ML20038C484
| ML20038C484 | |
| Person / Time | |
|---|---|
| Issue date: | 11/14/1981 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-1882, NUDOCS 8112110125 | |
| Download: ML20038C484 (40) | |
Text
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The ACRS Subcommittee on Advanced Reactors held a meeting on Ju 14g-1981 in Room 215, 5615 North Cumberland Avenue, Chicago, IL. The purpose of the meeting was to discuss matters relating to the LMFBR safety design c ri teria. The entire meeting was open to the public.
Notice of this meet-ing was published in the Federal Register on Tuesday, June 30, 1981. A copy of this notice is included as Attachment A.
A list of attendees for this meeting is included as Attachment B, the schedule for the meeting is included as Attachment C, and a list of all reference material for this meeting is included as Attachment D.
Attached E is a schematic diagram of DOE's Fast Reactor Safety Program. A complete set of handouts has been included in the ACRS Files.
Tnre were no written or oral statements from the public. The Designated Federal Employee for this meeting was Mr. Elpidio Igne.
INTRODUCTORY PRESENTATION ON DOE'S LMFBR SAFETY PROGRAM Mr. Gavigan (DOE) identified commercialization of LMFBRs as the goal of DOE's LMFBR Safety Program. The~ purpose of the Program is to proide a data base to assess the risk to the public of LMFBRs. The Program is not specif-ically about designing plants or designing saf1ty systems for plants. The Program will provide a technology base and supportive safety considerations in the areas of design, licensing, and economic optimization.
Mr. Gaviga6 stated that LMFBRs should be at least as " safe" (on a person rem per year basis) as light water reactors.
The risk values of WASH-1400 are being used for this comparison. Mr. Gav,igan pointed out the need for a quantitative acceptable risk criterion.
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ADVANCED REACTORS 7/14&l5/81 INTRODUCTORY PRESENTATION ON DOE'S LMFBR SAFETY PROGRAM DOE's LMFBR Safety Research and Development Program addresses both accident prevention and consequence mitigation. One of its controlling philosophies is to exploit inherent plant characteristics to resolve safety issues. The program is arranged so that information is supplied for the LMFBR design and evaluation effort.
Safety program elements will be integrated into an overall risk assessment structure. Cost benefit analyses will be performed to evaluate design options and new safety features. The DOE R&D Program is also forming a reliability data base based on U.S. and U.K. breeder reactors.
Mr. Gavigan provided a brief overview of the DOE LMFBR Safety Program.
He identifled the four lines of assurance (LOAs) or first-level products used by DOE in its Fast Reactor Safety Program.
Prevent Accidents (LOA-1)
- Limit Core Damage (LOA-2)
Maintain Containment Integrity (LOA-3)
Attenuate Radiological Consequences (LOA-4) 1.
PRESENTATION ON ACCIDENT PREVENTION (LOA-1)
Mr. Vaurio, DOE, stated that DOE's R&D objective relating to accident prevention is to demonstrate that LMFBRs can be designed, constructed, and operated so that they have extremely low probability of accidents (i.e., multiple fuel pin failures). He identified two general means to accomplish accident prevention:
The use of sound, conservative, intrinsically safe design features for normal operations and adequate margins for operational transients.
ADVANCED REACTORS 7/14815/81 The use of raliable, dedicated safety systems to assure that off-normal events can he prevented or accommodated safely.
Mr. Vaurio then presented the thrae parts (second-level products) of LOA-1.
The first oart relates to the first ceneral means to accomplish accident 9revention and is entitled " Reactor System Reliability." The second and third parts of LOA-1 relate to the second oeneral means to accomplish accident prevention and are entitled " Reactor Shutdown System Reliability" and " Shutdown Heat Removal System Reliability."
The Reactor' System Reliability effort is aimed at establishino cooperation between the designer and safety analyst.
This interface shouli assure that appropriate safety desian criteria and reliability requirements are reflected in the plant.
This should reduce the probability of occurrence of off-normal events that would necessitate the use of shutdown systems or shutdown heat removal systems.
The strategy for enhancing Reactor System Reliability includes demonstrating that plant and reactor systems are reliable, and demonstratino the effectiveness of the man-machine interface in enhancing the reliability of plant operations. The Reactor System Reliability effort is divided into four tasks (third-level products). The third-level products for LOA-1 are reactor core system, heat transport systems, auxiliary systems, and monitoring and control systems. Major program tasks which are
ADVANCED REACTORS 7/14415/81 ongoing in the Reactor System Reliability area include the Fast Flux Test Facility (FFTF) operational safety orogram at Hanford and the Experimental Breeder Reactor (EBR-II) operational safety program at Argonne National Laboratories ( ANL). There are several reliability analysis methodolooy tasks ongoing and several which are near comple-tion.
These reliability analysis tasks address plant status control system develonnent and some diagnostic /proqnostic techniques develop-ment.
In addition, there are seismic and material defect analysis work ongoinn at ANL and Westinqhouse.
The two areas highlinhted by Mr. Vaurio in his nresentation on reactor systan reliability were common-causa failure (CCF) and tha nan-machine interface (ti!11).
The objectives of the CCF progran are to develop an understanding of CCF mechanisms and to-apply the knowledge obtained to improve LMFBR safety. He stated that an explicit model as opposed to a " summary" model (one which tries to draw all the connon-cause failure into one model) would be used to evaluate common-cause failures.
The objective of the Mf11 orogram is to enhance operational safety through the proper application of human factors engineering to the design of systems, facilities, operational aids, orocedures, and environments.
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ADVANCED REACTORS 7/14815/81 The Reactor Shutdown System Reliability effort is aimed at demonstrating the ability to reliably shut the reactor down for all events within the design basis.
This will be done by conducting extensive out-of-reactor tests of primary and secondary control systems and by evaluating the shutdown system operations and failure experiences (by conducting failure modes and effect analysis (FMEA), common-cause failure analysis (CCFA),
and reliability studies) and then providing the analysis results to the designers. The Reactor Shutdown System Reliability effort is divided into three third-level products:
primary shutdown system; the secondary shut-down system; and shutdown system instrumentation, monitoring, and control systems. The major program tasks which are ongoing in the Reactor Shutdown System Relia,ility effort include primary control rod testing (W), secondary control rod testing (GE), plant production system testing (W), digital plant production system design and reliability analysis (W), and operating experience evaluation and digital flow analysis associated with control systems (W).
The Reactor Shutdown Heat Removal System Reliability effort consists of conducting reliability studies of SHR systems and components, of conducting tests of SHRS hardware at the system and component levels, and of evaluating operating experience.
Information acquired in this effort will be used l
to enhance the SHRS design. The Shutdown Heat Removal System Reliability l
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ADVANCED REACTORS 7/14 & 15/81 effort is divided into three third level products: main heat transport system, auxiliary heat transport systems, and monitoring and control systems. The major program tasks in this effort include reliability analyses of main heat transport systems (GE) and reliability analyses of auxiliary heat transport systems (W, GE, Atomics International).
1.1 Mr. S. Seeman of Hanford Engineering Development Laboratory (HEDL) made a presentation on Reactor System Reliability. He outlined a program developed by HEDL to enhance operational safety. The program is designed to prevent and accommodate plant accidents through optimization of the man-machine interface on the design of equipment, systems, facilities, operational aids, procedures, and environments. The program will then assess the relative benefit derived from these new designs.
This will be done by looking at cost benefit analyses where the benefit is actually a reduction in consequences. The probability aspect of the risk will be determined by testing with humans, LWR experience, human factors texts, etc.
An attempt will then be made to extrapolate the cost benefit to future reactors.
Mr. Seeman identified ongoing work in the following areas:
plant status control, testing method-ologies, and transition control. The master information and data acquisition system (MIDAS) will be used for plant status control.
MIDAS is an information management sistem being developed for testing and use on FFTF. This administrative system basically tracks work released to the plant and allows retrieval of detailed information
ADVANCED REACTORS 7/14 & 15/81 on equipment through linked file ctructure. MIDAS will maintain a work document control log, maintain that status log, provide all sorts of query and sort capabilities, and will integrate the plant components functionally.
Planned and potential MIDAS developments were discussed. A data display system is being set up to allow development and testing of the various operational aid concepts. This system will be used to evaluate man-machine interface methodologies through operator testing under simula-tions. The system can be driven by tapes from the FFTF simulator. A diagnostic / operational system is being developed to define a system for normal and off-normal transition reactor control. This system will provide a skeleton upon which advanced diagnostic and prognostic packages can be added and tested. The hierarchy of acceptable system states will first be formalized. Then, the state of the plant will be directly sensed / measured.
Consequently, when an off-normal state is realized a transition matrix can be found which describes the changes required to return to an acceptable state.
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1.2 Mr. R. Simonelli from the Advanced Reactors Division of Westinghouse Electric Corporation made presentations on the National Digital l
Reactor Shutdown System and on a reliability analysis that was done i
j for the SDHRS.
t In his presentation on the Digital Reactor Shutdown System, Mr. Simoneli stated that moving to digital systems would inherently improve reliability, availability, accuracy, power consumption, and self diagnosis. He stated that if there are to be two diverse shutdown systems, it would l
ADVANCE 0 REACTORS 7/14815/81 he nice if one was an analogue and the Other was a diQital system.
He related Westinahouse's intention to use the experiences with diaital systems acquired in the Liaht Water Reactor program and apoly this to the
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large breeder reactors as appropriate. Mr. Simonelli then identified the ornblem of using multiolexing and establishing a desinn basis for its use in the larae breeder. While multiplexing saves a lot of weight and power, it requires more and more electronic chips.
(M11tiplexino would require sionificant shared circuitry from a chip point of view and would provide for little separation.
Therefore, common mode failures would be more probable.)
1.3 Mr. Simonelli then addressed the Shutdown Heat Removal System Reliability assessment program.
This prooran's purpose is to develop an a priori method to determine if the LMFBR desion will meet safety coals and require-ments.
The proqram will develoo criteria for desian reliability, independ-ence, diversification, and single failure protection.
The current intention of Westinghouse is to have two independent safety grade !,hutdown heat re-moval systems, a Direct Reactor Auxiliary Coolina System (DRACS), and an Intemediate Reactor Auxiliary Coolant System (IRACS). Each of these sys-tems is to be redundant within itself and at least one of these is not to be too dependent on the primary heat transport system.
DRACS and IRACS will be sized to remove decay heat with one heat sink (loop) not available.
DRACS and IRACS designs will exceed by a considerable margin present SHRS reliability anals. The reason for this is:
the DRACS will be close to the heat source and by using natural circulation will not need to depend on a lot of active components.
DRACS will be a passive system that is not
ADVANCED REACTORS 7/14 A 15/81 affected by heat transnort system operation. The reliability nargin which has resulted is believed to be sufficient to acconnodate future desian nodifications, unforesee'n potential random or common-cause failures, and operatina restrictions.
The dominant contributor to the probability of SHRS failure is steam generator failure (sodiun/ water leak),
Loss of off site pcser is also a significant contributor to SHRS failures.
The quantitative approach to be used to determine system reliability will include: establishina study assumptions with the system engineers as to what the desinn can or cannot da, gathering of failure data from both domestic and foref on sources, modeling of the systems, and then assessing the model with a mathematical method / computer code.
2.
PRESENTATION ON ACCIDENT PREVENTION (LOA-1)
Mr. R. ft. Singer of Argonne National Laboratory stated that DOE's R&D objective is to demonstrate that the response of the reactor systems will limit core damage if initiation of an accident is not orevented.
This effort is aimed at confirming that the intrinsic capabilities of the LMFBR design (including the automatically initiated systems) will terminate events more serious than the design basis so that there is only limited core damage.
The success criteria established for limiting core damage are:
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Whole-core initiators (eg. seismic event, loss of electrical power) must not result in the clad meltina.
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Physical contact between molten fuel and coolant is prohibited.
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Local faults (eq. internal sub-assembly blockage) must not lead to gross coolant boiling in a reactor.
ADVANCED REACTORS 7/14&l5/81 These criteria were established to eliminate core coolability as an issue.
Mr. Singer pointed out that these criteria may change as a greater under-standing is acquired of the inherent behavior of the reactor systems and as the limitations of self-actuated systems are better identified. Mr. Singer then identified the three areas within LOA-2 where research is directed.
These second level products are directed towards faulted behavior in i
the shutdown, shutdown heat removal, and fuel (local fault only) systems.
The Reactor Shutdown System Fault Accommodation effort is aimed at demon-strating that faults within the normal shutdown system can be accommodated with a high probability by either a self-actuated shutdown system or by the normal system operating in a degraded mode. The reactor might also be shut down by inherent fuel and absorber motions.
The Reactor Shutdown System Fault Accommodation effort will ensure that basic structures have the necessary integrity to withstand all types of acci ents. The main effort in this area is the design and study of a cost-effective self activated shutdown system.
This system must be sufficiently diverse from the normal shutdown system to preclude its failure from a cause common to that which might cause the normal system to fail. The concept and testing schedule of a two latch self-actuated shutdown system was discussed. The two latch-system would be actuated by a curie-point electromagnet combined with a thermionic switch.
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- .l ADVANCED REACTORS
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The Shutdown Heat Removal. System (SHRS) Fault Accommodation effort will demonstrate that system fau3ts can be accommodated by proving shutdown heat removal through inherent and/or degraded mode system operation.1 This effort is divided into four subtasks (third-level products). DOE is looking
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at the ways of designing am' predicting the performance'of inherent SHRS, at the limits that exf st-on core coolability, at the structural integrity needed to insure' decay heat removal, a'n'd at the types of monitoring and
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controls needed to ensure reliable system operation. The overall strategy for SHRS Fault Accommodation is to develop a sufficient understanding o{
l the natural circulation phonomena in LHFBRs to support design and licensing activities.
Experimentally validated codes will be developed to provide an early input to plant designers regarding design options available, co're and vessel flow path requirements, and requirements for inherent and degraded mode SHRS operations.
SASSYS is the general system code which is being developed by Argonne National Laboratories. This code uses a modular approach and a detailed SAS4A core treatment. The testing programs in support of SASSYS and the overall SHRS Fault Accommodation effort were briefly discussed.
2.1 Mr. M. Cooper from the Advanced Reactors Division of Westinghouse Electric Corporation made a presentation on the Self-Actuated Shutdown System Development Program. The obje'ctive of this program is to develop an alternative shutdown system that is foolproof and that
I ADVANCED REACTORS 7/14 4 15/81 doesn't require an operator or plant protection system initiated
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signal. The alternative shutdown system design should eliminate most common node failure.
Mr'. Cooper provided background on the progran's development. The system under development consists of a drive line with electronagnetic coil, a nickeline (nickle-iron alloy) insert and the magnetic circuit that serves as a temperature-sensitive fuse, an articulated control assembly, and fuel pins outside of the control assembly which provide a signal to the temperature-sensitive alloy.
r The bene #its and limitations of the articulated control assembly were discussed.
The potential for an inadvertent release was addressed.
Mr. Cooper stated that initial in-sodium testing of the temperature sensitive electromagnet (TSEtt) and absorber assembly has been con-p'eted.
The system's conceptual viability has been demonstrated by
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testing.
Testing has been conducted in the areas of TSEM magnetic
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strength at various currents and temperatures, thermal and hydraulic char'acteristics of absorbs assembly and release mechanicm, effect 6f fage on the magnetic coil, and the effect of reactor environment (irradiation) on the magnetic properties of TSEM materials. Thermal transient testina, inlet orifice tests, seismic, prototype, and fast reacter tests are being considered.
2.2 Mr. E. Gluekler from General Electric Company discussed water-scale experiments that are being conducted at GE. These experiments will f
generate the data base for natural convection data in support of.
accident analysis code development and validation. A combination of
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ADVANCED REACTORS 7/14 & 15/81 sodium and water tests is required for code validation. The water tests provide detailed information on flow fields. The sodium tests provide information on heat transfer to structures.
In addition, the experiments will be used to evaluate the phenomena of flow stratifi-catio3 and flow mixing and their effect on both flow redistribution in the reactor core and on core coolability under abnormal conditions.
The effects of piping flow instabilities on core coolability will be analyzed. DRAC and check valve operation as they effect the reactor core will be investigated.
The assumptions used in the water-scaling models were discussed.
Modeling considerations used in the shutdown coolability tests include:
Richardson number similtude, friction
- coefficient similarity, prototypical flow area ratios and channel length, and simulation of decay power. Water-scaling is accomplished by synchronizing the time scales of the momentum and energy equations.
.The resulting coupling satisfied Dr. Carbon's concern that water tests might not adequately relate flow stratification and mixing in the upper
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plenum. Preliminary test results indicate that the flow distribution in the reactor core can be predicted reasonably well with a one-dimensional model. Some modificatio.is are required in the one-dimensional GE model to improve the predictions.
It was also observed that temperature peaks may occur during the transition from forced flow to natural convection flow. Continued testing will apply laser velocimetry to determine more detailed 4
ADVANCED REACTORS 7/14 & 15/81 flow fields in the upoer plenum region. Future efforts will evaluate both one-dimensional and three-dimensional nodels.
Mr. Gluekler stated that more advanced models are needed for low flow measurements.
2.2 Mr. J. Mills of Atomics Interational (AI) Made a presentation on scale-model sodium testing.
AI's program will begin with small-scale phenomenological tests in static sodium (eg. in-tank). Foll ow-on tests will be conducted with flowing sodium systems (e.g., interaction of the primary in-tank circulation with the external systems).
Eventually scale model testing of fully prototypic systems will be conducted.
The focus of AI's test program will be on DRAC type heat removal systems that involve in-vessel heat exchangers and looks at in-vessel natural circulation phenomena.
The program objectives will be the generation of 1ata for computer code validation, data which will provide a basis for comparison with similar tests in water.
A secondary objective of the sodium testing will be to provide a test base for the development of sodium flow instrumentation.
In analyzing LMFBR natural coryection scaling requirements, AI found that sodium models scale unifonily but are practical only for large sizes (for precise scaling).
AI also concluded that water models do not scale uniformly and are impractical for any size. Consequently, Al is l
examining app.o?.imate full scale models.
AI believes that a single-l l
noda-wall approximation represents a reasonable long-time solution since it is used commonly in analysis codes and because it is quite accurate for loss of flow scenarios with medium-to-slow transients.
Reasonable l
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ADVANCED REACTORS 7/14 & 15/81 scaling ratios for the DRAC loops were obtained with the nonuniform scale factors derived from the single-node-wall model (some compromises in dimensional and thermodynamic similarity were required).
2.3 Mr. Singer completed the discussion of LOA-2 with a brief discussion of Local Faults Accommodation. The basic objective of this effort is to demonstrate that local faults which may occur in the reactor core do not propagate and become a whole core event. The criteria used in Local Fault Accommodation is to prevent gross boiling in the reactor core.
The strategy here is to demonstrate that breached fuel elements can be detected and protective action can be taken prior to significant damage.
The three third-level products in Local Fault Accommodation are limitation of core damage, coolability of local faults, and local fault detection.
When a local fault is detected the operator has three options:
he can shut down to remove the damaged pin, he can replace the damaged pin at the next scheduled refueling, or he can continue operation. The R&D effort asst.tated with Local Fault Accommodation will deal with establishing the feasibility of and/or the safe operational limits associated with the last two options. This R&D effort will rely on a large body of domestic and foreign data to eliminate the concern of rapid fault propagation. The test program is planned to determine the extent, if any, of fuel failure propaga-tion.
It will characterize signals resulting from release of molten fuel and any subsequent propagation. The test program will characterize signals from fuels subjected to various faults at various stages of deterioration.
ADVANCED REACTORS 7/14 8 15/81 (There is a chemical reaction between the sodium and the oxygen-bearing material in the fuel, which tends to expand and perhaps enlarge a breach).
It will also examine the kinetics of sodium-fuel reaction. Finally, the test program will study the potential fuel loss and contamination spread from breached fuel pins.
3.
PRESENTATION ON MAINTENANCE OF CONTAINMENT INTEGRITY (LOA-3)
Mr. D. Ferguson introduced DOE's program to ensure that containment integrity is maintained.
This effort deals with core disruptive events.
This effort is separated into two second-level products.
Energetics Accommodation and Debris Accommodation.
Mr. Ferguson stated that the purpose of the Energetic Accommodation effort was to demonstrate a low probability of containment failure from core disruptive accident energetics-related threats.
Emphasis will be placed on demonstrating that energetic related threats can be adequately accommodated within the primary vessel without signifi-cantly modifying the reactor vessel. A relatively small amount of attention will be devoted to showing that sodium released to the containment will not result in a short-term containment building failure. Mr. Ferguson stated that the strategy in the Energetic Accommodation area is to demonstrate the inherent unlikelihood of a core disruptive accident resulting in significant loads on the reactor vessel and head. By implimenting cost effective design options
ADVANCED REACTORS 7/14 4 15/91 (primarily to the containment desian) this effort should enhance an
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LMFBR's ability to acconnadate energetics.
Design options that would allow a containment to accommodate the high pressures resulting from sodium fires in the head area will be considered. The Energetics Accom-modation task is divided into two third-level oroducts:
accomodation within the primary system boundry and accomodation within the contain-ment. Mr. Ferauson stated that as part of the Energetic Accomodation effort, DOE will attempt to show that there are inherent processes that make it very difficult for LMFBR fuel to be compacted at a signi-ficant rate.
Furthermore, DOE hopes to demonstrate that fuel melts resulting from an elevated power condition subsequent to rapid sodium voiding will produce motion that is sufficiently disbursive to pre-clude a sustained prompt critical burst of the type required to gene-rate large amounts of fuel vapor. With respect to the loss of flow accident, one of the two generic core disruptive accident types, DOE has focused on the heterogeneous LMFBR designs as a way of limiting the amount of total sodium void worth that the system could have.
This effectively reduces the amount and rate of fuel disbursiveness that would be required to avoid the sustained prompt super-critical condition.
(In order to allow for predictable consequences of an accident with variations in the model and variations in input parameters and in order to allow for a controlable LMFBR, a positive $2.5-$3.0 sodium void coefficient is desired.) Mr. Ferguson stated that foreign LMFBR programs largely ignore the sodium void coefficient value and core disruptive accidents.
Because they cannot absolutely argue that a failed
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7/14 g 18;/81 fuel event couldn't happen. they have placed instrumentation en every fuel subassembly and have hooked this instrumentation into their Dlant protective system.
DOE believes this to he excessive and believes that failed fuel can be detected in sufficient time using whole core detection techniques.
DOE will try to demonstrate that should an accident progress to core disruption, the accident will progress in a very incoherent or disbursive manner (particularly in large heterogeneous cores).
DOE will attenot to i
demonstrate that should an energetic burst develop which would lead to a sianificant amount of fuel vapor beino generated that the subsequent expulsion of fuel vapor and liquid through the upper internal structure would result in low head loadinn (SIMMER prediction).
DOE will also try to denons+ rate that if there were a large sodium spray fire into the con-tainment building that there would be a limit on the pressure developed as a result of complete sodiun burning. The extensive analysis and experi-mentation planned by DOE to demonstrate Energetic Accomodation was dis-cussed in detail.
Dr. Carbon and Dr. Mark questioned Mr. Ferguson on the efforts to study and the possibility of having a fuel-coolant interaction (FCI). Mr. Ferguson stated that the oxide-fuel sodiu9 system explosive interaction was precluded (in the range of temperatures that DOE calculates might exist through the full range of accident transients) by fundamental physical principles.
He explained theories which hypothesize that an FCI might occur. He explained that there is work underway to analyze the hydrodynamic and thermodynamic aspects of FCIs.
ADVANCED REACTORS 7/14 & 15/81 3.1 Mr. Ferguson was followed by Mr. D. Weber from ANL who elaborated on Energetic Accommodation by discussing accident analysis code development.
Mr. Weber stated that ANL is developing mathematical models to simulate both phenomenological and integrated responses of physical systems.
The methods and models at ANL are verified with analytical methods, inter-code comparison, and experimental (both in-pile and out-of-pile) informa-tion. The models developed by ANL are used to perform whole core analyses to assess energetics potential, identify critical phenomenology, and recommend future experiemental and analytical investigations. The results will be used to support design and licensing efforts in detemination of energetic potential and risk assessment.
Mr. Weber's presentation concentrated on the ANL integrated response models. He discussed hypo-thetical core dissruptive accident (HCDA) analysis codes for both the accident initiation and transition phases.
In the accident initiating phase, certain HCDA phenomena (eg. sodium voiding, cladding relocation, fuel relocation) could result in positive reactivity feedback. Conse-quently, there is a need to assess these phenomena to determine if a
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sustained prompt critical burst could occur, leading to large-scale fuel vaporization.
A large scale code such as SAS4A is being used to assess the themohydraulic and neutronic phenomena associated with the energetics potential. The various models comprising SAS4A were dis-I cussed in detail [e.g., steady-state and transient fuel pin behavior, 1
SSFUEL/ DEFORM-III; fuel motion and sodium voiding during transient over 1
l power (TOP) and transient undercooling over power (TUCOP) events, PLUTO 2; fuel motion during LOF and TUCOP, LEVITATE]. Detailed discussions identi-fied the objectives of each code, its principal characteristics, recent
s ADVANCED REACTORS 7/14 & 15/81 accomplishments, and near tern tasks.
TRANSIT, a whole core code for analyzing heterogeneous core phenomena in the HCDA transition phase was discussed (intact geometry and disrupted region analyses).
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3.1 Mr. J. Kramer from Argonne National Laboratories made a presentation on fuel behavior.
Specifically, he discussed fuel motion (in the initi-ating phase of an accident) and how that fuel motion may affect the subsequent course of an accident by way of reactivity effects. Fuel j
dispersal in a voided subassembly and fuel failure into unvoided sub-assemblies were addressed. The fuel motion analysis will investigate the response of clad pins to both TOP and LOF accident for cores with large sodium void worths. Fuel behavior program acitivities include modeling and model development (to look at physical phenomena) and code development.
Modeling efforts are addressing fuel disruption and cladding deformation and failure. Code development efforts will aid in fuel pin thermal-mechanical analysis and fission gas behavior analysis. Mr. Kramer pointed out that recent tests indicate that for very high-burnup fuels, fuel dispersal occurs before fuel melt. He also stated that cladding subjected to corrosive fission products is more susceptible to corrosion cracking than cladding that is not adjacent to fuel. Mr. Kramer identified the need for more experimentation in the area of fuel and clad behavior.
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r ADVANCED REACTORS 7/14 & 15/81 3.1 Mr. Klickman from Argonne National Laboratory made a brief presentation on the TREAT experiment progran.
He stated that TREAT was closely coordi-nated with other activities to perform integral tests and phenomenological experiments that satisfy the needs of both modelers and accident analysis peopl e.
(Integral tests are tests that lump together many phenomena.)
TREAT experiments look at the initiating phase of both TOP and LOF accidents.
TREAT tests include single fuel pin tests, seven pin tests, and 37 pin tests. The objectives of TREAT exoeriments are to determine the time and location of failures, to analyse fuel relocation, and to analyze bundle size effects.
Mr. Klickman described several specific tests conducted at TREAT and the test apparatus used.
Future test programs in TREAT will include a look at the accident transition phase.
3.2 fir. L. Baker from Argonne National Laboratories made a presentation on Dehris Accommodation.
He stated that the objective of this effort is to demonstrate a low probability of containment failure from core debris-related threats.
This will he accomplished by demonstrating an inherent Debris Acconmodation capability (e.g., inherent in-vessel retention.) An effort will be made to provide a wide range of design options to enhance Debris Accommodation. The Debris Accommodation effort is divided to evaluate accommodation of debris within the primary system boundry (reactor vessel),
within the inner cell (reactor cavity), and within the containmen't as a whol e.
Mr. Baker described eight of the major Dehris Accommodation pro-gram tasks which include: particulate debris bed coolability behavior, I
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ADVANCED REACTORS 7/14 8 15/81 molten debris interactions, sodium-concrete interaction, debris retention design options, hydrogen detection and control, gas release from concrete, and code development (e.g., containment analysis code system, CONACS; concrete structural code; spray fire code, SOMIX-2).
Dr. Mark pointed out that the concrete structural code, that addresses for the first time when and how concrete structures will respond to gradual heating in a post-accident situation, will be applicable to LWRs.
Mr. Baker then made a brief presentation on the core Debris Accommodation task force.
He stated that the task force had examined the inherent retention capabilities of LMFBR design and the potential for design improvements. He said that the task force identified potential safety issues and then determine what parts of each issue might become a matter of discussion at regulatory hearings. The task force was able to identify what R&D needed to be done.
Mr. Baker discussed several Debris Accommoda=
tion design options and the R&D work in progress and planned.
Mr. Bender suggested that Debris Accommodation tests should yield results which the LMFBR designers could use. Mr. Ferguson assured Mr. Bender that they did.
Mr. Lipinski noted that the LMFBR Safety program did not address accident recovery or containment of accident debris for an extended period of time.
His concern for this shortcoming in the LMFBR Safety Program was 1
shared by Mr. Bender.
Mr. D. Pederson from ANL made presentations on two tasks that are part of the Debris Accommodation effort:
particulate debris bed coolability and molten debris interactions.
ADVANCED REACTORS 7/14 8 15/81 In his presentation on particulate debris bed coolability Mr. Pederson provided a historical perspective on debris bed coolability experi-ments and the related predictive models.
He stated that there was a good data base already available in this area but added that there was little agreement among models and little variation in debris particle size.
Mr. Pederson pointed out several areas which need further study and testing. These areas, which were discussed in detail, include debris bed leveling, the mechanisms of debris bed fixation, and debris bed post dryout behavior (which includes coolability limits beyond fuel dryout, steel migration upon melting in the dried fuel zones, and the possibility of an FCI if fuel melting occurs).
Mr. Pederson made a presentation on molten debris penetration into sub-strates [e.g., miscible, non-gas releasing (Mg0 or AL 0 ); miscible, 23 gas-releasing (concrete); and immiscible, non gas-releasing (steel)]. He identified the various experiments that are going on in the area of simulate fluids which are being used to help develop models for molten debris pene-tration into various materials. The results of and outstanding questions from experiments to gather information about experiments molten UO2 penetration inta concrete and Mg0 were discussed.
3.2 Mr. Muhlestein of Hanford Engineering Laboratory made a presentation on sodium-concrete interactions.
He stated that the sodium-cencrete reaction provides a source term for energy and hydrogen generation, both of which might threaten containment integrity. The effects of the
ADVANCED REACTORS 7/14 & 15/81 following on the sodium-concrete reaction were discussed:
sodium tempera-ture, amount of sodium, sodium contact time with concrete, concrete type and size, concrete water content, and additives (eg sodium hydroxide).
Mr. Muhlestein stated that most sodium-concrete reaction tests have shown that chemical reactions are self-limiting.
He stated that most of ti:e reactions occurred in the first couple of hours of sodium-concrete contact and that under many conditions protective reaction product layers formed which separated sodium from unreacted concrete thereby terminating the reac tion. Tests indicate that the sodium-concrete reaction is signifi-cantly effected by the water content of the concrete. Test results pre-sented also indicated that limestone concrete was more resistant to reac-tion with sodium than basalt concrete. Mr. Muhlestein summarized his current understanding of the sodium-concrete reaction by stating that the sodium-water (contained in concrete) reaction produced sodium hydroxide and free hydrogen, that a sodium water reaction temperature threshold exists, that a sodium-concrete reaction will continue until a protective reaction I
product layer forms, and finally that liquid reaction products (sodium hydroxide, sodium carbonate) are more effective in forming a reaction product layer than are solid reaction products.
ADVANCED REACTORS 4.
PRESENTATION ON ATTENUATION OF RADIOLOGICAL CONSEQUENCES (LOA-4)
Mr. L. Baker provided an overview of DOE's Fast Reactor Safety R&D Program efforts to Attenuate Radiological Consequences. The objective of this effort is to demonstrate a high probability of obtaining a large attenu-ation of consequences from a radioactive release inside containment by inherent mechanisms and by engineered systems. This objective will be accomplished by developing an efficient, reliable vent / filter system and by demonstrating dose reduction by natural processes. Mr. Baker identified the major program tasks under attenuation of radiological consequences (ie. air cleaning systems, aerosol behavior, high density aerosols, leak plugging by aerosols, establishment of a sparging source term, and environmental attenuation).
4.1 Mr. Muhlestein then made a presentation on the Attenuation of the Radio-logical Consequences by engineered systems.
He identified attenuation within internal systems and attenuation within exhaust systems as this efforts two third-level products.
Internal systems are systems within containment that act to reduce the source term.
Internal systems are categorized ts either recirculating systems which circulate radioactive aerosols through filtration systems or as systems with direct acting features which enhance the aglomeration, settling, and plating out of radioactive aerosols within containment.
Exhaust systems collect radio-active aerosols as they are exhausted in a controlled manner from the
ADVANCED REACTORS 7/14415/81 containment to the environment.
Exhaust systems include both vent systems to prevent containment over-pressurization and purge systems to control the hydrogen concentration within containment. Mr. Muhlestein first discussed the development of containment air cleaning systems.
He then discussed several exhaust systems being considered for develooment. The active systems he mentioned included prefilters and HEPA filters as well as two-and three-stage aqueous scrubbers.
Mr. Muhlestein pointed out that while pre-filters and HEPA filters have a high removal efficiency, their mass loading capacity varies widely dependina on areosol moisture content and composition. He said that the long tern reliability of prefilters and HEPA filters for removal of Na0H aerosols was uncertain and that large filter banks would be required to provide for the requisite mass loading antici-pated under accident conditions.
The passive exhaust systems discussed include sand and gravel beds and submerged gravel scrubbers.
Each of these systems could be used for a filter / vented containment and each was discussed in detail. Regarding sand and gravel bed filters, Mr. Muhlestein stated that their loading capacity per unit area is too small for the anticipated l
l melt-through accidents and that the relative size and cost of such a system 1
would be prohibitive. The submerged gravel / fibrous scrubber air cleaning concept was presented as having a high removal efficiency for a wide range of sodium aerosol types.
This system can remove the mass loadings antici-l pated in postulated conditions. fir. Muhlestein concluded that the submerced gravel scrubber with high efficiency filter demister appears to be the best option for filter / vented containments. Mr. Bender pointed out that gas l
distribution in this system might be an issue.
ADVANCED REACTORS 7/14415/81 4.2 Mr. R. Amar of ANL-TMC/AI discussed Inherent Attenuation of the radiological consequences of an LMFBR accident. The objective of the program is to demonstrate that natural' processes substantially limit an accident's radiological consequences.
The significant activities in this program include studies of the sodium sparging phenomena, of aerosol behavior, and of self pluoging in containment leakage paths, as well as an assess-ment of accident consequences.
Mr. Amar pointed out that in order to determine the radiological consequences of an accident one needs to determine the radiological source term.
Consequently, the studies and modeling of the sodium spargine phenomena will attempt to determine the radioloaical source terms from boiling (or bubbling) pools of sodium.
As part of the sparging work, an attempt will be made to obtain quantative data on decontamination factors.
Mr. Amar stated that studies relating to aerosol behavior should demonstrate that natural phenomena such as sodium agoloneration and settling significantly reduce aerosol concentration.
As part of his discussion on aerosol behavior, Mr. Amar summarized the recent improvements to aerosol behavior models and codes.
Studies are underway to demonstrate that actual aerosol leakage through containment will be significantly less than is currently estimated due to self plugging (sodiun solidification) in containnent leakage paths.
Experiments on the leakage of aerosols through capillaries have indicated that leak paths tend to plug up.
This mechanism might significantly reduce the radio-logical consequences of an accident. Mr. Amar discussed consequence
ADVANCED REACTORS 7/14 & 15/81 assessment studies. He described a code (CRACOME) with an improved evacua-tion model. CRACOME allows for post-accident evacuation in the crosswind direction, allows you to put in a delay time for evacuation, and allows you to put in a warning time for evacuation.
5.
PRESENTATION ON PLANT CONSIDERATIONS
- None 6.
PRESENTATION ON R&D INTEGRATION Mr. Ferguson discussed DOE's Fast Reactor Safety Program R&D integration effort.
Its objectives are to perform integrated analysis, provide project support, and to manage the Breeder Safety Research and Development Program to ensure safety and an expeditious licensing process.
In the Integrated Analyses area, work is going on in three general areas; methodology and data base development, accident analysis, and risk assessment.
He then elaborated briefly on the major program tasks associated with the Integrated Analysis effort (e.g., code maintenance and user assistance, risk analysis j
methology, risk allocation, risk analysis, etc.).
Mr. Ferguson mentioned that the Project Support effort focuses on design and licensing support, and dissemination of the safety technology in support of licensing.
l Mr. Gavigan stated that granting of a limited work authorization for a CRBR might occur within fourteen to twenty-four months after Congressional I
funding and formal interaction with the NRC Staff (estimated July-August 1981).
l l
l
ADVANCED REACTORS 7/14 8 15/81 Mr. M. Temme of the Advanced Reactors Systems Department of General Electric Company made a presentation on Risk Assessment / Allocation. He discussed the ongoing work at GE in the area of risk analysis methods development and some of the applications thereof.
He also discussed the risk allocation methods that have been developed at Atomics International and at GE in somewhat separate programs. The primary objective of the risk assessment methods program is to develop a technical basis for performing credible risk assessments of breeder reactors. Mr. Temme said that the program would also develop and apply methods for R&D planning in support of the LOA strategy and will provide continued risk analysis support to the Fast Reactor Projects. The risk assessment methodology described will incor-porate system interactions.
7.
DISCUSSION Mr. Siegel asked if DOE had given any consideration to non-technical studies in the area of fast reactor safety. Mr. Gavigan responded that to date such studies have not a~ffected the R&D program for LMFBR safety.
He added that the R&D program does emphasize the ideas of man-machine interface and the development of design criteria for operating systems.
Mr. Lipinski asked if the TMI Lessons Learned were being applied to the LMFBR Safety program. He was told by Mr. Gavigan that they were not being applied within the R&D program but rather were being applied from a project viewpoint. Mr. Gavigan stated "Both CBR and the large developmental plants have looked at those specifically to assure that whatever we think needs to be done for breeders is being done."
ADVANCED REACTORS 7/14815/81 Dr. Carbon raised the question of whether a lot more experimental apparatus is needed to proceed with LMFBR R&D.
He wa s tol d by Mr. Gavigan that since there is no longer an LMFBR commercialization date there is no longer a need for rapid support.
Mr. Gaviaan stated that while the Carter Administration "put the brakes on the breeder program" there was still a lot learned about LMFBRs during that administration.
OPEN EXECUTIVE SESSION An open executive session was held in the afternoon on July 15, 1981.
LMFBR safety goals and program philosophy were discussed by the Subcom-mittee nenbers and its consultants.
The following items are some of the thoughts that were brought out during the executive session.
- The Subconnittee must reexamine what the public considers to be safe enough.
We may need outside help i.e., political scientists, to resolve this issue.
- The Subcommittee should expore what represents a reasonable reguladon so that power generated from advanced reactors remains viable.
The criteria should stress prevention rather than mitigation.
The issue between LMFBR vs. LWR is not safety but of proliferation i
potential.
- Can the ACRS encourage NRC to start some kind of licensing proceedings for LMFBRs?
The ACRS should be prepared if NRC is called upon to license LMFBRs.
l Homogenous vs. heterogenous issue must be resolved. - Is a core catcher needed for LMFBRs?
l
ADVANCED REACTORS 7/14 & 15/81 The basis for probabilistic reliability analysis must be determined.
What R&D is necessary?
Given that the UiFBR.is going to be licensed, what needs to be shown?
The advantaqes and disadvantages of loop vs. pot type reactor configu-ration must he determined.
- The Subcommittee should list important technical issues that need to be addressed.
- Does the ACRS know the margin of safety it is seeking? What is a balance approach of prevention vs. mitigation.
What is the role of the ACRS and industry in the LMFBR safety evaluation program.
- It was suggested that the Subcommittee discuss with the full ACRS the charge of the Subcommittee. A dialogue with the NRC Staff should be established soon.
- Instead of a set of general design criteria, a set of questions concern-ina LMFBR issues should be addressed to NRC after being approved by the ACRS.
- A set of questions relating to licensino issues of the LMFBR should be considered in lieu of a set of design criteria.
The ACRS should present a set of questions that needs to be answered to determine a research procram.
- A list of technical issues.should be made and addressed by the Subcom-mittee at the next meeting.
August 13/14, 1981 and September 17/18, 1981 were mentioned as possible l
dates for the next Advanced Reactors Subcommittee meeting.
l The meeting was adjourned at 5:00 p.m.
l NOTE:
For additional details, a complete transcript of the meeting is available in the NRC Public Document Room,1717 H. St., NW, Washington, DC 20555 or from Alderson Reporters, 300 7th St., SW, Washington, DC, (292) 554-2345.
l l
[
I m
Federal Regist:r / Vcl. 46. No.125 / Tuesday Jun2 30.1981 / Notices 33680 Company. which revised the license and impact and that pursuant to 10 CFR the Technical ifications for operation of At ansas Nuclear One, 31.5(d)(4) an environmentalimpact Administrative Judge Peter A. Morris.
statement,or negative declaration and U.S. Nuclear Regulatory Commission.
Unit No. 2 (the facihty) at steady state enfronraental impact appraisal need Atomic Safety and Ucensing Board reactor core power levels not in excess W be prepared in connection with Phne!. Washington, D.C. 30555 of 2815 megawatto thermal. in accordance with the provisions of the laauence of;his amendment.
Administrative Judge Decer H. Paris.
bcense and the. Technical Specifications.
For further details with respect to this U.S. Nuclear Regulatory Commission.
However, the facPty is temporarily sction. see (1) the applications for Atomic Safety and Ucensing Board amendment dated February 20 and Penel. Washington. D.C. 30655 seetricted from operating at full rated March 5.1961, as a plemented by Deted at methesda Maryland, this 22nd power pending completion of the staffs detailed review of the core protection references identifie in the related g,y g j,,, 3,,3 '
ar calculator system (CPCS) changea foi Safety Evalustion. (2) Amendment No.
g,g' g*
Ief Cycle 2 operation. The facility is located se to Ucense No. NPF-4 and (3) the D*' Ch*l""##-"O'#
Commission's related Safety Evaluation.
- ####"'# ' # #*"I# ##I'#f
! cte at the hcensee's site in Pope County.
All of these items are available for
"" *"* ' * " *****I Arkansas The license amendment is i
effective ae ofits date ofissuance blic inspection at the Commission's De amendment authonzes Cycle 2 blic Document Room l1717 H Street, operation at sevent (70) percent of the NW., Washington. D C. and at the bcensed power les of 2815 MWt with:
Arkansae Tech University. Russellville-Advloory Committee on Reector
- Changes in the Core Protection Arkansas 72801. A copy ofitems (2) and Safeguards Subcommittee on Calculator System (CPCS) to reflect (3) may be obtained upon request Advanced Reactors; Meeting l
utihzation of the CE-1 critical heat flux addrused to the U.S. Nuclear The ACRS Subcommittee on correlstion and associsted thermal Regulatory Commission. Washington.
i er bydraulic methodoloSt D C. 20555. Attention: Director. Division Advanced Reactors will hold a meeting on July 14 end 15.1e81. Rodeway Inn.
- Changes in the CPCS to reflect ofLicensing Board Room-215. 8615 North stihzation of the Statistical Combination Deted at Bethesda. Maryland shis 19th day Cumberland Avenue Chicage.lL to of Uncertainties (SCU) thermal divne sets.
discuss matters relating to the hydraube methodology for the Fde hr @ tory Coru siaen.
development ofliquid metal fast breeder combination of system parameter Robert A.ClaA reactor safety design criteria. Notice of uncertainties
'l.
'?#.
this meeting was published June 17.
- Changes in the RPS and ESFAS tnp A 8'8"#fh"8"'8 In accordance with the procedures setpoints to reflect a change in signal I" D" m-*'* Pade apei ses eel outlined in the Federal Register on transmitter design and to reflect staff s u mo coce nesaim October 7,1940. (45 FR 86535), oral or approwal of the hcensee's equipment tnp written statements may be presented by selpoints members of the public. recordings wtl
- ChanFes in the minimum requtred IDocset h 80 M]
be permitted only during those portions shutdown margin to lengthen the time of the meeting when a transcript is being avail ble for operator action during a Power Authortty of the State of New kept. and questions may be asked only boron dilution esent.
York, (Indian Point Station, Unit No. 3);
by members of the Subcommittee,its
- Changes required to maintain Reconstitution of Board consultants. and Staff. Persons desiring acceptable results for the steam!me Pursuant to the authority contained in to make oral statements should notify break snel sis
- Some demonstration fuel so CFR I 2.721. the Atomic Safety and the Designated Federal Employee as far 3
assembhes to test new fuel designs-Licensing Board forhetAuthority of in advance se practicable so that
- Numerous other misce!!aneous the State ofNew York (Indian Point app opriate arrangements can be made
[
ch.nge of a clarifying. editorial and S ation. Unit No. 3). Docket No. 30-286-to a low the necessary time during the l
edministrative nature.
OLA. is bereb reconstituted by saeeting for auch statements.
- Other changes in the Technical appointing th following Administrative
%e entire meeting will be open to Specification to incorporate judges to the Board: Ehzabeth S.
public attendance except for those requirements resulting from the detailed Bowers. Peter A. Morris, and Oscar H.
sessions during which the Subcommittee physics and thermal hydraulic analysis Paris. The former Board members were Anda it necessary to discuss proprietary of the Cvele 2 reload core.
Samuel W. Jensch, Chairman. R. Beecher information. One or more closed The applicatiens for the amendment Briggs, and Dr. Frankhn C. Daiber.
sessions may be necessary to discuss such information. [ Sunshine Act As reconstituted. the Board is comprfied of Exemption 4).To the extent practic qu r m a o the Atom c ergy Act m
sn thes of1954. as amended (the Act). and the Elizabeth S. Bowers, Chairman to m? closed sessions will be held immise inconvenience to members Commission's rules and regulations. The Peter A. Morris Commission has made a propriate Oscar H. Paris of the public in attendance.
ne agenda for subject meeting shall findings as required by e Act and the All conespondence, documents and be as follows:
Commission's rules and regulations in 10 other materials shall be flied with the CFR Chapter 1. which are set forth in the Board in accorderice with so CFR 2.701 Tuesday and Wednesday-July 14 and license amendment. Prior public notice (1980.ne addresses of the new Board gg,gasi of this amendment was not required mem rs are.
since the amendment does not involve a significant hazards consideration.
Administrative Judge Elizabeth S.
M####
f The Commission has determined that Bowers.U.S NuclearReguhtory
- Safety design bases for the the issuance of this amendment willnot Commission. Atomic Safety end Department of Energy (DOE) conceptual '
Licensing Board Panel. Washington, design of an advanced liquid metal fest result in any significant environmental D.C.20555 breeder reactor.
AUAcumtnT A"
=. -....
i 33690 Feder;l Regist:r / Vol. 46. No.12S / Tnsday. Jun? 30,1981 / Notic:s Dunns the initial portion of the krvel for that category by 37.2se eeta.
aneetieg the Subcommittee, along with MflectAng ma sin >wn permemble carryover of Sects of the tranian hostage situation any ofits consultants who may be abe shortfall am the presious restramt penod and relevant legal and administratae present.ma) exchange prehmmary and carr> forward of 20.000 sets from the prudents. it will consider what views regardmg matters to be additinMten or oral prwatahons coesidered during the balance of the
[ h*7 hc*
should be requested from othat sources.
d e
sectinit current penod will be deducted imm the to:a1 ne second meeting of the The Subcommittee will then heer lor the sveteedmg penod presentations by and hold dscussions Accordms'). pursunt to operstNo Commission la scheduled for July 18.
with representatises of the NRC S:aff.
parergh (si of proc!arnataon s?69 oflune 3o.
1961. from 600 p m. to 800 p m their consultants, and other interested Na )ou em bueb) repnied to mde the continuing on July 17 from 9.00 a.m. to persons regardmg this reslew fa af
,t.
cp a s E5 item 100 p.m. or later. It will also be held in
, g Room 1107 of the Department of State.
Furtherinformation rega dmg topics De remad rut.amt lesela for the 22nd and C Streets. N.W Washmgton.
to be discussed. whe:her the meettn3 apphcable periods w.D be D.C.
bas been cancelled or rescheduled. the ne esectings will be open to pu%c Chairmatis ruhng on requests for the C*"
observation. Written comments or owe opportunity to present oral statements Z
statements may be submitted at any and the time aUotted therefor can be s
time before or after the meeting and tsas en n obtained b) a pre;. aid telephone caU to
- 558 4-8+ "
senr*
the cognizant Deugnated Federal should be hmited to the substantive
(
Empfoy ee. Mr Elpidio Igne (telephor.e matters described above.
302/634-14141 betw een 815 a m. and his lette* m t!! be published in the rederet Approximately 60 seats wi!! be 500 p m.. EDT.
bsister and the octaon wC become effectaeavailable to the pubbc on a first come.
I have determined. in accordance with en the first worklos da) after pubbcattorBrst sers ed basis.'!he Commission Very tru!) yours Procee%s wiD k recorded and subsection 10(dl of the Federal
%%m L stock.
extensive minutes prepared which may Advisory Committee Act. that it may be 1
{
secersary to close some portions of this y m es.m c c o + e ees*et be examined subsequently at the sia n.c ecos sim.a'-*
Commission's ofLee. Room 2004 metteg to protect proprietst)
Department of State.
informanon The authorit) for suth Because the Commission has only closure is Exemtten (4! to the SunshirePRESIDENT"S COVMISSION ON the req)uirement to complete ther Act. 5 U.S C 152biclm HOSTAGE COMPENSATION Dated lune 2519c1 pobo C MD !e.
(Publ4c Notace Cal-4/4 t9]
Commssion's work by August 20.1981.
less thart fAfteen d3)s Dollce of the initial 3
Ad.s409 Comi stu c Afancir-nrert O"ge_.
Open Meetings meeting is being ghen grs tu, ese. s r e s s s o m For further informa tion. conta ct John mum.c ems rm.ei ne President's Commission on
+
Hos' age Compensetion will hold its fitst R. Dasis. Jr at (202) 632-3118 who can arrange admission to the buildang 1
meet 2ng on ju!) 6 and 7.1961 from 6.00 lames S. Dwight. lt.
OFFICE OF THE UNITED $TATES to 8.00 p.m. on July 8 and from 9 00 a.m.
CAcimian TRADE REpRE>ENTATIVE to 1:00 p m. or later on jul) 7. The I
meetmg will be held in Room 1107 of the le 25. teet Letter to trie Commiss'oner of Department of State. 22nd and C Streets. pm emme rv.o.am e4s t NM.
amisso caos atw.*=
Customa Adpstirs Rer.trair.1 Levels on cow Television Receivera from The Commission will operate under Repsbhc of Korea the authonty of Executive Order n:
M j
of January 19.1981. as amended The SECURITIES AND EXCHANGE i
g 3, Chairman of the Commission la Mr.
COMMI N
{
Pursuant to the authority deleg.ted to James S. Dwight. Jr. and it is composed i
i the United Stsies Trede Representafne of nine private citizen membus, a in tss2s.st248721 I
under Presidential Proclamation 4769 of ""*N' #"h m have formerly served june 30.1980. the following letter was in the legWative or executive branches Precious Metals Holdings,Inc.;F3ing sent to the Corr.rnsWoner of Custorm of Application rn adjustmg the th rd and four;h par;od
,','; n directed to addse l m 24.1981 restremt lese!s for color telensa.,n the President whether the United States Notice is Hereby Civen that Precious I
receis ers from the Repubhc of Korem should provide financial compensation
"*'" E 3*(k to United States nationals who beve Metals Hofdmas. Inc. (the "Appbcant")
UnirerfStores 7mde Representern*
been held in captivity outside the United 80 State Street Boston. Massachusetts
} m 21 1sa:
States, either (1) by or with the approval c2109. a closed 4nd, diversified.
management investment company Honorable Wimam T. Archeb of a foreign government, or Arsarr Commissioner. US Customs Setsice.of their status as employees (2) by reason registered under the Investment ofthe Ccmpany Act of1940 "Act"). Aledan apphcation on May a (1981. persaant to Deportmente rAe Treasury.
United States Government or as r
Washington. D C Ja229 dependents of such employees.
Section S(c)of the Act. requesting an l
Dear Mr. Commissioner-ne Cosernment At its iriitial session on the evening of Order exempting the Appbcant frein the July 6. the Commission willlimit itself to provisions of Section 2(a)(19) of the A e restrain e or li r a t adjusted under the carr)os er and Personal orientation and discussion of to the extent necessary to permit carr> forward prosisior's of the Order!
eMnWaN mh* & M L b M M M dhk b 4%4M8 Marketir4 Agreement on color telension Commission will plan its work schedule the Act. Allinterte n1 persona en 3
receivers. has would increase the restraint and receive briefings from Sieta enferred to the arp. cation on file edth Department officials concerning the b Commissior We a staternent aN O
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TENT ATIVE SCHEDULE ACRS SUECOMMITTEE MEETING ON ADVA',CED RE ACTORS ROADWAY INN, PARK RIDGE, IL JULY 14 & 15,1981 JULY 14,1981 PRESENTATION APPROXIMATE TIME SPEAKER TIME 8:30 a - 8:45 a CHAIRMAN'S OPENING STATEMENT M. Carbon 15 mins.
8:45 a - 9:00 a INTRODUCTION F. X. Gavigan 15 mins.
D. R. Ferguson 9:00 a - 11:00 a LOA PREVENT ACCIDENTS J. K. Vaurio 40 mins.
1.1 Reactor System Reliability Man-Machine Interface S. K. Seeman 40 mins.
1.2 Reactor Shutdown System Reliability Digital PPS R. D. Simonelli 10 mins.
1.3 Shutdown Heat Removal Syster Reliability CDS SHRS Analyses R. D. Simonelli 25 mins.
11:00 a 11:15 a BREAK 11:15 a - 12:05 p LOA LIMIT CORE DAMAGE R. M. Singer 20 mins.
2.1 Reactor Shutdown System Fault Acconrnodation WARD SASS Development M. Cooper 30 mins.
12:05 p - 1:00 p LUNCH 1:00 p - 2:15 p 2.2 Shutdown Heat Removal System Fault Accommodation f
GE Water Tests E.L.Gigler 30 mins.
Al Sodium Test Evaluation J. C. Mills 30 mins. '
. ~2.3 Local Fault Accommodation R. M. Singer 15 mins.
l MTALAMM T C.
JUL Y 7,19!.
TENTATIVE SCHEDULE ACF.5 5UBC0"?.;TTEE MEETING ON ADVANCED REA TOP.5 ROADWAY INA, PARK RIDGE, IL JULY 14 & 15,1981 s
JULY I4,1981 CONT'D
)
PRESENTATION APPR0yIMATE TIME SPEAKER TIME 2:15 p - 3:05 e LOA MAINTAIN CONTAINMENT D. R. Ferguson 5 mins.
INTEGRITY 3.1 Energetics A:conmodation D. R. Ferguson 15 mins.
Fuel Behavior J. N. Kramer 30 mins, 3:05 p - 3:20 p BPEAM 3:20 p - 4:30 p Code Development D. P. Weber 40 mins.
TPEAT Experineits A. K. Klicknan 30 mins.
~
JUL Y 15,19El a
8:30 a - 9:35 a 3.2 Debris A:connodation L. Baker, Jr.
15 mins.
Debris Sed Behavior D'. R. Pedersen 15 mins.
Material Interactions D. R. Pedersen 15 mins.
Na-Concrete Interactions L. t!uhlestein 20 mins.
9:35 a - 10:30 a LOA ATTENUATE RADIOLOGICAL L. Baker, Jr.
10 mins.
CONSEQUENCES 4.1 Engineered Attenuation L. Muhlestein 30 mins.
4.2 Inherent Attenuation R. C. Amar 15 mir.s.
10:30 c - 10:45 a BREAK 10:45 a - 11:35 a R&D Integration D. R. Ferguson 10 mins.
Risk Assessment / Allocation M. Temme 40 mins.
?
11:35 a - 12:00 n
, DISCUSSION
. w.
12:00 n - 1:00 p LUNCH 1:00 p - 4:00 p OPEN EXECUTIVE SESSION S:M l
1
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REFERENCE MATERIAL 4
ADVANCED R ACTORS SUBCOMMITTEE MEETING
,)
CHICAGO, IL JULY 14 8 15, 1981 1.
Mr. M. Bender letter to Dr. M. Carbon dated June 16, 1981 regarding LMFBR safety issues.
2.
" Goals for the Advanced Reactors Subcommittee LMFBR Design - Criteria
/
- Study," 'by Dr. P., Carbon and Dr. D. Okrent.
3.
Viewgraphs and handouts used during meeting.
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