scholarly journals IMPROVEMENT OF FUEL ELEMENT SHELL CONTROL METHODS TO INCREASE NUCLEAR REACTOR SAFETY

2020 ◽  
pp. 26-31
Author(s):  
P. F. Budanov ◽  
K. Yu. Brovko ◽  
Е. А. Khomiak ◽  
О. А. Tymoshenko
Keyword(s):  
Fuel Element ◽  
Nuclear Fuel ◽  
Geometric Parameters ◽  
Material Structure ◽  
Shell Material ◽  
Fuel Element Cladding ◽  
Fuel Rod ◽  
Inner Surface ◽  
Fuel Elements ◽  
Cladding Material

The analysis of the existing methods of control of the surface of the fuel element cladding material was carried out, which showed that their use for detecting surface and internal defects, such as local inhomogeneities, micro- and macropores, various cracks, axial looseness, etc. is characterized by low efficiency, is a laborious process that requires additional surface treatment, material of the fuel elements cladding. In addition, the investigated methods of controlling the surface of the fuel element cladding material make it possible to visually identify only rough external cracks and large slag inclusions, small cracks and non-metallic inclusions invisible under the slag layer. It is proposed to assess the quality of the surface of the shell material in case of its damage and destruction, the use of a computational apparatus based on the method of the theory of fractals. It is proposed to use the fractal properties of the shell material structure and a quantitative fractal value – the fractal dimension, which makes it possible to determine the degree of filling of the volume of the shell material structure during fuel element depressurization. A mathematical model of damage to the structure of the fuel element cladding material is developed depending on the simultaneous effect of high temperature and internal pressure caused by the accumulation of nuclear fuel fission products between the nuclear fuel pellet and the inner surface of the fuel element cladding, taking into account the fractal increases in the geometric parameters of the fuel element cladding. It is shown that damaged structures of the fuel rod cladding material depend on the pressure and temperature inside the fuel rod cladding, as well as the fractal increase in geometric parameters, such as: volume and surface area, outer and inner diameters, height and cross-sectional area, cladding length and height of nuclear pellets, gap between the inner surface of the cladding and nuclear fuel. A criterion for assessing the integrity of the fuel rod cladding is determined, which depends on the change in geometric values in the event of damage and destruction of the structure of the fuel rod cladding material. Practical recommendations are given on the use of the proposed method for monitoring the tightness of the fuel element cladding for processing information obtained from the computational module of the system for monitoring the tightness of the cladding for the automated process control system of the nuclear power plant power unit, which makes it possible to detect the depressurization of fuel elements at an earlier stage in comparison with the standard procedure.

scholarly journals Starting Point, Keys and Milestones of a Computer Code for the Simulation of the Behaviour of a Nuclear Fuel Rod

2011 ◽  
Vol 2011 ◽  
pp. 1-11 ◽  
Author(s):  
Armando C. Marino
Keyword(s):  
Nuclear Fuel ◽  
Extreme Case ◽  
Computer Code ◽  
National Commission ◽  
Fuel Rod ◽  
Safety Margins ◽  
Starting Point ◽  
Fuel Design ◽  
Advanced Fuel ◽  
Fuel Elements

The BaCo code (“Barra Combustible”) was developed at the Atomic Energy National Commission of Argentina (CNEA) for the simulation of nuclear fuel rod behaviour under irradiation conditions. We present in this paper a brief description of the code and the strategy used for the development, improvement, enhancement, and validation of a BaCo during the last 30 years. “Extreme case analysis”, parametric (or sensitivity), probabilistic (or statistic) analysis plus the analysis of the fuel performance (full core analysis) are the tools developed in the structure of BaCo in order to improve the understanding of the burnup extension in the Atucha I NPP, and the design of advanced fuel elements as CARA and CAREM. The 3D additional tools of BaCo can enhance the understanding of the fuel rod behaviour, the fuel design, and the safety margins. The modular structure of the BaCo code and its detailed coupling of thermo-mechanical and irradiation-induced phenomena make it a powerful tool for the prediction of the influence of material properties on the fuel rod performance and integrity.

Influence of Iron(II) and Iron(III) Ions on the Corrosion of MTR Fuel Element Claddings in Repository-Relevant Brines

1997 ◽  
Vol 506 ◽  
Author(s):  
S. Zschunke ◽  
J. Fachinger
Keyword(s):  
Corrosion Rate ◽  
Fuel Element ◽  
Fuel Cycle ◽  
Ph Value ◽  
Corrosion Mechanism ◽  
Federal Republic Of Germany ◽  
Spent Fuel ◽  
Material Test ◽  
Fuel Elements ◽  
Cladding Material

ABSTRACTAfter the USA decided in 1988 to no longer accept spent fuel elements from German material test reactors (MTR), a national back-end fuel cycle alternative was sought in the Federal Republic of Germany [1]. The aim is their direct final disposal in deep, stable geologic formations. The corrosion of material test reactor (MTR)-fuel element claddings (aluminium) in repository-relevant brines was examined. Before the aluminium cladding material can corrode, the POLLUX cask, containing the fuel elements, must be corroded. In this case, iron(II) and iron(III) ions are present in the brine. These ions decisively influence the corrosion of the MTR fuel element cladding material, therefore the mechanism responsible for this phenomenon should be identified. Tests were performed in which Fe(II) and Fe(III) salts were added to the brines. In these experiments, the percentage mass decrease of the aluminium cladding, the iron content of the brine, as well as the pH value were determined. As expected the results provided the information about the corrosion mechanism. The higher the concentration of iron ions in the brines, the higher the aluminium corrosion rate was for all three brines. Identical redox equilibria between Fe(II) and Fe(III) were formed in the brine, irrespective of whether Fe(II) or Fe(III) salt had been added. It is assumed that the acceleration of the corrosion rate is based on the fact that Fe(II) is reduced to metallic iron by absorbing the electrons produced during the oxidation of aluminium to Al(III). The aluminium cladding material does not function as a barrier for the release of radionuclides from the fuel elements. The results of this study show that the 0.38 mm thick aluminium cladding will corrode through after approximately four weeks.

scholarly journals Investigation of silver and iodine transport through silicon carbide layers prepared for nuclear fuel element cladding

2011 ◽  
Vol 410 (1-3) ◽  
pp. 24-31 ◽  
Author(s):  
E. Friedland ◽  
N.G. van der Berg ◽  
J.B. Malherbe ◽  
J.J. Hancke ◽  
J. Barry ◽  
...  
Keyword(s):  
Silicon Carbide ◽  
Fuel Element ◽  
Nuclear Fuel ◽  
Fuel Element Cladding ◽  
Nuclear Fuel Element ◽  
Iodine Transport

Enhancement of Heat Transfer Performance in Nuclear Fuel Rod Using Nanofluids and Surface Roughness Technique

2015 ◽  
Author(s):  
Kang Liu ◽  
Titan C. Paul ◽  
Leo A. Carrilho ◽  
Jamil A. Khan
Keyword(s):  
Heat Transfer ◽  
Surface Roughness ◽  
Heat Transfer Coefficient ◽  
Transfer Coefficient ◽  
Nuclear Fuel ◽  
Three Dimensional ◽  
Pressurized Water ◽  
Bulk Fluid ◽  
Fuel Rod ◽  
Dimensional Surface

The experimental investigations were carried out of a pressurized water nuclear reactor (PWR) with enhanced surface using different concentration (0.5 and 2.0 vol%) of ZnO/DI-water based nanofluids as a coolant. The experimental setup consisted of a flow loop with a nuclear fuel rod section that was heated by electrical current. The fuel rod surfaces were termed as two-dimensional surface roughness (square transverse ribbed surface) and three-dimensional surface roughness (diamond shaped blocks). The variation in temperature of nuclear fuel rod was measured along the length of a specified section. Heat transfer coefficient was calculated by measuring heat flux and temperature differences between surface and bulk fluid. The experimental results of nanofluids were compared with the coolant as a DI-water data. The maximum heat transfer coefficient enhancement was achieved 33% at Re = 1.15 × 105 for fuel rod with three-dimensional surface roughness using 2.0 vol% nanofluids compared to DI-water.

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.

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