Overview of TBM R&amp;D activities in India

Kumar, E. Rajendra; Jayakumar, T.; Suri, A.K.

doi:10.1016/j.fusengdes.2011.12.014

Cited by 35 publications

(3 citation statements)

References 7 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…A Mock-up of the Indian LLCB breeding blanket module [30][31][32] has Li 2 TiO 3 as tritium breeder. A neutronic model of this blanket module is constructed and inserted in the 3-D ITER neutronic model [33].…”

Section: Radiation Transport Simulationsmentioning

confidence: 99%

Multi-model quantification of defects in irradiated lithium titanate

Danani

Swami

Chaudhuri

et al. 2019

Fusion Engineering and Design

View full text Add to dashboard Cite

Lithium based ceramics (Li 2 TiO 3 , Li 4 SiO 4 etc.) in the packed pebble form, used in breeder blankets to breed tritium, are promising candidates for future fusion reactors. Neutron induced trap sites in the ceramic breeder affect the diffusion and inventory of tritium and thereby influence the quantity of tritium recovered. In the present work, a multi model framework has been employed to quantify the number of defects, which can be used to evaluate trap-sites, due to neutron irradiation in porous Li 2 TiO 3 pebbles. This framework is based on a three stage calculation viz. (i) radiation transport, (ii) obtaining the primary knock-on atom (PKA) energy spectrum and (iii) quantifying the ensuing damage by collision cascade simulations. The output of one stage is coupled to the next stage and finally the defect distribution is obtained. We use radiation transport calculations to obtain the neutron spectrum in ceramic pebbles. This is used as an input to SPECTER, a code that uses evaluated cross-sections, to obtain the Primary Knock-on Atom (PKA) energy spectrum. The PKA energy spectrum is then used to find the number and distribution of vacancies and displacements in porous Li 2 TiO 3 pebbles using binary collision approximation (BCA) based Monte Carlo simulations. The vacancy distribution and the number of vacancies in Li 2 TiO 3 have been evaluated for the ITER neutron source using the above frame work. The PKA energy spectrum of Li, Ti and O PKAs, have been obtained using SPECTER, and their average energies are 680, 120 and 320 keV respectively. The collision cascade simulations shows the expected Bragg peak for the number of defects created and has a range of 6-8 microns for O and Li PKA respectively and a fraction of a micron for Ti PKA. The total number of displaced atoms which is equal to the number of vacancies, at the end of the collision cascades, obtained using the above framework has also been used to estimate the displacements per atom (dpa) in Li 2 TiO 3 and this shows a good agreement with the dpa calculated using the NRT model.

show abstract

Section: Radiation Transport Simulationsmentioning

confidence: 99%

Multi-model quantification of defects in irradiated lithium titanate

Danani

Swami

Chaudhuri

et al. 2019

Fusion Engineering and Design

View full text Add to dashboard Cite

show abstract

“…India is developing a tritium breeding blanket for self-reliance in fusion reactor fuel supply [1][2][3][4]. The functions of blankets are tritium breeding, heat removal and radiation shielding.…”

Section: Introductionmentioning

confidence: 99%

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

Swami¹,

Sharma²,

Danani³

et al. 2022

Plasma Sci. Technol.

View full text Add to dashboard Cite

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.

show abstract

“…India has proposed the lead-lithium-cooled ceramic breeder (LLCB) and helium-cooled ceramic breeder (HCCB) as two breeding blanket options for DEMO. A module of the LLCB blanket is planned to be tested in ITER through the Indian Test Blanket Module (TBM) program [4][5][6]. The engineering [7] and nuclear analyses [8] of the Indian LLCB blanket module have already been performed.…”

Section: Introductionmentioning

confidence: 99%

Preliminary performance analysis and optimization based on 1D neutronics model for Indian DEMO HCCB blanket

Aggarwal

Danani

Youssef

2020

Plasma Sci. Technol.

View full text Add to dashboard Cite

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 (Li 2 TiO 3 ) 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% 6 Li 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 6 Li enrichment and the PF of Li 2 TiO 3 .

show abstract

Overview of TBM R&D activities in India

Cited by 35 publications

References 7 publications

Multi-model quantification of defects in irradiated lithium titanate

Multi-model quantification of defects in irradiated lithium titanate

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

Preliminary performance analysis and optimization based on 1D neutronics model for Indian DEMO HCCB blanket

Contact Info

Product

Resources

About