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Analysis of Kobayashi benchmark with indigenous Monte Carlo neutron transport code PATMOC

Physics open, 2023-12, Vol.17, p.100179, Article 100179

2023

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Evaluation of DNBR with neutronics calculation in LWR systems

Nuclear energy and technology, 2023-06, Vol.9 (2), p.77-83

2023

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Overview of the physics challenges in design of new energy systems

Nuclear and particle physics proceedings, 2023-11, Vol.339-340, p.38-42

Kannan, Umasankari

2023

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Simulation of random medium of fuel pebble with liquid salt coolant using PebMC code

Physics open, 2023-12, Vol.17, p.100190, Article 100190

2023

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Measurement of leakage neutron flux from reactor tank surface of AHWR-critical facility

Physics open, 2024-02, Vol.18, p.100193, Article 100193

2024

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Analysis for the use of thorium based fuel in LWRs

Annals of nuclear energy, 2022-09, Vol.174, p.109162, Article 109162

2022

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Operational characteristics of LWR core fuelled with thorium-based fuel

Annals of nuclear energy, 2023-08, Vol.188, p.109801, Article 109801

2023

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Fine mesh pin-by-pin analysis of IPWR equilibrium core

Nuclear and particle physics proceedings, 2023-12, Vol.341, p.6-11

2023

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Benchmarking exercise of indigenous Monte Carlo neutron transport code PATMOC for hexagonal full core reactor geometry

Nuclear and particle physics proceedings, 2023-11, Vol.339-340, p.62-65

2023

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Development of computer code ADWITA and data library for the solution of transmutation chain equations and application to the analysis of nuclear fuel cycles

Annals of nuclear energy, 2021-12, Vol.164, p.108619, Article 108619

2021

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Perspectives on nuclear data for advanced reactor design and analysis

Life cycle reliability and safety engineering, 2020-06, Vol.9 (2), p.135-146

Kannan, Umasankari

2020

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Transient analysis of AHWR core using implicit method of NDIFF3D

Nuclear and particle physics proceedings, 2024-02, Vol.342, p.10-13

2024

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Fuzzy Logics as an Integral Part of Evolutionary Algorithms

Nuclear science and engineering, 2019-10, Vol.193 (10), p.1160-1171

2019

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Design Considerations for Utilising Thorium in Advanced Energy Systems

Transactions of the Indian National Academy of Engineering (Online), 2024-06, Vol.9 (2), p.405-413

2024

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Study of VVER-1000 OECD LEU and MOX Computational Benchmark with VISWAM Code System

Nuclear energy and technology, 2016-12, Vol.2 (4), p.312-334

2016

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Investigation on the Use of the Monte Carlo Iterative k-Source Scheme for the Study of Neutron Subcritical Multiplication

Nuclear science and engineering, 2022-08, Vol.196 (8), p.927-942

2022

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Neutronic and burn up analysis of VVER 1000 benchmark using EXCEL and TRIHEX-FA code system

Nuclear engineering and design, 2023-12, Vol.414, p.112601, Article 112601

2023

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Comparative studies of iterative methods for solving the “optimally diffusive” coarse mesh finite difference accelerated transport equation

Annals of nuclear energy, 2021-07, Vol.157, p.108211, Article 108211

2021

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Estimation of safety parameters of equilibrium cycle of Indian Pressurized Water Reactor (IPWR)

Life cycle reliability and safety engineering, 2020-06, Vol.9 (2), p.129-134

2020

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Analysis of Integral Experiment of the AHWR Critical Facility with Diffusion-Based Monte Carlo Method

Nuclear science and engineering, 2023-04, Vol.197 (4), p.703-710

2023

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Reactivity effect of various radioisotopes in pressure tube type heavy water reactor using the computer code IGDC

Life cycle reliability and safety engineering, 2021-09, Vol.10 (3), p.217-233

2021

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Coupled neutronics–thermal-hydraulics analyses of unprotected transients in LBE-cooled Compact High-Temperature Reactor

Life cycle reliability and safety engineering, 2023-09, Vol.12 (3), p.259-268

2023

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DIAMOND – A Method of Characteristics neutron transport code using unstructured meshing

Annals of nuclear energy, 2020-03, Vol.137, p.107086, Article 107086

2020

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Convergence study of CMFD based acceleration schemes for multi-group transport calculations with fission source using fourier analysis

Annals of nuclear energy, 2021-09, Vol.160, p.108314, Article 108314

2021

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Initial core loading pattern optimization studies using estimation of distribution algorithm for VVER type cores

Nuclear engineering and design, 2023-10, Vol.412, p.112458, Article 112458

2023

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Analysis of Kobayashi benchmark with indigenous Monte Carlo neutron transport code PATMOC

Evaluation of DNBR with neutronics calculation in LWR systems

Overview of the physics challenges in design of new energy systems

Simulation of random medium of fuel pebble with liquid salt coolant using PebMC code

Measurement of leakage neutron flux from reactor tank surface of AHWR-critical facility

Analysis for the use of thorium based fuel in LWRs

Operational characteristics of LWR core fuelled with thorium-based fuel

Fine mesh pin-by-pin analysis of IPWR equilibrium core

Benchmarking exercise of indigenous Monte Carlo neutron transport code PATMOC for hexagonal full core reactor geometry

Development of computer code ADWITA and data library for the solution of transmutation chain equations and application to the analysis of nuclear fuel cycles

Perspectives on nuclear data for advanced reactor design and analysis

Transient analysis of AHWR core using implicit method of NDIFF3D

Fuzzy Logics as an Integral Part of Evolutionary Algorithms

Design Considerations for Utilising Thorium in Advanced Energy Systems

Study of VVER-1000 OECD LEU and MOX Computational Benchmark with VISWAM Code System

Investigation on the Use of the Monte Carlo Iterative k-Source Scheme for the Study of Neutron Subcritical Multiplication

Neutronic and burn up analysis of VVER 1000 benchmark using EXCEL and TRIHEX-FA code system

Comparative studies of iterative methods for solving the “optimally diffusive” coarse mesh finite difference accelerated transport equation

Estimation of safety parameters of equilibrium cycle of Indian Pressurized Water Reactor (IPWR)

Analysis of Integral Experiment of the AHWR Critical Facility with Diffusion-Based Monte Carlo Method

Reactivity effect of various radioisotopes in pressure tube type heavy water reactor using the computer code IGDC

Coupled neutronics–thermal-hydraulics analyses of unprotected transients in LBE-cooled Compact High-Temperature Reactor

DIAMOND – A Method of Characteristics neutron transport code using unstructured meshing

Convergence study of CMFD based acceleration schemes for multi-group transport calculations with fission source using fourier analysis

Initial core loading pattern optimization studies using estimation of distribution algorithm for VVER type cores

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