An overview of dual coolant Pb-17Li breeder first wall and blanket concept development for the US ITER-TBM design

C. P.C. Wong; S. Malang; M. Sawan; M. Dagher; S. Smolentsev; B. Merrill; M. Youssef; S. Reyes; D. K. Sze; N. B. Morley; S. Sharafat; P. Calderoni; G. Sviatoslavsky; R. Kurtz; P. Fogarty; S. Zinkle; M. Abdou

doi:10.1016/j.fusengdes.2005.05.012

An overview of dual coolant Pb-17Li breeder first wall and blanket concept development for the US ITER-TBM design

C. P.C. Wong, S. Malang, M. Sawan, M. Dagher, S. Smolentsev, B. Merrill, M. Youssef, S. Reyes, D. K. Sze, N. B. Morley, S. Sharafat, P. Calderoni, G. Sviatoslavsky, R. Kurtz, P. Fogarty, S. Zinkle, M. Abdou

Research output: Contribution to journal › Conference article › peer-review

104 Scopus citations

Abstract

An attractive blanket concept for the fusion reactor is the dual coolant Pb-17Li liquid (DCLL) breeder design. Reduced activation ferritic steel (RAFS) is used as the structural material. Helium is used to cool the first wall and blanket structure, and the self-cooled breeder Pb-17Li is circulated for power conversion and for tritium breeding. A SiC_f/SiC composite insert is used as the magnetohydrodynamic (MHD) insulation to reduce the impact from the MHD pressure drop of the circulating Pb-17Li and as the thermal insulator to separate the high temperature Pb-17Li from the helium cooled RAFS structure. For the reference tokamak power reactor design, this blanket concept has the potential of satisfying the design limits of RAFS while allowing the feasibility of having a high Pb-17Li outlet temperature of 700 °C. We have identified critical issues for the concept, some of which include the first wall design, the assessment of MHD effects with the SiC-composite flow coolant insert, and the extraction and control of the bred tritium from the Pb-17Li breeder. R&D programs have been proposed to address these issues. At the same time we have proposed a test plan for the DCLL ITER-Test Blanket Module program.

Original language	English
Pages (from-to)	461-467
Number of pages	7
Journal	Fusion Engineering and Design
Volume	81 A
Issue number	1-4
DOIs	https://doi.org/10.1016/j.fusengdes.2005.05.012
State	Published - Feb 2006
Externally published	Yes
Event	Proceedings of the Seventh International Symposium on Fusion Nuclear Technology ISFNT-7 Part A - Duration: May 22 2005 → May 27 2005

Funding

This work was supported in part by the US Department of Energy under DE-FC02-04ER54698, DE-FG02-99ER54313, DE-AC05-76RLO1830, DE-FG02-86ER52123, DE-AC07-05ID14517, W-7405-Eng-48.

Funders	Funder number
U.S. Department of Energy	DE-AC05-76RLO1830, DE-AC07-05ID14517, DE-FG02-99ER54313, DE-FC02-04ER54698, DE-FG02-86ER52123, W-7405-Eng-48

Keywords

Blanket
Dual-coolant
Helium-cooled
ITER-TBM
Pb-17Li breeder

UN SDGs

This output contributes to the following UN Sustainable Development Goals (SDGs)

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10.1016/j.fusengdes.2005.05.012

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Wong, C. P. C., Malang, S., Sawan, M., Dagher, M., Smolentsev, S., Merrill, B., Youssef, M., Reyes, S., Sze, D. K., Morley, N. B., Sharafat, S., Calderoni, P., Sviatoslavsky, G., Kurtz, R., Fogarty, P., Zinkle, S., & Abdou, M. (2006). An overview of dual coolant Pb-17Li breeder first wall and blanket concept development for the US ITER-TBM design. Fusion Engineering and Design, 81 A(1-4), 461-467. https://doi.org/10.1016/j.fusengdes.2005.05.012

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title = "An overview of dual coolant Pb-17Li breeder first wall and blanket concept development for the US ITER-TBM design",

abstract = "An attractive blanket concept for the fusion reactor is the dual coolant Pb-17Li liquid (DCLL) breeder design. Reduced activation ferritic steel (RAFS) is used as the structural material. Helium is used to cool the first wall and blanket structure, and the self-cooled breeder Pb-17Li is circulated for power conversion and for tritium breeding. A SiCf/SiC composite insert is used as the magnetohydrodynamic (MHD) insulation to reduce the impact from the MHD pressure drop of the circulating Pb-17Li and as the thermal insulator to separate the high temperature Pb-17Li from the helium cooled RAFS structure. For the reference tokamak power reactor design, this blanket concept has the potential of satisfying the design limits of RAFS while allowing the feasibility of having a high Pb-17Li outlet temperature of 700 °C. We have identified critical issues for the concept, some of which include the first wall design, the assessment of MHD effects with the SiC-composite flow coolant insert, and the extraction and control of the bred tritium from the Pb-17Li breeder. R&D programs have been proposed to address these issues. At the same time we have proposed a test plan for the DCLL ITER-Test Blanket Module program.",

keywords = "Blanket, Dual-coolant, Helium-cooled, ITER-TBM, Pb-17Li breeder",

author = "Wong, {C. P.C.} and S. Malang and M. Sawan and M. Dagher and S. Smolentsev and B. Merrill and M. Youssef and S. Reyes and Sze, {D. K.} and Morley, {N. B.} and S. Sharafat and P. Calderoni and G. Sviatoslavsky and R. Kurtz and P. Fogarty and S. Zinkle and M. Abdou",

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doi = "10.1016/j.fusengdes.2005.05.012",

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Wong, CPC, Malang, S, Sawan, M, Dagher, M, Smolentsev, S, Merrill, B, Youssef, M, Reyes, S, Sze, DK, Morley, NB, Sharafat, S, Calderoni, P, Sviatoslavsky, G, Kurtz, R, Fogarty, P, Zinkle, S & Abdou, M 2006, 'An overview of dual coolant Pb-17Li breeder first wall and blanket concept development for the US ITER-TBM design', Fusion Engineering and Design, vol. 81 A, no. 1-4, pp. 461-467. https://doi.org/10.1016/j.fusengdes.2005.05.012

TY - JOUR

T1 - An overview of dual coolant Pb-17Li breeder first wall and blanket concept development for the US ITER-TBM design

AU - Wong, C. P.C.

AU - Malang, S.

AU - Sawan, M.

AU - Dagher, M.

AU - Smolentsev, S.

AU - Merrill, B.

AU - Youssef, M.

AU - Reyes, S.

AU - Sze, D. K.

AU - Morley, N. B.

AU - Sharafat, S.

AU - Calderoni, P.

AU - Sviatoslavsky, G.

AU - Kurtz, R.

AU - Fogarty, P.

AU - Zinkle, S.

AU - Abdou, M.

PY - 2006/2

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N2 - An attractive blanket concept for the fusion reactor is the dual coolant Pb-17Li liquid (DCLL) breeder design. Reduced activation ferritic steel (RAFS) is used as the structural material. Helium is used to cool the first wall and blanket structure, and the self-cooled breeder Pb-17Li is circulated for power conversion and for tritium breeding. A SiCf/SiC composite insert is used as the magnetohydrodynamic (MHD) insulation to reduce the impact from the MHD pressure drop of the circulating Pb-17Li and as the thermal insulator to separate the high temperature Pb-17Li from the helium cooled RAFS structure. For the reference tokamak power reactor design, this blanket concept has the potential of satisfying the design limits of RAFS while allowing the feasibility of having a high Pb-17Li outlet temperature of 700 °C. We have identified critical issues for the concept, some of which include the first wall design, the assessment of MHD effects with the SiC-composite flow coolant insert, and the extraction and control of the bred tritium from the Pb-17Li breeder. R&D programs have been proposed to address these issues. At the same time we have proposed a test plan for the DCLL ITER-Test Blanket Module program.

AB - An attractive blanket concept for the fusion reactor is the dual coolant Pb-17Li liquid (DCLL) breeder design. Reduced activation ferritic steel (RAFS) is used as the structural material. Helium is used to cool the first wall and blanket structure, and the self-cooled breeder Pb-17Li is circulated for power conversion and for tritium breeding. A SiCf/SiC composite insert is used as the magnetohydrodynamic (MHD) insulation to reduce the impact from the MHD pressure drop of the circulating Pb-17Li and as the thermal insulator to separate the high temperature Pb-17Li from the helium cooled RAFS structure. For the reference tokamak power reactor design, this blanket concept has the potential of satisfying the design limits of RAFS while allowing the feasibility of having a high Pb-17Li outlet temperature of 700 °C. We have identified critical issues for the concept, some of which include the first wall design, the assessment of MHD effects with the SiC-composite flow coolant insert, and the extraction and control of the bred tritium from the Pb-17Li breeder. R&D programs have been proposed to address these issues. At the same time we have proposed a test plan for the DCLL ITER-Test Blanket Module program.

KW - Blanket

KW - Dual-coolant

KW - Helium-cooled

KW - ITER-TBM

KW - Pb-17Li breeder

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DO - 10.1016/j.fusengdes.2005.05.012

M3 - Conference article

AN - SCOPUS:31744433642

SN - 0920-3796

VL - 81 A

SP - 461

EP - 467

JO - Fusion Engineering and Design

JF - Fusion Engineering and Design

IS - 1-4

T2 - Proceedings of the Seventh International Symposium on Fusion Nuclear Technology ISFNT-7 Part A

Y2 - 22 May 2005 through 27 May 2005

ER -

An overview of dual coolant Pb-17Li breeder first wall and blanket concept development for the US ITER-TBM design

Abstract

Funding

Keywords

UN SDGs

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