scholarly journals Power Distribution and Possible Influence on Fuel Failure in WWER-1000

2008 ◽  
Vol 2008 ◽  
pp. 1-5
Author(s):  
Ján Mikuš
Keyword(s):  
Experimental Data ◽  
Fuel Assembly ◽  
Power Distribution ◽  
Power Reactor ◽  
Construction Materials ◽  
Static Loads ◽  
Fission Rate ◽  
Light Water ◽  
Fuel Failure ◽  
Fuel Pin

The work is focused on the influence of investigation of some core heterogeneities and construction materials on the space power (fission rate) distribution in WWER-1000-type cores, especially from viewpoint of the values and gradient occurrence that could result in static loads with some consequences, for example, fuel pin (FP) or fuel assembly (FA) bowing and possible contribution to the FP failure root causes. For this purpose, experimental data and their analysis from two earlier performed measurements on light water, zero-power reactor LR-0 were used, concerning the relative radial power distribution determined by measurements in a WWER-1000-type core containing single FPs with homogeneous gadolinium admixture () and the relative radial power distribution determined by measurements in FA situated on the periphery of a WWER-1000-type core neighbouring the baffle (thermal shielding).

Study on SCWR Fuel Assembly Design With Thermal-Hydraulic and Neutron-Physical Characteristics

2010 ◽  
Author(s):  
Zhenyang Li ◽  
Tao Zhou ◽  
Canhui Sun ◽  
Xiaozhuang Liu
Keyword(s):  
Fuel Assembly ◽  
Cross Section ◽  
Power Distribution ◽  
Neutron Cross Section ◽  
Physical Characteristics ◽  
Supercritical Pressure ◽  
Light Water ◽  
Assembly Design ◽  
Temperature Distributions ◽  
Parallel Channels

Physical characteristics of the coolant in the Supercritical-pressure Light Water Cooled Reactor (SCWR) vary greatly near the pseudo-critical point, which will cause variations of core neutron cross section and then bring about power perturbation. Further it will prompt the corresponding thermal parameters of supercritical water changed, and form feedback action, finally resulting in intensely coupled thermal-hydraulics and neutron-physical. Proper fuel assembly has been selected as research object, and the model of multiple parallel channels has been established. In view of this model, using DRAGON code for neutron-physical calculations and developing corresponding thermal-hydraulic programs, and then achieve coupling them through appropriate data interface, the calculation platform established. Finally the power distribution and the corresponding parameters temperature distributions in the model have been predicted. On account of deficiencies reflected in calculations, such as the heterogeneous power distribution, fuel assembly geometry has been changed, for instance the proper peripheral moderator wall has been added based on the preceding assembly, then do the coupling calculations and analyze the results. Comparisons between different results have been made, and the expected aim has been reached, which can prove the rationality of assembly modifications and meanwhile prove the usability of the calculation platform. Thus, modified assembly and the calculation platform could be the calculation foundation in future designs of SCWR.

Creep Inverse Problem of the Power Reactor Fuel Assembly Elements Contact Interaction

2014 ◽  
Vol 698 ◽  
pp. 643-648
Author(s):  
Michail Gusev ◽  
Vladimir Danilov
Keyword(s):  
Experimental Data ◽  
Fuel Assembly ◽  
Contact Interaction ◽  
Power Reactor ◽  
Wide Spectrum ◽  
Reactor Core ◽  
Grid Cell ◽  
Parameters Identification ◽  
Experimental Investigations ◽  
Core Elements

Creep inverse task solution for parameters identification of Zr1Nb alloy creep law is considered. Zr1Nb alloy is commonly used as a structural material for power reactor core elements. Strength and reliability justification of the fuel assembly (FA) components, manufactured from Zr1Nb alloy, requires the knowledge of the relationship between the intensity of creep strain rate, temperature and stress intensity. However, FA elements analyses performed using the known coefficients for the Zr1Nb alloy creep law, are not consistent with the experimental data. In this paper the results of parameters identification Zr1Nb alloy creep law are shown. The results are obtained using the numerical and experimental investigations of contact interaction between fuel rod and spacer grid cell. Numerical experiments, performed using verified creep law, are close to the experimental data. The results of this research can be used for a wide spectrum of thermal strength calculations of FA elements manufactured from Zr1Nb alloy.

scholarly journals Neutronics feasibility of using Gd₂O₃ particles in VVER-1000 fuel assembly

2016 ◽  
Vol 6 (3) ◽  
pp. 1-7
Author(s):  
Van Khanh Hoang ◽  
Phi Hung Hoang Thanh ◽  
Hoai Nam Tran
Keyword(s):  
Thermal Conductivity ◽  
Fuel Assembly ◽  
High Power ◽  
Power Distribution ◽  
Fuel Pellet ◽  
Power Peak ◽  
Calculation Results ◽  
Fuel Pin ◽  
High Burnup ◽  
Power Densities

Neutronics feasibility of using Gd2O3 particles for controlling excess reactivity of VVER-1000 fuel assembly has been investigated. The motivation is that the use of Gd2O3 particles would increase the thermal conductivity of the UO2+Gd2O3 fuel pellet which is one of the desirable characteristics for designing future high burnup fuel. The calculation results show that the Gd2O3 particles with the diameter of 60 µm could control the reactivity similarly to that of homogeneous mixture with the same amount of Gd2O3. The power densities at the fuel pin with Gd2O3 particles increase by about 10-11%, leading to the decrease of the power peak and a slightly flatter power distribution. The power peak appears at the periphery pins at the beginning of burnup process which is decreased by 0.9 % when using Gd2O3 particles. Further work and improvement are being planned to optimize the high power peaking at the beginning of burnup.

scholarly journals KYLIN-V2.0 CODE CALCULATION ABILITY VERIFICATION BASED ON VERA BENCHMAR

2021 ◽  
Vol 247 ◽  
pp. 10020
Author(s):  
Dongyong Wang ◽  
Yingrui Yu ◽  
Xingjie Peng ◽  
Chenlin Wang ◽  
Kun Liu ◽  
...  
Keyword(s):  
Fuel Assembly ◽  
Power Distribution ◽  
Nuclear Power ◽  
Neutron Transport ◽  
Monte Carlo Code ◽  
Neutron Transport Equation ◽  
Control Rod ◽  
Fuel Assemblies ◽  
Fuel Pin ◽  
Calculation Ability

Virtual Environmental for Reactor Analysis (VERA) benchmark was released by the Consortium for Advanced Simulation of Light water reactors (CASL) project in 2012. VERA benchmark includes more than ten problems at different levels, from 2D fuel pin case to 2D fuel assembly case to 3D core refuelling case, in addition, reference results and experimental measured data of some problems were provided by CASL. Fuel assemblies in VERA benchmark are various, including control rod assemblies, Pyrex assembly, IFBA assembly, WABA assembly and gadolinium poison assembly, and so on. In this paper, various fuel assembly models in the VERA benchmark have been built by using KYIIN-V2.0 code to verify its calculation ability from 2D fuel pin case to 2D fuel assembly case to 2D 3x3 fuel assembly case, and making a comparative analysis on the reference results in VERA benchmark, as well as the calculation results of the Monte Carlo code RMC. KYLIN-V2.0 is an advanced neutron transport lattice code developed by Nuclear Power Institute of China (NPIC). The subgroup resonance calculation method is used in KYIIN-V2.0 to obtain effective resonance selfshielding cross section, method of modular characteristics (MOC) is adopted to solve the neutron transport equation, and CRAM method and PPC method is adopted to solve the depletion equation. The numerical results show that KYLIN-V2.0 code has the reliable capability of direct heterogeneous calculation of 2D fuel assembly, and the effective multiplication factor, assembly power distribution, rod power distribution and control rod reactivity worths of various fuel assemblies that are calculated by KYLIN-V2.0 are in better agreement with the reference.

Neutronic Analysis of Accident-Tolerant Fuel Assemblies With Silicon Carbide Cladding in Pressurized Water Reactors

2017 ◽  
Author(s):  
Zhixiong Tan ◽  
Jiejin Cai
Keyword(s):  
Silicon Carbide ◽  
Fuel Assembly ◽  
Power Distribution ◽  
Safety Margin ◽  
Severe Accident ◽  
Pressurized Water Reactors ◽  
Fuel Pellet ◽  
Pressurized Water ◽  
Neutronic Parameters ◽  
Water Reactors

After Fukushima Daiichi Nuclear Power Plant accident, alternative fuel-design to enhance tolerance for severe accident conditions becomes particularly important. Silicon carbide (SiC) cladding fuel assembly gain more safety margin as novel accident tolerant fuel. This paper focuses on the neutron properties of SiC cladding fuel assembly in pressurized water reactors. Annular fuel pellet was adopted in this paper. Two types of silicon carbide assemblies were evaluated via using lattice calculation code “dragon”. Type one was consisted of 0.057cm SiC cladding and conventional fuel. Type two was consisted of 0.089cm SiC cladding and BeO/UO2 fuel. Compared the results of SiC cladding fuel assembly neutronic parameters with conventional Zircaloy cladding fuel assembly, this paper analyzed the safety of neutronic parameters performance. Results demonstrate that assembly-level reactivity coefficient is kept negative, meanwhile, the numerical value got a relatively decrease. Other parameters are conformed to the design-limiting requirement. SiC kinds cladding show more flat power distribution. SiC cases also show the ability of reducing the enrichment of fuel pellets even though it has higher xenon concentration. These types of assembly have broadly agreement neutron performance with the conventional cladding fuel, which confirmed the acceptability of SiC cladding in the way of neutron physics analysis.

scholarly journals Influence of void fraction in the power distribution for a GE-12 fuel assembly

2017 ◽  
Vol 792 ◽  
pp. 012051
Author(s):  
S J Castillo ◽  
G A Vargas ◽  
E del Valle Gallegos
Keyword(s):  
Fuel Assembly ◽  
Void Fraction ◽  
Power Distribution

scholarly journals Thermo-hydraulic Simulation of AP1000 Nuclear Reactor Fuel Assembly

2021 ◽  
Vol 31 (1) ◽  
pp. 60-71
Author(s):  
Leonardo Acosta Martínez ◽  
Carlos Rafael García Hernández ◽  
Jesus Rosales García ◽  
Annie Ortiz Puentes
Keyword(s):  
Fuel Assembly ◽  
Power Distribution ◽  
Capital Investment ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Geometric Model ◽  
Modular Construction ◽  
Construction Time ◽  
Hydraulic Simulation ◽  
In Series

One of the challenges of future nuclear power is the development of safer and more efficient nuclear reactor designs. The AP1000 reactor based on the PWR concept of generation III + has several advantages, which can be summarized as: a modular construction, which facilitates its manufacture in series reducing the total construction time, simplification of the different systems, reduction of the initial capital investment and improvement of safety through the implementation of passive emergency systems. Being a novel design it is important to study the thermohydraulic behavior of the core applying the most modern tools. To determine the thermohydraulic behavior of a typical AP1000 fuel assembly, a computational model based on CFD was developed. A coupled neutronic-thermohydraulic calculation was performed, allowing to obtain the axial power distribution in the typical fuel assembly. The geometric model built used the certified dimensions for this type of installation that appear in the corresponding manuals. The thermohydraulic study used the CFD-based program ANSYS-CFX, considering an eighth of the fuel assembly. The neutronic calculation was performed with the program MCNPX version 2.6e. The work shows the results that illustrate the behavior of the temperature and the heat transfer in different zones of the fuel assembly. The results obtained agree with the data reported in the literature, which allowed the verification of the consistency of the developed model.

Fully Coupled Simulation of Oxygen and Heat Diffusion for (U,Pu)O2 Fuel in Both Fast-Breeder Reactor and Light-Water Reactor

2015 ◽  
Vol 1 (4) ◽  
Author(s):  
Wenzhong Zhou ◽  
Rong Liu
Keyword(s):  
High Temperature ◽  
Power Distribution ◽  
Heat Diffusion ◽  
Light Water Reactor ◽  
Fast Breeder Reactor ◽  
Light Water ◽  
Water Reactor ◽  
Ratio Change ◽  
Overall Performance ◽  
Fully Coupled

Oxygen redistribution with a high-temperature gradient is an important fuel performance concern in fast-breeder reactor (FBR) and light-water reactor (LWR) (U,Pu)O2 fuel under irradiation, and affects fuels properties, power distribution, and fuel overall performance. This paper studies the burnup dependent oxygen and heat diffusion behavior in a fully coupled way within (U,Pu)O2 FBR and LWR fuels. The temperature change shows relatively larger impact on oxygen to metal (O/M) ratio redistribution rather than O/M ratio change on temperature, whereas O/M ratio redistributions show different trends for FBR and LWR fuels due to their different deviations from the stoichiometry of oxygen under high-temperature environments.

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