for the Fort Calhoun Station.
OPPD is participating as a member of the CE Owners Group for these analyses but due to some unique features of the Fort Calhoun Plant, much of the effort must be plant specific.
These unique features include the 24-inch stainless steel cold leg piping compared to the normal 30-inch carbonsteel piping.
Representatives of CE presented a slide presentation (copies attached) on: (1) the analysis approach being used generically for CE plants and for Fort Calhoun, and (2) the methods used in evaluating the fracture mechanics properties of the RCS piping.
The analyses performed of the Fort Calhoun Station indicate that: (1) a break in the cold leg is more limiting than a hot leg break, (2) the only load component near to a limiting load value is the RPV support foot under radial load, (3) the biological shield wall which carries the support foot may be overloaded under postulated events, and (4) the steam generator supports may have been underdesigned. The steam generator supports were modified by changing the size and materials used while retaining the original design during the current refueling outage. A description of these modifications to the steam generator supports will be submitted to the NRC in a Monthly Operating Report or in the Annual Report.
s The staff informed OPPD that the use of fracture mechanics to justify assuming less than a 2A (2 times the pipe crosssectional area) size break has not, as yet, been approved by NRC management. The staff requested OPPD to perform a " scoping review" assuming a 2A break and inve-tigate the possibility of adding restraints to prevent gross pipe movement, thereby limiting the effective leak size. OPPD agreed to perform such a review and provide the staff iisitial estimates of such items as hardware co;ts, personnel exposure and plant outage time necessary to perform the modification within approximately three weeks.
OPPD will also continue, as an independent parallel effort, the evaluation of the use of fracture mechanics for justifying a break of less than 2A size in the analyses. OPPD will include a decision in the report mentioned &bove on which type of analysis methods they intend to use for the Fort Calhoun Station.
(, l.),. L t.) U M
Philip C. Wagner, Project fianager Operating Reactors Branch #3 Division of Licensing
Enclosures:
As stated cc: See next page
]
4 ENCLOSURE 1 LIST OF ATTENDEES 5/7/80 MEETING WITH OPPD ASYMMETRIC LOADS Name Organization P. C. Wagner ORB #3/DL/NRC R. A. Clark ORB #3/DL/NRC i
S. B. Hosford NRC T. H. Liu NRC R. C. Iotti Ebasco C. B. Brinkman CE (Bethesda)
R. Gar,ble NRC E. Chelliah NRC J. Grant NRC B. J. Elliot NRC K. J. Morris OPPD T. J. McIuor OPPD R. R. Mills CE T. Natan CE F. C. Cherny MEB/NRC E. B. Blackwood IE/NRC 1
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