Neutronic performance of Indian LLCB TBM set conceptual design in ITER

Swami, H.L.; Shaw, A. K.; Mistry, A.N.; Danani, Chandan

doi:10.1016/j.fusengdes.2016.09.012

Cited by 19 publications

(7 citation statements)

References 11 publications

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“…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]. The ITER neutronic model along with zoomed view of blanket module mock-up is shown fig-2.…”

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

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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.

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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

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“…As with the irradiation scenario, the neutron spectrum of the front surface of a TBM port plug (equatorial port) is taken as a reference spectrum for irradiation of the materials. The neutron spectrum is shown in figure 1 [16].…”

Section: Reference Irradiation Scenario Neutron Flux Spectrum and Mamentioning

confidence: 99%

Activation characteristics of candidate structural materials for a near-term Indian fusion reactor and the impact of their impurities on design considerations

Swami

Danani

Shaw

2018

Plasma Sci. Technol.

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Activation analyses play a vital role in nuclear reactor design. Activation analyses, along with nuclear analyses, provide important information for nuclear safety and maintenance strategies. Activation analyses also help in the selection of materials for a nuclear reactor, by providing the radioactivity and dose rate levels after irradiation. This information is important to help define maintenance activity for different parts of the reactor, and to plan decommissioning and radioactive waste disposal strategies. The study of activation analyses of candidate structural materials for near-term fusion reactors or ITER is equally essential, due to the presence of a highenergy neutron environment which makes decisive demands on material selection. This study comprises two parts; in the first part the activation characteristics, in a fusion radiation environment, of several elements which are widely present in structural materials, are studied. It reveals that the presence of a few specific elements in a material can diminish its feasibility for use in the nuclear environment. The second part of the study concentrates on activation analyses of candidate structural materials for near-term fusion reactors and their comparison in fusion radiation conditions. The structural materials selected for this study, i.e. India-specific Reduced Activation Ferritic−Martensitic steel (IN-RAFMS), P91-grade steel, stainless steel 316LN ITER-grade (SS-316LN-IG), stainless steel 316L and stainless steel 304, are candidates for use in ITER either in vessel components or test blanket systems. Tungsten is also included in this study because of its use for ITER plasma-facing components. The study is carried out using the reference parameters of the ITER fusion reactor. The activation characteristics of the materials are assessed considering the irradiation at an ITER equatorial port. The presence of elements like Nb, Mo, Co and Ta in a structural material enhance the activity level as well as the dose level, which has an impact on design considerations. IN-RAFMS was shown to be a more effective low-activation material than SS-316LN-IG.

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“…The tritium breeding blanket is an important part of fusion reactor due to its functionality of tritium production which will provide the fuel self‐sufficiency to the reactor and also take out the energy deposited by neutrons to utilize in power production 11‐13 . There are many design concepts proposed worldwide for breeding blankets, 14‐25 India is also working on lead lithium cooled ceramic breeder (LLCB) blanket concept and helium cooled ceramic breeder (HCCB) concepts for Indian DEMO 25‐28 . HCCB concept has beryllium as neutron multiplier and lithium titanate as tritium breeder.…”

Section: Introductionmentioning

confidence: 99%

Helium cooled dual breeder blanket–preliminary design analyses of a candidate breeding blanket concept for near term IndianDEMOfusion reactor

et al. 2020

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Summary Helium cooled dual breeder (HCDB) blanket concept is designed for Indian DEMO fusion reactor and it is made of two tritium breeder materials PbLi and Li2TiO3. It has helium as a coolant and the India specific RAFMS as a structural material. High‐pressure helium first cools the plasma facing first‐wall and after that, it will extract heat from PbLi and ceramic breeder. Since PbLi is not used as a coolant, it therefore circulates with a low flow rate. It will overcome the corrosion and MHD issues associated with high temperature and high flow rate of PbLi. The idea behind the concept is to make a design which can be made using the existing blanket materials, extract high‐grade heat from the reactor and also enhance the availability. In HCDB blanket the role of neutron multiplier is done by PbLi, an alternative to the beryllium and it eliminates the issue associated with high toxic beryllium handling. It can be a potential tritium breeding blanket concept along with lead lithium cooled ceramic breeder (LLCB) and helium cooled ceramic breeder (HCCB) for near term Indian demonstration nuclear fusion power plant. In order to realize the HCDB conceptual design, preliminary estimations of tritium production, nuclear heat density have been carried out. The thermal behaviors of the HCDB blanket in Indian DEMO conditions have been also assessed and reported here. The assessment establishes the proof of HCDB blanket concept and supports it to be a good alternate blanket candidate for the Indian DEMO. The paper describes the HCDB concept along with analysis to verify the tritium self‐sufficiency and materials temperature limits.

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Neutronic performance of Indian LLCB TBM set conceptual design in ITER

Cited by 19 publications

References 11 publications

Multi-model quantification of defects in irradiated lithium titanate

Multi-model quantification of defects in irradiated lithium titanate

Activation characteristics of candidate structural materials for a near-term Indian fusion reactor and the impact of their impurities on design considerations

Helium cooled dual breeder blanket–preliminary design analyses of a candidate breeding blanket concept for near term IndianDEMOfusion reactor

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