Neutronic analysis of MSRE and its study for validation of ARCH code-Reference-Cited by-同舟云学术

Neutronic analysis of MSRE and its study for validation of ARCH code

Published:2017-08 Issue: Volume:320 Page:1-8
ISSN:0029-5493
Container-title:Nuclear Engineering and Design
language:en
Short-container-title:Nuclear Engineering and Design

Author:

Tadepalli Sai Chaitanya,Gupta Anurag,Umasankari K.

Publisher

Elsevier BV

Subject

Mechanical Engineering,Waste Management and Disposal,Safety, Risk, Reliability and Quality,General Materials Science,Nuclear Energy and Engineering,Nuclear and High Energy Physics

Reference26 articles.

1. URL www.polymtl.ca/merlin/libraries.htm.

2. Anurag, G., 2012. Arch: A 3d space-time analysis code in cartesian and hexagonal geometries. In: Proceedings of 19th National Symposium on Radiation Physics (NSRP-19), Mamallapuram, India.

3. Neutron multiplication studies in beryllium for fusion reactor blankets;Basu;Nucl. Sci. Eng.,1979

4. Benoist, P., et al., 1958. Critical and subcritical experiments on u-beo lattices. In: Proceedings of the Second United Nations International Conference on the Peaceful Uses of Atomic Energy, Geneva. vol. 12, p. 585.

5. A road map fot the realization of global-scale thorium breeding fuel cycle by single molten-fluoride flow;Furukawa;Energy Convers. Manage.,2008

Cited by 5 articles. 订阅此论文施引文献订阅此论文施引文献，注册后可以免费订阅5篇论文的施引文献，订阅后可以查看论文全部施引文献

1. Neutronics study of chloride fuel salts for experimental design of small Molten Salt Reactor;Nuclear and Particle Physics Proceedings;2023-11

2. Analysis of the Power and Temperature distribution in molten salt reactors with TRACE. Application to the MSRE;Annals of Nuclear Energy;2021-07

3. Development of a Python Module “SARRA” for Refuelling Analysis of MSR Using DRAGON Code;Proceedings of the 7th International Conference on Advances in Energy Research;2020-10-18

4. CFD Simulations of Molten Salt Reactor Experiment Core;Nuclear Science and Engineering;2019-07-08

5. Effects of fuel salt composition on fuel salt temperature coefficient (FSTC) for an under-moderated molten salt reactor (MSR);Nuclear Science and Techniques;2018-07-03