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Journal articles on the topic 'Prototype Fast Breeder Reactor (PFBR)'

Author: Grafiati

Published: 4 June 2021

Last updated: 11 June 2025

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1

Anuraj, Vijayan, Goverapet Srinivasan, and Kunhiraman Devan. "Stability characteristics of the 500 mw Indian PFBR." Nuclear Technology and Radiation Protection 30, no. 2 (2015): 113–23. http://dx.doi.org/10.2298/ntrp1502113a.

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After the successful operation of the fast breeder test reactor for over two decades, India is now nearing the completion of a 500 MW (electrical) prototype fast breeder reactor. This commercial scale power reactor is a sodium-cooled, pool-type, mixed-oxide fuelled fast reactor. The stability characteristics of the reactor are an important safety aspect to be studied. In the present work, linear stability of the prototype fast breeder reactor analysis is carried out using the transfer function method, while the stability of the system is checked via the Nyquist criteria. For the completeness of the study, transient analysis with various kinds of reactivity perturbations was carried out. The response of the system in both cases indicated that the system is stable.

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2

Kumar, Prabhat. "Welding and Fabrication of Stainless Steel Equipment for 500MWe Prototype Fast Breeder Reactor." Advanced Materials Research 794 (September 2013): 305–15. http://dx.doi.org/10.4028/www.scientific.net/amr.794.305.

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Austenitic stainless steel is the major structural material for the primary and secondary sodium systems (except for the steam generators) for the currently operating and planned fast reactors all over the world. The boundaries of sodium systems of Prototype Fast Breeder Reactor (PFBR) is designed so as to have an extremely low probability of leakage, rapidly propagating failure and gross rupture under the static & dynamic loads expected during various operating conditions.The degradation of material properties (e.g. effect of sodium, temperature and irradiation), transients, residual stresses, flaw size etc. are the important considerations, which shall be taken into account. The principal material of construction for PFBR is austenitic stainless steel of grade 316LN/304LN. The scope of welding and fabrication of PFBR components is too large due to versatile types of systems with varieties of components with complex constructional features. High operating temperature of various systems causing high stresses are to be minimized by designing thin walled structure. Most of the Nuclear Steam Supply System (NSSS) components are thin walled and require manufacturing in separate nuclear clean hall conditions to assure the quality.The welding with stringent tolerances along with high distortion in stainless steels due to high thermal expansion and low thermal conductivity makes the fabrication extremely challenging.The welding standards and acceptance criteria of PFBR equipment is stringent compared to other industrial specification. Manufacture of over dimensional components (diameter greater than 12m and thickness upto 40mm) such as MainVessel, Safety Vessel, Inner Vessel involves die pressing of large size dished end & conical petals. The solution annealing of cold worked petals is a mandatory requirement if strain exceeds 10%. Innovative welding techniques and many trials were conducted on mock up for establishing the process parameters. The forming techniques, bending methods and welding procedures were qualified with stringent non-destructive and destructive examinations and testing before taking up the actual job. Thermal Baffle has two large concentric cylindrical shells, inner and outer shells of about 12.4m diameter and fabrication is a challenging task. PFBR also involves dissimilar joint welding between carbon steel (A48P2) and austenitic stainless steel (316LN) at integration location of roof slab & main vessel. This welding is carried out by combination of Gas Tungsten Arc Welding (GTAW) & Shielded Metal Arc Welding (SMAW) processes using ER 309L & E 309-16 welding consumables with controlled heat input to minimize the dilution of carbon & distortion. The weld between primary pipe & grid plate cannot be accessed for in-service inspection and therefore requires extra-ordinary skilled welders. Space constraints & lack of accessibility makes the welding & inspection challenging. This paper highlights the welding and fabrication aspects of few major, over dimensional and critical equipment of 500MWe Prototype Fast Breeder Reactor. Keywords: Stainless Steel, Main Vessel, Safety Vessel, Inner Vessel, Grid Plate, PFBR, SS welding, distortion.

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3

Azhagarason, B., N. Mahendran, Tarun Kumar Mitra, and Prabhat Kumar. "Technological Challenges in Manufacturing of over Dimensional Stainless Steel Components of PFBR." Advanced Materials Research 794 (September 2013): 186–93. http://dx.doi.org/10.4028/www.scientific.net/amr.794.186.

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Prototype Fast Breeder Reactor (PFBR) is liquid sodium cooled, pool type nuclear reactor with generating capacity of 1250 MWt / 500MWe. Reactor assembly consists of many large dimensional components made of special grade austenitic stainless steel material. Safety vessel and Main vessel are torispherical dished end vessels with overall height of 12.8 m and 13.4/12.9 m diameter with thickness ranging from 20 to 40 mm. Vessels approx. 111 / 135 MT with running weld length of 500 & 540 m. Inner vessel and thermal baffles are the internals of reactor assembly made of SS 316LN. Forming of dished end petals, weld overlay on the inside surface, circumference matching between the cylindrical shells, cylindrical shell to dished portion was achieved within the tolerances specified. Due to limitations of transportation, these large sized components were manufactured at PFBR site. This paper discusses the experiences gained during the manufacturing of such over dimensional components at PFBR site in meeting the stringent tolerances on various dimensions and NDE requirements.

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4

Balaguru, S., K. Deenadayalan, Murali Vela, and P. Chellapandi. "Influence of Welding Speed over Dilution for Circular Grid Plate Hardfaced with Colmonoy-5." Applied Mechanics and Materials 565 (June 2014): 53–58. http://dx.doi.org/10.4028/www.scientific.net/amm.565.53.

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Colmonoy is a nickel based alloy hardfaced with SS-316 used in Prototype Fast Breeder Reactors (PFBR) to avoid self-welding and galling between nuclear reactor and grid plate. Colmonoy is coated on the base metal –SS 316 by Plasma transfer arc welding. Due to thermal cycling that exist in reactors de-bonding between colmonoy and base metal occurs which may lead to catastrophic failure of the Prototype fast breeder reactor. Hence, estimation of the fatigue life period of hardfaced colmonoy is mandatory. Before that, finding optimum parameters which influences fusion, weld strength, composition and dilution are mandatory. In order to evaluate the dilution characteristics with respect to welding speed, different specimens were prepared by holding other parameters constant. Prepared specimens were tested for hardness, microstructure and dilution to locate the difference between each specimen. Comparative studies were carried out on the specimen and their parameters like travelling speed, microscopic structure, Hardness and dilution percentage were tabulated. Based on the tabulated results, the best specimen is chosen with respect to its individual parameters.

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5

Dutta, N. G. "Helium Leak Detection of Vessels in Fuel Transfer Cell (FTC) of Prototype Fast Breeder Reactor (PFBR)." Journal of Physics: Conference Series 390 (November 5, 2012): 012080. http://dx.doi.org/10.1088/1742-6596/390/1/012080.

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6

Rangasamy, R. G., and Prabhat Kumar. "Quality Management during Manufacturing of High Tempertaure Thin Walled Austenitic Stainless Steel Sodium Tanks of Prototype Fast Breeder Reactor." Advanced Materials Research 794 (September 2013): 507–13. http://dx.doi.org/10.4028/www.scientific.net/amr.794.507.

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Austenitic stainless steels are the major material of construction for the fast breeder reactors in view of their adequate high temperature mechanical properties, compatibility with liquid sodium coolant, good weldability, availability of design data and above all the fairly vast and satisfactory experience in the use of these steels for high temperature service. All the Nuclear Steam Supply System (NSSS) components of FBR are thin walled structure and require manufacture to very close tolerances under nuclear clean conditions. As a result of high temperature operation and thin wall construction, the acceptance criteria are stringent as compared to ASME Section III. The material of construction is Austenitic stainless steel 316 LN and 304 LN with controlled Chemistry and calls for additional tests and requirements as compared to ASTM standards. Prototype Fast Breeder Reactor (PFBR) is sodium cooled, pool type, 500 MWe reactor which is at advanced stage of construction at Kalpakkam, Tamilnadu, India. In PFBR, the normal heat transport is mainly through two secondary loops and in their absence; the decay heat removal is through four passive and independent safety grade decay heat removal loops (SGDHR). The secondary sodium circuit and the SGHDR circuit consist of sodium tanks for various applications such as storage, transfer, pressure mitigation and to take care of volumetric expansion. The sodium tanks are thin walled cylindrical vertical vessels with predominantly torispherical dished heads at the top and bottom. These tanks are provided with pull-out nozzles which were successfully made by cold forming. Surface thermocouples and heaters, wire type leak detectors are provided on these tanks. These tanks are insulated with bonded mineral wool and with aluminum cladding. All the butt welds in pressure parts were subjected to 100% Radiographic examination. These tanks were subjected to hydrotest, pneumatic test and helium leak test under vacuum. The principal material of construction being stainless steel for the sodium tanks shall be handled with care following best engineering practices coupled with stringent QA requirements to avoid stress corrosion cracking in the highly brackish environment. Intergranular stress corrosion cracking and hot cracking are additional factors to be addressed for the welding of stainless steel components. Pickling and passivation, Testing with chemistry controlled demineralised water are salient steps in manufacturing. Corrosion protection and preservation during fabrication, erection and post erection is a mandatory stipulation in the QA programme. Enhanced reliability of welded components can be achieved mainly through quality control and quality assurance procedures in addition to design and metallurgy. The diverse and redundant inspections in terms of both operator and technique are required for components where zero failure is desired & claimed. This paper highlights the step by step quality management methodologies adopted during the manufacturing of high temperature thin walled austenitic stainless steel sodium tanks of PFBR.

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7

Balaguru, S., K. Saranraj, Murali Vela, and P. Chellapandi. "Variation in Residual Stresses due to Thermal Cycling Induced on the Hardfaced Grid Plate in PFBR." Applied Mechanics and Materials 591 (July 2014): 98–102. http://dx.doi.org/10.4028/www.scientific.net/amm.591.98.

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In sodium cooled fast breeder reactors, the high operating temperature necessitates hard facing of grid plate to avoid galling and to reduce the wear loss of the base material . Nickel-base cobalt-free alloy called Colmonoy-5 has been chosen as the hard facing material for the Prototype Fast Breeder Reactor (PFBR) grid plate. The grid plate (GP) which is a critical component made of 316 LN SS that holds the core subassemblies are hard faced, to prevent galling and also to minimize wear caused by subassembly insertion/removal and erosion due to high sodium velocity at 670 K. Thermal cycling of hard faced circular grid plate made by Plasma Transferred Arc Welding (PTAW) generates residual stresses due to differential shrinkage of the molten deposit and difference in coefficients of thermal expansion between the deposit and substrate material. In this project the effect of thermal cycling of a nickel-base hardfacing alloy deposited on an austenitic stainless steel grid plate was studied. Finite element analysis of hard faced circular grid plate is performed for obtaining residual stress which includes elasto-plastic analysis. Coupled thermo-mechanical analysis is done for thermal cycling of hardfaced circular gridplate to quantitatively estimate the residual stress.

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8

Perumalsamy, G., P. Visweswaran, D. Jagadishan, S. Joseph Winston, and S. Murugan. "Optimisation of helical geometry of inspection probe for steam generator tubes of the PFBR." Insight - Non-Destructive Testing and Condition Monitoring 63, no. 10 (2021): 585–91. http://dx.doi.org/10.1784/insi.2021.63.10.585.

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The steam generator (SG) tubes of the prototype fast breeder reactor (PFBR) located in Kalpakkam, India, need to be periodically inspected using the remote field eddy current (RFEC) technique. During the pre-service inspection of the SG tubes, it was found that the RFEC probes experienced frequent mechanical breakages. To avoid these failures, changes in the existing structural design of the RFEC probe were required. A helical groove design was proposed to obtain a smooth transition in the variation of stress across the probe during the inspection. It was difficult to calculate the flexural stiffness of the proposed helical geometry probe due to the varying cross-section along its length. In this paper, the smearing approach adopted to calculate the stiffness of the RFEC probe and the sensitivity analysis carried out to determine the optimal design of the probe are discussed. A probe was fabricated based on the helical groove design and tested to qualify its suitability for the SG inspection. The RFEC probe with helical grooves was employed for the pre-service inspection of the SG tubes of the PFBR. More than 200 tubes have been inspected using the proposed design and no mechanical failure of the probe has been observed.

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9

Balaguru, S., K. Saranraj, Murali Vela, and P. Chellapandi. "Evaluation of Stress Intensity Factor of 316 SS LN Circular Grid Plate Hardfaced with Colmonoy 5 for Various Loads." Applied Mechanics and Materials 431 (October 2013): 173–76. http://dx.doi.org/10.4028/www.scientific.net/amm.431.173.

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In sodium cooled fast breeder reactors, the high operating temperature necessitates hardfacing of grid plate to avoid galling of the base material. Nickel based cobalt-free alloy called Colmonoy-5 has been chosen as the hard facing material for the Prototype Fast Breeder Reactor (PFBR) grid plate. The grid plate (GP) which is a critical component made of SS316 that holds the core subassemblies are hard faced, to prevent galling caused by subassembly insertion/removal and erosion due to high sodium velocity at 670 K. Colmonoy 5 was hardfaced on circular grid plate along the annular groove by Plasma Transferred Arc Welding (PTAW). The process-induced residual stress generated due to difference in coefficients of thermal expansion between the deposit and substrate material was removed by annealing. In this paper, in order to evaluate Stress Intensity Factor (SIF) near the crack tip of the crack caused by the remote load or residual stresses generated due to thermal cycling owing to on/off of nuclear reactor, a grid plate model hard-faced with colmonoy 5 was prepared and then annealed to remove process-induced residual stress. Considering the plane strain condition to the same geometry of the model, a 3mm crack was considered at the vertical interface between the steel and the colmonoy. By supporting the specimen on the left and right corners of the upper side, load was applied exactly on the opposite side of crack in the positive direction, simulating mode I loading. After loading the model and getting the solution, KIwas obtained at the crack tip. This step was repeated for various loads to obtain the relationship between SIF and load for fracture characteristics.

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10

Kumar Krovvidi, S. C. S. P., C. S. Surendran, N. Chakraborthy, et al. "Experience with Material Combinations for Sliding Applications in PFBR In-Vessel Fuel Handling Machine." Advanced Materials Research 794 (September 2013): 705–13. http://dx.doi.org/10.4028/www.scientific.net/amr.794.705.

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Prototype Fast breeder Reactor (PFBR) is a pool type reactor of 500 MW(e) capacity using mixed oxides of Uranium and Plutonium as fuel and liquid sodium as coolant, which is currently under advanced stage of construction at Kalpakkam. In-Vessel handling of the core subassemblies (SA) is carried out by an offset arm type machine called Transfer Arm (TA). Different types of material combinations are utilized for various sliding pairs used in the machine for various motions. Criteria for the selection of these material combinations are decided by the compatibility of the respective machine element with the working environment, magnitude of the contact stress, working temperature, linear speed, availability of external lubrication, required life, required tolerance etc. Transfer of a SA is achieved by gripping/ungripping of SA using fingers, raising / lowering the gripper outer tube and rotation of TA. The drive for gripper finger operation is at ambient environment and finger actuation is in liquid sodium. An inner tube links the linear actuator to the finger actuator and is housed and guided inside the gripper outer tube. Relative movement of inner tube with respect to outer tube results in open / closing of gripper fingers. Initially, combination of material pairs at five nos. of guide locations was SS 304 LN for the outer tube and hardchrome plated SS 304 LN for inner tube. During testing in air after 20 cycles, jamming of inner tube with respect to outer tube was observed. This was solved by reducing number of guides to two, by changing the surface contact to line contact and by changing the material combination to SS 304 LN against colmonoy coated SS 304 LN. Similar failure was observed for sliding movement at guide locations between the outer tube and shielding sleeve during hoisting of the gripper. Initial material combination of SS 304LN and hardchrome plated SS 304 LN was changed to colmonoy coated SS 304LN and hardchrome plated SS 304 LN. The selected material combinations were validated by testing on a separate subassembly simulating the geometry & loading before actual implementation on the machine. Guide tube, which is used to guide the gripper is raised / lowered by means of a screw-nut mechanism. Initial material pair used for the screw and nut, which are working at ambient conditions was SS 304LN and SS 410 respectively to provide corrosion & galling resistance. However during initial performance testing, this material combination failed and the nut got jammed. Subsequently the problem was studied and overcome by changing the material of nut to phosphor bronze, which is relatively softer and hence provided uniform contact across the nut surfaces. Appropriate material selection and proper design of the geometry of guiding surfaces are very essential for the smooth operation of machine elements in sliding conditions. With improvements in the material choice and geometry of the guides, qualification testing of transfer arm was successfully completed in air and hot argon. Testing in sodium is under progress and the experience at high temperature has been encouraging.

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11

Ram Manda, Bhagwana, K. V. Sreedharan, S. Athmalingam, V. Balasubramaniyan, and P. Chellapandi. "Experience with Mechanical Seal of Prototype Fast Breeder Reactor’s Secondary Sodium Pump during Testing." Applied Mechanics and Materials 592-594 (July 2014): 1544–48. http://dx.doi.org/10.4028/www.scientific.net/amm.592-594.1544.

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Prototype fast breeder reactor’s (PFBR) secondary pumps have sealing between the rotating shaft and the stationary parts to prevent cover gas leakage into ambient air. This is achieved by means of triple mechanical seals with oil as buffer fluid. The oil for cooling the mechanical seals and bearing is supplied from an external source and is re-circulated after cooling and cleaning. During testing of secondary pumps several issues related to mechanical seal arrangement, material, oil temperature and oil leakage into process side were confronted. This paper discusses the successful resolution of such issues.

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12

KANI, Yoshio. "Safety Consideration for Prototype Fast Breeder Reactor “MONJU”." Journal of the Atomic Energy Society of Japan / Atomic Energy Society of Japan 45, no. 3 (2003): 190–94. http://dx.doi.org/10.3327/jaesj.45.190.

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13

Takahashi, Tadao, Osamu Yamaguchi, and Tetsuo Kobori. "Construction of the Monju Prototype Fast Breeder Reactor." Nuclear Technology 89, no. 2 (1990): 162–76. http://dx.doi.org/10.13182/nt90-a34343.

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14

Chetal, S. C., V. Balasubramaniyan, P. Chellapandi, et al. "The design of the Prototype Fast Breeder Reactor." Nuclear Engineering and Design 236, no. 7-8 (2006): 852–60. http://dx.doi.org/10.1016/j.nucengdes.2005.09.025.

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15

Jayalal, M. L., T. Jayanthi, S. A. V. Satya Murty, and P. Swaminathan. "Computational Intelligent Systems for Prototype Fast Breeder Reactor." Energy Procedia 7 (2011): 589–96. http://dx.doi.org/10.1016/j.egypro.2011.06.079.

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16

Bhaduri, A. K., S. K. Albert, V. Shankar, and S. K. Ray. "Welding and Hardfacing for the Prototype Fast Breeder Reactor." Indian Welding Journal 39, no. 2 (2006): 18. http://dx.doi.org/10.22486/iwj.v39i2.179200.

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17

Sakthivel, M., and K. Madhusoodanan. "Core Temperature Monitoring System for Prototype Fast Breeder Reactor." Nuclear Science and Engineering 170, no. 3 (2012): 290–93. http://dx.doi.org/10.13182/nse11-07.

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18

Kumar, Aniruddha, R. B. Bhatt, P. G. Behere, and Mohd Afzal. "Ultrasonic decontamination of prototype fast breeder reactor fuel pins." Ultrasonics 54, no. 4 (2014): 1052–56. http://dx.doi.org/10.1016/j.ultras.2013.12.008.

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19

Pandikumar, G., V. Gopalakrishnan, and P. Mohanakrishnan. "Multiple recycling of fuel in prototype fast breeder reactor." Pramana 72, no. 5 (2009): 819–32. http://dx.doi.org/10.1007/s12043-009-0074-z.

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20

Prakash, V., M. Thirumalai, M. Anandaraj, et al. "Experimental qualification of subassembly design for Prototype Fast Breeder Reactor." Nuclear Engineering and Design 241, no. 8 (2011): 3325–32. http://dx.doi.org/10.1016/j.nucengdes.2011.04.040.

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21

Ueda, K., M. Kimura, and S. Ohta. "The Reactor Building Structural Works at Site “Monju”, the Prototype Fast Breeder Reactor." Concrete Journal 27, no. 10 (1989): 34–41. http://dx.doi.org/10.3151/coj1975.27.10_34.

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22

Kumar, Aniruddha, R. B. Bhatt, Mohd Afzal, et al. "Laser-Assisted Decontamination of Fuel Pins for Prototype Fast Breeder Reactor." Nuclear Technology 182, no. 2 (2013): 242–47. http://dx.doi.org/10.13182/nt13-a16434.

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23

Wehmann, Udo, Hidehito Kinjo, and Takeshi Kageyama. "Studies on Plutonium Burning in the Prototype Fast Breeder Reactor Monju." Nuclear Science and Engineering 140, no. 3 (2002): 205–22. http://dx.doi.org/10.13182/nse02-a2257.

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24

Velusamy, K., P. Chellapandi, S. C. Chetal, and Baldev Raj. "Overview of pool hydraulic design of Indian prototype fast breeder reactor." Sadhana 35, no. 2 (2010): 97–128. http://dx.doi.org/10.1007/s12046-010-0022-0.

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25

Biswas, A., R. Srinivasan, P. Chellapandi, and S. C. Chetal. "Leak-before-break analysis of Prototype Fast Breeder Reactor main vessel." Transactions of the Indian Institute of Metals 63, no. 2-3 (2010): 653–56. http://dx.doi.org/10.1007/s12666-010-0098-6.

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26

RaniPatra, Subhra, R. Jehadeesan, H. Seetha, et al. "Event Identification in Prototype Fast Breeder Reactor Subsystem using Artificial Neural Network." International Journal of Applied Information Systems 4, no. 1 (2012): 1–9. http://dx.doi.org/10.5120/ijais12-450605.

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Pandey, G. K., I. Banerjee, G. Padmakumar, et al. "Development of Labyrinth-Type Sealing Devices for Prototype Fast Breeder Reactor Subassemblies." Nuclear Technology 175, no. 3 (2011): 692–99. http://dx.doi.org/10.13182/nt11-a12516.

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SAITO, Shinzo, Koji OKAMOTO, Isao KATAOKA, et al. "ICONE23-1505 SAFETY REQUIREMENTS EXPECTED FOR THE PROTOTYPE FAST BREEDER REACTOR "MONJU"." Proceedings of the International Conference on Nuclear Engineering (ICONE) 2015.23 (2015): _ICONE23–1—_ICONE23–1. http://dx.doi.org/10.1299/jsmeicone.2015.23._icone23-1_232.

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29

M., Ramakrishnan, Pramod Kumar Sharma, Bhuvana V., John Arul A., Mohanakrishnan P., and S. C. Chetal. "Insights from level-1 Probabilistic Safety Analysis of Prototype Fast Breeder Reactor." Nuclear Engineering and Design 250 (September 2012): 664–70. http://dx.doi.org/10.1016/j.nucengdes.2012.05.006.

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Glaser, Alexander, and M. V. Ramana. "Weapon-Grade Plutonium Production Potential in the Indian Prototype Fast Breeder Reactor." Science & Global Security 15, no. 2 (2007): 85–105. http://dx.doi.org/10.1080/08929880701609154.

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Ramakrishnan, M., A. John Arul, V. Bhuvana, P. Puthiya Vinayagam, and P. Chellapandi. "Accident Sequence Modeling Methodology for External Flood Probabilistic Safety Analysis of Prototype Fast Breeder Reactor." Applied Mechanics and Materials 592-594 (July 2014): 2460–64. http://dx.doi.org/10.4028/www.scientific.net/amm.592-594.2460.

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This paper compares two different accident sequence methodologies for external flood probabilistic safety analysis (EFPSA). It is shown that the two methodologies lead to identical expression for CDF with an example. Using accident sequences developed for internal events PSA is recommended for detailed external flood probabilistic safety analysis of prototype fast breeder reactor.

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32

Morgenstern, Friedrich H. "Kalkar Nuclear Power Plant (SNR-300)—A Sodium-Cooled Fast Breeder Reactor Prototype." Nuclear Technology 78, no. 3 (1987): 231–44. http://dx.doi.org/10.13182/nt87-a15989.

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Bhaduri, A. K., R. Indira, S. K. Albert, B. P. S. Rao, S. C. Jain, and S. Asokkumar. "Selection of hardfacing material for components of the Indian Prototype Fast Breeder Reactor." Journal of Nuclear Materials 334, no. 2-3 (2004): 109–14. http://dx.doi.org/10.1016/j.jnucmat.2004.05.005.

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Sharma, Pramod Kumar, V. Bhuvana, and M. Ramakrishnan. "Reliability analysis of Diesel Generator power supply system of Prototype Fast Breeder Reactor." Nuclear Engineering and Design 310 (December 2016): 192–204. http://dx.doi.org/10.1016/j.nucengdes.2016.10.013.

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35

Jayalal, M. L., L. Satish Kumar, R. Jehadeesan, et al. "Steam condenser optimization using Real-parameter Genetic Algorithm for Prototype Fast Breeder Reactor." Nuclear Engineering and Design 241, no. 10 (2011): 4136–42. http://dx.doi.org/10.1016/j.nucengdes.2011.08.023.

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36

Chellapandi, P., G. S. Srinivasan, and S. C. Chetal. "Primary containment capacity of Prototype Fast Breeder Reactor against core disruptive accident loadings." Nuclear Engineering and Design 256 (March 2013): 178–87. http://dx.doi.org/10.1016/j.nucengdes.2012.12.014.

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37

Vinod, V., S. Krishnakumar, R. Punniyamoorthy, et al. "Experimental evaluation of safety grade decay heat removal in prototype fast breeder reactor." Nuclear Engineering and Design 265 (December 2013): 1057–65. http://dx.doi.org/10.1016/j.nucengdes.2013.08.072.

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38

Devan, K., A. Riyas, M. Alagan, and P. Mohanakrishnan. "A new physics design of control safety rods for prototype fast breeder reactor." Annals of Nuclear Energy 35, no. 8 (2008): 1484–91. http://dx.doi.org/10.1016/j.anucene.2008.01.013.

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39

Bhaduri, A. K., I. Gowrisankar, V. Seetharaman, S. Venkadesan, and P. Rodriguez. "Development of transition metal joint for steam generator circuit of prototype fast breeder reactor." Materials Science and Technology 4, no. 11 (1988): 1020–29. http://dx.doi.org/10.1179/mst.1988.4.11.1020.

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40

Chandran, R. Krishna, Indranil Banerjee, G. Padmakumar, and K. S. Reddy. "Investigation of Thermal Striping in Prototype Fast Breeder Reactor Using Ten-Jet Water Model." Heat Transfer Engineering 32, no. 5 (2011): 369–83. http://dx.doi.org/10.1080/01457632.2010.483857.

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41

Joshi, V. B., R. V. Kolekar, and H. K. Dravid. "Neutron measurements using the SSNTD technique during prototype fast breeder reactor bulk shielding experiments." Radiation Measurements 36, no. 1-6 (2003): 741–45. http://dx.doi.org/10.1016/s1350-4487(03)00238-5.

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Sinha, N. K., and Baldev Raj. "Choice of rotatable plug seals for prototype fast breeder reactor: Review of historical perspectives." Nuclear Engineering and Design 291 (September 2015): 109–32. http://dx.doi.org/10.1016/j.nucengdes.2015.05.017.

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Sharma, Prashant, K. Natesan, P. Selvaraj, V. Balasubramaniyan, and P. Chellapandi. "Dynamic modeling of steam water system of prototype fast breeder reactor using RELAP code." Annals of Nuclear Energy 68 (June 2014): 209–19. http://dx.doi.org/10.1016/j.anucene.2014.01.025.

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Suchitra, J. Y., and M. V. Ramana. "The costs of power: plutonium and the economics of India's prototype fast breeder reactor." International Journal of Global Energy Issues 35, no. 1 (2011): 1. http://dx.doi.org/10.1504/ijgei.2011.039982.

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Kumar, Ashwin, and M. V. Ramana. "Compromising Safety: Design Choices and Severe Accident Possibilities in India's Prototype Fast Breeder Reactor." Science & Global Security 16, no. 3 (2008): 87–114. http://dx.doi.org/10.1080/08929880802565115.

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Rangasamy, R. G., Prabhat Kumar, and N. Manoharan. "Quality Management- The Key to Inherent Safety and Reliability of Prototype fast Breeder Reactor." Biosciences Biotechnology Research Asia 11, SE (2014): 167–71. http://dx.doi.org/10.13005/bbra/1406.

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Mathew, M. D., K. A. Gopal, S. Murugan, B. K. Panigrahi, A. K. Bhaduri, and T. Jayakumar. "Development of IFAC-1 SS: An Advanced Austenitic Stainless Steel for Cladding and Wrapper Tube Applications in Sodium-Cooled Fast Reactors." Advanced Materials Research 794 (September 2013): 749–56. http://dx.doi.org/10.4028/www.scientific.net/amr.794.749.

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Abstract:

Fuel cycle cost of sodium cooled fast reactors (SFRs) is strongly dependent on the in-reactor performance of core structural materials, i.e., cladding and wrapper tube materials of the fuel subassembly, which are subjected to intense neutron irradiation during service, leading to unique materials problems like void swelling, irradiation creep and helium embrittlement. In order to increase the burnup of the fuel and thereby reduce the fuel cycle cost, it is necessary to employ materials which have high resistance to void swelling as well as better high temperature mechanical properties. The Indian fast reactor program began with the commissioning of the 40 MWtFast Breeder Test Reactor (FBTR). The core structural material of FBTR is 20% cold worked 316 austenitic stainless steel (SS). For the 5000 MetPrototype Fast Breeder Reactor (PFBR) which is in an advanced stage of construction at Kalpakkam, 20% cold-worked alloy D9 (14Cr-15Ni-Ti SS) has been selected as the cladding and wrapper tube material for the initial core. The target burnup of the fuel is 100 GWd/t. Advanced austenitic stainless steel and oxide dispersion strengthened steels are being developed for achieving fuel burnup higher than 100 GWd/t. An advanced alloy D9 containing controlled amounts of titanium, silicon and phosphorous has been developed. This alloy named as IFAC-1 (Indian Fast Reactor advanced Clad-1) SS is aimed at thermal creep properties comparable to that of alloy D9, and superior void swelling resistance upto a target burn-up of about 150 GWd/t. The nominal chemical composition of IFAC-1 SS is 14Cr-15Ni-.25Ti-.75Si-.04P. The chemical composition has been optimized after extensive evaluation of the tensile, creep and microstructural stability of fifteen laboratory heats with different amounts of titanium, silicon and phosphorous. Void swelling behavior was studied using ion irradiation. IFAC-1 SS contains higher levels of low melting eutectic phase forming elements such as phosphorous, and so is susceptible to solidification cracking. Extensive pulsed TIG welding trials have been carried out on IFAC-1 SS/316LN SS weld joints with varied weld parameters to find out the feasibility of obtaining solidification crack-free welds and the optimum welding parameters have been established. This paper gives an overview of the development of this advanced core structural material for SFRs.

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Ramana, M. V., and Ashwin K. Seshadri. "Negligence, capture, and dependence: safety regulation of the design of India’s Prototype Fast Breeder Reactor." Journal of Risk Research 18, no. 8 (2015): 1030–50. http://dx.doi.org/10.1080/13669877.2014.1003958.

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Suzuki, Tohru, Yoshiharu Tobita, Kenichi Kawada, et al. "A preliminary evaluation of unprotected loss-of-flow accident for a prototype fast-breeder reactor." Nuclear Engineering and Technology 47, no. 3 (2015): 240–52. http://dx.doi.org/10.1016/j.net.2015.03.001.

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Gupta, Suman, Gagan Gupta, and P. Chellapandi. "Establishing Scaling Criteria of Prototype Fast Breeder Reactor Intermediate Heat Exchanger through Finite Element Modeling." Procedia Engineering 86 (2014): 701–8. http://dx.doi.org/10.1016/j.proeng.2014.11.072.

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