ML20010J395
| ML20010J395 | |
| Person / Time | |
|---|---|
| Issue date: | 08/21/1981 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-1878, NUDOCS 8109300448 | |
| Download: ML20010J395 (24) | |
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ISSUE DATE: 8/21/81 I
MINUTES OF THE ACRS SUBCOMMITTEE MEETING ON ADVANCED REACTORS WASHINGTON, DC JULY 7, 1981 The ACRS Subcommittee on Advanced Reactors held a meeting on July 1981 in Room 1167, 1717 H St., NW, Washington, D.C. to review the NRC Advanced Reactor Research program. The Subcommittee met with representatives of NRC, Sandia Laboratories, Argonne National Laboratory, Brookhaven National Labora-tories, DOE, and Los Alamos Scientific Laboratory.
Notice of this meeting was published in the Federal Register on June 19, 1981.
A copy of this notice is included as Attachment A.
A list of attendees is included as Attachment B.
The schedule for this meeting is included as Attachment C.
A complete set of handouts is attached to the office copy of the minutes as Attachment D.
The meeting began at 8:35 a.m. with a short executive session in which Dr. Carbon mentioned that the meeting was conducted in accordance with the provisions of the Federal Advisory Committee Act and the Government in the Sunshine Act. He stated that neither written statements nor requests for time to make oral statenents from any member of the public had been received.
The meeting was attended by Dr. M. Carbon, Subcommittee Chairman, Mr. J. Ray, Dr. C. Mark, Dr. C. Sicss, and Mr. M. Bender, Subcommittee members. Mr. E. Igne of the ACRS Staff was the Designated Federal Employee for this meeting.
m Dr. Carbon asked if there were any questions or comments. Therewerenope,
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so Dr. Carb'n called upon the lead speaker, Mr. Kelber.
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U Fast Reactor Program, C. Kelber, RES/NRC g
k Mr. Kelber nientioned that the fast reactor program had been in exist 4
19 since late 1973, and beginning in April of 1917 with President Carter's a
e announcement of a shift in national policy, the program was cut back and
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~ ADVANCED REACTORS 7/7/81 focussed on certain generic. issues. The current belief is that the CRBR will re-enter the. licensing phase and that a significant priority will be placed on the national program of research and development on fast breeder reactors.
Mr. Kelber stated that the objective was to revive the LMFBR program on a schedule-that would allow it' to contribute significantly in fiscal year 1982, to come up to full speed.n fiscal _ year 1983, and to enhance the responsive-ness to.NRR needs as '. hey are developed.
He identified some of the _possible candidates for high priority work as core-melt accident surveys for use in risk analysis and systems analysis.
Mr. Kelhar listed four areas of concern:
- 1) Heterogeneous Core - When licensing stopped Clinch River in 1977, the design contemplated a homogeneous core.
Now there is a pos-sibility that a het cogeneous core will be used.
- 2) Stability of the plant under load following, load dropping or shedding, or rapid ascent.
- 3) Equipment survival.
- 4) Plant instruments and emergency procedures.
Mr. Kelber postulated a new class of accident scenarios that he suggested should be considered. The basis would be an interuption of cooling or some I
reactivity insertion. As an example, he cited a Westinghouse pump on one of the PWRs that leaked a considerable amount of oil.
If a large oil leak found its way into the Clinch River core it would result in a positive reactivity
' input.
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ADVANCED REACTORS 7/7/81 Review of Safety and Licensing Issues, R. T. Curtis, NRC Mr. Curtis described some areas where NRC research will provide support. He listed the need for development of design and licensing criteria, a capability to review and evaluate computer codes, accident analysis, and experimental data which will be proposed by DOE and other applicants, and to assess the acceptability of these codes, analysis, and data for licensing review, and finally, some help in developing and structuring a data t,ase of physical data, engineering information, computer codes, and others to substantiate the technical basis for licensing.
Mr. Curtis identified some general issues.
Design criteria, guides, and standards have not been published, nor has there been a concentrated effort t) review existing rules for applicability in light of the lessons learned on TMI.
How to Demonstrate Reliability. The reliability of safety functions intended to prevent and mitigate accidents is a sensitive function of the system reliability, systems integration, systems interaction, and the scope covered in qualification tests, start-up tests, sur-vi:111ance tests, and other procedures.
Ho.* to ensure reliability.
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t; ADVA'NCED. REACTORS-7/7/81 Mr. Siess questioned the separation of demonstrating and ensuring reliability as two separate issues. Mr. Curtis stated that demon-strating reliability is the applicant's responsibility, ensuring reliat 's ity is -the responsibility of NkC.
Safeguards 'and the integration of the fuel cycle with the reactor.
Fuel. Design. The specific criteria to ensure that the core'has mechanical and neutronic stability.
Inservice Inspection and Testing. The LFMBR system operates in a regime of temperature which is beyond that of ASME Section 3 and is a special code case in that the temperature range is in the creep range.
It is basically a low pressure system.
It has radioactive coolant in the primary system which discourages in-service inspection.
Control Room Design and Human Factors.
Decay Heat Removal. One of the more serious problems is long-term decay heat removal.
Protect,f on, control, and electrical systems must be reviewed in light of new standards in this area.
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ADVANCED REACTORS 7/7/81 Pr. Siess questioned whether the protection and control systems included by inference so-called non-safety related systems. Mr. Curtis replied yes.
Some of the problems are: assessment of the need for a re:airdaent to de-tect reduced core coolant flow by direct measurement; the adequacy of sodium
~ level, pressure, temperature, and flow measurements systems for use in plant shut-down systems; assessment of capability of instrumentation to detect im-minent core coolant flow blockages; sodium boiling or void detection; the controlability of the system men challer.ged by control systes failures; and the adequacy of the shutdown system.
Mr. Curtis concluded with a list of seven questions which he said the re-search program has to answer.
1.
Can the decay heat be removed by natural conve Mn?
2.
What are the consequences of broken components *r 3.
What are the consequences of failure in the protection of control systems?
4.
How erergetic is the core-melt accident.
5.
Can containment integrity be maintained with sodium fires and core-melt accidents?
6.
How does the high temperature-low pressure sodium envirorment change the material requirements, designing, inspecting, and i
guaranteeing the liability of the system.
7.
To what extent doe-the large amount of plutonium in the fuel change the radiolo tcal source?
Accident Delq$ tion Study, M. C1auser, Sandia Mr. Clauser. defined the delineation study as a comprehensive and systematic delineation of various accident sequences that might be expected in a breeder
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o ADVANCED REACTORS 7/7/81 reactor'.
It deals with, or tries to deal with more or less t entire sequence of accidents.
It starts with initiation and progresses to many of the phenomena that' occur within the core region of the vessel, and continues to deal with the post-accide.* nhenomena.
Mr. Clauser mentioned that protected accidents, those in which scrams succeed, may well dominate the risk due to an LMFBR, yet these have received rather little study, certainly by comparison to the unprotected accidents.
Mr. Carbon inquired regarding the use of the word "may".
He asked if it means that we really have not looked, and do not know if it is "one chance in a hundred or 50 chances in a hundred?" Mr. Clauser replied that the probability of protected accidents are certainly higher than anything else.
Since relatively little study has been devoted to it, we cannot say for sure what the consequences of a protected accident are, and they may equal the consequences or the comparable consequer.ces of an unprotected accident.
Mr. Siess asked for an example of a protected accident.
lne reply was
'being unable to remove decay heat sufficiently and the possibility of a core mel t-down.
LOSS OF HEAT SINK - TRANSITION TO NATURAL CONVECTION Thermal-Hy*draulic Analysis, P. M. Wood, NRC Mr. Wood discussed the objectives of th SSC and C04 MIX Codes. He stated that the thermal-hydraulic analysis is aimed primarily at accidents at the point where the core is damaged.
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~ ADVANCE: 1 REACTORS 7/7/81 The. codes can handle pipe break accidents, loss of flow accidents, station blackout, locked pump rotors, and can also handle natural circulation.
Mr. Wood said they have 6 plant control model to investigate the effects of these failures on the plant. Another aspect of these programs is the history of temperature and pressure transients on components and ptets of the system, which will 2nable the people who 6 risk analysis to evaluate system integrity.
Status of the SSC Code Development and Verification, J. Guppy, Brookhaven In this study, pipe. protection systems predominate. To study operational transients, control system models are needed. The second program is code validation, and a new topic which has been added during this fiscal year is tt.e balance-of-plant modeling effort. Under the code validation program, the bottom line is to make comparisons against experiments so as to validate the computations or models you used.
A questiori was addressed to Mr. Cuppy:
Dces validation mean to you comparing a mathematical solution with the physical phenomenon?
Mr. Guppy replied, yes, but in lieu of that, there can be what is called FORTRAN checking, checking of the actual coding by running your code, making your code run comparable to another code.
Mr. Siess remarked that was just checking the algorithm of the code. Mr.
Guppy gave a brief history of the SSC Code. He mentioned that inere were several versions. Each version was desigrated by tacking a letter on the end, L for Loop, P for Pool, etc. Mr. Guppy discussed the status of the code development.
AD'VANCED' REACTORS T
_ Status of the Commix and Bodyfit Codes, W. Sha, Argonne The Commix Code, as described by Mr. Sha, is a three-dimmensional and time dependent, component and multi-component computer code for thermo-hydraulic analysis. Two examples were cited. The first was for pool analysis and the second was for the CDS decay heat removal system. Mr. Sha displayed slides showing how Commix can be used to plot velocity distribution and isothem plots for various reactor plans.
Licensing issues, T. Murley, NRC Mr. Murley said that little effort has been expended in NRR since President Carter put the Clinch River Project on hold in 1977.
NRR has. stat 2d that it could issue a safety evaluation report 15 to 18 months after receipt and acceptance of an updated SAR from the project. NRR has also stated that they could complete the hearing process 15 months after that. Mr. Murley said they are using June 1984 for construction pennit decision..
DOE has said they would like to have a limited work authorization by April 1983. Mr. Murley expressed doubt that an April 1983 LWA date could be met.
CDA Analysis Whole Core Accident Assessment and Containment Loads - Program Objectives,
- 5. D. Burson, NRC Mr. Burson talked about the CCA phase of things in which there now is the possibility of some sort of accidental reactivity reinsertion into the core.
He mentioned that with an LMFBR, because of the positive void coefficient of sodium, you will be involved with a positive reactivity transient, wnich can introduce large amounts of thennal energy into the system. He remarked that nearly all codes suffer from the problem of whether the m:: del really represents the physics, and then do the mathematical algorithms translate that model into usable answers. Finally, is the question of verification of the code by comparing it with experiment.
- h ADVANCED REACTORS Some Pproblems Connected with the Initiating Phase of CDA Accident in LMFBR's, Mr. Hummel, Argonne National Laboratory Mr. Hummel discussed the computer Codes SAS3D'which is the available model of the SAS family. The FCI model is used for a failure where the clad has beer previously intact and has not melted through. He described the FCI model as something that models the motion of the molten fuel out of the coolant channel and then subsequently the motion of the molten fuel and the sodium..
Mr. Hummel stated that what we are likely to do for the immediate future is to find out what fuel melt fraction criterion we could use and what would simulate this for the time being to give us something we could have some confidence in.
Mr. Bender indicated that if there ever was a place where we need to have some probabilistic understanding, it is of the way in which these fuel failures-are going to behave.
Status of the Simmer Code Development and Verification Mr. L. Smith, Los Alamos National Laboratory Mr. Smith discussed the' Simmer Development.
He said that calculations in the past year have concentrated on the conceptual oesign study. This looked primarily _at the beginning of life a-d then at the equilibrium cycle cores.
As part of these studies, separate-effects studies, were performed,'looking at the neutronics of the cores for various assorted core configurations, because they found that the heterogeneous core design behaved neutronically much different than what we found on the homogenous core.
F ADVANCED REACTORS Los Alamos has examined the capability of Simmer to do boilup and the capability of Simmer to predict blockages.
In general, very little Simmer model development has been done.
The objectives of these studies is to identify dominant transition phase phenomenology.
In doing this, some sensitivity studies have been performed, not probabilistic studies, but sensitivity studies, to identify which phenomena contributes to energetics basically by changing parameters partway through analysis.
'Mr. Smith mentioned that it was difficult to obtain consistent reactor design data when there is no strong LMFBR program and a real, other than paper design.
Mr. Bender inquired what was the ultimate output of this computation procedure?
Mr. Smith replied, distribution of materials, temperature of materials, pressures, neutronics levels of reactivity, and power level. Mr. Smith described differences between SAS and Simmer.
Simmer is a two dimensional versus the three-dimensional versus the three-dimensional geometry of SAS.
The fuci pin modc1 is much more detailed in SAS than it is in Simmer. The primary loop is modeled in SAS where it is not in Simmer.
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Status of the Contain Code Development, Mr. Clauser, Sandia Mr. Clauser mentioned the objectives of CONTAIN. He said that objectives were to model phenomena and system response within the containment building outside the reactor vessel in both existing and proposed containment systems for all types of reactors, through the original concentration and continued concentration in sodium - cooled LMFBR's.
ADVANCED REACTORS Following various types of accidents and, in particular, following a core-disruptive accident. The outcomes of the code calculations are basically the conditions that exist inside the containment that affect the containment building' itself to determine the fission product response to the outside environment, to determine the environment that may affect mitigation devices within containment, or the various equipment that is inside containment.
Mr. Bender and Mr. Clauser discussed the initiating conditions, for CONTAIN.
It basically starts with release of material from the primary containment system from the vessel into the containment building or the reactor cavity or wherever.
Mr. Clauser noted that the next stage of the development will be a fair amount of verification and testing effort and inclusion of some additional safety system model s. He stated that with an appropriate level of effort this work can be accomplished within a year. Mr. Clauser concluded with a discussion of addi-tional models that will be needed to deal with the sodium-cooled reactor and which were anticipated to be ready shortly.
Included in these were: sodium-concrete interactions are virtually complete, Debris Bed Dynamice is partially complete, and sodium sprays fire is a relatively minor task which will be completed by the end of summer.
Questions pirected to F. Gavigan, DOE, Concerning DOE, Concerning DOE Advanced Reactor Safety - Research Progbram Mr. Gavigan cited one of the ground rules that they used in meeting with this subcommittee. The rule is that they (DOE) would not talk about CRBR at the request of CRBR. CRBR feels that they are in a licensing environment and for D.O.E. to discuss what their safety problems might be would be out of place.
L ADVANCED REAC10RS',.
Mr. Gavigan noted that regarding cooperation with NRC, DOE would like to cooperate, but there are two problems. The first is that DOE wants to avoid a situation where the base money is by mistake or on purpose held hostage with respect to getting CRBR licensing. The second relates to being careful to asrure that licensing problems are realistically identified in an environment where the CRBR project office interacts with the regulators and defines what the issues are and that the project itself then goes out and funds those problems as part of the project operation, with DOE working with the project in identifying problems or helping solve the problems.
Regarding problems that had been previously mentioned during the meeting, Mr.
Gavigan noted:
Heterogeneous Core Stability in particular, is, as far as DOE is concerned, not a problem anymore, certainly for CRBR because of the core is rather small; in addition, because these problems have been looked at in great detail at Argonne National Lab-oratory, in the Applied Physics Division, for the past two years or so.
Void Coefficient A limit can be placed on the magnitude of positive void coefficient and avoid the instability problems.
Equipment Survival in Sodium 0xide and Sodium Hydroxide Environments The CRBR project is in the process of putting together an R&D program that is is going to address this issue.
Oil Leak From Pump Into Primary System This problem was looked and been analysed both fo, FFTF and CRBR, with the r
licensing people, and generally is not one of great significance.
ADVANCED REACTORS,
CDA Experiments - Primary System Integrity - Introduction, R. Wright, NRC Mr. Wright remarked on the experimental program on the CDA threat to the integrity of the primary system. He noted that the threat consists essentially of two parts and that these programs address both.
The.first is the threat of the CDA energetics to the primary system. The second is the post-accident core debris threat to the primary system.
For the energetics threat, the logic is relatively direct. For the system to be benign, there has to be either a benign termination of the initiating phase and a benign termination of the transition phase; or the work poten'tial of the prompt burst that results has to be small.
Regarding the debris threat, the need is simply that the core debris be coolable in the vessel. The primary case of interest in the U.S. is that the debris in the reactor vessel be coolable by the cooled sodium, that you have adequate heat removal from the sodium so that there is a stable system.
Mr. Bender and Mr. Wright' explored the capability to do the programs. Mr.
Wright said they had the capability to do a good job, they didn't need any major new facilities. However, they needed money to use the facilities they have.
Energetics Threat to Primary System - W. Camp, P. Pickard, Sandia W. Car.p Mr. Camp' delineated the kinds of experimental research needs that they have in the initiation and transition-phase areas, and he told the kinds of bases for prioritization that he or his group uses in deciding this.
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. Basically, in order of decreasing importance, it is a demonstration of key safety issues, code verification, phenomenology related to these key safety issues, and model development for separate effects type questions.-
He mentioned that to determine what the key safety issues are, they have used a number of sources. The primary source is the accident delination study.
Other sources have been, the safety study for the CRBR project of a number of years ago, the original licensing-type studies, various studies of heterogeneous core accident analyses, and some of-the HCDA code sensivity analyses which have been performed.
P. Pickard, Sandia Mr. Pickard discussed the major program, elements in the accident energetics program.
In the initiation phase area the questions tre basically those of fuel dynamics, coolant dynamics and clad motion. The only active experiment program in this aea is the fuel disruption program. These are visual in-pile experiments looking at the mode and timing of fuel dissruption cf both irradiating and fresh fuel under both of and prompt burst conditions.
Mr. Pickard listed some of the Experimental Work; In the disassembly phase the focus is basically to find the work potential. Given energy release what is the available work that can be done on containment.
As part of the prompt burst energetics program, they have also defined the coarse dispersed experiments, the fuel coolant interactions experiments, and the effective equation of state experiments.
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. The transition phase program currently consists of the tran i
fuel freezing amd streaming experiit. ants. This program has started in experimentation this year.
A diagnostic development effort which is the code apperature imaging system, a gamma ray imaging system to provide high resolution fuel motion information for initiating phase accident.
Core-Debris Threat to Primary Syste, R. Coast, Sandia Mr. Coats explained the core debris coolability part of the program. He mentioned that the core debris behavior program is really exam.ining a disrupted core over these various regimes, such as, formation of a bed or rubble and Sub-Dryout behavior. He discussed the term "dryout" and
- explained this meant voiding the bed of fluid.
Included in this program are dryout, post-dryout behavior, the onset of steel melt and its migration melt and migration, and the within a rubble or within a bed, the 002 ultimate vessel or containment attack.
He listed the important parameters of the Core-Debris Study:
For incipient dryout the depth of the bed is very important, as is particle site distribution and stratification. Strati-fication was defined as large particles at the bottom of a bed and progrossivcly smaller particles as you approach the top.
Post drvout - The dry zone site and its stability.
Conduc f vity (thermal) - The lack of conductivity information of a d y bed or bed of rubble.
On,ce steel. starts to migrate within one of these beds the ability to model it becomes very marginal.
Post-mel t.
The migration of melt material, the possibility of formation of voids within a bed and bridges of centered material over that void.
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W 4 Mr. Coats Continued with a Discussion of Some Future Experimental Work That Should be Done.
. Examine the uranate hypothesis.
Stratification needs further study.
Dispersal of particles in a sodium column should be studied.
Loss of a heat sink or long after the accident where sodium or bulk sodium in the reactor is at saturation.
The amount of debris, the location, the size of particles, toe stratific: tion, sodium temperatures and cooling mechanisms that are available.
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CDA Experiments - Containment Integrity Introduction, Mr. Curtis, NRC Mr. Curtis stated that the remaining items on the agenda related to the ex-perimental program in containment integrity. He mentioned the remaining speakers would be Dana Powers ano J..a Van DenAvyle.
Core Melt and Sodium - Concrete Interactions - D. Powers, Sandia Mr. Powers noted that the four programs he will present are:
1.
Fragmentation of molten core material by sodium.
2.
Sodium containment structure integrity program which deals with the sodium interactions with concrete.
3.
The ex-vessel core debris interations with concrete.
4.
The ex-vessel core debris interactions with core retention Laterials.
The first two of these pr0 grams asre LMFBR specific problems. The last two programs dealing with core debris interactions relate to both LMF8R and light water reactor safety concerns.
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- Mr. Powers related these programs to those previously discussed. He men-tioned that the four programs were all experimental in nature. The feed data to the containment program, and the fragamentation program feeds information to the debris coolable program.
He continued his presentation with a discussion of why this work is being done. One reason is to assess the threat posed to co.itainment by material interaction with core materials or the coolant that escapes the pressure vessel and begins to interact with structural materials in the reactor cavity.
Another reason is to determine when accidents have progressed.to the ex-vessel stage stop.
Elevated Temperature Design Assessment, J. Van Denavyle, Sandia Mr. ' Van Denavyle described this program as having three purposes:
1.
Quantification of elevated temperature failure modes. This involves a number of things ranging from looking at the metallurgy and the micro structure changes which occur within alloys as a res'ilt of elevated temperature deformation, It involves looking at fracture, the mechanics of fracture, the rate at which creep failures, fatique failures and combined creep fatique failures would occur.
2.
Assessment of the validity of design methods used for elevated temperature design.
3.
Explore advances in. flaw detection methods.
Mr. Van' DenAyyle noted that one of the major problems in elevated temperature design is trying to assure a design life of the order of 30 years or so.
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- One: factor, tur mentioned, which is not considered well at all in the current tests-that have been done to generate data for the design codes is to look at microstructural changes which do occur within the material as a result of time and temperatures and deformation.
Mr.' Bender and Mr. Van Denavyle explored the parameters affecting micro structural changes. The major parameters are time, temperature and cyclic stress. -Radiation, fluxes and the chemical environment have small effects.
The meeting was adjourned at 6:50 p.m.
A complete transcript of the meeting is on file at the NRC Public Document Room at 1717 H St., NW., Washington, D.C. or can be obtained from Alderson Reporters, 300 7th St., SW, Washington, D.C., (292)554-2345.
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ATTACBMENT A 32109 19, 1981 / Nottels l
Fed:ral Registir / Vcl. 46. No.118 / Friday, June am J Agenda. Open: 000 p.m. to 5. 0 p.m July 10.
these closed sessions will be held so as 0
Engineering and Applied Science. Roorn 1981--Discussion of program policy toward to minimize incons enlence to members
,f' of the publicin attendance.
53r, National Science Foundation.
Washington. D C. 20550. Teleptione: (.+02) budget cuts.
8
- hall Closed. 9:00 a.m. to 5 00 p.m july 9. and Th* """d* fUI "b Summary minutes. Comact Mrs. Mary osts 900 a.m. to LOO p.m July 10.1981-To e as follows: Resdo). July 7,1,981, &JO 35*-95r1 p
review and evaluate research proposals a.m. until the conclusion of business, g
at the above addresa as part of the selection process for During the Initial portion of the
- Y Purpose of advisory meeting To provide awards.
meeting. the Subcommittee, along with edisce. recommendations, and counsel on Reason for closing: ne proposals being
' any of its consultants who may be r
p major goals and pohcies pertaining te reviewed include information of a present, may exchange preliminary Engineenng progra.ns and actwities.
proprietary or confidential nature-y Agenda including technical information; financial views regarding matters to be data. Such as salaries; and personal considered during the balance of the Thursday./uly 8 information conceming individuals meeting.
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p00 am-Welcoming remarks. Dr. Slaughter:
associated with the proposals. These The Subcommittee will then hear Welcoming remarks and selection of matters are within exemptions (4) and (6) presentations by and hold discussions l
Chairperson. Dr. Sanderson of 5 U.S C 552b(c).Goverrunent in the with representatives the NRC Staff.
5 9 30 am-4ommittee discussion of Sunshine Act.
their consultants and otherinterested organization functions and work agenda Authority to Close Meeting: This I
1000 em-Coffee break determinations was made by the persons regarding this review.
1015 am-Presentation on national trends in Comimttee Management Officer pursuant Further information regarding topics a
engineering and studies of engineering to provisions of Section 10(d) of p.l.92-463.
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1&45 am-Overview of NSF Engineering delegated the authority to make such Chairman's ruling on requests for the programs: ECSE. Dr. Kahne; CPE. Dr. Lih; determinations by the Director. NSF. on opportunity to present oral statements MEAM. Dr. Strauss; CEE Dr. Butcher July 6. ters.
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the cognizant Designated Federal discussion of priorities 100 pm-Individual task group meetings p
susam rw s-w au mi Employee. Mr. Elpidio Igne (telephone 5.00 pm-Adjournment sowo coos rum-a 202/634-1414) between 8:15 a.m. and hiday./uly 10 5:00 p.m.. EDT.
I have determined in accordance with 900 am-Ceneral discussion NUCLEAR REGUI ATORY Subsection 10(d) of the Federal s
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""*h g,
g mpt on(4 to t e S shine Subcommittee on Advanced Reactors; 8
M.Re I r.
get,5 y,g_c,$$2ygeyg43, Metting Commi:ter Management Coordinator.
The ACRS Subcommittee on Dated. June 15.1961 e
In Duc auseMS FM saost. 646 =l Advanced Reactors will hold a meeting W C. Ho#*
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NW Washington.DC to discuss in Duc staam km 646 m]
Subco.nmittee for Economics of the Advanced Reactor Research Program for * * "8 C00E7'*
- Advisory Committee for Social and FY 1983. Notice of this meeting was Economic Mence published June 19.
In accordance with the procedures (Docket No.50-3641 "8
outlined in the Federal Register on
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In accordance with the Federal October 7,1980 (45 FR 66535)' oral or Advisory Committee, as amended the written statements may be presented by National Science Four dation announces members of the public. recordings will O e at ng nse o N o
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the following meeting:
be permitted only during those portions The U.S. Nuclear Regulatory of the meeting when a transcript is being Commission (the Commission) has Name. Subcommittee for Economics of thekept. and questions may be asked only issued Amendment No.1 to Facility Advisory Committee for Social and by members of the Subcommittee.its Operating License No. NPF-8. This Economic Science.
f Date and Time: July 0.1981: 900 a m. to 500 consultants, and Staff. Persons desiring amendment was issued to Alabama p.m.: and July 10.1981: 9.00 a.m. to 5.00 p.m.
to make oral statements should notify Power Company for the Joseph M.
dE"on v
t!it Designated Federal Employee as far Farley Nuclear Plant. Unit 2 to correct
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8 Str e la advancq as practicable to that typographical errors in the Technical r
Washington D appropriate arrangements can be made Specifications and to allow a 24 second Type of Meeting [C.Part Open-July 10.1981:
to allow the necessary time during the delay in eneryT.9 one emergency bus 100 p.m. to 500 p.m. Closed-July 9.1981:
meeting for such statements, for one diese' nanator, 990 a.m. to 500 p.m. and July 10,1981: 900 The entire meeting willbe open to The JoseMi i seley Nuclear Plant is Contact person: Dr. Daniel H. Newton-public attendance except for those located in Houston County, Alabama.
a m.1o 100 p.m.
Program Director. Economics. Room 312.
sessions during which the Subcommittee This amendmcat was effective on May i
finds it necessary to discuss proprietary h "8 13S7 6.1981.
[o.
le h n information.One or more closed The application for the amendment 7
- Purpose of subcommittee:To provide advice sessions may be necessary to discus's complies with the sandards and and recommendations concerning support suchinformation.(Sunshine Act requirements of the AtomicEnergy Act for tesearch in and research-related exemption 4).To the extent practicable, projects in Economics e
ATTACHMENT B
' EETING DATE:
JULY 7,1981 ADVANCED REACTORS SUBC0tHITTEE MEETING:
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1167 MEETING ROOM:
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i ATTACHMENT C e
MEETING AGENDA ADVISORY COMMITTEE ON REACTOR SAFEGUARDS Subcommittee on Advanced Reacters Subconmittee Members:
M. Carbon, Chairperson M. Bender W. Kerr H. Lewis C. Mark M. Plesset P. Shewmon 1717 H Street, N. W., Room id46 Washington, D. C.
PLACE:
DATE:
July 7,1981 TIME:
8:30 A.N.
SUBJECT:
_ Review of NRC Advanced-Reactor Research Program AGENDA 8:30 AM Opening:
M. Carbon, Chairperson (10 m) 8:40 AM
==
Introduction:==
C. N. Kelber, NRC (30m) 9:1 AM Review of Safety and Licensing Issues:
R. T. Curtis, NRC (45 m)
Accident Delineation 9:55 AM Comparative LMFBR Accident Risk:
M. Clauser. Sandia (25 m) 10:20 AM BREAK Loss of Heat Sink -- Transition to Natural Convection P. M. Wood, 10:30 AM Thermal-Hyd-sulic Analysis - Program Objectives:
NRC (5m) id:35AM Status of the SSC Code Development and ','erification:
J. Guppy. Brookhaven (25 m) 11:00 AM Status of the COMMIX and BODYFIT Codes:
W. Sha, Argonne (25 m)
CDA Analysis Whole-Core Accident Assessment ynd Containment Loads - Program i
11:25 AM Objectives:
S. 8. Burson, NRC ',5 m) s
1 4
11:35AM Comparative Studies with State-of-the-Art Accident Codes:
H. H. Hummel, Argonne (25 m)
Status of the SIMMER Code Development and Verification:
11:55 AM
.J.~ Scott, LASL (25 m) 12:20 PM Status of the CONTAIN Code Development:
M. Clauser, Sandia (20 m) 12:4dPM LUNCH 1:40 PM Licensiria Issues:
J. Miller, NRC, Office of NRD. (15 m)
Questions directed to F. Gavigan, DOE, concerning DOE Advanced 1:55 PM Reactor Safety-Research Program (30 m)
CDA Experiments -- Primary-System Integrity 2:25 PM
==
Introduction:==
R. Wright NRC (1dm) 2:35 PM Eaergetics Threat to Primary System:
P. Pickard/ W. Camp, Sandia (70 m) 3:45 PM BREAK 3:55 PM Core-Debris Threat to Primary System:
R. Coats, Sandia (30 m)
CDA Experin,ents -- containment Integrity 4:25 PM
==
Introduction:==
T. Walker, NRC (5 m) 4:30 PM Core-Melt Threat to Containment:
D. Powers, Sandia (20 m) 4:56PM Sodium-Concrete Threat to Containment:
D. Powers, Sandia (20 m) 5:15 PM Elevated-Temperature Materials integrity:
C. Carnes, Sandia (15 m) 5:25 PM Sam ary Remarks:
R. T. Curtis, NRC (10 m) 5:35 PM Adjournment e
e s
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ATTACHMENT D Expenditure Rate, C. Kelbor, (1 slide) 1.
2.
NRR needs for a RES Research Program, R. Curtis, 16 slides LMFBR phasing plan for FY82, C. Kelber (1 slide) 3.
Some technical pitfalls, C. Kelber (1 slide) 4.
Accident delineation study, M. Clauser (1 slide)
Thermal-Hydraulic Analysis - Program objectives, P. Wood (3 slides) 5.
6.
Super System Code (SSC) Development and Validation Programs at 7.
Brookhaven National Laboratory, J. Guppy, (15 slides)
Status of the COMMIX and BODYFIT codes, W. T. Sha, (45 slides)
Reactor-safety modeling and assessment, B. Burson (3 slides H. H. Humel, (10 slides) 8.
9.
Status of the SIMMER code development and verification, L. Smith,(28 slides) 10.
11.
CONTAIN Mr. Clauser, (9 siides) 12.
CDA Threat to primary system, R. Wright.(4 slides)
LMFBR initiation phases, experimental research needs,, W. Camp (24 slides) 13.
14.
Accident energetics program, P. Pickard (18 slides) 15.
CORE debris program, R. Coats, (25 slides) 16.
CORE - melt threat to containmert, D. Powers (38 slides)
Purpose of elevated temperature design assessment program, J. Van Donavyle 17.
18.
(10 slides) 4 k
ie
.