Search

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.

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.

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.}

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.

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.

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.

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.

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.

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 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.

The methods used in recovering, containing and controlling tritium in fusion reactor systems are the subject of the paper. Safety, cost, and breeding needs for tritium control are outlined.

The use of Li/sub 2/O pellets suspended in a stream of helium gas flowing through a fusion reactor blanket can serve the dual functions of tritium breeding and cooling. The neutronics properties, such as tritium breeding ratio and total nuclear heating, have been studied as functions of the enrichment of /sup 6/Li, the breeding and graphite zone thickness, and solid concentration in the suspension. The performance of the shield and the thermal reaction of the magnet in case of loss of coolant accident (LOCA) was also investigated. The radioactivity and afterheat in such blankets after shutdown were calculated.

A double-walled steam generator is used in the fusion power cycle to replace the intermediate loop. This will save the cost associated with a complicated system and the associated temperature degra...

A conceptual blanket design for UWMAK-II based on breeding in LiAlO 2 and helium cooling for a D-T fusion reactor is described. The reactor is a Tokamak with 316 stainless steel as the primary structural material, a major radius of 13 m and a minor radius of 5 m. The power output is 5000 MW(th) and the maximum temperature in the stainless steel structure is 650°C. This reactor design study is one of a series performed to evaluate the merits of various fusion reactor design concepts. In this paper the mechanical and the thermal hydraulics problem associated with the blanket for this reactor is described. Special attention has been given to the need for repairing and replacing the first wall of the blanket. Other problems which may arise from such a blanket design are also discussed.

STARFIRE is a design for a conceptual commercial tokamak electrical power plant based on the deuterium/tritium/lithium fuel cycle. In addition to the goal of being technologically credible, the design incorporates safety and environmental considerations. STARFIRE is considered to be the tenth in a series of commercial fusion power plants. STARFIRE has a 7-m major radius reactor producing 1200 MW of net electrical power from 4000 MW of thermal power, with an average neutron wall load of 3.6 MW/m 2 . The aspect ratio is 3.6 and a D-shaped plasma with a height-to-width ratio of 1.6 and average toroidal beta of 0.067 is used. The maximum magnetic field is 11T. Availability goals have been set at 85% for the reactor and 75% for the complete plant including the reactor. The major features for STARFIRE include a steady-state operating mode based on a continuous rf lower-hybrid current drive and auxiliary heating, solid tritium breeder material, pressurized water cooling, limiter/vacuum for impurity control, most superconducting EF coils outside the TF superconducting coils, fully remote maintenance, and a low-activation shield.

A gravity-flow bed of lithium oxide microspheres has been proposed as the tritium breeding and blanket cooling medium for both magnetic confinement and inertial confinement controlled fusion reactors using the D-T cycle. The use of the lithium oxide moving bed could help eliminate MHD, corrosion, and materials compatibility problems associated with the use of liquid lithium. Adequate particle transport mechanisms which alleviate particle attrition and erosion seem attainable. Moving bed heat transfer coefficients an order of magnitude greater than for fluidized beds are indicat, ed. The fabrication of microspheres is complicated by the aggressive caustic nature of the lithium oxide but should be possible for commercial-scale production.

This blanket concept uses a dilute suspension of fine solid breeder particles (Li/sub 2/O, LiAlO/sub 2/, or Li/sub 4/SiO/sub 4/) in a carrier gas (He) as the coolant and the tritium breeding stream. A small fraction of this stream is processed outside the reactor for tritium recovery. The blanket consists of a beryllium multiplier and carbon/steel reflector. A steel clad is used for all materials. A carbon reflector is employed to reduce the beryllium thickness used in the blanket for a specific tritium breeding ratio. The breeder particle size has to exceed few microns (greater than or equal to2 microns) to avoid sticking problems on the cold surfaces of the heat exchanger. The helium gas pressure is in the range of 2 to 3 MPa to carry the blanket and the heat exchanger loop. The solid breeder concentration in the helium stream is 1 to 5 volume percent. A high lithium-6 enrichment is used to produce a high tritium breeding ratio and to reduce the breeder concentration in the helium gas. At a lithium-6 enrichment of 90%, the local tritium breeding ratio is 2.03 based on a one-dimensional poloidal model. The total thickness of the helium stream is only 4more » cm out of the 50 cm total blanket thickness. The blanket uses a 35 cm of beryllium for neutron multiplication. A simple multi-layer design is employed where the blanket sector has the helium coolant flowing in the poloidal direction. The blanket concept has several unique advantages which are very beneficial for fusion reactors including ITER. 10 refs., 2 tabs.« less