K.R. Schultz

Overview of the Blanket Comparison and Selection Study

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

The ARIES-I Tokamak Reactor Study †

The ARIES research program is a multi-institutional effort to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. Three ARIES visions are currently planned for the ARIES program. The ARIES-I design is a DT-burning reactor based on modest extrapolation from the present tokamak physics data base; ARIES-II is a DT-burning reactor which will employ potential advances in physics; and ARIES-III is a conceptual D-3He reactor. The first design to be completed is ARIES-I, a 1000 MWe power reactor. The key features of ARIES-I are: (1) a passively safe and low environmental impact design because of choice of low activation material throughout the fusion power core, (2) an acceptable cost of electricity, (3) a plasma with performance as close as possible to present-day experimental achievements, (4) a high performance, low activation, SiC composite blanket cooled by He, and (5) an advanced Rankine power cycle as planned for near term coal-fired plants. The ARIES-I research has also identified key physics and technology areas with the highest leverage for achieving attractive fusion power system.

Status of inertial fusion target fabrication in the USA

This paper summarizes the current techniques used in the USA for fabrication of targets for inertial confinement fusion (ICF) experiments at the five ICF laboratories in the USA. It reviews the current target specifications that can be achieved and discusses directions for development of targets for ignition in the National Ignition Facility.

The ARIES-III D-3He tokamak-reactor study

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.<<ETX>>

SYNTHESIS OF HYDROCARBON FUELS USING RENEWABLE AND NUCLEAR ENERGY

Abstract In light of the current issues of carbon control and the desire to become less dependent on imported oil, we propose to apply non-carbon-based energy supplies (renewables and nuclear) to reduction of CO2 emissions and production of liquid synthetic fuels. To this end we have performed technical and economic analyses of systems ranging from hydrogen augmentation of coal-to-liquids processes, through the use of coal power plant CO2, to the extraction of atmospheric CO2 for the production of synthetic fuels. This paper emphasizes the use of nuclear power to provide the hydrogen and energy needed for utilization of coal power plant CO2 and points toward the closure of the carbon cycle by the ultimate use of atmospheric CO2.

Overview of the TITAN-I fusion-power core

The TITAN reactor is a compact (major radius of 3.9 m and plasma minor radius of 0.6 m), high neutron wall loading (~18 MW/m 2 ) fusion energy system based on the reversed-field pinch (RFP) confinement concept. The reactor thermal power is 2918 MWt resulting in net electric output of 960 MWe and a mass power density of 700 kWe/tonne. The TITAN-I fusion power core (FPC) is a lithium, self-cooled design with vanadium alloy (V-3Ti-1Si) structural material. The surface heat flux incident on the first wall is ~4.5 MW/m 2 . The magnetic field topology of the RFP is favorable for liquid metal cooling. In the TITAN-I design, 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. With the IBC concept the poloidal flow lithium circuit is also the electrical conductor of the toroidal-field and divertor coils. Three dimensional neutronics analysis yields a tritium breeding ratio of 1.18 and a molten salt extraction technique is employed for the tritium extraction system. Almost every FPC component would qualify for Class C waste disposal. The compactness of the design allows the use of single-piece maintenance of the FPC. This maintenance procedure is expected to increase the plant availability. The entire FPC operates inside a vacuum tank, which is surrounded by an atmosphere of inert argon gas to impede the flow of air in the system in case of an accident. The top-side coolant supply and return virtually eliminate the possibility of a complete LOCA occurring in the FPC. The peak temperature during a LOFA is 991 °C.

Contributions of the National Ignition Facility to the development of inertial fusion energy

The Department of Energy is proposing to construct the National Ignition Facility (NIF) to embark on a program to achieve ignition and modest gain in the laboratory early in the next century. The NM will use a {ge}1.8-MJ, 0.35-mm laser with 192 independent beams, a fifty-fold increase over the energy of the Nova laser. System performance analyses suggest yields as great as 20 MJ may be achievable. NIF will conduct more than 600 shots per year. The benefits of a micro-fusion capability in the laboratory include: Essential contributions to defense programs, resolution of important Inertial Fusion Energy issues, and unparalleled conditions of energy density for basic science and technology research. We have begun to consider the role the National Ignition Facility will fill in the development of Inertial Fusion Energy. While the achievement of ignition and gain speaks for itself in terms of its impact on developing IFE, we believe there are areas of IFE development, such as fusion power technology, IFE target design and fabrication, and understanding chamber dynamics, that would significantly benefit from NIF experiments. In the area of IFE target physics, ion targets will be designed using the NIF laser, and feasibility of high gain targets will be confirmed. Target chamber dynamics experiments will benefit from x-ray and debris energies that mimic in-IFE-chamber conditions. Fusion power technology will benefit from using single-shot neutron yields to measure spatial distribution of neutron heating, activation, and tritium breeding in relevant materials. IFE target systems will benefit from evaluating low-cost target fabrication techniques by testing such targets on NIF.

The Fusion Applications Study — “FAME”

Fusion has a wide spectrum of applications that appear technically possible and may become economically feasible. Near-term (approx. 2000) application for production of nuclear fuels and useful radioisotopes is an economically attractive possibility as soon as fusion is ready. Electricity production will remain a prime, large-scale application of fusion. In the longer term, as fossil fuels dwindle, production of hydrogen could become a major application. Additional applications some of which have not even been conceived of yet, will add to this potential richness and diversity of fusion. It is the purpose of the fusion applications study - FMAE - to innovate, investigate, and evaluate these potential applications.

Activation evaluation of fusion solid breeder materials

Solid lithium ceramic compounds are considered prime candidate breeder materials for D-T-based fusion reactors because of the potential advantages in safety, lack of electromagnetic effects, and relative ease of tritium extraction. Li/sub 2/O, Li/sub 4/SiO/sub 4/, LiAlO/sub 2/, and Li/sub 2/ZrO/sub 3/ are among the most plausible candidate solid breeder materials. These materials were studied to compare their activation characteristics after neutron exposure in a fusion reactor, including quantities relevant to safety and waste-disposal. Induced radioactivity, decay heat, biological dose rate, and the waste-disposal rating of these materials were considered. Impurity elements that appear in lithium such as Na, K, and Ca were also included in the analysis. The results show that Li/sub 2/ZrO/sub 3/ gives the highest values in every aspect of activation concerns except waste disposal, where LiAlO/sub 2/ dominates. LiAlO/sub 2/ and Li/sub 2/ZrO/sub 3/ will qualify for 10CFR61 low-level waste disposal only when these materials are located in the lower neutron fluence region in the blanket. Li/sub 2/O and Li/sub 4/SiO/sub 4/ show much lower (at least three orders of magnitude) induced activity than Li/sub 2/ZrO/sub 3/ and LiAlO/sub 2/ and satisfy all aspects of low activation considerations.<<ETX>>

Divertor and first wall armor design for the TIBER-II engineering test reactor

TIBER-II is a compact, high-power-density, steady-state current-drive engineering test reactor that uses a double-null divertor configuration, operating in the high recycle mode with low particle temperature. The nominal peak heat flux is 3.3 MW/m2, with off-normal condition peaks of up to 6.6 MW/m2. Plasma disruptions may cause peak energy deposition of up to 13 MJ/m2. The design uses water-cooled copper alloy tubes with brazed-on protective tiles. During early operating phases when disruptions may be frequent, these tiles will be made of carbon. During the nuclear testing phase, disruptions must be limited so that more erosion-resistant tungsten tiles may be used. The first walls of TIBER-II experience a heat flux of 0.23 MW/m2 with disruption energy density of 2.4 MJ/m2. They are protected by carbon-carbon composite armor tiles. The TIBER-II design requirements are challenging, and although the divertor and first wall armor designs successfully meet these requirements, significant issues must be resolved to verify their performance. These critical issues and the R&D required to resolve them are described.