High temperature reactor technology development in India

Progress in Nuclear Energy xxx (2017) 1e18

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High temperature reactor technology development in India I.V. Dulera*, Dr R.K. Sinha, A. Rama Rao, R.J. Patel Bhabha Atomic Research Centre, Mumbai, 400 085, India

a r t i c l e i n f o

a b s t r a c t

Article history: Received 21 July 2016 Received in revised form 17 April 2017 Accepted 21 April 2017 Available online xxx

High Temperature Reactor technology development programme was initiated in India with an aim to provide high temperature process heat for nuclear hydrogen production by splitting water. As high efficiency hydrogen production needs process heat at temperatures around 1123 K, a challenging technology development goal for the high temperature reactors was set to achieve coolant temperature of 1273 K. Currently development is in progress for a Compact High Temperature Reactor (CHTR), and a 600 MWth Innovative High Temperature Reactor (IHTR). Current design version of CHTR has 235U based TRISO (TRistructural-ISOtropic) coated particle fuel, Beryllium oxide (BeO) as moderator, graphite as reflector, and lead-bismuth eutectic (LBE) as the coolant. The design incorporates many passive safety features for reactor heat removal. Current design version of IHTR is based on pebble bed fuel configuration with molten salt as coolant. For both the reactors, reactor heat is removed passively by natural circulation of the coolant. Technology development for these reactors include development of TRISO coated particle fuel, lead-bismuth eutectic and molten salt coolant technologies, BeO and graphite, oxidation resistant coatings, high creep strength alloys compatible to these coolants, high temperature instrumentation for these coolants, as well as high efficiency hydrogen production and electricity generation technologies. © 2017 Elsevier Ltd. All rights reserved.

Keywords: High temperature reactors Hydrogen Liquid metal coolant Molten salt coolant High efficiency hydrogen production Passive safety

1. Introduction Due to exhausting world reserves of petroleum-based products and the environmental concerns with use of fossil fuel, it has become important to find an alternative energy carrier for transport applications. Hydrogen is considered an attractive alternative. Large numbers of institutions including R & D laboratories, educational institutes as well as many industries in India are involved in carrying out developmental work related to all the aspects of hydrogen energy. This encompasses hydrogen production, storage, utilisation in transport and power generation sector, development of safety codes and standards etc. While options for production of hydrogen from fossil fuel, such as steam methane reforming and coal gasification have been developed to satisfy demand during interim periods, nuclear energy based hydrogen production by splitting water is being developed as a sustainable and environmentally benign option. Indian High Temperature Reactor (HTR) technology development programme is aimed at nuclear hydrogen production by splitting water. Water splitting

* Corresponding author. E-mail address: [email protected] (I.V. Dulera).

hydrogen production processes, either by thermo-chemical or electrolysis routes, need process heat at high temperatures or electricity or both depending upon selection of the specific process. The efficiencies for these processes are usually higher at higher temperatures. Thermo-chemical processes operating in the temperature range of 823e1123 K have been reported to have efficiencies in the range 40e57% depending on the thermo-chemical process selected (Sinha and Banerjee, 2005; Prasad and Dulera, 2009) have described various processes and their energy requirements). Nuclear reactors have a great potential for supplying energy for these hydrogen production processes at required high temperature conditions in a sustained manner. Indian HTR programme is aimed (Sinha, 2013) at development of reactor systems capable of providing process heat at 1273 K. Currently, technology development program for HTR has been initiated to develop a small power Compact High Temperature Reactor (CHTR), and a 600 MWth Innovative High Temperature Reactor (IHTR). The CHTR would serve as technology demonstrator for technologies related to high temperature reactors. IHTR is aimed at producing large scale hydrogen, although it can as well be used for high efficiency electricity generation. In the subsequent paragraphs, the paper briefly reviews HTR concepts developed in the world. This is followed by CHTR and

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details of its design, its fuel, in-core and structural materials, reactor physics design, thermal hydraulics design, and passive systems of the reactor. Subsequently the paper describes IHTR and molten salt related thermal hydraulics and material compatibility studies. Finally paper very briefly describes developmental work for hydrogen production. 2. History of high temperature reactors As it is not the purpose of this paper to review history of HTRs and current developments, a very brief and indicative review of some of the HTRs has been provided to highlight similarities and differences of Indian reactors. (Brey, 2003), has briefly described evolution and future of developments of HTGRs): 2.1. Gas cooled reactors (GCRs) Gas cooled reactors were the initiator for all High Temperature Gas Cooled Reactors (HTGRs). Commercial experience with gas cooled reactors began in 1956 with generation of electricity from Calder Hall plant in United Kingdom (UK). Subsequently 26 Magnox and 15 Advanced Gas Cooled Reactors (GCRs) were operated by UK. In France, Japan and Spain also development of commercial GCRs took place. HTGRs were developed to improve upon the performance of GCRs. 2.2. Early HTGRs Development of HTGRs began in 1950s. HTGRs utilise ceramic particle fuel with coatings. Fuel is usually dispersed in graphite matrix. These reactors use graphite as the moderator material. Either prismatic type graphite moderator blocks are used or spherical fuel elements are used. Coolant Helium flows through coolant holes in the blocks or through the interstices in the pebble bed core. HTGRs can operate at very high core outlet temperatures because of all ceramic core. The early HTGR plant designs were based on steel primary system pressure vessels. Dragon in UK, AVR in Germany and Peach Bottom in USA were the first HTGRs. Some of the early HTGRs have been described in the following paragraphs: 2.2.1. Dragon It was a 20 MWth reactor built in UK. It first operated in 1965. It was an international project, started in 1959. The primary objective was to demonstrate the HTGR feasibility. The reactor used enriched uranium carbide fuel elements based on coated particle fuel. Helium was used as coolant. The core inlet and outlet temperatures were 623 K and 1023 K respectively. It did not produce any electric power, but served as a platform for development of helium gas cooled reactors and advanced fuel particle coatings. 2.2.2. AVR In Germany this 15 MWe Arbeitsgemeinshaft Versuchsreaktor (AVR) started operation in 1967. This was a pebble bed type of reactor, used particle fueled graphite spheres of 0.06 m diameter. This had a steel containment vessel. AVR had a coolant outlet temperature of 1223 K. The reactor was operated for about 1,22,000 h till 1988. 2.2.3. Peach Bottom This 40 MWe HTGR in USA operated from 1967 to 1974. This reactor utilised coated particle fuel. This reactor used two kinds of core. The fuel particles for core-1 had a single layer of anisotropic carbon to prevent hydrolysis of carbide fuel kernels and had several failed fuels. This lead to development of BISO (Buffer isotropic

pyrolytic carbon) coatings on the fuel kernels. 2.2.4. Fort St. Vrain This reactor commissioned in USA used Pre-stressed concrete as the primary system containing vessel. This used a core of hexagonal graphite block fuel elements and reflectors with fuel in the form of TRISO coated particle fuel, once through steam generator modules producing superheated and reheated steam, steam turbine driven helium circulators. This reactor served as a test-bed for demonstrating valuable technologies including the reactor core in the form of hexagonal graphite blocks with TRISO coated fuel particles, reactor internals, steam generators, fuel handling and helium purification systems. 2.2.5. THTR-300 This Thorium High Temperature Reactor 300 MWe plant was sponsored by Germany. This reactor also had its reactor vessel made of pre-stressed concrete. It operated during 1985 to 1988. Large HTGR steam cycle plant designs for the following reactors were worked out: 2.2.6. HTR-500 HTR-500 was a 500 MWe German design. It made considerable simplifications and optimisations based on the practical experience gained by THTR-300. This plant featured a simple design with the primary system components located within a single cavity Prestressed Concrete Reactor Vessel (PCRV). This was a pebble bed reactor. This reactor had a capacity of 1390 MWth to produce 550 MWe of electricity. 2.2.7. VG-400 This was 1060 MWth Russian pebble bed reactor design for cogeneration i.e. electricity production and supplying heat at core outlet temperature of 1223 K. Currently following two small power HTGRs are operating. 2.2.8. HTTR The High Temperature Engineering Test Reactor (HTTR) of the Japan Atomic Energy Research Institute (JAERI) is a graphitemoderated and helium gas cooled reactor with an outlet temperature of 1223 K and a thermal output of 30 MWth. The major objectives of the HTTR was to establish and upgrade the technological basis for advanced HTGRs and to conduct various irradiation tests for innovative high temperature basic researches. This was shut down since Fukushima Accident in 2011. JAEA, however, is under preparation of its operational re-start. 2.2.9. HTR-10 The 10 MWth High Temperature Gas-Cooled Reactor-Test Module is located at the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University in China. It is a pebble bed, helium cooled, graphite moderated modular HTGR. The HTR10 utilizes spherical fuel elements containing TRISO coated particle fuel. It attained first criticality in December 2000. Safety demonstration tests have been carried out for loss of forced coolant and control rod withdrawal without SCRAM demonstrating its safety aspects. Phase 1 (Steam Turbine Cycle: HTR -10ST) is continuing, and transitional works towards Phase 2 (Gas Turbine Cycle: HTR -10GT) are under way. Future HTRs include HTR-PM, under construction; and VHTR, under development: 2.2.10. HTR-PM This program in China is of HTGR plant demonstration and commercialization, and is based on experiences of HTR-10. HTR-PM

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demonstration plant comprising of 2 reactor modules started its construction in 2012 in Shidao Bay, Rongcheng City, Shandong Province, Its operational start is scheduled in late 2017. 2.2.11. VHTR The Very-High-Temperature Reactor (VHTR) is a next big step in the evolutionary development of HTGRs. The VHTR, being developed as a part of Gen-IV reactors, is a helium-gas-cooled, graphitemoderated, thermal neutron spectrum reactor with a core outlet temperature higher than 1173 K, and a goal of 1273 K, sufficient to support high temperature processes such as production of hydrogen by thermo-chemical processes. The reference thermal power of the reactor is set at a level that allows passive decay heat removal, currently estimated to be about 600 MWth. The VHTR is useful for the cogeneration of electricity and hydrogen, as well as for other process heat applications. It is able to produce hydrogen from water by using thermo-chemical, electro-chemical or hybrid processes. It may be seen that most of the high temperature reactors use graphite as moderator and helium as coolant. Indian designs differ especially with respect to coolant as emphasis has been given to natural circulation based passive cooling leading to use of molten metal and molten salt based coolants. For CHTR, BeO has been chosen as moderator material as well as reflector material partially. This has been done, as against graphite, to achieve a compact reactor core. For the larger IHTR, graphite has been selected as moderator and reflector material. As with most of the HTRs, the selected fuel is based on TRISO coated particle fuel. Although BARC is developing hydrogen production technologies for processes requiring heat at 823 K (Copper-Chlorine process) and greater than 1123 K (Sulfur-Iodine process), a decision was taken to carry out more challenging development for reactors operating at 1273 K so as to facilitate high temperature hydrogen production, expected to yield high efficiencies. This shall also be well suited for high efficiency electricity generation. In order to meet these applications, developments are also being carried out for the above mentioned hydrogen production processes as well as high efficiency electricity generation. Development of super-critical CO2 Brayton cycle (SCBC) is in progress for high efficiency electricity generation. 3. Compact High Temperature Reactor The design approach for CHTR is based on a vision to achieve goals such as high fuel burnup, natural circulation of coolant, compact design with safety based on inherent safety features and passive systems. Importance of the reactor lies in its uniqueness to address high temperature process heat requirement of various applications, including hydrogen production with complete passive safety, having practically no moving part in the reactor block and no fluid at high pressure. On account of the later attributes, the reactor is eminently suitable for operation in remote areas with minimal technical supervision of the experts. Design variants of CHTR concept is therefore also attractive as small nuclear power packs for supplying electricity in areas not connected to electricity grid. In view of this requirement, the reactor has been designed to be compact in weight and size for ease in its deployment in remote locations. Moreover there is flexibility to adopt systems operating at lower temperatures, alternate fuel cycle and coolants. The core of the reactor (Dulera and Sinha, 2008; Dulera et al., 2009) have described salient design features of the reactor) consists of nineteen hexagonal shaped Beryllium oxide (BeO) moderator blocks. These blocks contain centrally located graphite fuel tube which carries fuel inside longitudinal bores located on its wall. Central bore of the fuel tube is used for flow of LBE as primary coolant. This coolant has been chosen to suit natural circulation based passive

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cooling. This reactor provides flexibility to use different kinds of fuel. Reactor physics designs for 233U-Th as well as enriched 235U based fuel have been established. Currently design option based on enriched 235U based fuel is being pursued. Major design and operating parameters are listed in Table 1. TRISO coated particle fuel is the building block for the fuel compacts, facilitating high burnup and high temperature process heat production. The moderator blocks are surrounded by eighteen BeO reflector blocks. These blocks are in turn surrounded by graphite reflector blocks. These core components are contained in a reactor shell of a material (currently chosen material is Nb-1%Zr-0.1%C alloy) suitable for high temperature applications and resistant to corrosion against LBE coolant. Fig. 1 (Sinha et al., 2016) shows cross-section of the reactor core. Cover plates of the same material at top and bottom close the reactor shell. Cylindrical vessels, called coolant plenums, are provided above and below the reactor shell for coolant exit and entry into the core respectively. Natural circulation based flow of coolant between the two plenums remove nuclear heat passively from the reactor core. The reactor has heat utilisation vessels on top of the upper plenum to provide an interface to systems for high temperature process heat applications. A set of sodium heat pipes passively transfer heat from the upper plenum to the heat utilisation vessels. One more set of heat pipes has been provided to transfer heat to a heat sink in the case of any postulated accident. Primary shut down system consists of a set of six tantalum alloy shut-off rods, which fall by gravity in the six first ring coolant channels of the core. This system as well as the control system consists of tubes of tantalum-tungsten alloy. For both these systems, central ballast of tungsten pellets is provided to counteract the up-thrust of LBE. The ballast rod is made into separate pieces for ease of fabrication. Each of these pieces is located with the help of recesses provided in adjacent pieces. Dropping end of the shut-off rods is conical in shape. Drop-time estimation was carried out for the shut-off rods and was estimated to be about 0.7 s. Experimental validation is planned in near future. Secondary shutdown system is in the form of liquid poison injection system injecting poison in tubes located in BeO reflector region. Burn-up compensation rods are housed in the remaining six BeO reflectors. These are kept fully inserted in the beginning, and periodically moved out slowly. Layout of CHTR components is shown in Fig. 2. CHTR development is aimed to meet the objectives and requirements for the advanced nuclear energy systems. Some of the features which are being aimed through design include: a) No impact in public domain (passive safety) under postulated accident condition. This is achieved due to:
Natural circulation based heat removal.
Decay heat removal by passive means.
TRISO coated particle based fuel having high limiting temperature (1873 K).
High boiling point coolant (1943 K).
Ceramic core having high thermal inertia resulting in slow rate of temperature rise in case of any postulated event. b) No impact in public domain (passive safety) under postulated accident condition. This is achieved due to:
Natural circulation based heat removal.
Decay heat removal by passive means.
TRISO coated particle based fuel having high limiting temperature (1873 K).
High boiling point coolant (1943 K).
Ceramic core having high thermal inertia resulting in slow rate of temperature rise in case of any postulated event. c) Nuclear Waste

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I.V. Dulera et al. / Progress in Nuclear Energy xxx (2017) 1e18 Table 1 CHTR design and operating parameters. Attributes

Design Parameters

Reactor power Fuel Refueling time Fuel Burnup Coolant Inlet temperature Outlet temperature Moderator Reflector Heat removal by Primary shutdown system Secondary shutdown system Control system

100 kWth TRISO coated particle fuel 15 effective full power years 68,000 MWd/t of heavy metal Molten LBE 1173 K 1273 K BeO Partly BeO and partly graphite Natural circulation of coolant Six mechanical shut-off rods made of Ta alloy filled with Tungsten pellets Twelve liquid poison injection systems Twelve systems of Ta alloy filled with Tungsten pellets

Fig. 1. Core cross-section of CHTR (Sinha et al., 2016).


Reprocessing of nuclear graphite is aimed.
In future, reprocessing of TRISO coated particle fuel to close the fuel cycle. d) Economics
High temperature heat can be used for high efficiency electricity or hydrogen generation e replacement of fossil fuel based transportation. e) Flexibility in fuel cycle
Fuel could be based on 233U-thorium or enriched uranium. f) Siting
Can be operated at remote locations, with minimal infrastructure. Technology challenges for this reactor are being met through indigenous technology development and studies. Some of these are: a) Reactor physics design of long life compact core. b) High temperature liquid metal coolant e thermal hydraulics, coolant chemistry, equipment and instrumentation development. c) High temperature material development. d) High performance, high temperature nuclear fuel.

e) Development of ceramic in-core material e beryllia, graphite and carbon-carbon composite. f) Oxidation resistant coatings for graphite and structural materials. g) Passive means of heat removal by natural circulation of coolant under normal operation. h) Passive heat transfer from primary to secondary system by high temperature sodium heat pipes. i) Decay heat removal by passive means. j) High efficiency hydrogen production system. - Thermochemical processes: Sulphur-Iodine and Copper Chlorine processes. - High temperature steam electrolysis. Some of these developments have been described in the subsequent paragraphs. 3.1. CHTR fuel The philosophy of HTR fuel, based on TRISO coated particle fuel, is different than that of Pressurised Heavy Water Reactor (PHWR) and Light Water Reactor (LWR) fuel pellets and cladding. The fuel in an HTR core is contained in billions of coated particles, each of them

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i) All components must be designed to accommodate expected loads, which could occur during handling and shipping. j) The reactor can be brought to a safe state following any postulated accident. Safety design requirements: The various life limiting phenomena which are considered in safety design of coated particle are given below:

Fig. 2. CHTR components layout (Sinha et al., 2016).

have its own containment. The small kernels of fuel are each coated with layers of carbon and silicon carbide. The resulting particle is designed to withstand the pressure of the generated fission gases and to form an essentially leak tight barrier to fission product release. While LWR fuel cladding performs this function on a larger scale during normal operating conditions, coated particle fuel also requires this high level of integrity under accident conditions. Thus, the fuel particle is required to stay intact with high reliability during both normal operation and accident conditions. Fundamental design basis and functional requirements for the CHTR fuel is that the fuel does not fail during normal operating conditions and during anticipated operational occurrences. Failure here means breach of fission product containing barriers (coatings) and release of fission products from the coated fuel particle. The various functional requirements that have been established for CHTR fuel are given below: a) Fuel must generate the required power with negative fuel temperature coefficient for the desired exit burnup and long life. b) Retain fuel and fission gas products which are to be isolated from the primary coolant. c) Minimise parasitic neutron capture. d) Maintain fuel tube and fuel compact location in the core and distribute fuel kernel to satisfy the basic nuclear and thermal-hydraulic requirements. e) The fuel tube must direct reactor coolant through the core to achieve acceptable flow distribution, so that the heat transfer performance requirements can be met for all modes of operation. f) Resist chemical and irradiation-induced material damage. g) The mechanical design of fuel must assure that fuel damage is improbable during normal operation and anticipated operational occurrences. h) Provision for handling, shipping, and core loading of the fuel.

i) The coated fuel particles has been designed to avoid, in principle, failure considering irradiation induced damage and chemical attack through the entire service life. ii) The coated particles should not fail systematically and for this the design has considered the following: a) Penetration depth of Pd-SiC interaction (or corrosion depth) shall not exceed 50% of thickness of SiC layer. b) Kernel migration shall not exceed the thickness which is the sum of first two layers c) Internal pressure due to fission gas and CO generated shall not lead to failure of coatings. iii) Maximum fuel temperature shall not exceed 1873 K at any anticipated transient to avoid failure of coated fuel particle due to SiC coating degradation. iv) Initial failure fraction in coating layers shall be less than 0.2% in terms of the sum of exposed uranium and SiC defects. (limiting off-site exposure). v) The fuel compacts and the graphite blocks shall not be broken or cracked considering thermal stress and irradiation-induced damage. The CHTR fuel, based on TRISO coated particles, is designed to operate at high temperatures and withstand high burn up. Fertile material and the burnable poison make the fuel temperature coefficient negative, thus making the reactor inherently safe. Each fuel assembly carries fuel inside 12 equip-spaced longitudinal bores of 0.01 m diameter. These bores will be filled with fuel compacts of 0.01 m diameter and 0.035 m length made by embedding TRISO coated particles in graphite matrix. A typical TRISO coated fuel particle has a kernel (500 mm diameter) comprising of fissile, fertile, and burnable poison materials followed by four coating layers. The functional requirements and dimensions of these layers are given below: a) Low-density buffer pyrolytic carbon (BPyC) layer: This porous layer (90 mm thick) acts as an absorber and gives volume to accommodate fission products and kernel swelling. b) Inner high-density PyC (IPyC) layer: This layer (30 mm thick) serves as a barrier to gross diffusion of fission products and fission gases. This is to protect integrity of subsequent SiC layer. c) Silicon carbide (SiC) interlayer: This layer (30 mm thick) contains gaseous fission products released by the kernel and thus acts like a pressure vessel. This also acts as an additional diffusion barrier to metallic fission products. The thickness needs to be adequate to withstand the developed pressure and corrosion attacks by fission products. d) Outer high-density PyC (OPyC) layer: This layer (50 mm thick) as well as IPyC layer, on irradiation, puts SiC layer into compression to limit effective stresses. It also provides protection to SiC layer against corrosion from external media. It also provides bonding surface for making fuel compact. As very large number (about 14 millions) of TRISO particles comprise the core, manufacturing of fuel kernels, deposition of

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multilayered coatings on the particles, and characterisation of the coated as well as uncoated particles were the major challenges, which are being met through indigenous R&D. The technology related to oxide and carbide based micro sphere fabrication by Internal Gelation Process (IGP) was already established in BARC (Kumar et al., 2013) explains about the IGP process and microparticle production). This technique was selected since this has the natural tendency to form spherical particles since solution droplets are converted to gel microspheres in a chosen gelation medium. The required size can be obtained and desired density can also be achieved. Good control over surface morphology can be achieved by variation of chemical parameters. Moreover this process is adaptable for automation. Technology was developed (Sathiyamoorthy et al., 2009) describes about the process development for TRISO coatings) for providing coatings of PyC and SiC on fuel kernel by Chemical Vapour Deposition (CVD) process in spouted bed reactor. This reactor has the capability to fluidize high density particles by a specific and closed circuit manner resulting uniform contact between particles and gaseous precursor for CVD to take place. These are obtained by pyrolytic decomposition of hydrocarbon gas (for PyC layers) and Methyl TriChloro Silane (CH3SiCl3) vapour for (SiC layer). Different coatings are prepared at different temperatures. The specifications (Dulera and Vaze, 2009) describes specifications of the CHTR fuel) for the coatings were very stringent with respect to coating uniformity in shape, size and properties. This was a novel development. Four coating layers were required to be prepared in a single campaign at processing temperature ranging from 1623 K to 1723 K. Achieving and maintaining the uniformity of coating properties like thickness, density, porosity, isotropicity, crystallite size, hardness and phases for four successive coatings was a challenging task. During the process of formation of coatings, there is a continuous change in particle size and density requiring adjustment of parameters. Various process parameters like temperature, volume of the bed, composition and flow rates of the gases, etc. were optimized to meet these challenges. Analytical models and codes were developed for simulating the manufacturing process so as to optimize the parameters (Mollick et al., 2014) describes modeling and simulation of the coating process). Code was (Dutta and Vaze, 2012) provides model for studying behaviour and performance of the TRISO particles) developed for studying performance behaviour of TRISO particles under various operating conditions. These studies are important to understand the long-term behaviour of the fuel. In addition, fuel compact fabrication technology, with feasibility for automation, was also developed. Schematic of TRISO coated particle fuel and fuel compacts are shown in Fig. 3. CHTR fuel bed consists of prismatic BeO moderator block with centrally located graphite fuel tube carrying fuel compacts. Schematic of single fuel bed is shown in Fig. 4. Techniques used for characterization are: Transmission Electron Microscopy (TEM), Scanning Electron Microscopy (SEM), Raman, X-

Fig. 4. Schematic of a CHTR fuel bed (Kaushik et al., 2012).

ray Powder Diffraction (XRD), Small Angle X-ray Scattering (SAXS), Small-Angle Neutron Scattering (SANS), Brunauer, Emmett and Teller (BET) technique, Pycnometry, Nano-indentation etc. Fig. 5 shows SEM images of TRISO coated fuel particle and radiographs of fuel compacts. 3.2. Development of structural materials CHTR core materials (consisting of fuel tubes, moderator blocks and reflector blocks) are exposed to extreme environmental conditions, such as neutron fluence, high temperatures (around 1273 K) and corrosive environment of LBE coolant. CHTR core internal materials comprise high-density nuclear grade BeO moderator and reflector blocks, nuclear grade high density isotropic graphite fuel tubes, down comer tubes, and reflector blocks. Other

Fig. 5. SEM images of TRISO coated fuel particle and radiographs of fuel compacts (Sinha et al., 2016).

Fig. 3. Schematic of TRISO coated particle fuel and fuel compact (Dulera and Sinha, 2008).

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metallic structural materials are based on refractory metal alloys such as Nb-1%Zr-0.1%C, and tantalum-tungsten based alloy. Graphite and these alloys are coated with oxidation resistant coatings. Current status of development of these materials has been briefly covered in the following paragraphs: 3.2.1. BeO At high temperatures, candidate moderator materials were Beryllia (BeO) and graphite. High purity BeO was chosen due to its larger x (energy loss by neutron per collision) and lesser number of collisions for thermalisation as compared to graphite. It has low neutron absorption cross section, high moderating capacity, very high melting point, very low vapour pressure at high temperatures. Beryllia is produced by calcinations of beryllium hydroxide. Crude beryllium hydroxide of 98% purity was produced by fluoride route from Indian beryl ore. The major impurities present in the crude beryllium hydroxide are iron, magnesium, calcium, manganese, sodium, silica etc. The neutron cross-section areas of these impurities are relatively higher than that of beryllia. Due to this reason these impurities are not desirable for nuclear grade beryllia. This was further purified using hydro processing to 99.6%, which is essentially a nuclear grade beryllia. The sintered powder produced by this process was characterized by XRD to identify the crystallite phase. The chemical analysis of the powder was carried out using DC arc carrier distillation. The high purity beryllia powder so obtained can be sintered to achieve theoretical density more than 99% at 2073 K by vacuum hot pressing method. A typical sample of nuclear grade beryllia block of 99% density was produced by vacuum hot pressing technique. Subsequently technologies were also developed for (Ghanwat and Sathiyamoorthy, 2008) describes technology behind production of nuclear grade BeO) production of high density nuclear grade BeO blocks by cold pressing followed by sintering as well as vacuum hot pressing route. Sample pieces with high density (>99% Theoretical density) prototypes were made up of nuclear grade Beryllia. This is shown in Fig. 6. 3.2.2. Carbon based materials Fuel tubes, down comer tubes, and outer reflector materials are made of high density isotropic nuclear graphite. Developments include (Sinha and Dulera, 2010) explains technology challenges for TRISO coated particles as well as issues related to use of carbon based materials for nuclear application, (Dhami et al., 2006) describes development aspects of the special carbon based materials) components made of high density carbon-carbon composite and high density graphite. High density Carbon-Carbon (C-C) composite CHTR fuel tube (Fig. 7) samples were fabricated in collaboration with National Physical laboratory (NPL), New Delhi for demonstrating the technologies involved. Lab-scale facilities have been setup in BARC to process C-C composite coupons using resin impregnation and Chemical Vapor Infiltration (CVI) techniques. Prototype fuel tube of high density graphite using calcined petroleum coke & coal tar pitch as raw

Fig. 6. Cylindrical and annular hexagonal high density nuclear grade beryllia blocks.

Fig. 7. Carbon-carbon composite tubes.

material was manufactured by extrusion, carbonization and graphitization. In parallel, in house precision machining of fuel tube and other graphite components was initiated. Prototypes of fuel tube, moderator and reflector blocks, and down comer tubes, were manufactured. These are shown in Fig. 8. Facilities have been setup for impregnation and carbonizing so as to obtain high density material. 3.2.3. Niobium alloy The Nb-1%Zr-0.1%C alloy is one of the most promising refractory metal alloys, having an excellent combination of high temperature properties. It is suitable for several structural applications in the new generation advanced nuclear reactors operating at temperatures exceeding 1273 K. This alloy was developed (Tewari et al., 2011a) explains the novel technique used for development of alloy and metal working and other fabrication aspects) in collaboration with Nuclear Fuel Complex (NFC), India. The deformation behavior of as-cast material at different temperatures and strain rates, recrystallization behavior at different temperatures, time and evolution of microstructures at different processing conditions (ascast, deformed and recrystallized) were studied. The as-cast Nb alloys were deformed up to 35% at different temperatures. To determine the optimum recrystallization temperature and time for the alloy, several heat treatments were conducted by systematically varying temperature and time. The welding of this alloy was also a challenging task due to its high melting point and reactive nature. In comparison to the conventional welding processes, high energy density sources like laser and electron beam which can produce deep penetrations with narrow heat affected zones are more suitable to weld the components of this Nb-alloy. Studies were conducted for laser and electron beam welding (Tewari et al., 2011b) discusses studies on the laser and electron beam welding of the Nb-

Fig. 8. Machined prototype graphite components.

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1%Zr-0.1%C alloy) of the Nb alloy using both processes in bead on plate and butt joint configuration by varying the process parameters. Weld joints produced by both techniques were subjected to optical and electron microscopy, and the micro hardness profile across the width of welds was also studied. Subsequent to development of alloy, components of a loop for carrying out natural circulation studies on lead-bismuth coolant at 1273 K was fabricated and assembled. This is shown in Fig. 9. Oxidation resistant coating for the alloy was also developed and tested (Vishwanadh et al., 2013) describe studies on oxidation behaviour of Nb alloy and Silicide coated alloy). 3.2.4. Oxidation resistant coating for graphite Carbon based components are prone to oxidation at high temperatures. Therefore they need to be coated with oxidation resistant coating. Many techniques for various types of coatings on graphite have been developed. The technique includes plasma spray, CVD, plasma CVD etc. Coatings developed include SiC (Selvakumar et al., 2012) describes induction assisted CVD method), alumina, yttria, Yttria Stabilized Zirconia (YSZ) etc. (Padmanabhan et al., 2013) describes the plasma spray method of yttria coating on graphite for molten metal applications). For the plasma spray technique, simulation studies were carried out for process parameter optimization. Simulation code was developed and validated. Particle temperatures and velocity profiles for various particle sizes, input powers and gas flow rates were worked

out. A window of operating parameters for plasma spray deposition of yttrium oxide was arrived at based on the particle velocity and temperature profile provided by the simulation studies. Experimental studies were carried out in the range of parameters within this window to optimize the process. Optimization was done based on experimental studies with coating adhesion as the response parameter. Yttria was deposited on graphite substrates by plasma spray technique. Particle simulation in plasma jet, in-flight particle temperature and velocity and process optimization were carried out. Coating characterization was carried out by XRD and SEM. Differential thermal analysis (DTA) of yttria coated graphite was carried out at 1498 K. Specimen after Differential Thermal Analysis (DTA) was analyzed by SEM, Energy-Dispersive X-ray Spectroscopy (EDX) and XRD. Thermal shock resistance was carried out in flowing Ar-H2 atmosphere from 1273 K to room temperature. Graphite slabs of size 1.5 m  0.7 m were coated with yttria. Facilities to characterize these coatings have also been setup. In addition Indira Gandhi Centre for Atomic Research (IGCAR) (Sure et al., 2015) describes behaviour of yttria stabilized zirconia coating on graphite under thermal cycling) has carried out significant R&D and has set up facilities for coating on graphite as well as their characterization and qualification. In IGCAR coatings of partially stabilized zirconia (PSZ) and Y2O3 coatings were developed for protection of high density graphite (HDG). The PSZ coated HDG samples were exposed to molten LiCl-KCl salt at 873 K under Ultra High Purity (UHP) argon atmosphere. The PSZ and coated HDG exhibited excellent corrosion resistance in molten LiClKCl salt due to no morphological change and phase changes after 2000 h of exposure. PSZ coated HDG samples were subjected to thermal cycling at 873 and 1023 K up to 200 cycles to understand the durability of the coatings. The results of this study approved the choice of HDG crucibles with PSZ coating for corrosion protection in molten salt environments. In another study IGCAR group (Vetrivendan et al., 2015) has also studied and concluded that significant improvement was found in the thermal cycling life (upto 1723 K) of plasma sprayed yttria coating over HDG if an interlayer of SiC is provided). Fig. 10 shows coated graphite pieces. 3.3. Reactor physics design Reactor physics design of CHTR core with a very high operating temperature and very long core life was a challenging task. Use of non-standard materials and non-availability of nuclear data at high temperatures posed several challenges. Moreover, due to compactness of the core, designing the control system was a challenging task. The reactor physics design (Dwivedi et al., 2011, 2015) describe the reactor physics design of CHTR and 3D space time analysis of anticipated transient without scram in CHTR with fuel temperature feedback respectively) of the CHTR has been carried out with the following objectives: a) Negative fuel temperature reactivity coefficient;

Fig. 9. Schematic and photograph of High temperature liquid metal loop of Nb alloy manufactured at NFC (Sinha et al., 2016).

Fig. 10. PSZ coated high density isotropic graphite crucible (IGCAR) and Yttria coating on graphite crucible (BARC).

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b) Fuel capable of high temperature performance; c) High burn up fuel with long refueling interval. Initially the core was designed for pure 233U based fuel. During the analysis of the core physics, it was found that 100% 233U results in a positive fuel temperature reactivity coefficient (FTC). The coefficient was shown to become negative if fissile fuel is mixed with either fertile isotopes like 232Th and/or a burnable poison. A combination of 233U mixed with thorium and small amount of gadolinium (added only in central fuel tube) satisfies the reactivity control and other design requirements. Later on, a design based on enriched 235U was worked out and is being pursued. The reactor fuel consists of 8 kg of enriched 235U. For present fuel configuration in CHTR core, it has been estimated that this fuel would generate required power of 100 kWth for 15 effective full power years of continuous operation. The Doppler coefficient of reactivity was found to be satisfactorily negative. The physics design simulation of the CHTR has been carried out with 2-stage calculations. In the first stage, for each type of fuel assembly containing fuel pins, structural material, clad, coolant and moderator, a very fine detailed neutron transport calculations were performed. In this stage, the energy groups of the neutron spectrum were also condensed. In the second stage, the reactor core, constructed with these homogenized cell types, was simulated. The neutron diffusion equation is solved for such a core and flux, criticality etc. are computed. The 22-energygroup cell-homogenized cross-sections were obtained with collision probability code ITRAN. The code uses first flight Collision Probability/Interface Current methods. The calculations have been done using the 172-group ENDFB-VI library. These 22-group cell homogenized cross-sections were used in triangular/hexagonal mesh diffusion theory code TriHTR. Both the above codes were also used for burnup calculations. Preliminary transient analysis had been carried out with code called MRIF which is based on point reactor model with thermal-hydraulics feedbacks. It can be seen (Fig. 11) that the initial core reactivity is quite large. Burnable Compensation Rods (BCRs) have been used to bring down the initial reactivity. The lower curve shows the variation of keff with burnup for the configuration with all BCRs inside the core. Subsequently BCRs will be removed. The keff values for this fuel design, as a function of burn-up in terms of refueling period, are shown in Fig. 11. To control/adjust small excess reactivity and to shut down the reactor if required, CHTR is designed to have control and shut down systems to provide sufficient negative reactivity throughout

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the core life. As mentioned above, a number of in-house developed computer codes like ITRAN, TriHTR, and Monte Carlo based codes were used for the design. The effect of smaller single control rod worth and negative FTC can be gauged by analyzing a postulated accidental scenario in which an inadvertent fast withdrawal of a control rod in critical state is considered. A preliminary safety analysis based on point-reactor kinetics model with temperature feedback was carried out. This results in power rise and subsequent rise in fuel temperatures. A negative reactivity feedback is introduced by the rise in fuel temperature due to negative FTC and the power stabilizes at about 3 times the initial power. More importantly, the fuel and coolant temperatures stabilize to 1493 K and 1330 K respectively within the permissible limits. The worth of primary and secondary shutdown systems is found to be adequate to shutdown the reactor independently, even in case of one rod failure, in most reactive situations. Effect of a fast transient is shown in Fig. 12. 3.4. Thermal hydraulics design As mentioned earlier, LBE was chosen as the primary coolant for CHTR. Many features of the lead-bismuth eutectic alloy make it attractive for its application as coolant. Some of these have been mentioned below: a) High boiling point of 1943 K, with sufficient margin to ensure single phase operation. Moreover, coolant void reactivity is negative; b) Good heat transfer properties; c) Inertness in reaction with water and air without any explosion d) Low moderation and absorption of neutrons; e) Good reflecting properties; f) No g-activity; g) Good neutronic performance; h) Feasibility of reactor heat removal as well as reactor decay heat removal by passive means; i.e. natural circulation of coolant. The reactor heat is removed by passive natural circulation of this coolant. Following studies were carried out to establish and validate the design: a) Analytical studies and development of computer codes for natural circulation of LBE coolant;

Fig. 11. Variation of keff with burn-up (Sinha et al., 2016).

Fig. 12. Fast transient effectewith drawl of maximum worth control rod in 5 s.

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b) Experimental loops with LBE coolant for natural circulation studies. One loop is operating at 823 K since 2009 and another operating at 1273 K since April 2014. c) Freezing and de-freezing studies for LBE; d) Development of instrumentation for LBE; Detailed analytical and experimental studies (Borgohain et al., 2015) describes natural circulation studies in lead bismuth eutectic loops) were carried out to establish the design. Fig. 13 shows schematic of LBE loop operating at 823 K. The test loop mainly consists of a heated section, air heat exchanger, valves, various tanks and an argon gas control system. All the components and piping are made of SS316L. The LBE ingots are melted in a melt tank and then transferred to the sump tank. The LBE coolant in the sump tank is then pressurized by the argon gas system. Due to pressurization, molten LBE flows into the loop and subsequently fills up the loop. After filling, the loop is isolated from the sump tank by a valve. Natural circulation of the coolant takes place in the loop due to heating of the coolant in the heated section and cooling in the heat exchanger. Air is used as the secondary side coolant in the heat exchanger. After losing heat in the heat exchanger, LBE enters the heated section through a 15 mm (0.015 m) Nominal Bore (NB) pipeline. During the operation of the loop, high purity argon gas was used as cover gas in the vessels. Electrochemical oxygen sensor is used for monitoring the oxygen level in LBE. Intermittently, argon plus hydrogen gas is used to maintain the chemistry of the LBE coolant in the loop. All the vessels and piping of the loop are equipped with band heaters for trace heating. The loop is provided with adequate thermal insulation in order to reduce heat loss to the environment. In the main heated section of the loop, heat is generated by electrical heaters and transferred to the liquid metal coolant as sensible heat. It mainly consists of ‘U’ shaped mineral insulated immersion type heater elements mounted on a stainless steel flange. The expansion and contraction in the liquid metal due to heating and cooling is accommodated with the help of an expansion tank, partially filled with molten LBE, and located between the heated section and the heat exchanger. The cover gas provided over the surface of the LBE in the expansion tank acts as a cushion. The cover gas pressure is maintained with the help of a regulating valve provided in the cover gas system. Schematic of a

similar LBE natural circulation loop but operating at 1273 K is shown in Fig. 14. A computer code, LeBENC (Lead Bismuth Eutectic Natural Circulation) was developed (Borgohain et al., 2011a) describes experimental studies carried out in LBE loop, development and validation of computer code as well as oxygen sensor) to study the steady state and transient behaviour of liquid metal natural circulation loop. This is a one dimensional finite difference code and can handle uniform and non-uniform diameter piping in the loop, different working fluids (water and LBE) and accounts for axial conduction in the fluid and pipe wall. This was validated with single phase water loop natural circulation data. Comparison of experimental and analytical results is shown in Fig. 15. The correlation conforms to the experimental results. So, it can be concluded that, in spite of high thermal conductivity of LBE, the role of axial conduction of LBE in natural circulation is not considerable. Transient studies were also carried out to simulate postulated accident scenario in the reactor. The main objective was to observe the LBE coolant behaviour during various transient conditions. The transient experimental studies include startup of the loop from stagnation conditions, loss of heat sink and step power change. The studies showed good correspondence between analytical and experimental results. 3.5. Safety and reliability 3.5.1. Safety concept and design philosophy Since the CHTR is expected to be used as a power pack for remote areas, it incorporates many novel design and safety features for reactor operation with fewer operator interventions, thereby minimising skilled man-power requirements for operation. The CHTR strongly relies on inherent safety features and passive systems for reactor control, shutdown, and heat removal under normal and abnormal conditions. 3.5.2. Provisions for simplicity and robustness of the design Some of the provisions for simplicity and robustness of design are the high temperature capability of the fuel, low power density of the core, high thermal capacities and thermal conductivity of the components of ceramic core, high boiling point, chemically inert

Fig. 13. Schematic and photograph of LBE natural circulation loop operating at 823 K.

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Fig. 14. Schematic and photograph of LBE natural circulation loop operating at 1273 K (Sinha et al., 2016).

Fig. 15. Comparison of analytical and experimental results.

liquid metal coolant and the use of inherent and passive safety features. 3.5.3. Inherent safety features CHTR has many features, which make it inherently safe. These are listed below: a) A strong negative Doppler coefficient of the fuel for any operating condition results in reactor power reduction in case of fuel temperature rise during any postulated accident scenario; b) High thermal inertia of the all-ceramic core and low core power density (~500 W/litre) results in very slow temperature rise of the fuel and reactor core components during a condition when heat sink is lost; c) A large margin between the normal operating temperature of the fuel (around 1373 K) and the allowable limit of the TRISO coated particle fuel (1873 K) to retain fission products and gases resulting in their negligible release during normal operating conditions. This also provides a healthy margin to take care of any unwanted global or local power excursions;

d) A negative moderator temperature coefficient results in lowering of reactor power in case of increase in moderator temperature due to any postulated accident condition; e) Due to use of LBE coolant with high boiling point (1943 K), there is a very large thermal margin to its boiling, the normal operating temperature being 1273 K; f) Coolant itself is chemically inert, and is maintained in inert gas atmosphere. Even in the eventuality of accidental contact with air or water, it does not react violently with explosions or fires; g) Due to above atmospheric temperature coolant melting point (396 K), even in case of a primary system leakage, it solidifies and prevents further leakage; h) The thermal energy stored in the CHTR coolant, due to very low pressure, is small as compared to high pressure coolants in case of light and heavy water reactors, thus leading to smaller energy release in the event of a leak or accident; i) Very low pressure in the coolant, allows use of a graphite/ carbon based coolant tube having low neutron absorption cross section, thus improving neutronics of the reactor; j) For the well known problem of formation of polonium-210 in LBE based reactor, studies are proposed to be carried out as regards filtration as well as cleaning. k) In case of a leakage, the coolant retains iodine and other radionucleides. l) For LBE, the reactivity effects (void, power, temperature, etc.) are negative; thus reducing the reactor power in case of any inadvertent power or temperature increase.

3.5.4. Development of passive systems CHTR has many passive features as listed below: a) Passive core heat removal under normal operation: During normal operation of the reactor, the core heat is removed by natural circulation of LBE coolant. The main coolant-circulating loop comprises fuel tubes, down comers and top and bottom plenums. The coolant at 1173 K enters the fuel tube in lower plenum, takes the reactor heat, and at 1273 K it is delivered to the upper plenum. After transfer of heat to heat pipes, coolant returns to lower plenum through down comer tubes.

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b) Passive transfer of heat to secondary system: A set of twelve high temperature sodium heat pipes passively transfer heat from upper plenum of the reactor to a set of heat utilisation vessels which are kept directly above the upper plenum. Manufacturing and testing facilities for high temperature sodium based heat pipes have been set-up (Basak et al., 2011) describes design of heat pipes); (Panda et al., 2015) gives details of design as well as development of manufacturing process for high temperature sodium based heat pipes). Computer codes have been developed for design and analysis of these heat pipes (Panda et al., 2017) describes numerical simulation of high temperature sodium heat pipe). c) Passive heat removal under postulated accident conditions (Basak et al., 2005) describe conceptual designs of the proposed passive heat removal systems): CHTR has heat pipe based passive heat removal system to cater to postulated accident condition. This system, which is capable of removing neutronically limited power of 200 kWth, which is 200% of normal reactor power, prevents the temperature of the core and coolant from increasing beyond a set point. For the loss of load condition, when coolant circuit is intact, a system of six variable conductance sodium heat pipes dissipates heat to the atmosphere. Studies and development of carbon-carbon composite based high temperature heat pipes are in progress to remove power from the core, in case the coolant is lost.

3.6. Development of heat pipes Sodium based high temperature heat pipes have been developed for passive heat removal in molten metal/salt cooled high temperature reactors. Both theoretical and experimental studies have been carried out. Codes for parametric design calculations based on correlations have been developed. Simplified Finite Element Method (FEM) models of heat pipes were developed with commercial FEM code COMSOL. To examine the internal workings of the heat pipe, a full Computational Fluid Dynamics (CFD) model, based on OPENFOAM, has been prepared with user defined modules to model the interface mass flows at the wick-vapour interface, saturated nature of vapour phase, modified wick properties, as well as user defined material properties. A range of facilities have been commissioned for fabrication of sodium heat pipes. These include capacitive discharge spot welding machine, helium leak testing machine, outgassing furnace and residual gas analysers for degassing of heat pipes, inert gas glovebox for handling and charging of sodium, compact hydraulic press and high capacity welding machine for crimping and sealing. The antechamber of the glovebox (Fig. 16) can accommodate heat pipes upto 0.5 m long. An extension of the mini-antechamber has been provided so as to facilitate fabrication of longer (upto 1.5 m) heat pipes. Still longer heat pipes can be easily accommodated by attaching appropriately

sized antechamber extensions. Using this facility, a high temperature sodium heat pipe has been fabricated and operated upto 800  C and the experimental measured temperature profile was compared (Fig. 17) with the computed values from the simplified conduction FEM and the detailed CFD models. The graphs clearly show a good match between experimental results with those predicted by the developed CFD model. The resistance due to flow of vapour has been incorporated in the CFD model. The resistance due to wall and wick has been modeled in FEM. These lead to better match with experimental results. 3.7. Structure of the defence-in-depth Some of the major features of the CHTR design, structured in accordance with various levels of the defence in depth are listed below: 3.7.1. Level-1 Prevention of abnormal operation and failure: The CHTR design features contributing to this level are as follows: (i) Heat removal from the core under normal operating conditions is accomplished through natural circulation of the coolant, which essentially eliminates the hazard of a loss of coolant flow; (ii) The extent of overpower transients and their consequences are limited by: a) Low core power density; b)A highly negative Doppler (fuel temperature) coefficient, achieved through the selection of an appropriate fuel composition; c) Use of burn-up compensation rods to compensate for reactivity change with burn-up; d)Negative reactivity effects (void, power, temperature etc.) achieved with the use of LBE coolant; e) Use of all-ceramic core with high heat capacity and high temperatures margins; and f)The resulting low excess reactivity.

3.7.2. Level-2 Control of abnormal operation and detection of failures: The CHTR design features contributing to this level are the following: i)The use of two independent passively operating shutdown systems; ii) The use of a high heat capacity ceramic core to prevent fuel temperature from exceeding the design limits for a long time. The above mentioned design features are expected to result in the reactor operation and safety functions being fully passive and independent of operator intervention. 3.7.3. Level-3 Control of accidents within the design basis: Features of the CHTR that contribute to this level are:

Fig. 16. Details of inert gas glovebox for high temperature heat pipe fabrication.

i) The use of two independent shutdown systems, one comprising of mechanical shut-off rods and the other employing a liquid poison injection based passive shutdown system, altogether resulting in an increased shutdown reliability; ii) The use of two independent heat pipe based systems to transfer reactor core heat to the outside environment during abnormal conditions;

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Fig. 17. Comparison of observed heat pipe surface temperatures with those predicted by developed models (Panda et al., 2017).

3.7.4. Level-4 Control of severe plant conditions, including prevention of accident progression and mitigation of consequences of severe accidents: The features important for this level are: i) Excellent high temperature (up to 1873 K) performance of TRISO coated particle fuel, ensuring that the probability of the release of fission products and gases is very low. ii) Large heat capacity ceramic core, resulting in a slow fuel temperature rise with more than 50 min being available for a corrective action even when all heat sinks are lost; iii) The feasibility of erection of the reactor in an underground pit with sealed barrier of reinforced concrete and steel covers is foreseen to provide an additional barrier for the prevention of release of radionucleides.

3.7.5. Level-5 Mitigation of radiological consequences of significant release of radioactive materials.

of graphite properties in thermal Analysis of CHTR); (Kanse et al., 2012) have studied micro-mechanical studies on graphite strength prediction models). Rules for design of graphite components significantly differ from those of metallic components. Efforts (Basak et al., 2012) describe an approach for graphite component design for nuclear reactors) are being made to formulate the design rules for these components. CFD analysis were carried out with respect to flow of coolants in the complex geometries (Borgohain et al., 2011b) describe CFD analysis for heat transfer in low Prandtl number fluid flows). In order to establish safety objectives in case of an earthquake, seismic behaviour of structural components was investigated. Since CHTR core consists of loose blocks of beryllium oxide and graphite separated from each other by small gaps. This feature introduces non-linearity in the structure and analysis was done to understand the phenomenon under seismic event for such case. This was carried out to determine vibration characteristics of CHTR core and to design locking/anti-rotation features. Analysis was carried out using an explicit dynamic FEM code. Result of analysis for seismic behaviour of a single stack of CHTR is shown in Fig. 18.

i) Passive design features of the previous levels avoid any significant release of radioactive materials and necessity for evacuation or relocation measures outside the plant site. 3.8. Computer code development and analytical studies A number of computer codes have been developed for system design and modeling. These are HPDATA (Panda et al., 2017) describes numerical simulation of high temperature sodium heat pipe) for heat pipe design and analysis, TSDATA, for thermosyphon design analysis, code for gas gap filling system design and analysis (Basak et al., 2005) describes conceptual designs of the proposed passive heat removal systems and the code development for the design and analysis of these systems) BTRISO (Dutta and Vaze, 2012) provides model for studying behaviour and performance of the TRISO particles under various operating conditions and effect of various parameters e.g. physico-thermal and mechanical properties of the coatings) for modeling of TRISO particle fuel behaviour etc. A large number of thermal and structural analysis (Kaushik et al., 2014) describe safety studies for CHTR) and (Raza et al., 2015) describe thermal & safety analysis for CHTR) have been carried out and studies have been carried out with respect to structural design for graphite components (Kaushik et al., 2012) have studied effect

Fig. 18. Results of seismic analysis of CHTR core.

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The analysis showed initiation of whirling motion during such event. Hence design modifications were carried out to provide antirotation feature in the moderator and reflector blocks. A steady state two dimensional thermal analyses of the reactor core mid plane was carried out to determine the prevalent temperatures in various components of the reactor under normal operating conditions. The temperature contours are shown in Fig. 19. These results show that maximum temperature at mid-plane is 979  C (1252 K). The CHTR fuel, TRISO coated particle fuel, can sustain up to 1873 K. This ensures the availability of sufficient operating margins. BeO and graphite components were also analysed for thermal stresses and were found to be within acceptable limit. The influence of the neutron irradiation on the materials has been studied. Due to low neutron flux levels, it is expected that swelling would be limited. Moreover, sufficient clearances have been incorporated in the dimensions of the components to take care of differential swelling and thermal expansions.

3.9. Design and operating characteristics of systems for non-electric applications When used for high temperature process heat applications, the CHTR would include a suitable fluid circulated through the heat utilising vessels to transfer high temperature process heat through an interface heat exchanger to the two stages of the Sulfur-Iodine (S-I) process for hydrogen production. In this way, about 320 Nm3/day of hydrogen gas could be produced. In addition to this, the waste heat could be used to produce 1.5 m3/day of potable water. In addition BARC is also carrying out R & D (Kaushik et al., 2015) describes optimized process parameters for IHTR) for high efficiency production of electricity using SCBC. The specific advantage of supercritical CO2 is its high density, which is comparable to liquids resulting in low compressor work and component sizes. The sizes of SCBC components are considerably smaller than those of traditional Rankine cycle, Helium Brayton cycle and gas turbine power plants. SCBC is simple, compact, and less expensive and have shorter construction periods, thus improves overall economics. For the application as smaller power pack, development also includes thermo-electric generators (Basu et al., 2014) describes improved Si-Ge based thermoelectric generators) for direct conversion of heat to electricity.

4. 600 MWth Innovative High Temperature Reactor BARC is carrying out design of a 600 MWth reactor for commercial hydrogen production (Dulera et al., 2005) provides logic behind selection of pebble bed concept and molten fluid based coolant); (Vijayan et al., 2013) describes IHTR as well as other advanced reactors being developed in India as well as internationally). One of the considerations for design was passive removal of decay heat from the surface of the reactor vessel. Analysis showed requirements of a high surface area to volume ratio for the vessel containing the core leading to a slender design. Analysis for reactor physics showed that an annular core was preferable. Various design options as regards fuel configurations, such as prismatic bed and pebble bed were considered for thermal hydraulics and temperature distribution analysis. The options were analysed as regards temperatures seen by the components. In the prismatic bed configuration fuel compacts made of TRISO coated particle fuel are embedded in hexagonal shaped graphite blocks. These blocks have passages for coolant to flow. Pebble bed configuration has TRISO coated particle fuel mixed with graphite powder and given shape of a spherical pebble. Coolant flows through the gaps between the pebbles. For the core, an annular geometry was assumed. It was seen that the maximum temperature seen by the fuel was comparable for both of these fuel configuration. Coolant options such as molten lead, molten salt and gaseous medium like helium were considered. Besides these, other criteria such as ease in component handling, irradiation related material and fuel degradation, better fuel utilisation and passive options for coolant flow etc. were also considered. It was important to compare other considerations such as fuel handling, reshuffling, and removal criteria. Prismatic block would need special tooling and reactor outage for their shuffling, removal or new fuel loading. There are also chances of distortions of block due to variable irradiation causing difficulties in handling. In case of pebbles, fuel loading and removal are online with online monitoring facility for burn-up measurement. In case the burn-up seen by the pebble is not as per the design value, it is sent back to the core. This aspect of pebble bed fuel configuration shows a clear advantage over prismatic blocks. Molten salt as compared to helium was preferred due to its feasibility to provide natural circulation based passive cooling. Hence pebble bed reactor core with molten salt-based coolant was selected for further design. Fig. 20 shows schematic of 600 MWth Innovative HTR design. Major research and development for reactor components, coolant technologies, reactor safety, fuel and material development etc., are in progress. Table 2 shows proposed specification for this reactor. Cross sectional layout and schematic of components used for reactor physics analysis is shown in Fig. 21. In order to develop and demonstrate the technologies, a smaller power (20 MWth) reactor is being designed (Singh et al., 2015a) describes design of a fuel pebble for utilizing thorium in 20 MWth IHTR). Analytical modeling for circulating fuel is being developed for the same (Singh et al., 2015b) describes dynamic analysis of circulating fuel reactor). 4.1. Thermal hydraulic and material compatibility studies

Fig. 19. Temperature (degree centigrade) contour of CHTR core during normal operation.

A molten salt natural circulation loop (Dulera et al., 2013) describes thermal hydraulics and material compatibility aspects of molten fluoride salt coolant), as shown in Fig. 22, has been set up to carry out these studies. Fig. 23 shows comparison of analytical and experimental studies. In addition, an experimental facility to study corrosion behaviour of FLiNaK salt on the structural materials has also been set-up. Experiments on various materials using this and other salts have been initiated. This is shown in Fig. 24.

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an optimum mix of the technologies. The S-I process is a three-step process involving formation (Bunsen reaction) and decomposition of hydriodic acid (HI) and sulphuric acid (H2SO4). The decomposition reactions being endothermic need external heat. HTRs are well

Fig. 20. Schematic of 600 MWth IHTR (Sinha, 2013).

Table 2 Proposed specifications of IHTR (Sinha and Dulera, 2010). Reactor power

Coolant outlet/inlet temperatures Moderator Coolant Reflector Cooling mode Fuel Control Energy transfer systems

H2 production

600 MWth for following deliverables (Optimised for hydrogen Production)
Hydrogen: 80,000 Nm3/hr
Electricity: 18 MWe
Drinking water: 375 m3/hr 1273 K/873 K Graphite molten salt Graphite Natural circulation of coolant 233 UO2 & ThO2 based TRISO coated particle fuel Passive power regulation and reactor shutdown systems Intermediate heat exchangers for heat transfer to system for hydrogen production þ High efficiency turbo-machinery for electricity generation þ Desalination system for potable water High efficiency thermo-chemical/High temperature steam electrolysis processes.

Fig. 21. Core cross sectional and components layout.

5. Indian programme on hydrogen production technologies The programme is aimed at development of high efficiency processes of hydrogen production by splitting water. The processes include thermo-chemical as well as high efficiency steam electrolysis. Two promising thermo-chemical processes representing high and relatively lower temperature regimes are the SulphurIodine (S-I) and Copper-Chlorine (Cu-Cl) processes. Presently R & D in BARC is focussed mainly on the more challenging S-I process. Efforts have also been initiated for the Cu-Cl process. In thermochemical processes, water splitting is subdivided into different partial reactions, each occurring at a different temperature. The process is a sequence of thermally driven chemical reactions in which water and heat are the inputs, hydrogen and oxygen are the outputs and the chemicals and reagents are recycled in a closed cycle. Development of materials and technologies is also in progress for High Temperature Steam Electrolysis. Other options will also be considered before a large-scale deployment is planned so as to have

Fig. 22. Schematic of natural circulation loop for molten salt coolants (Sinha et al., 2016).

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Fig. 23. Comparison of analytical and experimental results for steady state heater inlet/outlet temperature at different Power.

Fig. 25. Integrated S-I closed loop of glass.

Fig. 26. Metallic S-I system. Fig. 24. Experimental facility for corrosion studies for molten salt coolants.

6. Conclusions suited for supplying heat to these reactions. Input to the process being only water and heat and output are hydrogen and by-product oxygen. The process is operated in a closed loop, and other chemicals used such as sulphur and iodine are recycled back to the process. Apart from feasibility of the process, efficiency, stability of close loop operation, safety, suitable materials, catalysts and integration aspects with a HTR are the key challenges. BARC has carried out R & D on the basic reactions (Shriniwas Rao et al., 2015) describes study of effect of high pressures and elevated temperatures on Bunsen reaction of S-I thermo-chemical process), development of catalysts for H2SO4 and HI decomposition and has demonstrated closed loop S-I process (Fig. 25) using glass/quartz as material for equipments and produced hydrogen at 30 normal litres/hour. India is the fifth country to achieve this feat. Metallic system (Fig. 26) for demonstrating high pressure operation is being setup for individual sections. In parallel, technologies for High Temperature Steam Electrolysis (HTSE) and Cu-Cl process are also being developed.

Nuclear option for hydrogen production is inevitable due to environmental concerns. A high temperature nuclear reactor can operate in the combined heat and power (CHP) mode to produce electricity, hydrogen as well as potable water. High temperature reactors have large potential to provide necessary high temperature process heat required for this purpose. BARC is carrying out developmental work related to all aspects of high temperature nuclear reactors. The design for CHTR is based on a vision to achieve goals such as high fuel burnup, natural circulation of coolant, compact design with complete passive safety, having practically no moving part in the reactor block and no fluid at high pressure. R & D work encompasses development of fuel and materials, oxidation, compatibility, and corrosion related studies, development of coatings, joining technologies, development of high burn-up - high temperature fuel, studies for irradiation behaviour of fuel and materials, development of characterisation techniques, development of passive systems for high temperature thermal management, and development of systems for reactor safety. These are in various phases of development within BARC and in some other DAE

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List of abbreviations AVR BARC BCRs

Arbeitsgemeinshaft Versuchsreaktor Bhabha Atomic Research Centre Burnable Compensation Rods

HTGRs HTR HTRPM BeO Beryllium oxide/ Beryllia HTSE BET Brunauer, Emmett and Teller technique, HTTR BISO Buffer Isotropic Pyrolytic Carbon IGCAR C-C Carbon-Carbon IGP CHP Combined Heat and Power IHTR CHTR Compact High Temperature Reactor INET Cu-Cl Copper-Chlorine JAEA CVD Chemical Vapour Deposition JAERI DC Direct Current LBE DTA Differential Thermal Analysis LeBENC EDX Energy-Dispersive X-ray Spectroscopy LWR FEM Finite Element Method NFC FTC Fuel Temperature Reactivity Coefficient NPL GCRs Gas Cooled Reactors PCRV HDG High Density Graphite PHWR HPDATA Heat Pipe Design and Transient Analysis PSZ

High Temperature Gas Cooled Reactors High Temperature Reactor High temperature gas cooled reactord pebble-bed module High Temperature Steam Electrolysis. High Temperature Engineering Test Reactor Indira Gandhi Centre for Atomic Research Internal Gelation Process Innovative High Temperature Reactor Institute of Nuclear and New Energy Technology Japan Atomic Energy Agency Japan Atomic Energy Research Institute Lead-Bismuth Eutectic Lead Bismuth Eutectic Natural Circulation Light Water Reactor Nuclear Fuel Complex, National Physical Laboratory Prestressed Concrete Reactor Vessel Pressurised Heavy Water Reactor Partially Stabilized Zirconia

units such as NFC and IGCAR. Design variants of CHTR concept are also attractive as small nuclear power packs for supplying electricity in areas not connected to electricity grid. For IHTR, pebble bed configuration with coolant has been chosen considering feasibility of on-line fuelling and natural circulation based passive cooling. BARC is also pursuing activities for high efficiency hydrogen production by splitting water. SCBC is being developed as an option for high efficiency electricity generation. References Basak, A., Dulera, I.V., Sinha, R.K., 2005. Passive Accident Condition Heat Removal Systems for CHTR, 16th Annual conference of Indian Nuclear Society, 15-18 November, 2005. Mumbai, India. Basak, A., Dulera, I.V., Vijayan, P.K., 2011. Design of high temperature heat pipes and thermosiphons for Compact High Temperature Reactor (CHTR). In: Proceedings of the 21st National & 10th ISHMT-ASME Heat and Mass Transfer Conference. IIT Madras, India, 27-30 December, 2011. Basak, A., Dulera, I.V., Vijayan, P.K., 2012. An Approach for Graphite Component Design for Nuclear Reactors. National Conference on Carbon Materials, November 2012, Mumbai, India. Basu, R., Bhattacharya, S., Bhatt, R., Ahmad, S., Singh, A., Navaneethan, M., Hayakawa, Y., Aswal, D.K., Gupta, S.K., 2014. Improved thermoelectric performance of hot pressed nanostructured n-type SiGe bulk alloys. J. Mater. chem. A 2, 6922e6930. Borgohain, A., Jaiswal, B.K., Maheshwari, N.K., Vijayan, P.K., Saha, D., Sinha, R.K., 2011a. Natural circulation studies in a lead bismuth eutectic loop. Prog. Nucl. Energy 53, 308e319. Borgohain, A., Maheshwari, N.K., Vijayan, P.K., Sinha, R.K., 2011b. CFD Analysis on Heat Transfer in Low Prandtl Number Fluid Flows, 14th International Topical Meeting on Nuclear Reactor Thermal hydraulics, NURETH-14, 2011. Borgohain, A., Srivastava, A.K., Jana, S.S., Maheshwari, N.K., Vijayan, P.K., 2015. Natural Circulation Studies in Lead Bismuth Eutectic Loops. Thorium Energy Conference (ThEC15), October 19-22, 2015, Mumbai, India. Brey H.L., The evolution and future development of the high temperature gas cooled reactor, Proceedings of International Conference on Global Environment, and Advanced Nuclear Power Plants, Paper 1194, (GENES4/ANP2003), September 15-19, 2003, Kyoto, Japan. Dhami, T.L., Mathur, R.B., Dakate, S.R., Aggarwal, R.K., Haldar, S.K., Sangal, A., Dasgupta, K., Sathiyamoorthy, D., 2006. Development of special carbon materials for novel nuclear reactors. In: Bahl, O.P., Tripathi, V.S., Goyal, Pradeep (Eds.), Indo Carbon 2006 Conference Proceedings. Shipra Publication, Delhi, India, pp. 242e254, 2006. Dulera, I.V., Sinha, R.K., 2008. High temperature reactors. J. Nucl. Mater. 383, 183e188. Dulera, I.V., Vaze, V.V., 2009. Fuel for high temperature reactors. IANCAS Bull. Adv. Nucl. Fuels VIII (2). April 2009. Dulera, I.V., Basak, A., Sinha, R.K., 2005. Options for Design of 600 MW(Th) Indian High Temperature Reactor for Hydrogen Production, 16th Annual conference of Indian Nuclear Society, November 15-18, 2005, Mumbai, India. Dulera, I.V., Sinha, R.K., Saha, D., 2009. High Temperature Reactors. Proceedings of International Conference on Peaceful Uses of Nuclear Energy, 29 September - 01 October, 2009, New Delhi, India. Dulera, I.V., Vijayan, P.K., Sinha, R.K., 2013. Indian Programme on Molten Salt Cooled Nuclear Reactor. Conference on Molten Salts in Nuclear Technology, Bhabha

SANS SAXS SCBC

Small-Angle Neutron Scattering Raman, Small Angle X-ray Scattering Super-critical CO2 Brayton cycle

SCRAM SEM S-I TEM THTR TRISO TSDATA UHP UK USA VHTR XRD YSZ

A rapid emergency shutdown of a nuclear reactor Scanning Electron Microscopy Sulphur-Iodine Transmission Electron Microscopy Thorium High Temperature Reactor TRistructural-ISOtropic Thermo Syphon Design and Transient Analysis Ultra High Purity United Kingdom United States of America Very-High-Temperature Reactor X-ray Powder Diffraction Yttria Stabilized Zirconia

Atomic Research Centre, 9-11 January, 2013, Mumbai, India. Dutta, B.K., Vaze, K.K., 2012. BTRISO, Internal Report, 2012. Dwivedi, D.K., Gupta, A., Krishnani, P.D., 2011. Reactor physics design of Compact High Temperature Reactor (CHTR). In: SMiRT Post Conference Seminar on Innovative Fast Reactor Design. IGCAR, Kalpakkam, India. Nov, 2011. Dwivedi, D.K., Gupta, A., Krishnani, P.D., 2015. 3D Space Time Analysis of Anticipated Transient Without Scram in CHTR with Fuel Temperature Feedback. Thorium Energy Conference (ThEC15), October 19-22, 2015, Mumbai, India. Ghanwat, S.J., Sathiyamoorthy, D., 2008. Nuclear Grade Beryllia Ceramic Through Beryllate Process. Proceedings of the National Symposium on Science & Technology of Glass and Glass-Ceramics (NSGC-08), 2008, Mumbai, India. Kanse, D., Khan, I.A., Bhasin, V., Vaze, K.K., 2012. Micro-mechanical Studies on Graphite Strength Prediction Models. National Conference on Carbon Materials November 2012, Mumbai, India. Kaushik, A., Basak, A., Dulera, I.V., Vijayan, P.K., 2012. Effect of Graphite Properties in Thermal Analysis of CHTR. A parametric study, National Conference on Carbon Materials, November 2012, Mumbai, India. AIP Conference Proceedings 1538, 11 (2013). Kaushik, A., Raza, S.K., Basak, A., Dulera, I.V., 2014. Safety Studies for Compact High Temperature Reactor (CHTR); New Horizons in Nuclear Reactor Thermal Hydraulics and Safety (IW-NRTHS), January 13-15, 2014, Mumbai, India. Kaushik, A., Dulera, I.V., Vijayan, P.K., 2015. Super-critical Carbon Dioxide Based Brayton Cycle for Indian High Temperature Reactors. Thorium Energy Conference (ThEC15), October 19-22, 2015, Mumbai, India. Kumar N., Aggarwal S.K., Suryanarayana S., Bamankar Y.R., Vittal Rao T.V., Radhakrishna J., Pai R.V., Deb A.C., Ganatra V.R., Vaidya V.N. and Mukerjee S.K., Preparation of UO2 Microspheres for SGMP by Internal Gelation Technique on 1 Kg/day Scale, XVIIth International Solgel Conference, August 25-30, 2013, Madrid, Spain. Mollick, P.K., Sathiyamoorthy, D., Rao, P.T., Venugopalan, R., Chakravartty, J.K., 2014. Process Modeling and Simulation to Develop UO2-TRISO Coated Particle for Indian High Temperature Reactors. BARC Newsletter- Special Issue, 2014. Padmanabhan, P.V.A., Chakravarthy, Y., Bhandari, S., Vandana, Das A.K., 2013. Plasma Spray Coating of Yttrium Oxide on Graphite Substrates for Compatibility Studies in Molten Metals. Proceedings of the twenty eighth national symposium on plasma science and technology: fusion science and technology - abstract book, p. 136e137; PLASMA, 2013. Panda, K.K., Basak, A., Dulera, I.V., 2015. Design and development of High Temperature Heat Pipes and Thermo Siphons for Passive Heat Removal from Compact High Temperature Reactor. Thorium Energy Conference (ThEC15), October 19-22, 2015, Mumbai, India. Panda, K.K., Basak, A., Dulera, I.V., 2017. Numerical Simulation of High Temperature Sodium Heat Pipe for Passive Heat Removal in Nuclear Reactors. accepted for publication in Nuclear Engineering and Design, 2017. Prasad, C.S.R., Dulera, I.V., 2009. Nuclear Hydrogen Production. Proceedings of International Conference on Peaceful Uses of Nuclear Energy, 29 September - 01 October, 2009, New Delhi, India. Raza, S.K., Kaushik, A., Kelkar, P.P., Dulera, I.V., Vijayan, P.K., 2015. Thermal & Safety Analysis of Compact High Temperature Reactor (CHTR). Thorium Energy Conference (ThEC15), October 19-22, 2015, Mumbai, India. Sathiyamoorthy, D., Rao, V.G., Rao, P.T., Mollick, P.K., 2009. Operational Aspects of Spouted Bed for the Quality Control of TRISO Coated Particles. International conference on Characterization and Quality Control of Nuclear Fuels (CQCNF), 2009. Selvakumar, J., Ramadurai, K., Mollick, P.K., Sathiyamoorthy, D., 2012. Silicon Carbide Thin Film by Induction Assisted CVD: precursor and Deposition Chemistry. National conference on carbon materials 2012, Mumbai, India. Shriniwas Rao, A., Sujeesh, S., Nafees Ahmed, V., Fani, H.Z., Tewari, P.K.,

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Gantayet, L.M., 2015. Study of effect of high pressures and elevated temperatures on Bunsen reaction of Iodine-Sulfur thermo-chemical process. Int. J. Hydrogen Energy 40, 5025e5033. Singh, I., Gupta, A., Krishnani, P.D., 2015a. Study on Design of Thorium Based Fuel Pebble for IHTR-20 MWth. Thorium Energy Conference (ThEC15), October 1922, 2015, Mumbai, India. Singh, I., Gupta, A., Krishnani, P.D., 2015b. A Dynamic Analysis of Circulating Fuel Reactor with Zero Dimensional Modeling. Thorium Energy Conference (ThEC15), October 19-22, 2015, Mumbai, India. Sinha, R.K., Banerjee, S., 2005. Nuclear Energy to Hydrogen. International conference on roadmap to hydrogen energy organised by INAE, 4-5 March, 2005, Hyderabad, India. Sinha, R.K., Dulera, I.V., 2010. Carbon based materials e applications in high temperature nuclear reactors. In: First Asian Carbon Conference (FACC 2009), 25the27th November 2009, 17. Indian Habitat Center, New Delhi, India e Published by Indian Journal of Engineering and Material Sciences, pp. 321e326. October, 2010. Sinha, R.K., Chellapandi, P., Srinivasan, G., Dulera, I., Vijayan, P.K., 2016. GenerationIV Concepts: India, Handbook of Generation IV Nuclear Reactors. Woodhead Publishing Series in Energy, pp. 413e452. Sinha, R.K., 2013. Nuclear power for energy security - Indian scenario. J. Sci. Cult. 79

(1e2), 5e16. January-February. Sure, J., Thyagarajan, K., Mallika, C., Kamachi Mudali, U., 2015. Thermal-cycling behavior of plasma-sprayed partially stabilized zirconia coatings on highdensity graphite substrate. J. Therm. Spray Technol. 24 (6). August 2015. Tewari, R., Vishwanadh, B., Srivastava, D., Dey, G.K., Vaibhav, K., Jha, S.K., Mirji, K.V., Prakash, B., Saibaba, N., 2011a. Development of Nb-1% Zr-0.1%C Alloy as Structural Components for CHTR. BARC Report- BARC/2011/E/020, 2011. Tewari, R., Badgujar, B.P., Vishwanadh, B., Viswanadham, C.S., Dey, G.K., Ali, M., Bagchi, A.C., Saha, T.K., Mirji, K.V., 2011b. Study of Laser and Electron Beam Welding of Nb-1Zr-0.1C alloy. BARC Report - BARC/2011/I/014, 2011. Vetrivendan, E., Rao Ch, J., Mallika, C., Kamachi Mudali, U., 2015. Development of silicon carbide interlayer for deposition of yttria over graphite for uranium melting application. In: Proceedings of International Corrosion Conference and Expo (CORCON 2015). Chennai trade centre, Chennai, India. November 19e21, 2015. Vijayan, P.K., Kamble, M.T., Dulera, I.V., 2013. Nuclear reactors for the future. J. Sci. Cult. 79 (1e2), 17e22. January-February. Vishwanadh, B., Naina, R.H., Majumdar, S., Tewari, R., Dey, G.K., 2013. A study on the oxidation behaviour of Nb Alloy (Nb-1% Zr-0.1%C) and Silicide-coated Nb Alloys. Metall. Mater. Trans. A 44A, 2258e2269.

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