Analysis on Pellet-Cladding Interaction of Fuel Rod During Power Ramp of NHR200-II

2018 ◽  
Author(s):  
Tianshan Kang ◽  
Songyang Li ◽  
Dingqu Wang ◽  
Yueyuan Jiang ◽  
Weihua Li
Keyword(s):  
Mechanical Model ◽  
Computational Models ◽  
Nuclear Reactor ◽  
Stress And Strain ◽  
Transfer Model ◽  
Analysis Model ◽  
Reactor Operation ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Calculation Results

In order to ensure the safety of fuel rods in nuclear reactor, it is necessary to consider the condition of the power ramp during reactor operation, which may cause breakage risk of fuel rod at the time of pellet-cladding interaction (PCI) appearing. To analyze this phenomenon and reduce the risk, a performance analysis model for fuel rod is developed to carry out the steady state and transient simulation using commercial software COSMOL. The full model has three main computational models, which are heat transfer model, mechanical model and internal pressure model. The calculation results show that the stress and strain of NHR200-II fuel rods has enough margins during the power ramp process, which indicates that the risks of fuel rod damage are very low.

Adaptation of FEMAXI-6 Code for Fuel Rods of RBMK-1500 and Employment of Uncertainty and Sensitivity Analysis

2009 ◽  
Author(s):  
Tadas Kaliatka ◽  
Ausˇra Jusevicˇiu¯te˙ ◽  
Eugenijus Usˇpuras
Keyword(s):  
Thermal Conductivity ◽  
Sensitivity Analysis ◽  
Reasonable Agreement ◽  
Normal Plant ◽  
Kurchatov Institute ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Code Model ◽  
Uncertainty And Sensitivity Analysis ◽  
Calculation Results

This paper provides information about possibility to apply FEMAXI-6 code for RBMK-1500. According to RBMK-1500 specification new thermal properties and models responsible for thermal analysis (thermal conductivity and heat capacity) were included in FEMAXI-6 code. Using adapted FEMAXI-6 code model of RBMK-1500 fuel rod was developed and tested by employing uncertainty and sensitivity analysis. The processes of fuel rods during normal plant operation were modelled. The received results were compared with calculations performed by specialists from Kurchatov Institute (designers of RBMK). The reasonable agreement of both calculation results shows that adapted FEMAXI-6 code and developed model are suitable for future analysis of processes in fuel rods of RBMK-1500.

scholarly journals New Strategies in the Code of Uncertainty and Sensitivity Analysis (CUSA) and Its Application in the Nuclear Reactor Calculation

2020 ◽  
Vol 2020 ◽  
pp. 1-16 ◽  
Author(s):  
Jiayu Du ◽  
Chen Hao ◽  
Ji Ma ◽  
Peijun Li ◽  
Xiaoyu Zhou ◽  
...  
Keyword(s):  
Sensitivity Analysis ◽  
Cross Sections ◽  
Nuclear Reactor ◽  
Analysis Model ◽  
Total Uncertainty ◽  
Practical Applications ◽  
Uniform Distributions ◽  
Uncertainty And Sensitivity Analysis ◽  
Calculation Results ◽  
New Strategies

Best-Estimation Plus Uncertainty (BEPU) analysis method can provide more information to improve the reliability of calculation results than the safety analysis with conservative assumption. And the statistical sampling-based uncertainty and sensitivity analysis methods are widely used in practical applications of the multiphysics, multiscale coupling nuclear reactor system. In this paper, a novel and efficient sampling method for inputs with normal and uniform distributions is introduced and a systematic theory for uncertainty and sensitivity analysis is established based on the classical statistical theory. Then the Code of Uncertainty and Sensitivity Analysis (CUSA) is updated based on these new strategies. For applications, the explicit and implicit effects for resonance and nonresonance isotopes are studied in depth, and a simple UO2 pin cell is considered to examine the performance of CUSA and the total uncertainty and sensitivity analysis abilities. The numerical results indicate that the implicit sensitivity analysis model and the uncertainty quantification functions developed in CUSA are correct and can be used for sensitivity and uncertainty analysis in nuclear reactor calculations. Even more important, the LHS-SVDC is recommended to propagate the uncertainties in multigroup cross sections.

Numerical Research on Fuel Rod Progression During Core Degradation Process Using MELCOR

2018 ◽  
Author(s):  
Tangtao Feng ◽  
Wenxi Tian ◽  
Ping Song ◽  
Jun Wang ◽  
Mingjun Wang ◽  
...  
Keyword(s):  
Numerical Study ◽  
Degradation Process ◽  
Severe Accident ◽  
Analysis Model ◽  
Experimental Conditions ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Numerical Research ◽  
Fuel Elements ◽  
Very High

Core degradation experiment will be conducted in Fuel Rod Melt Progression apparatus (FROMA) to investigate the distribution of mass and energy in designed fuel rod during a core degradation process. Numerical research on the core degradation experiment is mainly to evaluate the designed parameters of FROMA. The pre-numerical study was conducted using the widely accepted severe accident analysis software MELCOR. The fuel rods used in the experiment consisted of real reactor fuel elements, and the designed fuel rods were electrically heated with internal center tungsten-rhenium rod to a very high temperature. Numerical analysis model is described, and the input parameters are in accord with experimental conditions. In the transient period of core degradation, the production hydrogen and the distribution of mass and temperature are obtained. All the predicted MELCOR results will be compared with the experimental measurements.

scholarly journals Development of Equivalent Beam Model of High Burnup Spent Nuclear Fuel Rods under Lateral Impact Loading

Metals ◽  
2020 ◽  
Vol 10 (4) ◽  
pp. 470
Author(s):  
Sanghoon Lee ◽  
Seyeon Kim
Keyword(s):  
Nuclear Fuel ◽  
Structural Integrity ◽  
Spent Nuclear Fuel ◽  
Model Parameters ◽  
Beam Model ◽  
Spent Fuel ◽  
Analysis Model ◽  
Fuel Rods ◽  
Integrity Assessment ◽  
Fuel Rod

Spent nuclear fuel (SNF) is nuclear fuel that has been irradiated and discharged from nuclear reactors. During the whole management stages of SNF before it is, in the end, disposed in a deep geological repository, the structural integrity of fuel rods and the assemblies should be maintained for safety and economic reasons. In licensing applications for the SNF storage and transportation, the integrity of SNF needs to be evaluated considering various loading conditions. However, this is a challenging task due to the complexity of the geometry and properties of SNF. In this paper, a simple and equivalent analysis model for SNF rods is developed using model calibration based on optimization and process integration. The spent fuel rod is simplified into a hollow beam with a homogenous isotropic material, and the model parameters thus found are not dependent on the length of the reference fuel rod segment that is considered. Two distinct models with different interfacial conditions between the fuel pellets and cladding are used in the calibration to account for the effect of PCMI (Pellet-Clad Mechanical Interaction). The feasibility of the models in dynamic impact simulations is examined, and it is expected that the developed models can be utilized in the analysis of assembly-level analyses for the SNF integrity assessment during transportation and storage.

Effect of pipe expansion on the redistribution of residual stresses after molding

2020 ◽  
Vol 10 (2) ◽  
pp. 122-126
Author(s):  
Nikolay A. Makhutov ◽  
◽  
Dmitry A. Neganov ◽  
Eugeny P. Studenov ◽  
◽  
...  
Keyword(s):  
Residual Stresses ◽  
Strain Distribution ◽  
Strain State ◽  
Stress And Strain ◽  
Cylindrical Shape ◽  
Pipe Material ◽  
Pipe Expansion ◽  
Calculation Results ◽  
Residual Deformations ◽  
Stress And Strain Distribution

In the factory, pipes for trunk oil and oil product pipelines are obtained by molding and welding. To ensure a cylindrical shape and reduce technological residual stresses, expansion technology is used. Pipe expansion causes a significant change in the values of residual deformations and stresses. The article presents both the calculation results and graphs regarding stress and strain distribution during bending of the stock and their redistribution after expansion. Based on the calculation results, the final total values of residual stresses and residual deformations caused by bending and expansion were stated to be important components of the stress-strain state observed in pipelines being operated under cyclic loading, as well as those used in assessing how degradation affects the ductility of the pipe material. These factors were concluded as being reasonably taken into account when performing verification calculations regarding long-running pipelines if, based on their diagnostics and analysis, their state does not meet modern strength requirements.

Rod Bow Ultrasonic Measurements Possibilities

2019 ◽  
Vol 5 (3) ◽  
Author(s):  
Marcin Kopeć ◽  
Martina Malá
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Reactors ◽  
Research Centre ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Nuclear Fuel Assembly ◽  
Fuel Assemblies ◽  
Ultrasonic Measurements ◽  
History Of ◽  
Fuel Particles

The ultrasonic (UT) measurements have a long history of utilization in the industry, also in the nuclear field. As the UT transducers are developing with the technology in their accuracy and radiation resistance, they could serve as a reliable tool for measurements of small but sensitive changes for the nuclear fuel assembly (FA) internals as the fuel rods are. The fuel rod bow is a phenomenon that may bring advanced problems as neglected or overseen. The quantification of this issue state and its probable progress may help to prevent the safety-related problems of nuclear reactors to occur—the excessive rod bow could, in the worst scenario, result in cladding disruption and then the release of actinides or even fuel particles to the coolant medium. Research Centre Rez has developed a tool, which could serve as a complementary system for standard postirradiation inspection programs for nuclear fuel assemblies. The system works in a contactless mode and reveals a 0.1 mm precision of measurements in both parallel (toward the probe) and perpendicular (sideways against the probe) directions.

Application of the Burnable Poison in the Design of Space Nuclear Reactor

2017 ◽  
Author(s):  
Wang SanBing ◽  
Xie Qilin ◽  
He ChaoHui
Keyword(s):  
Neutron Spectrum ◽  
Nuclear Reactor ◽  
Operation Time ◽  
Life Time ◽  
Stable Operation ◽  
Burnable Poison ◽  
Calculation Results ◽  
Homogeneous Reactor ◽  
In The Beginning ◽  
Criticality Safety

The previous research showed that the application of burnable poison was helpful to improve the criticality safety of space nuclear reactor (SNR). In order to analyze the worth of burnable poison in the SNR’s design, a model of homogeneous reactor had firstly been built based on the design of SAFE400. Comparing its difference with the real design of SAFE400 through criticality calculation, the precise of our model had been verified. Then the influence of the criticality parameters and immersion accident character parameters for this model had been analyzed for the application of the different burnable poisons (such as samarium, europium or gadolinium). The calculation results had shown that the application of the most of the burnable poisons would soften the neutron spectrum and induced a decrement of reactor’s keff in the beginning of life. However, the immersion accident analysis gave out another result that only the reactor using gadolinium could ensure the criticality safety of reactor after it made its initial keff equal with the design value. Meanwhile, compared with the initial design of SAFE400, in one hand, the burn-up results had shown that the decrement of homogeneous reactor’s reactivity using gadolinium as burnable poison was deceased after the 10 years full power operation; in other hand, its neutron spectrum became more softer with the operation time; and what’s more important, the amount of the burnable poison was not decreased with burn-up during its service life-time. These results implied that the application of the burnable poison (especially for gadolinium) could highly ensure the criticality safety and stable operation of SNR.

The Mechanical Mechanism Analysis for Mortar Arch Framework Slope Protection Structure

2013 ◽  
Vol 859 ◽  
pp. 143-148
Author(s):  
Yang Xu ◽  
Ding Ling Li ◽  
Li Peng ◽  
Yan Xiao ◽  
Yi Hua Nie
Keyword(s):  
Finite Element Analysis ◽  
Mechanism Analysis ◽  
Analysis Model ◽  
Element Analysis ◽  
Finite Element Analysis Model ◽  
Vertical Loading ◽  
Total Displacement ◽  
The Finite Element Analysis ◽  
Calculation Results ◽  
Real Scale

The finite element analysis model was built as the real scale for mortar arch framework slope protection, and the displacement and strain at different points were collected by vertical loading pressure. So the mechanical mechanism can be studied, and the analysis was done between calculation results and testing results of solid miniature model. The studying results show that the point on the arch foot is the worst stress place for each arch, and the total displacement increase nonlinear as the distance from the slope top increases, and the bump phenomenon exists in the bottom of slope, the points are likely to be broken.

Seismic Response Analysis of Soil-Structure Interaction on Base Isolation Structure

2013 ◽  
Vol 663 ◽  
pp. 87-91
Author(s):  
Ying Bo Pang
Keyword(s):  
Finite Element ◽  
Seismic Response ◽  
Soil Structure ◽  
Base Isolation ◽  
Soil Structure Interaction ◽  
Response Analysis ◽  
Analysis Model ◽  
Seismic Response Analysis ◽  
Structure Interaction ◽  
Calculation Results

As an effective way of passive damping, isolation technology has been widely used in all types of building structures. Currently, for its theoretical analysis, it usually follows the rigid foundation assumption and ignores soil-structure interaction, which results in calculation results distortion in conducting seismic response analysis. In this paper, three-dimensional finite element method is used to establish finite element analysis model of large chassis single-tower base isolation structure which considers and do not consider soil-structure interaction. The calculation results show that: after considering soil-structure interaction, the dynamic characteristics of the isolation structure, and seismic response are subject to varying degrees of impact.

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