ML19260B265
| ML19260B265 | |
| Person / Time | |
|---|---|
| Issue date: | 09/08/1979 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-1665, NUDOCS 7912070490 | |
| Download: ML19260B265 (15) | |
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/4C45 - /66J DATE ISSUED: 9/8/79 INUTES OF THE ACRS SUBCOMMITTEE ON ADVANCED REACTORS WASHINGTON,:0C
^$ TOT AUGUST 7, 1979
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The ACRS Subcomittee on Advanced Reactors held an open meeting on August 7, 1979 in Room 1046, 1717 H St., N. W., Washington, D. C.
The purpose of thi.s 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 on July 23 and 26, 1979. Copies of these notices are included as Attachment A.
A list of the attendees is included as Attachment 8 and a schedule for this meeting is included as Attach-ment C.
Selected portions of the meeting handouts are included as Attachment D.
A complete set of handouts has been included in the ACRS file. No written state-ments of requests to give oral statements were received from members of the pub-lic. The meeting was attended by Dr. W. Kerr, Subcommittee Chairman, Dr. R.
Savio and Dr. T. G. McCreless, ACRS Staff; Dr. S. Siegel and Dr. R. Seale, ACRS consultants. Dr. R. Savio was the designated Federal Employee. The meeting was opened at 8:30 am with a short Executive Session during which Dr. Kerr su:xaarized the schedule and the goals for the day's meetin3 The meeting was held entirely in open, session and was adjourned at 5:45 pm on this day.
AEROSOL RELEASE AND TRANSPORT PROGRAM - M. SILVERBERG, RES/ARSR Mr. Silverberg gave a brief status report on the NRC sponsored Aerosol Release and Transport (ART) program. The objective of the ART program is to provida data and a verified methodology for radiological consequence assessments. The programs are directed towards providing improved assessments of the source term, i.e., radionuclide release, from the primary systems in a core disassembly accident and the transport and aerasol behavior in containment. Work directed towards the improved assessment of the source term from a core melt may be initiated in the future. The source term experiments are being performed in FAST and CRI-3 facilities at CRNL. The shakedown and calibration tests for the FAST facility have been completed and some of the underwater U02 vaporization tests have been perfarned. Comparis m of the character of aerosols produced in-pile in the ACPR at Sandia and aer;.:ls produced by capacitor discharge vaporization tests are being performed.
It is anticipated that FAST tests with water will be completed in FY S0 and that the FAST tests with sodium will be initiated in FY S0. The capacitor discharge vaporization (CDV; tests are 1515 117 g
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y; Advanced R'eactors August 7, 1979 directed towards characterizing the aerosols produced at various fuel temper-atures. They will be completed in FY 80.
It is anticipated that the acoustics instrumentation for bubble diagnostics will be tested in FY 80.
Development work on aerosol modeling and aerosol codes is being carried out at BCL. Comparisons of the CRAB and HAARM-3 codes are currently in progress.
The BCL aerosol property measurements involving 002 have been completed and the sodium-UO2 mixtures experiments are in progress. The HAARM-3 verification procedure has been released for peer review. The first high mass concentration Na2 x-fuel aerosol test in the NSPP facility has been completed. ARSR is also 0
participating in the aerosol test instrumentation assessment being carried out by the CSNI.
It is expected that the HAARM-3/ CRAB size distribution comparison and the NSPP coagglomeration experiments will be completed in FY 80.
The relationship of NRC's research to other US and foreign work was discussed.
The ARSR is currently comunicating, to some extent, with 00E, France, the United Kingdom, and the Federal Republic of Germany. Full collaboration is being hindered by the current status of the information exchange agreements.
The ACRS recomendations made in the 1978 report to the Congress were briefly discussed. The ARSR is in agreement with the recommandations but indicated that their ability to be responsive would be limited by the presently proposed funding levels.
FUEL FAILURES - P. PIZZICA, ANL ANL is currently engaged in model comparisons with European countries, para-mete: sensitivity studies, and some code development effort which is being carried out within the framework of the SAS code. Loss of flow and transient overpower analysis comparisons are being made in cooperation with the U.K.
Results of these comparisons have shown significant differences. The U. S.
analysis models include the reactivity feedback resulting from fuel motion J\\
Advanced Reactors August 7, 1979 after pin failure whereas the U.K. models do not. The sensitivity studies have indicated that, wnen gap conductance is held constant over the transient, the boiling and clad motion times are not sensitive to different values for gap conductance. Variations in the gap conductance during the transient were not addressed. Differences in the predicted values for the eteady state gas retention and transient gas release phenomena have occurred.. the U.S./U.K.
model comparisons.
Dr. Pizzica indicated that ANL's evaluation had indicated a need for improve-ments in the modeling of transient fission gas release, in the movement of the fission gas in the pin after release, better representation of the loading of the clad by solid fuel expansion, better data on irradiated clad properties, and better estimates of the initial gap sizes. The comparison of calculaticnal methods being made in the WAC programs has produced results similar to the U.S./U.K. comparisons. The foreign participants in this activity used fuel disassembly models of the VENUS type. A number of modeling improvements have been identified as being needed in the SAS 3-0/ EPIC code. They are:
1.
A better mechanistic modeling of clad failure 2.
Improved treatment of ejection of the fission gas under the conditions of weakened clad with little or no molten fuel 3.
Improvements in the treatment of clad melt-through for transients involving the prefailure ejection of the fission gas and for transients involving fresh fuel 4.
The need for the modeling of the thermal expansion of the fuel clad after failure 5.
Modeling of fuel conduction during slow transients and an improved clad temperature calculation during transients involving extended cladding failure and fuel pin disassembly 6.
The representation of plate-out and plugging 7.
Improved treatment cf the latter stages of the rapid assembly transient to the point of neutronic shutdown.
It was noted that the code is presently inade.quate for long transients where pin geometry is lost.
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Advanced Reactors August 7, 1979 Comparisons have been made between the SIMMER and SAS 3-D codes. Results are presently being analyzed and documented. The tentative conclusions are that thr codes in their present state of development yielu similar results.
FRAM CODE DEVELOPMENT WORK - R. CURTIS, NRC/RES Dr. Curtis discussed the work being sponsored at ANL which is directed towards the development of Fast Reactor Analysis Model (FRAM) computer code for the study of the initial transition phase of an HCDA in an LMFBR. Models will be developed to describe the gradual melting (transition) from an intact geometry to a disruptive state. This work is complementary to the DOE effort in the area of the SAS-4 code development. The FRAM calculations will provide input to SIMMER for studies of the long-term behavior of the disruptive state.
It is expected that the first version of FRAM will be a combination of the SAS-30 boiling and voiding models and EPIC and FCC-1. Work has begun on the coding of a new boiling / voiding model (BIFLO) which will include the computation of coolant bypass flow around the iniital boiling region.
It is expected that this model will serve as a basis for cladding relocation calculations.
It is expected that these models will make possible some analysis of the effects of incoherence in voiding and fuel relocation.
BODYFIT CODE DEVELOPMENT - W. SHAO, ANL Dr. Shao sumarized the progress that had been made on the COMMIX and BODYFIT code development programs. The objective of this work is to develop three dimensional transient codes which will make possible the rigorous thermal hydraulic analysis of reactor components under normal and abnormal operating conditions. Emphasis is being placed on the analysis of the natural circula-tion mode. These codes will be used to develop equivalent representations of system cmponents for input into SSC calculations. The COMMIX-1 program is operational and uses a three dimensional, two phase fluid model with non-equilibrium temperatures and inhomogeneous velocity distributions.
The BODYFIT-1 and BODYFIT-2 programs are parallel to the COMMIX-1 and COMMIX-2 programs but will incorporate a transform technique for converting a complex fuel subassembly geometry to a rectangular mesh geometry.
It is expacted that a very significant savings in ccmputer time will be achieved with the BODY"IT-1 and BODYFIT 2 codes. Use of COMMIX-1 and COMMIX-2 codes for design purposes becomes expensive.
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Advanced Reactors August 7, 1979 The COMMIX-1 code has been tested against the SLSF P-2 transient and the SLF P-2 natural circulation transient, German experiemnts involving a slow transient and a seven pin bundle, and in the W-1 LOP-1 pretest predictions.
Comparison with experiments has been quite good.
BNL THERMAL HYORAULIC EXPERIMENTS - T. GINSBERG, BNL Dr. Ginsberg sumarized the BNL experimental programs. The objective of these programs is to investigate thermal hydraulic phenomena of importance in fast breeder reactor safety analysis and to develop a data base which will be used in the validation of the SIMMER code. The work to date has involved the identification of key HCDA phencuena, scaling, and simulant techniques, and the analysis of microwave heating techniques. Experiments have been performed in the area of fuel dispersion, boiling pool heat transfer, multi-phase fue1 relocation, and the role of Taylor instabilities in HCDA energetics. The sig-nificant results of these experiments are summarized on pages 1 and 2 of Attachmant O.
The requirements for the equipment for the microwave heated boiling pool experiments have been defined and a supplier for the microwave system will be sought-in FY 79.
SSC CGDE DEVELOPMENT AND CODE VALIDATION PROGRAMS AT BNL - JAMES GUPPY (BNL),
CHARLES KELSER (NRC/ARSR), AND JOHN MEYER (MIT)
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Dr. Guppy discussed the LMFBR related portion of the SSC code development program. The SSC series of ccmputer codes is designed to simulate the thermal hydraulics of the entire plant system, and is intended to study ocerational and accident transients with particular emphasis on natural convection.
Work is currr.ntly carried out in three LMFBR related versions of the SSC ccde.
They re SSC-L which simulates short-term (up to one half-hour) transients and loop-type LMFBRs, SSC-P which simulates short-term (up to one half-hour) transients in pool-type LMFBRs, and SSC-5 which is intended to simulate intermediate to long-term transients. SSC-5 will simulate heat transfer modes and hydraulic systems not included in the SSC-L and SSC-P versions. Work has been initiated on an SSC-W code which will simulate short-term transients in LWRs. The SSC-L has been operational for nearly two years and is being used in U.S. and foreign LMFBR analysis. The first operating version of SSC-P is 1515 121
Advanced Reactors August 7, 1979 expected to be released in late 1979 and the first operating version of SSC-W is expected in early 1980. Work on the SSC-S code is in the early stages. A description of a typical SSC-L porblem is given on page 3 of Attachment D.
Dr. Kelber discussed the recent decision to produce an LWR version of the SSC code. The initial version of the SSC-W cnde will be directed towards PWRs and will treat only a single phasa in the primary loop. The first steam generators to be modeled will be the once-through designs with later work extending to the U-tube designs. The SSC-W will be managed by the ARSR. The TRAC code will be modified and later used as a systems analysis code with the capabilities for treating two phase flow. Work on the SSC-W code was begun in mid-May of 1979 and it is expected that the ccmpleted version of this code will be available long before the modified TRAC code.
Dr. Meyer discussed computational methods for analyzing natural circulation in the reactor pl:nt. He identified the need for treating low flow transients with a transient rather than with a quasi-steady state method and the need for more analysis of the stratification of flow in pipes in natural convection transients. He noted that axial conduction in the core was not significant except for very low flow (0.1% and lower) transients, and 'that it may prove to be very difficult, if not impossible, to treat heat transf~er between fuel subassemblies. He indicated that he believed that future work should be directed toward evaluating SSC methods, improving the numerical methods in physical models used in SSC code verification by comparisons to available experiments and com-putational codes, ari defining the range of applicability for the SSC code.
INFCE/NASAP PROGRAMS - R. FOULDS, NRC/RES Mr. Foulds briefly discussed the INFCE/NASAP work. Mr. Foulds noted that the RES/ARSR involvement in this work had been small as a re:ilt of the lack of funding. He noted that there are eight working groups in the INFCE dealing with fuel availability, enrichment availability, supply assurances, fuel repro-cessing, fast breeders, spent fuel management, waste management and disposal, and advanced fuel cycle concepts. Reports from the working groups are expected to be issued early in 1980. Six reactor types are being reviewed in the NASAP j g } C.
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Advanced Reactors August 7, 1979 work. They are the light water breeder reactor (three PWR/ breeder pairs),
the liquid metal fast breeder reactor (six variants), the heavy water reactor (a variant of the CANDU reactor); the low enrichment, high temperature, gas cooled reactor, and a gas cooled fast reactor. Comments on DOE draft reports on these reactor concepts were submitted in June of 1979. The DOE report to the President and the C7ngress on this work is expected to be issued on December 24, 1979.
EXECUTIVE SESSION Dr. Kelber briefly summarized the material which had been presented to the Subcommittee and indicated that he would like to discuss his actions on the ACRS recommendations with the Subcommittee and obtain clarification as to the intent of the Committee's recommendations. Dr. Kerr recommended that these be handled by the preparation of written material which would be submitted to the Ful' Comnittee. Scheduling of future meetings was briefly discussed. It was decided that a meeting would be held on November 29 and 30, 1979 in Albu -
querque at which the Sandia work would be discussed in some detail, a status report would be obtained on the LASL work, and a site tour would be arranged of the NRC sponsored experiments at the Sandia Laboratory.
The meeting was adjourned at 5:45 pm on this day.
15i5 i2:
l Federal Register / Vol. 44. No.145 / Thursday. July OS.1979 / Notices 43822 prematurely disclosed would interfere ha. an individual notice published i Subcommittee will discuss implications with the schievement of the purpose fo the eral Register approximately15 of(the accident, including the underlyir which they were prepared: and days r more) prior to the meeting.
cahses contnbuting to it. Not.ce of thi
. Thoso bcommittee and Workin mentting was published July 24.1979.
(4) Material contained in perscen md Dynamics. August 1S-17. Ip9, medical or similar files which is Group m etings for which it is San rancisco. CA.The Subcomnusee disclosed would constitute a clearly anticipate that there willbe a ortion unwarranted invasion of privacy.
or all of the eeting open to public willr view the status of the h!ar and are indicated an asterisk
). It is 11 Boi ng. Water Reactor Contair..ent edures for inspection of Recosds expected that t.. sessions the full Prograins.
Committee mee desig ted by an
- Cordbination of DramicI.o ds.
A public reading room will be /ily between 9 a.m. and 5 p.m %. pen xcept asterisk (*) will be. en whole or in August 22,1979. San Franciseg. CA.The Shturdats. Sundays and holida s. at 499 part to the public. AC. full Committee Ad Hochubcommittee will rekiew the Scyth Capitol Street. S.W, Se meetings begin at 8:3 a.
and topic of mbination of dyn ' icloads Flogr. Washington. D.C., for spection Subcommittee and 5 or Group on strue s, components, nd systems.
of t. nscnpts of Commission earings meetings usually b
'n at 8:3Q a.m. De
'Emerg 7cy Core Coo
- System, arid ineetings. The Commiss n's Public - exact time when ems listed the August U--
1979. Idahopalls. ID:The i
Parti 6ipation Plan desenbe plans for agenda will be cussed durin full Subcommit e will reviev NRC disseininating information ompiled by Committee me 'ngs and when Research Prhgrams on LOFT. Semiscale, the Colnmission for use ' ts study of Subcommitte and Working Grou BEACON. and RELnP air quality issues. Reque for these meetings wi start will be publishe
- Weste Mdnegemen August 23-29.
documeints and for perm /ssion to inspect pnar to ea meeting. Information as 1979 (Tentativ1-). Was, gton. D.C. The other records and matedals should be whether eeting has been firmly Subcommittee'wtll reyiew NRC Waste directed th the Office el Public Affairs.
schedule cancelled, or rescheduled, or hianagement Rkseerd Programs in 499 South Capitol Stredt. S.W.,
whethe changes have been made in the terms of their g6als. Sudgets. and Washington, D.C 20g03, telephone (202) agend for the August 1979 ACRS full prionttes.
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1 245-6355. The Assisttnt Director for Co ittee meettng can be obtained by Public Affails and.M! ministration will a pr paid telechone call to the Office of ACRS Full Comyt ee hieetings make the inilial determination on th ecutive Director of the Committee August 9-n.16?.-A.
- Evaluation of whether the ecord/can be identified and (telephone 202/634-3287. ATTN: 51ary E.
Licensee Event Reports.
whether disclosur( is permitted under Vanderholt) between 8:15 a.m. and 5.CO B. " Review of pfcposed operation of Westinghouse reactors of the Salem this policy. \\/
p.m EDT.
. uckar Generatindstation. Unit 2 class.
A person whose request to inspect a Subcommittee and Working Group C.,Three hill IsNnd Nuclear Station.
record is refusetf may in writing seek
'% g;,p Unit 2 Accident Review of underlymg reconsiderationAof the request by the
'Dree Mile Island. Unit 2 Accident-causes contnb ing tb and implications Director. whose, etermination wtll be Implications Re Nuclect Power Picat -
of. the accident
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Copying of Rel cords final Reactor-Evakiation. \\g-Water D. *La Crosse Boilin Design. July 26-27,1979. Wa shington.
D.C.The Ad Hoc Subcommittee will Every effo[t wt!kbe : ade to address the topic of further ACRS E.
- Pipe cracking in Boiling-Water accommodate requests for copies of review of pending applications for Reactors.
/
l records, su5 ject to the availability of the operating licenses as a result of the F. 'Resoluhon of Antic: pated Commissi n's limited fac:lities.
accident. and specifically. the Salem Transients Without Scram (ATWS) md Nuclear Power Station. Unit 2 which will generic maders related to Light. Water 1
Reactors. /C. ' Disc,tission of a mo{dified Dated; [ y 20.19~9.
serve as a prototype for the near. term wuilam H/ tems. Ir, Westinghouse operating license stage.
""'8 f cus the discusion. Notice of developmint of emergency plans in Director. /
n cor wexm rw-s.
.s -i this mee.
~,. m s r.eu -. c.
support of !!ght. water nuclear power ar
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alants (NMREG-0396). {
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'Advcaced Reactors. August 7.1979.
epte.#er M. :979. A nda to be i
Washington. D.C.The Subcommittee an ounged.
NUCLEAR REG'/LATCRY will contmue its review of matters COMMISSION i
related to NRC sponsored research on ar.o ced
/
l the safety of advanced reactor designs.
Date July :3. m9.
Adv?sory Committee on Reactor Notice of this meeting was published John Hoyle, Safeguards; Proposed eatings uly 23,1g 9.
A# ory C:mmittee.Vce egem :t C"!cer.
!) order to provide admnce m.,
....,.w......es. August 8.19 3, intrmation regatding proposed Washmgton. D.C.The Subccmmittee
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mletings of the ACRS Subcommittees will review proposed regulatory guides sumo coes isso.ai.as ahd Working Groups, andjof the full and revisions to existing regulatory Committee, the following greliminary guides: also,it may discuss pertinent Draft Regulatory Guide;!ssuance and chedule reflects the curretit situation.
activities which affect the current Availability aking into account addittdnal meetings licensing process and/or reactor which have been scheduled and operation. Notice of this meeting was The Nuclear Regulatory Comm:ssion meetings which have been ostponed or published July 24.1979.
has !ssued for public comment a drait of
,f cancelled since the last list f proposed
- Dree Mile Island. Unit 2 Accident-a proposed revision to a guide in its meetmgs published June 31979(44 FR Implications Re Nuclear Power Plant Regulatory Guide Series together with a Design. August S.1979 (Afternoon).
draft of the associated value/ impact 37368). Those meetmgs whi are defimtelv scheduled have or will Washmgton. D.C.The Ad Hoc statement. This series has been 44 I
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ACRS SUBCOMMITTEE ON ADVANCED REACTORS AUGUST 7, 1979 WASHINGTON, D. C.
ATTENDEES LIST ACRS NRC Dr. W. Kerr, Chairman R. Foulds, RES Dr. S. Siegel, Consultant C. Kelber, Dr. R. Seale, Consultant Dr. R. Savio, Staff
- Dr. T. G. McCreless, Staff
- Designated Federal Employee MISCELLANEOUS R. Leyse P. Pizzica, ANL H. Ocmanus, ANL W. Shao, ANL T. Ginsberg, BNL T. Johnson, Statco T. Taxelius, Statco J. Guppy, BNL J. Meyer, BNL 1515 125 ddacA mxd A
A G E fl D A August 7, 1979 - ACRS Review Meeting 8:30 am Introductory Statement by Chairman Kerr 3:45 am Status of Aerosol Release and Transport Programs Silberberg 9:30 am Atil Fuel Failure Studies Pizzica 10:30 am Atil-FRAM (HCDA Transaction Phase)
Curtis/Kelber 10:45 am Attl - Sodium Mixing (30 - Code Development)
Shao 12:00 noon Bril - Thermohydraulics Experiments Ginzberg 12:30 pm Lunch 1:30 pm Bril - SSC 'rograns Gup,ay/Meyer/Kelber 3:30 pm It:FCE/f1 ASAP Program Kelber/Foulds 4:00 pm Executive Session - ACRS Recommendations (Until close of business)
The meeting is anticipated to be completed by 6:00 pm.
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INFCE WORKING GROUPS 1.
FUEL AVAILABILITY 2.
ENRIGIMENT AVAILABILITY 3.
SUPPLY ASSURANCES 4.
REPROCESSING Pu llANDLING, RECYCLE 5.
FAST BREEDERS 6.
SPENT FUEL MANAGEfENT 7.
WASTE MANAGEMENT AND DISPOSAL 8.
ADVANCED FUEL CYCLE CONCEPTS
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NASAP REVIEW SCHEDULE i
e ALL MAINLINE REACTOR PRELIMINARY SAFETY AND ENVIRONMENTAL INFORMATION DOCUMENTS l
(PSEIDs) ARE DUE AT NRC BY 2/9/79 (THE LWR-VARIANT IS IN) e ROUND ONE COMMENTS (ON DOE DRAFT NASAP l'
REPORT) DUE 4/15/79 e
ROUND TWO COMMENTS (ON DOE NASAP REPORT)
DUE; 6/15/79
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NASAP
" MAINLINE" NASAP REACTORS TO BE
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REVIEWED
- 2. LIGHT WATER BREEDER REACTOR (LWBR) [THREE PREBREEDER/ BREEDER PAIRS) 3.
LIO.U I D METAL FAST BREEDER REACTOR (LMFBR)
(SIX VARIANTS)
- 4. HEAVY WATER REACTOR [HWR)(A C.E. VARIATION OF THE CANDU)
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- 5. HIGH TEMPERATURE GAS COOLED REACTOR (HTGR)
G (LOW ENRICHMENT FUEL)
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- 6. GAS COOLED FAST REACTOR (GCFR) w
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ALTERNATIVE SYSTEM RESEARCli PROGRAMS PROPOSED HWR PRESSURE TUBE MATERI ALS EVALUATION SEISMIC ANALYSIS LWBR CORE PERFORfWICE ASSESSMENT GCFR CORE RETENTION ASSESSMENT
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Y INFCE & NASAP ACTIVITIES CURRENT PARTICIPATE IN INFCElWG-8 ACTIVITIES COMMENT ON SAFETY (RESEARCH NEEDS) INVOLVED IN INFCE PROPOSALS f,
MONITOR NASAP ACTIVITIES CONTRIBUTE SAFETY RELATED INPUT (COMMENTS) TO ALTERNATE SYSTEM PROPOSALS DEFERRED EVALUATE SAFETY RESEARCH NEEDS FOR SPECIFIC G
PROPOSALS G
SCOPE OUT TENTATIVE REACTOR SAFETY RESEARCH PROGRAMS EVALUATE ANTICIPATED COST OF NEEDED SAFETY RE-
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SEARCH FOR PROJECTED CONCEPTS
.