Text

Rev 3 to Core Seismic Input Palisades Nuclear Power Plant
	ML18057A405
Person / Time
Site:	Palisades
Issue date:	09/30/1989
From:	Nakaki D, Salmon M, Wesley D; ABB IMPELL CORP. (FORMERLY IMPELL CORP.)
To:	;
Shared Package
ML18057A404	List: ML18057A405;
References
	11-0540-0020, 11-0540-0020-R03, 11-540-20, 11-540-20-R3, NUDOCS 9008300013
	Download: ML18057A405 (148)
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I IMPELL 0540-018-1831 11-0540-0020, Rev. 3 CORE SEISMIC INPUT PALISADES NUCLEAR POWER PLANT by D. A. WESLEY M. W *. SALMON D. K. NAKAKI Prepared for CONSUMERS POWER COMPANY Jackson, Michigan September 1989 OOB1b

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I REPORT APPROVAL COVER SHEET CLIENT: CONSUMERS POWER COMPANY.

PROJECT:

PALISADES FUEL SEISMIC INPUT 0540-018-1831 JOB NUMBER(S):

REPORT TITI..E:

CORE SEISMIC INPUT - ?ALISAOES tlUCLEAR POWER PLANT REPORT NUMBER:

11-0540-0020 REVISION RECORD REV.

PREPARED REVIEWED,,.

APPRO~D DATE 1

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I Section 1

2 3

4 5

6 REFERENCES 11-0540-0020, Rev. 3 TABLE OF CONTENTS Tit.le INTRODUCTION * * *. *.....

1.1 Scope..**..

1.2 Plant Description....

1.3 Organization of the Report.*.*..*.

STRUCTURE MODEL AND SOIL-STRUCTURE INTERACTION

2.1 Structure Model 2.2 Soil Compliance Functions NSSS MODEL 3.1 Steam Generator **..

3.2 Reactor Pressure Vessel 3.3 Reactor Coolant Pum~s....

3.4 Reactor Coolant Loop Piping 3.5 Coupled NSS~

3.6 Combined NSSS-SSI Model COM9INED NSSS-SSI MODEL 4.1 Combined NSSS-SSI Modal Frequencie~

4.2 Composite Modal Damping Ratios SEISMIC INPUT SEISMIC RESPONSE 6.1 Core Support Time-Histori~s 6.2 In-Structure Response Spectra 1-1 1-2 1-3 1-3 2-1 2-1 2-1 3-1

. 3-1 3-3 3-3 3-4 3-6 3..:a 4-l 4-1 4-1 5-1 6-1 6-1 6-2

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I Table 2-1 2-2 2-3 3-1 3-2 3-3 3-4 3-5 3-6 3-7 3-8 3-9 3-10 3-11 3-12 3-13 3-14 3-15 3-16 3-17 3-18 3-19 11-0540-0020, Rev. 3 LIST OF TABLES Title Member Properties of the Two-Stick Conta'ii1ment Building Model Shown in Figure 2-1.*

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Tributary Weights for the Two-stick Cont'ainment Building Model Comparison of the Soil Compliance Functio.ns for the Palisades Containment Building Steam Generator Member Properties.**.***

Steam Generator Bottom Support Core Influerice Coefficient Matrix Steam Generator Nodal Weights and Mass Moments of Inertia Steam Generator Component Frequencies Reactcr Pressure Vessel Member Properties Reactor Pressure Vessel Weights and Mass Moments of Inertia Reactor Pressure Vessel Support Influence Coefficient Matrix Reactor Pressure Vessel Component F~equencfes Reactor Coolant Pump Member Properties..**.

Reactor Coolant Pump Weights and Mass Moments of Inertia..........*..

Reactor Coolant Pump Support Influence Coef-ficient Matrix {Element 1)

Reactor Coolant Pump Support Influence Coef-ficient Matrix (Element 4)

Reactor Coolant Pump Co~ponent Fr~quencies

Primary Coolant Loop - Cold Leg (Loop "A")

Primary Coolant Loop - Cold Leg (loop "B")

Primary Coolant Loop - Cold Leg (Loop C")

Primary Coolant Loop - Hot Leg (Loop "D")

Loop Joint Loads Coupled NSSS Frequencies for Fixed-Base Support Conditions for All Components ii Paqe --

2-3 2-4 2-5 3-10 3-11 3-12 3-13 3-14 3-15 3-16 3-17 3-18 3-19 3-20 3-21 3-22 3-23 3-24 3-25 3-26 3-27 3-28'

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Table 3-21 4-1 4-2 4-3 4-4 4-5 5-1

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LIST OF TABLES (CONTINUED)

Title Coupled NSSS Frt!quencies for Sliding Support Conditions -for the Pumps and Steam Generators...

Comparison of RPV Response Combined NSSS-SSI Modal Frequencies (Hz)

Comparison of Combined.NSSS-SSI Modal Frequencies with Separate SSI and NSSS Frequenci.es (in Hz)

Median Soil Case Mbdal Damping Ratios for the Combined.NSSS-SSI Hodel - Upper Soil Case Hodal Damping Ratios for the Combined KSSS-SSI Hodel - Median Soi 1 Case Modal Oaniping Ratios for the Combined*NSSS-SSI Model - Lower Soil Case

.*..... ~.

Earthquake Records..... *........

iii 3-g9.

3-30 4-3 4-4 4-6 4-7 s~3.

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Figure 1-1 1-2 2-1 3-1 3-2 3-3 3-4 3-5 3-6 3-7 3-8 3-9 3-10 3-11 3-I2 3-I3 3-I4 3-I5 5-1 5-2 6-I 6-2 6-3 11-05.40-0020, Rev. 3 LI ST OF FIGURES Title Plan View of Containment Building. Auxiliary Building,* and Turbine Buildings Showing 'the Arrangement of Major Equipment Transverse Section of the Contairvnent and Turbine Buildings...**********

Two-stick, Lumped-Mass Model of the Containment Builqing and Internal Structure NSSS Plan View NSSS Elevation Steam Generator Lumped-Mass Hodel Reactor Pressure Vessel Lumped-Mass Hodel Reactor Pressure Vessel Internals Cross-Section.

Reactor Coolant Pump Lumped-Mass Model.

Pump P-50B to Steam Generator E-408 Pump P-50B to Steam Generator E-508 Pump P-50B to.Steam Generator E-SOB Pump P-50B to Steam Generator E-508 Palisades Primary Coolant Looo - Cold Leg -

Pumps IA and 2A 'to Vessel Palisades Primary Coolant Loop - Cold Leg -

Pumps IA and 2A to Vessel Palisades Primary Coolant Loop - Cold Leg -

Pumps IB and 2B to Vessel Palisades Primary Coolant Loop - Hot leg Palisades Primary Coolant Loop - Hot Leg Ground Response Spectra Comparison of Upper Bound, Median, and Site-Specific Spectra.......**. ~ * :.

Displacement Time-History at Upper Reactor Core Support for Taft S59E Record, Soft Soil Case..

Displacement Time-History at Lmrer ~eactor ~ore Support for Taft S69E Record, Soft Soil Case.*

Displacement Time-History at Upper Reactor Core Support for Coyote Lake Record, *Soft,Soi 1 Case iv Page*

1-4 1-5 2-6 3-31 3:-32 3-33 3:_34 3-35 3-36 3-37 3-3'8 3-39 3-41 3-42 3-43 3-44 3-45 5-4 5-5 6-4 6-5 6-6

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I Figure 6-4 6-5 6-6 6-7 6-8 6-9 6-10 6-11 6-12 6-13 6-14 6-15 6-16 6-17 6-18 6-19 11-0540-00200 Rev. 3 LIST OF FIGURES (Continued)

Title Displacement Time-History at Lower Reactor Core Support for Coyote Lake Record, Soft Soil Case Displacement Time-History at Upper Reactor Core Support for Castaic Record, Soft Soil Case Displacement Time-History at Lower Reactor Core Support for Castaic Record, Soft Soil Case

Displacement Time-History at Upper Reactor Core Support for El Centro S40E Record, Soft Soil Case G

Displacement Time-History at Lower *Re~ctor Core Support for El Centro S40E Record, Soft" Soi 1 Case Displacement Time-History at Upper Reactor Core.

Support for El Centro NOOE Record, Soft Soil Case Displacement Time-History at Lower Reactor. ~ore.

Support for El Centro NOOE Record, Soft Soil Case

Displacement Timt:-History at Upper. Reactor _Core Support for Taft SS9E Record, Mediari Soil Ca~e Displacement Time-History at Lower Rear.tor Core Support for Taft S69E Record, r1edian Soil Case Displacement Time-History at Upper Reactor Core Support for Coyote Lake Record, Median Soil Case Displacement Time-History at Lower Reactor Core Support for Coyote Lake Record, Median Soil Case Displacement Time-History at Upper Reactor Core Support for Castaic Record, Median Soil Case Displacement Time-History at Lower Reactor Ccre Support for Castaic Record, Median Soil Case Displacement Time-History at Upper Reactor Core Support for El Centro S40E Record, Median Soil Case Displacement Time-History at Lower Reactor Core Support for El Centro S40E Record, Median Soil Case Displacement Time-History at Up.per Reactor Core Support for El Centro NOOE Record; Median Soil Case v

6-7 6-8 6-9 6-10 6-11 6-12 6-13 6-14 6-15 6-16 6-17 6-18 6-19 6-20 6-21 6-22

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Figure 6-20 6-21 6-22 6-23 6-24 6-25 6-26 6-27 6-28 6-29 6-30 6-31.

6-32 6-33 6-34 6-35 LIST OF FIGURES (Continued}

Title Displacement Time-History at Lower Reactor Core Support for El Centro NOOE Record, Median Soil Case Displacement Time-History at Upper Reactor Core Support for Taft S69E Record, Stiff Soil Case..

Displacement Time-History at Lower Reactor Core Support for Taft S69E Record, Stiff Soil Case Displacement Time-History at Upper Reactor Core Support for Coyote Lake Record, Stiff Soil Case.*

Displacement Time-History at Lower Reactor Core Support for Coyote Lake Record, Stiff Soil Case Displacement Time-History at Upper Reactor Core Support for Castaic Record, Stiff Soil Case..**

Displacement Time-History at Lower Reactor Core Support for Castaic Record, Stiff Soil Case..

Displacement Time-History at Upper Reactor Core Support for El Cf.ntro S40E Record, Stiff Soil Case Displacement Time-History.at Lower Reactor Core Support for El Centro S40E Record, Stiff Soil Case Displacement Time-History at Upper Reactor Core Support for El Centro NOOE Record, Stiff Soil Case Displacement Time History at Lower Reactor Core Support for El Centro NOOE Record, Stiff Soil Case In-structure Response Spectra for Five Earthquake Records - Concrete Internal Structure (EL 616')

Median Soil Case - (5% Damping).

Median and Upper Bound In-Structure Response Spectra - Concrete Internal Structure (EL 616') -

Median Soil Case - 5% Damping Top of Base Mat (EL 590'), Median Spectral Acceler-ation, Soft Soil Case..

Top of Base Mat (EL 590~), Upper Bound Spectral Acceleration, Soft Soil Case........

Concrete Internal Structure (EL 616'), Median Spectral Acceleration, Soft Soil Case vi 6-23 6-24 6-25 6-26 6-27 6-28 6-29 6-30

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6-31 6-32 6-33 6-34 6-35 6-36 6-37 6-38

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LIST OF FIGURES (Continued)

I Figure Title Page I

6-36 Concrete Internal Structure (EL 616'), Upper Bound Spectral Acceleration, Soft Soil Case 6-39 6-37 I

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6-39 Concrete Internal Structure (EL 649'), Media:i Spectral Acceleration, Soft Soi.1 Case 6-40 Concrete Internal Structure (EL 649'), Upper Bound Spectral Acceleration, Soft Soil Case 6-41 Contairvnent Building (EL 730'), Median Spectral Acceleration, Soft Soil Case 6-42 I

6-40 Containment Building (EL 730'), Upper Bound Spectral Acceleration, Soft Soil Case 6-43 I

6-41 6-42 l

6-43 Top of Base Mat (EL 590'), Median Spectral Acceler-ation, Median Soi 1 Case 6-44 Top of Base Mat (EL 590'), Upper Bound Spectral Acceleration, Median Soil Case........

6-45 Concrete Internal Structure (EL 616'), Median Spectral Acceleration, Median Soil Case 6-46 I

6-44 Concrete Internal Structure (EL 616'), Upper Bound Spectral Acceleration, Median Soil Case 6-47 6-45 J

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6-47 Concrete Internal Structure {EL 649'), Median Spectral Acceleration, Median Soil Case 6-48 Concrete Internal Structure (EL 649'), Upper Bound Spectral Acceleration, Median Soil Case 6-49 Containment Building (EL 730'), Median Spectral Acceleration, Median Soil Case.........

6-50 1

6-48 Containment Building (EL 730'), Upper Bound Spectral Acceleration, Median Soil Case...........

6-51 6-49 I

6-50 I

6-51 Top of Base Mat {EL 590'), Median Spectral Acc~ler-ation, Stiff Soil Case..............

6-52 Top of Base Mat {EL 590'), Upper Bound Spectral Acceleration, Stiff Soil Case 6-53 Concrete Internal Structure {EL 616'), Median Spectral Acceleration, Stiff Soil Case 6-54 I

6-52 Concrete Internal Structure {EL 616'), Up.per Bound Spectral Acceleration, Stiff Soil Case 6-55 I

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LIST OF FIGURES (Continued)

I Figure Title Page I

6-53 Concrete Internal Structure (EL 649'). Median Spectral Acceleration, Stiff Soil Case *..

6-56 6-54 I

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6-56 Concrete Internal Structure (El 649'). Upper Bound Spectral Acceleration, Stiff Soil Case 6m57 Containment Building (El 730'), Median Spectral Acceleration, Stiff Soil Case

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6-sa Containment Building (EL 730'), Upper Bound Acceler-ation, Stiff Soil Case.

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11-0540-0020, Rev. 3 I

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1.

INTRODUCTION This report describes the analysis conducted to develop the seismic input to the reactor core of the Palisades Nuclear Power Plant.

The seismic input is defined by time-history displacement records at the upper and lower core supports.

This input can then be used by others to determine the seismic response of the core, including nonlinear response effects if any.

The seismic adequacy of the Palisades plant was reconfirmed.

as part of the systematic Evaluation Program (SEP) for a number o~

equipment items as well as the essential civil structures (Ref. 1).

The essential civil structures were analyzed for seismic loads developed from.soil-structure interaction models which included a broad range of soil prope~ties. New in-structure response spectra were developed at selected locations including the Reactor Pressure Vessel support.

During the SEP review, insufficient information was available to comment on the seismic adequacy of the Nuclear steam Supply system (NSSS) including the Reactor Coolant Pumps (RCP), steam Generators (SG), Reactor Piping Vessel (RPV) and vessel internals, and the Main Coolant L6op Piping.

However, the major NSSS components were designed for combiPed seismic horizonta~ and vertical static loads as shown below acting simultaneously.

Horizontal Vertical Component (g)

(g)

RPV 0.468

0. 31.2

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SG 0.2 0.2 RCP 0.55 0.55 1-1

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11-0540-0020, Rev. 3 No dynamic frequencies of the individual components were available for the SEP review nor was a seismic model of the overall NSSS apparently ever developed. However, in order to determine the appropriate seismic input to the core, it is.necessary to account for any dynamic amplification in the NSSS system which could result in increased accelerations at the reactor core support locations compared with the seismic response at the corresponding elevations of the concrete internal structure.

In order to account for. any dynamic amplification in the NSSS which could affect the RPV internals, a three-dimensional lumped mass model of the NSSS including the RPV, RCP, SG, and main coolant loop piping was developed as described in this report. This model was incorporated into the existing SEP SSI model of the reactor containment building and concrete internal structure. Time-history seismic responses at the core support locations within. the RPV were then obtained for the same broad range of soil properties used in the SEP evaluation.

1.1 Scope The analysis described in this report was devoted to establishing the expected seismi~ inputs to the RPV internals. This involved the development of a NSSS dynamic model of sufficient detail to characterize the effects of the NSSS components on the seismic response of the RPV and hence, the RPV internals.

The approach used and the criteria employed as well as the existing structure model and soil compliance functions (Ref. 1) were con-sistent with recommendations for SEP plants (Ref. 2)

An assessment of the seismic adequacy of any of the major NSSS components or their anchorage, or compliance with current design criteria such as the ASME Code (Ref. 3) o~ other codes and standards was not included in the current scope.

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11-0540-0020, Rev. 3 1.2 Plant Description The Palisades plant is owned and operated by Consumers Power Company (CPC} and is located on the east shore of Lake Michigan approximately 16 miles north of Benton Harbor, Michigan. The plant is a Pressurized Water Reactor which was designed to produce 845 MW of electrical power.

The NSSS was designed and supplied by combustion Engineering, Inc., and the Bechtel Corporation designed and supplied the remaining civil structures and equipment.

The NSSS is a two-loop system with four RCP' s. The overall plant orientation including the NSSS is shown in Figure 1-1. Figure 1-2 shows a transverse section of the containment and turbine buildings.

The containment building is a post-tensioned concrete structure founded on a separate base slab. The reinforced concrete internal structure provides the supports for the NSSS.

1.J organization of the Report In Chapter 2, the civil structure model of the reactor building and concrete internals structure is discussed, together with the range of soil properties, soil compliance functions, and SSI methods used in the analysis. Chapter 3 provides a description of the NSSS model developed for Palisades including a summary of the NSSS fundamental frequencies. Chapter 4 descri.bes the combined reactor building structure -

NSSS model including the composite modal damping ratios developed for the overall system.

The natural earthquake records selected for use in developing the seismic input to the reactor core are discussed in Chapter 5 including comparisons of their ground response spectra with the smoothed design spectra.

Chapter 6 presents the seismic response results at the core support locations to be used for the seismic evaluation of the reactor core.

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H.P. turbine Operating floor -

Water treatment building (below)

Water Intake structure (below)

Moisture separator Warehouse (below)

\\ Corridor Control room L. P. turbine Generator Exciter Spent fuel pool Aux: bay roof Railroad track Radwaste treatment (below)

Figure 1-1.

Plan View of Containment Building, Auxiliary Building, and Turbine Building, Showing the Arrangement of Major Equipment (Ref. 1) 6 U'I

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Water intake structure D

Circulation water pump Condense;-

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Transverse Section of the Containment and Turbine Buildings (Ref, 1)

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STRUCTURE MODEL AND SOIL-STRUCTURE IHTERACTIOH 2.1 Structure Model The dynamic model of the reactor containment building and concrete inte!nal structure described in the Palisades FSAR (Ref.

4) was reviewed during the SEP evaluation (Ref. 1). This model is a two-stick lumped mass model as shown in Figure 2-1. The structure model developed for the FSAR design analysis was used without change in the SEP.

Only the soil springs were modified during the SEP to more accurately reflect the expected seismic behavior of the structure on the Palisades site.

Table 2-1 shows the member properties used for the model and Table 2-2 shows the tributary weights associated with the lumped masses.

In this model, the NSSS weight was included in the concrete internals model. For the current analysis reported here, the inclusion of. a separate model of the NSSS was necessary in orde.r to obtain the dynamic responses at the core supports.

Consequently, the concrete internal model was modified by subtracting the mass of the NSSS from the mass of the concrete internal structure.

Table 2-2 also shows the weights of the concrete internal structure refle6ting the separation of the NSSS.

2.2 Soil Compliance Functions Figure 2-1 also shows the equivalent soil springs which were used in the FSAR analysis (Ref. 4).

These springs were based on an elastic half-space analysis using a shear modulus, G, of

6. 4x106 lb/ft2 and a Poisson's ratio of O. 25. No effects of embedment or any layered site characteristics were accounted for in the FSAR analysis.

As part of the SEP evaluation (Ref. 1), new soil compliance functions were developed for the containment building.

These frequency-independent compliance functions account for the lay~red site and the ernbedment of the structure. The SEP compliance functions 2-1

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11-0540-0020, Rav. 3 were developed for a 60-foot radius structure embedded to a depth of 17 feet, with a soil layer of 150 foot to bedrock.

A median shear modulus of G = 4.38x106 lb/ft2 was used for the soil layer which accounted for the reduction in soil stiffness expected at the O. 2g SSE.

Also, a Poisson's ratio of O. 4 5 W;ts used which was considered appropriate for saturated sands.

Using the simpli!ied approach developed by Kausel (Refs. 5 and 6), the embedment stiffening effects due to the lateral soil pressures developed during a seismic event were included.

However, since a cohesionless soil is unable to develop any significant tensile capacity, cracks may be expected to form between the vertical surfaces and the soil during an earthquake in the o. 2g range.

Consequently, only one-half the theoretical embedment effect was used for the horizontal and rocking compliance functions, and no embedment effect was used for the vertical compliance function. Embedment of the structure is shallow; thus, the kinematic coupling between the horizontal and rocking I

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1 cornpl iance functions was neglected.

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In order to account for uncertainties in the dynamic soil properties, the SEP evaluation included analysis cases using a shear modulus, G, of +/- 50% variation from the median case. This resulted in a range of soil modulus values from 30% to 90% of the low-strain value.

A value of 5% of critical damping was used for the soil hysteretic damping for all soil cases.

Table 2-3 shows the layered soil compliance functions used for the three soil cases in the SEP evaluation together with the original FSAR design values.

The same SEP values for all three soil cases were used in the current investigation.

Thus, with the exception of the reduction of mass in the concrete internals model to reflect the NSSS weight, the SSI model of the concrete structure and soil compliance functions used in the current investigation is identical with that used for the SEP evaluation, in~luding th~ same broad range of soil properties..

2-2

Member 1 -

5 6

7A 7B 8

9 -

11 12 -

13 14 11-0540-0020, Rev.

Table 2-1 Member Properties of the Two-stick Containment Building Model Shown in Figure 2-1 (from FSAR, Amendment 15).a Ax,ft2 Ay, ft2 Iz, ft4 103 E,

kip/ft2 1, 310 874 2,340,000 792 1,429 814 833,513 576 1,451.2 865 465,209 576 1,188.7 811 384,312 576 1,376 865 336,503 576 278,000 185,000 9,999,999 792 10.0 6.67 100.0 683 10.0 6.67 100.0 1,840 103 G,

kip/ft2 338 246 246 246 246 338 aAbbreviations:

A, shear area; Iz, moment of inertia; E, modulus of elasticity; G, shear moduluf=.

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11-0540-0020, Rev. 3 Table 2-2 Tributary Weights for the Two-stick Containment Building Model Elevation Weight c ioJk)

Weight ( 103k)

Node Locatiora (ft.)

(Orig. Model)

(New Model) 1 Base Slab 590 26.6 SAME 2

Containment 608.75 7.39 SAME 3

Containment 646.25 7.39 SAME 4

Containment 683.75 7.39 SAME 5

Containment 721.25 7.39 SAME 6

Containment 765 16.73 SAME 7

Concrete Internals 606 9.51 7.55 8

Concrete Internals 622 7.48 5.95 9

Concrete Internals 649 6.437 5.024 TOTAL ~SSS Weight = 4.90x103 kip 2-4 I

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Table 2-3 comparison of the Soil Compliance Functions for the Palisades Containment Building (Refe_rence 1)

Original Design Lower Bound Median Soil Upper Bound Soil Soil Spring Constantsa kh, lb/ft l.B5x109 0.96x109 1.92x109 2.a1x109 A: 0, lb-ft/rad 4.92x1012 3.27x1012 6.59x1012 9.8ox1012 ky, lb/ft l.37x109

1. 45x109 2.B7x109 4.34x109 Damping Constantsb Dh, %

34 34 34 D~, %

1 1

1 Dv, %

6 6

6 a

Subscripts h,

~, and v, represent horizontal, rocking and vertical modes, respectively.

b 5% soil damping was used for all modes.

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Figure 2-1 Elevation, ft 765-Containment 1 vessel

. 721.25 - 5 2

4 683.75-J 646.25-3 4

2 608.75-5 L549 Concrete 6

8 internal

_ 622 structure 17 -606 a

KH = 1.54 X 108 lb/in.

. 9 1 120 ft 11 14 13 Kv = 0.57 x 108 lb/in.

Two-Stick, Lumped-:*1ass Model of the Containment Ruilding and Internal Structure (from FSAR, Amendment 15) 2-6

11-0540-0020, Rev. 3

3.

NSSS HODEL l

The Palisades NSSS is a two-loop PWR system consisting of J

two ste.am generators, four reactor coolant pumps, the reactor pressure vessel and a pressurizer.

The configuration of the pressurizer and its mounting is such that essentially no effects of the pressurizer seismic response will be felt by the RPV and the RPV internals.

Lumped-mass models of the SG,

RPV, and the RCP were developed.

These models are described in subsequent sections together with their uncoupled response frequencies and mode shapes.

An overall three-dimensional model of the NSSS consisting of two steam generators, four reactor coolant pumps, and the reactor pressure vessel connected by the main coolant piping was then developed.

The response frequencies for the combined NSSS model

'ncluding the appropriate component boundary conditions are also hown.

Figures 3-1 and 3-2 show the plan view and elevation of the coupled NSSS, respectively.

3.1 Steam Generator A nine-node, lumped-mass model of the steam generators was developed as shown in Figure 3-3.

The vessel is supported by a support cone (Element 1) which rests on Lubrite pads.

The base is free to slide in the x-direction (see Figure 3-1).

However, it is restrained against translation in the z-direction and against rotation in both directions by anchor bolts in slotted holes in the base (Elevation 615 '-1"). The upper steam generator lateral response is resisted by snubbers in the x-direction at Elevation 652 1 -6" and by trunnions in the z-direction at Elevation 659 '-6". This restraint system allows thermal expansion in the x-direction (approximately E-W) radially from the RPV but provides positive restraint in the z-direction, tangential to the RPV.

3-1

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I 11-0540-0020, Rev. 3 The lumped-mass model of the SG shown in Figure 3-3 was developed with beam elements representing the outer shell properties for Elements 2 through 9.

Element 1 is modelled by an influence coefficient matrix developed for the support cone which is not accurately represented by a beam. The beam properties for the model are shown in Table 3-1, and Table 3-2 shows the global influence coefficient matrix for the support cone.

The units of the matrices are in per lb, in per in-lb, radian per lb or radian per in-lb.

Lumped masses at Nodes 2, 3, 4, 5, 7, and 9 were developed which include the outer shell, internal steel. components, and enclosed water.

Nodes 1, 6, and 8 are massless support locations.

Table 3-3 shows the nodal weights and diametral mass moments of inertia.

Note that different weight distributions were used to model the horizontal and vertical response characteristics.

This was done since the majority of the weight of the water tends to be reacted at the bottom head for vertical response rather than being more uniformly distributed as in the case of horizontal response.

The lower response frequencies of the uncoupled steam generator model are shown in Table 3-4.

These frequencies were developed from the lumped-mass model previously described but with a fixed-base boundary condition in both the x-and z-directions rather than the sliding base condition which actually exists. Also, al though the tributary mass of the reactor coolant piping is included (Node 2), no stiffness associated with the pipe is included.

consequently, the frequencies presented in Table 3-4 may be con-sidered as lower. bounds of the frequencies which could be expected at very low response levels (i.e., at levels such that the friction in the Lubrite pads is sufficient to prevent sliding), but are lower I

bounds for this condition since the stiffening effect of the reactor coolant piping is omitted.

3-2

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11-0540-0020, Rev. 3 3.2 Reactor Pressure vessel The reactor pressure vessel (RPV) is a symmetric vessel supported by massive supports under three of the nozzles which are rigidly embedded in the concrete internal structure.

The RPV was modelled as a nine-node, lumped-mass modc;.l as shown in Figure 3-4.

Beam elements were used for all elements of the RPV except for the support system for which an influence coefficient matrix was developed (Table 3-7).

The beam properties are shown in Table 3-5.

Table 3-6 shows the lumped weights and corresponding diametral mass moments of inertia.

Again, a different weight distribution was used for the horizontal and vertical response models since the majority of the weight of the water, is reacted at the bottom head for vertical response. A vertical cross-section of the RPV showing the vessel internals is shown in Figure 3-5. The core barrel extends from approximately Node 2 (Elevation 598 '-11") to Node a (Elevation I

624 1 1/2").

The region of primary interest for the fuel reload analysis is the fuel assembly region which extends from Node 2 to Node 4 (Elevation 614'-9").

The frequencies of the uncoupled RPV were also developed in a similar fashion to those of the steam generators.

Table 3-8 lists the lower three frequencies for the uncoupled RPV.

Since the RPV is symmetric-, the uncoupled lateral response frequencies in the x-and z-directions are identical.

The uncoupled model includes the tributary weight of the connected primary coolant loop piping but no stiffness associated with th~ piping.

3.3 Reactor coolant Pumps The reactor coolant pumps are supported on a four-column frame constructed of 24-inch diameter pipes resting on Lubrite pads at Elevation 608 1 -6 11

The pump drive motor is supported by a steel cone with two large cut-outs attached to the pump as shown in Figure 3-6. The pump body and the motor frame were assumed to bP. essentially rigid.

Thus, the flexibility in the system consists of the pump 3-3

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I 11-0540-0020, Rav. 3 support frame, the motor support cone, and the attached reactor coolant loop piping. Influence coefficient matrices were developed for the support frames and motor support cone as shown in Tables 3-11 and 3-12, respectively.

As can be noted from Table 3-12, the stiffness of the motor support cone is substantially different in the two principal directions due to the two large cut-outs required for access to the drive shaft coupling.

The orientation of these cut-outs was not known at the time this investigation was conducted.

Presumably, the global orientation could vary from pump to pump.

In order to evaluate the maximum ef feet of the seismic response of the remainder of the NSSS on the RPV internals, it was decided to orient the assumed cut-out locations all in one direction such that the lowest frequency (and hence, highest response) of all four pumps corre-sponded to the maximum expected ex-direction) response of the steam generators (i.e., in the sliding direction).

3.4 Reactor coolant Loop Piping.

Flexibility matrices were generated for four piping loops of the primary coo!dnt system by Consumers Power Company (Ref. 7).

The coordinate system employed for each of the evaluations is the same and is shown in Figures 3-1 and 3-2. The matrices were generated by imposing a unit load in the global coordinate system directions noted in the figures. The resultant deflect.ions constituted columns of the resultant flexibility matrix. The units of the matrices are in per lb, in per in-lb, radian per lb or radian per in-lb.

The first column in each table represents displacements and rotations induced by a unit force in the positive x-direction. The second and third columns are the response to a unit force in the y-and z-directions.

The fourth through sixth columns are the response to unit inch-lb moments imposed in a positive sense about the x, y, and z axes.

3-4

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11-0540-0020, Rev. J In order to develop an association between Figures, Tables and Plots, the following correspondence is identified.

Loop A joins P-lB (Pump P-SOB) and E-50B (steam generator

1).

Table 3-14 represents the flexibility matrix, and Figures 3-7 through 3-10 characterize the geometry.

The flexibility matrix was determined at Node 40 - the steam generator joint. This matrix was used (with use of symmetry) to reflect the flexibility of all four pumps to steam generator loops.

Loop B joins Pump lA (2A on the other side) to the reactor vessel.

Table 2-15 represents the flexibility matrix and Figures 3-11 and 3-12 characterize the geometry. The matrix was calculated at Node 5 -

the pump joint.

This matrix (with symmetry) was used to represent two loops.

Loop C joins Pump lB (2B on the other sid&) to the reactor vessel.

Table 3-16 represents the flexibility matrix and Figure 3-13 reflects the loop geometry.

The matrix was calculated at Node 5 -

the pump end.

This matrix (with symmetry) was used to represent two loops.

Loop D joins the vessel to steam generator 1 (or E-50B).

The geometry is reflected in Figures 3-14 and 3-15.

Table 3-17 contains the hot leg matrix.

Node 30 is the joint at which the matrix was calculated - the generator ends. This matrix was used (with symmetry) for both hot legs.

oeadweight analyses have also been conducted on all four lcops. _

These analyses have been conducted simply to arrive at joint loads.

Table 3-18 shows the loads which the loops impose.

It is noted that there are essentially two loops "A" on e':\\ch side of the vessel.

3-5

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I 11-0540-0020, R9V. 3 3.5 coupled NSBB A coupled three-dimensional, lumped-mass model of the HSSS was developed using the models of the various individual components as described above.

The boundary conditions of these components were modified as appropriate in the coupled model to conservatively reflect the expected seismic response of the components at the 0.2q SSE level. As noted previously, the steam generators, while symmetric vessels, are allowed to slide at the base only in the x-direction (Figure 3-1).

Also as previously noted, the orientation for the reactor coolant pump motor support core was assumed so as to result in the lowest response frequency and consequently the highest seismic response in the x-direction.

Although the uncoupled pump models are identical, as are the cold leg Loop A pipes, Loops B and C are different. This occurs because the pump rotations for all pumps are the same.

Therefore, the coupled system results iri the same frequencies and mode shapes for Pumps lA and 2A and also for Pumps lB and 2B, but Pumps lA and 2A are different than Pumps lB and 2B.

The tributary weight of all piping loops was lumped into appropriate nodes at either the RPV, RCP, or SG.

The tributary weights of the piping were determined from the deadweight analysis results shown in Table 3-18.

Thus, while the mass of the reactor coolant piping is included in the coupled NSSS model, and the stiffness of all piping loops is included, the model does not yield the frequencies 0f the pipes themselves.

This is considered to be unimportant to the overall seismic response of reactor core since the frequencies of all pipe loops are in the rigid range.

Fixed-end loop frequencies are greater than 75 Hz as a design condition.

In order to determine the appropriate boundary conditions to use for the NSSS response expected at the O. 2g SSE level, several iterative analyses were conducted.

Initielly, the bases of the steam generators and pumps were assumed fixed.

This condition is 3-6

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11-0540-0020, Rev. 3 appropriate for very low levels of seismic response as expected when the friction restraints provided by the Lubrite pads are not overcome by the seismic shear loads at the bases of the components, and the seismic overturning moments developed are less than the deadweight restoring moments in the pump supports.

Table 3-19 shows the response frequencies for ~he coupled NSSS system with fixed base conditions for the pumps and steam generators.

As can be readily *observed, the addition of the additional stiffnesses from the reactor coolant loop piping serves to somewhat increase the response frequencies of the various individual components with fixed-based conditions.

Note that the tributary mass of the piping was included in the individual (un-coupled) component frequencies. For instance, the 9.1 Hz fundamental frequency in the x-direction is increased to 10. 4 Hz in the x-direction for Pumps lA and 2A and to 11.4 Hz in the x-direction for Pumps lB and 2B.

Similarly, the uncoupled pump frequency of 9.9 Hz in the z-direction is increased to 10.9 Hz and 12.4 Hz for Pumps lB and 2B, and lA and 2A, respectively.

Less variation is noted for the steam generators where the uncoupled z-direction fundamental of 27.2 Hz increases to 27.4 Hz, the x-dircction fun-damental increases from 28.1 Hz to 29.8 Hz, and the vertical from

44. 8 Hz to 44. 9 Hz. The reactor pressure vessel fundamental frequency of 36.9 Hz is virtually unchanged in the z-direction and slightly increased to 37.1 Hz in the x-direction.

A conservative estimate of the expected response of the NSSS was obtained by conducting a response spectrum analysis of the fixed-base component system using the in-structure response spectra developed as part of the SEP evaluation for the median soil condition in Reference 1. These in-structure response spectra were generated using a numerically generated time-history record which produced ground response spectra which envelop the TJSNRC Regulatory Guide 1.60 {Ref. 8) spectra. Composite modal damping ratios in Reference 3-7

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I 11-0540-0020, Rev. 3 1 were limited to 20% of critical. The NSSS damping was initially conservatively assumed to be 3% of critical based on the recom-mendations of Reference 2 for welded steel structures at about one-half yield stress. Based upon this analysis, it was determined that sliding of the steam generator bases (in the x-direction) could be expected, and that sliding and support frame uplift of the pumps could also be expected at the 0.2g SSE.

since the objective of this analysis was to determine the maximum seismic loads which could be transmitted from the pumps and steam generators to the RPV, and consequently, the maximum seismic response which could be expected within the RPV, it was decided to model the pump supports in both horizontal directions and the steam generator supports in the x-direction as frictionless.

Also, the pump supports were not restrained against base rotation.

One or more of the pump support legs can lift, but not all at the same time.

This model results in the lowest frequency system with correspondingly the highest seismic response, and also results in all friction loads which would actually exist to be resisted through the main coolant loop piping to the RPV rather than by friction.

The response frequencies for the coupled NSSS model with frictionless supports for the RCP and SG are shown in Table 3-20.

A comparison of the frequencies presented in Table 3-19 for the fixed-base components indicates substantial reductions in the pump frequencies (i.e., from the 10.4 to 12.4 Hz range for the fixed-base fundamentals to 3.8 to 7.2 Hz for the sliding support conditions.

The fundamental horizontal frequencies of the steam generators are decreased from 27.4 to 25.0 Hz in the x-direction and from 29~8 to 27.4 Hz in the z-direction.

Again, the fundamental response fre-quencies of the reactor pressure vessel are virtually unaffected by the boundary conditions of the pumps and steam generators.

3-8

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11-0540-0020, Rev. 3 3.6 combined Nsss-ssr Hodel In order to determine whether it was necessary to include a complete NSSS model in the combined soil-structure -

NSSS model, or whether a model of the RPV only was adequate to provide a conservative input to the core, a simple time-history analysis was conducted using an isolated RPV and the coupled NSSS model.

The time-history waveform selected was a 5 Hz, approximately one-quarter g input which is representative of the second mode contribution for the median soil condition at the RPV support location (see Table 4-2).

Table 3-21 shows a comparison of the accelerations and displacements of the core support locations (Nodes 2 and a in the RPV) for the isolated RPV and coupled NSSS models. As can be readily noted, neither the accelerations nor displacements are significantly increased when the effects of the pumps and steam generators are included as compared to an isolated RPV.

This occurs because the main coolant pipes are attached at nearly the same elevation as the RPV supports (i.e., the supports are located on the nozzles), and the fact that the steam generator snubbers resist most inertia loads resulting from the additional pump and steam generator masses.

Since the dynamic response of the core support locations for the isolated RPV and coupled NSSS models is so close, it would probably be adequate to use only an isolated RPV model combined_ with the existing SSI model to generate the core seismic input.

However, in order to generate slightly more conservative input, it was decided to include the coupled NSSS model in the c~mbined NSSS - SSI model for the time-history analyses.

3-9

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11-0540-0020, Rav. 3 Table 3-1 Steam Generator Member Properties Area Shear Moment of Member (in2)

Coef!icient Inertia (in4) l*

2 and 3 3450 0.53 l0.66x106 4 and 5 2130 0.53 6.81.xl.06 5, 7, a, and 9 3600 0.53 24.s2x106 Influence Coefficient Matrix (Table 3-2) 3-10

Px t-:.X

1. ox10-B L\\Y

!'1 z ex 8Y ez 11111'7***~- -*-'

Table 3-2 I

I rm -

Steam Generator Bottom Support Core Influence Coefficient Matrix Px Py Mx 0

0 0

l.93x10-9 0

0 l.Oxlo-8 4.71x10-ll Symmetric 4.6Jx10-12 My 0

0 0

0 5.47x10-12 3-11 Mz

-4.71x10-ll 0

0 0

0 4.6Jx10-12 J

xE-6 I

0 U1

~

0 I 8 N

0

w

,~

I

,~*

'~

I~}

__________...........,,~._ _____ _

Node 1

2 3

4 5

6 7

8 9

Table 3-3 Steam Generator Nodal Weights and Mass Moments of Inertia Weight (lb)

Diametral Mass Moment of Horizontal Vertical Inertia (in-lb-sec2) 58,000 58,000 199,100 383,170

1. 35x106 406,900 323,160 6.24x106 127,400 88,880

1. 06x106 273,570 211,830 4.55x106 101,700 101,700 l.23x106 3-12 J

I 0 tn

~

0 I

0 0

N 0.

a,.

r I

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1 l

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2 3

4 11-0540-0020, Rev. 3 Table 3-4 Steam Generator Component Frequencies Uncoupled Frequencies (Fixed-Base) - Hz 27.2 z-direction 28.1 x-direct.ion 44.8 vertical 48.7 x-direction 3-13

Mel'lber 1

2 3

4 5

6 7

8

,......, ~~- - - - -

Table 3-5 Reactor Pressure Vessel Member Properties Area Shear Area Moment of Inertia (in2)

(in2)

(in4) 2481 1316 20.ox109 4820 2558 19.67x106 4820 2558 19.67x106 6172 3275 25.85x106 6958 3692 34.47x106 6958 3692 34.47x106 6172 3275 25.asx106 4017 2131 20.ox109 3-14 I

0 U'I

~

0 I g

~

w

Node 1

2 3

4 5

6 7

8 9

Table 3-6 Reactor Pressure Vessel Weights and Mass Moments of Inertia Weight (lb)

Diametral Mass Moment of Inertia Horizontal Vertical (in-lb-sec2) 127,512 426,586 0.67x106 377,126 337,327 0.97x106 224,885 145,286

1. a2x106 301,392 213,293 2.7Bx106 92,900 92,900 368,626 320,326 2~ox106 238,409 195,132 l.87x106 3-15 I

0 UI

~

8 N

0

if w

I

11r

.-* 1111111 r-

-~ mll.., -

r~ -

Table J-7 Reactor Pressure Vessel Support Influence Coefficient Matrix Px Py Pz Mx My t.X 7.95x10-lO 0

0 0

0

~y l.54xlo-10 0

0 0

~z 7.95xlo-10 l.29x10-lJ 0

ex 2.14x10-14 0

8Y Symmetric 6.sxio-14 ez 3-16 Mz

-1.29x10-13 0

0 0

xE-6 0

2.14x10-14 b

U'I,,.

0 g N

0

w 1

I t*.

o

_)

'I

~,

~

l

<lj t

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I 11-0540-0020, Rav. 3 Table J-8 Reactor Pressure Vessel Component Frequencies Mode Uncoupled Frequency (Hz) 1 36.9 x and z response 2

55.2 x and z response 3

85.0 vertical response 3-17

I--

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11-0540-0020, Rev. 3 Table 3-9 Reactor Coolant Pump Member Properties Member l

(Table 3-11) 2 Rigid 3

Rigid 4

(Table 3-12) 5 Rigid

Influence Coefficient Matrix 3-18

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Node 1

2 3

4 5

11-0540-0020, Rev. 3 Table 3-10 R~actor Coolant Pump Weights and Mass Moments of Inertia Diametral Mass Moment Weight of Inertia (lb)

(in-lb-sec2) 13,100 85,900 2.42x105 10,300 55,150 44,850 7.92x105 3-19

-.. - - - - _........., -~-u- - -,... -,,-*

Table 3-11 Reactor Coolant Pump Support Influence Coefficient Matrix (Element 1)

Px Py Pz Mx My Mz i\\X l.393xlo-7 0

0 0

0

-5.73Xl0-10

~}'

1.ax10-9 0

0 0

0

~z l.393x10-7 s.1Jx10-10 0

0 ex Symmetric s.ox10-12 0

0 xE-6 8}'

4.47x10-ll 0

ez s.ox10-12 3-20 I

0 U1

~

0 I g N

0.

~

w

. I t ',

1.

- -..,...-,.,;J ~ -

Table 3-12 Reactor Coolant Pump Support Influence Coefficient Matrix (Element 4)

Px Py Pz Mx My Mz

!:lX l.09x10-7 0

0 0

0

-2.41xio-7 6Y

1. 21x10-7 0

0 0

0 6Z 2.56xio-7 i.12x10-9 0

0 ex 2.67x10-ll 0

0 er Symmetric 5.6x10-ll 0

ez l.24x10-ll 3-21 xE-6 0

C1'I

~

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?:?

w

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11-0540-0020, Rev. 3 Table 3-13 Reactor Coolant Pump Component Frequencies (Fixed-Base)

Mode Uncoupled Frequency (Hz) 1 9.10 x-axis 2

9.94 z-axis 3

27.7 x-axis 4

28.3 z-axis 5

63.2 vertical 3-22

6.X 6Y 6Z ex

(:)}'

ez

- - - - *-... -..... ~ - - - - -,_,

Px 5.2244

-2.6708 2.4733

-0.0072

-0.0349

-0.0274 Table 3-14 Primary Coolant Loop :- Cold Leg (Loop "A")

P-lB to E-50B - Flexibility Matrix Py Pz Mx My

-2.6708 2.4733

-72.15E-4

-349.03E-4 8.6736 2.3806 411. 34E-4 10.62E-4 2.3806 4.2732 224.36E-4

-261. 71E-4 0.0411 0.0224 5.06E-4 0.28E-4 0.0011

-0.0262 0.28E-4 4.61E-4 0.0285 0.0062 0.45E-4 0.23E-4 3-23 Mz

-273.73E-4 385.04E-4 61.52E-4 xE-6 0.45E-4 0.23E-4 5.20E-4 I

0

'U'I A

0 I g N

0.

~

w

.. /.. -

r l

Table 3-15 Primary Coolant Loop -

Cold Leg (Loop "B")

Pumps lA and 2A to Reactor Vessel - Flexibility Matrix Px Py Pz Mx My t.X

1. 314 0

2.369 0

-160.570E-4 t\\Y 0

5.634 0

172.264E-4 0

t\\Z 2.369 0

4.437 0

-302.4656E-4 ex 0

172.264E-4 0

2.7925E-4 0

ey

-160.570E-4 0

-302.4656E-4 0

2.618E-4 ez 0

291.1209E-4 0

0 0

3-24 Mz 0

291.1209E-4 0

0 0

2.618E-4 xE-6 b

U'I

~

0 I

0 0

N 0

0

~

Table 3-16 Primary Coolant Loop -

Cold Leg (Loop "C")

Pumps lB and 2B to Reactor Vessel - Flexibility Matrix Px Py Pz Mx My

!'J.X

.0912 0

-.391 0

19.5302E-4 6Y 0

6.194 0

-65.6244E-4 0

c..z

-.391 0

6.70 0

-410.676E-4 ex 0

-65.6244E-4 0

2.9496E-4 0

8}'

'19.5302E-4 0

-410.676E-4 0

3.0892E-4 ez 0

377.3402E-4 0

-.2793E-4 0

J-25 Mz 0

377.3402E-4 0

xE-6

-.2793E-4 0

2.9671E-4 I

0 c.n

~

0 b

0

~

f I I*,

I O

O O O

'I,

]

~

"' R O

0 0

o f

O

~

4 f

w------

Table 3-17.

Primary Coolant Loop -

Hot Leg (Loop no 11 )

Pumps lB and 2B to Reactor Vessel - Flexibility Matrix Px Py Pz Mx My Mz

.6. x

.0681

-.1993 0

0 0

-19.635E-4

.6.Y

-.1993 1.0866 0

0 0

74.S954E-4 tJ.Z 0

0 1.1441 20.6822E-4 -75.0492E-4 0

xE-6 ex 0

0 20.6822E-4

.8029E-4

-.05236E-4 0

ey 0

0

-75.0492E-4 -.05236E-4 0.8203E-4 0

ez

-19.6350E 7t..5954E-4 0

0 0

.8378E-4 3-26 0

~

w

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l 1

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I Loop A

B c

D 11-0540-0020, Rev. 3 Table 3-18 Loop Joint Loads Generator Load Vessel Load Pump Load (Node)

(Node)

(Lbs.)

-13,057

-16,000

-12,365

~12,104

-12,102

-12,322

-23,415

-21,999 3-27

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I 11-0540-0020, Rev. 3 Table 3-19 Coupled NSSS Frequencies for Fixed-Base Support Conditions for All Components Frequency Component and Direction of (Hz)

Peak Response 10.4 Pumps lA and 2A, x-direction*

10.9 Pumps lB and 2B, z-direction 11.4 Pumps lB and 2B,. x-directio!l 12.4 Pumps lA and 2A, z-direction 27.4 Steam Generators, z-direction 29.8 Steam Generators, x-direction 30.l Pumps lA and 2A, z-direction 30.2 Pumps lB and 2B, z-direction 34.5 Pumps lB and 2B, x-direction 36.3 Pumps lA and 2A, x-direction 36.9 Reactor Pressure Vessel, z-direction 37.1 Reactor Pressure Vessel, x-direction 44.9 Steam Generators, Vertical x-direction approximately E-W:

z-direction approximately N-S:

(See Figures 1-1 and 3-1) 3-28

I I

11-0540-0020, Rev. 3 Table 3-20 Coupled NSSS Frequencies for Sliding Support Conditions for the PUmps and Steam Generators component and Direction of Frequency Peak Response (Hz) 3.8 Pumps lA and 2A, z-direction*

4.3 Pumps lB and 2B, z-direction 7.0 Pumps lB and 2B, x-direction 7.2 Pumps lA and 2A, x-direction 13.l Pumps lB and 2B, z-direction 14.8 Pumps lA and 2A, z-direction 25.0 Steam Generators, x-direction 27.4 Steam Generators, z-direc:tion 30.0 Pumps lB and 2B, x-direction 34.2 Pumps lA and 2A, x-direction 36.9 Reactor Pressure Vessel, z-direction 37.0 Reactor Pressure Vessel, x-direction 44.9 Steam Generator, Vertical z-direction approximately N-S; x-direction approximately E-W; (See Figures 1-1 and 3-1) 3-29

Isolated RP'l Node 2 Accel.

Displ.

Accel.

(g)

(in.)

(g) 0.247 4.98E-6 0.223 j_:.,,/

'1L........U Table 3-21 Comparison of RPV Response Coupled NSSS Node 8 Node 2 Displ.

Accel.

Displ.

Accel.

(in.)

(g)

(in.)

(g) 2.0SE-6 0.248 6.03E-6 0.236 3-30.

Node a Displ.

(in.)

4.13E-6 I 0

~

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I Loop B Figure 3-!

~1sss Plan Vie1*1 3-31 11-0540-0020, Rev. 3 Loop D Loop A Loop C

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I Figure 3-2

~ISSS Elevation 3-32

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Y (Vert.)

z }-x

~ Lumped-Mass

()i Massless Node I k Element Number g'

9

--659' - 6" (Upper Support

t-- -

658' -l/2"

---652' - 6" (Snubbers) 648'-9-l/2" 5

-- 634' 1/2"

-- -521

3/8"

--- 617' 3/4" 6l5' - l" Figure 3-3 Steam Generator Lumped-Mass Model 3-33

I I

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I

\\ I I

y )--x z

~ '.........

-~

11-0540-0020, Rev. 3 Lumped-Mass Q.

l Massless Node Element Number 627' 4-7/16"

624' 1/2" 621' - 7" 618' 2-1/2"

--- 615' 7"

614' 9"

613' 2"

I 3 3

606 I 10" I 2 2

598' - 11" 1

/------ 595 I 2"

Figure 3-4 Reactor Pressure Vessel Lumped-Mass Model 3-34

~

11-0540-0020, l\\ev. 3 I

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'!(f' ID I

N..['T NCZZU:

\\I...

I, Ii I I I I I

~

I Figure 3-5 Reactor Pressure Vessel Internals Cross-Section 3-35

__._J.$1~11.-.~*lll*.*-1 1.11. !!Ill,-.~

.. *--~.l!I!

.... ~1111!.211141!!U!A!llllE*. "'!t¢~."ll'f!lll.1'1

_glgtl4141)&1. 1.x1.;1_g;1;1_:;;;1.1

... 1"\\'!£1J&JEI:.:

1

...,,1

.*. **ww.11

... *.--*... *.. *£till.Jf>'1;"'!".: --

-~ ~ *-

-.~.J'

.""\\::;,;*;_:*~~~;!~****.-,~t~W *;i....,.

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4.

COMBINED NSSS-SSI MODEL The model used for the time-history analyses to develop the core seismic input was developed by combining the three-dimensional NSSS model of the RPV, SG, and RCP previously described with the existing soil-structure model of the containment building.

As previously noted, the concrete internal structure mass was modified slightly from the FSAR/SEP structure model by subtracting the weight of NSSS from the concrete.

The same soil properties as were used in the SEP evaluation were retained for the present analysis.

In addition to the reduction in mass of the concrete internal structure nodes to account for the NSSS weight, several massless nodes were added to the concrete internal structure stick model at the support locations for the NSSS components. Thus, massless nodes were added at Elevation 608' -6" (pump base support), Elevation 615 1 -1" (RPV and steam generator base supports) and Elevations 652'-6" and 659'-6" (for the upper steam generator snubbers and uppec truPnion supports, respectively).

The NSSS components were connected to these nodes using the appropriate boundary conditions described in the previous section to develop the final combined NSSS-SSI model.

4.l combined NSSS-SSI Modal Frequencies The combined NSSS-SSI modal frequencies for the first 12 modes are shown in Table 4-1 for the three soil cases previously described.

Due to the change in mass of the concrete internal structure and the addition of the NSSS model, additional modes corresponding to the NSSS model are introduced and minor changes are noted in the soil-structure modes when compared with the SSI model results without the NSSS.

Table 4-2 shows a comparison of the modal frequencies for the combined NSSS-SSI model with those computed for the SSI model only (median soil case) as well as the modal frequencies of the NSSS only. Similar close comparisons were observed for the other soil conditions.

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11-0540-0020, Rev. J 4.2 composite Modal Damping Ratios Composite modal damping ratios were developed for the combined NSSS-SSI model for the three soil cases usinq the same energy proportioning method (Ref. 9) that was used for the SEP evaluation (Ref. 1). In this approach, the composite modal damping ratios are determined based on two terms.

One terms accounts for the energy assumed to dissipate in a viscous form, and the second term is based on the energy assumed to dissipate in a hysteretic form. In the SEP evaluation (Ref. 1) as well as the current analysis, the viscous portion was assumed to consist of the horizontal and vertical components of the soil damping.

The remaining portion of the soil damping and structural damping was proportioned on the assumption that the energy was dissipated hysteretically.

The structural damping values recommended in NUREG/CR-0098 (Ref. 2) were used to compute the composite modal damping ratios used in the current analysis.

Based on the SEP evaluation (Ref. 1), 3% of critical was used for the containment building due to the low stresses calculated in the pres tressed concrete.

Similarly, a review of the expected seismic stresses in the NSSS indicated stresses less than one-half yield are expected for the 0.2g SSE.

Thus, 3 % of critical was used for the NSSS in developing the composite modal damping.

As with the SEP evaluation, a maximum composite modal damping ratio of 20% of critical was used in the current analysis.

Tables 4-3 through 4-5 show the composite modal damping ratios calculated for the upper, median, and lower soil cases.

A comparison of the composite modal damping ratios computed for the modes which are primarily s~ructure response for the median soil case (Table 4-4) with those developed for the median soil case of the soil structure model used for the SEP evaluation indicates identical damping ratios.

4-2

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11-0540-0020, Hev. 3 Table 4-1 Combined NSSS-SSI Modal Frequencies (Hz)

Upper Median Mode Soil Case Soil Case 1

2.42 (s}*

2.05 (s}

2 3.99 3.99 3

4.31 4.31 4

5.78 ( s, vert.)

4.79 (s,. vert.)

5 6.33 (s) 5.75 (s) 6 6.67 6.43 7

6.94 6.76 8

12.99 (s) 12.61 ( s}

9

13. 36 13.36 10 15.42 15.42 11 17.92 (SI vert.) 17.74 (s, vert.)

12 20.15 (s) 19.97 ( s)

(s) denotes primarily a structure mode; other modes are primarily equipment 4-3 Lower Soil Case l.5l(s) 3.44 (s, vert.)

3.99 4.28 (s) 4.32 6.40 6.75 12.22 {s) 13.36 15.42 17.56 (SI vert.)

19.78 (s}

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11-0540-0020, Rev. 3 Table 4-2 comparison of Combined NSSS-SSI Modal Frequencies with Separate SSI and NSSS Frequencies (in Hz) -

Median Soil Case combined SSI NSSS Mode NSSS-SSI Model Model Model 1

2.05 2.06 2

3.99 3.8 3

4.31 4.3 4

4.79 4.79 5

5.75 5.78 6

6.43 7.0 7

6.76 7.2 8

12.61 12.85 9

13. 36 13.1 10 15.42 14.8 11 17.74 17.74 12 19.97 20.02 4-4

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11-0540-0020, Rev. 3 Table 4-3 Modal Damping Ratios for the Combined NSSS-SSI Model -

Upper Soil Case Frequency Damping Mode (Hz)

(\\ of Critic3.l) 1 2.42 8

2 3.99

)

4.31 3

4 5.78 10 5

6.33 7

6 6.67 11 7

6.94 20 8

12.99 13 9

13. 36

)

10 15.42

)

11 17.92 4

12 20.15 7

4-5

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I 11-0540-0020, Rev. 3 Table 4-4 Modal Damping Ratios for the Combined NSSS-SSI Model -

Median Soil Case Frequency Damping Mode (Hz)

(\\ Of CritL::al) 1 2.05 B

2 3.99 3

3 4.31 3

4 4.79 10 5

5.75 20 6

6.43 5

7 6.76 4

8 12.61 10 9

13. 36 3

10 15.42 3

11 17.74 4

12 19.97 6

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I 11-0540-0020, Rev. J Table 4-5 Modal Damping Ratios for the Combined NSSS-SSI Model -

Lower Soil Case Frequency Damping Mode (Hz)

(% of Critical) l

1. 51 9

2 3.44 10 3

3.99 3

4 4.28 20 5

4.32 16 6

6. 40 3

7 6.75 3

8 12.22 7.5 9

13. 36 3

10 15.42 3

11 17.56 4

12 19.78 5

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I ll-0540-0020, R*v. 3 S.

SEISMIC INPUT The seismic input used for the analysis consisted ot five recorded earthquake records selected as being representative of an actual earthquake of o. 2g which could be experienced at the Pallaades site. A suite of natural *earthquakes was selected since the response of the core is expected to be highly nonlinear, and it is recognized that a single earthquake record, even an artificial earthquake record whose response spectra envelop smooth, broadband ground response spectra, may not always produce conservative response results for a nonlinear system.

This occurs primarily due to the phasing of the input and the nonlinear response cycles.

Although peaks and valleys occur in response spectra of natural earthquak11s, by selecting a suite of appropriate records, it can be assumod that no frequency regions are not excited by at least one or mora Of the records and that enou_gh excitation records are used in ordar to determine the worst phasing and hence, conservative responsa of the core.

Table 5-1 shows the records selected for this analygis.

Those records encompass earthquakes of lot:al magnitudes, ML, varying between 6.4 and 7.2 with peak ground accelerations above O.lBg.

All records were scaled to 0.2g peak ground acceleration tor this analysis.

The five records were recorded on site condition& which varied from deep alluvium to rock.

Figure 5-1 shows the 5' damped response spectra of the five earthquake records.

As is apparont, different spectral accelerations occur for different earthquake records over the frequency range of interest.

Figure 5-2 shows a plot of the median of the 5' damped response spectra for the five earthquake records as well aa the upper bound of response spectra for the same five records.

It should be emphasized that the median response spectrum is tho modian 5-1

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11-0540-0020, Rev. 3 of the five records (i.e., at any frequency, the spectral accel-erations of two records lie below the median curve and two records have spectral accelerations which lie above the median), and is not a median spectrum over the entire frequency range from a single earthquake record.

Similarly, the upper bound of the response spectra represents the maximum spectral acceleration resulting from any of the five records, and different records produce the maximum response in different frequency ranges.

However, as is apparent from Figure 5-2, the median of the spectra lies fairly close to the SEP site-specific spectrum (Ref.

10) scaled to o. 2g, and the upper bound of the five spectra envelops the site-specific spectrum over the entire frequency range of interest.

Thus, it is assumed that the peak seismic response of the SS! model analysis from the suite of five records is well above that which would be predicted from the use of a single artificial earthquake record whose response spectra envelops that of the site-5pecific spectra. Also, the fact that a number of time-history records were developed assures that the worst case phasing of the response cycles for the important SSI modes will maximize the nonlinear response in the core.

5-2

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Earthquake Date Kern County 1952 Morgan Hill 1984 San Fernando 1971 Imperial Valley 1979 Imperial Valley 1979

=

11-0540-0020, Rev. 3 Table 5-1 Earthquake Records Magnitude Recording Station Component Ms = 7.7 Taft S69E ML = 7.2 Lincoln School Ms = 6.1 Coyote Lake S15W ML = 6.5 Ms = 6.6 Castaic N69W ML = 6.4 Ms = 6.9 Differential NOOE ML = 6.6 Array Ms = 6.9 El Centro S40E ML = 6.6

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09 FEB 1971. CASTAIC.

COMP N69W.

CA.

15 OCT 1979. EL CENTRO.

COMP NOOE.

CA.

15 OCT 1979. EL CENTRO #4.

COMP S40E.

24 APR 1984. COYOTE LAKE DAM.

COMP S15W.

.JULY 21.

1952 -

TAFT.

COMP S69E 5X damping 5X damping 5X damping 5X damping 5% damping SAN FERNANDO, CA.

IMPERIAL VALLEY.

MORGAN HILL.

CA.

KERN COUNTY, CA, 7....--~~~~~--~~~~~~~- -~~~~~~~~~~~~~~~~~

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11-0540-0020, Rev. 3 The Coyote Lake record is a very conservative record compared to the expected earthquake motion expected for the Palisades site as indicated by the fact that the 5% damped spectral acceleration is the 1. 5 Hz range is nearly 20 percent greater than the site-specific spectral acceleration at the same frequency (see Figures 5-1 and 5-2).

Although the seismic response of the core may be expected to be highly nonlinear, it is clear that the use o! the maximum displacement records developed here will provide a conservative input for the core analysis.

6. 2 In-structure Response Spec.tr a In addition to the displacement time-histories at the core support locations, acceleration time-histories were also determined at several locations for each of the five records and three soil conditions. These locations were the top of the base mat (Elevation 590'~. approximately the RPV support location (Elevation 616 1 } the top of the concrete internal struc::.cire (Elevation 649 1 )

and a location high on the containment building (Elevation 730 1 ). These are the same locations for which spectra were developed for the SEP eyaluation (Ref. 1).

In-structure response spectra for 2, 3, 5, and 7 percent of critical damping were developed for each location.

Individual spectra were developed for each earthquake record and each soil condition.

A median and upper bound of the five spectra for each soil case and damping ratio were then developed.

Figure 6-31 shows an *example of the 5% damped spectra for the five time-history records at Elevation 616 feet on the concrete internal structure for the median soil case.

Figure 6-32 shows the median and upper bound of these in-structure response spectra.

A similar procedure was employed for each of the other equipment damping I

ratios, soil cases~ and locations.

Plots of the median and upper 6-2 I

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11-0540-0020, Rev. 3 bound in-structure response spectra for the four damping ratios for each location and soil condition are shown in Figures 6-33 through 6-56.

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FIGURE 6-11 Displacement Time History at Upper Reactor Core Support for' Taft Record. Median Soil Case

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FIGURE 6-12 Displacement Time History at Lower Reactor Core Support for Taft Record. Median Soil Case i:~ **. *'.i.tA,..._... '.-. *.'..,.~,

I

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-IGURE 6-13 Displacement Time History at Upper Reactor Core Support for Coyote Lake Record. Median Soil Case I

0 U1 0

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FIGURE 6-14 2.00 24.00 Displacement Time History at Lower Reactor Core Support for Coyote Lake Record. Median Soil Case

.00 I

0 c.n 0

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FIGURE 6-15 20.00 24.00 Displacement Time History at Upper Reactor Core Support for Castaic Record, Median Soil Case c *' * t '" '...

jo*. ',.~. "* w~-, '

2.00 r.t "

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FIGURE 6-16 Displacement Time History at Lower Reactor Core Support for Castaic Record. Median Soil Case 2.00 I

0 U1

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4.00 1.00 1.oo TIME (sec) 2.00 24.00 F.tGURE 6-17 Displacement Time History at Upper Reactor Core Support for El Centro S40E Record. Median Soil Case 2.oo I

0 01

~

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w

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FIGURE 6-18 Displacement Time History at Lower Reactor Core Support for El Centro S40E Record. Median Soil Case I

0 U'I

~

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FIGURE 6-19 Displacement Time History at Upper Reactor Core Support for El Centro NOOE Record. Median Soil Case I I

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.4.00 B.00 1.00 TIME 1.00 (sec) 2.00 24.00 2.00 FIGURE 6-20 Displacement Time History at Lower Reactor Core Support f or' E l Centro ~JO 0 E Record. Med i an So i l Case I

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FIGURE 6-21 Displacement Time History at Upper Reactor Core Support for Taft Record. Stiff Soil Case 28.00 I

0 U1 0

I 0

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w

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FIGURE 6-22 20.-00 24.00 Displacement Time History at Lower Reactor Core Support for Taft Record. Stiff Soil Case 28.00 I

.0

<.n

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--:-IGURE 6-23 Displacement Time History at Upper Reactor Core Support for Coyote Lake Record, Stiff Soil Case I

0 C.1'I

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FIGURE 6-24 Displacement Time History at Lower Reactor Core Support for Coyote Lake Record Stiff Soil Case

.. ~'!'~

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FIGURE 6-25 Displacement Time History at Upper Reactor Core Support for Castaic Record, Stiff Soil Case

... 1.......

... '*. ';.,-.~~**:.*..

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0 U'I

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FIGURE 6-27 Displacement Time History at Upper Reactor Core Support for El Centro S40E Record. Stiff Soil Case I

0 U'I

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FIGURE 6-28 Displacement Time History at Lower Reactor Core Support for El Centro S40E Record. Stiff Soil Case I I I

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FIGURE 6-30 Displacement Time History at Lower Reactor Core Support for El Centro NOOE Record. Stiff Soil Case

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l 11-0540-0020, Rev. J REFERENCES
1. Nelson, T. A., R. C. Murray, D. A. Wesley, and J. D. Stevenson, Seismic Review of tha Palisades Nuclear Puver Plant Unit 1 as Part of the Systematic Evaluation Program, NUREG/CR-1833, January, 1981.
2. Newmark, N. M., and W. J. Hall, Development of Criteria tor seismic Review of Selected Nuclear Power Plants, NUREG/CR-0098, 1977.
J. American Society of Mechanical Engineers, Boiler and Pressure Vessel Code,Section III, 1983.
4. Consumers Power Company, Final Safety Analysis Report, Jackson, 1'
Michigan, 1978, (plus appendices).
J l..
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5. E. Kausel, R. V. Whitman, F. Elsabee, ~nd J. P. Morray, Dynamic Analysis of Embedded Structures, in Transactions.
Fourth International Conference on Structural Mechanics in Reactor
echnoloqy, San Francisco, Paper K2/6, Vol. K(a), 1977.
6. E. Kausel, and R. Ushijima, Vertical and Torsional stiffness ot cylindrical Footings, MIT Research Report R79-6, Dept. of Civil Engineering, Massachusetts Institute of Technology, February, 1979.
7. Jenkins, R.
B. to D.
A. Wesley, Private Communication, RBJ 65-88, September 7, 1988.
8. U.S. Nuclear Regulatory Commission, Design Response Spectra tor Seismic Design of Nuclear Power Plants, Regulatory Guide 1.60, 1973.
9. J. M. Roesset, R. V. Whitman, and R. Dobry, Modal Analysis for Structures with Foundation Interaction, J. Str. Diy,, ASCE, Vol. 99, No. ST3, 399-416, March, 1973.
10. Project Calculations, C-1356-018-1 through -11, October, 1988.
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I 11-0540-0020, Rev. 3 APPENDIX l\\
CORE VERTICAL SEISMIC DISPLACEMENT TIME-HISTORIES Auqust 1989
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11-0540-0020, Rev. J APPENDIX A CORE VERTICAL SEISMIC DISPLACEMENT TIME-HISTORIES
1.
INTRODUCTION This appendix presents the time-history analysis methods and results for the vertical seismic input to the core.
In view of the rigidity of both the core support system and fuel bundles, it was originally envisioned that the vertical loads in the core would be develope':i using a linear elastic model. of the core together with the vertical response spectra developed during the systematic Evaluation Program (Ref. Al).
However, subsequent analysis by ANF has indicated that the core uplift forces developed by the coolant flow coupled with the lack of core hold-down springs at Palisades may result in core levitation at high 3eismic inputo.
Although the core bundle alignment will be maintained, momentary uplift of the bundle from the lower core support plate necessitated the development of a nonlinear vertical core analysis model which required time-history vertical motion as input.
A three-dimensional lumped-mass model has been developed for generating representative horizontal seismic inputs to the reactor pressure vessel (RPV) internals.
The analysis performed in creating the model is described in Section 3 of this report.
The model consists of several lumped-mass sticks (beam elements) which model the response of the reactor containment building and the concrete internal structure coupled with the three-dimensional response of the nuclear steam *supply system (NSSS) components including the reactor pressure vessel (RPV), the steam generators (SG), and the reactor coolant pumps (RCP).
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11-0540-0020, Rev. 3 The vertical analysis used the same model and range of soil properties and equivalent soil springs which were used in the horizontal analysis. In order to bound the possible soil conditions which may be present at the Palisades Site, three different sets of spring rates were used in the analysis. The various spring rates were chosen to correspond to a +/- 50% variation in shear modulus from the median case of C = 4.38x 10 6 1 b2
fl In order to use the same lumped-mass model to develop the vertical motion, a confirmatory analysis was conducted to verify that no amplification of the RPV vertical motion through the core support system would occur.
2.
VERTICAL RESPONSE OF REACTOR INTE::WALS In evaluating the vertical response of the reactor internals, it was first necessary to verify that there will be no dynamic amplification of the vertical motion due to the relative flex~~ility of the core internal suppcrt structure.
The potential for dynamic amplification can be characterized by the natural frequency for vertical vibration.
If the frequency falls within the amplified region of the vertical response spectrum, there will be some dynamic amplification.
However, if the frequency falls in the rigid range of the spectrum, there will be no amplification and the internals support system can be t.reated as rigid.
Therefore, to determine whether the vertical response of the internals will lead to amplification, a calculation was performed to estimate the vertical frequency of the core internal support structure.
The components of the reactor internals include: the core barrel, the fuel bundles, the core shroud, the core support plates, and the core support columns and beams that support the fuel bundles.
The core barrel is a circular cylinder with a typical inside diameter of 149.75 inches and a thickness that varies between 1 and 2 inches (Ref. A4).
The barrel has two diametrically opp~sed, reinforced A-2 J
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I 11-0540-0020, Rev. 3 openings near the top for the outlet nozzles.
The fuel bundles, the core shroud, and the core support plates are housed within the barrel. The bottom plate and stiffener beams, located at the bottom of the core barrel, support the core and carry the vertical load out to the barrel.*
Therefore, all of
~he vertical loads are transferred to the barrel.
For vertical motion, the core barrel is supported only at the top by the barrel flange, which is seated on a ledge of the reactor pressure vessel.
The core barrel lower support (snubbers) allow vertical translation and only resist tangential forces.
The structural elements are fabricated from ASME SA-240 stainless steel.
The temperature of the core barrel during operation was taken to be 550° F.
In order to develop a model to represent the vertical vibration of the core barrel, the stiffness and mass contributions of the different components were evaluated.
There ara essentially three main contributors to the vertical stiffness: the core barrel flar:;e, tl-e core barrel, and the bott*Jm plate and stiffener beams.
The vertical stiffness contributed by the flange was estimated by including the effects of bending and shear deformation of the flange as well as the bending interaction between the flange and the core barrel shell.
The vertical stiffness of the barrel was evaluated by calculating the axial stiffness of each of the different segments where the cross-section properties changed.
The seven beams, supporting the fuel bundles, span across the bottom of the core barrel. The beams are welded to the core barrel shell and therefore, have boundary conditions that are intermediate to pinned and fixed conditions.
As a conservative measure leading to a more flexible model, the vertical stiffness of the bottom beams was based on the most flexible beam which spans across the diameter of the barrel with the assumption of pinned support conditions. The mass properties of the system were tabulated from the dimensions of the components.
The majority of the total mass comes from the fuel bundles.
There A-3
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11-0540-0020, Rev. 3 are a total of 204 bundles, each weighing 1,320 lbs.
An additional weight of 114 lbs. per bundle was also included to account for the virtual fluid mass participating with the fuel bundle as it moves.
Using the estimated mass and s*ciffness properties, a
mul~i-degree-of-freedom lumped parameter model was developed to represent the vibration of the internals supported only by the flange at the top of the barr~l. The resulting vertical frequency was estimated to be 12. 9 Hz.
Note that this is a conservative frequency estimate due to the modeling assumptions used for the bottom beams.
The frequency of 12.9 Hz is greater than 10 Hz, which is the lower bound frequency for the rigid range of the spectrum for the soft soil condition as shown in Figure A-1.
Therefore, it was concluded that, for the soft soil case, the reactor internals can be treated as rigid in the vertical direction and will not lead to additional amplification of the vertical motion.
The soft soil condition controls both the vertical and horizontal response of the fuel ~undJ~s since it results in the l~rgest displacements in both directions, and hence the greatest impact loads.
A-4
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3.
ANALYSIS OF NSSS-SSI HODEL TO VERTICAL MOTION The peak horizontal response of the core, as determined in the previous analysis, was found to occur for the Coyote Lake record from the 1984 Morgan Hill earthquake. Figure 5-1 shows the ensemble ground response spectra, including the Coyote Lake record.
From this figure, it may be noticed that the Coyote Lake record governs the response at the lower frequencies (1 -
2 Hz).
The maximum horizontal displacement at the upper core support for the soft soil Coyote Lake record was about 0.6J inches.
Since the response of the reactor internals is governed by the horizontal ~otion, only one case need be run for developing vertical response time-histories.
The previous results show that the soft soil case coupled with the Coyote Lake record will govern the horizontal response; hence, this case was selected for use in obtaining the vertical response as well.
The maximum peak horizontal acceleration of the site-specific spectra for Palisades is 0.20g.
All of the records used in the horizontal response analysis were scaled to peak horizontal acceleration values of 0.20g.
For the vertical response, it has been recommended (Ref. AJ) that design motions should be taken as 2/J of the value in the horizontal direction across the entire frequency range. This recommendation is based on statistical studies of natural earthquake motions. For the Coyote Lake vertical record, the vertical acceleration was scaled to:
213(0.20g) = 0.133g as recommended by Reference A3.
A-6
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11-0540-0020, Rev. 3 Since it was determined that no amplification in the core suppo,rt system for the vertical direction exists, the three-dimensional, lumped-mass NSSS-SSI model was used without modifi-cation to determine the vertical input to the core. The displacement time-history at the lower core support location (Node 2 in Figure 3-4) was determined for the Coyote Lake record.
This time-history was superposed with the vertical displacement time-history at Node 2 obtained from the horizontal analysis.
Figure A-2 shows the plot of the resulting displacement time-history for this case. The peak displacement is about o. 14 inches up and o. 15 inches down and occurs at about 4.5 seconds into the record.
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A2.
A3.
11-0540-0020, Rav. 3 REFERENCES Nelson, T. A., et al, seismic Review of th* Palisades HUclear Power Plant Unit 1 as Part of th* systematic Bvaluation Proqram, NUREG-CR-1833, u.s. Nuclear Regulatory Commission, Washington, D.C., 1981.
Porter, L.
D., Ragsdale, J. T., and McJunkin, R.
D.,
Processed Data from the stronq-Motion Records of the Santa Barbara Earthquake of 13 August 1978, Final Results 1979, California Division of Mines and Geology, Sacramento, CA, 1979.
Newmark, N. M., and Hall, W. J., Developm*nt of Criteria for Seismic Review of Selected Nuclear Power Plants, NUREG-CR 0098, U.S. Nuclear Regulatory Commission, Wash-*
ington, D.C., 1977.
A4.
CE Drawing 2966-SJ-1752 Rev. 2, September 1, 1988.
AS.
CE Drawing 2966-J-2402 Rev. W, August, 1969.
A-9

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