Blanket Thickness Research Articles

Critical heat flux in subcooled flow boiling of fluorocarbon liquid on a simulated electronic chip in a vertical rectangular channel

Experiments are performed to correlate and model CHF during subcooled flow boiling of dielectric fluorocarbon (FC-72) liquid on a smooth 12.7 × 12.7 mm 2 simulated electronic chip inside a vertical rectangular channel. Two distinct CHF regimes are established based on flow visualization and experimental data. At low flow velocities, CHF is accompanied by the formation of a continuous vapor blanket and the dryout of a liquid sublayer between the blanket and the chip. Higher velocities reduce the thickness of the vapor blanket and maintain partial contact of subcooled liquid from the core with the chip, leading to the formation of discrete vapor blankets much smaller than the boiling surface of the chip. Experimental data obtained in the lower velocity regime display excellent agreement with predictions of a new CHF model which accounts for the effects of the liquid velocity profile, liquid subcooling and flow geometry.

Minimum thickness blanket-shield for fusion reactors

A lower bound on the minimum thickness fusion reactor blankets can be designed to have, if they are to breed 1.267 tritons per fusion neutron, is identified by performing a systematic nucleonic optimization of over a dozen different blanket concepts which use either Be, Li17Pb83, W or Zr for neutron multiplication. It is found that Be offers minimum thickness blankets; that the blanket and shield (B/S) thickness of Li17Pb83 based blankets which are supplemented by Li2O and/or TiH2 are comparable to the thickness of Be based B/S; that of the Be based blankets, the aqueous self-cooled one offers one of the most compact B/S; and that a number of blanket concepts might enable the design of B/S which is approximately 12 cm and 39 cm thinner than the B/S thickness of, respectively, conventional self-cooled Li17Pb83 and Li blankets. Aqueous self-cooled tungsten blankets could be useful for experimental fusion devices provided they are designed to be heterogenous.

A new 3He gas loop blanket for fusion reactor systems

A new concept of tritium breeding blankets is introduced for fusion reactor systems using a 3He gas loop combined with neutron multipliers and moderators. To verify the tritium breeding capability of this 3He based blanket, several neutronic calculations have been made in one-dimensional cylindrical geometry and compared with those for the current Li based blankets. The results show that the tritium breeding ratio of the 3He gas loop blanket can exceed unity with a reasonable blanket thickness. In particular, the possibility of a “thin” tritium breeding blanket is suggested in a homogenized system of Be and 3He. With these neutronic survey calculations, the T-3He fuel cycle has been modelled to estimate the required amounts of tritium and 3He supplies. The feasibility of the 3He gas loop blanket system is discussed through considerations on the merits of this T-3He fuel cycle.

Design analysis and optimization of self-cooled lithium blankets and shields

Abstract A study of self-cooled lithium blankets was carried out to define the performance of these blankets and to determine the potential to operate at the maximum possible values of the performance parameters. The main design parameters considered during the course of the study were the tritium breeding ratio, the blanket energy multiplication factor, the energy fraction lost to the shield, the total blanket thickness, the reflector material selection, and the compositions of the different blanket zones. Another study was carried out to determine materials, compositions, arrangements, and thickness of the shield zone for the reference blanket. Helium and water-cooled shields were optimized for the inboard and outboard sections of the reactor. Based on the above two studies, the reference blanket and shield configurations were developed for the ANL Tokamak Power Systems Study. The helium-cooled shield was selected for use with liquid metal blankets to reduce safety concerns related to lithium—water reactivity. This helium-cooled shield provides shielding characteristics similar to a conventional water-cooled shield. The analyses and results from these studies are the subject of this paper.

Experimental Study of the Large-Scale Axially Heterogeneous Liquid-Metal Fast Breeder Reactor at the Fast Critical Assembly —Power Distribution Measurements and Their Analyses

Power distributions of the large-scale axially heterogeneous liquid-metal fast breeder reactor were studied by using the experiment results of fast critical assemblies XI, XII, and XIII and the results of their analyses. The power distributions were examined by the gamma-scanning method and fission rate measurements using /sup 239/Pu and /sup 238/U fission counters and the foil irradiation method. In addition to the measurements in the reference core, the power distributions were measured in the core with a control rod inserted and in a modified core where the shape of the internal blanket was determined by the radial boundary. The calculation was made by using JENDL-2 and the Japan Atomic Energy Research Institute's standard calculation system for fast reactor neutronics. The power flattening trend, caused by the decrease of the fast neutron flux, was observed in the axial and radial power distributions. The effect of the radial boundary shape of the internal blanket on the power distribution was determined in the core. The thickness of the internal blanket was reduced at its radial boundary. The influence of the internal blanket was observed in the power distributions in the core with a control rod inserted. The calculation predicted the neutron spectrum hardermore » in the internal blanket. In the radial distributions of /sup 239/Pu fission rates, the space dependency of the calculated-to-experiment values was found at the active core close to the internal blanket.« less

Cholera enterotoxin-induced mucus secretion and increase in the mucus blanket of the rabbit ileum in vivo

The in vivo rabbit ileum was used to study the relationship of cholera enterotoxin-induced water and electrolyte secretion and mucus secretion and to determine whether the enterotoxin influenced the intestinal mucus blanket. In experiments in which luminal fluid viscosity was used to assess mucus secretion, it was found that while cholera enterotoxin induced a sustained secretion of water and electrolytes, the toxin-induced mucus hypersecretion was short lived (3 to 5 h) and subsequent exposure of the mucosa to cholera enterotoxin or prostaglandin E1 did not stimulate mucus secretion further. Dibutyryl cyclic AMP and theophylline caused a modest mucus secretion in ileal loops which differed from that of cholera enterotoxin in both magnitude and in the fact that the mucus secretion occurred 2 to 3 h after the onset of water and electrolyte secretion. An oral replacement solution was used in the ileum to reduce the enterotoxin-induced loss of water and electrolytes into the lumen. While such a solution slowed the appearance of acidic glycoprotein in the intestinal lumen, it did not change the amount of glycoprotein secreted over a 7-h period, suggesting that toxin-induced mucus secretion was not simply due to a flushing action of the experimentally caused diarrhea. To assess mucus blanket thickness, neutral glycoprotein was recovered from the blanket of rabbit ileal loops 7 h after exposure to cholera enterotoxin and the thickness of the mucus blanket was measured directly 4 and 18 h after toxin exposure. Both methods indicated that even though cholera enterotoxin-induced mucus hypersecretion had subsided and there was histological evidence of goblet cell mucin depletion, there was a sustained increase in mucus blanket thickness that was detectable for at least 18 h after mucosal enterotoxin exposure.

Connections between physics and economics for Tokamak fusion power plants

A simplified physics, engineering, and costing model of a tokamak fusion reactor is used to examine quantitatively the connection between physics performance and power-plant economics. The material contained herein was generated as part of a broader study of the economic, safety, and environmental impact of fusion based on a range of confinement schemes, fusion fuels, blanket/shield configurations, power-conversion schemes, and commercial end products. Only a DT-fuelled tokamak reactor that produces electricity through an intermediate heat exchange and a conventional thermal-electric conversion cycle is considered; a self-cooled lithium-metal blanket with vanadium-alloy structure, steel shield, and superconducting magnets is used for all cases studied. An optimistic extension of Troyon scaling is applied to a high-elongation (κ = 2.5) and low-safety-factor (q ψ =2.3) plasma with β=0.1 and efficient (I φ P CD =0.2 A/W) current drive. This 1200-MWe (net) power plant provides an economically competitive base case with which to compare other approaches to tokamak fusion power. The base case chosen for comparisons represents an optimistic extrapolation of present tokamak physics and technology. Troyon scaling with a coefficientβ B φ a/I φ equal to 0.04 is applied; the impact of an ad hoc but pessimistic scaling that diminished the Troyon coefficient with plasma elongation was also examined. Additionally, a constant current-drive efficiency, ϒ=nI φ R T /P CD =0.2 A/W, atT=10 keV plasma temperature is assumed; although representing an aggressive R&D target relative to present experience, the realization of bootstrap currents for the basecase, and especially for the second-stability-region tokamak, can significantly reduce this problem. The impact and reoptimization for a constant normalized current-drive efficiency, ϒ=nI φ R T/P CD, was also examined. Although the focus of this study has been the optimistic basecase tokamak, comparisons are made with tokamaks based on (a) operation in the second-stability region (β=0.2, increased aspect ratio, reduced elongation), (b) super high-field but low-beta operation, (c) very low aspect ratio and highly elongated spherical torus, and (d) a direct application of the present database using a long-pulsed, low-beta tokamak. The economic impact of a range of base-case parameters and operating variables is examined, including current-drive efficiency, beta, stability limits, advanced magnets, economy of scale, blanket/shield lifetime, blanket thickness, and plant lead time. It is found that a range of tokamak options, relative to the optimistic base case selected for this study, may provide economically competitive power plants. Areas where physics and technology advances are needed to achieve this attractive end product are quantitively elucidated for all tokamak options considered.

Optimal Use of Beryllium for Fusion Reactor Blankets

The use of beryllium as a neutron multiplier for fusion reactor blankets has been analyzed. The analysis has been performed based on designs for a helium-cooled ceramic breeder and a self-cooled liquid-metal blanket, which have both been suggested for the Next European Torus reactor. It is shown that the use of beryllium in a ceramic breeder blanket is best in a “sandwich-type” arrangement, where a beryllium block is embedded between a thin ceramic layer and the thick main breeding zone, or in a homogeneous mixture of beryllium and breeding ceramics. The sandwich-type solution needs only a minimum of beryllium inventory. Monte Carlo calculations show that heterogeneity effects in such a blanket are negligible. Therefore, the “homogeneous” solution can be achieved in a more heterogeneous arrangement like slabs of beryllium with the breeding ceramics in between.The use of beryllium also provides a benefit for liquid-metal blankets, using either LiPb or lithium metal as breeding material, since neutron multiplication and the tritium breeding ratio are enhanced in such a way that it is possible to reduce the blanket thickness considerably or to replace the inboard breeding blanket by a simple neutron reflector. It turns out that in such a blanket the use of lithium metal as breeding material is superior to that of LiPb.

Dynamics of lip formation for explosion craters

A large fraction of the material ejected by a near-surface detonation impacts just beyond the crater to form the crater lip. For certain combinations of explosive energy, charge configuration and cratering medium, the accumulating ejecta blanket in the lip region becomes so thick that the crater walls become unstable. A model that includes the post-impact redistribution of impacting material to maintain slope stability is presented. This model has three components: a source and ejecta trajectory model, a true crater/upthrust scaling law and a slope failure model. Calculated apparent crater contours and ejecta blanket thickness distributions in the lip region are in good agreement with results of field tests.

A Thin LiPb/Be Helium Cooled Blanket for MINIMARS: Neutronics Analysis

This work presents a thin blanket design for the commercial tandem mirror reactor MINIMARS using the reactor parameters to date. Future studies may call for minor changes in the parameters to incorporate the idea of central cell wall stabilization and to achieve the inherent safety goal. Although the study was carried out for a 17Li-83Pb system, the trends observed here hold true for other breeding materials. Recent studies for minimizing blanket thickness used idealized blankets with semicontinuously varying composition and no allowance for cooling. Here we present a realistic thin blanket design that satisfies all requirements of neutronics, safety, thermal-hydraulics, and mechanical design.

Thin Blanket Design for MINIMARS - A Compact Tandem Mirror Fusion Reactor

Recent fusion power reactor designs have shown a trend toward lower power, lower cost, higher mass utilization compact configurations with inherent safety, in order to improve the economic aspects of fusion and make them more competitive with other energy sources. Since the blanket thickness directly impacts the size and mass of the remaining reactor components, it is prudent to minimize its thickness while ensuring adequate neutronic and thermal performance. This paper describes the blanket for the MINI-MARS compact tandem mirror fusion power reactor. The blanket which utilizes HT-9 ferritic steel structure, LiPb breeder, Be multiplier/moderator and He gas cooling is only 17 cm thick and is backed up by a steel reflector. Helium gas cools the blanket and reflector in series and the outlet temperature of 575/sup 0/C gives a gross thermal power cycle efficiency of 42.7%.

Blanket Optimization Studies for Cascade

A nonlinear, multivariable, blanket optimization technique is applied to the Cascade inertial confinement fusion reactor concept. The thickness of a two-zone blanket, which consists of a BeO multiplier region followed by a LiA10/sub 2/ breeding region, is minimized subject to constraints on the tritium breeding ratio, neutron leakage, and heat generation rate in Al/SiC tendons that support the chamber wall.

Optimal Blanket Concepts for D-T Fusion Reactors

The maximum blanket energy multiplication and the minimum thickness D-T fusion reactor blankets can be designed to have are estimated with the aid of the optimization code SWAN, using simplified blanket models. It is found that energy multiplications as high as 2 might be attainable. The minimum effective thickness of the breeding blanket identified is 7 cm or 8.5 cm, when C or PCA are used, respectively, for the reflector. All of these blankets use beryllium. Also performed is a preliminary partial feasibility study of a number of novel blanket concepts, including blankets using 6Li17Pb83, W, Zr and Cu as a primary constituent.

Advanced Commercial Tokamak Optimization Studies

Our recent studies have concentrated on developing optimal high beta (bean-shaped plasma) commercial tokamak configurations using TRW's Tokamak Reactor Systems Code (TRSC) with special emphasis on lower net electric power reactors that are more easily deployable. A wide range of issues were investigated in the search for the most economic configuration: fusion power, reactor size, wall load, magnet type, inboard blanket and shield thickness, plasma aspect ratio, and operational ..beta.. value. The costs and configurations of both steady-state and pulsed reactors were also investigated. Optimal small and large reactor concepts were developed and compared by studying the cost of electricity from single units and from multiplexed units. Multiplexed units appear to have advantages because they share some plant equipment and have lower initial capital investment as compared to larger single units.

Impact experiments: 1. Ejecta velocity distributions and related results from regolith targets

Velocity distributions are determined for ejecta from 14 experimental impacts into regolithlike powders in near-vacuum conditions at velocities from 5 to 2321 m/sec. Of the two powders, the finer produces slower ejecta. Ejecta include conical sheets with ray-producing jets and (in the fastest impacts at V imp ⪸ 700 m/sec) high-speed vertical plumes of uncertain nature. Velocities in the conical sheets and jets increase with impact velocity (Sect. 6). Ejecta velocities also increase as impact energy and crater size increase; a suggested method of estimating ejecta velocity distributions in large-scale impacts involves homologous scaling according to R/ R crater, where R is radial distances from the crater (Sect. 7). The data are consistent with Holsapple-Schmidt scaling relationships (Sect. 8). The fraction of initial total impact energy partitioned into ejecta kinetic energy increases from around 0.1% for the slow impacts to around 10% for the fast impacts, with the main increase probably at the onset of the hypervelocity impact regime (Sect. 9). Crater shapes are discussed, including an example of a possible “frozen” transient cavity (Sect. 10). Ejecta blanket thickness distributions (as a function of R) vary with target material and impact speed, but the results measured for hypervelocity impacts agree with published experimental and theoretical values (Sect. 11). The low ejecta velocities for powder targets relative to rock targets, together with the paucity of powder ejecta in low-speed impacts ( < 1 projectile mass for V imp ≈ 10 m/sec) enhance early planetary accretion effeciency beyond that in some earlier theoretical models; 100% efficient accretion is found for certain primordial conditions (Sect. 12).

Helium-Cooled Solid Breeder Blanket Design for a Tokamak Fusion Reactor

A preliminary design for a helium-cooled solid breeder blanket for a tokamak fusion reactor has been developed, and its performance looks quite good. The design is capable of bearing a 4 MW/m2 neutron wall load, and the ideal pumping power required for the whole primary helium loop including the steam generators is only 2.5% of the total thermal power. The maximum blanket thickness including the helium duct work is only 860 mm, the minimum thickness is only 730 mm. The design work was focused on the thermalhydraulic aspects, which represent the key problems associated with using helium as a coolant. The present work demonstrates that the potential disadvantages helium has, due to its limited heat transfer capabilities, can be avoided or minimized by an appropriate thermal-hydraulic design. As a result, helium with its many advantages remains a promising fusion blanket coolant.

Evaluation of experimental magnetic printing device using weak dipole development process

An experimental magnetic printing device employing a novel low background development process is described. Printing characteristics of patterns with a resolution of 10 dots/mm (254 dpi) were evaluated as a function of the system variables. Line and pattern print was used to establish the operating margins of recording parameters (such as drive current and medium thickness} and developing parameters (such as toner medium interference and toner blanket thickness). Print quality could be improved by applying a vacuum clean-up step on the developed image.

Inertial Confinement Fusion with Direct Electric Generation by Magnetic Flux Compression

A high-power-density laser fusion reactor concept is investigated in which directed kinetic energy imparted to a large mass of liquid lithium--in which the fusion target is centrally located--is maximized. In turn, this kinetic energy is converted directly to electricity with, potentially, very high efficiency by work done against a pulsed magnetic field applied exterior to the lithium. Because the concept maximizes the blanket thickness per unit volume of lithium, neutron-induced radioactivities in the reaction chamber wall can be many orders of magnitude less than is typical of D-T fusion reactor concepts.

Crater ejecta scaling laws: Fundamental forms based on dimensional analysis

A model of crater ejecta is constructed using dimensional analysis and a recently developed theory of energy and momentum coupling in cratering events. General relations are derived that provide a rationale for scaling laboratory measurements of ejecta to larger events. Specific expressions are presented for ejection velocities and ejecta blanket profiles in two limiting regimes of crater formation : the so‐called gravity and strength regimes. In the gravity regime, ejecta velocities at geometrically similar launch points within craters vary as the square root of the product of crater radius and gravity. This relation implies geometric similarity of ejecta blankets. That is, the thickness of an ejecta blanket as a function of distance from the crater center is the same for all sizes of craters if the thickness and range are expressed in terms of crater radii. In the strength regime, ejecta velocities are independent of crater size. Consequently, ejecta blankets are not geometrically similar in this regime. For points away from the crater rim the expressions for ejecta velocities and thickness take the form of power laws. The exponents in these power laws are functions of an exponent, α, that appears in crater radius scaling relations. Thus experimental studies of the dependence of crater radius on impact conditions determine scaling relations for ejecta. Predicted ejection velocities and ejecta‐blanket profiles, based on measured values of α, are compared to existing measurements of velocities and debris profiles.

Stripping of Keanakakoi tephra on Kilauea Volcano, Hawaii

Morphological characteristics, erosional processes, and effects of burial and exhumation by debris mantles on basaltic volcanic landforms have been evaluated through field study of the Keanakakoi Formation, a basaltic tephra formed in 1790 by phreato-magmatic eruptions from Kilauea caldera, Hawaii. The upper coarse lithic, intermediate fine vitric, and lower mixed members of this formation play different roles in the creation of micro-terrain elements during stripping of tephra from the underlying bedrock. Of the seven micro-terrain elements defined, bedrock, scabby upland surfaces, and lag gravels are the most distinctive and widely distributed. Different proportions and combinations of micro-terrain elements define five zones of progressive deterioration of the Keanakakoi tephra blanket southwestward from Kilauea caldera into the Kau Desert. Fluvial and eolian processes operate on different time scales and at different locations, governed by blanket thickness, debris caliber, and the formation of case-hardened crusts. Stripping of an entire mantle is probably not possible; however, materials trapped within depressions form the only clearly discernible morphological expression of previously more extensive debris blankets.