Fusion Reactor Blanket Designs Research Articles

Prospects for Low-Activity Aluminum Structures

High-purity, low-activity powder metallurgy aluminum alloys can be developed for use in a fusion reactor at 300 to 400/sup 0/C using helium as a heat transfer medium. Hot water as a coolant may limit aluminum to 200/sup 0/C. From a heat transfer point of view, based on the dual- or two-temperature design approach, commercial fusion reactor blanket designs appear to be feasible. To meet all of the blanket design requirements feasibility requires quantification of thermal hydraulics, materials, neutronics, and material responses. Also, radiation damage and lifetime analyses are key issues for design qualification. Based on tests performed to date, aluminum appears well suited for experimental fusion reactors operating at wall temperatures below 200/sup 0/C.

Economic Design Optimization of the LiPb Blanket for the Mirror Advanced Reactor (MARS)

A self-consistent procedure has been established for economic design optimization of the lithiumlead (LiPb) blanket for the MARS tandem mirror reactor. The procedure is necessarily iterative and enables progress in blanket design to be assessed in terms of the minimization of an economic figure of merit F for the complete reactor system. Typical economic design questions regarding blanket and central cell parameters such as tritium breeding ratio, neutron energy multiplication factor, thermal cycle efficiency, blanket radial thickness, magnet radii, etc., can then be addressed in terms of their influence on overall system costs. This procedure is not necessarily specific to MARS and has general applicability to fusion reactor blanket design optimization. Application of the procedure resulted in a blanket with small ( about 38 cm) radial thickness, highly enriched (90%) lithium, adequate tritium breeding ratio (1.14) and a neutron energy multiplication and thermal efficiency approaching those for blankets of considerably larger radial dimensions.

Neutron Multiplication Studies in Beryllium for Fusion Reactor Blankets

Beryllium is used as an efficient neutron multiplier in several fusion reactor blanket designs. In the framework of the experimental research program on the neutronics of fusion reactor blanket designs established at the Institut für Reaktorentwicklung der Kernforschungsanlage Jülich GmbH, measurements of the neutron multiplication in beryllium produced by 14-MeV neutrons were carried out to check basic nuclear data. The measurements were made in rectangular geometry as a function of beryllium thicknesses of up to 20 cm. The experimental values of the neutron multiplication were found to be 25% lower than the calculated values for all thicknesses. The low value of the multiplication casts doubts as to the suitability of beryllium as a neutron multiplier in fusion reactor blankets to yield useful tritium breeding ratios.

Measurements of the Neutron Spectra from Materials Used in Fusion Reactors and Calculations Using the ENDF/B-III and -IV Neutron Libraries

Proposed fusion reactor blanket designs bring into focus a large number of problems dealing with the interaction of 14-MeV neutrons with different materials. Carbon, oxygen, aluminum, titanium, and iron are among the materials used in the blanket. To have confidence in fusion reactor blanket calculations, a necessary prerequisite is that the transport code correctly describes the interaction of 14-MeV neutrons with the materials of the blanket. Spherical assemblies of the above materials ranging from 1 to 5 mean-free-paths in thickness have been bombarded with a centered nominal 14-MeV neutron source. The emitted neutron energy spectra were measured using time-of-flight techniques (3-nsec full-width-at-half-maximum system resolution) in a geometry where the flight path (7 to 10 m) is long compared to the dimensions of the spherical targets. The spectra have been calculated with the Monte Carlo neutron transport code TART using the ENDF/B-III and -IV neutron libraries and compared with measurements.

Comparative Study of Radioactivity and Afterheat in Several Fusion Reactor Blanket Designs

Comparative Study of Radioactivity and Afterheat in Several Fusion Reactor Blanket Designs

New Concepts for Controlled Fusion Reactor Blanket Design

New Concepts for Controlled Fusion Reactor Blanket Design

A Comparative Study of Several Fusion Reactor Blanket Designs

A study of the nuclear performance of several recently reported fusion reactor blanket designs is presented. In particular, the nuclear heating, the tritium breeding ratio, and the charged-particle production rates in the various systems are reported. It is found that the total nuclear heating can be overestimated by as much as 30%, that ∼20 MeV per fusion is a typical value for the energy production capability of most blankets, and that 22.4 MeV per fusion is a more maximum than nominal value for blankets without fissile materials. The tritium breeding ratio in lithium blankets is high, and uncertainties in nuclear data are unlikely to prevent such systems from breeding. Flibe blankets are marginal in this regard, and uncertainties can prevent breeding in these systems. Hydrogen and helium production rates are fairly large in all systems; they are highest in sintered aluminum product and in the PE-16 alloy, and lowest in niobium, with stainless steel in between. However, much of the required nuclear data on charged-particle-producing reactions is unavailable, and the need for cross-section measurements in this area is discussed.

The 93Nb( n, 2n) 92Nb cross-section at 14 MeV extracted from experimental neutron emission spectrum

Informations on the 93Nb( n, 2tn) 92Nb reaction at 14·6 MeV have been obtained from measurements of neutron emission spectra and its analysis taking into account the compound nucleus emission and emission during the precompound phase of the reaction. The result is a value of ≈ 1200 mb for σ n, 2 n confirming recent analysis that the activation cross-section of ≈ 450 mb should be considerably increased with important consequences for design of fusion reactor blankets, where Nb is a preferred material.