Lead Lithium Ceramic Breeder Research Articles

Entropy generation minimization-based heat transfer and mass transfer study of rectangular packed-bed filled with spherical particles

The study focuses on tritium-extraction and heat-extraction characteristics of the lead lithium ceramic breeder (LLCB) blanket module, an Indian tritium breeding blanket concept design in the International Thermonuclear Experimental Reactor (ITER). The present study focuses on the mass transfer phenomenon and entropy generation minimization (EGM)-based heat transfer analysis of the packed bed of a fusion reactor considering tritium isotopes in helium gas. For temperature-dependent convection-diffusion of tritium in helium gas and its permeation through the solid structure, a multi-region model in OpenFOAM using the locally varying porous medium approach is developed for studying the coupled multi-physics phenomena. For heat extraction behavior, a parametric study-based analysis is done using the EGM approach that involves irreversibility due to heat transfer, mass transfer, viscous, porosity effect, and heat generation rates. It is found that a lower Reynolds number (Re), and higher non-dimensional heat generation number (β), and non-dimensional species generation number (ϛ) result in a higher tritium extraction. In contrast, due to minimum total entropy generation, higher Re, and lower β and ϛ are preferred. Random packing arrangement as compared to BCC and FCC arrangements gives better tritium extraction and uniformity. However, the FCC arrangement is found as a better choice based on entropy minimization. Diagonally-opposite as compared to same-side inlet-outlet configuration extracts tritium more effectively and also found to be a better configuration from the EGM approach. The present work is significant since, for the breeder blanket, it presents the implementation of Sievert law boundary conditions and the EGM model for the first time in the literature.

Development of flow regime maps for lead lithium eutectic–helium flows

Abstract Institute for Plasma Research (IPR), Gandhinagar (India) is currently involved in the design and development of its Lead-Lithium Ceramic Breeder (LLCB) module for testing in the International Thermo-nuclear Experimental Reactor (ITER). In order to fulfill the ITER safety requirements, some postulated events need to be analyzed. Among them, the internal loss of coolant accident is being studied using RELAP/SCDAPSIM/MOD4.0 code. To this aim, RELAP/SCDAPSIM/MOD4.0 capabilities of modeling lead lithium eutectic (LLE) and helium flows are being extended nowadays by a collaboration of three institutions, namely IPR, Innovative System Software (ISS) and UPC-BarcelonaTech. The current study is part of this effort and is focused on the adjustments of existing RELAP/SCAPSIM/MOD4.0 flow regime maps in order to deal with the LLE-helium pair. Due to the lack of experimental data, the current study is based on a numerical assessment of LLE-helium flows. The volume of fluid method implemented in OpenFOAM toolkit has been chosen to characterize the flow regime at different mixture velocities and helium fractions. The paper includes a description of the followed methodology and its validation. Results of the systematic simulations of both horizontal and vertical upward flows are shown and the proposed flow diagrams presented. It is worthwhile to mention that the numerically obtained flow regime maps must be validated by an experimental campaign. However, the current approach allows the use of RELAP/SCDAPSIM/MOD4.0 code to properly initiate the study of breeding blanket off-normal events postulated for Lead-Lithium Ceramic Breeder, Helium Cooled Lithium-Lead and Dual Coolant Lithium-Lead designs.

Performance assessment of the Helium cooled First Wall mock-up in HELOKA facility

The Lead-Lithium Ceramic Breeder (LLCB) Test Blanket Module (TBM) being developed by India for testing in ITER adopts various well developed design engineering and manufacturing technologies. The First Wall (FW), which is directly exposed to the incident heat flux, is designed for high pressure helium flow, high operating temperature (up to 100 bar and 550 °C) and considerable thermal stress and electromagnetic disruption loads. In order to check the thermal performance of the FW and ensure its structural integrity, a mock-up of the FW fabricated in India was tested in HELOKA test facility at KIT, Germany. Both normal and accidental operating conditions were investigated under ITER-like surface heat fluxes. Based on these results the thermal performance of the FW were validated.

One-dimensional nuclear design analyses of the SST-2

Steady State Tokamak-2 (SST-2) will be an intermediate fusion machine before Indian DEMOnstration power reactor (DEMO) development to realise the reactor technologies. It is designed for fusion gain $$Q=5$$ and fusion power in the range of 100–300 MW. Nuclear design analyses of SST-2 machine have been carried out to support the conceptual design work. Analyses have been carried out for two breeding blanket concepts: Indian lead–lithium ceramic breeder (LLCB) and helium-cooled ceramic breeder (HCCB). The analyses assess the tritium production and radiation shielding capability of the machine referring to the engineering design parameters. In this study, one-dimensional radiation transport calculations have been performed to assess the SST-2 nuclear responses for 1 full power year (FPY) operation. Nuclear responses such as tritium breeding ratio (TBR), various radiation loads to toroidal field (TF) coil have been calculated to obtain the radial build-up of SST-2 capable of breeding tritium and satisfying the shielding requirements. The assessment has been made using the ANISEN code and FENDL 2.1 cross-section library. It is observed that the TBRs with LLCB and HCCB blankets are 0.85 and 0.94, respectively. Shielding calculations confirm that the radial build is sufficient to protect the superconducting TF coils for 1 FPY.

Neutronic design optimization of ITER TBM port#2 bio-shield plug

Indian Lead Lithium Ceramic Breeder (LLCB) Test Blanket Module (TBM) will be tested in one-half of the equatorial port #02 of ITER. The TBM systems demand the opening through bulk bio-shield boundary for installation & removal of TBMs from the port. In order to serve the requirement of TBM system, a Bio-Shield Plug (BSP) is placed at biological shielding location of the equatorial port. The neutronic design of BSP is important because it serves the purpose of biological shield boundary of ITER port and has to limit the ITER occupational radiation exposure behind the bio-shield up to 10 μSv/hr 1 day after shutdown. The BSP has to be removed and reinstalled during the removal and installation of TBM Sets from the port. The design of BSP should be compatible with TBM set installation & removal plans. It should also allow the TBMs coolant and instruments pipes to pass through. The shielding capability of BSP has to be adequate enough to limit the exposure. The engineering constraints like the feasibility of BSP removal/reinstallation & passage for pipes affect the shielding capability of BSP. The scope of this paper is to investigate the limitation of existing BSP and assess the better options for shielding capability improvement. The shielding capability of advanced shielding materials like B4C and Ferro-boron for BSP application is also assessed. The neutronic analyses have been performed using the MCNP radiation transport code and FENDL-2.1 nuclear cross section data library. The Activation code FISPACT-2007 has been employed to estimate the contact dose rates. The outcome suggests that B4C and Ferro-boron would be better candidate materials for the bio-shield plug of TBM port.

Design update and thermal-hydraulics of LLCB TBM first wall

The First Wall (FW) of Indian Lead Lithium Ceramic Breeder (LLCB) Test Blanket Module (TBM) is a U-shaped structure made of India specific Reduced Activation Ferritic Martensitic Steel (IN-RAFMS). As the FW is directly exposed to high heat flux from plasma, it is actively cooled by high pressure and high temperature helium gas. Following the Conceptual Design Review (CDR) at ITER, some improvement on the design of LLCB TBM has been envisaged, this includes the FW design to reduce the operational load on helium cooling process systems. The optimization studies have been carried out selecting suitable channel size, shape and no. of flow circuits of FW to keep the FW temperature and pressure drop within the acceptable limits. The parametric study has been performed with different mass flow rates and channel configurations of FW. This analysis resulted in the reduction of mass flow requirement by about 38% of the current value. The optimized parameters have been used to carry out the thermal-hydraulic and thermo-mechanical analyses of the two selected configurations of FW using ANSYS code. The temperatures and stresses are found to be within the required limits. Finally, the detailed thermal-hydraulics of final selected design of TBM FW have been performed using ANSYS CFX to estimate temperature distribution along with the validation of analytically calculated pressure drop and Heat Transfer Coefficient (HTC). It is observed that the temperature distribution on FW obtained from the flow analysis in CFX and that obtained by performing thermal analysis in ANSYS using HTC from engineering correlations are similar. The detailed analysis of the variation of HTC along the flow length of the heat flux zone has also been discussed. This paper also discusses transient and accident thermal analysis of the updated design for some of the ITER scenarios and the temperatures have been found to be within the required limits.

Data-acquisition, control & interlock system design for corrosion experiments of IN-RAFM steel with flowing Pb-Li in presence of magnetic field

Indian Reduced Activation Ferritic Martensitic Steel (IN-RAFMS) has been selected as the reference structural material for Lead-Lithium Ceramic Breeder Test Blanket Module (LLCB TBM). Qualification requirements for the selected structural material include assessment of compatibility, in TBM-like conditions, against flowing eutectic lead-lithium (Pb-Li) liquid metal, which is primary coolant for TBM internal structures as well as a tritium breeder and neutron multiplier. In this view, corrosion experiments are planned to be conducted at relevant temperatures and flow conditions in a dedicated pump driven experimental loop fabricated at Institute for Plasma Research (IPR). Such an experimental facility requires continuous operation over long duration along with constant monitoring and control of critical process parameters like temperature, pressure and flow. In addition, to ensure occupational safety and investment protection of associated loop components, alarms and interlocks are required to notify any off-normal condition(s) arising during loop operation and to mitigate possible accidental scenarios. To perform the intended functions, a data-acquisition, control and interlock system is designed using PXI-express technology and LabVIEW software platform. Functionality of developed system has been successfully validated in an integrated manner. This paper discusses definitions and implementation methods of control and interlock logics. Detailed design of developed system including performance of pressure control system and electromagnet interlock module is presented.

Fabrication feasibility studies for first wall of Indian LLCB TBM

The conceptual design of the Indian Lead Lithium Ceramic Breeder (LLCB) Test Blanket Module (TBM) First Wall (FW) presents a single U-shaped component made of IN-RAFM steel having 65 square channels of 20 mm × 20 mm running along the toroidal and radial direction. There are 13 circuits; each circuit consisting of 5 channels with inlet on the 1st channel of one side and the outlet on the 5th channel of other side. The conventional machining route with innovations in the existing technique was explored to develop the method to make channel mock-ups for FW application. This developed technique includes initial pre-machining and milling of higher thickness plates, making square channels by gun drilling followed by wire-cut EDM and final milling on the external surface and then hot bending into U-shape. Modified 9Cr-1Mo steel was the structural material used for this mock-up which has two flow circuits (each circuit of 5 channels). The other dimensions of this mock-up were chosen in such a way that functional requirements of full-sized FW would be fulfilled. During manufacturing, this mock-up was subjected to a series of inspection and non-destructive tests such as helium leak test and hydro-test at 150 bar pressure as a part of in-stage and final inspections. This paper describes the main activities carried out for the fabrication and factory acceptance tests of this U-shaped 10-channels TBM FW mock-up.

Analysis of trace levels of impurities and hydrogen isotopes in helium purge gas using gas chromatography for tritium extraction system of an Indian lead lithium ceramic breeder test blanket module.

In the fusion fuel cycle, the accurate analysis and understanding of the chemical composition of any gas mixture is of great importance for the efficient design of a tritium extraction and purification system or any tritium handling system. Methods like laser Raman spectroscopy and gas chromatography with thermal conductivity detector have been considered for hydrogen isotopes analyses in fuel cycles. Gas chromatography with a cryogenic separation column has been used for the analysis of hydrogen isotopes gas mixtures in general due to its high reliability and ease of operation. Hydrogen isotopes gas mixture analysis with cryogenic columns has been reported earlier using different column materials for percentage level composition. In the present work, trace levels of hydrogen isotopes (∼100ppm of H2 and D2 ) have been analyzed with a Zeolite 5A and a modified γ-Al2 O3 column. Impurities in He gas (∼10ppm of H2 , O2 , and N2 ) have been analyzed using a Zeolite 13-X column. Gas chromatography with discharge ionization detection has been utilized for this purpose. The results of these experiments suggest that the columns developed were able to separate ppm levels of the desired components with a small response time (<6min) and good resolution in both cases.

Experimental measurement and numerical modeling of the effective thermal conductivity of lithium meta-titanate pebble bed

The effective thermal conductivity (keff) of lithium meta-titanate (Li2TiO3) pebble beds under fusion relevant environments is an important property for the design of IN LLCB TBM (Indian Lead Lithium Ceramic Breeder Test Blanket Module). The transient hot wire technique was used to examine the thermal property of the Indian made Li2TiO3 material. The hot wire is used as both the heating element as well as for the temperature measurement. The keff of Li2TiO3 pebble bed has been investigated from room temperature to 800 °C. Experiments were performed on uncompressed Li2TiO3 pebble bed in stagnant helium gas filled at ambient pressure. A clear dependence of the keff on the temperature of the pebble bed was observed. The pebble bed has pebbles of 1 ± 0.15 mm diameter and packing fraction of 63%. The experimental results showed that the keff increased from 0.903 W/m°C to 1.204 W/m°C with the increase of bed temperature from 34.3 °C to 785.4 °C. The random close packing of poly dispersed Li2TiO3 pebble bed has been generated using discrete element method and then numerical modeling has been performed using finite element method to estimate keff. The numerically determined keff of the Li2TiO3 pebble bed agrees reasonably well with the obtained experimental data. The experimentally achieved keff results are also compared with the reported experimental results elsewhere and also with Zehner–Schlunder correlation.

Comparative studies for two different orientations of pebble bed in an HCCB blanket

The Indian Test Blanket Module (TBM) program in ITER is one of the major steps in its fusion reactor program towards DEMO and the future fusion power reactor vision. Research and development (R&D) is focused on two types of breeding blanket concepts: lead–lithium ceramic breeder (LLCB) and helium-cooled ceramic breeder (HCCB) blanket systems for the DEMO reactor. As part of the ITER-TBM program, the LLCB concept will be tested in one-half of ITER port no. 2, whose materials and technologies will be tested during ITER operation. The HCCB concept is a variant of the solid breeder blanket, which is presently part of our domestic R&D program for DEMO relevant technology development. In the HCCB concept Li2TiO3 and beryllium are used as the tritium breeder and neutron multiplier, respectively, in the form of a packed bed having edge-on configuration with reduced activation ferritic martensitic steel as the structural material. In this paper two design schemes, mainly two different orientations of pebble beds, are discussed. In the current concept (case-1), the ceramic breeder beds are kept horizontal in the toroidal–radial direction. Due to gravity, the pebbles may settle down at the bottom and create a finite gap between the pebbles and the top cooling plate, which will affect the heat transfer between them. In the alternate design concept (case-2), the pebble bed is vertically (poloidal–radial) orientated where the side plates act as cooling plates instead of top and bottom plates. These two design variants are analyzed analytically and 2D thermal-hydraulic simulation studies are carried out with ANSYS, using the heat loads obtained from neutronic calculations. Based on the analysis the performance is compared and details of the thermal and radiative heat transfer studies are also discussed in this paper.

Status of India-specific Reduced Activation Ferritic-Martensitic steel and fabrication technologies development for LLCB TBM

India is developing the Lead Lithium Ceramic Breeder Test Blanket Module (LLCB TBM) for in-situ experiments in ITER. India-specific Reduced Activation Ferritic-Martensitic (IN-RAFM) steel is the chosen structural material for the fabrication of LLCB TBM. IN-RAFM steel plates with different thickness are being produced from commercial heats. Presently, characterization for mechanical and physical properties is in progress in order to generate material properties database from these commercial heats.In order to avoid welds on plasma facing surface of TBM First Wall (FW), scaled down mock-ups of FW are being fabricated using Hot Isostatic Pressing (HIP). Electron Beam (EB) welding, Laser/laser-hybrid welding and Narrow Gap-Tungsten Inert Gas (NG-TIG) welding, are being developed for fabrication of other TBM components. These welding processes have distinct advantage over conventional welding processes because of their high depth to width ratio, minimal HAZ, lower weld volume, reduced distortion and lower residual stresses in the weld joints. Mock-ups of LLCB TBM components are being fabricated using Grade 91/IN-RAFM steel to investigate the accessibility for the envisaged welding processes and to identify and develop the non-destructive evaluation (NDE) techniques. In addition, this would also help to establish the assembly sequence of LLCB TBM. Ultrasonic based methodologies are under development to assure integrity of these joints in the fabricated components.Details of these developments in IN-RAFM steel as structural material, fabrication technologies and mock-ups fabrication towards realization of the LLCB TBM are discussed in this paper.

Nuclear analyses of neutron activation system for Indian TBM

Measurement of neutronic responses in Lead Lithium Ceramic Breeder (LLCB) Test Blanket Module (TBM) is an essential part of Test Blanket Module Program in ITER. It contributes to the validation of neutronic design and the estimation of uncertainties associated in neutronic calculations. Selection of neutronic measurement locations inside LLCB TBM has been made in consideration of two criteria. First, the location of measurements should be such that the spatial variation of neutronic responses in TBM is minimum. And second, the number of measurement locations should be based on the neutron spectra variation along with the radial depth of TBM. The Neutron Activation System (NAS) is proposed for measuring the neutronic responses inside LLCB TBM. It measures the absolute neutron flux with information on the energy spectrum at selected locations in the TBM. The nuclear analyses have been carried out for NAS arrangement inside LLCB TBM. The paper covers the identification of suitable location & neutronic measurement points inside TBM by estimating the 3D neutronic responses mapping inside LLCB TBM and radial profile of neutron flux. The impact of NAS arrangement inside TBM on the neutronic responses is analyzed by estimating the neutronic responses at the particular locations inside TBM with and without NAS arrangements. To conceptualize the NAS design, estimations of activity and reaction rates in candidate activation foils & materials have been done. The neutronic analyses have been performed using the MCNP radiation transport code and FENDL-2.1 nuclear cross section data library. The activation code FISPACT-2007 has been employed to estimate the activation responses.

Estimation of Tritium Release from LLCB TBM and Its Ancillary Systems and Tritium Management in Different Locations of ITER

The Indian Lead Lithium Ceramic Breeder (LLCB) Test Blanket Module (TBM) is to be installed in one half of equatorial port #2 for testing in ITER machine. Liquid Pb-Li and solid Li2TiO3 are the tritium breeder materials in LLCB TBM. Tritium permeates through structural materials in particular at higher temperatures, which is a major operational and safety concern. Therefore, tritium flows in different locations of ITER Tokamak complex have been estimated.Tritium transport from LLCB TBM and its ancillary systems into process rooms has been studied and analyzed in this work. A steady state diffusion limited permeation model neglecting surface effects has been used for the analysis. Tritium permeation to the Vacuum Vessel, Pipe Forest Area, Port Cell, Pipe Chase Area, Tokamak Cooling Water System Vault Annex (TCWS-VA) and Tritium Process Room in L-2 level has been estimated.The requirement to be fulfilled in each equatorial port cell is that the tritium concentration in the port cell during maintenance operations should be below the admissible limit for human access (regulatory maximum allowable value < 1 DAC = 3.4 × 105 Bq/m3, Derived Air concentration). The presence of the Detritiation System (DS) in the Port cell has to be taken into account.This admissible limit for human access has to be reached in a sufficiently short time (target = 12 h) after plasma shutdown. Additional release during maintenance and radiological zoning with recommended <10 µSv/h need to be considered. Management of concentration of permeated tritium in different locations considering above requirement has also been discussed in this paper.

Nuclear Analyses of Indian LLCB Test Blanket System in ITER

Heading towards the Nuclear Fusion Reactor Program, India is developing Lead Lithium Ceramic Breeder (LLCB) tritium breeding blanket for its future fusion Reactor. A mock-up of the LLCB blanket is proposed to be tested in ITER equatorial port no.2, to ensure the overall performance of blanket in reactor relevant nuclear fusion environment. Nuclear analyses play an important role in LLCB Test Blanket System design & development. It is required for tritium breeding estimation, thermal-hydraulic design, coolants process design, radioactive waste management, equipment maintenance & replacement strategies and nuclear safety. The nuclear behaviour of LLCB test blanket module in ITER is predicated in terms of nuclear responses such as tritium production, nuclear heating, neutron fluxes and radiation damages. Radiation shielding capability of LLCB TBS inside and outside bio-shield was also assessed to fulfill ITER shielding requirements. In order to supports the rad-waste and safety assessment, nuclear activation analyses were carried out and radioactivity data were generated for LLCB TBS components. Nuclear analyses of LLCB TBS are performed using ITER recommended nuclear analyses codes (i.e. MCNP, EASY), nuclear cross section data libraries (i.e. FENDL 2.1, EAF) and neutronic model (ITER C-lite v.l). The paper describes a comprehensive nuclear performance of LLCB TBS in ITER.

Preparation and analysis of helium purge gas mixture to be used in Tritium Extraction System of LLCB TBM

Hydrogen isotopes are extracted from the Ceramic Breeder (CB) and liquid Lead Lithium (Pb-Li) breeder of Lead Lithium Ceramic Breeder (LLCB) Test Blanket Module (TBM) with Helium purge gas. 1000 ppm of hydrogen gas is mixed with the purge helium gas to facilitate improved extraction of hydrogen isotopes from the breeder zones by hydrogen swamping reactions [1]. An experimental set up is developed for making up the purge gas mixture with a composition similar to the purge gas composition to be used for extraction of hydrogen isotopes from CB and Pb-Li of LLCB TBM. This is achieved by introducing different ppm levels (1000 - 5000 ppm) of hydrogen in helium gas by flow control mechanism. The analysis of the purge gas mixture is performed using a highly sensitive Gas Chromatography (GC) system. This paper describes the detailed design of the experimental set-up and results for the analysis of different concentrations of hydrogen in helium purge gas.

Neutronic performance of Indian LLCB TBM set conceptual design in ITER

Tritium breeding blanket testing program in ITER is an important milestone towards the development of the fusion reactors. ITER organization is providing an opportunity to the partner countries to test their breeding blanket concepts. A mock-up of Indian Lead Lithium Ceramic Breeder (LLCB) tritium breeding blanket known as LLCB Test Blanket Module (TBM) will be tested in ITER equatorial port no. 2. LLCB blanket consists of lead lithium (PbLi) as a neutron multiplier & tritium breeder, ceramic breeder (Li2TiO3) as a tritium breeder and India specific Reduced Activation Ferretic Martinic Steel (IN-RAFMS) as a structural material. A stainless steel block which is cooled by water, called as shield block, is attached with TBM to provide neutron shield to ITER TBM port. A comprehensive neutronic performance evaluation is required for the design of the LLCB TBM set (TBM+shield block) and associated ancillary systems in ITER. The neutronic performance of the conceptual design of TBM set in ITER has been carried out and reported here. In order to carry out the neutronic performance evaluation, the neutronic models of the LLCB TBM set along with TBM frame have been constructed and inserted in the equatorial port of ITER reference neutronic model C-lite. Neutronic responses such as tritium production rate, nuclear heating, neutron flux & spectra, gas production & DPA in the LLCB TBM set are calculated considering 500MW fusion power & fluence level of 0.3MWa/m2. Radiation transport code MCNP6 and FENDL 2.1 nuclear cross-section data library are used to perform the neutronic analyses of the LLCB TBM set. Neutronic analyses provide the important inputs for the LLCB TBM set engineering design, nuclear safety assessment & other associated ancillary equipment design.

Numerical modelling for the effective thermal conductivity of lithium meta titanate pebble bed with different packing structures

The effective thermal conductivity (keff) of lithium meta-titanate (Li2TiO3) pebble beds is an important parameter for the design and analysis of IN LLCB TBM (Indian Lead Lithium Ceramic Breeder Test Blanket Module). The keff of Li2TiO3 pebble beds under stagnant helium gas have been determined numerically using different uniform packing structures and random close packing (RCP) structures. The uniform packing structures of Li2TiO3 pebble bed are modelled by using the simple cubic, body centered cubic and face centered cubic arrangement. The packing structure of the RCP bed of Li2TiO3 pebbles is generated with the discrete element method (DEM) code. keff of Li2TiO3 pebble beds with different packing fractions have been reported as function of temperature; keff of the RCP Li2TiO3 pebble bed is compared with reported experimental results from literature. The numerically determined keff of the Li2TiO3 pebble bed agrees reasonably well with the experimental data.

Overview of LLCB TBM design and R&D activities in India

The Lead-Lithium Ceramic Breeder Test Blanket Module (LLCB-TBM) is the Indian TBM representing the DEMO breeding blanket, to be installed in ITER radial port no-2 for testing in ITER. The conceptual design of LLCB TBM systems and their engineering design are under progress. The neutronic analysis and engineering design of LLCB TBM set (TBM+TBM Shield) is undergoing optimization. The LLCB TBS auxiliary systems; Helium cooling systems, Lead-Lithium Cooling System and Tritium Extraction Systems design are conceptualized and detail design is under progress. The system arrangements in port cell area, Tokamak Cooling Water System (TCWS) vault and Tritium building along with interface requirements have been worked out within the allocated space. LLCB TBS related R&D activities are under progress at Institute for Plasma Research (IPR), Gandhinagar in association with Bhabha Atomic Research Centre (BARC), Mumbai and Indira Gandhi Centre for Atomic Research, Kalpakkam. The major research and development areas are liquidmetal technologies, lithium ceramic pebbles, lead-lithium eutectic alloy, India specific Reduced Activation Ferritic Martensitic Steels (IN-RAFMS) development and fabrication technologies development. This paper will highlight the current LLCB TBM set and axillary systems design and status of R&D activities in various areas.

Preliminary corrosion studies of P-91 in flowing lead–lithium with and without magnetic field for Indian lead–lithium ceramic breeder test blanket module

To study the corrosion of P-91 (9% chromium and 1% molybdenum) material with lead–lithium (Pb–Li) eutectic, two experiments were carried out in a forced convection loop, at eutectic temperature of 550 °C. The first experiment was carried out at a velocity of 15 cm s−1 for 1000 h and the second experiment, at a velocity of 30 cm s−1 for 2700 h. In both the experiments, P-91 sample coupons were exposed to Pb–Li flow in the presence and absence of magnetic field. Samples were analyzed using an optical microscope, scanning electron microscope (SEM) and electron probe micro-analyzer (EPMA). Micro-Vickers hardness testing was also carried out. Dissolution of elements into liquid metal is the main corrosion mechanism. Iron and chromium were selectively getting leached out from the near-surface region (∼4 µm) in the first experiment and molybdenum and manganese were also found leaching from a greater depth in the second experiment. The samples kept in the magnetic field showed a higher corrosion rate (∼320 µm/year) as compared with the corrosion rate (∼200 µm/year) of the samples kept in non-magnetic field regions. Hardness of the exposed samples was lower than the unexposed samples in both the experiments. Hardness was found to be low in the near-surface region for all the samples in both the experiments.