ML20079J564

Summarizes Findings of 821207-08 Plant Tour Re Use of PORVs or Depressurization Scheme for Removing Decay Heat.Addl Info Requested Includes General Arrangement & Isometric Drawings of Piping Near San Onofre Pressurizer

ML20079J564
Person / Time
Site:	Palo Verde, San Onofre, 05000000
Issue date:	12/16/1982
From:	Berry D SANDIA NATIONAL LABORATORIES
To:	Marchese A Office of Nuclear Reactor Regulation
Shared Package
ML20079E633	List: ML20079J564 ML20081C062 ML20081C069
References
FOIA-83-168 NUDOCS 8212280269
Download: ML20079J564 (3)

Text

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Sandia Nati0nal Lab 0ratories A :. u e : t r. e v. ve..n 5 7 H December 16,- 1982

, A. R. Marchese

! U. S. Nuclear Regulatory Commission . ,

Generic Issues Branch Office of Nuclear Reactor Regulation Washington, DC 20555

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Dear Andy:

On December 7 and 8, 1982, we accompanied you on a~ tour of the San Onofre 3 and Palo Verde 1 nuclear power olants to gather information related to the use of PORVs or a similar depressuri-zation scheme for purposes'of removing decay heat. Based on these tours, I am writing this letter to summarize some of our l

. findings and to document some of the areas requiring further l study by Sandia as related to TAP A-45's evaluation of the need ~

for depressurization capability at San Onofre 2. (Re: letter from D. Berry to A. J5rchese, dated 10/18/82.)

Findings (for both plants, unless noted otherwise)

1. The Palo Verde 1 plant is much more compartmentalized than the San Onofre 3 plant, with the result that Palo Verde appears to be less vulnerable to special emergencies than does San Onofre 3. '

,' 2 . Pneumatic supplies to the atmospheric dump valves appear to y # be available despite the unavailability of electrical I power. The actual design limitations of the pneumatic supply needs to be determined.

3. HVAC support 'for diesel generators and other system + -

equipment appears to be designed sufficiently independent to minimize common-mode failures resulting from the loss of one

.HVAC train. 9

74. Cold shutdown systems have been designed and built to safety-grade standards for independence and reliability, o I 5. The Palo Verde 1 plant has 100% load rejection capability withou,t reactor scram, while San Onofre 3 must scram given q 100% load rejection,

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/ A. ~R. Marchese, USNRC December 16, 1982 Areas Recuiring Further Study for San Onofre 2

1. The design, ef fectiveness, and reliability of the auxiliary spray and the auxiliary spray bypass features must be deter-mined, with consideration being given to the operat.ional

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experience of St. Lucie.

2. The utility-prepared fire analyses for ths auxiliar'y feedwater "O pump room, cable spreading room, and high pressure injection pump rooms must be reviewed.

'c 3. The pumping characteristics of the HPI pumps must be-evaluated

( , in the context of their ability to support feed and bleed operations.

4. Tne, safety benefits of,having electrical Class IE heat,ers must be determined.

5. The technical bases for and quantitative estimates of the failure probability of steam generator tubes must be identified and evaluated.

6. The operational experience at St. Lucie regarding advers'e contamination and containment environments following the lifting of pressurizer safety valves must be reviewed.

7. The contention by Combustion Engineering that estimated

',- radioactive material releases during a steam generator tube rupture are equivalent for PORV or auxiliary spray mitigation schemes must be confirmed.

C 8. The technical specification limitations (i.e., stress

}N' limitations) which prevent Ginna from using auxiliary s' pray under certain accident conditions must be determined and contrasted with San Onofre. '

9. The significance of different PORV sizes (2:1 ratio) for St. Lucie 2 and Calvert Cliffs must be established.

To assist in the resolution of the above areas of concern, we need the following information from you:

a. General arrangement drawings of San Onofre 2 and isometric drawings of piping in the vicinity .of the San Onofre 2 pressurizer.

A b. The latest copy of Combustion Engineering emergency JF - operator guidelines, CEN152, Rev. 1.

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, - - A. K. Marchese, USNRC December 16, 1982

c. Technical specifications for the electrical, auxiliary feedwater, auxiliary spray, normal spray, steam. generator, and 'high pressure injection systems,

d. Plant operating procedures for station blackout, steam s generator tube rupture, and loss of feedwater accidents.

c-e. The utility's submittal on fire protectiorr requ-irements.

f. The utility's evaluation of auxiliary feedwater system reliability.

g. Information on St. Lucie's auxiliary spray and pressurizer safety valve experience.

Please call me or Don Gallup if you need any diarification of our needs.

Sincerely, QOn Dennis L. Berry Nur' lear Facility Analysis Division 9414 Copy to:

9410 D. J. McCloskey

• 9414 G. B. Varnado '

9414 G. A. Sanders 9414 D. R. Gallup ,,

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. j sarcy UNITED STATES N g 'e NUCLEAR REGULATORY COMMISslON

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V' % . . J/ AUG 2 319S2 MEMORANDUM FOR: Themis P. Speis,. Assistant Director. . .

for Reactor Safety Division of Systems Integration FROM: 0. E. Bassett, Director Division of Accident Evaluation - - - -

Office of Nuclear Regulatory Research

SUBJECT:

FEED AND BLEED EXPERIMENTS IN SEMISCALE As you are aware, RES has performed a Semiscale feed and bleed experiment (S-SR-2) at the request of NRR. Results of this test indicate that '

difficulty was experienced in maintaining 'a steadf-state feed and bleed condition without uncovering the heater rod bundles. Several members of -

your staff have had questions as to how this relates to the PWR feed and bleed operations being purposed for several plants. These questions are now being addressed by EG&G Idaho, Inc., while taking into account the ~

atypicalities of Semiscale, as they might affect the feed and bleed behavior.

As a stact, EG&G have provided a letter report of several steady-state calculations they have performed in an effort to help understand the Semiscale. data (see enclosure). These calculations included a study of.

the sensitivity of the results to core power, break flow quality, surge line and pressurizer. geometry, and availability of equipment, and were extended to parametric values typical of a commercial PWR. The report concludes that there are large uncertainties in predicting the satisfactory per'formance of feed and bleed in the steady-state, but that the Semiscale experiment does not point to the existence of a definite problem regarding '

the satisfactory performante of feed and bleed in a PWR. "The Semisdale results still must be analyzed to determine the extent that the atypicalities effect the results and to put them in proper perspective. This analysis work is currently in progress and should be completed in September 1982.

The Division of Accident Evaluation has concluded that the Semiscale results have not produced new and unique results that indicate a PWR would have a definite problem regarding feed and bleed. Therefore, we do not' recommend a board notification. We have concluded that the Semiscale results should be carefully analyzed to determine that relevance to PWR feed and bleed transients and this work is now in progress. We

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should be able to provide you with a more complete answer to the relevance of the Semiscale experiment and our investigation into the feed and bleed transient in a PWR by late September 1982.

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/MA Bassett, Director Division of Accident Evaluation Office of Nuclear Regulatory Research

Enclosure:

Ltr fm North to Tiller dtd 08/06/82 .

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L. H. Sullivan, EPB R. R. Landry, EPB C haron, NRR ,

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. August 6, ISB2 Mr. R. E. Tiller. Director -

Reactor Operations and Programs Division

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Idaho Falls. ID B3415 l l

PRI,MRY C00 TANT SYSTEM FEED AND BLEED - PM-137.-E2 - - -

Dear Mr. Tiller:

At the request of the Nuclear Regulatory ComissToh the semiscale Program recently conducted experiments designed to investigate the feasibility of prt.sary coolant system (PCS) feed and bleed as a me.ans of r-ejecting decay heat in the absence of steam generator haat removal.

. g The results and preliminary analysis of the experiments sugges.ted that a reasonable uncertainty may exist in the ability to effect stable -

PCS feed and bleed. Since current pressurized wat.er reactor emergency operating guide.1ines call for priory feed and bleed uncer certain abnormal conditions, it was cor41dered of some importance tFut the general subject of feed and bleed be studied in some depth aaf ttiat the 5'emiscale results be carefully analyzed so that they might be inter-preted in the proper perspective. To this end, the Semiscale Program is currently engaged in an extensive analysis effort invobirs both full-scale plants and experimental results (i.e., Semiscale and LOFT). .

~ The purpose of this letter is to provide a brief overview of our analy -

sis of feed 'a.nd bleed to date, including the recent Sectiscale results.

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=-----..-=..-....-.....e m ..aa.w. d fritsry coolant system feed and bleed in a pressurized water reactor

. becoros a necessary decay heat reeoval mechanism in the unlitely ' event '

that all secondary heat renoval capability is lost. While there exist

- ntoerous scenarios that could lead to this sitsation, the focus of the present analyses .Ts the._ feasibility of achieving a favorable cool * ... . s

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. C. the stea= generator secondaries cre c<npletely depleted of coolant

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PK-137-C2 t,tnust 6, 1982 Psce Z E.

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L_ t.ht.bigh. pressure injection system (ECCS).fs.opern.tve to-. .._

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. an undegi-a.ded condition

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, 4. the pressurizer hestert tre inactive

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. re. ifer valves (s) (PORVs) are operative 3 6. primary recirculation ptraps are off .

Feed and bleed would cocraence een the PpRYN) were 6pene~d. {ble'ed) and high dia.- the PORV(pressure s) provides injection for began (feed).- The passage of steam out a

tiers of CCCS coolant pro..the, rejectionfor ofthe decay heat while the,introduc vides sakeup resultant coolant loss. .

The rewinder of this le'tter exarnines the general aspects of primary W -

feed a.nd bleed operation and the therul-hydrauf fe phenocneed that gover6 it.

The pr' edicted fystem Wsponse is outliried first had the

- effects of uncertainties plant parameters. Next the are illustrated edth an exa ple ft:n actual key therzai-hydrauf fe phenomer.a that in-e fluence the uncertainties are discussed. Finally, the data from 1 5esiscale experirvents is briefly presented to deconstrate system signa-tures and e esponse during an actus1 experiment.

.: )!' -THEORETICAL FEED AN M.EED OPEP.ATING PRESStfRE R -

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A siniple examination of the mass and energy transfer pathways associ-in.. . .

tied with feed shd theid Ycsulti'th thicandluit6h that' feed isd blei!E ~

4 -.is theoreticc11y possible within a d certain- band of pressure (te.e Figur MT'?T:.WW.cJ) d ir ' -

• dieey tes.t Jhe.poverning

'l%I/eTPHPIFTiiMi .pareeters, thfch,d.stermine 6 tis 7%f 'PMWYitshn'd enththis press,um f DF :rg-l.'

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,- tha' ceters' 6 prs,t are ing 6tnfunctio prescats (dtherg$ minhaspriiery f. :ifstdd pretsure The. 'pi-estbre..f Wer 66u6d it dich.the PORY.3 4fHn~T.. '

can' pass en6 ugh stema

~~ (wit 3 the coolsnt replaced b'y anbient tec:perkturt

',. .rzterl:tp .remva cuff.lcientienergy' ftpa the systen Steady-59te-:. ',:-fi.c ='.

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hr,ot.ponible. . OperatioGt4 pressure ebove the to.edv beed vr.ay& '.44

be 2.ccv.plished by cycling the FORY opert and closed within a desired pressure band. - --

An upper pressure bound to the steady-state. operating band is defined L- by a balance between the PORY average coolant r.emoval rate and the RPIS coolant injection rate. The aver e PORY coolant remval rate is simply defined as the core power, div ded by the difference between'

, inlet and outlet enthalpies:

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Base through case conditions asstree that.100f quality steam t5 discharged the PORY. The affect of neied quality is ext.:::ined later,

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PN-137-82 .

August 6.1982 '

Page.3

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• O core out ' Iin I (ihiseesstr>es with that thewater.

bient terperature coolant reueved through the P_0RY is replaced '

. . . Actually a coolant deficit exists at pressures higher than the t.jper be.md and a steady-state condition camot exist due to a continual loss of system coolant inventory.)

Below the upper bound the system mass inventory can theoretically be maintained HPIS or cycling withinit aondesired and off.operating range by either throttling the O UNCERTAINTIES ASSOCIATED WITM STEADY-STATE OPEP_

f e SeEnot 9el.1,Afefined "QTE il __2RgeveraJ.un,certaihtiesge

.,-.w.J3 pfeCtIQ the .cdrYei[ Subjects'Cdised 3ny@.yhW.PCRV cess; redovaT. tn'd enerpy'. rem;.Eg@W,J(s are the sre the PORY is dependent on the fluid conditions at the top of th surizer.

4

- POM will be a mixture of liquid and Tapor.If the pressurizer is near li At a given systes pressure this results in great'er mass flow than for saturated stea:s f1w through the PORY.

The result of having two-phase flow through the PORY is therefore f to twer the upper bound pressure, Uncertainty phase flow. also arises in the PORY energy removal curve due to two-

- decreases while the mass discharge rate increases.With Rependirg upon decreasing the quality the energy removal rate at a given pressure may be less

- than or greater than that for saturated steam.

the opercting. band wi.II. Wry.'acco,rdingly..._ l'he lower bound of

>- Another significant variable that affects the width of the opercting be:d PORY. concerns the actual heat lead that must be rejected through the Core decay heat decreases continually with time af ter shutdcun. . ,

Also, if some additional heat sink exists, such as envirore: ental heat loss or residual water in the steam generator secondaries, t.he heat I load required to be rejected through the PORY will decrease. '

to Figure 1, this results in loweririg the core power line end thuspreferring lowering the bottom end of the operating band, and also lew: ring the POW ing band.average nass flw curve, thus raising the upper end of the operat-s -. . ... .. .= ..

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ih'e finai Uncertainty addressed here is the uncertainty associsted E

I kith the HP15 injection curv&. The elfects here are rather clear; a lever injection rate will Iwer the upper end. of the oper.rting band.

The quentitai;ive effects of the uncertainties and/or. variances discus-sed above are illustrated in Figures 2 through 6. For these examples the curves Wern generating plant. 49gted using date obtainen from the Zion ! nuclear A 3411 HW(t) pressurized water reactor, Figure N.. 2 shews a primary feed and bleed map for n 22 cecay neat powr level.

A steady-state opercting band is seen to egist between 7.5 amr 1( MPs.  ;

A decay heat level of 2t of full pcmer is typical of the tire. period  ;

fr::e about 10 c:in to 20 min after shutdown. Figure 3 is a similar

" curire, but here no makeup punp injection is assumed:- only the HPIS

'. pu ps were assumed to be operating. lhe HPIS pusps are shown to deed-head at ahnut 10.3 iga. For this cbse no nperating t>and exists, s,ince at the rainimuu pressure where the PORY can re.~. ave the energy t.hcre

~ ' is a mass deficit between the PORY coolant ret: oval and the HPIS injec-

. tion capacity. '

' Figure 4 chowc the primary feed ared bleed map re:c l-1/25 rnit power,

. h decay heat level typical of the period frc: 1/2 to 1 hr after shut-down, and for only HPIS injection. Cceparison to Figure 3 Ifws that the reduction in core power and corresponding PORY average cast flow both act to establish a steady-state operating band. .

The" above curves are based upon the assumption t. hat 100% quality steam exists at the PORY. Figure 5 shows the sensitivity of the PORY energy

~' removal curve to Icwer qualities as determined with the HTM fIcu encel.

Since the energy recoval per unit mass decreases while the ness flow .

rete increases the energy re=nval rate initially decreases with decreas-

. ing quality. Hewever, since the mass flow rate increases tdstantiall

- .. with decreasing quality the energy removal rate eventually increases. y The effect on the lower operating bound pressure is not larget hoever the large increase in PDRY. mass flow rapidly levers the upper end of .

the band. . As en ext =ple, for the conditions used in figure 4 the operating band does not exist at qualities belcw approximately 75%

(see Figure 6).

The foregoing enalysis is useful, in that it provides a basis for exanining the feasibility of feed and bleed and for quantit.stively assessing the effects of uncertainties or variations in the bounding

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7' earimeters. H:ncever, it does not. ' address transient !,enavstr that may

" fiave an ir;portant bearing on the ultimate viability cf prhary feed

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and bleed. In particular, it' should be evide st thst there exists sc:e uncertainty regarding the ability to safely bring the primary coolant, system to within the -feasible" operating pressure band witN:ut sustain-ing unacceptable coolant luss in t.he process. Factors whhh beor on -

l' this transient process include'the priury coolent sistem state et

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the initiation cf Ja attempt to t eed and bleed, and the nat.ure of t.he

- t,uvient. disc.harged through the PORy(s) in ceores surizing %e system '

to within the operating band. These questions can only be addressed threugh experimentation ond the use of etnnputer code analyses.

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TACTORS AFFECT,ING,,PORV DISCHARGE v a:: .

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'.'. Of the (Actors previously diseu: cod ino largest uncertainty affecting the feed and bleed operating band arises from the influence of two-phase PORY ficw. The esss flow through the PORY is de st' ream fluid conditions at the top of the pressurizer. pendent on factors up-Several contribute to establishing pressurizer fluid conditions. The ones

.. discussed here are: transient vs steady-state behavior, primary cool-L. nnt system conditions orientation. pressuriter/ surge line geometry. and surge line -

5, ;.

travs tent, ys Steady-State Conditions and Priury Inveng If i'eed and bleed is not initleted soon after loosing the secondary heat sink the primary liquid seell will fill the pressurizer end col-le ase the steam bubble. Severti conditions may form or sustain a vapor e

bubble at the top of the pressurizer. A vapur bubble can be produced c.

by loss of pressurizer liquid inventory, heeting of the fluid to saturn-f tien, and/or depressurization. . In the present study the pressurizer heaters fore precluded. are assumed to be non-operational and direct heating is there-In a transient depressurization. liquid flashin2 in the pressurizer will tend to create a high quality region nect the top ts long es the fluid in the pressurizer is t.he bettest in the '

system. However, the liquid swell that accompanies bulk flashing will 2

tend to decrease the quality at the top of 1.he pressurizer. For either a quest-steady-state situation, er in a transient once tho original pressurizer inventory has been replaced with coolant from the hot leg.

the PORY fluid conditions are dependent upon the conditions in the ,

hot teg. If the coolant lost through the PORY is replaced by low r ty fluid the rass discharge out the PORY will remain fairly high quali-This will occur until the pri=ary system inventory is, reduced enough

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to cause significant voiding in the hot Teg. .Once significant hot leg voiding occurs pressurizerl sur .

1 etw into pley as described bela:e. ge line geocetry and orientation

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' R. E. Tiller '

PN-137-82

• 6 .

Au::ust 6,1982 Paje6

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kressurizer/ Surge Line beome_try

, For d giveri vapor vulume a pressur'irer with c large tenoth-to-diameter retto would have e " tall" voic height relative, to e. pressurizer with a sr. aller ratio. in addition to also having a smaller cross-section.

A steam bubble of greater height would tend to enhance searration from the vapor of liquid droplets created by bubbles brea' ring'through the liquid surf ace, due to the greater wall surfece area and re'.uced .

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.. potential for droplets being thryn upriard into the high npor velocity area near the PORY line entrance . linwever since vapor must by

- - ne'cessity pass through the pressurizer 11guld from the surge line to the PORY a laroe t./D would tend to promote liquid swell and dreplet

. entrainment'due to the smaller cross-sectional area.

. In any case, the influence of the pressurizer geometry may be clinimized by the preclusion .of counter-current flow in the surge line. Even if a if quid / vapor separation mechanism did exist in the pressurQer, typical surge line velocities are well above flooding limits. There-fore, the 11guid could not drain back to the loop and would continue to be stored in the pressurizer until the PORY discharge quality self-

adjusted to acconodate removal of the cacc. It theref cre appears necessary to have high quality steam supplied from the hot leg in order to have high quality PORY discharge.

h5rge Line Orientation ,.

? If hot leg voiding does occur, the orientation of the surce 'ine would influence the primary system inventory 6t which high quattiy steam Surge line-to-hot leg connections of various '

entered the pressurizer.

orientations, from horizontal side entrance to vertical tc? entrance, are used in current PWR's. With the top entrance line, and quiescent hot icg conditions, minimal hot leg voiding is necessary to ano.< high quality surge line flow. With a side entrance line the hot leg pipe liquid level

.- must drop much Tower before high quality flow begins. In either case

.o- the surge line flow c:ay.still be varied significantly if non-quiescent

'ke conditions exist that disrupt stratified f1me, such as when primary' .

recirculation ptcpt are turned on or a transient depressurtrat. ion is occurring. -

s. For typical PWR pressurizer dimensions ,the vapor velccity (due to an open PORV) in a vapor filled cross-section is on the. order of 1 1 f t/s which presents little chante of dropict entrainment.

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PN-137-82 ,

August 6.1982 .

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CONCLUSIONS BASED Oh 5114Pt trIED AtmLYSIS

.' ' Based on i.he' foregoing discussion it is concluded that a sinlified p.

approacn to chterraining the fessibility of primary fer d and hiced in a pressurtzed water reactor lies in the mapping of energy and mass f1cnes.- l l

'4oreover, this technique can be used to quantitatively assess the sensitivity of the operating pressure band ta variations in the l

[- boundary conditions of ECCS flow PORr flow, and decay heat.

' It is evident that plausible variations and uncertaintics .in these cartmeters can lead to the climination of a steady-state operating pressure range.

p. - Principal among these uncertainties is the coolant discharge through J the PORY. The predictability of this single parameter is su! Ject to b' much greater uncertainty than either decay heat or ECCS flow, l

i Irresnective of the existorice of a th' eoretically feasible operating pressure band, there remains tne question as to whether the reactor' system can be safely maneuvered into this pressure range. In this i

regara it is clear that a dependence'must be placed on computer code analyses (with suitable verification) and adequate suppcrting experi-mental data. Such analyses and/or experimer.ts should examine tne plausible scenarios which lead the ov.rator to comence prksry feed and biced, since the initial condition of the primary coolar.t system L (perticularly inventory) will have a significant effect on the cutcc:ne.

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.f u.rthermore, T it would appear to be i useful exercise to er.a=Inc the ~

eperating clap that results for each set of individus1 PWR phr4. para .

meters. The operating map represents an ultimate statement as to whether '

b, fee d and bleed is possible. end is the starting point for ext.atning specific design features that bear on the operating bounds.

AfSULTS FR0!4 SEHTSCALE EXPERIMENTS .-

1,

9. ' An experiment was conducted in the Semiscale. Mod-2A facilitr to evaluate J tystem behavice during priraary feed and bleed operatiens, figuht 7 '

' shows the primary feed and bleed operating map representing boundary conditions used for the experiment. It is seen that these parameters de. fine a steady-state operating band between 7,1 and 3.2194. Several uttempts were made to establish steady s. tate feed sad bleed within the operating hand. While it was possibic to maintain pressure control by cycling the PORY, measurements showed a continuous loss of primary coolant inventory due to a low quality dischcree out the FORY. The phenomena that led to this behavior are descri5ed below. .

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singie-phash, 15.2 MPa. At initial conditiont the pressurizer heaters were used toliqu tontrol pressure' with a small steam bubble in the pressurizer, So:e subcooling existed in parts of the loops due to the lack of natural circulation resulting from having tepty secondaries. At 15000 seconds  :

(test time) pressurizer heaters were turned off and. the SORY was latched l

'ophn. Figure B shtms that the system rapidly depressurized down to appror.bately 8 Mpa. This corresponded closely to the saturation pressure

.. trf the teldest fluid in the loops. As seen from the pressurizer collapseo liquid level curve, flashing of the hot pressurizer fiufd initially t-esulte'd in substantial voiding of the pressurjzer. Re.ferr~ing to rig'ure 9 it is seen that this is reflected in the PORY cass discharge rate.

^ Following a brief initial mass riow surge the flow out the FORy agreed closely with the predicted steam flow rate frar 1000 quality. The steam

,- bubble depleted efter approximately 250 s (rigure B) and the PORY cass ficu rate increased to appro:imately 5 times the steam flow rate (Figure 9). As seen in Figure 9, the mass flow rate out the PORY appeared to be dependent upon the conditions in the hot leg. Once substantial voiding

'^ of the hot leg occurred the flow out the PORY began to agree with the predicted steam flow rate, in spite of the fact that the pressurizer L

t emined nearly liquid full (Figure 8). (The pressuriter sur .

in the Hod-2A system is connected to the side of the hot leg.ge ) line

' At th6 time when sufficient pri. mary coolant inventory was finally lost to as to void the hot leg the core was still adeque'tely covered and .

' ' cooled. As teen in Ngure 10, there was still a small deficit in the

• rnass injected into the system with the HPIS relative to the PORY cass

~ discharge rete. The result was then a very slow continued loss of mass kbich led to eventual uncovery of the core at about 17000 seconds. -

~ '

ih5 imporf 6nce of the $em{ scale resuits lies in demonstrating the dominance y bf the PORY ditcharge rate on primary feed and bleed capability and the dependence of the PORY discharge on hot leg conditions, and consequently syste= inventory. The inability to eatntain system inventory cince the PORY hass flW rate dropped to reflect steam flow is subject to eberimental bncertainties, since the steady-state operating band of figure 7 is rather narrow. Uncertainties exist in the actual PORY orifice charactegis-tics, fiPIS injection rate and the measurement th'ereof, system heat lost ,

hnd fluid leakage.

. ths observed FORY discharge relation to hol leg conditions however, ites outside the effects of uncertainties discussed above. The questions

.[ that need to be addrested in interpreting &nd extrapolating the results

are largely related to the geometry effects. The Mod-2A system has a thort pressurizer relative to the desired scaling of L/D. The L/D effects I on liquid vapor separation A:ust be analyzed. The surge line needs to be

evaluated also, insinly with regard to the, influence of the side entry to the hot leg as opposed to other designs.

I 1

f i 'A. Core power was augme.nted to compensate for the best estimate system heat loss.

l

i'...-- :. , . ,

P. ,h.t. Tiller '

P -

pu-127-n? s -

[* :..' . A0 gust 6. 1932 '

L. Psee 9 . ..

~

4. . . ' -

00Nttt 51*JH5 BASED OM SEMISCA_LE !.XPERI]ir:gS

- In and of Ahecselves, the result.$ nom the se. ifs ate cz,<crimts do

,,p.

not point to the existente of a definitu problem regarding primary feed and bleed. But. they do tend to support & concern about the reldtive tenuousnest of the process. Further antlysis atitcpting to.

?.

7/

quantify the potenttal er.perimental distortions aric their effect on

~ the re.tults is now in progress. These analyses, along with the results of analysis of er.istent LOTT data, and computer etde calcula-

' , tionr, of 1.he Semiscale experiment and a full-scale plant,u!;I be documented in September. '

. Very truly yours, -

'*- " t

/!., sc i. & 6 C .

. .[. .

?

P. Ntrth. Nanager

' Water Reactor Research 9

., Test Facilities Dt.vtsion

• 05:ld a

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R. R. L andry, HRC . 2 W. R. Young, DOE-lD .

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lnfomalloh bbl 6$ned frcra lhe reference file from study the BE/D:

.',. . Et the Idaho Hattonal Engineering 1.abordt.ory. ~

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