ML19294C162
| ML19294C162 | |
| Person / Time | |
|---|---|
| Issue date: | 01/18/1980 |
| From: | Stampelos J Advisory Committee on Reactor Safeguards |
| To: | Plesset M Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-SM-0169, ACRS-SM-169, NUDOCS 8003070249 | |
| Download: ML19294C162 (22) | |
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E ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 0,
WASHINGTON, D. C. 20555
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January 18, 1980
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Dr. Plesset, ECCS Subcommittee Chairman C0iMENTS 04 SDilSCALE, TLTA, AND THE LYNN 30 SECTm ECCS TEST FACILITIES
'Ihis report was prepared at your request to provide a short description and coments on the subject facilities. tis information was obtained from written reports and from conversations with the NRC Staff. We comments represent opinions based on these information sources.
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ACRS Fellow
Attachment:
as stated cc: ACRS Members ACRS Technical Staff 8003070 @
- .. wy Semiscale Description t e Semiscale program consists of a continuing series of thermal-hydraulic experiments having as their purpose the generation of experimental data that can be applied to the development and assessment of analytical models describing IDCA phenomena in IMR power plants. Dnphasis is placed on acquiring system effects data that characterize the most significant thermal-hydraulic phenomena likely to occur in the primary coolant system of a nuclear plant during the depressurization (blowdown) and mergency core cooling phases of a LOCA. We experiments are performed with test systems that simulate the principal physical features of a nuclear power plant but are much smaller in volume. Nuclear heating is simulated in the experiments by a " core" comprised of an array of electrically heated rods, each of which has dimensional and heat-flux characteristics similar to those of nuclear fuel rods.
h e Semiscale system is a one-dimensional representation of a PWR. Semi-scale's original design was limited in simulating phenomena in vessel plenums, downcomer, and pipes because of its small size and one-dimensional characteristics. We facility has two loops which include two U-tube steam generators, two reactor coolant pumps, a pressurizer, and a reactor vessel.
Both hot and cold leg breaks can be simulated in one of the loops. A schematic of the system is shown in Figure 1.
%e intact loop is designed to represent the three unbroken loops of the referenced PWR. %e broken loop is representative of a single loop of a four-loop PWR in which a hypothesized LOCA occurs. Semiscale system characteristics are listed in Table 1.
We reactor core is shown in Figure 2.
We reactor vessel (Figure 3) includeu an external downcomer. Additional apparatus for test-ing Westinghouse's upper head injection emergency core cooling system can be included. We Semiscale Facility is currently funded for about $10 million in FY 80.
Coments 1.
Semiscale is toc ill. We justification of proper scaling factors to a full sized PWR may not be realizable when trying to snulate life sized reactor plants. %gpresentSemiscalepressurizerisaboutthesizeofa basketball (1.}9Ft total volume). %e liquid volume of the entire sys-tem is 6.65 Ft. % e operating loop and blowdown loop piping are respec-tively3"and11/2"indiayter. W e total liquid volume of the referenced full-sized PWR is 11,344 Ft. Phenomena such as capillary action may be important at Semicale irrespective of the best scaling factors.
For example, heat transfer in the Semiscale pressurizer may be important whereas an equili-brium adiabatic mdel for the full scale PWR may be appropriate in predictirg transients.
,. =y 2.
Best estimate computer codes are not available. Nst thermal-hydraulic computer codes used in EECS analysis are one-dimensional and sploy the conservative assumptions of 10 CFR 50, Appendix K.
A great deal of diffi-culty was encountered in obtaining best estimate calculations for the reactor coolant pump trip problem. Codes should be developed that predict transient.s at Semiscale so that they may later be used as an aid in under-standing full scale PWRs. Semiscale can be used to benchmark and develop best estimate codes since multi-dimensional effects are eliminate 1; one can use this one-dimensional facility with the one-dimensional computer code models.
3.
W e testing of the upper-head injection system in a pencil-thin reactor vessel is not a good idea. Nlti-dimensional effects are probably very important when evaluating a new ECCS spray system when counter-current steam flow up the coolant channels may be present. We non-scaled heat flux (resulting from the extra heat capacity of the vessel wallt.) presents un-realistic heat transfer. Iaidenfrost boiling may be exaggerated as the spray is expelled from the vessel.
4.
Smiscale pump flow characteristics and head degradation in two-phase flow may not be representative. %e semiscale coolant pumps are designed to create a flow path geometry similar to full scale pumps. A c mparison of the specific speed of the pumps shows that Semiscale pumps are radial flow t ereas the full scale pumps are mixed flow.
(A description of specific speed is included as the last attachment of this section.)
5.
Small break modeling may be impossible. %e blowdown loop piping is 1 1/2" in diameter. A small break scaled to this facility would be a pin-hole. Many questions must be answered, e.g.:
Will a 2-phase mixture ever appear through the nozzle for this type of break at Semiscale? Will you see a trannition from subcooled or 2-phase flow to superheated steam?
It is unknown if a small break loss of coolant transient can be performed on a small scale model.
6.
Recent planned improvements may not help this system. S e NRC Staff proposed (Summer 1979) the addition of a once-through-steam-generator (OTSG) and a U-tube steam generator in addition to a more elaborate secon-dary (feedwater) system. Rese improvements probably will not increase the utility of Semiscale since they do not address its fundamental deficiencies.
Any event witnessed at Semiscale must be interpreted with considerable caution for application to the full scale plant. % e NRC Staff did not have additional information on the upgrading for Semiscale or the instru-mentation developnent program for any of the three facilities (Semiscale, TLTA, Lynn) at this time. The latest information on the upgrading plans remains the Subccanittee minutes, summer of 1979.
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% Total Liquid Voltane Liquid Volume reactor vessel downcomer 1107 9.75 0.649 9.76 reactor vessel upper head 795 7.00 0.466 7.01 reactor vessel upper plenum 617 5.44 0.362 5.44 reactor vessel core 646 5.69 0.379 5.70 total reactor vessel 4476 39.45 2.624 39.46 operating loop (total) 5421 47.79 3.658 55.01 hot 1og 237 2.09 0.362 5.44 cold leg 255 2.25 0.312 4.69 steam generator 3231 28.48 1.485 22.33 pressurizer (liquid volume) 1080 9.52 0.484 7.28 blowdown loop (total) 1447 12.76 0.848 12.75 hot leg 79 0.70 0.046 0.69 cold leg 85 0.75 0.050 0.75 steam generator 1077 9.50 0.631 9.49 Total Oystem Liquid Volume 11344 100.00 6.650 107.22 3"
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as =wint. Row machine is best. suited for low heads and hieh d'schartes, while r
a radial-flow machine is bc<tautted for higher he de mad lower A,eh im Time parameter which is used to pick the type of pump (or turbine) best. suited for a given application is sp:ciSc speed n,. SpeciSc speed is obtained by combiams both C, and C, in such a==nnee that the diameter D is elimi==n a-C, sis (g/,ps)us
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It should be noted that the spectSc speed tr=dicia==Hy ased for pumps in the United States is de6ned as N, = NQus/AH3'* Here the speed # is in revolutions per minute, O is in gallons per minute, and.1His in feet. This form is not di= dan-less; therefore,its values are much larger than those found for n,(the conversion ihetor is 17,200). Thus most texts and references published before the introduction of the Si system of units use the traditional de6sition for speci6c speed.
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-g The Two Loop Test Apparatus (TLTA)
Description his facility is used to gather infomation on the BWR IDCA from the start of the accident through reflood of tM core. Current hardware limitations restrict the simulation to core blowdown only. Core refill has not been simulated.
We 'No Ioop Test Apparatus includes a single, electrically-heated, full-size, 64 rod bundle within a pressure vessel, and both normal and emer-gency cooling systems.
'No loops circulate water to jet pumps within the pressure vessel which also contains a steam separator. Dnergency cooling systems include high and low pessure core spray, low pressure coolant in-jection, and automatic depressurization (HPCS, LECS, LPCI, and ADS). We TLTA system schematic is shown in Figure 1.
System blowdown is initiated by actuation of quick-opening valves in a recirculation line.
We current TLTA facility is owned by General Electric and cost about
$1 million. We current test program (Blowdown /ECC Injection) using the TLTA is jointly funded: 1/4 from General Electric,1/3 from the Electric Power Research Institute, and 5/12 by the Nuclear Regulatory Commission.
W e program authorizes $10 million over a five-year period ending in 1981.
Comnents 1.
Some shortecznings of the presant facility are important when performing a small break transient simulation (recirculation line break). We jet pumps in the current facility are too short. Eis limitation does not allow for full reflood capability (see Figure 2).
Because the jet pumps are too short and because of the scaling of the annulus enclosing the jet pumps, this annulus must be overfilled to support the core water height prior to the start of the accident. Were are additional problens with the scaling of the plenum below the core. We jet pump diffuser's are too short.
When the plenum level drops sufficiently steam can vent through the jet pumps bypassing the core. These difficulties are important in a review of general perfomance. Computer code predictions have been used to modify the apparatus so that it has the same timing as in the real life situation.
2.
Good computer codes are not available. General Electric uses a better estimate group of codes (no + 20% ANS, otherwise an evaluatien model) to predict transients in TLTA. she NRC uses 'ELAP-4/ MOD-6 for best estimate evaluations.
(RELAP has orgy been anployed to calculate blowdown - no refloor, or ECC injection.) The NRC Staff indicated that REIAP-4/ MOD-6 is a " state of the art code" but probably not very good.
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Modelirg of the fuel rod heat flux and fuel rod heat capacity may be unrealistic. Se current TLTA facility has directly heated rods.
'Ihe current ;essing through the rod claddirg heats the rods.
4.
Se valve control system does not operate well and a programmed closing sequence would not produce repeatable results. S e steam line pressure control valve (current TLTA facility) simulates the closing of the main steam isolation valves. Se valve shuts immediately for a large TACA. Se small break scenario calls for the valve to shut 30 seconds followirn 'a break initiation. S e backpressure effect of the main steam on the core is important since it reduces the swell level of the core and higher steam pressures should increase the break discharge rate.
5.
Se proposed trodifications for TLTA include:
three full size (8 x 8), full length electrically simulated a.
bundles with typical channels and surroundirg bypass region; b.
functional jet pumps mounted on extended tailpipes to provide full core reflood; indirectly heated rods (heated by an internal filament), and c.
d.
improved modelling in the steam separators, upper and lower plenums.
W e modifications are further described in the attachment.
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Background
At the outset of the p%a.a there was a desire by all parties to run ocuplete
- Ibwever, integral tests - fran time of break occurrence through bundle reflood.
er of technical problerts identified with such a test for seich there were a rnx
'Ihese problems were no satisfactory resolut2on could be fcux! at the t2me.
generally asscciated with (1) the short jet px:ps and the need to unintain the proper distribution of liquid in the downccrer, and (2) the unavailability of a heater brxile design with 1:oth rated heat flux (at treak initiation) and high tar.perature capability (prior to renW).
swia.a plan was thus develepa$ to address the ICCA by time phase 'using
'Ibe t:bo test configurations with scne charx3es in koth the bundle and jet pmps tcing
'Ibe first phase, called ED/EX:C-1A, was to made tatween the two test phases.
address the initial blowdbwn, flow coastdown vr1 early spray interaction period.
III injection period out to and
'De secorr$ phase was to examine the langer terh: Separate effects tests were also planned to including reficed of the burrile.
ate the effects of an alternate axial power shape bundle (in the 1-B phase), and to evaluate heat transfer in a nore typical gecnetry including a thin-valled channel Provisions were also rnade for 'a with a concentric core bypass region (BD/EXI-II).
slert test phase to evaluate blowdown /EXI perfornance in earlier BWR gecrretries (i.e., ET s without jet ptnps).
Since the fo=:ulation and approval of this plan (early in.the progra:n),
new infornation has teen developed. Data fran the 8x8 BmT and BD/ECC-1A test phases indicate that the short jet pt=ps have a significant inpact an fluid distribution and thus intzt: duce scr:e atypicality of perfornance in the 'ILn.
Additionally, theoretical scaling and systan perfarnance evaluations have further (1) parallel channels with scme nulti-dinensicnal errphasized the need for:effect cim1ation in the upper plenun, (2) nere prototypical hea the burrile/bypasn region, (3) inproved r=#1W gecr:etry with full height static Ivwh, ar3 (4) cc:plete, integral experiments cavering the entire IDCA transient.
Such censiderations led to evaluation of a ranter of alternate cx:nfiguraticns (1) an external path in W iel with the bundle and/or a partitioned including:
bundle to evaluate potential parallel channel effects, (2) alternate heater arx5 burrile designs to provide nere typical tundle to bypass heat transfer, (3) a two-piece downccrner regian to allow full jet ptrp reflood height in conjuncticn with yw liquid inventcry distribution, and (4) canbinations of the almve.
'Ihese cancepts have been reviewed by the PfG and their relative merits have baen 64 mmca.
Separately, during the ER Refill /Mficni Progra:n ccnsiderations, a certmit-
, ment was made by all parties to consider incorporating a parallel channel
'!his cx2:mitnent, in addition to the above capability into the BD/III Fi@ia.u. described sinulation inprovsment needs of the T of an inproved gecnetry for the TLE in the BD/III-1B phase.
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Ctnsidera+hus and Criteria
'Ihe following features or capabilities were considered essential:
integral blowdown /Er injection capability frcrn tire zero to reflood, 1.
2.
full sized, full pcNer heated bundles, 3.
typical bundle / bypass heat transfer, 4.
full height reflood gecr:etry.
An additional constraint inposed was that the total cost would ret exceed the Other features identified as high "mnts" current contract funding limit.
incitx5ed full height in the lower plenun and prototypical b.Indle inlet and outlet ccnfigurations.
Two basic options were considered:
(1) constructing a new maltL%~dle facility near the ATIAS power supply, utilizing mach of the ATU6 Irop equi;rrnt, and (2) ncdifying the TLTA in its present location. A new facility adjacent to the AHAS Heat Transfer Icop was ruled out because it would far exceed the The latter option has been selected as our recomenda-inposed cost constraint.
tion, and is de.scribd below.
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'Ihe TL'IA configuration rea:rrended for the BD/ECC-1B phase is as follows:
General - Ctrplete integral blo,down/Er injeion capability fran break ini-tiaticn through core reflood. Figure 1 shws the general layout of the vessel and interrals.
Ore Regicn
'Ihree, full size, full p>er (typically peak, average, and low 8x8 bundles with prototypical channels and a surreurding bypass region.
power)
See Figure 2.
Jet Pu::ps - Functional ptrJps ntnmted on extended tailpipes to provide full he.tght core reflood capability.
Vessel /Downocruer Region
'Ihe vessel muld te approx 1rately be inches larger in dianeter than the current vessel to accordate the additicral tundles. 'Ihis swdoning provides a scaled cross-secdonal area and allcus the "norral" water level to be at full reactor height.
Up er Plenun and Stem. Separator - A full height upper plenun and steam separator would be used. Baffling would be provided in the upp plenun with separate Hr injecticn capabilities in each zone above the brdles. A prototypical stea:n separator my be used in place of a scaled separator.
Iower Plenu:n
'Ibe icwer plenan w:uld be approximately 50% of full height with scale:1 volumes mintained in both the open and guide tube regions.
Beaters and Bundles - An ind.trect (filament) beater rod design wvald be used as slown in Figure 3.
A cosine axial guer shape would be obtained by the interior tapered wall element. Current return through the outer sheath eliminates the need for heavy tuswork ateve the bundles ard allows use of prototypical upper tieplates and other hardware as s.%wn in Figure 4.
Prototypical grid em will also be used in the burdle. Figure 5 sh:ws a typical bundle inlet cxmnecticn and also shows the p>er cxnhons. Theum. couples are inMMed in ~
the outer sheath and the 'IC leads would also be taken out at the LutLL of the hardle.
TL TR
The heater rod design is very sirrilar to a design developed for AHAS test-ing. HNever, minor changes will be made to u: prove the ease of fabrication.
Furthernore, this design has not been used previously for this particular appli-caticn, so sme qualification testing will be done.
Utilization of Dcisting Eqai; rent Existing egai; rent will be use extensively. '1he TLTA facility will be used ircluding the basic structure (cnly mirer nodifications regaired), water and power services, controls, data acqaisition systs, feed. rater systm aM ECC syste s.
'No acMitional SCR's at the AHAS p:w_r supply will be cmW to the existing pwer cable. The additional pwer regaird will be tranmitted by operating the existing tranrission cables at a higher voltage potential (plus and mnus 180 volts, respectively - as ccrpared to the present operation at zero and 180 volts).
Cost and Schedale The schedule f r irple enting the m1tibundle n.TA, sFhn, testing, and data evaluation is shwn in the tar-chart of Figure 6.
tbte that the schMule for ccrpletire the p2%2am is noved frur, the first gaarter to the third gaarter of 1981. Table 1 prwides a breakdown of costs for the nultitundle configuration. Table 2 indicates the a:nmt and disposition of runining program funds. An apprc:>ti rate cost breakdown by gaa:ur is provided in Table 3.
Trade-offs The pvpused nodificatims wald significantly irprove the reactor simitation accuracy of the BD/dr experirents. As indicated in Table 2, the rtralrung aan-tract funds are suf ficient to ccrplete the facility rrr"fication ($3.03M) and also to ccrplete an effective test series ($1.40v.).
7t> allcv for these improved sirulation experirents within the existirs p2%2am bdget, the separate stuiies cn alternate axal power shape and non-jet p=p plant configuration will be eliminated.
H3mer, the total p2w2am gaality irprove-ent strongly favors this trade-off.
03nclusion
'Ihe above rectruendaticns will address all of the substantial concerns identified with the current TLTA. kloption of these rewa.a-dations will greatly enhance the value of the BD/KC Program.
(This proposal is taken from a letter written by G. W. Burnette, GE External Programs Manager, to W. D. Beckner of NRC's Division of Water Reactor Safety Research, and dated March 30, 1979) l A TE S T ///Fo As oF
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