ML20148S738

Minutes of the 780814 Meeting of the Nrc/Acrs Subcomm on Emergency Core Cooling Sys(Eccs)Re Review of the Status of NRC Exper Res Prog on ECCS-LOCA Incl LOFT,Semiscale,Two- Phase Flow Instrumentation & Pbf Thermal Hydraulic Aspects

ML20148S738
Person / Time
Issue date:	10/06/1978
From:	Advisory Committee on Reactor Safeguards
To:	Advisory Committee on Reactor Safeguards
References
ACRS-1573, NUDOCS 7812040030
Download: ML20148S738 (122)

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MINUTES OF THE ACRS ECCS SUBCOMMITTEE

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AUGUST 14, 1978 PDA fh/l7Y The ECCS Subcommittee of the ACRS met on August 14, 1978, at the Westbank Motel, Idaho Falls, Idaho. The main purpose of the meeting

- was to review the status of IIRC experimental research programs on ECCS-LOCA. These programs included LOFT, Semiscale, Two-Phase flow in-strumentation, and the thermal hydraulic aspects of PBF.

Notice of the meeting was published in the Federal Register on Friday, July 28, 1978.

Copics of the notice, meeting attendees, and the schedule are included as Attachments A, B and C, respectively.

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The Designated Federal Employee for the meeting was Dr. Andrew Bates.

No requests for time to make oral statements were received from members of the public and no written statements were recieved.

INTRODUCTORY STATEMENT BY SUBCOMMITTEE CHAIRMAN Dr..!sbin, Subcommittee Chairman convened the meeting at 8:30 a.m.,

sintroduced the ACRS members, staff, and consultants, who were present, and' indicated that.the meeting was being conducted under provisions of the Federal Advisory Conmittee Act and the Government in the Sunshine Act.

Dr. 'Isbin indicated that the ACRS was reviewing NRC Reactor Safety Research Programs in response to a requirement for an annual report to Congress. One imper' ant phase of the report will cover LOCA-ECCS and

- many of these research programs are being carried out at INEL.

Dr. N. Ka'ufman, Director of the LOFT Program, briefly reviewed the 7 status of the LOFT schedule and the three year forecast of tests (Attachment D).

As of. August 1978, the project is several months

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ahead' of schedule and power range testing is under way with the 2

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ECCS 8/14/78 reactor. The first nuclear powered test at 8 KW/ft peak power should take place in November or December 1978.

Future tests will take place about once every 5 months.

This schedule will depend somewhatlupon.the_

amount of fuel damage experienced and the time required to decontaminate the system.

Mr. S. Naff, EG&G, reviewed the LOFT L1 Test series (Attachment E).

The last test, L1-5, in the series was completed on April 28, 1978.

The 1.1 test series include not and cold leg breaks with various ECC injection methods. The tests were successful in determining the facility characteristic during blowdown refill and reflood.

Conclusions reached.

after this first series of tests include:

(1)

ECC Delivery to the core inlet is rapid, (2) The lower plenum is not voided, (3) Non-axisymmetric downcomer behavior was obsei ved, (4) 110t wall effects are small, (5) Lower Plenum boiling and entrainment effects are small, (6)

ECC bypass is as predicted for cold leg injection.

In response to a question from Dr. Plesset on the amount of ECC bypass in a 5.5' core vs a 12' core, Dr. Ybarrondo and Mr. Naff indicated that they believe they can make pretty good predictions on the larger core.

Recent tests in Semiscale Mod-3 with a 12'f t core compa,re well with calculations used on the Semiscale Mod-l - 5.5' core.

Dr. Leach pointed out that the LOFT downcomer is 12 ft long and he would not expect a great deal of difference in PWR's with a similiar downcomer length.

Diameter arid gap sizes are more important to the bypass effects.

Mr. Naff reported on the details of the L1-5 test.

This was the first LOFT test with a nuclear core installed.

Other differences from previous tests included a reduced system volume, decreased hydraulic resistance in the simulated pump, and no IIPIS delay.

The test showed good agreement

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'ECCS: 8/14/78 with predictions' implying that the present analytical models are representative of the physical phenomena during a LOCE.

lion-symmetric flow effects were observed in the reactor core during reflood.

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was reflected in. differences in thermocouple readings under the intaM loop and broken loop sides of the core.

1 In response to a question from Dr. Catton, Mr. flaff indicated that there is no question that using Relap to represent multi-dimensional flow in the downtomer presents problems because of the 1-dimensional nature of the code. They have however been favorably impressed with the code resul ts.

In response to Dr. Plesset, Dr. long indicated that he would be going to Germany in September to discuss 360 tests in the Upper 1

' Plenum Test Facility. These tests would help provide data on asymmetric flows. Mr.11aff also indicated that because of the smaller size of LOFT and the stiffness.of the core barrel and support. s.truc_tur.e he_did not..._

believe that data frc,a LOFT hydrodynamic phenomena could be applied to commerical PWR's. Their predictions with the WHAM code agree reasonally well with measurements, with some overprediction of loads and deflections.

Dr. L. Leach, EG&G, reviewed the upcoming LOFT test plans and analysis t

that will be performed (Attachment f).

Tests will be conducted at 2613I/m, 39131/m and 52131/m with various types of ECC injection and assuming various conditions of ECCS functionability. The last test of the L2 series, will be run with some pre-pressurized fuel in the center fuel bundle.

An analysis program has been setup which includes BE and EM pre-test and

. post test predictions with helap-4, Frap-T, Moxy-Score, and TRAC.

All of the tests will have the pre and post test analysis with a BE. version of Relap-4. Tests at the 391;W/m (12 KW/ft) power level will have EM predictions made. At one time there were plans to perform a nuclear power test at 13131/m (4 KU/ft),, this has been cancelled.

The sequence of tests to be run has been based on running low power tests where fuel damcge is not expected prior to tests where damage may occur.

This is to minimize the need for new fuel and down time to cleanup the

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llo tests of UllI are presently planned in LOFT.

A new head structure would have to be built to do the tests, along with the accumulator system.

111EL also expects to make some ' probability studies for the LOFT tests' in order to produce an expected range of temperatures.

It is expected that a consistent model will be used between all of the BE predictions.

A second effort aimed toward code verification and development will be less rigid in control of the computer models,.

Use of the EM calculations will help to determine the amount of conservatism in the models.

All of the conservatisms of Appendix K will be included in these calculations.

The SRL WRAP model will be used for the EM calculations.

Post test EM 3

calculations will be done with the actual initial test conditions and 1.0 Af15 decay heat.

This will help to determine how much conservatism th.e 20% excess decay heat introduces.

The experiment probabilistic analysis will attempt to quantify the degree of uncertainty introduced due to model uncertainties, initial coridition uncertainties, model convergence, and uncertainties in the plant physical description. Multi-dimensional modeling of the LOFT core with Mary / SCORE will also be carried out to determine effects of 3-D flows

.and radiation heat transfer during blowdown and reflood.

Each of the tests' to'be" conducted on LOFT will. have an er.perimental safety Analysis perroteca for it in order to define technical specification limits for the test.

Various system failures-will be postulated and snalyses will have to show sufficient margin and no unacceptable results. The LOFT syst em has a supplementary coolant injection system in case one of the llPB or LPIS systems fa.i'l p

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ECCS 8/1/1/78 Dr. D. McPherson,11RC LOFT Program lianager, summarized the LOFT achievements for 1976-1978 (Attachment G). These included the completion of the L1 test series, core loading and initial criticality, and a great deal of research information which will be covered in a RIL to be issued The RIL should also d'iscuss scaling of Semiscale and LOFT data to commercial PWR's sizes.

Mr. D. Olson, EG&G, presented a brief introduction and status report on the Semiscale program (Attachment II).

Semiscale is contributing to the Research program by providing model verification and development data, it is helping with the LOFT test planning, and it aids in the evaluation of physical scaling effects. The Semiscale Mod-l program provided signi-ficant results on the effect of cold rods in the core, alternate ECC concepts, and the. effect of steam generator tube ruptures.

The Semiscale Mod-3 system has been installed and tests on this system will determine the effects of a 12 foot core,1 and 2 external downcomers, and active versus passive broken loop components as well as provide data on a UllI system.

UllI sensitivity studies should start in December 1978.

In response to a question from Dr. Isbin, Mr. Olson indicated that they were Es5Fe~6f'the~ limitations of the Ilod-3 systeii(with r.espect to the actual, UHI, system in pWR's. These include the single downcomer pipe, a large surface to volume ratio for heat transfer, and a distorted upper head geometry.

Dr. Plesset cautioned that the results of the Semiscalc UHI tests could be quite different than the results one would see in a UllI

' P.WR.

In response to a question from Dr. Catton, Mr. Olson indicated that he belleved ihat 1 continued to have some concerns over the Mod-3 tests due to the distorted geometric ratios in the small system. R may also be concerned over how the data is used and the amount of j

interpretation that may be necessary to try to~ apply it to the 5,-

UllI system.

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ECCS. 8/14/78-Mr.'J. Cozzaul, EG&G, reviewed the results of the steam generator tube rupturetestsinSemiscale(AttachmentI).

These tests were run in order to determine the effect of broken steam generator tubes in ECCS performance and to provide data for evaluating the analysis methods.

Failed steam generator tubes were simulated by injecting hot water in the intact hot leg at the beginning of refill or reflood.

Flow rates were sized to represent a range of 6 to 60 broken S.G. tubes.

It was

-found that from 16-20 broken tubes resulted in high peak clad temperatures due to steam binding and delayed reflood.

The PCT values remained below U

1300 K (2200 F = 1478 K).

A smaller number of brok'en tubes did not significantly delay reflood, a larger number of tubes produced prolonged reverse core flow and good core cooling with the low quality steam generator flow.

Mr. Cozzaul indicated that the Flood-4 code had been used to attempt to model the system.

Results obtained after modifying the code to handle the problem were fairly good.

Mr. T. Larson, EG&G, reviewed the first tests conducted in the Mod-3 Semiscale facility (Attachment J).

These tests are intended to investi-gate the performance of the Mod-3 system and to establish a baseline performance for the UHI tests to be conducted.

Test series 7 will include blowdown-refill tests, reflood tests, integral blowdown-reflood tests, and a small break test.

Tests beyond the series 7 tests will include the UHI tests.

The tests thus far have provided significant data on the system performance.

The results are similiar to the Mod-l tests, however there are some changes which are attributable to the different design.

The study of these differences are providing insight into the core hydraulic phenomena. The initial Mod-3 tests show the effect of the upper head draining into the core during blowdown (no UllIinjection).

The fluid contained in the upper head appeared to provide some core cooling.

Mr. D llanson, EG&G, reviewed the Semiscale Mod-3 UHI testing program to be conducted ( Attachment K).

The objectives of this test program

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~7-8/14/78 include characterizing the thermal-hydraulic phenomena associated.

j uith UllI and determining-the effectiveness of UllI core cooling.

The

.Semiscale - Mod-3 system is designed to represent a ji UllI system.

It includes guide tube, support columns, injection [yorts, and volumes scaled to the UllI system.

Extensive instrumentation is included to determine the system ef fects. Mr. Ilanson described the hardware in detail (see attachment).

The tests will be designed to look at thermal stratification in the upper head, effects of the injection flow rate and quantity,. the initial pressure of injection and the cold leg ECC injection. The first test will be conducted in December 1978.

Pre and post test predictions will La conducted.

Computer codes available for the UllI analysis cre limited.

Relap-4 Mod-5 and Mod-G are not capable of doing the top quenching problem. ' A modified Relap-4 Mod-5 code at Sandia is. having problems with the running time and initialization of the probbi. TRAC and a two fluid COBRA model are just becoming l

available and will be used as soon as possible.

Dr. Catton indicated that the high height to diameter ratio in the upper head and the closeness of the guide tubes and support columns may influence the re-circulation patterns in the Ulti flows.

Mr. Ilanson indicated that they hope to be abic to pick up such.information from the instrumentation available.

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Mr. M. Stanley provided the Subcontnittee with an overview of the Il!EL Instrumentation Calibration Activities (Attachment L).

The Instrumenta-tion Division at IllEL consists of five Jifferent groups which develop, test, build, assemble, and calibrate the instructious for Lorr, semisenic,

'BF, the 2D/3D program, and also for programs at llEDL, BCL, and ORNL.

The Division has 158 employees sith about 1/3 of them at the BS, MS, and pilD levels. The Fabrication and Assembly Labs meet Class A cican room conditions and RDT standards.

Calibration activities concentrate on measurement of two phase fluid mass, energy and momentum transfer i

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-B-8/14/78 with itBS traceable standards, production calibration, and performance calibration.

NBS traceable calibration for single phase flows are based on secondary standards for flow, pressure, and temperature.

Calibration facilities inicuds ARA III at IrlEL, a Semiscale air-water loop, a steam water loop at Karlsruhe and }i Canada (WCL), and a large scale steam-water facility at WYLE.

The flRC has established a Two-Phase Flow Standards Committee to help institute basic standards for 2-phase fl ow.

IllEL has two members on this group.

In response to a question from Dr. Isbin regarding the capabilities and needs for better instru-mentations, Mr. Stanley indicated that he thought that the present measurements of average properties was pretty good; however, some of the detailed local conditions that would need to be measured for the.

advancco codes were still under instrumentation development.

Mr. R. Wesley, EG&G, reviewed the instruments under development at INEL (AttachmentM).

These include an ultrasonic densitometer for use in determining liquid levels; void fraction, and density measurements; mass flow instrumentation using drag disk-turbine meters, pitot tubc.r, and drag screens; s-ingle, double and triple beam gamma densitomater's with stronger source and a photon energy analysis system to allow use in radiation fields; and the development of a optical probe using a high temperature _Stori'jen_s]lj_th' cooling sys. tem foF the opt.ics~The.' Storz _.C 1rns system is also being developed at LASL.

These instruments will be applied to a wide range of experimental programs including the 20/3D tests, LOFT, Semiscale, and others.

Other concepts for instrumentation

.are un er cons ideration and basic development.

In response to.a question d

from Dr. Catton, Mr. Wesley indicated that the response of many of the j

the instruments was flow regime dependent and that a great deal of effort in the calibration progrdm goes into identifying the flow regime in which the instrument is operating.

It is also important to recognize.

that the measurement device itself may disturb the flow and alter the s

flow pattern.

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'ECCS 8/14/78 INEL is not presently looking at neutron densitometers due to the difficulty in-obtaining a strong enough neutron source.

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Mr. J. Anderson reviewed the INEL instrument test results (Attachment N).

The results of calibration tests in the various available flow facilities show that the newly developed devices generally compare quite favorably 1

4 with the reference measurements of flow, density, etc.

Mr. Anderson i

indicated that generally 5% accuracy in the flow measurement is the limit of the.Icipability of the system due to the uncertainties in the test loop flows, temperatures, and fluid densities.

I Mr. J. Droughton, EG&G, reviewed the PBF Test LOC-11C thenual-hydraulic results and predictions for the Subcommittee (Attachment 0).

The LOC-ll tests was run to evaluate PWR fuel behavior during a cold leg break LOCA.

The experiment consisted of 4 single rods, 2 pre-pressurized to 350 and 700 PSIA, and 2 at 15 PSIA.

The rods were unirradiated and of standard 15x15 fuel geometry.

The tests was conducted in the PBF loop with a loop heatup period, power calibration and fuel preconditioning, decay heat buildup, and a blowdown and quench of the fuel.

The test was run 0

at a power of 68.57 KW/m at a coolant temperature of 597 K.

LOC-llc was preceded by LOC-11A and 118 in which temperatures never went above 950 F.

LOC-llc reached a PCT value of about 1000 K.

Pre-test predictions were made with Relap-4 Mod-5 and Relap-4 Mod-6 using the Condre-Bengston transition boiling and film boiling correlations.

Post test Relap-4 calculational improvements were also made to improve the analysis answers.

These changes involved improved system noding and revised gap conductance models in the fuel.

Some of the calculated flows and pressures agree quite well with the data, other calculations are not'so good, in general the PCT values were overpredicted by 100 K.

The major reeson for the overprediction appears to be the calcula'tional logic involved in the i

change over from a low quality, high flow regime to a high quality, low j

- flow regime.

Radiation heat transport may also influence the calculation I

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~10-8/1/l/78 and test results. Additional LOC tests will be conducted,jsome of them will be rod bundle tests.

3 Mr. A. Mekner, EGSG, reviewed the results of the PCM-1 test (Attachment P).

- This teht was intended to drive a rod to failure under stabic film boiling and included a possibility for a molten fuel coolant interaction.

The intent was to produce clad temperatures in the beta phase region of 1200-1500 K without melt and at the some time have fuel melting equal to about 60% of the fuel radius.

Rod peak power reached about 78 KW/m.

DNB occurred at a power level of about 58 KW/m, it had been calculated to occur at 66.5 KW/m (B&W-2) and 66.6 KW/m (W-3).

Clad temperatures reached approximately 1500 K.

It is believed that the rod failed after about 8 minutes in DNB.

It was found that there was relatively little fission gas release from the fuel and apparently there was little communication between the break area and the gas plenum in the fuel while it was hot.

The expansion of the molten fuel probably provided a good sealant.

After shutdown the oxidized fuci clad franmented due to quenching.

Mr. p. MacDonald reviewed the PCM-5 test (Attachment Q).

The PCM-5 test was a 9-rod cluster test in PbF that ran the fuel in a film boiling node.

Protest predications were made using TODEE COBRA IV, and FRAP-T4.

The onset of DUB in this test also occurred about 12% below the expected value calculated by the B&W-2 model.

The actual power may not be known as well' as they would like, cladding temperaturco reached about 1300 K during the test.

The rods went into film boiling at different times and stayed there for from 1 to 11 minutes.

The first rod appeared to fail after about 5 minutes in DNB.

There was no evidence of any pressure pulses or failure propagation.

Rod bowing may also have contributed to the premature DNB during the

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1 ECCS 8/14/78 Dr. Isbin thanked the participants and adjourned the meeting at 5:45 p.m.

More details are available in the meeting transcript available from ACE Federal Reporters, 444 North Capital St., N.W., Washington, D.C., or from the NRC Public Document Room,1717 H Street, N.W., Washington, D.C.

A complete copy of the slides used is on file with the ACRS record copy of the mintues.

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26162 cnd 30631, Jun 16 cnd July 17 a all file sasupplerrant to the petiti:n The petit!tn:r requests th2 C:mmiss nterv;n? whicti must include c ist kn to tm:nd sectirn 31.11, cen:ral If/.1978, respectiv;ly.

of

'2 contenti ns which ars soug to cansa fir 'use cf byprIduct matIr131 In accord nc2 with th2 procedures be ' Ligated in th3 m*.tter, and th3 fo certain in vitro clinical cr labop, outlinid in tha FrDcRAL Rrcisten on base. for each. All petitions w[Il be to testing. to include veterinari;tns October 31; 1977 (42 FR 56972), eral or acted upon by the Commission 9r the as eneral licensees. The petiti der written statements may be presented Lice g Board, designated ts stateh that:

f by members of the public. recordings Commi si:n or by the Chafr "y the

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will be permitted only during those an of h

portions of the meeting when a tran-the A mic Safety and I ensing u

teri ar et cib e er script is being kept, and questions may Board I nel. Timely petitionf will,be ister o orm AEC-483.Jor in vitro esting consid:r to determirn 'wpether a under th[ie terms of the general licen[se pro. be asked only by rnembers of the sub-hearing s ould be notlad of another Wied for"In section 31.11 of 10 CFR Part 31.

committee, its consultants, and staff.

approprial order issued te 'rding the Rather vdterinarians must request A specif. Persons desiring to make oral state.

dispositir i the petitions ic byprodu'g,t material license on fofm AEC-ments should notLIy the designated.

313. It is also my understanding that the fee Federal employee as far.tn advance as In th7 4 L that a hea ing is held f r the specific byproduct licedse will be practicable so that appropriate ar-and a persor s permitted o intervene, that person %ecomes a party to the 8fs unt(tYalntNg c$1 pa$

rangements can be made to allow the p

necessary time during the meeting for proceeding has a right to partic!-

and upgradg{e their diagnostic faculties considerably.. beneve it isg hindrance to such statements.

pate fully in t e conducJ of the hear-ing. Far ex le, that person may progress to recutre a differynt license thm The agenda for subject meeting pr:sent eviden.e and/ cross-examine that extended, to physicians. The small shall be as follows:

Ia Monder, Agusf u, afa; a:n c.m. untu tAe yfn

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hy ic sp Wnesses.

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. conclusion e/ business, i

A copy M D ERAr. Rmsm dioimrnunoassayh veten(f irnply a similar The subcommittee may meet in executive wou1

'I N tice is avallat(e i public inspec-type !! censure fo narians. Would you tion c.t the Commlssi n's Public Docu* please consider hwingJhis type of licensure ausion, with any of its consultants who may mint Room. 17 fH Street NW for veterinarians ilsott

.be present, to explore and exchange their

% 7 preliminary opinions regarding matters Washington, D.C and at the local ing is available for,etition for rulemak-A copy of theAp which should be considered during the tooms at the Public public inspection in meeting and to formulate a report and ree-Public Document Library cf Charlottdand Mecklenburg the Commissiori) Public Document ommendations to the full committee.

C unty. 310 NortJi Tryon Street, Room.1717 H Stzeet, NW., Washing-At the conclusion of the execut!ve session.

Charlttte, N.C. 28202, between the ton, D.C. A copy Qf the petition may the subcommittee will hear presentations by be obtained b writing to the Rules and hold discussions with representatives of

.m. weekdays,9 h urs of 9 a.m. an 9

and Procedur[s Bfanch, Dirlslon of the NRC Staff, the Idaho National Engf.

a.m. and 6 p.m. on a rday and 2 p.m.

neering Laboratory (INELL and their con-cnd 6 p.tn. on S n

, and at the Rules and Re ortb, Office of Adminis. sultants, pertinent to the above topics. The

- Spring Street, Wa}!brhry, 201 South Oconn County I tration, U.S. Nuclear;Jtegulatory Com-subcommittee may then caucus to deter-lhalla. S.C. 29691.

missi n.WasJtington.-D.C. 20555.

mine whether the matters identified in the between the hou *s of'h an.m. and 9 10 a All persods who sire to submit initial seulon have been adequately coveyed,

p.m. en Monday D a.

d 5 p.m.

written comments or ggestions con-and whether the project is ready for renew Tu;sday throug Fridqy, and 9 a.m.

cerning th,e petition pr rulemaking by the full comm!stee.

and 12 noon on jSaturday. The Com-should sei;1d their co!'ments to the Further information regarding i

misslin has arranged ib,r other docu-Secretaryjof the Comminion, U.S. Nu-topics to be discussed. whether the l

ments and corresponden'ce relating to clear Regulatory Commission, Wash-meetirig has been cancelled or resche-i thi proposed ar6endmerit to the Spe-ington D'C. 0555. Attention: Docket-duled, the chairman's rulin; on re-clal Nuclear Material License No.

Ing and. Service Branch. Bk September quests for the opportunity to prasent j

LNM-1773 to bd kept at e same loca-oral statements and the time allotted tJ;ns.

26,197.

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therefore can be obtained by a prepa!d i

Da at Washington, D.

this 21st telephone call to the designated Ped-Dated at 11ver Spring, Md., this day off Ji.tly,1978-14th day of uly,1978.

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eral employee for this meeting. Dr.

Fo the Nuclear Regulatory Com* Andrew L. Bates, telephone 202-634-For th2 uclear Esgu[atory Com-

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h m!ssion-g SA!GII. J. Cut '

'tICHARD W. STAROsTEcKT, Secretaryoffhe Commisst chi Tuel Reprocestiny and Re*

I-Doc. 78-20898 Filed 7-27-18; 8:45 aml

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Dated July 26,1978.

cy le Branch Diviston of Fuel Jomt C. Hoit.t.

l Advisory Committee Cy le and b'sterici Si ety.

o EF11 D 78-20753 Mled 7-27-1 : 8:45 aml

[7590-01]

Afanagement Officer.

ADVISORY COMMITTEE 'ON REACTOR SAFE.

• IFR Doc. 78'-21132 Filed 7 27-78; 9:09 an$3 WARDS, SU8COMMnTEE ON EMERGENCY

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CORE COOLING SYSTEMS (ECCS) i

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(Docket'No. PRM-31-31 ge, ting l

FICE OF MANAGEME

• AND L

R. F. H ACHREINER The ACES Subcommittee on Emer-BUDGET

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gency he QN W W an open 14.1978 at the; Filing of Petition for Rulemakin

. meeting on August EARANCEO EPORT5 N tice is hereby civen that Dr.lR. F.

Westbank Motel Coffee Shop, 475 I

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NzcTreiner by letter dated Jun*e 19, River Parkway, ' Idaho Falls. Idaho 197. has filed with the Nuclear llegu-83401 to review the status of research The followjn. ' a list of requests for la {ry Commission a petition for %ule. projects related to LOFI'. SEMIS-clearance pf repo intended for use

[

ma g to amend the Commission's CALE thermal-hydraulic aspects of in collepting info ation from the i

• • General Domestic the Power Burst Facility (PDF). and 2-publiefeceived by th ffice of Ma>

for Byproduct Material,"(L1-r; lati:n 10 -hase flow instrumentation. Notice of agement and Budget o uly 24.1976 i

ce 3

P:rt 31.

, this meeting was p'ublished at 43 FR (4gU.S.C. 3509). The p ose of pub-FEDERAL REGISTER, VOL. 43, NO.14--TRIDAY, JULY 28,1978

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ECCS SUBCOMMITTEE MEETING AUGUST 14,1978 IDAHO FALLS, ID ATTENDEES LIST _

ACRS NRC H. Isbin, Chairman P. Strom M. Plesset, Member L. S. Tong W. Mathis, Member R. Smith H. Etherington, Member W. D. Lanning I. Catton, Consultant W. A. Paulson T. Wu, Consultant R. R. Landry F. Zaloudek, Consultant G. D. McPherson K. Garlid, Consultant A. Bates, Staff

• US/ DOE J. Austin, Staff i

T. McCreless, Staff W. R. Young A. J. Pressesky

• Designated Federal Employee J. E. Solecki P. E. Litteneker EG&G W. N. Bixby M.' L. McCormick-Barger JAPAN UNIVAC D. L. Reeder K. Saito L. P. Leach W. J. Quapp H. J. Zeile L. Winters A. C. Peterson J. R. White L. J. Ybarrondo S. A. Naff J. Ernest Wilkins J. C. High M. E. Kline J. V. Anderson e

ATTACHMENT B o

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q AGENDA ACRS MEETING 3

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J AUGUST 18,.1978' WESTBANK BANQUET ROOM IDAHO FALLS, IDA110

c M0tIDAY, AUGUST 14, 1978-s H.'Isbin

-8:30 am Executive. Session L. S. Tong 8145 am Introduction to NRC Programs 9:00 am LOFT

.I (15 min.) - introduction and Plant Status N. C. Kaufman S. Naff (40 min.) - L1 Series Summa.ry L. P. Leach i

(40 min.) - L2 Test Series, Analysis Pl us and Schedule (including EM plant)

D. McPherson (10 min.) - Long Term Test Plans 10:45 am

. BREAK 11:00 am Semiscale D. J. Olson (10 min.)-IntroductionandCurrentStatus J. Couzoul.

(20 min.) - Steam Gen'erator Tube Rupture Test Series Summary T. Larson (30 min.)"- MOD 3 Baseline Test Series Results to Date D. J. Hanson (30 min.) - M00'3 UHI Test' Series -and Schedule 12:30 pm LUNCH 1:30 pm Two Phase Flow I.nstrumentation s

M. Stanley (20 min.) - Overview of INEL Instrumentation and Calibration Activities B. Wesley (20 min.) - INEL Instrumentation Development Activities including Semiscale, LOFT, 3D, and Two Fluid Measurements -

)

i J. Anderson (40 min.) - Instrumentation Test Results 2:50 pm BREAK e

e

l

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MONDAY, AUGUST 14, 1978 2

i 3:05 pm DBF

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(30 min.) - LOC 11 A, B, C Hydraulic Performance J. Broughton (20 min.) - PCM-1 P. E. MacDonald (20 min.)-PCM-5 P. E. MacDonald

-l 4:15 Pm DISCUSSION 4:45 pm ADJ0 URN l

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FY 77 FY 78 FY 79~

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~L1-5 power L2-1 sequence load checkout sequence range sequenc6 initial critical and zero power physics testing b

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LOFT 3 Year Sudget Planning i orecast

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(12-78)

(9-79) a (4-30)

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(4-79)

Test

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module module l

period

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period INEL-S-12 406 -

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LOFT NONNUCLEAR

~ T EXPERIMENT SERIES L1 9

BREAK SYSTEM EXPERIMEIJT BREAK BREAK OPENING SYSTEM ECC PRESSURE DESIGNATION SIZE TYPE TIME AP INJECTION (BAR) h i

L1-1 1/2 FULL HOT 17.5 maec LOW COLD 93 BREAK AREA LEG LEG L1-2 FULL BREAK COLD 17.5 msec HIGH COLD 155 AREA LEG LEG (DELAYED)

L1-3 FULL BREAK COLD 17.5 msac LOW LOWER 155 AREA LEG PLENUM L1-4 FULL BREAK COLD 17.5 msec LOW COLD 155 AREA LEG LEG L1-5*

FULL BREAK COLD 17.5 maec LOW COLD 155 AREA LEG LEG

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• WITH NUCLEAR CORE Idah N2ional f'!

Engineering Laboratory I

$ -ff 7Q

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LOFT NONNUCLEAR EXPERIMENTS ECC DEL.lVERY CONCLUSIONS

1. ECC. DELIVERY TO CORE INLET IS RAPID

2. LOWER PLENUM IS NOT VOIDED

3. NON-AXISYMMETRIC DOWNCOMER BEHAVIOR i

IS OBSERVED

4. HOT WALL EFFECTS ARE SMALL

5. LOWER PLENUM BOILING AND ENTRAINMENT i

EFFECTS ARE SMALL 3

6. ECC BYPASS IS AS PREDICTED FOR COLD LEG j

INJECTION INEl.-S-5776 i

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9 Differences Between L1-5 and Other L1 Series Experiments

• Core installed

• Decreased system volume

• PCS pumps left running

• Pump simulator hydraulic resistance decreased

• No HPIS delay INEL-S-11232 y.

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Comparison of Primary System Pressure -

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LOFT LOCEs L1-5 and L1-4, Semiscale S-01-6, RELAP4

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20 T5 M!

- L1-5 pretest RELAP4 1

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L1 -5

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LOFT LOCE.L1-5 Pump Simulator Differential Pressure l

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l L1-4 test data

--- L1-5 test data 53 j

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INEL-S-11244

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LOFT REACTOR VESSEL

-O LIQUID INVENTORY 2500 i

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____L1-4 L1-5 p 2000

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20 30 40 50 60 70 80 90 100 TIME AFTER RUPTURE (s)

INEL-S-10 286

m

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LOFT LOCE L1-5

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j Fuel Clad Temperature at 0.66 m Elevation j

560 l

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540 g

! 520 5

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Thermocouple near intact loop hot leg

-.-. Thermocouple near broken loop hot leg 420 1

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0 10 20 30 40 50

-10 Time after rupture (s)

INEL-S-12 265 j

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Highlights of LOFT Experiment i

s L1-5

~

i

• Control rod insertion more rapid than required

• Core mechanical response less than predicted

• Core instrumentation functioned well

• Decompression faster ~than previous tests (due to smaller system volume and lower pump simulator resistance)

• Bowncomer voiding more rapid than-previous tests

• Refill and reflood.similar to L1-4 j

• ~ Asymmetric flow observed in core

,j i

inet-s,o 2 7

Important LOFT Test Results 1!

q o Good agreement between LOFT and Semiscale implies

=l scaling rationale correct

• Good agreement with predictions implies present analytical models are representative of actual physical phenomena existing during LOCE 7

B, u..

a Important LOFT Test

~?

L~

Results (cont'd) 1

• ECCS data indicates A. No hot wall delay in LOFT 1

B. Condensation not-instantaneous

~l C.. ECC delivery occurs prior to end-1 of-bypass D. Lower plenum does not empty T

INEL-S-11236 e

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'l LOFT Nuclear Tests L2-2 26 kW/m L4-1 26 kW/m Lower plenum ECC L2-3 39 kW/m L2-4 52 kW/m

)

k L2-5 39 kW/m Power-loss i

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L2-6 39 kW/m Pressurized fuel n _ -3 so,'s <

INEL-S,11922 d

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LOFT Nuclear Tests Variables held constant:

1. All double ended cold leg breaks

2. Loss of one LPIS and HPIS 9

3. PWR temperature program d

i INEL-S-11923 3

1

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r8 -

~

Analysis Summary LOFT LOCE's Planned Analysis L2-2 L2-3 L2-4 L2-5 L2-6 L4-1 Best estimate experiment predictions d

with RELAP4, & FRAP-T Evaluation model pretest predictions f

y y

y Evaluation 'model pr 'est analysis y

oor"

'Ith

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q

.ev

.~ analysis y.

y

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9 4{

h.un.-dimensional core modeling with I

d MOXY-SCORE I

Experiment safety analysis y

y y

y y

y j

' TRAC experiment predictions y

y y

y INEL-s-10 378 J

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Experiment Prediction Analysis Scheme RELAP4/ MOD 6 RELAP4/ MOD 6 Hot pin model System Blowdown Model 7-RELAP4/ MOD 6 h(t)

(

-> FRAPT4 B

wdown Medium Pin Model Tg(t)-

n r RELAP4/ MOD 6 RELAP4/ MOD 6 T(t)

System Reflood Cool Pin y Blowdown Model

~

Model

& Reflood INEL-S-12 205

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9 Evaluation Model. Analysis of LOFT LOCE L2-3

)re-Test Analysis Objective:

Quantify the conservatism resulting from application of Apperidix K conservatisms to LOFT Dost-Test Analysis Objectives:

Quantify how m'uch of this conservatism results from the assumptions of initial LHGR,. decay heat and break flow INEL 3-10 363 6

b.

e 6

O' YO C

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Evaluation Model Analysis of k.

LOFT LOCE L2-3 (Cont'd)

Pre-Test Analysis (Appendix K Calculation)

• Uses Savannah River Laboratories' WRAP Evaluation Model' Package

~

~

• Assumes LOCE is conducted without any single failures

• Worst initial conditions including control band.and instrument uncertainties, worst

' breakflow (up to Moody x 1.0), and ANS + 20% decay heat for infinite operation f

INEL-S-12 207 i

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Evaluation Model Analysis of LOFT LOCE L2-3' (Cont'd)

Post-Test Analysis (Intermhdiate Model Calculation)

Two additional calculations will be performed which will be identical to the pre-test analysis except:

o Actual initial conditions 'and best estimate decay. heat will be. used in both calculations o The calculations will have (a) Best estimate

.(b) Worst breakflows INEL-S-10 364

=

LW

=

i ne LZ-4 UUICK LOOK MepOH

--I--L2-4 Experirnental data

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L2-4 BE~ Prediction i

L2-4 EM Prediction x,N'N

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Multi-Dimensional Modeling of the LOFT Core lMOXY/ SCORE was selected because:

e it is available at INEL o it'in~cludes a realistic fuel rod model o it includes a radiation heat transfer model o it includes the RELAP4/ MOD 5 heat transfer package o it includes three-dimensional modeling capibility 11 runs as fast as other available three'-

~

o dimensional codes INEL-S-10 362

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h,

(

l Muiti-Dimensional Modeling of the LOFT Core Analysis Objectives o De'termine maximum guide tube

~

temperatures during blowdown and reflood o Determine core liquid level detector temperature response o Determine the effects of radiation heat transfer on peak cladding temperature during.

blowdown and reflood Determine the effects of three-dimensional c:

flows on peak cladding temperature during-blowdown and.reflood INEL S-10 361 w

Results of MOXY/ SCORE L2-4.

Calculations

-1300 i

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M 1200

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j1100 hia.

- E 1000

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mE 800 I

x No radiation heat transfer, no cross flow ~

m

$ -700

---

No radiation heat transfer, cross flow.

- - -Radiation heat transfer, cross flow 600 I

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0 2

4 6

8 10 12 14 16 18 20 Time after rupture (s)

INEL-S-12 h10 n

s

PRESENTATION TO ACRS ECCS SUBCOMMITTEE r

AususT 14, 1978 p

BY DR. G. D. McPHERSON., LOFT PROGRAM MANAGER ACHIEVEMENTS FY 1976 - 1978 LONG - TERM PROGRAM 0F NUCLEAR EXPERIMENTS 91 f

p y

1

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LOFT ACHIEVEi1ENTS- -

FY 1976 - 1978 COMPLETED ALL NON-IIUCLEAR LOSS-OF-COOLANT EXPERIMENTS (THE ll SERIES)

. - LOADED FIRST lIUCLEAR CORE AND ACHIEVE CRITICALITY-DEMONSTRATED GENERALLY GOOD AGREEMENT WITH' SEMISCALE COUNTERPART IESTS (SCALING)

LOCA CODE CONSERVATISMS IDENTIFIED

^

NO UNEXPECTED PHENOMENA IDENTIFIED

'BY END FY73 ALL START-UP IESTING WILL BE COMPLETED AND i-LOFT WILL BE READY FOR THE FIRST NUCLEAR LOCE (THE L2-2 EXPEBIMENT)

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t ACHIEVEMENTS-FY 1976 - 78 (CONT.)

o ISSUED.A RESEARCH INFORMATION LETTER ON THE L1 SERIES COVERING TOPICS:

DOWNCOMER FLulD BEHAVIOR LOWER PLENUM SWEEPOUT ECC BEHAVIOR AT INJECTION LOCATION

~

RELAPl!, HHAM6 COMPARISONS WITH EXPERIMENTAL DATA SEMISCALE-LOFT-LPWR SCALING:

EVALUATIONS'& TRENDS' SUPPRESS' ION SYSTEM PHENOMENA i

PUMP'AND PRESSURIZER Il0DELING CRITICAL FLOW 4

0 p

i Y

= -

=

LOFT EXPERIMENT PROGRAfi &' EXPECTED SCHEDULE TEST SEQUENCE NO. ExeTS.

PERIOD POWER ASCENSION 5

~

~

11/78 - 2/81 (FULL $1ZE BREAK)

INTERMEDIATE &

3 7/81 - 6/82 St1ALL COLD LEG BREAKS ALTERNATE ECCS 3

1/83 - 3/84

!!0T LEG 3REAKS 2

1984 ATWS 2

1985 STEAM GENERATOR 2

1986 TUBE RUPTURE DISt1ANTLE 1987 9

1 1

Semiscale MOD-1 Program Has Provided Significant Results On

1. Effect of cold rods

2. Alternate ECC concepts

3. Effect steam generator tube ruptures

[?

masu os 1

e

Semiscale MOD-3 Test Program Baseline tests Current status

• Blowdown (3 tests) 3 tests complete o Reflood (2 tests) 1 test complete

• Integral (2 tests?

• Lower plenum irijection (2 tests)

Objectives

~

• Determine effects of configuration change on system behavior.

~

9 12-foot versus 5.5.-foot core

• Active versus passive broken loop

• External versus internal downcomer T

• Establish MOD-3 operational capabilities INEL-S-12 487 s

..--.. A.

e

e MOD-3 Test Schedule

~

u FY-78 and FY-79

• MOD-3 baseline tests - June-Nov 1978

• MOD-3 UHI sensitivity tests - Dec 1978-April 1979

• 2 pipe characterization tests - April-July 1979 (Tentative)

• MOD-3 2 pipe sensitivity tests - Sept 1979-Feb 1980 (Tentative)

INEL-S-12 486 I

---

J------- ----- -.

I

~

[

l MAJOR OBJECTIVES OF STEAM GENERATOR TUBE

~

RUPTURE TESTS i

e EVALUATE THE SENSITIVITY OF PEAK CLAD TEMPERATURE TO THE MAGNITUDE OF SIMULATED TUBE RUPTURE FLOW l

e PROVIDE DATA FOR EVALUATING THE CAPdBILITY 0FCURRENT ANALYSIS METHODS TO PREDICT THE CONSEQUENCES OF A STEAM GENERATOR y.

TUBE RUPTURE F

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EFFECTS OF TUBE-RUPTURE AT REFILLI g

_ON PEAK CLADDING TEMPERATURE 1.

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600 I

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0 10 20 30 40 50 60 70 NUMBER OF TURE RilPTilRFS mn : r..nn-

EFFECTS OF A RELATIVELY LARGE NUMBER OF TUBE RUPTURES AT

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REFILL ON CORE THERMAL RESP.ONSE

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1000 i

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800 E

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NITROGEN FLOW INTIATED g

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ROD REMOVED FROM INTACT I 600 h

LOOP HOT LEG SIDE.

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ROD CLOSE TO INTACT 500 -

LOOP HOT LEG SIDE,

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400 O

O

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'3000 100 200 300 400 500 600 u

TIME AFTER RUPTURE (s) ma.s.ssos N.

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FECTS OF TUBE RUPTU.RE TIME ON AK CLADDING TEMPERATURE AFTER REFLOOD "1500.

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RUPTURES AT~ REFILL 1400 A

RUPTURES AT REFLOOD 1300 s

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800 1

700 9

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-10 I U B E_ R U P T U.R E S..._.....

CONCLUSIONS

~

• EXTREMELY SMALL RANGE OF TUBE RUPTURES PROVIDES POTENTIAL FOR HIGH CLADDING TEMPERATURES NO TEST DATA INDICATES e

CLADDING TEMPERATURES WOULD EXCEED 1350 K

. COLD LEG INJECTION OF WATER o

OR GASEOUS NITR. OGEN

~

INFLUENCES SYSTEM RESPONSE TO RUPTURE FLOW ma.s.eeo; y

+

L Semiscale MOD-3 Test Series 7 Objectives:

• Investigate the specific performance characteristics of the MOD-3 system during blowdown, refill, and reflood

~

• Establish the baseline performance characteristics of the MOD-3 system IN'EL-S-12 502 L

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Response for-Test S-07-1

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Three-Dimensional Views of: Core Thermal.Li J

Response for Test S-07-1 (con't)

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Measured Densities in Upper Head.

Test S-07-1

~1000 g

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Time after rupture (s) inet-s-i2 sos 4

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%GV-154 Comparison of m=

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High Power Zone Cladding Temperatures 1200 MOD-1

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OD-3 S 07-1 (184 cm)

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20 40 60 80 C

Time after rupture (s) ma.-s-n su

Core Thermal Response on Heater Rod C3

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During Test S-07-4 850 3

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800 L...,,

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30 60 90 120 150 g Time after reflood initiation (s) 4)

INEL-S-12 501 N

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FLECHT-SET Test 3105B and MDD-3 f.j Test S-07-4 400 g

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Time after reflood initiation (s) inet-s 12 si4 0('

= - -

Conclusion.s MOD-3 Semiscale system is providing y

important thermal-hydraulic data In-core instrumentation is providing

~

insight to core hydraulic phenomena

• MOD-3 blowdown results are similar to MOD-1 blowdown results i

• MOD-3 reflood results are similar to FLECHT-SET 3105B results INEL-S-12 516 '

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SUPPORT UPPER CORE TUBE SUPPORT

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(SUPPORT HERMOCOUPLE.

%' q' g [ PRESSURE'fNSTRUMENT ROBE PROBE W

(u p

HOUSING) f.v

.i p! =

UPPER PLENUM INEL-S 6415

btiVilbLALC D Y d I CIVI

' MOD-3 VESSEL INTERNALS R-v COLD LEG

'k

,[

HOT LEG

'yh n o

o

, go 4,

T' N

[YU COLD LEG m

HOT LEG

~

/o

'~~:p 1l $

-s LIOUlD y$

}{Y LEVEL S

N!

DOWNCOMER

_ 4:

PROBE HOT 7._.y ANNULUS w.

LEG j

T=

NOZZLE j

11 TURBINE I H i COLD LEG =g 0

P etOwMerER ti NOZZLE 3 4 'a DRAG 4

I s '} / SCREEN SUPPORT DEVICE fI TUBE \\, CONTROL e/U e ROD GUIDE ~* 6 TU BF.:

1. 4 PRESSURE /

p UPPER PLENUM' PROBES UPPER CORE = 4 PLATE >[ LIQUID ,'I Y LEVEL PROBE Ti HEATER ROD DOWNCOMER = g f!" GROUND ' qpo ah.,l.., 2 a INEL-S-6416

N *- SEMISCALE SYSTEM MOD-3 VESSEL INTERNALS 4 4 'T N v. i TURBINE g_ a o 9 FLOWMETER - -f' t ' PROBE = co O O' SG SCREEN G l DEVICE i _ CORE HOUSING TF i ASSEMBLY $SSURE PROBE !g:M ~ 1 <\\; g N. ce l ,4 \\ PRESSURE y i / PROBES i AL BEAM = u

j MMA g:g

/,/ MSITOMETER ~ DRAG l ~ DOWNCOMER e \\ l !!l SCREEN 4 ff \\; L' DEVICE M DIAGONAL l!d j DENSITY ~ ]J g i PORT g \\ .; ij 7l i 'p v 3: LOWER PLENUM THERMOCOUPLE } g [, \\ RACK O O a i DIAGONAL THERMOCOUPLE [#s -T'M ~ (x DENSITY PORT y PORT = gl{' . LIQUID LEVEL HEATER ROD PROBE INEL-S-6417

h e e Application of Analytical Methods in Semiscale Testing Program

• Test specification i
• Test prediction
• Post test analysis-3

'/ ' 9;7, y &. - g ,li.* Q: D INEL-S-1,2 4,45]?.i{ ! { ~ . ~,?0.QR'(($ f' ~ 'INE;$

ll lll l:]l ; 1i11 ~ h s 1 44 2 1 e S -L r E o o N I t a e f d n v ee d i I t eH et o c a Um namhI t u ggo t Hd l s istin e uf~ mUo mo e se r s1 d eh 3 p y s vp y t i r e, n f miI d e a o ru i t u a yH n oy tu t m U eMt r da/ l e eu glo a d o m gn cg l f i at r n o' n pir c nr t t t o ir u i s ao gel ai l S 3 o sl t i r v nmr t v i r u a oa t at och d ph sr n l f o e eao f ae Mimb Tpc Ecb

r instrumentation Division r ~ l l Development & Evaluation instrumentation Assembly Laboratory Laboratory WRRD Advanced Measurement Project instrument Systems ]nstrumentation Branch Engineering. Engineering Branch e 8 PhD's o 25 MS 11 BS 37% electronics o 158 employees 14% mechanics a I 19 Master Techs , 33% sciences o 17 Sr. Techs e 12 Techs 15 Assoc. Techs e 6 Tech. Trainees inet-s-12.247 e l-N

L-1 h-l' f Ml7s. R e us 4-C p$ u m y ac 3 sw gg LWSE I'N, 3m a.., $, 'gt ,.;a. e~ oV Q ra y m _u m ~ ~ wd W q a a arm 4 eV W .=- Q N mc igi ' 1-m ,cc y., u o,, m @g,.. k ..v. g.:3 i o O h A m eu u ca m

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.s 8

?R M .. u O UM g ,b, S u l 4 [;4c i.% f!, (b.R.x ,, ;*!;I) ,.,,,t;;; m N T W A,p i,5: g e w.. = g y,y;, g. m,,,,,7 _..g,,g. 3 .y. c a

y ,_ v 4, l j instrument Production t .i. ~ Calibration ~ ~ L _ _ _.= _ _ _ _,_. _ z _ _ : _. J The establishment of coefficients unique to an individual instrument Spring constants e l I Temperature dependence I o Sensitivity o INEL-S-12 243 {L

__ _ _ ) ~

D instrument Performance i

~ c i Calibration

L m rw-rx-,w,._n.w.- - w.--

.:- - - ~ L.n-u t r _ : .1- - '"EL*f ' 7 8M A*". Y f ; T, i"WmV LMT Ca. ? ' h m W EW,./U A'-

• F3E"-- MET M 133M1U M' I_. 8

'? # U-I EY'M'- -l _. I.1-7. I '_ L-<*. Im s. E l Objectives: Verify the instrument type meets j e F specifications j l Range o e Accuracy Sens.t..ty g e i ivi i ? F Qualify for intended environment 0 e ( e Allow mo'dification for specific geometry l effects due to measurement location s t___.-___m-_.m._.____._._--_ INEL-S-12 244 r-I 4 ~

..i Ultrasonic Densitometer L ng Short Short Helix Bar Bar - I Bar - Il - Principle Attenuation Propogation time 1 Transient 8"*'8" ' Test Steady state steam Autoclave steam Conditions 8I*8I8' water water Liquid level Vessel Liquid level Density / Void fraction Application Void fraction INEL-S-8440 b a r f

a Mass Flow Instrumentation o DTT Pitot Tubes Drag Screen-rag g dy Streamiine Drag body Principle Turbine Stagnat, ion Gamma Gamma Support None instruments Densitometer Densitometer Tested Steady state air water and steam water Conditions Transient steam water Vessel and-Core and P..iping Applicat. ions piping piping l INEL-S-8441 N i va .. ~

H -3 1 l i ^ _i P l , p) - -j JL p.f Qp'&j1 f j,ft)

1. 1. gg,.J f~uel bundle lower 7

tie plate j ) pF*' l d=)P s _._-__R ~,._irip? /,.-\\ \\ r l p./ \\ o r47 fj',,x\\,/ / e y, v ~7 )\\ / \\ O e' /O '\\';ef. -)I r 3$':c;2gj x, (! jf !e L 9 f m _. .gf Drag-disc Utitrasonic ' density 4-e )"@@Q turbine detector l if transducer INEL.S.11 M2 j

^ Gamma Densitometer Semiscale LOFT 3 Beam Semisale Single 2 Beam Beam Non nuclear Nuclear Cs 37 Cs37 CoS8 Cs37 30 Ci 15 Ci 20 Ci ~ "I

• 20 Cl-1 3

4 2 Detector Nai Nal Nai Nal 5x5 cm - 5x5 cm 2.5x2.5 cm 5x5 cm Operating 13 5 9 status procurement INEL-S-8442 4 I N

Gamma Densitometer Semiscale I 3-D PBF Scanning 3 Beam 3 Beam 3 Beam Low Energy Low Energy Low Energy Cd " Cs 37 Cd'" ~ Source Fess Am2*' 60 Ci 0.05 Ci Am24 Gd53 1 3 1 1 Moving Detector Ge Nal Ge Si 1x1.5 cm 2.5x2.5 cm 1.6x0.7 cm 1x1.5 cm Operating 4 2 12 2 Status INEL-S-8445 I i Vg

3D -JAERI Design ^ Preamp

Detector,

~ Collimator. LN DEWAR 6 in. pipe size k i ~ r i y Low energy 3 source ' "k piug ,(erviiium e e INEL-S-8457

1 i L L l Optical Probe Development ~ i Phases tr TM:' : a z r as.sr-v.un t _7 -- U:r.".

cm _ ;

-1.

m. - 1 m w u a v E _-

c i Phase Temp. C Pressure MPa Radiation Application-LASL A 175 0.7 No 3-D EG&G B 350 0.7 No 3-D EG&G C 350 15.2 No Semiscale EG&G D 600/1000 15.2 No Semi /3-D EG&G E 600/1000 15.2 Yes LOFT Dimensions: length - 3 m; O.D. - 25 mm mn-seas Temperature transient AT = 240 C in 50 see Pressure transient AP = 15.3 MPa 2 %)

Mechanical View of.3-meter Long Hopkins Rod Lens Cooled Endoscope ling gas O tet -j -i i i l ,W I If If If II }I ) (- 3, I )I )( II I 8o g a,, m V-ll MT Q _/ [ [ Protective / Fine rod window transition optic's Objective ' Eyepiece end end 3m working length

e m ~ Instruments Not ~ Presented implemented -In development Planned Liquid level Radio-tracer Tomography MDTT rakes PNA Laser Doppler velocimeter Fuel rod instruments Stagnation probe Gamma scattering Basic measuremenis Hydrostatic bearings Holography Conductivity probe Impedance probes 2. ) d INEL-s-12 343

~ Comparison Data for Blowdown 3F.~ 1.1 1 [ - - m-S g0.9 G .tf d 1 m:0.7 E V +0 - Single Beam a [0.5 Ultrasonic. density i, Densitometer = detector 0; 0.3 'h O c i J 0.1 1,,,, c, t -0.1 l 0 1 2 3 4 / ~ R Time (s? ' ""- S - 8

• 1 to

Temperature Sensitivity in Air 350 i i i i O L 300 $ 250 / ..g 200 3 ./A 150 O'A $ 100 50-[ A Sensor 1 1 () O Sensor 2 0 0 50 100 150 200 ~ Relative sensitivity (oio air-water ) '""-S-8 2 i

Screen anci Lassahn Continuous Density Combination ~ j 10 s sa .bd 8 e% e L. B 8 _o 6 5 m m E 4 p 3[* y 2n m 2 mOE nf ~ 0 2 4 6 8 10 ? Reference mass flow rate (kg/s) INEL-S-9371

Typical Ramapo Disc Mass Flow Measurement 14 i i i i i i rh = A(Rp)% ~ R=(p V2) drag disc f Pa Range m

1. 12 Ps a

10 .$g o e 0.4 - 0.7 U . g!: 8 .9 A A 0.04-0.08~ e E m 6 E o (4 [ ~~ a O 2 2 e _, o Ub ^ O' 0 2 4 6 8 10 12 14 t Reference mass flow rate (kg/s). 4 c INEL-S-9370

tr-s. 4 10 0 i i i i M 8 t

3. 3 I

L O 7 ?": UEwa e O.3 m3 0r s m, C I "Q X_ i o G O-O

• _o ot a-e(

3 o

• m Cm I

3 B

• e

-g g oa m ~ -c og 3 oO e 3 c.. mx G '3 r - C +3 l 9 D! O "I ' O i i i O. 1U O n( O 1U O or ot a a )s/gk( etar wolf ssam derusaeM I -l v_

~ ~ ~ ~ ~ yg PITOT TUBE RAKE DIMENSIONS n 0.06350 m U 1 1 0.003810 m h j, 0.01270 m n n 0.01651 m 0.06670 m J i - 1/8 OD SS TUBING h 0.01651 m u I-e 0.01270 m U 0.00381 m i o T h d h0.01110m INEL-S-8116 = = 0.02540- m -

N-7 25 i s i i 20 7 s en x

• 15 E

Ow cc D 10 en i <w 2 5 hb 0"g 0 5 10 15-20 25 REFERENCE, rh (kg/s) INEL-S-8119 e k

LOFT Five Inch Tost Sectiorr 5 inch double extra strong pipe area = 0.0083647 m2 4 e t ? d g s O' Y' 384.56 68.89 S6 T4 T2 T1 T3 S7 S9 g .07_ I 8.2 45.5 S S-77.3 48.4 Y Pipe A 15.8l 6 a O.D.14.12 densitometer I.D.10.32 DTT dia. 3.81 Flow n" n n n n m rir, n r, h J '] 'g u h DTT Pipe O.D. 5.0 Radioactive Radioactive h tracer tracer Density R4, DP9 Velocity (Q4) injections detectors R5,R6 C D3 Momentum (F5) Temp (T10) Test section pressure All dimensions are in centimeters Transversing impedence O e7 O P3 probe INEL-S-11775 i

l 4 LOC-11 Experiment Objectives

• Evaluate PWR fuel behavior during l

postulated PWR cold-leg-break LOCA conditions INEL-S-10 042 { e( o

~ Experiment Description Single rod test - 4 test rods, each surrounded ~ by individual coolant flow shrouds Standard 0.422 in PWR fuel rod geometry -i s" s'

• Unirradiated fuel rods 2 pressurized fuel rods
• 350 and 700 psia 2 unpressurized fuel rods
• 15 psia

~ INEL-S-10 041 \\ R

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o..3 i

i NEUTRON DETECTOR _. WATER l. WMf3h%$jF#g, .idgg@' w kea h .. 4,t.. o.e,. 3., %m. MNiM%,,o:., %s c .y 4Bjs...,., as@A2Wh.b,$ijiR% #Ep "943 ~ s: ):, !pp::%+q>m, e ?dw%:tp> %g-n v;}r;,g ge ss ....w .h. y@M

V.1

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• ~d je cf.C.w, Q;64 g <.

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pqq, 36 %

%q @q $ ? M E F F " $ p %q q n :g A. p n? g %kg.p#d @ i ak 4.ivgy g?". T@spsy myW MD wspy wng ,&ta &- 2k :n i m.yN.uf u v a) e w. .f@.l;..-. pI. W $ Nik...%.fhN,.1; d df 'h;' p&p k n.

v s

v p Ygk [u[g if f Y .4 i '^ D, n .agn.hewhY$gng: W...nudR.$ff,pg ?WQ y h p ps . &@h % AR @RG)W W kW[t %'% $..,c e.,.?5& @.ex? W" @ '..l g~f"fl @bk $Qh 41] Mb&&fkhk fyElMROD@W.:kp l . Ai% l hA s ..ea i Y,Jf f[.N, ..~.,Wgp#4p f fi$g' ive w W'M M Q g g g #g #.,tY dIfsifd i r I ZlRCALOY FLOWT INSIDE EDGE OF j SHROUD IN-PILE TUBE ~ j 4 '--b.-.

t w i LOC-11C l Postlest RELAP-4 Improvements ^

• Break flow multipliers set to 1.0 l
• Nozzle temperature set to 318 K
• Additional piping heat transfer (hanger rods, instrumenation, upper and lower pienums and flow shrouds)

INEL-S-9911 Q ) i

LOC-11C 1 .1 Postlest RELAP-4 1' ~ L Improvements q (cont'd) T Noding of lower plenum - 2 volumes Noding of downcomer - 2 volumes Revised FRAP-T gap conductanc~e values in MOD-5 Use of BEFRAP to calculate gap conductance in MOD-6 t Minor improvements in piping geometry Minor improvements in valve timing ~ ? INEL-S-9910 = =

Twg-:,[ H;.tleg l . Hat ik spool: l b///) - ^~~ y g // ////, '//// //// i /l 4ll' c h. / ////' M 'l 4 T cold I.g '//// //// //// //// 7-/ ,// = z z Cold leg spool a 7 f o 7777 l / 9 /,/ f ~ r / / / / /'/} / / f 7 u t k Ed E New RELAP4 Model of the LOC-T '//)/ k1 . j d siowdown system-r 4 s' s 94 6 p9 3_,,0o 8s Volume 4 4 n g.l i 4 //// Heat slab )_ g INEL-S-8943 p a ////////

~ Ratio of Mass in Core to Mass in System l PBF LOFT Semiscale Core 1 1 1 Upper plenum 96 1.94 3.1 Lower plenum 20 2.25 1.5 Downcomer-110 2.40 2.8 Bypass region 99 0.17 INEL-S-8940 D i N

8 Comparison of RELAP CalculatcJd and ~ l Measured Cold Leg Pressure Data ~- l from LOC-11C 20 i i i i i i 15 er RELAP MOD 6 and DATA / E E RELAP MODS-old 2 10 7.s o 9N k RELAP MODS ') s -,s 5 ~ DATA. s t %.s x.~ H*" "' ""'" 0 o -5 0 5 10 15 20 25 30 i INEL-S-8930 Time (s) ~

0 Comparison of RELAP Calculated and o Measured Hot L~eg; Flow Rate' Data! ~ from LOC-11C 70 i i i i i ,i i i f \\\\ 60 RELAP MOD 5 gI and MOD 6 ./;,A 4 =_ k ( DATA + i 50 q ) i 3 RELAP MOD 6 3 N: I* O-7 40 ?" J ,[../' 30 ~,- f ' / p i, / d 3: 11 f .9 20 I! I if u e' 1 / g 4 / */ i 5 .10 RELAP MOD 5 1 gi x 0 q RELAP MODS-old ..t......- -10 0~ -5 0 5 10 15 .20 25 30 INEL-S-8925

Comparison of RELAP Calculatsd and Measured 0.6 m,90 Clad Temperature Data from LOC-11C 1500 i i i i i. i 1400 RELAP MODS /'"i s,,,,/,,,, s.~~ 1300 2 1200 / ELAP MODS-old f ss' .'X d. -_ _ ,- s _ /-- 2 1100 / a f ./ !/ .- / ..~ ~ ^ = = ^ - 4 1000 8. l l ~i '. t u RELAP MOD 6 E' 900 f/ e l-- 2-l 800 t DATA ((< 700 600 = . ~ ' 500 g -5 0 5 10 15 20 25 30 iNEL-S-8928 Time (s)

v; q 1: =. Conclusions 'Ii

P.

In general RELAP4/ MOD 6 wel predicted.the LOC-11C thermal-hydraulic behavior Predicted and measured coolant break flows were in good agreement Measured core volumetric flows.were slightly different than predicted- ~ Coolant quality was apparently greater than predicted CHF was primarily dependent upon the duration of the initial coolant flow spike INEL-S-10 044 'f ? O

6 Conclusions (cont'd) ~ ~ 4 Measured cladding temperatures were 50 to 100 K less than predicted Radiation heat transfer was significant late in-the transient Incipient cladding ballooning and collapse were observed ~ - l O 1. c' a

~ BPF-LOCA Test Description Summary Blowdown /Heatup ~ ~ Test LO C-3 LO C-5 LOC-6 LOC-7 Rods 4 4 4 4 Irradiated rods 2 2 2 2 4 Internal pressure 350 350 350 350 (psia) 700 700 700 700 Flux shape Flat Flat Flat Flat Type Blowdown Blowdown Blowdown Blowdown j #cil Case D A B C ~ Peak power 17 17 17 17 (kW/ft) Test date March June September March 79 79 79 80 Q INEL-S-9915 s

PBF-LOCA Test Description ~~ Summary Blowdown /Heatup (cont'd) Test LOC-8 LOC-11 LOC-12 LOC-14 i Rods 4 4 4 16 Irradiated rods 2-0 2 0 Internal pressure 350 15 350. 350 350 (psla) 700 700 ~ 700 Flat Cosine Flat Flat Flux shape Type Blowdown Blowdown Blowdown Blowdown Heatup Heatup Case E D E D Peak power 17 17 17 17-(kW/ft) Q f Test date July January October February 4 ~ 80 78 80 81 (- INEL-s-9916 1

e, PCM-1 Test Objectives Extend PCM fuel behavior data base to ~ include: f a Rod failure under stable film boiling at ~ power e The possibility for a molten fuel coolant-t interaction. INEL-S-9629 -f p

PCM-1 2 Specification Constraints Rod power: Range of previous PCM tests & PWR maximum anticipated. power (45-80 kW/m) Clad temperatu~re: S phase (1200-1500 K), no melt Fuel temperature: Substantial molten fuel core (60% of radius) w INEL-S-9627 1--..

' T.HE.RMAL FUELS BEHAVIOR PROGRAM TEST CONFIGURATION SCHEMATIC P-3 COOLANT THERM 0 COUPLE g -g ^ ~ INTERNAL PRESSURE V TRANSDUCER FLOW SHROUD a FUEL R0D lLADDING SURFACE HERM0 COUPLES \\\\ \\ \\ A BYPASS FLOW h A REGION

COOLANT FLOW y

? PATH INEL-S.-4316 BOTTOM OF THE FUELR00 "M LVDT D TURBINE FL'0WMETER k

Test PCM-1 DNB Data ~ ~~ 1 Calculated ~ B&W-2 W-3 Actual Rod power at onset of DNB 66.5 kW/m. 66.6 kW/m 58 kW/m ~ Extent of film 0.41 - 0.41 - 0.25 - boiling 0.91 m 0.91 m 0.85 m FRAP-T calculations Groeneveld film boiling correlation (5.9 eq.) Axial power profile of 1.35 peak to average INEL-S-12 222 p Q I -s s -%w. ,9 4= ,,*r --c-D- -r-e-

Gross Gamma count, i ~ Cladding Elongation and ~ Rod Internal Pressure 16 800-4- i i i i ~ e 15 Cladding elongation ,y i 3 700-p q4g %)E, 2 t 'a. n e

s E

1 l 13 2 f 7 i 81 600- 5.1 i t 12Ei m .c I E .9 o h g r E 500-0 11 e E_1 E. m .o Initial rod failure 10 3 ~ m -2 Rod pressure I g *h 400-S G C \\ l E-e

6 F N "

• W ef" N4 > W

.2 300-y -3 e 8o E 5 f 0 7* Tu 0 -4 pe_ m _ -- - -_m_ x ^ C ~# Gross gamma count =3 ~ 1 1 I I 5 100- 5 0 5 10 15 20 Time (min) ? ,,,t. "\\ l + w---~ --.~,m ~

---

..a...

M

BUILD-5 Calculations of Reaction Laysr i ~ Thickness-0.5-2500 g i ,i i i i l Original cladding thickness = 0.61 mm Cladding temperature at 0.378 mm 0~4 ' 2000 N hot spot (0.41 m) \\ E g . f 1 1747 K - 0.3 e 1500 Ei Total thickness E \\ 0.217 mm O2 stabilized alpha 0.2.$ $1000 thickness \\ . 61 mm F e F-0.1 500 ZrO2 oxide thickness 1 I I I I I I I 0-- g 0 0 1 2 3 4 5 6 7 8 1 h Time (min) inet-s-. 2s 1 -- x.

PCM-1 ~ ~ New Information from Test Cladding did not fail until about 3 minutes after the 17% embrittlement criterion was met. At-power failure location was isolated from the plenum. e - At-power failure was less violent than shut down failure. j No significant molten fuel-coolant interaction occurred. INEL-S-9611 y

.I 1 Test PCM-5 Objectives ~

• Scoping test to provide data on testing rod clusters in the PBF
• To determine the integral behavior of a fuel assembly operating in film boiling
• To determine the behavior of a single rod operated in film boiling surrounded by other fuel rods operating in film boiling.
• To operate the center rod in film boiling

~ L above the beta phase transformation temperature for 2 to 5 minutes. ki INEL-S-11968 I e ~ <w'"-

~ ~' Test Design ~

• Unirradiated 15x15 PWR-type fuel rods i

about 1-m long 1

• 9-rod assembly with PWR lattice spacing i

i l

• PWR environment in the PBF in-pile test loop
• Coolant temperature - 590 K i
• Coolant pressure - 15.1 MPa D

I INEL-S-11967, ' f-

i i 1 Test Design 1 p 1 e l Unirradiated 15x15 PWR-type fuel rods j about 1-m long i I j 9-rod assembly with PWR lattice spacing 1 i L l i ] l PWR environment in the PBF in-pile test j loop f

• Coolar* 'emperature - 590 K I

i e Co< aressure - 15.1 MPa i D I \\ INEL-S-11967 N w _ ,,.-c.-n.__._-......_.-__,,n-_

Sc' hematic of Test PCM-5 Fuel ssembly oc l In-pile tube Flowmeter '> C O -*-flow shroud D 1 Pressure Flow 014 l transducer shroud o4 Thermocouples 55? t 3 ; 4 Flux wires i ! { d SPND > 3 d l Grid ~ < z' 4 spacers t $ % 9-fuel rod ' . t E, s t > R test assembly 6 { ! t t i thermocouples ,,,a-w Flowmeters >CO CO "hermocouples x-o 1 Pressure G# transducer b ~ INEL.-S-11 964 shroud flow Bypass flow

G-Y THERMAL FUELS BEHAVIOR PROGRAM PCM-5 TEST ROD lNSTRUMENTATION 0.480-m 0.580 m W2 U U ] 0.780-m 0.680 m i U@@ LVDT LVDT A . =... bk$-11971

i ns;nman_ rucLa Ucr1A ~ PCM-5 PRETEST PREDICTIONS 1 RAFFLE SET' ENRICHMENT v 1 COBRA 111 C SIZE SHROUD 5 RAFFLE POWER DISTRIB0 TION i.l TODEE . HEAT FLUX R- 0 DISTRIBUTIONS y COBRAIV COOLANT CONDITIONS Y, FRAP-T4 FUEL R0D ? BEHAVIOR g,;;;;;,,,, s '" ZEC's'Te;s ' g!

a.9 THERMAL FUELS BEHAVIOR PROGRAM ~PCM-5 POWER DISTRIBUTION WATER DENSITY 433 Kg/m3 0.988 ~ 1.39 1.167 1.13 1.22 1.31 1.03 1.28 1.12 0.90 1.22 i, f' 0.95 O.80 1.13 0.82 0.78 1.03 i 0.73 0.833 800 0.80 0.90 0 0.988 0.99 1.00 0.833 0.95 1.12 1.167 1.01 0.82 1.31 0.99 0.73 1.39 .dink::.o.. INEL-S-1696

Calculated Cladding Temperature Distribution - Test PCM-5 f; 2060 K 1617 K 1960K 1957 K 1906 K Side rod Corner rod ~ 1617 K 1486 K 1397 K ~ 1486 K 1908 K 1889 K Center rod -Side rod 2060 K O i INEL-S-12 214 ~ - -

i. e

e ~ Onset of DNB for Test PCM-5 L Calculated B&W-2 Actual Coolant temperature (K) 605 590. Coolant pressure (MPa) 15.2 15.1 J'A Rod peak power (kW/m) 52.5 45 - Coolant flow (1/s) 1.9 2.0 INEL-S-12 213

. ' t r

d e e @:b W

~ e -9 Corner Rod Segment 1.0 ii, ,i i I i i 0.9 I D TA i l 1 O O.8 lO! O '/ // r-O.7 3 0 DA i 'l./ i g ' p; oO T- ^ 0.6 E i ? @9a. I L c .eo.s E o DA l $o,4 N O LOFT-1 9g O W-3 i g D^ A B&w-2 l ig lo O A O.3 l s s\\ O DA i b 'A i l N l N 0.1 3 IDNBR = 1.2 gN j o i i, iir i i i i o 0 1 2 3 4 5 '" " -s-12 223 DNBR-

.a-,e i Center Rod i, i; i i i i i i, j 1.0 l O l 040. A 09 O 0.8-I / 0.7 A i l i

0) A

- 0.6 E 5 9k o 0.5 O A z 11 i m O A E 6A O LOFT-1 1 E 0.4 l h O W-3 la6 h .A B &W-2 1% \\ i O A O.3 \\ 1 'h A i 6 'A N j 0.2 g 3 N, h

• • • ==

N i o A I 0.1 l DNBR = 1.2 C "*'Do r i!i i r i i i i o 0 1 2 3 4 5 DNBR INEL-S-12 224 L .i

Rod 1 ] Cladding Temperature Versus Time 1500' 0.68 m ~y v e1300 a A s 1 -y 3 11 l \\ y in s y ft '\\ l D \\ f \\ a 11 1 \\ g (i 1 E 1100 11 i .l ri i e 0.58 m 2: lce}/.%. L Y a \\ I m Y ff. 1 o I jL1 \\'d 00 I 9 t g j v.A g tn I 0.78 m ,g c 1 1 o 700 1 ( } 0.48 m I I' 500' -5 0 5 .10 q Time from first indication of DNB (min) mu.s.u us i i:dM ' ao 5. y'f. r>. r = = ~

~ Rod 1 b Rod Failure Indications 1 Fission product O.. .i 3 gamma / l x ---Cladding l /- (A%g .j elongat,on I. i ......... R o d 1 M9 '!I 4 M ? pressure l \\ I ) i / i I li M t i r m,rc I %,hh i i I ,......................................................,p, I { i .s. i D i ..i. -5 0 5 10 I Time from first indication of DNB (min) INEL-S-11 963 N

e s ~ ca-n Rod Bowing Test PCM-5 I V, ( U~ ~s 's; / ~ s__/ \\s/ . Fractured \\ m pn one m [ k 4 '?\\ l & 's,_ &}6 i 5 t i i 9 7 8 i ' Fractured INEL-S-11 997 I t

~ Summary

• DNB occurred during power calibration at. higher flows and lower powers then planned.
• Center rod was successfully tested at i

beta-phase temperatures, surrounded by 6 fuel rods also in film boiling

• Considerable variation in duration and extent of film boiling occurred.
• Rod failure propagation did not occur.

9 INEL-S-12 355 -}}