MHD thermohydraulics analysis and supporting R&amp;D for DCLL blanket in the FNSF

Smolentsev, Sergey; Rhodes, Tyler; Pulugundla, Gautam; Courtessole, Cyril; Abdou, Mohamed; Malang, Siegfried; Tillack, Mark; Kessel, Chuck

doi:10.1016/j.fusengdes.2017.06.017

MHD thermohydraulics analysis and supporting R&D for DCLL blanket in the FNSF

Journal Article · Fri Jun 16 04:00:00 EDT 2017 · Fusion Engineering and Design

DOI:https://doi.org/10.1016/j.fusengdes.2017.06.017· OSTI ID:1610064

Smolentsev, Sergey ^[1]; Rhodes, Tyler ^[2]; Pulugundla, Gautam ^[2]; Courtessole, Cyril ^[2]; Abdou, Mohamed ^[2]; Malang, Siegfried ^[3]; Tillack, Mark ^[4]; Kessel, Chuck ^[5]

Univ. of California, Los Angeles, CA (United States); DOE/OSTI
Univ. of California, Los Angeles, CA (United States)
Independent consultant (Germany)
Univ. of California, San Diego, CA (United States)
Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States)

A Fusion Nuclear Science Facility (FNSF) has been recognized in the fusion community as a necessary facility to resolve the critical technology issues of in-vessel components prior to the construction of a DEMO reactor (Abdou et al., 1996) [1]. Among these components, development of a reliable, low-cost and safe blanket system that provides self-sufficient tritium breeding and efficient conversion of the extracted fusion energy to electricity, while meeting all material, design and configuration limitations is among the most important but still challenging goals. In the recent FNSF study in the US (Kesel et al., 2015) [2], a Dual-Coolant Lead-Lithium (DCLL) blanket has been selected as the main breeding blanket concept. This paper summarizes the most important details of the proposed DCLL blanket design, presents the MHD thermohydraulic analysis for the PbLi flows in the blanket conduits and introduces supporting R&D studies, which are presently ongoing at UCLA. We also discuss the required pre-FNSF R&D in the area of MHD Thermofluids to support the further work on the DCLL blanket design & analysis and its integration into the fusion facility.

View Accepted Manuscript (DOE)

Research Organization:: Univ. of California, Los Angeles, CA (United States)

Sponsoring Organization:: USDOE Office of Science (SC), Fusion Energy Sciences (FES)

Grant/Contract Number:: FG02-86ER52123

OSTI ID:: 1610064

Alternate ID(s):: OSTI ID: 22850338

Journal Information:: Fusion Engineering and Design, Journal Name: Fusion Engineering and Design Journal Issue: B Vol. 135; ISSN 0920-3796

Publisher:: ElsevierCopyright Statement

Country of Publication:: United States

Language:: English

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Fusion Blankets and Fluoride-Salt-Cooled High-Temperature Reactors with Flibe Salt Coolant: Common Challenges, Tritium Control, and Opportunities for Synergistic Development Strategies Between Fission, Fusion, and Solar Salt Technologies Forsberg, Charles; Zheng, Guiqiu (Tony); Ballinger, Ronald G. Nuclear Technology https://doi.org/10.1080/00295450.2019.1691400	journal	December 2019
Integrated Liquid Metal Flowing First Wall and Open-Surface Divertor for Fusion Nuclear Science Facility: Concept, Design, and Analysis Smolentsev, Sergey; Rognlien, Thomas; Tillack, Mark Fusion Science and Technology, Vol. 75, Issue 8 https://doi.org/10.1080/15361055.2019.1610649	journal	May 2019
Alternative Plasma-Facing-Material Concepts for Extreme Plasma-Burning Nuclear Fusion Environments Allain, J. P. Fusion Science and Technology, Vol. 75, Issue 7 https://doi.org/10.1080/15361055.2019.1647030	journal	October 2019

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MHD thermohydraulics analysis and supporting R&D for DCLL blanket in the FNSF

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