Scholarly article on topic 'Materials and Manufacturing Technologies for Sodium Cooled Fast Reactors and Associated Fuel Cycle: Innovations and Maturity'

Materials and Manufacturing Technologies for Sodium Cooled Fast Reactors and Associated Fuel Cycle: Innovations and Maturity Academic research paper on "Materials engineering"

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Energy Procedia
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{"Fast Breeder Reactor" / manufacturing / PFBR / sodium / "fuel cycle"}

Abstract of research paper on Materials engineering, author of scientific article — Baldev Raj

Abstract Fast Breeder Reactors (FBRs) are emerging as vital source of power generation in India towards meeting energy security and sustainability for the growing economy of India. The success of FBR programme depends on continuous operation of reactor system and reprocessing and waste management facilities to ensure economy and also to achieve sustained fissile material production for several FBRs planned in future. Extensive research and development in domains of materials and manufacturing technologies are demanded towards development of FBRs and their associated fuel cycle technologies. This paper highlights the work and the approaches adopted at the Indira Gandhi Centre for Atomic Research, Kalpakkam for the successful development of materials, manufacturing and inspection technologies for both reactor and reprocessing facilities of the current and future Indian FBR programme.

Academic research paper on topic "Materials and Manufacturing Technologies for Sodium Cooled Fast Reactors and Associated Fuel Cycle: Innovations and Maturity"



Materials and Manufacturing Technologies for Sodium Cooled Fast Reactors and Associated Fuel Cycle: Innovations and Maturity

Baldev Raj

Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102, India


Fast Breeder Reactors (FBRs) are emerging as vital source of power generation in India towards meeting energy security and sustainability for the growing economy of India. The success of FBR programme depends on continuous operation of reactor system and reprocessing and waste management facilities to ensure economy and also to achieve sustained fissile material production for several FBRs planned in future. Extensive research and development in domains of materials and manufacturing technologies are demanded towards development of FBRs and their associated fuel cycle technologies. This paper highlights the work and the approaches adopted at the Indira Gandhi Centre for Atomic Research, Kalpakkam for the successful development of materials, manufacturing and inspection technologies for both reactor and reprocessing facilities of the current and future Indian FBR programme.

© 2011 Published b y IEl sevier Ltd.Selection and/or peer-review under responsibility of Indra Gandhi Centre of Atomic Research

Keywords: Fast Breeder Reactor; manufacturing; PFBR; sodium; fuel cycle 1. Introduction

Fast Breeder Reactors (FBRs) are vital towards meeting security and sustainability of energy for the growing economy of India [1]. India has successfully operated Fast Breeder Test Reactor (FBTR) since October, 1985. Forty years of R&D, technology development of demanding equipments by manufacturing and testing of full size components and absorbing 390 reactor-years of international experience of sodium cooled fast reactors has enabled us to embark on 500 MWe Prototype Fast Breeder Reactor (PFBR) with confidence and courage. The construction of PFBR is in advanced stage of completion. The main objective of the construction of PFBR is to demonstrate the techno-economic viability of sodium cooled fast reactors for the commercial deployment. In addition, it will provide a demonstration of comprehensive closed fuel cycle technologies such as fuel fabrication, reprocessing, waste management and waste immobilisation. PFBR will also enable validation of first-of-a-kind design concepts and the experience garnered during operation and maintenance of mechanisms, components and sensors in the sodium environment and reactor operating temperatures (820K) will lead to standardisation for the adoption in the series of future reactors.

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Energy Procedía 7 (2011) 186-198

Asian Nuclear Prospects 2010

1876-6102 © 2011 Published by Elsevier Ltd. Selection and/or peer-review under responsibility of Indra Gandhi Centre of Atomic Research doi:10.1016/j.egypro.2011.06.025

It is planned to construct six more 500 MWe Commercial FBRs (CFBRs), starting in 2014, with improved economy and enhanced safety. Beyond this, 1000 MWe metallic-fuel based FBRs will be constructed in India starting in 2025. The development of sodium cooled fast reactors necessitates extensive research and development in domains of materials, manufacturing and inspection technologies in close association with almost all the disciplines of science and technology.

The operating environment in sodium-cooled fast reactors demands radiation resistant in-core materials satisfying stringent requirements [2]. The in-core materials operating at a temperature of 673-973K in sodium environment experience high neutron flux levels (8*1015 n/cm2/sec). The neutron flux at the core centre with neutron above 0.1 MeV energy is 4.7*1015 n/cm2/sec. Hence, the in-core materials need to possess limited void swelling, desired creep resistance and acceptable resistance to fracture at end of life and during fuel handling at the back-end of the fuel cycle. Keeping these stringent property requirements of in-core material in view, a 15Cr15Ni2.2Mo-Ti austenitic stainless steel (Alloy D9) has been chosen as the in-core material for cladding tube and hexagonal wrapper for PFBR [3]. However, for achieving higher fuel burn up in future CFBRs, improved materials for clad tubes are being developed, which include Alloy D9I with higher phosphorus and silicon contents, and ferritic/martensitic oxide-dispersion strengthened (ODS) steels. Plain 9Cr-1Mo steel for the wrapper and modified 9Cr-1Mo steel for the cladding are also being developed for FBRs with metallic fuels.

The elevated operating temperatures of PFBR demand the structural materials to have adequate creep and fatigue properties as well as good corrosion resistance. In addition to this, the structural materials used in sodium circuit side require compatibility with sodium whereas the material to be used in steam generator requires the compatibility with both sodium and water media. 316L(N) austenitic stainless steel (SS) has been chosen as the major structural material for reactor vessels and primary and secondary loops of PFBR whereas modified 9Cr-1Mo steel is the main structural material for steam generators. Other improved materials being developed for the CFBRs include nitrogen-enhanced 316LN SS for increasing the design life of reactor components, and type-IV cracking resistant boron-added modified 9Cr-1Mo steel for the steam generators.

The entire range of manufacturing technologies for PFBR components has been successfully developed and implemented in collaboration with academic and research establishments and Indian Industries. In the case of reactor vessel, to achieve the strict quality standards of formed parts in terms of geometrical tolerances, residual stresses and microstructural defects, FEM-based modelling and experimental validation was carried out for estimation of spring-back during forming of multiple curvature thick plates. This model enabled design of the tooling for manufacture of reactor vessels with desired dimensional accuracy. Optimisation of grain boundary character distribution in Alloy D9 used for cladding tubes and wrapper, was carried out by adopting the grain boundary engineering approach to reduce radiation induced segregation.

Extensive welding is involved in the fabrication of reactor vessels, piping, steam generators, fuel subassemblies etc. Activated Tungsten Inert Gas Welding process along with activated flux developed at IGCAR has been successfully used in fabrication of dummy fuel subassemblies (DFSA) required for testing. To improve the resistance to high temperature wear, especially galling, hardfacing by weld deposition over surfaces of various components was necessary. Based on radiation dose rate and shielding considerations during maintenance, handling and decommissioning, nickel-base hardfacing alloy was chosen to replace the traditionally used cobalt-base Stellite alloys.


V 650" 6008 550-a> . tn 500-

I 450>- "

400350300200 300 400 500 600 700 800 900 1000 1100

Test Temperature (K)

Fig. 1. Yield stress value of alloy D9 in 20% cold worked condition

20% cold worked condition

A number of advanced non-destructive evaluation procedures were developed and implemented for inspecting various manufactured components, notable among them being the tube-to tube-sheet welds in steam generators, the clad tube and the end-cap welds, the weld between the hexagonal wrapper and the foot of the fuel sub-assemblies, etc. This paper highlights the work and the approaches adopted for the successful development of materials, manufacturing and inspection technologies for the in-core and structural components of current and future Indian Fast Breeder Reactor Programme.

2. Materials and manufacturing technology for fast breeder reactors

2.1. Alloy development

Clad and wrapper materials for sodium cooled fast reactors (SFRs) have undergone continuous evolution over the years. The first generation materials belonged to austenitic stainless steels 304 and 316 grades. The next development of radiation resistant austenitic stainless steels involved increasing the nickel content and decreasing the chromium content in comparison to standard 304 SS and 316 SS

Fig.2. Process flow sheet for manufacturing ODS clad tube.

along with addition of titanium Simultaneously, solute elements like silicon, phosphorous, niobium, boron and carbon have been found to strongly influence the swelling behaviour. This realisation has led to the development of advanced core structural materials such as alloy D9 which is currently used in PFBR. The melting and processing variables for commercial productions of alloy D9 were optimized based on the experience with laboratory heats [3]. Commercial scale melting of three heats with Ti/C ratios of 4, 6 and 8 have been made at MIDHANI, Hyderabad, from virgin raw materials through VIM with ladle adjustment to achieve desired Ti/C ratio. This was followed by vacuum arc re-melting. Rounds of 30 mm diameter were produced through hot forging and hot rolling. These rounds were reduced to 11 mm diameter by cold swaging. Tensile tests were carried out using 20% cold worked samples for this commercial scale melting. The value of yield stress of the commercially melted alloy D9 with Ti/C of 6, at various test temperatures is shown in Fig.1. The yield stress of the 20°% cold worked sample with Ti/C of 6 is well within the acceptable limits of PFBR specifications (575-750 MPa at room temperature and 425-625 MPa at 723K).

Cold worked alloy D9 has reached a record dose of 140 dpa without excessive deformation. In the light of recent international alloy development efforts, and in-reactor experience, a modified alloy D9 is being developed to realise still higher swelling resistance, by adding phosphorous in the range of 0.025 to 0.04 wt% and silicon in the range of 0.7 to -0.9 wt% to the alloy D9 composition. The long-term solution for CFBR core structural materials is considered to be 9-12% Cr ferritic-martensitic (F-M) steels. These steels promise excellent swelling resistance to doses even upto 200 dpa. However, the increase in ductile to brittle transition temperature (DBTT) due to irradiation is a cause of concern during fuel handling operations. A significant increase in toughness can be realised by avoiding the formation of delta-ferrite and ensuring fully martensitic structure as well as by optimising the austenitizing temperature to refine the prior austenite grain size with strict control on embrittling tramp elements. These materials are promising for wrapper applications in CFBRs. The creep resistance of FM steels decreases significantly above 823K and hence these steels are not suitable for clad applications. Oxide dispersion strengthening is a promising means of improving the creep resistance of F-M steels. This has led to the design of oxide dispersion strengthened (ODS) F-M alloy, with 0.3 to 0.4 % yttria (Y2O3) particles of 3-5 nm size. Based on a thorough literature survey describing international experience related to manufacturing of ODS alloy, the manufacturing route has been formulated (Fig.2). The ODS alloy clad tubes have been manufactured with the help of The International Advanced

Processing conditions Processing condmons

Fig.3. Frequency distributions of (a) Z 3 boundaries [average grain size (with and without twins) in each condition is also indicated]; (b) Z9 and Z27 boundaries [(Z3/Z9+Z27) is also shown to assess random boundary connectivity] in 15% deformed and annealed samples.

Research Centre for Powder Metallurgy and New Materials (ARCI) and Nuclear Fuel Complex (NFC). The comprehensive evaluation of the produced tubes is currently under investigation.

2.2. Grain boundary engineering in alloy D9

Radiation-induced segregation (RIS) is an important consequence of irradiation at elevated temperatures where point defects are mobile [4]. RIS is a non-equilibrium process that happens due to the unequal participation of solutes in the vacancy and interstitial defect fluxes [5]. The grain boundary misorientation significantly influences RIS. It has been established in austenitic stainless steels under 5 MeV proton irradiation at elevated temperature that the RIS is insignificant at a small angle tilt boundary or low 2 coincidence site lattice (CSL) boundary but not so at a random grain boundary [6].

Grain boundary engineering in alloy D9 was realized through a one step thermo-mechanical treatment [7], in an attempt to reduce RIS. The experimental methodology adopted in this investigation was based on the strain-annealing approach in which small amount of strain (5-15°%) was imparted on solution annealed (SA) sample. The cold deformed samples were subsequently annealed at various

Fig.4.(a) annealed at 1273K for 1 hour with 15% pre-strain (Colour code: Z 3 - red; Z 9 - blue; Z 27 - green; Z1 -yellow; other low Z CSL - purple, random HABs - black), (b) grain boundary map showing only the random HABs; (c) Kernel average misorientation (KAM) map of GBE alloy D9 after 4th iteration .

temperatures (1173-1273K) for different time periods (0.5-2h). It has been observed that annealing following 5°% deformation results only in a moderate increase in proportion of S3 boundaries and its variants. The grain boundary character distribution (GBCD) data in terms of S3n boundaries in samples annealed in the temperature range 1173-1273K with 15% pre-strain is shown in Fig.3. It could be observed that a very high fraction of S3n boundaries have been achieved after annealing at 1273K with 15% pre-strain (Fig.3(a) and Fig. 3(b)). The increase in S3 boundaries could be attributed to the formation of new S3 boundaries by geometrical interaction between pre-existing S3s (i.e. multiple twinning) and/or formation of new twins during annealing following deformation. According to the CSL rule, interaction between two S3 boundaries may form a S9 boundary according to the relationship S3 + S3 = S9 [8]. When a S9 boundary generated in this way encounters another S3 boundary, it results in either a new S3 boundary based on the relationship S3 + S9 = S3, or a S27 boundary following the relationship S3 + S9 = S27. However, from the statistics of the S3, S9 and S27 (Fig. 3(a) and Fig. 3(b)), it could be inferred that S3 + S9 = S3 occurs more frequently than S3 + S9 = S27 [9]. These regenerated S3, S9 and S27 boundaries take part in the reconfiguration of the existing grain

boundary network that eventually breaks down the random high angle boundaries (HABs) connectivity (Fig.4(a)). The ratio of 23 to (29 + 227) in these conditions is shown in Fig. 3(b). A lower 23 to (29 + 227) ratio in the samples annealed at 1273K (Fig. 3(b)) signifies that a significant disruption of random HABs connectivity has happened (Fig. 4b).

Apart from the GBCD, the degree of cold work which influences dislocation density is an important parameter determining the void swelling behaviour and RIS of austenitic stainless steels. A 20% cold work (CW) has been found to be optimum in terms of void swelling and phase stability in alloy D9.

Fig.5. Hardness profiles for Colmonoy Fig.6. Hardfacing of grid plate sleeve (ID ~80 mm) deposits by GTAW and PTAW welding using indigenously designed PTAW torch. processes

Cold working up to ~10% in the GBE alloy D9 has not found to alter the misorientation of the 23 boundaries much since majority of the 23 boundary (~90°%) are within 2° from their ideal misorientation1. To achieve higher retained strain (similar to 20°% CW) in the micros tructure, a four-step iterative thermo-mechanical treatment has been adopted for alloy D9. In each processing cycle,

Fig.7. Grid plate mounted on the furnace with hardfacing in progress.

rolling with a thickness reduction of 10°% was carried out at ambient temperature, and the subsequent annealing was performed at 1273 K for a shorter annealing time (~30 min) [11]. This has resulted in very high fraction (~63°%) of 23 boundaries with substantial retained strain as revealed by kernel average misorientation (KAM)2 map (Fig. 4(c)). The KAM map can be used to evaluate localized

1 The allowable deviation for 23 boundaries as per Brandon's criterion [10] is 8.6°.

2 The color-scale-coded KAM map is plotted on the basis of first-neighbor kernel parameter with a maximum misorientation angle of 5For a given data point, it calculate the average misorientation between the data point and all of its neighbors (exclude misorientations greater than some prescribed value - 5° in this case).

plastic strain in materials [12]. The higher KAM values in the grains correspond to higher local misorientations (i.e. higher retained strain and dislocations) and vice versa. Such a GBE microstructure with further -10%% CW is likely to reduce RIS and void swelling in alloy D9 [6]. The detailed electron microscopy and irradiation study is under progress.

2.3. Hardfacing of the reactor components

The liquid sodium, coolant used in fast breeder reactors with controlled oxygen (<3 ppm), has a tendency to remove the oxide film on the component surfaces and expose virgin metal. Many of these components would be in contact with each other or would have relative motion during operation, and their exposure at high operating temperatures (typically 823K) coupled with high contact stresses could result in self-welding of the clean metallic mating surfaces. In addition, the relative movement of mating surfaces could lead to galling, a form of high-temperature wear, in which material transfer occurs from one mating surface to another due to repeated self-welding and breaking at contact points of mating surfaces. Further, susceptibility to self-welding increases with temperature for AISI 316 stainless steel [13]. Hardfacing of the mating surfaces has been widely used in components of liquidsodium cooled FBRs to avoid self-welding and galling [14]. Many challenges have been undertaken at IGCAR to evolve a robust hardfacing strategy for the components of PFBR. At first, based on radiation dose rate and shielding considerations during maintenance, handling and decommissioning, Ni-base Colmonoy was chosen to replace the traditionally used Co-base Stellite alloy for hardfacing [15]. Next challenge to adequately qualify and characterise this Ni-base alloy, having poorer weldability and high-temperature stability than the Co-base alloy, was also addressed in a comprehensive manner. The choice of appropriate deposition process is a must for success. Metallurgical studies revealed that the hardness and microstructure of the gas tungsten arc welding (GTAW) deposit of the Ni-base hardfacing alloy is significantly affected by dilution from the base metal with the width of the softer dilution zone (of up to 2 mm) often exceeding the final desired hardface deposit thickness. Also, certain components like grid plate sleeves of about 80 mm ID, that required hardfacing deep inside the inner surface of the sleeve were not amenable for hardfacing by conventional processes like GTAW process unless major design concessions were permitted. Hence, the more versatile plasma transferred arc welding (PTAW) process was chosen so that the width of the dilution zone could be controlled to as low as 0.2 mm (Fig. 5), by optimising the deposition parameters. An indigenous fabricator (M/s Omplas, Chennai), who carried out the hardfacing of components for the 500 MWe FBR Project, designed and developed a suitable miniature PTA torch (Fig. 6) for hardfacing the inner surface of grid plate sleeves.

2.4. Hardfacing of bottom plate of grid plate assembly

Grid plate of PFBR is a massive structure consisting of two plates (top and bottom of ~6.5 m in diameter). Grid plate houses a large number of sleeves in which foot of the sub assemblies rest. The grid plate assembly in turn rests on core support structure that also acts as boundary between cold and hot sodium in the reactor. Both the grid plate assembly and the core support structure are made of AISI 316L(N) stainless steel and immersed in flowing sodium and remain in contact throughout the reactor life (60 years). Hence, there should not be any self welding between these two components at their contact location. Hardfacing is carried out on two annular grooves machined on the bottom plate of the grid plate. These grooves are located towards the periphery of the grid plate and hence, diameters of these groves are close to that of the grid plate itself and total length (circumference) of single hardfaced deposit was close to 21 m

Fig.8. Ultrasonic surface wave based methodology for non-destructive inspection of hexcan welds of PFBR subassemblies (a) hexcan weld (b) a schematic of the mock up weld, and (c) a mock up weld with reference defects (top) and the corresponding ultrasonic images of the defects (bottom).

During technology development of the grid plate, extensive cracking of the deposit is observed when deposition was carried out as per the procedure finalized initially based on trails carried out on 80 mm thick 1000 mm diameter plate. Subsequently, a detailed review of the design of the groove, welding process, procedure, heat treatment etc. was taken up in which designers, materials engineers, hardfacing agency and manufacturer of the grid plate participated. The groove width was reduced from 45 to 20 mm and the groove angle increased from 30° to 60°. This enabled us to carry out hardfacing in single layer and single pass of deposition. Preheat temperature for deposition was increased from 773K to 923K and the furnace for preheating and stress relieving heat treatment was modified to ensure that the temperature variation across the component during heating or cooling is reduced considerably. It was also decided to carry out hardfacing continuously using four PTAW machines positioned on a circular track which was concentric to the bottom plate. Machine controls were suitably modified to have smooth deposition between starting and ending locations of the deposit, which is found to be more prone to cracking than the other locations of the deposit. The process has been successfully employed for the bottom plate of the grid plate assembles of PFBR. After bottom plate reached preheat temperature, the entire operation of hardfacing of the two grooves of ~6.5 m dia. each took only a few hours. Figure 7 shows the grid plate mounted on the furnace bed with hardfacing operation in progress. Neither cracks, nor de-bonding nor surface porosities were observed on the deposits.

2.6. Non-destructive evaluation

The structural integrity of reactor core and other safety components including fuel assemblies, reactor vessels, steam generators etc. need to be ensured through stringent quality evaluation and periodic inservice inspections. Advanced ultrasonic and eddy current based NDE procedures have been developed for inspection of structural and core components of the reactor and steam generator components, as part of fabrication quality assessment.

Due to the complex geometry of hexcan weld of the PFBR fuel subassembly, radiography testing of the weld cannot provide required sensitivity during the fabrication stage. Hence, a new ultrasonic methodology has been developed for testing of the hexcan weld (Fig. 8). The developed methodology involves testing the weld from the thinner plate side using 1 MHz Rayleigh wave, which penetrates the complete thickness (3 mm) of the weld (approx one wavelength). This methodology is quite fast and can be used for detection of both axial and circumferential defects.

Fig.9. Integrity Assessment of TTS Weld Joints in Steam Generators of PFBR (a) rod anode based micro-focal radiography set-up and (b) Radiography image of a typical TTS weld joints showing penetrameter wires and porosity.

The importance of high integrity welds in steam generators is due to risks arising out of sodium-water reaction in the event of breaching the integrity of steam and water boundary. Tube to tube-sheet weld joints are the regions where the possibility for a leakage path is highest. The main defects in the weld joints are porosities, concavities or convexities. It is expected that the radiography technique should be able to detect single porosities of size at least of the order of 50 (im In the entire weld, the total pore count must be such that the sum of diameters of all the pores visible is less than 2.54 mm This leads to the necessity of high sensitivity defect detection in these welded joints. In this regard, rod-anode based micro-focal radiography (Fig. 9(a)) technique and appropriate digital image processing methodology has been developed and standardized [20]. More than 100 trial weld joints were radiographed to establish the procedure for quality assurance. Results on the trial welds of these tubes have shown that it is possible to resolve a 32 micron diameter steel wire placed on the inside of the tube (Fig. 9(b)). This corresponds to a sensitivity of 1.3%-1.6% of the wall thickness of the tube. Radiographs taken during the developmental stage of the welding have given feedback to arrive at correct weld parameters for acceptable weld joints. This procedure has been successfully employed for the steam generators for PFBR.

3. Materials and coatings technology for back end of fuel cycle

In the spent nuclear fuel reprocessing plant, the integrity, safety, economy and uninterrupted operation depends on the quality and performance of critical engineering components used for vessels and piping [21-23]. A typical flow sheet adopted for a Purex process based fast reactor reprocessing plant is shown in Figure 10.

Fig.10. Overview flow sheet of FBR PUREX process.

Nuclear reprocessing plants involve equipment/vessels such as fuel dissolvers, evaporators for various purposes and high active raffinate waste storage tanks [1,2]. Nitric acid is used in various conditions from dilute (1-4 N) to concentrated (10-14 N), room temperature (solvent extraction) to intermediate (warm, raffinate waste storage tanks) to boiling temperature (dissolver, evaporator). The presence of redox species in dissolvers, partitioners etc) is an added dimension

and the corrosion problems are more severe during dissolution of fast reactor fuel due to large yields of fission products and trans uranium in the spent fuel as well as the presence of complex organic species after dissolution (specific to carbide fuels). The reprocessing plants are designed with the objective of zero incident failures as leakages in pipes, vessels and equipment could considerably delay the restarting of the operation since the repair and maintenance in high activity area is never a planned activity. The selection of a material for nuclear reprocessing plant is a complicated process involving many parameters such as corrosion resistance, mechanical properties, weldability, availability, cost, etc. The presence of intense radiation, particularly gamma, and of radioactive contamination of surfaces, usually alpha radiation emitting particles, during dissolution of spent fast reactor fuel is an additional factor of concern. The performance of various materials of construction in reprocessing plants and also the new materials and coating technology being developed for use in future FBR reprocessing plants are briefly highlighted.

3.1 Materials selection issues for closed fuel cycles

In the context of closed fuel cycle, any designed increase in burn-up or fuel alters the chemistry of the spent fuel and hence, the reprocessing procedure or waste management strategies. For severe conditions of dissolution involved in the head end process, the materials of construction and electrodes

Fig.11. AFM topography showing surface morphology after passive film dissolution in 304L SS [24].

play a major role in meeting the demand of zero failure concept and minimum maintenance of the plants. For the dissolution of spent (U, Pu) C fuel from fast breeder test reactor (FBTR) at Kalpakkam, boiling 11.5 N HNO3 with Ag+/Ag2+ redox ion under electrolytic condition is employed. Used of advanced nitric acid grade (NAG) SS with controlled chemical composition of impurities like S, B, P etc., and with higher Si, Cr etc. have also been found to undergo IGC under such aggressive nitric acid conditions [21, 22]. The kinetics of such surface dissolution, passive film formation, growth and destruction (Figure 11) depends on the concentration of HNO3 employed in reprocessing applications [24]. Studies have been carried out to explore valve metals (Ti, Zr, Hf, Nb and Ta) and its alloys (Ti-5%Ta, Ti-5%Ta-1.8%Nb etc) as alternate to austenitic SS for use in nitric acid application. A Ti-5%Ta-1.8%Nb alloy has been developed as a candidate material for structural applications, in view of its excellent resistance to three phase corrosion in highly oxidizing media (Figure 12).

TiTaNb alloy three phase corrosion rates

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830WQ 810AC 950WQ 950AC 1000WQ 1000AC WELD

below beta transus above beta transus H Liquid H Vapour □ Condensate

Fig. 12. Corrosion rate of Ti-5%Ta-1.8%Nb alloy in different material conditions in three phase corrosion studies.

Similarly, the weldability of the Ti-Ta-Nb alloy was excellent and the weldments showed corrosion resistance «< 1mpy. The Ti-5%Ta-1.8%Nb alloy consisted of a micro structure of alternate lamellae phases, with the solutes Ta and Nb repartitioned into the phase. It was necessary to distribute the high temperature P phase uniformly in the matrix of polygonal a, by suppressing the martensitic or Widmanstatten or Massive transformations in the alloy. Based on the extensive research and the international experience, Ti and Ti-5%Ta-1.8%oNb alloy have been chosen in India as candidate material for electrolytic dissolver applications in fast reactor spent fuel reprocessing plants [21, 22, 25]. Zirconium is highly resistant to corrosion attack in nitric acid environments. With a proper design, components made of zirconium can withstand highly concentrated nitric acid. For severe corrosive environments of dissolvers in 11.5 M HNO3 Zircaloy-4 has also been proposed for the construction of future dissolver based on the vast experience on zirconium based alloys in French reprocessing plants. Corrosion rate data generated for both Zircoloy-4 base and weld samples indicated the superior corrosion resistance in all three conditions compared to Ti alloys (Figure 13).

In addition, efforts were made to deposit nanostructured titanium and TiO2 by magnetron sputtering process. Preliminary results indicated improved corrosion resistance after coating in nitric acid medium Efforts were made to develop nanotube TiO2 coatings on titanium for enhancing the catalytic activity electrodes used in reprocessing plant. Parameters to develop such nanotube coatings with high surface area have been optimized and the developed surface has been characterized. Similarly, attempts were made to develop superhydrophobic surfaces resistant to nitric acid environments over titanium (Figures 14 a & b). Also, nanoporous surface with high surface area has been prepared with bulk metallic Zr-based alloys for nanofiltration and catalytic purposes. Electrochemical and surface investigation of zirconium based metallic glass Zr59Ti3Cu2oAl1oNi8 alloy in nitric acid and sodium chloride media was also studied as a candidate coating material.

3.2. Dynamic nitric acid loop and zircaloy-4 high temperature corrosion testing system

A dynamic nitric acid loop (NAL) of 400 L capacity was designed at IGCAR, Kalpakkam, made of AISI type 304L SS, with flowing 6 N HNO3 (inactive conditions) at different temperatures for evaluating the corrosion performance of materials over long operating periods. The typical flow sheet of the dynamic nitric acid loop used is shown in Figure 15a. The corrosion behaviour of AISI type 304L SS used in the demonstration fuel reprocessing plant (DFRP) and the nitric acid grade (NAG) type

Fig.13. The average corrosion rate of candidate materials in liquid, vapour and condensate phases in 11.5M boiling nitric acid for 240 h.

304L SS are evaluated at different temperatures (40°C, 60°C, 80°C, 107°C (boiling) and vapour phase). Experiments were carried out for 100, 250, 500 and six 1000 h, and the results indicated that the DFRP alloy exhibited a corrosion rate of 13.8 mpy after 8000 h in boiling condition compared to 7.5 mpy of NAG alloy. The study is being continued for upto 10000 h total duration in order to generate the corrosion rate data. The availability of such a reliable and useful data will be helpful in predicting the remnant life of the components used in the reprocessing plants by developing suitable analytical and modeling tools of the corrosion processes.

Zircaloy-4 based high temperature corrosion testing system of 10 litres capacity (Figure 16b) was made with provisions for testing under liquid, vapour and condensate conditions [26, 27]. The system has been operated for 3000 h with 11.5 M nitric acid. Studies conducted so far indicated corrosion rate values below 1 mpy for Zircaloy-4 samples (Figure 12).

Fig.14. (a) SEM micrograph of SHP coated Ti and (b) Nyquist plot behavior of measurement in 0.5N HNO3

Based on the results obtained a prototype dissolver vessel similar to the design of the operating dissolver at CORAL plant was manufactured for long term corrosion evaluation under simulated dissolver conditions. Dissimilar joining of Zircaloy-4 with type 304L SS was achieved using friction joining process and was characterized for integrity and corrosion resistance. The better corrosion behaviour of friction welded Zircolloy-4 to type 304L SS joints was observed when there are no intermetallic compounds at the interface [21, 17].

3.3. Materials development for molten salt based pyrochemical reprocessing applications

For reprocessing of spent metallic fuels, electrorefining method and a fuel fabrication system with a metal injection casting method have been chosen internationally as the promising route.

Fig. 15. (a). Schematic flow sheet showing the various parts of dynamic nitric acid loop and (b) Zircolloy-4 high temperatures corrosion testing system.

The electrorefining process has been carried out in an electrorefiner that contains a molten chloride salt (LiCl-KCl) floating on liquid cadmium operating at 773 K under an argon atmosphere. The sheared spent fuel charged in anode baskets of the electrorefiner was refined in the molten salt while uranium was recovered at a solid cathode and U-Pu-MA (minor actinides) were collected at a liquid cadmium cathode. Corrosion resistance of the materials in molten chloride salts at high temperature is of prime importance for equipments like electrorefiner and salt purification vessel in pyrochemical reprocessing plants. In order to overcome the corrosion of metallic materials in molten chloride environment, efforts were made to select materials and coating technology useful for various unit operations like salt

Fig.16. (a). Corrosion attacked observed in 316L SS after exposer in LiCl+KCl and (b): Insignificant corrosion attacked observed in TBC-YSZ coated sample.

The various materials like 410 SS, 430 SS, 316L SS, Inconel 600, 625 etc. are tested for salt preparation, electrorefining, cathode processing, waste processing etc. as containers and associated accessories, and C-based high density graphite, pyrographite etc. as crucible and electrode materials. Casting of type 316L SS main vessels, electroforming of Ni and Ni-W coated crucibles, ceramic coatings of yttria stabilized zirconia, and PVD based nanostructured nitride coatings are some of the developments carried out in association with industry, academic and R&D institutions. Molten salt test assembly, double modular glove box, thermal cycling system, high temperature electrochemical system etc. have been established for testing purpose. Based on the literature, coating of 7-9 wt% yttria-stabilized zirconia on type 316L SS has been choosen as one of the candidate materials for the salt purification vessel and electrorefiner. Plasma spray zirconia coatings applied over type 316L SS have been chosen for corrosion investigations in a molten chloride medium [28]. Also attempts to laser remelt the as-coated plasma sprayed partially stabilized zirconia (PSZ) surface on type 316L SS was carried out at varying laser powers to achieve a smooth and pore free surface and characterized for the properties. The plasma sprayed yttria stabilized zirconia coating on type 316L SS exhibited better corrosion resistance (Figure 17 a&b), and laser remelting was established to eliminate microstructural inhomogeneities like pores and voids present in the coating [28]. Carbon-based coatings are also being developed for various applications.

4. Conclusion

A few illustrative examples pursued for the successful deployment of materials, manufacturing and inspection technologies for the current and future Indian Fast Breeder Reactor Programme have been addressed. Research, development and deployment of materials, manufacturing and inspection technologies has been enabled by robust design, extensive and innovative modelling, characterisation, testing and evaluation of materials, improvements in manufacturing processes, and validation of technologies by demonstration with real-life structures and components. The extensive collaboration among academic, research and industrial organisations with 'science-based technology' approach is the guiding principle for the successful development of materials and manufacturing technologies for the Indian Fast Breeder Reactor Programme.


Author is thankful to a large number of colleagues in Metallurgy Group and collaborators from academic, research and industrial organizations for their contributions.


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