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1
Wulff, Wolfgang. "Computer simulation of two-phase flow in nuclear reactors". Nuclear Engineering and Design 141, n.º 1-2 (junio de 1993): 303–13. http://dx.doi.org/10.1016/0029-5493(93)90108-l.
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Bakhshayesh, Moshkbar y Naser Vosoughi. "A simulation of a pebble bed reactor core by the MCNP-4C computer code". Nuclear Technology and Radiation Protection 24, n.º 3 (2009): 177–82. http://dx.doi.org/10.2298/ntrp0903177b.
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Lack of energy is a major crisis of our century; the irregular increase of fossil fuel costs has forced us to search for novel, cheaper, and safer sources of energy. Pebble bed reactors - an advanced new generation of reactors with specific advantages in safety and cost - might turn out to be the desired candidate for the role. The calculation of the critical height of a pebble bed reactor at room temperature, while using the MCNP-4C computer code, is the main goal of this paper. In order to reduce the MCNP computing time compared to the previously proposed schemes, we have devised a new simulation scheme. Different arrangements of kernels in fuel pebble simulations were investigated and the best arrangement to decrease the MCNP execution time (while keeping the accuracy of the results), chosen. The neutron flux distribution and control rods worth, as well as their shadowing effects, have also been considered in this paper. All calculations done for the HTR-10 reactor core are in good agreement with experimental results.
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Salcedo, L. L., E. Oset, M. J. Vicente-Vacas y C. Garcia-Recio. "Computer simulation of inclusive pion nuclear reactions". Nuclear Physics A 484, n.º 3-4 (julio de 1988): 557–92. http://dx.doi.org/10.1016/0375-9474(88)90310-7.
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Lee, Kim, Moon, Lim y Cho. "Heat-Absorbing Capacity of High-Heat-Flux Components in Nuclear Fusion Reactors". Energies 12, n.º 19 (3 de octubre de 2019): 3771. http://dx.doi.org/10.3390/en12193771.
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Nuclear fusion energy is a solution to the substitution of fossil fuels and the global energy deficit. However, among the several problems encountered for realizing a nuclear fusion reactor, the divertor presents difficulties due to the tremendous heat flux (~10 MW/m2) from high-temperature plasma. Also, neutrons produce additional heat (~17.5 MW/m3) from collisions with the materials’ atoms. This may lead to unexpected effects such as thermal failure. Thus, a comprehensive investigation on the divertor module is needed to determine the heat-absorbing capacity of the divertor module so to maintain the effect of incident heat flux. In this study, using an analytical approach and a simulation, the quantitative effect of heat generation on the thermophysical behavior, such as temperature and thermal stress, was analyzed while maintaining the incident heat flux. Then, a correlated equation was derived from the thermal design criteria, namely, the maximum thimble temperature and the safety factor at the vulnerable point. Finally, on the basis of the thermal design criteria, the heat-absorbing capacity of a nuclear fusion reactor in operating conditions was determined. This study contributes to the understanding of the divertor’s effects in nuclear fusion reactors for high-heat-flux and high-temperature applications.
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Okunev, V. S. "Fundamentally New Composite Materials of Fast Reactors Made on the Basis of Nanotechnology". Key Engineering Materials 887 (mayo de 2021): 159–64. http://dx.doi.org/10.4028/www.scientific.net/kem.887.159.
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The main goal of the work is to identify the advantages of fast reactors when using nanotechnology in the manufacture of core materials. The research methods are based on the adaptation of known technologies (including powder metallurgy) to the design of fast reactors and on the numerical simulation of physical processes carried out using computer programs for the analysis of emergency conditions of fast reactors (including anticipated transient without scram - ATWS). The results of the research show that the use of structural materials based on steels hardened by nanooxides in combination with fundamentally new types of fuel based on composite materials can significantly improve the safety of nuclear technics. Sintered mixtures of ceramic microgranules (oxide, nitride) and nanoadditives of metallic beryllium or uranium are considered as nuclear fuel. Such composite nuclear fuel improves reactor safety and power. The following types of composite fuel were analyzed: mixed oxide with additives of a beryllium or uranium nanopowder, mixed mononitride with additives of a beryllium or uranium nanopowder. Most preferably, a ceramic-metal pellet fuel based on mononitride microgranules and uranium metal nanopowder. The use of such fuel (with a volume fraction of metallic uranium up to 20%) significantly increases the safety of the reactor, combining the advantages of metal and ceramics and completely neutralizing their disadvantages. The proposed materials are of practical importance in the development of new concepts of nuclear technics, in the transition to large-scale nuclear power and high-power reactors. The use of a new cermet-based composite fuel increases the power of the reactor and significantly increases the safety of the reactor.
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Sadek, I. S. y R. Vedantham. "Optimal control of distributed nuclear reactors with pointwise controllers". Mathematical and Computer Modelling 32, n.º 3-4 (agosto de 2000): 341–48. http://dx.doi.org/10.1016/s0895-7177(00)00139-4.
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Khorshidi, Abdollah. "Accelerator-Based Methods in Radio-Material 99Mo/99mTc Production Alternatives by Monte Carlo Method: The Scientific-Expedient Considerations in Nuclear Medicine". Journal of Multiscale Modelling 11, n.º 01 (14 de enero de 2019): 1930001. http://dx.doi.org/10.1142/s1756973719300016.
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Some accelerator technologies are already used for commercial [Formula: see text]Mo-99mTc production, as the economic criteria are considered representative of the main differences between diverse technologies including accelerators and reactors. This study has provided a review of known and potential [Formula: see text]Mo production using conventional medical facilities. Accelerator-based method in 99mTc production via ([Formula: see text], [Formula: see text]) direct reaction on [Formula: see text]Mo was simulated using 18[Formula: see text]MeV proton beam. Meanwhile, a conceptual design for indirect [Formula: see text]Mo production via [Formula: see text]Mo([Formula: see text])[Formula: see text]Mo and [Formula: see text]Mo(n,[Formula: see text]2n)[Formula: see text]Mo reactions was investigated when an electron source of 35[Formula: see text]MeV by accelerator is used. These indirect reactions were explored via inserted [Formula: see text]Mo samples at different positions inside the lead region. Furthermore, Adiabatic Resonance Crossing (ARC) method based on proton accelerator via transmutation in [Formula: see text]Mo([Formula: see text]Mo was examined when the 30[Formula: see text]MeV proton beam is used. Saturation activity and yield were investigated using alternative proposed methods. The potential proliferation risk associated with accelerator technetium production is minimal. While accelerators could be turned into neutron sources which could in turn be used to irradiate [Formula: see text]U to breed plutonium, and centrifuges used to enrich [Formula: see text]Mo for targets could conceivably be turned to enriching uranium, this would result in very tiny global production capability particularly compared with research or power reactors. The potential of the fresh methods could provide a replacement or complement over current reactor-based supply sources in various radioisotopes production purposes.
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Gabbar, Hossam A., Muhammad R. Abdussami y Md Ibrahim Adham. "Micro Nuclear Reactors: Potential Replacements for Diesel Gensets within Micro Energy Grids". Energies 13, n.º 19 (5 de octubre de 2020): 5172. http://dx.doi.org/10.3390/en13195172.
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Resilient operation of medium/large scale off-grid energy systems, which is a key challenge for energy crisis solutions, requires continuous and sustainable energy resources. Conventionally, micro energy grids (MEGs) are adopted to supply electricity and thermal energy simultaneously. Fossil-fired gensets, such as diesel generators, are indispensable components for off-grid MEGs due to the intermittent nature of renewable energy sources (RESs). However, fossil-fired gensets emit a significant amount of greenhouse gases (GHGs). Therefore, this study investigates an alternative source as an economical and environmental replacement for diesel gensets that can reduce GHG emissions and ensure system reliability. A MEG is developed in this paper to support a considerably large-scale electric and thermal demand at Ontario Tech University (UOIT). Different sizes of diesel gensets and RESs, such as solar, wind, hydro, and biomass, are combined in the MEG for off-grid applications. To evaluate diesel gensets’ competency, the diesel genset is substituted by an emission-free generation source named microreactor (MR). The fossil-fired MEG and MR-based MEG are optimized by an intelligent optimization technique, namely particle swarm optimization (PSO). The objective of the PSO is to minimize the net present cost (NPC). The simulation results show that MR-based MEG could be an excellent replacement for a diesel genset in terms of NPC and selected key performance indicators (KPIs). A comprehensive sensitivity analysis is also carried out to validate the simulation results.
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Prošek, Andrej y Marko Matkovič. "RELAP5/MOD3.3 Analysis of the Loss of External Power Event with Safety Injection Actuation". Science and Technology of Nuclear Installations 2018 (2018): 1–14. http://dx.doi.org/10.1155/2018/6964946.
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The code assessment typically comprises basic tests cases, separate effects test, and integral effects tests. On the other hand, the thermal hydraulic system codes like RELAP5/MOD3.3 are primarily intended for simulation of transients and accidents in light water reactors. The plant measured data come mostly from startup tests and operational events. Also, for operational events the measured plant data may not be sufficient to explain all details of the event. The purpose of this study was therefore besides code assessment to demonstrate that simulations can be very beneficial for deep understanding of the plant response and further corrective measures. The abnormal event with reactor trip and safety injection signal actuation was simulated with the latest RELAP5/MOD3.3 Patch 05 best-estimate thermal hydraulic computer code. The measured and simulated data agree well considering the major plant system responses and operator actions. This suggests that the RELAP5 code simulation is good representative of the plant response and can complement not available information from plant measured data. In such a way, an event can be better understood.
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Pakari, O., V. Lamirand, B. Vandereydt, F. Vitullo, M. Hursin, C. Kong y A. Pautz. "Design and Simulation of Gamma Spectrometry Experiments in the CROCUS Reactor". EPJ Web of Conferences 225 (2020): 04016. http://dx.doi.org/10.1051/epjconf/202022504016.
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Gamma rays in nuclear reactors, arising either from fission or decay processes, significantly contribute to the heating and dose of the reactor components. Zero power research reactors offer the possibility to measure gamma rays in a purely neutronic environment, allowing for validation experiments of computed spectra, dose estimates, reactor noise and prompt to delayed gamma ratios. This data then contributes to models, code validation and photo atomic nuclear data evaluation. In order to contribute to aforementioned experimental data, gamma detection capabilities are being added to the CROCUS reactor facility. The CROCUS reactor is a two-zone, uranium-fueled light water moderated facility operated by the Laboratory for Reactor Physics and Systems Behaviour (LRS) at the Swiss Federal Institute of Technology Lausanne (EPFL). With a maximum power of 100W, it is a zero power reactor used for teaching and research, most recently for intrinsic and induced neutron noise studies. For future gamma detection applications in the CROCUS reactor, an array of four detectors - two large 5”x10” Bismuth Germanate (BGO) and two smaller Cerium Bromide (CeBr3) scintillators - was acquired. The BGO detectors are to be arbitrarily positioned in the core reflector and out of the vessel for measurements at arbitrary distances. The CeBr3 detectors on the other hand are small enough to be set in the guide tubes of the control rods for in-core measurements. We present a study of the neutron and gamma flux in the core and reflector using the MCNP 6.2 and Serpent 2 Monte Carlo codes for coupled neutron and photon transport criticality calculations. More specifically, we investigate and compare predicted spectra as well as reactivity worth of different envisioned experimental setups. We further predict pulse height spectra as well as doses to the crystals with and without cadmium shielding to estimate allowable reactor powers with respect to detector radiation hardness. The results serve as basis for calibration and aid in the design and regulatory approval of the experiments.
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Al'perovich, M. N. y L. D. Ivanov. "Computer simulation of a reactor experiment". Soviet Atomic Energy 69, n.º 4 (octubre de 1990): 861–62. http://dx.doi.org/10.1007/bf02046022.
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Khater, Hany, Talal Abu-El-Maty y El-Din El-Morshdy. "Thermal-hydraulic modeling of reactivity accidents in MTR reactors". Nuclear Technology and Radiation Protection 21, n.º 2 (2006): 21–32. http://dx.doi.org/10.2298/ntrp0602021k.
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This paper describes the development of a dynamic model for the thermal-hydraulic analysis of MTR research reactors during a reactivity insertion accident. The model is formulated for coupling reactor kinetics with feedback reactivity and reactor core thermal-hydraulics. To represent the reactor core, two types of channels are considered, average and hot channels. The developed computer program is compiled and executed on a personal computer, using the FORTRAN language. The model is validated by safety-related benchmark calculations for MTR-TYPE reactors of IAEA 10 MW generic reactor for both slow and fast reactivity insertion transients. A good agreement is shown between the present model and the benchmark calculations. Then, the model is used for simulating the uncontrolled withdrawal of a control rod of an ETRR-2 reactor in transient with over power scram trip. The model results for ETRR-2 are analyzed and discussed.
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Bousbia-Salah, Anis, Fabio Moretti y Francesco D’auria. "State-of-the-art and needs for jet instability and direct contact condensation model improvements". Nuclear Technology and Radiation Protection 22, n.º 1 (2007): 58–66. http://dx.doi.org/10.2298/ntrp0701058b.
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There is a common understanding among thermal-hydraulic experts that the system analysis codes have currently reached an acceptable degree of maturity. Reliable application, however, is still limited to the validated domain. There is a growing need for qualified codes in assessing the safety of the existing reactors and for developing advanced reactor systems. Under conditions involving multi-phase flow simulations, the use of classical methods, mainly based upon the one dimensional approach, is not appropriate at all. The use of new computational models, such as the direct numerical simulation, large-eddy simulation or other advanced computational fluid dynamics methods, seems to be more suitable for more complex events. For this purpose, the European Commission financed NURESIM Integrated Project (as a part of the FP6 programme), was adopted to provide the initial step towards a Common European Standard Software Platform for Modeling, recording and recovering computer data for nuclear reactor simulations. Some of the studies carried out at the University of Pisa within the framework of the NURESIM project are presented in this paper. They mainly concern the investigation of two critical phenomena connected with jet instabilities and direct contact condensation that occur during emergency core cooling. Through these examples, the state-of-the-art and the need for model improvements and validation against new experimental data for the sake of getting a better understanding and more accurate predictions are discussed.
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Pacheco de Carvalho, J. A. R., C. F. F. P. R. Pacheco y A. D. Reis. "Computer Simulation and Depth Profiling of Light Nuclei by Nuclear Techniques". Advanced Materials Research 107 (abril de 2010): 123–28. http://dx.doi.org/10.4028/www.scientific.net/amr.107.123.
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This article involves computer simulation and surface analysis by nuclear techniques, which are non-destructive. The “energy method of analysis” for nuclear reactions and elastic scattering is used. Energy spectra are computer simulated and compared with experimental data, giving target composition and concentration profile information. The method is successfully applied to depth profiling of 18O and 12C nuclei in thick targets through the 18O(p,α0)15N and 12C(d,p0)13C reactions, respectively. Similarly, elastic scattering of (4He)+ ions is applied to determination of concentration profiles of O and Al for a thick target containing a thin film of aluminium oxide.
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Sharma, M. P. y A. Moharana. "Simulation of turbulent mixing rate in simulated subchannels of a reactor rod bundle". Kerntechnik 86, n.º 3 (1 de junio de 2021): 210–16. http://dx.doi.org/10.1515/kern-2020-0089.
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Abstract Subchannel analysis codes are widely used for the thermal-hydraulic design of nuclear reactor rod bundle. The effectiveness of subchannel analysis codes depends on turbulent mixing between these subchannels. Turbulent mixing has no direct contribution to the axial mass flow rate through subchannel but it will cause exchange of momentum and energy between the neighboring subchannels. Thus, it is important to evaluate the turbulent mixing coefficient for reactor rod bundle as it is a significant factor in the lateral energy and momentum equation for subchannel analysis codes like COBRA IIIC, COBRA-IV and MATRA LMR-FB. With the rapid developments in computational fluid dynamics and computer performance, three-dimensional analyses of turbulent flows occurring in the nuclear rod bundle have become more prominent. Several numerical analyses have already been attempted to investigate the flow behavior in rod bundles of different reactors. Much of these are dedicated to find out the structure of turbulence in rod bundle but a few analyses has been done to evaluate the magnitude of the turbulent mixing coefficient. In view of this, CFD analyses were carried out to determine the turbulent mixing coefficient in the simulated sub-channels of the reactor rod bundle. Previous studies on the structure of turbulence reveals that it is highly anisotropic. Hence, the Reynolds Stress Model (RSM), finer mesh and near wall distance ( y + ≤ 2) is required to capture turbulent mixing phenomena. The validation of results is done by comparing with subchannel mixing experiments.
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Cohn, Charles E. "Programming for a nuclear reactor instrument simulation". SIMULATION 50, n.º 1 (enero de 1988): 25–27. http://dx.doi.org/10.1177/003754978805000104.
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Orszulik, Magdalena, Adam Fic, Tomasz Bury y Jan Składzień. "A model of hydrogen passive autocatalytic recombiner and its validation via CFD simulations". Archives of Thermodynamics 34, n.º 4 (1 de diciembre de 2013): 257–66. http://dx.doi.org/10.2478/aoter-2013-0042.
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Abstract Passive autocatalytic recombiners (PAR) is the only used method for hydrogen removal from the containment buildings in modern nuclear reactors. Numerical models of such devices, based on the CFD approach, are the subject of this paper. The models may be coupled with two types of computer codes: the lumped parameter codes, and the computational fluid dynamics codes. This work deals with 2D numerical model of PAR and its validation. Gaseous hydrogen may be generated in water nuclear reactor systems in a course of a severe accident with core overheating. Therefore, a risk of its uncontrolled combustion appears which may be destructive to the containment structure.
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Li, Hua Long, Jianlong Lin y Jerzy A. Szpunar. "Software for Simulation of Oxidation Processes". Defect and Diffusion Forum 237-240 (abril de 2005): 189–94. http://dx.doi.org/10.4028/www.scientific.net/ddf.237-240.189.
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A methodology for discrete simulation has been developed that incorporates many structural characteristics of polycrystalline material properties, such as: texture, grain boundaries, microstructure, phase composition, chemical composition, stored energy, and residual stresses. The computer models that have been developed to study oxidation processes are based on a quantitative description of the oxide and substrate structure. That description allows for the simulation of the transport of metal and oxygen ions along interfaces and bulk portions of material and the formation of oxide structure. The proposed model can help researchers and engineers to understand the physical mechanism of oxidation in order to predict material behavior and optimize material processing and properties. In this paper, the results on the simulation of the oxidation process are presented on different substrates of Zr-Nb alloys, which are used for the manufacturing the pressure tubes used in the CANDU nuclear reactors. The effects of substrate texture, microstructure, grain boundaries, and beta phase distribution on oxidation kinetics and hydrogen permeation are demonstrated.
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Chaiko, Mark A. y Michael J. Murphy. "COTTAP: A Computer Code for Simulation of Thermal Transients in Secondary Containments of Boiling Water Reactors". Nuclear Technology 94, n.º 1 (abril de 1991): 44–55. http://dx.doi.org/10.13182/nt91-a16220.
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Pacheco de Carvalho, J. A. y A. D. Reis. "Aplicaciones de simulación por ordenador, reacciones nucleares y difusión elástica al análisis de superficies de materiales". Boletín de la Sociedad Española de Cerámica y Vidrio 47, n.º 4 (30 de agosto de 2008): 252–57. http://dx.doi.org/10.3989/cyv.2008.v47.i4.186.
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Luo, Run, Shripad T. Revankar y Fuyu Zhao. "Comparative Safety Analysis of Accelerator Driven Subcritical Systems and Critical Nuclear Energy Systems". Applied Sciences 11, n.º 17 (3 de septiembre de 2021): 8179. http://dx.doi.org/10.3390/app11178179.
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The accelerator driven subcritical system (ADS) has been chosen as one of the best candidates for Generation IV nuclear energy systems which could not only produce clean energy but also incinerate nuclear waste. The transient characteristics and operation principles of ADS are significantly different from those of the critical nuclear energy system (CNES). In this work, the safety characteristics of ADS are analyzed and compared with CNES by a developed neutronics and thermal-hydraulics coupled code named ARTAP. Three typical accidents are carried out in both ADS and CNES, including reactivity insertion, loss of flow, and loss of heat sink. The comparison results show that the power and the temperatures of fuel, cladding, and coolant of the CNES reactor are much higher than those of the ADS reactor during the reactivity insertion accident, which means ADS has a better safety advantage than CNES. However, due to the subcriticality of the ADS core and its low sensitivity to negative reactivity feedback, the simulation results indicate that the inherent safety characteristics of CNES are better than those of ADS under loss of flow accident, and the protection system of ADS would be quickly activated to achieve an emergency shutdown after the accident occurs. For the loss of heat sink, it is found that the peak temperatures of the cladding in the ADS and CNES reactors are lower than the safety limit, which imply these two reactors have good safety performance against loss of heat sink accidents.
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Liu, Xing, Ryuji Mukai, Xiao Hu Deng y Dong Ying Ju. "Computer Simulation of Heat Treatment Process for Support Plate of Nuclear Reactor". Advanced Materials Research 314-316 (agosto de 2011): 380–83. http://dx.doi.org/10.4028/www.scientific.net/amr.314-316.380.
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This paper focuses on the thermal stress and deformation analysis of the support plate of a nuclear reactor during the quenching process. A 3D finite element model of the support plate is incorporated into nonlinear coupling analysis that considers temperature, stress, and deformation. To verify the effect of cooling rate on the thermal stress and deformation of the model, we applied the heat transfer coefficients of water and heat treatment oil, depending on temperature variations, into heat conduction analysis. This analytical method enables the determination of the maximum deformation and residual stresses, so that the strength of the support plate can be identified.
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Shriwise, Patrick C., John R. Tramm, Andrew Davis y Paul K. Romano. "TOWARDS CAD-BASED GEOMETRY MODELLING WITH THE RANDOM RAY METHOD". EPJ Web of Conferences 247 (2021): 03023. http://dx.doi.org/10.1051/epjconf/202124703023.
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The Advanced Random Ray Code (ARRC) is a high performance computing application capable of high-fidelity simulations of full core nuclear reactor models. ARRC leverages a recently developed stochastic method for neutron transport, known as The Random Ray Method (TRRM), which offers a variety of computational and numerical advantages as compared to existing methods. In particular, TRRM has been shown to be capable of efficient simulation of explicit three dimensional geometry representations without assumptions about axial homogeneity. To date, ARRC has utilized Constructive Solid Geometry (CSG) combined with a nested lattice geometry which works well for typical pressurized water reactors, but is not performant for the general case featuring arbitrary geometries. To facilitate simulation of arbitrarily complex geometries in ARRC efficiently, we propose performing transport directly on Computer-Aided Design (CAD) models of the geometry. In this study, we utilize the Direct-Accelerated Geometry Monte Carlo (DAGMC) toolkit which tracks particles on tessellated CAD geometries using a bounding volume hierarchy to accelerate the process, as a replacement for ARRC’s current lattice-based accelerations. Additionally, we present a method for automatically subdividing the large CAD regions in the DAGMC model into smaller mesh cells required by random ray to achieve high accuracy. We test the new DAGMC geometry implementation in ARRC on several test problems, including a 3D pincells, 3D assemblies, and an axial section of the Advanced Test Reactor. We show that DAGMC allows for simulation of complex geometries in ARRC that would otherwise not be possible using the traditional approach while maintaining solution accuracy.
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Tseng, Cheng-Min y Rudy M. Lepp. "Dynamic simulation of the SLOWPOKE-3 nuclear heating reactor". SIMULATION 44, n.º 4 (abril de 1985): 181–88. http://dx.doi.org/10.1177/003754978504400403.
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Pacheco de Carvalho, J., C. F. R. Pacheco y A. D. Reis. "Applications of Nuclear Techniques, Computer Simulation and Microscopy to Surface Analysis of Materials". Microscopy and Microanalysis 19, S4 (agosto de 2013): 133–34. http://dx.doi.org/10.1017/s1431927613001281.
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There is a wide range of surface analysis techniques which are, generally, complementary. Nuclear and non-nuclear techniques have been available. Nuclear techniques, which are non-destructive, provide for analysis over a few microns close to the surface of the sample, giving absolute values of concentrations of isotopes and elements. They have been applied in areas such as scientific, technologic, industry, arts and medicine, using MeV ion beams. Nuclear reactions permit tracing of isotopes with high sensitivities. We use ion-ion nuclear reactions, elastic scattering and the energy analysis method where, at a chosen energy of the incident ion beam, an energy spectrum is recorded of ions from nuclear events, coming from several depths in the target. Such spectra are computationally predicted, giving target composition and concentration profile information. A computer program has been developed in this context, mainly for flat targets. The non-flat target situation arises as an extension. Successful applications of the method are given using the 18O(p,α0)15N reaction and elastic scattering of (4He)+ ions. SEM and TEM are used as useful complementary techniques.Two types of samples were prepared containing thick and thin oxides, respectively. The first sample (S1) was obtained by high temperature oxidation of austenitic steel in C 18O2 gas. Weight gain measurements had given a 4.2 μm thick oxide. SEM has shown a reasonably flat oxide (Figure 1 (a)). The second sample (S2, also labelled Al/Al2O3) was obtained by anodization of high purity aluminium at 100V in an aqueous solution of ammonium citrate. An oxide thickness of 0.1370 μm was expected. TEM has given an oxide film thickness of 0.1340 μm (Figure 1 (b)). The 18O(p,α0)15N reaction at Ep=1.78 MeV and 165º was used to analyse sample S1. Figure 2 (a) shows a good computed fit to data. A 18O step concentration profile was found, corresponding to a thick 18O oxide with thickness X1=4.4 μm. Sample S2 was analysed by elastic scattering of α particles at Eα=2.0 MeV and 165º. Figure 2 (b) shows a good computed fit to data. A thin oxide film thickness of X1=0.1350 μm was found, close to the TEM value. The fit also shows a ratio of atomic densities of O and Al slightly above 1.5. The combined use of nuclear techniques, SEM and TEM microscopy has proved to be very important for surface analysis of materials. The reported results would be difficult to obtain by other techniques.Supports from University of Beira Interior and FCT (Fundação para a Ciência e a Tecnologia)/PEst-OE/FIS/UI0524/2011 (Projecto Estratégico-UI524-2011-2012) are acknowledged.
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Estienne, Magali, Muriel Fallot, Lydie Giot, Loïc Le Meur y Amanda Porta. "Recent advances in beta decay measurements". EPJ Nuclear Sciences & Technologies 4 (2018): 24. http://dx.doi.org/10.1051/epjn/2018034.
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Three observables of interest for present and future reactors depend on the β decay data of the fission products: the reactor decay heat, antineutrinos from reactors and delayed neutron emission. Concerning the decay heat, significant discrepancies still exist between summation calculations in − their two main ingredients: the decay data and the fission yields − performed using the most recent evaluated databases available. It has been recently shown that the associated uncertainties are dominated by the ones on the decay data. But the results subtantially differ taking into account or not the correlations between the fission products. So far the uncertainty propagation does not include as well systematic effects on nuclear data such as the Pandemonium effect which impacts a large number of nuclei contributing to the decay heat. The list of nuclei deserving new TAGS measurements has been updated recently in the frame of IAEA working groups. The issues listed above impact in the same way the predicted energy spectra of the antineutrinos from reactors computed with the summation method, the interest of which has been recently reinforced by the Daya Bay latest publication. Nuclear data should definitely contribute to refine and better control these calculations. Lastly, a lot of nuclear data related to delayed neutrons are missing in nuclear databases. Despite the progresses already done these last years with new measurements now requiring to be included in evaluated databases, the experimental efforts which still need to be done are significant. These different issues will be addressed here before to comment on recent experimental results and on their impacts on the quoted observables. Some perspectives will also be presented. Solving the issues listed above will require to bring together experimental, simulation, evaluation and theoretical activities.
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Li, Shiyang, Lang Zhou, Jian Yang y Qiuwang Wang. "Numerical Simulation of Flow and Heat Transfer in Structured Packed Beds with Smooth or Dimpled Spheres at Low Channel to Particle Diameter Ratio". Energies 11, n.º 4 (15 de abril de 2018): 937. http://dx.doi.org/10.3390/en11040937.
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Packed beds are widely used in catalytic reactors or nuclear reactors. Reducing the pressure drop and improving the heat transfer performance of a packed bed is a common research aim. The dimpled structure has a complex influence on the flow and heat transfer characteristics. In the present study, the flow and heat transfer characteristics in structured packed beds with smooth or dimpled spheres are numerically investigated, where two different low channel to particle diameter ratios (N = 1.00 and N = 1.15) are considered. The pressure drop and the Nusselt number are obtained. The results show that, for N = 1.00, compared with the structured packed bed with smooth spheres, the structured packed bed with dimpled spheres has a lower pressure drop and little higher Nusselt number at 1500 < ReH < 14,000, exhibiting an improved overall heat transfer performance. However, for N = 1.15, the structured packed bed with dimpled spheres shows a much higher pressure drop, which dominantly affects the overall heat transfer performance, causing it to be weaker. Comparing the different channel to particle diameter ratios, we find that different configurations can result in: (i) completely different drag reduction effect; and (ii) relatively less influence on heat transfer enhancement.
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de Carvalho, J. Pacheco, C. F. R. Pacheco y A. D. Reis. "Applications of Nuclear Techniques, Computer Simulation and Microscopy to Depth Profiling of Light Nuclei". Microscopy and Microanalysis 18, S5 (agosto de 2012): 83–84. http://dx.doi.org/10.1017/s1431927612013074.
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There is a wide range of surface analysis techniques which are, generally, complementary and provide target information for depths near the surface. Nuclear techniques, which are non-destructive, provide for analysis over a few microns close to the surface giving absolute values of concentrations of isotopes and elements. They have been applied in areas such as scientific, technologic, industry, arts and medicine, using MeV ion beams. Nuclear reactions permit tracing of isotopes with high sensitivities. We use ion-ion reactions and the energy analysis method. At a suitable energy of the incident ion beam, an energy spectrum is recorded of ions from the reaction, coming from several depths in the target. Such spectra are computationally predicted, giving target composition and concentration profile information. Elastic scattering is a particular and important case. A computer program has been developed in this context, mainly for flat targets. The non-flat target situation arises as an extension.
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Zhang, Guo Duo, Xu Hong Yang, Dong Qing Lu y Yong Xiao Liu. "Research on Pressurizer Pressure Control System Based on BP Neural Network Control of Self-Adjusted PID Parameters". Applied Mechanics and Materials 291-294 (febrero de 2013): 2416–23. http://dx.doi.org/10.4028/www.scientific.net/amm.291-294.2416.
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The pressurizer is an important device in nuclear reactor system, and the traditional PID regulator is usually used to control pressure system of pressurizer in modern reactors. However, it is difficult to get precise parameters of traditional PID controller, and the PID control method is relied on the precise mathematical model badly. And the response of PID controller is often shown by the large amount of overshoot and long setting time which are not the desired results. For such a large inertia and complex time-varying control system, the tradition PID controller can not obtain the satisfy control results. A controller based on BP neural network in this paper has a simple structure, and the parameters of PID controller can be tuned on-line by the neural network self-learning characteristics. The computer simulation experiment demonstrates that the BP neural network PID controller performs very well when compared with the tradition PID regulator in minimal overshoot and more quick response.
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Merk, Bruno, Mark Bankhead, Dzianis Litskevich, Robert Gregg, Aiden Peakman y Craig Shearer. "On a Roadmap for Future Industrial Nuclear Reactor Core Simulation in the U.K. to Support the Nuclear Renaissance". Energies 11, n.º 12 (16 de diciembre de 2018): 3509. http://dx.doi.org/10.3390/en11123509.
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The U.K. has initiated the nuclear renaissance by contracting for the first two new plants and announcing further new build projects. The U.K. government has recently started to support this development with the announcement of a national programme of nuclear innovation. The aim of this programme with respect to modelling and simulation is foreseen to fulfil the demand in education and the build-up of a reasonably qualified workforce, as well as the development and application of a new state-of-the-art software environment for improved economics and safety. This document supports the ambition to define a new approach to the structured development of nuclear reactor core simulation that is based on oversight instead of looking at detail problems and the development of single tools for these specific detail problems. It is based on studying the industrial demand to bridge the gap in technical innovation that can be derived from basic research in order to create a tailored industry solution to set the new standard for reactor core modelling and simulation for the U.K. However, finally, a technical requirements specification has to be developed alongside the strategic approach to give code developers a functional specification that they can use to develop the tools for the future. Key points for a culture change to the application of modern technologies are identified in the use of DevOps in a double-strata approach to academic and industrial code development. The document provides a novel, strategic approach to achieve the most promising final product for industry, and to identify the most important points for improvement.
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Tramm, John R., Andrew R. Siegel, Amanda L. Lund y Paul K. Romano. "A COMPARISON OF STOCHASTIC MESH CELL VOLUME COMPUTATION STRATEGIES FOR THE RANDOM RAY METHOD OF NEUTRAL PARTICLE TRANSPORT". EPJ Web of Conferences 247 (2021): 03021. http://dx.doi.org/10.1051/epjconf/202124703021.
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The random ray method is a recently developed neutron transport method that can be used to perform efficient full-core, general-purpose, high-fidelity 3D simulations of nuclear reactors. While Tramm et al. have so far documented the new random ray algorithm in several publications, one critical detail has not yet been published: how to best determine the volume of each source region (or cell) of the simulation. As the “true” analytical constructive solid geometry cell volumes are typically not known a priori they must be computed by the application at runtime, which is not straightforward in TRRM as different rays are used each power iteration such that the sampled volume of each cell also changes between iterations. In the present study, we analyze two different on-the-fly stochastic methods for computing the cell volumes and quantify their impacts on the accuracy of scalar flux estimates. We find that the “na¨ıve” stochastic volume estimator (which arises naturally from the derivation of the Method of Characteristics), is highly biased and can result in over 1,000 pcm error in eigenvalue. Conversely, we find that the “simulation averaged” estimator is unbiased and is therefore equivalent to the use of analytical cell volumes even when using a coarse ray density. Thus, the new simulation averaged method is a critical (and as yet undocumented) component of the TRRM algorithm, and is therefore vital information for those in the reactor physics community working to implement random ray solvers of their own.
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Seydaliev, M. y D. Caswell. "CORBA AND MPI-BASED “BACKBONE” FOR COUPLING ADVANCED SIMULATION TOOLS". AECL Nuclear Review 3, n.º 2 (1 de diciembre de 2014): 83–90. http://dx.doi.org/10.12943/anr.2014.00036.
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There is a growing international interest in using coupled, multidisciplinary computer simulations for a variety of purposes, including nuclear reactor safety analysis. Reactor behaviour can be modeled using a suite of computer programs simulating phenomena or predicting parameters that can be categorized into disciplines such as Thermalhydraulics, Neutronics, Fuel, Fuel Channels, Fission Product Release and Transport, Containment and Atmospheric Dispersion, and Severe Accident Analysis. Traditionally, simulations used for safety analysis individually addressed only the behaviour within a single discipline, based upon static input data from other simulation programs. The limitation of using a suite of stand-alone simulations is that phenomenological interdependencies or temporal feedback between the parameters calculated within individual simulations cannot be adequately captured. To remove this shortcoming, multiple computer simulations for different disciplines must exchange data during runtime to address these interdependencies. This article describes the concept of a new framework, which we refer to as the “Backbone,” to provide the necessary runtime exchange of data. The Backbone, currently under development at AECL for a preliminary feasibility study, is a hybrid design using features taken from the Common Object Request Broker Architecture (CORBA), a standard defined by the Object Management Group, and the Message Passing Interface (MPI), a standard developed by a group of researchers from academia and industry. Both have well-tested and efficient implementations, including some that are freely available under the GNU public licenses. The CORBA component enables individual programs written in different languages and running on different platforms within a network to exchange data with each other, thus behaving like a single application. MPI provides the process-to-process intercommunication between these programs. This paper outlines the different CORBA and MPI configurations examined to date, as well as the preliminary configuration selected for coupling 2 existing safety analysis programs used for modeling thermal–mechanical fuel behavior and fission product behavior respectively. In addition, preliminary work in hosting both the Backbone and the associated safety analysis programs in a cluster environment are discussed.
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Garcia, Humberto E. y Azim Houshyar. "Discrete Event Simulation of Fuel Transfer Strategies for Defueling a Nuclear Reactor". SIMULATION 70, n.º 2 (febrero de 1998): 104–18. http://dx.doi.org/10.1177/003754979807000203.
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Frick, Konor, Shannon Bragg-Sitton y Cristian Rabiti. "Modeling the Idaho National Laboratory Thermal-Energy Distribution System (TEDS) in the Modelica Ecosystem". Energies 13, n.º 23 (1 de diciembre de 2020): 6353. http://dx.doi.org/10.3390/en13236353.
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Integrated energy systems (IES)—including the intimate coupling between thermal generators, the grid, ancillary processes, and energy storage—are becoming increasingly pertinent to the energy grid. To facilitate a better understanding of IES, Idaho National Laboratory (INL) has developed the experimental Thermal Energy Distribution System (TEDS) to test the interoperability of nuclear reactors, energy storage, and ancillary processes in a real-world setting. This paper provides an overview of the development of TEDS within INL’s Modelica dynamic process modeling ecosystem as part of the IES initiative. The model will bridge the gap between lab-scale experimental results and desired grid-scale energy solutions. Two simulation sets were run. The first was a 5-h test simulating a facility shakedown test, putting the facility through five potential operating modes and showcasing the ability of the valving, control sensors, and component controllers to meet system demands. The second case imposed a typical summer day demand on the system from a region with mixed commercial and residential electrical needs. In this case, the generator alone could not meet peak demand but instead required the thermal-storage unit to act as a peaking unit.
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Merk, Bruno, Anna Detkina, Seddon Atkinson, Dzianis Litskevich y Gregory Cartland-Glover. "On the Dimensions Required for a Molten Salt Zero Power Reactor Operating on Chloride Salts". Applied Sciences 11, n.º 15 (21 de julio de 2021): 6673. http://dx.doi.org/10.3390/app11156673.
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Molten salt reactors have gained substantial interest in the last years due to their flexibility and their potential for simplified closed fuel cycle operation for massive expansion in low-carbon electricity production, which will be required for a future net-zero society. The importance of a zero-power reactor for the process of developing a new, innovative rector concept, such as that required for the molten salt fast reactor based on iMAGINE technology, which operates directly on spent nuclear fuel, is described here. It is based on historical developments as well as the current demand for experimental results and key factors that are relevant to the success of the next step in the development process of all innovative reactor types. In the systematic modelling and simulation of a zero-power molten salt reactor, the radius and the feedback effects are studied for a eutectic based system, while a heavy metal rich chloride-based system are studied depending on the uranium enrichment accompanied with the effects on neutron flux spectrum and spatial distribution. These results are used to support the relevant decision for the narrowing down of the configurations supported by considerations on cost and proliferation for the follow up 3-D analysis. The results provide for the first time a systematic modelling and simulation approach for a new reactor physics experiment for an advanced technology. The expected core volumes for these configurations have been studied using multi-group and continuous energy Monte-Carlo simulations identifying the 35% enriched systems as the most attractive. This finally leads to the choice of heavy metal rich compositions with 35% enrichment as the reference system for future studies of the next steps in the zero power reactor investigation. An alternative could be the eutectic system in the case the increased core diameter is manageable. The inter-comparison of the different applied codes and approaches available in the SCALE package has delivered a very good agreement between the results, creating trust into the developed and used models and methods.
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Hamdani, Ari, Satoshi Abe, Masahiro Ishigaki, Yasuteru Sibamoto y Taisuke Yonomoto. "Unsteady Natural Convection in a Cylindrical Containment Vessel (CIGMA) With External Wall Cooling: Numerical CFD Simulation". Energies 13, n.º 14 (15 de julio de 2020): 3652. http://dx.doi.org/10.3390/en13143652.
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In the case of a severe accident, natural convection plays an important role in the atmosphere mixing of nuclear reactor containments. In this case, the natural convection might not in the steady-state condition. Hence, instead of steady-state simulation, the transient simulation should be performed to understand natural convection in the accident scenario within a nuclear reactor containment. The present study, therefore, was aimed at the transient 3-D numerical simulations of natural convection of air around a cylindrical containment with unsteady thermal boundary conditions (BCs) at the vessel wall. For this purpose, the experiment series was done in the CIGMA facility at Japan Atomic Energy Agency (JAEA). The upper vessel or both the upper vessel and the middle jacket was cooled by subcooled water, while the lower vessel was thermally insulated. A 3-D model was simulated with OpenFOAM®, applying the unsteady Reynolds-averaged Navier–Stokes equations (URANS) model. Different turbulence models were studied, such as the standard k-ε, standard k-ω, k-ω shear stress transport (SST), and low-Reynolds-k-ε Launder–Sharma. The results of the four turbulence models were compared versus the results of experimental data. The k-ω SST showed a better prediction compared to other turbulence models. Additionally, the accuracy of the predicted temperature and pressure were improved when the heat conduction on the internal structure, i.e., flat bar, was considered in the simulation. Otherwise, the predictions on both temperature and pressure were underestimated compared with the experimental results. Hence, the conjugate heat transfer in the internal structure inside the containment vessel must be modeled accurately.
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Plott, Beth M., Shelly Scott-Nash, Bruce P. Hallbert y Angelia L. Sebok. "Computer Modeling of a Nuclear Power Plant Operating Crew to Aid in Analysis of Crew Size Issues". Proceedings of the Human Factors and Ergonomics Society Annual Meeting 39, n.º 18 (octubre de 1995): 1214–18. http://dx.doi.org/10.1177/154193129503901814.
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An analytical approach to addressing the implications of nuclear power plant shift sizing is needed as an augmentation to the classical empirical approach. The research reported in this paper was to evaluate the feasibility and validity of one potential analytical approach as a means of evaluating the consequences of crew reduction on crew performance in a nuclear power plant setting. The approach selected for analysis was task network modeling and simulation using a tool named Micro Saint. Task network modeling allows the human factors engineer to extend the information from a task analysis and generate a computer simulation of crew performance that can predict critical task times and error rates. Through modeling, the current and proposed processes can be evaluated and analyzed in order to understand, identify, and test opportunities for process improvement or reengineering. For this effort, models of a conventional nuclear power plant during four extremely demanding scenarios were developed. Task analysis and timing data were collected at the Imatran Voima Nuclear Power Plant at Loviisa, Finland. The task analyses were collected over a two week period by interviewing reactor operators, reviewing procedures, and conducting walk-throughs. We then refined the models and incorporated workload modeling constructs. At the completion of the modeling effort, the models were executed and the data collected were used to predict crew performance in varying staffing conditions.
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Wang, Li, Wentao Sun, Jie Zhao y Dichen Liu. "A Speed-Governing System Model with Over-Frequency Protection for Nuclear Power Generating Units". Energies 13, n.º 1 (31 de diciembre de 2019): 173. http://dx.doi.org/10.3390/en13010173.
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Overspeed is more likely to occur in the process of load rejection or large disturbances for nuclear steam turbines due to the large parameter range and low steam parameters, as well as the power of the low-pressure cylinder accounting for a high proportion of the total power. It is of great significance to study the overspeed characteristics of nuclear power plants (NPPs) to ensure the safe and stable operation of the unit and power grid. According to the characteristics of NPPs, the overspeed protection model and the super-acceleration protection model were established, which were added to the speed-governing system model. The response characteristics of the reactor, thermal system, steam turbine and speed-governing system in the process of load rejection or large disturbances of the power grid were analyzed and simulated. The results were compared using the simulation software personal computer transient analyzer (PCTRAN). The simulation results showed that quickly closing both the high and medium pressure regulating valves could effectively realize frequency control when load rejection or a large grid disturbance occurred. The over-acceleration protection cooperates with the super-acceleration protection to avoid the repeated opening/closing of the valves due to overspeed protection. This could effectively reduce the impact of large disturbances on the reactor, thermal system, and turbine.
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Heidari Sangestani, Soroush, Mohammad Rahgoshay, Naser Vosoughi y Mitra Athari Allaf. "Study of Fast Transient Pressure Drop in VVER-1000 Nuclear Reactor Using Acoustic Phenomenon". Science and Technology of Nuclear Installations 2018 (2018): 1–11. http://dx.doi.org/10.1155/2018/7862847.
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This article aims to simulate the sudden and fast pressure drop of VVER-1000 reactor core coolant, regarding acoustic phenomenon. It is used to acquire a more accurate method in order to simulate the various accidents of reactor core. Neutronic equations should be solved concurrently by means of DRAGON 4 and DONJON 4 coupling codes. The results of the developed package are compared with WIMS/CITATION and final safety analysis report of Bushehr VVER-1000 reactor (FSAR). Afterwards, time dependent thermal-hydraulic equations are answered by employing Single Heated Channel by Sectionalized Compressible Fluid method. Then, the obtained results were validated by the same transient simulation in a pressurized water reactor core. Then, thermal-hydraulic and neutronic modules are coupled concurrently by use of producing group constants regarding the thermal feedback effect. Results were compared to the mentioned transient simulation in RELAP5 computer code, which show that mass flux drop is sensed at the end of channel in several milliseconds which causes heat flux drop too. The thermal feedback resulted in production of some perturbations in the changes of these parameters. The achieved results for this very fast pressure drop represent accurate calculations of thermoneutronic parameters fast changes.
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González-Pintor, S., D. Ginestar y G. Verdú. "Updating the Lambda modes of a nuclear power reactor". Mathematical and Computer Modelling 54, n.º 7-8 (octubre de 2011): 1796–801. http://dx.doi.org/10.1016/j.mcm.2010.12.013.
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Dong, Zhe, Miao Liu, Di Jiang, Xiaojin Huang, Yajun Zhang y Zuoyi Zhang. "Automatic Generation Control of Nuclear Heating Reactor Power Plants". Energies 11, n.º 10 (16 de octubre de 2018): 2782. http://dx.doi.org/10.3390/en11102782.
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A nuclear heating reactor (NHR) is a typical integral pressurized water reactor (iPWR) with advanced design features such as an integral primary circuit, self-pressurization, full-power-range natural circulation, and hydraulic control rods. Through adjusting its electric power output according to the variation of demand, NHR power plants can be adopted to stablize the fluctuation of grid frequency caused by the intermittent nature of renewable generation, which is useful for deepening the penetration of renewables. The flexibility of an NHR power plant relies on the automatic generation control (AGC) function of the plant coordination control system, whose central is the AGC law. In this paper, the plant control system with AGC function is designed for NHR plants, where the AGC is realized based on the stabilizers of grid frequency and main steam pressure. Then, the AGC problem is transferred to the disturbance attenuation problem of a second-order dynamic system, and an active disturbance attenuation control (ADRC), which is just the addition of a feedback control given by a proportional‒integral (PI) law and a feedforward control driven by a disturbance observer (DO), is then proposed. Finally, this ADRC is applied to realize the AGC function for NHR-200II reactor power plant, and numerical simulation results show the implementation feasibility and satisfactory performance.
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Rössler, K. "Laboratory Simulation of Chemical Interactions of Accelerated Ions with Dust and Ice Grains". International Astronomical Union Colloquium 85 (1985): 357–63. http://dx.doi.org/10.1017/s0252921100084918.
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AbstractEnergetic ions or atoms in space may undergo hot chemical reactions upon penetration into interplanetary or interstellar dust grains, ice layers, cometary matter, and surfaces of planetary moons. The mechanistic pathways can be different from those of classical ion molecule interactions, photolytical and radiolytical processes. The kinetic energy of the hot reactant facilitates endothermic reactions and those with high energy of activation, among them atom-molecule interactions. The conditions of hot cosmic chemistry are simulated in laboratory experiments in order to obtain insight into the nature of chemical products and the reaction mechanisms of their formation. This paper reviews the methods of ion implantation, nuclear recoil in situ, nuclear recoil implantation, secondary knock-on processes and computer simulation of collision cascades. Carbon and nitrogen impact in frozen H2O, NH3 and CH4 is shown to lead to the formation and radiolytic permutation of a series of organic molecules, among them e.g. formaldehyde, methanol, methylamine, cyanamide, formamidine and guanidine which may act as precursors for biomolecules.
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Cherezov, Alexey, Jinsu Park, Hanjoo Kim, Jiwon Choe y Deokjung Lee. "A Multi-Physics Adaptive Time Step Coupling Algorithm for Light-Water Reactor Core Transient and Accident Simulation". Energies 13, n.º 23 (2 de diciembre de 2020): 6374. http://dx.doi.org/10.3390/en13236374.
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A new reactor core multi-physics system addresses the pellet-to-cladding heat transfer modeling to improve full-core operational transient and accident simulation used for assessment of reactor core nuclear safety. The rigorous modeling of the heat transfer phenomena involves strong interaction between neutron kinetics, thermal-hydraulics and nuclear fuel performance, as well as consideration of the pellet-to-cladding mechanical contact leading to dramatic increase in the gap thermal conductance coefficient. In contrast to core depletion where parameters smoothly depend on fuel burn-up, the core transient is driven by stiff equation associated with rapid variation in the solution and vulnerable to numerical instability for large time step sizes. Therefore, the coupling algorithm dedicated for multi-physics transient must implement adaptive time step and restart capability to achieve prescribed tolerance and to maintain stability of numerical simulation. This requirement is met in the MPCORE (Multi-Physics Core) multi-physics system employing external loose coupling approach to facilitate the coupling procedure due to little modification of constituent modules and due to high transparency of coupling interfaces. The paper investigates the coupling algorithm performance and evaluates the pellet-to-cladding heat transfer effect for the rod ejection accident of a light water reactor core benchmark.
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Valinčius, Mindaugas, Tadas Kaliatka, Algirdas Kaliatka y Eugenijus Ušpuras. "Modelling of Severe Accident and In-Vessel Melt Retention Possibilities in BWR Type Reactor". Science and Technology of Nuclear Installations 2018 (1 de agosto de 2018): 1–14. http://dx.doi.org/10.1155/2018/7162387.
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One of the severe accident management strategies for nuclear reactors is the melted corium retention inside the reactor pressure vessel. The work presented in this article investigates the application of in-vessel retention (IVR) severe accident management strategy in a BWR reactor. The investigations were performed assuming a scenario with the large break LOCA without injection of cooling water. A computer code RELAP/SCDAPSIM MOD 3.4 was used for the numerical simulation of the accident. Using a model of the entire reactor, a full accident sequence from the large break to core uncover and heat-up as well as corium relocation to the lower head is presented. The ex-vessel cooling was modelled in order to evaluate the applicability of RELAP/SCDAPSIM code for predicting the heat fluxes and reactor pressure vessel wall temperatures. The results of different ex-vessel heat transfer modes were compared and it was concluded that the implemented heat transfer correlations of COUPLE module in RELAP/SCDAPSIM should be applied for IVR analysis. To investigate the influence of debris separation into oxidic and metallic layers in the molten pool on the heat transfer through the wall of the lower head the analytical study was conducted. The results of this study showed that the focusing effect is significant and under some extreme conditions local heat flux from reactor vessel could exceed the critical heat flux. It was recommended that the existing RELAP/SCDAPSIM models of the processes in the debris should be updated in order to consider more complex phenomena and at least oxide and metal phase separation, allowing evaluating local distribution of the heat fluxes.
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Liu, Chuan, Ying Luo, Min Yang y Qiang Fu. "Effects of material hardening model and lumped-pass method on welding residual stress simulation of J-groove weld in nuclear RPV". Engineering Computations 33, n.º 5 (4 de julio de 2016): 1435–50. http://dx.doi.org/10.1108/ec-08-2015-0216.
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Purpose – The purpose of this paper is to clarify the effect of material hardening model and lump-pass method on the thermal-elastic-plastic (TEP) finite element (FE) simulation of residual stress induced by multi-pass welding of materials with cyclic plasticity. Design/methodology/approach – Nickel-base alloy and stainless steel, which are used in J-type weld for manufacturing the nuclear reactor pressure head, can easily harden during multi-pass welding. The J-weld welding experiment is carried out and the temperature cycle and residual stress are measured to validate the TEP simulation. Thermal-mechanical sequence coupling method is employed to get the welding residual stress. The lumped-pass model and pass-by-pass FE model are built and two materials hardening models, kinematic hardening model and mixed hardening model, are adopted during the simulations. The effects of material hardening models and lumped-pass method on the residual stress in J-weld are distinguished. Findings – Based on the kinematic hardening model, the stresses simulated with the lumped-pass FE model are almost consistent with those obtained by the pass-by-pass FE model; while with the mixed hardening material model, the lumped-pass method has great effect on the simulated stress. Practical implications – A computation with mixed isotropic-kinematic material seems not to be the appropriate solution when using the lumped-pass method to save the computation time. Originality/value – In the simulation of multi-pass welding residual stress involved in materials with cyclic plasticity, the material hardening model should be carefully considered. The kinematic hardening model with lump-pass FE model can be used to get better simulation results with less computation time. The results give a direction for welding residual stress simulation for the large structure such as the reactor pressure vessel.
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Deng, Baiquan, Zaixin Li, Jinhua Huang y Tao Yuan. "Computer Simulation of Tritium Inventory and Leakage of the Fusion Experimental Reactor FEB-E". Fusion Science and Technology 46, n.º 4 (diciembre de 2004): 548–60. http://dx.doi.org/10.13182/fst04-a590.
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Paladino, Domenico, Max Huggenberger y Frank Schäfer. "Natural Circulation Characteristics at Low-Pressure Conditions through PANDA Experiments and ATHLET Simulations". Science and Technology of Nuclear Installations 2008 (2008): 1–14. http://dx.doi.org/10.1155/2008/874969.
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Natural circulation characteristics at low pressure/low power have been studied by performing experimental investigations and numerical simulations. The PANDA large-scale facility was used to provide valuable, high quality data on natural circulation characteristics as a function of several parameters and for a wide range of operating conditions. The new experimental data allow for testing and improving the capabilities of the thermal-hydraulic computer codes to be used for treating natural circulation loops in a range with increased attention. This paper presents a synthesis of a part of the results obtained within the EU-Project NACUSP “natural circulation and stability performance of boiling water reactors.” It does so by using the experimental results produced in PANDA and by showing some examples of numerical simulations performed with the thermal-hydraulic code ATHLET.
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Botha, Frikkie, Robert Dobson y Thomas Harms. "Simulation of a syngas from a coal production plant coupled to a high temperature nuclear reactor". Journal of Energy in Southern Africa 24, n.º 2 (1 de mayo de 2013): 37–45. http://dx.doi.org/10.17159/2413-3051/2013/v24i2a3128.
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In light of the rapid depletion of the world’s oil reserves, concerns about energy security prompted the exploration of alternative sources of liquid fuels for transportation. One such alternative is the production of synthetic fuel using an indirect coal liquefaction process or coal-to-liquids (CTL) process. In this process, coal is gasified in a gasifier in the presence of steam and oxygen to produce a synthesis gas or syngas consisting mainly of hydrogen and carbon monoxide. The syngas is then converted to liquid fuels and a variety of useful chemicals in a Fischer Tropsch-type synthesis reactor. However, the traditional process for syngas production also produces substantial amounts of carbon dioxide. In fact, only about one third of the carbon in the coal feedstock ends up in the liquid fuel product using traditional CTL technology. If more hydrogen was available than the hydrogen produced during the gasification step, the carbon utilisation of the process could be improved significantly. The high temperature reactor (HTR) is a gas cooled Generation IV nuclear reactor ideally suited to provide power and high temperature heat for carbon neutral production of hydrogen via high temperature electrolysis. The integration of an HTR into a CTL process therefore provides an opportunity to improve the thermal and carbon efficiency of the CTL process significantly. This paper presents a possible process flow scheme for a nuclear assisted CTL process. The system is evaluated in terms of its thermal or syngas production efficiency (defined as the ratio of the heating value of the produced syngas to the sum of the heating value of the coal plus the HTR heat input) as well as its carbon utilisation. If the hydrogen production plant is sized to produce only enough associated oxygen to supply the needs of the gasification plant, syngas is produced at about 63% thermal efficiency, while 71.5% of the carbon is utilised in this process. It was found that the optimum HTR outlet temperature to produce hydrogen with a high temperature steam electrolysis process is 850°C. If enough process heat and power are available and process equipment capacities are sufficient, the carbon utilisation of the process could be improved even further to values in excess of 90%.
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Ramadhani, Amanda Dhyan Purna, Susilo Susilo, Irfan Nurfatthan, Yohannes Sardjono, Widarto Widarto, Gede Sutresna Wijaya y Isman Mulyadi Triatmoko. "DOSE ESTIMATION OF THE BNCT WATER PHANTOM BASED ON MCNPX COMPUTER CODE SIMULATION". JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA 22, n.º 1 (25 de marzo de 2020): 23. http://dx.doi.org/10.17146/tdm.2020.22.1.5780.
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Cancer is a malignant tumor that destroys healthy cells. Cancer treatment can be done by several methods, one of which is BNCT. BNCT uses 10B target which is injected into the human body, then it is irradiated with thermal or epithermal neutrons. Nuclear reaction will occur between boron and neutrons, producing alpha particle and lithium-7. The dose is estimated by how much boron and neutron should be given to the patient as a sum of number of boron, number of neutrons, number of protons, and number of gamma in the reaction of the boron and neutron. To calculate the dose, the authors simulated the reaction with Monte Carlo N Particle-X computer code. A water phantom was used to represent the human torso, as 75% of human body consists of water. Geometry designed in MCNPX is in cubic form containing water and a cancer cell with a radius of 2 cm. Neutron irradiation is simulated as originated from Kartini research reactor, modeled in cylindrical form to represent its aperture. The resulting total dose rate needed to destroy the cancer cell in GTV is 2.0814×1014 Gy.s (76,38%) with an irradiation time of 1,4414×10-13 s. In PTV the dose is 5.2295×1013 Gy.s (19,19%) with irradiation time of 5.7367×10-13 s. In CTV, required dose is 1.1866×1013 Gy.s (4,35%) with an irradiation time of 2.5283×10-12 s. In the water it is 1.9128×1011 Gy.s (0,07%) with an irradiation time of 1,5684×10-10 s. The irradiation time is extremely short since the modeling is based on water phantom instead of human body.Keywords: BNCT, Dose, Cancer, Water Phantom, MCNPX
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Kuzevanov, Vyacheslav S. y Sergey K. Podgorny. "Temperature field in gas-cooled reactor core in transient conditions under different approaches to mass flow profiling". Nuclear Energy and Technology 5, n.º 4 (10 de octubre de 2019): 297–303. http://dx.doi.org/10.3897/nucet.5.48392.
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Positive effect of profiling the gas-cooled reactor core within the framework of the GT-MHR project was investigated in (Podgorny and Kuzevanov 2017, Kuzevanov and Podgorny 2017, 2018). The necessity arises to supplement already implemented analysis of equilibrium conditions of core operation with investigation of effects of profiling on the temperature field in transient modes of reactor core operation. The present paper is dedicated to the investigation of development of transients in gas-cooled nuclear reactor core subject to the implementation of different principles of core profiling. Investigation of transients in reactor core represents complex problem, solution of which by conducting direct measurements is beyond the resources available to the authors. Besides the above, numerical simulation based on advanced CFD software complexes (ANSYS 2016, 2016a, 2016b, Shaw 1992, Anderson et al. 2009, Petrila and Trif 2005, Mohammadi and Pironneau 1994) is also fairly demanding in terms of required computer resources. The algorithm for calculating temperature fields using the model where the reactor core is represented as the solid medium with gas voids was developed by the authors and the assumption was made that heat transfer due to molecular heat conductivity can be described by thermal conductivity equation written for continuous medium with thermal physics parameters equivalent to respective parameters of porous object in order to get the possibility of obtaining prompt solutions of this type of problems. Computer code for calculating temperature field in gas-cooled reactor in transient operation modes was developed based on the suggested algorithm. Proprietary computation code was verified by comparing the results of numerous calculations with results of CFD-modeling of respective transients in the object imitating the core of gas-cooled nuclear reactor. The advantage of the developed computer code is the possibility of real-time calculation of evolution of conditions in complex configurations of gas-cooled reactor cores with different channel diameters. This allows using the computer code in the calculations of transients in loops of reactor facility as a whole, in particular for developing reactor simulators. Results are provided of calculations of transients for reactor core imitating the core of gas-cooled nuclear reactor within the framework of GT-MHR project performed for different approaches to profiling coolant mass flow. Results of calculations unambiguously indicate the significant difference of temperature regimes during transients in the reactor core with and without profiling and undeniable enhancement of reliability of nuclear reactor (Design of the Reactor Core 2005, International Safeguards 2014, Safety of Nuclear Power Plants 2014) with profiling of coolant mass flow in the reactor core as a whole.