ML20140B458

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Forwards Addl Info for Snupps FSAR Re Mechanical Engineering,In Response to NRC 810825 Request.Info Will Be Incorporated Into FSAR Revision 7
ML20140B458
Person / Time
Site: Wolf Creek, Callaway  Wolf Creek Nuclear Operating Corporation icon.png
Issue date: 09/09/1981
From: Petrick N
STANDARDIZED NUCLEAR UNIT POWER PLANT SYSTEM
To: Harold Denton
Office of Nuclear Reactor Regulation
References
SLNRC-81-95, NUDOCS 8109140191
Download: ML20140B458 (9)
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September 9, 1981 SLNRC 81 95 FILE: 0541 SUBJ: NRC Request for Information-Mechanical Engineering Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Docket Nos.: STN 50-482, STN 50-483, and STN 50-486 Ref: NRC (Youngblood) letter to UE (Bryan) and KGE (Koester), dated August 25, 1981, Same subjact

Dear Mr. Denton:

The referenced letter requested additional information for the SNUPPS FSAR in the area of mechanical engineering.

The enclosure to this letter provides the requested information and will be incorporateu in Revision 7 to the SNUPPS FSAR.

Ver,truly yours,

\\ 4h%M\\ C

,C-Nicholas A. Petrick RLS/dck/3a29 Enclosures cc:

J. K. Bryan, UE G. L. Koester, KGE D. T. McPhee, KCPL W. A. Hansen, NRC/ Cal T. E. Varidel, NRC/WC l

D. F. Schnell, UE

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8109140191 810909 PDR ADOCK 05000482 A

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v Q210.2 The applicant states that all circumferential breaks in the RCS piping are assumed to result in a limited separation such that the maximum flow area is less than a full break area.

The appli-cant must provide the design information assumed for each location where limited break areas are postulated including gap size, restraint stiff-ness, blowdown force, and maximum restraint deflection.

The results of the time-history analysis (if used) should include the break area vs. time and mass flux rate vs. time which were used to calculate the subcompartment pressuriza-tion.

In addition, all restraint locations on the RCS piping must be shown.

RESPONSE

In the reactor coolant loop analysis described in Section 3.6, limited break areas are assumed at the reactor vessel inlet and outlet nozzles.

At these locations, the break is limited because of the physical constraint built into the plant design.

At this location, the reactor coolant piping restraints located in che shield wall annulus limit the break opening area.

A description of these restraints specifically for the SNUPPS units is contained in Section 5.4.14.

In the reactor coolant system analysis, all other circum-ferential breaks are assumed to be double-ended.

However, because of the physical configuration of the plant, these breaks are also limited in area, The specific restraint configurations which limit the break opening are described in Section 5.4.14.

Refer to revised Sections 3.6 and 5.4.14.

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210.2-1 9/81 i

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7 M66 SNUPPS Q-%l0.7 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, ac determined by the dynamic RCS analysis or the location of pipe restraints, are such that the separation results in a limited flow area.

However, in the reactor coolant loop analysis, limited break areas are only postulated at the reactor vessel inlet and outlet nozzles.

At these locations, the break area is limited to approximately one square foot.

This reduced break area is justified based on the configura-tion of the plant.

Specifically, reactor coolant piping restraints located in the shield wall annulus (as descr: bed in Section 5.4.14) limit the movement of the reactor coolant pipe such that a full double-ended break could not develop.

For all other break locations in the reactor coolant loop, full double-ended break locations are assumed.

Longitudinal breaks are assumed to have an opening area equal to one flow area of the pipe.

2.

Pipe breaks are postel.ited to occur in the following locations in Class 1 piping runs or 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 equatiors (12) and (13) of the ASME l

B&PV Code,Section III, Subsection NB, exceed 2.4 Sm, where Sm is the design stress intensity, as given in the ASME B&PV Code, ana 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 3.6-10 9/81 1

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SNUPPS NIdb G. AI O. 2.

(d)

Additional locations of maximum stress intensity or cumulative usage factor to assure a minimum of two break locations between ter;ninal ends.

A complete discussion of the reactor coolant loop break location is provided in Reference 1.

b.

ASME B&PV code,Section III - Class 2 and 3 Piping Within Protective Structures 1.

Breaks are postulated to occur at terminal ends, including:

(a)

Piping-pressure vessel or equipment nozzle intersection (b)

High-energy / moderate-energy boundary (c)

Pipe to anchor intersection 1

Rev. 7 3.6-10a 9/81

2 10. 2 SNUPPS generator outlet nozzle.

This restraint is attached to the secondary shield wall and extends horizontally to the vertical run of the crossover leg pipe.

b.

Hot leg A restraint is located near the 50-degree elbow in the hot leg to prevent exceseive displacement of the hot leg following a pos+ t ~ ated guillotine break at the steam generato

t nozzle.

This restraint consists of stru tural steel members which transmit loads to tb macrete structure.

This restraint is shown i.

_gure 5.4-20.

c.

Hot leg and cold leg lateral restraints A restraint on each reactor coolant system hot leg and cold leg is located near the reactor vessel safe-end to reactor ccolant system piping weld with the reactor vessel primary shield wall to prevent excessive displacement of either the hot leg or the cold leg follcwing a postulated guillo-tile break at the reactor vessel safe-end to piping weld.

These restraints are shown in Figure 5.4-21.

5.4.14.3 Design Evaluation Detailed evaluation ensures the design adequacy and structural integrity of the reactor coolant loop and the primary equipment i

supports system.

This detailed evaluation is made by comparing l

the analytical results with established criteria for acceptability.

l Structural analyses are performed to demonstrate design adequacy for safety and reliability of the plant in case of a large or small seismic disturbance and/or LOCA conditions.

Loads which the system is expected to encounter often during its lifetime (thermal, weight, and pressure) are applied, and stresses are compared to allowable values as described in Section 3.9(N).1.4.

The safe shutdown earthquake and design basis LOCA, resulting l

in a rapid depressurization of the the system, are required design conditions for public health and safety.

The methods used for the analysis of the safe shutdown earthquake and LOCA conditions are given in Sections 3.9(N).1.4.

5.4.14.4 Tests and Inspections NondestrLctive examinations are performed in accordance with the procedures of the ASME Code,Section V, except as modified by the ASME Code,Section III, Subsection NF.

Rev. 7 5.4-57 9/81

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