Neutronic analysis of Indian helium-cooled solid breeder tritium breeding module for testing in ITER

Abstract

India has proposed the Helium Cooled Solid Breeder blanket concept as a tritium breeding module for testing in ITER. The module has lithium titanate for tritium breeding and beryllium for neutron multiplication. Beryllium also enhances tritium breeding. A design of the module is prepared to analyze in more detail. The neutronic analysis is performed to assess the tritium breeding rate, neutron distribution, and heat distribution in the module. The tritium production distribution in sub-modules is evaluated to support the tritium transport analysis. The tritium breeding density in the radial direction of the module is also assessed for further optimization of the design. The heat deposition profile in the entire module is generated to support the heat removal circuit design. The estimated neutron spectrum in radial direction also provides a more in-depth picture of the nuclear interactions inside the material zones. The total tritium produced in the HCSB module is around 13.87 mg for an ITER Full day operation considering the 400 seconds ON time and 1400 seconds dwell time. The estimated nuclear heat load on the entire module is around 474 kW, which will be removed by the high-pressure helium cooling circuit. The heat deposition in the TBM is huge (around 9 GJ) for the entire day operation of the ITER, which demonstrates the scale of power that can be produced through a fusion reactor blanket. As per the Brayton cycle, it is equivalent to 3.6 GJ of electrical energy. In terms of power production, it is will be around 1655 MWh annually. The evaluation is done using the Monte Carlo Radiation transport code and ENDF-VIII nuclear cross-section data. The HCSB TBM neutronic performance demonstrates the tritium production capability and high heat deposition.