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The 3 GJ target with low repetition rate and thick liquid wall chamber presents the mainline choice for the Z-Pinch power plant. An engineering scoping assessment has been developed for two candidate breeders (Flibe (F 4 Li 2 Be) molten salt and Li 17 Pb 83 liquid metal) to identify the design requirements and optimize the components' dimensions. Several important engineering features have been incorporated to improve the Z-Pinch performance. For instance, an advanced high-temperature steel-based structure could operate near 800°C, an advanced power cycle could achieve high thermal conversion efficiency approaching 50%, a low-activation F82H-based steel with controlled impurities will generate only low-level waste, and an innovative idea has been developed to establish jet flow using a sluice valve. This paper identifies self-consistent reference parameters and documents an interesting comparison between the candidate breeders, highlighting the fundamental differences in performance and the benefits and drawbacks of each breeder.

A blanket concept made of the low electrical conductivity SiC/SiC composite and utilizing Li 17 Pb 83 as coolant and tritium breeder has been developed and integrated with the magnetic diversion system. Neutronics issues related to tritium breeding adequacy particularly with the area lost to the dump plates at the ring and point cusps were addressed. Radiation damage and lifetime considerations for the SiC/SiC structural material were also addressed. Another issue of concern is providing adequate shielding for the superconducting cusp magnets. Detailed neutronics analyses show that tritium self-sufficiency can be achieved. A 0.5 m thick water-cooled steel shield that doubles as the vacuum vessel is a reweldable lifetime component and will provide adequate shielding for the magnets.

The initiation of the Global Nuclear Energy Partnership includes nuclear-based transmutation devices to recycle the spent fuel. Fusion can offer an alternative to the use of fast reactors for the transmutation of actinides. At a modest fusion power of 20 MW, a Z-Pinch driven sub-critical blanket can burn actinides and produce power. Several engineering issues have been examined: the effect of the sub-critical blanket and its internal fission neutrons on tritium breeding, radiation damage to structure, energy deposition and extraction, and chamber activation. Our initial assessment indicates the Z-Pinch could be an attractive option for burning actinides, but special attention should be paid to the challenging engineering issues.

The ARIES-AT power core was evolved with the overall objective of achieving high performance while maintaining attractive safety features, credible maintenance and fabrication processes, and reasonable design margins as a rough indication of reliability. The blanket and divertor designs are based on Pb-17Li as coolant and breeder, and low-activation SiCf/SiC as structural material. Flowing Pb-17Li in series through the divertor and blanket is appealing since it simplifies the coolant routing and minimize the number of cooling systems. However, Pb-17Li provides marginal heat transfer performance in particular in the presence of MHD effects and the divertor design had to be adapted to accommodate the peak design heat flux of 5 MW/m2. The blanket flow scheme enables operating Pb-17Li at a high outlet temperature (about 1100 °C) for high power cycle efficiency while maintaining SiCf/SiC at a substantially lower temperature consistent with allowable limits. Waste minimization and additional cost savings are achieved by radially subdividing the blanket into two zones: a replaceable first zone and a life of plant second zone. Maintenance methods have been investigated which allow for end-of-life replacement of individual components. This paper summarizes the results of the design study of the ARIES-AT power core focusing on the blanket and divertor and including a discussion of the key parameters influencing the design, such as the SiCf/SiC properties and the MHD effects, and a description of the design configuration, analysis results and reference operating parameters.

The ARIES-AT study was initiated to assess the potential of high-performance tokamak plasmas together with advanced technology in a fusion power plant and to identifying physics and technology areas with the highest leverage for achieving attractive and competitive fusion power in order to guide fusion R&D. The 1000-MWe ARIES-AT design has a major radius of 5.2 m, a minor radius of 1.3 m, a toroidal β of 9.2% (βN = 5.4) and an on-axis field of 5.6 T. The plasma current is 13 MA and the current-drive power is 35 MW. The ARIES-AT design uses the same physics basis as ARIES-RS, a reversed-shear plasma. A distinct difference between ARIES-RS and ARIES-AT plasmas is the higher plasma elongation of ARIES-AT (κx = 2.2) which is the result of a “thinner” blanket leading to a large increase in plasma β to 9.2% (compared to 5% for ARIES-RS) with only a slightly higher βN. ARIES-AT blanket is a simple, low-pressure design consisting of SiC composite boxes with a SiC insert for flow distribution that does not carry any structural load. The breeding coolant (Pb–17Li) enters the fusion core from the bottom, and cools the first wall while traveling in the poloidal direction to the top of the blanket module. The coolant then

A multi-institutional study HAPL (High Average Power Laser) is investigating a relatively near term conceptual design of a laser driven inertial confinement reactor. A primary focus of the study is the protection of the first wall (FW) from the target emanations. This paper gives a brief analysis of one of several possible blankets that can be integrated with the chosen FW protection scheme. The structural material is conventional ferritic steel (FS) F82H cooled with liquid lithium. The maximum average temperature is constrained to 550 deg. C. The chamber radius is 6.5 m at midplane, tapering to 2.5 m at the ends, and is surrounded by a cylindrical vacuum vessel. The first wall (FW) is 0.35 cm FS, which has a 0.1 cm thick layer of tungsten bonded to it facing the target. The FW is cooled with Li admitted at the bottom of the blanket, flows through a gap between 0.25-0.5 cm to the top, then returns through the center of the blanket channel to the bottom. There are 60 laser beam ports situated around the chamber. The tritium breeding ratio (TBR) is 1.124. A Brayton cycle is envisaged with an efficiency in the range of 42-44%.

The long-range goal of the Z-Pinch IFE program is to produce an economically-attractive power plant using high-yield z-pinch-driven targets (~3GJ) with low rep-rate per chamber (~0.1 Hz). The prese...

The Advanced Power Extraction (APEX) project has been exploring concepts for power producing blankets that can enhance the potential of fusion energy. The pursued concepts cover both liquid and solid wall designs. The solid wall blanket branch of the project has been concentrating on the use of nano-composited ferritic (NCF) steel structure coupled with Flibe (Li 2 BeF 4 ) coolant. The present blanket is a solid wall design with an innovative coolant scheme, which allows part of the coolant to be re-circulated in order to enhance the outlet temperature, and thus improve the power cycle efficiency. The structure is 12YWT, an oxide dispersion strengthened (ODS) ferritic steel, which has a maximum operating temperature of 800 °C and a compatibility with Flibe up to 700 °C. Several methods for fabricating the complex geometry of the first wall (FW) and blanket are presented. Preliminary coolant routing is proposed with solutions offered for minimizing heat losses and simplifying assembly and maintenance. Even though this blanket is somewhat complicated, its forward looking aims, that of maximizing nuclear performance to achieve a high thermal conversion efficiency, are well worth striving for.

As an element in the U.S. Advanced Power Extraction (APEX) program, the solid first wall and blanket design team assessed innovative design configurations with the use of advanced nano-composite ferritic steel (AFS) as the structural material and FLiBe as the tritium breeder and coolant. The goal for the assessment is to search for designs that can have high volumetric power density and surface heat-flux handling capability, with assurance of fuel self-sufficiency, high thermal efficiency and passive safety for a tokamak power reactor. We selected the re-circulating flow configuration as our reference design. Based on the recommended material properties of AFS we found that the reference design can handle a maximum surface heat flux of 1 MW/m2, and a maximum neutron wall loading of 5.4 MW/m2, with a gross thermal efficiency of 47%, while meeting all the tritium breeding, structural design and passive safety requirements. This paper will cover the results of the following areas of assessment: material design properties, FW/blanket design configuration, materials compatibility, components fabrication, neutronics analysis, thermal-hydraulics analysis including MHD effects, structural analysis; molten salt and helium closed cycle power conversion system; and safety and waste disposal of the re-circulating coolant first wall and blanket design.

As an element in the US Advanced Power Extraction (APEX) program, we evaluated the design option of using advanced nanocomposite ferritic steel (AFS) as the structural material and Flibe as the tritium breeder and coolant. We selected the recirculating flow configuration as our reference design. Based on the material properties of AFS, we found that the reference design can handle a maximum surface heat flux of 1 MW/m2, and a maximum neutron wall loading of 5.4 MW/m2, with a gross thermal efficiency of 47%, while meeting all the tritium breeding and structural design requirements. This paper covers the results of the following areas of evaluation: materials selection, first wall and blanket design configuration, materials compatibility, components fabrication, neutronics analysis, thermal hydraulics analysis including MHD effects, structural analysis, molten salt and helium closed cycle power conversion system, and safety and waste disposal of the recirculating coolant design.

The Advanced Power Extraction (APEX) program is exploring concepts for blanket designs that can enhance the potential of fusion while using both liquid and solid walls. In that context, an innovative blanket design of dry solid wall configuration, with potential for high neutron wall loading has been proposed. The blanket utilizes nanocomposited ferritic (NCF) steel structure (designated 12YWT), which has a maximum operating temperature of 800 deg. C. The cooling/breeding material is Flibe (Li{sub 2}BeF{sub 4}), a low viscosity version of this molten salt, which has a melting temperature of 465 deg. C and is compatible with ferritic steels up to 700 deg. C. The dimensions for this study have been taken from ARIES-AT. The blanket module, which extends 0.3 m in the toroidal direction at mid-plane, is equipped with spiraling discs ramping from the bottom to the top. The coolant enters on the bottom at 500 deg. C, then travels on the spiral discs, from the rear of the module to the front, then back to the rear, all the way to the top where it exits from the module at 590 deg. C. On its way up, the coolant velocity is amplified at the first wall (FW) bymore » centrifugal action, providing a high heat transfer coefficient for dissipating the high surface heating. The 3 mm thick FW is scalloped with semicircular projections inclined in the flow direction. This facilitates unimpeded smooth flow at the FW, while at the same time stiffening the FW against pressure, obviating the need for welded reinforcements. The discs are made of two halves assembled together with a Be pebble bed enclosed as a neutron multiplier. Preliminary neutronics analysis for the outboard blanket has shown a local tritium breeding ratio (TBR) of 1.36, and an energy multiplication (M) of 1.26 including contribution from a secondary module, using natural Li. The pressure drop of 0.56 MPa in the module produces a modified primary bending stress of 124 MPa where the allowable stress for this material at 750 deg. C is 142 MPa. Without optimization, this blanket is capable of dissipating an average neutron wall loading of 6.4 MW/m{sup 2}, with a peak value of 9.6 MW/m{sup 2} and a peak surface heating of 1.3 MW/m{sup 2}. The coolant picks up energy in the secondary blanket exiting the reactor at 600 deg. C. Assuming the use of a supercritical steam power cycle, an efficiency of 48% can be expected.« less

ARIES-ST is a 1000 MWe fusion power plant based on a low aspect ratio ‘spherical torus’ (ST) plasma. The ARIES-ST power core was designed to accommodate the unique features of an ST power plant, to meet the top-level requirements of an attractive fusion energy source, and to minimize extrapolation from the fusion technology database under development throughout the world. The result is an advanced helium-cooled ferritic steel blanket with flowing PbLi breeder and tungsten plasma-interactive components. Design improvements, such as the use of SiC inserts in the blanket to extend the outlet coolant temperature range were explored and the results are reported here. In the final design point, the power and particle loads found in ARIES-ST are relatively similar to other advanced tokamak power plants (e.g. ARIES-RS [Fusion Eng. Des. 38 (1997) 3; Fusion Eng. Des. 38 (1997) 87]) such that exotic technologies were not required in order to satisfy all of the design criteria. Najmabadi and the ARIES Team [Fusion Eng. Des. (this issue)] provide an overview of ARIES-ST design. In this article, the details of the power core design are presented together with analysis of the thermal–hydraulic, thermomechanical and materials behavior of in-vessel components. Detailed engineering analysis of ARIES-ST TF and PF systems, nuclear analysis, and safety are given in the companion papers [4] , [5] , [6] , [7] .

A reassessment of the SOMBRERO laser driven fusion power plant that was designed in 1990–91 has been conducted. New information and analysis has confirmed most of the original design decisions except that the tritium inventory in the blanket may be larger than originally calculated. Possible methods of lowering the tritium inventory are described along with a discussion of the critical issues that still remain 10 years after the original design was completed.

The ARIES-AT blanket has been developed with the overall objective of achieving high performance while maintaining attractive safety features, simple design geometry, credible maintenance and fabrication processes, and reasonable design margins as an indication of reliability. The design is based on Pb–17Li as breeder and coolant and SiCf/SiC composite as structural material. This paper summarizes the results of the design study of this blanket. © 2001 Elsevier Science B.V. All rights reserved.

This paper summarizes the results of the design study of the ARIES-AT blanket and divertor, which have been developed with the overall objective of achieving high performance while maintaining attractive safety features, credible maintenance and fabrication processes, and reasonable design margins as an indication of reliability.

Using xenon as a fill gas in IFE chambers results in generating a large amount of radioactive xenon, iodine, and cesium isotopes. The cesium isotopes, 134 Cs and 136 Cs, and the iodine isotope, 126 I, would produce a high level of off-site dose (62.13 rem) at the plant site boundary if they were released entirely to the environment during an accident. The xenon gas is pumped out of the chamber (recycled) during operation to remove unburned T 2 and D 2 . Removing the Cs and I isotopes from the Xe gas during this recycling process will reduce the Cs and I inventories inside the chamber to negligible values. This process limited the off-site dose caused by the accidental release of Xe from the chamber of the SOMBRERO power plant to 29.9 mrem, which is less than the 1 rem no-evacuation limit. Using krypton, argon or neon as fill gases will result in producing lower levels of off-site dose than xenon.

ARIES-ST is a 1000 MW fusion power plant conceptual design based on a low aspect ratio ‘spherical torus’ (ST) plasma. The plasma-facing components include the inboard and outboard first wall, divertor plates, and plasma stability plates. Several unique aspects of ST plasma influence the engineering design. The limited inboard space precludes a full inboard divertor slot, such that substantial power and particle flows are expected to impact the inboard first wall. Strong requirements on plasma vertical stability require close-in conductors. The use of He-cooled tungsten for the stability plates allows them to act as plasma-interactive components operating at high temperature and moderately high (1–2 MW/m2) heat flux. High plasma core radiation fraction and a natural tendency for low inboard transport losses help to alleviate these inboard problems, such that the peak power loads are not expected to exceed the capability of He-cooled tungsten and steel structures. The main features of the plasma-facing components are summarized here together with the analysis of their thermal hydraulic and thermomechanical performance.

The advanced limiter-divertor plasma-facing systems (ALPS) program was initiated in order to evaluate the potential for improved performance and lifetime for plasma-facing systems. The main goal of the program is to demonstrate the advantages of advanced limiter:divertor systems over conventional systems in terms of power density capability, component lifetime, and power conversion efficiency, while providing for safe operation and minimizing impurity concerns for the plasma. Most of the work to date has been applied to free surface liquids. A multi-disciplinary team from several institutions has been organized to address the key issues associated with these systems. The main performance goals for advanced limiters and divertors are a peak heat flux of \ 50 MW:m 2 , elimination of a lifetime limit for erosion, and the ability to extract useful heat at high power conversion efficiency (40%). The evaluation of various options is being conducted through a combination of laboratory experiments, www.elsevier.com:locate:fusengdes

A plasma based, deuterium and tritium (DT) fueled, volumetric 14 MeV neutron source (VNS) has been considered as a possible facility to support the development of the demonstration fusion power reactor (DEMO). It can be used to test and develop necessary fusion blanket and divertor components and provide sufficient database, particularly on the reliability of nuclear components necessary for DEMO. The VNS device can be complement to ITER by reducing the cost and risk in the development of DEMO. A low cost, scientifically attractive, and technologically feasible volumetric neutron source based on the spherical torus (ST) concept has been conceived. The ST-VNS, which has a major radius of 1.07 m, aspect ratio 1.4, and plasma elongation three, can produce a neutron wall loading from 0.5 to 5 MW m−2 at the outboard test section with a modest fusion power level from 38 to 380 MW. It can be used to test necessary nuclear technologies for fusion power reactor and develop fusion core components include divertor, first wall, and power blanket. Using staged operation leading to high neutron wall loading and optimistic availability, a neutron fluence of more than 30 MW year m−2 is obtainable within 20 years of operation. This will permit the assessments of lifetime and reliability of promising fusion core components in a reactor relevant environment. A full scale demonstration of power reactor fusion core components is also made possible because of the high neutron wall loading capability. Tritium breeding in such a full scale demonstration can be very useful to ensure the self-sufficiency of fuel cycle for a candidate power blanket concept.

The scientific parameters and the technology issues for a modest size spherical torus (ST) at 10 MA plasma current are discussed. This class of devices includes a DT-capable ST experiment (DTST, R0 = 1.2 m) for extended plasma performance tests for limited pulse lengths and neutron fluences, and a volume neutron source (VNS, R0 = 1.1 m) for steady state energy technology testing to high neutron fluences. The scientific issues of interest for DTST include non-inductive ramp-up of plasma current on a limited timescale (~30 s), the confinement needed for high Q burn, the behaviour of energetic particles, the physics and techniques to handle intense plasma exhaust, and the possibility of high performance plasma regimes free of disruptions or large disruption impact. Of further interest for the VNS would be steady state operation using large external current drive, possibly at a modest Q (~1-2), achieving significant neutron wall loading (~1 MW/m2) and a configuration relatively amenable to remote maintenance. A much longer timescale would be permitted in a VNS for non-inductive current ramp-up. The centre leg of the toroidal field coils, possibly multiturn for DTST and necessarily single turn for a VNS without significant nuclear shielding, presents technical and material issues of unique importance to the ST. Positive ion neutral beam injection and high harmonic fast wave (~80 MHz) heating and current drive systems already available are likely to be adequate for DTST following pulse length extension to ~50 s. Given an adequate physics database, the remaining enabling technologies needed for the VNS appear largely similar in nature to those of the ITER EDA design.

The mechanical and thermal design issues of the centerpost (CP) for a spherical torus based volumetric neutron source (ST-VNS) is described. It is intended to provide a test bed for developing nuclear technologies as well as for qualifying blanket designs for use in fusion devices. As scoped out, the VNS is capable of staged operation from a neutron wall loading of 0.5–5.0 MW/m 2 as the physics and engineering design assumptions are raised from modest to aggressive levels. Design margins are ensured since operation will be adequate at a wall loading of 2 MW/m 2 . The device has a major radius of 1.07 m, a minor radius of 0.77 m for an aspect ratio of 1.4, an elongation of 3, triangularity of 0.6 and has a naturally diverted plasma. It can be driven with neutral beams (NB) or ICRF. In the NB driven case, the plasma current is 11.1 MA, and in the ICRF case it is 13.2 MA. The magnetic field at the plasma major radius is 2.13 T and 2.55 T for the NB and the ICRF cases, respectively. This ST-VNS utilizes a single turn unshielded normal conducting CP fabricated from dispersion strengthened Cu, is 15.5 m long, has a diameter of 0.55 m at the midplane, a diameter of 1.2 m at the extremities and, in the ICRF driven case, carries a current of 13.6 MA. For the same case, the ohmic heating at the start of operation is 153 MW, increasing to 178 MW after 3 full power years. Thermal hydraulics, stress analysis and effects of transmutation on the CP have been determined, and a procedure for replacing the CP with minimal down time is described.

Progress is given on the investigation of a low cost, scientifically attractive, and technologically feasible volumetric neutron source (VNS) based on the spherical torus (ST) concept. The ST-VNS h...

Mechanical, thermal and neutronics design aspects of the toroidal field coil centerpost (CP) for a spherical torus based volumetric neutron source (ST-VNS) are presented. It is being investigated w...

The purpose of the ALPS program is to identify and evaluate advanced limiter/divertor systems that will enhance the attractiveness of fusion power. The highest priority goals at present are achievi...

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.

With the worldwide development of fusion power focusing on the design of the International Thermonuclear Experimental Reactor (ITER), developmental strategies for the demonstration fusion power plant (DEMO) are being discussed. A relatively prudent strategy is to construct and operate a small deuterium–tritium fuelled volumetric neutron source (VNS) in parallel with ITER. The VNS is to provide, over a period less than 20 years, a relatively high fusion neutron fluence of 6 MW year m−2 and wall loading of 1 MW m−2 or more, over an accessible blanket test area of more than 10 m2. Such a VNS would complement ITER in testing, developing, and qualifying nuclear technology components, materials, and their combinations for DEMO and future commercial power plants. The effort of this study has established the potential of the spherical tokamak as a credible VNS concept that satisfies the above requirements.

A systems model of a laser-driven IFE power plant is being developed to assist in design trade-offs and optimization. The focus to date has been on modeling the fusion chamber, blanket and power conversion system. A self-consistent model has been developed to determine key chamber and thermal cycle parameters (e.g., chamber radius, structure and coolant temperatures, cycle efficiency, etc.) as a function of the target yield and pulse repetition rate. Temperature constraints on the tungsten armor, ferritic steel wall, and structure/coolant interface are included in evaluating the potential design space. Results are presented for a lithium cooled first wall coupled with a Brayton power cycle.

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

The ARIES-RS divertor system is selected and analyzed. A radiative divertor approach using Ne as the radiator is chosen to reduce the maximum heat flux to B 6M W m 2 . A 2 mm W layer is used to withstand surface erosion allowing a design life close to 3 full-power-years. This W coating on the V-alloy structure is castellated to meet structural design limits. A detailed description of the calculated heat flux distribution, thermal-hydraulics, structural analysis, fabrication methods and vacuum system design are presented. An innovative design using adjustable bolts is utilized to support the divertor plates, withstand disruption loads and allow adjustment of alignment between plates. With the exception of the concentration of Ne at the divertor, it is found that this divertor system design can satisfy all the design criteria and most of the functional requirements specified by the project. © 1997 Elsevier Science S.A.

LIBRA-SP is a 1000 MWe light ion beam driven inertial confinement fusion power reactor design study which utilizes a self-pinched mode for propagating ions to the target. It is driven by 7.2 MJ of ...

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.

Engineering design issues of a volumetric neutron source (VNS) based on a steady state low aspect ratio DT tokamak are presented. At the present the major radius is 0.8 m, the minor radius 0.6 m for an aspect ratio of 1.33, the plasma current is 10.1 MA, the toroidal field at the major radius is 1.8 T, and the fusion power is 39 MW giving an average neutron wall loading of 1.0 MW/m{sup 2} on the outboard side with an available testing area of 10 m{sup 2}. Two neutral beams delivering more than 20 MW are used to drive the steady state fusion plasma. A single turn unshielded water cooled dispersion strengthened (DS) Cu centerpost is used in conjunction with a conducting Cu bell jar which acts as a vacuum boundary and the return legs for the toroidal field (TF) coils. The centerpost is 9 m long, carries 7.2 MA and is specially shaped to minimize ohmic heating, which is calculated using temperature dependent DS Cu properties and increases in resistivity due to nuclear transmutations are accounted for. A naturally diverted plasma scrapeoff layer dominated by pressure-driven instabilities is assumed giving a peak heat flux of 5.2 MW/m{sup 2} onmore » the diverter plates. Fabrication approaches for the centerpost and its replacement time lines have been estimated to be feasible and reasonable. 5 refs., 6 figs., 2 tabs.« less

The design of a 1,000 MWe light ion driven fusion power plant is given. Considerable progress has been made in analyzing the target performance and. its impact on the cavity design. Recent declassi...

The recent U.S. effort on the ITER (International Thermonuclear Experimental Reactor) shield has been focused on the limiter module design. This is a multi-disciplinary effort that covers design layout, fabrication, thermal hydraulics, materials evaluation, thermo- mechanical response, and predicted response during off-normal events. The results of design analyses are presented. Conclusions and recommendations are also presented concerning, the capability of the limiter modules to meet performance goals and to be fabricated within design specifications using existing technology.

The use of light ion or electron beams to compress matter to the densities required for fusion has been proposed for more than 20 years. In the past ten years, a series of light ion beam power plan...

This paper describes the design of a 1000 MWe inertially confined fusion power reactor utilizing near symmetric illumination provided by a KrF laser. The nominal laser energy is 3.4 MJ, the target ...

Ion channeling is a recent idea for improving the performance of fusion reactors by increasing the fraction of the fusion power deposited in the ions. In this paper the authors assess the effect of ion channeling on D-T and D-{sup 3}He reactors. The figures of merit used are the fusion power density and the cost of electricity. It is seen that significant ion channeling can lead to about a 50-65% increase in the fusion power density. For the Apollo D-{sup 3}He reactor concept the reduction in the cost of electricity can be as large as 30%.

The neutron yield in a D-{sup 3}He reactor is much lower than that in a D-T reactor of equivalent power. Therefore, the rate of neutron damage and gas production in the first wall of D-{sup 3}He reactors is lower by more than an order of magnitude. Whereas different structural materials proposed for use in commercial fusion reactors will last the reactor lifetime of 30 full power years in a D-{sup 3}He reactor, frequent replacement of the first wall and blanket will be required during the lifetime of a D-T power reactor. The blanket modules may require 30 replacements depending on the material used and the maximum allowable damage level. The down time required for replacement of the first wall and blanket in a D-T reactor will impact the reactor availability and consequently the cost of electricity. It appears that a D-{sup 3}He reactor should have a 10% advantage in availability over a D-T reactor.

Magnetohydrodynamic (MHD) effects of the liquid metal self-cooled blanket proposed for ITER are discussed in this paper. Scoping calculations of heat transfer, MHD pressure drop and structure stresses at a steady state for the self-cooled lithium/vanadium inboard blanket design have been performed in order to show if the blanket option can meet the prescribed design criteria, or if modifications are required. The finite element computer code ANSYS TM (DeSalvo and Gorman, ANSYS User Manual, Swanson Analysis System, Inc., 1989) is used to compute two-dimensional temperature and stress distribution in the inboard blanket. The results of the investigation indicate that the ITER self-cooled lithium/vanadium blanket can satisfy the design criteria from the standpoint of heat transfer, MHD pressure drop and stresses. A comfortable safety margin can be obtained if insulting materials are used to decouple the conductive walls from the eddy currents resulting from the flow of liquid metals across magnetic fields.

This paper reports that a combination of man-made and natural resources on earth could provide sufficient {sup 3}He fuel for the technological development of D-{sup 3}He fusion reactors. Helium exists in natural gas wells; however, at the current rate of natural gas usage, this resource would provide {lt}5 kg/yr of {sup 3}He. The radioactive decay of {sup 3}H produced in fission production reactors could yield 110 kg of {sup 3}He by the year 2000 if it were retained. Apparently, a large amount of {sup 3}He exists within the earth's mantle, but it is inaccessible. A significant quantity of {sup 3}He, which could be imported to supply a fusion power industry on earth for hundreds of years, exists on the moon. The solar wind has deposited {gt}1 million tonnes of {sup 3}He in the fine regolith that covers the surface of the moon. The presence of this solar wind gas was confirmed by analyses of the lunar regolith samples brought to earth. A strong correlation is noted between the helium retained and the TiO{sub 2} content of the regolith; consequently, remote-sensing data showing high-titanium-bearing soils in the lunar maria areas have been used to locate potentially rich sites for helium extraction.more » Surface photographs of Mare Tranquillitatis have shown that nearly 50% of this mare may be minable and capable of supplying {approximately}7100 tonnes of {sup 3}He. A mobile mining vehicle is proposed for use in the excavation of the soil and the release of the helium and other solar wind gases. The evolved gases would be purified by a combination of permeators and cryogenic techniques to provide a rich resource of H{sub 2}, helium, CO{sub 2}, H{sub 2}O, and N{sub 2}, followed by helium isotopic separation systems.« less

The key features of Apollo, a conceptual D-3He tokamak reactor for commercial electricity production, are summarized. The 1000-MW(electric) design utilizes direct conversion of synchrotron radiatio...

The SOMBRERO inertial fusion reactor conceptual design study is of a 1000 MWe KrF laser driven near symmetric illumination system which utilizes a Li2O solid breeder moving bed in a blanket made en...

The ARIES research program is a multi-institutional effort to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. Four ARIES visions are currentl...

A detailed safety analysis was performed for the inertial confinement fusion reactor OSIRIS. The radioactivity induced in the carbon fabric chamber concrete shield and Flibe breeder is very low allowing for their disposal at the end of the reactor life as Class A low level waste (LLW). The biological dose rate after shutdown behind the reactor biological shield shield is very low (0.11 µmrem/hr) allowing only for hands-on maintenance. A total of 91.5 Ci/day are routinely released to the environment producing an off-site dose to the maximally exposed individual (MEI) of 2.43 mrem/yr at the reactor site boundary. Only a small fraction (0.2%) of the reactor first wall would be mobilized during a loss of coolant/loss of flow accident. The decay heat generated in the concrete shield is very low such that its temperature would only increase by less than 2 degrees during such an accident OSIRIS contains 660 tonnes of liquid Flibe as a coolant and breeder. A severe accident including a breach of the react...

A magnetic lens has been designed which will focus 30 MeV Li+ ions emitted from a diode onto a DT filled target as required in the LIBRA-LiTE fusion reactor study. Thirty such lenses are located 2....

A comparison of the key features of the D-3He Apollo and the DT Aries fusion power reactor designs is made. The reduction in neutron production from the D-3He reaction has a major effect on the per...

ARIES-III is a D-He{sub 3} reactor design study. It is found that the organic coolant is well suited for the D-He{sub 3} reactor. This paper discusses the unique features of the D-He{sub 3} reactor, and the reason that the organic coolant is compatible with those features. The problems associated with the organic coolant are also discussed. 8 refs., 2 figs., 6 tabs.