ML19256E763
| ML19256E763 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 10/31/1979 |
| From: | Burwell S Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7911150159 | |
| Download: ML19256E763 (6) | |
Text
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7EM j o arouq(c, UNITED STATES g_
NUCLEAR REGULATORY COMMISSION o
3_
_E WASHINGTON, D. C. 20555
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g3y Docket Nos: 50-445 And 50-446 APPLICANT: TEXAS UTILITIES GENERATING COMPANY FACILITY:
COMANCHE PEAK STEAM ELECTRIC STATION, UNITS 1 & 2
SUBJECT:
SUMMARY
OF OCTOBER 17, 1979 MEETING ON STAFF ANALYSES OF DYNAMIC ASYMMETRIC LOADS ON PRIMARY COOLANT SYSTEM COMPONENT SUPPORTS Summary A meeting was held with representatives of the Texas Utilities Generating Company at the Comanche Peak Steam Electric Station near Glen Rose, Texas. The purpose of the meeting was to gather information to permit the NRC staff and its consultants (Idaho National Engineering Laboratory - EG&G Idaho, Inc.) to perform an independent analysis of dynamic loads and forces on the supports for a representative Westinghouse 4-loop pressurized water reactor. The meeting consisted of a description (scope and objectives) by the NRC staff and EG&G Idaho of the dynamic asymmetric load program, a review of the additional infor-mation requested by the consultant through the NRC staff, and a tour inside the containment to examine the primary coolant system component suppurts and nearby structures. The additional information requested is provided as Enclosure 1, and attendance list is provided as Enclosure 2.
Background
The NRC program for dynamic asymetric loads on reactor coolant system component supports is described in NUREG-0410, "NRC Program for the Resolution of Generic Issues Related to Nuclear Power Plants," Report to Congress, January 1,1978.
The information sought in this meeting, and in our letter of May 31,1979, is to be used in the conduct of the contract described in Category A Technical Activity Task No. A-2, paragraph 4.b. of NUREG-0410.
Meeting Details After introductions, we described the purposes of the NRC dynamic asymmetric load program and the analyses being perfomed by our consultant EG&G Idaho, Inc.
These are detailed in the above references and will not be repeated. We noted that this effort is not a part of the CPSES safety review, but rather relates to the development of NRC staff capabilities to independently verify results reported by applicants for similar pressurized water reactors.
1336 093 7911150 /N R
A
, (M 3 ~ 1979 Messrs. Saffell and Thinnes of EG&G Idaho, Inc. described their analyses in more detail using viewgraph slides. The presentation reviewed the problem definition, development of loads, structural analysis computer codes, finite element models, load application, results of the analyses, and use of the results.
I have placed a copy of the viewgraph slides used in this presenta-tion in the LPM Project Files, available for review on request. The analyses using CPSES as a representative plant will include loads on the steam generator supports and reactor coolant pump supports. Loads are developed from hypoth-esized loss-of-coolant accidents. Loads developed from simultaneous seismic events are not included in this analysis.
We agreed to keep the applicant advised on the progress and results of our consultant's analyses. We may request future meetings or conference calls to confirm that the modeling is accurate and representative of the CPSES design.
Also, we requested that the applicant provide comments on the analyses consistent with our objectives of making the verification model a more useful and accurate tool for confirming the results reported for the Westinghouse 4-loop pressurized water reactor.
The attendees then reviewed the information listed in Enclosure 1 in detail to assure an understanding of our consultant's request. All items will be fur-nished by applicant letters within the next two to three weeks.
The consultants and NRC staff then toured the inside of each containment to examine the primary coolant system component supports and nearby structures.
The visit inside both containments was requested because of access restrictions imposed by construction activities, and the advanced state of construction on Unit 1.
L rhfutt
> ~ (L ' L Spottswood B. Burwell Licensing Project Manager Light Water Reactors Branch, #2 Division of Project Management
Enclosures:
1.
Information Needed For Asymmetric Loading Meeting 2.
Attendance List ces w/ enclosures:
See next pages 1336 094
/
Mr. R. J. Gary OCT31jgyg Mr. R. J. Gary Executive Vice President and L.
General Manager-Texas Utilities Generating Company I
2001 Bryan Towers Dallas, Texas 75201 Nicholas S. Reynolds, Esq.
Debevoise & Liberman 1200 Seventeenth Street Washington, D.C.
20036 Spencer C. Relyea, Esq.
Worsham, Forsythe & Sampels 2001 Bryan Tower Dallas, Texas 75201 Mr. Homer C. Schmidt Project Manager - Nuclear Plants Texas Utilities Generating Company 2001 Bryan Tower Dallas, Texas 75201 Mr. H. R. Rock Gibbs and Hill, Inc.
393 Seventh Avenue New York, New York 10001 Mr. A. T. Parker Westinghouse Electric Corporation P. O. Box 355 Pittsburgh, Pennsylvania 15230 Rir'iard W. Lowerre, Esq.
Assistant Attorney General Environmental Protection Division P. O. Box 12548, Capitol Station Austin, Texas 78711 Mrs. Juanita Ellis, President Citi:ans Association for Sound Energy 1426 South Polk Dallas, Texas 75224 Geoffrey M. Gay, Esq.
West Texas Legal Services 406 W.T. Waggoner Building 810 Houston Street Fort Worth, Texas 76102 1336 095
s Mr. R. J. Gary g 3g g Ms. Nancy Holdam Jacobson Citizens for Fair Utility t
Regulation p
1400 Hemphill Fort Worth, Texas 76104 Mr. Richard Fouke 1668-B Carter Drive Arlington, Texas 76010 9
e 1336 096
ENCLOSURE 1 XT 4 2
IGFORMATION NEEDED F39 ASYMMETRIC L0f D'i., MEET";'
REQUESTED BY R. NATTU 1.
A reproducible isometric drawing of the Primary Coolant System.
2.
Elevation view of reactor cavity with di.rensions.
3.
Any lead deflection curves for component supports in tne primary system or elastic load limit for each of these component supports.
4.
Stiffnesses of reinforced concrete portions of these component supports.
5.
Stiffness and load limit of the snubbers near the top of the steam generators.
6.
Hot dimensions of the vessel supports. Shoe and nozzle support dimensions are needed to determine support gaps.
7.
A breakdown of water volume and densities within designated regions of the Reactor Pressure Vessel.
8.
Fuel Assembly drawings and first three structural frequencies of the fuel in and out of water.
9.
A breakdown of reactor pressure vessel internals weights.
- 10. Primary coolant piping wall thicknesses.
- 11. Operating temperature and pressure in RPV, hot legs, cold legs, the primary and secondary sides of the steam generator.
- 12. Operating component support temperatures.
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- s ENCLOSURE _2_
E 32159 ATTENDANCE LIST MEETIliG ON STAFF..NALYSES OF DYNAMIC ASYMMETRIC LOADS ON PCS COMP 0NENT SUPPORTS NRC - STAFF S. B. Burwell R. K. Mattu A. J. Cappucci, Jr.
EG&GIDAHO,INCI B. F. Saffell, Jr.
G. Thinnes TUSI C. K. Feist J. Marshall B. Dacko J.Ryan(tour)
WESTINGH0USE D. R. Fraser P. VanTeslaar D. A. Ferg GIBBS & HILI.
C. N. Yeh R. E. Heim R. E. Ball.trd 336 098