ML20148D682
| ML20148D682 | |
| Person / Time | |
|---|---|
| Issue date: | 05/25/1980 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-1695, NUDOCS 8010090178 | |
| Download: ML20148D682 (71) | |
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'"{' I THE ACRS SUBCOMMITTEE ON ADVANCED REACTORS DATE ISSUED:
ALBUQUERQUE, N. M. M 4y ', .' b,.-
hg Lp My,1 4 November 29-30, 1979 W ,' g g The ACRS Subcommittee on Advanced Reactors held a meeting on November 29-30, 1979 at Sandia Laboratories. The purpose of this meeting was to continue the Subcommittee review of matters related to the NRC sponsored research on the safety of advanced reactors. Notice of this meeting was published in the Federal Register of November 14, 1979. A copy of this notice is A list of attendees is included as Attachment B, included as At?achment A.
Selected and a schedule for this meeting is included as Attachment C.
A complete portions of the meeting handouts are included as Attachment D.
set of the meeting handouts is included in the ACRS files. No written statements or requests to give oral statements were received from members of the public. The meeting was attended by Dr. W. Kerr, Subcommittee Chairman; Dr. M. Carbon, Dr. J. C. Mark , Dr. M. Plesset, and Dr. P. ShewTnon, Subcommittee members; Dr. Savio, Dr. T. G. McCreless, ACoS staff, and the ACRS consultants , Dr. I. Catton , Dr. T. Theofanous. Dr. R. Savio was the Designated Federal Employee. The meeting was opened at 8:30 a.m. on November 29 with a short encutive session during which Dr. Kerr summarized the schedule and the goals for the day's meeting. The discussions on November 29 were recessed at 6:15 p.m. and reconvened at 8:30 a.m. on A
November 30. The meeting wcs adjourned at 5:45 p.m. on November 30.
closed session was held between 5:45 p.n. and 6:15 p.m. on November' 29 to i discuss the proposed FY 1981 budget. The remainder of the meeting was l i .j held entirely in open session.
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THIS DOCUMENT CONTAINS POOR QUAUTY PAGES 8 010 0 90 //[f
INTRO,DVCTION - C. KELBER, NRC/RES Dr. Kelber summarized the material which was to be discussed in this two day meeting. ne noted that the accidents of interest for LMFBRs might be divided into.four classes. These are (1) accidents occurring out of the reactor, such as fuel handling and sodium fires and spills, (2) interruptions in heat transport, l (3) systems interactions tha't give rise to unsafe conditions and (4) neutron'per-l turbations/or instabilities. l Accidents involving sodium fires / spills are an area of great interest. The Sandia work includes experiments involving sodiu'm-concrete interaction. LMFBR designers are, in dealing with this problem, avoiding the inerting of containment and mini- ;
mizing the use of steel liners over concrete. The French are building a full scale test facility (ESMERALDA) to demonstrate methods for sodium fire control.
The NRC is currently trying to obtain this information under exchange agreements.
LMFBR designers favor the use of natural convection as a backup heat removal system.
It appears that it would also be advantageous to use natural convection as a normal shutdown decay heat removal process in order to minimize the thermal shock to the system. Natural convection tests have been performed in EBR-2 and two tests are planned for the FFTF reactor. The SSC and COMMia codes are being developed as analytical tools. The capabilities of the SSC code are being extended to treat-single phase heat transfer in water cooled reactor systems.
It appears there are advantages in heterogeneous core arrangements. These arrange-ments increase breeding gains and lower requirements for plutonium inventory. The physics of these cores ha"e not been thore.jhly analyzed and it is expected that further work will be done in this area.
Dr. Kelber noted a good deal of the material to be discussed in this two day meeting would describe fuel testing. Dr. Kelber stated that he believes that an independent NRC effort is required in this area. The perfcrmance of tests and interpretation of the test results are extremely difficult and it is only through the exercise of a high degree of objectivity that an applicant will be able to put the proper interpretation upon his own test results. The NRC reeds to have
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.; s the capability for performing independent analysis and investigation of the copli-cant's claims as to the performance of his fuel design. This capability should permit'the NRC licensing staff to specify clear and unambiguous tests which will scope the performance of specific fuel design. ,
It'is expected that the NRC will enter into cooperative efforts with soad foreign
- researchers. These will incluae involvement in the CABRI project, the Monju licensing review, and cooperation with the French using the information obtained from the 0-series simulated debris bcd test which was carried out during design of an internal core catcher for the Super Phenix.
OVERVIEW 0F THE SANDIA PROGRAM - J. WALKER, SANDIA Dr. Walker indicated that the Sandia program was directed toward providing a technology base for the licensing of advanced reactors in the United States.
Experimental data will be developed and used to describe the governing phenomena, to develop descriptive models, and to improve and verify existing models and codes. ,
The initiating conditions, the potential for energy release, the ultimate coola-bility of the core, and the capability of the containment to withstand damage are key questions. A more detailed listing of the phenomenological uncertainties is given on page 1 of Attachment D.
Sandia is and expects to continue participating in foreign programs. It is involved with the United Kingdom, the French, the Germans, and the Japanese. A more detailed description of the;e programs is given on page 2 of Attachment D.
The areat into which bandia's work is being directed are summarized on pege 3 of Attachment D. The greater part of the effort is going into fuel dynamics studies, debris bed / molten core technology, and containment structural integrity work.
Sandia's experimental capabilities are sumarized on page 4 of Attachment O.
The Annular Core Research Reactor (ACRR) and the core melt sodium melt test faci-lities are the principal experimental facilities.
OVERVIEW 0F 'HE NRC FUEL MELT RESEARCH PROGRAM - M. SILBER 8 ERG, NRC/RES I
Mr. Silberberg noted that the current NRC position with regard to the design l
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w- el j e a e-.4--4.a ar 4L,a ==S4-.e--*l4, a- Jw - 44 i-m<i46 +. aum-.ev Ju-4 :'a 4r ---.a w8 4- u # 8-6 of LMFBRs calls fer. containment features to mitigat'e the consequences.of core melts, e.id.for the~use of steel liners to protect concrete from a. sodium attack during sodium. spills.- These positions have come oat of the Staff reviews of the The NRC fuel melt research program will treat fuel debris bed ,
FFTF'and CRBR. !
behavior, melt interactions, fuel coolant interactions, the radioactive source '
term, systems analysis codes, mitigation feature concepts, and integrated systems '
analysis. The research program will treat both LWR and advanced re ctor phenone-Contain-nology. Work is being performed at Sandia, ORNL, and Battelle, Columbus.
ment systems codes are Sing developed.- It.is ARSR's goal to develop generalized-containment etdes, which will have a broad range of applicability. I OVERVIEW OF THE SANDIA CONTAINMENT IWTEGRITY PROGRAM - R. COATS, SANDIA The Sandia containment integrity program treats debris bed / molten core material ,
behavior, sodium concrete interactions.and development cf analytical methods f The ;
.for analyzing the behavior of the containment under accident conditions. .
The D-4 current program would treat debris bed samples of up to 25 kilograms.
test has recently been completed. Data obtained for this test were used in an The molten core analysis of postulated debris bed behavior in the TMI reactor.
technology subtisk wi'd treat concrete erosien and gas evolution and will look at Funding permitting, ..
the capability of materiils for retaining the molten core.
larger scale facilities will be constructed and used to perform experiments in Sie future.
POST ACCIDENT HEAT REMOVAL STUDIES - R. C0ATS, SANDIA Dr. Coats stated that the object of the PAHR program was to provide verified models of disruptive materials behavior in post accident environments which could be used for safety assessment. The program will address fragmentation and settling, fragmented debris cooling, sintered and molten debris behavior, and high temper-ature behavior and materials interaction. Fragmentation experiments are being The description of the apparatus is given .
performed in a thermite melt apparatus.
on page 5 of Attachment D. Sodium will be poured into the thermite melt. The The summary behavior of the materials under thes3. conditions is being studied.
of the thermite melt fragmentation experiments piirformed to date and the results are given on pages 6 and 7 of Attachment D.
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1 The behavior'.of reactor-materials at high temperatures will be studied in an in-pile (MP'-series) experiment and out-c#-core studies. The D-5 debris bed f experiment nas.been completed ~and some preliminary results obtained. Results of l this experiment are summarized'on page 8 of. Attachment D.
Thedebhisbedexperiments'haveindicatedthatdryoutoccursnearthebottomof the bedI, and uat the dryout threshold decreases with the' increased bed-loading. ,
The-significant time delay prior to'dryout near threshold has been noted.. Dryout threshold has not been foufid to be sensitive to the amount of initial subcooling of the' bed. The tempsrature rise during dryout is much less than adiabatic and >
decreases with time. Equilibrium is approachs2 well below the steel melt tem-perature at threshold power levels. A summary of the debris experiments performed ,
i to date and a description of the appratus are given on pages 9-11 of Attachment D.
The ACRR has the capability for testing ten centimeter diameter debris beds at up to'4.6 kilowatts per kilogram steady state power. !
The MP series experimental apparatus and the test results are summarized on pages 12-14 of Attachment D.
INTERPRETATION OF TEST RESULTS AND FUTURE FLANS - R. LIPINSKI, SANDIA Dr. L1pinski noted that the Sandia program was directed towards expanding the present state-of-the-art to identify unanticipated phenomena and to establishing the requirenents for a test matrix to resolve PAHR issues. Dryout neat flux ;
decreases with bed depth for shallow beds and then becomes essentially a constant ,
value and is insensitive to total bed depth for deep beds. Correlations used to predict the dryout heat flux in shallow beds are given on page 15 of Attachment D. .
The relationship shown on page 15 was developed at Sandia Lab. Dryout for a deep bed is predicted by the relationship shown on page 16 of Attachment D. The dryout l heat flux obtained in the D-1 and 0-4 experiments is shown on page 17 of Attach-
! ment D. One of these experiments cerformed was a small initial subcooling. The other three were performed with large initial subcooling. A factor of 5 reduction in.the dryout heat flux was indicated. lt is believed that this is due to the ,
bed disturbance which results from the initial large subcooling. Effects of top and bottom cooling observed in these experiments are shown on pages 18 and 19 of Attachment D. These modes of cooling have significant effects on the dryout flux.
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j The proposed design for the Super Phenix core retention device is shown on page 20 of Attachment D.
Dr. Lipinski indicated that thev experiments had brought to light unanticipated phenomena. He also noted that Lirdia believes that to be meaningful, experiments must use internal particle heating.
The O series and MP series will be continued. The plant tests are described on pages 21 and 22 of Attachment D. The relationship of these tests tt unresolved PAHR issues-is tummarized on page 23 of Attachment D.
M0LTEN CORE TECHNOLOGY - D. POWERS AND T.Y. CHU, SANDIA-Dr. Chu and Dr. Powers discussed the out-of-pile Sandia core melt program. Melt 0
sizes are from 10 to 220 kilograms vith melt temperatures from 1100 C to 28000C.
A steel /corium mixture is used for the melt material. The concrete crycib'es are basalt, limestone, and CRBR concrete mixture. Reaction times of up to two hours are studied. Cylindrical and hemispherical crucible geometries are used.
The results of these experimetns are sumarized on pages 24 and 25 of Atcachment l
- 0. Gas sources in the experiments have been bound and free water and carbonates-in the concrete. Hydrogen and methane have been observed. Sacrificial material ~
(borax, basalt, and metals) and refractory materials (magnesium oxide, high alumina cement, urania, and firebrick) have been used in the experiments. A surrary of some qualitative results is given on page 26 of Attachment D. .
The future werk will include large scale tests which will lead to a quantitative description of the effects of the core retention materials ud comparisons of in-pile and out-of-pile experiments. The emphasis will be on tests with mag-nesium oxide, basalt concrete and limestone concrete. The large melt facility will have the capability for testing 100 to 500 kilogram oxide melts at tempera-tures from 26000C to 27000C.
SODIUM CONTAINMENT AND STRUCTURAL INTEGRITY - R. ACTON, SANDIA Dr. Acton discussed the sodium concrete interaction work being carried out. This work is directed toward identifying the basic phenomena, quantizing the rate of erosion of the concrete, and identifying the reaction products. Other work is
1 being ca ried out at the Hanford Engineering Development Laboratory and at t',e (
Argonne National Laboratury. The program was initially intended -as a confirma-
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tory research effort. The experiments, however, did not confirm-the earlier findings but indicated that the chemistry of the reaction was more complex than ,;
indicated by earlier work and that there was scaling effects. As a result the program was redirected to be more explcr. tory. The results obtained have been used in the CRBR and FFTF licensing actions. The. work performed to date indi-cates that energetic reactions can occur and that panetretion of the concrete is sodium limited. Generation of large quantities of hydrogen gas and heat have been observed. The reaction products were such that deformation of a steel liner could result. .The experiments done to date have involved up to 250 kilograms of sodium at temperatures up to 1073 degrees. Larger tests (up to 6400 kilograms of sodium) are planned. The test results are summarized on pages 27,28, and 29
-of Attachment 0.
CONTAINMENT CODE DEVELOPMENT - P. PICKARD, SANDIA Dr. Pickard discussed the scope of the CONTAIN code development. CONTAIN is a containment systems code for core melt down accidents which analyzes the post accident sequences from the breach of the reactor vessel to release from the secondary containment. It can treat energetic disassembly /rcactor vessel accidents, and sodium spills between containment cells. The code includes mass and energy transper: in the reactor containment, treatment of the debris bed heat transfer, materials interactions, the effects of cell liners, concrete, and sacrificial material, and the effects of engineered safety systems, such as fire suppression systems, heat removal systems, and vent / filtered devices. It is intended that the code will be a general purpose code and will be applicable to a wide range of containment designs.
1 There are currently a number of codes which treat particular aspects of thc contain-ment system. A summary of '.hese codes is given on paces 30 and 31 of Attachment D.
The status of this development, and descriptions of the individual code elements are given on pages 31 and 36 of Attachment D.
ELEVATED TEMPERATURE PRIMARY SYSTEM INTEGRITY PROGRAM
SUMMARY
- T. WALKER,NRC/RES Dr. Walker summarib d the elevated temperature primary systes integrity program.
He stated that the NRC/ARSR effort has at a minimal level and that research of an expanded secpe was being conducted ta the DOE program. High strain creep tests are'being conducted on 31655 and lower' strain tests (requiring up to three years to run) will be initiated when funding permits. Dr. Von Riesemann discussed the results that have been obtained.at Sandia. Creep / fatigue failure are being run on 2-1/4 CR-M0 and 3165S. The work is directed towerd measuring the creep and fatigue failure. These materials are typical of those likely to be used in the primary and secondary coolant loops on LMrBR demonstration reactors. The work does not duplicate the DOE programs and is cirected toward identifying key safety issues rather than' generating a comprehensive data base.
Uniaxial testing has been completed on 2-1/4 CR-M0 and on 31655 samples (base metal and weldments). These tests were to monitor the damage and to correlate the results with non-destructive examinations.
Eddy current and ultrasonic testing methods were found to have.too low a sensi-tivity and positron annihilation methods were found to be insensitive to disloca-tion effects in 31655 at elevated temperatures. No correlation .etween micro-structural evolution and design margin (fraction of life remaining) could be
' discovered. ,
FY 81 SUDGET DISCUS 710N (CLOSED SESSION) - C. KELBER, NRC/RES Dr. Kelber indicated that no changes in the FY 81* budget had occurred since the
- submission of.the Commissioner's budget to DMB. Input from OMB on the proposed Presidential budget was expected in about a month. The Advanced Converter program is highly uncertain. It is expected that a small amount of money will be appro-priated to the program and that this funding will be too small to conduct a meaningful research effort. The funding received is likely to be used to continue '
the programs that had.been initiated in FY 78 and to provide assistance to the Fort St. Vrain licensing effort.
The discussions were recessed after this sestion to begin on November 30 at 8:30 am,
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-g-ARSR /CCIDENT CNERGETICS PROGRAM OVERVIEtl - R. WRIGHT, NRC/RES Mr. Wright summarized the Sandia in-pile' accident energetic safety research program. He indicated that the program was not directed toward prototypical
. fuel testing-and involved fuel geometries of a smaller scale. The research is i directed towards identifying and measuring key safety phenomena in experiments involving neutronic heating. It is expected that che experiments will result in data.which can be' used for model development ar.e verification and provice a basis for the planning of larger more prototypical experiments. A stamary of the Sandia program and the relationship to key LMFBR safety questions is given on pages 37 and 38 of Attachment'D. Prompt burst experiments in the ACRR using single pins and fresh fuel are planned fur FY 80 and FY 81. Similar experiments utilizing irradiated fuel arc planned for FY 81 and FY 82. Seven pin capsule tests are planned fe; FY 82, and tests in a flowing sodium loop are planned for FY 83.
INTRODUCTION AND TECHNICAL BASES OF THE SANDIA ACCIDENT ENERGETICS PROGRAM -
W. CA"P, SANDIA >
Or._ Camp indicated that this program was directed to identifying the key phenomena in core disruptive accidents. He identified areas of fuel dynamics work potential and transition phase, the work is related to prompt burst.and loss of flow acci-dents. The fuel dynamics experiments treat the behavior of the fuel when sub-jected to the transient. The initial state of the fuel is the intact design con-figuration. The work potential research addresses accidents which do not termi-nate in a benign fashion but lead to an energetic disruption of the core, eith a resultant tnreat to the containment. The transition phase work treats the ccre '
after a loss of design g2cmetry and e loss of coolability-with the potential for .
energetic recriticality. The mechanics of fuel dispersal and fuel coolant inter- <
actions are key que:tions.
FUEL FAILURE AND DISRUPTION M00E - G. CANON AND R. 0 STENSEN, SANDIA Dr. Canon summarized the FO-1 fuel disruption experiments. The mode of fuel l
dispersal is the determinirg factor in the lost of flow accident sequences. This work is desigt.ed to identify key phenomena. The mode and time scale and the fuel disruption as determined by history, initial conditions, and ramp rates are investigated. These are in-pile experiments. They will utilize n'gh speed cinematography and photographic radiometry experimental apparatus show on page 39
e of Attachment D. A typical high power history is shown on page 40 of Attachment D.
Films'of FD series tests were shown. The test pellets fail in the swelling mode and undergo subsequent fragmentation. Neither " dust cloud" or froth formation was observed in these tests. The rapid large scale fuel swelling which occurred is believed by Sandia to be the dorminant early fuel disruption mechanism in a loss of flow accident. The fuel swelling observed is not well predicted by the current FRAS-type fission gas modeling. The FISGAS code has been writ'en to provide improved modeling.
FUEL DYNAMICS PROGRAM - W. CAMP AND T. STALKER Dr. Camp and Dr. Stalker summarized the Sandia Fuel Lynamics Program. The effec-tiveness of the fission gas and sodium vapors in fuel dispersal, the axial fuel motion, and the sodium-fuel dynamics' are key issues. These issues had been pre-viously discussed during NRC licensing actions on 00E reactors. A comparison of the CRBR project and NRC positions is given on page 41 of Attachment D. In the transient overpower accidents, the key uncertainties are the failure location and mode and the fuel sweepout/ freezing and plugging mechanisms. It is noted that a large ramp rate at prompt critical results during a loss of flow accident if a '
large inventory of sodium remains in the low power channels as prompt criticality is approached and if this sodium is preferentially re.aoved from the core relative to fuel following pin failure in a low power channel. To avoid this, a large scale axial expansion, immediate fission gas driven coolant and fuel dispersal, and an efficient fuel sweep out mest be demonstrated. It is NRC's position that these phenomena have not bee adequately demonstrated. ARSR Selieves that the DOE :
program does not address these issues and that the program is formulated from the DOE's position as a reactor developer. The Sandia program is formulated to address these issues. The surmary of the importance of verious transient overpower / loss of flow test parameters is given on pages 42 -44 of Attachment T. The relationship of the Sandia test program to the issues which are to be addressed is summarized .
on pages 45 and 50 of Attachment D.
Fuel motion ditgnostic systems are being develmed at Sandia. In-core detectors and a coded aperature imaging system are being worked on. A prototype of the coded aperature imaging system has been developed and tested. Sandia believes that the technique is sound and is capable of providing high resolution diagnostics.
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The full potential of the system will not be realized until after improvement in the detectors and shielding apparatus. These modifications are underway. Steady state testing of the improved apparatus is expertt:d in February 1960, with fuel motion tests expected in the spring of 1980.
ADVANCED REACTORS - NOVEMBER 29-30, 1979 MEETING - WORK ENERGY POTENTIAL RESEARCH PROMPT BURST ENERGETICS - K. RAIL Dr. Rail sumarized the scope of work potential research meeting conducted at Sandia. The work includes prompt burst energetics experiments, equation-of-state experiments, high ramp rate experiments, and fuel coolant interaction experiments.
The ;.ork on prompt burst energetics is directed toward identifying and charac-terizing the cressure sources (fuel vapor, fuel coolant interactions, and fission product), investigating the work potential, and investigating any other disassem-bly phenomena. The goal is to develop analytical models which describe the observed phenomena. Work is currently directed towards models of fuel pin failure, ,
fuel coolant interaction, and effective equation-of-state. To date only single pin experiments have been performed at Sandia. The apparatus for typical experi-ments are shwon on pages 45 aad 47 of Attachment O. Three experiments have been performed utilizing fresh UO2 in helium, thirteen utilizing fresh UO2 in sodium, and three utilizing fresh UC in sodium. The upgraded ACRR has the capability for testing irradiated fuel pins in bundle sizes up to seven pins. The summary of the already performed prompt burst exp3riments is given on page 48 of Attachment D. It is noted that some potential for fuel coolant interaction has been observed with both the oxide and the carbide system. The fuel vapor. appears to have a significant effect in limiting work potential of the hot fuel and the sodium vapor.
In this size experiment, sodium vapor expansion appet.rs to dominate the work potential. Clad failure has been observed to be due to stress rupture rather than melt through.
The apparatus is used for the equation-of-state experimetns and a summary of the resuits obtained is given on pages 49 and 50 of Attachment D.
Experiments described as coarse predispersed mixtura experiments arc proposed for investigating guel coolant interaction phenomena. The apparatus that would be used is described exceptionally well on pages 51 and 52 of Attachment D. It is ARSR's belief that an.adeouate fuel coolant interaction model does not exist at this time-
1 TRANSITION PHASE - R. OSTENSEN, SANDIA Dr. 0 stensen described the transition phase experiments currently underway at Sandia. These experiments are studying freezing and penetration phenomena.
Crust formation, heat transfer under these conditions and the behavior of the slurry are the key phenomena being studied. Simulant materials are being used in addition to the uranium experiments. Materials interfacs behavior appears to be the area that will require the most investigation. Models have been developed which explain some of the data which has been obtained. Extrapolation from simu-lant materials to actual reactor material appears to be difficult. Serious uncertainties remain in the liquid entrainment phenomena, crust stability phenomena and slurry freezing phenomena.
ACCIDENT DELINEATION - R. CURTIS, NRC/1ES AND P. p!CKARD, SANDIA Dr. Curtis and Dr4 Pickard summarized this part of the work. They noted that the accident delineation was a small effort (two to three man years) and was intended to provide basic guidance to the rest of the program rather than a detailed accident delineation. An effort was initiated in FY 78 at a three man year per level. Phase I of the work which was due to construct event trees for a repre-sentative spectra of LMFBR cecident sequences has been completei Phase II which '
is directed towards establishing the relative importance of the elements of the accid ut sequence and sequence outcome is currently underway.
SIMMER PROGRAMS AND STATUS REPORT - J. SCOTT AND B. WILLIAMS, LASL Dr. Scott and Dr. Williams summarized the status of the LASL work on the SIMMER code. They indicated that SIMMER was a unique effort directed toward describing I mechanistically the transition and termination phases of the core disruptive )
accident. The SIMMER code also describes the initiation phase of the accident and in this respect can be compared with the SAS series codes. The transition phase development of the SIMMER code is currently underway and some parts of this analysis are operational. Some studies treating the recriticality potential of the disruptive core have been carried out. The SIMMER code is currently in use in a number of pigces in the U.S. and in foreign countries. A list of the users is given on page-52'of the Attachment,0. The code verification work is proceeding.
A summary of the status of the verification analysis is given on pages SG-54 of Attachment D. i~
4 ATTACHMENT A explon and eneksase their pr twy'
. epinions regarthag metten which aboah be considered during b meeting and to brmulete a report and MCommenttations to the fuD f"anm!ttes.
At the condusion af the Executive Sasion. Se Subemmittee wiU hear NUCLEAR REQULATORY Presentetions by and hold discussions ColdMISSION - - with repruenutives of the NRC Staff Adytoory CommMtw ea Meac4N and their consultants, pertinent to b Safeguards, SubcommMtee on above topics.
Advanced Reactors, Meeting in edition.it may be necruey to e hold one or more closed sessions as the ,
he ACRS Subcommittee on Mcommitta wul be conside%
f.dvanced Reactors will bold a meeting Portions of the budget and program of ',
on November 29-30.1979 at the San &a ' the Office of Nudear Regulatory Laboratories. Albuquerque.NM to , Rasearts. Since the NRC bodget ,
discuse NRC-sponsored research on the PNP"'I' an ww a part of the -
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.sefery of advanced reactors at the President's budget-eot yet submitta'd to Sandia endIm Alamoslaboratories. Congrew-pubbe diodoenn of y" Notice of this meeting was published budaetary information la not permitted.
See OMB Qreular No. A-la he ACRS.
October 16. str79 In accordance with the(44 FR 80178) dures proce bowevet. is required by Section 5 of tha outlined .n the Federal Register on , d 1878 NRC Authorization Ad to review f t
October 1.1978 (44 FR 56408). oral or the NRC noearch program and budget 5 written statements may be presented by and report the results of its review to members of the public, recordings will Congreaa. In order to parform this be pennitted only during those portions sview, the ACRS must be able to of the meeting when a transcript la being engage to franl; discussion with kept. and qv-stions may be asked only i members of the NRC Staff. For the by members m the Subcommittee,its wr.::n just stated, a discussion would , j consultants, and Suff. Persons dwirtag to enske oral statementi should actify not be possible if held la public sowlon.
I have do; ermined, thersfact, that it is the Designated Federal Employee as far ,
asceseary to does one oc man seesions is advance as practicable so that j at this meeting to pavent frustration of ,
appropriate arrangements can be made ,
this aspect of the ACRS* statutory to allow the neceuary time durmg the l
, l posponsibilities. In accordance with meeting for such statements.
The agenda for subject meetlog shau Exemption (e)(3) to the Government in i the Sunshine Act (552b(c)(9)(B)). :
be as fouows. l Purthn information ngarding teples
%redsy and Fr;dsy. November 2-3c.tre to be discussed whether b muting u e en until the coru.lusica of busines, has been canceUed or rescheduled, the j each day, Chh's % on aqunts for the i ne Subcofninittee may meet in opp trunity to present oral statements Executive Session, with any of its and the time allotted therefon can be g ,
consultants who may be present, to obtained by a prepaid telephone call to ,
h the Designsted Federal Employee for '
this meeting. Dr. Richard Sevio (telephone anlax/ue7) beeween tis a.m. and 500 p.m EST.
- Dated. November 7.scs.
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- I Jata C.&yle.
A%ory comunlan uanarwi.nt o1%.e. .
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ATTACRMENT B MEETING DATE: November 30, 1979 SUBCOMtiITTEE MEETING: Advanced Resctors
, ATTENDANCE LIST Gilbert L. Cano Sandia L. H. Rib ARSR/NRC E. L. Gluekler FRSTMC-R. W.. Wright ARSR/NRC Raymond W. 0 stensen Sand?a J. V. Walker Sandia Thomas J. Walker NRC/ARSR M. Silberberg ARSR/NRC
'ARSR/NRC R. T. Curtis NRC/ARSR C. Kelber K. Terry Stalker Sandia B. Burson ARSR/NRC LASL W. Brierburg XFK/Sandia R. Alcouffe M. Young Sandia E. W. Ptrts LASL David A. McArthur Sandia M. Corradini Sandia John Brammer Sandia R. Coats Sandia Milton Clausen Sandia J. E. Smaardyk Sandia Elaine Bergeron Sandia Jim Meyer NRC/NRR Steven A. Wright Sandia T. R. Schmidt Sandia Ron Lipinski SLA J. H. Scott LASL Gerry Mitchell SLA Bill Camp SLA Ken Murata Sandia Al Marshall SLA Ahti Suo-Auttila LASL ,
J. T. Hitch re: S Lt. ,
Dave Williams Sandia C. J. Cagliostro LASL Mike Stevenson LASL M. Schwarz CEA/Sandia ,
L. L. Smith LASL J. Kelly SLA SLA D. J. Cagliostro LASL X. O. Reil l
C. R. Bell LASL D. H. Worledge UKAEA/Sandia l
l M. Senglaub Sandia E. X. Beauchamp SLA l
R. A. Sallach SLA L - - -
a%...-... -
Advisory Comitte? on Reactor Safeguards Subcomittee on Advanced Reactors Meeting .
November 29 December 1. 1979 l W Sandia Laboratories Ccror. ado Club - Kirtland AFB East
- Deta* led Agenda -
- k Nove-ber 29 (Thu-sday1 Welcone and Administrative Matters 0815 - 0830 Executive Session 0830 - 0845 Dr. Kelbu, (NRC) c ARSR Introdu tion and Overview 0845 - 0920 of FYB1 Dudget (CLOSE0 SESSION-Exemption 9)
Dr. Walker Sat.dia Prcere.m Management 0920 - 0950 Braak 0950 - 1000 FUEL MEl.T PiSEARCH - EFFECTS Oil CONTAIMENT Mr. Silberberg, (NRC) 1000 - 1030 Overview of NRC Fuel Melt Research PtsgrsW c Dr. Coats L- Organization of Sandia Containment 1030 - 1045 Program
%. f. A1181 Mr. Rivard 1 Core Debris Behavior:
1045 - 1215 Procra? arid Results to Date Dr. Lipinski 1045 - 1130 . Theoretical Interpretation 1130 - 1215 A. and Future Plans Lunch - Coronado Club
'1215 - 1300 Molten Core Technology: A1218 Dr. Powers .
1300 - 1430 Program and Results to Cate Dr. Keltner I 4
1300 - 1340 Large Melt Facility Dr. Powers 1340 - 1400 Fa ure Program Plans 1400 - 1430 .
Break 1430 - 1445 Mr. Acton Sodium Containment and Stre:.tural I 1445 - 1530 Integrity:A1054
- p ;. A1198 Dr. Pickard ,
CONTAIN Code Development:
{J;l 4
1530 - 1615 Dr. Pickard )
i
. Sandia Containment Program Sumary
,,' 1615 - 1630
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l 2-t Break 1630 - 1645 Elevated Temperature PrMary System Dr. Walker , (NRC) 1645 - 1730 Integrity: All72 ARSR Program Relationships 1645 - 1655 Dr. VonRiesemanr.
Sandia Progra,Results 1655 - 1725 Discussion and Executive Session 1725 ,Coronado Clut No Host Cocktails 1800 e*
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-3 November 30 (Friday 1 ACCIDENT ENERGETICS RESEARCH Executive Session 0800 - 0815 Mr. Wrinht, (NRC)
ARSR Accident Energetics 0815 - 0835 Program Overview Dr. Camp 0835 - 0850 Introduction and Technical Basis of Sandia's Accident Energetics Program: A1016 Fuel Failure and Disruption Modes Dr. Cano 0850 - 0950 Program and Results to Date Dr. 0 stensen 0850 - 0900 Theoretical Interpretation and f 0900 - 0930 Future Picns break 0950 - 1000 Dr. Camp Fuel Dynamics:
1000 - 1130 Technical Justification, Program Needs and Plans, Facilities and Diagnostics 1130 - 1215 Lunch - Coronado Club Dr. Reil Work Potential Research:
1215 - 1315 Prempt Burst Energetics Dr. 0 stensen 1315 - 1345 Transition Phase Dr. Walker Summary of Sandia's Accident 1345 - 1400 Energetics Program Accident Delineation: A1197 Dr. Curtis, (NRC) 1400 - i445 AP.SR Program Structure Dr. Pickard Sandia Program Status Break 1445 - 1500 LASL SIMMER Program Update: A7015 1500 - 1700 Executive Session - CLOSED SESSION- Exemption 9 1700 - 1800 Adjourn 1800 -
I December 1 (Saturday 1 Sandia Facilities Tour 0800 - 1100 l
. .. l c asiuso w AMACNy ACCIDDIT EllERGETICS PHEN 0MEll0 LOGICAL UtiCERTAINTIE i LOSS-OF-FLG E
- EFFECT OF FlSSION GAS RSALON FUEL DISFER
- EFFECT 0F !!A VAPOR STREAMING ON
~
- AX1 AL FUEL EXPA"SIO:i
- MOTICM OF SODIUM VS. !!0T10N OF FUEL
- INTRA PIN FUEL MOTION IPI"S1E'1T O'!ERP0' DER
- FUEL-PIN FAILURE POSITION
- FUEL-PIN FAILURE DYNAMICS 3 - FUEL SWEEFOUT IN CHANNELS
- POTENTIAL FOR FUEL FREEZING AND PL .
S RECRITICElTY "THERE ARE VIRTUALLY NO MEC FOR RECRITICALITY EVENTS,"
EECHA'liCALD!S ASSE"2LY A';D WORK E' ERGY '
C, - INTERFACE OF VENUS WITH SAS3A NOT MODELED
- REACTIVITY AUGMENTATION-MITIGAT
- REACTIVITY EFFECT OFUERELATIVE TO HEAT FU
- AUGMENTATION OR MlTIGATION OF _.
TRANSFER TO SODIUM AND STEEL WNS IN THIS FURTiiER COMPLICATING THE EFFEC ,
C 3aga iF 7;;t tacg cp ex7tain n13t vegiFiCAT ,
VENUS-EXPA"$10N."
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- ELEVATED IEMP. INT. i T 29 ,i 4
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MATERIALS OBJECTIVES, TEST NO.
MPL' DEMONSTRATE CONTAll(MEf4T UO2 PARTICULATE AND TEftPERATURE DIAGNOS-TICS IN EXPERIMEf4T CAP-SULF UP TO If4CIPIENT U02 MELT.
MP2' PRODUCE A SMALL. MOLTEN-UO2 PARTIC'JLATE EXAMINE 002 FOOL REG 10ld.
({ M E L T I flG , FORfiAT1014 0F CRUSTS AllD VOID REG 10f45 MP3* UO7 PARTICULATE / MELT STEEL SUBSTRUCTURE SOCID : TEEL DISK BY FiSS10N-HEATil1G OVER-LYING U02 EXAMINE MOLTEN STEEL BEHAVIOR.
MP4 UO7 PARTICULATE / INVESTIGATE THE EROSION Sotle MG0 DISK RATE OF h30 SUPPORTif4G A FISSION-HEATED BED OF UO2 PARTICULATE.
MP5 MIXED 00 2/ STEEL MELT STLEL PARTICULATE PARTICULATE AND STUDY STEEL AGGLufi-ERAT 10N AND MIGRATION UNDER A TEMPERATURE G R A D I E lit .
MP6 002 PARTICULATE MELT APPROXIMATELY 0 5 KG OF 002 If4VESTIGATE HEAT FLUX P A R T I T 1 0 f4 1 N G A f4 D CRUST FORMATION.
d[ MP7 002 PARTICULATE DETERMlf1E THE THERMAL CONDUCTIVITY OF DRY .
UU2 / STEEL DEBRIS. <
MP8 N!XED UOo/ STEEL DETF.RMlNE THE THERMAL PARTICULIsTE CONDUCTIVITY OF DRY UU2 DEBRIS.
'IEST COMPLETED.
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SUMMARY
DF PAHR UNRESOLVED ISSUES 91NCIPIENT DRYOUT ljNIFORM N A-UO2 BE0 PARTICLE-HEATED SATURATED DRYOUT D3, D9, D11, D14, D15, D16 DEEP g D1, D2, D4-DB, D10, D12, D13 SHALLOW TOP SUBC00 LING D2, D4-D8, DIO, D12, D13 PRE-D!STURBEo D2, D4-118, D10, D12, D13 POST-D1STURBED D9, D11, D13-16 BOTTOM SuBC00 LING D6, D12, D15 STRATIFICATION ePosT-DRYOUT (PRE-MELT)
DRY ZONE THICKliESS D3, D4, D5-D16
) VERSUS POWER AND I!ME DRY 20!iE STABILITY D5, D11, D13-D16 AGAINST S001UM REENTRY MP1, MP2, MP7, MP8 DRY DEBRIS CONDUCTIVITY .
ePOST-MELT MPS, D14, D15 SfEEL MIGRAT:0N gg MP2, D11, F13, D16 V0lo FORMAT 1014 :
MP6, D11, D13, D14 V0la STABILITY MP6, D11, D13, D16 FCI P0TEf4TI AL MP6, D16 .
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TA -- ISSUES L 2, 3, 4 SINGLE PIN MEDIUM STEADY STATE POWER AtID BURt4UP VARY RAMP RATE -
I TB -- ISSUES L 2, 3, 4-SINGLE PIN I
CONSTANT RAMP RATE
' VARY STEADY STATE POWER AND BURNUP l
TC -- ISSUES L 2, 3, 4 .
Sit 4GLE FIN INVESTIGA'E PARAMETERS FOUND MOST SENSITIVE IN TESTS TA AND TB l TD -- ISSUES 5, 6, 7, 8 MEDillM STEADY STATE PCWER, BURNUP AND RAMP COMPARE 1 PIN TO 7 PIN (6 FRESH)
G ,l TE -- ISSUES 5, 6, 7, 8 SINGLE PIN
- MEDIUM RAMP COMPARE HIGH AND LOW GAS CONTENT e -
/
p IASSUMES THAT TD COMPARISON IS SUCCESSFUL
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' ASSUMES THAT ID COMPARIsoft IS SUCCESSFUL O :
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L
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1 O 1 TO BE DETERMlilED TG m
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SINGLE PIN VARY STEADY STATE POWER AND BURNLP LTB -- ISSUES 1, 2, 3 SINGLE PIN INVESTIGATE PARAMETERS FOUND MOST
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COMPARE 1 PIN TO 7 PIN LID -- ISSUES 4, 5, 6 O SINGLE eiN-COMPARE HICH AND LOW GAS CONTENT LTE -- ISSuss 4, 5, 6 SINGLE PIN
- INVESTIGATE SENSITIVE PARAMETERS IF NECESSARY LTF -- iSSus 7 7 PIN MEDIUM BURNUP AND STEADY STATE POWER kh ' ASSUMES THAT LIC COMPARISON IS SUCCESSFUL dk l
0 3
LOF-D-TOP TEST MATRIX EURNUP, %
- Pins S.S. POWER
- SERIEt:.
4 1 M LTA :
H I 1
3 1 L 1 8
1 H 8
1 L 1 TO BE DETERMlllED LTB :
M 4 LTC 1 4
7 .M (O
L 8 LTD 1 '
H I 1 ;
1 TO BE DETERnli1ED LTE 4
LTF 7 M l i
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l
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LA -- ISSUES 2, 5, 6 7 PIN I
MEDIUM BURNUP AND STEADY STATE POWER VARY POWER LEVEL AT FUEL FAILURE LB -- ISSUES 2, 5, 6 7 PIN VARY GAS CONTENT AND POWER LEVEL AT FUEL FAILURE LC -- ISSUES 2, 5, 6 7 PIN INVES*1 GATE SENSITIVE PARAPETERS BASED ON LA AND L3 TESTS (OR MORE PINS?)
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