The European Journal of Physics N (EPJ-N)
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Thermal decomposition analysis of simulated high-level liquid waste in cold-cap
No full textThe cold cap floating on top of the molten glass pool in liquid fed joule-heated ceramic melter plays an important role for operation of the vitrification process. A series of such phenomena as evaporation, melting and thermal decomposition of HLLW (high-level liquid waste) takes place within the cold-cap. An understanding of the varied thermal decomposition behavior of various nitrates constituting HLLW is necessary to elucidate a series of phenomena occurring within the cold-cap. In this study, reaction rates of the thermal decomposition reaction of 13 kinds of nitrates, which are main constituents of simulated HLLW (sHLLW), were investigated using thermogravimetrical instrument in a range of room temperature to 1000 °C. The reaction rates of the thermal decompositions of 13 kinds of nitrates were depicted according to composition ratio (wt%) of each nitrate in sHLLW. It was found that the thermal decomposition of sHLLW could be predicted by the reaction rates and reaction temperatures of individual nitrates. The thermal decomposition of sHLLW with borosilicate glass system was also investigated. The above mentioned results will be able to provide a useful knowledge for understanding the phenomena occurring within the cold-cap
Measurements of the effective cumulative fission yields of
No full textThe effective Neodymium cumulative fission yields for 235U have been measured in the fast reactor PHENIX relatively to the 235U fission cross-section. The data were derived from isotope-ratio measurements obtained in the frame of the PROFIL-1, PROFIL-2A and PROFIL-2B programs. The interpretations of the experimental programs were performed with the ERANOS code in association with the Joint Evaluated Fission and Fusion library JEFF-3.1.1. Final results for 143Nd, 145Nd, 146Nd, 148Nd and 150Nd were 5.61%, 3.70%, 2.83%, 1.64% and 0.66%, respectively. The relative uncertainties attached to each of the cumulative fission yields lie between 2.1% and 2.4%. The main source of uncertainty is due to the fluence scaling procedure (<2%). The uncertainties on the Neodymium capture cross-sections provide a contribution lower than 1%. The energy dependence of the fission yields was studied with the GEF code from the thermal energy to 20 MeV. Neutron spectrum average corrections, deduced from GEF calculations, were applied to our effective fission yields with the aim of estimating fission yields at 400 keV and 500 keV, as given in the International Evaluated Nuclear Data Files (JEFF, ENDF/B and JENDL). The neutron spectrum average correction calculated for the PROFIL results remains lower than 1.5%
Estimation of the radionuclide inventory in LWR spent fuel assembly structural materials for long-term safety analysis
No full textThe radionuclide inventory of materials irradiated in a reactor depends on the initial material composition, irradiation history and on the magnitude and spectrum of the neutron flux. The material composition of a fuel assembly structure includes various alloys of Zircaloy, Inconel and stainless steel. The existing impurities in these materials are very important for accurate determination of the activation of all nuclides with a view to assessing the radiological consequences of their geological disposal. In fact, the safety assessments of geological repositories require the average and maximum (in the sense of very conservative) inventories of the very long-lived nuclides as input. The purpose of the present work is to describe the methodology applied for determining the activation of these nuclides in fuel assembly structural materials by means of coupled depletion/activation calculations and also to crosscheck the results obtained from two approaches. UO2 and MOX PWR fuels have been simulated using SCALE/TRITON, simultaneously irradiating the fuel region in POWER mode and the cladding region in FLUX mode and aiming to produce binary macro cross-section libraries by applying accurate local neutron spectra in the cladding region as a function of irradiation history that are suitable for activation calculations. The developed activation libraries have been re-employed in a second run using the ORIGEN-S program for a dedicated activation calculation. The axial variation of the neutron flux along the fuel assembly length has also been considered. The SCALE calculations were performed using a 238-group cross-section library, according to the ENDF/B-VII. The results obtained with the ORIGEN-S activation calculations are compared with the results obtained from TRITON via direct irradiation of the cladding, as allowed by the FLUX mode. It is shown that an agreement on the total calculated activities can be found within 55% for MOX and within 22% for UO2, whereas the latter is reduced to 9% when more accurate irradiation data are used (core-follow flux data instead of life-average flux data)
Selection of a tool to decision making for site selection for high level waste
No full textThe aim of this paper is to create a panel comparing some of the key decision-making support tools used in situations with the characteristics of the problem of selecting suitable areas for constructing a final deep geologic repository. The tools addressed in this work are also well known and with easy implementation. The decision-making process in matters of this kind is, in general, complex due to its multicriteria nature and the conflicting opinions of various stakeholders. Thus, a comprehensive study was performed with the literature in this subject, specifically in documents of the International Atomic Energy Agency (IAEA), regarding the importance of the criteria involved in the decision-making process. Therefore, we highlighted six judgment attributes for selecting a decision support tool, suitable for the problem. For this study, we have selected the following multicriteria tools: AHP, Delphi, Brainstorm, Nominal Group Technique and AHP-Delphi. Finally, the AHP-Delphi method has demonstrated to be more appropriate for managing the inherent multiple attributes to the problem proposed
Modelling of powder die compaction for press cycle optimization
No full textA new electromechanical press for fuel pellet manufacturing was built last year in partnership between CEA-Marcoule and ChampalleAlcen. This press was developed to shape pellets in a hot cell via remote handling. It has been qualified to show its robustness and to optimize the compaction cycle, thus obtaining a better sintered pellet profile and limiting damage. We will show you how 400 annular pellets have been produced with good geometry's parameters, based on press settings management. These results are according to a good phenomenological pressing knowledge with Finite Element Modeling calculation. Therefore, during die pressing, a modification in the punch displacement sequence induces fluctuation in the axial distribution of frictional forces. The green pellet stress and density gradients are based on these frictional forces between powder and tool, and between grains in the powder, influencing the shape of the pellet after sintering. The pellet shape and diameter tolerances must be minimized to avoid the need for grinding operations. To find the best parameters for the press settings, which enable optimization, FEM calculations were used and different compaction models compared to give the best calculation/physical trial comparisons. These simulations were then used to predict the impact of different parameters when there is a change in the type of powder and the pellet size, or when the behavior of the press changes during the compaction time. In 2016, it is planned to set up the press in a glove box for UO2 manufacturing qualification based on our simulation methodology, before actual hot cell trials in the future
Assessment of stainless steel 348 fuel rod performance against literature available data using TRANSURANUS code
No full textEarly pressurized water reactors were originally designed to operate using stainless steel as cladding material, but during their lifetime this material was replaced by zirconium-based alloys. However, after the Fukushima Daiichi accident, the problems related to the zirconium-based alloys due to the hydrogen production and explosion under severe accident brought the importance to assess different materials. In this sense, initiatives as ATF (Accident Tolerant Fuel) program are considering different material as fuel cladding and, one candidate is iron-based alloy. In order to assess the fuel performance of fuel rods manufactured using iron-based alloy as cladding material, it was necessary to select a specific stainless steel (type 348) and modify properly conventional fuel performance codes developed in the last decades. Then, 348 stainless steel mechanical and physics properties were introduced in the TRANSURANUS code. The aim of this paper is to present the obtained results concerning the verification of the modified TRANSURANUS code version against data collected from the open literature, related to reactors which operated using stainless steel as cladding. Considering that some data were not available, some assumptions had to be made. Important differences related to the conventional fuel rods were taken into account. Obtained results regarding the cladding behavior are in agreement with available information. This constitutes an evidence of the modified TRANSURANUS code capabilities to perform fuel rod investigation of fuel rods manufactured using 348 stainless steel as cladding material
Evaluations of Mo-alloy for light water reactor fuel cladding to enhance accident tolerance
No full textMolybdenum based alloy is selected as a candidate to enhance tolerance of fuel to severe loss of coolant accidents due to its high melting temperature of ∼2600 °C and ability to maintain sufficient mechanical strength at temperatures exceeding 1200 °C. An outer layer of either a Zr-alloy or Al-containing stainless steel is designed to provide corrosion resistance under normal operation and oxidation resistance in steam exceeding 1000 °C for 24 hours under severe loss of coolant accidents. Due to its higher neutron absorption cross-sections, the Mo-alloy cladding is designed to be less than half the thickness of the current Zr-alloy cladding. A feasibility study has been undertaken to demonstrate (1) fabricability of long, thin wall Mo-alloy tubes, (2) formability of a protective outer coating, (3) weldability of Mo tube to endcaps, (4) corrosion resistance in autoclaves with simulated LWR coolant, (5) oxidation resistance to steam at 1000–1500 °C, and (6) sufficient axial and diametral strength and ductility. High purity Mo as well as Mo + La2O3 ODS alloy have been successfully fabricated into ∼2-meter long tubes for the feasibility study. Preliminary results are encouraging, and hence rodlets with Mo-alloy cladding containing fuel pellets have been under preparation for irradiation at the Advanced Test Reactor (ATR) in Idaho National Laboratory. Additional efforts are underway to enhance the Mo cladding mechanical properties via process optimization. Oxidation tests to temperatures up to 1500 °C, and burst and creep tests up to 1000 °C are also underway. In addition, some Mo disks in close contact with UO2 from a previous irradiation program (to >100 GWd/MTU) at the Halden Reactor have been subjected to post-irradiation examination to evaluate the chemical compatibility of Mo with irradiated UO2 and fission products. This paper will provide an update on results from the feasibility study and discuss the attributes of the coated Mo cladding design to meet the challenging requirements for improving fuel tolerance to severe loss of coolant accidents
An attempt for a unified description of mechanical testing on Zircaloy-4 cladding subjected to simulated LOCA transient
No full textDuring a Loss Of Coolant Accident (LOCA), an important safety requirement is that the reflooding of the core by the emergency core cooling system should not lead to a complete rupture of the fuel rods. Several types of mechanical tests are usually performed in the industry to determine the degree of cladding embrittlement, such as ring compression tests or four-point bending of rodlets. Many other tests can be found in the open literature. However, there is presently no real intrinsic understanding of the failure conditions in these tests which would allow translation of the results from one kind of mechanical testing to another. The present study is an attempt to provide a unified description of the failure not directly depending on the tested geometry. This effort aims at providing a better understanding of the link between several existing safety criteria relying on very different mechanical testing. To achieve this objective, the failure mechanisms of pre-oxidized and pre-hydrided cladding samples are characterized by comparing the behavior of two different mechanical tests: Axial Tensile (AT) test and “C”-shaped Ring Compression Test (CCT). The failure of samples in both cases can be described by usual linear elastic fracture mechanics theory. Using interrupted mechanical tests, metallographic examinations have evidenced that a set of parallel cracks are nucleated at the inner and outer surface of the samples just before failure, crossing both the oxide layer and the oxygen rich alpha layer. The stress intensity factors for multiple crack geometry are determined for both AT and CCT samples using finite element calculations. After each mechanical test performed on high temperature steam oxidized samples, metallography is then used to individually determine the crack depth and crack spacing. Using these two important parameters and considering the applied load at fracture, the stress intensity factor at failure is derived for each tested sample. This procedure provides an assessment scheme to determine experimentally the fracture toughness of the prior-β region in the mid-wall of the oxidized samples. The obtained fracture toughness for CCT and AT samples are thus compared, confirming that the linear elastic fracture mechanics is a relevant tool to describe the strength of LOCA embrittled cladding alloys
Storage of thermal reactor fuels – Implications for the back end of the fuel cycle in the UK
No full textFuel from UK's Advanced Gas-Cooled Reactors (AGRs) is being reprocessed, however reprocessing will cease in 2018 and the strategy for fuel that has not been reprocessed is for it to be placed into wet storage until it can be consigned to a geological disposal facility in around 2080. Although reprocessing of LWR fuel has been undertaken in the UK, and this option is not precluded for current and future LWRs, all utilities planning to operate LWRs are intending to use At-Reactor storage pending geological disposal. This strategy will result in a substantial change in the management of spent fuel that could affect the back end of the fuel cycle for over a century. This paper presents potential fuel storage scenarios for two options: the current nuclear power replacement strategy, which will see 16 GWe of new capacity installed by 2030 and a median strategy, intended to ensure implementation of the UK's carbon reduction target, involving 48 GWe of nuclear capacity installed by 2040. The potential scale, distribution and timing of fuel storage and disposal operations have been assessed and changes to the current industrial activity are highlighted to indicate potential effects on public acceptance of back end activities
Eddy current testing system for bottom mounted instrumentation welds
No full textThe capability of eddy current testing (ECT) for the bottom mounted instrumentation (BMI) weld area of reactor vessel in a pressurized water reactor was demonstrated by the developed ECT system and procedure. It is difficult to position and move the probe on the BMI weld area because the area has complexly curved surfaces. The space coordinates and the normal vectors at the scanning points were calculated as the scanning trajectory of probe based on the measured results of surface shape on the BMI mock-up. The multi-axis robot was used to move the probe on the mock-up. Each motion-axis position of the robot corresponding to each scanning point was calculated by the inverse kinematic algorithm. In the mock-up test, the probe was properly contacted with most of the weld surfaces. The artificial stress corrosion cracking of approximately 6 mm in length and the electrical-discharge machining slit of 0.5 mm in length, 1 mm in depth and 0.2 mm in width given on the weld surface were detected. From the probe output voltage, it was estimated that the average probe tilt angle on the surface under scanning was 2.6°