APEX advanced ferritic steel, Flibe self-cooled first wall and blanket design

Abstract

APEX Advanced Ferritic Steel, Flibe Self-Cooled First Wall and Blanket Design As an element in the U.S. 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.

Safety assessment of two advanced ferritic steel molten salt blanket design concepts

“…Molten salt thermal properties and heat transfer coefficients for these equations were taken from Refs. [2,3]. The pre-exponential constant and diffusion activation energy for tritium in the molten salt were taken from Ref.…”

“…Two blanket concepts are currently under consideration by the APEX study that employ a molten salt both as a breeding and coolant material, and an AFS as a structural material [2,3]. The first design concept has a solid first wall (FW), and is designated in this article as the solid FW design.…”

“…Conservative weather conditions are Pasquill-Gifford stability class F atmospheric conditions for a wind velocity of 1 m/s 2. The specific dose for a compound (molten salt or AFS oxide) is the isotope-by-isotope summation of the dose associated with radioactive elements within the compound.…”

Preliminary Design of a Helium-Cooled Ceramic Breeder Blanket for CFETR Based on the BIT Concept

“…The tritium breeding blanket is one of the key invacuum vessel components for fusion reactors. It is required to perform tritium breeding to fuel DT plasma operation based on both 6 Li(n, α)T and 7 Li(n, α) reactions between the neutrons from the plasma chamber and the lithium in the blanket; to remove the neutron heat deposited in the blanket to demonstrate fusion power generation at higher coolant outlet temperatures; and to provide sufficient radiation protection as one part of the in-vessel shielding system, which is comprised of the tritium breeding blanket, the shielding blanket and the vacuum vessel.…”

“…FLiBe) [6] . Usually, the lithium component needs to be 6 Li-enriched to improve the TBR. In addition, a neutron multiplier (e.g.…”