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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.

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.

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.

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.

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.

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. >

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.

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.

Results of the self-cooled, liquid-metal blanket design from the Blanket Comparison and Selection Study (BCSS) are summarized. The objectives of the BCSS project are to define a small number (about three) of blanket concepts that should be the focus of the blanket research and development (RandD) program, identify and prioritize the critical issues for the leading blanket concepts, and provide technical input necessary to develop a blanket RandD program plan. Two liquid metals (lithium and lithium-lead (17Li-83Pb)) and three structural materials (primary candidate alloy (PCA), ferritic steel (FS) (HT-9), and vanadium alloy (V-15 Cr-5 Ti)) are included in the evaluations for both tokamaks and tandem mirror reactors (TMRs). TMR is of the tube configuration similar to the Mirror Advanced Reactor Study design. Analyses were performed in the following generic areas for each blanket concept: MHD, thermal hydraulics, stress, neutronics, and tritium recovery. Integral analyses were performed to determine the design window for each blanket design. The Li/Li/V blanket for tokamak and the Li/Li/V, LiPb/LiPb/V, and Li/Li/HT-9 blankets for the TMR are judged to be top-rated concepts. Because of its better thermophysical properties and more uniform nuclear heating profile, liquid lithium is a better coolant than liquid 17Li83Pb. From an engineeringmore » point of view, vanadium alloy is a better structural material than either FS or PCA since the former has both a higher allowable structural temperature and a higher allowable coolant/structure interface temperature than the latter. Critical feasibility issues and design constraints for the self-cooled, liquid-metal blanket concepts are identified and discussed.« less

The requirements of tritium technology are centered in three main areas, i.e., (1) fuel processing, (2) breeder tritium extraction, and (3) tritium containment. The gaseous tritium stream from the breeder tritium extraction system is significantly different from the plasma exhaust stream and, therefore, may have an important impact on the operation of the fuel processing system. For some blankets, such an aqueous solution blanket, the blanket tritium stream may dominate the fuel processing system in terms of component size and power consumption. The importance of the blanket interface to a fuel processing experiment, such as TSTA, has been identified. The initial work to define the blanket processing system, which is proposed to be added as part of TSTA, will be discussed here.

A water cooled solid-breeder blanket concept was developed for ITER. The main function of this blanket is to produce the necessary tritium for the ITER operation. Several design features are incorporated in this blanket concept to increase its attractiveness. The main features are the following: (a) a multilayer concept which reduces fabrication cost; (b) a simple blanket configuration which results in reliability advantages; (c) a very small breeder volume is employed to reduce the tritium inventory and the blanket cost; (d) a high tritium breeding ratio eliminates the need for an outside tritium supply; (e) a low-pressure system decreases the required steel fraction for structural purposes; (f) a low-temperature operation reduces the swelling concerns for beryllium; and (g) the small fractions of structure and breeder materials used in the blanket reduce the decay heat source. The key features and design analyses of this blanket are summarized in this paper.

This paper describes the mechanical and thermal design features of the MARS (Mirror Advanced Reactor Study) blanket and presents a credible concept for maintaining the central cell components of the reactor.

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

We have initiated a study to define a blanket processing mockup for Tritium Systems Test Assembly. Initial evaluation of the requirements of the blanket processing system have been started. The first step of the work is to define the condition of the gaseous tritium stream from the blanket tritium recovery system. This report summarizes this part of the work for one particular blanket concept, i.e., a self-cooled lithium blanket. The total gas throughput, the hydrogen to tritium ratio, the corrosive chemicals, and the radionuclides are defined. The key discoveries are: the throughput of the blanket gas stream (including the helium carrier gas) is about two orders of magnitude higher than the plasma exhaust stream;the protium to tritium ratio is about 1, the deuterium to tritium ratio is about 0.003;the corrosion chemicals are dominated by halides;the radionuclides are dominated by C-14, P-32, and S-35;their is high level of nitrogen contamination in the blanket stream. 77 refs., 6 figs., 13 tabs.