Heat transport and afterheat removal for gas cooled reactors under accident conditions
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This new type of power reactor has a great deal of inherent variability implicit in its design. We have emphasized a particular design variant with great inherent safety realized by automatic power control and afterheat removal. Moreover, we have focused attention on one which generates high-temperature heat, believing that a unit of K heat indeed has an economic worth twice that of a thermal unit at the K temperature typically delivered by LWRs, and that the likely-somewhat-greater costs of a high-temperature heat-supply will be significantly exceeded by the assuredly-greater economic benefits of combined-cycle generation.
It must be noted that major variations on the fundamental theme sketched above may be feasible. For instance, we have evaluated a K water-heating design of appropriately minimized intra-core coolant volume which functions in nucleonic, structural and most heat transfer senses quite similarly to the helium-heating one just discussed.
We have also examined uranium-fueled cores, and have verified that they function basically the same as thorium-fueled ones in nucleonic senses though the latter have superior high-temperature mechanical properties, avoiding Pu and its low melting-point. A major design variable is the peak specific fuel power. Doing so lowers the intra-fuel peak pressure occurring at maximum burnup, albeit at some loss in the fuel's thermal transport qualities, and lowers worst-case structural requirements.
Operating at lower peak specific power in the fuel or with significantly different initial fuel mean density offers the prospect of reactor cores - and thus reactors - of quite different configurations. From a national energy systems standpoint, reactors of this new type act to supply high-temperature heat-on-demand throughout their entire operational lives, and thus constitute a base for reliable, large-scale electrical energy supply at a predictable, capped cost over multi-decade intervals.
Since they have no day-to-day operator controls and are never re-fueled or serviced, they generate no requirements for highly-trained personnel or special materials e. Their exceptionally efficient use of unenriched fissionable material - thorium or natural or depleted uranium - obviates all long-term fuel supply issues. Since they are never re-fueled, these new types of reactors present no spent-fuel handling, transport, reprocessing, or disposal issues, and they greatly reduce military diversion concerns.
This last point may be of the greatest importance. The high-grade, high-pressure heat produced-on-demand by this new type of reactor is intended to support combined-cycle electricity generation, with its especially favorable economic and environmental impact indices, e. Because of its extraordinarily great, multi-layered set of safety features, this new type of reactor may ultimately be suitable for underground siting in urbanized areas. Such siting would also substantially reduce capital charges for transmission systems, as well as make reasonably high-grade waste heat readily available, e.
Thermal oxidation of nuclear graphite: A large scale waste treatment option
High quality manufacture of this new type of power reactor, either for indigenous use or for export, is within the capabilities of any technologically developed country. Moderately enriched uranium for use in the ignitor modules of reactor cores can be competitively procured in any needed quantity from Russian, European or American sources by any NPT-member nation. Enriched lithium isotopes for use in filling the thermostating modules are similarly available in pertinent quantities.
At the present time, the economic pace-setting technology for large-scale central station electricity production is natural gas-fired combined-cycle generation. The principal avoided-cost of the new type nuclear reactor-energized variant relative to the gas-fired variant of combined-cycle generation would be the dominating one of the natural gas fuel-cost; in contrast, initial nuclear fuel costs are capitalized and nuclear refueling costs would be zero, as refueling is never done.
The capital cost of the system to reliably dump afterheat is a small fraction of total plant capital costs. However, careful engineering and economic analysis of a complete design will be necessary to develop significant confidence in these estimates. Working with additional collaborators, we hope soon to have a reasonably detailed design available for consideration by the entire nuclear engineering community. Large-scale, central-station electricity production using nuclear heat sources has become an unattractive prospect in a number of technically advanced countries.
This situation has persisted for two decades in some cases. It is perhaps not a severe exaggeration to suggest that the crisis facing nuclear power is a fundamental, structural one, compounded of substantial technical and economic components. Even more important is the need to deal with public perception. If popular opinion is to accept - let alone invite - re-introduction of nuclear power generation in the majority of the OECD nations, the several existing and largely independent arguments against its usage must be effectively answered.
Minor variations on widely deployed nuclear power technologies may not suffice. In token of the seriousness with which we view the current situation, we are attempting to develop one such fundamentally new approach, which we have sketched above. In a number of salient respects, it substantially simplifies the obviously safe winning of high-temperature heat from nuclear power reactors for use in modern thermal engines. See Figure 7.
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This new approach offers the prospect of assured proper reactor utilization not just when the best-trained and most highly-motivated technicians are its operators, but also when the least-trained and most careless operators may be in charge. It permits nuclear power generation capabilities to be made available with high confidence regarding materials diversion to countries which may not have highly stable political arrangements.
It fully addresses nuclear fuel supply issues, even when intensive, world-wide nuclear electrification is considered. It potentially doubles the economic value of a unit of nuclear-derived heat, by delivering it at substantially higher temperature for conversion to electricity with combined-cycle technology. We expect that this set-of-features may significantly lower both the economic and the non-economic costs of long-term, large-scale nuclear electricity production, reducing such costs to highly competitive levels.
Consequently, we believe that this new approach, or one similar to it, may satisfy via nuclear power much of humanity's requirements for electricity in the 21 st century.
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Indeed, some such novel approach may be necessary , if nuclear power is to fulfill its potential. In order to investigate and quantify physical-technological feasibility, we have employed digital computer-based performance simulation of several different types of models of the new class of reactors which we discuss.
These models support detailed studies of the nucleonics , the structural and the heat-transport aspects of reactor functioning. For maximum design flexibility and fidelity-of-modeling, we have examined the neutron transport and nuclear reactions in our model reactor designs with Monte Carlo-based means. To perform this function, we have employed the general-purpose TART95 neutron and gamma-ray three-dimensional transport- and reaction-modeling code-set developed and distributed by the Lawrence Livermore National Laboratory LLNL. This software package represents a development effort whose scale is of the order of 10 2 man-years and an associated code-validation effort of the order of man-years in size.
TART95 and its ancestors have very frequently been employed for calculation of the reactivity of critical assemblies. However, they lack time-dependence, in that they generate snapshots-in-time of the transport phenomena which they model. We have used the latest-released version of the LLNL ENDL Evaluated Nuclear Data Library as the physical data source for this code, which we have employed exclusively in the neutron energy-group mode, with TART95's most recently upgraded thermal scattering and resonance cross-section multi-band-averaging features.
Model reactor designs in our studies are resolved into several hundred spatial zones, usually possessing axial symmetry, and a few hundred different materials. Sixteen isotopes are usually carried in each zone, representing both fertile and fissile isotopic components of nuclear fuel, in addition to reflector and coolant elements, structural materials, and various neutronic poisons including fission products, carried as an ENDL-standard mix. Thus we ensure proper reactivity dependence on temperature and accurate representation of the course of long-term, possibly high fuel-burnup reactor operation.
Our models often employ homogenized materials, whenever the physical scale-lengths of different materials are less than or equal to neutronic mean-free-paths for neutron energies of interest. Indeed, we constrain zone dimensions to be less than a neutronic mean-free-path for any principal reaction e. We employ different spatial zones, many carrying a unique material composition, in our models to account for substantially different material or isotopic compositions throughout the reactor. The behavior of the isotopic fractions in each zone of a model problem are integrated in time with a fourth-order Runge-Kutta integration scheme, which couples the standard fissile and fertile isotopes of the actinide elements to each other and to fission products, using the reaction rates just calculated for the conditions in each of the spatial zones of the problem by TART Neutron absorption on all non-actinide isotopes is implicitly accounted for.
The newly-updated isotopic abundances in each zone are then inputted as a newly-reformulated ''problem'' to the TART95 code for another cycle of neutron transport and reaction calculations, thereby completing the basic set of operations of a single integration time-step. As would be expected, this 'critical value' is typically the fissile isotopic or the fission product concentration in the leading edge of the nuclear deflagration-wave propagating into the unenriched fuel-charge.
Between and time-steps suffice for an integration simulating 3 decades of reactor operation, depending on choices made regarding initial conditions and time-step controls. The top level of our nucleonics modeling program-set, which we call BURNBRED for 'burn' and 'breed' , combines the TART95 neutron transport-and-reaction package with the isotope kinetics integration package, and provides input, control and editing functions. In one simulation, several trillion floating-point arithmetic operations are performed and several billion bytes of intermediate neutronic reaction data-sets are processed using the computing system's hard-disk memory.
Such calculations often include use of a heat-transport feature in the BURNBRED package which permits the modeling of thermostatic module control of the time-dependent neutronic reactivity variation of the reactor's core, relative to a priori specifications of coolant flow through the core. The human time-to-assimilate the results of such a problem-run and specify the design of the next problem to be modeled is usually not much smaller than the duration of the run itself; thus a far faster computer could not be effectively employed.
Indeed, the computing system which we currently use centered on a Pentium MHz chip interfaced to Kbytes of pipeline-burst cache memory has been benchmarked to be roughly twice as fast as a single CPU of the fastest supercomputer generally available, the CRAY-YMP, when executing the extremely memory-reference-rich and highly scalar e. The present baseline-design core configurations, discussed in Appendix C , have been largely designed and analyzed with semi-analytic methods organized into spreadsheets.
These analyses determine the coolant pumping requirements by balancing the fluid drag in the coolant-tubes threading the core's fuel-charge which tubes dominate total loop losses against the available pressure-head. The fluid drag is modeled using turbulent pipe friction formulae, while the pressure-heads are either specified for the pumped primary coolant flow or calculated for the thermosyphoned secondary coolant flow from the thermal-gradient-derived density profiles. The heat transfer from the fuel into the coolant pipes is studied with finite element codes discussed below.
The transfer of heat from a pipe wall into the coolant-stream is calculated using turbulent boundary-layer heat-transfer coefficients. The stress levels encountered in the coolant tubes are readily found, once the pressures and wall thicknesses are known. The former are determined by the pumping-power analysis just sketched, while the amount of tube-wall material is limited to levels that can be tolerated within the core's neutron economy.
These relations are then combined to develop and analyze specific reactor designs. The coolant mass-flows required are set by the inlet and outlet temperatures, while the size-scale of the coolant tubes is determined by the desired temperature drop across the fuel i. The aggregate area of coolant tubes and the fuel's specific volume are set by desired values of the overall neutronic thickness of the fuel-charge i.
These factors then determine the inlet pressures needed in both the primary and secondary loops in order to circulate the coolant. These pressures and the quantities of coolant-pipe material available dictate the stress levels in the pipe walls. The transfer of heat from the porous, burning fuel-charge into the coolant-pipes threading through it is studied with detailed finite-element thermal codes.
These analyses employ volumetric heat production in the fuel, thermally nonlinear conduction through the fuel and the walls of the coolant tubes, and boundary-layer heat-transfer coefficients to finally deliver the heat into the bulk of the coolant-gas. These detailed analyses are particularly useful when considering the performance of the 2-D lattice of coolant-tubes during situations when some of the loops comprising the lattice are inoperative.
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Other, earlier designs which we have considered involved coolant tubes and fuel configurations with much more complex cross-sections; these systems have been studied with coupled thermal and mechanical finite-element codes. These codes remain available for more detailed analyses of future reactor designs, whether these are based upon the present simple configurations or are ones utilizing more complex layouts.
The first phase of the nucleonics of the new type of reactor likely is familiar to students of fast breeder reactors.
A centrally-positioned nuclear ignitor moderately enriched in U has a neutron-absorbing material e. Local fuel temperature rises to the design set-point and is regulated thereafter by the local thermostating modules. Neutrons from the fast fission of U are mostly captured at first on local U and Th Uranium enrichment of the ignitor may be reduced to levels not much greater than that of LWR fuel by introduction into the ignitor and the fuel region immediately surrounding it of a radial density gradient of a refractory moderator such as graphite; high moderator density enables low-enrichment fuel to burn satisfactorily, while decreasing moderator density permits efficient breeding to occur.
The optimum ignitor design involves trade-offs between proliferation robustness and the minimum latency from initial criticality to the availability of full-rated-power from the fully-ignited fuel-charge of the core; lower ignitor enrichments require more breeding generations and thus impose longer latencies. Even though such material could be diverted to use as relatively high-performance feedstock for isotopic enrichment to weapon-useful material, the complexity of doing so is not qualitatively smaller than an ab initio effort using natural uranium. The core's maximum unregulated reactivity slowly decreases in the first phase of the ignition process because, although the total fissile isotope inventory is increasing monotonically, this total inventory is becoming more spatially dispersed.
By proper choice of initial fuel geometry, fuel enrichment vs.