4 research outputs found
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One- and two-dimensional radiation analysis of the Compact Ignition Tokamak
The Compact Ignition Tokamak (CIT) is being proposed as the next major fusion experiment to follow the operations of the Tokamak Fusion Test Reactor (TFTR). CIT is a compact deuterium-tritium-burning tokamak device that is designed to achieve ignition. The high neutron wall loadings, 7 to 9 MW/m/sup 2/, associated with the operation of this device require that neutronics-related issues be considered in the overall system design. Radiation shielding is required for the protection of both device components and personnel. A close-in igloo shield, 1.8 m nominal thickness, has been designed around the periphery of the tokamak structure to permit personnel access into the test cell after shutdown and to limit the total activation of the test cell components. This paper discusses the major neutronics issues related to the design of the Compact Ignition Experiment, describes the methodologies used to quantify these concerns, and presents the results of radiation transport and activation scoping studies
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Activation analysis of the compact ignition tokamak
The US fusion program has completed the conceptual design of a compact tokamak device that achieves ignition. The high neutron wall loadings associated with this compact deuterium-tritium-burning device indicate that radiation-related issues may be significant considerations in the overall system design. Sufficient shielding will be requied for the radiation protection of both reactor components and occupational personnel. A close-in igloo shield has been designed around the periphery of the tokamak structure to permit personnel access into the test cell after shutdown and limit the total activation of the test cell components. This paper describes the conceptual design of the igloo shield system and discusses the major neutronic concerns related to the design of the Compact Ignition Tokamak
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MCNP capabilities at the dawn of the 21st century: Neutron-gamma applications
The Los Alamos National Laboratory Monte Carlo N-Particle radiation transport code, MCNP, has become an international standard for a wide spectrum of neutron-gamma radiation transport applications. These include nuclear criticality safety, radiation shielding, nuclear safeguards, nuclear well-logging, fission and fusion reactor design, accelerator target design, detector design and analysis, health physics, medical radiation therapy and imaging, radiography, decontamination and decommissioning, and waste storage and disposal. The latest version of the code, MCNP4C, was released to the Radiation Safety Information Computational Center (RSICC) in February 2000.This paper described the new features and capabilities of the code, and discusses the specific applicability to neutron-gamma problems. We will also discuss the future directions for MCNP code development, including rewriting the code in Fortran 90
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Simple contour analysis of ignition conditions and plasma operating regimes in tokamaks
Contour plots of ignition, auxiliary power requirements, heating and operating windows, optimal path to ignition, ignition margin, etc., are generated analytically in terms of a small number of parameters (aB/sub 0//sup 2//q/sub */, R/sub 0//B/sub 0/, , etc.) for classes of devices with equivalent performance. Numerical studies are carried out to map the physics design space. Considering both the Murakami density limit (approx.B/sub 0//R/sub 0/) and the Troyon beta limit (approx.I/aB/sub 0/), results from analytic calculations indicate that in a standard tokamak geometry (A approx. 2.5 to 3.5, kappa = b/a approx. 1.6 to 1.7, q/sub psi/ approx. 2.6) devices with aB/sub 0//sup 2//q/sub */ approx. 20 should be ignitable provided confinement does not degrade with heating (ohmic + alpha + auxiliary, etc.) power; however, aB/sub 0//sup 2//q/sub */ approx. 30 (25) may be required for minimal ignition for a typical L- (H-) mode confinement scaling. Increased plasma elongation (kappa approx. 2) may help to reduce these requirements