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 A method for uncertainty and sensitivity analysis in fuel performance codes

2021 ◽  
Vol 9 (3) ◽  
Author(s):  
Ana Carolina Bortoletto Dantas ◽  
Antonio Teixeira e Silva
Keyword(s):  
Sensitivity Analysis ◽  
Confidence Interval ◽  
Input Data ◽  
Monte Carlo Sampling ◽  
Fuel Rod ◽  
Inner Pressure ◽  
Uncertainty And Sensitivity Analysis ◽  
Steel Aisi ◽  
Line Temperature ◽  
Stainless Steel Aisi

The present study proposes a method for the execution of uncertainty and sensitivity analysis on TRANSURANUS code, adapted for the use of stainless steel AISI-348 as the cladding material for a PWR reactor fuel rod, thus allowing to determine which input data are more relevant to the TRANSURANUS models, as well as a confidence interval for the results. The analysis was made through Monte Carlo sampling, where input values related to the geometry and composition of the fuel rod were taken from a normal distribution truncated around fabrication tolerance values. The generated samples were used as TRANSURANUS input data, and after numerous executions of the code, the results pertaining to the fuel center line temperature, fuel rod inner pressure and cladding strains were used to obtain a confidence interval and to make a variance based sensitivity analysis, showing that the models used in TRANSURANUS are additive in nature, and input interactions are not relevant to the code.

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.

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.

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.

scholarly journals Propagation of uncertainty and sensitivity analysis in an integral oil‐gas plume model

2016 ◽  
Vol 121 (5) ◽  
pp. 3488-3501 ◽  
Author(s):  
Shitao Wang ◽  
Mohamed Iskandarani ◽  
Ashwanth Srinivasan ◽  
W. Carlisle Thacker ◽  
Justin Winokur ◽  
...  
Keyword(s):  
Sensitivity Analysis ◽  
Plume Model ◽  
Propagation Of Uncertainty ◽  
Uncertainty And Sensitivity Analysis ◽  
Oil Gas
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