ML19296B196
| ML19296B196 | |
| Person / Time | |
|---|---|
| Issue date: | 10/19/1979 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-1670, NUDOCS 8002200242 | |
| Download: ML19296B196 (101) | |
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~: ATE ISSUED: 10/19/79 MDIUTES OF THE MEET 2G CF SIE
- CES SUBCCMMI7"ZE Of WI%TCY CCRE CCCL2iG S*.GTSS AUGUST 27-28, 1979 IDAHO F.'-J LS, IDAHO "he ACRS Subecemittee en hergency Core Cooling Syste-s met in Idaho Falls, Idaho, to develop information for consideration by the ACR5 in its review of NRC Research Pregrams.
In particular this meeting was devoted to revies-ing ECCS related research programs relating to LCET and Semiscale and to development of the RE!.AP and SEACQi codes. W.e notice of d e meeting appeared in the Federal Register of August 10, 1979. A copy is included as Attachment A.
?.e Ccemittee received no written request to make an oral statement nor did it receive a written statement. A list of attendees is included as Attachment 3.
Te.e Subcomittee did not prepare any decments during the meeti.g.
AUGUiT 27, 1979 Dr. Plesset, Subecmittee Chairman, called the meeting to crder. He explained the purpose of the meeting and identified the principal parti-ci : ants. Dr. McCrele'ss was named as the Designated Federal hployee.
Dr. Plasset recuested de censultants to prepare written ecmments on the topics to be discussed and to submit thent to Dr. McCreless.
Dr. Plesset discussed some concerns of members of the ECCS Sebecmmittee regarding de research program for Semiscale. He stated that the Sub-committee recognites that Samiscale is a verf useful and important facility that has, on occasisn, been misused in the past. He said that this misuse was not so much. y researchers but more likely by cecple in-volved in licensing wherein tr.1 facility is regarded as a valid integral facility rather dan a facility where ene can study important effects.
Dr. Plesset said that he was pleased and favorably impressed by seme comments dat Or. Teng had provided in a letter in which he said that particular emphasis was to be given to de scale effect prior w a Semiscale test.
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ECCS Meeting 3/27-28/79 PRESENTATICN SY CR. LONG SUN TCtG Or. Tong of the NRC Staff said that the ACES recorrr.erdatien (to Chairman Hendrie en May 16, 1979, that safety research on the behavior of light water reactors during ancraleus transients te initiated as seen as pessible) is new the goal of NRC research. He explained -he secpe of small break and reactor transients research. Bis is to include re-search on behavior of GiRs during anomalous transients including postulated and exploratory accidents and the development of capability to simulate wide range of postulated transients and accident conditions including small, medium and large breaks.
Dr. Tong said that NRC can not wait to construct new facilities or to develop new codes bu: must make use of existing codes and facilities.
Dr. Tong said that RES has divided its 'erk effort into three tasks; namely: general approach, scaling criteria, and test plan. De approach to small break and reactor transient research is shown in Attachment C.
We approaches has been divided,into engineering analysis and experimentation. D e cbjectives are also indicated. We interfaces between engineering analysis and experimentation are included.
Dr. Tong said that both the NRC Staff and its contractors must concen-trate on the analysis and understanding of the physical phencmena and not on Jreparing eceputer input and reading ccmputer printout.
Dr. Tcng discussed the scaling criteria. Bis infocnation is included as Attachment D.
Bis attachment indicates the physical phencmena to be preser/ed and the parameters to be controlled for toch small and large breaks.
Dr. Tong cemcared small and large break *CCAs. Bis infor atien is included as Attachment I.
Sample break size, significant heat scurce, significant heat sink, heat transfer in steam generator, primarf si'de pressure, flow behavior in priracy side, ICCS, and plant recover / are all indicated in this attachment.
i ECOS ?.eeting 8/27-29/79 t e necessary modification to de Semiscale test facility in order to acccc:nedate the small break test are shown in Attachment F.
Dr. Tcng presented a bree-dimensional plot of end state of an acci:en as a functicn of total break side and failures or cperator errors. Sis plot is included as Attachment G.
Dr. Tong discussed de testing plan.
Stegrei system tests are planned to test natural circulation, primary and secondar/ loop cooling echanisrs, symptoms of small break and reactor transient, plant recovery techniques and nuclear feedback in reactor transients including ALIS. Separate effects tests are planned for pressuri::er and relief valve. cperation, heat transfer in the steam generator, heat transfer in uncovered core, ficw blockage in core, two-phase ficw patterns, and valves. W e planned tests and the test facility are shown in Attachment H'.
We pressure capability of each test facility is also included. Se last four columns represent foreign test facilities.
Dr. Tong discussed code availability for predicting reactor transients and small breaks. Bis information is sur:rnarized in Attachment I.
Be code, lab, availability of code and strength and weaknesses of the codes are included.
In respense to a questien from Dr. Plesset, Dr. Tong explained that the E.CBI test facility in Gecnany is slightly larger than Semiscale and has 64 reds, as compared to 23. As in Semiscale the facility uses electrically heated reds.
ECB: has a steam generator however.
Dr. Tong said the.
Germans have premised to try a small break experiment but that a test date has not been see.
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a ECCS Meeting 8/27-29/79 Dr. Plesset asked if it were possible for a small break to eccur that would provide a LCCA and the pressure remain up encugh so dat high pressure injectien is not called for.
Dr. Teng said dat dis will be investigated.
Dr. Catton noted that tne general approach taken by Research is better but he recemmended in regard to scaling criteria that similitude analysis should be undertaken since geccetrical similarity is violated. He said that a more detailed investigation including sensitivity analysis is needed.
Dr. Theofanous. cautioned that he believed the RES effora should be focused and that perscnnel frem Semiscale should do analyses as well as experimentatien. He recommended that RES have at least three people to develop comprehensive and cohesive research plans.
Dr. Lipinski recemmerEfed dat automatic actions to prevent core damne should be made automatic. Dr. Theofanous noted that the location of ECC injection and resultant fluid mixing ara stratification effec *a could be extremely important in scaling aspects as system size dif,fer. Dr. Tong indicated that this should be dene in future scaling sttdies.
PRESDTTATICN BY CR. ERNEST WILKDIS Dr. Wilkins welcomed de Subecmmittee to Idaho Falls. He identified other programs at DIEL dat de ACRS might like to discuss.
Cf dose that are related to de NRC, he explained that de dermal fuel, cede assessment, 3-D and licensing assistance programs might be of interest. Cf de TE programs, he suggested that de ACES might be interested in de Test Reacter, Waste Management programs and in non-reactor programs en geothermal, hydrepewer and fusien. Dr. Wilkins cuoted de statement by Dr. Ybarrendo last year in which he had said that NRC and de rest of nuclear ccmmunity should reassess de objectives of the LCCA-ECCS program in order to estaclish a well defined end point for de LCCA-ECOS
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ECCS Meeting 9/27-2949 work.
Cr. Wilkins said dat this basic philosophy is still true today.
He discussed the acccmplishments of DIEL in providing analyses as re-quested by NRC resulting fr:m de ""aree Mile Island 2 accident.
PRESENTATICN BY CR. N. C. KAUF? DAN Dr. Kaufman, Director of LCET, provided a general overview of the water reactor safety program at DIEL. De characteristics of.the program are included in Attachment J.
Cr. Kauhan also discussed the characteristics of large scale experimental programs and DIEL's involvement in industry factions. These are shewn in Attachments K and L.
Dr. Kaufman reviewed the LCFT Test Program. He explained dar the testing schedule of 1978 had been accelerated by about 6 mont..s as a result of the data obtained in Test L2-2.
It was decided that Test L4-1 was not necessarf as this was another test to be run at 8 kilo-watts per foot maximum linear heat generatien rate. As a result of the Bree Mile Island accident and of de similarity of L2-3 with L2-2 the program was changed substantially. He discussed the signi-ficant changes and they are. included on Attachment M.
Test serie's L3, i.e. small break test program, was started with Test L3-0 on May 30,1979.
Dr. Kaufman said that DIEL will emphasi::e the entire accident scenario and not just the decompression phase.
me LCET 3-year test plan is shown in Attachment N.
In response to a question frem Dr. Lipinski, Dr. Kau ban explained dat the tests have been selected so E.ct ene dees not necessarily rely en de results of another.
FCISES'Ir!CN SY J. H. LDTEBARGER
.ir. Linebarger discussed the analysis of LCET L2-2 and L2-3 Tests.
He compared dese two experiments. He said dat both experiments w re
ECCS Meeting 9/27-29/79 blowdown dcminated, i.e. the peak clad temperature occurred during blewdown.
3cth exteriments wre self limitirg, i.e. water flowed dcwn core before ECC water entered and caused early rewet of de core.
~n both tests, de core der =al respense was very tightly ceupled t: de hydraulics.
- n both tests essentially the same sequence of events occurred but the timing differed. Be initial conditiens of the t*,e tests are shown in Attachment O.
Mr. Linebarger said that the difference in results between the two tests is principally due to the difference in power between the two tests.
A a: vie was shewn dat depicted as a function of time the predicted and measured cladding temperatures versus core elevation.
In :esponse to questien from Dr.. Catten, Mr. Linebarger explained th:t to maintain the same T between the cold and hot legs, it is necessary to run the intake pu=ps,at higher speed for L2-3 because of the higher power.
Mr. Linebarger discussed the typical cold leg mass flow as a function of time. A plot of this info:mation is included in Attachment P.
Mr. Linebarger explained that a.ecurce of fluid for early rewet is f::ca the intace cold leg, prior to ECC, and he suggested that the desirability to augment this source of coolant with what he called "self-initiated core cooling" he considered.
Mr. Linebarger compared the ECCS information for the t.c L2 Tests and L1-5.
He noted that delivery of fluid to the core for final _ quench during ECC is not dramatically changed. Bis ccmparative inferr ation is included in Attachment Q.
He explained Sat for de less-of-ccolanc accidents :.:n to date dey have been blowdown dominated and self-limiting. He said dat during blewdcwn, dermal response of de core is tightly coupled to hydraulics and during bicwdewn the phencmena of relative magnitudes and timing are con-sistant, with de initial core p:wer and C cf de cc.re.
ICCS Meeting 3/27-28/79 Mr. Linebarger discussed de prototypicality of the LCF"' test. He compared the LCET data to a REIAP study en the Zion plant. He ex-plained that de Zion data were scaled by the total volume ratic cf de two systems. Or. Catton suggested that the proximity of ce loops in a four loop plant such as Zion to de broken leg might affect de data when compared to LCET with three leops lumped 190 F frem the break. Mr. Linebarger c.. cluded that for the tests run to date, the LCET results realistically, if not conservatively, scale the-dominant hydraulic phencmena during blewdown and provide a realistic, if not conservative, indication of de fuel clad temperature.
In response to.a question frem Dr. Plesset, Mr. Linebarger explained that the Zion calculation showed early quench.
(This was discussed in greater detail in a presentation by Mr. Ralph Nelson.)
Dr. Catton said that he thought the movement of the stagnation ~ point would be dependent en'the hot leg flew resistance, pump flew and core resistance.
Mr. James White said that de :u=p speed and broken 10cp resistance will be varied between IfFT tests L2-3 and L2-5.
Dr. Catten noted that REIAP is a 1D code and some aspects of the problem require a 3-D treatment, such as ficw in de annulus. He said that detracts from making very strong conclusions.
Mr. Linebarger said that the driving force is so streng that it keeps things going in a ene-dimensional way.
FCISDN"ATICN SY RALPH NELSCN Mr. Nelsen discussed the analytical results frem the L2-2 and L2-3 tests.
He exclained that he would discuss the various aspects of how these calculations do or do not predict rewet. He noted that L2-3 was Standard Problem No. l'J.
He presented a plot of upper plenum pressure
.m o m
ECOS Meeting 3/27-23/79 and of broken 1 cop cold leg flew as a 61.netien of time after :upture.
Bis information is included as AttacMents R and S.
He noted that in both ccmparisens dat REIAP-4/MCD-6 predictions were de best.
Mr. Nelsen also presented a plot of L2-3 test predictions versus actual results. 21s is included as Attachment T.
He concluded dat de ccmparison of the hydraulics of RFJ.AP to dose which existed in :.2-3 experiment are very good. He said that dey feel those predictions are valid.
A movie was shown. A 3-D heat transfer surface showing wall temperature, heat flux and fluid quality was depicted as a function of time. Mr. Nelson explained that the movie showed why RE!.AP-4/MCC-6 did not predict rewet in LCET L2-3..He explained the boiling curves used in GAC and TAAC clus Ilceje. He concluded that hydraulic phenomena observed in L2-3 and L2-3 have been adequately predicted with REIAP-4/MCD-6 ard that L2-2 and L2-3 types of rewet can be successfully predicted by several different methods.
He said dat he has been successful in this prediction by changing Biasi
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in REIAP and Iloeje in TAAC, but that he has not decided the proper mechanism or phencmenon to support the changes.
Dr. P.esset requested a written surmary of the information as presented by Mr. Li: ebarger and Mr. Nelson. He explained that he thought this informa-tion would be of great interest to the ACRS.
Cr. Kaufman said cat he would provide the su= mary.
PRESclTATICM BY D. 3. JARRELL Mr. Jarrell discussed de p:wer operated relief valve (PCRV) isothemal transient tests on May 31. He explained dat this is not a direct simula-tion of te TMI incident. He caid te test cbjectives were:. to check instrument adequacy, to benchmark de ccmputer modeling and to determine needed cperator training. He described the initial conditiens for experiment
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o ECCS Meeting 8/27-29/79 L3-0.
Sese are listed in Attachment U.
He also described de system configuration. This is shown in AttacMent V.
Mr. Jarrell characterized de experiment by examining the pressurizer level. He explained that the pressuri er level fell from the time of de break to about 47 seconds and that at this time de system pressure reached saturation and that vapor formed in de primary system. The vapor formation rapidly raised de licuid level in de pressurizer to beyond its indicated range. ':his occurred between 47 secords to about 73 seconds after the time of rupture. Beycrd 1400 secords, the pressurizer level reentered the indicated range and shewed a great deal of tsc-pnase activity.
The level centinued to fall until the experiment was terminated. Censito-meters in de hot and cold legs indicated that vapor generation in de cold leg hacpened about 70 seconds and in the hot leg about 170 seconds into de transient. A plot of the pressurizer liquid level is included as Attachment W.
He noted that an adequate supply of water to es pressurizer did exist frem units with elevations higher than that of the pressurizer.
Mr. Nelson compared the blind pretest predictions with measured primary system pressure, collapsed licuid level, ard pressurizer liquid level.
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These ccmparisons are shewn in Attachments X, Y and 2.
Dr. iheofanous cautiened that care must be exercised in decidirq what is conservative and what is non-conservative.
Cr. Fabic said that the liquid level ccmparison with measured level is like ccmparing apoles and oranges.
Cr. Theofancus was concerned about de pressure difference between de measured and predict,ed values.
Mr. Jarrell attributed $1s to de nedalization problem and break ficw =cdeling. He exclained dat this problem does not exist in de cede per se but in de way de code was secuo ard run.
Cr. Kaufman attributed it to de uncer ainties in knowning de break characteristics.
Cr. Theofancus expressed his concern. He said that he considers de disagreement is very significant and noted dat if significant difference occurs with minor perturbations de cuestien naturally arises as how much of a difference can be expected wid measured perturbations.
ECCS Meeting 3/27-28/79 Mr..Jarrell drew de f 11 ewing conclusions from the L3-0 experiment.
He concluded that de pressurizer filled ccepletely and thus gave de operator no indication for scme 13C0 sacends during de-transient.
He concluded dat de available instrumentation characteri:ed rajor phenomena.
He concluded that the computer rodels predicced the major phencmena that were ac ually enceuntered. He cencluded ~. hat the improved pressurizer instrumentation was needed for Test L3-4.
Dr. Catton rececmended tat censideration be given to de length of de surge line.
PRESENTA"!CN SY T. K. SAMUEI.S Mr. Samuels discussed LCET experiment planning. He explained Sat the test cbjectives were to provide assistance to NRC in defining the s all
. break accident scenario. He explained that the L3 series will censider break flows that are greater dan, less than, and equal to de high pressure injection system ficw.
It will also include stuck open PCRV valves. Cther EcET test objectives are: to provide data to evaluate cceputer codes, to identify unexpec*ed phenemena, to investigate typical instruments as used in pressurized water reacters, to evaluate plant recovery methods, to investigate operator actions, to investigata safety valves operation and to study the impact of non-cendensible gases.
Mr. Samuels discussed the LCET scaling criteria. He explained that de high pressure injbetion system (HPIS) will be scaled to de ratic cf *.CE" hcwer to dat Of a typi.:a1 EWR. He noted that dis was volume scaled for the. L2 series but.dat power scaling will decrease de HPIS ficw and is a more severe accident. He said dat de accumulator liquid will be volume scaled and dat de accumulater gas will be maintained at $e gas /licuid ratio. He explained that the low pressure injection system will be scaled
ECOS Meeting 9/27-23/79 by the core flood area. He said that de core power for LCF" shculd be run at 73 megawatts for volume scaling but noted dat maximt.n power for LCET is only 50 megawatts. ne break. area will be volume scaled.
The test conditions for de L3 series are shown in Attactent AA.
Se break location will be in the cold leg for three tests and in de PCRV for the fourth test. We break size, location, and flew is shown in Attac.5ent AB.
Mr. Samuels presented a plm. of crimary and secondary pressure as a function of time resulting from experiment L3-1.
This information is included as Attachment AC.
Se break size in experiment L3-1 was 0.1 sq ft.
For this experiment there was no auxiliary feedwater.
The conditions for experiment L3-2 are included as Attachnent AD.
In this experiment both the break flew and HPIS ficw are about equal during, de saturated portion'of bicwdown. Mr. Samuels presented a plot of de REIAP data for eis experiment. A:;ain there was no auxiliary feedwater available. Se plot is included as Attach:ent AE.
Mr. Samuels explained that for transients where 2e HPIS ficw was greater than the break ficw during the saturated portion of blowdewn, the primary system would not repressurize until it went liquid full. He explained dat the concept was later checked with Cembustion Engineering and Westing-house and in both cases Seir analyses confi:ms the same infor. nation. He said dat presently DIEL is replanning this particular experiment.
Mr. Samuels discussed the planned ccnditiens for Experiment L3-4.
See Attachment AF.
Plot of primary and secondarf system pressures as cal-culated by REIAP are shown in Attachment AG.
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ECCS Meeting 8/27-29/79 Mr. Samuels discussed de use of de tests data. He said that 5ey would use the data to evaluate small break scenarios, to assess cedes and to evaluate current instru=ent ability to track a mall break.
Mr. Samuels discussed plant recovery operation. He explained cat during saturated blowdown it was deir objective to maintain the steam generator He noted th' t no operator action water level above de top of the, tubes.
a was required for Experiment L3-1.
He said that they plan :o manually initiate plant cooldown at 3600 seconds for Tests.L3-2, L3-3 and L3-4.
It is planned to initiate steam flow at 36C0 seconds while maintaining steam generator water level. The ecoldown rate will be limited to 70-90 F per hour.
He explained how decay heat is being treated in LCE.
With the exception of those instrument changes noted in Attachment AH, the existing instruments installed for Test L2-3 will be used for,small break analysis.
Mr. Samuels summarited the small bre'ak test program for LCFT. The major features of the program are shown in Attachment AI.
A comparison of the small break site for the various experiments being conducted in I4N, Semiscale and in the audit calculations is shewn in AttacF=ent M.
Mr. Samuels discussed de LCN operational transient tests. ':his series is known as de L6 Test Series. He explained dat ce purpose of de test series is to provide data en the dermal hydraulic behavior for use in evaluating and assessing. computer codes. The tests will also be' used to evaluate overall plant response and severity of anticipated transients with partialar emphasis en be most probable events. He cutlined scme censiderations for transient testing. :+a identified available ecmputer codes; namely, LCN hybrid, German "ALv.CD" and EPRI R.5"aAN.
He said
E005 Meeting 8/27-29/79 that DJEL does not have the RETW Ccde at this time. He indicated dat plant availability for LCCE has necessitated that plant da= age in de L5 series be a minimt:m and that recovery rapid. He said that other con-siderations for transient testing include ce availability of proper instrumentation, safety analysis and test harJware.
Mr. Samuels listed scce advantages and disadvantages of the LCFT L6 transient test series. 2ese are snown in Attachcent AK.
He said that the anticipated transients to be analyzed are: loss of feedwater, centrol red widdrawal at pawer, excessive lead increase (feed ficw matching steam flow), less of primary flow with coastdown, less of steam load, and uncontrolled boren dilution. He revided a su:trary scenario of four transients. tese were: the loss of primary flow with coastdown transient, less of feedflew transient, loss of feedwater transient and centrol red withdrawal transient. He presented the calculational results. for these four transients as decennined by the LCFT hybrid code.
PRESENTATICN BY 7. K. CHIR Dr. Chir reported en the experimental results of finned and unfinned der:.ccouples to affect cuench. He explained that the experiments were supported by EPRI and were c:nducted at UCLA. He said the views that he expressed were entirely his own and not 2ose of IFRI.
He dascribed the test apparatus that were used in his experiments. He said that ce system can cbtain very high temperatures with induction heating and has the acility for visual observation. Attachment AL shows de cuench front behavior as a function of ficodi:h velccity for both finned and unfinned rods. With velocities of greater than 10 centimeters per second the scalleged shape was cbserved. He presented resula of a nt=ter of experiments with Sermoccuples mounted in different locations.
He concluded that de presence of external ther=ccouples does induce faster red cuenching.
ECCS Meeting 3/27-28/79 PRESENTATICN BY R. C. G D G Mr. Gattula discussed the effect of the thermccouples being :rcented en de cladding surface fer de LCE" fuel reds.
He said the pu pese of de
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study was to detamine if the measurements are valid and -aid =*1a 2ere are adverse effects en core behavior because thermecouples are ex-ternally mounted. He described the tyoical attachment of thetzccouples on the LCET fuel. He showed some results frem tests. He concluded that the LCFT surface dermoccuples measured cladding temperature to within 20 C except for short durations during Ch'd and quench. Measurements of peak clad temcerature were wichin 5 C during bicwdown and no large time difference between rods with embedded thermoccuples and those with surface dermoccuples, as in LCET, ducing CNB and quench.
Mr. Gottula described seme analytical computations made with a finite elenent code called CCUPLE. 'Ihe calculational results indicated a difference between embedded and surface ther=ccouples of less than 20 C during film toiling and 50 C at !%B and quenc.k.
- He said that there seems to be no reason to believe that external thermoccuples have a significant effect en early quench in LCET tests. He explained that dese results were discussed without disagreement during an April 1979 meeting in Cenver, Colorado, in which rewet phenomena was discussed. He acknowledged however, that scme lead test rod data at PSF indicates that under some cir-cumstances there may be an effect. Mr. Gottula provided a sum.ary of programs to evaluate de effect of LCET external termecouples en cuench behavier
~tese are shown in Attachment AM.
Mr. Gottula said that while Cr. Chir snowed scme effect during low flooding conditiens, dat INEL does not feel there is any reason to believe dat a significant effecc will exist under conditiens sUch as in LCE" during blewdewn.
ECCS Meeting 8/27-28/79 PP6.SENTATICN BY NCRTH Mr. North discussed scme cede developmen: efforts a-D:EL. He said dat REIAP-4A'.CD-7 is an equilibrium, hemegenecus, one-di~ensical cede and dat MCD-7 is the propcsed final version. He explained dat de new version is intended to circumvent scme of the 1 imitations cf de equilibrium, hemo-geneous assumption model as it allows some st.bccoled ECC injection. Se new capabilities of the code include self initiali:stien and centinuous LOCA calculatien from blowdown to reflood, He said the self initialization model is now working and that renedali:ation is being checked cut. 2e ccepleted code is expec ed to be ready for release by the end of 1979.
Mr. North discussed the WRAP Code. StP stands for Water Reactor Analysis Program. His is a Code that is a collection of ecdes intended to do licens-ing calculations. We objective of the cede is to provide licencing style LOCA analysis capability. REIAP-4 is the basic vehicle. Many other codes are coupled with this REIAP Code. Se actual assembly of the cede.cackage will be done at Savannah River laboratories. Mr. North explained that the models have been provided to Savannah River taboratory and checkout and demonstration is currently in progress.
Mr. North discussed the ':RAC-B Code. "his Code is a non-hemcgeneous, non-equilibrium, advanced code. B indicates that the cede is adaptive to a SWR.
We objective of de '!RAC-3 Code is to provide a best-estimate ccde of LCCA event for a ShR which is capable of analyzirg cperating transients. He said de development is just getting underway. He identified the phenctera of importance to a ShR design basir, accident and this infoc ation is shown in Attachment AN.
He explained dat these phenemena need special attention if an accurate ShR analysis can be made. He said dat such develegment will be coordinated vid CASL, GE, and EPRI.
Mr. North discussed REIAP-5. "his is a non-e<iuilibrium non-hemogenecus analysis tool aimed at transient analysis of the primary system. PIIAP-5 is a best-estimate computer program.. A '/ersion of de cede has been re-leased to de National Energy Software Canter. Additional developtent is underway.
n ECCS Meeting 3/27-29//9 De SEACCN Code is a non-scuilibrium, non-heecgenecus de: mal-hydraulics code for analysis of de contaiment syscem.
Mr. North explained dat de objective of de program is to provide a short-term, best-esticate containment transient cede that will allow understanding of conservatisms in licensing calculations. He noted dat substantial gradient may exist between varicus rocms in contaiments. He said that the structure at the beginning of a transient is cold and has a potential for heat transfer.
He explained dat the steam ficw passage may be very imccrtant in accident analysis. Hesaidthatthepreliminaryversionof$.hecodehasbeen released to de National Energy Software Center and that de final version of de code is expected to be ready by tne end of 2e year. Checkout and testing will follow.
PRESENTATICN SY S. R. BEHLDIG Mr. Behling described in greater detail the DIEL work en WRAP, REIAP-4/
MCD-7 and mI related, support activities.
Mr. Behling explained that for de WPAP Code, the cbjective is to incorporate all Appendix K requirements into a single computer pr: gram.
DEL will checkout the various pieces to assure that they are correctly incorporated by Savannah River. Mr. Behling presented de analysis scheme for WRs and ShRs. tese are shown in Attachment AC.
"he DEL schedule calls for the mR demonstration to be ecmpleted by Septerter 13, 1979 and for EhR check to be completed by September 25, 1979 with de-monstration to be c:mpleted by Septerier 30, 1979.
Mr. Behling discussed the DiEL efforts relating to REIAP-OHCD-7. He explained dat de new medel is intended to everccme shortcemings of MCD-6 wie dese major changes: subccoled liquid can co-exist wi e saturated vapor; i=preved vertical slip medel; improved reactor kinetics; new ANS decay heat standard in de decay heat medel and a new critical heat flux correlatien.
D
ECCS Meeting 9/27-28M9 Mr. Behling described the SI-related sucport activities cenducted cy IEL.
He divided the support into a study of natural circulation and cf transient analyses. He explained that 2e natural c'- **cn study censisted of core bleckage st'Mies, a steam generator study, a HPI-letdewn st'My, and a change over to natural circulation study. Calculations were done with PSI.AP-4.
Forty calculations of core blockage and various steam generator cperating conditions were conducted. None snowed core stagnation. 2e HPI-letdown study was cenducted to see if, the steam generator was lest, could te core be cooled with high pressure injection and the letdown systen. " hey found it could help natural circulation. We natural circulation study was used to determine if stagnated flew would occur in a leg when the p:=ps wre turned off.
He, indicated that the calculations matched ceserved data.
He classified the transients study as a parameteric study of "what ifs".
'Ihis study was acecmplished with REI.AP-4/MCD-7. He presented scme results of calculations obtained. 'Ihese are shown in Attachment AP.
He said that INEL has learned that small break analyses are more difficult than anticipated but that a good comparisen with measured data was found.
Dr. Plesset asked if the Japanese had made a similar type of analysis of the t ree Mile Island accident. Mr. Behlirq said yes and that they had used REIAP-4/MCD-5 with good results.
Dr. K. Soda of JAERI said that he would see if he could send the Japanese analysis to the ACRS.
(He has subsequently provided de informatien which has been distributed separately.)
Or. Theofancus asked if a discrepancy had been observed with the rate
- of mass ficw frem de reactor. Mr. Sehling said that they had difficulties wi2 the problem because the boundary conditions wre urJ<newn.
Cr. Plesset cuestioned how long the ecmputer runs wre and Mr. Sehlirs explained that some 47,0C0 cceputer-seconds en C'C-7600 were necessary for 1650 transient-seconds.
ECCS Meeti.ag 9/27-28/79 FRESDT"AT'CN SY V. H. RANSCM Mr. Ransem discussed REIAP-5. He explained that the cbjec-ives of dis cede was to prtduce an econcmic, advanced best-estimate cede. Previcus objectives included an EM licensing code as well but this objeccive has been d,reeped.
Mr. Ransem explained dat de apprcach taken was co improve user convenience, to include advanced two-phase modeling and to rake the numerical scheme run faster. The features of the REIAP-5 Code ara included in Attachment AQ.
REIAP-5/MCD-0 is a derral-hydraulic systems code. It is limited to a single compenent, i.e. steam and water.
Mr. Ransom explained cac later versions will incorporate non-condensibles. He said dat the code was released to de National Energy Sof tware Center in May 1979. MCD-1 version is being developed in support of LCET and Semiscale testing.
Added capabilities were discussed. He described the structure of the code. 'Ihis structure is shown in Attachment AR.
He presented plots of REIAP calculations and Semiscale data for pressure, density and mass flew and discussed de nodalization for Semiscale S-07-6 and for an upper head drainage problem. He also discussed results of ccuntercurrent calculations made by Intermountain Technology, Inc. He concluded frem these examples that ncn-equilibrium is significant at scme portiens of ECC and to upper head injection predictions.
He explained that temperatures in de downecmer shew that the wall temcerature cuenches and this effect plays a significant role in downecmer water depletien.
The mechanistic abrupt area change medel appears premising for CCEI.
predictions.
He said dat REIAP-5 had been applied to de first of de small break test, L3-0.
He presented results of predicted liquid level in de core area and in de dcuncemer. He also shewed scme gest-test calculatiens of system and pressuri::er pressure. He explained dat de only cha ges made in de pre-and post-prediction calculations were made in break flew area, by changin; it by 10%, and by including a loss of factor for de elbcw in de surge line.
ECCS Meeting 9/27-29/79 He discussed several eversights made in de medeling, exclaining Sat had $1s been included he wculd had expected de results to have been improved. He 'su=arized de results as indicating dat the pressurizer almost filled with licuid and that de core did not uncover. & said it probably would not have uncovered even if the steam buccle had been released. Fe said dat =inor flew paths can be very imporat in small break analyses. te calculations ran about twice as fast on the CDC-7600 as the real time transient.
In respense to a question from Dr. Wu regarding RSIAP-5's ability to calculate countercurrent ficw in horizontal surge lines, Dr. Fansem ex=lained that REIAP-5 did not; have the capability at this time but it is expected to be added this year.
PRESD(PATICN BY J. TRAPP Mr. Trapp discussed the REIAP-5 Simulatiion of the Marviken Test. te Marviken facility, a discharge pipe connected to a full-scale pressure vessel, is used to study reactor vessel blewdewn. Flew through a half-meter no::le aids in the study of subccoled choking and the development of critical flow models for de subcooled flow regions. te em-phase mixture in the vessel depressurizes in about 50 seconds for 'this series of tests. Se objective of these tests was to study the flew rate for a given depressurir.atice rate.
Critical flow models were used in de computer code since rate-dependent correlatiens for interphase mass transfer and interphase drag have not been developed. Nonequilibritml and em-dimensional geccetric effects complicate ce analysis of critical flew through the no::le. S e use of extremely s=coth ideal no::les would eliminate gecmetric effects and aid in the analysis of be experimental results.
e.
FCCS Meeting 8/27-29/79 Approximately 15% vena centracta effect appeared in the no::le. Iquilibritmi conditiens charactari:ed abcut 90% of the flew. Se critical pressure ratio (compared te the stagnation pressure in the tank) was abcut 0.6.
Se validity of tne mechanistic flew medel was questiened cecause of the inter-action of geccetry and renequilibrium effects and because of doubt of the knowledge of the throat pressure. 21s series of tests were inconclusive.
An idealized medel with modifications to take into account non-equilibrium effects was recommended as a model for critical ficw for reactor calculations.
PRESEMTAT*CN 3Y C. 3RCACUS Mr. C. Broadus discussed de SEACCN Code (descriptien and develcpment).
Se curpose of the SEACCN Code is to provide a best-estimate containment analysis capability in order to evaluate the actual transient or phenomena associated with a reactor containment. The current licensing philosophy of using conservative analyses may have drawbacks: we do not know the degree of conservatism involved in the analysis, conservative or h mcgeneous-type codes may cover L:p important problems or phencmena, and the calculation of high pressures and temperatures from conservatively Icw input data may not really be conservative as far as safety =argins are concerned. 2e SEACCN Code may be used for European type licensing analyses where safety factors are censidered after a best-estimate calculatien is performed.
The latest version is SEACCN MOD 2-A.
The SEACCN Code includes:
1.
2-dimensional, 2-phase, nonequilibrium effects; 2.
heat transfer to walls and structures;,
3.
water preperties tables; 4.
dynamic storage reutines; 5.
mass, mcmentum and energy source model; 6.
wall-film medel; 7.
special heat transfer correlations for centainment analysis;
ICCS Meetirs 8/27-28/79 9.
.edular programming organizacien; 9.
lumped parameter regica model; 10.
restart cacability;
- 11. partial flow blockage capacility between cells;
- 12. handles 2-ccmponent air and water mixtures; and, 13.
facility for cc= plex gec=etries.
Results of the P.attelle-Frankfurt D-15 test (basis for the containment analysis standard problem, CASP) using SEACCN were discussed. "he results of these calculations are included in de handcuts. SEACCN, because of its fini:e-difference structure, cannot calculate the passage of a steam front through a rocm. We finite difference construction tends to diffuse the front over a number of cells. We air fraction, in, addition, is not purged out of the cell as rapidly as in the real situatien. An apparent delay time in thermo-couple response, as calculated by SEACCN, resulted in the re-examination of actual thermeceuple response in a still-air envirement.
SEACCN MCD 3 is presently being developed.
It will include 3 new models; best-estimate correlations for interphase heat, mass,and momentum transfer, an improved form and friction loss model which will have the capability of handling wall friction, and an improved gecmetic modeling capability. This improved version will be used to study separate-effects-type problems (entrainment, 2-D flew, etc.) and integrated contalment problets. De capability to analyze an ice condenser contalment or the~ effect of a hydrogen burn in the containment has been considered for future wrk.
PRESENTATICN 3Y M. SAHCTA Mr. Sahota discussed de SEACCN Ccde (thermal-hydraulics). He said $e SEACCN Code centains a rigorous treatment of discersed droplet-type flow and bubbly flow. !he interphase exdanges for a simple droplet are con-sidered' TK exchange rate is extrapolated ever various droplet sizes
ECCS Meeting S/27-23/79 after consideration of the distribution of the bubbles.
For void frac-tiens close to 0.5, atere bubbles may be censidered to be in a licuid or de droplets may be distributad in a gas, steady-state correlations are used to predict de exchange rates. S e dreplet appreach is used instead of a more global view, The droplets are modeled as solid spheres in this approach. Crag coefficients for deformed droplets may be quite higher and should be considered. Se conduction of heat away frem the surface interface and net de change Of phase frem gas to liquid is the limiting step in de gas te 1.iquid transition.
PRESDrrA"'ICN SY CR. 2. RCSZ"CCZY Dr. Posetoccy discussed NRR Research needs.
Cr. Pes::teccy presented the important licensing data needs **hich related to LCPr, Semiscale, and other short term programs to be completed within the next year or tao.
Dr. Fcseteccy stated that Westinghouse was not ecmpletely pleased with the upper head injection (UHI) test program. Westinghouse is not doing the pretests for the program.
If meaningful data is obtained, Westing-house will perform all of the calculations and blind analyses. CCSPA has been merged with iFAC to perferm UHI calculations.
(Previcusly, CCBRA used the boundary conditions generated by SATAN for dese calcula-tiens.)
Th, results of the CCBRA/ TRAC run will be available shortly.
The tao main results frem the small break LOCA tests (run at SD4ISCAI.I in de Scring of 1973) *aere ccmparatively large uncertainties in de calculations for de depressurir.ation rate and in de intreduction of cold ICCS water into de primary system. Because Of d e uncertainty.
concerning de validity of 2e data, additienal :ests and predicticns by venders have been crdered. Se new test was ccepleted in Cecember of 1978. te calculations frem the vendors are not in yet.
ECCS Meeting 9/27-29/79 Small break LCCAs are occurring with a higher frecuency ten previously considered (4 events within 30 reactor years for de S&W plants). Plant response to very small breaks is different to the response frem the pre-vicusly examined breaks. Very small breaks are not larye enough :: re.cVe the decay heat from the core. Natural circulation is more important. Se pressurizer may not empty. Se NRC/NRR has requested vendors to perform tests that demonstrate the basic behavior of very small breaks, the de-pressurization, the pressure hang-up, and de repressurization.
(These tests are listed in NUREG-0579.)
SD4ISCAI.E Test S-07-5 was run and voiding occurred in the dcwncemer during the test. Se SD4ISCALE downcomer is one-dimensional. S e peak centerline temperatures were not significantly changed. We large effect of the quenen time and subsequent oxidation (Zr-H O reaction) are 2
quite different from those predicted in cede calculations. REIAP-4 and TRAC (1-D) did ret predict this test result. REIAP-5 may be able to predict this result next year. Bis increased quench time for SC4ISCALE cannot be calculated new. M C (3-D) has predicted this downcomer voiding for an actual ER.
TRAC (1-D) uses a reflux model where TRJC (3-0) uses a two-phase flew model. te 3-D model is considered more appropriate.) This extended quench time may lead to large uncertainties (by a fact of 2 or more) in the amount of the Zr-H O reacted. De 2
phencmenen involved (void production in a subecoled system, pressure reduction in the void, more toilirg) can occur in a large system.
It is an unstable condition.
This voiding (in de downcemer) has been observed in every single apparatus on which tests have been conducted.
Dr. Tong cautioned that the results of this test were influenced by de poor insulatien in the dcwneceer.
NRR believes the phenecena is real and thac it is not limited to small sized test apparatus. Mditional insulation has been added to de SO4ISC.M.E downcemer to scale down the heat flux for additional testing of dcwncocer voiding during small break LCCAs.
ECO3 Meeting 3/27-28/79' The last group of data requirements includes transients and accidents (loss-of-feedwater, red withdrawal, turbine trip, etc.). A test at Peach Sottem (SWR) in 1977 showed dat de void collapse tighc increase the reactivity of the core. Bis void cellapse was not correctly treated by the available computer ccdes. General Electric has developed a new code and is in the process of evaluating these transients.
Calculatienal techniques have not been goed enough to predict phencmena discussed in this presentatien. Rese pienemena were discovered experi-mentally.
NRR reports to the licensing boards all infonnation that could significantly affect the licensing of a particular plant. A member of the Subccemittee felt that information gained from a facility like SE4ISCALE should carry qualifications and reservations in its applicability to PWRs.
PRESECATICN BY T. f.AhtSCN Mr. Larsen discussed scaling of Semiscale. He said geccetric and dynamic similarity in scaled experiments are difficult to maintain since different scaling criteria (surface area-volume ratics, steam generation rates) are more important in various transients. Distortions exist in systems ~here multidimensional effects are present especially when elevatiens and volumes are involved. Bree techniques of scaling (linear, dimensionless numhers, and volume) were addressed.
We L/D ratics frem the referenced system are maintained in linear scaling. Bis technique has been used in rall-scale steady-state systems somewhat successfully. Time scales for rate processes change but accustic ti=es can be maintained given proper scaling factors.
Linear scaled systems of 2 megawatts (Semiscale) require heaters de si::e of piano wire.
ECCS Meetirq 8/27-28/79 The controlling dimensionless nt=bers may not be the same throughout the whole transient. He multitude of dimensionless numbers complicrees this scaling technique (e.g. C*E-14 dimensienless numbers).
Volume scaling is used to scale LCFT. De time scale relative to the referenced plant is preserved.
Proper ~ hydraulic resistance distributions can be maintained. Elevation, volume, length-area, and resistance relation-ships cannot be maintained simultaneously. Accustic transit time relation-ships (-hich effect subcooled dec::xnpression leads) are not maintained.
SEMISCALE MOD 3 (scaled to the Westinghouse Trojan plant) was or.ginally designed to lock at ucper head injection (UHI) censiderations. Volume scaling was employed for this facility. Elevations and piping sdeface-area-to-volume ratios are distorted. t e large L/D ratios cause the system to act primarily one-dimensionally. Mr. Larson felt that uncertainties in the SEMISCALE and LCFT data were about 1 20%. Sci! SCALE is experimenting.
with an on-line ecmputerized rod bundle p wer control in order to sLdate the proper heat flux during transients. Additional insulatien has been added to reduce the structural surface area problem.
- he leops use hydraulic resistano, volume distribution, and pump suction leg depth scaling criteria. Tao stea:u generators are used in the facility. Cne is basically scaled for LC?r while the other is scaled for a par. De PnR type is scaled for full elevations and pressure dreps. t e broken loop, however, has an oversi::ed secondary volume.
Calculations have been performed hich indicate that 2e '. eat flux in the SC4ISCALE downcemer was three or more times greater relative to a BR during Test S-07-6 (dewncemer voiding). H e sur~ face area-to-volume ratics are about eight times greater in the S&.ISCALE downcomer than in a ER.
SCCS Meeting 81/27-2S/79 SC4ISCALE Test Series 5 censidered the effect of ptmp trip versus no pump trip during LOCAs. hhen the pump was tripced, CNS was delayed in the core.
?=p degradation has been recorted for small and large reactor ecclant pe=ps (SE4ISC'd.I and LC:FT) at between 5 and 7 seconds after a large break LCCA.
Tests run in the S24ISCALE MCD 1 series indicate that the steam generator did not affect cold leg mass flew rates significantly during the blewdown period of a large break LCCA.
It has become apparent by studying "MI-type situations dat losses are extremely imprtant because of the large percentage of decay heat removed by this mechanism. Bis is a problem in scale facilities since as facility size decreases the percent of core p::wer lost increases significantly.
Increased use of insulation and heat tape should alleviate this problem.
The relative mass of pipas does not appear to be a problem where fluid temperatures undergo step changes (e.g., loss of feedwater) when pipes are well insulated.
te small reactor coolant pumps used in the SD4ISCALE facility cannot replicate the par pumps. Small pumps suffer two-phase degradation faster than large pumps. t e scecific speed of small pc=ps is about 1500 whereas in the par pt=p it is about 5200. 2e SU4ISCALE pumps start to degrade at void fractions of 2C%. Cata and calculations frem "MI-2 indicate cat large reactor coolant pumps pump well until the void fracticn is abcut 4C%.
Se scaling of critical flew phenomena is another area of cencern.
A scaling factor of 0.48 has been found to give censistant results as a correction factor for the vena centracta effect.
te pressurizer surge line at SD4ISCALE is 0.37 inches in diameter (B&W par is about 3.5 inches). te surge line volume and resistance
ECCS Meeting 8/27-23/79 have been scaled. Surface tension effects in the SEMISCALE surge line are not considered a problem. A cc=parison of critical gas velecities and actual velocities in FeiR and de SC4ISCALE surge lines may be in-adecuate for ce case where steam is entering or leaving de pressuriter via the surge line. Be absolute length of the surge line and elevation of the pressurizer with respect to de dog leg will make a difference in the flow rates. We question, under what circumstances 'muld steam flew prevent water from flcwing out of the pressurizer is being investigated further.
Dr. Posztoczy discussed calculations and experimental data for countercurrent steam flew and fleeding in long horizontal pipes.
Subcccmittee members -ere concerned aceut the increased friction due to pipe length and the rate of change of the steam velocity in stepping the penetration of water into and along the pipe. Ebrther examinatien of this problem is required.
The large L/D ratics virtually make Sc4! SCALE cne-dimensional. Multi-dimensional effects, siach as the het downcemer, may not beccme apcarent.
Be scaling for the WI experiment is considered a gravity dcminated situation. Se three-dimensional effect is thought to be less than in a large break LOCA.
PRESECATICN SY D. J. CLSCN Mr. Olson discussed "I 80-81 Plans for SE4ISCALE. Small break testing will continue until March of 1980. Tests will be performed which relate the results of SD4ISCALE to LCFT. We second stage.of small break testing will address the separate effects issues discussed earlier in dis meeting (e.g. surge line questions). Test S-0-76 will be repeated with the new downcemer insulation.
Loss of feedwater tests will start in April of 1980. d ese results should be available in September of 1980.
Following this series of tests, WI sensitivity studies will be performed. Bis erk has been delayed because of the lack of capability for ce ';HI calculation.
ECOS Meeting 8/27-28/79 PREsDfrATICN BY E. HL:.VE00 Mr. Harvego discussed test plans and upgrading of Semiscale to conduct transient tests. Dr. Catton indicated that he believed that B&W and Mr.
Michelson had perfoced an analysis of the s:1all break LTA as a series of steady-state (cuasi-static) processes. me SmISCM.E tests will not spect-fically examine this aspect as part of its cd 'uthentification program.
Five small break tests have been planned.
These tests will simulate four-inch to ene-inch FAR breaks.
me la t test s
will look at the effects of ECC injection into the u=per head. Natural circulation and pump trip /no trip tests have been suggested as supple!ntal tests.
A plan has been procesed to use the event tree analysis frem WASH-1400 in running loss of feedwater supplemental tests. T* ce SEMISCALE is a non-nuclear facility, severe limiting transients can be investigated.
A full range of operational type transients have been planned. Che long-term test, ATds, will require the inclusion of a reactivity feedback model into the rod heat control. Secondar/ feedback effects will require the implementation of a closed-loop secondary. De adequacy of HPIS to provide sufficient core cooling for varicus break sizes (as depicted en several event trees) was questioned by the Subcomnittee members.
PRESDTTATICM BY G. JCHNSDI Mr. Johnsen discussed the 04I simulations at SmISCALE. me SDtISCALE facility ran tests to simulate the bubble in the 24I core. We bubble size calculations were not independently checked by de staff at de facility.
2e cbjectives of these tests were to decemine if recovery by depressuri-
- ation was viable and to determine whether de pressurizer level at 24I was a valid indication of de system licuid inventor /.
Many changes had to be diade in the facility to medel de 04I configura-tion. Se heat losses in de facility were the primary mcde of heat i
ECCS Meeting 9/27-28/79 rejection instead of the steam generator. The steam generators in the facility wre U-tce type.
One test showed dat de prima:/ system could be depressurized wi2 the gas bubble in 2e upper plenum region of de core wideut uncover-ing tne core to the paint here the residual heat removal system could be activated. Both SC4ISCALE pumps could not be operated with the non-condensible gas in the system. Nei der pump could generate sufficient net Insitive suction pressure.
If one pump wre cperated, however, it would force the gas into the 'other leop and the cperating pump could then circulate cooling flow into the core.
System parameters in the SD4ISCALE simulation (e.g., core temperature, pressures) seemed to be in general agreement with the TMI transient results. Validity of de results was discussed in earlier presentations.
After the pumps wre secured at SD4ISCALS, the oscillatory behavior of the core licuid level ceased. Water from de hot legs, steam generators, and upper portiens of the systen drained into the core to balance the boil-off. The core level remained steady. The core liquid level dropped as 2e coil-off continued. The pressurizer remained full during this
't transient. Additional tests indicate that significant core damage could have occurred at T4I-2.
The graphs and chara from these tests of plant parameters (e.g. core mixture level, pressures, temperatures) Wre included as handcuts.
G'i.NE9AL C'.CSDC CCW.ENTS In response to a cuestien frem Dr. Lipinski, Dr. Ros:tec y indicated Sat the SRC is considering running sized tests @.ere risk is relatively icw and de plant dees not exceed CE. The example of General Electric's code verificatien of de voiding reactivity effect (in a full sized ShR) was addressed again.
e ECCS Meeting !!/27-28/79 Dr. Ecsztec:y stated dat both the SRC and S&W believe dat natural circulation can be interrupted in the once-through steam generator plants either by a bubble in the top of de candycane or when 2e feedwater level in the steam generator falls to a low level. Calculations indicate that the U-tube steam generator does not have these difficulties unless the steam generator drys out. The Sa4!SCALI facility, in order to test these beliefs, should include each of these types of steam generators.
Meeting a s adjourned.
NMES:
1.
For additional details a complete transcript of this meeting is available in the NRC Public Document..com,1717 H Street, N.W.,
Washington, D.C. 20555, or from Ace-Federal Reporters, Inc. 444 North Capitol Street, N.W., Washington, D.C. 20001.
2.
Materials provided to the Subcomittee at this meeting are on file in the ACRS office.
9
Federal Resister / Vol. 44. No.156 / Friday. August 10. 1979 / Notices 47191 NUCLEAR REGULATCRY fannulate a nycrt and reed =tions to CCMMISS:CN es M f> mittee.
At the cmictusion of is Executive Session.
Adytsory Cotr9fttee en Reacter de sube
- ttee w'.H hear presentations by Safeguartis. Subcommittee on and hold discuseions with representatives of Emergency Core Cocilng Systems the NRC StaH. and dear consultants.
(ECCS); Meeting perenent to the above topics.no w~-ttee may then caucus to determme The ACRS Subcommittee on whether de matters idenufled in de imnal Emergency Core Cooiing Systems will session have been adequately covered and hold an open meeting on August 27-;8.
whether the project is ready for review by the 1979 at the Westhank Motel Coffee full Cmun.
Shop. 475 River Parkway. Idaho Falls. ID Furtberinfor ation regarding topics 83401 to review NRC Research Programs to be discussed. whether $e =est:ng on LOFT. Sociiscale. BEACCN. and has been cancelled or rescheduled. de RELAP. Nocce of this meeting was Chair =an's n: ling on requests for the announced July llS.1979 (44 FR 438::::}.
opportunity to present oral statements In accordance with the procedures and the time allotted therefore can be outlined in the Federal Register on obtained by a prepaid telephone call to
, October 4.1978. (43 FR 459:5), oral or the Designated Federal E=ployee for wntten statements may be presented by this meeting. Dr. Andrew L Bates members of the public recordings will (telephone 02/834-3::87) between 8:15 be per=itted only dunng those portions a.m. and 5.00 p.m EDT.
of the meeting when a transcr:pt is being Dated: August S t9 3.
kept. and questions may be asked only Samuel J. m by members of the Subecc:=ittee,its g
f,3, gg,,,
consultanis. and Staff. Perscus desiring to make oral statements should notify the Designated Federal E=ployees as far in advance as practicable so that appropnate arrangements can be made to allow the necessary time during the meecng for such state. 'ents.
The suends for subfet tnzeeting shall be as follows:
Mcndcy and Tuesday. August 27-:S 1979 tx a.or unal the conclusion of business ecch day, ne Subco nmittee cray meet in Execunve Session. wttb any ofits consultants who may be present. to explore and exchange their prelinunary opuuons regarding matters which should be considered dancg the meenng and to ATTACHMENT A
E SUSC2MIMEE.'d2CDJG CN IMEPGDCY CCRE CCCLDG SYST9dS -
AUGUST 27-25, 1979 IrAHO FALIS, ICAHO AUDICEES LIST ACRS En&G NRC M. Plasset, Chairman G. W. Johnsen S. Fabic W. Mathis D. J. Olsen P. O. Strom I. Cat':en N. C. Kaufman-L. S. Tong
.T. Theofancus J. E. Wilkins Z. Ros::tocrf T. Wu J. L. LaChance K. Garlid P. Nort KFK GERMANY /INEL F. Zaloudek C. Broadus W. Lipinski R. K. McCardell S. J. Dagbjartsson R. Allemann N. C. Kauf=an T. McCreless, Staff
- D. Snider CF.",.GAE/I'.'EL A. E. Peterson
" Designated Federal D. L. Batt G. H. Weimann Employee R. J. Wagner R. W. Shumway D. W. Crocher B&W J. Trapo H. Bailey USCCE EG&G W. R. Young S. A. Naff W. W. Sixby S. T. Kelppe K. R. Peters R. Nelson J. E. Solecki T. Sudch M. Akimoto h7.STINGHCUSE D. Jarrell J. Lineba.ger S. Kellman E. C. Anderson D. L. Reeder CCFPCNWEALUI EDISCN D. D. Miller E. A. Harvego C. W. Solbrig J. A. Cearten S. R. Behlirx; ENEPGY INCCRPCRATED W. J. Cuacp
- 3. C. Gottula J. H. McCleskey T. K. Larson D. J. Hanson UNIV CF CA L. Winters P. E. MacDonald A. Presperetti P. J. Schally V. K. Chir K. Soda T. K. Shmuels CRtR.
J. R. White V. H. Ransem J. D. %hite M. Sahota ATTACHMENT B
).
APPROACllES OF SHAlL BREAK AND REACT 0lt TRANSIENT RESEARCil ENGINEERING ANALYSIS EXPERIMENTATION Develop and.lustify the scaling criteria ai.
Review existing data (in test facility and In reactor) and infonnation (veridor's sub-ratiosiale, idelitifying questlogiable area pilssion, HRC audit & ACRS reconnendations) needing parametric study dild identify the Wedkness of existing Codes
& linprovanents required Plan experiment to meet the needs, includie e
Infonnation lastrinnentation requirunent and f acility Analyze reactor tratistent and abnonnal oper.
g
+
allonal events and associated consequences 7
anoill fications Heques based on physical behavior of plants Establish test matrix and expected results Establish the nature arid extent of new e
phenonena to be studied and the data required Construct & modify test facilities
', Perform detailed sensitivity study by computer codes
,g improve or develop undels in the codes y
Validate engineert'ng analysis results by e Data evaluation and error analysis experiniental data and/or code calculations e Physical interpretation of experimentally Predict reactor behavior during various
,Informa tion measured separato effect and system inteur.
e transients T
hehavior when considered interims af react Supply simulation during postulated translevet Discover any aa?ety deficiencies, such
(~
as unforeseen potential reactor accident, e Organize the data into a useful format suci.
poor control, luproper operation pro-correlations, nodels, tion-diniensional grou, A
cedures, etc.
L V
y Eapability of analyzing wide range Characterization of systemi response of postulated transients and accident during various postulated transients conditions in LWRs for use of sinulating (WRs for confirniing our duditing Licensee's calculations understanding and exploring further postulations a
O O
[
SCALING CRITERIA Preservation of Physical Phenomena Small Break Large Dreak Correct time of systen pressure Power /voltane = constant Power /voltane = constant of flashing 2-$ fluid Break arca/voltane = constant Break area /voltana = constant Fluid inventory and distributton Initial loop'tanp. & core AT main.
d Correct history of pressure distr l-Elevation and flow resistance Flow length and resistance bution throughout the systun simulated coefficient simulated Infllal Core AT, RC pians flow and Core quench during bloudown N/A loop resistance maintallied i
ITeTiood heat transfer Core-height maintained Core lielght maintained
~
I Natural circulation rate, lleight of steani generator and fl/A fraction of core uncovery loop geometry saaintained Reactor thennal-hydraulic transient Both primary and secondary of N/A steam generator simulated.
Secondary coolant. temperature R
and level simulated.
L Aux 111ary FW systen simulated ECC Bypass N/A powncaner size, geanetry and rate counter-current steam flow mainta-Downcaner swelling and time Wall heat flux / volume = constant Wall heat flux / volume = constant constant of plant thennal transient s
Time of core uncovery in S.B.
Upper plentan volume and distance Internals and nozzles in upper N
and steam binding in L.D.
between the top of core to the plenian to be simulated br top of exit nozzle maintained I
~
s COMPARIkON OF SHALL BREAK AND INtGE BREAK LOCAs i
t.
f Small Break LOCA large Break LOCA
(
2 2
Sample break size 0.02 ft 4.0 ft Significant lleat Source Decay heat Stored and decay heat
(
Sigolficant lleat Sink Break Flow and lleat Transfer thru Dreak flow and ECC water S.G. to Secondary sidd i
f Psec > P ri. Ilsec 't Ilpri Ileat Transfer in S.G.
Ppri > P ece llpri + lisec p
s AFW Significant AFW Insignificant
(
1 Irlmary Side Pressure liigli pressure maintained because Fast depressurization by blowdown I
of slow draining Flow behavior in Primary l.
Strattfled flow 1.
Bubbly or droplets dispersed flow i
2.
Separationofnon-condenslIsles 2.
Ilomogeneous with flow i
Side at 131 11 Spot i
9 3.
Gravitational. force control 3; ikwnent tan
- control l
4 Core may uncover by flashing 45 Core enytted and recovered quickly l
[q S.
Pressurizer effect significant 5.
Pressurizer has less offect-g ECCS 1.
Chargirk p sup & ilPSI 1.
Acctanulator most effective 2.
Ef,fectiveness. depends on the 2.
Effectiveness depends on the pressure for initiation of initiation pressure and location l
Injection of injection 3.
.In. cold leg break LOCA, core niay 3.
In cold leg break LOCA, there l
have to be partially uncovered may be stears hinding and l~CC to vent steam thru loop seal.
bypass.to slow down reflooding.
I 1.
Afil f4 flatural Circulation for wet Plant recovery S.G.
1.
Acctanulator & Iteflooding 2.
llanual open all PORVs to lower the 2*
Continuous LPSI or illlR pressure for llPSI, Accumulator,
/
LPSI & Rillt wheli sleau, disnp is not avallalle
MODIFICATI0lls til SEMISCALE FOR SMALL llREAK TEST SCALING UEFICIENCIES IN MOD 3 MODIFICATIONS I'
WALI AllEA TO VOLUHE RATIO TOO LARGE IN DOWN-1.
ADD INsutATION TO All. TilESE.
COMER, CORE BAltitEL, UPPER AND LOWER PLElluMS-LOCATI0tlS 2.
TOO SHALL A BREAK OPENING AND Tile SIZE OF 2.
DIScilARGE FLON RATE AT SullGE LINE Ill SCAllilG DOWN A SMALL DREAK LOCA OPEilING.WILL DE PRE-CALIBRATED 3.
IWO DISSIMILAR STEAH GENERATORS 3.
REPLACED BY SlHILAR STEAH GENERATORS
(
lj,
ll0 SEC0tlDARY SIDE PLANT TRANSIENT SIHULATI0ff 11.
ADD A CLOSED LOOP Oil SECollDARY SIDE WITil PROGRAHHEI) CollT110L 5.
ilo TWO EQUAL ACTIVE RC LOOPS AND PUHPS 5.
ADD 'A HATCillflG PilHP IN INTACT
.00P 6.
UPPER PLEilUM SlHULATION HEEDS HORE STUDY-6.
DE-Et4TRAltlHEllT EFFECT IN UPPER PLEiluH COULD DE OBTAlHED FROH CCTF
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l 0/14/79
~
TESTING MATRIX LOFT Saniscale TilTF FLECilT ILTA pKl.
CCIF SCTF 2250 2250 2250 60 1000 500 do 80 Technical Subjects for Testing psi psi psi psi psi psi psi pst 1,
Integral Syston Tests a.
Natural Circulation Loop geometry & canponent effects X
X X
X X
Syston Volding & S.G. condensation X
X X
X Effect of non-condensibles X*
X X
X ECCS Injection & location X
X X
Effect of break locations X
X X
Effect of upper plentan size & exit nozzle X
X height Effect of running itC planps part time into X
X accident b.
Primary Loop Cooling Hechanisms Break flow cooling, w/o S.G..
X X
Effect of 11ll11 delay, w/o S.G.
X X
0Wil Jet ptsap effect X
c.
S'econdary Loop Cooling Hechanisms S.G. perfonnance & control X
X X
X Secondary simulation of reactor transient X
X d.
S)suptivus'of small break & reactor transient tilagnostic display & detecting instrinnentation X
X Leak signal arid location X
1 y
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1.
0/14/79 TESTING HATitlX LOFT Semiscale TitTF FLECllT TLTA PKL CCTF SCTF TP1F 2250 2250 2250
~60 1000 500 80 80 2300 Technical Subjects for Testing psi psi psi psi psi psi psi psi pst 1.
Integral System Tests (continued) e.
Plant llecovery Techniques Adjust ECC activation pressures X
X X
Manual opening of PORVs X.
X Iteturn to natural circulation X
X X
Operator intervention X
X f.
Nuclear feedback in reactor transient X
including ATUS in long tenn 2.
Separate Effect Tests a.
Pressurizer and relief valve operation X
X b.
lleat Transfer in S.G. (UTSG or OTSG)
(
Primary condensation and CCFL X
X X
Secondary liquid level effect on ll.T.
X X
X t
X Auxiliary FW inlet locations c.
lleat Transfer in uncovered Core Hlxture level swelling X
X X
X X
X X
Ste&u cooling with eioisture & flow X
X.
X X
X X
X Stagnant steam natural convection X
X X
X 3 g
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CODE LAB WilEN STRENGTils AND WEAKHESSES AVAILABLE TRAC-PIA LASL NOW Sniall Dreaks: Excessive Running Time. (6Ifrs.)
Consequences of collapsing nodes to reduce running time not yet assessed.
Trips an(t controls not adequate for Reactor Transients analyses.
RELAP-S INEL NOW Small Dreaks:
Insufficient assessment as of now, concerning small break capability.
Trips and controls not adequate for Reactor Transients analyses, r
{
TRAC-Pfl LASL Decenher 1979 Good, fast running code, calable of addressing PUR small breaks and reactor ~ transients.
i (PWR)
Trigs,and controls n per RETRAN.
r Deceuber 1980 (DWR)
REIRAN BNL, INEL August 1979 Not adequate for small breaks.
(from Good for Redctor Transients (PWR and bWR), not requiring multidimensional neutronics EPRI)
Teedback. Good trips and control logic. Long running time.
RELAP-3B DNL NOW Adequate' for many transients. flot for small breaks. Control logic not as good f
as that of RETRAN. Long running time.
RAMONA-Ill DNL lecenher 1979 Good for many DWR transients and accidents. Not for small breaks. Fast running.
Some control system by Decenher 1979; Complete controls by Septenher 1980.
IRT DNL NOW for PWit transients, not for Small Dreaks.
Improvement needed in S.G. modelin3 Fast running time. Controls and trips not as gogi as in IIETRAll.
RELAP-4/
INEL NOW Hore adequate than RELAP-4/H0D 6 for Pult small break analysis. Long
[
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Inadequate controls and trips for reactor transients.
7 l
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n LAB 0HAIDHV CilARACTERISTICS OF 1NEL PROGRAMS AND FACILITIES PROGRAMS ARE ORIENTED TOWARD IMPROVED UNDERSTANDING OF PROCESSES AND METil0DS IMPORTANT TO WATER REACTOR SAFETY CAN AND llAVE CONTRIBUTED TO Tile UNDERSTANDING OF PilYSICAL PROCESSES IN LWR SYSTEMS AND OF METil0DS AND TEClill100ES FOR DESIGN, ANALYSIS, OPERATION, AND EVALUATION g
OPPORTUNITIES INilERENT IN UNIQUE COMBINATION OF FACILITIES A
AND ANALYTICAL CAPABILITIES OPPORTUNITIES INilERENT IN DIVERSITY OF FACILITY DESIGN AND OPERATION f hi?888 lasha,Inc.
1
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(ADMAmiv CilARACTERISTICS OF Tile CONDUCT OF LARGE-SCALE EXPERIMENT PROGRAMS 1
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9 DESIRE TO BE RESPONSIVE TO INDUSTRY NEED DEGREE OF COMPLETENESS OF PREPARATION AND ANTICIPATION 9
INFORMATIONAL VALUE OF EXPERIMENT UTILIZAIION OF FLEXIBILITY AND OPPORTUNITIES E
BASE FACILITY AND RELATED SUPPORT COSTS
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INCREMENTAL EXPERIMENT COSTS
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UNIVERSITY CONTRACTS AND PRESENTATIONS WORKSil0PS AND SEMINARS TECilNICALSOCIETYPARTICIPATIdN dasseww.i.w.
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1 LA80RAIDAY i
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}i SIGNIFICANT LOFT PROGRAM CHANGES ACCELERATE L3, L6 TEST SERIES PERFORMED L3-0 TEST E
L INCREASED EMPilASIS ON PROCESS INSTRUMENT BEllAVIOR, OPEP.ATOR ASSISTANCE, PLANT CilARACTERIZATION
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LOFT Three-Year Plan FY-79 FY.-80
_ FY-61 Q N D J F MIAEAl.IJ JlhliSID N D Ji FiM W M 3: J A S a N D J FlM{APA J J A S lL2-2 l lL3-0 L3-4 LS-Xl' L6-5 i
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LOCE L2-2 1.0EE_L2-1 i[
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MAXIMUM LINEAR llEAT 26.fi KW/M
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PRIMARY COOLANT MASS 19tl.2 KG/S 199.8 KG/S j
FLOW
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l ll0T-TO-COLD LEG TEMP-22.7 K
32.2 K
ERATURE DIFFERENCE L
l.
COLD LEG TEMPERATURE 557.7 K
560.7 K
i SYSTEM PRESSURE 15.6ft MPA 15.06 MPA i
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I FIRST INDICATION IN 2 11 2 11 25 DOWNCOMER OF ECCS FLull) (S)
REFLOOD RATE (M/S) 0.12 1 0.02 0.12 1 0.02 0.10 1 0.02 l
(A BYPASS (%)-
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CORE VOLUME REFLOODED (S) 59 55 55 t
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AXIAL REWETTING PATTERN BOTTOM-T0-TOP UOTTOM-TO-TOP BOTTOM-TOP-MIDDLE LAST REWET (S) 11 5 25 55 1
MAXIMUM CLADDING TEMP-PERATURE DURING 515 665 850 REFLOOD (K)
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LADOHAlOlly EXPEltlMENT L3-0 INITIAL C0llDITIONS Pl!!?VlilY COOLANT SYSIEtt tie 6SbHED_YALUE MASS FLOW RATE (KG/S) 201.0 1
6.3 PilESSilRE (MPA) 111. 711 1 0.07 INTACT 110T LEG TEMPERATURE (K) 556.7 1
3.0 INIACT COLD LEG TEMPERATURE (K) 559.7 1
3.0 I
DECAY HEAT GENERATION (KW) 11. 2 i
1.0 D
AMillENT llEAT LOSS (KW) 2fl8.0 i 60.0 i
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IDAHO N A HONAL
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EXPERIMENT L3-0 INITIAL CONDIIIONS (CONTINUED) l I'
!!RESSURIZER llEASURED VALUE
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3.0 SlEAM GENERALOR SECollDARY_SillE WATER TEMPERATURE (K) 557.3 1
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LAltGE NUhBER OF INSTRt!MENTS e
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