ML19296D748
| ML19296D748 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 02/22/1980 |
| From: | Kintner L Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8003130105 | |
| Download: ML19296D748 (4) | |
Text
c p* " E G ti UNITED STATES
+(( 3 e, 'h NUCLEAR REGULATORY COMMISSION 3.,
E WASHINGTON, D. C. 20555
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FEB : 2 1980 Docket No. 50-341 APPLICANT: Detroit Edison Company FACILITY:
Enrico Fermi Atomic Power Plant, Unit 2
SUBJECT:
EUMMARY OF JANUARY 10 AND 11, 1980 MEETING REGARDING ASYMMETRIC LOCA LOADS ON REACTOR PRESSURE VESSEL SUPPORT STRUCTURES The NRC staff met with representatives of the applicant and General Electric to discuss the methods used to calculate loads on the Fermi 2 reactor pressure vessel, internals and support structures resulting from postulated ruptures taken individually at the recirculation outlet, inlet, and feedwater inlet piping weld to the applicable vessel nozzle safe end.
The reactor vessel and its support were designed and built by Combustion Engineering. General Electric (GE) computed the responses to the pipe break dynamic loads on the vessel and its support using the DYSEA code.
Nuclecr Utility Services (NUS) computed the transient pressure distribution in the annulus between the sacrificial shield and the reactor vessel. Sargeant &
Lundy used a separate model to calculate the seismic responses of the reactor building, reactor vessel and support structures. Calculatei responses include those resulting from the reactor vessel averturning moment, the blowdown reaction force of the pipe and the annulus pressure on the sacrificial shield, the reactisn forces of the jet inside the reactor and the jet from the pipe, the jet impingement load on the shield wall, and seismic loads. Various LOCA loads we. combined using time-history methods.
Results for components have been presented in the FSAR and by reference reports from contractors (GE, NUS, Sargeant and Lundy). The staff requested that results be presented in one document, including the calculated loads on each component (e.g., skirt) which when combined result in calculated stress levels most closely approaching the ASME Section III faulted condition Code allowable stresses. The applicant said it is currently preparing revised results, due to revisions in the DYSEA Code, and will file the results in April 1980.
It will provide the requested summary of results including a qualitative summary of conservative factors inherent in the analysis, a description of the seismic model, which is different from the model used in the DYSEA Code analysis, and the methods of node selection and loading application of various pipe break loads on the mathematical model used for the DYSEA analyses.
300310
. FEB : 2 ;ggg The applicant reviewed the analysis of the reactor vessel support skirt in detail. Buckling loads and stresses for the Fermi 2 support skirt were evaluated using a simple model.
The applicant described a more detailed method of calculation that was used to show that the maximum critical buckling stress for the Hatch 2 support skirt was within the allowable critical buckling stress. The effect of using this Hatch 2 calculational method on Fermi 2 was estimated, and it appeared that maximum critical buckling stress for the Fer ~ ' support skirt was also within the allowable critical buckling stress.
The
... cant agreed to describe its stress evaluation of the support skirt in more detail than is currently provided in the FSAR, Page 3.9-8c.
A drawing of the support skirt, showing the longitudinal welds, weld to the reactor vessel, and access openings through the skirt was mailed to NRC subsequent to the meeting.
In calculating the pressure in the annulus between the sacrificial shield and the reactor vessel following a pipe rupture at the vessel nozzle safe end, the applicant used the efflux of coolant from a variable pipe break area.
The break area utilized was the gap between the end of the ruptured pipe and the vessel nozzle which varied from zero at the time of pipe separation from the nozzle to a maxin.um area equivalent to the pipe flow area.
The motion of the end of the ruptured pipe was calculated by considering the force of the coolant acting on the pipe, neglecting the time to propagate the crack around the pipe and neglecting the restraining forces of the pipe supports.
For the recirculation inlet pipe break, the break area was calculated to be equal to the pipe flow area at 9.5 milliseconds after pipe rupture. The staff said that the applicant's method of computing the break area appeared to be acceptable.
Applicant said the space beneath the reactor and within the support skirt was not included in the model for calculating pressure distribution around the reactor following a LOCA because the pressure buildup in this space would result in a negligibly small force on the vessel for pipe breaks in the annulus. This conclusion was based on the fact that openings in the support skirt providing access to the space underneath the reactor have a total area less than 0.1 times the total area of those openings in the reactor pedestal which provide exhaust from this space. The staff stated they would check applicant's conclusion and requested a drawing to show openings in the reactor pedestal.
(This drawing was subsequently provided).
f f
A. a kl C L. L. Kintner, Project Manager Light Water Reactors Branch No. 1 Division of Project Management
Enclosure:
List of Attendees cc:
See next page
Dr. Wayne H. Jens
~
cc.
Eugene B. Thomas, Jr., Esq.
David E. Howell, Esq.
LeBoeuf, Lamb, Leiby & MacRae 21916 John R 1333 New. Hampshire Avenue,, N. W.
Hazel Park, Michigan 48030 Washington, D. C.
20036 Pei.er A. Marquardt, Esq.
Co-Counsel Mrs. Martha Drake The Detroit Edison Company 230 Fairview 2000 Second Avenue Petoskey, Michigan 49770 Detroit, Michigan 48226 Mr. William J. Fahrner Dr. Wayne H. Jens Assistant Vice President Project Manager - Fermi 2 Engineering & Construction The Detroit Edison Company Detroit Edison Company 2000 Second Avenue 2000 Second Avenue Detroit, Michigan 48226 Detroit, Michigan 48226 Mr. Larry E. Schuerman i
Licensing Engineer - Fermi 2 Detroit Edison Company 2000 Second Avenue Detroit, Michigan 48226 Charles Bechhoefer, Esq., Chairman Atomic Safety & Licensing Board Panel U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Dr. David R. Schink Department of Oceanography Texas A & M University College Station, Texas 77840 fir. Frederick J. Shon Atomic Safety a Licensing Board Panel U. S. Nuclear Regulatory Comission Washington, D. C.
20555 Mr. Jeffrey A. Alson 772 Green Street, Building 4 Ypsilanti, Michigan 48197
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ENCLOSURE NRC-DECO MEETING 1/10/80 GE SAN JOSE NRC g
L. L. Kintner G. L. Davis K. Desai C. Child T. Huang R. L. Smi th C. M. Johnson (Part Time)
S. Arain R. Villa Detroit Edison E. Lusis W. Street M. Batch L. Schuerman DECO-NRC MEETING 1/11/80 FERMI-2 NRC GE L. Kintner C. M. Johnson K. Desai R. L. Smith T. Huang G. L. Davis R. Villa DECO E. Lusis W. Street M. Batch L. Schuerman