Development of fast breeder reactor technology in India

Progress in Nuclear Energy xxx (2017) 1e24

Contents lists available at ScienceDirect

Progress in Nuclear Energy journal homepage: www.elsevier.com/locate/pnucene

Development of fast breeder reactor technology in India P. Puthiyavinayagam*, P. Selvaraj, V. Balasubramaniyan, S. Raghupathy, K. Velusamy, K. Devan, B.K. Nashine, G. Padma Kumar, K.V. Suresh kumar, S. Varatharajan, P. Mohanakrishnan**, 1, G. Srinivasan 2, Arun Kumar Bhaduri Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamilnadu, India

a r t i c l e i n f o

a b s t r a c t

Article history: Received 23 December 2016 Received in revised form 10 March 2017 Accepted 13 March 2017 Available online xxx

India is pursuing a three stage nuclear power program employing its natural uranium and Thorium reserves. FBRs form the second stage of the program linking the first phase with natural U and third phase with Th fuels. The saga of FBR technology development in India is presented in this paper. The valuable experience in design and operation of Fast Breeder Test Reactor (FBTR) is recounted briefly. The R&D for the techno-economic demonstration of Prototype Fast Breeder Reactor (PFBR) is explained including technology development for addressing manufacturing challenges and engineering full-scale qualification. Finally a brief projection is made on the developments for FBR600 with improved safety and economics as well as metal fuel technology for future experimental and power reactors to follow. © 2017 Elsevier Ltd. All rights reserved.

Keywords: Nuclear energy Fast reactor Breeder Neutronic performance Fuel cycle Metal fuel Technology development Sodium facilities Safety studies Material development

1. Introduction The indispensability of Fast Breeders for the sustained growth of nuclear energy in India was realized by Dr. Homi Bhabha, even in the 1950s. He recognized that the limited reserves of Uranium in India cannot sustain the nuclear energy programme for long, and gave out the breeder route for the efficient use of Uranium. In the

Abbreviations: FBR, Fast Breeder Reactor; FBTR, Fast Breeder Test Reactor; PFBR, Prototype Fast Breeder Reactor; DAE, Department of Atomic Energy; IAEA, International Atomic Energy Agency; CEA, French Alternative Energies and Atomic Energy Commission; IGCAR, Indira Gandhi Centre for Atomic Research; LHR, Linear Heat Rating; DHR, Decay Heat Removal; DND, Delayed Neutron Detection; SGDHR, Safety Grade Decay Heat Removal System; OGDHR, Operational Grade Decay Heat Removal System; IFTM, Inclined Fuel Transfer Machine; SADHANA, SAfety Grade Decay Heat removAl loop in Natrium; CDA, Core Disruptive Accident. * Corresponding author. ** Corresponding author. E-mail addresses: [email protected] (P. Puthiyavinayagam), kpmkris@gmail. com (P. Mohanakrishnan). 1 Formerly with IGCAR, Kalpakkam. 2 Raja Ramanna Fellow, IGCAR.

long run, Thorium, which India has in abundance, was slated to play a major role in India's Nuclear Programme (Fig. 1) (Kakodkar, 2008). The Department of Atomic Energy (DAE) has been actively pursuing the well known ‘three stage nuclear programme’ chalked out by Dr. Bhabha (Venkataraman, 1994). Even when the major thrust of R&D activities in Bhabha Atomic Research Centre (BARC) was on the first stage of Indian nuclear power program, studies on the second stage had been initiated in parallel. Fast Breeder Reactor (FBR) technology in India had its roots in the several theoretical studies carried out in 1968 on several design options for fast reactors in the Indian context. It was concluded that for FBRs, carbide or metal fuel will be best suited for high breeding in U-Pu fuel cycle. Before launching on a large scale indigenous programme on fast reactors, however, there was an obvious need for a test reactor aimed at providing experience in fast reactor operation and large scale sodium handling. The reactor was also required to serve as a test bed for irradiation of fast reactor fuels and materials. The need for comprehensive R&D on all aspects of fast reactors, up to the closing of the fuel cycle, was also recognized.

http://dx.doi.org/10.1016/j.pnucene.2017.03.015 0149-1970/© 2017 Elsevier Ltd. All rights reserved.

Please cite this article in press as: Puthiyavinayagam, P., et al., Development of fast breeder reactor technology in India, Progress in Nuclear Energy (2017), http://dx.doi.org/10.1016/j.pnucene.2017.03.015

2

P. Puthiyavinayagam et al. / Progress in Nuclear Energy xxx (2017) 1e24

Fig. 1. Nuclear energy scenario of India.

1.1. Indira Gandhi Centre for Atomic Research An R&D centre, exclusively devoted to the development of Sodium-cooled Fast Reactor (SFR) technology in India was established in 1971 at Kalpakkam, where construction of the first indigenous twin units of 235 MWe Pressurized Heavy Water Reactors was in progress. It was initially christened as 'Reactor Research Centre’. At the heart of the centre was proposed a sodium cooled test reactor, named Fast Breeder Test Reactor (FBTR), which would serve as a test bed for irradiation of fuels and materials and provide experience in large scale sodium handling and reactor operation. An agreement was signed with French Alternative Energies and Atomic Energy Commission (CEA) for transfer of the design of the Rapsodie reactor, training of personnel in Rapsodie and transfer of manufacturing technology of critical components. Parallel to the construction of Fast Breeder Test Reactor, engineering halls with sodium loops for component testing, hot cells for Post-Irradiation Examination and laboratories with advanced facilities for materials and metals research, safety studies, sodium and fuel chemistry studies, development of instrumentation and R&D on reprocessing were also established. Thus the entire gamut of R&D for the second stage of our programme was included in the mission of the centre. The Reactor Research Centre (RRC) was renamed as Indira Gandhi Centre for Atomic Research (IGCAR) in 1985. This paper focuses primarily on the reactor design and technology development. Other aspects such as fuel fabrication, reprocessing and waste management are covered in the companion papers (Setty et al., 2017; Natarajan, 2017; Wattal, 2017). 2. Fast Breeder Test Reactor FBTR is a loop type, sodium cooled fast reactor (Suresh Kumar et al., 2011). Though adapted from Rapsodie, FBTR has several design modifications. The major change was the incorporation of a steam-water circuit in place of the sodium-air heat exchangers deployed in Rapsodie. The steam-water circuit has four oncethrough steam generator modules of the design used in the French Phenix Reactor. The construction of FBTR was started in

1972, and civil works were completed by 1977. Most of the components were installed in 1984. The reactor was made critical on 18th October 1985. Fig. 2 shows the schematic flow sheet of the heat transport circuits and the temperatures shown are as per the original design at the rated conditions of operation. Heat generated in the reactor is removed by two primary sodium loops and transferred to the corresponding secondary sodium loops through Intermediate Heat Exchangers. Each secondary sodium loop is provided with two once-through steam generator modules. Steam from the four modules is fed to turbine-generator (TG). A Dump Condenser of 100% capacity is also provided in the steam water circuit for continued reactor operation when TG is not available. 2.1. The carbide fuel Rapsodie used MOX fuel with 30% PuO2 and 70% UO2 (with the latter enriched to 85%) as the driver fuel. This design was originally proposed for the core of FBTR. In view of the non-availability of highly enriched uranium, studies on MOX fuel with 70% PuO2 & 30% UO2 (natural uranium) were carried out, which indicated poor performance in terms of the linear heat rating, swelling and compatibility with sodium. Uranium mono-carbide fuels had been tested, on a limited scale, in many fast reactors all over the world. Mixed carbide had, however, not been used in any of the reactors as a driver fuel. Mixed carbide with high Pu content had not been tested in any reactor. Nevertheless, based on out-of-pile studies, it was decided to go in for Pu rich carbide MK-I fuel (70%PuCþ30%UC) as the driver fuel for the first core of FBTR. Being a fuel without any irradiation data, the core was made small. The first criticality was achieved with a small core with 23 subassemblies of MK-I carbide fuel as against the originally intended core with 65 MOX fuel subassemblies (Srinivasan et al., 2006). The initial target burn-up was set at 25 GWd/t. The operating Linear Heat Rating (LHR) of the MK-I fuel was conservatively set at 250 W/cm. The reactor power was limited to 8 MWt for the 23 subassemblies (SA) core. Out-of-pile tests by electrical resistance heating up to 620 W/cm showed no sign of melting, giving the scope for raising the LHR to 320 W/cm. With the

Please cite this article in press as: Puthiyavinayagam, P., et al., Development of fast breeder reactor technology in India, Progress in Nuclear Energy (2017), http://dx.doi.org/10.1016/j.pnucene.2017.03.015

P. Puthiyavinayagam et al. / Progress in Nuclear Energy xxx (2017) 1e24

3

Fig. 2. Heat transport circuits of FBTR.

increased LHR, the reactor power was raised to 10.2 MWt in 1993. Post-irradiation examination (PIE) at 25 GWd/t indicated that the fuel swelling rate is similar to the values reported in literature for low Pu carbide fuels (20% PuC). The burnup limit was progressively increased through detailed analysis supported by PIE conducted with SA discharged at 50 GWd/t, 100 GWd/t and 155 GWd/t peak burnup. As shown in Fig. 3, PIE at 155 GWd/t indicated that the fuel had entered a restrained swelling phase, with the porosities exhausted. This value was set as the limiting burn-up for the MK-I fuel at 320 W/cm. The reactor operation demonstrated that MK-I fuel could operate at 400 W/cm peak LHR after fuel-clad gap closure (Srinivasan et al., 2010a). One SA was irradiated up to 165 GWd/t and it withstood this high burnup without failure. Though it may be possible to raise the burn-up slightly more, considering the fact that the clad ductility is low, it was decided to stop the endurance test. As a safe measure, it has been decided to stick to 155 GWd/t as the allowable burn-up for the MK-I fuel. So far, except for the failure of one pin at a burn-up of 146 GWd/t, more than 1400 pins have reached 155 GWd/t without any clad failure. Incidentally, cold worked SS-316 & 316 L have been taken to the highest levels of damage vis- a-vis published literature (~80 dpa).

2.2. Evolution of FBTR core and power with time FBTR has the unique feature of being a reactor where the excellent burn-up capability of its driver fuel has been demonstrated (Varatharajan et al., 2010). It was originally planned to discharge the MK-I core at 25 GWd/t burn-up and enlarge the core with all MK-II fuel (55%PuCþ45%UC), to enable raising the reactor power close to the design value of 40 MWt. Induction of MK-II SA was started in 1996, when one MK-I SA reached a burn-up of 25 GWd/t. However, due to the progressive extension of the burnup limit of the MK-I fuel to a level of 155 GWd/t, the option of going in for a MK-II core was given up. Thirteen MK-II SA had been loaded

till 2006. Eight high Pu MOX fuel (44%PuO2þ56%UO2) SA were also loaded in 2006, to validate the fabrication process of the MOX fuel for PFBR. The current policy is to progressively discharge the MK-I & MK-II SA after attaining burn up of 155 & 100 GWd/t respectively and replace with fresh MK-I. The core and power evolution of FBTR is indicated schematically in Fig. 4. The current core has 36 MK-I, 7 MK-II and 8 MOX SA, and the current power rating of the core is 26.1 MWt. In order to realize operating temperatures close to the design values, three out of the seven water tubes in each of the Steam Generator modules were cut and blanked in 2008. The sodium outlet temperature reached is ~490  C, as against the design intent of 515  C. The design flux of 3.1  1015 n/cm2/s has been realized, with a harder spectrum compared to the MOX design of Rapsodie, making FBTR an excellent irradiation facility for materials.

2.3. Reactor performance FBTR completed 30 years of operation in Oct 2015. During the three decades, it has provided valuable operating experience. As of Dec 2015, FBTR has generated 524 GWh of thermal energy and 35 million units of electrical energy, logging a cumulative high power operation time (i.e. with the SG in operation with feed water) of ~35,000 h. The four sodium pumps have been in trouble-free cumulative service of more than 800,000 h. Sodium chemistry has been well maintained, with the plugging temperature maintained below 105  C (corresponding to an oxygen impurity below 1 ppm). As an irradiation test facility, FBTR has been used for the testing of MOX fuel for power reactors, Zr-Nb for PHWRs, alloys of core structural materials, low dose irradiation of 304 L N & 316 L N, materials for Compact High Temperature Reactors being developed by BARC, sol-gel based vibro-compacted fuel, metallic fuel pins etc. Many of the irradiations are long term, on-going ones. The feasibility of producing Sr89, a medical isotope used in bone cancer therapy has been studied by irradiating Yttria pellets and the

Please cite this article in press as: Puthiyavinayagam, P., et al., Development of fast breeder reactor technology in India, Progress in Nuclear Energy (2017), http://dx.doi.org/10.1016/j.pnucene.2017.03.015

4

P. Puthiyavinayagam et al. / Progress in Nuclear Energy xxx (2017) 1e24

Fig. 3. Fuel cross-sections of MK-I fuel at various burn-up levels.

Fig. 4. Evolution of FBTR core and Power over the years.

results are satisfactory. Several tests have been conducted, mostly between 1995 & 2000. These include study of the plant dynamics during normal

transients, primary coast-down and battery take over tests, natural convection decay heat removal capability with the secondary pumps alone tripped and then with the primary pumps alone

Please cite this article in press as: Puthiyavinayagam, P., et al., Development of fast breeder reactor technology in India, Progress in Nuclear Energy (2017), http://dx.doi.org/10.1016/j.pnucene.2017.03.015

P. Puthiyavinayagam et al. / Progress in Nuclear Energy xxx (2017) 1e24

tripped, decay heat removal capability of the steam generators by opening the trap door of the cabin housing them, void coefficient measurements at various core locations (Srinivasan et al., 2010b), response of the DND system during any clad failure (Srinivasan et al., 2014), measurement of the replacement of worths of fresh fuel (MK-I) replacing spent fuel (MK-I & II) at various locations. Over the years, several major incidents have been faced. These include a fuel handling incident in 1987, primary sodium leak of about 75 kg inside a nitrogen-inerted cabin, water leaks from the Biological Shield Cooling coils around the reactor, the core cover plate housing the core thermocouples getting permanently stuck at a higher elevation (leading to core temperature anomalies in the MK-II SA at peripheral locations), three reactivity transients of different nature and reactivity anomalies between estimated and measured during large scale core changes. All these have been effectively addressed, except the cause of the reactivity transients. These are suspected to be due to thermal deformation of the fuel subassemblies due to the large temperature gradients inherent in a small core. With the expansion of the core, these transients have not recurred. 2.4. Residual life As a part of the periodic safety review by the Regulatory Body, residual life assessment was carried out. It is seen that the sole factor limiting the reactor life is the residual ductility of the grid plate supporting the core. Based on sample irradiation, the allowable dpa for a residual uniform ductility criterion of 10% is ~4.5 dpa. Measures have been initiated to reduce the neutron dose on the grid plate by modifying the bottom reflectors in the fuel subassemblies (Kasiviswanathan et al., 2011). It is estimated that the reactor can be safely operated upto ~5 Effective Full Power Years as of now. It is proposed to keep the reactor in operation till 2030 by operating in a mission mode with ~30% annual availability factor. Several Post-Fukushima retro-fits have been carried out, especially to avoid entry of water into the complex, based on the reassessed Design Basis Flood levels, including tsunami. Seismic re-evaluation of the plant was carried out. All the minor fixes and most of the major retro-fits have been completed. It is planned to install two additional DG sets in a seismically qualified Flood safe Building by 2017. 3. Prototype Fast Breeder Reactor A steering group was setup by DAE in December 1979 for drawing up a plan for a Prototype Fast Breeder Reactor (PFBR). The Steering Group submitted its report in July 1980 and a supplementary report in November 1980 in which it recommended a pool type of reactor of 500 MWe capacity as this plant size was found technologically viable in India and balance of plant like turbogenerator can be manufactured indigenously. Further, scale up from FBTR was considered feasible and unit energy cost lower compared to a 250 MWe plant. Subsequently, based on the recommendation, a Prototype Fast Breeder Reactor Working Group (PFBR-WG) was formed in March 1981 to prepare a feasibility report for the project. The Working Group recommended a design in which preferred major parameters have been indicated in order to start with the design process. Some of the choices were re-visited later on and based on the state of art technology, economic analysis, other aspects with respect to its adoption, those choices were revised and firmed up. The choice of reactor system concepts and major design features were discussed in-depth considering its merits, technology availability, suitability in the Indian context based on which the conceptual design was finalized (IGCAR, 1983). World wide experience

5

on MOX use and expertise in India of oxide reprocessing led to the choice of MOX fuel. Pool type reactor with two primary and two secondary pumps was found to result in cost saving and simpler layout. Operating temperatures in different parts of the plant have been established on the basis of the following (i) Structural requirements of hot leg components, in particular, control plug (ii) Limiting clad hot spot temperature to 973 K (iii) High temperature properties of steam generator material, ferritic 9Cr- 1Mo modified steel (T91) (iv) Optimization of cost of heat exchangers (IHX and SG) (v) Sodium pumping cost. Based on all these, the design temperature chosen are: at the hot pool 820 K, at the cold pool 670 K, at the IHX inlet 628 K, at the IHX outlet 798 K and steam temperature at the turbine inlet 763 K. 3.1. Industrial capability PFBR equipment can be broadly classified into two categories i.e. Nuclear Steam Supply Systems (NSSS) and Balance of Plant (BoP). NSSS components are quite large in size and weight. Based on the manufacturing requirements and methodologies, NSSS components can be broadly classified as follows: (i) In-house capability within DAE existed for the manufacture of fuel pin and assembly components such as clad tubes and hexagonal shaped wrapper tubes. (ii) Relatively thin walled, large diameter vessels like Main Vessel (MV), Safety Vessel (SV), Inner Vessel (IV) and thermal baffles. These components involve SS fabrication to stringent form tolerances. These components have to be assembled at the site. (iii) Long SS components, requiring precision machining, deep-hole boring and/or vertical turning and boring of large diameters, machining to good surface finish and assembly by welding. Stringent geometric tolerances have to be achieved on the assembly. Grid Plate (GP), Core Support Structure (CSS), Control Plug (CP), Control and Safety Rod Drive Mechanism (CSRDM), Diverse Safety Rod Drive Mechanism (DSRDM), Transfer Arm (TA) and Inclined Fuel Transfer Machine (IFTM) fall in this category. Hard-facing using colmonoy deposit is required for many parts (iv) Box like structures made of carbon steels (A48 P2) involving stringent forming tolerances, thick section welding, post weld heat treatment and precision machining. Roof slab, Large Rotatable Plug (LRP) and Small Rotatable Plug (SRP) fall under this category (v) Heat exchange equipment of the shell and tube type, which include Intermediate Heat Exchanger (IHX), Steam Generators (SG) and sodium-air and sodium-sodium in the Safety Grade Decay Heat Removal System (SGDHR) (vi) Sodium pumps, which require fabrication of the vessel and 11 m long shaft. Heavy castings of the impeller, suction casing and discharge bowl are required. The pump is to be assembled in clean conditions (vii) Heavy flasks involving design and fabrication of drives and heavy-duty fabrication with handling facility. Only limited industries in India could take up the manufacture of NSSS components. An extensive survey of the potential of the Indian industries to undertake the challenges of manufacturing the components of PFBR was carried out. The exercise confirmed the capacity and capability of the Indian industries to take up the manufacture of PFBR NSSS components. All components of the BoP, like turbine, generator, condenser, low pressure & high pressure heaters, boiler feed pumps, and transformers and switchgears are conventional in nature, being used in fossil fired thermal power plants. 3.2. Description of PFBR Based on the detailed studies, the final design of the PFBR (Chetal et al., 2006) is arrived at which is described in brief in the following section. PFBR is a pool type reactor with 2 primary and 2

Please cite this article in press as: Puthiyavinayagam, P., et al., Development of fast breeder reactor technology in India, Progress in Nuclear Energy (2017), http://dx.doi.org/10.1016/j.pnucene.2017.03.015

6

P. Puthiyavinayagam et al. / Progress in Nuclear Energy xxx (2017) 1e24

secondary loops with 4 S G per loop. The overall flow diagram comprising primary circuit housed in reactor assembly, secondary sodium circuit and BoP is shown in Fig. 5. The nuclear heat generated in the core is removed by circulating sodium from cold pool at 670 K to the hot pool at 820 K. The sodium from hot pool after transporting its heat to four IHX mixes with the cold pool. The circulation of sodium from cold pool to hot pool is maintained by two primary sodium pumps and the flow of sodium through IHX is driven by a level difference (1.5 m of sodium) between the hot and cold pools. The heat from IHX is in turn transported to eight Steam Generators (SG) by sodium flowing in the secondary circuit. Steam produced in SG is supplied to turbo-generator. In the reactor assembly (Fig. 6), the main vessel is the important component which houses the entire primary sodium circuit including core. The sodium is filled in the main vessel with free surfaces, blanketed by argon. The inner vessel separates the hot and cold sodium pools. The reactor consists a total of 1757 subassemblies comprising fuel, blanket, absorber, reflector, different types of shielding assemblies and internal storage locations meant for spent fuel assemblies etc. Of these, 181 constitute the driver fuel assemblies. The CP, positioned just above the core, houses mainly 12 absorber rod drive mechanisms (ARDM). The top shield covers the MV and supports the Primary Sodium Pumps (PSP), IHX, CP and fuel handling systems. For the core components, 20% cold worked D9 material (15% Cr- 15% Ni with Ti and Mo) is used to have better irradiation resistance. Austenitic stainless steel type 316 L N is the main structural material for the out-of-core components and

modified 9Cr-1Mo (grade 91) is chosen for SG (Chetal et al., 2006). PFBR is designed for a plant life of 40 y with a load factor of 75%. 3.3. Extensive R&D program PFBR is designed wholly indigenously and it requires large R & D efforts to ensure safe and reliable performance of the reactor. The concepts for various systems of the reactor have been chosen after studying the concepts adopted in various operating reactors and considering the performance of these systems from operational and safety considerations. The concepts chosen for PFBR have also been reviewed by the Novatome, France and OKBM, Russia. However, to ensure reliability of operation, it is required to undertake a detailed R&D program. Due consideration has been given to design and safety experiments conducted in other research laboratories. The R & D program has been broadly classified under different heads: physics and shielding, chemistry of sodium, materials development, thermal hydraulics, structural mechanics, component development, fuel development, inspections techniques, instrumentation and control, civil engineering and safety. For example, for fuel subassembly, the R&D tasks carried out are: pressure drop measurements, flow induced vibration studies, cavitation studies, testing of pressure drop devices, lifting force measurements, flow blockage study, irradiation testing of 37 fuel pin-bundles in FBTR for performance validation, material specimen irradiation for swelling & creep data, out of pile test data on mechanical properties of fuel clad and hexcan, Out of pile test data on fuel thermal conductivity & melting point, fresh SA enrichment

Fig. 5. PFBR flow sheet.

Please cite this article in press as: Puthiyavinayagam, P., et al., Development of fast breeder reactor technology in India, Progress in Nuclear Energy (2017), http://dx.doi.org/10.1016/j.pnucene.2017.03.015

P. Puthiyavinayagam et al. / Progress in Nuclear Energy xxx (2017) 1e24

7

Fig. 6. Reactor assembly.

detection system, neutron transport through axial shield, neutron streaming through top/bottom gas plenum. The reactor design is a complex and multidisciplinary subject and requires expertise from various sources. The efforts are therefore both in-house and as well as from various national academic and research institutions. The R&D on structural mechanics domain have been well covered in the references (Chellapandi et al., 2006, 2009 and 2010a,b). In this paper, details of R&D in other areas are briefly covered. 3.4. Reactor physics modelling in FBR At the time of FBTR design and construction, reactor physics technical capability was developed initially with collaboration of CEA, France. This capability was systematically enhanced keeping in mind that in FBR, even though the core is small, average neutron energies reach thermal in the shield in contrast to about 100 keV in the core and blanket. A number of computer codes for nuclear data processing, core, shielding and neutron monitoring design as well as safety analyses were developed indigenously. Through International Atomic Energy Agency (IAEA) based collaborations, nuclear data related to neutron interactions for all isotopes relevant in reactor operation, fuel burnup, fuel storage, reprocessing and recycling were obtained. These data bases were processed in our laboratory (Cullen, 2001). Collaboration was established with Institute of Physics and Power Engineering (IPPE), Russia for validation of core and shielding computations by benchmark analyses. Recently collaborations with CEA was revived in reactor safety (Devan et al., 2012). Indian physicists also participated in many IAEA coordinated research projects related to FBR (IAEA TECDOC-731, 1994, IAEA TECDOC-1623, 2010, IAEA TECDOC1626, 2010). Starting from 1-D and 2-D models, more exact 3-D hex-Z computer code was developed for core simulation with fuel burnup and

refuelling (Mohanakrishnan, 2008). Compared to thermal reactors, it is essential to use larger number of neutron energy groups in FBR core simulation. Monte Carlo modelling of core was chosen for FBTR due its small size and short term core changes for experimental irradiations (Raghukumar et al., 2013). The in-vessel shield design is perhaps the most complex problem faced by reactor physicists in FBR as the shield requirements conflicts with neutron monitoring requirements. High temperatures and compact core makes in-core monitoring difficult while neutron and gamma flux attenuations by about 10 decades in shield make exact computations difficult for present day computers. A midway coupling method (as shown in Fig. 7) of using result from discrete ordinate transport method as input source for Monte Carlo model was developed for IHX shield design. A new 175 neutron group 42 gamma-group coupled crosssection library was also developed for this purpose. Technical capability for safety analysis and Probabilistic Safety Analysis (PSA) is now essential for any reactor design. Certain features of FBR like tightly coupled core, prompt negative reactivity feed backs make reactor operation as well as transient modelling manageable despite low neutron life time and low delayed neutron fractions. Considering radial and axial zones in core and blanket, a code was developed for transient modelling and validated up to initiation of coolant voiding in accident cases (Harish et al., 1999; IAEA TECDOC-1139, 2000). For ensuring adequate safety, system reliability studies have been performed for PFBR during the design of important safety critical systems such as shutdown and safety grade decay heat removal systems and based on the study, improved features have been incorporated to meet the respective reliability targets of 1E-6/ry and 1E-7/ry for these two systems. Subsequently, for events at full power, Level-1 PSA has been performed. Sixteen initiating event groups were considered along with twelve safety and safety support systems. These safety system models were integrated with accident sequence models to quantify

Please cite this article in press as: Puthiyavinayagam, P., et al., Development of fast breeder reactor technology in India, Progress in Nuclear Energy (2017), http://dx.doi.org/10.1016/j.pnucene.2017.03.015

8

P. Puthiyavinayagam et al. / Progress in Nuclear Energy xxx (2017) 1e24

Fig. 7. Mid-way coupling of deterministic and stochastic methods for IHX shield design.

the core damage frequency. The study establishes that the likelihood of severe core damage is < 1E-6/ry and there are no identifiable unbalanced risk contributors (Ramakrishna et al., 2012). 3.5. Thermal hydraulics modelling in FBR Design of a fast breeder reactor demands detailed knowledge of thermal hydraulic parameters prevailing in the reactor at various conditions. Coolant sodium has a high heat transfer coefficient and can remove very high heat fluxes (~2 MW/m2). The high boiling point of sodium (880  C) leads to low pressure systems. Thus, the main loads on FBR structures are of thermal origin, viz., high operating temperature (creep), large temperature gradient (thermal stress) and large number of cyclic variations in temperature (thermal fatigue) due to various transients taking place in the plant. In FBR system, the Biot number is of the order of unity, demanding multi-physics heat transfer studies for accurate prediction of structural temperatures. Wide operating temperature range coupled with a high volumetric expansion coefficient of sodium is favourable for natural convection heat removal. However, a large buoyancy force in sodium pools leads to high Richardson number ~1, posing the risk of thermal stratification and associated temperature fluctuations and thermal fatigue in the structures. Sodium jets from fuel bundles at differing temperatures entering into a common reactor pool can lead to high-cycle thermal fatigue as a consequence of jet instabilities and large heat transfer coefficient of

sodium. Nuclear heat is generated in fuel pin bundles with helical wire spacers housed inside a hexagonal wrapper. All the sub-channels of the bundle are not of identical hydraulic resistance, which leads to flow mal-distribution and circumferential temperature variation in fuel clad. Prediction of inter-wrapper heat transfer during onset of natural convection in emergency decay heat removal, calls for multi-scale modelling of the entire primary system. The main drawback of sodium is its violent chemical reaction with air and water. Hence, inert argon gas is maintained above sodium free surfaces to avoid sodium-air contact. Large free surface velocity in compact reactor pool leads to argon gas entrainment in sodium and the associated risk of reactivity fluctuations in the core, which is a serious safety issue. The above-discussed thermal hydraulic problems can be analyzed either by experimental tests or by numerical simulations. Experiments with liquid sodium is costly and time consuming due to high temperature, possibility of reaction with air and water and opaqueness. Water can simulate hydraulic but not heat transfer behavior of sodium. Even to simulate thermal hydraulic processes taking place in cover gases, such as argon and nitrogen, numerical simulation is cheaper and faster especially when the number of parameters involved is large. For a successful design, a judicious combination of experimental and numerical approaches is a must. The method adopted in design of PFBR is based on performing many large-scale water experiments and limited sodium

Please cite this article in press as: Puthiyavinayagam, P., et al., Development of fast breeder reactor technology in India, Progress in Nuclear Energy (2017), http://dx.doi.org/10.1016/j.pnucene.2017.03.015

P. Puthiyavinayagam et al. / Progress in Nuclear Energy xxx (2017) 1e24

9

experiments (described elsewhere in the paper) to validate computer codes and then using these validated computer codes for numerical prediction of sodium flow and temperature distributions in the reactor. 3.5.1. Thermal hydraulics codes Technical capability in modelling phenomena in Computational Fluid Dynamic (CFD) has been achieved by a combination of commercial CFD codes and in-house CFD codes. For multidimensional fluid flow and heat transfer analyses, two computer codes THYC-2D and THYC-3D have been developed in house. They respectively solve 2 and 3-dimensional Navier-Stokes and energy equations in Cartesian and cylindrical coordinate systems (Hughes and Gaylord, 1964) using a control volume based discretisation method (Patankar, 1980). Boussinesq approximation is used to take care of buoyancy effects while turbulence is modelled using the standard k eε model (Launder and Spalding, 1974). Porous body formulation is used to model submerged small-scale structures. Commercial CFD codes, viz., PHOENICS and STAR-CD have also been extensively used for component level thermal hydraulic analyses. For modelling the whole plant behavior under various events/transients, system level plant dynamics codes, viz., DYANA-P and DHDYN have been developed, respectively for short term and long term transient studies. These codes have been validated against numerical/ experimental results published in literature and experiments carried out in house/tests carried out in FBTR. 3.5.2. Pool thermal hydraulic studies and plant dynamic studies In pool type FBR, coexistence of hot and cold pools within the main vessel, in vessel storage of spent fuel subassemblies and large size of pool coupled with large volumetric expansion coefficient of sodium lead to thermal stratification. In PFBR, to break the thermal stratification, an anti-stratification porous shell has been designed and provided below the core cover plate in the control plug (Fig. 8) for which pool thermal hydraulics were investigated in detail by 2D and 3-D CFD analyses (K. Velusamy et al., 2010). Its porosity has been optimized such that there is no significant dilution in measuring the sodium outlet temperatures from various subassemblies. The phenomenon of thermal striping which leads to random temperature fluctuations in the structures such as core cover plate, facing non-isothermal sodium streams has been investigated by multi-level CFD modelling (Velusamy et al., 2005). In the first level, 3-D CFD simulations are carried out to identify zones of hot pool prone to temperature fluctuations using standard turbulence models. In the second level Direct Numerical Simulation is carried out in the localised zones to quantify the frequency and amplitude of temperature fluctuations in sodium as well as in structures (Fig. 9). This novel coupling of modelling levels led to the optimization of distance between the thermocouple tip and SA top (Velusamy et al., 1990; Maity et al., 2011). By 3-D CFD simulations (Satpathy et al., 2013) an optimum passive anti-gas entrainment device was developed by identifying critical free surface velocity of sodium to be 0.5 m/s. To avoid gas entrainment during free fall of sodium in main vessel cooling system, the profile of overflow weir shell has been optimized by experimental simulations (Fig. 10). Flow distribution inside grid plate was predicted to be circumferentially uniform and the maximum cross flow velocity was seen to be 9 m/s while the pressure difference in grid plate is 4.6 mlc from 2 and 3eD CFD studies employing a porous body model (Natesan et al., 2006). For an unlikely accidental condition of rupture in one of the feeding pipes, sodium bypasses the core. Adequacy of core cooling and safe shutdown of the reactor, were ensured under this condition by a coupled plant dynamic and CFD study. Simulation of coolant flow with the fuel pin bundles calls for complete 3-D CFD based

Fig. 8. Flow field in hot pool at full power (Design with skirt).

calculations which demands very high computing power with validated software for achieving prediction capability of reliable hot channel and hot spot factors. For PFBR, such simulations were performed for the fuel subassembly with 217 pins using 84 node parallel computers (Naveen Raj and Velusamy, 2016) (Fig. 11). The heat removal capability of inter-wrapper flow has been investigated by multi-dimensional CFD analyses and it was found that the temperature limits of core and pool are satisfactorily met under natural convection decay heat removal condition (Partha Sarathy et al., 2012). Technical capability to design Delayed Neutron Detection (DND) system required 3-D transient hydraulic analysis of hot pool for predicting the evolution of fission product concentration in the pool after a fuel SA failure (Natesan et al., 2007). It was found that clad failure in any subassembly can be detected by DND within 69 s, considering dilution in the concentration of fission products in hot pool both by mixing as well due to their radioactive decay (Fig. 12). By studying the neutronic and thermal hydraulic responses of the plant for various possible events, by a comprehensive plant dynamics study using DYANA-P code, an optimum list of SCRAM parameters and the maximum permissible response time for various instruments have been established (Natesan et al., 2012). As an example, the evolution of core temperatures during one primary sodium pump trip event is depicted in Fig. 13.

3.6. Technology development for the major reactor components PFBR component dimensions are large compared to FBTR. The principal material of construction is 20% cold worked D9, SS 316 LN, 9 Cr-1Mo (Grade 91) and carbon steel equivalent to A516 Grade 65. In view of large scale-up factor and different construction features as compared to FBTR, manufacturing technology development of

Please cite this article in press as: Puthiyavinayagam, P., et al., Development of fast breeder reactor technology in India, Progress in Nuclear Energy (2017), http://dx.doi.org/10.1016/j.pnucene.2017.03.015

10

P. Puthiyavinayagam et al. / Progress in Nuclear Energy xxx (2017) 1e24

Fig. 9. Global temperature (K) distribution in sodium and peak-to-peak sodium temperature fluctuation.

Fig. 10. Weir e Crest testing full scale water model (Left) & water film thickness being measured (Right) by a special conductance probe.

critical components was taken up through the Indian industries. Manufacturing challenges associated with the major components are briefly summarized below: (i) Manufacture of MV, IV and SV involves die pressing of large size dished end petals to close profile with 0.2% tolerance on radius and rolling of shells to 0.2% tolerance on radius. Solution annealing of petals is called for if the cold work exceeds 10%. (ii) GP is a box type structure with top and bottom plates of 6.8 m diameter and interconnected by sleeves of 1 m height. There are 1758 sleeves fixed between top and bottom plates. The manufacture of GP involves welding of thick plates

Fig. 11. Temperature field in fuel bundle exit.

(80 mm) of smaller widths to get a wider plate of 6.8 m diameter, its solution annealing at 1050 C and machining of 6.8 m dia plate to a flatness requirement of 0.05 mm, drilling of large number of holes on top and bottom plate with H7/g6 tolerance and within 0.1 mm location tolerance, stress relieving heat treatment of plates for dimensional stability, colmonoy deposit on the plates and on internal surfaces of sleeves at a depth of around 400 mm from opening edge and its subsequent stress relieving at 850 C.

Please cite this article in press as: Puthiyavinayagam, P., et al., Development of fast breeder reactor technology in India, Progress in Nuclear Energy (2017), http://dx.doi.org/10.1016/j.pnucene.2017.03.015

P. Puthiyavinayagam et al. / Progress in Nuclear Energy xxx (2017) 1e24

11

Fig. 12. Clad failure detection time for different FSAs estimated by CFD studies.

(vi)

(vii)

Fig. 13. Evolution of core temperatures during one PSP trip event.

(iii) Roof slab is a box type structure. It has to be made in sectors in shop floor and then integrated at site assembly shop for final assembly. The material of construction is special carbon steel to avoid lamellar tearing, as it has a number of radial stiffeners and ‘T’ welds. A 75 sector of roof slab was manufactured as part of development. (iv) The PSP are vertical type mechanical, centrifugal pumps with a free sodium level and inert cover gas and are driven by an AC variable speed drive motors. Components requiring special manufacturing techniques are primary pump shaft, impeller casting and hard faced hydraulic bearing. The pump shaft requires a highly controlled cylindrically during manufacturing, involving process of welding, heat treatment at 1050 C under inert atmosphere and final machining. The long shaft of 630 mm dia. and 11 m length is dynamically balanced to ISO grade better than G 2.5 to minimize vibrations. (v) SG is one of the critical components of FBR. SG is a shell and tube type counter current flow heat exchanger with sodium on shell side and water on the tube side. Each tube is of 23 m length and is having a bend to accommodate differential thermal expansion between the tubes and shell and among the tubes. The material of construction is ferritic 9Cr-1Mo modified steel (T91). Based upon the operating experience in other reactors, the type of weld joint chosen for the tube to

(viii)

(ix) (x)

tube sheet is a bore welded type with a raised spigot. The SG was manufactured successfully by two industries. The IHX is a shell and tube type heat exchanger. The material of construction is SS 316 LN. The tubes (0.8 mm wall thickness) are attached to the tube sheet (160 mm) by rolled and weld joint. The weld with the tube sheet is an autogenous weld. The tube to tube sheet weld joint and the sleeve valve mechanism were developed through the industries. There are two independent, fast acting, diverse drive mechanisms of absorber rods, called CSRDM and DSRDM. The mechanisms along with their respective absorber rods operating in fail safe mode ensure high reliability. Each mechanism is made up of about 300 machined and standard items that need stringent tolerances and tight quality control. Both the mechanisms have been manufactured indigenously and tested in air and sodium. There are two mechanisms for primary handling of core subassembly. These are called Transfer Arm (TA) and Inclined Fuel Transfer Machine (IFTM). The gripper mechanism of TA consists of long slender components which are required to locate the specified subassembly blindly under sodium. A typical positional accuracy of 30 mm at the head of the fuel subassembly is required for reliable fuel handling operation. IFTM is used to transfer subassembly from reactor assembly to the fuel transfer cell. TA, 23.5 m tall and weighing 23.5 t, has been manufactured and successfully tested in air at site. Manufacture of TA involved precision machining and grinding of long slender components to H7/g6 tolerance, deep hole boring of 3.4 m long hollow bars, its heat treatment at 1050 C, hard chrome plating, development of colmonoy bushes and distortion control during welding. Annular linear induction pump of 70 m3/h, 415 V, 34, 50 Hz was manufactured and successfully tested at 560 C. Large diameter pipe fittings (800 NB, 550 NB pipe bends, tees, reducer), Louvre type air dampers of 2.87 m  2.75 m with 99.9% leak tightness were also manufactured.

Fig. 14 shows a few components, which have been manufactured prior to launching PFBR. The same strategy was adopted during the course of manufacture of other challenging components, such as GP, primary sodium pipes and header, CSS, CP and sodium pump (Fig. 15).

Please cite this article in press as: Puthiyavinayagam, P., et al., Development of fast breeder reactor technology in India, Progress in Nuclear Energy (2017), http://dx.doi.org/10.1016/j.pnucene.2017.03.015

12

P. Puthiyavinayagam et al. / Progress in Nuclear Energy xxx (2017) 1e24

Fig. 14. Components manufactured under Technology development exercises.

3.7. Experimental studies for performance evaluation of fast reactor components Engineering qualification of systems and components of reactor is an essential part of development of fast reactor technology. In order to test, qualify and establish the design characteristics of reactor components, many experimental facilities in sodium and water have been constructed and operated in IGCAR. 3.7.1. Sodium facilities Sodium facilities have been constructed mainly from the point of view of meeting the testing and qualification requirements of FBTR and PFBR. One of the earliest facilities is the sodium loop with a heater capacity of 500 kW, which served as a test bed for FBTR components and for the initial developmental activities in sodium technology. Heat transfer studies in sodium to sodium heat exchangers and sodium to air heat exchanger have been carried out in this facility apart from the performance assessment of centrifugal pump and electromagnetic pumps (Prahlad et al., 1990). Engineering calibration of eddy current flow meters (Sharma et al., 2010) and permanent magnet flow meters have also been carried out in this facility. Full scale testing of large components of reactor in sodium has been carried out in large component test rig housed inside a 43 m tall building. This facility has four test vessels of different capacities with a maximum operating temperature of 600 C. Important features of this facility are: flat linear induction pump for sodium circulation, permanent magnet flow meters for sodium flow measurement, air cooled cold trap, plugging indicator, nickel tube samplers for sodium sampling, wire type leak detectors, continuous and discontinuous level probes, sodium aerosol detectors and leak collection trays. This facility has provided excellent experience on operation of sodium system for more than two

decades. Thermal hydraulics studies on roof slab and CP models have been performed in this facility. Shutdown mechanisms of PFBR have been qualified for 18 years of reactor operation through tests carried out in this facility (Rajan Babu et al., 2010). This facility has contributed significantly in the development of ultrasonic under sodium scanner (Sylvia et al., 2013) qualification of electromagnetic pumps (Nashine and Rao, 2014), qualification of permanent magnet flowmeters (Vijayakumar et al., 2011a,b) and sodium cleaning of large components (Ignatius Sundar Raj and Sreedhar, 2013). In order to study the phenomenon of selfwastage and adjacent tube wastage characteristics of sodiumwater reaction with respect to the tube material used for SG (Kishore et al., 2012; Kishore et al., 2016), a dedicated facility called Sodium Water Reaction Test Rig (SOWART) was established. This facility (Fig. 16) has been instrumental in testing and development of various types of hydrogen sensors in sodium and in argon (Sreeramamurthy et al., 2014) and qualification of cold trap design (Hemanath et al., 2010). Moreover, experimental trials for development of acoustic based sodium water reaction sensors are also being carried out. In order to validate the design of long, straight vertical, once through type SG, a dedicated Steam Generator Test Facility (SGTF) for establishment of thermal hydraulics characteristics has been set up (Suresh Kumar et al., 2012). Apart from performance assessment (Vinod et al., 2014), studies in the area of flow instability, flow induced vibration, depressurization and leak detection by acoustic method have also been carried out on the model SG (5.5 MWt with 19 tubes, 23 m long) (Ramakrishna et al., 2016). Apart from engineering development, several experimental studies have been carried out towards safety demonstration of systems. An important one in this category is the establishment of thermal hydraulic behavior of SGDHR of PFBR in SAfety Grade

Please cite this article in press as: Puthiyavinayagam, P., et al., Development of fast breeder reactor technology in India, Progress in Nuclear Energy (2017), http://dx.doi.org/10.1016/j.pnucene.2017.03.015

P. Puthiyavinayagam et al. / Progress in Nuclear Energy xxx (2017) 1e24

13

Fig. 15. PFBR components under manufacture.

Fig. 16. Flow sheet of SOWART.

Decay Heat removAl loop in Natrium (SADHANA) facility. This is a 1:22 scale down model in power with 355 kW thermal power capacity based on Richardson number similitude (Padmakumar et al.,

2013). Steady state and transient experiments have been carried out in this facility to validate the design (Vinod and Chandramouli, 2014). In order to study the behavior of structures immersed in

Please cite this article in press as: Puthiyavinayagam, P., et al., Development of fast breeder reactor technology in India, Progress in Nuclear Energy (2017), http://dx.doi.org/10.1016/j.pnucene.2017.03.015

14

P. Puthiyavinayagam et al. / Progress in Nuclear Energy xxx (2017) 1e24

sodium during transient conditions, thermal shock test facility, In sodium test facility - creep and fatigue loops (Shanmugavel et al., 2011) have been established. Thermal shock test facility was used to assess the integrity of dissimilar weld joints of electromagnet immersed in sodium for the DSRDM (Patri et al., 2014) and for development of temperature sensitive magnetic switch. Low cycle fatigue studies and creep fatigue interaction studies on structures in contact with flowing sodium have been carried out in the fatigue loop. This facility consists of tribometer section, fretting wear chamber, low cycle fatigue section and thermal striping chamber. A creep loop (Rajasundaram et al., 2013) with a maximum testing temperature of 625 C is under operation for studying the creep properties of FBR materials under the influence of flowing sodium. Another important area in which experimental studies have been carried out is the physical, chemical and metallurgical aspects of sodium and sodium service components. These studies have been carried out in facilities, viz., bi-metallic loop SILVERI NAtrium (SILVERINA) and LEak Experiments on NAtrium (LEENA). Important studies carried out in these facilities include (i) carbon transfer behavior and its influence on the mechanical properties of materials used in the secondary sodium circuit (Sivaibharasi et al., 2012), (ii) self welding susceptibility experiments on different combinations of material specimens (Meikandamurthy et al., 2012), (iii) sodium aerosol compatibility on inflatable seal rubber material, (iv) performance testing of TA bearing in sodium, (v) sodium aerosol characterization in cover gas region (Kumar et al., 2015) (vi) tribological behavior studies on reactor component material specimens (Kumar et al., 2013) (vii) qualification trials on layout of sodium leak detectors (Vijayakumar et al., 2011a,b). Moreover, a test facility was constructed to study the atmospheric dispersion characteristics of sodium aerosols in open atmosphere. In this facility, aerosols resulting from sodium burning are released at 10 m height using a chimney. Various sampling techniques are used for characterization of the dispersed sodium aerosols. 3.7.2. Water facilities In order to establish the pool hydraulic behavior of sodium in primary circuit from the point of view of gas entrainment, free level fluctuation and flow induced vibration of immersed structures, 1/ 4th scale model of primary circuit of PFBR based on Froude number similitude has been established. This facility known as SAMRAT (Fig. 17) has two pumps, each with 1200 m3/h capacity and simulates both hot and cold pools of reactor. Rod type heaters are provided in the core to simulate core heat generation for carrying out studies on dilution in measurement of core outlet temperature. Studies in the area of gas entrainment, thermal stratification and thermal striping were carried out in this facility (Banerjee et al., 2013). A new facility with 5/8th of reactor size is also made to study the gas entrainment phenomena in reactor pool with better simulation of governing forces. For studying the gas entrainment phenomena in the surge tank of secondary sodium circuit and to arrive at suitable devices for jet breaking, experimental studies have been carried out on 1/12 scale down model (Ramdasu et al., 2006). A special test facility for subassembly was established for catering to the requirements of testing subassembly, orifices and labyrinth separately. Geometrical configuration of labyrinth and orifices for PFBR subassemblies were arrived through extensive experimental program conducted in this facility (Pandey et al., 2011&, 2013). This loop has a flow induced vibration test setup specifically meant for testing subassembly component vibration and lift force studies. Separate test section is also provided to carry out studies related to core flow monitoring mechanism. This facility has also contributed significantly in the pressure drop and cavitation studies of PFBR and FBTR subassemblies (Ramdasu et al.,

2005). 3.8. Safety Research 3.8.1. Severe accident studies The event of a severe core melt down accident resulting in significant relocation of the active core, is categorised as a beyond design basis event. As an accident mitigation measure, a passive safety device called Core Catcher (CC) is positioned in the lower plenum of the reactor. The CC provides a sturdy platform to hold the heat generating corium debris and ensure passive removal of decay heat. Long term decay heat removal is ensured by natural circulation through SGDHR. In order to assess the effectiveness of the Post Accident Heat Removal (PAHR), it is important to understand signature of the heat generating debris bed. The key factors affecting coolability of debris bed are bed porosity, morphology of fragmented particles and degree of spreading/heaping of the debris on the CC. Various studies have been initiated to have insight of the involved processes in PAHR and to develop validated models for application to reactor conditions. Molten Fuel Coolant Interaction (MFCI) program is undertaken in stages from lower mass to larger mass of molten fuel using various experimental facilities (Das et al., 2013). Realistic debris characterization has been taken up using highly sophisticated facility called Sodium Fuel Interaction Facility (SOFI) using uranium e sodium system, representing the actual scenario as shown in Fig. 18(a). Series of MFCI experiments with high temperature ceramic simulants in water at different parametric conditions to give insight of the complicated phenomena involved are being performed. Basic melt fragmentations, bed forming characteristics, size spectrum experiments are carried out in the initial stage of investigation, using setups in various geometrical scale upto 1:10, using simulant Woods metal e water system as shown in Fig. 18(b). Effective heat removal capability of CC assembly with internals and heat generating debris bed configuration is being performed in 1:4 geometrically scaled water model setup called the post accident thermal hydraulics facility (Gnanadhas et al., 2011). 3.8.2. Studies on sodium fires Experimental facilities have been developed to investigate sodium droplet burning, particle size distribution during sodium spray, effect of oxygen concentration in sodium burning and passive extinguishment of sodium fire. The ignition behavior of sodium droplets in the atmospheric air has been studied numerically with two available models of pre-ignition stage combustion. Surface reaction is very important in the pre-ignition stage, and the different reaction rate-controlling processes involved in this stage are explained using the shrinking core model (Muthu Saravanan et al., 2011). 3.8.2.1. Studies on small sodium spray fires. A series of experiments were conducted with small sodium inventories (2-5 g) for understanding of spray fire scenario and validation of Reactor Containment Building (RCB) pressure rise under Core Disruptive Accident (CDA) resultant fire. The facility consists of a highly instrumented quartz cylindrical chamber and associated systems as shown in Fig. 19. The particle size distribution of ejected sodium droplet from a nozzle of 1.6 mm was measured by processing the recorded images of high speed camera (Ponraju et al., 2013). A typical particle size distribution is shown in Fig. 20. The life time of single burning droplet was measured by processing the image captured during combustion. The variation in flame diameter of burning droplets with time was used as input for modelling spray fire. The percentage of sodium combustion is measured to be about

Please cite this article in press as: Puthiyavinayagam, P., et al., Development of fast breeder reactor technology in India, Progress in Nuclear Energy (2017), http://dx.doi.org/10.1016/j.pnucene.2017.03.015

P. Puthiyavinayagam et al. / Progress in Nuclear Energy xxx (2017) 1e24

15

Fig. 17. Schematic of SAMRAT facility.

Fig. 18. (a). SOFI Experimental Setup. (b). Woods metal-water experimental set up.

Please cite this article in press as: Puthiyavinayagam, P., et al., Development of fast breeder reactor technology in India, Progress in Nuclear Energy (2017), http://dx.doi.org/10.1016/j.pnucene.2017.03.015

16

P. Puthiyavinayagam et al. / Progress in Nuclear Energy xxx (2017) 1e24

Fig. 21. Schematic of MINA facility.

Fig. 19. Experimental setup.

Fig. 20. Particle size distribution for a typical sodium spray experiment.

80%. Sodium oxide aerosol produced during sodium combustion was allowed to settle down and analyzed for its size distribution using laser scattering technique.

3.8.2.2. Studies on medium scale sodium spray fires. Sodium spray fire experiments have been carried out with 2 kg of sodium in Mini sodium experimental facility (MINA), by ejecting it through a rectangular slit simulating a fine crack into test chamber of 140 m3 volume schematic diagram of MINA facility is given in Fig. 21 and the fire scenarios are given in Fig. 22. Results indicated considerable reduction of air temperature from about 900 C at the vicinity of spray to 120 C at 1 m apart from the spray location. However, no significant rise in ambient temperature and pressure of the test chamber are observed. Experiment on vertical sodium spray fire simulating sodium ejection during CDA has been conducted in SOCA test facility. About

2 kg of sodium at 500 C was ejected though multiple nozzles in to a steel test chamber of 20 m3 volume. The experimental facility eSOCA and spray fire scenario recorded during the sodium fire experiment are shown in Fig. 23 and Fig. 24 respectively. The fire scenario was observed to be random and fire penetrated aluminum cladding wrapped over the heaters of ring header and interacted with the insulating material. These experimental data are being used for validation of containment code being developed for calculating the impact of sodium fire on RCB (Muthu Saravanan et al., 2016). 3.8.3. Studies on passive extinguishment of sodium fires Studies were conducted to assess the fire suppression effectiveness of sodium Leak Collection Trays (LCT), which are provided below secondary sodium equipment and piping for collecting leaking sodium during accidental conditions (Diwakar et al., 2008; 2011). The LCT works by draining sodium into a collection vessel, where oxygen supply is limited and fire gets suppressed by oxygen starvation. Full scale leak collection system with segmented drainage circuit having dump tank was tested by simulating a leak of 100 kg sodium at 400 C through 1 cm2 opening area. Schematic of leak collection system is shown in Fig. 25 and the sodium leak collection scenario during experiment is shown in Fig. 26. Experimental results indicated that LCT is effective in sodium collection. 3.8.3.1. Development of carbon microspheres - A novel sodium fire extinguisher. Carbon in the form of Carbon Micro-Sphere (CMS) possesses potential application in extinguishing sodium fire and is being developed at IGCAR (Snehalatha et al., 2013). Scanning Electron Microscope (SEM) image of CMS is shown in Fig. 27. Synthesis of CMS involves step wise carbonization of styrene divinyl benzene resin in inert atmosphere. It possesses high thermal conductivity, low density (0.7 g/cc), chemical inertness and excellent flow characteristics and could be directed on to fire using extinguishant with conventional nozzle. It extinguishes the fire by formation of blanket layer over burning metal. Once the fire is extinguished, the metal can be easily recovered since no actual reaction occurs between CMS and the metal (Fig. 28). 3.9. Materials development program Three classes of steels, namely, austenitic stainless steel, ferritic

Please cite this article in press as: Puthiyavinayagam, P., et al., Development of fast breeder reactor technology in India, Progress in Nuclear Energy (2017), http://dx.doi.org/10.1016/j.pnucene.2017.03.015

P. Puthiyavinayagam et al. / Progress in Nuclear Energy xxx (2017) 1e24

17

Fig. 22. Spray fire scenario in MINA facility.

Fig. 23. Schematic of SOCA facility.

steel and carbon steel are used for the manufacture of different NSSS Components. The selection of materials is based on the information available in the literature and the design codes, as well as the knowledge and understanding gained from R&D activities at IGCAR. In PFBR, the fuel clad tubes experience temperatures in the range of 673e973 K under steady state operating conditions. Under transient conditions, the temperatures can rise up to 1273 K. For target burnup of 100,000 MWd/t, the maximum neutron dose for is 85 dpa. Major loads experienced by the fuel clad are the internal pressure due to accumulated fission gases released from fuel matrix (~5 MPa) and moderate fuel-clad interaction. Other loads may be due to temperature gradients and irradiation induced swelling gradients. The hexagonal sheath of the core subassembly operates at relatively lower temperatures than the fuel clad. The typical operating temperature range is 673e873 K which incidentally falls within the peak swelling temperature range. The peak neutron dose is about 85 dpa similar to that for fuel clad. Major loads on the hexcan are the internal pressure due to sodium coolant (~0.6 MPa) and the interaction loads at the contact pads due to bowing of the subassemblies under temperature and swelling gradients. Slight thermal stresses due to transients are also present because of the higher thickness of the wrapper. Alloy D9 in 20% cold worked condition has been chosen for clad and wrapper tubes of PFBR for the initial core in view of its higher swelling resistance as compared to SS 316. D9 is compatible with fuel and sodium coolant, and the international experience on its in-reactor behavior is satisfactory. Chemical composition of alloy D9 is based on modifying that of SS 316 from considerations of better swelling and irradiation creep behavior (Raj, Dec. 2008). The desired composition is achieved by controlled additions of silicon and titanium, increasing nickel content and lowering the chromium. Minor elements having strong

Please cite this article in press as: Puthiyavinayagam, P., et al., Development of fast breeder reactor technology in India, Progress in Nuclear Energy (2017), http://dx.doi.org/10.1016/j.pnucene.2017.03.015

18

P. Puthiyavinayagam et al. / Progress in Nuclear Energy xxx (2017) 1e24

Fig. 24. Sodium spray fire scenario.

Fig. 25. Leak collection system.

neutron absorption cross-section and impurities affecting weldability have been kept to a minimum. Small addition of boron is made to improve creep ductility. Permissible inclusion contents are stringent so as to minimize radiation embrittlement and sodium attack (as the cladding tube wall is extremely thin). Advanced austenitic stainless steel and oxide dispersion strengthened steels are being developed for achieving fuel burnup higher than 100 GWd/t in future core reloads (Mathew et al., 2013). The reactor assembly consists of core, GP, CSS, MV, SV, top shield and ARDM. The minimum sodium temperature in the primary pool of PFBR during normal operation (core inlet temperature) is expected to be about 670 K. The mean above-core temperature (core outlet temperature) will be about 820 K during operation, and 923 K under plant transient conditions (arising due to failure of pumps, rupture of pump to GP pipe, uncontrolled withdrawal of control rod etc. and credible combinations of incidents are also considered {for example: combination of seismic event and power failure event}). The environment of operation is liquid sodium or argon with sodium vapor or nitrogen gas depending upon the location of the component. Except for the components near the core, like GP which may see a fast fluence of 1021 to 1022 n/cm2, for all other components such as IV, MV, SV, IHX and PSP, radiation is not a consideration. Liquid sodium coolant with strict control of chemistry, particularly of elements responsible for liquid metal embrittlement of stainless steel (As, Sb, Bi) and also of carbon and oxygen, does not pose any embrittlement problem. In the primary sodium circuit, the primary stresses are low. However, secondary stresses of thermal origin are quite significant. These stresses are both steady and transient in nature. The secondary circuit will see large primary stresses of transient nature during any accidental big leak in the SG.

The design life of the structural components is 40 years. Creep, low cycle fatigue and creep-fatigue interactions are important considerations in the choice of materials (Mannan et al., 2003). Sodium streams coming from the core channels are at different temperatures and these, on impinging upon the component surface, produce local variations in the surface temperature with associated stress fluctuations. This phenomenon called thermal striping would lead to high cycle fatigue damage and this is another important consideration in choosing materials for structural components above the core. Based on the above considerations, austenitic stainless steel (Grade 304 LN and 316 LN)) has been chosen for reactor assembly components which are in contact with sodium. Other major advantages of austenitic stainless steels include availability of vast data base on mechanical properties including very long term creep data, ease of availability and fabrication, and above all, the availability of design data in the nuclear codes (RCCMR and ASME). For SG, a range of materials starting from ferritic steels (2.25Cr1Mo, Nb stabilised 2.25Cr-1Mo, 9Cr-1Mo (grade 9), Mod. 9Cr-1Mo (grade 91)), austenitic stainless steels (AISI 304/316/321) and alloy 800 were examined. In view of poor resistance to aqueous stress corrosion cracking (SCC), austenitic stainless steels of 300 series were not considered. Alloy 800 shows better resistance than austenitic steels, but it is not immune to SCC in chloride and caustic environments. Therefore, ferritic steels are most preferred for SG applications (Mannan et al., 2003). For top shield components, carbon steels meant for pressure vessel applications at moderate and low temperatures with very good notch toughness meets all the operating requirements including those of cost and availability. Austenitic stainless steels

Please cite this article in press as: Puthiyavinayagam, P., et al., Development of fast breeder reactor technology in India, Progress in Nuclear Energy (2017), http://dx.doi.org/10.1016/j.pnucene.2017.03.015

P. Puthiyavinayagam et al. / Progress in Nuclear Energy xxx (2017) 1e24

19

Fig. 26. Sodium collection Scenario on LCT.

Fig. 27. SEM image of CMS.

would also meet all these requirements, though cost is much higher than that of the carbon steel. Therefore, special grade (low carbon) A48P2 steel specified in French code RCC-MR is chosen as the material of construction for the top shield of PFBR. In summary, the materials selected are: (i) Alloy D9 stainless steel has been chosen for the clad and wrapper tubes for the initial cores of PFBR (ii) For structural components forming part of the reactor assembly, 304LN and 316LN stainless steel is recommended; 304LN is to be used for components which will see lower temperatures where creep is not important (iii) Mod. 9Cr-1Mo steel has been selected for the SG components (iv) For top shield components including the roof slab, carbon steel grade A48P2 is recommended. Apart from component manufacturing development, the suppliers of critical raw materials in the form of plates, forgings, tubes and welding consumables have been developed indigenously by manufacturing the critical materials such as D9 for fuel cladding tubes and hexcans; 23 m long seamless tubes, tube sheet and dished end forgings for SG; 8 m long SS 316 LN tubes of 19 mm OD x 0.8 mm wall thickness for IHX; plates of 8 m length x 30 mm thickness in SS 316 LN and welding consumables (stainless steel electrodes, 16-8-2 and Gr.91 filler wires). More details on the development and challenges in the materials domain can be had from the references (Raj, 2007; Raj, Aug. 2008 and Raj, Dec. 2008). 4. PFBR project

Fig. 28. Sodium fire and its mitigation by CMS.

The PFBR project activities were started in 2003 and a public sector undertaking called “Bharatiya Nabhikiya Vidyut Nigam Limited (BHAVINI)” was incorporated in 2003. While IGCAR is responsible for the design, R&D, development and safety clearance for design, BHAVINI is the responsible organisation for the

Please cite this article in press as: Puthiyavinayagam, P., et al., Development of fast breeder reactor technology in India, Progress in Nuclear Energy (2017), http://dx.doi.org/10.1016/j.pnucene.2017.03.015

20

P. Puthiyavinayagam et al. / Progress in Nuclear Energy xxx (2017) 1e24

construction, commissioning and operation. The consent for siting and construction of the reactor involves clearances from various regulatory bodies. The consent for siting involves clearance from Atomic Energy Regulatory Board (AERB). The consent for construction involves environmental clearance from Tamil Nadu state government, Ministry of Environment and Forests (MoEF) and clearance from AERB. A site selection committee was constituted by the DAE during 1983 to select suitable sites for nuclear power plant in the southern electricity region considering criteria such as site geology, general hydrology, water availability, environmental factors like population density, meteorological factors, seismic conditions, transportation problems etc. One of the sites considered for appraisal and found suitable was Kalpakkam. The Govt. Of India approved the selected site in May 1987. It meets all the site selection criteria stipulated for location of a nuclear power plant. The AERB also duly approved the site in October 2000. 4.1. Construction and erection challenges Components like MV, SV, IV, roof slab etc., cannot be transported in a single piece from manufacturer's shop to the project site due to their large dimensions. The over-dimensional components have to be manufactured in parts in shops and later on integrated at site in Site Assembly Shop (SAS). Accordingly, the individual petals/transportable subassemblies were manufactured in shop floor and transported to SAS for integration. The SAS consists of high bay and low bay. High bay dimensions are 80 m  26 m x 34 m and low bay dimensions are 72 m  24 m x 24 m. Two cranes (45 t and 20 t) are provided in the high bay where simultaneous manufacture of MV, SV, IV took place. A 60 t crane is provided in low bay for manufacture of roof slab and thermal baffles. The concept of inter-connected building has been adopted for nuclear island of PFBR. The nuclear island extends for 100 m  92 m area with tall buildings. The highest among these is the RCB which is about 72 m tall. With the base raft thickness of 3.5 m, the civil construction of Nuclear Island inter-Connected Buildings (NICB) involves pouring of 35,000 m3 of concrete (Fig 29). It is required to carry out civil construction and equipment erection in parallel which involves state-of-art erection equipment and construction methodologies and highly optimized construction sequences. Thanks to the experiences gained through technology development exercises, the large diameter shell structures viz. SV (Fig. 30), MV and IV have been fabricated with the tolerances on radius lower

Fig. 29. Reactor Vault and NICB construction.

than the values specified (<½ of thickness) except at few locations (Kumar, 2014). Erection of large dimensioned and slender FBR components with stringent dimensional accuracies (for eg. horizontality of ±1.4 mm over 13.5 m dia. support flange) is a typical challenging task, carried out for the first time in the country. Transportation of thin shell structures from the site assembly to the final location is another challenging activity in the construction. With systematically planned mockup trials, confidence to erect components with the specified tolerances has been achieved. The confidence has been demonstrated through successful erection of SV, MV, IV, baffles, roof slab etc., sequentially (Fig. 31). Details of erection and sequence were finalized after a comprehensive analytical and virtual simulation studies. 5. Role of other organizations The MOX fuel manufacture has been the responsibility of the Advanced Fuel Fabrication Facility (AFFF) of the Bhabha Atomic Research Centre (BARC). In addition, special materials such as Beryllium for the neutron source was supplied and enriched B4C for the absorber were supplied by BARC. Further, manufacture and supply of neutron detectors of various sensitivities and the design of Inclined Fuel Transfer Machine were the responsibility of BARC. Nuclear Fuel Complex (NFC), a unit of Department of Atomic Energy, manufacture the core subassemblies for FBTR as well as for PFBR, except fuel pins. They have a stainless steel seamless tube plant and special materials plant. For PFBR, NFC has successfully developed the technology for manufacture of fuel clad tubes in D9 alloy and 23 m long SG tubes in modified 9Cre1Mo alloy. They have excellent facilities for tube drawing, vacuum annealing, ceramic fuel processing and pelletisation, welding and associated inspection. For PFBR, all the subassemblies hardware and the final assembly, including the manufacture of radial blanket pellets were carried out by NFC (Reddy et al., 2013). 6. Research and development for FBR600 India plans to construct six more FBRs (FBR 600 MWe) with improved economy and enhanced safety beyond PFBR as a part of future FBR programme (Puthiyavinayagam, 2015). Improved economy is brought by the way of design optimization, simplified design, reduction of specific material quantities and using improved/alternate materials. Twin unit design with sharing of non-safety related facilities, enabling integrated manufacture and erection leading to reduced construction time. Several safety features are being developed to meet the evolving safety criteria after Fukushima accident and Gen-IV reactor design criteria, particularly to address all beyond design basis events including prolonged station blackout conditions, events resulting in severe core damage and large radioactivity release to the public. The design is focussed towards practically eliminating severe accident scenarios. Among several measures taken to meet these requirements, an important one is that sodium void reactivity value should be kept lower than 1 $ or near zero, which is in-line with international development in FBR technology. The proposed homogeneous FBR600 core configuration is shown in Fig. 32. It has two enrichment zones of mixed oxide (PuO2-UO2) fuel, followed by blanket and reflector. After in-vessel storage, Ferro-Boron (Fe-B) sub-assemblies are provided as the outer shield. Based on R & D conducted at IGCAR (Sunil Kumar et al., 2010), Fe-B is used as an improved shield material. Fe-B has good compatibility with sodium, good neutron shielding properties and lower density compared to steel, resulting in smaller core size and reduced load on core support structure. Table 1 summarizes the basic parameters of the core. Reduction of sodium void coefficient

Please cite this article in press as: Puthiyavinayagam, P., et al., Development of fast breeder reactor technology in India, Progress in Nuclear Energy (2017), http://dx.doi.org/10.1016/j.pnucene.2017.03.015

P. Puthiyavinayagam et al. / Progress in Nuclear Energy xxx (2017) 1e24

21

Fig. 30. Critical components manufactured with excellent tolerances.

Fig. 31. Installation of major reactor assembly components of PFBR.

compared to PFBR is made possible by increasing the neutron leakage and by reducing the power density in the inner core. To enhance the neutron leakage from the core in axial direction, the upper axial blanket is replaced by a fission gas plenum. Breeding ratio is maintained by increased radial blanket thickness and increased axial lower blanket length (Bachchan et al., 2015). The present design gives a total breeding ratio of 1.11. For higher reliability of Shut Down Systems (SDS) compared to PFBR, following technological developments are underway. Two types of SDS are provided in the design. The first SDS comprises 9 Control and Safety Rods (CSRs) and 3 Hydraulically Suspended Absorber Rods (HSARs). CSRs are used for both power control as

well as shutdown whereas HSARs are used exclusively for shutdown. They get triggered by active means due to a SCRAM signal or by passive means when there is a flow reduction through the core. The second SDS comprises a bank of 3 Diverse Safety Rods (DSRs). DSRs have the function of reactor shutdown only and are kept above the core during normal operation. Other than enhancing reliability, the technology development aims to mitigate the consequences of the Anticipated Transients Without Scram (ATWS) in the design extension conditions. To improve the reliability of SDS (at least by one order with respect to PFBR) further, it is planned to use (a) stroke limiting device in CSRDM to limit the consequences of uncontrolled withdrawal of CSRs and (b) the temperature

Please cite this article in press as: Puthiyavinayagam, P., et al., Development of fast breeder reactor technology in India, Progress in Nuclear Energy (2017), http://dx.doi.org/10.1016/j.pnucene.2017.03.015

22

P. Puthiyavinayagam et al. / Progress in Nuclear Energy xxx (2017) 1e24

economy, the SG length has been increased to 30 m. The availability and reliability of Decay Heat Removal (DHR) systems will be enhanced during fuel handling, in-service inspection, design basis events, design extension and post accident conditions. The SGDHR is envisaged to have four DHR circuits, each with 10 MWt heat removal capacity. Two circuits will be fully under natural convection while two circuits will be under forced convection. The forced convection circuit is being developed in such a way that each will be able to remove 10 MWt even under natural convection. In addition, to meet the DHR requirement during post accident situations, the concept of vessel cooling is being developed. The CC is designed to take care of whole core melt down with B4C spikes introduced to prevent re-criticality. To take care of high temperature molten corium attack, the CC materials/coatings are under development. 7. Metal fuel program

Fig. 32. Core configuration of 600 MWe FBR.

Table 1 Core parameters of FBR-600. Parameter

Value

Power (MWe) Pu enrichment (wt %) Active fuel length (mm) Bottom axial blanket length (mm) Fuel pin/Spacer wire diameter (mm) Radial blanket pin diameter (mm) No of pins per SA in core/blanket Peak Linear Power (W/cm) Peak discharge burnup (GWd/t) Pu inventory (t) Cycle length (efpd) Residence time Maximum core DT (C) Sodium void coefficient ($) Breeding ratio Reactor doubling time (years)

600 19.9/26.5 1000 400 6.6/1.6 14.3 217/61 450 100 ~2.5 210 3 cycles 150 <1 1.11 24

sensitive magnetic switches based on Curie point in series with the power supply circuit of electromagnet for DSRDMs. Additionally an Ultimate Shutdown System (USS) comprising of 6 modules; each containing neutron absorbers like B4C granules is under development. The function of USS is to add sufficient negative reactivity to the core, with some delay, under ATWS type of events for ensuring ultimate sub-criticality and preventing energetic CDA. The design envisages 3 PSP and 4 IHX. To enhance the safety and

The initial FBRs would employ MOX fuel due to its proven experience. However, in order to achieve high growth rate in U-Pu cycle in the second stage, FBRs with metal fuels, having high breeding ratio and low doubling time are essential. This warrants deployment of metal fuelled fast reactors at the earliest. (The same may not be applicable to FBR with Th bearing fuel, as the breeding ratio change from ceramic to metal fuel is minimal). Before launching the metal fuelled power reactors, it is imperative to understand the metal fuel behavior and master the metal fuel reactor technology along with closed fuel cycle (Grover and Chandra, 2006). The strategy for metal fuel development is through pin level irradiation in FBTR and then assembly level irradiation followed by loading substantial part of FBTR core with metal fuel assemblies. Direct data generation is essential as the experience internationally is also very limited and restricted. This necessitates the need for a dedicated metal fuelled test reactor before going for power reactor scale of 1000 MWe (Devan et al., 2011). While designing the test reactor, it is preferred to have the same fuel pin design, which can be directly adopted in the subsequent metal fuelled power reactors. Further, the design ratings, viz. linear heat rating (LHR) and temperatures are preferred to be the same as that of the power reactors, to the extent possible. This means that the active fuel column height be near 1000 mm, fuel pin diameter be in the optimum range of 8e9 mm and LHR be in the range of 400e450 W/cm. Also, the neutron flux should be as high as possible to have faster irradiation tests. From the cost and fissile inventory considerations, it is desired to design the test reactor with the lowest power possible and less fissile inventory. With the approach enumerated above, preliminary conceptual design of test reactor core are being evolved by IGCAR. The test reactor would be of 100 MWt capacity which is likely to employ ternary metal fuel with a composition of Natural U-23 wt % Pue6 wt% Zr. The reference design is of sodium bonded type, though the choice is open for mechanical bonded concept also. Apart from serving its primary objective of generating experience with the metallic fuels, it would also cater to other important applications such as isotope production, material irradiations and testing, sensor calibrations, etc. over a period of about 60 years. It is expected to serve as a replacement for FBTR, which has got a limited residual life of around 12e15 calendar years. With the establishment of this reactor, India will continue to have a fast neutron spectrum research reactor. 8. Summary Fast Breeder reactors form the 2nd stage of the Indian Nuclear

Please cite this article in press as: Puthiyavinayagam, P., et al., Development of fast breeder reactor technology in India, Progress in Nuclear Energy (2017), http://dx.doi.org/10.1016/j.pnucene.2017.03.015

P. Puthiyavinayagam et al. / Progress in Nuclear Energy xxx (2017) 1e24

Power program. It is vital for the nuclear capacity growth considering the available nuclear resources in India. Development of FBR technology was spearheaded by Department of Atomic Energy with the establishment of a dedicated Research Centre called IGCAR (originally known as RRC) and launching of a test reactor called FBTR. FBTR has been operating for over 30 years with a unique Plutonium rich mixed carbide fuel. The successful operation of FBTR gave the confidence for launching the power reactors of higher capacity. In terms of reactor power and other performance related parameters such as LHR and burnup, FBTR evolved gradually beginning from a modest 10 MWt, 250 W/cm and 25 GWd/t respectively. The peak burnup has reached 165 GWd/t and LHR has been raised to 400 W/cm. FBTR performance has been very satisfactory and many irradiation tests have been carried out. In parallel, activities started on the design of 500 MWe PFBR, a techno economic demonstration of Fast breeder technology. Since the scale up is high, the choice of fuel and reactor systems were chosen systematically based on a thorough review of the experience gained world wide and fitting within the Indian industrial capacity and capability. The design of PFBR was accomplished, backed by an extensive R&D program, reactor physics analysis, detailed numerical analysis in the area of structural mechanics and thermal hydraulics and plant dynamic studies. Due to large size reactor components, a large technology development exercise was undertaken addressing the manufacturing challenges and to establish the manufacturing methods to the required precision. Subsequent to the design and manufacturing technology development, engineering tests were undertaken on full scale as well as scaled down models to ascertain the performance qualification. Those tests were undertaken in both sodium and water facilities. Further, safety experiments were carried out to understand the sodium fire behavior and to evolve the fire mitigation system. Also, experiments were conducted on the severe accident scenario including post accident heat removal aspects. In parallel, an indepth materials development program was also undertaken on the class of steels employed in the PFBR NSSS. PFBR project was launched in 2003 by BHAVINI and the construction challenges in the project are briefly covered. Subsequent to PFBR, incorporating the rich experience that has been obtained through PFBR, design of 600 MWe FBR with improved economy and enhanced safety is being undertaken. Salient aspects of FBR 600 are explained. Subsequent to MOX fuelled reactors, metal fuel based reactors are going to be launched. To enable this, metal fuel development is undertaken and to get experience, a test reactor is planned. Acknowledgement This paper brings out the work carried out at IGCAR on the development of FBR technology over few decades. The contributions of all those who participated in the development program is acknowledged. Also, the efforts and help rendered by several colleagues in the preparation of this paper is acknowledged. References Bachchan, Abhitab, Devan, K., et al., 2015. In: Conceptual Design of Future Fast Reactor Core with Enhanced Safety and High Breeding, Proceedings of the 20th National Symposium on Radiation Physics (NSRP-20), RPNI-2, 535. Mangalore University, Mangalore, India. Banerjee, I., et al., 2013. Development of gas entrainment mitigation devices for PFBR hot pool. J. Nucl. Eng. Des. 258, 258e265. Chellapandi, P., Srinivasan, R., Chetal, S.C., Raj, Baldev, 2006. Experimental creep life assessment of tubular structures with geometrical imperfections in welds with reference to fast reactor plant life. Int. Jl. Press. Vessel Pip. 83. March 2006. Chellapandi, P., Chetal, S.C., Raj, Baldev, 2009. Investigation of Structural Mechanics

23

Failure Modes in FBR’, Pressure Vessels and Piping : Codes, Standards, Design and Analysis. Narosa Publishing House, New Delhi. Chellapandi, P., Chetal, S.C., Raj, Baldev, 2010a. Structural integrity assessment of reactor assembly Components of A Pool type sodium fast reactor under core disruptive accident - Part 1: development of computer code and validations. J. Nucl. Technol. 172 (1). Oct. 2010a. Chellapandi, P., Chetal, S.C., Raj, Baldev, 2010b. Structural integrity assessment of reactor assembly Components of A Pool type sodium fast reactor under core disruptive accident - Part 2: analysis for a 500 MWe prototype fast breeder reactor. J. Nucl. Technol. 172 (1). Oct. 2010b. Chetal, S.C., Balasubramaniyan, V., Chellapandi, P., et al., 2006. The design of the prototype fast breeder reactor. Nucl. Eng. Des. 236 (Issues 7e8), 852e860. Cullen, D.E., 2001. The ENDF Preprocessing Codes”, IAEA-NDS-39, Nuclear Data Section. International Atomic Energy Agency, Vienna, Austria. PREPRO-2000 obtained from IAEA Nuclear Data Section on CDROM. Das, S.K., Sharma, A.K., et al., 2013. Post-accident heat removal: numerical and experimental simulation. Nucl. Eng. Des. 265, 1246e1254. Devan, K., Bachchan, Abhitab, et al., 2011. Physics design of experimental metal fuelled fast reactor cores for full scale demonstration. Nucl. Eng. Des. 241 (Issue 8), 3058e3067. Devan, K., Alagan, M., et al., 2012. A comparative study of FARCOB and ERANOS 2.1 neutronic codes in predicting the Phenix control rod withdrawal end-of-life experimental results. Nucl. Eng. Des. 245, 89. Diwakar, S.V., Sundararajan, T., et al., 2008. Development of mathematical model for optimization of sodium leak collection tray. Nucl. Eng. Des. 238 (10), 2684e2692. Diwakar, S.V., Mangarjuna Rao, P., et al., 2011. Numerical prediction of fire extinguishment characteristics of sodium leak collection tray in a fast breeder reactor. Nucl. Eng. Des. 241, 5189e5202. Gnanadhas, Lydia, Sharma, A.K., Malarvizhi, B., et al., 2011. PATH e an experimental facility for natural circulation heat transfer studies related to post accident thermal hydraulics. Nucl. Eng. Des. 241, 3839e3850. Grover, R.B., Chandra, Subhash, 2006. Scenario for growth of electricity in India. Energy Policy 34 (Issue 17), 2834e2847. Harish, R., Sathiyasheela, T., et al., 1999. KALDIS: a Computer Code System for Core Disruptive Accident Analysis of Fast Reactors, IGCe208, IGCAR Report (Kalpakkam). Hemanath, M.G., Meikandamurthy, C., et al., 2010. Theoretical and experimental performance analysis for cold trap design. Nucl. Eng. Des. 240, 2737e2744. Hughes, W.F., Gaylord, E.W., 1964. Basic Equations of Engineering Science. McGrawHill publications, N.Y. IAEA TECDOC-1139, 2000. Transient and Accident Analysis of a BN-800 Type LMFR with Near Zero Void Effect. CRP TECDOC. IAEA, Vienna. IAEA TECDOC-1623, 2010. BN-600 Hybrid Core Benchmark Analyses, CRP TECDOC. IAEA, Vienna. IAEA TECDOC-1626, 2010. Advanced Reactor Technology Options for Utilization and Transmutation of Actinides in Spent Nuclear Fuel, CRP TECDOC. IAEA, Vienna. IAEA TECDOC-731, 1994. Evaluation of Benchmark Calculations on a Fast Power Reactor Core with Near Zero Sodium Void Effect, CRP TECDOC. IAEA, Vienna. IGCAR, 1983. Conceptual Design Report of PFBR, IGCAR Internal Report. Ignatius Sundar Raj, S., Sreedhar, B.K., 2013. Sodium removal from sodium wetted under sodium ultrasonic scanner. Nucl. Eng. Des. 254, 120e128. Kakodkar, Anil, 2008. Evolving Indian Nuclear Programme: Rationale and Perspectives. Indian Academy of Sciences, Bangalore. July 4, 2008. Kasiviswanathan, K.V., Karthik, V., et al., 2011. Performance assessment of fuel and core structural materials irradiated in FBTR. Energy Procedia 7, 129e139. Kishore, S., Ashokkumar, A., et al., 2012. An experimental study on impingement wastage of Mod 9Cr-1Mo steel due to sodium water reaction. Nucl. Eng. Des. 242, 49e55. Kishore, S., Beauchamp, Francois, et al., 2016. Impingement wastage experiments with 9Cr-1Mo. Steel. Nucl. Eng. Des. 297, 104e110. Kumar, Prabhat, 2014. An overview of welding and fabrication aspects during manufacture of nuclear reactor components for 500 MWe prototype fast breeder reactor. Nucl. India 56 (No. 1e3). Kumar, H., Ramakrishnan, V., et al., 2013. Tribologicalbehaviour of AFNOR grade Z6 NCT DV 25-15 precipitation hardening austenitic stainless steel with Ni-Br coating in high temperature liquid sodium, Surface Effects and Contact Mechanics XI. WIT Trans. Eng. Sci. 78, 36e45. Kumar, Amit, Subramanian, Venkatesan, V., et al., 2015. Characterisation of sodium aerosols in cover gas region of SILVERINA loop. Aerosol Air Qual. Res. 15, 1813e1822. Launder, B.E., Spalding, D.B., 1974. The numerical computation of turbulent flow. Comput. Methods Appl. Mech. Eng. 13, 269e289. Maity, R.K., Velusamy, K., et al., 2011. Computational fluid dynamic investigations of partial blockage detection by core-temperature monitoring system of a sodium cooled fast reactor. Nucl. Eng. Des. 241, 4994e5008. Mannan, S.L., Chetal, S.C., et al., 2003. Selection of materials for prototype fast breeder reactor. Trans. Indian Inst. Metals. ISSN: 0019-493X 56 (2), 155e178. Mathew, M.D., et al., 2013. Development of IFAC-1 SS: an advanced austenitic stainless steel for cladding and wrapper tube applications in sodium-cooled fast reactors. Adv. Mater. Res. 794, 749e756. Meikandamurthy, C., Kumar, H., et al., 2012. Self Welding susceptibility of cold worked alloy D9 and 316 LN austenitic Steels in flowing sodium. Material Sci. Technol. 28 (No.12). Mohanakrishnan, P., 2008. Development and validation of a fast reactor core

Please cite this article in press as: Puthiyavinayagam, P., et al., Development of fast breeder reactor technology in India, Progress in Nuclear Energy (2017), http://dx.doi.org/10.1016/j.pnucene.2017.03.015

24

P. Puthiyavinayagam et al. / Progress in Nuclear Energy xxx (2017) 1e24

burnup code- FARCOB. Ann. Nucl. Energy 35, 158. Muthu Saravanan, S., Mangarjuna Rao, P., et al., 2011. Numerical investigation of sodium droplet ignition in the atmospheric air. Int. J. Nucl. Energy Sci. Technol. 6 (4), 284e297. Muthu Saravanan, S., Mangarjuna Rao, P., et al., 2016. NAFCON-SF: a sodium spray fire code for evaluating thermal consequences in SFR containment. Ann. Nucl. Energy 90. Nashine, B.K., Rao, B.P.C., 2014. Design, in-sodium testing and performance evaluation of annular linear induction pump for a sodium cooled fast reactor. Ann. Nucl. Energy 73, 527e536. Natarajan, R., 2017. Reprocessing of spent nuclear fuel: present challenges and future program. Prog. Nucl. Energy. Natesan, K., Kasinathan, N., et al., 2006. Thermal hydraulic investigations of primary coolant pipe rupture in an LMFBR. Nucl. Eng. Des. 236 (No.11), 1165e1178. Natesan, K., Velusamy, K., et al., 2007. Thermal hydraulic study on random failure of fuel by delayed neutron detection system. Nucl. Eng. Des. 237 (23), 2219e2231. Natesan, K., Kasinathan, N., et al., 2012. Plant dynamics studies towards design of plant protection system for PFBR. Nucl. Eng. Des. 250, 339e350. Naveen Raj, M., Velusamy, K., 2016. Characterization of velocity and temperature fields in a 217 pin wire wrapped fuel bundle of sodium cooled fast reactor. Ann. Nucl. Energy 87, 331e349. Padmakumar, G., Vinod, V., et al., 2013. SADHANA facility for simulation of natural convection in the SGDHR system of PFBR. Prog. Nucl. Energy 66, 99e107. Pandey, G.K., et al., 2011. Development of Labyrinth type sealing devices for PFBR subassemblies. J. Nucl. Technol. 175 (No.3), 692e699. Pandey, G.K., et al., 2013. Development of honeycomb type orifices for flow zoning in PFBR. Nucl. Eng. Des. 262, 63e71. Partha Sarathy, U., Sundararajan, T., et al., 2012. Thermal hydraulic design of safety grade decay heat removal system for a fast breeder reactor. Nucl. Eng. Des. 250, 480e499. Patankar, S.V., 1980. Numerical Heat Transfer and Fluid Flow, Hemisphere. Patri, Sudheer, Ruhela, S.P., et al., 2014. Experimental evaluation of structural integrity of scram release electromagnet. Nucl. Eng. Des. 274, 90e99. Ponraju, D., Hemanth Rao, E., et al., 2013. Experimental and Numerical Simulation towards Sodium Safety in SFR, International Conference on Fast Reactors and Related Fuel Cycles. Safe Technologies and Sustainable Scenarios (FR-13), Paris, France. Prahlad, B., Kale, R.D., et al., 1990. Thermal performance tests on a sodium-tosodium heat exchanger. J. Nucl. Sci. Technol. 27 (Issue 6), 547e553. Puthiyavinayagam, P., 2015. 48th annual meeting of TWGFR, IAEA,IPPE, obninsk. Russ. Fed. 25e29. Raghukumar, G., Reddy, C.P., et al., 2013. In: Effects of Spatial Variation in Burnup of Fuel SA on the Reactivity in FBTR Irradiation Campaigns, Proceedings of Advances in Reactor Physics: Simulation Techniques and Analysis Methodologies, ARP-2013. BARC, Mumbai. Raj, Baldev, 2007. Status of Sodium Cooled Fast Reactors with Closed Fuel Cycle in India. International Congress on Advances in Nuclear Power Plants, Nice, France. May 13-18, 2007. Raj, Baldev, Aug.2008. Science and Technology of Fast Breeder Reactors: an Indian Success Story. Professor Brahm Prakash Memorial Lecture. The Indian Institute of Metals, Bangalore Chapter, The Indian Institute of Science, Bangalore, 21st August 2008. Raj, Baldev, Dec.2008. Science and Technology of Fast Breeder Reactor Programme in India: Challenges and Achievements. Prof. Jai Krishna Memorial Award 2008. Annual Convention of Indian National Academy of Engineering, Goa. December 5-6, 2008. Rajan Babu, V., Veerasamy, R., et al., 2010. Testing and Qualifciation of control & safety rod and its drive mechanism of fast breeder reactor. Nucl. Eng. Des. 240, 1728e1738. Rajasundaram, P., Vijayaraghavan, S., et al., 2013. Experience in Long Term Operation of Sodium Loop for Creep Experiments in Dynamic Sodium at High Temperature, Procedia Engineering, vol. 55. Elsevier, pp. 842e849. Ramakrishna, M., Sharma, P.K., et al., 2012. Insights from Level-1 probabilistic safety analysis of prototype fast breeder reactor. Nucl. Eng. Des. 250, 664. Ramakrishna, Ranga, Kishore, S., et al., 2016. Experimental studies on acoustic leak detection in steam generators of fast breeder reactors. J. Maintenance Eng. 1, 254e266. Ramdasu, D., et al., 2005. Experimental Verification of the Hydraulics of Mixed Oxide Fuel Subassembly for Fast Breeder Test Reactor ; 32nd National

Conference on Fluid Mechanics and Fluid Power (NCFMFP e32) (Osmanabad, Maharashtra). Ramdasu, D., et al., 2006. Devices to mitigate gas entrainment in the surge tank of PFBR. Paper No. ICONE14-89095 ASME Digit. Collect. 2, 55e61. Reddy, A.V.G., et al., 2013. Challenges in the manufacture of subassemblies for the Indian fast reactors; IAEA paper for technical Meeting on design. Manuf. Irradiat. Behav. Fast React. Fuel; Obninsk Russ. Fed. IAEA-TECDOC-CD-1689, 30 May - 3 Jun 2011. Satpathy, K., Velusamy, K., et al., 2013. Numerical simulation of liquid fall induced gas entrainment and its mitigation. Int. J. Heat Mass Transf. 60, 392e405. Setty, D.S., Kapoor, K., Saibaba, 2017. Nuclear fuel cycle-developments and challenges in fuel fabrication technology in India. Prog. Nucl. Energy. Shanmugavel, M., Vijayaraghavan, S., et al., 2011. Operating experience of High temperature sodium loop for material testing. Energy Procedia 7, 609e615. Sharma, Prashant, Suresh, S.K., et al., 2010. Development, computer simulation and performance testing in sodium of an eddy current flowmeter. Ann. Nucl. Energy 37, 332e338. Sivaibharasi, N., Thyagarajan, K., et al., 2012. Evaluation of microstructural, mechanical properties and corrosion behavior of AISI type 316LN stainless steel and modified 9Cr-1Mo steel exposed in a dynamic bimetallic sodium loop at 798 K (525 _C) for 16,000 hours. Metallurgical Mater. Transactions-A 43A, 561e571. Snehalatha, V., Ponraju, D., et al., 2013. Synthesis and characterization of carbon microsphere for extinguishing sodium fire, CARBON MATERIALS 2012. AIP Conf. Proc. 1538, 89. Sreeramamurthy, A., Ashokkumar, A., et al., 2014. Performance of semiconducting oxide based hydrogen sensor for argon cover gas in engineering scale sodium facility. Nucl. Eng. Des. 273, 555e559. Srinivasan, G., et al., 2006. Fast breeder test reactor. Nucl. Eng. Des. 236, 796e811. Srinivasan, G., Varatharajan, S., et al., 2010a. The performance of the carbide fuel. INS Newsl. 7 (No.3), 24e27. July e Sep. Srinivasan, G., Reddy, C.P., et al., 2010b. Engineering & physics tests in FBTR. INS Newsl. 7 (No.4), 18e22. Srinivasan, G., Keshavamurthy, R.S., et al., 2014. Delayed neutron measurements with a natural uranium fission product source in fast breeder test reactor. Ann. Nucl. Energy 65, 314e319. Sunil Kumar, D., Keshavamurthy, R.S., et al., 2010. A feasibility study of ferro-boron as in-core shield material in fast breeder reactors. Nucl. Eng. Des. 240, 2972. Suresh Kumar, K.V., Babu, A., et al., 2011. Twenty five years of operating experience with the fast breeder test reactor. Energy Procedia 7, 323e332. Suresh Kumar, V.A., Noushad, I.B., et al., 2012. Steam generator test facility-a test bed for steam generators of indian sodium cooled fast breeder reactors. Nucl. Eng. Des. 248, 169e177. Sylvia, J.I., Jeyan, M.R., et al., 2013. Ultrasonic imaging of projected components of PFBR. Nucl. Eng. Des. 258, 266e274. Varatharajan, S., Sureshkumar, K.V., et al., 2010. In: Progressive Evolution of the Core of the Fast Breeder Test Reactor, Proceedings of the 18th International Conference on Nuclear Engineering (ICONE 18) (Xian, China). Velusamy, K., Ramadevi, A., et al., 1990. Thermal hydraulic analysis for the locations of core cover plate and core monitors in PFBR. J. Energy Heat Mass Transf. 12, 163e169. Velusamy, K., Sundararajan, T., et al., 2005. Investigations of thermal striping in primary circuit of prototype fast breeder reactor. Proc. 13th Int. Conf. Nulear Engg. 16e20. Beijing, China, May. Velusamy, K., Chellapandi, P., et al., 2010. Challenges in pool hydraulic design of indian prototype fast breeder reactor. SADHANA 35 (Part 2), 97e128. Venkataraman, G., 1994. Bhabha and His Magnificent Obsessions. Universities Press (India) Ltd., ISBN 978-81-7371-859-5 Vijayakumar, G., Nashine, B.K., et al., 2011a. Development and testing of PM flowmeter with samarium cobalt magnet assembly. J. Energy Procedia 7, 630e637. Vijayakumar, G., Rajan, K.K., et al., 2011b. Performance evaluation of PFBR wire type sodium leak detectors. Nucl. Eng. Des. 241, 2271e2279. Vinod, V., Chandramouli, S., 2014. Experimental study on the transient response of passive decay heat removal system. Nucl. Eng. Des. 280, 564e569. Vinod, V., Shivakumar, L.S., et al., 2014. Experimental evaluation of heat transfer performance of sodium heated once through steam generator. Nucl. Eng. Des. 273, 412e423. Wattal, P.K., 2017. Back end of indian nuclear fuel cycle e a Road to sustainability. Prog. Nucl. Energy 017.

Please cite this article in press as: Puthiyavinayagam, P., et al., Development of fast breeder reactor technology in India, Progress in Nuclear Energy (2017), http://dx.doi.org/10.1016/j.pnucene.2017.03.015