•Thorium-Transuranic fuel deployment in a LW-SMR has been investigated.•Neutronics, fuel cycle and kinetics parameters of a LW-SMR have been analyzed.•Sub-channel analysis for Thermal-hydraulics ...parameters of Th/TRU fuel rods has been conducted.•Transmute the current spent fuels and simultaneously using the Thorium fuel to burn in reactor and breed U-233.
One of the main challenges faced by nuclear power plants is the scarcity of U-235, the fissile isotope used in conventional nuclear reactors. While there is enough U-235 in the world to fuel current reactors, the demand for energy is increasing rapidly, especially in emerging economies like China and India, and this could lead to shortages in the future. One potential solution to these challenges is to use a mixture of Thorium and Transuranic (TRU) elements. TRU elements are produced in spent fuel from conventional Pressurized Water Reactors (PWRs) and pose a long-term storage challenge due to their long half-lives and radiotoxicity. However, they can be used as a fuel option for energy production. Additionally, Thorium is much more abundant than Uranium, and it is estimated that there is enough Thorium in the world to supply nuclear power for thousands of years. Combining Thorium and TRU elements in Small Modular Reactors (SMRs) can therefore offer a sustainable and cost-effective solution to the challenge of nuclear fuel. By transmuting TRU elements, the long-term storage challenge of spent fuel can be reduced, while the use of Thorium as a fuel can significantly reduce the demand for Uranium. This would not only provide a more sustainable source of energy but also help to address concerns over the scarcity and environmental impact of conventional nuclear energy. In this regard, current study investigates the potential application of Thorium/TRU fuel in near deployment Light water SMRs (LW-SMRs). The world Thorium resources and projects and also the reprocessing methods of U-233 and Transuranic isotopes have been reviewed comprehensively. Then the neutronics, fuel cycle, kinetics and thermal hydraulic analysis of conceptual LW-SMR loaded by TRU and Thorium fuels have been conducted and the results have been compared with reference SMR data.
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•The VERA benchmark problem #2 is developed using the Condor v2.8.05 cell-code.•Results are compared with reference ones from benchmark in terms of reactivity and pin power distributions.•A good ...agreement with benchmark results, assessing the capabilities of Condor to model highly heterogeneous modern PWR fuel assemblies.•Additional comparisons with Serpent v2.32 models are provided.•In-depth analysis of diverse key aspects of cell level models is developed and valuable insights are provided.
Condor is a modern and versatile code that applies multi-group collision probabilities with heterogeneous response coupling to solve cell-level neutronic calculations. Its suitability to handle realistic PWR calculations is in this work analyzed using the VERA numerical benchmark problem #2. It consists of a series of two-dimensional lattices representative of a modern PWR, aimed to address the code’s capabilities to capture the spatial and spectral distributions. Results using Condor v2.8.05b code are obtained and compared with reference values in terms of system reactivities and power distributions. Further investigations are also provided, such as the impact of diverse Nuclear Data Libraries, the use of problem’s symmetries and the effect of integration and inter-cell coupling parameters. Moreover, comparisons with analogous models using Serpent v2.1.32 Monte Carlo particle transport code are also included, which allows to contrast both the computational performance and the overall accuracy of the Condor code.
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•A different absolute calibration procedure for the 3He detector at ASDEX Upgrade.•Reproducible geometry and better count statistics.•A detailed Monte Carlo simulation of the calibration using the ...Serpent code.•Varying discrepancy factor with components and source position.•The factor is sensitive to moderator thickness and neutron scattering.
The neutron production in ASDEX Upgrade (AUG) neutral beam injection (NBI) heated discharges is dominated by beam-target fusion reactions. Hence, the neutron rate (NR) and energy distributions are footprints of the fast ion distribution. This motivates to establish a reliable neutron rate calibration. Comparisons at AUG between the experimental NR and the one predicted by the TRANSP code show systematic variations from campaign to campaign. Potential reason for this is the delicate absolute calibration of the neutron detectors. Therefore, a different calibration technique was performed, enabling longer acquisition time, uniform geometry, better statistics and thus less uncertainty. A toy train carrying a radioactive source (238Pu/B) over two radial positions on the equatorial plane shows a periodical NR on the epithermal 3He neutron detector. The calibration results are compared to a neutron transport simulation using the Monte Carlo (MC) code Serpent. Preliminary comparisons for one source position on the outer railway track show a discrepancy factor of about 130 in the position of least material inside the simulation, in the direct line of sight to the detector. For a better understanding of these results, two additional measurements were performed. The results were again compared to a detailed Serpent simulation. This paper describes the calibration set-up for the neutron measurements in AUG, provides a brief simulation background on reaction rate estimations and a survey on the comparison between the measured and calculated neutron rates.
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The TEPLATOR is a new type of nuclear reactor which the main purpose is producing heat for district heating. It is designed as a special thermal reactor with 55 fuel channels for fuel assemblies, ...which is moderated and cooled by heavy water and operated around atmospheric pressure. The TEPLATOR DEMO is designed for the use of irradiated fuel from PWR or BWR reactors. Using heavy water as the moderator and coolant in this reactor concept allows to use natural uranium as an alternative fuel in case that the irradiated fuel is not available for some reason. This solution is suitable because of the price of natural uranium and the absence of costly fuel enrichment. This article is focused on deeper analyses of alternative suitable fuel for TEPLATOR based on natural uranium and new fuel geometries. This work builds on previous research on alternative fuel material and geometry for the TEPLATOR. It is mainly concerned with the neutronic development of fuel assemblies, the possibility of manufacturing of developed fuel types, and optimization of fuel management and uranium consumption. This article contains predetermined candidates for suitable fuel geometries and new untested fuel geometry types with some new advantages. Finally, optimization of the whole reactor core and number of fuel channels was made in terms of increased safety and higher fuel burn-up. Presented calculations were performed by Monte Carlo code Seprent.
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This work aims to introduce a new neutrons source technology for research and medical applications, by the conceptualization of the Micro Research Reactor cooled by Heat Pipes (MRR-HP). This is ...achieved by developing a 3D-detailed neutronics model using the Monte-Carlo Serpent-2.1.31 code, and investigating the neutronics behavior and define the design parameters of the MRR-HP. The MRR-HP is a cylindrical monolithic block (Magnesium-Oxide) that includes; fuel rods (Uranium-Nitride), Dummy Rods (Beryllium-Oxide), Water and Potassium heat pipes for cooling. The monolithic block is surrounded by reflector (Beryllium-Oxide), where moveable reflective bent plates and a cylindrical block are inserted for control/shutdown purposes. The MRR-HP features: 1- Fast neutron spectrum. 2- Compact (1 m diameter, 1.73 m length, mass: 1.982 tons), which permits a high degree of portability and establishment in small facilities. 3- Passive cooling. 4- Operating without refueling/maintenance for ~12.5 years. 5- High-Assay-Low-Enriched-Uranium = 19.75%. 6- Neutron flux (
). 7- Inherently safe. Based on the obtained results of the neutron flux and spectrum, the MRR-HP is a multi-purpose reactor that has the potential to perform: Boron Neutron Capture Therapy, Neutrons activation analysis, Testing and Calibration, Transmutation, Neutron Imaging, Education and Training, Prompt Gamma Neutrons activation analysis, Isotope Production, Fast Shielding.
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•A full 3-D model for the OPAL Research Reactor using Serpent 2 MC Code is developed.•First six operation cycles were calculated without the aim of other calculation codes.•Very good agreement with ...experimental results was found.•Code Performance and scalability analysis was carried out.•Full 3-D Serpent 2 models for Research Reactors arise as a viable approach.
Monte Carlo (MC) neutron transport codes have been extensively used for more than three decades to perform criticality calculations and to solve shielding problems due to their capability to model complex systems without major approximations. The irruption of affordable low-cost high performance computer resources in the last decade allows to consider some initially unexpected applications, such as full core burnup calculations or cell level modeling for few group parameters calculations. In this work the concern of the potential use of Monte Carlo codes to perform full 3-D calculations including burnup for a state of art Research Reactor is analyzed, regarding aspects related to accuracy, performance and resources requirements. For such purpose Serpent 2 v.1.24 Code, developed by VTT Technical Research Centre of Finland is used for full core burnup calculations of the 20MWth OPAL Research Reactor. This code is the second version of a brand-new Monte Carlo code designed to perform burn dependent cell-level and full 3-D core calculations using optimized schemes to diminish the computational effort. In past works the first version of Serpent Code was tested as a cell-level-code to model the MTR-type fuel assemblies from OPAL Research Reactor, obtaining fairly good results. Further works were developed for full 3-D models, where several parameters such as critical configurations, in-core thermal neutron flux profiles and effective delayed neutron parameters were obtained and compared to experimental data and other codes results, showing a very good performance. In the present work, a full 3-D model is developed using specifications and high quality experimental data from IAEA Technical Report Series. This model is used to perform full-core 3-D calculations including burnup and refueling for the first six operating cycles without the aim of an external calculation code. To perform such task, an ad hoc code to manipulate Serpent 2 restart files was developed in order to model the overall full core burnup problem without the help of any other calculation code. The results were compared with the experimental data available showing a very good agreement. Finally several aspects of the computational issues related with this modeling such performance, scalability and resources requirements are discussed, showing that the use of full core 3-D MC models including burnup for small cores represents nowadays a feasible alternative for specific calculations.
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The use of a new Monte Carlo Serpent code for the calculation of water-cooled
reactors is presented and a calculation scheme of the fuel assembly for
VVER-1000 reactors developed. The calculation of ...neutron-physical
characteristics for the fuel assembly of VVER-1000 is carried out for
different states and the results obtained by the Serpent model compared with
the results of other reactor codes. The analyses of these results are
presented in the paper submitted here. Based on this article, the Monte
Carlo Serpent code could be used for neutron-physical calculations of
VVER-1000 reactors.
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Description of computation model for Kyoto University Critical Assembly (KUCA) developed with the help of Monte-Carlo Serpent code was presented in this paper. The simulation of criticality and ...subcriticality condi-tions for KUCA was carried out. The effective multiplication factors for different critical experiments were calcu-lated for KUCA. The presented obtained results were considered and compared with the experimental data and with computation results from other Monte Carlo codes.
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