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

Steady-state analysis of neutronic and thermal-hydraulic for Bushehr nuclear reactor’s fuel assemblies using nodal expansion and single heated channel method

Document Type : Research Paper

Authors

D. Naghavi Dizaji
A. Kolali
N. Vosoughi

Department of Energy Engineering, Sharif University of Technology, P.O.BOX: 14565-1114, Tehran - Iran

https://doi.org/10.24200/nst.2020.548.1369
Abstract
For the neutronic and thermal-hydraulic analysis of nuclear reactor cores, it is necessary to develop nuclear computing software to calculate neutronic and thermal-hydrodynamic parameters for their safe operation. In this paper, S4HC software was developed for steady-state thermal-hydraulic core calculations using a single heated channel method. To analyze the Bushehr reactor core, after calculating neutron parameters by the nodal expansion method, a thermal-hydraulic analysis of fuel assemblies was performed using S4HC software. After thermal-hydraulic calculations for the fuel assemblies, including the hot fuel assembly, it was concluded that all the coolant thermal-hydraulic parameters are within their allowed ranges and the reactor has sufficient saturation margins.

Highlights

  1. Naghavi Dizaji D. Investigating the Propagation of Thermal-hydraulic Noise in PWRs in Two phases. MSc Thesis, Sharif University of Technology. Tehran, Iran. 2018 [In Persian]. Available at: https://ganj.irandoc.ac.ir//#/articles/6652df6a65abbfee32f6168bf4782a19.

 

  1. Kolali A, Naghavi Dizaji D, Vosoughi N. Development of the S3-HACNEM Simulator Program in order to Solving the Forward and Adjoint Neutron Diffusion Equation for Rectangular Geometry Reactor Cores. Journal of Nuclear Science and Technology. 2024;45(2):21-28 [In Persian]. https://doi.org/10.24200/nst.2023.469.1319.

 

  1. Naghavi Dizaji D, Vosoughi N. Thermal-hydraulic Investigation of Bushehr Nuclear Reactor in Two-Phase mode by Single Heating Channel method. 25th Iran Nuclear Conference, Bushehr, Iran. 2019 [In Persian].

 

  1. Kolali A, Naghavi Dizaji D, Vosoughi N. Development of the SH3-ACNEM Simulator Program in order to Solving the Forward and Adjoint neutron Diffusion Equation for Hexagonal Geometry Reactor Cores. Journal of Nuclear Science and Technology. 2024;44(1):103-110 [In Persian]. https://doi.org/10.24200/nst.2023.436.1298.

 

  1. Todreas N.E, Kazimi M.S. Nuclear systems I&II. Taylor & Francis. 1990.

 

  1. BrkiÄ D. Determining friction factors in turbulent pipe flow. 2012.

 

  1. El-Wakil M.M. Nuclear Heat Transport. 1971.

 

  1. AEOI, Final Safety Analysis Report (FSAR) of BNPP-1. 2007.

 

  1. Hosseini S.A, Vosoughi N. On a various noise source reconstruction algorithms in VVER-1000 reactor core. Nuclear Engineering and Design. 2013;261:132-143.

Keywords

Neutronic and thermal-hydraulic analysis
Single heated channel
High-order nodal expansion method
S4HC
Bushehr nuclear reactor
  1. Naghavi Dizaji D. Investigating the Propagation of Thermal-hydraulic Noise in PWRs in Two phases. MSc Thesis, Sharif University of Technology. Tehran, Iran. 2018 [In Persian]. Available at: https://ganj.irandoc.ac.ir//#/articles/6652df6a65abbfee32f6168bf4782a19.

 

  1. Kolali A, Naghavi Dizaji D, Vosoughi N. Development of the S3-HACNEM Simulator Program in order to Solving the Forward and Adjoint Neutron Diffusion Equation for Rectangular Geometry Reactor Cores. Journal of Nuclear Science and Technology. 2024;45(2):21-28 [In Persian]. https://doi.org/10.24200/nst.2023.469.1319.

 

  1. Naghavi Dizaji D, Vosoughi N. Thermal-hydraulic Investigation of Bushehr Nuclear Reactor in Two-Phase mode by Single Heating Channel method. 25th Iran Nuclear Conference, Bushehr, Iran. 2019 [In Persian].

 

  1. Kolali A, Naghavi Dizaji D, Vosoughi N. Development of the SH3-ACNEM Simulator Program in order to Solving the Forward and Adjoint neutron Diffusion Equation for Hexagonal Geometry Reactor Cores. Journal of Nuclear Science and Technology. 2024;44(1):103-110 [In Persian]. https://doi.org/10.24200/nst.2023.436.1298.

 

  1. Todreas N.E, Kazimi M.S. Nuclear systems I&II. Taylor & Francis. 1990.

 

  1. BrkiÄ D. Determining friction factors in turbulent pipe flow. 2012.

 

  1. El-Wakil M.M. Nuclear Heat Transport. 1971.

 

  1. AEOI, Final Safety Analysis Report (FSAR) of BNPP-1. 2007.

 

  1. Hosseini S.A, Vosoughi N. On a various noise source reconstruction algorithms in VVER-1000 reactor core. Nuclear Engineering and Design. 2013;261:132-143.

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