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28 results on '"Athmalingam, S."'

2. A Study on Rayleigh Plateau Instability in Slender Jets of Nuclear Materials

Author

Chandra Kumar, M., Jasmin Sudha, A., Subramanian, V., Venkatesan, R., Athmalingam, S., Venkatraman, B., Chaari, Fakher, Series Editor, Gherardini, Francesco, Series Editor, Ivanov, Vitalii, Series Editor, Cavas-Martínez, Francisco, Editorial Board Member, di Mare, Francesca, Editorial Board Member, Haddar, Mohamed, Editorial Board Member, Kwon, Young W., Editorial Board Member, Trojanowska, Justyna, Editorial Board Member, Xu, Jinyang, Editorial Board Member, Sivaram, N. M., editor, Sankaranarayanasamy, K., editor, and Davim, J. Paulo, editor

Published
2023
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5. Investigation of CT Linearity and Establishing the Density Calibration Curve for the Gamma Ray CT System

Author

Ramar, R., Menaka, M., Subramanian, V., Venkatesan, R., Athmalingam, S., Venkatraman, B., Cavas-Martínez, Francisco, Series Editor, Chaari, Fakher, Series Editor, di Mare, Francesca, Series Editor, Gherardini, Francesco, Series Editor, Haddar, Mohamed, Series Editor, Ivanov, Vitalii, Series Editor, Kwon, Young W., Series Editor, Trojanowska, Justyna, Series Editor, Mandayam, Shyamsunder, editor, and Sagar, Sarmishtha Palit, editor

Published
2022
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8. Theoretical simulation on evolution of suspended sodium combustion aerosols characteristics in a closed chamber

Author

Sujatha Pavan Narayanam, Amit Kumar, Usha Pujala, Subramanian V., Srinivas C.V., Venkatesan R., Athmalingam S., and Venkatraman B.

Subjects
Sodium aerosols, Aerosol modeling, HAARM code, Aerosol size growth, Method of moments, Nuclear engineering. Atomic power, TK9001-9401
Abstract

In the unlikely event of core disruptive accident in sodium cooled fast reactors, the reactor containment building would be bottled up with sodium and fission product aerosols. The behavior of these aerosols is crucial to estimate the in-containment source term as a part of nuclear reactor safety analysis. In this work, the evolution of sodium aerosol characteristics (mass concentration and size) is simulated using HAARM-S code. The code is based on the method of moments to solve the integro-differential equation. The code is updated to FORTRAN-77 and run in Microsoft FORTRAN PowerStation 4.0 (on Desktop). The sodium aerosol characteristics simulated by HAARM-S code are compared with the measured values at Aerosol Test Facility. The maximum deviation between measured and simulated mass concentrations is 30% at initial period (up to 60 min) and around 50% in the later period. In addition, the influence of humidity on aerosol size growth for two different aerosol mass concentrations is studied. The measured and simulated growth factors of aerosol size (ratio of saturated size to initial size) are found to be matched at reasonable extent. Since sodium is highly reactive with atmospheric constituents, the aerosol growth factor depends on the hygroscopic growth, chemical transformation and density variations besides coagulation. Further, there is a scope for the improvement of the code to estimate the aerosol dynamics in confined environment.

Published
2022
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20. Effect of Internal Decay Power and Surface Cooling on Nuclear Fuel Droplet Solidification During MFCI in a SFR—A Numerical Study

Author

Kumar, M. Chandra, Sudha, A. Jasmin, Subramanian, V., Athmalingam, S., and Venkatraman, B.

Abstract

AbstractMelting of the nuclear core is one of the severe accident scenarios in a Sodium-cooled Fast Reactor (SFR). During such an event, molten corium may come into contact with the coolant sodium. This interaction of the molten fuel and the coolant is commonly termed molten fuel–coolant interaction (MFCI) in the nuclear industry. In this study, a numerical analysis is carried out to study the solidification of a molten fuel droplet in the liquid sodium pool. In the first part of the study, the effect of constant internal heat generation on the solidification of the droplet is evaluated with convective heat dissipation prescribed at the droplet surface. The internal heat generation (decay power) and the heat transfer coefficient are varied as parameters, and the time required for complete solidification of the molten droplet is obtained. Based on the results, the freezing of the droplet is categorized into three regimes: conduction limited, transition, and internal heat generation dominated regimes. It is observed that the solidification process of nuclear fuel droplets generated during MFCI is not influenced by internal heat generation and lies in a conduction-limited regime for decay power level prevailing in a medium-sized SFR. Hence, in the next part of the study, the numerical analysis is carried out by incorporating the time-dependent decay power and the temperature-dependent heat transfer coefficient in the computational model by developing user-defined subroutines depicting a realistic scenario of an accident. The results of the analysis show that because of the high subcooling of sodium, film boiling is ruled out; nucleate boiling with a maximum heat transfer rate occurs briefly. The heat transfer coefficient then declines as the interface temperature between the droplet and the sodium decreases rapidly until the natural convective regime is reached. A parametric study on the droplet diameter is also carried out by varying the diameter from 0.5 to 10 mm, spanning the typical particle size spectrum expected during MFCI.

Published
2023
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22. The IAEA Coordinated Research Project on Sodium Properties and Safe Operation of Experimental Facilities in Support of the Development and Deployment of Sodium-cooled Fast Reactors (NAPRO)

Author

Azpitarte, O., Japas, M., Chocrón, M., Vázquez, E., Villanueva, A., Long, B., Xu, C., Xie, C., Latgé, C., Gerschenfeld, A., Anderhuber, M., Chacon, L., Matteo, L., Cavaro, M., Féron, D., Trabuc, P., Bubelis, E., Perez-Martin, S., Stieglitz, R., Eckert, S., Krauter, N., Selvaraj, P., Raghvachary, S., Athmalingam, S., Chellapandi, P., Ohira, H., Lee, J., Park, S., Lee, Y., Hong, J., Jayaraju, S., Roelofs, F., Marinenko, E., Zagorulko, Y., Passerini, S., Gerardi, C., Grandy, C., Monti, S., and Kriventsev, V.

Subjects
sodium properties, sodium safety, sodium facilities design
Abstract

The International Atomic Energy Agency (IAEA) recently established a Coordinated Research Project (CRP) on “Sodium Properties and Safe Operation of Experimental Facilities in Support of the Development and Deployment of Sodium-cooled Fast Reactors - NAPRO”, to be carried out in the period 2013 – 2017. Eleven institutions from ten Member States participate in this CRP. The complete scope of this CRP is covered by three work packages. A specific work package (WP1), under the coordination of ANL (USA), is focused on the compilation and expert assessment of data sets of Na physical and chemical properties, as well as correlations for pressure drops and heat transfer in Na facilities. A second work package (WP2), under the coordination of IPPE (Russian Federation), addresses the compilation, evaluation and development of best practices and guidelines for the design, operation and maintenance of Na facilities. Finally, Work Package 3 (WP3), coordinated by CEA (France), concentrates on the compilation and development of guidelines and rules for the safe operation of Na facilities, including the prevention, detection and mitigation of Na leaks and fires. Finally, this work presents some examples of the principal results obtained so far by WP 1.

Published
2017

23. IAEA NAPRO Coordinated Research Project : Heat Transfer and Pressure Drop Correlations for Sodium Cooled Systems

Author

Bubelis, E., Perez-Martin, S., Passerini, S., Gerardi, C., Grandy, C., Jayaraju, S., Ferry Roelofs, Latge, C., Gerschenfeld, A., Anderhuber, M., Cachon, L., Matteo, L., Athmalingam, S., Selvaraj, P., Monti, S., Institute of Biomedical Engineering [Karlsruhe], Karlsruhe Institute of Technology (KIT), Argonne National Laboratory [Lemont] (ANL), Nuclear Research and Consultancy Group, CEA-Direction de l'Energie Nucléaire (CEA-DEN), Commissariat à l'énergie atomique et aux énergies alternatives (CEA), CEA Paris Saclay, Nuclear and Radioanalytical Chemistry, IGCAR, Kalpakkam-603102, Tamilnadu, International Atomic Energy Agency [Vienna] (IAEA), The authors wish to thank the International Atomic Energy Agency for the support, guidance and leadership on the NAPRO Coordinated Research Project., ANDERHUBER, Marine, and CEA-Direction des Energies (ex-Direction de l'Energie Nucléaire) (CEA-DES (ex-DEN))

Subjects
[PHYS.NUCL] Physics [physics]/Nuclear Theory [nucl-th], SFR, [PHYS.NUCL]Physics [physics]/Nuclear Theory [nucl-th], [PHYS.NEXP] Physics [physics]/Nuclear Experiment [nucl-ex], correlations, Na properties, IAEA, [PHYS.NEXP]Physics [physics]/Nuclear Experiment [nucl-ex]
Abstract

International audience; In 2013 the International Atomic Energy Agency (IAEA) established a Coordinated Research Project (CRP) on “Sodium properties and safe operation of experimental facilities in support of the development and deployment of Sodium Cooled Fast Reactors”, the socalled CRP-NAPRO project to be carried out in the time period of 2013–2017. This activity has the aim toestablish a common database of sodium properties and related correlations, as well as other important sodium technology related issues, thus contributing to the enhanced safety of future sodium cooled systems. The CRP work package WP 1.2, under the leadership of Karlsruhe Institute of Technology (KIT), is focused on the collection and assessment of heat transfer and pressure drop (friction factor) correlations for sodium cooled systems. In the end result, this activity will lead to arecommendation of qualified correlations for conditions occurring in sodium systems, such as forced convection and natural convection for circular tubes, rod bundles, wire-wrapped tubes, etc., which will be published in the form of separate chapter(s) of a general handbook. This work is carried out by five participating organizations from IAEA Member States through the review and evaluation of the existing correlations and the development of recommendations for experts working on sodium cooled systems. The implemented methodology for WP 1.2 is described, as well as the heat transfer and pressure drop (friction factor) correlations included in WP 1.2 and their classification. Major findings to date related to WP 1.2 are presented in this paper. The last section of the paper also includes preliminary conclusions from WP 1.2, as well as a list of the main correlations used by the participating organizations when simulating sodium cooled fast reactors and other sodium cooled systems.

Published
2016

26. Reliability analysis of safety grade decay heat removal system of Indian prototype fast breeder reactor

Author

Arul, A.J, Chandran, Senthil Kumar, Athmalingam, S., Singh, O.P., Suryaprakasa Rao, K., Arul, A.J, Chandran, Senthil Kumar, Athmalingam, S., Singh, O.P., and Suryaprakasa Rao, K.

Abstract

The 500MW Indian pool type Prototype Fast Breeder Reactor (PFBR), is provided with two independent and diverse Decay Heat Removal (DHR) systems viz., Operating Grade Decay Heat Removal System (OGDHRS) and Safety Grade Decay Heat Removal System (SGDHRS). OGDHRS utilizes the secondary sodium loops and Steam–Water System with special decay heat removal condensers for DHR function. The unreliability of this system is of the order of 0.1–0.01. The safety requirements of the present generation of fast reactors are very high, and specifically for DHR function the failure frequency should be less than 1E-7/ry. Therefore, a passive SGDHR system using four completely independent thermo-siphon loops in natural convection mode is provided to ensure adequate core cooling for all Design Basis Events. The very high reliability requirement for DHR function is achieved mainly with the help of SGDHRS. This paper presents the reliability analysis of SGDHR system. Analysis is performed by Fault Tree method using "CRAFT" software developed at Indira Gandhi Centre for Atomic Research. This software has special features for compact representation and CCF analysis of high redundancy safety systems encountered in nuclear reactors. Common Cause Failures (CCF) are evaluated by beta-factor method. The reliability target for SGDHRS arrived from DHR reliability requirement and the ultimate number of demands per year (7/y) on SGDHRS is that the failure frequency should be <=1.4E-8/de. Since it is found from the analysis that the unreliability of SGDHRS with identical loops is 5.2E-6/de and dominated by leak rates of components like AHX, DHX and sodium dump and isolation valves, options with diversity measures in important components were studied. The failure probability of SGDHRS for a design consisting of 2 types of diverse loops (Diverse AHX, DHX and sodium dump and isolation valves) is 2.1E-8/de, which practically meets the reliability requirement.

Published
2006
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28. Experience with Mechanical Seal of Prototype Fast Breeder Reactor’s Secondary Sodium Pump during Testing

Author

Manda, Bhagwana Ram, Sreedharan, K.V., Athmalingam, S., Balasubramaniyan, V., and Chellapandi, P.

Abstract

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.

Published
2014
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