ML17277A513

Effects of Hydrodynamic Loads on Safety-Related Equipment & Piping Outside Containment, Technical Rept

ML17277A513
Person / Time
Site:	Columbia
Issue date:	03/28/1983
From:	Ettouney M, Fialkow M, Odonnell J BURNS & ROE CO.
To:
Shared Package
ML17277A512	List: GO2-83-260, Forwards Effect of Hydrodynamic Loads on Safety-Related Equipment & Piping Outside Containment, Technical Rept ML17277A513
References
NUDOCS 8304060282
Download: ML17277A513 (22)

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EFFECT OF HYDRODYNAMIC LOADS ON SAFETY 'RELATED EQUIPMENT AND PIPING OUTSIDE CONTAINMENT TECHNICAL REPORT prepared by BURNS AND ROE, INC.

for application to

• WASHINGTON PUBLIC POWER SUPPLY SYSTEM NUCLEAR PROJECT NO.

2 I

Prepared by:

M. N. Fialkow

/r Reviewed by:

~ /I P t.Mz M. M. Ettouney Reviewed by:

J

. O'Don ell

/

Approved by:

B. Bedrosian Submitted by: J. P. Verderber Date:

i',

" ~8304060282 830328 I

PDR ADOCK 0500039>

PPR

TABLE OF CONTENTS PAGE l.

.Introduction

2. 'omparison of In-Structu're Acceleration

Response

Spectra 3.

Comparison of Peak Building Accelerations n

4:

Test Results for Peak Building Accelerations 3

5.

Summary. and Conclusions 6.

References 11

LIST OF FIGURES FIGURE NO.

TITLE PAGE 1

Horizontal Acceleration vs. Frequency -, El.

567 Original OBE and Revised OBE + SRV.

2 Horizontal Acceleration vs.

Frequency El.

567 Original SSE and Revised SSE

+ SRV + Chugging 3

Horizontal Acceleration vs.

Frequency - El.

435 Original OBE and Revised OBE + SRV 4

Horizontal'cceleration vs.

Frequency El.

435 Original SSE and Revised SSE

+ SRV + Chugging 5

Vertical Acceleration vs.

Frequency - El.

470 Original OBE and Revised OBE + SRV Vertical Acceleration vs.

Frequency - El.

470 Original SSE and Revised SSE

+ SRV + Chugging Horizontal Building Accelerations Peak Values 8

Vertical Building Accelerations Peak Values 12 13 14 15 16 17 18 19

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INTRODUCTION The design for dynamic loads of the piping and equip-ment located in the reactor building outside primary contain-ment is based on the building response to seismic loads determined using a lumped mass-spring model.

Subsequent to this design, building.responses due to hydrodynamic loads originating in the wetwell (SRV actuation and chugging) and due to seismic loads were calculated using a finite'element model. Hereinafter, the building responses used as the basis of design which were developed with the lumped mass-spring model are referred to by the term "original" and the building responses calculated with the finite element model are referred to by the. term "new".

Previous"investigations of the hydrodynamic loads (References 1,

2), have indicated that the structural response accelerations in the reactor building outside primary contain-ment caused by these loads are not significant for equipment and piping design.

The objective of this report is to provide

/

further quantitative evidence for this position.

In brief, the original design specification relates the basis of dynamic design to the dynamic characteristics of the equipment and. the loading.

When the fundamental (lowest natural) frequency of the equipment, is large enough to qualify the equipment in relation to the dynamic loads as "rigid," the dynamic design depends on the peak building

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accelerations, (horizontal and vertical) at the location of the equipment.

Otherwise, the dynamic design depends'n the values of the acceleration response spectra corres-

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'I ponding to the natural frequencies of the equipment.

Xn v'iew of the above, this report includes the follow-

Lng,

a. 'A comparison of calculated in-structure acceleration response spectra in Section 2.

b.

A comparison 'of calculated peak building response accelerations in Section 3.

c.

A summary of in-plant 'SRV actuation test results for peak building response accelerations in Section 4.

Summary and Conclusions are presented in Section 5 with References in Section 6.

2.

COMPARISON OF IN-STRUCTURE ACCELERATION RESPONSE 'SPECTRA Comparison of the original seismic response spectra with new comparable response spectra (due to seismic and con'current hydrodynamic loads) involves consideration of the following cases:

a.

Original OBE vs.

new (OBE + SRV)

'I b.

Original SSE vs.

new-(SSE

+ SRV + chugging).

Structurally, the linkage of the reactor building outside primary containment i.e., of the secondary contain-ment portion of the building, to the primary containment

.portion of the building is limited:

in the horizontal direction this linkage is provided through the basemat and the connection at the stabilizer truss level (Elevation 567) while vertically, only through the basemat.

These structural connecti'ons are pertinent to the transmission of forces or, more generally, bui:lding responses caused by the hydrodynamic loads originating in the wetwell.

Consequently, the locations of these connections are representative for building responses caused by hydrodynamic loads.

In line with the preceding, response spectra for the loadings in the above'ases are developed for the locations in the reactor building outside containment corresponding to the levels of-structural linkage.

Horizontal acceleration response spectra are developed at Elevations 567 and i

435 and are shown in Figures 1 to 4.

Vertical acceleration

response spectra are developed at Elevation 470, an ejeva-tion close to the base where Vertical acceleration data are readily available; the vertical response spectra are shown in Figures 5 and 6.

Combination of spectral acceler-ations due to seismic, SRV actuation and chugging (where applicable) loads, is done by the square root, of the sum of the squares

method, Review of these figures shows.that the original acceler-ation response spectra equal or exceed th'e new comparable spectra except for the vertical direction in a relatively 1

narrow range of frequencies.

Attention is directed to specific results discussed below.

a.

Horizontal acceleration response spectra Except for th'e non-significant frequencies of 1 Hz or less, the original horizontal spectra due to the SSE equal or exceed the new horizontal spectra due to the

SSE, SRV actuation, and chugging at al.l frequencies and at both elevations.

Likewise, the original horizontal spectra due to the OBE exceed at all frequencies the new horizontal spectra due to the OBE and SRV actuation.

b.

Vertical acceleration response spectra The orig-inal vertical spectrum due to the SSE'exceeds the new spectrum due to the

SSE, SRV actuation, and chugging at all frequencies except for the non-significant frequencies of 1., Hz or less.

Similarly, the original spectrum

.due to the

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OBE exceeds, except at the isolated frequencies of 10 I

and ll Hz, the new spectrum due 'to the OBE and SRV actu-ation.

With regard'o the'apparent deficiency at 10 and ll Hz, it is noted that 'the design vertical load, which includes dead weight effects as well, has a minimum value equal to the'gravity (dead) load multiplied by the factor of on'e plus the original acceleration in g's.

Consequently, the design load def'iciency 'at the frequencies of 10 and ll Hz is. less than 7 percent of the original vertical design load.

Such a variation is considered to be within the nor-mal design tolerance, i.e.; equipment designed for the origina'1 OBEloads is adequate to sustain the new loads.

3.

COMPARISON OF PEAK BUILDING ACCELERATIONS Original and new peak building accelerations in the reactor building outside primary containment are compared for. the same.two cases as in preceding Section 2, namely a.

Original OBE vs.

new (QBE + SRV),

, b.

Original SSE vs.

new (SSE

+ SRV + chugging).

Profiles showing the variation of the peak a'cceleration with elevation in the building are developed from readily available data for each of. the four loadings in the above cases.

The profiles of horizontal peak acceleration are shown in Figure 7 and the profiles of vertical peak accel-eration are shown in Figure 8.

Review of these figures shows that at all elevations, the original peak building accelerations are greater than the new comparable peak accelerations.

Thus," Figure 7

shows that a.

The original horizontal peak accelerations due to OBE exceed the new combined horizontal peak accelerations

'ue to OBE and SRV at all elevations.

b.

The original horizontal peak accelerations due to F

SSE exceed the new combined horizontal peak accelerations due to SSE plus SRV and chugging at all elevations.

Figure 8 demonstrates the same findings with respect to the 'vertical peak accelera'tions.

As previously noted, the dynamic design of the general category of relatively rigid equipment is controlled by peak building acceleration.

The results of this section demonstrate that. the addition of hydrodynamic loads has no impact. on the design of such equipment.

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4.

TEST. RESULTS FOR PEAK BUILDING ACCELERATIONS

, SRV actuation test programs have been implemented in two foreign BWR plants of Mark II containment configuration.

These are the Caorso plant in Italy and the Tokai-2 plant in Japan.

The program at the Tokai-2 plant is especially significant as the primary containment at this plant is a steel vessel, similar to the one at the subject plant.

A description of the test programs and results is given in References 3 to 6.

A summary of the maximum test acceler-ations in the plants is given in the Burns and Roe proprietary technical report (Reference 1).

The maximum building. accelerations recorded at the outside building walls at Caorso is equal to 0.007g.

Similar results were obtained at Tokai-2.

Inasmuch as the subject plant is structurally similar to those at Caorso and Tokai-2, maximum accelerations of the same small order of magnitude are anticipated in the reactor building outside primary containment due to SRV actuation.

Such small acceler-ations are not considered to be significant in design of equipment and piping..

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5.,

SUMMARY

AND CONCLUSIONS The study has investigated comparatively horizontal and vertical response accelerations in the reactor building outside primary containment.caused by the original (seismic)

J and the new (seismic and hydrodynamic SRV actuation and chugging) loads.

It has been determined that the original acceleration response spectra for the seismic (OBE/SSE) loads exceed the corresponding new spectra for the comparable combinations of seismic (OBE/SSE) plus hydrodynamic loads, except for the vertical direction in a narrow range of'requencies.

It is shown that in this range the increase in design load due to addition of hydrodynamic loads is small (less than 7%)'nd within the normal design tolerance.

Also, the study indicates that the original peak build-ing accelerations due to seismic loads exceed at all elevations the corresponding new peak accelerations due to the comparable combinations of seismic and hydrodynamic loads.

It is further noted in the report that peak building accelerations actually recorded during in-plant.

SRV actuation tests at reactor buildings similar to the reactor building at C

WPPSS Nuclear Project No.

2 are too small to be'onsidered significant=in design of equipment and piping.

Based on these findings, the loadings used for the design of the piping and equipment in the reactor 'building outside containment, namely the original seismic loads, are, effectively more severe than, or at least as severe as, the comparable loadings due to combinations of new seismic and hydrodynamic loads.

Consequently, it, is not necessary to reanalyze this piping and equipment as to capacity under the latt'er set of loads.

6..

REFERENCES (1)

"SRV Loads - Improved Definition and Application Methodology for Mark II Containments,"

Technical Report, Proprietary, Burns and Roe, Inc.,

July 1980.

(2)

(3)

"Chugging Loads Revised Definition and Application Methodology to Mark II Containments,"

(Based on 4TCO Test Results),

Technical Report, Burns and Roe, Inc., July 1981.

General Electiic Company Report, "Caorso SRV Discharge Tests Phase I Test Report,"

NEDE-.25100-P, May 1979.

(4)

General Electric Company Report, "Caorso SRV Discharge Tests Phase II ATR Report,"

NEDE-25118, August 1979.

'I r

(5)

Japan Atomic Power

Company, "Tokai-2, Main Steam

'Safety Relief Valve Operational Test Report,"

August 1978.

(6)

Japan Atomic Power

Company, "Reactor Building

Response

to Actuation of Main Steam Line Safety-Relief Valve(s) at Tokai-2 Nuclear Power Plant,"

TOK2-,RV-O 6 30.

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~ 005 HER'BE +.'RV, 'DAMPING =

. 02 16 32 40 48 FREQUENCY:

CPS HORIZONTAL ACCELERATION vs.

FREQUENCY EL.

567'ESPONSE SPECTRA OUTSIDE CONTAINMENT HPPSS NUCLEAR PROJECT NO.

2 REACTOR BUILDING FIGURE 1

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435'ESPONSE SPECTRA OUTSIDE CONTAINMENT

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2 REACTOR BUILDING FIGURE 3

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2 REACTOR BUILDING FIGURE 5

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