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.

The Defect Microstructure and Element Composition in Denuded Zones of Stainless Steels Irradiated in BN-350 Nuclear Reactor

MRS Advances ◽  
2016 ◽  
Vol 2 (21-22) ◽  
pp. 1209-1215 ◽  
Author(s):  
Oleg V. Rofman ◽  
Kira V. Tsay ◽  
Oleg P. Maksimkin
Keyword(s):  
Grain Boundaries ◽  
Stainless Steels ◽  
Nuclear Reactor ◽  
Polycrystalline Materials ◽  
Spent Fuel ◽  
Long Term Storage ◽  
Radiation Induced ◽  
High Degree ◽  
Induced Defects

ABSTRACTIt is known that microstructure of metallic polycrystalline materials irradiated with neutrons is often characterized by a high degree of heterogeneity in distribution of radiation-induced defects. Depleted zones are located along grain boundaries and their width is not only determined by irradiation temperature and damage dose, but also by migration of point defects and dislocations integrity, that makes it more difficult to interpret experimental results of this phenomenon. At present, denuded zones are still objects for investigation as they influence both operation characteristics of reactor materials and their safe long-term storage. In this work, denuded zones in hexagonal ducts of spent fuel assemblies constructed from 0.08C-16Cr-11Ni-3Mo and 0.12C-18Cr-10Ni-Ti stainless steels from BN-350 fast nuclear reactor were investigated by TEM. There were determined some irradiation parameters affecting the development of denuded zones and their width; void size distributions in near-grain boundary regions are presented. There was shown redistribution of alloying elements at grain boundaries using Energy-dispersive X-ray spectroscopy (EDS).

scholarly journals Cluster Dynamics Modeling of Materials: Advantages and Limitations

2007 ◽  
Vol 129 ◽  
pp. 51-58 ◽  
Author(s):  
Alain Barbu ◽  
Emmanuel Clouet
Keyword(s):  
Monte Carlo ◽  
Point Defects ◽  
Source Term ◽  
Kinetic Monte Carlo ◽  
Short Review ◽  
Metallic Materials ◽  
Dynamics Modeling ◽  
Term Evolution ◽  
Primary Damage

The aim of this paper is to give a short review on cluster dynamics modeling in the field of atoms and point defects clustering in materials. It is shown that this method, due to its low computer cost, can handle long term evolution that cannot, in many cases, be obtained by Lattice Kinetic Monte Carlo methods. Indeed, such a possibility is achieved thanks to an important drawback that is the loss of space correlations of the elements of the microstructures. Some examples, in the field of precipitation and irradiation of metallic materials are given. The limitations and difficulties of this method are also discussed. Unsurprisingly, it is shown that it goes in a very satisfactory way when the objects are distributed homogeneously. Conversely, the source term describing the primary damage under irradiation, by nature heterogeneous in space and time, is tricky to introduce especially when displacement cascades are produced.

Corrosion Behavior of Pre-oxidized High Burnup Spent Fuel in Salt Brine

2003 ◽  
Vol 807 ◽  
Author(s):  
Andreas Loida ◽  
Bernhard Kienzler ◽  
Horst Geckeis
Keyword(s):  
Dissolution Rate ◽  
Spent Fuel ◽  
Salt Solutions ◽  
Fuel Surface ◽  
Fuel Sample ◽  
Salt Brine ◽  
High Burnup ◽  
Interim Storage ◽  
The Impact

ABSTRACTDuring long-term interim storage of spent fuel, pre-oxidation of the UO2-matrix may not be ruled out completely. This can happen if air could find access to the fuel in the case of cladding failure. The aim of this work is to study the impact of pre-oxidation of the fuel surface on the UO2 matrix dissolution rate and the associated mobilization or retention of radionuclides in highly concentrated salt solutions. The tests were performed with samples that suffered pre-oxidation during up to seven years. The dissolution rate of a fuel sample contacted by small quantities of air-oxygen was found to be roughly a factor of 10 higher in comparison to non oxidized samples, but concentrations of radionuclides, especially Pu and U were hardly affected. The majority of dissolved radionuclides, especially Pu, U appear to have been reimmobilized on the fuel sample itself.

CASTOR® BR3: Design of a Transport and Storage Cask for Spent Fuel From a Belgian Nuclear Power Plant

2001 ◽  
Author(s):  
Ulrich Knopp
Keyword(s):  
Nuclear Power ◽  
Storage Period ◽  
Nuclear Research ◽  
Ductile Cast Iron ◽  
Radioactive Material ◽  
Research Centre ◽  
Spent Fuel ◽  
Interim Storage ◽  
And Storage

Abstract The CASTOR® BR3 cask has been designed and manufactured to accomodate irradiated fuel (U and MOX) from the BR3 test reactor at the nuclear research centre SCK/CEN in Dessel near Mol, Belgium, which is currently being dismantled. The CASTOR® BR3 is designed as a Type B(U)F package for transport and will be licensed in Belgium. In addition, the CASTOR® BR3 needs a license as a storage cask to be operated in an interim cask storage facility. To obtain these licenses, the cask design has to observe the international regulations for the safe transport of radioactive material as well as the special requirements for the cask storage. The CASTOR® BR3 is a member of the CASTOR® family of spent fuel casks, delivered by the German company GNB. In this way, the cask has such typical features as the following: • monolithic cask body made of ductile cast iron; • double-lid system consisting of primary and secondary lid for long-term interim storage of the fuel. This family of casks has been used for over 20 years for transport and storage of spent fuel. In this paper, the IAEA regulatory requirements for transport casks are summarized and it is shown by selected examples how these requirements have been converted into the cask design and the analyses performed for the cask. Finally, the cask features for an interim storage period of up to 50 years will be spotlighted. Main topics are the evaluation of the long term behaviour of selected cask components and the cask monitoring system for the surveillance of the leak tightness of the cask during the storage period.

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.

Safety Aspects of Long-term Dry Interim Storage Of Type-b Spent Fuel and Hlw Transport Casks

2004 ◽  
Vol 15 (3-4) ◽  
pp. 207-214 ◽  
Author(s):  
D. Wolff ◽  
U. Probst ◽  
H. Völzke ◽  
B. Droste ◽  
R. Rödel
Keyword(s):  
Type B ◽  
Spent Fuel ◽  
Safety Aspects ◽  
Interim Storage

Dose Rate Effects on the Accumulation of Radiation Damage

2007 ◽  
Author(s):  
V. V. Rondinella ◽  
T. Wiss ◽  
J.-P. Hiernaut ◽  
D. Staicu
Keyword(s):  
Dose Rate ◽  
Radiation Damage ◽  
Alpha Decay ◽  
Lattice Parameter ◽  
Time Interval ◽  
Spent Fuel ◽  
Dose Rates ◽  
Dose Rate Effects ◽  
Doped Materials ◽  
Interim Storage

During storage, spent fuel and other waste forms accumulate alpha-decay damage (and He). The dose rates and the temperatures experienced during storage are lower than during in-pile operation: however, the duration of the storage is much longer (of the order of up to a few hundred years if extended interim storage concepts are considered); if final disposal in the repository is considered, the time interval in which radiation damage accumulates is open-ended. In order to simulate within timeframes suitable for laboratory experiments long-term accumulation of alpha-decay damage, the so-called alpha-doped materials can be used, i.e. materials loaded with short-lived alpha-emitters (like e.g. Pu-238, U-233, etc.). The question is often posed if the accelerated accumulation of decay damage and He obtained using alpha-doped materials does cause some artefact related to the rate of accumulation rather than by the integrated dose. This work presents evidence that, at least within the range of alpha-activities considered, there is no dose rate effect. By comparing property evolution as a function of accumulated dpa for alpha-doped materials with activities of ∼1010 and ∼108 Bq/g, respectively, the same trends and levels of alteration are observed. In particular, macroscopic properties like hardness (measured by Vickers indentation) or swelling (evolution of lattice parameter derived from XRD), and microstructural formation and accumulation of defects in the lattice of the alpha-doped material are investigated, showing a remarkable similarity of behaviour vs. dpa independently not only from the dose rate, but also from the composition (namely, Pu and U are considered).

scholarly journals Long-Term Radiotoxicity Evaluation of PWR Spent Uranium and MOX Fuel and Highly Active Waste

2020 ◽  
Vol 207 ◽  
pp. 01024
Author(s):  
Petar Paunov ◽  
Ivaylo Naydenov
Keyword(s):  
Nuclear Power ◽  
Spent Nuclear Fuel ◽  
Potential Danger ◽  
Spent Fuel ◽  
Highly Active ◽  
Active Waste ◽  
Interim Storage ◽  
The Given ◽  
Nuclear Power Production

One of the main concerns related to nuclear power production is the generation and accumulation of spent nuclear fuel. Currently most of the spent fuel is stored in interim storage facilities awaiting final disposal or reprocessing. The spent fuel is stored in isolation from the environment in protected facilities or specially designed containers. Nevertheless, spent fuel and highly active waste might get in the environment in case the protective barriers are compromised. In such a case, spent fuel may pose risk to the environment and human health. Those risks depend on the concentration of the given radionuclide and are measured by the value of potential danger. The potential danger is called also ’radiotoxicity’. The paper examines spent uranium and MOX fuels from a reference PWR, as well as the highly radioactive wastes of their reprocessing. The radiotoxicity of the four materials is examined and evaluated for a cooling time of 1000 years. The contribution of different radionuclides is assessed and the cases of reprocessing and no reprocessing of spent fuel have been compared.

Annealing Of Radiation Damage in Tungsten

1970 ◽  
Vol 28 ◽  
pp. 412-413
Author(s):  
Robert C. Rau ◽  
John Moteff
Keyword(s):  
Electron Microscopy ◽  
Transmission Electron Microscopy ◽  
Radiation Damage ◽  
Thermal Annealing ◽  
Neutron Fluence ◽  
Dislocation Loops ◽  
Defect Clusters ◽  
Polycrystalline Tungsten ◽  
Transmission Electron ◽  
Radiation Induced

Transmission electron microscopy has been used to study the thermal annealing of radiation induced defect clusters in polycrystalline tungsten. Specimens were taken from cylindrical tensile bars which had been irradiated to a fast (E > 1 MeV) neutron fluence of 4.2 × 1019 n/cm2 at 70°C, annealed for one hour at various temperatures in argon, and tensile tested at 240°C in helium. Foils from both the unstressed button heads and the reduced areas near the fracture were examined.Figure 1 shows typical microstructures in button head foils. In the unannealed condition, Fig. 1(a), a dispersion of fine dot clusters was present. Annealing at 435°C, Fig. 1(b), produced an apparent slight decrease in cluster concentration, but annealing at 740°C, Fig. 1(C), resulted in a noticeable densification of the clusters. Finally, annealing at 900°C and 1040°C, Figs. 1(d) and (e), caused a definite decrease in cluster concentration and led to the formation of resolvable dislocation loops.

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