ML19345B546
| ML19345B546 | |
| Person / Time | |
|---|---|
| Issue date: | 09/22/1980 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-1766, NUDOCS 8012010599 | |
| Download: ML19345B546 (8) | |
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MINUTES OF THE ADVANCED REACTOR SUSCOF;41TTEE 0TM Ig 3 JULY 9, 1980 WASHINGTON, DC The ACRS Advanced Reactor Subcommittee met cn July 9, 1950 to discuss the NRC sponsored research on advanced reactors. The principle topics of discussion were the NRC sponsored research on advanced reactors at ANL, BNL, and th?.
aerosol release and transport work being conducted at ORNL and BCL. Notice of this meeting was published in the Federal Register on June 24, 1950. A copy of this notice is included as Attachmert A.
A schedule ~for this meeting is included as Attachment 3 and a list of attendees for this meeting included as Attachment C.
The complete set of materials provided to the Subcommittee is in the ACRS fil,es. No oral statements were given by members of the public nor were there any requests for time to give oral statements.
No written state-c.c.ts were submitted by members of the public. The ACRS members in attendance me Dr. M. Carson, Subcomittee Chairman and Dr. W. Kerr. The meeting was also attended by Dr. M. First, ACRS consultant and Mr. P. Boehnert, ACRS Staff.
The Designated Federal Employee was Mr. P. Boehnert.
INTRODUCTION - C. Kelber, NRC/RES Dr. Kelber incicated that he viewed the previous ACRS recommendations as being able to be categorized into three groups. They were:
- 1) Analyze a broader spectrum of accidents with less emphasis on the CDA and the CRSR 2)
Increase investigation of the natural convection process
- 3) Review the fuel safety tests needs, program, and testing capabilities.
RES has responded to this in a number of ways. The scope of the accident dura-tion work has been expanded to include the 00E Conceptual Design System reactor.
The SSC effort has been shifted from code development to applicaticn to the analysis of reactor accidents with scram. Studies comparing the characteristics cf heterogeneous and homogeneous cores are currently underway. An increased emphasis has been given to the application and development of SSC and COMMIX to the analysis of the transition to the natural convection flow and to investigations as to the need for a natural convection test facility.
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' Advanced P.eactors July 9, 1950 Dr. Kelber stated that he would support the construction of an impreved natural convection test facility. Fuel safety test needs and existing test capabilities are being reviewed within the context of parameter sensitivity studies. Communications with 00E hase been increased. Dr. Kelber emphasized the accumulated investment in the program and in the development of technology teams and stressed the need for continued funding of these programs.
REACTOR SAFETY MODELING AND ASSESSMENT - H. Hunmel and P. Garner, ANL Dr. Hummel and Dr. Garner summarized the ANL activities in the reactor safety modeling and assessment program. The modeling development involves the EPIC (fuel and sodium motivi.) code and the SIFLO (sodium boiling and voiding) code, cooperative studies with the UK and WAC, large reactor whole core analysis, and fuel pin failure study. The SIFLO work is directed toward the assessment of the radial temperature variation in LMF5R assemblies which could lead to incoherence in boiling and voiding fo the reactor core subassemblies. The EPIC code has been incorporated into the SAS system and is in use in the US and Europe. Future improvements which are being considered are the addition of slate-out and plugging models, the improvement in the axial motion disassembly model and a mechanistic calculation of the clad ripping. The UK/NRC bilateral program involves a comparison of US (SAS/ EPIC) and UK (FRAX-2) codes for the prediction of whole core accidents. This work is essentially complete. The next phase of the work will focus on the prediction of individual accident phenomena. Code comparisons of benchmark problems and transient-over-power and less-of-ficw problems are being performed within the WAC studies. The benchmarking of the BIFLO code has involved comparisons with COBRA-4 and SAS codes and the TH GS and SLSF experiments. COMMIX-1A (single-phase) was also-used to a lesser extent for the benchmarking process. Development is proceeding on the COP;4IX and BODYFIT series codes. The COMMIX-1 and COMMIX-1A (single-phase) analysis codes have been issued. The COMMIX-2 code (two-phass) is under development. The 60DYFIT-1 code is essentially completed and is scheduled to be released by August 1951.
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Advanced Reactors July 9, 19S0 4
SSC DEVEL0pEMNT AND CODE VALIDATION PROGRAMS - J. Guopy, SNL Dr. Guppy sumarized the SSC development effort in the code validation programs at the Brookhaven National Laboratory. The SSC code is designed to model the plant system fluid flow and thermodynamic behavior. The code is one-dimen-sional in nature and includes models of the plant control and protection systems.
It is intended that the code will be validated against tests performed in EER-H, FFTF, and SRN-300. The data from component tests perfomed at LMEC, ANL, and the SRN-3C0 prototype steam generator will also be used. Ccmparisons will also be made with the analytical results from the OEMO, IANUS, COMMIX, and COSRA coder. Work is proceeding on four versions of the codes, they cre:
- 1) SSC-L - simulates short-term (up to 1/2 hour) transients in loop type LMF5Rs
- 2) SSC-P - simulates short-term transients in pool type LMFSRs
- 3) SSC-W - simulates shcrt-term transients in LWRs
- 4) SSC simulates intermediate to long-term (beyond 1/2 hour) transients.
The SSC program has been funded since 1976. The SSC-L code was operational in Septes er 1977, the SSC-P code was operational in September 1979, and the SSC-W code was operational in March 1920. The SSC-S effort is proceeding.
f AEROSCL RELEASE AND TRANSPORT - T. S. Kress, ORNL Dr. Kress summarized the aerosol release and transport progarm being conducted at the Oak Ridge Naticnal Laboratories. The studies are directed toward under-standing the behavicr of fuel in sodium aerosols in the primary vessel and in the containment. Experiments are;. presently being conducted in the FAST facility. This is an approximate one-tenth scale simulation of a LMFSR primary vessel. Small samples of UO2 (typically abcut 20 grams) are electrically heated via capacitor discharge to molten energy states (approximately 4000 joules per gram). The vapcr bubble characteristics dynamic behavior and condensation / transport characteristics in the sodium liquid are to be observed.
Tests have been performed with the fuel vapor source submerged in sodium are i
planned. The argon and water tests have provided the increased understanding of the phenomena and when compared with the sodium tests should provide a means for identifying critica.1 phenomena.
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Advanced Reacter July 9, 1930 The Experiments are also being performed in the NSPP and in the CRI-2.
experiments involve simutaneously burning uranium in sodium to produce airborne mistures of fuel and sodium aerosols. The experiments are designed to provide an imprcved ynderstanding of basic phencmena and to provide a means for experi-mentally verifyin'c *.ne HAARM-3 aerosol code. Experiments performed to date indicate that th( HAARM-3 code is adequate for describing single compor.ent U0s and sedium aerosols. Coagglomeratien of the fuel and sodium aeroscis is evident for mixtures in all proportions and particle sizes. The basic assumptions for scaling experimental results to larger pressure vessels appear to be satisfied.
Future work would be in part directly toward studies of aercsol interactions with vented / filtered containment components and containment coolant systems.
AEROSCL CODE DEVELOPMENT AND OUALIFICATION - J. A. Gieseke, Battelle-Columbus Labs.
Mr. Gieseke described the aerosol code development and verification work being ccnducted at the Battelle-Columbus Laboratory. Work is directed towards the PVRM-3 code, the ZONE code, the CRAB code, and the QUICK codes. The HAARM-3 code has been developed as a reference code. The other code vehicles have been developed to investigate the models of the HAARM-3 code. The IONE code provides the improved geccetric representation. The CRAB and QUICK codes use particle distributions other than the log-normal distribution used in the HAARM-3 code.
The ZONE code models provide for separate mixing regions rather than a single completely mixed region model. Significant special effects were found during the initial part of the transient. Very little special effect was found for longer times. The HAARM-3, CRAB, and QUICK codes produce essentially equiva-lent results. The future efforts will be directed in part towards the study of mixed aerosols and augmented code verification.
BNL LMFBR EXPERIMENTAL PROGRAM - T. GINSBERG, BNL The work Mr. Ginsberg described the experimental work being sponsored at BNL.
is directed toward investigating thermal hydraulic phenomena in fast breeder The presentation focused on the assessment of the transition phase reactors.
It has been concluded that transition phase recriticality phenomenology.
Further events cannot be ruled out on the basis of early fuel disperal.
study needs to be directed at predicting the phenomena associated with the bottled-up-core.
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on file in the Nudeer Regulatory UNEMPLOYWENT COMPENSATION Room 440. Rosslyn. Vtrgitua :200F. (703)
Comm:ssion's Pubhc Document Room located at 1717 H St N.W., WastEngton.
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The Twen*y Sesenth meeting of the Signed at Washirgen. D C tbs tath day Dated this day. lune 16.1seo. at Bethesda-National Commission on Unemployment of June 1960 Maryland.
Compensation is scheduled to be held in James M. Rosbrow.
For the Nuc!rtr Regulatory Commissior.
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During the initial portion of the Advisory Committee on Reactoe meeting. the Subcommittee along with Washington DC to discuss a proposed Safeguards, Subcommittee on 8 any ofits consultants who may be fire protection rule for nucIest power facibties operating prior to January 1.
I Advanced Reactors; Meeting present, will exchange prehmmary He ACRS Subcommittee on views regarding matters to be 1979 (SECY-a>-88). Notice of this Advanced Reacters will hold a meeting considered dunng the balance of the seeiing was published June 20.
In accordance with the procedures on July 9.1950 :a room 1t8?.1717 H St.,
meeting NW. Wa.$.:ngton. DC to continue its The Subcommittee will then hear outlir,ed in the Federal Register on rece-of the NRC funding and prog +am presentations by and hold discussiens Cctober 1.19 9. (44 FR 56408}. oral or direction or prog am termmation. as with representstives of the NRC StafL written statements may be presented by appropriate.for the annua! ACRS their consultants, and other laterested members of the public, recording) will be permitted only during those portions reports to NRC and Cong ess. Notice of this meeting was published June 20.
Further information regarding topics of the meeting when a transcript is being persons.
In accordance with the procedures to be discussed. whether the meetmg kept snd questions may be asked only has been cancelled er rescheduled. the by members of the Subcommittee. its outlined in the Federal Register on Chairman's ruling on requests for the consultants, and Staft Persons desinng October 1.19*9. (44 FR 56408J. oral or opportunity to present oral statements to make oral statements should notify written statements may be presented by and the time allotted therefor can be the Designated Federal Employee as far inembers of the public. recordings will obtained by a prepaid telephone call to in advance as practicable so that I
be permitted ofdy during those portions the cognizant Designated Federal appropricte arrangementa can be made of the meetmg when a transcript is being Employee. Mr. Paul Boehnert (telephone to allew the necessary time during the kept and questions may be asked only 202/634-3:87) between 8:15 a.m. and meeting for such statements.
hy rabers of the Subcommittee.ita He entire meeting will be open to a/ i.s.Itants and Staff. persons desiring 500 pan EDT.
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to make oral statements should notify Dated. June is,19e0.
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Wednesday.fuly 9.1960,1:00 pan.
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SCHEDULE SEQUOYAH SUCCOMMITTEE MEETING ' JULY 9, 1980 8:30 - 8:45' 1. Executive Session 8:45 - 10:00 2. Status of the NPn0 Review (NRC/TVA) 5 min a. Review Schedule 10 min b. Status of non-TMI Open items 15 min . c. Status of TMI Open Items 10 min d. Su=.ary of Design Features for Protection Against Floods 5 min Single Unit /Two Unit Operation Design Features e. 15 min f. Repert on the Special Low Power Test Program (NRC/TVA) 15 min g. TVA Response 10:00 - 11:00 3. Implementation of TM: and NTOL P,ecommendations 60 min (TVA/NRC) 30 min 11:00 - 11:30 4. Report on Sequoyah Nozzle Cracks (TVA) 90 min 11:30 - 1:00 5. Status Report on Ice Condenser Risk Assessment Studies-(NRC-NRR & RES/TVA) Discussion of PAS Draft Study on Ice a. Condansers (NRC) b. Discussion of TVA Risk Assessment Work (TVA) Plans for future Work (TVA/NRC) c. 1:00 - 2:00 LUNCH 60 min 2:00 - 3:00 6. Status Report on Hydrogen Control Studies Discussion of TVA's Work and Proposed a. Interim Measures for Enhancing Hydrogen Control Capabilities (TVA) l b. NRC Response / Status Report on NRC Work 60 min 3:00 - 4:00 7. Status Report 'on Vented Filtered Containment Studies (NKO 30 min-4:00 - 4:30 8. Executive Session 4
s-DRAFT AGENDA J O'3' b i @ d ju O)Ju.1 N ARSR PRESENTATION D 9 T)D I -s a TO ACRS WG-6 ON JULY 9, 1930 8:30 - 8:40 executive SESSION 8:40 - 9:00 INTRODUCTION - C. KELBER, NRC 9:00 - 9:30 REACTOR SAFETY MODELING AND ASSESSMENT - HUMMEL, ANL 9:30 - 10:00 3-D CODE DEVELOPMENT - $HA, ANL 10:00 - 10:45 SSC CODE DEVELOPMENT AND IESTING - GUPPY, 3NL 3REAK (10 MINUTES) 10:55 - 11:15 THERMALHYDRAULIC LMF3R SAFETY EXPERIMENTS - 61NSEURG, ENL 11:15 - 11:40 AEROScL MEASUREMENTS AND N0DELING FOR FAST REACTOR $AFETY - GIESEKE, BCL 11:40 - 12:30 AEROSOL RELEASE AND IRANSPORT FROM LMFBR Fust - KRESS, ORNL %}}