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The luminescence arising from lanthanide cations offers several advantages over organic fluorescent molecules: sharp, distinctive emission bands allow for easy resolution between multiple lanthanide signals; long emission lifetimes (μs -ms) make them excellent candidates for time-resolved measurements; and high resistance to photo bleaching allow for long or repeated experiments. A method is presented for determination of nucleosides using the effect of enhancement of fluorescence of the easily accessible europium(III)-TNB in presence of different nucleosides. The latter coordinates to Eu(III) -TNB and enhances its luminescence intensity as a result of the displacement of water from the inner coordination sphere of the central metal. A similar method for the determination of DNA based on the quenching of Eu(III)-TNB has been established. The interaction of Eu(III)-4,4,4 trifluoro-1-(2-naphthyl)1,3-butanedione (TNB) complex with nucleosides (NS) (guanosine, adenosine, cytidine, inosine) and DNA has been studied using normal and time-resolved luminescence techniques. Binding constants were determined at 293 K, 298 K, 303 K, 308 K and 313 K by using Benesi-Hildebrand equation. A thermodynamic analysis showed that the reaction is spontaneous with ΔG being negative. The enthalpy ΔH and the entropy ΔS of reactions were all determined. The formation of binary and ternary complexes of Eu(III) with nucleosides and TNB has been studied potentiometrically at (25.0 ± 0.1) °C and ionic strength I = 0.1 mol.dm(-3) (KNO3) . The formation of the 1:1 binary and 1:1:1 ternary complexes are inferred from the corresponding titration curves. Initial estimates of the formation constants of the resulting species and the protonation constants of the different ligands used have been refined with the HYPERQUAD computer program. Electrochemical investigations for the systems under investigations have been carried out using cyclic voltammetry (CV), differential pulse polarography (DPP), and square wave voltammetry (SWV) on a glassy carbon electrode in I = 0.1 mol/L p-toluenesulfonate as supporting electrolyte.

This work describes the application of time resolved fluorescence in microtiterplates and electrochemical methods on glassy carbon electrode for investigating the interactions of europium-3-carboxycoumarin with pesticides aldicarb, methomyl and prometryne. Stern-volmer studies at different temperatures indicate that static quenching dominates for methomyl, aldicarb and prometryne. By using Lineweaver-Burk equation binding constants were determined at 303 K, 308 K and 313 K. A thermodynamic analysis showed that the reaction is spontaneous with ΔG being negative. The enthalpy ΔH and the entropy ΔS of reactions were all determined. A time-resolved (gated) luminescence-based method for determination of pesticides in microtiterplate format using the long-lived europium-3-carboxycoumarin has been developed. The limit of detection is 4.80, 5.06 and 8.01 μmol L(-1) for methomyl, prometryne and aldicarb, respectively. This is the lowest limit of detection achieved so far for luminescent lanthanide-based probes for pesticides. The interaction of the probe with the pesticides has been investigated using cyclic voltammetry (CV), differential pulse polarography (DPP), square wave voltammetry (SWV) and linear sweep voltammetry (LSV) on a glassy carbon electrode in I = 0.1 mol L(-1) p-toluenesulfonate at 25 °C. The diffusion coefficients of the reduced species are calculated. The main properties of the electrode reaction occurring in a finite diffusion space are the quasireversible maximum and the splitting of the net SWV peak for Eu(III) ions in the ternary complex formed . It was observed that the increase of the cathodic peak currents using LSV is linear with the increase of pesticides concentration in the range 5 × 10(-7) to 1 × 10(-5) mol L(-1). The detection limit (DL) were about 1.01, 2.23 and 1.89 μmol L(-1) for aldicarb, methomyl and prometryne, respectively. In order to assess the analytical applicability of the method, the influence of various potentially interfering species was examined. Influence of interfering species on the recovery of 10 μmol L(-1) pesticides has been investigated.

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

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.

In this article we explore some of the safety issues associated with the Advanced Power Extraction (APEX) liquid metal first wall (FW) concept called the Convective Liquid Flow First Wall (CLiFF) design. In particular, we examine the chemical reactivity of and site boundary dose from three liquid metals being proposed for the CLiFF FW during a worst case confinement-boundary-bypass accident. The liquid metals considered are: lithium, tin, and gallium. The accident analyzed is a loss-of-vacuum accident (LOVA), during which air enters the CLiFF vacuum vessel (VV) from an adjoining room through a connecting diagnostic port or plasma-heating duct. The resulting lithium fire was analyzed and the energy released from this fire was found to be manageable for the CLiFF blanket design presently under consideration. The estimated dose at the site boundary is less than the no-evacuation limit of 10 mSv for ground level releases if plant isolation occurs within several hours.

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.

A tritium-breeding blanket design is investigated for a D-T Field-Reversed Configuration (FRC) scoping study. The thrust of our initial effort on the blanket has been to seek solutions as close to present-day technology as possible, and we have therefore focused on steel structure with helium coolant, The simple FRC cylindrical geometry has allowed us reasonable success due to the low FRC magnetic field and relatively easy maintenance. In this design the breeder is Li 2 O tubes. The design is modular with 10 modules each 2.5 m long. The inner radius of the first wall is 2.0 m and the FW/blanket/shield thickness is about 2 m. The surface heat flux will be radiation dominated, fairly uniform, and relatively low, because most of the charged particles follow the magnetic flux tubes to the end walls. The neutron wall loading is 5 MW/m 2 . In this design the surface heat flux equals 0.19 MW/m 2 . The maximum Li 2 O tube temperature is 1003°C, The helium exit temperature from the heat exchanger is about 800°C which allows a thermal efficiency of about 52%. The local tritium breeding ratio (TBR) equals 1.1 and is sufficient because in the FRC geometry the plasma has nearly full coverage. The helium pumping power is 1 MW. The coolant routing is optimized to limit the steel maximum temperature to 635°C. The same concept would be applicable to a spherical torus arid spheromak.

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

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.

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

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 ...

LIBRA-SP is a conceptual design study of an inertially confined 1000 MWe fusion power reactor utilizing self-pinched light ion beams. There are 24 ion beams which are arranged around the reactor cavity. The reaction chamber is an upright cylinder with an inverted conical roof resembling a mushroom, and a pool floor. The vertical sides of the cylinder are occupied by a blanket zone consisting of many perforated rigid HT-9 ferritic steel tubes called PERITs (PEr-forated RIgid Tube). The breeding/cooling material, liquid lead-lithium, flows through the PERITs, providing protection to the reflector/vacuum chamber so as to make it a lifetime component. The neutronics analysis and cavity hydrodynamics calculations are performed to account for the neutron heating and also to determine the effects of vaporization/condensation processes on the surface heat flux. The steady state nuclear heating distribution at the midplane is used for thermal hydraulics calculations. The maximum surface temperature of the HT-9 is chosen to not exceed 625{degree}C to avoid drastic deterioration of the metal`s mechanical properties. This choice restricts the thermal hydraulics performance of the reaction cavity. The inlet first surface coolant bulk temperature is 370{degree}C, and the heat exchanger inlet coolant bulk temperature is 502{degree}C. 4 refs., 6 figs.,more » 2 tabs.« less

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 ...

The thermal-structural behavior and performance of the SIRIUS-P power reactor first wall concept is analyzed. The SIRIUS-P conceptual design study is of a 1.0 GWe laser driven inertial confinement fusion power reactor utilizing near symmetric illumination of direct drive targets. Sixty laser beams providing a total of 3.4 MJ of energy are used at a repetition rate of 6.7 Hz with a nominal target gain of 118. The spherical chamber has an internal radius of 6.5 m and consists of a first wall assembly made from carbon-carbon composite material, and a blanket assembly made of SiC composite material. The chamber is cooled by a flowing granular bed of solid ceramic materials, non-breeding TiO{sub 2} for the first wall assembly and breeding Li{sub 2}O for the blanket assembly. Helium gas (P = 0.15 MPa) is used in a fluidized bed outside the reactor to return the particles to the top of the reactor. A moving bed is chosen over a fluidized bed because of its superior heat transfer capability. The heat transfer in a moving bed depends on the level of agitation and on the effective thermal conductivity of the solid material and the interstitial gas, whereas in a fluidized bed,more » it is entirely dominated by the thermal conductivity of the carrier gas. This work describes the three-dimensional thermo-structural steady state analysis of the first wall coolant tubes. The performance of the first wall depends, under normal operating conditions, on the thermal loading conditions and internal coolant pressure loading conditions. The analysis utilizes a commercial finite element analysis code with complete 3-D modeling. The analysis shows that the stresses are dominated by bending due to the internal pressure of the He gas; modifying the shape of the tube from purely elliptical at the midplane, while keeping the flow area constant, reduces the stresses. A comparison between the results of this 3-D model with a previous 2-D study shows a pronounced effect on the temperature distribution.« less

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.

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...

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...

Three different startup scenarios, one using pure D-3He, one using pure D-T to assist reaching the D-3He operating point, and one using a mixture of D-T-3He, have been analyzed, for the startup of ARIES-III. ARIES-III is a conceptual D-3He tokamak fusion power reactor operating in a second stability configuration. The process of starting the plasma up and bringing it to the desired operating point has been optimized to minimize the need for auxiliary ICRF heating during startup. In the second and third startup scenarios, seeding the plasma with tritium during startup reduces the amount of ICRF power required, but leads to a 14 MeV neutron pulse. Neutronics calculations have been performed to generate the nuclear heating profiles in the first wall and shield. The neutronics results were scaled with the neutron power to determine the nuclear heating profiles at different times during the startup phase. In this work, a two-dimensional transient thermal analysis is performed for the startup phases and...

The results from this study indicate that light ions can be a competitive factor in the race to commercial fusion power. The relatively simple and near-term driver technology is particularly attractive compared to higher cost laser and heavy ion schemes. The cavity design and engineering operations can be tailored such that Utilities could envision a reliable and maintainable power plant. The major problem to be faced now is the method of beam propagation to the target. The LIBRA-LiTE design reveals that ballistic transport may be more attractive from a physics standpoint, but the severe neutron environment presents a challenge to materials scientists. Continued experimentation and research is needed to develop a truly attractive ICF power plant.

A breeding blanket design for the tritium production ICF reactor SIRIUS-T is presented. The blanket consists of alternating layers of beryllium multiplier and lithium breeder/coolant with vanadium alloy structure. A tritium breeding ratio of 1.9 can be achieved for the optimum layered configuration with lithium enriched to 90% 6Li. An optimum coolant routing scheme is utilized to achieve adequate thermal hydraulics performance. The maximum structural material temperature is 650°C.

SIRIUS-T is a study of an advanced tritium production facility which utilizes direct drive symmetric illumination inertial confinement fusion provided by a KrF laser. Symmetrically illuminated reac...

The US blanket design activity has focused on the developments and the analyses of a solid breeder blanket concept for ITER. The main function of this blanket is to produce the necessary tritium required for the ITER operation and the test program. Safety, power reactor relevance, low tritium inventory, and design flexibility are the main reasons for the blanket selection. The blanket is designed to operate satisfactorily in the physics and the technology phases of ITER without the need for hardware changes. Mechanical simplicity, predictability, performance, minimum cost, and minimum R D requirements are the other criteria used to guide the design process. The design aspects of the blanket are summarized in this paper. 2 refs., 7 figs., 3 tabs.

Aqueous lithium salt blankets have some very attractive features which make them likely candidates for use in the next generation of fusion devices such as ITER and beyond. The main attraction derives from the fact that such a blanket can go from a non-tritium-breeding state to a tritium breeding state by the simple dissolution of a lithium salt into the cooling water. Such a blanket is also relatively simple to design and analyze, and is amenable to accommodating penetrations. On the other hand, the salt solution raises concern with respect to radiolysis, electrolysis and corrosion, as well as safety issues related to them. This paper describes the design of the US entry of an aqueous salt blanket for ITER.

A solid wall blanket utilizing advanced ferritic steel and re-circulating Flibe has been scoped out under the Advanced Power Extraction (APEX) program. This paper describes the engineering aspects of the design. The new alloy is the nano-composited ferritic (NCF) steel 12YWT and the breeding/cooling material is Flibe (Li/sub 2/BeF/sub 4/). The blanket has a solid first wall, but has an innovative coolant scheme, which allows part of the cooling fluid to be re-circulated to enhance the outlet temperature and thus improve the power cycle efficiency. Another innovation is the use of Pb as a neutron multiplier. The coolant outlet temperature is 681/spl deg/C allowing a conversion efficiency of 47%. Material issues, structural analysis, fabrication aspects and coolant circuits are covered.

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.

An aqueous lithium salt blanket is designed as a candidate for the tritium production blanket of ITER (International Thermonuclear Experimental Reactor). There are two rows of flat tube aqueous salt flow channels placed between beryllium blocks. The inboard and outboard blanket configurations have been optimized to maximize tritium breeding. A local outboard TBR (tritium breeding ratio) of 1.44 is obtained using LiOH at a concentration of 5 g/100 cm/sup 3/. The overall TBR determined from 3-D neutronics calculations is 0.88. The aqueous-salt flow parameters have been optimized to minimize the salt pressure and inventory, leading to a pressure of 1.15 MPa and tritium inventory of 104 g. There are no crevices in the aqueous-salt channels, and a redundancy has been provided against a salt-to-water leak potential. The first wall, blanket, and shield qualify for shallow land burial. The highest temperature in the blanket after a two-week period without coolant is only 440 degrees C. The first wall and blanket are capable of withstanding the worst disruption load. >

A two-dimensional thermal transient analysis, r-/spl theta/ model, has been performed to study three cases of total LOCA in the inboard components for the ITER CDA design using realistic assumptions. The study shows that the first wall can withstand the surface heat flux without active cooling for an extended period of time longer than the 10 seconds normally assumed for the plasma to continue operation after the onset of LOCA. No melting at the first wall or at the blanket is expected, but the lead at the back of the vacuum vessel reaches the melting point within about two days after the onset of LOCA without active cooling. At this time, the front of the vacuum vessel will reach about 600/spl deg/C. The temperature of the vacuum vessel will rise above 800/spl deg/C in a week.

LIBRA-LiTE is a conceptual design for a 1300-MWe power plant using light ion inertial fusion. LIBRA-LiTE differs from the LIBRA design in its use of ballistically focused light ions to drive the target. Focusing magnets are positioned 2.05 m from the target, which, to mitigate neutron damage effects, has required a novel magnet design using liquid lithium as a conductor. A sacrificial film of liquid lithium protects the magnets, the target chamber side walls and bottom from the X-rays and debris released by the target microexplosion. The target neutrons deposit in a tritrium breeding blanket of liquid lithium confined to woven metal tubes on the sides and in a pool on the bottom. The top of the target chamber is a metallic dome removed far enough (16 m) from the target to be a lifetime component. >

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

Two 1000-MW/sub e/ power plants-one driven by a KrF laser and the other by a heavy ion beam induction accelerator-are being designed. The goal is to make both designs as attractive as possible in terms of technical credibility, environmental impact, safety, cost, and operability. For the laser driver, a gas-protected dry-wall chamber with uniformly illuminated direct-drive targets has been selected. The chamber is made of a carbon/carbon composite, and the breeding blanket is a flowing bed of Li/sub 2/O ceramic granules. The breeding material flows through the chamber by gravity and is moved through heat exchangers using fluidized transport. The HIB-driven plant design uses a wetted-wall chamber with indirect-drive targets. This reactor features a molten salt (2LiF-BeF/sub 2/) breeder that is guided through the chamber in porous woven-carbon-fabric flow channels. Progress in the conceptual design of the KrF- and HIB-driven power plants is reviewed. The design features, advantages, and issues for the two reactors and drivers are highlighted. >

The Apollo-L3 conceptual reactor design utilizes D-/sup 3/He fuel in a first stability tokamak configuration with direct conversion of synchrotron radiation to electricity as well as thermal conversion of surface heat and neutron energy. An update of the parameters of the design is given, and analyses of certain critical issues of Apollo-L3 are reported. These include plasma startup, loss of fast ions by magnetic ripple, safety analyses, and tritium storage. The update has led to an increase in the major radius by about 40 cm and an increase in the cost of electricity by about 10%. An alternative to direct energy conversion of the synchrotron radiation is to convert the energy using a thermal cycle; this increases the cost by another 10%. Startup of the plasma using a small amount of tritium eliminates the need for auxiliary heating other than that provided by the current drive system. Analysis of accident scenarios led to a possible offsite early dose of 3.6 rem for a LOCA without burning of the organic coolant and 103 rem if the organic coolant burns. On the basis of the expected temperature of the shield in a LOCA accident and the radioactive inventory, Apollo-L3 qualifies for a level of safety rating of one. >

The impact of the HARD (high aspect ratio design) option on the blanket and shield performance as concluded from CDA (conceptual design activity) is examined. The neutron wall load distributions were calculated for the different operating modes of HARD. The average neutron wall load over the test ports is 30 to 70% higher than CDA, which increases the testing capability of ITER. The net tritium breeding capability of HARD was calculated based on the use of CDA blanket design. The net tritium breeding ratio is 8% less than CDA. However, the increased average neutron wall load over the test sections overrides the impact on the external tritium supply because of shorter operating time for the same neutron fluence. The shielding analysis of HARD suggested several modifications to meet the design limits of CDA. The total inboard thickness at midplane should be restored to 84 cm with the vacuum gap between the shield and the vacuum vessel reduced to 2 cm as in CDA, or 86 cm inboard thickness with the current gap thickness of 4 cm. In addition, it is necessary to extend the upper parts of the side modules to provide extra shielding for the upper divertor region. >

SOMBRERO, a conceptual design study of a 1000-MWe KrF laser driven IFE (inertial fusion energy) power reactor utilizing direct drive targets with near symmetric illumination, is described. The chamber is constructed of a low-activation carbon/carbon composite and the blanket consists of a moving bed of solid Li/sub 2/O particles flowing through the chamber by gravity. The particles are transported through the intermediate heat exchanger and around the loop in a fluidized state by He gas at 0.2 MPa. There are 60 laser beams in near-symmetric distribution; the laser energy is 3.4 MJ; the target gain is 118; and the rep-rate is 6.7 Hz. The chamber has a cylindrical central region with conical ends, a minimum radius of 6.5 m, and a blanket/reflector thickness of 1 m at the midplane giving a tritium breeding ratio of 1.25 and an overall energy multiplication of 1.08. A grazing incidence metallic mirror is located 30 m from the target, and the final focusing dielectric coated mirror which is shielded from direct neutrons is located at 50 m. The inlet temperature of the Li/sub 2/O particles is 550 degrees C, the equilibrated outlet temperature is 740 degrees C; and the power cycle efficiency is 47%. Using a laser efficiency of 7%, the driver power is 325 MWe. The gross power output is 1360 MWe, giving a net power output of 1000 MWe, with 360 MWe used for the driver and auxiliary equipment. >

The ARIES-III conceptual reactor design utilizes D-/sup 3/He fuel in a second stability tokamak configuration. The process of starting up the plasma and bringing it to the desired operating point was analyzed using a time-dependent particle and power balance code in conjunction with MHD (magnetohydrodynamic) equilibrium and stability, and neutral beam current drive calculations. Two different startup scenarios, one using D-T to assist reaching the D-/sup 3/He operating point and one using only pure D-/sup 3/He, were analyzed. In order to minimize the amount of ICRF (ion cyclotron range of frequencies) auxiliary heating required, one must program the plasma current so that the plasma enters second stability as soon as practical; it was assumed that confinement improves greatly when the plasma is in second stability. Seeding the plasma with tritium during startup reduces the amount of ICRF power required, but leads to a 14-MeV neutron pulse. The blanket and magnet heating due to this neutron pulse are within acceptable limits. The DT assisted startup was chosen as the reference procedure. Pure D-/sup 3/He startup of ARIES-III can be achieved with 200 MW of heating power in addition to that available from the current drive system. >