ML20010G664

Forwards FSAR Page 3.6-10,revising Page Submitted w/810909 Response to Mechanical Engineering Branch Request for Info. Page Will Be Included in Next FSAR Revision

ML20010G664
Person / Time
Site:	Wolf Creek, Callaway
Issue date:	09/16/1981
From:	Petrick N STANDARDIZED NUCLEAR UNIT POWER PLANT SYSTEM
To:	Harold Denton Office of Nuclear Reactor Regulation
References
SLNRC-81-104, NUDOCS 8109220372
Download: ML20010G664 (2)

Text

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fk7 Nicholas A. Petrick Exocutive Director 1301;ses4010 September 16, 1981 SLNRC 81- 104 FILE: 0541 SUBJ: MEB Review Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Connission Washington, D. C. 20555 Docket Nos. STN 50-482, STN 50-483, and STN 50-486

Reference:

SLNRC 81-95, dated September 9, 1981, NRC Request for Information - Mechanical Engineering ,

Dear Mr. Denton:

The referenced letter provided information that was requested by the NRC's Mechanical Engineering Branch. In discussions with Dr. Gordon Edison it was determined that changes to that information were required. The attached FSAR page 3.6-10 replaces the same page included with the reference. This page will be included in the next FSAR revision.

Very truly yours, 5 f%c

- Nicholas A. Petrick i RLS/mtk

, Enclosure cc: J. K. Bryan UE i D. F. Schnell UE 5 G. L. Koester KGE /

D. T. McPhee KCPL //[

T. E. Vandel NRC/WC i W. A. Hansen NRC/ CAL 8109220372 010916 PDR ADOCK 05000482 A PDR

SNUPPS C 'M 01 in order to verify the design basis break loca-tions in the reactor coolant loop noted therein.

At all postulated circumferential break loca-tions, the maximum loop, piping displacements, as determined by the dynamic RCS analysis or the location of pipe restraints, are such that the separation results in a limited flow area.

- For all postulated circumf erential breaks, hydraulic forcing f unctions associated with full double-ended breaks are as .amed in the RCS structural analysis except for the reactor vessel inlet and outlet nonle breaks. At these locations the break area is limited to approximately r:ne square foot. This reduced break area is justified based on the configuration cf the plant. Specifically, reactor coolant piping restraints focated in the shield wall annulus (as described in Sectm S.4.14) Timit the movement of the reactor coolant pipe auch that a full double-ended break could not develop.

When performing other plant analyses such a the RCS piping jet impingement analyses and containment mass and energy release calculations, limited break areas are assumed at all other postulated circumferential break locations in the RCS. The application of limited break areas is based on RCS piping restraint design, primary compor.ent support design, and maximum calculated RCS displacements. The Westinghouse-designed RCS restraints and primary component supports physically limit RCS displacement following a postulated pipe break. Generic analyses performed by Westinghouse were used tri determine conservative upper bound values of maximum break opening areas at each postulated break location in the Westinghouse RCS. These limited break areas were used in the SMIPPS jet impingement and containment mass and energy release analyses. Longitudinal breaks are assumed to have an opening area equal to one flow area of the pipe.

2. Pipe breaks are postulated to occur in the following locations in Class 1 piping runs or I branch runs outside the primary reactor coolant loops and pressurizer surge line as 'follows:

(a) The terminal ends of the piping or branch run. ,

(b) Any intermediate locations between the terminal ends where stresses, calculated using equations (12) and (13) of the ASME B&PV Code,Section III, Subsection hT, exceed 2.4 Sm, where Sm is the design stress intensity, as given in the ASME

. B&PV Code, and the stress range calcu-lated, using equation (10) of the ASME B&PV code, exceeds 2 4 Sm.

(c) Any intermediate locations between ter-minal ends where the cumulative usage factor, derived from the piping fatigue

- analysis, under the loadings associated with the OBE and operational plant condi-tions, exceeds 0.1.

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Rev. 7 4 3.6-10 9/81 1 1

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