ML20136H102

Nonproprietary Comparison of Data for Beaver Valley Power Station,Unit 2,W/WCAP-9736 Data

ML20136H102
Person / Time
Site:	Beaver Valley
Issue date:	11/30/1985
From:	Bhandari D, Takeuchi K, Wills M WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML19276D141	List: ML20136H090 ML20136H102
References
WCAP-11005, NUDOCS 8601090249
Download: ML20136H102 (72)

Text

. . _ _ . - - - - - - - -

WCAP-11005 i

COMPARIS0N OF DATA FOR BEAVER VALLEY POWER STATION, UNIT 2 WITH WCAP - 9736 DATA Prepared for NRC Review in Conjunction with Review of WCAP-9736, Docket Number 50-412 NOVEMBER 1985 by D. R. Bhandari K. Takeuchi M. E. Wills WESTINGHOUSE ELECTRIC CORPORATION NUCLEAR ENERGY SYSTEMS P. O. BOX 355

_ PITTSBURGH, PENNSYLVANIA 15230 hD 0 12 A PDR

l i This package is submitted by Westinghouw Ehttric Corporation on behalf of Duquesne Light Company's Beaver Valley Power Station, Unit 2 (BVPS-2). Review of this package by the Nuclear Regulatory Commission is requested in conjunction with +.he submittal of WCAf-9736 on the BVPS-2 Docket, Number 50-412. The antent of this transmittal is to supplement the information I I

presentea in WCAP-9736, address b.g requests for BVPS-2 plant-specific data and to resolve A NRC comments which aro'se 'during the review meeting of September if, d 1985 in'ecthesda, Maryland, atter,.1ed by personnel from the NRC, Duquesne Light Company,'and Westinghouse Electric Corporation.

The text and figures enclosed are divided into seven attachments as follows:

1 1

l. CompLrison of Data for Beaver Valley Power Station, Unit 2 with Data Presented in WCAP-9736.

2. Stiffnesses and Frequencies for Beaver Valley Power Station, Unit 2

3. Initial Pressure Differentials and Initial Displacements

4. Treatment of 3-0 Structural Motion and Data of Horizontal X- and Z-Directional Quantities

5. Verification of Non-Linear Boundary Conditions

6. External Loads

7. Relative Displacements Between Barrel and Vessel 1

92960:10/112285

Attachment 1 COMPARISON OF DATA FOR BEAVER VALLEY POWEK STATION, UNIT 2 WITH DATA PRESENTED IN WCAP-9736 2

92960:10/112285

.]

01fferences between the MULTIFLEX 3.0 input data for Beaver Valley Power St'ation, Unit 2 (BVPS-2) and the Three-loop Neutron Pad plant presented in Table 7-1 of WCAP-9736 have been examined. The following actions wert taken on behalf of BVPS-2 in variation to the referenced plant:

(1) Modified Printout stations for plot coordinates for Card Group #5. This produced no effect.

(2) Increased the total number of legs (NOLEG) in the system from (

a,c

) ,

(3) Invoked the NTPLOX option, which is used to write intermediate leg and

~

node plot stations to tape. [ ]a,c (4) Recalculated the upper head volurre for leakage element #10 to reflect P BVPS-2 configuration.

(5) Recalculated the following thermal parameters for BVPS-2:

(a) Enthalpy - ENTHOT, ENTCLD (b) Saturation pressure - PSATH0T, PSATCLD (c) Fluid density - DENSHOT, DENSCLD l (d) Fluid temperature - TEMPHOT, TEMPCLD (6) Modified Card Group #21 to obtain a better steady state mass balance'.

(

l l

3 92960:10/112285

(7) Modified the loop geometry data (Card Group #11) to accurately model BVPS-2, and in particular, to locate the 8-inch bypass line properly.

This necessitated the following modifications:

~

(a) addition of [' ]a,c legs to the Broken Loop hot leg (b) addition of [ ]a,c legs to the Broken Loop crossover leg (c) aadition of [ ]"'" legs to the Unbroken Loop hot leg (d) addition of [ ]C legs to the Unbroken Loop crossover leg (e) addition of [ ]' legs to the Unbroken Loop cold leg (f) deletion of [ ]"'C extraneous legs from the downcomer This accounts for the net difference of [ ]a,c legs in the hydraulic model.

(8) Modified the loss coefficients (Card Group #12) to incorporate the BVPS-2 specific loss coefficients.

(9) Modified slightly the noding for BVPS-2 downcomer model as noted in item (7)(f) above, resulting in a cosmetic renumbering of the horizontal

'flowpaths. The internals and barrel / baffle regions remain unchanged, with the same leg-numbering scheme.

(10) Changed loop flowrate (WLOOP) to reflect BVPS-2 configuration.

(11) Modified reactor coolant pump data to reflect BVPS-2 specific data.

(12) Modified reactor power level (POWER 0) to reficct BVPS-2 specific data.

(13) Modified the structural input to be BVPS-2 specific.

9206Q:10/112085 4

. i Figure 1-1 through 1-5 present the leg and node numbering scheme for BVPS-2, which is similar to that presented in the reference WCAP, Figures 7-1 through 7-5. The interface modeling between the flexible walls and the Hydraulic piping Network for BVPS-2 is the same as shown in Figure 7-6 of the WCAP.

Table 1-1 provides a listing of the BVPS-2 MULTIFLEX 3.0 input deck, similar to the input listing shown in Table 7-1 of the WCAP.

The basis of Beaver Valley Power Station, Unit 2 (BVPS-2) structural data used

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in the three-dimensional system finite element model is the three-loop standard (generic) plant model data modified for the changes from the inverted top hat to the old-style deep beam. Also, there is a modification made for the BVPS-2 specific loop and vessel support stiffnesses. 1he effect of these modifications can be assessed by comparing the vibrational response of the BVPS-2 reactor internals with that of a three-loop standard plant given in the topical report (Table 7-1, page 7-17). Tables 1-2 and 1-3, respectively, represent the structural input data used in the MULTIFLEX Code (MFX 3.0) for BVPS-2 and the generic three-loop standard plant. It is seen from Tables 1-2 and 1-3 that the f requency response of the BVPS-2 plant is similar to that of the three-loop plant used in the topical report, WCAP-9736.

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Attachment 2 4

STIFFNESSES AND FREQUENCIES FOR BEAVER VALLEY POWER STATION, UNIT 2

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3 9206Q:10/112085 g

_ . . mus The linearized and non-linear stiffnesses at the top and bottom of the Beaver Valley Power Station, Unit 2 barrel-vessel interfaces are identical to those used in the topical report for a three-loop standard plant.

TOP BOUNDARY Linearized Stiffness = [ ] lb/in Nonlinear Stiffness:

k=[ ]' lb/in; gj=[ ]"

k=[ ]a,c lb/in; g2 *E 3 BOTTOM BUUNDARY Linearized Stif fness = [ ]a,c lb/in Nonlinear Stiffness:

]"'" lb/in; gj=[ ]a,c k) = [

k2=[ ] lb/in; g2*E 3 AP) = [ ]"'" lbf; x j=[ ]"'C AP2=[ ] lbf; x 2

• E 3 The load-deflection curves for the stiffnesses applicable to BVPS-2 are shown in Figures 2-1 and 2-2.

The core barrel fundamental f requency (in water) calculated from the time-history displacement shown in Figure 2-3 is approximately [ ]a c which is in good agreement with the value shown'in the presentation to the NRC.

9206Q:10/120585

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l Attachment 3 INITIAL PRESSURE DIFFERENTIALS AND INITIAL DISPLACEMENTS l

9206Q:10/112085 34

Equation (2-33) in the NRC-approved NCAP-8709 NP/A is a misprint, because the term of initial pressure differentials is missing. The correct one is equation (2-11) in WCAP-9736. In any event, the effect of the initial ..

pressure differentials and the resultant barrel displacements is a

insignificant. In fact, conservative estimates of these values are [ J ,c psia and [ ] mils, respectively, as detailed in the following paragraph.

In a typical 3-loop plant, primary coolant flow enters three inlet nozzles and impinges on the core barrel wall. A small fraction of this flow goes into the outlet nozzles and the upper head region. The majority of the flow turns and flows down the downcomer and into the lower plenum. As a result, this downcomer region has large variations in flow velocities. Experimeatal tests indicate that for this downcomer/ lower plenum flow path a great deal of mixing occurs. In addition, these tests show that [

a J .c However, if the loop characteristics are not the same and the loop flow rates are not exactly the same [

]a,c This was found to be the case when a review of the 3XL hydraulic flow test report was performed.

The results of this review indicated that the circumferential pressure drop a

was as high as [ J ,c psi in the downcomer (i.e, in the area of the downcomer between the nozzle and the top of the panel) with all loops operating. This value was obtained by reducing the pressure tap data.

l The core barrel displacements due to this initial pressure differential are

! made based on very conservative assumptions. Assuming that this maximum s Ap = [ ] psi is acting along the total length of the core barrel, then for a simply-supported barrel, the maximum deflection is [

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35 9206Q:1D/120585

Attachment 4 TREATMENT OF 3-D STRUCTURAL MOTION AND DATA 0F Z-DIRECTIONAL QUANTITIES l

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9206Q:1D/112085 36

l In the three-dimensional system structural model, each node has three-translational and three-rotational degrees of freedom. In order to l

compute system dynamics response during a loss-of-coolant-accident (LOCA),

each node is assigned six (6) degrees of freedom, and therefore the forces and displacements in the horizontal, lateral and vertical directions are obtained from the system response.

The total hydraulic forces in the horizontal X- and 2-directions (see the figure below) for the core berrel and the vessel are shor in figures 4-1 through 4-4. Also shown in Figures 4-5 through 4-8 are the horizontal X- and Z- relative displacements between the barrel and the vessel. It is shown that every quantity in the Z-direction is substantially less than [ ]a,c ,7 the respective X-directional valves. Therefore, it is reasonable to consider that the Z-directional fluid-structure interactions can be ignored.

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Attachment 5 VERIFICATION OF NON-LINEAR BOUNDARY CONDITIONS i

i 9206Q:1D/112085 46

l WESTINGHOUSE PROPRIETARY CLASS II l

Verification of the non-linear boundary conditions has been provided by comparison of the MULTIFLEX 3.0 model vs. WECAN calculation as shown in WCAP-9736 and in the presentation to the NRC staff. The same comparison is attached below. Good agreement of both computed results verifies the treatment of the non-linear boundary conditions.

As shown in WCAP-9736, the non-linear boundary conditions are generic to the plant with the same number of loops. Those used in the verification are for a three-loop plant. Therefore, the method of non-linear boundary conditions applies to Beaver Valley Power Station, Unit 2. '

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92960:10/112285

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FIGURE 6-1 Barrel / Vessel Relative Displacements At The Top Level i

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FIGURE 5-2 Barrel / Vessel Relative Displacements At The Inlet Nozzle Elevation 49

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FIGURE 5-3 Thermal Shield / Vessel Relative Displacements At The T.S. Middle Elevation 50

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ML171 FLEX 3.0 0,C FIGURE 5-4 Thermal Shield / Vessel Relative Displacements At The T.S. Bottom Elevation 51

Attachment 6 EKTERNAL LOADS 11206Q:10/112085 52

_. -. . . . _ _ . - = - - -. _ . . . - - . - - . - . _ . - - .

The loads on the reactor pressure vessel (RPV) and internals that result from the depressurization of the system and from the pressurization of the area around the break may be categorized as:

l (1) reactor internal hydraulic loads (vertical and horizontal):

l (2) reactor coolant loop mechanical loads

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(3) reactor cavity pressurization leads (only for breaks at the RPV safe end

!ocations)

(4) jet impingement loads (only for RPV nozzle safe end, 8-inch bypass line locations) i All the loads are calculated individually and combined in a time-history manner. The analytical methods used for the calculations are discussed in sections 6-1 through 6-4.

6.1 REACTOR COOLANT LOOP MECHANICAL LOADS l The reactor coolant loop mechanical loads are applied to the RPV nozzles b'y the primary coolant' loop piping. For guillotine pipe separations, the loop l mechanical loads result from the release of normal operating forces present in the pipes prior to the separation as well as transient hydraulic forces in the reactor coolant system. The magnitudes of the loop release forces are I

determined by performing a reactor coolant loop analysis for normal operating i

I loads (pressure, thermal, and dead-weight). The loads existing in the pipe at the postulated break location are calculated and are " released" at the i initiation of the LOCA transient by application of the loads to the broken piping ends. .

Loop release loads for Beaver Valley Power Station, Unit 2 supplied by Stone j and Webster Engineering Corporation are applied in a time history manner.

Resultant loads of each nozzle inlet and outlet of the broken loop are combined and translated to the RPV centerline. To develop loop release loads t i

j 9206Q:10/112085 53 i

normal operating condition loads were subtracted from the loop transient time histories. Plots of these total loop release forces on the RPV are given in Figures 6-2 through 6-7, and the coordinate system is shown in Figure 6-1.

6.2 REACTOR PRESSURE VESSEL INTERNAL HYORAULIC LOADS Af ter a postulated break at the RPV inlet nozzle, the depressurization path for waves entering the reactor vessel is through the nozzle which contains the broken pipe and into the reaion between the core barrel and reactor vessel.

This region is called the towncomer annulus. The initial waves propagate up, around, and down the downcomer annulus, then up through the region circumferential1y enclosed by the core barrel, that is, the fuel region.

The region of the downcomer annulus close to the break depressurized rapidly, -

but because of restricted flow areas and finite wave speed (approximately

[ ] feet per second), the opposite side of the core barrel remains at a high pressure. This results in a net horizontal force on the core barrel and RPV. As the depressurization wave propagates around the downconer annulus and up through the core, the barrel differential pressure reduces and, similarly, the resulting hydraulic forces drop.

The internal hydraulic loads on the BVPS-2 reactor pressure vessel and the core barrel calculated from the MULTIFLEX 3.0 code are shown in Figures 4-1 through 4-4. For the sake of brevity, only the total horizontal thrust on the reactor vessel and the core barrel in both the X- and Z-directions are shown.

6.3 REACTOR CAVITY PRESSURIZATION LOADS Reactor cavity forces arise from the steam and water which are released into the reactor cavity through the annulus around the broken pipe. These forces occur only for postulated breaks at the RPV nozzle safe end locations. The reactor cavity is pressurized asymmetrically, with high pressure on the side adjacent to the break. The horizontal differences in pressure across the reactor cavity result in horizontal forces' en the reactor vessel. Vertical forces on the reactor vessel arise from similar varia'tions in pressure on the upper and lower head and the tapered parts of the reactor vessel.

92060:10/120585 54 1

l Reactor cavity loads were calculated for a 150-square-inch guillotine break opening at the cold leg nozzle safe end. Loads were developed at the RPV nozzle centerline and primary shield wall. Plots of these forces are given in Figures 6-8 through 6-13. i 6.4 JET IMPINGEMENT LOADS The jet impingement load is an axial force along the broken pipe centerline which is caused by the pressure of the escaping jet of coolant acting on the exposed pipe cross-section at the break location. The jet force is calculated by multiplying the [

l l

a

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a The magnitude of jet force is [ Jc kips for the RPV inlet nozzle break.

9206Q:10/ll2085 55

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FIGURE 6-2 Loop Force FX - RV inlet Break 57

o,e FIGURE 6-3 Loop Force FZ - RV Inlet Break 58

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FIGURE 6-5 Loop Woment MX - RV inlet Break 60

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FIGURE 6-7 Loop Woment MZ - RV Inlet Break 62

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FIGURE 6-8 Cavity Pressure Horizontal Force (Fx) On RPV in x - Direction 63

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RGURE 6-9 Cavity Pressure Lateral Force (Fz) On RPV in z - Direction 64

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F10VRE 6-10 Cavity Pressure Vertical Force (Fy) On RPV in y - Direction 65

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FIGURE 6-13 Cavity Pressure Vertical Moment (Wy) On RPV in y - Direction 68

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l Attachment 7 RELATIVE DISPLACEMENTS BETWEEN BARREL AND VESSEL 9206Q:10/112085 69

The time history relative displacements between the barrel and vessel are shown in Figures 4-5 through 4-8, Attachment 4.

92060:1D/112085 70