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

Author: Grafiati

Published: 4 June 2021

Last updated: 25 July 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 o

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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 stre

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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 dishe

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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

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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

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

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7

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 constructi

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8

Sankar, Bindu, K. Vijaya kumar, j. Rajan, and s. Rajeswari. "DESIGN AND DEVELOPMENT OF HMI IN SIMULATOR FOR FRESH FUEL HANDLING SYSTEM." International Journal of Engineering Applied Sciences and Technology 8, no. 1 (2023): 162–68. http://dx.doi.org/10.33564/ijeast.2023.v08i01.030.

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KALBRSIM operator training simulator is used for training Prototype Fast Breeder Reactor (PFBR) nuclear power plant operators. In simulator, using virtual panel HMI which is an exact replica of actual hardware panels, all the essential fresh fuel operations can be done. Visualization power is also provided by 3D models which aid in quicker understanding of the complex process involved in Fresh Fuel Handling System (FFHS). Operation using Virtual Panel is an excellent tool for system developers to test the logic and process models of FFHS in KALBRSIM before integration with hardware panels. Als

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9

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 st

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10

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 Pl

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11

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 elemen

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12

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

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13

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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14

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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15

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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16

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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17

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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18

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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19

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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20

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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21

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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22

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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23

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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24

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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25

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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26

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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27

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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28

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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29

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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30

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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31

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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32

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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33

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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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 Ind

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34

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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35

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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36

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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37

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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38

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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39

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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40

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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41

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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42

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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43

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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44

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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45

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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46

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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47

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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48

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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49

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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50

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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