24 research outputs found
Development of an Optimization Code for the Logistics of Spent Fuel Assembly Loading into Final Disposal Canisters
No full textStructural performance and mechanical properties investigation of spent nuclear fuel rods under static and dynamic bending loads
No full textCurrently, most Spent Nuclear Fuel (SNF) is kept safely in storage either at on-site facilities or at centralized interim storage sites. Moreover, many countries face delays in implementing their waste management programmes for SNF and high-level waste disposal. As storage pools approach their capacity, an increasing demand for interim dry-storage solutions is foreseen in the near future and, is associated with emerging regulatory issues. Storage prolongation in interim facilities for time periods far beyond than what originally envisaged, requires license renewal of the facility, as well as of transport and storage casks. At the same time, the safety of SNF operation activities (transportation and handling) at surface encapsulation facilities must be ensured.
In order to evaluate the SNF response under normal or accident scenarios, in different stages of the back-end of the nuclear fuel cycle, it is essential to assess the mechanical properties of fuel rods, as well as their evolution during and after irradiation. Several previous studies have identified mechanisms that could affect such properties during dry-storage. However, data generated from experiments with irradiated SNF are extremely rare. This PhD complements previous studies by generating new data for the structural performance of SNF, in areas where technical gaps have been identified. In particular, the mechanical properties of SNF as function of burnup have been investigated both experimentally and numerically, under quasi-static and dynamic bending loads.
The experimental activities were conducted at the hot-cell facilities of JRC-Karlsruhe and include three-point bending and gravitational impact tests performed at room temperature on pressurized samples from commercially used PWR UO2 rods. Their dynamic response was studied by successfully applying a new Image Analysis methodology developed in this work. Preliminary tests were conducted on surrogate rodlets consisting of fresh and hydrogenated Zircaloy-4 cladding tubes filled with alumina pellets. The results showed that the sampleâ s ductility decreases with increasing hydrogen concentration in the cladding.
Five SNF rods, with a wide range of burnup, were selected for the tests on irradiated fuel. High burnup samples showed higher toughness due to irradiation damage on the cladding, and less ductile behavior. Overall, strain rate had marginal influence on ductility for high-burnup samples, whereas for low burnup ones, ductility decreased significantly with strain rate. The flexural mechanical properties of the claddings as function of burnup were derived from the 3-point bending tests using the Euler-Bernoulli beam theory for beam elements with hollow-circular profile. A series of post-test examinations concerning the rod failure processes showed that fuel release in case of rod fracture is very limited, amounting to less than the mass of one pellet.
Finite Element Analysis was performed in order to simulate the experiments and the rodsâ bending behavior under the examined load conditions. An extensive sensitivity analysis was performed to minimize modeling uncertainties and the final model was calibrated against experimental data from the 3-point bending tests. Good agreement was observed between numerically predicted and experimentally derived mechanical properties, and the model can be used to simulate the mechanical behavior of SNF rods under different loading configurations.LR
Numerical simulation of spent fuel segments under transport loads
No full textPackages for the transport of spent nuclear fuel shall meet the International Atomic Energy Agency regulations to ensure safety under different Transport conditions. The physical state of spent fuel and the fuel rod cladding as well as the geometric configuration of fuel assemblies are important inputs for the evaluation of package capabilities under these conditions. Generally, the mechanical behavior of high burn-up spent fuel assemblies under Transport conditions shall be analyzed with regard to the assumptions which are used in the containment and criticality safety analysis. In view of the complexity of the interactions between the fuel rods as well as between the fuel assemblies, basket, and cask containment, the exact mechanical analysis of such phenomena is nearly impossible. The gaps in information concerning the material properties of cladding and pellet behavior, especially for the high burn-up fuel, make the analysis more complicated additionally. As a result, enveloping analytical approaches are usually used by BAM within the safety assessment of packages approved for transport of spent nuclear fuel.
To justify the safety margins of such approaches additional analyses are necessary. In this paper, numerical simulations of a segment of a spent fuel assembly are presented. The segment modeled represents the part of a generalized BWR fuel assembly between two spacers. Explicit dynamic finite element calculations are performed to simulate the spent fuel behavior under regulatory defined accident conditions of transport. A beam element formulation is used for the modeling of the fuel rods representing the compound consisting of claddings and fuel pellets. The load applied is gathered from experimental drop tests with spent fuel casks performed at BAM. A hot cell bending test performed at JRC Karlsruhe is the basis for obtaining the material behavior of the fuel rods. The material properties are determined by
simulating the test setup of JRC and optimizing the results to fit the
experimental load deflection curve. The simulations of the fuel Assembly segment are used to get a better understanding about the loads on fuel rods under accident conditions of transport
Logistics optimization code for spent fuel assembly loading into final disposal canisters
No full textOptimization and calibration of the X-ray Fluorescence (XRF) arrangement, for the analysis of environmental samples
No full text267 σ.Η παρούσα Διπλωματική Εργασία ασχολείται με τη μελέτη της τεχνικής φθορισμού των ακτίνων-Χ (X-Ray Fluorescence analysis-XRF), που χρησιμοποιείται ευρέως στο Εργαστήριο του τομέα Πυρηνικής Τεχνολογίας του Εθνικού Μετσόβιου Πολυτεχνείου (ΕΠΤ-ΕΜΠ). Στόχος αυτής της ΔΕ είναι η αναβάθμιση και βαθμονόμηση της διάταξης φθορισμού των ακτίνων-Χ για ανάλυση δειγμάτων περιβαλλοντικής σημασίας, μέσω πειραματικών τεχνικών και τεχνικών προσομοίωσης Monte Carlo PENELOPE. Οι αναλύσεις των φασμάτων φθορισμού των ακτίνων-Χ, πραγματοποιήθηκαν με το λογιστικό πρόγραμμα QXAS (Quantitative X-ray Analysis System). Βάση της νέας διαδικασίας αναλύσεων και ποσοτικού προσδιορισμού, υπήρξε συμμετοχή σε Άσκηση Διαβαθμονόμησης του ΔΟΑΕ (Διεθνής Οργανισμός Ατομικής Ενέργειας), τα αποτελέσματα της οποίας παρουσιάζονται στην εν λόγω Διπλωματική Εργασία.This diploma thesis, deals with the research of X-Ray fluorescence (XRF) Analysis, which is widely being used in the Department of Nuclear Engineering of the National Technical University of Athens (NED-NTUA). The aim of this thesis is the optimization and the calibration of the XRF arrangement, for the analysis of environmental samples with the use of experimental techniques and Monte Carlo simulations (PENELOPE). The analysis of the XRF spectrum had been done with the quantitative X-ray software (QXAS) package. Through the new analysis and quantitative method which has been adopted, the NED-NTUA participated in a Worldwide Open Proficiency Test for X-Ray Fluorescence Laboratories, offered by the IAEA (International Atomic Energy Agency). The results of this Intercomparison Test, are presented in this diploma thesis.Ευστάθιος Π. Βλασσόπουλο
Toward Reanalysis of the Tight-Pitch HCLWR-PROTEUS Phase II Experiments
No full textThe HCLWR-Proteus Phase II experiments were conducted from 1985 to 1990 in the zero-power reactor Proteus at PSI in Switzerland. The experimental program was dedicated to the physics of high conversion light water reactors and in particular to the measurement of reactor parameters such as reaction rate traverses, spectral indices, absorber reactivity worths and void coefficients. The HCLWR experiments are especially interesting because they generated knowledge in the epithermal range of the neutron flux spectrum, for which little integral experimental data is available.
In an effort to assess the interest of this experimental data to validate modern nuclear data and improve their uncertainties, a preliminary re-analysis of selected configurations was conducted with Monte-Carlo codes (MCNP6/SERPENT2) and modern nuclear data libraries (ENDF/B-VII.0, JEFF-3.1.1 and JENDL-4.0). The spectral ndices, flux spectra and sensitivity coefficients on k∞ were calculated using cell models representative of the tight-pitch measurement configurations containing 11% PuO2-UO2 fuel rods in different moderation conditions (air, water and dowtherm).
Spectral index predictions using the three nuclear data libraries agreed within two standard deviations with the measured values. The only exception is the Pu-242-capture-to-Pu-239-fission ratio, which was overestimated with all libraries by more than four standard deviations, i.e. 13%, in the non-moderated configuration. In this configuration, Pu-242 captures are few since the flux spectrum in the Pu-242 capture resonance region (between 1eV and 1keV) is small making this spectral index hard to measure. Sensitivity coefficient predictions with both MCNP6 and SERPENT2 were in good agreement
Toward Reanalysis of the Tight-Pitch HCLWR-PROTEUS Phase II Experiments
No full textThe HCLWR-Proteus Phase II experiments were conducted from 1985 to 1990 in the zero-power reactor Proteus at PSI in Switzerland. The experimental program was dedicated to the physics of high conversion light water reactors and in particular to the measurement of reactor parameters such as reaction rate traverses, spectral indices, absorber reactivity worths and void coefficients. The HCLWR experiments are especially interesting because they generated knowledge in the epithermal range of the neutron flux spectrum, for which little integral experimental data is available.
In an effort to assess the interest of this experimental data to validate modern nuclear data and improve their uncertainties, a preliminary re-analysis of selected configurations was conducted with Monte-Carlo codes (MCNP6/SERPENT2) and modern nuclear data libraries (ENDF/B-VII.0, JEFF-3.1.1 and JENDL-4.0). The spectral ndices, flux spectra and sensitivity coefficients on k∞ were calculated using cell models representative of the tight-pitch measurement configurations containing 11% PuO2-UO2 fuel rods in different moderation conditions (air, water and dowtherm).
Spectral index predictions using the three nuclear data libraries agreed within two standard deviations with the measured values. The only exception is the Pu-242-capture-to-Pu-239-fission ratio, which was overestimated with all libraries by more than four standard deviations, i.e. 13%, in the non-moderated configuration. In this configuration, Pu-242 captures are few since the flux spectrum in the Pu-242 capture resonance region (between 1eV and 1keV) is small making this spectral index hard to measure. Sensitivity coefficient predictions with both MCNP6 and SERPENT2 were in good agreement
Mechanical integrity studies on spent nuclear fuel rods
No full textThe consequences of potential accidents causing spent fuel rod failure may involve fuel particles release and dispersion. This paper presents recent results from spent fuel experimental studies performed at JRC-Karlsruhe addressing handling/transportation and long-term storage issues. An impact test using a hammer drop device in hot cell was performed on a spent fuel segment from a UO2 PWR rod with a burnup of ~67 GWd/tHM. The segment was not defueled and was repressurized to 40 bar before the test. Similarly to what observed in previous impact tests, only the fuel volume directly affected by the rod fracturing was released. In addition to the fuel material released during the impact, neither further particles release nor "flow-out" type of behaviour was observed by further tapping on the fractured segments after the test. Preliminary particle size distribution analysis of the fuel particles deposited on a second stage filter of the testing chamber collecting particles with size ≤8 m indicates a log-normal distribution with main particle size of 2.4 m and standard deviation of 1.1 m. A few sub-micron particles were detected. The detailed analysis of the results, including finer particle fractions, is still ongoing. The final goal of these investigations is to determine criteria and conditions governing the response of spent fuel rods to impact loads and other thermo-mechanical solicitations corresponding to normal and off-normal conditions that may be experienced by the rod during handling, transportation, storage and after extended storage. In addition to impact and other mechanical loading tests, property measurements as a function of accumulated radiation damage are performed on spent fuel and alpha-doped analogues to determine the long term evolution and the potential effects of ageing processes on the mechanical integrity of the spent fuel rod.JRC.G.III - Nuclear Decommissioning Department (Karlsruhe
Numerical Simulation of Spent Fuel Segments under Transport Loads
No full textPackages for the transport of spent nuclear fuel shall meet the International Atomic Energy Agency regulations to ensure safety under different transport conditions. The physical state of spent fuel and the fuel rod cladding as well as the geometric configuration of fuel assemblies are important inputs for the evaluation of package capabilities under these conditions. Generally, the mechanical behavior of high burn-up spent fuel assemblies under transport conditions shall be analyzed with regard to the assumptions which are used in the containment and criticality safety analysis. Considering the complexity of the interactions between the fuel rods as well as between the fuel assemblies, basket, and cask containment, the exact mechanical analysis of such phenomena is nearly impossible. The gaps in Information concerning the material properties of cladding and pellet behavior, especially for the high burn-up fuel, make the analysis more complicated additionally. As a result, enveloping analytical approaches are usually used by BAM within the safety assessment of packages approved for transport of spent nuclear fuel.
To justify the safety margins of such approaches additional analyses are necessary. In this paper, numerical simulations of a spent fuel assembly Segment are presented. The segment modeled represents the part of a generalized BWR fuel assembly between two spacers.
Dynamic and quasi-static finite element calculations are performed to simulate the spent fuel behavior under regulatory defined accident conditions of transport. Beam elements are used for the modeling of the fuel rods representing the compound consisting of claddings and fuel pellets. The dynamic load applied is gathered from an experimental drop test with a spent fuel cask performed at BAM. A hot cell bending test performed at JRC Karlsruhe is the basis for obtaining the material behavior of the fuel rods. The material properties are determined by simulating the test setup of JRC and optimizing the results to fit the experimental load deflection curve. The simulations of the fuel assembly segment are used to get a better understanding about the loads on fuel rods under accident conditions of transport
Investigation of nuclide inventory of cladding material irradiated in the Goesgen PWR core
No full textA characterization of the spent nuclear fuel (SNF) for its radionuclide (RN) inventory is vital for various back-end stages of the nuclear fuel cycle. It concerns both the fuel and the metallic (i.e., cladding and structural material) components of the spent fuel assemblies, where different calculation approaches and methods should be deployed for their characterization. This study concentrates on fuel traces and other impurities within the cladding. During the operating cycles, the Zircaloy cladding is exposed to a considerable amount of irradiation. The impact of the exposure should be checked to assure the integrity of the cladding and thus the safety of the stored spent fuel. Within the work package “Spent Nuclear Fuel Characterization and Evolution until Disposal” (SFC) of the EURAD project, dedicated samples were produced, irradiated and the radionuclide inventory of the cladding was analysed and compared. In parallel a blind test was performed, in which different partners used different codes to simulate the irradiation quantity. The blind test showed good agreement between most of the codes, in particular in view of the small amount of the evolved fuel traces. Furthermore, the presence of actinides, caused by precipitation of uranium on the inner surface of the cladding during manufacturing, was found to be negligible in comparison to precipitation of traces of fuel pellets on the cladding during operation. The good agreement between the simulating codes enables to depict further the initial amount of alloying elements of the cladding material itself in a better manner. In particular specific isotopes of cobalt, nickel and iron, which are directly connected to the unique properties of each cladding material can be better identified based on the accurate measuring techniques used in this study.JRC.G.II.6 - Nuclear Data and Measurement Standard
