An Overview of the US DCLL ITER TBM Program

Journal Article · Fri Jan 01 04:00:00 EST 2010 · Fusion Engineering and Design

DOI:https://doi.org/10.1016/j.fusengdes.2010.02.021· OSTI ID:1024707

Wong, C P. C. ^[1]; Abdou, M ^[2]; Dagher, M ^[2]; Katoh, Yutai ^[3]; Kurtz, Richard ^[4]; Malang, Siegfried ^[5]; Marriott, Edward ^[6]; Merrill, Brad ^[7]; Messadek, K ^[2]; Morley, N B ^[2]; Pint, Bruce A ^[3]; Sawan, M ^[8]; Sharafat, S ^[2]; Smolentsev, Sergey ^[2]; Sze, Dai-Kai ^[9]; Willms, Scott ^[10]; Ying, Alice ^[2]; Youssef, M Z ^[2]

---

1. General Atomics, San Diego
2. University of California, Los Angeles
3. ORNL
4. Pacific Northwest National Laboratory (PNNL)
5. Consultant
6. University of Wisconsin, Madison
7. Idaho National Laboratory (INL)
8. University of Wisconsin
9. University of California, San Diego
10. Los Alamos National Laboratory (LANL)

Under the US Fusion Nuclear Science and Technology program, we selected the Dual Coolant Lead Lithium (DCLL) concept as our primary Test Blanket Module (TBM) for testing in ITER. The DCLL blanket concept has the potential to be a high-performance DEMO blanket design with a projected thermal efficiency of >40%. Reduced activation ferritic/martensitic (RAF/M) steel is used as the structural material. Helium is used to cool the first wall and blanket structure, and the self-cooled Pb-17Li breeder is circulated for power conversion and for tritium extraction. A SiC-based flow channel insert (FCI) is used as an electrical insulator for magnetohydrodynamic pressure drop reduction from the circulating Pb-17Li and as a thermal insulator to separate the high-temperature Pb-17Li ({approx}650-700 C) from the RAF/M structure, which has a corrosion temperature limit of {approx}480 C. The RAF/M material must also operate at temperatures above 350 C but less than 550 C. We are continuing the development of the mechanical design and performing neutronics, structural and thermal hydraulics analyses of the DCLL TBM module. Prototypical FCI structures were fabricated and further attention was paid to MHD effects and the design of the inboard blanket for DEMO. We are also making progress on related R&D needs to address key areas. This paper is a summary report on the progress and results of recent DCLL TBM development activities.

• Research Organization: Oak Ridge National Laboratory (ORNL)
• Sponsoring Organization: SC USDOE - Office of Science (SC)
• DOE Contract Number: AC05-00OR22725
• OSTI ID: 1024707
• Journal Information: Fusion Engineering and Design, Journal Name: Fusion Engineering and Design Journal Issue: 7 Vol. 85; ISSN 0920-3796
• Country of Publication: United States
• Language: English

Related Subjects

36 MATERIALS SCIENCE
70 PLASMA PHYSICS AND FUSION TECHNOLOGY
BUILDING MATERIALS
COOLANTS
CORROSION
DESIGN
ELECTRICAL INSULATORS
FIRST WALL
HELIUM I
LITHIUM
MAGNETOHYDRODYNAMICS
PRESSURE DROP
STEELS
TESTING
THERMAL EFFICIENCY
THERMAL HYDRAULICS
TRITIUM