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In some settings, systems may not fail completely but instead undergo performance degradation, leading to reduced efficiency. A significant concern arises when a system transitions into a degraded state without immediate detection, with the degradation only becoming apparent after an unpredictable period. Undetected degradation can result in failures with significant consequences. For instance, a minor crack in an oil pipeline might go unnoticed, eventually leading to a major leak, environmental harm, and costly cleanup efforts. Similarly, in the nuclear industry, undetected degradation in reactor cooling systems could cause overheating and potentially catastrophic failure. This paper focuses on reliability modeling for systems experiencing degradation, accounting for time delays associated with undetected degraded states, self-repair mechanisms, and varying operating environments. The paper presents a reliability model for degraded, time-dependent systems, incorporating various aspects of degradation. It first discusses the model assumptions and formulation, followed by numerical results obtained from system modeling using the developed program. Various scenarios are illustrated, incorporating time delays and different parameter values. Through computational analysis of these complex systems, we observe that the probability of the system being in the undetected degraded state tends to stabilize shortly after the initial degradation begins. The model is valuable for predicting and establishing an upper bound on the probability of the undetected, degraded state and the system's overall reliability. Finally, the paper outlines potential avenues for future research. [ABSTRACT FROM AUTHOR]

Copyright of Atomic Energy Science & Technology is the property of Editorial Board of Atomic Energy Science & Technology and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

This review delves into the quest for optimal efficiency in the secondary cooling process of nuclear reactor water plant coolant systems. Modeling secondary cooling nuclear processes is hardly performed. Thus, Neural networks with traditional statistical methodologies are integrated to innovate a hybrid model to revolutionize nuclear reactor cooling systems' performance, reliability, and safety. A total of 63 indexed papers were reviewed in the nuclear field that analyzed critical research gaps, including the need for uncertainty modeling and resilience against external hazards. Insights into sensor technologies, data analytics, and real-time monitoring underscore the importance of continuous optimization and predictive maintenance were reviewed. A descriptive analysis for a month of sampling data was presented for the parameters of temperature for TT003 and TT004 and pressure for PT002 and PT003 of the secondary process. The confidence level of 95.0% is identified for the temperature and pressure parameters. The lowest standard error was recognized at 0.00032 and 0.01691, respectively. The review culminates with a forward-looking perspective, recognizing the pivotal role of hybrid machine learning models in shaping the future of secondary cooling processes for nuclear reactor water coolant plants to improve the efficiency and sustainability of nuclear reactor systems. [ABSTRACT FROM AUTHOR]

The main motivation behind this numerical analysis is to disclose the salient magneto-thermo characteristic features of two-dimensional viscous incompressible magnetized Soret and Dufour effects on the boundary layer flow of Walter’s-B fluid over a stretching sheet under the influence of viscous dissipation. The novel physical effects such as nonuniform heat source/sink and magnetic Ohmic heating are included in the governing equations. Further, prescribed surface temperature condition is introduced in the boundary conditions to describe the thermal transport features. However, the deployed non-Newtonian Walter’s-B fluid rheology model finds its innumerable applications in numerous areas of science and engineering including electronic cooling, polymer processing, nuclear reactor cooling, extraction of polymer sheet, plastic sheet drawing, chemical engineering, metallurgy, injection molding and many more. Therefore, authors have motivated with these applications and advantageous of considered fluid model. Hence, an effort is made to describe the dynamic characteristic features of non-Newtonian Walter’s-B rheology model about a stretching sheet. However, present physical flow model produces highly nonlinear coupled two-dimensional partial differential equations and which are not acquiescent to available analytical methods. Owing to this, a robust MATLAB-based BVP4C technique is deployed to produce the appropriate numerical solutions through suitable similarity transformations. The graphical representations of velocity, temperature and concentration profiles along with skin-friction coefficient, Nusselt and Sherwood numbers are presented. Magnifying magnetic number decays the velocity profile and enhances the thermal and concentration fields. Rising viscoelastic parameter diminishes the thermal and concentration fields and magnifies the velocity field. Rising the radiation parameter diminish the thermal field. Magnifying Soret number raises the concentration field. Rising the space and temperature dependent heat source/sink coefficient amplifies the thermal diffusion field. Magnifying Dufour and Eckert numbers amplifies the thermal field. Finally, the guarantee and accuracy of the investigated problem is presented through a comparison with previous studies. [ABSTRACT FROM AUTHOR]

Current theoretical and computational investigation within the Darcy–Forchheimer medium through electromagnetic fields reveals the heat and mass transportation characteristics for the flow of Williamson–Sutterby nanofluid under the effects of Cattaneo–Christov double diffusion, radiation heat flux, magnetic dipole and convective boundary over a stretchy surface are taken into account. Cattaneo–Christov heat mass flux model is applied to shape the thermal and nanoparticle concentration equations. The prime purpose of the current study is to enhance the rate of heat transfer in industrial processes. Mechanism of thermal insulation, pharmacological processes, pace technology, crushing, nuclear reactor cooling, pharmaceutical procedures, geothermal reservoirs to enhance oil recovery with chemically reactive systems has great importance in mass transport. Moreover, the mass transportation is made by Arrhenius activation energy and binary chemical reaction. Additionally, the influences of bioconvection of self-propelled microorganisms are considered. The nanofluids with swimming microorganisms have great significance in microfluidics devices, medicine, cancer therapy, biotechnology applications such as biofuels and enzyme biosensor. The amalgamation of Sutterby–Williamson nanoparticles and microorganism with slight homogeneous diffusion exhibit the novelty of present work. Nonlinear PDEs are transformed into coupled ODEs by using similarity functions. The attained equations are then solved numerically with the help of software. Hartmann, Sutterby Reynolds, Sutterby Deborah, bioconvection Rayleigh numbers, mixed convection, porosity, ferrohydrodynamic interaction, buoyancy ratio and inertial resistance parameters slow down the flow of fluids. But reverse effects were observed for temperature, concentration and motile density profiles for Hartmann, porosity and ferrohydrodynamic interaction parameters. These theoretical outcomes play a vital role in industrial applications to enhance the heat/mass process, biomedical flows, heating and cooling systems, chemical processes and mining industries. [ABSTRACT FROM AUTHOR]

This work objective focuses on studying the combined influences of variable thermal conductivity, chemical reaction, and magnetohydrodynamics (MHD) on the flow of a tangent hyperbolic nanofluid flow over an exponentially stretching surface, considering a first-order velocity slip condition. Additionally, thermophoresis and Brownian motion impacts are taken into account. The phenomena of heat transfer are analyzed considering several factors such as thermal radiation, Joule heating and nonlinear heat source. On the other hand, mass transfer is explored under the effect of chemical reaction. Tangent hyperbolic fluid is an important branch of non-Newtonian fluids known for its ability to describe shear thinning effects. Understanding fluid flow on exponentially stretched surfaces is of great significance due to its applications in various industrial processes. These applications include fluid film condensing methods, plastic production for making plastic covers, fiber manufacturing (where it is used to spin fibers), glass blowing, metallurgical procedures, and the paper industry. The concept of magnetohydrodynamics (MHD) is significant due to its various engineering applications, such as MHD generators, flow meters, heat reservoirs, small components in different devices, and cooling systems for nuclear reactors. To analyze the system, using similarity transformations, the governing equations of continuity, velocity, and concentration are transformed into non-dimensional differential equations. The numerical solution is obtained using the shooting technique. The study presents the physical significance of all the fluid parameters involved, focusing on the velocity, temperature, and concentration profiles. These profiles are presented graphically and discussed in detail. The results show that the fluid velocity profile increases with enhancing values of the We and the magnetic number M. The thermal profile increases with higher Nt and Rd The concentration profile decreases with higher values of Q t and Nb. [ABSTRACT FROM AUTHOR]

Production processes become more and more complex in terms of engineering implementation. During the operation of equipment for nuclear power plants or research reactor facilities, it is advisable to consider big data volumes, their change dynamics and deviations from the normal state. The IRT-T reactor facility at the School of Nuclear Science and Technology in National Research Tomsk Polytechnic University, is used for machine learning. [ABSTRACT FROM AUTHOR]

The present study investigates the three-dimensional convective rotating hybrid nanofluid flow over a linear stretching/shrinking sheet under the influence of dissipative heat. The behaviour of nanofluids, which are mixtures of Ag and MoS 4 nanoparticles in water in this study, is considered in the occurrence of convective flow and the impact of rotation. The linear stretching / shrinking sheet represents a common model in many industrial processes. The analysis also accounts for the effects of dissipative heat that arise due to viscous dissipation and Ohmic dissipation. A well-equipped set of transformations is used for the conversion of the non-dimensional form of the governing equations. Moreover, the adaptation of the Hamilton–Crosser conductivity model enhances the study due to enhanced physical properties. Further, the combined effect of nanofluid properties, convective flow, rotation and dissipative heat on the fluid flow and heat transportation features is examined followed by the numerical solutions adopted to find the solution of the transformed set of equations. The results obtained from this investigation will have potential implications for enhancing the use of nanofluid in industrial applications where efficient cooling of nuclear reactors, electronics cooling, automotive cooling systems, etc. are crucial. Further, use of nanofluid in optimising heat transfer processes in relevant engineering systems is vital. [ABSTRACT FROM AUTHOR]

This research addresses the implications of slip on the magnetohydrodynamic transport properties of Newtonian nanofluids. The study employs Buongiorno’s theoretical framework and focuses its attention on a cylindrical object that experiences tensile deformation inside a medium characterized by porosity. In contrast to the often observed boundary conditions that presume a constant temperature and concentration, the current research utilizes hydrodynamic and thermal slip conditions. The recognition of many significant elements, including the Arrhenius activation energy, magnetic field, and viscous dissipation, also has important significance. The phenomena of interest pertaining to fluid flow is first defined and afterward converted into a non dimensional form by including relevant commonalities. In order to do a thorough examination of the current situation, the shooting algorithm and the bvp4c technique are used. Graphical and tabular representations are used to demonstrate the implications of several emerging features on the velocity, temperature, and dispersion of nanoparticles. The observed data suggests that a rise in chemical reaction parameters leads to a drop in nanoparticle concentration distributions, but the activation energy parameter exhibits an opposite tendency. It is discovered that the boundary layer thickness under slip flow circumstances differs from that seen without slip flow. The Sherwood number rises when the chemical reaction parameter gets higher, whereas the activation energy parameter escalates in the opposite direction. The present research covers a broad variety of applied sciences applications, with a special emphasis on thermal oil recovery, geothermal reservoirs, chemical engineering, and nuclear reactor cooling. [ABSTRACT FROM AUTHOR]

Reactors produce noise that negatively impacts the environment during their operation. This study focused on vibration and noise reduction in shunt oil reactors through an enhanced design, aiming to mitigate these environmental consequences. Additionally, the precision of measuring vibroacoustic characteristics was enhanced by employing the finite element method. This study utilized the 110-750 kV reactor, POM-60000/525-У1. To minimize noise and vibrations from the active part to the tank, the rubber gaskets' thicknesses, dimensions, and placements were carefully chosen, considering the magnetic system's design specifics and pressing forces within the reactor frame. The reactor's design underwent improvements through modeling and optimization. In the revised design, the number of radiators in the cooling system on both sides and the stiffeners on the reactor tank's walls were increased, while the reactor's overall dimensions were preserved. The isosurfaces derived from vibroacoustic testing on the reactor tank model facilitated the processing and visualization of results, thereby validating the model's accuracy and effectiveness. The average background noise levels of the enhanced reactor were significantly lowered, with the noise difference between the reactor and its surroundings being nearly double that of its existing counterpart. Since noise from wall vibrations and the cooling system impacts reactor lifecycles, reducing these vibrations is crucial for extending the lifespan of the improved reactor. [ABSTRACT FROM AUTHOR]

In the research field on severe accidents in nuclear power plant, a specific scenario correspond to accident with oxidative conditions for which damaged fuel can be highly oxidised with significant releases of ruthenium oxides. Ruthenium chemistry is complex, and the current knowledge has to be deepened to better assess ruthenium source term with potential radioactive releases to the environment as volatile ruthenium tetroxide. In this work, experimental studies are focused on ruthenium behaviour along a stainless-steel thermal gradient tube, with maximum temperature of 1200 °C, simulating the reactor cooling system in oxidizing conditions with mainly steam/air gas mixtures. Results showed that few % of ruthenium oxides (< 10% with SS tube) can reach low temperature, representative of containment temperature, even with low oxygen content in the carrier gas. The ruthenium revaporization process from the Ru deposits along the tube on mid-term was studied. Influence of carrier gas composition (steam %), flow rate and NOx feed are discussed. [ABSTRACT FROM AUTHOR]

Pyrolysis converts biomass into other biofuels and is an alternative for residue management. This study aimed to investigate the kinetics of pyrolysis of açaí seeds and to bring insights into the slow pyrolysis of their briquettes. The thermogravimetric analysis was performed in an N2 atmosphere with a final temperature of 1173 K. It tested heating rates of 2, 5, 10, 15 and 20 K min−1. We determined the kinetic triplet using the isoconversional models of Friedman, Ozawa–Flynn–Wall, modified Coats–Redfern, Starink, and Vyazovkin, and the master plots method. In addition, slow pyrolysis occurred in a fixed-bed reactor coupled with a cooling system. The bio-oil yield was optimized via response surface methodology, and the compounds were identified with the support of gas chromatography coupled with mass spectrometry. The isoconversional model of Ozawa–Flynn–Wall showed activation energy of 165.59 kJ mol−1, with the slightest deviation value. The average percentage deviation between theoretical and experimental master plots corresponded to second-order random nucleation. At last, the global activation energy value was 134.76 kJ mol−1, and an Arrhenius pre-exponential factor was 1.11 1010 s−1, with R2 = 0.96. For slow pyrolysis, it observed an inflection point at 715 K and 11.93 K min−1. We concluded that the second-order reaction mechanism represents the pyrolysis of the açaí seed. In addition, the bio-oil yield maximizes at milder temperatures with high heating rates or high temperatures with low heating rates. These production parameters affect the bio-oil composition. [ABSTRACT FROM AUTHOR]

In a wide variety of mechanical and industrial applications, e.g., space cooling, nuclear reactor cooling, medicinal utilizations (magnetic drug targeting), energy generation, and heat conduction in tissues, the heat transfer phenomenon is involved. Fourier's law of heat conduction has been used as the foundation for predicting the heat transfer behavior in a variety of real-world contexts. This model's production of a parabolic energy expression, which means that an initial disturbance would immediately affect the system under investigation, is one of its main drawbacks. Therefore, numerous researchers worked on such problem to resolve this issue. At last, this problem was resolved by Cattaneo by adding relaxation time for heat flux in Fourier's law, which was defined as the time required to establish steady heat conduction once a temperature gradient is imposed. Christov offered a material invariant version of Cattaneo's model by taking into account the upper-connected derivative of the Oldroyd model. Nowadays, both models are combinedly known as the Cattaneo-Christov (CC) model. In this attempt, the mixed convective MHD Falkner-Skan Sutterby nanofluid flow is addressed towards a wedge surface in the presence of the variable external magnetic field. The CC model is incorporated instead of Fourier's law for the examination of heat transfer features in the energy expression. A two-phase nanofluid model is utilized for the implementation of nano-concept. The nonlinear system of equations is tackled through the bvp4c technique in the MATLAB software 2016. The influence of pertinent flow parameters is discussed and displayed through different sketches. Major and important results are summarized in the conclusion section. Furthermore, in both cases of wall-through flow (i.e., suction and injection effects), the porosity parameters increase the flow speed, and decrease the heat transport and the influence of drag forces. [ABSTRACT FROM AUTHOR]

The High Temperature Test Facility (HTTF) at Oregon State University is an integral system test facility to simulate postulated reactor transients of prismatic high-temperature gas-cooled reactors(HTGRs). A series of test campaigns was launched, providing abundant test data that could be used to benchmark reactor system analysis codes like the System Analysis Module (SAM). In this study, a SAM model of the facility is developed based on the two-dimensional (2D) ring model approach. All components including the ceramic matrix, graphite heaters, helium coolant channels, core barrel, upcomer, pressure vessel, and reactor cavity cooling system are modeled as concentric cylindrical rings. The model is used to simulate one of the benchmark problems—Pressurized Conduction Cooldown (PCC)—within the scope of the Organisation for Economic Co-operation and Development Nuclear Energy Agency International HTTF Benchmark. The simulations consist of two parts. In the first part, operating and boundary conditions as well as thermophysical properties of materials are specified for the benchmark problem. Results from the first part will be used in code-to-code comparison. In the second part, the SAM model is used to simulate Test PG-27, which is the first PCC test carried out in the HTTF, with only two of the ten heater banks activated. The results in the second part are used for code-to-data comparison. Because the helium coolant flow rate is not measured in this facility, it is estimated using the input power and inlet/outlet coolant temperatures. Additionally, radial heat flow in the ceramic blocks is complicated by hundreds of cylindrical coolant channels and heater rods embedded in them. As such, it is necessary to deduce an effective thermal conductivity for the ceramic to analyze the core thermal behavior. SAM predictions of the helium coolant and ceramic temperatures are compared with test data measured in three equivalent sectors. Overall, the SAM results agree reasonably well with test data within the variation of data among the three sectors, which demonstrates SAM's capability in capturing transient effects in HTGR using the simplified 2D ring model. [ABSTRACT FROM AUTHOR]

There are several industrial applications where boiling is used, for example boilers, refrigeration systems, nuclear reactor cooling, and microelectronic chip cooling. Experimental research has been carried out to determine the flow boiling heat transfer capabilities of copper-alumina-coated surfaces for application in heat transfer equipment. De-ionized (DI) water is used as the coolant for experimentations in a minichannel with dimensions 10 × 1.5 × 10 mm. Copper surfaces coated with thin copper-alumina nanocomposite films are created using the electrodeposition process. The coated layer created using an electrochemical technique offers strong adhesiveness with the base copper and is therefore anticipated to be suitable for real-world heat transfer appliances as part of the ongoing scientific development in subcooled flow boiling. The electrochemical technique offers easier control over its various parameters, such as current density, duration and electrolyte composition, making it possible to easily achieve a variety of surface characteristics, such as crystallinity, wettability and porosity. as required in the coated surfaces. Additionally, the copper-alumina is a hydrothermally stable oxide material that is well suited for use in boiling heat transfer devices. The boiling (subcooled flow) heat transfer tests are carried out at various mass flows. The improvement in the two-phase heat transfer coefficient (HTC) and critical heat flux (CHF) can reach up to 90 % and 93 %, respectively. The coated surfaces have improved CHF and HTC because of improved wettability, increased surface roughness, and the existence of active nucleate sites in high-density. [ABSTRACT FROM AUTHOR]

The present study deals with electrically conductive fluid flow across a heated circular cylinder to examine the oscillatory magnetic flux and heat transfer in the presence of variable surface temperature. The proposed mathematical formulation is time-dependent, which is the source of the amplitude and fluctuation in this analysis. The designed fluctuating nonlinear computational model is associated with the differential equations under specific boundary conditions. The governing equations are converted into dimensionless form by using adequate dimensionless variables. To simplify the resolution of the set of governing equations, it is further reduced. The effects of surface temperature parameter β , magnetic force number ξ , buoyancy parameter λ , Prandtl number Pr, and magnetic Prandtl parameter γ are investigated. The main finding of the current study is related to the determination of the temperature distribution for each inclination angle. It is seen that a higher amplitude of the heat transfer rate occurs as the surface temperature increases. It is also noticed that the oscillatory magnetic flux becomes more important as the magnetic Prandtl number increases at each position. The present magneto-thermal analysis is significantly important in practical applications such as power plants, thermally insulated engines, and nuclear reactor cooling. [ABSTRACT FROM AUTHOR]

Peristalsis of Ree–Eyring non-Newtonian fluid is significant to investigate the rheological features of biological fluids such as blood, saliva, intravascular fluids, intracellular fluids, and interstitial fluids. As a consequence of this development, the present article explores the cross-diffusive magnetohydrodynamic peristaltic transport of a Ree–Eyring fluid conveying tiny particles through a flexible porous channel under the influence of activation energy. A lubrication approach is adopted to reduce the complexity of the system. The analytical solution is achieved for the velocity field. In addition, the Runge–Kutta-based shooting technique is employed to solve the temperature and fluid concentration equations. Dual solutions are executed for the Newtonian and Ree–Eyring fluid cases and discussed through tabular and graphical findings for several sets of pertinent parameters. In this investigation, it is perceived that an enhancement in the Darcy number optimizes the velocity field. The fluid temperature rises with elevated values of the Brinkman and Dufour numbers whereas the reverse trend is noticed in the fluid concentration field for Soret and activation energy parameters. Moreover, the obtained outcomes are applicable to a variety of fields in the medical sciences and engineering, such as the radiosurgery, the spectroscopy, the optoelectronics, the power conversion devices, and the nuclear reactor cooling. [ABSTRACT FROM AUTHOR]

The present study addresses briefly the safety and design requirements of nuclear power plants from earthquakes and tsunamis that may affect the structure or cooling systems of their reactors, and which may result in additionally and longer term destructive impacts on nearby communities and marine life due to the additional release of radioctivity - as was the case with the 11 March 2011 Fukushima Daichi nuclear plant in Japan, as well as with the release of radioactivity by other nuclear power plants by tsunamigenic earthquakes in other parts of the world. The vulnerability of nuclear power plants to earthquakes and tsunamis was specifically examined by the author in conducting a comprehensive study of historic earthquakes and tsunamis, as well as by an extensive air and land field survey of Southern California, undertaken under contract with the U.S. Nuclear Regulatory Agency (NRC), and the U. S. Army Coastal Engineering Research Center (CERC), in connection with the licencing of the San Onofre Nuclear Power Plant near San Clemente in California, and of subsequent attempts to licence the additional Units 2 and 3 of the same facility of the Southern California Edison Company (the licensee). The present evaluation is also based on historical records extended back in time for determining earthquake and tsunami events when California was still under Spanish control under Gaspar de Portolá, the Spanish military officer from Catalonia in Spain, the first governor of Upper California, and founder of Monterey and San Diego, before California was annexed by the United States as a State of its Union. Also researched were archives in Seville, Spain. [ABSTRACT FROM AUTHOR]

Copyright of Nuclear Safety is the property of Nuclear & Radiation Safety Center and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Copyright of Nuclear Safety is the property of Nuclear & Radiation Safety Center and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

To increase energy efficiency, the flow of fluids containing nanoparticles is crucial in industrial applications notably in nuclear reactors and nuclear system cooling. In light of this, this study examines the flow of a water‐based ternary hybrid nanofluid (graphene, single‐walled carbon nanotubes, and titanium dioxide) across a curved stretching sheet with suction. The non‐Fourier heat flux model is also considered in the modeling. The existing partial differential equations are converted into ordinary differential equations through the use of similarity variables. These ordinary differential equations are then numerically solved using the Runge–Kutta–Fehlberg fourth‐ and fifth‐order method along with a shooting approach. The collection of graphical findings for the key variables on the temperature and velocity profiles is investigated. Results reveal that the heat transport in ternary hybrid nanoliquid rises as the heat source/sink parameter rises. The Biot number influences the thermal profile positively, whereas the increasing curvature parameter values reduce heat transport. The curvature parameter has a positive impact on skin friction but the suction parameter has a negative impact on skin friction. [ABSTRACT FROM AUTHOR]

Reactor TRIGA PUSPATI (RTP) is the only research nuclear reactor in Malaysia. Maintenance of RTP is crucial which affects its safety and reliability. Currently, RTP maintenance strategies used corrective and preventative which involved many sensors and equipment conditions. The existing preventive maintenance method takes a longer time to complete the entire system's maintenance inspection. This study has investigated new predictive maintenance techniques for developing RTP predictive maintenance for primary cooling systems using machine learning (ML) and augmented reality (AR). Fifty papers from recent referred publications in the nuclear areas were reviewed and compared. Detailed comparison of ML techniques, parameters involved in the coolant system and AR design techniques were done. Multiclass support vector machines (SVMs), artificial neural network (ANN), long short-term memory (LSTM), feed forward back propagation (FFBP), graph neural networks-feed forward back propagation (GNN-FFBP) and ANN were used for the machine learning techniques for the nuclear reactor. Temperature, water flow, and water pressure were crucial parameters used in monitoring a nuclear reactor. Image marker-based techniques were mainly used by smart glass view and handheld devices. A switch knob with handle switch, pipe valve and machine feature were used for object detection in AR markerless technique. This study is significant and found seven recent papers closely related to the development of predictive maintenance for a research nuclear reactor in Malaysia. [ABSTRACT FROM AUTHOR]

Natural circulation is the important way for taking away decay heat in pool-type sodium-cooled fast reactor (SFR) when all primary pumps trip. Three-dimensional (3D) unrestricted flow and thermal stratification phenomena exist in sodium pool during natural circulation. 3D modeling of whole reactor system with complex secondary or third circuit may generate huge numbers of grid, it may cause the difficulty of performing simulation. 3D computational fluid dynamics (CFD) method is mainly used for the thermal hydraulic analysis in primary system. The power variation of heat exchanger is unclear under natural circulation condition, which is key boundary condition in 3D CFD simulation. It is necessary to develop the numerical method for realizing the boundary more closed to the actual condition. In this paper, a "1D system code + 3D CFD" simulation method combining the simulation of 3D CFD and system code SAC-IRACS was proposed. The steady and transient simulations with 24000 s of PHENIX natural circulation test were performed. Flow paths in PHENIX primary system during the test were identified and mass flow rate were obtained. 3D thermal stratification phenomena in sodium pool during natural circulation test were captured. Variation trends of key thermal hydraulic parameters are basically consistent with the test. There is the reverse flow in RVCS under natural circulation conditions. At 24000 s, mass flow rate of reactor core, primary pumps and RVCS are 63.13 kg/s, 58.04 kg/s, −5.11 kg/s respectively. The decay heat in primary system can be taken away by intermediate reactor auxiliary cooling system (IRACS). Obvious thermal stratification phenomena appear again in sodium pool with the efficient cooling. This "1D system code + 3D CFD" simulation can provide reasonable results, and can be applied to natural circulation analysis in other pool-type SFRs. [ABSTRACT FROM AUTHOR]

• A novel combined passive cooling system for reactor and SFP is proposed. • The C-type heat exchanger and two-phase loop thermosyphon are utilized. • Two application schemes are proposed for comparative analysis. • A set of 3D numerical simulation method is established for this novel system. To evaluate the safety and reliability of a novel combined passive cooling system for tank-in-pool reactors during Station Black-Out (SBO) accidents, a numerical analysis employing ANSYS FLUENT is conducted. A model including the Reactor Pool (RP) and the Spent Fuel Pool (SFP) is established. The porous media approach and User Defined Function (UDF) are utilized to simulate the thermal–hydraulic processes of the coupling system of the primary loop, Passive Residual Heat Removal Heat Exchanger (PRHR HX), Two-Phase Loop Thermosyphon (TPLT), and spent fuel. The principal thermal–hydraulic parameters such as thermal power, heat transfer coefficients, temperature, fluid velocity, and mass flow rate are obtained via numerical computations. The results demonstrate that the system can effectively cool the reactor and spent fuel, preventing saturation boiling within pools. Consequently, this study proves its feasibility and potential advantages for tank-in-pool reactors, thereby providing important theoretical substantiation for reactor design and cooling system optimization. [ABSTRACT FROM AUTHOR]

• A Direct Reactor Auxiliary Cooling System (DRACS) has been designed to remove the core decay heat efficiently after transient accidents for a 10 MW LBE-cooled SMR. • The DRACS features a dual heat sink design, with water serving as the cooling source soon after accidents. As the water evaporated, the cooling source finally transitioned to be air. • Using RELAP5, the capability of the DRACS has been verified under the PLOFA, which effectively removed core decay heat while ensuring that the liquid lead–bismuth was not condensed. • Sensitivity studies have been conducted to verify the robustness of the DRACS by varying initial conditions. In this paper, a Direct Reactor Auxiliary Cooling System (DRACS) has been designed to remove the core decay heat after accidents such as the core scram to ensure the safety of a 10 MW pool-type Small Modular Lead-based Reactors (SMR). The DRACS is characterized by dual heat sinks, i.e., water and air. The protect loss of flow accident (PLOFA) has been simulated by RELAP5 to illustrate the thermal–hydraulic feathers in both the primary circuit system and the DRACS. The natural circulation was established in the flow path of the primary circuit where the DRACS was in operation. The complete evaporation of water caused obvious variation of LBE temperature, however, affected LBE mass flow rate marginally. Moreover, the DRACS was efficient in removing core decay heat and preventing LBE solidification under different initial conditions. The conclusions from the current work would help with the designs of decay heat removal systems for SMR. [ABSTRACT FROM AUTHOR]

• "SCTRAN/CO2-SIMULINK, a simulation platform, is developed and validated in this study." • Control systems are designed to regulate S-CO2 Brayton cycle-cooled reactor system during operational transients. • The control system uses integral time absolute error as the performance metric, minimized by a Real Coded Genetic Algorithm. • The control system's performance is tested using linear and abrupt load-following transients. The Supercritical CO 2 (S-CO 2) Brayton cycle-cooled nuclear reactor system is characterized by its high efficiency, small size, and ability to adapt to changes in load swiftly. The use of modern control methods to enhance its stability and efficiency necessitates using a simulation tool that enables their implementation. In this study, an integrated simulation platform, "SCTRAN/CO2-SIMULINK", is developed and verified. Various control systems are designed, and integral time absolute error is adopted as a performance metric for the analysis. A real-coded genetic algorithm is introduced to optimize the performance. The control system's performance is analyzed by applying linear load-following and abrupt load transients. Analysis of simulation results indicates that the developed platform enables the application of control and optimization methods from MATLAB/SIMULINK to S-CO2 Brayton cycle-cooled nuclear reactor system. Regulation of the inlet pressure of the reactor core is proposed in order to prevent it from pressure-induced strains during the transients. [ABSTRACT FROM AUTHOR]

• An analysis program for the Brayton cycle system of the small helium-xenon cooled nuclear reactor was developed independently. • Analyzed the factors that limit system power and final parameter, and concluded that system require larger size air cooler. • The effects of different reactivity insertion ways on the transient response characteristics of the system are compared and analyzed. • The influence of constant compressor inlet temperature on system transient response characteristics is compared and analyzed. The helium-xenon cold nuclear reactor system has a good prospect in aviation power because of its simple system, small size and lightweight. In this paper, based on the Modelica language, we independently developed a Brayton cycle system analysis program for a small helium-xenon cooled nuclear reactor, and verified the key models. Firstly, a design analysis was conducted on the flow and heat transfer enhancement of the reactor coolant channel, and the factors limiting the system power and final parameters were analyzed through simulation calculations. Then, the transient operating conditions of the system under different reactivity introduction methods and controlling the constant inlet temperature of the compressor were simulated, and the dynamic operating characteristics of the system and the coupling effects of the main equipment were studied. The simulation results show that the main factor limiting the power increase and final parameter reduction of the system is that the low density of air at high altitude cannot provide sufficient low-temperature cooling air to take away a lot of waste heat. Increasing flight speed and reducing flight altitude can increase the cooling power of the air cooler and reduce the final system parameters. By introducing the same reactivity through different reactivity insertion ways, the ultimate steady-state operating state of the system is the same. By controlling the compressor inlet temperature to be constant, the system output power can be improved. The relevant system analysis results provide a reference for the design of aeronautic electric propulsion system of helium-xenon cooled nuclear reactor. [ABSTRACT FROM AUTHOR]

• Designed a multiuser lumped cooling water system for ECRH and ICRF testing bench, which is different from the traditional one system corresponding to one cooling water system. • The water supply test of two users with 400t/h flow regulation is realized to verify the operation mode of the future fusion reactor. • The integrated cooling water system can greatly reduce the construction and maintenance costs of the system and the redundancy of the cooling water capacity. In EAST(Experimental Advanced Superconducting Tokamak), the heat produced by ECRH (Electron Cyclotron Resonance Heating) and ICRF (Ion Cyclotron range of Frequencies Heating) should be respectively removed through independent cooling water system. Their capacities are typically proposed based on the full power of the ECRH and ICRF. Indeed, the operational duration of the wave system is limited, which necessitates a larger cooling water capacity redundancy, occupies a significant amount of space, and requires a substantial budget. In this paper, the cooling water system for ECRH and ICRF of CRAFT (Comprehensive Research Facility for Fusion Technology) testing bench is designed according to its requirements.Initially, the optimization of the cooling water process is carried out in accordance with the specified parameters. Thereafter, the entire process is simulated employing the AFT Fathom code.Finally, a multiuser integrated cooling water system is designed, which is different from the traditional majority one-to-one mode. And the results show that all the parameters can meet the system requirements, meanwhile the redundancy of the cooling water capacity can be greatly reduced by flow regulation.The cooling water integral structure will provide a new idea for the design of fusion reactor cooling water systems. [ABSTRACT FROM AUTHOR]

A simple reactor design for the conversion of CO2 methanation into synthetic methane based on free convection is an interesting option for small-scale, decentralised locations. In this work, we present a heat-management design of a multi-tubular reactor assisted by CFD (Ansys Fluent®) as an interesting tool for scaling-up laboratory reactor designs. The simulation results pointed out that the scale-up of an individual reactive channel (d = 1/4′, H = 300 mm) through a hexagonal-shaped distribution of 23 reactive channels separated by 40 mm allows to obtain a suitable decreasing temperature profile (T = 487–230 °C) for the reaction using natural convection cooling. The resulting heat-management configuration was composed of three zones: (i) preheating of the reactants up to 230 °C, followed by (ii) a free-convection zone (1 m/s air flow) in the first reactor section (0–25 mm) to limit overheating and, thus, catalyst deactivation, followed by (iii) an isolation zone in the main reactor section (25–300 mm) to guarantee a proper reactor temperature and favourable kinetics. The evaluation of the geometry, reactive channel separation, and a simple heat-management strategy by CFD indicated that the implementation of an intensive reactor cooling system could be omitted with natural air circulation. [ABSTRACT FROM AUTHOR]

A numerical analysis of an unsteady radiative and viscous dissipative fluid flow across a semi-infinite oscillating vertical plate with a constant temperature and mass diffusion is examined here. The fluid under consideration is optically thin gray, emitting, heat-retaining, and non-dispersing. A finite-difference approach known as the Crank−Nicolson strategy was applied to find the solution to the limitless governing equations. Applications of such engineering problems can be found in fields such as aerospace, solar power, the cooling of nuclear reactors, and chemical and mechanical engineering. Based on the computational solutions, the impacts of distinct physical parameters, such as velocity, concentration, temperature, skin friction, Nusselt number, and Sherwood number profiles, are portrayed pictorially and imparted. The numerical solution of the velocity profile developed in this study fits extremely well with what was previously published and validated. Furthermore, we discovered that plate oscillation, radiation, and viscous dissipation parameters surprisingly influence the flow pattern. [ABSTRACT FROM AUTHOR]

Modern deterministic codes for modeling the safety-dominant nuclear power plant accident groups with reactor circuit leaks (including inter-circuit leaks in steam generators) do not determine the conditions and consequences of the processes of hydrodynamic instability of the coolant. Work on preventing conditions of hydrodynamic instability was carried out for individual safety systems of the reactor installation in which specific equipment is installed. The occurrence of hydrodynamic instability of the coolant changes the conditions of heat exchange in the active zone of the reactor, leads to cyclic high-amplitude hydrodynamic loads on the internal body structures and other negative effects. The purpose of the presented work is to create methods for modeling conditions of hydrodynamic instability of the coolant during accidents with reactor circuit leaks. The method of implementation of the work is based on the development of a thermodynamic method for determining the conditions of hydrodynamic instability in the conditions of accidents with reactor circuit leaks. On the basis of the approximate method, the minimum dimensions of the reactor circuit leaks from the HPP in the area of hydrodynamic instability of the coolant were determined: for the emergency cooling system of the reactor active zone with high-pressure pumps - 400 mm; for the system of emergency cooling of the active zone of the reactor with low-pressure pumps - 100 mm. The occurrence of hydrodynamic instability of the coolant changes the conditions of heat exchange in the active zone of the reactor, leads to cyclic high-amplitude hydrodynamic loads on the internal body structures and other negative effects. The original method of determining the criteria for the occurrence of hydrodynamic instability of the coolant at the stage of starting the pumps of safety systems in the event of an accident with reactor circuit leaks is presented. [ABSTRACT FROM AUTHOR]

In this presented study, the cooling problem of the I.T.U. Triga Mark-II reactor has been handled and analyzed, and solutions were proposed. First of all, a thermal model of the reactor, heat exchanger, and cooling tower trio was established in the reactor. With this model, which was obtained with the help of experimental data, the parameters affecting the change of reactor water temperature over time were determined, and significant findings were obtained by investigating the possibilities of increasing the cooling power of the existing system. Then, using these mathematical equations, the effects of parameters that can affect the power of the reactor cooling system are investigated. The parameters affecting the cooling power are the cooling water flow rates in the second cooling circuits and the deposited layer that may exist as a result of numerical calculations. Different models have been created with machine learning algorithms (page regression, decision tree) to estimate the effect of the deposit layer. The mathematical and predictive models obtained with the experimental data for the heat transfer coefficient of the deposit layer, hbd, were compared. The pace regression algorithm modeled the hbd values with the least error rate (RMSE: 1.66) among the models. It has been calculated that the average tank water temperature will decrease by approximately 3.5 °C if the deposits layer is cleared. [ABSTRACT FROM AUTHOR]

• The proposed LFR-180 design presenting a compact, passive, and long-life LFR. • LFR-180 incorporates H-OTSG for steam generation and DRACS for decay heat removal. • LFR-180 is a fully passive reactor requiring no intervention in 149.1 h after shutdown. • H-OTSG serve as active safety system heat exchangers can effectively control the PCOT. • LFR-180 startup process is straightforward with no need for auxiliary heat. The lead–bismuth eutectic cooled fast reactor (LFR) emerges as a cutting-edge technology in nuclear reactor design. This study introduces a compact, passive, and long-life LFR reactor LFR-180. The LFR-180 incorporates modular helical-coiled once-through steam generators (H-OTSG) and direct reactor auxiliary cooling systems. This integration enables the generation of superheated steam and passive safety features of the LFR-180, eliminating the necessity for a water-steam separation device and contributing to enhanced power generation efficiency. A detailed transient thermal hydraulic analysis is conducted with tailored models for LBE solidification and H-OTSG thermal hydraulics. Results demonstrate that LFR-180 is a fully passive reactor, requiring no intervention up to 149.1 h after shutdown. The H-OTSG can serve as the heat exchangers of the active safety systems that can effectively reduce the peak core outlet temperature from 1003.6 K under the SBO accident to 724.2 K after shutdown. [ABSTRACT FROM AUTHOR]

High-temperature heat pipes, utilizing alkali metals as working fluids, are essential heat transfer components in cooling systems of nuclear reactors. To assess the long-term isothermal behavior and heat transfer performance variation of high-temperature sodium heat pipes, this research presents the design and construction of a long-duration experimental test rig for high-temperature heat pipes. A 100-hour experimental investigation was conducted under operating conditions of 900 °C. The results demonstrate that the heat pipes can operate stably for extended periods after startup. The average temperature at the condenser section gradually increased, while the overall temperature difference fluctuation decreased. The magnitude of temperature difference reduction was measured to be 2.3 °C, and the effective thermal resistance of the heat pipe decreased to 0.0639 K/W. These results suggest an enhancement in both the isothermal performance and heat transfer capability characteristics of the sodium heat pipe after long-term testing. This study provides valuable insights for the design and assessment of high-temperature heat pipe systems in nuclear reactor cooling applications. [ABSTRACT FROM AUTHOR]

Carbon-fiber-reinforced carbon matrix (C f /C)-Mo30Cu brazed joints play an important role in the cooling systems of thermonuclear reactors. However, the limited contact area of heterogeneous interfaces severely limits the heat transfer efficiency. To overcome this drawback, we prepare three-dimensional porous interfaces by pre-oxidizing the C f /C composite. Results show that circular gaps are formed between the carbon fibers and the pyrolyzed carbon after pre-oxidization at 600 °C in air. Fster braze penetration in the C f /C composite is achieved, and the heat-transfer area across the interface is dramatically increased. The room-temperature thermal conductivity of the joints reaches a maximum value of 146 W·m−1·K−1 at a pre-oxidation time of 2 min; this value is 30 % higher than that obtained without treatment. The enhancement in thermal conductivity is mainly attributed to the increased contact area at the interface between the brazing seam and the C f /C matrix, which provides more channels for heat transfer. This method of significantly improving the thermal conductivity is an important guide for the thermal management of thermonuclear reactors. [Display omitted] • Pre-oxidation process constructs 3D structures on the surface of C f /C composite. • The effective contact area between the brazing seam and the C f /C composite is significantly increased. • Enhanced interfacial microstructure improves thermal conductivity and strength of joints. • Thermal conductivity of joints as high as 123 W·m−1·K−1 even when tested up to 500 °C. [ABSTRACT FROM AUTHOR]

• Models of a scaled Water-cooled Reactor Cavity Cooling System (RCCS) were developed using STAR-CCM + and the TRACE system thermal hydraulics code and were validated against two experimental data sets. • Radiation heat transfer was modeled as the primary heat transfer mechanism in the TRACE RCCS model, and subsequent heat flux profiles were adapted to provide boundary conditions to the CFD model. • CFD modeled conjugate heat transfer from the cooling panel to the internal water coolant and the subsequent buoyant forces within the RCCS. • Results showed asymmetric flow profiles due to the geometry of the RCCS and significant thermal stratification was found in the Low Reynolds CFD case which could lead to higher thermal stresses within the RCCS. High Temperature Gas-cooled Reactors (HTGRs) have garnered interest for use in commercial energy production and for providing high temperature process heat for industrial applications. One of the passive safety systems suggested for decay heat removal in accident conditions is known as the Reactor Cavity Cooling System (RCCS). Two steady-state experimental data sets with varying Reynolds numbers (Re = 2490, 10279) conducted with the TAMU RCCS experimental facility were modeled and simulated using the computational fluid dynamics code STAR-CCM + and the TRACE system thermal hydraulics code for code validation. It was found that both the CFD and system code models showed good agreement with respect to the global conditions of the internal flow within the RCCS. Comparisons of the local temperature measurements showed that the TRACE code tended to overpredict temperatures toward the bottom of the RCCS cooling panel but showed good agreement above the midpoint of the RCCS risers. The CFD model found significant thermal stratification and recirculation in the Low Reynolds case due to the asymmetric heating profile applied to the riser panel. The bottom and lower manifold geometries of the RCCS design were also found to produce asymmetrical flow conditions which led to differences in local riser temperatures. [ABSTRACT FROM AUTHOR]

The chemical behavior of cesium molybdate (Cs2MoO4) in light water reactors during severe nuclear accidents remains unexplored. This study demonstrated the deposition behavior of Cs2MoO4 on Type 304 stainless steel (SUS304) at 1530 to 530 K under dry (Ar) and humid (Ar + H2O) conditions. Cesium molybdate was partially decomposed on the SUS304 surface, thereby inducing the oxidation of iron (Fe) and chromium (Cr) under the dry condition. Molybdenum (Mo) metal and molybdenum dioxide (MoO2) were detected on the surface, while Cs coexisted with chromium in the oxide layer at 1500 K. Both Cs2MoO4 and Mo metal were identified on the SUS304 surface at 1230 K. Under the humid condition, the oxidation of the SUS304 was affected by Cs2MoO4 vapor. Molybdenum was detected in the form of spots in the iron oxide layer, while cesium was not detected above 1500 K. Molybdenum metal was detected on the surface of SUS304 oxide at 1230 K. Cesium molybdate was deposited on the SUS304 at 730 to 530 K under both the dry and humid conditions. The results are discussed in relation with the thermodynamic model of the Cs-Fe-Cr-Mo-O system. Thus, the chemical behavior of Cs2MoO4 at the interior of the reactor cooling system is elucidated. [ABSTRACT FROM AUTHOR]

Hydraulic transient is a flow situation where the velocity of the flow and the pressure change rapidly with time in pipes networks filled with fluid. Water hammer is one of the consequences that may occur accidentally in the reactors cooling system. The pressure level can exceed the limits given by the manufacturer, and it leads to the failure of the pipes integrity or nuclear fuel elements fixation in the core. Cooling system involves many components, and when a flow control component changes abruptly its status (for example, a valve closing or pump trip), it causes rapid changes in velocity and propagates pressure pulses in short time. The cooling system is equipped with flapper valve and flywheel which overcomes the transient issues. A pump or check valve may be stopped or closed abruptly and may cause bad impacts on the system which is simulated with soft trip using flywheel and without flywheel in case of its failure for any accidental reasons. In addition the abrupt valve closure is deemed as a major issue in the system if occurred. AFT Impulse 5.0 software is used to simulate and study the transient of each case. The simulation appears that the reactor cooling system may suffer from surge that may occur during the sudden closure of check valve but in other hand it can be avoided by regular maintenance of flapper valve to open in timely manner. [ABSTRACT FROM AUTHOR]