The Indian test blanket module（TBM） program in ITER is one of the major steps in the Indian fusion reactor program for carrying out the R＆D activities in the critical areas like design of tritium breeding blankets ...The Indian test blanket module（TBM） program in ITER is one of the major steps in the Indian fusion reactor program for carrying out the R＆D activities in the critical areas like design of tritium breeding blankets relevant to future Indian fusion devices（ITER relevant and DEMO）.The Indian Lead–Lithium Cooled Ceramic Breeder（LLCB） blanket concept is one of the Indian DEMO relevant TBM,to be tested in ITER as a part of the TBM program.Helium-Cooled Ceramic Breeder（HCCB） is an alternative blanket concept that consists of lithium titanate（Li_2TiO_3） as ceramic breeder（CB） material in the form of packed pebble beds and beryllium as the neutron multiplier.Specifically,attentions are given to the optimization of first wall coolant channel design and size of breeder unit module considering coolant pressure and thermal loads for the proposed Indian HCCB blanket based on ITER relevant TBM and loading conditions.These analyses will help proceeding further in designing blankets for loads relevant to the future fusion device.展开更多
For the solid blanket concept of helium cooled ceramic breeder (HCCB) demonstration fusion power plant (DEMO), a feasible blanket structure with configuration 2×X is proposed as considering relatively low tempera...For the solid blanket concept of helium cooled ceramic breeder (HCCB) demonstration fusion power plant (DEMO), a feasible blanket structure with configuration 2×X is proposed as considering relatively low temperature limit of neutron multiplier beryllium pebbles. Based on that, preliminary design for the typical blanket module of HCCB DEMO has been carried out and verified by thermal-hydraulic analysis and structural analysis. Furthermore, the specific relationship of maximum temperature depended on the surface heating of blanket key part first wall (FW) is also analyzed.展开更多
India,under its breeding blanket R&D program for DEMO,is focusing on the development of two tritium breeding blanket concepts;namely the lead-lithium-cooled ceramic breeder and the helium-cooled ceramic breeder(HC...India,under its breeding blanket R&D program for DEMO,is focusing on the development of two tritium breeding blanket concepts;namely the lead-lithium-cooled ceramic breeder and the helium-cooled ceramic breeder(HCCB).The study presented in this paper focuses on the neutronic design analysis and optimization from the tritium breeding perspective of the HCCB blanket.The Indian concept has an edge-on configuration and is one of the variants of the helium-cooled solid breeder blanket concepts proposed by several partner countries in ITER.The Indian HCCB blanket having lithium titanate(Li2TiO3)as the tritium breeder and beryllium(Be)as the neutron multiplier with reduced-activation ferritic/martensitic steel structure aims at utilizing the low-energy neutrons at the rear part of the blanket.The aim of the optimization study is to minimize the radial blanket thickness while ensuring tritium self-sufficiency and provide data for further neutronic design and thermal-hydraulic layout of the HCCB blanket.It is found that inboard and outboard blanket thicknesses of 40 cm and 60 cm,respectively,can give a tritium breeding ratio(TBR)>1.3,with 60%6Li enrichment,which is assumed to be sufficient to cover potential tritium losses and associated uncertainties.The results also demonstrated that the Be packing fraction(PF)has a more profound impact on the TBR as compared to 6Li enrichment and the PF of Li2TiO3.展开更多
文摘The Indian test blanket module（TBM） program in ITER is one of the major steps in the Indian fusion reactor program for carrying out the R＆D activities in the critical areas like design of tritium breeding blankets relevant to future Indian fusion devices（ITER relevant and DEMO）.The Indian Lead–Lithium Cooled Ceramic Breeder（LLCB） blanket concept is one of the Indian DEMO relevant TBM,to be tested in ITER as a part of the TBM program.Helium-Cooled Ceramic Breeder（HCCB） is an alternative blanket concept that consists of lithium titanate（Li_2TiO_3） as ceramic breeder（CB） material in the form of packed pebble beds and beryllium as the neutron multiplier.Specifically,attentions are given to the optimization of first wall coolant channel design and size of breeder unit module considering coolant pressure and thermal loads for the proposed Indian HCCB blanket based on ITER relevant TBM and loading conditions.These analyses will help proceeding further in designing blankets for loads relevant to the future fusion device.
基金supported by the National Special Project of China for magnetic confined nuclear fusion energy(2015GB108004)
文摘For the solid blanket concept of helium cooled ceramic breeder (HCCB) demonstration fusion power plant (DEMO), a feasible blanket structure with configuration 2×X is proposed as considering relatively low temperature limit of neutron multiplier beryllium pebbles. Based on that, preliminary design for the typical blanket module of HCCB DEMO has been carried out and verified by thermal-hydraulic analysis and structural analysis. Furthermore, the specific relationship of maximum temperature depended on the surface heating of blanket key part first wall (FW) is also analyzed.
文摘India,under its breeding blanket R&D program for DEMO,is focusing on the development of two tritium breeding blanket concepts;namely the lead-lithium-cooled ceramic breeder and the helium-cooled ceramic breeder(HCCB).The study presented in this paper focuses on the neutronic design analysis and optimization from the tritium breeding perspective of the HCCB blanket.The Indian concept has an edge-on configuration and is one of the variants of the helium-cooled solid breeder blanket concepts proposed by several partner countries in ITER.The Indian HCCB blanket having lithium titanate(Li2TiO3)as the tritium breeder and beryllium(Be)as the neutron multiplier with reduced-activation ferritic/martensitic steel structure aims at utilizing the low-energy neutrons at the rear part of the blanket.The aim of the optimization study is to minimize the radial blanket thickness while ensuring tritium self-sufficiency and provide data for further neutronic design and thermal-hydraulic layout of the HCCB blanket.It is found that inboard and outboard blanket thicknesses of 40 cm and 60 cm,respectively,can give a tritium breeding ratio(TBR)>1.3,with 60%6Li enrichment,which is assumed to be sufficient to cover potential tritium losses and associated uncertainties.The results also demonstrated that the Be packing fraction(PF)has a more profound impact on the TBR as compared to 6Li enrichment and the PF of Li2TiO3.
