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 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.

The Gasification of Graphite Matrix in Pebble Bed Reactors

2009 ◽  
Author(s):  
Xinli Yu ◽  
Suyuan Yu
Keyword(s):  
High Temperature ◽  
Corrosion Rate ◽  
Fuel Element ◽  
Pebble Bed ◽  
The Core ◽  
Graphite Matrix ◽  
Matrix Graphite ◽  
Element Matrix ◽  
Fuel Elements ◽  
High Temperature Gas

This paper mainly deals with the simulations of graphite matrix of the spherical fuel elements by steam in normal operating conditions. The fuel element matrix graphite was firstly simplified to an annular part in the simulations. Then the corrosions to the matrix graphite in 10 MW High Temperature Gas-cooled Reactor (HTR-10) and the High Temperature Gas-cooled Reactor—–Pebble-bed Module (HTR-PM) were investigated respectively. The results showed that the gasification of fuel element matrix graphite was uniform and mainly occurred at the bottom of the core in both of the reactors in the mean residence time of the spherical fuel elements. This was mainly caused by the designed high temperature at the bottom. The total mass gasified in HTR-PM was much greater than the HTR-10, while it did not mean much severer corrosion occurred there. As it is known the core volume of HTR-PM is much larger than the HTR-10, which will result in much greater consumed graphite even for the same corrosion rate. The steam only lost about 1 to 3 percent after flowing through the cores in both reactors for different steam conditions. The corrosion of graphite became worse when the steam concentrations increased in helium coolant. The results also indicated that the corrosion rate of fuel element matrix graphite tended to increase slightly with the prolonging of the service time.

scholarly journals Impact of Radiation-Induced Microstructures on the Integrity of Spent Nuclear Fuel (SNF) Elements in Long-Term Storage

2021 ◽  
Vol 1 ◽  
pp. 17-18
Author(s):  
Neslihan Yanikömer ◽  
Rahim Nabbi ◽  
Klaus Fischer-Appelt
Keyword(s):  
Mechanical Properties ◽  
Monte Carlo ◽  
Radiation Damage ◽  
Spent Fuel ◽  
Radiation Induced ◽  
Primary Damage ◽  
Interim Storage ◽  
Fuel Elements ◽  
Cladding Material

Abstract. The current safety concept provides for a period in the range of 40 years for interim storage of spent fuel elements. Since the requirement for proof of safety for to up to 100 years arises, the integrity of the spent fuel elements in prolonged interim storage and long-term repositories is becoming a critical issue. In response to this safety matter, this study aims to assess the impact of radiation-induced microstructures on the mechanical properties of spent fuel elements, in order to provide reliable structural performance limits and safety margins. The physical processes involved in radiation damage and the effect of radiation damage on mechanical properties are inherently multiscalar and hierarchical. Damage evolution under irradiation begins at the atomic scale, with primary knock-on atoms (PKAs) resulting in displacement cascades (primary damage), followed by the defect clusters leading to microstructural deformations. In this context, we have developed and applied a multiscale simulation methodology consistent with the multistage damage mechanisms and the corresponding effects on the mechanical properties of spent fuel cladding and its integrity. Within the improved hierarchical modelling sequence, the effect of the radiation field on the fuel element cladding material (Zircalloy-4) is assessed using Monte Carlo methods. A molecular dynamics method is employed to model damage formation by PKAs and primary damage defect configurations. The formation of clusters and evolution of microstructures are simulated by extending the simulation sequence to a longer time scale with the kinetic Monte Carlo (KMC) method. Transferring the calculated radiation-induced microstructures into macroscopic quantities is ultimately decisive for the structural/mechanical behaviour and stability of the cladding material, and thus for long-term integrity of the spent fuel elements. Results of the multiscale modelling and simulations as well as a comparison with experimental results will be presented at the conference session.

Filtration Technology Research of Graphite Dust Produced in Spent Fuel Transportation Process in HTR-PM

2018 ◽  
Author(s):  
Jinhua Wang ◽  
Bing Wang ◽  
Bin Wu ◽  
Yue Li ◽  
Haitao Wang
Keyword(s):  
High Temperature ◽  
Fuel Element ◽  
Nuclear Power ◽  
Storage System ◽  
Spent Fuel ◽  
Graphite Matrix ◽  
Fuel Storage ◽  
Fuel Elements ◽  
And Storage ◽  
High Temperature Gas

With the continuous development of the nuclear power technology in the world, all countries in the world are becoming more and more interested in the inherent safety of nuclear power technology, while the research and development of the spherical bed type high temperature gas cooled reactor nuclear power technology in China has formally catered to this demand. As a major national science and technology project, since the construction of the high temperature gas cooled reactor demonstration project (HTR-PM) since 2012, the civil construction of the nuclear island has been basically completed, the installation of equipment has been carried out orderly, and many process systems have entered debugging and operation stage gradually. As an important auxiliary process system, fuel handling and storage system for online refueling of the pebble bed high temperature gas cooled reactor, plays an important role in relation to the stable operation of the reactor. The main functions of the fuel handling and storage system are loading the fresh fuel elements and unloading the spent fuel elements which has reached its target burnup continuously for reactor operation, the spent fuel elements would be discharged into the spent fuel canister firstly, when the spent fuel storage canister is full of spent fuel, the canister would be sealed through welding method, and then the spent fuel canister would be transferred and stored in the spent fuel storage silo with the ground crane system. The fuel element of the pebble bed high temperature gas cooled reactor is spherical fuel element with graphite matrix, the fuel elements will have friction and collision with the inner wall of the pipeline in transporting process, which will produce graphite dust, the graphite dust should be removed continuously though filtration method, so as not to affect the fuel elements transportation in pipeline. This article focus on the production mechanism and filtering method of the graphite dust in graphite matrix fuel element transporting process in pipeline, to study the graphite dust removal technology, and then we could provide theoretical guidance for the design and operation of the key system and equipment for HTR-PM.

Research on Surface Corrosion of the Adaptive Expandable Inner Bracing Screen

2010 ◽  
Vol 146-147 ◽  
pp. 1260-1264
Author(s):  
Jian Min Ma ◽  
Yong Hong Liu ◽  
Hang Li ◽  
Zhi Fei Liu
Keyword(s):  
Corrosion Rate ◽  
Calcium Ion ◽  
Ph Value ◽  
Chloride Ion ◽  
Effect Of Temperature ◽  
Corrosion Mechanism ◽  
Surface Corrosion ◽  
Sand Control ◽  
Chlorine Ion ◽  
Screen Surface

The surface corrosion mechanism of the adaptable and expandable inner bracing screen is introduced, experiment and analysis was made for the surface corrosion with the effect of temperature, chloride ion, calcium ion, pH value, this offer help to the application of adaptable and expandable sand control screen. The results show that surface corrosion rate is expedited with the temperature increase during 30 °C～80 °C, it reach the highest spot at 80 °C and then declined. The screen surface occur dot-corrosion easily in solution which contain chlorine ion, corrosion rate decline with the concentration increase of chlorine ion. Calcium ion can restrain the screen surface corrosion rate. Corrosion rate is decline with the pH value increase during the pH value of 3～6.

Investigation of Diffusion of UO2Cl2 in Saturated NaCl Solutions at Various Temperatures

1981 ◽  
Vol 11 ◽  
Author(s):  
Harry Murso ◽  
Bodo Plewinsky ◽  
Dieter Leopold ◽  
Gonter Marx
Keyword(s):  
Borosilicate Glasses ◽  
Federal Republic Of Germany ◽  
Transport Mechanisms ◽  
Spent Fuel ◽  
Waste Storage ◽  
Water Filling ◽  
Diffusion And Convection ◽  
High Level ◽  
Fuel Elements ◽  
Extraction Of Radionuclides

Even in the early 1960's it had already been decided to place high-level radioactive waste in the salt formations of the Federal Republic of Germany. However a major concern is the formation of highly corrosive brines in the event of water filling the repositories. In order to estimate the risks of waste storage in salt repositories it is necessary to investigate the physicochemical properties of the waste materials and of those radionuclides extracted into brine. The extraction of radionuclides from borosilicate glasses or spent fuel elements has been discussed, taking into consideration the two transport mechanisms: diffusion and convection.

Unique Autoclave Stress Induced Failure Mechanism

2005 ◽  
Author(s):  
John Butchko ◽  
Bruce T. Gillette
Keyword(s):  
Corrosion Rate ◽  
Grain Boundary ◽  
Failure Analysis ◽  
Failure Mechanism ◽  
Metal Corrosion ◽  
Corrosion Mechanism ◽  
Reliability Testing ◽  
Process Change ◽  
Transistor Reliability ◽  
Pass Through

Abstract Autoclave Stress failures were encountered at the 96 hour read during transistor reliability testing. A unique metal corrosion mechanism was found during the failure analysis, which was creating a contamination path to the drain source junction, resulting in high Idss and Igss leakage. The Al(Si) top metal was oxidizing along the grain boundaries at a faster rate than at the surface. There was subsurface blistering of the Al(Si), along with the grain boundary corrosion. This blistering was creating a contamination path from the package to the Si surface. Several variations in the metal stack were evaluated to better understand the cause of the failures and to provide a process solution. The prevention of intergranular metal corrosion and subsurface blistering during autoclave testing required a materials change from Al(Si) to Al(Si)(Cu). This change resulted in a reduced corrosion rate and consequently prevented Si contamination due to blistering. The process change resulted in a successful pass through the autoclave testing.

scholarly journals The Need for Integrating the Back End of the Nuclear Fuel Cycle in the United States of America

MRS Advances ◽  
2018 ◽  
Vol 3 (19) ◽  
pp. 991-1003 ◽  
Author(s):  
Evaristo J. Bonano ◽  
Elena A. Kalinina ◽  
Peter N. Swift
Keyword(s):  
United States ◽  
Nuclear Fuel ◽  
United States Of America ◽  
Spent Nuclear Fuel ◽  
Fuel Cycle ◽  
Nuclear Fuel Cycle ◽  
The United States ◽  
Spent Fuel ◽  
Dry Storage ◽  
The Us

ABSTRACTCurrent practice for commercial spent nuclear fuel management in the United States of America (US) includes storage of spent fuel in both pools and dry storage cask systems at nuclear power plants. Most storage pools are filled to their operational capacity, and management of the approximately 2,200 metric tons of spent fuel newly discharged each year requires transferring older and cooler fuel from pools into dry storage. In the absence of a repository that can accept spent fuel for permanent disposal, projections indicate that the US will have approximately 134,000 metric tons of spent fuel in dry storage by mid-century when the last plants in the current reactor fleet are decommissioned. Current designs for storage systems rely on large dual-purpose (storage and transportation) canisters that are not optimized for disposal. Various options exist in the US for improving integration of management practices across the entire back end of the nuclear fuel cycle.

Studying the Effect of Fuel Elements Structural Materials Corrosion on their Operating Life

2021 ◽  
Vol 1038 ◽  
pp. 108-115
Author(s):  
Yuliana Hapon ◽  
Maksym Kustov ◽  
Volodumur Kalugin ◽  
Alexander Savchenko
Keyword(s):  
Experimental Data ◽  
Mass Transfer ◽  
Fuel Element ◽  
State Level ◽  
External Factors ◽  
Economic Losses ◽  
Operating Life ◽  
Hydrodynamic Mode ◽  
Internal And External Factors ◽  
Fuel Elements

The paper deals with experimental data regarding the effect of internal and external factors on the corrosion decay of Zr1Nb alloy fuel elements. Based on the analysis results, losses of zirconium that transfers to oxide or coolant as per the fuel element wall weight and thickness as well as economic losses from their corrosion decay have been theoretically calculated. To avoid a state-level emergency occurrence, an increase in the fuel element wall thickness up to 660 μm is proposed, which can increase the operating life under the conditions of trouble-free coolant mass transfer hydrodynamic mode.

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