Identified Gap Areas in the Knowledge Domains
SODIUM COOLED FAST REACTORS:
1. 1 SODIUM TECHNOLOGY
(Contact person: Shri M.G. Hemnath, Dr. R. Sridharan)
Cold-trap is used to remove the oxygen and hydrogen which ingresses into sodium system. Figure shows the details of cold trap. A part of sodium from the main loop enters into the cold trap. The sodium will be cooled inside the cold trap. The impurities like oxygen and hydrogen will be precipitated in the wire mesh provided inside the cold trap as oxides and hydrides of sodium. The pure sodium outlet from cold trap will be circulated back to the main loop. In order to achieve higher trapping capacities by efficiently removing sodium oxide and hydrides detailed investigation to understand the hydrodynamics and mechanism for chemical kinetics inside the cold trap are required. The work involves study of the activation energy, order of nucleation and growth, mechanism of crystallization and precipitation for oxides and hydrides of sodium over the cold walls and wire mesh provided inside the cold trap. These provide the inputs for developing the mathematical model to investigate mass transfer of oxygen and hydrogen from sodium so as to improve the design of large capacity cold trap. The large capacity cold trap developed based on mathematical model has to be experimentally validated in sodium.
1.1.2 Investigation of physical and chemical properties of sodium nanofluids for applications
Liquid sodium is the coolant of choice for fast breeder reactors although its chemical reactivity with air/water is one area of concern. Liquid sodium containing nano-metallic particles dispersed in it, called as nano-fluid, has been recently reported to exhibit low chemical activity with no significant impairment to its heat transfer characteristics. Actual choice of the metal for dispersing in sodium depends on its influence on the heat transfer and kinematics characteristics of the nano-fluid as well as on the cost of the dispersed metal. The proposed work involves the identification of the metallic element or its alloys, development of suitable chemical methods for synthesis of this chosen component in the nanocrystalline form (preferably 3- 30 nm), characterization of these nanoparticles using electron microscopy, BET surface area analysis and development of methods for dispersion of the nano powders in sodium. This work would also involve measurement of the various physical parameters (specific heat, thermal conductivity, viscosity etc.) of the nanao-fluid, its durability with time as well as the comparison of its chemical reactivity with water against that of pure sodium with water.
(Contact person: Dr. V. Jayaraman)
Semiconductor oxide are useful in monitoring trace levels of pollutants in air. Development of selective compositions for sensing CH4, I2, NH3, NOx, etc in process streams of our interest would be taken up in the proposed work. This investigation involves the identification of suitable base metal oxide compositions along with dopants and additives to impart selectivity for the species mentioned. Following are the work elements: a) preparation and physicochemical characterization of a series of metal oxide compositions by both soft-chemical and solid state methods b) studying the gas sensing characteristics of them in porous structures c) thin film deposition of select compositions by pulsed laser deposition d) studying the gas sensing characteristics of thin films. Sensor arrays will be built and studied for their characteristics using the select compositions that would emerge from these investigations and those have been established previously.
(Contact person: Shri S. Chandramouli)
PFBR is designed to operate with four failed fuel pins (as gas leakers) in the core. The presence of these leakers implies that the cover gas over the sodium pool would be contaminated with fission gases such as Xe and Kr. The major consequences due to this radioactivity in the cover gas is that the leakage from the reactor vessel into the Reactor Containment Building (RCB) would raise the dose level inside the RCB to levels higher than permitted and would increase the amount of radioactivity being discharged through the stack. The fission gases needs to be flushed out of the reactor plenum and sent to delay tanks and then to Cover Gas Purification System (CGPS) for decontamination. The purified cover gas can thus be safely vented out to atmosphere. However before sending the fission gases to the purification system, it is essential to strip out the sodium vapors in the cover gas because the vapors can condense in relatively cooler locations and choke the lines of the cover gas purification circuit. The concentration of sodium vapor in the cover gas space is about 25g/m3. A sodium vapor condenser is used to remove the sodium vapors from argon. This is a device wherein large surface area is provided in the form of random metal packing, such that the sodium vapors can condense on the packing material. Though the present design of vapor condenser has been effective, there is scope for improving its efficiency. A new design of the sodium vapor condenser has been conceived. The cover gas laden with sodium vapors enters at the bottom of the condenser. The cover gas is stripped off the sodium vapor in the condenser and sodium free argon comes at outlet of the condenser. The condensed sodium drains back to the reactor vessel. Effective removal of sodium vapor from cover gas requires a graded temperature drop along the height of the condenser. The present study aims at analyzing the performance of this new design by developing a mathematical model for the new design and setting up small scale experimental model to validate the mathematical model.
1.2 ALTERNATE COOLANTS FOR FAST REACTORS, ADS AND FUSION
Lead and lead-bismuth eutectic alloys are considered as alternate coolant for fast breeder reactor and are the candidate coolants of the advanced nuclear facilities such as Accelerator Driven Systems (ADS). These alloys are highly corrosive to structural steels but this corrosion can be minimized by forming a passive oxide layer on the steels. An detailed understanding of the conditions required for the formation this passive layer and its stability requires a complete knowledge on data on Pb-M-O and Bi-M-O systems (M=alloying component in steels) including their phase diagrams and thermochemical data on the relevant ternary oxides present in these systems. The envisaged work would involve study of the above mentioned ternary phase diagrams, measurement of chemical thermodynamic data by high temperature, vapour pressure and emf techniques, and development of high temperature oxygen sensors to be employed in these liquid metal systems.
1.3 SODIUM COMPONENTS
(Contact person: Shri B.K. Sreedhar)
In the primary sodium pump of a fast reactor, the rotor is supported in sodium by means of a hydrostatic bearing and above sodium by means of journal bearing (for radial action) and Kingsbury thrust bearing. The bearings above sodium are lubricated by oil. Although chances of any oil leak from the bearing cooling circuit has been nullified by suitable features in design, any oil leak into the primary circuit can result in extended shutdown of the reactor, as has occurred in the Prototype fast reactor (PFR) in the UK. It is therefore preferable to eliminate the oil circuit completely in future designs. Active Magnetic bearings (AMB) make this possibility come true because they are contact less bearings. Other advantages of AMB include (i) the ability to internally process information and thus monitor the health of the machine and take remedial measures (smart machine technology) (ii) absence of mechanical wear. Active Magnetic Bearing (AMB) development for vertical centrifugal sodium pumps is currently underway with M/s National Aerospace Laboratories (NAL), Bangalore. The project, which is for a modest rotor weight of 100 kg – stage 1 of a 3 stage program with the final aim of fabricating a bearing for the prototype pump, is currently in the testing stage. Future works in this area include : (i) Design, analysis and experimental study of auxiliary / catcher bearings for large rotors (ii) Design optimization of Active Magnetic Bearing.
(Contact person: Shri B.K. Sreedhar)
In the primary sodium pump of PFBR, the active argon cover gas is sealed from the atmosphere by means of a triple mechanical seal arrangement. This arrangement consists of two mechanical seals mounted in tandem between the pump cover gas space and the Kingsbury type thrust bearing and one mechanical seal between the thrust bearing and the atmosphere. The seals are colled by an independent oil circuit. Although chances of any oil leak from the bearing cooling circuit has been nullified by suitable features in design, any oil leak into the primary circuit can result in extended shutdown of the reactor, as has occurrred in the Prototype fast reactor (PFR) in the UK. It is therefore preferrable to eliminate the oil circuit completely in future designs. A ferrofluid seal makes this possible. A ferrfluid seal consists of an annular axially polarised permanent magnet in contact with two stationary pole pieces, the magnetic fluid and a magnetically permeable shaft/shaft sleeve. The assembly creates a closed magnetic ciruit with the magnetic force holding the ferrofluid in the gap thus forming a liquid sealing ring that adheres to the polepieces and the shaft surface. Development of ferrofluid seal is in progress in IGCAR. The project is in its infancy and tests have been completed on a 25 mm shaft up to a maximum speed of 2500 rpm and a differential pressure of 100 mbar(g). Topics in this area are: Scaling up and design optimization for large sized shaft : Includes static and dynamic modeling of seal, fabrication and testing of seal, determination of radial stiffness of seal, effect of shaft inclination on seal performance, study of long term dynamic operation, temperature effect etc. on ferrofluid sealing properties, etc.
In a typical pool type Fast Reactor (FR), the Intermediate Heat Exchanger (IHX) penetrates the inner vessel (IV) that separates the hot and cold pools of the reactor. The primary hot coolant flows on the shell side of IHX to the cold pool, exchanging heat with the secondary coolant that flows on the tube side. The IHX is specifically kidney shaped from the considerations of reducing overall reactor block dimensions. A sealing is required at the IHX penetration in IV in order to avoid / minimize the flow of coolant from hot to the cold pool resulting in by-pass of the IHX, which is not desirable. A mechanical seal is preferred option from reactor safety considerations. The complex kidney shaped IHX profile together with the need for accommodating the relative multi-axial movements and tilts between both IHX and IV during various operating states of the reactor poses challenges and necessitates development of a novel design and confirmation through detailed theoretical CFD modeling towards assessing the coolant leakage to meet the sealing requirements.
Vertical sodium pumps use hydrostatic radial bearing at the bottom end and conventional oil lubricated bearing at the top of the shaft. The study requires assessment of different options available for hydrostatic bearing, its modeling, analysis and evaluation of leakage and verifying the analysis using a full scale model in vertical and inclined shaft . Effect of jet formation at the outlet of the bearing is also to be studied by analysis and verified in the model. Further, HSB surfaces are hard faced to minimize the wear during start up and stop. Analysis and experimental study to be conducted to evaluate the wear resistance with different hard facing materials and with different methods / processes for deposition.
Sodium is used as coolant in fast reactors due to its excellent heat transfer characteristics. But one of the problem associated with sodium is it is highly reactive with water. Hence there is always a possibility of sodium-water reaction in the steam generator which causes either a plant shutdown or reduction in the operating power due to non-availability of a steam generator module. The aim of the project is to develop an advanced steam generator concept either from those available in the literature or a totally new innovative concept where there is no risk of sodium-water reaction.
(Contact person: Shri B.K. Sreedhar)
The future FBR’s will be of larger capacity (1000 MWe) than PFBR and the pumps in these reactors will have to handle larger flow rates at higher pressure drops. The size of the pump, however, cannot be increased proportionally as this will increase the capital cost (main vessel dimensions, sodium inventory etc.). In order to limit the pump dimensions it may be necessary to operate the pump with some degree of cavitation. Cavitation in pumps can result in material erosion depending upon the degree of cavitation and time of operation with cavitation. In addition to pumps, the other areas where cavitation can occur are in hydrostatic bearing, foot/outlet of SA. It is therefore valuable to study the relative cavitation erosion resistance of materials presently in use, such as SS304, 304L, 304 LN, 316, 316L, 316 LN, as well as materials planned in future reactors such as Cr-Mo steels, with / without surface treatment, using a vibratory cavitation device. This method will enable accelerated cavitation erosion testing of various materials. It is planned to do the study in sodium under conditions similar to that in the reactor and establish relationship between liquid properties, flow properties, material properties and erosion rate. The work will involve (i) design, fabrication and assembly of a vibratory cavitation sodium test loop (ii) testing of material samples, with/without surface treatment, using an ultrasonic vibratory device, (iii) modeling of the ultrasonic cavitation field, and (iv) examination of eroded specimens and analysis of results to study effect of material and fluid (cover gas pressure, temperature) on cavitation erosion rate.
Fuel handling is carried out in fast rectors under reactor shut down conditions (Offline refueling) and hence the refueling duration directly influences reactor capacity factor. Emphasis is now placed on design of fast breeder reactors with improved economy, improved safety and high capacity factor. To improve the capacity factor, innovative design measures are contemplated in order to reduce the time for refueling. The fuel handling scheme of present reactors envisages handling of a single subassembly at a time. Multiple subassembly handling requires innovative design features of fuel handling machines and suitable scheme has to be worked out.
In the LMFBR Steam generators liquid sodium is used to heat the water to produce superheated steam. Sodium (~0.4 MPa) and water (17.2 MPa) are separated by a single wall tube. In case of a tube leak, water/steam rushes to shell side sodium, causing sodium-water reaction. Sodium-water reaction is an exothermic reaction, and produces sodium hydroxide, sodium oxide, sodium hydride and hydrogen. Sodium-water reaction causes wastage of tubes due to corrosion, erosion and over heating. This causes water/steam leak rate to increase with time leading to further damage propagation and so on till action is initiated through leak detection system. The aim of the project is to develop a code to predict the evolution of leak rate and tube wastage based on experimental results available.
1.4.1 Radiation Damage
Development of the reliable interatomic potentials suitable for large scale simulations is an important requirement for our research works involving large scale MD and MC simulations. At present, we employ semiempirical potential based MD approach. The Finnis Sinclair and the Embedded Atom methods are being pursued for the purpose of generating reliable potentials. Extensions to the formalism that take care of the changes in the immediate environment of the atoms due to the presence of large scale disorder are required so as to carry out simulation on the fly. So an activity on the generation of ab-inito potentials, for materials of our interest, can be a research problem to be pursued by a JRF.
The nanostructured ferritic alloy is characterized by very high number density (~1023–1024/m3) of nanosized (~2–5 nm) oxygen rich Y-Ti-O nanoclusters dispersed uniformly in ferritic matrix. These nanoclusters act as obstacles to dislocation motion thereby reducing creep rate and exhibiting high resistance to radiation damage. Nanoclusters in ODS have core/shell structures in which the core is enriched in Y, O, Ti and the shell is enriched in Cr. Ab-initio DFT calculations of Y, Ti, O, Cr and their clusters in bcc-Fe has been performed to understand the core/shell structure of the oxide nanoclusters. Our calculations show strong binding of Y, Ti, O clusters in the presence of vacancy leading to enrichment of these elements at the core of the nanocluster. The binding energy of the defect clusters increases when we replace Ti with Zr, which could have effect on the dispersion of nanoclusters. Further, the interaction of Cr with Y and Ti is found to be repulsive, resulting in the depletion of Cr in the core of the nanocluster. There is a considerable interest in the Radiation stability of these nanoclusters that require calculations of interactions of the precipitates with radiation induced cascades.
Nanoscale precipitates, such as TiC in austenitic stainless steel reduce void swelling. We have studied variety of steels with different Ti compositions using experimental positron annihilation measurements and followed the TiC precipitate formation in these systems. In general, these metal carbides suffer from carbon off-stoichiometry. We have embarked on computing positron annihilation parameters to substantiate our experimental observations. We have computed self consistent ionic relaxation around vacancies in TiC using VASP DFT code and used atomic superposition method for positron calculations. We have also identified interfacial defects of vanadium carbide as positron traps to explain measured positron lifetime results in 9Cr reduced activation ferrtic-martensitic steel. This fruitful area of research, combining experiments and simulation needs to be pursued further in the context of ODS alloys with Yttria precipitates.
(Contact person: Dr.B.K. Panigrahi)
Correlating the ion damage with neutron damage, calls for controlled irradiation experiments in FBTR and accelerator irradiation. This is in addition to the insight that may be obtained from computer simulation. An experimental programme that utilizes both irradiation sources, followed by investigation of physical properties, such as swelling, changes in microstructure etc.
The response of materials to radiation, leading to deleterious changes in material properties is truly a multi-scale phenomenon. The final state is influenced by processes which straddle huge dynamical range of length- and time-scales. The mechanism through which macroscopic material properties emerge through these scales is ill understood, and there is a World-wide effort towards establishing a multi scale simulation programme, and there is an initiative at MSG in this direction. Expertise in KMC methodology and quasi-continuum methods will be developed. Modeling at the multiscale will be taken up at first to develop the protocols for seamless parameter passing and communication between different scales.
It is proposed to introduce metallic fuel in FBR’s after 2020 to achieve higher breeding ratio. Ternary alloys of U-Pu-Zr, containing 10 wt% Zr has been used in the past. Addition of Zr has two objectives: ( 1) increase in the solidus temperature and (2) to inhibit the inter- diffusion of clad and fuel components, which could form low melting eutectic alloys. Studies carried out on U-Pu-Zr with varying Zr content has shown that Breeding Ratio increases with reducing the Zr content. Optimising the Zr content is a key issue in the development of metallic fuel. It is proposed to examine the phase stability of these alloys as a function of composition and temperature using cluster expansion and Monte carlo methods using ATAT ( Alloy theory Automated Toolkit) in conjunction with ab-initio methods.
(Contact person: Dr. B. K. Panigrahi Dr. Amrthapandian , Dr.M. Vijayalakshmi)
It is well known that void swelling depends sensitively on the evolution of phases in austenitic stainless steels which has a dominant influence on irradiation creep behavior, mechanical strength and ductility. Variations in chemical composition and thermo-mechanical treatment influences the microstructure and in turn void swelling and irradiation creep. Our recent results on D9 and D9I show the strong influence of chemical composition on void swelling behaviour. Thus the solution to this challenge lies in judicious choice of composition and tailoring of microstructures. The thesis work will encompass study of irradiation induced microstructure evolution using ion irradiation and TEM. The modeling of change in microstructure represented in terms of distribution of vacancies, interstitials, impurities defect cluster and dislocations will also be carried out using rate theory approaches.
The evolution of microstructural features under fast neutron irradiation is governed by the synergestic effects of the displacement damage and helium production by (n, a) reaction. Hence a proper simulation of fast neutron induced effects using accelerator requires simultaneous irradiation of the samples with heavy ions (displacement damage) and helium ion implantation (helium generation). A dual beam irradiation facility for carrying out such experiments is being setup at the Materials Physics Division. This facility will make use of heavy ion beams from the tandetron accelerator and helium ions from the 400 kV accelerator that is being installed. The work involves design and fabrication of UHV beam lines and ion optical systems for delivering the beams from the two accelerators on to the target situation inside an UHV irradiation chamber. In addition to the development activities associated with the installation of the dual beam irradiation facility, the scientist will also be actively involved in the experimental research.
Development of the reliable inter atomic potentials suitable for large scale simulations is an important requirement for our research works involving large scale MD and MC simulations. At present, we employ semi empirical potential based MD approach. The Finnis Sinclair and the Embedded Atom methods are being pursued for the purpose of generating reliable potentials. Extensions to the formalism that take care of the changes in the immediate environment of the atoms due to the presence of large scale disorder are required so as to carry out simulation on the fly.
Defect parameters, such as the formation, migration energies, and interaction energies with impurities are crucial for the investigation of evolution of defects. A variety of experimental methods, such a positron annihilation spectroscopy, SIMS are capable of addressing these issues. An integration of these experimental activities with computer simulation would provide insight. At present, we have a programme on DFT based calculations for materials like iron, nickel, chromium and calculated the equilibrium lattice parameter, bulk modulus, defect formation energy, binding energy etc. These have to be extended to alloy systems. The defect parameters from the ab-inito calculations will form an input to rate theory modeling on the evolution of defect structure.
Currently, classical Molecular Dynamics simulations are being carried out to simulate the cascae structure in simple systems such as Fe and Ni, for low PKA energy of 10 keV. These need to be extended to more complex systems, and for larger PKA energies of ~ 100 keV, that requires larger computational effort. An in-house general purpose code for structural pattern recognition of defects in the cascade structure has been developed and successfully used for analyzing damage cascades in bcc-Fe. The studies on cascade structure are of importance not only for alloy systems for development of radiation resistant structural materials, but also in the context of oxides and glasses, in the context of long term radiation waste storage. Several studies in pyrochlore structures are being carried out internationally, and we need to enter into this.
kMC is a technique that enables realistic modeling of the kinetic path followed by the microstructure. The microstructure can be represented in terms of the distribution of vacancies, interstitials, impurities, dislocation lines, grain boundaries, and defect clusters which are identified by their positions. Processes like vacancy diffusion, dislocation climb and kind nucleation, void growth and grain growth are characterized by a sequence of activated events. In the simplest kMC approach, transition state theory is used to identify the set of escape pathways for each of the entities in their present state, and the corresponding rates. The system’s present state is modified by choosing one escape path randomly from the set of pathways, and the time elapsed is incremented in a manner consistent with the average time for escape from that state. An expertise in this area needs to be developed to investigate the time evolution of the spatial distribution of nanoparticles and evolution of microstructure, which is of importance in the context of study of ODS steels.
Nano-dispersions of carbides, nitrides and oxides of transition metals have strong effect on the phase stability of structural steels. Vanadium nitrides constitute the key precipitates for creep strengthening of Fe-Cr steels. DFT calculations are employed to investigate phase stability of alloys and compounds of interest to our fast reactor technology. As an example, investigations on vanadium nitrides reveal that vacancies on the nitrogen sublattice lowers the formation enthalpy of VN and V2N in relation to respective stoichiometric phases, in agreement with experiments. The influence of nitrogen vacancies on the phase stability is, however, not well understood. A study of phase equilibrium of vanadium nitrides is proposed to be continued combining DFT calculations with Cluster Expansion-Monte Carlo methods of alloy theory (ATAT). We have established the in-house expertise for using alloy theoretical methods towards modeling thermochemical properties and phase equilibria in vanadium nitrides.
Secondary Ion Mass Spectrometry offers unique advantages for analyzing depth dependence of concentration of chemical constituents with nanometric depth resolution. This analytical capability allows quantitative evaluation of concentration gradient across interfaces and atomistic transport phenomena in condensed phase. Measurement of diffusion coefficients as small as 10-24 m2/sec is possible using SIMS depth profiling. With the availability of UHV compatible thin film deposition units at TFCS/SND, it is now possible to prepare suitable specimens for diffusion measurements using thin marker layer geometry. In addition, we can also use low energy (200-400 keV) heavy ion bombardment to study Radiation Enhanced Diffusion (RED). A new SIMS machine is under procurement.
This Medical cyclotron facility coming up at VECC can provide a proton beam of 30 MeV, 350 mA, and a Materials Science Beamline is being established. With this facility, radiation damage up to 10 dpa, at a rate of 2x10-5 dpa /S within a specimen of ~ 1mm can be produced. This will be of value in carrying out irradiation experiments in bulk state of alloys of interest to FBRs, viz., D9, D9i, Ferritic Steels etc. Further, investigations on damage rate dependent phenomena, insitu experiments, post irradiation experiments and correlation of microstructure with mechanical properties can be carried out.
With the coming in of Grid Computing and National Knowledge Network, a programme to keep abreast with the latest developments in parallel computing facilities need to be in place.
While the above topics have emphasized on computer simulations, with emphasis on material properties, it may also be worthwhile to nucleate an activity on Modeling of complex phenomena and processes. A group of scientists, with a mathematical bent of mind, may be encouraged to look at complex, non-linear, coupled models that may have applications in diverse areas from sensor networks, propagation of signals etc. This activity will place more emphasis on the development of physically motivated models, and less on large scale computations.
An array of sophisticated experimental facilities are becoming available at various synchrotron facilities around the world. These can be used for studies on structure and microstructure of materials of our interest. For example, the 3D-XRD beam line at the ESRF facility, opertated by RISO laboaratories. Here, experiments can be carried out at elevated temperatures that provide a unique opportunity to investigate in-situ, the transformations in a individual grain at elevated temperature, and the associated nucleation of nano precipitates. The recent developments in Anomalous Small-Angle X-ray Scattering techniques at BESSY Synchrotron, operated by HMI opens up new opportunities to analyse the composition and structure of nano-sized precipitations in materials ie it gives element sensitive the structure of nano sized phases in materials, Information concerning sizes, size distributions, surface properties and volume fractions and element sensitive composition fluctuations can also be obtained from the measured set of scattering curves near below the absorption edges of elements containing in the samples.
The three-dimensional (tomographic) atom probe (3DAP) combines time-of-flight type of mass spectrometry with a multi-anode detector. This instrument permits the elemental reconstruction of a small volume of metallic material of the specimen with near atomic resolution. It enables a three-dimensional reconstruction of the analysed volume in real space which are essential for obtaining internal structure of nano precipitate. Initially, collaboration with HMI can be considered, after which we can set up these facilities at IGCAR.
1.4.2 In-core Materials
220.127.116.11 Creep anisotropy studies on ODS cladding as a function of chemical composition and particle morphology studies on both high chromium ferritic ODS and austenitic SS (Biaxial creep) (Contact person: Dr. K. Laha)
To increase the burnup of the fuel in sodium cooled fast reactors (SFRs), it is necessary to develop advanced clad and wrapper materials that are (i) more resistant to irradiation induced void swelling and (ii) having adequate creep strength. High chromium (9-12 wt %) ferritic martensitic steels have inherent void swelling resistance unlike austenitic stainless steels (SS) like 316 SS and 14Cr-15Ni-Ti grade SS. For clad application, the creep strength of (9-12 wt %) ferritic martensitic steels is not adequate at temperatures of 873 and higher although, it has very good swelling resistance. High chromium ferritic steels derive their creep strength from (i) solid solution strengthening by Mo and W, (ii) precipitation strengthening from intergranualr and intragranular precipitates and (iii) transformation induced dislocation substructure. Microstructural instability resulting from coarsening/dissolution of precipitates, reduction in solid solution strengthening due to precipitation of intermetallic precipitates and recovery of dislocation substructure lead to decrease in the creep strength of ferritic martensitic steels drastically at temperatures greater than 873 K. Creep strength of (9-12 wt %) ferritic martensitic steels can be increased substantially by oxide dispersion strengthening (ODS) with highly inert refractory metal oxide particles introduced through powder metallurgy route. There are several challenges in the areas of production and characterisation of ODS clad tubes and understanding of the creep properties of ODS steels. These include (i) Understanding the influence of particle size, morphology and distribution on creep deformation and rupture strength behaviour. (ii) ODS steels have generally a dual phase structure consisting of tempered martensite and small amount of d-ferrite. The effect of d-ferrite on the creep deformation and rupture strength and the optimization of d-ferrite content is not known. (iii) Anisotropy in creep rupture strength and ductility. (iv) Formulation of creep constitutive equations for design.
Austenitic stainless steels are being used as clad material for fast reactors world wide. Ferritic steels and ODS alloys with enhanced void swelling resistance are being developed as potential clad materials for fast reactors to increase the burnup. Austenitic stainless steel clads having excellent corrosion resistance in boiling nitric acid, do not pose any problem during dissolution process of the PUREX spent fuel reprocessing. However, the ferritic steels with lower chromium content may pose heavy dissolution of the clad and affect the purity of the feed for solvent extraction. The present study aims to study in detail the corrosion behaviour of ferritic steels and ODS alloys in boiling nitric acid, with different microstructural conditions simulating the radiation exposure and higher burn up in the FBR. Understanding the deleterious effect of ferritic steel clad in FBR dissolution condition of reprocessing and mitigation of such effects is the objective of the work.
ODS alloys for FBR clad tubes are produced through powder metallurgy route and the typical size of oxide particles in these alloys are of the order of 3-5nm. Fusion welding processes are not recommended for joining clad tubes made from ODS alloys to alloys to end plug as the oxide particle would dissolve in the matrix thus making the weld considerably softer than the base metal. Pressure resistance welding and magnetic-pulse welding are the two processes currently adopted for welding of ODS clad tubes to end plug. Hence two separate PhD topics can be proposed on this subject. (a) Procedure development for welding ODS clad tube to end plug using resistance pressure welding and characterization of the weld joint. (b) Procedure development for welding ODS clad tube to end plug using magnetic pulse welding and characterization of the weld joint. The research methodology would involve the following: 1. Study of effect of welding parameters on the quality of the weld joint. 2. Microstructural characterization of the fusion zone and HAZ. 3. Development of welding procedure and inspection methods for end cap welding. 4. Preparation of acceptance criteria for the joint. 5. Characterization of the weld joint for its mechanical properties
(Contact person: Dr. M.D. Mathew)
(i) Effect of intermediate heat treatment temperature on out-of-pile creep behaviour of 20% CW D9I.
(ii) Effect of cold work on creep properties of D9I cladding tubes
The selection of clad and wrapper materials for sodium cooled fast reactors is primarily based on their resistance to neutron irradiation induced void swelling and this limits the achievable burn-up of the fuel. Thermal creep resistance is also an important design requirement for the fuel clad tubes. For the PFBR, alloy D9 SS, a 14Cr-15Ni-Ti modified SS, has been selected as the clad and wrapper material for the initial cores and the target burn-up is 100,000 MWd/t. In order to achieve higher fuel burn-up (upto 150,000 MWd/t) in subsequent cores of PFBR, an improved version of D9 SS, called D9I SS, has been developed by optimising the amount of phosphorous, silicon and titanium in D9 SS wrt to vosid swelling and creep strength. D9I SS will be used in 20% coldworked condition for better swelling resistance. Conventionally, this steel is given a solution treatment in the temperature range of 1050-1150 C before the final tube making operation which consist of cold drawing to achieve 20% coldwork. This solution treatment temperature does not dissolve the primary titanium carbides which form during solidification of the melt. These carbides being coarse, do not contribute to the creep properties or swelling resistance of the material. On the other hand, the amount of titanium in solution is reduced and thus the amount of fine secondary titanium carbides produced during subsequent tube processing and during service, is reduced. Fine titanium carbides are beneficial for swelling resistance and creep strength. The scope of the study includes: 1. Studying the effect of cold work on creep properties of D9I SS cladding tubes so as to optimise the degree of cold work. The creep studies will be carried out over the range of 12 to 25 % cold work at various creep test conditions. 2. Understanding the effect of intermediate heat treatment temperature on out-of-pile creep behaviour of 20% CW D9I. D9I SS will be subjected to 2-step heat treatment consisting of a higher temperature treatment to dissolve the primary carbides and a lower temperature treatment to re-precipitate fine titanium carbides and the heat treatment conditions will be optimised. Material treated with the optimised conditions will be subjected to 20% cold work and the creep properties will be evaluated at various test conditions.
(iii) Weldability of 20% CW D9I cladding to end plug
(Contact person: Dr. A.K. Bhaduri)
(i) Oxide dispersion strengthened (ODS) steels have emerged as candidate materials for structural applications in future fast reactors. The basic powder metallurgy process for synthesis of a 9Cr-ODS ferritic alloy and the manufacturing process for a clad tube of dimensions required for the PFBR have been demonstrated. It is now required to optimise the composition with respect to both major (Cr content) and minor alloying elements and dispersoid content. Studies on high-Cr varieties as well as austenitic grades of ODS are already under way. An essential component in this exercise is to evaluate the stability of the strengthening components, namely, the nano-sized dispersoids under long-term ageing and irradiation conditions through microstructural and mechanical tests. The stability of nano-sized phases depends on a number of factors including dispersoid stoichiometry, solutes in matrix, dispersoid - matrix interface structure and strain. It is proposed that a comprehensive high spatial resolution structural imaging (HRTEM) combined with advanced chemical imaging techniques using transmission electron microscopy study will be carried out on selected compositions under specific heat-treatments together with extensive simulations to examine the effect of composition, processing and environmental parameters. The study will result in an improved understanding of the basic materials science relating to nanostructures in engineering materials and contribute to improved design principles for advanced materials.
(Contact person: Dr. M. Vijayalakshmi)
Titanium modification of the standard AISI 316L austenitic stainless steel has been demonstrated to considerably delay the onset of void swelling under fast reactor operating conditions in addition to other favourable properties, and hence has been the material of choice for the PFBR. The suitability of the alloy for core structural applications is due to a uniform distribution of fine-scale titanium carbide precipitates in the austenitic matrix that serve to trap vacancies at the precipitate - matrix interfaces. This point-defect -precipitate interaction is critically dependent on the structure of the precipitate - matrix interface and the stoichiometry of the precipitate. The long-term performance of the alloy in the reactor is thus dependent on the stability of these nanoscale precipitates with respect to phase structure and crystallite size, which in turn, is also decided by the interfacial structure and chemistry. Composition tailoring of Alloy D9 has been envisaged to gain property enhancements with respect to void swelling behaviour, microstructural stability, tensile and creep properties. Such compositional changes will affect the kinetics of precipitation and the stability of microstructure. It is also known that under neutron irraidation at the reactor operating temperatures, irradiation induced precipitation of carbide, carbonitrides, phosphides can occur, depending on the alloy composition with respect to minor alloy elements. In order to assess the efficacy of such compositional adjustments, a study of the effect on interfacial composition and structure is required. Thus, it is essential to conduct a comprehensive high spatial resolution study on the structure and chemistry of interfaces between various nano-scale precipitates such as titanium carbide and iron phosphide, and the austenite matrix. It is proposed to carry out detailed high-resolution TEM studies using advanced structural and chemical imaging techniques on a range of alloys under various long-term thermo-mechanical treatments. Such a study will lead to a better selection of an engineering alloy based on sound materials science principles.
1.4.3 Out of Core Structural Materials
Defects are unavoidable during welded fabrication. Various NDT techniques are employed to ensure that these defects are detected and only welds are acceptable quality is put in service. However, it is possible that some of the defects can remain undetected. Further, defects which are accepted for various reasons (size is within the acceptable limit, repair is difficult, inspection is not possible etc.) may have adverse effect on the performance of the joint. Hence, it is proposed to study influence of weld defects on the performance of the joint using fracture mechanics approach. Defects of various size and shape would be produced in the weld and effects of these defects on various modes of loading (fatigue, creep) would be studied using fatigue and creep crack growth studies. It is proposed to conduct two separate studies on austenitic stainless steel 316LN and Grade 91 steel weld joints.
Grade 91 steel is supplied and used in normalized and tempered condition. The normalizing heat treatment currently specified for this steel is 1040-1095C/1h and the tempering heat treatment is 78015C for 1h for 25 mm thickness with a minimum duration of 60 minutes. This means material with different thicknesses would be subjected to tempering heat treatment for different durations. Further during fabrication of the components, it is quite possible that the component may undergo some cold work and this cold work can affect the properties of the material. Components subject to more cold work (>10%) has to undergo normalizing and tempering heat treatment again. Thus components made of Gr. 91 used in SFR would have undergone different heat treatments depending on their thickness and fabrication history. In this project, it is proposed to study effect of various normalizing and tempering heat treatments on mechanical properties (tensile, impact and creep) of the Gr. 91 steel. It is also proposed to study the effect of cold work on the properties of normalized and tempered Gr.91 steel. Project would also involve study microstructural evolution in the steel during heat treatment and creep testing.
The post weld heat treatment (PWHT) currently specified for P91 steel weld joints is at 76015C with a duration of 2.5 min/mm with minimum duration being 30 min. Hence, PWHT given to the weld joint would vary depending on the thickness of the weld joint. Further, for large components, in which materials of different thicknesses are used, PWHT is determined by the weld joint with largest thickness. Often, if there is a long delay between the welding and final PWHT for the component, localized PWHT may also have to be carried out. Under such situation weld joint would undergo more than one PWHT and total duration would exceed more than the minimum specified. Though the PWHT specified currently for FBR components is 760C, most of codes permit PWHT to be carried out at lower temperatures (ASME section III specify 675-760C). Hence, there is a possibility of conducting PWHT at various temperatures for different durations during fabrication of components from P91 steel. Effect of these variations on the mechanical properties of the base metal and weld joints of Grade 91 steel is not readily available. In this proposed project a systematic study on the effect of PWHT duration and temperature on mechanical properties (tensile, impact and creep) properties of the Grade 91 steel weld joint would be carried out. These properties would be correlated to the microstructural changes in the weld joint. Further, variation in PWHT time and temperature on long term service performance of the weld joints would be assessed.
Like any other Cr-Mo steels, Grade 91 steel is also susceptible to Type IV cracking, fracture in the HAZ during creep testing or in high temperature service with a rupture life considerably lower than that of the base metal. Width of the HAZ and welding groove angle are among many factors that influence the Type IV cracking. Hot wire GTAW process, used for the fabrication of the steam generators of FBR produces a narrow HAZ and fusion boundaries that are almost straight. Both these are known to improve the creep strength of the weld joint, though the fracture is expected to take place by Type IV mode. It is proposed to conduct creep test on Grade 91 steel weld joint produced by hot wire narrow gap GTAW process and compare the performance of the joint with those produced by conventional GTAW or SMAW process. It is also proposed to determine the strength reduction factor for weld joints produced by narrow gap GTAW process, which can be used in design of the components made using this welding process.
18.104.22.168 Material design data for 316LN SS with higher nitrogen content
(i) Effect of nitrogen on creep behaviour (Contact person: Dr. M.D. Mathew)
(ii) Effect of nitrogen on fatigue behaviour (Contact person: Dr. R. Sandhya)
(iii) Effect of nitrogen on creep-fatigue interaction diagram (Contact person: Dr. R. Sandhya)
Creep, low cycle fatigue (LCF) and creep-fatigue interaction (CFI) are major considerations in the design of high temperature structural components of sodium cooled fast reactors. High temperature structural components of PFBR are made of 316L(N) SS containing 0.02-0.03 wt% carbon and 0.06-0.08 wt% nitrogen. The design life is 40 years. In order to increase the economic competitiveness of fast reactors, there is a strong desire to increase the design life from the current level of 40 years to 60 years. Efforts are therefore underway to develop a high nitrogen grade of 316LN SS. The influence of nitrogen on LCF, CFI and creep properties will be studied using 316LN SS containing four different nitrogen levels, namely, 0.07, 0.11, 0.14 and 0.22 wt% with a view to optimise the nitrogen content in terms of these properties. The scope of the project involves: 1. Characterisation of the creep properties at different temperatures on all the 4 heats with different nitrogen contents, microstructural and substructural studies to understand the creep strengthening role of nitrogen, and develop creep life prediction models. 2. Characterisation of the low cycle fatigue and creep-fatigue interaction behaviour of all the 4 heats with different nitrogen contents, study the effect of hold time, dynamic strain ageing behaviour, understanding the fatigue strengthening role of nitrogen, and develop fatigue life prediction models.
Most of the predictive softwares used for determination of diffusion distances in multi-component systems, like the nuclear structural materials, require the interatomic potentials, which are not available for all constituents of the multi-component systems. Hence it is proposed to evaluate experimentally these values as useful inputs to prediction of mass transfer in dissimilar joints in nuclear systems.
The processing of advanced austenitic and ferritic steels consists of a number of thermomechanical treatments at various stages of fabrication. The formation of texture during deformation and transformation or recrystallisation can influence the properties of the steel. Using Electron Back Scattered Diffraction the different types of texture will be studied. Modeling will be used as a tool to predict the texture evolution in these systems.
1.5 THERMAL HYDRAULICS
In fast reactors, argon gas is maintained above primary sodium pool to accommodate volume changes in sodium and to avoid sodium-air contact. Agitation of free surface due to large sodium velocities can lead to entrainment of argon gas within sodium and its transportation to core through IHX. Other sources of gas entrainment in primary sodium are main vessel cooling system with an overflow weir with large plunging velocity and rotating pump shaft which is partially submerged in sodium. Passing of argon through the core leads to reactivity oscillations. Hence, adequate care needs to be taken in design to avoid argon entrainment, in terms of adequate submergence of IHX window, low free surface velocity, optimum overflow height for weir and suitable vortex breakers in pump vessel. The objectives of the research are: (i) development of Computational Fluid Dynamic models for gas entrainment, (ii) experimental validation of these models, (iii) investigation of various gas entrainment mechanisms, and (iv) assessment of various mitigation devices for their effectiveness.
1.5.2 Decay Heat Removal System - Investigation of decay heat removal in Na-air heat exchanger under high wind conditions, incorporating appropriate model for the intermediate loop (Contact person: Shri P. Selvaraj)
In PFBR to remove the decay heat a dedicated system called safety grade decay heat removal system has been provided. This system consists of a sodium to sodium heat exchanger and a sodium to air heat exchanger linked by an intermediate circuit. Heat generated in the core is transferred to the intermediate circuit and the air heat exchanger transfers heat from the intermediate circuit to air. A tall stock on the top of air heat exchanger provides sufficient buoyancy head for the air flow. Since the air flow is by natural circulation the changes in wind conditions at the inlet of air heat exchanger and at the outlet of stock influences the air flow into the heat exchanger. The heat exchanger capacity depends on the air flow. The effect of wind conditions, especially under cyclone on the air flow and hence the heat exchanger capacity need to be evaluated incorporating appropriate model for the intermediate loop.
In the primary sodium system of a FBR, sodium streams of vastly differing temperatures emanate from a large number of subassemblies and mix in a sodium pool. The outlet temperatures of these streams are continuously monitored by thermocouples supported by control plug. The accuracy of measurement of these temperatures depends on the location of the thermocouple, flow and temperature profile at the outlet of subassembly and dilution by the flow from the neighboring subassemblies. At the fuel – breeder subassembly interface, the temperature difference between the sodium streams is very large. The control plug makes the flow at the core periphery nearly radial whereas the flow comes out of various subassemblies axially. Apart from this, sodium streams from various subassemblies are also subjected to oscillation due to turbulence and jet instability. Hence, the thermocouples at the core periphery are located in a region of large flow fluctuations. Detailed knowledge of temperature fluctuations is essential for design of core monitoring system and control plug.
In a fast reactor, argon cover gas is maintained above sodium pools to accommodate temperature dependent sodium volume changes and also to serve as blanket between sodium and ambient air. The sodium free level (interface between sodium and argon gas) is susceptible to fluctuations either if the sodium has large surface velocity or the diameter of the vessel is large. There exists significant temperature difference between liquid sodium and cover gas. Thus, the free level is a region of large axial temperature variation in the vessel walls as well as in structures partially immersed in sodium. Due to free level fluctuations, components partially dipped in sodium will alternately see hot sodium and cold argon, leading to thermal fatigue. Detailed investigation of this thermal hydraulics / thermo-mechanical phenomena is important.
(Contact person: Dr. K. Velusamy)
Deviations in the coolant flow in a fuel subassembly are possible by internal blockages inside the subassembly which may be due to the entry of foreign particles or due to the defect in the fuel pin components giving rise to loose particles which may block the flow partially. It may cause local sodium boiling, dry-out and melting of fuel pin and cladding.
A good knowledge on the local blockage becomes essential for safe operation of the reactor. The aim of these studies is to
· Estimate the temperature profile behind the blockages as a function of flow, power, leakage flow through the blockage and inlet temperature,
· Establish the margin between the onset of boiling and onset of dry-out behind the blockages and characterize the types of boiling events,
· Find out the limitation of the detection limits of blockages in relation to the measuring devices
· Demonstrate whether the blocked subassembly can be cooled during power operation and by natural convection.
Fast breeder reactor core contains nuclear heat generating fuel in the form of pellets which are packed into slender leak tight tubes. A group of such pins are bundled together in a hexagonal tube which forms a subassembly through which the coolant sodium is forced for heat removal. The pins are separated by spacer wires which are helically wound around all the pins individually. The clad material, which is usually an austenitic SS, forms the first physical barrier for the radioactive fission products and hence its integrity is important. From safety consideration, the subassembly should never become devoid of coolant flow. Otherwise fuel temperature would increase damaging the clad which might propagate to neighbouring pins before the reactor protection system comes into play. Even though design measures are provided in the subassembly to preclude flow blockage, following a defense in depth principle, it is postulated that the event of total instantaneous blockage of a fuel subassembly should not lead to total core melting. Hence, it becomes necessary to estimate the rate of propagation of damage in order to ascertain the time available to safely shutdown the reactor and to establish the safety margin. This phenomenon calls for detailed thermal hydraulic modeling of affected fuel subassembly along with its internals, flow through SA and consequent coolant temperature increase due to fission heat generation and reactivity effects, increase in the temperature of fuel and clad due to blockage, melting of fuel / clad, boiling of sodium, relocation of molten fuel / clad material within the affected subassembly, heating up of the neighboring subassemblies and the response of the core monitoring thermocouples. This is a coupled transient thermal hydraulic analysis including the neutronic effects.
(Contact person: Shri U. Parthasarathy)
Decay heat exchangers (DHX) are dipped coolers located in hot pool of a fast reactor. When primary sodium pumps are not in service, sodium flow through the core is by natural convection. Under this condition, cold primary sodium exiting DHX-outlet windows flows down to the bottom of the hot pool due to adverse buoyancy. This cold sodium penetrates the narrow inter-wrapper gaps between the various subassemblies (SA) and absorbs a significant amount of decay heat generated in the core. The interaction among (i) cold primary sodium exiting DHX, (ii) stratified hot pool, (iii) inter-wrapper flow (IWF) and (iv) core flow through SA, is a complex thermal hydraulic process. This is a multi-scale transient phenomenon involving IWF characterized by a few mm and hot pool characterized by a few meters with sodium flow changing from forced convection to natural convection. Quantification of benefits offered by IWF is very important for safety assessment of the reactor. The objectives of the research are: (i) characterization of IWF for friction factor and Nusselt number, (ii) development of thermal hydraulic models coupling IWF and flow through SA, (iii) coupling these models with commercial CFD codes for pool hydraulics, and (iv) quantification of IWF for various events in the reactor.
Due to difference in heat generation rates of fuel, breeder and storage subassemblies, temperatures of sodium jets leaving these subassemblies are vastly different. These streams mix in hot sodium pool of the reactor. Random temperature fluctuation at the interfaces between these streams, arising out of jet instability, combined with large heat transfer coefficient of sodium lead to temperature fluctuations in the structures. This phenomenon, known as thermal stripping, results in high cycle fatigue and crack initiation in the structures. Frequency and amplitude of temperature fluctuation are important parameters governing the damage to the structures. There is a need to characterize the thermal stripping behavior in terms of jet velocity, temperature difference between jets, distance of structures from jet, size of jets etc. to design the reactor systems with suitable protection.
In once through type steam generators water enters as single phase liquid (sub cooled) gets evaporated and superheated and leaves as super heated steam. In these types of steam generators two phase length and superheater length co-exists along with single phase liquid length. These types of steam generators are prone to flow instabilities. Typical instabilities are static or Ledinegg instability and dynamic or density wave oscillations. To avoid failures the operating conditions of the steam generator has to be selected such that it is in the stable operating region. The aim of the project is to develop a code to predict stable regime of operation of steam generator under various operating conditions including low pressures, low flow rates and low powers.
One of the severe accidents considered in FBR design is ‘complete loss of pumping power associated with failure of all the shutdown systems’. This scenario is considered as defense in depth in spite of the provision of three different classes of power supply for pumps and two diverse shutdown systems. During this event, coolant flow through the core reduces resulting in sodium boiling within fuel subassemblies. Coolant boiling will generally start from the top of the core and propagate to the bottom. Two phase flow through the subassemblies results in increase in the pressure drop and hence further deterioration of flow rate in the boiling channels. Sodium boiling is associated with strong reactivity effects resulting in power changes and ultimately leading to a large power excursion in the core known as ‘core disruptive accident’. In order to simulate the entire scenario of the event for evaluating the consequences thermal hydraulic models for the core with multiphase momentum and heat transport behavior have to be developed and validated against representative sodium experiments.
Hydrogen is generated due to sodium-water reaction in the steam generator of LMFBR. There is a possibility of leakage of this hydrogen into the steam generator building in case of a failure/crack/hole in the steam generator shell. This hydrogen mixes with the air in the steam generator building. Since hydrogen density is lower it can start accumulate at some top portion and explode at some point of time. The aim of the project is to develop a model for estimating the pressure and temperature rise inside the steam generator building due to hydrogen vapour cloud explosion. The pressure and temperature rise are need to be calculated for different volumes and concentrations of hydrogen air mixture.
One of the severe accidents considered in reactor design is the uncontrolled withdrawal of a control rod with failure of all the shutdown systems. During this event, reactor power increases leading to fuel melting. Melting starts at the middle of active core and propagates towards top and bottom of the core. The clad, inside which the fuel pellets are packed, will generally be able to maintain its integrity due to the availability of cooling at its outer surface. Because of the hydrodynamic effects caused by the fission gas released from the molten fuel, the melt is pushed up and down (fuel squirting) to less reactive locations of the core. This introduces negative reactivity and power reduction leading to a new steady operating state. The fuel squirting behavior depends on operating power and burnup conditions of the core. Numerical simulation of this phenomenon to demonstrate reactor safety requires models that can simulate fuel melting and its relocation under the action of hydrodynamics effects.
1.6 STRUCTURAL INTEGRITY
The liquid free surfaces in large vessels having many internals undergo complicated motions under seismic base excitations. Typical example is the behaviour of sodium free levels in pool type fast reactors. While free vibration of surfaces can be formulated based on linear aquatic wave propagation theory, the motions under seismic excitation are non-linear and random in nature, for which limited publications are available. The scope of the project is formulation of governing differential equations, solutions using analytical (simple cases), numerical and experimental techniques. Model tests can be carried out on the 10 t capacities 3x3 m shake table and also 100 t capacity 6x6 m shake table facilities available at IGCAR. The effects of internals should be quantified distinctly.
1.6.2 Stability of Thin shells under seismic loadings – Development of a code for determining the dynamic stability of thin shells with fluid structure interactions (Contact person: Dr. P. Chellapandi)
Large diameter thin walled shell structures, in particular with large inertia due to liquid filled in have to be investigated critically for the seismic loadings. They undergo buckling under seismic forces, and moments as well as dynamic pressure experienced at the fluid – structure interfaces. These apart, they would also be unstable under the special mechanism called ‘Parametric instability’. These two major instabilities should be investigated in this project. The work contents of this project are determination of seismic induced pressures, forces and moments, formulation of buckling and parametric instabilities problems and analytical and numerical solutions. The results have to be validated using shake table tests. Two tables : 5 t capacity 3x3m size and 100 t capacity 6x6 m size, available of IGCAR can be used.
(Contact person: Dr. P. Chellapandi)
The primary sodium pump in a pool type fast breeder reactor has a long shaft with thrust bearing on the top shield and pocket type hydrostatic bearing (HSB) at the bottom immerged in sodium. The pump has a variable speed drive to operate over the speed range 200-600 rpm. The critical aspect is to ensure that pump would not seize, i.e no mechanical interactions between stationary and rotating parts in the vicinity of HSB under all operating speeds. This implies than there should be liquid film all the time. This requirement becomes critical under seismic induced forces when pump operates at low speed. The academic components of the project are: formulation of governing differential equations for HSB based on fluid dynamic principles, deviation of bearing stiffeners and damping as a function of speeds and seismic excitations, formulation of instability – equations, numerical and analytical solutions and finally validation by shake table tests. A special facility built for validating the stiffeners and damping characteristics of full scale HSB and also shake table tests facilities to investigate the stability under combined resonance and whirling phenomena can be employed for the experimental works.
(Contact person: Dr. P. Chellapandi)
A few literature addresses above problem combined action of pressure and axial force may lead to either plastic collapse characterized by symmetric deformation pattern and buckling characterized unsymmetrical deformation. Further the buckling pattern under fast transient loadings are characterized by many number of lodes resulting in higher buckling strength. The formulation and solution of large size torispherical loads subjected to static and transient loadings are the scope of the project. The existing computer codes can be used for the static buckling. However, the problem for transient loadings need to be formulated a fresh. The facility is available at IGCAR for carrying out static tests. Tests can be conducted for generating transient loadings for which facilities can be built. The object of the project is to develop a new code rules for the same.
(Contact person: Shri S. Jalaldeen)
Fast reactor structures are characterized by large diameter thin walled shells, large liquid mess effects, presence of confined liquid in the narrow annular spaces, sloshing of various liquid surfaces and strong coupling between fluid and structures. The main sources of vibrations are mechanical forces induced by pumps and fluid dynamic forces induced by the self excitation of fluid and structure systems. The investigation of vibrations of various shells in the pool type reactor calls for identification / quantification of various fluid forces and pump induced forces, random analysis formulation of dynamic responses in frequency domain and validation of results by carrying out a few characteristic tests.
Ratcheting referred here is a progressive deformation caused by moving sharp temperature gradients along the shell surface. The problem is complicated with possibility of thermomechanical coupling and creep effects. Till date, there is no robust material constitute models are available. Hence, development of constitute models, formulation of ratcheting problems, identification of analytical to numerical solutions by introducing the complexities sequentially and finally experimental validations are the main scope of the project. A dedicated test facilities built at IGCAR for simulating thermal ratcheting in scaled down models of FBR shells can be effectively utilized. The project ultimately is aimed at recommend improved design rules for thermal ratcheting of thin shells relevant to FBR to be incorporated in RCC-MR, the French design code. The material of construction is austenitic stainless steel type SS 316 LN.
For the large diameter stainless steel pipes of SFR, leak before break (LBB) argument is applied comfortably. However, for the smaller diameter pipes, with the best available leak detection system, it is difficult to justify the LBB argument. It is required to quantify the possibly smallest diameter pipes for sodium application for which, LBB can be applied confidently is the main scope of the project. The project calls for many advanced fracture mechanics concepts such as developing charts for asymptotic crack lengths, tearing instabilities, leakage areas and leak rote estimations. These apart, since pipe lines are operating at high temperature for which creep effects are important, appropriate creep crack growth models need to identified and fine tuned. The creep and fatigue growth behaviour under compressive loads and negative rations are front-line areas in the domain of fracture mechanics. The project involves extensive experimental works, for which the test facilities for fatigue and creep crack growths, tearing instabilities and collapse load estimations, which have been built by putting efforts over a decade can be used. The object is to develop new rules to be incorporated in RCC-MR: A16.
1.7 PLANT LIFE
Under sodium ultrasonic NDT is similar to under water inspection. However major differences such as high temperature, lack of direct visibility, wetting and coupling parameters, and electrically conductive environment have to be studied in detail to improve confidence in under sodium ultrasonic inspection. Effect of each of these aspects must be studied to understand and characterize the ultrasonic signals encountered in under sodium inspections.
An important aspect in ultrasonic inspection is the local visibility which will help in navigation and correct placement of the probe on the component for scanning for possible defect detection. Thus there is a need for compact under sodium ultrasonic viewing system. The imager thus developed should be suitably integrated with corresponding remote controlled manipulator The ultrasonic viewing system, must operate in very harsh conditions including high temperature (160oC – 450 °C), thermal gradients, chemical activity of liquid metal and strong gamma radiation (up to 30 kGy/h). These conditions significantly determine the design/operation/performance of the ultrasonic visualization system. Detailed study of these parameters will help in assessing on the resolution, visibility and dimensional reliability of ultrasonic imaging system.
316L(N) SS is the structural material for the primary sodium side components of PFBR (Main vessel, inner vessel, intermediate heat exchanger, etc). Mod.9Cr-1Mo ferritic martensitic steel (Gr. 91 steel) is the steam generator material for PFBR. The design life of these permanent components is 40 years. Design life of future reactors will be 60 years. The mechanical properties of these materials are well established in the as-received condition, i.e. 316L(N) SS in the solution annealed condition and Gr.91 steel in normalised and tempered condition. During long term service at high temperature, these materials undergo microstructural changes. These changes in microstructure will cause changes in the mechanical properties. The effect of very long term thermal ageing, on the time-independent mechanical properties of these materials is not well established. Therefore there is a need to simulate the effect of thermal ageing for 60 years on the tensile and impact properties of these materials, for validation of the design data. 316L(N) SS in the solution-annealed condition contains nitrogen in supersaturated solid solution. Thermal aging causes microchemical repartitioning of nitrogen and major alloying elements leading to the formation of nitrides, complexes and ordered precipitates. Thermal ageing of Gr.91 steel leads to recovery of the tempered martensite substructure, replacement of the lath structure with subgrains, changes in dislocation substructure, and formation of complex intermetallic phases. All these changes are a function of the time and temperature of thermal ageing. The scope of this work involves thermal ageing of 316L(N) SS and Gr.91 steel to various time-temperature conditions, evaluation of the tensile and impact properties of the aged materials and establish empirical and mechanistic models to predict these mechanical properties for 60 years of ageing.
1.8 ELASTOMERS DEVELOPMENT
1.8.1 Development of Sodium Aerosol Compatible, Fluorocarbon Elastomer Formulation(s) Suitable for Manufacture of Inflatable Seals by Cold Feed Extrusion & Continuous Cure(Contact person: Shri N.K. Sinha)
Objective is to develop one or two aerosol-compatible formulations (with minimum life of ten years) based on suitable blends of Advance Polymer Architecture (APA) Fluorocarbon of 70 ± 5 0 Shore A hardness, elongation at break of 150% (minimum) and minimum tensile strength of 3.5 MPa (measured at 1200C and 1500C) which is (are) suitable for manufacture of inflatable seals for Fast Reactors by cold feed extrusion and continuous cure using microwave, hot air and infrared. Major areas of work include survey, study of formulation processibility, identification of ageing model, accelerated-ageing and life-assessment of standard ASTM/ISO/BS samples supported by studies on rheology and physico-mechanical properties using rubber process analyzer and related microscopic investigations.
1.8.2 Development of Sodium Aerosol Compatible, Silicone Elastomer Formulation(s) Suitable for Manufacture of Inflatable Seals by Cold Feed Extrusion & Continuous Cure (Contact person: Shri N.K. Sinha)
Objective is to develop one or two aerosol-compatible formulations (with minimum life of ten years) based on Silicone of appropriate hardness, elongation at break and tensile strength that is (are) suitable for manufacture of inflatable seals for Fast Reactors by cold feed extrusion and continuous cure using microwave, hot air and infrared. Major areas of work include survey, study of formulation processibility, identification of ageing model, accelerated-ageing and life-assessment of standard ASTM/ISO/BS samples supported by studies on rheology and physico-mechanical properties using rubber process analyzer and related microscopic investigations.
Objective is to develop a constitutive model for elastomer that combines the synergistic effects of radiation, sodium aerosol and temperature on the kinetic theory and the phenomenological aspects of rubber elasticity and can be used in Finite Element Analysis (FEA) to predict life of elastomers for inflatable seals of Fast Reactors. Major areas of work include survey, ageing studies on standard ASTM/ISO/BS samples, determination of properties of the samples and FEA in commercial codes using possible constitutive models for validation of the models against the empirical data.
1.8.4 Development of Miniature Test Rigs to Assess and Validate Life of Aerosol Compatible Elastomers and Inflatable Seals Combining Tests and Finite Element Analysis (Contact person: Shri R.Veerasamy/Shri N.K. Sinha)
Objective is to develop miniature test rigs, which can determine contact pressure and stress fields in inflatable seals or seal sections and can test the seals in sodium aerosol at 120/1500C under static and dynamic conditions of rotary oscillation to assess and validate the life of aerosol compatible elastomers and seals. Major areas of work include survey, conceptualization, design, testing in the rigs and stress as well as leakage modeling in finite element codes for assessment and validation.
1.9 REACTOR PHYSICS
In FBTR, for irradiating U-Pu metallic fuel at the required heat rating, either the fuel has to be enriched in U-235 or irradiation to be done at higher flux levels. In thermal reactors, use of materials with high scattering cross-section, high thermalising property and low absorption cross-section for neutrons have been used to get higher flux. In a FBR, the neutrons should not be allowed to slow down. So a material having high scattering cross-section, very low slowing down capability and low absorption cross-section of neutrons is to be identified and analysis performed in FBTR geometry.
Pyro-processed fuel retains part of fission products (Yttrium & lanthanides) and whole of minor actinides. It is estimated that equilibrium concentrations with recycle can reach 0.5 wt % for fission products and 0.8 % for minor actinides. The effects of these isotopes in breeding ratio & doubling time and safety related parameters like coolant void reactivity effect and temperatre coefficients are to be investigated.
Shields around core and blankets form major part of reactor assembly in fast reactors. Neutrons leaking from core and blankets are energetic and high in number. They can not only cause radiation damage to permanent structures and components but also pose radiological safety problems by activation and streaming through gaps of components and structures. Reduction of the neutron and gamma radiation to acceptable levels has always been a challenging problem in view of the high attenuation and streaming involved both from experimental and theoretical points of view. Development of suitable simulation models employing faster methods of solution such as Monte Carlo methods and deterministic methods using asymmetric quadrature are required. Validation of the modeling will be undertaken in research reactors like APSARA. The spectroscopic tools for the experiments also require refinements towards measurement of neutron and gamma spectra.
As passive systems are designed not to depend on external electrical power sources and human initiative for successful operation, extraneous elements are effectively decoupled from system performance and have good potential for improved reliability. However the reliability of such passive systems depends more on the process conditions arising out of natural convection phenomenon and its interaction with the hardware aspects of the design. Studies are required to establish an adequate methodology to quantify the reliability of such systems considering process and hardware aspects comprehensively.
1.11 REACTOR SAFETY
Prototype Fast Breeder Reactor (PFBR) is U-PuO2 fuelled sodium cooled Pool type Fast Reactor and it is currently under advanced stage of construction at Kalpakkam, India. Prototype Fast Breeder Reactor is equipped with two independent, fast acting and diverse shutdown systems. A shutdown system comprises of sensors, logic circuits, drive mechanisms and neutron absorbing rods. The two shutdown systems of Prototype Fast Breeder Reactor are capable of bringing down the reactor to cold shutdown state independent of the other. The first shutdown system comprises absorber rods which are used to control the reactor power and also to shutdown the reactor on receiving scram signal and hence they are referred to as Control and Safety rods. The absorber rods of the second shutdown system of Prototype Fast Breeder Reactor are called as Diverse Safety rods (DSR) and their drive mechanisms are called as Diverse Safety Rod Drive Mechanisms (DSRDM). DSR are normally parked above active core by DSRDM. On receiving scram signal, Electromagnet of DSRDM is de-energised and it facilitates fast shutdown of the reactor by dropping the DSR in to the active core. The response time of an Electromagnet is one of the very important safety parameters. The response time is defined as the time elapsed between the initiations of scram order to the time DSR is released from the Electromagnet. Presently it is measured using scram signal initiation and the decay curve of the EM. The response time of the Electromagnet is found to vary depending on the alignment between EM and armature of mobile DSR. The aim of the present study is to perform a combined theoretical electromechanical analysis of the Electromagnet and the armature simulating the alignment and also forces and moments at the junction between them. A computer code is to be developed to predict the coupled magnetic and mechanical behavior and this is to be validated by experimental results.
Presently PFBR core catcher is designed to take 7 fuel SA melt. Criticality analysis has established that up to 55 SA falling on the core catcher as heap remain sub-critical. For future FBR, as defense in depth, innovative design of core catcher is to be found which can ensure enough separation the fuel melt heaps and maintain subcriticality. Also use of neutron absorber materials in the core catcher is to be investigated. Finally a design where coolability is ensured to be found.
Core Disruptive Accident (CDA) is very low probability event in FBR. However, its consequences are analysed as a part of safety evaluation, wherein post accident heat removal plays an important role. The core materials continue to produce heat even after the accident due to the decay of fission products. After a CDA, the molten core moves downwards and gets fragmented to fine debris of particle size ~ 200 - 400 mm, due to quenching on contact with sodium. This debris settles on core catcher provided at the bottom of the reactor assembly as a porous bed of 30 % – 40 % porosity. Cooling of this heat generating bed by natural convection (with limited local boiling in the bed) by suitable innovative decay heat removal system without exceeding the temperature limits is the main issue to be addressed.
To handle the fluid dynamics part of the core disruptive accidents for FBR's, Lagrangian codes like REXCOH are presently used. These codes need rezoning. This is due to the fact that the material interfaces undergo deformation due to high pressures. Due to these deformations, the derivative calculation is riddled with large errors. To offset these errors, the material boundaries need to be redrawn periodically so that the geometry is well defined and finite difference approximations are valid. This process of redrawing the boundaries consistent with mass conservation is called the rezoning. Presently, we are employing a version of rezoning that was developed at RSD. We need to have a better version that will address a larger class of deformations.
The comprehensive CDA model consists of 3 key elements, namely fluid dynamics, neutronics and structure. The fluid dynamics in turn involves multi-component, multi-phase heat and mass transfer modules. It also needs a reasonable modeling of the interface dynamics with pool and channel flows. A typical model invokes heat transfer involves something like 53 binary contacts, 33 vaporisation/condensation paths and 22 melting/freezing paths. This modeling as a whole is complex and needs a chemical engineering background.
1.11.6 Heat and Mass Transfer - Development of a model to predict heat and mass transfer behaviour in argon cover gas of a sodium cooled fast reactor with effect of irradiation - Sodium aerosol deposition in annular gaps of top shield of a sodium cooled fast reactor
(Contact person: Dr. R. Bhaskaran)
Pool type Sodium Cooled Fast Reactor (SFR) has a large surface area of sodium free surface which is at high temperature blanketed by argon cover gas. The complicated phenomena involved in this system is the heat and mass transfer from the free surface. There exists a fourth mode of heat transfer is due to evaporation and condensation of sodium aerosols dispersed in argon cover gas space. Limited information is available on this subject without considering effects of presence of ‘r’ radiation field. The present project deals with the effects of ‘r’ on the heat and mass transfer behaviour in the argon cover gas space. A dedicated facility is under construction of IGCAR to investigate this problem. In this, the sodium free surface at varying temperature of ~ 1 m diameter is conceived with the varying argon cover gas space and top shield with varying heat removal, capabilities, facility of apply ‘r’ field, sophisticated measurement system for characterizing the aerosols and their behaviour with the presence of ‘r’ field and other instrumentation and control system. The important aspect of the project is that it contains science, engineering and technology components more than one PhD theses are possible.
Interaction between liquid sodium and concrete is a realistic possibility in Fast Breeder Reactors (FBR). Large sodium spillage on the Steam Generator Building (SGB) floors of the FBR is a Category 4 Design Basis Event. Extensive physical investigations at IGCAR have led to the specification of limestone aggregate and fines based concrete as a sacrificial layer on the SGB floors of PFBR. In order to have even better and more cost effective sodium resistance, fundamental mechanisms governing the concrete damage need to be addressed using microstructural investigations. Detailed micro-structural evaluations of the virgin and sodium interacted concretes: through a combination of various spectroscopic methods such as X-Ray Diffraction (XRD), Fourier Transform Infra-Red (FTIR) etc. and petrographic and scanning electron microscopy, are to be carried out. Naturally occurring dunite rocks also has shown good resistance to hot liquid sodium. Therefore a proposed project can aim to study both limestone and dunite based concretes. Combination of limestone concrete and dunite slab over it can also be investigated. Influence of the type of cement and water to cement ratio (w/c) and effects of deleterious materials present in the limestone or dunite aggregate can also be investigated.
(Contact person: Shri B. Krishnakumar & Dr.P. Swaminathan)
At present response time of core thermocoupl is 6 sec with uncertainity of 2 sec. Due to large variation in response time, core temperature distribution pattern gets disturbed during “lowering of rod” in FBTR or “power setback” in PFBR. This results in spurious scram order from digital signal processing system. Hence there is need to develop fast response temperature measurement system with as minimum dispersion as possible.
At present response time of core thermocoupl is 6 sec with uncertainity of 2 sec. KALMAN filter technique will be studied and tuned for fast reactor temperature measurement system towards improving the response time)
(Contact person: Dr.P. Swaminathan)
During the intermediate power range of fast reactor, fluctuation in pulse channels is processed. The standard deviation of the signal is a pointer to the reactor power. The response time of core thermocouple of the central assembly is only 200msec. The fluctuation increases with reactor power. Analysis can be attempted to extract reactor power information from standard deviation of core temperature signal.
(Contact person: Shri N. Murali & Dr.P. Swaminathan)
Analog Instrumentation and Control(I&C) systems have become obsolete and Digital I&C systems are increasingly used. Definite procedure exits for calculation of reliability of digital hardware systems. But software reliability assessment and improvement is not yet very well defined. Scope of research is to develop mathematical software reliability model for safety critical applications.
(Contact person: Shri S.A.V. Satya Murthy & Dr.P. Swaminathan)
Distributed Digital Control System (DDCS) is used for supervision and Control of Nuclear Reactor. Nearly 15000 process signals are processed by DDCS. The information consisting of the value of the process signals, alarm messages, trip messages etc are transmitted to control room through optical fibre. Control room operator should be protected from Information overload. Scope of Research is to develop optimum Information Management System such that operator feels comfortable.
(Contact person: Shri N. Murali & Dr.P. Swaminathan)
At present DDCS is increasingly used for supervision and control of Nuclear Reactor. DDCS of PFBR consists of nearly 100 computer systems which are physically distributed throughout the Plant. At present standard commercially available TCP/IP protocol is used for communication between the computer systems. This protocol has large overhead which is not necessary for control applications. Scope of research is to develop appropriate protocol such that overhead is minimum. This ensures higher reliability.
(Contact person: Dr.P. Swaminathan)
At present Synchro to digital converter chip is increasingly used for precise measurement of rotary position, such as position of control rod, guide tube, gripper assembly etc. This IC is not available in India. Due to import restrictions, non availability will affect the I&C systems of Fast Reactor programme. Scope of research is to design and manufacture the SDC-IC chip.
(Contact person: Shri S.A.V. Satya Murthy & Dr.P. Swaminathan)
At present large amount of copper cable is used for transmission of signal for sensors to real time computer systems. Signal is affected by electromagnetic noise present at the site. Cost of the cable is considerable and cable may also propagates fire. Scope of research is to develop both hardware and software for fault tolerant wireless sensor network for supervision and control of nuclear reactor.
(Contact person: Shri S.A.V. Satya Murthy & Dr.P. Swaminathan)
Alarm in the control room always come in groups. Operator would like to know the root cause of the event in plant. Plant specific expert subsystem need to be developed to provide fault tree analysis. Scope of research includes development of general purpose expert system which is connected to Plant Local Area Network. Plant specific expert rules need to be developed for all possible events for each type of Reactor.
(Contact person: Dr.P. Swaminathan)
At present digital design is increasingly used for I&C of Nuclear Reactor. No proven methodology exits for Verification and Validation of digital design. Scope of Research includes development of Verification and Validation methodology for digital design along with associated tools.