scholarly journals Simulation of pressurized water reactor to find the best control solution

2021 ◽  
Vol 7 (2) ◽  
pp. 126-135
Author(s):  
Taia Petik ◽  
◽  
Viktoriia Vataman ◽  
Konstantin Beglov
Keyword(s):  
Nuclear Reactor ◽  
Reactor Core ◽  
Distributed Model ◽  
Thermal Processes ◽  
Pressurized Water Reactor ◽  
Physical Processes ◽  
Control Solution ◽  
Pressurized Water ◽  
Water Reactor ◽  
Boric Acid Concentration

Since it is impossible to simultaneously control all technological and thermal parameters of the nuclear reactor, a vertically distributed model of a power unit was developed, which allows determining the axial offset and analyzing the behavior of thermal and neutron-physical processes of individual zones in core during boron control and movement of rods, as well as regulating the power of the nuclear reactor. For ten vertically distributed zones, neutrons and thermal processes in the reactor core were analyzed, as was the change of the axial offset during the discharge of the reactor load under the influence of the movement of the rods and the increase in the boric acid concentration. The purpose of this publication is to develop a vertically distributed model of the facility and use information technology to find the best solution for the control of a pressurized water reactor.

CFD Analysis of the Coolant Mixing within the Upper Plenum of a Pressurized Water Reactor (PWR)

2013 ◽  
Vol 444-445 ◽  
pp. 411-415 ◽  
Author(s):  
Fu Cheng Zhang ◽  
Shen Gen Tan ◽  
Xun Hao Zheng ◽  
Jun Chen
Keyword(s):  
Temperature Distribution ◽  
Fluid Dynamic ◽  
Characteristic Temperature ◽  
Reactor Core ◽  
Flow Distribution ◽  
Sensitivity Study ◽  
Cfd Model ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor

In this study, a Computational Fluid Dynamic (CFD) model is established to obtain the 3-D flow characteristic, temperature distribution of the pressurized water reactor (PWR) upper plenum and hot-legs. In the CFD model, the flow domain includes the upper plenum, the 61 control rod guide tubes, the 40 support columns, the three hot-legs. The inlet boundary located at the exit of the reactor core and the outlet boundary is set at the hot-leg pipes several meters away from upper plenum. The temperature and flow distribution at the inlet boundary are given by sub-channel codes. The computational mesh used in the present work is polyhedron element and a mesh sensitivity study is performed. The RANS equations for incompressible flow is solved with a Realizable k-ε turbulence model using the commercial CFD code STAR-CCM+. The analysis results show that the flow field of the upper plenum is very complex and the temperature distribution at inlet boundary have significant impact to the coolant mixing in the upper plenum as well as the hot-legs. The detailed coolant mixing patterns are important references to design the reactor core fuel management and the internal structure in upper plenum.

Depletion Calculation for a Nodal Reactor Physics Code

2010 ◽  
Author(s):  
Antonio Carlos Marques Alvim ◽  
Fernando Carvalho da Silva ◽  
Aquilino Senra Martinez
Keyword(s):  
Diffusion Equation ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Expansion Method ◽  
Design System ◽  
Pressurized Water Reactor ◽  
Reactor Physics ◽  
Pressurized Water ◽  
Nodal Method ◽  
Orthogonal Polynomial Expansion

This paper deals with an alternative numerical method for calculating depletion and production chains of the main isotopes found in a pressurized water reactor. It is based on the use of the exponentiation procedure coupled to orthogonal polynomial expansion to compute the transition matrix associated with the solution of the differential equations describing isotope concentrations in the nuclear reactor. Actually, the method was implemented in an automated nuclear reactor core design system that uses a quick and accurate 3D nodal method, the Nodal Expansion Method (NEM), aiming at solving the diffusion equation describing the spatial neutron distribution in the reactor. This computational system, besides solving the diffusion equation, also solves the depletion equations governing the gradual changes in material compositions of the core due to fuel depletion. The depletion calculation is the most time-consuming aspect of the nuclear reactor design code, and has to be done in a very precise way in order to obtain a correct evaluation of the economic performance of the nuclear reactor. In this sense, the proposed method was applied to estimate the critical boron concentration at the end of the cycle. Results were compared to measured values and confirm the effectiveness of the method for practical purposes.

State space modeling of reactor core in a pressurized water reactor

2014 ◽  
Author(s):  
A. Ashaari ◽  
T. Ahmad ◽  
Mustaffa Shamsuddin ◽  
Wan Munirah W. M ◽  
M. Adib Abdullah
Keyword(s):  
State Space ◽  
Reactor Core ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Space Modeling ◽  
State Space Modeling

Research on the Loading Pattern of Reactor Core With MOX Fuel in the First Cycle

2012 ◽  
Author(s):  
Tianqi Zhang ◽  
Shihe Yu ◽  
Xinrong Cao
Keyword(s):  
Cross Section ◽  
Reactor Core ◽  
Fuel Utilization ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Loading Pattern ◽  
Mox Fuel ◽  
Initial Cycle ◽  
Calculation Code

In order to research the performance of Pressurized Water Reactor (PWR) with 1/3 MOX fuel in the initial cycle, this paper serves Qinshan II reactor core as the reference core to design suitable MOX assemblies and study relevant core properties. The analyses documented within use assembly cross section calculation code CASMO-4 and core calculation code SIMULATE-3 studied by Studsvik. The purpose of this paper is to demonstrate that the Qinshan II reactor is capable of complying with the requirement for MOX fuel utilization without significant changes to the design of the plant. Several impacts on key physics parameters and safety analysis assumptions, introduced by MOX, are discussing in the paper.

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