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{lATTACHMENT A.

ONITED STATES CF APEUCA -ila

|?'; NUCLEAR REGULATORY COMI1ISSION f| F..

II a|| BETCRE THE ATCMIC SAFETY AND LICENSING SCARD E'-

: E'f E!! E!! ;

in the Matter of g.;

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y! SOUTH CARCtINA ELECTRIC & GAS T1 r

CCMPANY and SOUTH CAROLINA PUBLIC ESERVICE AUTHORITY ) h

Docket No. 395-OL h(Virgil C. Sumer Nuclear | -@

Station, Unit 1) ) _he, .

EAFFIDAVIT OF LA'a'RENCE E. HOCHRE!TER y

||

| My na:ne is Lawrence E. Hochreiter, and my qualifications can be found in pAttachment A. I am an Advisory Engineer at 'iestinghouse Nuclear Energy .):<

- Systems. I have read the affidavit and testimony of Dr. Michio Xaku .$-| (Tr. 3573 - 3555, 3670 - 3764). Before discussing Dr. Xaku's affidavit $,!

f or testiaany, I would like to make some general statements: f' ..y; c.;

1.) 'd uses NRC approved licensing models and computer co' des gj!( for loss of Ccolant Accident (LOCA) analysis. The codes jI and models used for the design basis accident conform to @

fthe Appendix X criteria. The computer codes used to -

:ij analyze the design basis transient (LCCA) for V. C. Sumer kw

)tplant met the Accendix X requirements and were a: proved by .g '

-'[ the NRC. (Staff SI:R page 6-23).

Y i:Er 7.

'

2.) Dr. Kaku at several points claims that comouter codes have yc;

|| not been verified. The computer codes a<.1 codels used in 4)I the design basis LCCA transients have been verified j

[!' against test data from secarate effects tests such as jjj FLECHT, .and integral systems tests of different scale such ,

-

'i 8108120095 810807 as se'liscale and LCFT. The validaticns of the codes have Y; PDR ADOCK 05000395 AI c PDR been reviewed and a: proved by the '4RC. The Westinghouse j-

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ccecuter mdels have also been used for blind tact credic- [.tions in the .'iRC standar. problem program as another [s

ethod of cada verification. Oi

5 :

With specific references to Dr. Xatu's affidavit, I offer the following*obser ntions. I have used Dr. Kaku's Roman and Arabic numbering system f

for easf reference. rEs i&

. .c3Sect. !) Evacuation and Accident Hazards At the V. C. Sumer Plant y

:.if,

iPara. 2) As explained above, the LCCA analyses are performed using -

j: ,

evaluation mdels which satisfy ICCFR50 Appendix X and b.

which have been reviewed and accepted by the NRC. Tne

maximum acceptable temperature using such models was set .iby 10CFR50.46(b)(7) at 2200cF after extensive rulemaking h..

*

j .

I' hearings in which reliability of the computer codes as %

iL well as all other known uncertainties were considered. 4

xThis temoerature includes the margin of safety for such #

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| uncertainties found to be adequate in the ECCS hearings. . jf

} The use of these analyses to establish the peaking factor]s|

does not constitute retrofitting the data. Dr. Xaku 4.1

apparently does not understand that with respect to limit- g

| ing potential peak clad temerature folicwing a loss-of- $rei

) coolant accident operatico of a nuclear reactor is con- E

trolled by limiting the ceaking factor (peak divided by iI

I| average linear kilowatts / foot) in the ccre together with y

max!=um ocwer during normal operation and it is to these

ilmitations which ICCFR50.16(a)(1) refers when it states .jL

that confermance with the criteria -'.ay require that fi

restrictions te imposed on reactor operation. When one jperforts a LOCA calculation, the peaking factor is (,

increased until the calculated peak clad temperature :j! ecuals 2200cF for the maxinu . pcwer rating. This estab- j

lishes the maxit,2m acceptable peaking factor under ,

10CFR50.46. Tnis .aximum acceotable peaking f actcr 7A

ccrees::c, ding to 22CCc; and the ,admu, ccwer level >

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- are include'! in the technical specificatiens which become $j part of the operating license and with which tha operstar j

{]{ cust either cocoly ce shut down his facility. To the

?! extent that ;ening factors do not reach the maximum yacce table valve and calculated peak clad te.eperatures do f

j not reach 22C0cF there is cargin over and above the

!! minin;m safety urgin found to be acceptable in the exten- gsive ECCS hearinos. 5

i~

221

fiThe. design of the emergency core cooling systes incorpor-'

5ates suitable redundancy, interconnections, leak detec-'

xtion, isolatten and containment capabilities as required .-

5;y General Design Criteria 35 so tnat the ECCS safety . .:5

function is accccplished assuming any single failure. T.g.:

Single failure for this purpose means an occurence which jresults in the loss of capability of a cceponent to per- ijform its intended safety functica with respect to protec- gtion against a LOCA and includes multiple failures result- $i

cing from a single occurrence. The examle cited by p"Dr. Ksu of a pressurizer PORY failure transforming a hmajor primary cold leg break fran a relatively harmless j-

Class 8 accident into a more sericus Class 9 accident is ji

without merit. The consequences of a major cold leg bre d [wculd be unaffected by the failure of a pressurizer PORY. 8=

LThe additional ficw area resulting fecm such a failure r1

5would be miniscule in cocparison to the area of the double pended been of the 30 inch reactor coolant pipe. I

E

I disagree strongly with the statement indicating the kresults presented in Chapter 15 of the FSAR are based en {pure speculation. Specific accidents are analyzed to see [hcw they challenge the design an the safety systeo }response. Ccnservative assu-etions are cade on the [

$ response characteristics of the plant such that the design .

$ I is fully challenged. All these conservative ass'sptiens &2 .-

.f.are designed to F $e the CalCUIated dCCident IflosieSE -[

,,

eTore severe, and thus protide increasad s1fety targin.,

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k! j .'1I In sumary, I wculd expect anycne who claims to be N

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' . ' familiar with LOCA analysis to be aware of the matters N|q [

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j; covered in the ECCS Rule:taking 2.nd in 10CFR5u Appendix X. j.Tl

t: r:||j Para. Se.) As stated earlier the cmouter codes used for LOCA .h

ra

j analysis satisfy 10CFR50.46 and 10CFR50 Accendix X and g,i

j have been verified and approved by the NRC. With regard dj| to LOFT, LOFT experi=.ients are scaled tes's with known . N- s.

differences between the test and a contercial PWR. Whenj; the ifcensing codes are used to model LCFT, tha code ;g! calculations indicate mch higher ternperatures than do the 3i, :x:j, experi=ents indicating that margin exists in the code jj- calculations. Best estimate codes which mre accurately -@l mj model the two-phase ficw and do not use the conservative @[| model assumptions specified in Appendix X, compare much ff{

| more favorably with the LOFT data. Dr. Kaku apparently jj does not understand how test results are used in LG'A $ms

. fii analysis. -

s :-

I. jf.I Para.Sh.) Dr. Kaku apparently does not understand the calculated h

large' break accident. The eargency core cooling systra,

'

a is designed to recover the core after a large break. 5a

ei During the blowdown phase, emergency core cooling water is .Er:

J initially assumed to be swept cut the break with the RCS .g.

water. As blowdown continues and the system depres- j|g.-

j@.psurizes, the ECCS water is calculated to penetrate the

k"'downcorer, fill the lower plenum and initiate core r.

1 reficoding. At this point the conservative Appendix K gi requiremnts wculd show that more ECCS wat:er is injected 5

F!, into the system than can ficw up through the core. There- p

j{j fare, the ECCS water injection flow is larger than the

| break flow.t-<.T-1,

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g"} Cr Xtku stated that the British have grave doubts on the

|[ purchase of U.S. reactors because of concerns of pressure f, ;: veswl integrity. Af ter an extensive safety review of the }|1 Westinghouse pig design concept, the British Nuclear . $.; ; Installation inspectorate found the Westinghouse PWR to be fi [ licensable in Great Britain. Any concerns they may have $

i .had regarding pressure vessel integrity were ratisfactor- ift

f fly addressed. Since that ti:r,e the British have entered hm.); ) into a licensing and technology exchange agreement with '

-

| ! Westinghouse for the introduction of the PWR into Great {.f[ Britain. I am aware of the British inquiry into this area :Y.

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t t :2j,, because I an involved in technology transfer to thel

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licensee under this agreeent. ;d,

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'! Para.7.) With regard to TMI-2, Westinghouse performed calculations Zt &

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assuming that the PCRV did not close. Our calculations, j|

! which used the NSAC-1 flow history and plant data, indi- $x

j cated no fuel melting in the core. Also later, NSAC y,! revised its estimates of core inventory and net inficw to $

the reactor which resulted in a larger core inventory and L-f_,,

j ] greater makeup flows. If these ficws would be used in our 4! calculations we would again not predict any fuel melting ;[

l and the tdp of the core would benefit from ftproved steam fj' cooling. We should have expected Dr. Kaku, before ma'<ing I

j}i'q!=! the argument which he does, to have researched the liter-

ature and the NRC's public document roon in which these1'j calcuations are described.

[i'

! j-

I' Para. 8&9.) In my opinion Dr. Kaku does not give a balanced view of Cr

f' the industry resconse to the accident at IMl-2. The {t'

,

{' Xemeny Cocnission did indicate deficienc!es in the manage- ,i

( ment of TMI-2 which shculd be applied to all reactors, ii L

]! The industry has ocunted a vigorous program thecugh AIF, ;1-

| i-

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_ _ _ _ _ _ .

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EINP0 and NSAC to address these concerns. Similar concerns E

.y.were identified by -the NRC in its post TMI reports (NUREG- @

rj . C660, NUREG-0737) and the industry is working to incor- }.:

porate the new requirennts and regulations. Clearly the {industry has respcNed to learn the lessons of TMI 2 '$

-lit

P ara. 11.) With regard to this itemi all sides were heard frcm in the ij4.:

emergency core cooling hearings dich lasted two years. {'.Technical concerns expressed by different parties includ- :(ing those quoted by Dr. Xaku were addressed on a technical fbasis such that a meaningful conclusion could be reached. fAlso the NRC staff is chartered with the responsibility of 4assessing the Appendix X requirements to the light of any Inen experimental data and ensuring that the Appendix X )nodels and assumptions remain valid. In this regard the j

2:NRC gave advance notice of a proposed rulenking on TDecetter 6,1978 but has taken no further action since the hclose of the coment period. We should have expected 3

.5:Jr. Kaku to have researched dether or not these pointi 1!were considered (Aich th.ey were) in the rulemaking,.

ot:gh

- Para.12.) The fact that the EC has a-list of unresolved probles:s is &w

noteworthy since it means that they are fulfilling theirrole as the industry regulator such that industry generic 5

mkey issues are identified. The industry can then focus t|hits attention on those issues to resolve the NRC concern. I':

pFurthermore, the NRC has taken any necessary action on the jjitems to ensure that no imediate safety concern exists ;hwith respect to plants in operation. With resoect to Dr. M.

_P.Xaku's comment that "we are a quarter of a century into Sim

= the nuclear age and still have NRC concerns does not ]j

reflect well en the NRC and industry' should be taken with :i!t:

i a grain of salt. The turnaround time on a nuclear design 'jj_

;in the US is about 12 years. We are building a opera- %l

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tional information base on the second generation P'dR's in jj:e.

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the US. This operating infomation is being fed back into -t'gjthe design prccess to imrove existing and future

})Immvements are back-fitted to existing designsdesigns.

as apprcpriate. (Pages 1-12 and 1-13 of the Staff SER, f@

NLHEG-0717)- :.5

ilth regard to specific ccricents on Dr. Xaiu's testimony on accident Janalysis. The following shcuid be noted:

bI"'I bGadoliniu:s is a poisen sometimes added to the C02 ctr. 3616 ??mixture as an oxide for reactivity control but ifquid

-!$gadolinium is not adJed to the coolant as Dr. Xaku aopears e'

Sto believe. ti

?!

is approximately 51CO F where hO

tr. 3616 The melting point of 002 $as the :elting point of pure uranium metal is only2069"F. A person reasonably familiar with accident

(ij

analyses involving potential damage to the, core should {'?.

know which is higher even if he did not knew the exact Ey

tumbers, -g.5-iOr. Kaku has testified that'he read Chapter 15 of the 7.htr. 3638

C. Sumer F5AR ind the NRC's evaluation of the piant FSAA

(tr. 3633) and to be familiar with PWR accident analysis, ff5

yet on pgs. 3633 and 3639 he apparently has all of the.h

different accidents confused. :}-3?i

.

First of all the FLECHT data and recorts all refer to the I-,

?

-

experimental reflood data and resuiting heat transfer .cdels hichff are used for tha large break 1.CCA calculations (design basis

'

f The LCFTRAN code Dr. Xaku refers to is not used for5,-

accident).the design basis LCCA calculations and has no relationship

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! |whatsoever to the.FLECHT rescrt and analysis. T

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4;

LCFTRAM is used for ANS Class I, II and sece Class III transients-

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|r, calculations but not (cr the design basis accident (f.CCA)

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. calcutations, and as such is not required to -est the Appendix <.;r

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f: tr. 3675 - 3575 -{.1

5 With regard to the Appendix X ifmit 'of 22009 ceak clad i.s: :

j; temperature, the concerns expressed by Dr. Kaku en the j,.,

,I I zirconfus.<ater reacticn effects were identiffe.d and [.

j addressed in the tw-year e::ergency ccre cooling hear- [,

f| Ings. The Appendix X rule requires the use of the Baker - {t Just correlation. Data presented at the core cooling L

| hearine; indicatec that 22004 was a conservative limit, h..; .

|acceptable in ter=s of the zirconiu=-water reaction p

!! effects. Newer exner! mental data has shown this calcu- ),'..r.

[| lated reactica rate to be censervatively high by approxi- j,ii - mdtaly a factor of Oce-third. I W uld have expected [

s-Dr. Xaku to have been familiar with the literatt.re en this {,

| subject. Ej| ee,'. l.i...

I tr. 3673: W'th regard to a full scale ECCS test: A full scale test .h

| can be less demar. ding en the system being tested than a {f well devised series of component tests coupled with a good yb r

f program of analytical predictions cf the effect .' system j.|.

| failures en coe.ponents. In a full scale test, one takes y5 what one gets. One has little er no control of influences E'

y.,I a particular coeponent will experience. In individual p

|| tests, one can subject individual cor ocnents to conditions Iy..

{j,| two, three, or even ten times worse than they might pp: !' receive in *he full scale test. It is also true, of 7| m,

',! course, that a single fu}) teale test provic'as informatica ii., ~

ii for only one set of conditions. On the other hand, a 1

4.p: .

; series of wil devised engineering tests of individual.

|[: cxponents and groups of cocpopnents can cover a broad $}i '.

range of accident conditions. Nrther:vare, the results of 'dh:c.

hese tests can be matched ua with a thorough analytical ..y1 pecgran allowing confident prediction of the results of 1..s.[ system failures under a wide variety of conditions. Here {f; aoain, Dr. Xaku does not de cnstrate an understanding of f[; the role of test data. 3e cY

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, .ptr. 3632 - 3683 p

The NRC safety concerns list has been discussed earlier in [, the affidavit. (See Page 6 of this document.) 1.

|-tr. 3694 - 3695, 3697, 3693 3-

] Its apparent that Dr. Kaku does not understand the heat [transfer modes of a PWR in either steady state operation

or during calculated transients in spite of the fact that f'he has spent 1/3 of his time on accident analysis. In 4-

i[p

,

Chapter 15 of the FSAR, for the Class I, II and III acci-

}| dents the main concern is on DNS (departure fecn nucleate

boiling) <ince this can lead to a loss of fuel rod cooling I- .t

3 capability and could also lead to fuel failure. The gi initials CH3 and the concept of departure from nucleate f

boiling are so fundamental as to be famfif ar to anyone ~.[

| involved in accident analysis. Dr. Kaku should have also E|'.Eknown that nucleate boiling is preferred cooling mode as jcormared to film boiling Aich occurs after DNS. Y

lI i|: With regard to steam bir. ding, we do not regard it as a 1! uncharted area of thermal hydraulics sirca it caly con- l! .cerns calculations of heat transfer in the stem generator '.}

'

i;r ,

;, and pressure drops in the reactor coolant icop. These {matters were thoroughly examined in the emergency core jcooling hearings and conservative Appendi.c X type calcu- p1ations can be adequately cerformed in these areas to

maximize the steam binding effects and thereby increase f(r

the calculated peak clad tec9erature. Dr. Xaiu should p

have been familiar with the outline of these calculationssince they are discussed in the FSAR, NRC's 5ER and the

~

Appendix X rule. Dr. Kaku's publication in the Technology fReview is an editorial covering a broad range of issues. *f

It does not indicate the depth of understanding of either J$

thermal hydraulics or accident analysis one wculd expect $!C-

of one to be relied upon in these areas. Dr. Kaku ig!!es3

that there h acm cort of closed network of scientists in .few

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ 1

,, _ _ _ _ _ _ _ ,

'

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national laboratories Ao prevent h!.m from getting his }'

g articles published in scientific journals. In 'y [g _

; experience in conducting reviews ti sapers being pubitshed Yd>

j in =y areas, the objective is to ensure that the caper is 4' . .; . f actual and the resulting data supports the pacer's j

L conclusions. [F 4.

l EY tr. 3698 Dr. Xaku should have been aware of the DNS margin for .h

Class I II, and III transients, if he read Chapter 15 of $! .. the FSAA. T! i!.

cy .

tr. 3698: Dr. Xaku's lack of '<newledge on what peak lic.er heat rate Tclrre,

('<ilcvatt/ foot) and hot channel factors are further demn- g[ strates his lack of essential *(nowledge in the thermal- {} hydraulics and accident analysis arer. The key parameter .j| in the large break t.CCA calculations .s the peak liner )! heat rate for the hot channel. It is this charinF which J

Liireaches 22000F for the calculated LOCA. -

$x

tr. 3730 - 3731 h| Again Dr. Kaku refers to WCAP - 7907 which is the LOFTRM h

A

{ report as being necessary for the Appendix X calcula-

[ tiens. The LCFTRM code is not used in the LOCA Appendix 3I

..X calculations for Westinhouse PWR plants. Nowhere in (,

,

? this report is it stated that it is used for Appendix X @j

i analysis. As clearly pointed out by the NRC en page 15-10 I,

of the Staff SER NUREG-0717, the reports WCAP-7907 and i..'

.

i WCAF-9230 describe the -ndels used in the evaluation of [feednater system pipe breaks and not loss-of-coolant acci- [

'

dents in the reactor coolant system.

t iDr. Xaku refers to the FLECHT crogram on 99 3730 as a iconcern. It should be noted that the FLECHT program and ..:data was fully debated in the Emegency Core Cooling hear- [ings and was accroved for use in the Appendix X analysis. g

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| Cr. Kaku also ties the Flecht prograta to '4AFS - 7907 5

[ (LCFTRMI and '4AP - 9230 feedline ruoture). There is no ,

)- connect ion. Both quoted '4 CAPS are for Class I, II and III 1J,<

transients which are not design basis LCCA's, whereas the y;.

.

rLECHT tests were designed to understand reflood for a E.gi.

i. Iarge LOCA.gb[A=

vtr. 3742: Dr. taku infers that the thermal-hydraulics of a Class 9 Eytransient can be obtained frera a standard text book. T

.t

Nothing could be further tron the truth. The Class 9 Yl::

*ccident is an extremely complex coupled heat transfer,x$) ;

fluid mechanics, and materials problem n ich requires a gi

high level of expertise in a wide range of disciplines. |fkfI @'

rAnyone can oostulate accident scenarios however to estabitsh the extent 'T

kto which varicus accident scenarios are credible is an entirely dif-.v[ ferent matter. In order to determine the credibility of any given event @

ene must have access to detailed knowledge in all the related technical jdisciplines. Therefore, when one evaluates the credibility of nuclear f

Uaccident scenaries one must have the ;ssistance of others who have

detailed faciliarity with just abcut all the technical disciplines 5involved in the design, construction and operation of a nuclear plant. f[{For exa@le, taking the simple assu@ tion of a pipe break, to establish il

,

f ts credibility one would have to '<new the details of the provisions for fleak detection, and the sensitivity of those orovisions (Fluid Systems f

f and Instrumentation Engineering). Then one would have to know how large ,y. a crack would have to be in order to have sufficient leakage to be f

detected (Therro-Hydraulics, Radiation Effects, and Fluid Systems [$"- -

cEngineering). j;

S:

Then one nould have to '<new the crack size which could cause rupture of fthe pipe (Stress Analysis, Metalurgical Engineering and Fracture f

k Med.anics) . Then one would have to '<new the fength of time for the -[le s:rallest detectable crack to grow to the critical size eich could cause ff $

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"Jpture under the existing stress fields during oceration (Stress Analy- M{;'

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- " sis, System Engineering, Metalurgical Engineering). Only then would a Icompartson beteen leak detection ti:e and the time for a crac'< to oro- h!

e.pagate to rupture allow a detemination of the likelyhoed of the opera- gi

ter detecting the leak and shutting dcwn (Detailed kncwledge of Tech- [nical Specificatiens and Operating Precedures) before the crack could jj,

pecpagate to the critical size , ditch could cause rupture. Even:after [the credibility of the size rupture is established it is necessary to f.c.detemine the potential effects of such a rupture and the credibility of Ithose effects. This requires a knowledge of the function of the pipe in "h.

connection with other systems (Fluid Syster::s Engineering) and the proVf- j~

sfons within. the plant to mitigate or prevent adverse effects (control .)jand protection and Nuclear Safety Engineering). Then one would have to spj

evaluate the consequences of each scenario (Fluid Mechanics, Th&mo- @.m

Hydraulics and Mechanical Effects Analysis) and the credibility of the -j:consequenc'es of any overall scenario would be the ccebined credibility - ]$|

:=of all the events required to occur in ceder to lead to those con- gsequences. While one can postulate scenarios and run computer programs $jto generate consequences based on the assumed scenario, this of itselfwill not provide any insight to the credibility of the consequences fcalculated. $

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AITACFFEAT A 25E-c.-!;:

gEducational and Professienal. Qualification ,.n.

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!.awrence E.*dochretter .,::. . . .

Mtiscry Engineer-Safecutch Engineering TwNuclear Safety Decartrent gj

Westinghouse Nuclear Energy Systects .%|A- .

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N!y.3E ducation a-* ,:

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[p,8.S. Mechanical Engineering, University of Buffalo,1963 .

| M.S. Nuclear Engineering, Ourdue University,1967g@-Ph'3 Nuclear Engineering, Purdue University,1971

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li'4estinghouse short courses on canagerent techniquesHIT short course on beo-phase ficw and heat transfer,1973 $

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Professional Experience 1-Si

From 1953 to 1971 ' worked on my H.S. and Ph'D in Nuclear Engineering in -fthe areas Jf liquic metal heat transfer .and turbulent flow fluid mechanics. yI also taught a undergraduate course on heat, acr entu;r, and,cass transfer. j'

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From 1971 to the present I have been associated with Westinghouse Nuclear yEnergy Systens and have held different technical and ranagerial positions jof increasing responsibility. From 1971 to 1972 I was a senior engineer -@

in Ther al-Hydraulic Design and worked on developing thersal hydrau".ic y)design methods fob PWR cores. From 1972 to 1977 I was manager of the $

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Safeguards DeveTy, ent group who had the responsibility of planning, j,analyzing, and) utilizing themal-hydraulic data from Westinghouse and jother experiments to aid in the develo;nent and verification of thermal- 5

1.

,~hydraulic models for Westinghouse safety analysis codes. Programs which .

;

were under ey direction were the REOiT program, FLECHT-SET program, steam '!

Vater Mixing program and other Internal Westinghouse programs. I was lp!- appointed Advisory Engineer in 1977 and am the principal investigator on ([

) the NRC/EPRI/ Westinghouse RECHT-SEASET progran. I also participated in @f support efforts for the TMI-2 Accident, Westinghouse Kar,eny Coenission jI

j support, and the Industry Task Force on TMI-2 clean-up activities. $i.T.*1 -;

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Since 1976 I have been an adjunct professor in Nucle 4r Engineering at $Carnegie-Mellon University and have taught or tean t' aught graduate level (

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courses in thermal-fluids design of Nuclear Reactors Advanced Topics $I in Nuclear Safety and Two-Phase flow and Heat Transfer. I am a renber of }r[

the ASME, X-13 (Neutronics Heat Transfer Group), and the Pittsburgh section :'[:- of the A'iS. I am a reveiwer for the Nuclear Safety Journal, ASME, and Journal Y

~" of Nuclear Technology. A list of my publications is attached. -fx

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| PUBLICATICUS_ .i.i? $:c . x-

1. Prickett, W. , N6edi, A. , Hechrei ter, L.E. , Hategger, L. , Foad, H. , Arcella , F. , 1-

"The Design and Evaluation of a 1000 S'e Central Station Power Plant with a J.i;

Circulatir.g fuel Gaseous Cent Feactor,' AEC Contract No. AT(ll-1)-1373 $:

! (Graduate Nuclear Engineering Cesign Seminar,1964). f. i il

r 2. Hochreiter, L.E., and A. Sesonske, " Turbulent Te perature Fluctuations in ~ dii Flcwing Sodium-Fotassf uci * Transacticns 4.N.S. San Diego Meeting, June 1957. - . i!!

E3. Hochreiter, L.D. and A. Sescoske, 'Tharsal Turbulence Characteristics in si

Sodit=-Petcssic:," International _ Journal of Heat.5ts: Trans fer, Vol. 12, 4!pp. 114-118, 1969.

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4. Hochreiter, L.E. and ?.. Sesenske, "'ielocitj Profile and Pressure Drop ligMeasurerents in Fully-Developed Mercury Pipe F1cw." Transactions ANS fi

Winter Meeting, Vol.13, No. 2,1970. @#

m5. Xudva, A. , Hochreiter, L.E. , Sesonske, A. , "Turbuler.te Scales and Eddy 4

Diffusivities," / aper presented at ihh National ASME-AlChE Meeting, hPhilade!; % cenry!mia, Aug st 1968. .qi

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}$6. Chu, P. T., L. E. Hochreiter, H. Chele-er L. H. Boran, L. S. Icng,,_

g "THINC-!Y, A New Ther-al Hydraulic Ccde for PE Ther al Des.ign,"Transactions A'iS Washingten Winter Meeting, Yol.15, No. 2.1972. ?.,

5g7. L.E. Hochreiter, A. Sesonske, " Turbulent Structure of Isothen al and $qj Non-Isother al liquid Hetal Pipe Flow", Int. J. Heat. Mass Transfer, '$

q Yol.17, pp.113-123 (1974). j.f 8. J. A. Blaisdell, F.F. Cadek, L.E. Hochreiter, . A.P. Suda, J. Waring. .j, .le

i "Fl.ECHT - Systems Effects Tests - Phase A Results" Transactions ANS q} San Francisco Heeting Vol.16, No. 2,1973. .h

3:Y. 9. J.P. Waring, E.R. Rosal . L.E. Hochreiter, "FLECHT - Systems Effects Ay Tests. - Phase 51 F.esults , Transactions A.';5, Washington D.C. Meeting, 4a

II .Yol. 19, 1974. .

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10. H. C. Teh, L. E. Hochreiter, "An Analysis of Oscillations in Siculated A|

Reficed Experi:mnts", Transactions Nis Vol . 22, 1975. fa

11. G.P. Lilly, L. E. Hoch eiter, " Mixing of Er ergency Core Coolant with Steam j{ in a PWR Cold Leg". Transactions AHS Yol 22, 1975.. *s

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! 12. E. R. Rosal. L. E. Hochreiter, "FLECHT Low Flooding Rite Test Series," f| Transactions ANS, Vol 22, 1975.

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13. C. P. Lilly, L. E. Hechreiter, "Exa.,inatico of Deflecd Heat Transfer 3; Nchanis-.s using Flecht Da ta*, Transactions A'iS, Yol. 24, pg. 3M (1976). .i'

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[ PUBLICATIONS (2) I- 1

E14. H. C. Yeh, L. E. Hochreiter, " Mass Effluence Curing FLECHT Forced Reflood h

Experiments". Transactions MS, Vol. 24, pg. 301 (1976) 42:

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j%15. H. Chelecer, L. E. Hochreiter, L. H. Bocan, ard P. T. Chu, "An improved

Thermal-Hydraulic Analysis Method for Rod Bunale Ceres", NuclearEngineering and Design, Vol. 41, No. 2, April 1977. di i

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16. J. P. Waring and L. E. Hochreiter, "FLECHT-SET: Systers Effects On yReflooding", presented at the Manchester Conference en Thereal 'TReactor Safety (Sectester 1977). .. ,. - /. s t 7f

17. L. E. Eachreiter, K. Riedle, "Reflood Heat Transfer In Pressurized Water 5Reactor", Tnemal and Hydraulic Aspects of Nuclear Reactor Safety, Vol.1: yLight Water Reactors, ASME Atlanta ceeting 1977. g

18. L. E. Hochreiter, E. R. Rosal, R. R. Fayfich, "Downflow Fil:n Boiling in da Rod Bundle At low Pressure" Presented at the ANS/ ENS Nuclear Power Reactor ESafety Meeting, Brussels (1978). j

19. H. C. Yeh, C. E. Dodge, L. E. Hochreiter, " Generalized Reflood Heat Transfer hCorrelation" Presented at the ANS/E:ts Nuclear Power Reactor Safety Meeting, 5Brussels (1978). , yj

20. A. G. Stephens, M. A. Ecery., L. E. Hochreiter, " Local Density HeasurementsIn A Steam-Water Mixing Zone Using the photon Attenuationlechnique",

d@Topics In Two-Phase Heat Transfer and Flow, S. G. Bankoffi ed.', Trans ASMEpresented at 1978 Winter Meeting, San Francisco. p.x , j

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Radioa;Yao, L. E. Hochrsiter, C. E. Dodga, "A Simple Method for Calculating ,' 1S. C.

ctive Heat Transfer in Rod Bundles with Croplets and Yapcr as the .5!

| Absorbing Media", Journal of Heat Transfer, TRAh5 ASME, Vol. 101 (1979) -$.i x

Nuclear Technology Vol. g. Dececher 1979.H. C. Yeh, C. E. Dodge, L. E. Hochreiter, "Reflood Heat Transfer Correlation ', $22. E

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- v .e:23. H. C. Yeh, L. E. Hochreiter, 8 Mass Effluence Curing FLEchT Forced Reflood W

Experiments", Nuclear Engineering and Design, Vol. g,1980. ys.

24. D. L. Burman, L. E. Hochreiter, S. E. Jacobs, D. F. Paddleford, S. C. Shell, EH. C. Yeh, " Degraded Core Cooling Calculations for THI-2" Trans MS Vol. 34, ilpg. 869, Las Vegas Meeting, June 1980. I {

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m25. D. L. Surman, P. DeSchutter, L. E. Hochreiter, P. J. Kuchfrka, " Comparison f-:

of Westinghouse LOCA Eurst Test Results with ORNL and t7ther Programs Results", gCSSI Specialist meeting on Fuel Eehavior under Accident Conditions, Sept.1 - =4, 1980. Helsinki, Finland. $

.?!26. L. E."hochreiter, S. Weng, "A Model For Dispersed Flcw Heat Transfer In A 11 .

Rod Bundle During Reflood" Paper presented at the 19th National Heat Trans- pIfer Conference Orlando, Florida, hTD-Vol. 7 (1980). F.

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27. R. C. Howard and L. E. Hochreiter, " Experimental Det$rnination of the Two-Phase ili'-c;

Ficw Distribution at the Tube Sheet cf a Ndel PWR Stea= Generator", @Proceedings cf the ANS/ASME/3RC Internation! hoical Meeting on Nuclear ;;%Reactor Thermai-Hydraulics, NUREG-CP-00i?, Vol .1,1930. }

=.-28. L. E. Fechref ter, S. E. Jacobs, N. J. Liparulo D. F. Paddlaford, "PWR 5

Contaircent At cspheric Response for a Postulated Class 9 Accident:, MTrans MS, Vol. g, Washington, D. C. , Novecter 1980. j

w.<29. S. Wong ar.d L. E. Hochreiter, " Low Reynolds Nr.ber Forced Convection StesafiCooling Heat Transfer in Rod Sundles", Trans ASME, paper 80-VA/HT-S9, Winter yr.eeting Chicago,1980.3E30. S. C. Yao, l.. E. Kochreiter, W. J. Leech, ' Heat Transfer Augmenta- ig

tion in Rod Eundles Xear Grid Spacers", Trans ASME, paper 80-WA/hT-62, n::Winter ceeting, Chicago 1980. 7d

is31. R. C. Mcward, L. E. Hochreiter, H. F. McGuire, "U-Tube PWR Steaa Generator ;:5.]Heat Transfer During a large loss of Coolant Accident Transient", Trans ASME, ;iE!peper S0 '#/HT-61 Winter =eeting, Chicago,1980. iiEq

32. N. Lee, S. Veng, L. E. Hochreiter, " Heat Transfer Mechanisms During Rod Sundle 3:Reflooding" presented et the'Yhted CS'il specialist Meeting on Transtent Two- $~Phase Ficw, March 1981. jp

.=:33. M. J. Loftus. L. E. Hochreiter, C. E. Ccr.way, E. R. Rosal, A. H. Wenzel, .35"Non-Equilibriun Yapor Temperature Measurecents in Rod Sandle ud Steam j'.i

Generator Two-Phase Flows", Presented at the, Third CSN! Sptcialif t meeting Xcn Transient Two-Phase Flow, March 1981. G4:134. D. Squarer, A. T. Pieczynski, L. E. Ecchreiter, "Dryout In large Particle Deep 'j

Debris Beds * Trans MS, Vol. 38_, June 1981.,

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&- J 5. D. Squarer, A. T. Pieczynski, L.. E. Hochreiter "Effect of Debris Sed YPressure, Particle Size, and Olstribution, on Degraded Core Coolability" ?.(To be presented at Winter ASME meeting and accepted .for publication in 'ENuclear Technology). ,{

:-36. D. L. Burcan, D. L. Hagrean, L. E. Hochrei ter, N. Lee, M. J. Loftus , ?

" Inhibition of Coplanar Rod Bundle Blockage By Cross Ficw Coolf ag*, Presented Iat the ASS /E'IS Topical Meeting on Reactor Safety Aspects of Fuel Eehavior },August 3 - 6, 1981.

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WCAPS, }

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- E1 . 2:1. H. Chelecer, P.T. Chu, L.E. Hochreiter, "THINC-IV - An Improved Program 4

for Thereal Hydraulic Analysis of Red Sundle Cores", WCAP-7956 (1973). -gE! 2. L.E. Hochreiter, H. Chelecer, " Application of the THINC-IV Progra:n to y

l FWR Cesign", WCAP-8054 (1973).

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3. J. A. Blaisdell, L.E. Hochreiter, J. Warirg, "RECHT-SET Phase A Report", .yWCAP-8238(1974)... I

N4. 6.P. Lilly, A.G. Stephens, L.E. Hochreiter, "l/14 Scale Steam / Water y

MixingReport",WCAP-8307(1974). E. y5. G.P. Lilly, L.E. Hochreiter, "l/3 Scale Steam / Water Mixing Report", ,[

WCAP-8423(1974). ym

6. W.F. Cleary, J.P. Waring, E.R. Rosal, L.E. Hochreiter, M.J. Slafka, .$M.F. McGuire H.J. Fix, "RECHT/ SET Phase 8 Systen Design Description", IWCAP-8410(UL-784)(1974). _j

57. J.P. Waring, E.R. Rosal, L.E. Hochreiter, "RECHT/ SET Phase 8 Dati gReport",WCAP-8431(1974). E

$8. J.P. Waring, L.E. Hochreiter, "FLECHr-SET Phase 8 Evaluation Report", %WCAP-8SS3(1975). 3

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'(' 9. E.R. Rosal, H. McGuire, L.E. Hochreiter, M. Krepinevich, " Low Flooding E.

sRate FLECNT Data Re;crt", WCAP-8651 (1975).

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E 10. Centrib<rtien to WCAP-6622. Topical Report W ECCS Evaluation Model, October I) 1975 version. --

jE,vi 11. L. E. Foe.reiter, 8. A. .v Intyre, E. R. Rosal, M. Y. Young, R. R. Fayff ch, Ic

i R. P. VI.iu', "G-2,17x17 Defill Feat Transfer Tests and Analysis", h<

j WCAP-8703-P(1976).[,

! 12. G. P. Lilly, H. C. Yeh, L. E. Hochreiter, N. Yamagaucht, "PWR f' 2CHT j.

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] Costne Lew Ficoding Rate Test Series Evaluation Report", WCAP-8838 (1977).[P

i,j 13. Centributice.s to 't.AP - 9183, "RECHT Lcw Flooding Rate Skcv Series ?J Evaluatico Report", (Novecber 1977). .j ..

1 3? 14. C. E. Conway, L. E. Hochreiter, H. C. Krepinevich, ri. W. Massie, Jr., )J E. R. Rosal, R. C. Howard, "PWR RECriT Separate Effects And Sys-

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Effects Test (SEASET) Progran Plar.", iiRC/EPRI/Jestir.gho:.se Report .';o.1,p

|| :'ececher 1977.{., '

| 15. L. c. noct.reiter, H. W. Massie, Jr. , R. C. Ecuard,11. J. Loftus, ' . IJj Xavalkevich, :t. C. Krepinevich, H. C. McGuire, A. E. Tcr'e. "PWR iREC 5T SEASET Stea- Generatcr Separate Effects Task Task Plan Repcrt", ge

| .iRC/EPRir*estir.ghcuse Pe;srt do. 2, March 1978. J'

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1Y' ::fi. 16. L. E. Hochreiter, C. E. Conway, C. E. Dodge, H. C. Krepinevich, H. W.$ Hassie, Jr., E. R. Rosal. T E. Sobek H. M. Valkovic, "PVR FLECriT

d. SEASET Unblocked Bundle, Fo:ced And Gravity Reflood Task: Task PlaniiJ Plan Rep 0rt, HRC/EPP.INest'nghouse ?.eport No. 3, March 1978.w

!h 17. R. C. Howard, M. F. McGuire L. E. Hxhreiter, "PWR FLECHT SEASET Steam

b|FGenerator Separate Effects Tast Data Report". RECHT-SEASET Progra

13 HRC/EPRI/destinghou.e Report No. 4 January 1930.wn

18. L. E. Rochreiter, R. A. Basel, R. J. Dennis , N. Lee, H. V. Masste, Jr.,ip[ M. J. Loftus, E. R. Rosal M. H. Valkovic, "PWR RECHT SEASET 21-Rodl.;1.! Flcut Blockage Task Task Plan Report, RECHT SEASET Prograra NEC/EPRl/

M Westinghouse Report No. 5 hUREG/CR-1370 NP-1382 WCAP-95S3, March 1980..L3 19. L. E. Hochreiter. H. Lee, M. F. McGuire, H. W. Massie, Jr., M. J. Lof tus ,jIJ M. M. Yalkovic, "PWR FLECHT SEASET 161-Rod Bundle flow Blockage Task: -

;[9 Task Plan Report". FLECHT SEASET Progra HRC/EPRI/ Westinghouse P.eport No. 6;E NUPIG/CR-1531 EPRI hP-14SS, WCAP-9692 July 19S0. .

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f. 20. M. J. Lof tus, L. E. Hochreiter, C. E. Conway, C. E. Dodge, A. Tong, E. R.4 Rosal, M. M. Valkovic, S. Wong, "PWP. FLEtriT SEASET Unblocked Bundle, Forced

4 Anc Gravity Reflood Task Bata Report', FLECHT SEASET Prograc NRC/EPRl/ West- iS inghouse Report No. 7 NUREG/CR-1532, EPRI NP-1459 WCAP-9599, June 1980. -

p@ 21. S. Wong, L. E. Hochreiter, " Analysis Of The FLECHT SEASET Unblocked Bundle !

!E Stean Cooling And Boiloff Tests *, FLECHT SEASET Progran HRC/EPRI/ Westinghouse ;

y Report No. 8 NUPlG/CR-1533 EPRI NP-1450, WCAP-9729, January 1981. -

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i; I heareby certify that the foregoing information is true and 4-| '}'

correct to the best of my kncwledge and belief. s5?y.

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6th AugustSubscribed and sworn to before :ne this . day cf .

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