diablo canyon, units 1 and 2 - response to open items in … · 2017. 1. 31. · x regulatory...

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REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS) g7DISTRIBUTION FOR INCOMING MATERIAI /323

REC: STOLZ J F ORG: CRANE P A DOCDATE: 06/30/78NRC PACIFIC GAS 8c ELEC DATE RCVD: 08/29/78

DOCTYPE: LETTER NOTARIZED: NO COPIES RECEIVEDSUBJECT: LTR 1 ENCL 40FORWARDING RESPONSE TO OPEN ITEMS (AS LISTED> IN SUPPLEMENT NO 7 TO THE STAFFSAFFTY EVALUATION REPT.

PLANT NAME: SIMBA~~ UNIT ~< qDIABLO CANYON .— UNIT 2

REVIEWER INlTIAL: XgMDISTRIBUTER INITIAL:gQ/

DISTRIBUTION OF THIS MATERIAL IS AS FOLLOWS

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FPIIIIPTIIIIIIIIIIFIrer I IIE PIIRIPPLC IF 1 C GA.S A.ND 'LZ'CT R-l~C '-<CtO. P@~'T77 BEALE STREET, 31ST FLOOR ~ SAN FRANCISCO, CALIFORNIA 94106 ~ (415) 781-4211

LIOHN C ~ MORRI SS BYVICE PRESIDENT AND OENERAL COUNSEL

MALCOLM H. PUR BUSHASSDCIATC OENERAL COUNSEL

CHARLCS T, VAN OCUSCNPHILIP A. CRANC, JR.HCNRY J, LSPI ANTSRICHARD A ~ CLARHC

JOHN S OISSONARTHUR L, HILLMAN,J R,

ROSCRT OHLSACHCHARLCS W THISSCLL

AESIETASI SESESAL EOUSICL

June 30, 1978JOSHUA OAS LEVROSEST L OOIIDONLCISH O DAS ~ IDVOESHASD J, DELLAOANTAwILLIANN, CDHA1DSJO ~ EPH O CNSLtsfJONN M PsvtPATSIC«O. OototoPCTCS W, MANSOHENJUA ~ II JATOP, RONALD LAUPHEINES~Icstr C LIPSDHJANCS D, LOSSOONRIOHASD L, NEIS ~DOUSLAS A, DottsofJ, NIONAEL 1tlDCNSASHIVOR C OANSONOut ANN LEVIN ~ CNIPPDAVIDJ WILLIAN~ OMOSUCC R WOSTNINOTON

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J, PCTES OAUNOA1THCS~TEVEH P OU1CCI ANELA OHAPPtLLEOSIAN O DENTOMOASV P CNOINASDONALD Color ~ ONDAVID D OILstsfANNETTC OATENRostsf L, NASSI ~KESNIT R KUSITETHEODOSE L LINDOESO, JS,TIIOHASD P LOOSEMA11V W LONO J1RISHASO M EIOS ~RODES J, PCTES ~RDSEOT R lllortffOHISLCTA OANOE1SONJACC W. Ouusr~ I~ 11LCV A WOO

OILS EAT L, MASS ISC CDHASO J NOOANHEVOLENN WE ~ T Jo DANIEL C OI~ SONDAN OSATSOM LU~ SOOC JO ~ EPH I KELLVJADr P. FALLIN.JS. HOWASO V, OOLU ~

~ ENIOS OOUN ~ EL

Mr. John F. Stolz, ChiefLight Water Reactors Branch No. 1Division of Project ManagementU. S. Nuclear Regulatory CommissionWashington, D. C. 20555

Re: Docket No. 50-275-OLDocket No. 50-323-OLDiablo Canyon Units 1 6 2

gf TIE~5Dear Mr. Stolz:

The following open items in Supplement No. 7 tothe Staff Safety Evaluation Report are addressed in thissubmittal: SER CONCLUSIONS, Section 22, item (3), sub-items 3.9.3.9 (1), (2), and (4) through (17) (pp. 3-69~3-70) .

'

Five copies of this submittal have been sentdirectly to Mr. Dennis Allison, one copy has been sentto Mr. Robert. Bosnak, and one copy has been sent toDr. Chester P. Siess in care of Mr. John McKinley.

Kindly acknowledge receipt of this material onthe enclosed copy of this letter and return it to me inthe enclosed addressed envelope.

Very truly yours,

Enclosures (40)CC w/enc.: Service List

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SER CONCLUSIONS 22.0, ITEM (3)

Section 3.9.3.9

( 1) "Submittal of a demonstration of the similarity between the Diablo Canyon

reactor coolant system and the reactor coolant system that was tested to

justify 4 percent damping."

Response: During the past several years, Westinghouse performed studies aime d

at determining the damping values for the reactor coolant loop system. These

studies included data search to evaluate previous data as well as the state ofthe art and to conduct tests of a prototype plant to obtain actual damping

measurements under different conditions (this test program was conducted atthe Indian Point II Nuclear Power Plant). The results of all these studieshave been compiled and organized to show in a simple manner the real damping

characteristic of a reactor coolant loop. This work was presented at the

National Topic Meeting on Water Reactor Safety, sponsored by the American

Nuclear Society, and published by the US AEC as CONF 73304. A copy of the

paper is enclosed as Attachment I. After this work was published, others have

independently published more work on the subject (1) which confirms the

results obtained by Westinghouse. As a result of these studies performed by

Westinghouse, the NRC staff authorized the Westinghouse Pressurized Water

Reactors Division to use 4% damping (2) for the seismic SSE and/or the LOCA

dynamic analyses. This decision was based on the very restrictive and

conservative approach of not allowing any data extrapolation despite the clearevidence that damping increases with system response beyond the 4X damping

approved.

The information provided in Attachment I shows that Westinghouse reactorcoolant loops behave in a very similar manner under seismic excitation and

following a modal pattern. The Diablo Canyon loop piping, primary equipment,

and primary equipment supports are similar in configuration to those in the

(1) Nuclear Engineering and Design, Vol. 25, No. 1 June 1973.(2) WCAP 7921-AR "Damping Values of Nuclear Power Plant Components" accepted

by the NRC ~

D/C SER Supp. 7

SER CONCLUSIONS 22.0, ITEM (3) (CONTINUED)

Section 3.9.39 (Continued)

plant used by Westinghouse for the Indian Point tests. The primary piping forDiablo Canyon is made from similar material and has approximately the same

overall dimensions, dynamic response, and structural stiffness as the testplant. The reactor coolant pump, steam generator, and reactor vessel forDiablo Canyon are respectively manufactured from the same material, using thesame fabrications methods, and have similar dynamic responses to the testplant. primary components. The primary equipment supports are attached to the

equipment at the same'relative locations in the same fashion. The support forDiablo Canyon, like the test plant, consist of bolted and welded structuralsteel elements and are designed to perform the same function, i.e., allow forfree thermal expansion but restrict the motion caused by other loads. The

functional similarities of supports in the design is such that the loop pipingand components respond dynamically in a similar manner for both cases. The

modal damping is presented as the losses produced essentially by the main

heavy components (vessel, steam generator pumps) vibration on their supports.The Indian Point II shaker tests have demonstrated that the magnitude ofdamping increases with amplitude independently of the frequency changes thatthe system could have.

From the information enclosed. in the Attachment I, it can be seen that most ofthe data provided was obtained for extremely low stress levels and wi,thouthaving damping present from bumpers and snubbers., Also it can be seen thatthe nonlinearities between components did not produce losses similar to theones occurring during earthquakes because of the low vibration magnitude ofthe available data. The enclosed Table A provides a summary of approximately.maximum displacements given by the analysis and the corresponding deflectionsat the top of the steam generator correlation at 4X damping.

In Figure A, the line plotted in Attachment I has been extrapolated to thedeflections obtained at the top of the steam generator for Hosgri event

analysis of Diablo Canyon. It shows that the reactor coolant loop has lossesmuch larger than the value adopted for the Hosgri analysis. The line extra-

D/SER Supp. 7

SER CONCLUSIONS 22.0, ITEM (3) (CONTINUED)I

Section 3.9.3.9 (Continued)

polated in Figure A shows 14X damping for 2" deflection using data, which as

stated above, does not include several other phenomena which will absorb

energy: stresses, external supports, and nonlinearities. Figure A also shows

the level of the San Fernando earthquake in the San Onofre Plant with the

proposed line which was obtained from regression analysis based on small

deflection data.

The magnitude of the earthquake measured in the free field of the San Onofre

,Site was .012 g with steam generator damping values between 3 and 5X (see

Attachment I). The present Hosgri input to the Diablo Canyon Plant is more

than 30 times larger, which clearly indicates that the damping values adopted

are extremely low and corresponding to a very restrictive criteria to insure

additional extreme conservatism. The issue of higher inputs from the floordue to lower damping adopted for the building damping is shown in Figure. B

where the following assumptions are made:

1. The building damping is 5% instead of 7X.

2. The magnitude of free field of earthquake in Diablo Canyon variesfrom .6 g to .75 g.

The plot indicates what damping of the main coolant loop needs to be changed

to obtain the same stress level and consequently, the same margin.

As shown in the Figure, the reduction of interior concrete damping from 7X to5% would require that the reactor coolant system damping be increased to6-1/2X if the maximum ground acceleration remains the same (i.e., 0.75 g).

On the other hand, should the damping value of the reactor coolant system

remain at 4X then it is necessary that the maximum ground acceleration be

reduced from 0.75 g to 0.63 g to compensate. for the reduction in the interiorconcrete building damping.

D/C SER Supp. 7

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TABLE A

MAXIMUMDEFLECTIONS DURING FAULTED CONDITIONS

Deflection Location

~adinca

HoslDEX

At Top of Steam Generator

At Top of RCL Pump

Maximum Test Amplitude

0.5,

0.5

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PRESENT AT THE NATIONAL TOPICMEETING WATER REACTOR SAFETY OF THE

AMERICAN NUCLEAR SOCIETY, SALT LAKE

CITY, March 26-28, 1973

DhtPlNG FOR DYNAMIC ANALYSIS OF REACTOR COOLANT LOOP SYSTEMS

George J. Bohm

Vestinghouse Electric CorporationPittsburgh, Pennsylvania

ABSTRACT

The behavior of reactor coolant loop components under dynamic loads is nor»mally obtained analytically by assuming that the system has an energy dissipationmechanism which can 'be represented by equivalent viscous damping coefficients.In this paper, the results of tests in past experimental'work are correlated withthe concept of losses due to material and structural damping as well as to thevarious modes excited. Damping ratio values are proposed for use when dynamicelastic analysis is performed to obtain system response under seismic excitationand loads due to postulated accidents. The dependence of damping values on themagnitude of component response also is discussed.

I

INTRODUCTION

The effect of damping on the dynamic response of structural systems has beenrecognized in the past, and studies have been performed to quantitatively estimatethis effect. It is Known that vibrating structures have energy losses which de-pend on numerous factors, such as material characteristics, stress levels, andgeometric configurations. This dissipation of energy or damping effect occurs

, because energy is transformed into heat, sound waves, stress waves, and otherforms. The important problem is to establish a reliable method of measuringenergy losses in a way that will represent the system characteristics and thatsimultaneously can be used in dynamic analysis to predict system response undergiven forcing functions.

In complex systems, such as a reactor coolant loop, the determination of theenergy losses occurring during vibrations becomes a very difficult task. In thissystem, several large components, i.e., vessels, pumps, and steam generators, areconnected by pipes and supported by concrete structure through various types ofsteel supports. Figure 1 shows a typical configuration of a pWR with the reactorcoolant loop components; for clarity, supports are not included in the drawing.These large components have complex internal structures with small clearances be~parts. The impact of the internals increases the energy loss when the componentvibrates.'lippage between components and their supports will also contributeto the energy loss. Qhen vibration amplitude increases, losses will increase ina manner difficult to predict. For most physical systems, it is difficult to gi"a mathematical treatment of the damping mechanism.

608

STEAMQENERATOR

PRESSURIZER

MAINCOOLANTPUMP

REACTOR

Fi . 1 — Main Coolant Loops of a PMRgo

dels and values representing the da p gm in of theThe adoption of realistic mode s anOf articular interestsystem is necessary when dynam yamic anal sis is performed. par

icto the designer of a nuclear steam s pp yearn su 1 system s t e eh d termination of the dynam ct loo s stem under sign can sifi t seismic loads and/orresponse of the reactor coolant loop y

of one of the pipesthe dynamic effect produced by a postulated sudden rupture of e of these components can(b1cwdown accident) . Zn.both cases the response o some o

b hi h damping is introduced«tain relatively large magnitud s.udes. The manner y w c artant consequences on t e a eqh de uacy of the results.4 the analysis will have important q

the analysis of a «reactorT establish and recommend damp gin values for use n e*

0u e data obtained exper men ai t lly. Data available~oo>ant loop it is necessary to use a i tal work''as been performedthis t e are scarce. Exper men a

h i b i fo ddetermine the dynamic characteriseristics of the pr mary sy(EGCR) in Oak Ridge ixphe E erimental Gas-Cooled Reactor

t"e San Onofre PWR( ~ 4~ ), and the Indian PointMicM an(>) and the EGCRar Power Plant, Monroe, c gan

Plant(2i 8). Most of this wor asrk has been directed towar s excibl b cause there is evidence~1th the largest possible forces.e . This is desirable ecause

h 1 rge forces and displace-t"at structures behave differentlyt when sub)ected to t e ad i e rupture loads, thanagents caused by strong-motion earthqh uakes or postulate p pe

<uring small vibration.he San Onofre reactor oh f tor loop components from the

970 d h F do h uake(9 10)f February 9, 1971, have been published > . These stu

n d~mping characteristics during aactual earthquakes.

609

The present paper ~discusses the type of damping to be adopted fana

or the dynan

and tlysis of reactor cooEant loop systems the methods used to me do measure amping,

an the recommended va9ues to be used in order to obtain conservati ve antically acceptable responses. Effects of response amplitude d ion amp ng valuesto interpret the resul't of tests performed on some plants are also discussed

DYHAHIC ANAL'YSIS AND DA1'PING CHARACTERISTICS

Several causes produce energy dissipation during vibration. Enerp s a result of internal friction within the material are called iroduced a

n. nergy losses

dam in dca e nternal

nnec ons etweanp g, and energy diwsxpation due,.to,irelative movements at connecti belements, joints, and interfaces is "called external dampin Thg ese types togeth~~are usually called structural damping, because they represent the sum of all lpaa~ao a vibrating structure in air. Losses of other origins are t ia e present n reactorcoolant loops. When structures, vibrate in a fluid, or with h draulic da

p' 's called viscous damping.'coustic radiation damping produced bythe energy dissipation Zm the surrounding air is also viscous. Dam i b imof xntermittent contact between closely positioned vibrating parts of the systemis another factor contrihuting .to the losses. The dynamic analysis u t id

yp o non-conservative system in order to provide reliable results. Therelation between the input energy and the energy dissipated from all sources willcontrol the response amplitude,

The dynamic analys'is of the reactor coolant loop is performed using the follavL~qmathematical approach. Applying a finite element and/or lumped mass technique,a three-dimensional model is built consisting of masses interco n t d ith

s mo e s then subjected to dynamic loads in case of. blowdown or accelerationtime history at the supports in case of seismic excitation. With this mathematicalapproach, as in the case of the inertia forces, damping is considered actin indiscrete points. It is mcognized, however, that, in practice, the damping inreal systems is distributed.

For blowdown cases., where the system is excited by time-varying hydraulicforces, the equations o'f mtion can be written as:

~ 0

MU + KU + D ~ F(t)where M is the mass matrix, K is the stiffness matrix, 'D is the damping forcevector, U is the displacement vector, and F(t) is the column vector of equivalentexternal forces.

Equation (1) requires the adoption of adequate boundary conditions which app>lconstraints at certain Locations of the system.

For seismic excitation, the vector F(t) is zero, and the displacements U a<aforced to vary in a defined manner (prescribed functions of time) at the

supporta'he

differential equations represented by. equation (1) are coupled by the off-diagonal terms of M and ~X.

If the damping vector, D, of the system is defined, the system of different~a~equations (1) can be integrated numerically with a digital computer to obtain t»atime-history response of the system. With adequate consideration, this procedu<allows the inclusion into the analysis of non-linearities such as ga s and impactbetween corn onent nponent and supports - - and permits the treatment of the matrices> Mi(ll) -'

ps

D, and K, as variables with time.

610

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< ~

/

k

\This procedure of time integration is usually performed for blowdown analysis

because of the complexity of the forcing functions, which depend on the break locationassumed, and the geometric non-linearities of the system which cannot be readilytreated by other mathematical methods.

For linear equations of motion, other mathematical methods, such as modalanalysis, are generally used to obtain the response of the system. Besides theobvious requirement for the matrix M and K to be constant, the form of linearitythat is required will only „be preserved .if during harmonic motions the dampingforces themselves vary harmonically. If viscous damping is assumed, D~CO, thenthe connection betwden messes is represented by dempers producing forces propertional to veloci.ty acting in parallel with the. springs. It is known 1 p that(12 13

if -it is assumed that C~2AK {where X is a constant) then the system of differentialequations can be uncoupled and <n ~ Xo>n is the critical damping ratio of eachmode with natural frequency mn. By assuming that C~2BM, the system can be also

Buncoupled and a critical damping ratio is given by gn ~ —. In'bo'th cases,,the damping of the system can be established with only one constant.

Finally, the damping matrix can be assumed to be proportional to criticaldamping, i.e. the coefficients of the matrix are given as a percent damping

'nfor each mode. If C ~ 2BM,. then <n ~ —,where Bn are the elements of the diagonal"n

B matrix and the values gn are the modal damping coefficients adopted in the,analysis (13) .

With this assumption, linear analysis of the structure can be performed bysolving an eigenvalue problem to determine natural frequencies and normal modes.The response of the system can be then obtained by modal superposition using the

'available techniques of linear analysis such as Duhamel integrals and responsespectrum. The adoption of viscous damping is useful to perform modal resp'onseanalysis and also when time integration is performed, because most of the existingcomputer codes for dynamic analysis of this magnitude use viscous elements.

Damping forces encountered in the vibrating system are not necessarily linear~ functions of velocity or displacement. From the analytical point of view it is

very useful to replace the damping forces by a equivalent viscous damping causingthe same amount of energy dissipation. Structural Pamping due to {a) internalfriction within the materials, and {b) losses at the connection between elementsmay be prepresented by a force proportional and in quadrature to the elastic force.This assumption is based on the hypothesis that the energy loss per cycle duringharmonic. motion is independent of the vibration frequency and,is proportional tothe square of the vibration amplitude. Physical damping representing the energylosses present during vibration in a fluid, shock absorber and hydraulic dashpotsis proportional to velocity. The energy loss per cycle of motion is also proportionalto the square of the amplitude. For harmonic motions, therefore, structural damping

.can be treated as viscous with an equivalent viscous damping coefficient.Material damping qr losses due to internal friction of solids has been extensively

studied by Lazan{ "p >, and='nformation has been published on the correlationof the magnitude of damping with the stress amplitude and the stress history. Themethod is based on the energy loss per cycle of vibration and clearly indicatesthat the losses will increase with the stress amplitude. For higher stresses,energy dissipation by plastic action is a major factor of damping. The stress-strain hysteresis loop is the same for static and dynamic loads, thus, independent

611

0

l p'

of frequency. Specific values of material damping are available for differentstress levels, stress distribution and geometries. For specimen with no-non-un formstress distribution, the measured losses must be considered adequatelmanner to obtain the material damping(16).

a e y n an averqz4gy

looAll these types of dissipation forces are present when the reactor 1oz coo ant

oop is excited by seismic and blowdown forces. When the system is representedy a three-dimensional complex model'of springs and masses, these losses cannot

be accounted for individually at all the locations where it occurs. Componentsare modeled by elements which may be limited in number by the capacity of computerprograms and the model must lump the losses at certain locations. There will besome areas where the internal friction within the material will be predominant(higher stress locations). In another case, sliding between connecting elementswill be important and certainly' large amount of internals compone t i hvessels and steam generator will be excited. Some will'mpact and/or slide withineach other; dissipating energy. Losses produced by Coulomb damping (dry friction)where the damping force is practically constant can not be described usin lie ua 'o

q ti ns. Pipe insulatxon has been reported to increase the pipe dampin ratio(1?),The water circulating inside the pipes may also contribute to the losses(~) duringvibration. The resulting effects will not have, in'eneral, viscous characteristica,particularly when the response of the components's relatively large.

DAMPING MEASUREMENT METHODS

Various methods have been used to measure damping of reactor coolant loops.The accuracy varies with the conditions and type of excitation. The results ofthese measurements are .the basic sources of information(18 available to inputinto the analysis. These methods are:1) A sinusoidal force is applied as a source of excitation and the responseacceleration is measured throughout the loop. By changing the frequency of excitatioa,maximum accelerations (and displacements) can be obtained at the natural frequencyof the system. Damping is then obtained by assuming viscous characteristics andmeasuzing the band width of the peaks at half power points of the peak amplificationusing the expressions

aaha) 12m . 2Q

(2)

whereha> ~ band width of the displacement response curve at the half-power pointun ~ resonance frequency

~ percent: of crieical dampingQ ~ magnification factor at resonance.The method of measuring the band width of the peaks was used by VCLA to

obtain damping in the Gas~oled Reactor(19) and the Sari.Onofre Reactor(<) whichwere excited by sinusoidal forces applied to the building and operating floor,respectively.2)) The system is excited at natural frequency by half "beats", and the free decayinSvibration is measured. To obtain the natural frequencies of the system, a preliminary

ow-amplitude oscillatory load is applied at various locations (top and bottomsteam generator and pump, end pipes) during a frequency sweep. In the neighborhood

612

r

f each natural frequency, sevexal half "beats" are applied to the component andthe response amplitudes are recorded. The force is suddenly cut off at its peaklit de., With this method adequate build-up of response is obtained, and the

(20)~plification of the 'system is restricted to a short period of time . After<he force is interrupted, the vibration decays and damping is calculated by measuring~Mo peaks, n cycles apart, with the expression

( ~ —Rn(—)1 12' n

'erich is reasonably valid for values of g < 0.2. (6)'~his";method was used in measurements of damping for the indian Point PWR

3) The snapback method which consists in applying an initial static deflectionto the system and suddenly releasing the external force producing the deflection.The system tends to return to the equilibrium posit'ion with a decaying amplitude,gnd damping is obtained in a similar manner as for the beat method, by recordigg

(1, 17)the decaying amplitude of the accelerometers installed on the components4) The system is excited by explosions as in the case of the Experimental Gas»C oled xeactor. Underground explosions will excite the system and the rate ofdecay measurements or the use of the half power point width of the Fourier Modulus

OO ~

(1)can be used to compute the damping.5) The response of the system is measured, using "ambient" vibration techniques,when the system is operating with pumps running and when they are still. A setof conveniently distributed accelexometers provides information regarding the systemres onse. The test records so obtained contain naxrow band random response atespstructural natural frequencies and, when the pumps operate, nearly sinusoidal respons eat pump-speed related frequencies and their harmonics. The random response ischaracterized through the use of statistical theory. The power spectral densityfunction (PSDF) is obtained by feeding the random response U(T) into a spectrumanalyzex'nd will contain a number of resonance peaks corresponding to the systemnatural frequencies. The modal displacement and damping represented by each ofthese peaks can be evaluated by treating the individual peak as a response of asingle degree of freedom oscillator. Damping values may be obtained from the resonancepeak by the bandwidth method. An alternative method is to calculate the autocorrelationfunction of the response based on the center frequency of the peaks and a suitablenarrow bandwidth. The autocorrelation function essentially describes the dampingduring free vibxation of the system. Damping can be obtained by the logarithmicdecrement method ( ). This method was used to estimate damping on the Indian Pointd2 plant(6). The same basic method can be used by measuring the component xesponsewith etta hed accelexometers when a seismic excitation occurs at the plant site.c

heDampin has been estimated for the San Onofre Nuclear Generating Station for t eSan Fernando Earthquake using the Fast Fourier-transform( and the PSDF

g af9) (18)

The comparison of the results provided by several or by all of these methodsmust take into account that the system is non-linear. For the "ambient vibration"technique, or the seismic disturbance measurements, the input to the componentcan be assumed to be a white noise or a bandlimited white noise. The K3S responsecan be obtained by adopting fox damping the initial slope of the autocorrelationfunction, because using averages or the bandwidth of the PSDF would provide somewhatlower damping values (6, 22),

613

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In Pigures 2 and 3, damping values obtained from ambient vibration tests andamplitude decay tests are plotted as a function of modal vibration amplitude. Forcomparison purposes, in Figure 5, the vibration amplitude corresponding to thedamping values obtained from the ambient vibration, is assumed to be 1.4a, whereo is the RHS displacement.

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SHAKER TESTS (150OF)

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0.00.1 1.0 10.0 100.0

DEFLECTION IIN X 10 ~)

llFig. 2 — Steam Generator Damping Ualues, Radial Nodes (From Ref. 6)

7.0

6.0

«c 5.0I

4.0z

g 3.0

> 2.0Cl

1.0

LEGEND:4.5 Hz TANGENTIAL) SHAKER TESTS

0 5.3 Hz RADIAL f (150'F)~ 4.0-4.9 Hz, PREOPERATION TESTS

{270-530'F) {3zJ )

8 4.0-4.9 Hz PREOPERATION TESTS1270 530'F) {IzJ )

rrAA

', 0.1 1.0 'IO.O

DEFLECTION (IN X 10'100.0

Fig. 3 - Reactor Coolant Pump Damping Ualues (Prom Ref. 6)

614

RESULTS OF MEASUREMENTS AND RECOMMENDED DAMPING

Measurements of damping in the reactor coolant loop have been reported ingo literature for various plants. The reactor coolant loop of these plants has,

general, different characteristics, but in all cases their loops consist essentially0f the same type of components: vessel, pumps, steam generator, and pipes. The

loop components are supported in various ways and this is reflected in the differenceog the natural frequency measured in the tests, and also in the corresponding damping

@glues obtained'he difference exists not only between each plant that has been

reseed, but also for the different loops ~3~ on the same plant. Tlus ds because

tha dynamdc charactertstic of the system util be affected by a difference in theorientation, geometries, and the particular environment. Notwithstanding, thereis a common pattern of heavy equipment with complex internal structures, whichjnherently have an important capability of dissipating energy during vibration.

The modal response study of this type of system indicates that most of thegodesp and especially the lower modes, consist of the main vibrations of certaincomponents and, sometimes, the same component in two different directions. Theresults are identified by labels, such as "steam generator normal mode in the north/south direction". For that specific mode, the most important displacem'ent concernsthe steam generator in the north/south direction, but including other componentsdriven by the steam generator, particularly the connecting pipes, oscillatingsimultaneously. In that case, the total energy dissipated is a result of all thevibrating components. s

Figure 4 shows a plot of experimental and analytical results obtained fromthe test at the Indian Point Plant(6». Mhen coupling exists between components,special care is required in attributing damping to certain components(

In Figure 5, damping results from different sources have been plotted. Thereis a definite tendency 'for the damping to increase with deflection. The plottedpoints have been fitted with a curve. The data plotted has been selected fromthe test performed at the Indian Point Reactor (6~, the San Onofre generatingplant( p "p 5» and the Experimental Gas-Cooled Reactor(lp

Not all values reported in the literature are plotted because, in some cases ',

it was not possible to extract a clear relationship between damping and displacement.The other information not plotted tends to confirm the order of magnitude of thedamping given in Figure 5. Consequently, the following data shouM be consideredin addition to the data plotted in Figure 5.1) Steam Generator

a) Tests on the EGCR reactor indicate the following:"For small snapback and forced vibration tests, dampings are typically0.5-1.5% of critical, while at higher deflections, dampings of 3-7%

are in evidence" (p. 97, Ref. 8) ~ The steam generator frequencieswere in the range of 5.8 - 6.0 Hz and the maximum snapback test had

an amplitude of 0.075 inches.For blast tests "values of damping measured .at the top of the steamgenerator ranged from 1% to 3%, depending on the size of the chargeand the direction" (p. 95, Ref. 2).

b) Measurements taken on the San Onofre Plant during the Lytle Creekearthquake provide damping values between 3.4% <"> and 5%( 8). The

lowest value is plotted in Figure 5.

615

l J

I

2) Pressurizera) From the Lytle Creek earthquake records, the damping appears to be about

2% and 9% of critical( 0) with maximum response acceleration between0.04g and 0.08g.

3) PipingResults of piping damping between 3.8X and 5% have been reported for theIndian Point Plant(6). Measurements taken at the Tsuruga Nuclear Plantusing the snapback method provide dqmping ratios between 3.2% to 8.6X on16 in. and 14 in. diameter pipes( > ~

The maximum ground acceleration during the San Fernando earthquake was 0,018g,and, as reported in Reference 9 (p. 84), damping values for the steam generatorand pump are between 2% and 4%, with possible variations of plus or minus 1%. Thepressurizer damping was 2% with the same estimated variation. For comparison pur-poses, it should be mentioned that the lowest typical values of maximum seismicground acceleration presently used for nuclear plants are O.lg for OBE and 0.15gfor DBE.

REFERENCES : TEST 49SHIMS: OUT~ SHAKER POS.: SGTR. 46R

SCALE:O O IN

EXPERIMENTAL, 2.9 Hz——THEORETICAL, 2.92 HZ

Fig. 4 - Steam Generator 1st Bending Mode, Radial(Shims Out) (From Ref. 6)

616

e

to987

el

3cC

I—2 He

ci:j~ Q

'Q

I—

CL'.9

O.S

0.7

0.6

0.5

0.4

TiP.J

~ ~ ~ 1:

:J ~'pt JJj ~;

'

~ i ~ ~ ~ ~

' ~ iX

09

0.2

A IPP Shaker8 IPP ShakerC IPP ShakerD IPP ShakerE IPP ShakerF IPP ShakerG IPP ShakerH IPP ShakerI IPP Sha'ker

J IPP Shaker

T. 2.9 Ht St. Gen. RodentT. 6.5 Hz Sl. Gen. Red~atT. 2.48 Ht St. Gen. Rad~st

T. 3.15 Hz Sl. Gen. Rad«tT. 4.5 Ht Pump Tan<atT. 5.3 Ht Pump Radio>

T. 2.6 Ht St. Gen. Tan<at

T. 2.43 Ht St. Gen. Tan«tT. 3.15 Hz Sl. Gen. TanllT. 6.9 Hz St. Gen. Tan+I

K . EGCR Shaker T. 5.92 Hz 8 5.82 Hz St. Gen. EW 8L NSttetl EGCR Blast T. 5.8 Hz St. Gen. NSOl

M SONGS Shaker T. 5.87 Ht Ik 3.18 Hz Pump NS 8L EWlslN SONGS Shaker T; 2.87 Ht Pressurizer NS«t0 SONGS Shaker T. 1.87 Hz St. Gen. EWl~ lP SONGS Litle Creek Earth. 2.0 Hz St. Gen. EWttt'

SONGS S. Fernando Earth. 1.9 Hz 6. 2.85 Hz St. Gen. Tan 8L RadttetR. IPP Preop. T. 2.4 Hz 6 3.25 Ht St. Gen Rad 11.40)tatS'. IPP Preop. T. 2.4 Hz Ik 3.25 Hz St. Gen. Tan II.4u)tat

. ~ T . IPP Preop. T. 4.0 Hz 8 4.9 Hz Pump I).4u)<at

0.1

.001

DEFLECTION (IN)0.01

Fig. 5 — Percent of Critical Damping for Reactor Coolant Loop ComponentsDeflection (in)

The plots presented in,Figures 2, 3 and 5 could be questioned since they includedifferent component frequencies and components with different geometries. Theintent is to show a trerA and provide a basis such that, for response levels inthe range of interest, ~ping ratio values can be conservatively adopted. Porthese reasons, the best fit line in Pigure 5 was traced. This plot indicates notonly that the measured damping values can be fitted with a curve, but also that,despite expected scattering, they fall inside a characteristic band which covers allthe various test, methods and components tested.

For a better understanding of the appLication of the modal damping ratiosto the seismic and blowdown dynamic analysis, a discussion of the system behaviorunder these loads is presented. When the reactor coolant loop is subjected toseismic and/or blowdown loads, the main response is characterized by the vibrationof the heavy components, vessel, pumps, pressurizer, and steam generator in theirlover modes. The contribution of higher modes of these components is insignificantcompared with the former, i.e., the steam generator will respond mainly in modesrelated to the lower frequencies of cantilever shapes with components in the directionof these low modes (see Pigure 4). These were mainly the modes excited by theSan Fernando earthquake on the San Onofre plant and most of the modes obtainedduring the Indian Point tests. The results in Figure 5 correspond to these lowermodes. Figure"5 also shows that the main displacement of the components for a4% damping corresponds to a maximum displacement of the order of .060 inches, whichis well below the disp'Xacements calculated for blowdown and seismic loads. Highermodes with very small Cisplacements are of no interest because of their small contri»bution to the total deflection. They will not be underestimated because they arerelated to larger energy losses than the ones indicated in Figure 5, as it isusual for real structures. The higher mode of the steam generator excited duringthe Indian Point II .tests show much higher damping, as indicated in Figure 2 ~It should also be recognized that for larger amplitudes the characteristicsof the nonlinearities will change when compared with the nonlinearity present atlow amplitude. Pipes, pumps, and steam generators in some plants, have hydraulicsnubbers and/or bumpers that will act only when large dynamic oscillations occur.In such cases, the nonlinear effects will be enhanced. The proportional contributionof small gaps and dry friction effects on the total energy losses during largevibration amplitudes vill change in a manner difficult to predict. Por large deforma-tions, the effect of stress .cycles will increase the energy losses; these typeof losses were negligible during the tests plotted in Figure 5.

Dynamic elastic analysis, performed for various PMR primary loops with 0.5%and 1% critical damping, indicates that maximum deflection of the steam generatorwill be about 0.5 inches under blowdown loads, above 2 inches for high level DBE's,and around 0.2 inches for ~the „lowest defined DBE. The maximum pump deflectionwill be about 0.5 inch for blowdown and ranges between 0.30 and 0.70 inches forDBE, depending on DBE intensity and support conditions. These figures indicatethat 4% modal damping ratio for the design of reactor coolant loop under theseloads will provide aa ample margin of safety and ensure a conservative design-It is acknowledged that due to the conservatism of this proposed damping, the ana>VN<~

will provide responses much larger than the real ones. Purther test work wouldbe desirable to measure damping for higher response so that less conservative appr«~~can be incorporated into the analysis, by considering models which more accuratelydepict the non-linear behavior of the structure experiencing large deflectionsand stresses <

618

Furthermore, for the combined blowdown and DBE, adoption of higher dampingyagues can be g ust ified by considering the unl ikely probabi 1 ity of simul taneousdecurrence of the peaks in time and place. Values of 4% to 5% of modal dampingratio are recommended, at this time, for this combined event because, in absenceof data for large deflections, any extrapolations beyond measured values basedon )udgment can not be proposed in view'f the severity of the assumed accident.

The adoption of one value for modal critical damping ratio means that g*

~ constant, and, consequently, the appropriate 9 values must be introduced ninto analysis depending on natural frequencies.

Once the damping ratio for the DBE is adopted, the recommendation for thepBE depends on the degree of margin of safety that the designer is willing toincorporate, for a transient that has a higher probability of occurrence than thpBE, In t at respect, it should be kept in mind that for this event the allowableh

an e

stresses given by the AQfE Section III Code and the permitted deflections are much>ower. Consequently, the safety of the plant will be completely assured by adoptinga fraction of the DBE damping for these cases and simultaneously requiring compliancewith the Code limits. For these reasons, the modal damping ratio of 2% is proposedfor the design of the reactor coolant loop under Upset conditions.

This somewhat arbitrary yet conservative approach could, in particular cases,lead to seemingly unrealistic analytical results, e.g. a larger displacement responsefor OBE than for DBE. This result should not be considered as an anomaly of theproposed damping ratio, but a consequence of the arbitrary margins of safety adoptedfor both events. With the intent of increasing the margin of safety, when thesmall earthquake occurs, the damping adopted is so low that it produces for designpurposes, a larger response than that of the DBE (which is assigned a more realisticdamping).

hihseWhen similar plants are built in high and low seismic areas the OBE f th

g seismic, plant could have the same intensity as the DBE of the low seismic0 e

site. In this case both plants will be designed with the same seismic input butfor different seismic responses because of the damping value adopted. The allowableatresses, however, will not be the same.

Designing a plant for very small seismic excitation input using the proposeddamping values is of no concern, even though the response will be below the deflectionwhere damping is less than 2/, as shown in Figures 2, 3 and 5. In such a case

flections and stresses are so small that they will be inconsequential foratructural adequacy considerations. Any increase of response vill automaticallyincrease the true physical damping and maintain the system within acceptable limits.

CONCLUSIONS

From the study of the test results o'btained from different sources the followinmodal d1 damping rat@os are proposed for analysis of Reactor Coolant Loop Systems:For Upset Condition, 2/; For Faulted Condition, 4%.

r

The following comments support the adequacy and summarize the basis of theapproach adopted.1,~ he test results reported and used as a basis for this study have been concernedwithh such small amplitude vibrations that the material or internal damping wasaegligible. Lov stress (well below yield) was the main characteristic of these<<sts. Under the DBE and/or DBA loads, the stresses in the components are much

619

4

higher, and, consequently, the material damping will become an additi 1a onal factorto be considered xn computing the total energy dissipation. The allowable stresare permitted to be well above yield for Faulted Condition (which includes DBEand/or DBA) per Section III of the AQ1E Code.

order2 ~ The possibility of adopting higher damping values sho ld bs ou e considered inor er to obtain more realistic responses by including the dcorrespon ing non-linea<

is known that for cases with damping proportional to the square

hthe velocity, for example, the transmissibility f tho e system s reduced for botigh and low frequencies compared with the equival t i dva en v scous amping case.

o )ec ve o improving the modeling of the damping force could be achievedtesting at large amplitudes.

e ac eved by adequate

3. The use of adequately conservative damping values for design is of s eciainterest for plants located in high seismics mic areas. n er the San Fernando earquake with a maximum ground acceleration of 0.018 '(9). h

himeasured steam generator damping was about, 3/ (Fi 5)

g .at t e site, the corresgure q. A typical DBE for a

ponding

gh sexsmic area is 0.4g which is an order of ma nitude laearthquake.

magn u e arger than the San Fernandos'

REFERENCES

2.

C. B. Smith and R. B. Matthiesen, "Vibration Testing and Earthquake Res onseof Nuclear R'eactors," Nuc. A l. Tech. Vol 7 7 1 J 1 , 6, go., u y, 6 — 34, (1969),

J. Chrostowrostowski, et al., Simulating Strong Motion Earthquake Effects on19

February, 1972.Nuclear Power Plants Using Explosive Blasts " UCLA-34P 193-10 ; UCLA-ENG-7119,

gr3. P. Ibanez, R. B. Matthiesen, C. B. Smith, G. S. C. Mang, "San Onofre Nuclear

Generating Station Vibration Tests " UCLA Sebo 1 f EUCLA-ENG-703?,, August, 1970.

c oo o ngineering, Report

4. RGenerati

. B. Matthiesen, P. Ibariezg L. G. Seine, C. B. Smith "S Onan ofre Nuclearenerating Station Supplementary Vibration Tests," UCLA School of Engineer-

ing, Report UCLA-ENG-7095, December, 1970.

5. R. D. Shanmahanman, P. Ibanez, R. Vasudevau, R. B. Matthiesen, and C. B. Smith," an o re Nuclear Generating Station Series IIIVibration Test," UCLA,School of Engineeiingp UCLA-ENG-7220, February, 1972.

~ 6. B. E. Olsen, N. R. Singleton, and G. J. Bohm, "Indian Point No. 2 PrimaryLoop Vibration Test program," VCAP-7920, Vestingbonse gleotrio CorporationgPittsburgh, Pennsylvania, September, 1972, or "Deterud.nation of theDynamic Characteristics of a Pressurized Mater Reactor Primary Loop" tobe presented by the same authors at the Second International Conferenceon Pressure Vessel Technology, October 1 - 49 1973 'n San Antonio, Texase

7. B. Bleiweis G. Har. Hart, and C. B. Smith, Enrico Fermi Nuclear Power Plant99

(1970).'Dynamic Response During Blasting," Trans. Amer. Nucl. Soc. 13 231 - 232~ ~ ~ ~

620

8. The EGCR Blast Test Program Group, "Blast" (Interim Report), UCLA, Schoolof Engineering, Report 8UCLA-ENG-7081, October, 1970.

9.

10.

R. Vasudevau, P. Ibanez, R. D. Sh'anman, R. B. Matthiesen, and C. B. Smith,"Modeling of the San Onofre Nuclear Generating Station for Seismic Response,"UCLA School of Engineering and Applied Science, UCLA-ENG-7251, July, 1972.

C. B. Smith, P. Ibanez, R. B. Matthiesen, and R. Vasadevau, "Response ofthe San Onofre Nuclear Generating Station to Earthquakes," UCLA, Schoolof Engineering, UCLA-ENG-7151, July, 1971.

G. J. Bohm and A. N. Nahavandi, "Dynamic Analysis of Reactor InternalStructures with Impact Between Components," Nuc. Sci. ~En .: 47, 391 —408,(1972). „

12. J. S. Przemieniecki, "Theory of Matrix Structural Analysis," McGraw-Hill,New .York, 1968.

13.

14.

M. F. Rubinstein and W. C. Hurty, "Dynamics of Structures," Prentice-Hill,Inc., New Jersey, 1964.

B. J. Lazan and L. E. Goodman, "Material and Interface Damping," Chapter 36,"Shock and Vibration Handbook," Vol. 2, by C. M. Harris and C. L. Crede,McGraw-Hill, New York, 1961.

15. B. J. Lazan, "Damping of Materials and Members in Structural Mechanics,"Pergamon Press, New York, 1968.

16. E. R. Podnieks,and B. J. Lazan, "Analytical Methods for DeterminingSpecific Damping Energy Considering Stress Distribution," WADC TechnicalReport 56-44, June, 1957.

17. K. Akino, M. Kato,Power Plant -'artJ. At. Ener Soc.Japanese). Ni on

and S. Tamura, "Aseismic Design in the Tsuruga NuclearIII, Design of Reactor Internals, Equipment and Piping,"Ja an, Vol. 13, No. 3, pp. 139 — 146, (1971) (inGenshir oku Gakkaishi.

18. A. Morrone, "Damping Values of Nuclear Power Plants Components," WCAP-7921,November, 1972.

19.- C. B. Smith and R. B. Matthiesen, "Forced Vibration Tests of the Exper-imental Gas-Cooled Reactor (EGCR), "UCLA Report No. 69-42, August, 1969.

20. E. G. Fischer, "Sine Beat Vibration Testing Related to Earthquake ResponseSpectra," The Shock and Vibration Bulletin,'ulletin 42, Part 2,1 « 9 January, 1972.

21. S. Cherry and A. G. Brady, "Determination of Structural Dynamic Propertiesby Statistical. Analysis of Random Vibrations," in "New Zealand WorldConference on Earthquake Engineering 3rd Proceedings," pp. 50 — 67, 1965.

7 44

A'

22. H. A. Cole, Jr., "On-I,ine Analysis of Random Vibrati "A1AA/0Ils ~ /ASM89th Stkructures, Structural Dynamics and Materials Co fon erence, palm.prings, California April 1 - 3, 1968, AIAA Paper No. 68-288,

23. USAEC Socket 50-361-24, "San Onofre Nuclear Gener ti S

2enera ng tation Units

and 3. Methods of Direct Application of Element Dampin ." BeCorp., January, 1972.

n amp ng. Bechte~

,—~

4

4/ 622

SER CONCLUSIONS 22. 0, ITEM (3) (CONTINUED)


(2) "Submittal of additional information about the sensitivity of non-linear1

systems in the reactor coolant system (parts of the reactor internals and

control rod drive mechanisms) to the synthesized time history motions used foranalysis and the significance'f the time history sel'ected."

Response: ., The staff has requested information concerning the effect ofsynthesized time-histories on non-linear system analysis. Several points are

important in the response to this item.

l. It is not valid to consider variations of time-histories at the floorelevation where equipment is excited. Any variation in the

time-history must be applied at the ground and amplified through the

building. Zt is expected that the building provides an effectivefilter such that the character of the time-history seen by thenon-linear system would be similar for variations in the time-historyintroduced at the base mat.

2. Two items were analyzed with non-linear time-history analysis forDiablo Canyon. The first was the CRDM's; the stresses in the CRDM's

were 65X of the allowable stress. Any reasonable change in stresswill not cause the conclusion of acceptability to change.

The second item analyzed by the non-linear time-history analysismethod was the fuel. The grids have been postulated to deform a

maximum credible amount in the analysis and the coolability of the

fuel has been demonstrated to be acceptable. This eoolable geometry

.analysis, as well as the conservative way in which the loads were

obtained, indicate that reasonable variations in the time-historyinput at the ground and filtered to the fuel will not change the

conclusion of the safety of the fuel. The question raised will notaffect the conclusion of the adequacy for Diablo Canyon as described.

D/C SER Supp. 7

~ R



(4) "Submittal of an assessment of the differences, if any, in the reactorcoolant system main loop analysis for Unit 2 as opposed to the Unit 1 analysisthat was submitted for our review."

Response: Diablo Canyon Unit 2 contains certain design differences in the

reactor vessel internals in comparison to Unit 1. The Unit 2 vessel has the

neutron pad design between the core barrel and,vessel shell, whereas Unit 1

has a thermal shield. There are also minor differences in the configurationof the upper and lower support plates. These differences will introduce a

variation in the transient internal hydraulics forces for the RPV LOCA

analysis between the two units.

h

The effect of this variation has been evaluated for Unit 2 in a manner

identical to the analysis performed for Unit 1 as described in WCAP 9241. Allinputs to the analysis were the same with the exception of the RPV structuralmode3. which reflected the above~oted differences and the applied internalhydraulic forces. The maximum RPV displacements and support loads were

essentially the same for the two units. No significant changes were found inthe response of the reactor coolant system to the vessel motion for Unit 2.

Thus the differences in the internals between the two units has no

significance in the applicability of Unit 1 results to Unit 2.

D/C SER Supp. 7

c

SER CONCLUSIONS 22 ~ 0, ITEM (3) (CONTINUED)


(5) "Submittal of additional information about support capacities for reactorcoolant system branch piping in order to assess the significance to those

supports of using the RG 1.92 method of combining responses."

Response: The results of the Criteria Implementation Review Meeting (received2/13/78) included "Enclosure 2 items 6. —LOCA plus Hosgri of RCL Branch

'ines." The- instruction in part was ".....the results presented should

include.....support loads versus design loads, or support ratings....."

The Applicant has requested Westinghouse to run two typical branch lines using

Reg. Guide 1.92 methods and present those loads to PGandE for support evalua-

tion. Any supports with large load increases will be reviewed by consideringthe actual support capacities. It is not expected that any problems will,arise requiring any design changes 'to 'the supports.

D/C SER Supp. 7

Ig ~ I



(6) "Submittal of a study to assess the effects on other piping of using the

RG 1.92 method of combining responses, including consideration of closelyspaced modes."

(7) "Submittal of a study to assess the combined effects on other piping ofactual support stiffnesses, seismic anchor movements, and response

combinations."

Response to (6) and (7): The applicant will conduct a sensitivity study to

compare the analysis techniques as described in the FSAR Adm. 50 and subse-

quent amendments with a series of representative piping systems analyzed with

Reg. Guide 1.92, closely spaced frequencies, support stiffnesses', and

available differential seismic movements. This evaluation will 'compare

support loads and piping stresses.

We expect that the results will be as shown in all of the other sensitivitystudies to be within the load capabilities of the supports, and stress

allowables of the piping. Presentation of the data to the Staff will be made

directly.

The Staff has requested an assessment of the combined effects on piping "of

actual support stiffnesses, seismic anchor movements and response

combinations." The Applicant has submitted, on March 17, 1978, additionalcopies of sensitivity reports entitled "Seismic Anchor Motions", "Piping

Stresses and Support Reactions due to 2-Directional vs. 3-Directional Methods

of Analysis", and "Effect of Support Stiffnesses on the Pipe Stresses and

Support Reactions." The conclusions of these studies were that the pipingstress was less than the design allowables, and piping support reactions were

acceptable. The Staff desires an evaluation of the combination of these

effects, and the recently added consideration for closely spaced modes.

PGandE is preparing this information; the results are expected to be available

during July 1978, and to indicate acceptability of the systems.

D/C SER Supp. 7

ig ~ )

4

~ ~ 1

p

SER CONCLUSIONS 22 'j ITEM (3) (CONTINUED)


(8) "Submittal of information about the upgrading of supporting for Design

Class 2 piping as necessary to ensure that Design Class 1 piping will performits intended safety function and to ensure the systems relied upon for coldshutdown will perform their functions."

Response: The supports of certain Class 2 piping which could be used as

alternate sources of auxiliary feedwater are being reviewed. The techniquesused in the support review are the same as those used to design Class I pipesupports. Hanger designs will preclude seismic loads from piping which has

not been analyzed seismically from contributing to the loads being analysedfor the subject Class 2 piping. All necessary hangers to beyond an isolationvalve will be included within the design upgrade, such that these additionalpiping systems will resemble all other design Class 'i piping/supplementalsystems. Design Class l to design Class 2" interfaces are accomplished. atclosed valves.

D/C SER Supp. 7

't ~ 7

'k I



(9) "Completion of our evaluation of the applicant's piping snubber study."

Response: The SER states that the staff has yet to complete their evaluation

of the Applicant's piping snubber study ("Stress Evaluation of Piping Systems

Assuming Single Snubber Failure", which was submitted with Amendment 59 inFebruary 1978). The submittal considered the failure of the snubber to lock

up during a seismic event and the failure of the snubbers to be free during

thermal expansion. The probability of attaining a pipe break'n a seismic

event due to snubber malfunction was demonstrated in the submittal to be

extremely small. The resulting piping stresses are within acceptable limits.Additionally, upon examination of the piping support loads and design

characteristics, subsequent failure of other piping supports is not expected.

The Applicant feels that due to the unlikely occurrence of these postulated

events, the subsequent acceptability of the piping systems, and the amount of

information provided to the Staff, this open item does not affect the overallsafety and acceptability of the plant. The Applicant considers that this item

should not delay the licensing process.

D/C SER Supp, 7

g rp

I I

SER CONCLUSIONS 22 ', ITEM (3) (CONTINUED)


(10) "Submittal of information concerning the results of requalification by

, testing (Section 3.9.3.7)."

Response: The staff has been provided with copies of test reports for the

boric acid tank, liquid hold-up tank, C02 tank, component cooling water heat

exchanger, component cooling water surge tank, six piping systems and tenhangers, and three valves on line. Still 'outstanding is an addendum to the

component cooling water heat exchanger test report which will be submitted tothe staff by. the end of July 1978. Upon submittal to the staff of thisaddendum, this open item shall be closed.

D/C SER Supp. 7

<'P ~

,) 4, r)



(ll) "Submittal of additional information regarding the in-situ testingprogram (Section 3.9.3.7)."

Response: The applicant elected to perform in-situ testing, which was not a

requirement of the staff (see section 3.9.3.2, para 5) or of any current

regulatory guidelines. Thus, it should not be the subject of an open issue

in the SER. The subject of in-situ testing programs were initiated by PGandE

and'Westinghouse to verify analytical techniques, obtain information foradditional evaluations, to support completed evaluations, or to demonstrate

the capability of components to perform their required functions while

subjected to Hosgri loadings. We felt this approach desirable in view ofthe magnitudes of the postulated Hosgri loads and decreases in margins from

those typically encountered for lower seismic

evaluations.'Nevertheless,

we are currently completing the related reports and will submit

them to the staff for their information and review. Since this testing is not

required, submittal of these reports for staff review is not a standard

practice and is not essential to their review. It should not be the source

for any further delay on this application and does not warrant identificationas an open issue in the SER.

D/C SER Supp. 7

0 C>P

j W Ig

~ L


Section 3.'9;3.9 (Continued)

(12) "Submittal of additional justification for the test inputs to be used in

testing a 14-inch motor operated gate valve (Section 3. 8.3.7)."

Response: Following extensive discussions with the staff, the test procedure

,,for the subject test was revised 'incorporating comments and suggestions

provided by the staff. The test response spectra included in the revised

procedure were generated in accordance with IEEE 382, April 1978, and have

been clearly shown to be more severe than those required for any of the valves

for which this test is intended. Copies of the revised test procedure were

submitted to the staff on 5/26/78.

D/C SER Supp. 7

l- t» w

4,)



(13) "Completion of our review of the'applicant's reports concerning quali-fication of Limitorque valve actuators and their applicability to Diablo

Canyon (Section 3.9.3.7)."

\

Response: The staff has been provided with copies of a Limitorque report1

covering the shaker table qualification of, Limitorque valve operators. This

report summarizes the qualification oP the Limitorque operators to the

requirements of IEEE — 344, 1975. All Diablo Canyon Limitorque operatorsidentified in Tables 7-7 and 7,-7A are in the generic range covered by these

tests. The staff has reviewed the Limitorque report and has accepted it forthe seismic accelerations listed in these tables.

D/C SER Supp. 7



(14) "Submittal of information concerning the material frequency of one valver

and concerning the qualification of several other valves (Section 3.8.3.7)."

Response: Th'ese valves 'are being qualified using essentially the same methods

and criteria as used for all other valves including those the staff reviewed

during the January 30, 1978, meeting. Extracts of appropriate sections of the

design report for valve HCV 637 were submitted to the staff on 4/10/78. A

complete copy of the design report, including detailed frequency calculations,was made available for the staff's review in San Francisco during the week of6/5/78.

Several design reports of the subject valves were reviewed with the staff in a

meeting on June 29, 1978. Two remaining design reports have been requested by

the staff and will be made available to them.

Summary results for these valves will be provided as noted in the response toItem 16.

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SER CONCLUSIONS 22 0$ ITEM (3) (CONTINUED)


(15) "Submittal of information concerning the ability of active valves to

operate during a seismic event as well as afterwards (Section 3.9.3.7)."

Response: The requested information was discussed with the staff in a meeting1

held June 29, 1978. The staff requested that a report be submitted describingthe in situ static deflection valve operability test and that certainadditional information be submitted concerning operability of the valves which

are the subject of Item 14 (of Section 3.9.3.9). This report is not yetcomplete. It and the requested information are being compiled at this time.

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SER CONCLUSIONS 22 ~ 0 j ITEM (3) (CONTINUED)


(16) "Submittal of updating amendments for Tables 7-5 through 7-8 of

Amendment 50 to the operating license application (Section 3.9.3.7)."

Response: The status of these tables at this time (June 1978) is as follows:

~ Table 7-5: All final data is available except for item 18, the

component cooling water heat exchangers. This table is being amended

to include this data.

Table 7-5A: This table is complete.

'able 7-6: This table is being amended to include final data on

items 7 and 16. Two remaining items, 31 and 41, are still being

analyzed.

Table 7-7, 7-7A, and 7-8, the valve tables, remain to be completed.

The data for Tables 7-7 and 7-7A are scheduled to be available during

August 1978, and the tables will be amended in September.

Table 7-8 will require more time.

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Section 3.9.3 ' (Continued)

(l7) "Submittal of information concerning modifications to reduce nozzle

loadings on the steam driven auxiliary feedwater pump (Section 3.9.3.8)."

Response: The nozzle loadings on the steam driven auxiliary feedwater pump

turbine have been reduced to acceptable values by adding a seismic restraintto the steam exhaust line. The reduced loads are within the guidelines

specified by the turbine manufacturer.

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