OVERVIEW
Transactions of The Indian Institute of Metals Vol. 62, Issue 2, April 2009, pp. 89-94
Development of cladding materials for sodium-cooled fast reactors in India Baldev Raj, Divakar Ramachandran and M. Vijayalakshmi Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamil Nadu-603 102, India. Email:
[email protected] Received 25 January 2009 Revised 11 February 2009 Accepted 13 March 2009 Online at www.springerlink.com © 2009 TIIM, India
Keywords: sodium-coled fast reactors, grain boundary engineering; D9; modelling: "InD9"
Abstract Sodium-cooled fast reactors are an important component of India’s move towards energy security. The Fast Breeder Test Reactor (FBTR) has been successfully operated for over 22 years, and the construction of a Prototype Fast Breeder Reactor (PFBR) has been taken up. Future development of Fast Reactor technology depends to a large extent on the economics, safety and waste management issues. The targets important for economics have been identified as reduction in capital costs, higher burn-up fuels, higher temperatures of operation, longer life and higher breeding ratios. These targets can be achieved with significant improvements in performance of materials, especially of Alloy D9 (15Cr-15Ni-0.2Ti), a titanium modified austenitic stainless steel, currently used for fuel cladding. Through research programmes implemented to improve the void swelling resistance of Alloy D9, it has been established that certain solute additions are essential in association with thermomechanical treatments for conferring swelling resistance. Research work conducted on refinement of Alloy D9 composition and thermomechanical treatments aimed at understanding irradiation effects, corrosion, mechanical properties and weldability at IGCAR Kalpakkam along with modelling studies for the prediction of weld behaviour is reported in the present paper. It is concluded that a significant rise in fuel burn-up can be achieved with the use of a new generation “InD9” alloy for fuel cladding.
1. Introduction Nuclear energy has re-emerged as an important alternative green energy source. It has crucial importance to supply mega resources of energy in combination with fossil fuels and renewable energy sources. Today 16% of the world’s electricity is generated in nuclear reactors of various designs. The Chernobyl disaster of 1984 had caused a world-wide slow-down in the share of nuclear-generated electricity for nearly two decades. However, countries like France, Russia, Japan, Korea, China and India have pursued active nuclear power programmes. Nuclear energy is now being recognised as the inevitable option for a sustainable energy with the potential to avoid serious climate change. In terms of CO2 emissions, nuclear energy is one of the most environmentally friendly options. Internationally, the thrust in the future is to develop nuclear power which can guarantee nonproliferation, improved safety and better waste management. In the Indian context, complete utilisation of uranium as nuclear resource material coupled with availability of a large thorium resource, the high energy potential of nuclear fuel, possibility of shortage of fossil fuels under conditions of difficult supplies and the negligibly small CO2 footprint of nuclear energy imply that only the nuclear option with closed fuel cycle can ensure energy security. Accordingly, India has undertaken and made rapid strides in the building and commissioning of fast reactors for energy generation, having opted for a closed nuclear fuel cycle that ensures economy of fissile materials and better waste management strategy by reducing radiotoxicity on a long term basis. The nuclear programme of the country is being implemented in three
stages: pressurised heavy water reactors of the CANDU type; sodium-cooled fast reactors; thorium-based nuclear reactors. The vast thorium reserves in the country demand that the third stage that can make use of it be introduced as early as feasible. Design efforts on accelerator driven systems are being taken up to ensure in situ incineration of longlived fission products and production of fissile U233 from fertile U232. At present, the transition from the fast test reactor stage where a 40 MW (thermal) was successfully operated for over 22 years, to a 500 MW(e) prototype reactor is being made. In India, fast reactors have a minimum power potential of 530 GW(e). This stage is crucial since it will provide the necessary experience and fuel for the third stage and beyond, with a power generation potential that is estimated to be in excess of 150 trillion watts. The chosen technology for this crucial 2 nd stage is sodium-cooled fast neutron spectrum reactors for generation of electricity. In order that the nuclear option is sustainable it is necessary that improved design targets be achieved: reduction in capital cost, high burn-up fuel, longer lifetime of the reactor, higher operating temperature and higher operating performance. Innovative approach is required to achieve these targets. Materials science, engineering and technology play a significant role when coupled with robust design, engineering development, instrumentation and control and robust safety approaches. Over the years, since the discovery of radioactivity, there has been a paradigm shift in the approach to materials selection for nuclear applications, closely following the maturity of metallurgy from an empirical art to science. Whereas early reactor designers chose materials from those
Baldev Raj et al. : Trans. IIM Vol. 62, Issue 2, April 2009
90
available based on suitability of properties and empirical but breakthrough insights, currently several programmes are devoted to develop new materials such as high burn-up materials or reduced activation ferritic martensitic (RAFM) steels specifically tailored for use in future fission and fusion reactors. A concentrated effort for the development of advanced materials, their characterisation and corresponding advances in materials technologies includes the following fields:
evaluation of design criteria for advanced materials
modelling to predict materials behaviour
knowledge based design of materials
extension of materials modelling to industrial scale phenomena
evaluation of in-service behaviour by simulation and laboratory experiments
evaluation of limiting performance conditions for industrial components
remnant life evaluation
remote repair with robotics
post-irradiation examination
continuous improvement in the characterisation techniques
continuous development of material codes and their refinement
The following section will describe the Indian scenario with respect to the fast reactor technology and the materials challenges in the sodium-cooled fast reactors. This will be followed by a few case studies related to materials development where significant experimental and modelling research conducted on Alloy D9 at IGCAR, Kalpakkam is highlighted. The first of these is about fuel clad materials that are required be resistant to irradiation induced swelling and embrittlement, resistant to sodium corrosion, have adequate weldability, and end-of-life creep strength and ductility. Next, thermomechanical treatments that result in a grain boundary structure optimisation leading to increased resistance to sensitisation of stainless steel welds is discussed. The role of modelling in the prediction of materials performance is then demonstrated through applications to welds. The last section reviews briefly the next generation nuclear materials.
2. Sodium-cooled fast reactors: Indian scenario The development, construction and operation of sodiumcooled fast reactors is an important component of India’s move towards energy security. Having operated a Fast Breeder Test Reactor (FBTR) successfully for over 22 years, and embarked up on the construction of a Prototype Fast Breeder Reactor (PFBR), future development of Fast Reactor technology depends to a large extent on the economics, safety and waste management issues. The medium-term targets important for economics have been identified as reduction in capital costs, higher burn-up fuels, higher temperatures of operation, longer life and higher breeding ratios. The long-term goals are to achieve significantly higher thermal efficiency with coolants such as liquid Pb and higher breeding ratio, requiring development of newer materials.
Broadly, nuclear plant components are classified into three classes: (i) core components that are installed in the reactor main vessel, (ii) out of core components that include part of primary sodium circuit that is outside the main vessel and (iii) balance of plant components that comprises the conventional power generation systems. Of these the core components comprise of (a) the core which includes the fuel, clad and wrapper and (b) the support structures for the core. The cladding tube requires a material that is resistant to radiation damage and is compatible with the fuel and fission products. Core component materials, especially the fuel clad and wrapper materials are subject to the most severe environments in the reactor and correspondingly have high demands placed on their properties and performance. Austenitic stainless steels are chosen as the major structural materials for the currently operating and planned Fast Reactors all over the world in view of their adequate high temperature mechanical properties, acceptable radiation resistance, compatibility with liquid sodium coolant, good weldability, availability of design data and satisfactory experience in use. Improvement of the economics can be achieved with significant improvements in performance of these structural materials, especially Alloy D9 (15Cr-15Ni0.2Ti), a titanium modified austenitic stainless steel, currently used for fuel cladding.
3. Structural materials issues for sodium-cooled fast reactors Materials in sodium-cooled fast reactors need to be capable of operating at higher temperatures in a more severe radiation environment as compared to materials in thermal nuclear reactors. The presence of sodium presents additional challenges to maintain and monitor low levels of oxygen and nitrogen dissolved in the liquid sodium. Thus the challenges for core components in sodium-cooled fast reactors revolve around radiation resistance, high-temperature mechanical properties and chemical compatibility with the fuel as well as the liquid sodium coolant. 3.1 Optimisation of austenitic stainless steels for clad tubes The material of choice for this function is a titanium modified SS316, also known as Alloy D9 (15%Cr-15%Ni0.2%Ti), in the 20% cold-worked condition. Austenitic stainless steels are still favoured for fuel pin cladding and other core component applications since they possess the required strength characteristics up to 923 K. Early studies on creep properties [1] of alloys with titanium to carbon ratio between 4 and 6 showed that titanium content strongly influences the creep rupture life. Alloys with Ti/C ~ 4 showed the best creep rupture life at 973 K. However, the rupture ductility was found to be poor. On the basis of metallographic analysis of the samples it was suggested that this is due to the intragranular precipitation of titanium carbides in the cold-worked matrix that led to the formation of creep cracks. Thus, it was recognised that the propensity for carbide formation needed control and it was recommended that the composition be optimised such that the Ti/C ratio was limited to ~ 4. However, the limiting factor at moderate reactor operating temperatures of up to ~ 873 K is void swelling which ultimately limits life of the fuel pin leading to a reduced burn-up of about 100 GWd/tonne (Gigawatt-days per tonne is a measure of the energy extracted from a metric ton of nuclear fuel. 1 Gigawatt-day corresponds to 86.4 x 1012 J or 24 million units of electricity (1 unit = 1 kW.hr)). Hence, for
Baldev Raj et al. : Trans. IIM Vol. 62, Issue 2, April 2009
clad tubes of the fuel pins the driving force for development of new structural alloys has been the required improvement in void swelling resistance. Based on mechanical characterisation, two prospective austenitic stainless steels with compositions differing in Ti content have been selected for evaluation of void swelling resistance. Ion-irradiation with 5 MeV Ni2+ ions after prior He implantation to mimic fast reactor conditions has been adopted [2] to evaluate radiation damage in these compositions. The damage rate in ion irradiation experiments is about three orders of magnitude higher than that typically experienced by materials under neutron irradiation in a fast reactor. This enables screening of new materials for void swelling behaviour through reasonably fast experiments. The equivalence of the high rate damage to the much slower damage rate under neutron irradiation in a fast reactor can be expressed in terms of a temperature shift. The magnitude of this temperature shift can be written as [3],
, where TNi+
and T n represent equivalent temperatures under Ni + and neutron irradiation respectively,
, with Q
representing the vacancy migration energy, and GNi+ and Gn being the damage rates under Ni+ and neutron irradiation respectively. Assuming the vacancy migration energy to be 30 kcal/mol and
, the temperature shift
corresponding TNi+ = 823 K, for example, is calculated to be 225 degrees. Temperature shift values thus calculated can be used to determine the equivalence to neutron irradiation. TRIM [4] calculations have been used to determine the fluence and irradiation time required to produce ~ 100 dpa (displacements per atom represents the average number of displacements an atom of the target material undergoes under a given fluence of irradiating particles. It is used as a measure of radiation damage.) as the damage peak for the 5 MeV Ni2+ ions in stainless steels. The degree of void swelling resulting from the irradiation is measured in terms of the step height between masked and unmasked regions of a 5 mm x 12 mm sample surface. Damage rates ~ 7 x 10 -3 dpa/s could be
91
achieved with a 1.7 MV Tandetron accelerator. Figure 1 shows the results of such irradiation experiments on two candidate alloys with Ti/C ratios of ~ 6 and ~ 4 in 20% cold-worked state, irradiated with 5 MeV Ni2+ ions after He implantation to a concentration amounting to ~ 30 appm, in the temperature range 723 – 973 K. It is seen that the alloy with the Ti/C ~ 6 (Ti ~ 0.25 wt%) exhibits significantly lower swelling of ~ 4% compared to nearly 15% for the alloy with Ti/C ~ 4 (Ti ~ 0.15 wt%). The peak swelling temperature is also significantly lower at 823 K for the former alloy, 100 degrees lower than the corresponding temperature for the latter alloy. Based on the temperature shift estimation to allow for the much smaller damage rate in a fast reactor, the peak swelling temperatures for Alloy D9 with Ti/C ratios of 4 and 6 are estimated to be 649 K and 598 K respectively. The chief influence of titanium in this alloy is through uniform distribution of fine secondary precipitates of TiC stabilising the cold-worked dislocation structure. The fine precipitates have a high degree of lattice mismatch with the austenite matrix and the interfacial defects that result can be monitored using a technique such as positron annihilation spectroscopy. The positron life-times were measured as a function of the isochronal annealing temperatures for the above two alloys [2]. The results are interpreted as showing a higher number density of TiC precipitates and a lower temperature of onset of precipitation in the alloy with a higher Ti content. Thus the reduced swelling in this alloy can be correlated to a higher number density of fine TiC precipitates that trap voids at the matrix – precipitate interface, while the lower peak swelling temperature is attributed to the effect of titanium in solid-solution on the effective diffusion coefficient of vacancies in the austenite lattice. Minor elements such as Si, Ti and P have a major influence on the void swelling behaviour [5] of Alloy D9. In an effort to further optimise the alloy composition around the nominal Alloy D9 levels and identify an improved alloy D9, a series of alloys were produced by varying the concentrations of Ti, Si and P around their nominal values in standard Alloy D9. As a result of ion irradiation studies on these alloys, an optimised austenitic steel based on 15Cr15Ni-Ti (Alloy D9) with Si, and P additions (“InD9”) are proposed for fuel pin cladding applications. The InD9 alloy with optimum composition of minor elements is expected to allow safe operation up to ~ 150 dpa for fuel clad material. 3.2 Grain boundary optimisation through thermomechanical treatments
Fig. 1 : Void swelling behaviour of Alloy D9 with (a) 0.25 wt% Ti, and (b) 0.15 wt% Ti
Sensitisation is known to occur in austenitic stainless steels as a result of precipitation of Cr-rich carbides at grain boundaries (Fig. 2(a)) which leads to a Cr-depleted zone near grain boundaries (Fig. 2 (b)), eventually leading to intergranular corrosion related material failures. The availability of a technique to determine statistically significant grain boundary character distribution over a large umber of grains has enabled derivation of a number of parameters that can be correlated to materials properties. Such correlations are essential for establishing a scientific basis to materials development. One such example in the materials development for fast reactors is the determination of effective grain boundary energy (EGBE) and its correlation to sensitisation of AISI 316L(N) stainless steels. The time to sensitisation generally reduces with cold-work up to a level of 15%, and increases with a decrease in carbon and increase in nitrogen in the steel. It has been known for some time that corrosion behaviour can also be influenced by grain boundary
92
Baldev Raj et al. : Trans. IIM Vol. 62, Issue 2, April 2009
(a)
(b)
character, which in turn can be modified through a tailoring of the thermo-mechanical treatments. However, until fairly recently, this was essentially an empirical observation that was applied without a detailed understanding of the basis. The difficulty was in defining a single measurable parameter that could represent the averaged state of grain boundary character in a sample. Developments in electron backscattered diffraction (EBSD) has enabled the rapid determination of grain boundary character for a large number of grain boundaries in a reasonable time scale. However, it is recognised that the grain boundary character determines the grain boundary energy, which in turn plays a role in interfacial reactions including grain boundary precipitation. It is necessary to define a quantity that describes the overall state of grain boundary energy in a material to correlate the results of an EBSD investigation with a macroscopically observed effect such as sensitisation. EGBE is defined as[6] , where
(c)
is the energy of
grain boundary of class i with a CSL notation Σi, fi is the fraction of such boundaries, Δθ i is the average deviation from exact CSL configuration, θmax is the maximum deviation allowed for that boundary, d is the grain size and γmax is the energy of random boundaries. It can be determined using fi and Δθi derived from an EBSD experiment. It is possible to increase the percentage of coincidence site lattice (CSL) boundaries up to 70% by controlled thermo-mechanical treatments. For a material that has a large fraction of grain boundaries with CSL misorientations, the EGBE has low values while a material with maximum randomised grains will have a high EGBE. It was possible to alter or delay sensitisation by an appropriate thermomechanical treatment. The degree of sensitisation (DOS) and the EGBE has been determined in AISI 316L(N) as a function of thermal treatments [7,8]. As seen from Fig. 2(c), at low as well as very high effective grain boundary energies, the susceptibility to sensitisation is greatly reduced. These states correspond to large fractions of grain boundaries being of low energy CSL type or very high angle grain boundaries. A number of thermomechanical treatments can be designed to achieve particularly low effective grain boundary energies for the material. It is found that there can be a ten-fold increase in the time to initiate sensitisation by suitably engineering the grain boundary state in a material. This is of high importance in practical welding situations commonly encountered during fabrication or repair of nuclear reactor components. 3.3 Modelling in prediction of materials performance
Fig. 2 : (a) Formation of Cr-rich carbides near the grain boundary in a AISI 316 stainless steel in a 5% coldworked state after treatment at 1073 K for 15 minutes, (b) energy dispersive X-ray spectroscopy (EDXS) results from the sample in (a) showing the degree of Cr depletion at the grain boundary, (c) Degree of sensitisation as a function of the effective grain boundary energy (EGBE) in an AISI 316LN stainless steel showing particularly low values for low energies (CSL boundaries) and high energies (random boundaries). EGBE is normalised to γ max and hence is represented as a dimensionless quantity.
Modelling lies at the core of any materials engineering project. The ability to model a given material property or process enables innovative solutions to engineering problems that saves a lot of experimentation time. Many times development of reliable and robust models is essential to predict material behaviour for domains where experiments cannot be performed, for example, for time scales ~ 60 years. In this section, we describe two specific case studies in welding, one where a conventional diffusion model has been applied to compute the solute redistribution across a dissimilar weld, and one where novel neural network model has been adapted to predict ferrite number. Welding being an important on-site fabrication technique used in the construction of nuclear reactors, such studies are of immense importance in ensuring that materials technology designed in
Baldev Raj et al. : Trans. IIM Vol. 62, Issue 2, April 2009
Fig. 3 : Comparison between the carbon concentration profile determined experimentally using Electron Probe MicroAnalysis (EPMA) as a function of distance from an AISI 316LN steel surface exposed to liquid sodium at 823 K for 16,500 hours.
laboratory clean environments is translated into sound structural components at the construction site. 3.3.1 Hard zone formation in dissimilar weldments of Cr-Mo steels Dissimilar metal welds are common in nuclear reactors, especially in the balance-of-plant components. The steam generator, for example, requires welding of Cr-Mo steels of different Cr contents. Post-weld heat-treatment (PWHT) of such joints results in the formation of a carbon-depleted soft zone on the low chromium side of the weld and a corresponding precipitate-rich hard zone on the high chromium side. Micromechanisms responsible for the formation of these zones have been identified based on a detailed microstructural and microchemical study of dissimilar welds between 9Cr-1Mo steel and 2¼Cr-1Mo steel subjected to PWHT at 1023 K [9]. It was found that the width and hardness of the soft and hard zones are influenced by the duration of the heat-treatment. C diffuses to the high Cr side and gets bound into carbides leaving behind a relatively soft ferrite matrix. Mo does not participate in the solute redistributions that take place. To better understand and mitigate the effect of hard and soft zones formation, a numerical procedure based on finite difference method has been used to simulate the zone formation during hightemperature exposure [10]. The possible different barrier materials were identified and confirmed by experiments [10]. Similar calculations were extended to predict the thickness of carburised layer in an AISI 316LN exposed to sodium containing 25 ppm of carbon for 16,500 hours (Fig. 3). Based on validation of the composition for short time with concentration profiles determined using an electron probe microanalyser (EPMA), thickness of layer could be predicted for 40,000 hours of exposure. 3.3.2 Bayesian neural network model for ferrite number prediction in SS welds A minimum ferrite content is necessary to ensure hot cracking resistance in austenitic stainless steel welds, while an upper limit on the ferrite content is essential to avoid
93
sigma phase embrittlement. The ferrite content results from the microstructural evolution during the welding process. Traditional models of ferrite content, quantified as Ferrite Number (FN) use linear expressions in terms of Cr-equivalent and Ni-equivalent concentrations. This has been found to be inadequate to represent the complex relationship between composition and FN since inter-solute interactions are ignored. In this context, an accurate Artificial Neural Network (ANN) based predictive model that accounts for the effect of the various alloying elements has been developed [11]. The Bayesian neural network models were employed to relate thirteen compositional variables to a single FN output. For this about half of 1020 datasets were used to train the network and the remaining half used to test performance of the network. Using this mode, it was possible to establish varying non-linear contributions of individual elements to FN depending on the base composition. Another important consideration in austenitic stainless steel welding is the solidification mode. The weld metal composition has to be tailored to promote a primary ferritic mode of solidification for minimising solidification cracking susceptibility and to reduce the amount of slag formation. Bayesian classification neural network has been applied to classify solidification modes based on composition [12]. Based on this model, it is shown that Ni, Cr, Mn and N are the main elements whose concentrations influence the solidification mode. The model has achieved a predictive accuracy better than 81% on an independent dataset. This degree of reliability of the prediction of solidification mode, given an alloy composition is of great practical significance. Correspondingly, with close control of Ni, Cr, Mn and N it is possible to obtain primary ferritic solidification mode and hence reduce the propensity for cracking during solidification and eliminate the slag produced by arc welding.
4. Emerging nuclear core component materials Future trends in the global fast reactor industry are towards higher operating temperatures, higher burn-up (200 GWd/t), higher breeding rations (~1.4) and longer lifetime for reactor (60 – 100 years). These goals require several developments in materials science and technology across all components of nuclear plants, especially for fixed core component materials. In terms of breeding ratio and sustainable growth of nuclear energy, metallic fuels are envisaged for the next generation of fast nuclear reactors in India. The fabrication, use and reprocessing of these fuels poses several challenges that are being studied currently. The fuel will have to be fabricated under remote operation in an inert atmosphere. Candidate alloying additions such as Zr to the pure metal are being considered and evaluated. Suitable corrosion resistant coatings and refractory container materials will have to be selected to minimise fuel – clad interactions. Metallic fuel cycle requires new waste treatment strategies to be developed. Pyrochemical reprocessing route is being developed involving molten salts at high temperatures, as opposed to the currently well-established aqueous route for oxide fuels. Increase in reactor operating temperature and thermal efficiency require better coolants than the currently used liquid sodium. Cooling by gas such as He and Pb-based liquid alloys will have to be considered. This demand requires verification of compatibility of fuel and the clad material and thus affects the choice of core structural materials. Further, increased burn-up considerations at the higher operating
94
Baldev Raj et al. : Trans. IIM Vol. 62, Issue 2, April 2009
temperatures requires novel fuel design concepts such as the annular fuel pellet. Most importantly, the current limitation on fuel burn-up, namely, void swelling of the core structural materials will have to be reduced or further delayed. Compared to currently used austenitic stainless steels, ferritic steels have a much better void swelling resistance and are capable of burn-up ~ 200 GWd/t as clad material. However their use is rendered difficult due to their poorer tensile and creep strengths at temperatures higher than ~ 873 K. Development of higher temperature tensile and creep strengths in these alloys will enable working the reactor at higher temperatures and to longer burn-ups, thus improving the economics of nuclear power production. Commercial ferritic-martensitic steels based on 9-12 % Cr compositions exhibit the highest swelling resistance. Such alloys therefore appeared ideal for fast reactor applications, but their reduced strengths above ~ 798 K has restricted their use to certain low stressed components such as sub-assembly wrappers, used to support clusters of fuel pins. To circumvent this limitation, programmes are being implemented to explore ferriticmartensitic oxide dispersion strengthened variants, which can possess good strength properties up to 923 K. Conventional alloy melting routes will have to be abandoned in favour of powder metallurgy techniques of ball-milling, hot isostatic pressing and hot extrusion for the synthesis of these alloys. Process optimisation for the development of 9Cr based ferritic / martensitic steels strengthened by a fine dispersion of yttria nanoparticles has been completed. The irradiation response of dispersoids is being studied. Safety and ease of handling spent fuel requires that the activity of the fuel assemblies on discharge from the reactor is reduced. One approach to this issue is to develop variants of the current structural materials where alloying additions that result in high activation are replaced with alternate elements to have reduced activity in the spent fuel. Solutes such as Mo and Ti are being replaced with W and Ta. The primary use envisaged for these reduced activation ferritic / martensitic steels alloys is in the fusion reactors where the radiation environment is much more severe. However, with improvements in creep rupture strength these steels can be used for future fast reactors [13].
the design of these materials have been discussed. Materials challenges for future reactors require development of new materials through sound design principles, validation with modelling and experimental measurements, fabrication technologies and in-service inspection methods to monitor their in-reactor performance. The current trends in materials development through intense international collaborations would certainly reduce the time and cost of alloy development for future reactors.
Acknowledgements One of the authors (DR) acknowledges useful discussions with Dr. M. Vasudevan, Dr. N. Parvathavarthini and Prof. I. Samajdar.
References 1. 2.
3. 4. 5. 6. 7.
8.
9. 10.
5. Summary Materials science, engineering and technology form an important ingredient for the safe and economic fast reactors. A number of materials and technologies that contribute to achieve the best performance using advanced materials have been highlighted in the present paper. The principles behind
11.
12. 13.
Latha S, Mathew M D, Rao K B S, Mannan S L, Trans. IIM, 49 (1996) 587 David C, Panigrahi B K, Balaji S, Balamurugan A K, Nair K G M, Amarendra G, Sundar C S, and Raj Baldev, J. Nucl. Mater., 383 (2008) 132 Straalsund J L, J. Nucl. Mater., 51 (1974) 302 Biersack J P, Haggmark L G, Nucl. Instrum. Meth., 174 (1980) 257 Maziasz P J, J. Nucl. Mater., 200 (1993) 90 Wasnik D N, Kain V, Samajdar I, Verlinden B and De P K, Acta. Mater., 50 (2002) 4587 Dayal R K, Parvathavarthini N, Mulki S and Samajdar I, Development of a very high resistance to sensitisation in austenitic stainless steels through special heat treatment resulting in grain boundary microstructural modification, European Patent Application no. 08159613.2-1215 dated 29/09/2008 Parvathavarthini N, Mulki S, Mani K V, Dayal R K, Samajdar I and Raj Baldev, Sensitisation control in AISI 316 L(N) austenitic stainless steels: defining the role of grain boundary structure, (2009) under communication Sudha C, Terrance A L E, Albert S K, and Vijayalakshmi M, J Nucl Mater, 302 (2002) 193 Anand R, Sudha C, Karthikeyan T, Terrance A L E, Saroja S, and Vijayalakshmi M, J Mater Res, 44 (2009) 257. Vasudevan M, Murugananth M, Bhaduri A K, Raj Baldev, and Prasad Rao K, Science and Technology of Welding and Joining, 9 (2004) 1 Vasudevan M, Bhaduri A K, Raj Baldev, and Prasad Rao K, Materials Science and Technology, 23 (2007) 451 Klueh R J, Nucl Mater, 378 (2008) 159
