scholarly journals Development of a Coupled Code for Steady-State Analysis of the Graphite-Moderated Channel Type Molten Salt Reactor

2018 ◽  
Vol 2018 ◽  
pp. 1-10 ◽  
Author(s):  
Long He ◽  
Cheng-Gang Yu ◽  
Wei Guo ◽  
Ye Dai ◽  
Hai-Ling Wang ◽  
...  
Keyword(s):  
Steady State ◽  
Density Distribution ◽  
Molten Salt ◽  
Power Distribution ◽  
Three Dimensional ◽  
Flow Distribution ◽  
Hydraulic Model ◽  
Molten Salt Reactor ◽  
Fuel Density ◽  
Channel Type

The molten salt reactor (MSR) is one of the six advanced reactor concepts selected by Generation IV International Forum (GIF) because of its inherent safety and the promising capabilities of TRU transmutation and Th-U breeding. In this study, a three-dimensional thermal-hydraulic model (3DTH) is developed for evaluating the steady-state performance of the graphite-moderated channel type MSR. The coupled code is developed by exchanging the power distribution, temperature, and fuel density distribution between SCALE and 3DTH. Firstly, the thermal-hydraulic model of the coupled code is validated by RELAP5 code. Then, the mass flow distribution, temperature field, keff, and power density distribution for a conceptual design of the 2MWt experimental molten salt reactor are calculated and analyzed by the coupled code under both normal operating situation and the central fuel assembly partly blocked situation. The simulated results are conductive to facilitate the understanding of the steady behavior of the graphite-moderated channel type MSR.

Steady Thermal Hydraulics Analysis for a Graphite-Moderated Channel Type Molten Salt Reactor

2008 ◽  
Author(s):  
Dalin Zhang ◽  
Suizheng Qiu
Keyword(s):  
Molten Salt ◽  
Flow Distribution ◽  
Axial Direction ◽  
Theoretical Models ◽  
Outer Wall ◽  
Molten Salt Reactor ◽  
Thermal Hydraulics ◽  
The Core ◽  
Fuel Salt ◽  
Channel Type

The Molten Salt Reactor (MSR) is one of the six GENIV systems capable of breading and burning. In this paper, a graphite-moderated channel type MSR was selected for conceptual research. For this MSR, a ternary system of 0.15LiF-0.58NaF-0.27BeF2 was proposed as the reactor fuel solvent, coolant and also moderator simultaneously with ca.1 mol% UF4 dissolving in it, which circulates through the whole primary loop accompanying fission reaction only in the core. 169 hexagonal graphite elements, each with a central fuel channel, are arranged in the core symmetrically by 30° angles. The theoretical models of the thermal hydraulics under steady condition are conducted in one-twelfth of the core and calculated by the numerical method. The DRAGON code is adopted to calculate the axial and radial power factors. The flow and heat transfer models in the fuel salt and graphite are founded basing on the fundamental mass, momentum and energy equations. The calculated results show the detailed mass flow distribution in the core; and the temperature of the fuel salt, inner and outer wall in the calculated elements along the axial direction are also obtained.

Coupled Neutronics/Thermal-Hydraulics for Analysis of Molten Salt Reactor

2013 ◽  
Author(s):  
Zhangpeng Guo ◽  
Yao Xiao ◽  
Jianjun Zhou ◽  
Dalin Zhang ◽  
Khurrum Saleem Chaudri ◽  
...  
Keyword(s):  
Molten Salt ◽  
Power Distribution ◽  
Flow Distribution ◽  
Iteration Process ◽  
Analytic Solutions ◽  
Molten Salt Reactor ◽  
Thermal Hydraulics ◽  
Multiple Channel ◽  
Channel Analysis ◽  
Reactor Types

The Generation IV international Forum (GIF) selected molten salt reactor (MSR) among six advanced reactor types. It is characterized by a liquid circulating fuel that also serves as coolant. In this study, a multiple-channel analysis code (MAC) is developed and it is coupled with MCNP4c to analyze the neutronics/thermal-hydraulics behavior of Molten Salt Reactor Experiment (MSRE). The MAC calculates thermal-hydraulic parameters, such as temperature distribution, flow distribution and pressure drop. MCNP4c performs the analysis of effective multiplication factor, neutron flux and power distribution. A linkage code is developed to exchange data between MAC and MCNP to implement coupling iteration process until the power convergence is achieved. The coupling calculation can achieve converged solution after a few iterations. The results are in reasonable agreement with the analytic solutions from the ORNL. This work was helpful for further design analysis and operation of MSR.

A Small PWR-Core Physical Calculation Based on PWR-Core Analysis Code CORAL

2021 ◽  
Author(s):  
Wen Yang ◽  
Lun Zhou ◽  
Junrong Qiu ◽  
Yun Tai
Keyword(s):  
Density Distribution ◽  
Power Distribution ◽  
Channel Model ◽  
Three Dimensional ◽  
Least Square ◽  
Object Oriented Programming ◽  
Core Analysis ◽  
Fuel Temperature ◽  
Fuel Assemblies ◽  
Calculation Results

Abstract Three dimensional PWR-core analysis code CORAL is developed by Wuhan Second Ship Design and Research Institute. This code provides basic functions including three-dimensional power distribution, fine power reconstruction, fuel temperature distribution, critical search, control rod worth, reactivity coefficients, burnup and nuclide density distribution, etc. CORAL employ nodal expansion method to solve neutron diffusion equation, and the least square method is used to achieve few group constants, and sub-channel model and one-dimensional heat transfer is used to calculate fuel temperature and coolant density distribution, and burnup distribution and nuclide nuclear density could be obtained by solving macro-depletion and micro-depletion equation. The CORAL code is convenient to update and maintain in consider of modular, object-oriented programming technology. In order to analyze the computational accuracy of the CORAL code in small PWR-core and its capability to deal with heterogeneous, calculation analysis are carried out based on the material and geometry parameters of the SMART core. The core has 57 fuel assemblies, with 8, 20 or 24 gadolinium rods arranged in the fuel assemblies. In this paper, a quantitative comparison and analysis of the small PWR problem calculation results are carried out. Numerical results, including effective multiplication factor, assembly power distribution and pin power distribution, all agree well with the calculation results of OpenMC or Bamboo at both hot zero-power (HZP) and hot full-power (HFP) conditions.

scholarly journals Thermal hydraulic model of the molten salt reactor experiment with the NEAMS system analysis module

2019 ◽  
Vol 126 ◽  
pp. 59-67 ◽  
Author(s):  
Adrian M. Leandro ◽  
Florent Heidet ◽  
Rui Hu ◽  
Nicholas R. Brown
Keyword(s):  
Molten Salt ◽  
System Analysis ◽  
Hydraulic Model ◽  
Molten Salt Reactor ◽  
Reactor Experiment

Numerical analysis of the passive heat removal system for molten salt reactor at steady state

2016 ◽  
Vol 102 ◽  
pp. 1337-1344 ◽  
Author(s):  
Xiangcheng Wu ◽  
Changqi Yan ◽  
Zhaoming Meng ◽  
Kailun Chen ◽  
Shaochuang Song ◽  
...  
Keyword(s):  
Steady State ◽  
Numerical Analysis ◽  
Molten Salt ◽  
Heat Removal ◽  
Molten Salt Reactor ◽  
Heat Removal System ◽  
Removal System

scholarly journals Core Flow Distribution from Coupled Supercritical Water Reactor Analysis

2014 ◽  
Vol 2014 ◽  
pp. 1-8
Author(s):  
Po Hu ◽  
Paul P. H. Wilson
Keyword(s):  
Steady State ◽  
Supercritical Water ◽  
Power Distribution ◽  
Reactor Core ◽  
Flow Distribution ◽  
Core Model ◽  
Core Flow ◽  
The Core ◽  
Extended Code ◽  
Reference Design

This paper introduces an extended code package PARCS/RELAP5 to analyze steady state of SCWR US reference design. An 8 × 8 quarter core model in PARCS and a reactor core model in RELAP5 are used to study the core flow distribution under various steady state conditions. The possibility of moderator flow reversal is found in some hot moderator channels. Different moderator flow orifice strategies, both uniform across the core and nonuniform based on the power distribution, are explored with the goal of preventing the reversal.

Study on Calculation Difference of the SCWR Core Steady-State Analysis Code

2020 ◽  
Vol 6 (3) ◽  
Author(s):  
Lianjie Wang ◽  
Lei Yao ◽  
Ping Yang ◽  
Di Lu ◽  
Wenbo Zhao
Keyword(s):  
Steady State ◽  
Cross Section ◽  
Power Distribution ◽  
Group Structure ◽  
Three Dimensional ◽  
Thermal Hydraulics ◽  
Steady State Analysis ◽  
Energy Group ◽  
The Cross ◽  
State Analysis

Abstract The three-dimensional code system supercritical water-cooled reactor (SCWR) coupled neutronics/thermal-hydraulics analysis (SNTA) code is developed for SCWR core steady-state analysis by coupling neutronics/thermal-hydraulics (N/T). This paper studies the calculation difference between the SNTA code and the standard reactor analysis code (SRAC). By using the impacts exclusive method, it is confirmed that the calculation difference between these two codes is caused by different feedback of the cross section. The cross section data and the energy group structure of the SRAC code differ from the SNTA code, and the density coefficient of reactivity calculated by the SRAC code is higher, which means the feedback of the density and power distribution is bigger and the axial power distribution varies rapidly. The SNTA code with finer energy group structure is suitable for the performance analysis of SCWR core which has strong N/T coupling characteristics.

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