Journal of Nuclear Science, Engineering and Technology (JONSAT)
In cooperation with the Iranian Nuclear Society

Thermal-hydraulics safety analysis of neutron flux trap channel of Tehran research reactor for fuel irradiation experiment

Document Type : Research Paper

Authors

S. Karbalai 1
M.H. Esteki 1
S.M. Mirvakili 2
ُS. Safaei Arshi 2

1 Department of Nuclear Engineering, Faculty of Physics, Isfahan University, P.O.BOX: 817467-3441, Isfahan – Iran

2 Reactor and Nuclear Safety Research School, Nuclear Science and Technology Research Institute, AEOI, P.O.BOX: 14399-51113, Tehran – Iran

https://doi.org/10.24200/nst.2024.1636
Abstract
In the research conducted on the potential use of the Tehran Research Reactor (TRR) for domestic fuel testing, a compact core configuration has been developed. This configuration includes a central neutron flux trap channel to achieve the desired linear heat rate for fuel testing. In this study, a safety analysis of loading the fuel irradiation capsule in the neutron flux trap channel is performed from a thermal-hydraulics perspective under steady-state conditions. The differences in coolant channels between the irradiation capsule and adjacent fuel plates cause a change in coolant flow distribution in the core after loading the fuel irradiation capsule in the neutron flux trap channel. The study investigates the effect of loading the fuel irradiation capsule on heat removal from the fuel plate adjacent to the neutron flux trap channel, which is the hottest fuel plate in the core. ANSYS Fluent software is used for the analysis, allowing for simultaneous simulation of rod-type fuels within the irradiation capsule and the fuel plates adjacent to the neutron flux trap channel. The simulation results show that the maximum clad temperature of the fuel plates adjacent to the irradiation capsule will be approximately 370K, which is lower than the saturation temperature under the reactor operating pressure and below the maximum permissible clad temperature to avoid corrosion. The maximum fuel temperature is 375.9K, which falls within permissible limits. These results indicate safe operation of the TRR core from a thermal-hydraulics perspective if the fuel irradiation capsule is loaded in the flux trap channel.

Highlights

  1. Safaei Arshi S, Khalafi H, Mirvakili S.M. Assessment of safety aspects of first rod-type fuel irradiation at Tehran research reactor, Part I: Neutronic analysis. Progress in Nuclear Energy. 2015;79:56-63.

 

  1. Safaei Arshi S, Khalafi H, Mirvakili S.M. Preliminary thermal-hydraulic safety analysis of Tehran research reactor during fuel irradiation experiment. Progress in Nuclear Energy. 2015;79:32-39.

 

  1. Safaei Arshi S, Mirvakili S.M, Khalafi H, Ezati A, Tajbakhsh A. Experimental validation of a modified RELAP5 model for transient analysis of Tehran research reactor mixed-core during fuel irradiation experiments. Progress in Nuclear Energy. 2017;100:11-21.

 

  1. Safaei Arshi S, Jozvaziri A, Mirvakili S.M, Keyvani M. A methodology to enhance thermal neutron flux in Tehran Research Reactor core for domestic fuel test purposes. Progress in Nuclear Energy. 2021a;136:103726.

 

  1. AEOI. Final Safety Analysis Report of Tehran Research Reactor. 2018.

 

  1. Safaei Arshi S, Mozafari M.A, Jozvaziri A, Mirvakili S.M. Investigation of safety aspects during steady state operation of Tehran research reactor fuel test loop. Progress in Nuclear Energy. 2021b;140:103895.

 

  1. ANSYS, Inc., Release 19.2. ANSYS Fluent User's Guide. 2018 August.

 

  1. Basile D, Beghi M, Chierici R, Salina E, Brega E. COBRA-EN, an Upgraded Version of the COBRA3CIMIT Code for Thermal-Hydraulic Transient Analysis of Light Water Reactor Fuel Assemblies and Cores. Report No. 1010/l. 1999.

 

  1. Abbate P. TERMIC V 4.1: A Program for the Calculus and Thermal-Hydraulic Design of Research Reactor Cores. INVAP. 2003.

 

  1. NUREG/CR-0497. MATPRO-Version 11, A handbook of materials properties for use in the analysis of light water reactor fuel rod behavior. Idaho National Engineering Laboratory, Department of Energy. 1979.

 

  1. IAEA. Thermophysical Properties of Materials for Nuclear Engineering: A Tutorial and Collection of Data. VIENNA. 2008.

Keywords

Neutron flux trap channel
Tehran research reactor
Irradiation capsule
Thermal-hydraulics safety
  1. Safaei Arshi S, Khalafi H, Mirvakili S.M. Assessment of safety aspects of first rod-type fuel irradiation at Tehran research reactor, Part I: Neutronic analysis. Progress in Nuclear Energy. 2015;79:56-63.

 

  1. Safaei Arshi S, Khalafi H, Mirvakili S.M. Preliminary thermal-hydraulic safety analysis of Tehran research reactor during fuel irradiation experiment. Progress in Nuclear Energy. 2015;79:32-39.

 

  1. Safaei Arshi S, Mirvakili S.M, Khalafi H, Ezati A, Tajbakhsh A. Experimental validation of a modified RELAP5 model for transient analysis of Tehran research reactor mixed-core during fuel irradiation experiments. Progress in Nuclear Energy. 2017;100:11-21.

 

  1. Safaei Arshi S, Jozvaziri A, Mirvakili S.M, Keyvani M. A methodology to enhance thermal neutron flux in Tehran Research Reactor core for domestic fuel test purposes. Progress in Nuclear Energy. 2021a;136:103726.

 

  1. AEOI. Final Safety Analysis Report of Tehran Research Reactor. 2018.

 

  1. Safaei Arshi S, Mozafari M.A, Jozvaziri A, Mirvakili S.M. Investigation of safety aspects during steady state operation of Tehran research reactor fuel test loop. Progress in Nuclear Energy. 2021b;140:103895.

 

  1. ANSYS, Inc., Release 19.2. ANSYS Fluent User's Guide. 2018 August.

 

  1. Basile D, Beghi M, Chierici R, Salina E, Brega E. COBRA-EN, an Upgraded Version of the COBRA3CIMIT Code for Thermal-Hydraulic Transient Analysis of Light Water Reactor Fuel Assemblies and Cores. Report No. 1010/l. 1999.

 

  1. Abbate P. TERMIC V 4.1: A Program for the Calculus and Thermal-Hydraulic Design of Research Reactor Cores. INVAP. 2003.

 

  1. NUREG/CR-0497. MATPRO-Version 11, A handbook of materials properties for use in the analysis of light water reactor fuel rod behavior. Idaho National Engineering Laboratory, Department of Energy. 1979.

 

  1. IAEA. Thermophysical Properties of Materials for Nuclear Engineering: A Tutorial and Collection of Data. VIENNA. 2008.

Home