Influence Analysis of Coupled Neutronics and Thermo-Hydraulics on Steady-State Characteristics of Supercritical Water-Cooled Reactor

2012 ◽  
Vol 472-475 ◽  
pp. 278-283
Author(s):  
Juan Chen ◽  
Tao Zhou ◽  
Zhou Sen Hou ◽  
Wan Xu Cheng ◽  
Can Hui Sun
Keyword(s):  
Steady State ◽  
Supercritical Water ◽  
Cross Sections ◽  
Power Distribution ◽  
Coupled Model ◽  
Calculation Model ◽  
Coolant Temperature ◽  
Coupling Condition ◽  
Coupling Calculation ◽  
Supercritical Water Cooled Reactor

the coulped neutronics and thermo-hydraulics model for supercritical water-cooled reactor (SCWR) is developed by internal coupling method. It is based on the two group neutron diffusion equations and the one-dimensional junction thermal analysis mode, in which the cross sections used for SCWR are generated by Dragon tool. Compared with the calculation results based on the non-coupling calculation model, the steady state characteristics under coupling calculation condition are detailed analyzed by considering parameters feedback at each axial node. The results show that, as coupled model is chosen its axial power distribution would give an obvious deviation from the cosine function that used for non-coupled model. Although the cladding temperature at most of the axial nodes rises with a shifted power peak, the maximum cladding temperature is finally decreased. For the above coupling condition, the maximum cladding temperature would appear at the external assemblies with lower coolant temperature but not at inner assemblies with higher coolant temperature. As the detailed description for coupling characteristics of supercritical water-cooled reactor is given, a certain theory reference for its system safety could be provided.

Improving on Assembly of SCWR Using MOX Fuel

2011 ◽  
Vol 347-353 ◽  
pp. 1633-1636 ◽  
Author(s):  
Can Hui Sun ◽  
Tao Zhou ◽  
Zhou Sen Hou ◽  
Meng Ying Liu ◽  
Feng Luo
Keyword(s):  
Supercritical Water ◽  
Power Distribution ◽  
Security Issue ◽  
Original Design ◽  
Fuel Rod ◽  
Simple Power ◽  
Mox Fuel ◽  
Higher Temperature ◽  
The One ◽  
Supercritical Water Cooled Reactor

A calculation is made for certain Supercritical Water Cooled Reactor (SCWR) using UO2 fuel and MOX fuel respectively. The results indicate that MOX fuel has a simple power distribution with UO2 fuel, but there is a larger power uneven factor when using MOX fuel, and using MOX fuel including weapon grade Pu has larger power uneven factor than using MOX fuel including reactor grade Pu. However, in the case of same power distribution, the fuel rod using MOX fuel has a higher temperature than the one using UO2 fuel. Therefore with the more uneven power distribution, the fuel in SCWR using MOX fuel has a higher temperature. This will result in a big security issue when using MOX fuel in original design of SCWR. Through analyzing the result of power distribution, an improved assembly of SCWR is presented. It can reduce the power uneven factor and increase the security of fuel rod using the improved assembly of SCWR.

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 Optimization Design for CSR1000 Core

2017 ◽  
Vol 4 (1) ◽  
Author(s):  
Lianjie Wang ◽  
Ping Yang ◽  
Di Lu ◽  
Wenbo Zhao
Keyword(s):  
Steady State ◽  
Power Distribution ◽  
Optimization Design ◽  
Code System ◽  
Fuel Cladding ◽  
Steady State Analysis ◽  
Control Capability ◽  
Supercritical Water Cooled Reactor ◽  
Conceptual Core ◽  
Steady State Performance

An optimization design of China supercritical water-cooled reactor (SCWR) with the rated electric power of 1000 MWe (CSR1000) conceptual core is proposed. Steady-state performance of the proposed core is then studied with the SCWR core steady-state analysis code system SNTA. These key parameters such as burnup performance, reactivity control capability, power distribution, maximum fuel cladding temperature, and maximum linear power density are analyzed. The relative coolant flow rate of the second flow path, which is suited with assembly power, is also presented. The study shows that the refueling cycle of CSR1000 core can be extended effectively under the optimization design.

Axial Power and Coolant-Temperature Profiles for a Non-Re-Entrant Pressure-Tube Supercritical Water-Cooled Reactor Fuel Channel

2015 ◽  
Vol 2 (1) ◽  
Author(s):  
Vitali Kovaltchouk ◽  
Eleodor Nichita ◽  
Eugene Saltanov
Keyword(s):  
Cross Sections ◽  
Power Distribution ◽  
Pressure Tube ◽  
Coolant Temperature ◽  
Outlet Temperature ◽  
Thermal Hydraulics ◽  
Fuel Temperature ◽  
Fuel Channel ◽  
Technical Specifications ◽  
Institute Of Technology

The axial power and coolant-temperature distributions in a fuel channel of the Generation IV pressure-tube super-critical water-cooled reactor (PT-SCWR) are found using coupled neutronics-thermal-hydraulics calculations. The simulations are performed for a channel loaded with a fresh, 78-element Th-Pu fuel assembly. Neutronics calculations are performed using the DONJON diffusion code using two-group homogenized cross sections produced using the lattice code DRAGON. The axial coolant temperature profile corresponding to a certain axial linear heat generation rate is found using a code developed in-house at University of Ontario Institute of Technology (UOIT). The effect of coolant density, coolant temperature, and fuel temperature variation along the channel is accounted for by generating macroscopic cross sections at several axial positions. Fixed-point iterations are performed between neutronics and thermal-hydraulics calculations. Neutronics calculations include the generation of two-group macroscopic cross sections at several axial positions, taking into account local parameters such as coolant temperature and density and average fuel temperature. The coolant flow rate is adjusted so that the outlet temperature of the coolant corresponds to the SCWR technical specifications. The converged axial power distribution is found to be asymmetric, resembling a cosine shape skewed toward the inlet (reactor top).

Burnable Poison Design for Supercritical Water Cooled Reactor With Annular Fuel

2017 ◽  
Vol 4 (1) ◽  
Author(s):  
Zhao Chuanqi ◽  
Wang Kunpeng ◽  
Cao Liangzhi ◽  
Zheng Youqi
Keyword(s):  
Rare Earth ◽  
Fuel Assembly ◽  
Supercritical Water ◽  
Power Distribution ◽  
Rare Earth Oxide ◽  
Burnable Poison ◽  
The Core ◽  
Power Peak ◽  
Supercritical Water Cooled Reactor ◽  
Control Rods

Burnable poison (BP) is used to control excess reactivity in supercritical water cooled reactor (SCWR). It helps reduce the number of control rods. Over all BP designs, the design in which rare-earth oxide mixes with fuel is widely used in SCWR. BP has influence on fuel assembly neutronics performance. After comparing four kinds of rare-earth oxide, Er2O3 is chosen as BP for the annular fuel assembly. The effect of different BP loading patterns on assembly power distribution is analyzed. The safety of annular fuel assembly is estimated with different BP contents. Core performance with and without BP is compared. The results had shown that the core radial power peaking factor decreased after introducing BP. It was also shown that the core axial power peaking factor increased, and the power peak moved toward the top of the core. The reason of this effect was studied. Two optimizations were given based on this study: decreasing the temperature of lower plenum and increasing the gradients of axial enrichments. By applying these optimizations, core axial power peaking factor and maximum cladding surface temperature decreased.

scholarly journals Design Analyses for Full Core Conversion of The Dalat Nuclear Research Reactor

2014 ◽  
Vol 4 (1) ◽  
pp. 10-25
Author(s):  
Ba Vien Luong ◽  
Vinh Vinh Le ◽  
Ton Nghiem Huynh ◽  
Kien Cuong Nguyen
Keyword(s):  
Steady State ◽  
Power Distribution ◽  
Research Reactor ◽  
Nuclear Research ◽  
Coolant Temperature ◽  
Thermal Hydraulics ◽  
Kinetics Parameters ◽  
Nuclear Research Reactor ◽  
Enriched Uranium ◽  
Full Core

The paper presents calculated results of neutronics, steady state thermal hydraulics and transient/accidents analyses for full core conversion from High Enriched Uranium (HEU) to Low Enriched Uranium (LEU) of the Dalat Nuclear Research Reactor (DNRR). In this work, the characteristics of working core using 92 LEU fuel assemblies and 12 beryllium rods were investigated by using many computer codes including MCNP, REBUS, VARI3D for neutronics, PLTEMP3.8 for steady state thermal hydraulics, RELAP/MOD3.2 for transient analyses and ORIGEN, MACCS2 for maximum  hypothetical accident (MHA). Moreover, in neutronics calculation, neutron flux, power distribution, peaking factor, burn up distribution, feedback reactivity coefficients and kinetics parameters of the working core were calculated. In addition, cladding temperature, coolant temperature and ONB margin were estimated in steady state thermal hydraulics investigation. The working core was also analyzed under initiating events of uncontrolled withdrawal of a control rod, cooling pump failure, earthquake and MHA. Obtained results show that DNRR loaded with LEU fuel has all safety features as HEU and mixed HEU-LEU fuel cores and meets requirements in utilization as well.

Power Flow Calculation Realization in Software for Grids with Four-Phase Power Lines

2014 ◽  
Vol 698 ◽  
pp. 694-698 ◽  
Author(s):  
Dmitry I. Bliznyuk ◽  
Pavel Y. Bannykh ◽  
Alexandra I. Khalyasmaa
Keyword(s):  
Steady State ◽  
Power Flow ◽  
Calculation Model ◽  
Power Lines ◽  
Power Flow Calculation ◽  
Steady State Analysis ◽  
Flow Calculation ◽  
Analysis Methods ◽  
Electrical Grids ◽  
Three Phase

The paper is devoted to the problem of power flow calculation and steady state analysis methods adaptation for four-phase electrical grids. These methods are based on developed models of four-phase power lines and phase convering transformers. The basis of research is nodal voltages equations for three-phase, four-phase and mixed (combined by three-and four-phase elements) grids. Algorithm of four-phfase elements parameters automized adaptation for power flow calculation model of "RastrWin" software have been developed.

Sign in / Sign up

Export Citation Format

Share Document