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Deltarasin is a recently identified small molecule that can inhibit KRAS–PDEδ interactions by binding to a hydrophobic pocket on PDEδ, resulting in the impairment of cell growth, KRAS activity, and RAS/RAF signaling in human pancreatic ductal adenocarcinoma cell lines. Since KRAS mutations are the most common oncogene mutations in lung adenocarcinomas, implicated in over 30% of all lung cancer cases, we examined the ability of deltarasin to inhibit KRAS-dependent lung cancer cell growth. Here, for the first time, we document that deltarasin produces both apoptosis and autophagy in KRAS-dependent lung cancer cells in vitro and inhibits lung tumor growth in vivo. Deltarasin induces apoptosis by inhibiting the interaction of with PDEδ and its downstream signaling pathways, while it induces autophagy through the AMPK-mTOR signaling pathway. Importantly, the autophagy inhibitor, 3-methyl adenine (3-MA) markedly enhances deltarasin-induced apoptosis via elevation of reactive oxygen species (ROS). In contrast, inhibition of ROS by N-acetylcysteine (NAC) significantly attenuated deltarasin-induced cell death. Collectively, these observations suggest that the anti-cancer cell activity of deltarasin can be enhanced by simultaneously blocking “tumor protective” autophagy, but inhibited if combined with an anti-oxidant.

Objective To study the clinical path system of stroke Chinese medicine syndromes, to explore the classification and identification methods of TCM syndromes of stroke, to collect high quality data and expert knowledge, and to provide some methods for the standardization study of TCM syndromes of stroke And basis, to improve the clinical efficacy of stroke Chinese medicine, support Chinese medicine scientific research information. Methods A unified four diagnostic information collection scale was used to collect the stroke-related cases from the clinical data. SPSS20.0 statistical analysis software factor analysis was used to qualitatively and quantitatively analyze the data in the clinical path of stroke. The vector method was used to model the relationship between TCM clinical information and syndromes of stroke, and the accuracy of prediction was observed. Results The support vector machine can provide some objective evidence in the judgment of TCM syndromes of stroke. The clinical path system of TCM syndrome can be used to provide reasonable and reliable information support for clinical decision-making of stroke.

One of the critical issues of Flibe/V-alloy blanket with REDOX control by Be is a large tritium inventory in V-alloy structures. Among the possible solutions to this issue would be to control REDOX not by Be but by addition of MoF 6 or WF 6 enhancing the reaction from T 2 to TF. The present study investigated feasibility of this procedure by thermodynamic and neutronics calculations. Using the blanket dimensions of Force Free Helical Reactor (FFHR), tritium inventory in V-alloy structure and Flibe were estimated based on the calculated equilibrium partial pressures of T 2 and TF in various cases of REDOX control by MoF 6 or WF 6 . Also carried out were neutronics examinations for the impact of Mo or W doping in the blanket. The results showed that the tritium inventory in the blanket area would be less than 100g at the TF level of 0.1 and 1 ppm in Flibe with addition of WF 6 and MoF 6 , respectively. WF 6 doping is far more advantageous than MoF 6 doping for low activation purposes.

Multi-walled carbon nanotubes with homogeneous diameters (40–60 ran), produced by chemical vapor deposition of hydrocarbon gas, are purified by nitric acids. Infrared and Raman studies indicate that oxygen containing surface groups, which are predominately carboxylic, phenolic and lactonic groups, are introduced into purified carbon nanotubes. Then three kinds of block-form porous tablets of carbon nanotubes are fabricated as electrodes in electrochemical double-layer capacitors. Using mounded mixture comprising carbon nanotubes and binder powders provides these tablets. Comparison of the effect of different processing on the structural performance of the capacitors is specifically investigated. Using chemically treated electrodes, electrochemical double-layer capacitors with a specific capacitance of about 33 F/g are obtained with 38 wt% H2 SO4 as the electrolyte.

An assessment of Flibe thermo-physical properties relevant to the prompt x-rays ablation of the liquid wall is presented with emphasis given to the equilibrium vapor composition and vapor pressure....

Neutronics calculations were performed for blanket designs using the molten salts Flibe and Flinabe to determine the transmutation rates of constituent elements and the rates of production of other elements. At least from mass balance considerations no free fluorine will be left provided that the recombination reactions with freed Be, Li, Na, and tritium are fast enough. However, more than 95% of the tritium bred will be in the form of TF. In addition, O and N are produced. A REDOX reaction needs to be established to control the TF activities. The Be used for neutron multiplication can be used for the REDOX control to reduce TF to T 2 . The thermodynamics for the reaction between TF and Be is an important process to be demonstrated.

Transmutation rates of Li, Be and F are calculated for a typical flibe blanket. The results concluded that the transmutation rate of F is more than double that of Be. Because of the high destruction rate of fluorine, there will be no free fluorine in the molten salt. Therefore, experimental program to address the chemistry control of flibe does not have to worry about the issues associated with free fluorine as long as kinetics are favorable (likely). Also, this calculation defines the chemical state of flibe after irradiation. This chemical state needs to be simulated closely for the flibe chemistry control experiments.

The nuclear performance of the thin Convective Liquid Flow First Wall (CLiFF) concept is investigated. Liquid walls offer the advantage of protecting solid structure behind them from excessive damage from neutrons originated in the plasma and thus have the capability for high power density applications; the central research focus of the Advanced Power Extraction (APEX) study. In the present parametric and scoping work, several combinations of liquid breeder and structure type where investigated. The aim is to maximize local tritium breeding ratio (TBR), power multiplication, and ensuring that the vacuum vessel and toroidal coils are protected from excessive radiation. The candidate liquid breeders considered are Li, Flibe, and Sn-Li. Vanadium-alloy is deployed with Li while either Ferritic steel or SiC is deployed with Flibe and Sn-Li. Deployment of other refractory alloys and their impact on TBR was also studied. The introduction of a beryllium multiplier zone in the blanket was shown to enhance tritium production capability, particularly for those liquid breeders whose TBRs are marginal.

Blanket design is in progress in helical-type compact reactor FFHR-2. A localized blanket concept is proposed by selecting molten-salt Flibe as a self-cooling tritium breeder from the main reason of safety: low tritium solubility, low reactivity with air and water, low pressure operation, and low MHD resistance which is compatible with the high magnetic field design in force-free helical reactor (FFHR). Numerical results are presented on nuclear analyses using the MCNP-4B code, on thermal and stress analyses using the ABAQUS code, and heat exchange efficiency from Flibe to He. R&D programs on Flibe engineering are also in progress in material dipping-tests and in construction of molten salt loop. Preliminary results in these experiments are also presented.

Thermonuclear fusion power stations based on the deuterium-tritium reaction require breeding blankets to produce the tritium (T) fuel consumed in the plasma. This paper resumes the state-of-the-art of the T related technology from the initial T production in the lithium-bearing breeder material to a T stream which is ready for re-injection into the plasma. The remaining development issues are outlined and conventional techniques are confronted with advanced methods requiring more R&D effort but promising certain advantages in return.

The ARIES-ST (Spherical Tokamak) is to investigate the attractiveness of a low-aspect-device as the confinement concept for a fusion power plant. The key driven force of the ST design is caused by the center column conductor. The design selected is a water-cooled Cu normal conductor. This selection has a major impact on the blanket design and selection, tritium breeding and over-all power balance. The blanket selected is a dual coolant concept, partially decided by the characteristics of the center conductor. The final blanket design is modified from the dual coolant concept, which developed under the EC DEMO program. The reason for this selection and the design issues are summarized in this paper.

The main feature of FFHR is force-free-like configuration of helical coils, which makes it possible to simplify the coil supporting structure and to use high magnetic field instead of high plasma beta. The other feature is the selection of molten-salt Flibe as a self-cooling tritium breeder from the main reason of safety. Collaboration works based on the LHD project have made great progress in the reactor studies by focusing on engineering aspects of the high magnetic field and Flibe system design. Encouraging positive results are shown on ignition access, mechanical stress in coils supporting structures, improvement in the blanket system including materials selection and tritium recovery. Critical issues on fundamental safety analysis and maintainability of reactor components are also discussed, and many subjects are pointed out as future works.

The ARIES-RS fusion power plant design study is based on reversed-shear (RS) physics with a Li/V (lithium breeder and vanadium structure) blanket. The reversed-shear discharge has been documented in many large tokamak experiments. The plasma in the RS mode has a high beta, low current, and low current drive requirement. Therefore, it is an attractive physics regime for a fusion power plant. The blanket system based on Li/V has high temperature operating capability, good tritium breeding, excellent high heat flux removal capability, long structural life time, low activation, low after heat and good safety characteristics. For these reasons, the ARIES-RS reactor study selected Li/V as the reference blanket. The combination of attractive physics and attractive blanket engineering is expected to result in a superior power plant design. This paper summarizes the power core design of the ARIES-RS power plant study.

The purpose of the ITER Blanket Testing Module is to test the operating and performance of candidate blanket concepts under a real fusion environment. To assure fuel self-sufficiency, the tritium breeding, recovery and processing have to be demonstrated. The tritium produced in the blanket has to be processed to a purity which can be used for refuelling. All these functions need to be accomplished so that the tritium system can be scaled to a commercial fusion power plant from a safety and reliability point of view. This paper summarizes the tritium processing steps, the size of the equipment, power requirements, space requirements, etc. for a self-cooled lithium blanket. This information is needed for the design and layout of the test blanket ancillary system and to assure that the ITER guidelines for remote handling of ancillary equipment can be met.

The ARIES-RS tokamak is a conceptual, D‐T-burning 1000 MWe power plant. As with earlier ARIES design studies, the final design of ARIES-RS was obtained in a self-consistent manner using the best available physics and engineering models. Detailed analyses of individual systems together with system interfaces and interactions were incorporated into the ARIES systems code in order to assure self-consistency and to optimize towards the lowest cost system. The ARIES-RS design operates with a reversed-shear plasma and employs a moderate aspect ratio (A4.0). The plasma current is relatively low (Ip11.32 MA) and bootstrap current fraction is high ( fBC 0.88). Consequently, the auxiliary power required for RF current drive is relatively low ( 80 MW). At the same time, the average

This letter notes the progress in plasma physics and device engineering required to make fusion power a viable option for electric power generation in the 21st century. (AIP) {copyright} {ital 1997 American Institute of Physics.}

To evaluate the anti-cancer effects of folic acid at the molecular level, we determined plasma folic acid concentration by radioimmuno-assay and the degree of total genomic DNA methylation by incubating DNA with 3H-S-adenosylmethionine (3H-SAM) in the presence of a methylase, and analyzed the methylation status of the c-myc and c-Ha-ras oncogenes by Southern blotting in 21 patients with advanced gastric cancer. The degree of total genomic DNA methylation of cancerous tissues was significantly lower than that of paracancerous and non-cancerous tissues; c-myc and c-Ha-ras oncogenes from cancerous (10/21, 5/10) and paracancerous (13/21, 4/10) tissues were hypomethylated. The plasma folic acid concentration in patients who showed hypomethylation was lower than that patients showing normal methylation. These findings suggest that a decrease in folic acid, and the subsequent DNA hypomethylation, may be involved in human gastric carcinogenesis.

This paper summarizes the current status of the Demo blanket test systems and how the ITER reactor design and operations are being accommodated. The US blanket program is planning to develop a liquid metal breeder and a solid breeder blanket for testing and evaluation. The test blanket modules will have prototypical components, materials, and coolants representative of power reactor systems. The modules are to be located in the ITER horizontal test ports and installed/removed with special remote handling equipment. Adjacent ITER blanket neutronic and temperature conditions suggest the use of an isolation frame surrounding the test blanket modules or submodules. This frame will also provide additional shielding to protect the adjacent vacuum vessel. The frame and blanket module are attached to the surrounding backplate to transfer static and dynamic loads. All coolants and tritium-bearing fluids will be routed out of the midplane port to special heat exchangers and tritium separation systems. Special remote handling equipment is being designed to install and extract the test blanket modules. Dedicated transporters will be used to move the blanket and shielding modules to dedicated hot cells. Special facility areas will be provided immediately outside the port areas for the heat exchangers, pumps, and tritium-separationmore » systems. 1 ref., 6 figs.« less

The Starlite team was asked to develop a power plant study for the US Demo. To define the mission of the Demo, a Utility Advisory Committee (UAC) was organized to establish the mission and requirement for the Demo power plant. Based on this input, the Starlite team outlined a set of top level requirements based on the advice provided by the UAC. With the mission and requirements thus established, the Starlite engineering team investigated various combinations of the structural material, breeding material and coolant for the blanket and shield. The reference design selected was with V-alloy as the structural material and Li as the coolant and breeder. The ability of this blanket to satisfy the top level requirements was also assessed. 11 refs., 1 fig., 1 tab.

Force-Free Helical Reactor. FFHR. is a conceptual design of helical fusion power reactor. Flibe is chosen as the coolant/ breeding material in this reactor mainly because of safety considerations. ...

A concept to recover tritium from lithium, based on a cold trap, has been developed as part of the U.S. contribution to ITER. The cold trap process can only reduce the tritium concentration to about 400 appm, which is far above the ITER design goal of reducing the tritium concentration in lithium to about 1 appm. To achieve this lower goal, protium is added to the lithium to a concentration higher than the saturation concentration of the hydrogen isotope at the cold trap temperature. Thus, LiH and LiT will precipitate out together at the cold trap. The tritium from the cold trap can be recovered by heating the Li(H + T) to 600 °C for decomposition. The H and T then can be separated by a cryogenic distillation process.

A series of reactor design studies based on the Tokamak configuration have been carried out under the direction of Professor Robert Conn of UCLA. They are called ARIES-I through IV. The key mission of these studies is to evaluate the attractiveness of fusion assuming different degrees of advancement in either physics or engineering development. This paper discusses the directions and conclusions of the blanket and related engineering systems for those design studies. ARIES-1 investigated the use of SiC composite as the structural material to increase the blanket temperature and reduce the blanket activation. Li{sub 2}ZrO{sub 3} was used as the breeding material due to its high temperature stability and good tritium recovery characteristics. The ARIES-IV is a modification of ARIES-1. The plasma was in the second stability regime. Li{sub 2}O was used as the breeding material to remove Zr. A gaseous divertor was used to replace the conventional divertor so that high Z divertor target is not required. The physics of ARIES-II was the same as ARIES-IV. The engineering design of the ARIES-II was based on a self-cooled lithium blanket with a V-alloy as the structural material. Even though it was assumed that the plasma was in the second stability more » regime, the plasma beta was still rather low (3.4%). The ARIES-III is an advanced fuel (D-{sup 3}He) tokamak reactor. The reactor design assumed major advancement on the physics, with a plasma beta of 23.9%. A conventional structural material is acceptable due to the low neutron wall loading. From the radiation damage point of view, the first wall can last the life of the reactor, which is expected to be a major advantage from the engineering design and waste disposal point of view. « less

The TITAN reversed-field-pinch (RFP) fusion-reactor study has two objectives: to determine the technical feasibility and key developmental issues for an RFP fusion reactor operating at high power density: and to determine the potential economic (cost of electricity), operational (maintenance and availability), safety and environmental features of high mass-power-density fusion-reactor systems. Mass power density (MPD) is defined as the ratio of net electric output to the mass of the fusion power core (FPC). The FPC includes the plasma chamber, first wall, blanket, shield, magnets, and related structure. Two different detailed designs TITAN-I and TITAN-II, have been produced to demonstrate the possibility of multiple engineering-design approaches to high-MPD reactors. TITAN-I is a self-cooled lithium design with a vanadium-alloy structure. TITAN-II is a self-cooled aqueous loop-in-pool design with 9-C ferritic steel as the structural material. Both designs use RFP plasmas operating with essentially the same parameters. Both conceptual reactors are based on the DT fuel cycle, have a net electric output of about 1000 MWe, are compact, and have a high MPD of 800 kWe per tonne of FPC. The inherent physical characteristics of the RFP confinement concept make possible compact fusion reactors with such a high MPD. The TITAN designs would meet the U.S. criteria for the near-surface disposal of radioactive waste (Class C, IOCFR61) and would achieve a high Level of Safety Assurance with respect to FPC damage by decay afterheat and radioactivity release caused by accidents. Very importantly, a “single-piece” FPC maintenance procedure has been worked out and appears feasible for both designs. Parametric system studies have been used to find cost-optimized designs. to determine the parametric design window associated with each approach, and to assess the sensitivity of the designs to a wide range of physics and engineering requirements and assumptions. The design window for such compact RFP reactors would include machines with neutron wall loadings in the range of 10–20 MW/m2 with a shallow minimum COE at about 18 MW/m2. Even though operation at the lower end of the this range of wall loading (10–12 MW/m2) is possible, and may be preferable, the TITAN study adopted the design point at the upper end (18 MW/m2) in order to quantify and assess the technical feasibility and physics limits for such high-MPD reactors. From this work, key physics and engineering issues central to achieving reactors with the features of TITAN-I and TITAN-II have emerged.

The TITAN reversed-field-pinch (RFP) fusion-reactor study has two objectives: to determine the technical feasibility and key developmental issues for an RFP fusion reactor operating at high power density; and to determine the potential economic operational, safety, and environmental features of high mass-power-density (MPD) fusion-reactor systems. Parametric system studies have been used to find cost-optimized designs. The design window for compact RFP reactors includes the range of 10–20 MW/m2. The reactors are physically small, and a potential benefit of this “compactness” is improved economics. The TITAN study adopted 18 MW/m2 in order to assess the technical feasibility and physics limits for such high-MPD reactors. The TITAN-I design is a lithium self-cooled design with a vanadium-alloy (V-3Ti-1Si) structural material. The magnetic field topology of the RFP is favorable for liquid-metal cooling. The first wall and blanket consist of single pass poloidal-flow loops aligned with the dominant poloidal magnetic field. A unique feature of the TITAN-I design is the use of the integrated-blanket-coil (IBC) concept. The lithium coolant in the blanket circuit is also used as the electrical conductor of the toroidal-field and divertor coils. A “single-piece” FPC maintenance procedure is used, in which the first wall and blanket are removed and replaced by vertical lift of the components as a single unit. This unique approach permits the complete FPC to be made of a few factory-fabricated pieces, assembled on site into a single torus, and tested to full operational conditions before installation in the reactor vault. A low-activation, low-afterheat vanadium alloy is used as the structural material throughout the FPC in order to minimize the peak temperature during accidents and to permit near-surface disposal of waste. The safety analysis indicates that the liquid-metal-cooled TITAN-I design can be classified as passively safe, without reliance on any active safety systems. The results from the TITAN study support the technical feasibility, economic incentive, and operational attractiveness of compact, high-MPD RFP reactors. Many critical issues remain to be resolved, however. The physics of confinement scaling, plasma transport and the role of the conducting shell are already major efforts in RFP research. However, the TITAN study points to three other major issues. First, operating high-power-density fusion reactors with intensely radiating plasmas is crucial. Second, the physics of toroidal-field divertors in RFPs must be examined. Third current drive by magnetic-helicity injection must be verified. The key engineering issues for the TITAN I FPC have also been defined. Future research and development will be required to meet the physics and technology requirements that are necessary for the realization of the significant potential economic and operational benefits that are possible with TITAN-like RFP reactors.

Materials recycling aspects including contact gamma dose rates and cooling times were investigated for the first wall, blanket, and shield components of future fusion power reactors. Candidate stru...

The magnetohydrodynamic (MHD) effects can present a feasibility issue for a self-cooled liquid metal blanket of magnetically confined fusion reactors, especially inboard regime of a tokamak. This pressure drop can be significantly reduced by using insulated wall structure. A self-healing insulating coating has been identified, which will reduce the pressure drop by more than a factor of 10. The future research direction to further quantify the performance of this coating is also outlined.

About three percent of the fusion energy produced by a D-{sup 3}He reactor is in the form of neutrons. Those neutrons are generated by D-D and D-T reactions, with the tritium produced by the D-D fusion. The neutrons will react with structural steel, deuterium, {sup 3}He and shielding material to produce tritium. About half of the tritium generated by the D-D reaction will not burn in the plasma and will exit as a part of the plasma exhaust. Thus, there is enough tritium produced in a D-{sup 3}He reactor and careful management will be required. The tritium produced in the shield and plasma can be managed with an acceptable effect on cost and safety. 3 refs., 2 figs., 3 tabs.

Virtual human with intelligent behavior plays an important role in distributed virtual environment. The paper studies path planning that is an essential intelligent behavior of virtual human. According to the features and requirements of distributed virtual environment, we propose a method which combines global static planning and local dynamic planning. Global static planning method builds ‘reachable point’ based on the abstract modeling of virtual environment, then adopts A* or Dijkstra algorithms to find optimal path and builds a matrix in advance. Local dynamic path planning takes advantage of ‘virtual sensing’ of the virtual human to realize obstacle-avoiding path adjustment in dynamic environment. Finally, we present the approach that implements the walking animation of web-based virtual human using VRML& H-Anim standards. We applied our approach to implement a virtual guide in a web-based virtual environment and got satisfying results.

The requirements for vacuum pumping and fuel processing for deuterium/helium (D/3He) burns in the physics operating phase for the International Thermonuclear Experimental Reactor (ITER) were assessed. These burns are expected to have low fusion power (100 MW), short burn times (≤ 30 s), limited operation (2000 shots), and a fractional burn ∼0.3%. For the physics phase, the fuel processing system will include several units to separate deuterium and helium (activated charcoal bed, SAES getter and a Pd/Ag diffuser), as well as an isotopic separation system to separate 3He and 4He. The needed vacuum system's cryorption surface area may be as large as 10 m2 if the burn time is ∼ 200 s, the fractional burn is 100 MW.

The ARIES design effort has sought to maximize the environmental and safety advantages of fusion through careful selection of materials and careful design. Three goals are that the reactor achieve inherent or passive safety, that no public evacuation plan be necessary and that the waste be disposable as 10CFR61 Class C waste. The ARIES-I reactor consists of a SiC composite structure for the first wall and blanket, cooled by 10 MPa He. The breeder is Li2ZrO3, although Li2O and Li4SiO4 were also considered. The divertor consists of SiC composite tubes coated with 2 mm of tungsten. Due to the minimal afterheat of this blanket design, LOCA calculations indicate maximum temperatures will not cause damage if the plasma is promptly extinguished. Two primary safety issues are the zirconium in the breeder and tungsten on the divertor. Li2ZrO3 was chosen because of its demonstrated high-temperature stability. The other breeders have lower afterheat and activation. Use of zirconium in the breeder will necess...

A key safety concern in a D-T fusion reactor is the tritium inventory. There are three components in a fusion reactor with potentially large inventories, i.e., the blanket, the fuel processing system and the plasma facing components. The ARIES team selected the material combinations, decided the operating conditions and refined the processing systems, with the aiming of minimizing the tritium inventories and leakage. The total tritium inventory for the ARIES-I reactor is only 700 g. This paper discussed the calculations and assumptions we made for the low tritium inventory. We also addressed the uncertainties about the tritium inventory. 13 refs., 2 figs., 3 tabs.

During the ITER design phase, the conceptual design of the fuel processing cycle has been established. The fuel processing cycle is designed to be able to handle all the tritium containing streams of the ITER. These streams include plasma exhaust, blanket tritium recovery, pellet propellant, neutron beam exhaust, water coolant detritiation, waste water from the room air detritiation system. The design is very conservative, i.e., the flow rate of each stream is high and the detritiation factor required is very high. A preliminary optimization study has been carried out to simplify the ITER fuel cycle design. We investigated: The throughput and composition of the input tritium containing streams from various components to the fuel processing cycle. The fraction of those streams needed to be detritiated. The required detritiation factors required for each of the streams. The results of the investigation determined that the major input tritium containing steams can be reduced by at least a factor of 10. The required detritiation factor can be reduced from a factor of 100 to 10{sup 6}. The size of the fuel processing cycle, the tritium inventory and the complexity of this system can, therefore, also be reduced.

A 2:1 mixture of LiF and BeF{sub 2} (FLIBE), is a potential tritium breeder material for fusion reactors, in particular, the Advanced Safe Pool Immersed Reactor (ASPIRE). A limited experimental campaign was conducted in an effort to test the postulates of the ASPIRE concept: namely, that MoF{sub 6} is effective in controlling the tritium species by maintaining the FT form and that MoF{sub 6} can serve as a source to plate out Mo on surfaces, thereby making the FLIBE system compatible with the corrosive FT. It was demonstrated experimentally that successive additions of MoF{sub 6} achieved quantitative (i.e., greater than 99.7%) conversion of H{sub 2} to HF. Thus, MoF{sub 6} is effective in controlling the tritium species. The degree of conversion of H{sub 2} to H demonstrates that H does not attack MO to form H{sub 2}. This supports the postulate that the system is compatible with Mo. Thus, if it were possible to plate out and maintain a coating of Mo on all surfaces in contact with the FLIBE system, the ASPIRE concept could work. Thermodynamic calculations also confirmed that MoF{sub 6} should be capable of quantitatively (>99.9%) converting H{sub 2} to HF. There is both experimental and theoretical more » evidence that a number of MoF{sub x} species are present in both the gas phase and the FLIBE solution. 17 refs., 3 figs., 3 tabs. « less

This paper describes the results of an exploratory study of blanket concepts based on SiC/SiC structure and LiPb breeder. An assessment of the performance of these concepts for advanced power plant application is presented, key issues are identified, and constraints relating to the SiC/SiC properties are discussed.

The design integration of the fusion-power core (FPC) and the reactor subsystems of the ARIES-I (Advanced Reactor Innovation and Evaluation Study) are summarized. Details such as support of the toroidal-field coils, divertor module access, blanket access, design and support of the RF antennas, and location of the primary vacuum and cryostat vacuum boundaries are considered. The maintenance procedure being considered for ARIES-I is a modular approach. With this type of maintenance, a module consisting of the first-wall, blanket, shield, divertor module, and toroidal-field coil is replaced as a single unit at the end of the module's life. Rapid replacement of the irradiated FPC components is expected. In-situ blanket submodule repair and replacement schemes addressing the possible failure of the first-wall or blanket before its designed lifetime are described. Replacement of the divertor plate assemblies is simplified by providing a direct-access path through which damaged plates can be removed and new plates installed without interfering with the other FPC components. >

A design window concept is developed for a He-cooled fusion reactor blanket and divertor design. This concept allows study of a parameter regime under which a possible design exists with different design requirements, such as allowable pumping fraction. The concept identifies not only the required parameter regime, but also investigates the robustness of the design, i.e., the validity of the design with change of design parameters and requirements. Some recent directions of helium cooled design for ITER and for divertor can also be explained by this design window concept.

A description of the ARIES-III research effort is presented, and the general features of the ARIES-III reactor are described. The plasma engineering and fusion-power-core design are summarized, including the major results, the key technical issues, and the central conclusions. Analyses have shown that the plasma power-balance window for D-/sup 3/He tokamak reactors is small and requires a first wall (or coating) that is highly reflective to synchrotron radiation and small values of tau /sub ash// epsilon /sub e/ (the ratio of ash-particle to energy confinement times in the core plasma). Both first and second stability regimes of operation have been considered. The second stability regime is chosen for the ARIES-III design point because the reactor can operate at a higher value of tau /sub ash// tau /sub E// tau /sub E/ approximately=2 (twice that of a first stability version), and because it has a reduced plasma current (30 MA), magnetic field at the coil (14 T), mass, and cost (also compared to a first-stability D-/sup 3/He reactor). The major and minor radii are, respectively 7.5 and 2.5 m. >

The thermal-hydraulic and structural design of the ARIES-III divertor plate is presented. The divertor plate is made of small-diameter W-3Re tubes laid along the radial direction and coated with 4 mm of plasma-sprayed tungsten on the plasma-facing side to withstand one hard disruption. The plate is contoured to have the constant heat flux of 5.44 MW/m/sup 2/ on the entire surface. The total divertor thermal power of 629 MW is removed by the organic coolant HB-40 with the same inlet/exit temperatures (340 degrees C/425 degrees C) as in the first-wall/shield coolant circuit. The principal mode of heat transfer is by subcooled flow boiling. The inlet pressure is 5.34 MPa and the exit pressure is 4.3 MPa, which, by passing through an orifice, is reduced to 1 MPa, equal to the first-wall/shield exit pressure. The total coolant flow rate is 3.35 m/sup 3//s and the circulation power is 18 MWe. The maximum plate temperature is 821 degrees C. The safety factor with respect to the critical heat flux is >or=2, and it is approximately 3 with respect to the maximum allowable plate temperature. Maximum equivalent total, thermal, and pressure stresses are also given. >

Through its successful operation, the US Fusion Demo must be sufficiently convincing that a utility or independent power producer will choose to purchase one as its next electric generating plant. A fusion power plant which is limited to the use of currently-proven technologies is unlikely to be sufficiently attractive to a utility unless fuel shortages and regulatory restrictions are far more crippling to competing energy sources than currently anticipated. In that case, the task of choosing an appropriate set of engineering technologies today involves trade-offs between attractiveness and technical risk. The design space for an attractive tokamak fusion power core is not unlimited; previous studies have shown that advanced low-activation ferritic steel, vanadium alloy, or SiC/SiC composites are the only candidates we have for the primary in-vessel structural material. An assessment of engineering design options has been performed using these three materials and the associated in-vessel component designs which are compatible with them.

The need for a high flux, high energy neutron test facility to evaluate performance of fusion reactor materials is urgent. An accelerator based D-Li source is generally accepted as the most reasonable approach to a high flux neutron source in the near future. The idea is to bombard a high energy (35 MeV) deuteron beam into a lithium target to produce high energy neutrons to simulate the fusion environment. More recently it was proposed to use a 21 MeV triton beam incident on a water jet target to produce the required neutron source for testing and simulating fusion material environments. The advantages of such a system are discussed. Major concerns regarding the feasibility of this system are also highlighted.

A review of corrosion and environmental effects on the mechanical properties of candidate structural alloys for use with liquid metals in fusion reactors is presented. The corrosion/mass transfer behavior of austenitic and ferritic steels and vanadium-base alloys is evaluated to determine the preliminary operating temperature limits for circulating and static liquid-lithium and Pb-17Li systems. The influence of liquid-metal environment on the mechanical properties of structural materials is discussed. Corrosion effects of nitrate and fluoride salts are presented. Requirements for additional data are identified.

A preliminary concept for a heavy-ion beam driven inertial confinement fusion power plant is presented. The high repetition rate of the RF accelerator driver is utilized to serve four reactor chambers alternatingly. In the chambers a novel first-wall protection scheme is used. At a target gain of 83 the total net electrical output is 3.8 GW. The recirculating power fraction is below 15%. The main goal of the comprehensive HIBALL study (which is continuing) is to demonstrate the compatibility of the design of the driver, the target and the reactor chambers. Though preliminary, the present design is essentially self-consistent. Tentative cost estimates are given. The costs compare well with those found in similar studies on magnetic confinement fusion reactors.

The technical issues, development problems and required experiments and facilities for fusion nuclear technology have been investigated. The results have been used to develop a technical framework for a test plan that identifies the role, timing, characteristics and costs of major experiments and facilities. A major feature of this framework is the utilization of non-fusion facilities over the next 15 years, followed by testing in fusion devices beyond about the year 2000. Basic, separate effect and multiple interaction experiments in non-fusion facilities will provide property data, explore phenomena and provide input to theory and analytic modelling. Experiments in fusion facilities can proceed in two phases: (1) concept verification and (2) component reliability growth. Integrated testing imposes certain requirements on fusion testing device parameters; these requirements have been quantified. The nuclear subsystems addressed in the study are: (a) blanket and first wall; (b) tritium processing system; (c) plasma interactive components; and (d) radiation shield. The two generic classes of liquid and solid breeder blankets have significant engineering feasibility issues, and new experimental data must be obtained before selection of an attractive design concept. Liquid metal blanket issues are dominated by problems related to momentum, heat and mass transfer, which can be addressed in non-neutron test facilities. Solid breeder blanket issues are, however, dominated by the effects of radiation, including heating, transmutation and damage, which can be reasonably addressed in fission reactors. The tritium processing uncertainties are primarily related to the control and recovery systems, and most can be addressed in existing and planned non-neutron facilities. A dominant feature of plasma interactive components is the strong interrelation to both plasma physics and nuclear technology. Required facilities include thermomechanical test stands and confinement devices with sufficiently long plasma burn. The radiation shield poses no feasibility issues, but improved accuracy of predictions will reduce design conservatism and lower costs.

A comprehensive Blanket Comparison and Selection Study was conducted to evaluate proposed D-T fusion reactor blanket concepts and to identify those concepts that offer the greatest potential for fusion reactor applications. The multilaboratory study was led by Argonne National Laboratory and included support from thirteen industrial, national and university laboratories; six primary subcontractors and seven specialized contributors. The primary objectives of the program were (1) to identify a small number (approx. 3) of the blanket concepts that should be the focus of the blanket R and D program, (2) to define and prioritize the critical issues for the leading blanket concepts, and (3) to provide technical input for development of blanket R and D programs. A blanket concept is generally defined by the selection of the component materials, viz., breeder, coolant, structure, and neutron multiplier, and specification of the geometrical configuration. Blanket concepts were evaluated for both the tokamak and tandem mirror reactor configurations using the STARFIRE and MARS reactor designs as a basis, with appropriate modifications to reflect recent advances in technology.