ML20054K934
| ML20054K934 | |
| Person / Time | |
|---|---|
| Site: | Clinch River |
| Issue date: | 06/30/1982 |
| From: | Ami Agrawal BROOKHAVEN NATIONAL LABORATORY |
| To: | |
| References | |
| BNL-NUREG-31438, NUDOCS 8207060311 | |
| Download: ML20054K934 (3) | |
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. BNL-NUREG-31438 SIMULATION OF IMBR*
A. K. Agrawal Department of Nuclear Energy Brookhaven National Laboratory Upton, New York 11973, USA Phone 516-282-2618 Introduction the entire plant. In addition, there is a paper on the engineering simulator from Japan, but it is to be given The title of this session is taken to imply the in another session systsa-vide thermohydraulic simulation of liquid metal In the following, the author has attempted to give f ast breeder reactors (MTBR). One is interested in an overview of the current state-of-the-art and future predicting the temperatures, pressures, and the coolant flow rates throughout the entire plant including the directions. These primarily reflect the results ob-reactor core, the primary and seconda.y sodium heat tained within the United States.
transport circuits, the steam generating systes as well as other auxiliary circuits. Such a simulation is S ta te-o f-t he-Art nsedsd for 1) scoping studies (i.e., in the pre-design phess of a plant), 2) detailed design developsent, 3) The subject of dynamic simulation of RFBRs has the safety analysis (post-design development phase), recently been reviewed by Agrawal and Khatib-Rahbar and 4) the operator training and plant operation. The (1). They have, in this paper, discussed thermohy-dra d ic models that are being used in the fast breeder required degree of sophistication will, of course, be nactor simulation for operational transients as well different for these phases. For example, the design.of structural components in the upper plenum of the re- as for the shutdown heat removal. The latter subject actor tank may require accurate characterization of the has also be m the topic of an international special-tempstature field in the three spatial and temporal ist's meeting at the Brookhaven National I,aboratory in sptca, while, from the cperating point of view this 1980 (2).
level of sophistication is not needed. It is, there-fore, evident that a flexible simulation approach is Sone of the earlier U.S. efforts in fast reactor ecssatial.
simulation were dedicated specifically for a plant.
For example, the IANUS code (3) was developed for the Fast Flux Test Facility, DEMO (4) for the Clinch River The types of conditions / transients that need to be simulated include normal operation, operational events, Breeder Reactor (CR3R) and NAIDEMO (5) for the EER-II.
These codes provided overall characterization of their design basis events and beyond design basis events.
The normal operation of the plant deals with start-up, respective plant. Detailed hot-spot temperatures in power operation and load following. The operational core assemblies were then obtained by employing another events include transients such as reactivity insertion computer program in tandez. This two-stage technique or undercooling transients, possible sodium-wster in. has worked acceptably, but it does suffer from some caraction in the steam generating system, interruption drawbacks. More pertinent drawbacks are 1) the plant response code is not affected by the outcome of in ths off-site power supply or the feedwater supply detailed sub-component analysis, 2) need to perform system. The design basis events are essentially what the plants are designed to acco=modate and the last calculations in tandem and 3) cannot handle situations category primarily includes the core disruptive . events. where the coupling between the plant and detailed codes
' All of these transients are part of the standard safety is not weak (such a situation may arise if there is analysis.
coolant flow reversal or coolant boils in some of the
' assemblies). In addition, these codes generally do not account for flow-dependent fiction losses in the This session emphasizes the simulation of 1.MFBRs reactor core and the plant. Finally, these codes are for only two key categories of the above-sentioned hard-wired, hence cannot be used without significant transients: operational disturbances or events and the post-shutdown decay heat removal. There are two papers modifications. These difficulties have been rectified fros France on Superphenix 2, cue from West Germany on in the SSC series of codes (6). The $$C-L code is being extensively used by a number of users both at their experience with the SSC code for $NR-300, one home and overseas. It is also being used in connectica froaa Argonne on EBR-II control simulation, one from with the CR3R licensing.
3rookhaven on the thermohydraulic modeling and simula.
tion, and the last one from Argonne on their effort to Some coments on the sodium boiling are in order.
adapt the SAS (whole-core accident code) to simulate For most operational events, sodium boiling, even in
' the hottest channel, is precluded by a substantial margin. On the other hand, for some postulated events,
- Nork performed under the auspices of the Unites States even with the plant protection system operating per Nuclear Regulatory Commission.
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dssign, there exists a possibility for boiling in hot o Finally, the international cooperation in ex-fuil cr blanket channels. Typically, the reactor power changing data may get further suppqrt since ex-is et decay heat levels (< 8% of full flows) corre- periments are expensive and time consuming.
eptading to the natural circulation. This situation is chsrteterized as low flux-low flow. The mechanism of References boiling under these conditions is quite different from thosa ancountered in the whole-core disruptive acci- 1. Agrawal, A. ~~., and Khatib-Rahbar, M., " Dynamic dints (CDAs). Coolant boiling for the CDAs are char- Simulation of LMFBR Systems", Atomic Energy Review scetrized as high flux-high flow. Thus, the slug or 18, 329-552 (1980).
multiple slug model for sodium boiling used, for ex- -
ample, in the SAS code (7) would not be appropriate for 2. Decay Heat Removal and Natural Convection in Fast th3 shutdown conditions. A homogeneous boiling model, Breeder Reactors, Agrawal, A. K., and Guppy, J. G.,
thersfore, was incorporated in the SSC code (8). Eds., Hemisphere Publishing Co., New York (1981).
Unresolved Issues 3. Additon, S. L., McCall T. B., and Wolfe, C. F.,
" Simulation of the Overall FFTF Plant Performance",
There seem to be two key unresolved issues: code Hanford Engineering Development Laboratory, varification and the accident management. The ecmputer EEDL-IC-556 (1976).
codse in use today are quite sophisticated and large.
Tha simulation of physical processes is done by making 4. Alliston, W. H., et al., " Clinch River Breeder Re-a numb 2r of convenient approximations and assumptions. actor Plant: INFBR Demo Plant Simulation Model it is therefore mandatory to verify the computer codes (DEMO)*, Westinghouse Electric Corporation Report to escsetain the modeling and programming adequacies. CRBRP-ARD-0005 (1978).
This cen be accomplished by applying either the code or ths models used in the code to pre- and post-prediction 5 Mohr, D. and Feldman, E. E., "A Dynamic Simulator of ths pertinent tests (whole plant as well as separata of the EBR-II Plant During Natural Convection with effset tests). At the same time, an acceptance crite- the NATDEMO Code", in Decay Heat Removal and rion should also be defined. A close cousin of the Natural Convection in Fast Breeder Reactors coda vstification task is to identify sensitive and Agrawal, A. K. and Guppy, J. G., Eds., demisphere important parameters. Finally, the code verification Publishing Co., New York (1981).
is a continuing process - each application to a test escults in a data point and as these data become large s 6. Agrawal, A. K., et al., "An Advanced Thermohy-ans asymptotically obtains a verified code
- draulie Simulation Code for Transients in LMFBRs (SSC-L Code)", Brookhaven National Laboratory, The accident management is a relatively never ices BNL-NUREC-50773 (1978).
in ths list of simulation. So far, most of the whole-plant simulation effort is geared for containing the 7. Ferguson, D. R., et al., "The SAS4A LMFBR Accident constquences of an event. But, in view of the TMI-2 Analysis Code System: A Progress Report", in Proc.
expstisace, some simulation for events which may not b* International Meeting on Fast Reactor Safety and
" contained" within the plant boundaries must be under- Related Physics, Chicago, Illinois, October 5-8' taksn. One is interested in simulating, for example, 1976, CONF-761001, U.S. Energy Research and De-the effects of accidental release of radioactivity
- velopment Administration (1977).
Future Trend 8. Khatib-Rahbar, M. , and Cazzoli, E. C. , " Analysis of Low Heat Flux Natural Convection Sodium Boiling in Sone of the important directions that are like17 LMFBRs*, Brookhaven National Laboratory. (Report in to ha pursued in the near future are highlighted here* press).
Thssa are:
o The need for a passive mode of decay heat re-moval from 13FBRs is likely to dominate the safety issues. Therefore, the predictability of the shutdown heat removal from an intact (nominal) as well as partially damaged plant should be assured with a high level of con-fidence.
o Many laboratory tests as well as in-plant tests may be condacted to verify (and possibly modi-fy) computer codes.
o Some probabilistic risk assessment techniques may be utilized to estimate frequencies of oc-currence of normal and of f-nor al events.
Then, with the help of deterministic analysis for these events, the cumulative load on struc-ture may be quantified.
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MEM0 FOR: Jim McKnight, DMB
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FROM: P. Larkins, TIDC J. Resner, TIDC 1 I
SUBJECT:
Transmittal of Speeches !
l Attached are two copies of a speech to be i sent to the POR and TERA. We have filed i the NRC Form 426. ,'
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