ML19329D702

Forwards Reactor Vessel Support Evaluation for LOCA Loadings in Response to NRC 751114 Request for Review of Design Basis

ML19329D702
Person / Time
Site:	Crystal River
Issue date:	12/19/1975
From:	Rodgers J FLORIDA POWER CORP.
To:	Schwencer A Office of Nuclear Reactor Regulation
Shared Package
ML19329D703	List: ML19329D702 ML19329D704
References
NUDOCS 8003160320
Download: ML19329D702 (3)

Text

NRC .bTRIBUTION FOR PART 50 DOCl; . f.' ATERI AL

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FROM: Florida Power Corp DATE OF DOC DATE REC'D LTH TWX RPT OTHER St. Petersburg, Fla. 33733 3.T. Rodeers 12'-19-75 12-22-75 'IX

- TO: . ORIG CC OTHEH SENT1mc PDR H Mr . A . Schwencer 1 signed SENT LOCAL PDR XX CLASS UNCLASS PROPINFO INPUT NO CYS REC'D DOCKET N05 XXX 'l 50-302 .

DESCRil' TION: Ltr trans the fol:.owing: ENCLOSURES: Info on Reactor Vessel Support i Evaluagion for LOCA L0ading...

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in MIT10,y0 PLANT NAME: Crystal River Unit 3. -

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\b December 19, 1975

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.s i Mr. A. Schwencer, Chief  %

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Director of Nuclear Reactor Regulation V\v' Branch No. 2-3 U.S. Nuclear Regulatory Commission N. .

.h' ashing ton , D. C. 20555 \ lc, , , .

In Re: Florida Power Corporation Crystal River Unit #3 Docket No. 50-302

Dear Mr. Schwencer:

In your November 14, 1975, letter you requested Florida Power Corporation (FPC) to review the design bases for the Reactor Vessel Support System for Crystal- River Unit

1. 3 to determine whether the three transient loads described in the enclosure of your letter were taken into account appropriately in the design.

In response to your request, FPC and B5W have reviewed the design basis for the Reactor Ve..sel Support System for the Crystal River Unit #3 Plant. Attachment 1 shows the model used to calculate the Reactor Vessel Support System loads resulting from a loss-of-coolant accident.

These loads were then combined with normal operating loads and seismic loads to determine the resulting total stresses on the Reactor Vessel Support System.

As noted in Attachment 1, the blowdown jet forces at the location of the break were considered in the design of the Reactor Vessel Support System. The remaining two effects; (1) transient differential pressures in the annular region between the vessel and the shield wall; and, (2) transient differential pressures across the core barrel within the reactor vessel, were not considered.

14135 General Office 3201 Tnitty-fourtn streu sourn . P O Box 14042. St Petersburg. Florida 33733 e 813-866-5151

Mr. A. Schwencer -

2- December 19, 1975 Although all of the transient effects were not utilized, the method of analysis described in Attachment I was considered to be the " State-of-the-Art" at the time this analysis was performed. In addition, no additional analyses of the Reactor Vessel Support System for Crystal

. River Unit #3 was required by the Commission as evidenced by their Safety Evaluation Report for Crystal River Unit #3, dated July 5, 1974.

Should additional disqussion concerning the design basis of the Reactor Vessel Support System for Crystal River Unit #3 be required, please feel free to contact this office.

Very truly yours, e

/

J. . Ro gers Asst. Vice President JTR/iw Attachment l

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