ML20062N024

Provides Status Update on Sqrt Program & Info Requested During 811130 Conference Call.Forwards Sargent & Lundy Rept, Equipment Qualification of LPCS Pump. Related Documents Encl

ML20062N024
Person / Time
Site:	LaSalle
Issue date:	12/15/1981
From:	Sargent C COMMONWEALTH EDISON CO.
To:	Schwencer A Office of Nuclear Reactor Regulation
Shared Package
ML20062N026	List: ML20062N024 ML20062N047
References
3033N, NUDOCS 8112180351
Download: ML20062N024 (47)

Text

r N Commomwealth Edison i / one First Nttional Plaza. Chictgo, libnois -

Address Reply to: Post Office Box 767 u

Chicago, lihnois 60690 -

Decembe r 15, 1981 0Yll ry a

/

+ <8> 'N($. 's Mr. A. Schwencer, Chief

~

Licensing Branch #2 I;

'l;.;

7 Division of Licensing di F

-7

'N

'I U.S. Nuclear Regulatory Commission f

Washington, D.C.

20555

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g s

O

Subject:

LaSalle County Station Units 1 and 2 SQRT Status Report and NRC Requested Information NRC Docket Nos. 50-373 and 50-374 Reference (a):

C. E. Sargent letter to A. Schwencer dated November 16, 1981.

Dear Mr. Schwencer:

The purpose of this letter is to provide a status update on the SQRT program as of December 7, 1981, and additionally to provide information requesteo by Mr. A. Bournia during our conference call on November 30, 1981.

SQRT Volume 10 Status report (Reference (a)) indicated that six LaSalle items of equipment were still undergoing testing to exteno the frequency coverage above the 33 Hertz of the initial test series and-to more completely test some items for record purposes.

The status is as follows:

~1.

Testing completed satisfactorily or reports being prepared:

C51-K002 A/H Voltage preamplifier C71-K14-A Contactor i

D18-N009 A/D Sensor Converter D18-K609 A/D Indicator or Trip Unit -

2.

Testing underway or scheduled:

l~

C51-K601 A/H Intermediate range monitor - test completion - December llth.

OOf r

i Cll-F0ll 2" Air Operated Globe - test completion - January 25th 3

Attachment (d) provides the justifications for interim operation //f ~

i for the equipment still undergoing test.

These justifications for interim operation are provided to satisfy any safety concern i

.regarding the operational significance of these devices and to facilitate closure of this item in the SER to allow fuel load.

8112180351 811215 i

PDR ADOCK 05000373 A

PDR L

A. Schwencer Decembe r 15, 1981 Also provided in this transmittal are documents requested by the NRC staf f on November 30, 1981.

The following documents are enclosed:

Attachments (a)

ASME-TGDA draft document dated 10/09/73, on fatigue evaluation for seismic loading.

(b)

A topical report by S&L on " Fatigue Considerations in Equipment Qualification."

This report was handed out to the NRC-SQRT audit members during the SQRT audit of LaSalle in 1980.

(c)

Fatigue Analysis Reports on LPCS pump consisting of:

(1) LPCS dynamic Analysis Report (from which the loads are taken)

(11) The fatigue evaluation report.

Please return the " Fatigue Analysis for the LPCS Pump" (attachment c.1) upon completion of your review.

With the summary of status and the accompanying interim operation justification for the two untested components, the necessary information to allow the SQRT issue to be closed with regard to granting the operating license has been provided.

We will continue to keep you appraised of any changes in the schedule as outlined, and will submit updated information for the final closecut items as it becomes available.

If there are any questions in this regard, please contact this office.

Very truly yours, eeR y C.E.

Sa rgent Nuclear Licensing Administrator Enclosures cc:

NRC Resident Inspector - LSCS (w/o Attachments)

Mr.

J.

N.

Singh w/o Attachments lm 3033N L

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DRAFT 10/9/73

6. 'l

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k*n%' CHI 4DIT ?

CHAPTER VI FATI@E EVALUATIO:i FOR SEISMIC LOADING Sid.,

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i I.

IllTRODUCTI0'l

}l hen a seismic enleation is perfomed, the analysis is directed touards a prediction of the r:aximum load which a particular congonent will experience during the entire ti.ma the earthquake is occuring, as is described in other sections cf this report. -This maximua load is then ccmpared to the lead uhich the cc.;onent can uithstand to indicate whether or not the design.

is adeq;ata. I!s;.2ver, when consideration is mde of tha entire beha'vior, i.e. time history, of the load which a cc:cponent experiences during an

^

carthquake, the tcc'nic characteristics come to light, and perhaps mare tl>an a pura ncn-cyclic load evaluation may be necessary in viewir;g tr.c -jows.

Upon insp2ctica of tha ntatcr of times in ubich the seismic loads change taagnitudo and diz action, one suspacts tItat ccrgenents my require-fatigue consideritions.

It is the purpose of this section to set fcrth 1

l the number of fatigue design cycles during a given earthquake and the num,';ar 1

of earthquakes which may be expected during the life-time of'a power plant.

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-Finally, soma discussion i.s Trade of the use of th.is infctmation for, fatigue

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evaluaM an.

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II FATICUE CYC!.ES FOR EAcil EARTilQ' JAKE To evaluate the am:.ber of cycles which c:4.ist withia a given earthquake, a

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typical boiling unter reactor building-reactor dynamic model was excited by three different recorded time histories. May 10, 1910, El. C,entro tis compc.nant c.

29.4 sec t 1932., Taf t, ?! 69 U cm1ponent, 30 seci.an'! Harch 1957, Golden

DRAFT 10/9/73 6.2

(

Cate S80E compc:ient,13.2 sec.

The modal response was truncated such n,that the respcasc of three different frequency bandwidths could be studied, 0+-1J Hz,10-20 Hz, and 20-50 Hz.

This was done to give a good approximatica to the cyclic behavior expected from structures with different frequcecy content.

The question of defining a fatigue cycle was met in the following manner:

The resultant tice history is assumed to be linear between the points at which responsas are calculated. As long as the sign of the slope of the line connecting the'se points remains the same, the loading is cons'idered to be continucily increasing (ordecreasing)inabsolutevalue.Howc?er, i,.,

if the slope of this line changes sign, then a stress reversal has, occured.

1 Since a cycle irglies that the loading hits come full circle, i.e. back to the original slope, two consecutive stress reversalr. c*onstitute a stress

~

I t

j cycle.

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.I -

Enveloping tha results frcm the three cat thquakes and avereging the results from several different points of the dynamic model, the following cyclic behavior was formed:

.s i

Table 1: Numbar of Dynamic Response Cycles Expected Durina a Seismic Event _

Frequency Sard 0+ - 10 10 - 20 20 - 50

!!o. of Seismic cycles 168 359 643

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DRAFT 10/ 9/73 6.3 l

The amplitudes of stress reversals is also important. Evaluation of the ground motion sh;us an earthquake to have r'elatively few peaks of high amplitude, in fact approximately 90-95% of the peaks in the referenced time histories are belcu 50% of the maximum peak. Therefore, one may expect the response to have approximately the same cyclic makeup.

Tables 2, 3, and 4 demonstrate the make-up of some selected compcnents of the dynamic nodel.

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Percentage of Stress Cycles j

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-Mith -Stress Amplitudes Below 50% of the Maximum Load X

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. lFrequencyRar.gr

. Hz.

10-20 Hz 20-50 Hz

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_ pft gra,,

pntro Taft Gate

-. Earthquake

! Ceptre ' Taft Gate a

t Duration, sec l__29.4-30.0 13.2 t 29.4 30.0 13.2 l 29.4 30.0 13.2 c

m 1 Component PERCENTAGES A

99.2 98.,0

'99.9 94.7 95.6 90.8 94.7 99.2 96.8 i

j 99.2 97.1 99.9 5 S7.9 97.9 96.8 97.6 99.1 97.7 9

,B 95.4 96.4 92.8 99.6 99.8 99.3

< C 99.0 95 8 99.9 j

ID.

99.1 96.9 99.5 96.2 S7.2 94.1 99.2 99.6 98.9

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99.1

.95.8 99.9,'

95.3 95.8 -

91.3 95.7 99.8 99.1 l

98.5 95./,

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99.8 h95.0 95.7 93.4 97.1 99.4 98.3

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Time. History Input Cycles Below 50% of Peak c-

' TI Centro 93%

Taft 90%

95%

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!! Frequency _Ra.nge 0

10 Hz 10 - 20 Hz 20 '50 Hz M

,l El Golden

'l.El Golden l

q Eli Goiden jl~entro

I Earthquake Centro Taft Gate Centro Taft

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C Taft Gate

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Duration, sec

,- 29.4

-30.0 13.2 f[.29.4 30.0 13.2

' 29.'4 30.0 13.2 j

Component LP E R C E N T A G E e

n A

85.5 85.4 96.2 80.6' 77.6 81.3

' 81.9 89.4 87.5 B

85.8 84.9 95.2

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80.4 78.0 80.6 82.6 80.3 89.3 s

C 93.1 81.2' 99.4 82.3 77.3 81.9-92.1 92.8 94.0 D

M.0 8110 98.0 84.3 79.6 83.4 90.0 93.0 91.4 1

4 3

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.3 79.4 97.6 82.0-79.1 80.6 80.9 9I.2 90.4' I

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30.6 79..5 92.2 84.4 78.9 83.3 7,9. 5 89.0 89.0

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, Average 87.7 81.9 96.6 82.3 78.4 81.9 84.5 89.3 90.3 (Over-all average 85.9) x, Tim: History Input Below 25% of Peak

.% =

5

't El Centro 78%

- Taft 70%

-. Golden Gate 90%

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1 T;J1LE 4 DETAILED STUDY OF STRESS REVERSAL PERCENTAGES FOR 0-10 liz RN!GE LOAD RANGE C0!iPO? LENT 0-p/4 p/4-p/2 p/2-P A

85.4 12.6 2

I' C

.81.2 14.6 4.2 3

D 81.0 16.0 '

.3.0

[i-AVERSGE G2.5 14.5.

3.1 i

I<V!t i

Intcrpolatin9 those results:

I p/2-3p/4 contains 2.56% of stress reversals 3p/4-P contains.44% of stress reversals i

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4.5% of Total Area-f5 f' BS". o ~ 4,S%,, -

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. @ j#f,. Total-Area

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hamneavea.6:./,4'2 Total Area $)kyfy8, 2778,7/.Dygg'f,m.a._._

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%,.n.ww..r.,b.=,. e,.. Mt.4 ~a me aw as a.

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26 50 75 100 Percent of Peak Value ll.

PIG. 1 DENSITY OF STRESS REVERSALS

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DRAFT 10/9/73

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From these tables it can be conservatively stated that, independent of earthquake o.- component frequency, 99.5% of the stress reversals occur below 75% of the maximum stress level', and 95% of the reversals lie below

. 50% of the maxirum stress level. Th'is relationship is graphically shown i

in' Figure 1.

In sumary, the cyclic behavior number of fatigue. cycles of a ccmponent M

'during an earthquake is found in the following manner:

J. The fundamental frequency and peak seismit loads are found by a standard seismic analysis.

L The number of cycles which the compo~nent experiences are found from 2.

C-Table 1 according td the frequency range within which the fundanental ll frequency lies.

For fatig:e evaluation, one-haif percent (.005) of these cycles are 3.

j conservatively assumed to be at 'the peak ictd 4.5 percent (.045) at three-quarter peak. The remainder of the cycles uill have negli<jible',

1 contributien to fatigue usage. The use of three numbers is discussed i t I.

in Section IV o'f this chapter.

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DRAFT 10/9/73 6.8 g

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, III.

PREDICTIO!! 0F NUMBER OF EXPECTED EARTtiQUAKES f

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' Now that tne nunber of cycles expected during an~ earthquake has been defined, 1

there remains the question of how many earthquakes may be expected during the postulated 40 year 1.ifetime of a plant. The two levels of earthq'iake intensity are examined separately.

. The safe shutdot:n earthquake has the highest level of response. However,

~

the encounter probability of the SSE is so small that it is not.necessary l

to postulate the possibility of'more than one SSE during the 40 year life 1

of a plant.

Fatigue evaluation due to the SSE is not necessary since it b,

is a faulted condition and thus not required by ASME Section III.

T)$e half-SSE is an upset condition' and t!ierefere raist be included in fatigue z.

cvaluations acc~ording to ASi!E Section III. -Investigation of seinaic histories

?l l

Q in.PSARSofrunyplants(SeeChapter3)*showthatduringa40yearlife J

It is probable that five earthquakes with intensities one-tenth of the SSE 1

intensity, one earthquake approximately 20% of the proposed SSE intensity,

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Therefore,theprobabilityofevenaha,1f-SSEisextemejylow.

pill. occur.

To cover the comtiined effects of these earthquakes and the cumulative).

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effects

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of even lesser' earthquakes, one half-SSE intens'ity earthquake is postulated A

j for fatigue evaluatlon.

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It should be noted that,tbe relationships between tha number of cycles, i

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fatigue usage factors, and peak stress is highly nonlinear.

Equivalent

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' number of earthquakes are estimated from the relationship given in ASME.

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Section III between fatigue cyclic test multiple factors for stress level,

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DRAFT 10/9/73 6.9 x

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' K, and number of cycles K *

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.g = (;f,$)% w j

%ueb For example, if the quarter-SSE eartdquake (QSSE) has a stress level half K csss ' X K' tass ~e S

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l that of the half-SSE (llSSE),.

K n a i 2.Ks455C:-

nQSSTs..(Ks QSSEl4.3 (EsOSSE)4.3 - 1/20 K

)

i 7'n llSSE J s liSSE sQSSE z

g7 w

w That is gp3m g y l

Kn HME 2.0

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n liSSE 20.n QSSE Kn I405#

• E0 G GL556 K

=

or one half-SSE Ns the same cumulative fatigue damage as 20 quarter-SSE's.

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Frca this, it is e/ident that the conservatively chosen number of cycles for the one halt-SSE postulated to occur, tiill core than account for the fatigue usage factors accumulated during smaller earthquake events. k

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DRAFT 10/9/73 6.10 IV. Tile USE OF FATIGUE CYCLE NUMDERS The amount of cumulative fatigue damage that results from the stress cycling is found by use _of Itiner's c'riteria in which it is assumed that cyc'les at if 11) cyclas t.ould produce failure at a stress level S, then n j

j the same stress level would use up the fraction n /fi of tha total fatigue j

j life.

Failure cccurs when the cumulative usage factors equals to one.

Therefore,lline,r's Criteria states s

n j

I -<1 1 N y

(,' )

v 1

.for.no fatigue failure.

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3 To simplify the analysis, seismic cycling is postulated to be composed of three parts (See Section II). One half percent of the seismic cycles are said to exist at the level of the peak load previously calculated, another

.i-4.5 percent occurs at three quarter peak level and the rest (9b%) at.

Astables2,3.,andI. demonstrate,thes'evaluesare\\

half, paak.leyel.

con

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'servatively chosen.

The equation used to ensure that no seismic fatigue failures occur becomes '

3

.005 ff

.045 it.g

.tJ5 !!

T T

<1 l

II)

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2 3

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DRAFT 10/9/73 6.11

.f

.z

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f!N

- Total number of cycles from Table 1 T

i,

' lj N,

- Allowable cycles at. peak loading U

N

- Allowable efcles at three quarter peak loading 2

N

- Allowabic cycles at half of peak loading N, N, N from appropriate 3

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3

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U llowever, the last tena becomes insignificant, so the above inequality may-bh roduced to 3? C '.-

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DRAFT 10/9/73 6.12

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i REFERENCES 1.

CRITERIA 0F THE ASME BUILER AND PRESSURE VESSEL CODE f0R CESIGN BY ANALYSIS IN SECTIONS IV AND VIII, DIVIS!0N 2, A:iERICAN SOCIElY OF MECHANICAL ENGINEERS, 1969.

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- g: >t*.

• . ',?Y'$ llr ' a ~,b,'l)" h y.

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• i

t

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?, ;$ '. '. :.#? r.;,'.. *

'; y '..[

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,, 2 : a,# ;-, s* *J.e. '. '.,(.',',.

• ..**c.

f,,

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FATIGUE CONSIDERATIONS IN EQUIPMENT QUALIFICATION I.

INTRODUCTION

()

With the identification of the SRV and LOCA hydro dynamic loads the effect of the large number of cycles associated I

with these events has become an important issue.

Our approach in addressing this issue is to test and analyze representative. pieces of equipment to evaluate the,ir sensitivity to fatigue.

Presented here is the methodology

__ {

f and philosophy in defining our fatigue test program.

f, !

II.

METHOD OF FATIGUE TESTING I

Two of the most widely used and accepted dynamic test methods were considered for this fatigue test program.

These methods are the single frequency continuous sine

(])

test and the multiple frequency random motion test.

Each

2 -

of these methods possess certain advantages and dis-219 advantages which are outlined below.

j li Single Frequency Continuous Sine Tes-t Advantages:

_l t

1)

Assures complete stress revernal or full stress range for each cycle

}

2)

Able to define number of equiva'ent cycles j

l t

using statistical me.thodology of ASME Task i

i Group on Dynamic An: lysis.

Disadvantages:

l 1)

Input only at sing 1c-Ircquency, will no' I,

consider participation of other modes.

l

}

I

)

i o

4 Multiple Frequency Random Motion Test j Advantages:

1)

More closely represents the actual nature of the loads 2)

Considers participation o'f'all modes Disadvantages:

1)

Difficulty in defining test duration introduces uncertainty 2)

With accelerated testing cannot assure complete 2

stress reversal of full stress range between occurrences.

After evaluating these advantages and disadvantages it was

.r decided to use the-multiple frequency random motion test.

a:

x.

III.

DYNAMIC LOADS Table 1 identifies all of the dynamic events and the number of cycles associated with each event.

The number of cycles un associated with the SRV case are not considered separately

_4 gggy for fatigue evaluation as they are already included in'the SRV case.

It is important to' recognize that the dynamic gg mechanical loads of the SRV are the same as with the gggy

'SRV and SRV FA SA cases.

(i.e. It is Just another pop of the j

safety relief valves differing only that the valves fail to close).

The annulus pressurization case is not considered for fatigue

(

evaluation because it is a single event, of very short 4

duration, limited only to the area of the biologica1' shield wall'.

=

a.-.

In considering_the' cases asso'ciated with the LOCA event i-

[.

(CO and CHG) it is important to evaluate the impact of each of the postulated accidents ~(DBA, IBA,~and-SBA)'.

For fatigue purposes the SBA was chosen because of its large number of cycles ~ compared 'to the IBA.

Below is provided a.ccmparison of the IBA'al.d SBA pos'tulated accidents.

Postulated Number of Equivalent-Cycles Accident col or CO2 CHG IBA 525 90 r

SBA 0

3000 IV.

FATIGUE TEST PROGRAM i

).

Tlie fatigue test program consists of four separate biaxial

~

random motion tests.

These tests and the loads used to define the required response spectra for each' testi are:

1)

Upset Condition Test OBE + Envelope (SRVFA SA 2)

Emergency Condition' Test.

~

SSE.+ Envelope (SRVpg & SRVfg) + CHG 3)

SRV Fatigue Test s

SRV gg 4)

SRV + LOCA Fatigue-Test

~

SRVgg + CHG D

s I

The spectra was combined by the absolute sum method.

For i

the upset and emergency condition tests high frequency l

V j white noise background vibration conesponding to a ll

/

constant velocity of 0.12 and 0.20 ft/sec respectively,

- ll

!)

was add'ed to the design basis curves to allow for any future-increases'in th'e high fre_quency region and to I

all,ow for inve'stigation of impact of high. frequency loads.

The duration of each test is:

Test Number of Total Test Duration Tests Per Number of Duration (Sec)

Orientation Orientation (Sec)

Upset 15 2

5 150 Emeig.

15 2

1 30 SRV 350 2

'l 700 SRV + LOCA 150 2

1 300 The method used to determine the duration of each test is presented in Table 2.

o O

e 9

O O

+

y O

e O

f 9

0 e

e e

0

TABLE 1 Determination of Number of Equivalent Cycles' Penk Nunber of Number of Number of Total Number of Distribution Even. Description Design Occurrences Cycles per Number of Equivalent U

Events per Event Occurrence Cycles Cycles Fa to (Note 1)

(Note 2)

(Note 3)

(Note 2)

I N_o_t e---3)

OBE 1

1 10 50 1.00 50 SSE 1

1 10 10 1.00 10 s

i SRV 253 1

10 2530 0.30 759 first Actuation SRV 169 34 10 57460 0.20 11492 subsequent Actuation I

Ystuck open relief valve 3

1 10 30 Not defin ed col or CO2 1

Not Def ined Not Defined N( t Defined 525 CIIG 1

1000 3

30000 0.10 3000 AP 1

1 1

1 Not De ?ined

. Notes: 1) Per Zimmer Design Crlteria A51-70 Rev. 01

2) The peak distribution reduction factor was determined using the method suggested ASME B&PV Special Working Group on Dynamic Analysis

3) Number of equivalent cycles;- total number of cycles x peak distribution' reduction factor 4

l l

I I

r.,

TABLE 2 Worksheet for Determining Test Durations Number of Number of Number of Number of Patigue Tes t Time (Sec) Time (Sec)

Event Equivalent Cycles Cycles Cycles Duration Considered Remaining ycles Mult Considered Considered Remaining Required in Case 1 for' Case 2 Description Cycles by 1.5 in.Emerg.

In Upset for Fatigue Seconds (Note 5)

(Note 5)

(No_te 1)__

TestJN,ote_21 [es t_(No,'

31 Tgst QQ3g_4)

OBE 50 75 2250 0

0 J

SSE 10 15 450 0

0 SRV 759 1139 450 2250 0

0 gg SRV 11492 17238 450 2250 14538 969 300 669 g3 SBA case is considel ed more ses ere than IB i case as SHA case has many more c ycles.

col or CO2 For SBA

here is no CO.

CHG

~ 3000 4500 450 4050 270 300 Notes:

1) Factor of 1.5 is a safety factor to account for uncertainties

2) This is based upon using an average of 15 CPS for the 30 Sec duration of the emergency condition test.

3) This is L' sed upon using an average of 15 CPS for the 150 Sec duration of the

. upset condition test.

4) This is based upon using an average of 15 CPS

5) Two cases are considered in accounting for potential fatigue effects Case 1: SRVg3 + CHG

~

Case 2: SRV gg 9

Attachment (d)

LaSalle County Station, Units 1 & 2 Justification for Interim Operation SQRT Program Ten status reports have been issued on the LaSalle SQRT Program to record the completion of equipment qualifications over the past two years.

A test program to extend the frequency coverage to 100 bz was added to the arogram after the chugging load definition was accepteo.

Inis component and assembly testing was additional to the in-situ equipment tests done in the plant during 1980.

This external testing is nearing completion and in fact would be completed now except for that difficulties during this wrap-up phase when improper techniques ard misrouted test equipment caused delays.

As indicated on the cover letter for this eleventh status report, only two items have not finished testing to the higher frequencies; they are:

a)

Intermediate range monitor C51-K601A/H b)

Two-inch air-operated globe valve Cll-F0ll.

The last status report included SQRT summary forms for these incom-plete test items for the older -33 Hz threshold.

Even though test completion is expected prior to actual fuel load, a specific functional justification for interim operation is included here-in to satisfy any safety concern regarding the operational significance of these devices during a seismic and dynamic loading event at LaSalle.

Part A Intermediate - Range Monitor (C51-K601 A/H) This device was originally called a wide-range neutron monitor in early BWR's and it was originally listed that way in the LaSalle SQRT Program for test identification purposes.

It is essentially the intermediate-range neutron monitoring system (IRM), however, and its functional use is referenced in this operations analysis, therefore a more specific designator is appropriate.

The IRM system provides an alternate scram to that provided by the APRM (15% setting) during startup operations.

The Intermediate Power Range Monitor (IRM) system is a neutron measuring system used during reactor start-up.

The system monitors neutron fluence in the reactor core by eight fission chambers (A through H) inserted in dry wells.

The output of these chambers is signal conditioned by a mean square voltage system (Sensor Converter) to discriminate against gamma radiation.

The function of the system is to provide the operator a rate-of-power-increase

, indication to assure reasonable and efficient power increase from startup at source level to about five percent neutron power.

At this power the Local Power Range Monitors and hence the Average Power Range Monitor (APRM) is sufficiently on-scale to indicate reactor power.

The rate of power ascension is monitored by ranging o f the linear readout of the IRM.

This equipment prevents rod withdrawal when any non-bypassed IRM channel is down scale and initiates scram if one channel in each safety division goes above the high level satpoint.

The linear readout is manually ranged (Range Switch) for each of the eight channels from the source range level to the power range in one-half decade steps.

The safety function of the system is to limit reactivity changes via manual rod withdrawal to that rate which can be observed and followed by the operator by appropriate ranging.

During the reactor startup, the reactor core power level protection against overpower (MCPR) is achieved by the APRM.

The APRM trip level is set down to 15% power (per Technical Specification) when the reactor mode switch is in the START-UP mode.

Justification for this is provided in the basis for the technical specification.

The SRM does not perform an essential automatic safety action during start-up; rather, the APRM trip governs this via this set down to 15% power; and the IRM provides the alternate start-up trip function.

If a seismic event occurs during reactor operation in the startup range, the reactor core protection is provided by the seismically quallified APRM trip in the set-down mode.

The IRM need not function as the operator will not attempt further reactivity withdrawal by control rods whenever a seismic event occurs but will shutdown the reactor.

This IRM trip is routinely removed via mode switch action when the RUi4 mode is entered at about five percent neutron power level.

Acknowledgement should be made that the IRM equipment was qualified to the 33-Hz level which was the original SQRT criteria and that such qualification is adequate for the start-up mode of operations (before switching to RUN via the mode switch) with respect to any seismic event.

The incomplete test on this IRM device is the extension of frequency coverage to the higher frequencies (to 100 Hz) ascribed to hydrodynamic (chugging) loads.

Therefore, for the initial fuel loading and ascent to power at LaSalle, from the combination of the alternative start-up scram provided by the APRM's (Chapter 15.4) and the original IRM qualification for seismic event frequencies.

It is concluded that sufficient justification exists to exempt this SQRT requirement until test coveraie is completed to 100 Hz.

Part B Two-inch Air Operated Globe Valve (C11-F011)

This justification for interim operation covers the drain valve on the scram discharge volume, which valve is an ASME Class 2 valve currently undergoing dynamic qualification testing as part of the LaSalle/Zimmer SQRT Program.

The drain valve (Cll-F0ll) on the scram discharge volume (SDV) is normally open but closes to retain CRD exhaust water inside the SDV piping during and following a scram.

It reopens on resetting of the scram; when open, it drains the SDV fluid which is reactor water to the reactor building equipment drain tank from where eventually it goes to the radwaste system for treatment.

The draining ection assures that adequate SDV space is available for the ejecteo water from subsequent scrams.

The LaSalle SDV is sized to take two repetitive scrams without opening this drain valve.

The SDV has an instrumented chamber with level sensors which prevent rod withdrawal if the volume is above an intermediate level; they also initiate scram when the level reaches a limiting value above which a scram might be slor9d in rod insertion rate.

During normal reactor operations the SDV would at any point in time be at such a level that successful scram can be accomplished.

CRD leakage flow normally passes through the SDV at low pressures.

The drain valve need not be qualified to operate during or after a seismic event; it need only meet passive requirements of pressure integrity according to ASME Code.

The reactor can be scrammed and remain safely shut-down without regard to the initial position of this valve and without subsequent operation of this drain valve.

Its primary safety purpose is to preserve pressure barrier integrity during the brief interval (less than 0.2 percent o f the time) when the SDV is pressurized during a scram.

It is an intermittent pressure boundary device, whose dynamic qualification does not af fect reactor safety but retains reactor water inside the SDV which is inside the reactor building.

(A separate NUREG 0803 evaluation addresses the structural integrity of the SDV piping and equipment from the containment point of view).

The operation of Cll-F011 merely enables scram / leakage flow to arrive at the reactor building equipment drain tank as directed by the operator (reset);

it does not open the pressure boundary to the outside environment, therefore the qualification issue is separate from the 0803 concern.

_4 _

From the above reasons it is evident that no safety significance is involved in the operation of LaSalle prior to completion of SQRT testing to 100 Hz of SDV drain valve Cll-F0ll.

3033N

SARGENT & LUNDY

) l[; ;'t)7 &

[/

-T ENGINEERS

,s t._..'

Summary Report for Faticue Analysis for LPCS Pump I.

Introduction This report identifies the method and results associated with the fatigue analysis of the LPCs pump / mot'or assembly when subjected to static and dynamic loads.

The analysis is performed to assess the adequacy of the pump / motor assembly design with respect to fatigue, when subjected to static, seismic and as well as various hydrody.amic loads per project specification.

A three dimensional lumped mass beam finite element model of the pump / motor assembly and its support is developed and dynamically analysed using the response spectrum analysis method.

The same model is analysed due to static nozzle loads, pump thrust loads, and dead weight.

Critical location stresses and usage factors are evaluated and results are presented.

II.

Method of Analysis

/

As stated in the previous section, a three din ensional finite element model is developed for the pump-motor assembly and used i n the analysis.

The merits of using this approach are three fold: (1) it represents the correct distribution of mass and stiffness for the determination of natural frequencies, including higher nodes, (2) a complete dynamic analysis may be performed to determine the loads using this model, and (3) it correctly determines how the loads are distributed through each load path in the statically indeter-minate portions of the structure.

1

W:

SARGENT & LUNDY ENGINEERS CHICACO The pump / motor assembly with some details are shown on Figure 1.

The schematic representation of the model is provided in Figure 2.

Essentially, the model is a three dimensional lumped m'ss beam element model, capable of a

accepting loads from the vertical.and two horizontal directicns simultaneously. -Model has a total number of 105 nodes, 97 beam elements and 16 boundary elements.

S&L SAPIV computer program is utilized for both dynamic and static analysis of the model.

The analysis is performed in two distinct steps.

First, a modal analysis is performed using the Subspace Iteration Technique and all eigenvectors are saved on a permanent file for subsequent analysis with various dynamic loading conditions.

Twenty-five modes'were considered in the

~

analysis, with the highest frequency of 140.5 cycles /second.

This frequency is high enough to cover the ZPA of all applicable response spectra (both seismic and. hydrodynamic-series).

Table 1 on Page 5, gives the frequencies of the model.

As can be seen from this table, the fundamental-frequency.of the equipment is 7.82 cycles /second,. justifying i

the need for detailed dynamic analysis for this flexible equipment.

Second step of the analysis consists of applying various dynamic and static loads to the model and investigating the structural behavior of the pump / motor assembly.

Response Spectra Analysis method was selected for 'the dynamic analysis of the assembly.

Response spectra curves for this analysis are conservatively developed by enveloping the EW and NS curves for horizontal spectra and wall and slab curves for vertical spectra at elevation 673'-4" of reactor building.

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SARGENT & LUNDY EN GIN E E RO CHICAGO r

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[

SARGENT & LUNDY ENGINEERS cMicAso This enveloping process is performed for all seismic and

. hydrodynamic spectra curves, namely SRV,. CHUGGING,

~ Condensation Oscillation, ODE and SSE.

Response spectra curves for'SSE are given on pages'7 - 9 as an example.

For the static analysis of the pump / motor assembly, action of the following loads are investigated:

(1)

Nozzle loads, obtained from the analysis of the piping'sub-systems attached to the suction and discharge nozzles, (2)

Weight of all the metal' components and the water' inside the pump, (3)

Inte,rnal design pressure acting on pressure boundary components,

-(4)

Pump thrust load.

The forces and. moments are obtained for each element as'a res' ult of all the dynamic and static loads.

Element loads are reconbined in the following categories using absolute sum method:

a)

Operating loads b)

SRV i

c)

CHUGGING d)

Condensation Oscillation e)

OBE f)

SSE

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ENGINEERS CHICAGO

.(

Thesl ccmbined loads are then used to calculate the component' stresses at critical locations.

Evaluation of stresses were done according to the general criteria-set by subsection article NB-3200 of Section III Code for service levels B&C.whenever applicable.. Specialized stress analysis techniques are also used whenever necessary.

Number of equivalent cycles for each. dynamic load case was taken from Reference 1.

Using these equivalent cycles, fatigue evaluation were done by the methods of-NB3222.4 (e).

Equivalent cycles for pressure variations were estimated n

from Pressure-Temperature-Transient histories of the LP02 and RH07 Subsystems (Ref. 2).

TII.

Results

,c Five representative locations were selected for fatigue

'f,'

usage factor computations; namely, discharge head bolting, stuffing box-discharge elbow connection, discharge column,:

pump first stage casing at minimum section and motor stand. -

In the calculation of usage factor for bolting, a fatigue-y strength reduction factor of 4 is used.

Usage factors are shown in Table 2.

j Table 2 I

' Fatigue Usage Factors

.x.

/

Location Eui s

Discharge Head Bolting 0.221 Sttyffing Box-Discharge Elbow Connection 0.141 Discharge Column 0.042 Pump First Stage Casing at Min.Section 0.012

^

Motor Stand 0.011 i

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SARGENT & LUNDY ENGINCCRS CHICAGO Summary information for usage factor computations are given in Tables 3-7.

IV.

Concluding Remarks Based on this analysis, the total fatigue usage factor does not exceed 0.221 for bolting and 0.141 elsewhere.

Therefore, the LaSalle LPCS pump is qualified for fatigue s

due to both dynamic and static loads.

V.

References 1.

" Number of Equivalent Cycles due to Dynamic Loading",

handout from M. Hassaballa, 07/07/80.

2.

" Pressure - Temperature - Transient History - Subsystems LP-02, RH-07", EMD File No. 023028, Rev. O, 12/26/80.

3.

Welding Council's Bulletin WRC-107.

4.

" Qualification Report for LPCS Pump, E21-C00,1, LaSalle 1",

EMD File No. 028197, 02/03/81.

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