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

H L SWAMI; Deepak SHARMA; C DANANI; P CHAUDHARI; R SRINIVASAN; Rajesh KUMAR

doi:10.1088/2058-6272/ac5b3f

Plasma Science and Technology > 2022 > 24(6): 065601. > DOI: 10.1088/2058-6272/ac5b3f

H L SWAMI, Deepak SHARMA, C DANANI, P CHAUDHARI, R SRINIVASAN, Rajesh KUMAR. Neutronic analysis of Indian helium-cooled solid breeder tritium breeding module for testing in ITER[J]. Plasma Science and Technology, 2022, 24(6): 065601. DOI: 10.1088/2058-6272/ac5b3f

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

H L SWAMI^1,,
Deepak SHARMA^1,,
C DANANI^{1, 2},
P CHAUDHARI^{1, 2},
R SRINIVASAN^{1, 2},
Rajesh KUMAR^{1, 2}

1.
Institute for Plasma Research, Gandhinagar 382428, India
2.
Homi Bhabha National Institute, Anushaktinagar, Mumbai 400094, India

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Author Bio:
H L SWAMI, E-mail: hswami@ipr.res.in
Received Date: December 19, 2021
Revised Date: March 03, 2022
Accepted Date: March 06, 2022
Available Online: December 12, 2023
Published Date: May 18, 2022

Graphical Abstract

Abstract

Abstract

India has proposed the helium-cooled solid breeder blanket concept as a tritium breeding module to be tested in ITER. The module has lithium titanate for tritium breeding and beryllium for neutron multiplication. Beryllium also enhances tritium breeding. A design for the module is prepared for detailed analysis. Neutronic analysis is performed to assess the tritium breeding rate, neutron distribution, and heat distribution in the module. The tritium production distribution in submodules 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 of the entire module is generated to support the heat removal circuit design. The estimated neutron spectrum in the 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 per full day of operation of ITER, considering the 400 s ON time and 1400 s 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 test blanket model (TBM) is huge (around 9 GJ) for an entire day of operation of 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, this would be around 1655 MWh annually. The evaluation is carried out using the MCNP5 Monte Carlo radiation transport code and FEDNL 2.1 nuclear cross section data. The HCSB TBM neutronic performance demonstrates the tritium production capability and high heat deposition.
- HCSB,
- TBM,
- tritium,
- nuclear heat,
- ITER

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Acknowledgments

Mr H L Swami pays sincere tribute to the late Dr R Srinivasan. He was a good mentor and supervisor. His contribution towards the development of Indian nuclear fusion research always remains with us.

References (26)

References

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