No abstract available

[en] It is pointed out that plasma parameters for a fusion reactor have been fairly accurately defined for many years, and the real plasma physics objective must be to find the means of achieving and maintaining these specifiable parameters. There is good understanding of the generic technological problems: breading blankets and shields, radiation damage, heat transfer and methods of magnet design. The required plasma parameters for fusion self-heated reactors are established at ntausub(E) approximately 2.10¹⁴ cm^-3sec, plasma radius 1.5 to 3 m, wall loading 5 to 10 MW cm^-2, temperature 15 keV. Within this model plasma control by quasi-steady burn as a key problem is studied. It is emphasized that the future programme must interact more closely with engineering studies and should concentrate upon research which is relevant to reactor plasmas. (V.P.)

[en] This paper covers the results of a scoping study to assess the possible application of magnetic diversion to a laser inertial fusion reactor. The purpose of magnetic diversion is to steer the ions away from the chamber wall. Its impact on the engineering design and performance of the reactor is discussed and key issues are identified. Magnetic diversion allows more robust choices for the chamber armor as the ions can be directed to external collector plates. However, issues associated with the design and size of the plates as well as the magnetic field impact on the design complexity and choice of coolant must be addressed. Magnetic diversion also opens up the attractive possibility of trying to convert the ion energy to electricity with much better efficiency than that obtained by conventionally transferring the ion energy to a power cycle fluid through a heat exchanger. Since the recycling power to the laser represents a high fraction of the electrical output from a conventional power plant (up to about 25%), this would help in improving the overall plant efficiency appreciably

[en] Four conceptual designs of hybrid inertial fusion systems based on the wetted or magnetically protected first-wall protection schemes are compared. Each first-wall protection scheme was developed with a high- or low-power-density blanket to evaluate power and fissile fuel production levels. The comparison shows that a magnetically protected-wall reactor with a low-power-density blanket provides maximum annual fissile fuel generation with an attractive fuel cycle time

[en] Fusion Next Step Device (FER) plans to experiment Deutrium-Tritium (D-T) reaction, remote handling and other fusion engineering issues. The fast neutron of 14 MeV caused by D-T reaction does not only activate the structural components inside the vacuum vessel, but also damages some first walls. The technique to remove the armour tiles of first walls by simple and quick operation is a key technology for the D-T burning Next Step Device. To establish the rational remote tile handling technology, consideration of consistency between the reactor structure and remote equipments should be made. The report comprises mainly the joint structures of armour tiles, design conditions (electro-magnetic force, cooling systems and so forth) and remote equipments. In addition, it is referred in shape memory alloy (SMA) applications, transportation of damaged tiles from the vacuum vessel and inspection systems for the first wall integrity. Hereafter, furthermore study in depth for the tile handling must be made in parallel with verification of remote systems and tile attachment structures using partial mockups. (author)

[en] This document provides a systems description of the Reference Design for The Next Step (TNS) evolved at Oak Ridge National Laboratory (ORNL) during FY 1978. The description is presented on the basis of 24 individual device and facility systems. Additional information on these systems, the Reference Design, and the FY 1978 Oak Ridge TNS activities can be found in the associated technical memoranda, ORNL/TM-6720 and ORNL/TM-6722--ORNL/TM-6733

[en] Pulse Star is a pool-type ICF reactor that emphasizes low cost and high safety levels. The reactor consists of a vacuum chamber (belljar) submerged in a compact liquid metal (Li₁₇Pb₈₃ or lithium) pool which also contains the heat exchangers and liquid metal pumps. The shielding efficiency of the liquid metal pool is high enough to allow hands-on maintenance of (removed) pumps and heat exchangers. Liquid metal is allowed to spray through the 5.5 m radius belljar at a controlled rate, but is prohibited from the target region by a 4 m radius mesh first wall. The wetted first wall absorbs the fusion x-rays and debris while the spray region absorbs the fusion neutrons. The mesh allows vaporized liquid metal to blow through to the spray region where it can quickly cool and condense. Preliminary calculations show that a 2 m thick first wall could handle the mechanical (support, buckling, and x-ray-induced hoop) loads. Wetting and gas flow issues are in an initial investigation stage

[en] UTLIF(2) is a conceptual design study on a light ion beam driven fusion reactor based on a concept of rod-bundle blanket. Survivability and maintainability of the first wall and the blanket are regarded as of major importance in the design. The blanket rod is composed of a thick tube which has enough stiffness, a thin wrapping wall which receives high heat flux, and liquid lithium which breeds tritium and removes generated heat. The rod can be pulled out from the outside of the reactor vessel, hence the replacement is very easy. Nuclear and thermal analysis have been made and the performance of the reactor has been shown to be satisfactory. (author)

[en] This report presents the thermal-hydraulic analysis of three ex-vessel Loss-of-Coolant Accidents (LOCAs) in the first wall/blanket cooling system of the alternative SEAFP reactor design. The LOCAs are caused by a rupture of the pump suction pipe, an inlet header, and an outlet header respectively. The ex-vessel LOCAs considered result from a rupture of a cooling pipe located outside the plasma vessel. In order to determine the worst case LOCA conditions, no plasma shutdown and no other counteractions have been assumed. The analyses have been performed using the thermal-hydraulic system analysis code RELAP5/MOD3. Special attention has been paid to the transient thermal-hydraulic behaviour of the cooling system and the temperature development in the first wall and blanket. For the LOCAs considered, the temperature development in the first wall is more critical than the temperature development in the blanket. In the LOCA caused by a rupture of the pump suction pipe, melting at the midplane of the outboard first wall starts about 56 s after break initiation. In the LOCA caused by a rupture of an inlet header, melting at the midplane of the outboard first wall starts about 62 s after break initiation, whereas in the LOCA caused by a rupture of an outlet header melting at the midplane of the outboard first wall starts about 74 s after break initiation. (orig.)

No abstract available