A neutronic experiment to support the design of an Indian TBM shield module for ITER

A shield module is associated with an Indian Test Blanket Module (TBM) in ITER to limit the radiation doses in port inter-space areas.The shield module is made of stainless steel plates and water channels.It is identified as an important component for radiation protection because of its radiation exposure control functionality.The radiation protection classification leads to more assurance of the component design.In order to validate and verify the design of the shield module,a neutronic laboratory-scale experiment is designed and executed.The experiment is planned by considering the irradiation under a neutron source of 14 MeV and yields of 1010 n s-1.The reference neutron spectrum of the ITER TBM shield module has been achieved through optimization of the neutron source spectrum by a combination of steel and lead materials.The neutron spectrum and flux are measured using a multiple foil activation technique and neutron dose-rate meter LB 6411 (He-3 proton recoil counter with polyethylene),respectively.The neutronic design simulation is assessed using MCNP5 and FENDL 2.1 crosssection data.The paper covers neutronic design,irradiation and the outcome of the experiment in detail.

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

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 316 LN ITER-grade（SS-316 LN-IG）,stainless steel 316 L 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-316 LN-IG.

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

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.

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

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.