ML19241B539

Forwards Request for Addl Info as Result of Mechanical Engineering Branch Review of Fsar.Requests Amend to FSAR to Include Requested Info

ML19241B539
Person / Time
Site:	Comanche Peak
Issue date:	06/08/1979
From:	Baer R Office of Nuclear Reactor Regulation
To:	Gary R TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC)
References
NUDOCS 7907180892
Download: ML19241B539 (5)

Text

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g NUCLEAR REGULATORY COMMISSION

.; e WASHINGTON, D. C. 20555 E

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Docket Nos. 50-445 and 50-446 Mr. R. J. Gary Executive Vice President and General Manager Texas Utilities Generating Company 2001 Byran Towers Dallas, Texas 75201

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION FOR COMANCHE PEAK TEAM ZLECTRIC STATION, UNITS 1 AND 2 Enclosed is a request for additional information which we require to complete our evaluation of your application for operating licenses for Comanche Peak Steam Electric Station, Units 1 and 2.

This request for additional information is the result of our review by the Mechanical Engineering Branch of the informa-tion in your FSAR.

At this time we have not completed the review of the informa-tion submitted in FSAR Amendment 5, and our continuing review may result in additional requests or NRC staff positions by the Mechanical Engineering Branch.

Please amend your FSAR to include the information requested in the Enclosure.

Your schedule for resynding to the enclosed request for additional information should be submitted within three weeks.

Based on your schedule for response and our workload, we will determine any licensing review schedule adjustmants and inform you of any significant changes.

Sincerely,

\\

f Robert L. Baer, Chief Light Water Reactors Branch No. 2 Division of Project Management

Enclosure:

Request for Additional Information ccs w/ encl,sure:

See next page U16 011 70071'80 SM

Texas Utilities Generating Company Jun o a tg7c ccs:

Nicholas S. Reynolds, Esq.

Debevoise & Liberman t.-

1200 ' <enteenth Street

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Washin3.an, D.C.

20036 Spence: C. Relyea, Esq.

Worsham, Forsythe & Sampels 2001 Bryan Tower Dallas, Texas /5201 Mr. Homer C. Schmidt

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Project fdanager - Nuclear Plants Texas Utilities Generating Company 2001 Bryan Tower Dallas, Texas 75201 f<r. H. R. Rock Gibbs and Hi11, Inc.

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393 Seventh Avenue New York, New York 10001

.i I-f<r. A. T. Parker e'

Westinghouse Electric Corporation P. O. Box 355 L

Pittsburgh, Pennsyl vania 15230

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ENCLOSURE

-REQUEST FOR ADDITIONAL INFORMATION COMANCHE PEAK STEAM ELECTRIC STATION, UNITS 1 AND 2 TEXAS UTILITIES GENERATING COSDANY DOCKET f405. 50-445/50-146 110.0 MECHANICAL ENGINEERINr, RANCH i

112.25 Previot, analyses for other nuclear plants have shown that certain (3.9.3) reactor system components and their supports may be subjected to (3.9.2) previously underestimated asytTnetric loads under the conditions that result from the postulation of ruptures of the reactor coolant piping at various locations.

It is therefore necessary to reassess the capability of these reactor system components to assure that the calculated dynamic asynnetric loads resulting from these postu-lated pipe ruptures wil' be within the bounds necessary to provide high assurance that the reactor can be brought safely to a cold shutdown conditions. The reactor system components that require reassessment shall include:

a.

Reactor pressure vessel b.

Core support and other reactor internals c.

Control rod drives d.

ECCS piping that is attached to the primary coolant piping, e.

Primary coolant piping f.

Reactor vessel, steam generator, pressurizer, and pump supports.

The following information should be included in the FSAR abcut the effects of postulated asymetric LOCA loads on the above mentioned reactor system components and the various cavity structures.

1.

Provide arrangement drawings of the reactor vessel support systems in sufficient detail to show the geome*ry of all principal elements and materials of constructicn.

2.

If a plant-specific analysis will not be submitted for your plant, provide supporting infomation to demonstrate that the generic plant analysis under consideration adequately bounds the postulated accidents at your facility.

Include a comparison of the geometric, structural, mechanical and themal-hydraulic smilarities between

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your facility and the case anaiyzed. Discuss the effects of any differences.

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8 112-2 1979 3.

Consider all postulated breats in the reactor coolant

_e piping system, including the following locations:

a.

Reactor vessel hot and cold leg nozzle to piping terminal ends.

b.

Pump suction and discharge nozzles to piping terminal ends.

c.

Steam generator inlet and outlet nozzles to piping terminal ends. If

. rovide an assessment of the effects of asyrnetric pressure differentials 2/ on the systems and components listed aboie in combination with all external loadings including eafe shut-down earthquake loads and other faulted condition loac, for the postulated breaks described above. This assessnent may utilize the following mechanistic effects as applicable:

a.

limited displacement break areas b.

fluid-structure interaction c.

actual time-dependent forcing function d.

reactor support stiffness e.

break opening times.

4 If the results of the assesscent in item 3. above indicates loads leading to inelastic action in these systems or displacement exceeding previous design limits, provide an evaluation of the inelastic behavior (including strain nardening) of the material used in the system design anc the effect on the load transmitted to the backup structures to which these systems are attached.

1/ Postulated steam line breaks may control the design of certair steam generator supports. and therefore must also be considered ir, support design.

-2/ Blowdown jet forces at the location of the rupture (reaction

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• forces), transient differential p.essures in the annular region between the component and the wall, and transient differential pressures across the core barrel within the reactor vessel.

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t-8 IN 112-3 5.

For all analyses perforced, include the o. hod of analysis, the structural and hydraulic computer code =moloyed, drawings of the models employed, and comparisons of the calculated to allowable stresses and strains or deflections with a.

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basis for the allowable values.

6.

Demonstrate that active components will peform their safety function when subjected to the combined loads resulting from the loss-of-coolant accident and the safe shutdown earthquake.

7.

Demonstrate the functional capability of any essential piping when subjected to the combined loads resulting from the loss-of-coolant accident and the safe shutdown earthquake.

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