Nuclear Reactor Field Research Articles

Sound waves tracing method in temperature–velocity coupling field of nuclear reactor

Turbulent coolant flow in a nuclear reactor’s primary loop leads to temperature non-uniformity in the heat pipe section. Utilizing sound wave temperature measurement techniques mitigates calculation errors caused by temperature variations. However, complex temperature–velocity coupling induces significant bending of sound wave paths, affecting measurement accuracy. This study proposes a triangular ray tracing method to accurately track sound wave paths considering coolant temperature gradients and flow dynamics. Experimental validation using finite element simulation data from the Hua-long Pressurized Reactor primary loop demonstrates improved reconstruction accuracy by up to 0.59% compared to straight lines and by up to 0.04% compared to other ray tracing methods.

The Usage of Micro Modular Nuclear Reactors on Military Headquarters as a Prospective Solution to Achieve Energy Security and Improve National Defense

Recently, there has been a trend of developing Micro Modular Reactors in the Small Nuclear Reactor field that can produce up to 15 MW(e) of power. Micro Modular Reactor is the fourth-generation technology which is included in the sub-category of Small Modular Reactor or Small Modular Reactor. This reactor produces electricity and heat that is safe, clean and economical which is believed to be better than large-scale Nuclear Power Plants. Micro-reactors have several technological advantages in that they can be manufactured in a factory setting, transported and assembled more easily, and can be integrated with renewable micro-power grids. Through the literature study method, this research analyzes the prospect of using Micro Modular Reactor technology for military headquarters in order to support national defense. The results of the analysis of this study conclude that MMR technology is worth trying to be applied in Indonesia, especially for electricity generation in state institutions, but it is necessary to carry out a further feasibility analysis of the possibility of state losses as a form of the ideas put forward in this study.

A Review of the Latest Developments in the Field of Additive Manufacturing Techniques for Nuclear Reactors

This review paper provides insights the into current developments in additive manufacturing (AM) techniques. The comprehensive presentations about AM methods, material properties (i.e., irradiation damage, as-built defects, residual stresses and fatigue fracture), experiments, numerical simulations and standards are discussed as well as their advantages and shortages for the application in the field of nuclear reactor. Meanwhile, some recommendations that need to be focused on are presented to advance the development and application of AM techniques in nuclear reactors. The knowledge included in this paper can serve as a baseline to tailor the limitations, utilize the superiorities and promote the wide feasibilities of the AM techniques for wide application in the field of nuclear reactors.

Impact of bioconvection on the free stream flow of a pseudoplastic nanofluid past a rotating cone

AbstractIn the current work, the repercussions of Brownian motion and thermophoresis on the three‐dimensional free stream flow of tangent hyperbolic (pseudoplastic) nanofluid past a rotating cone are explored. The tangent hyperbolic model expresses the characteristics of a shear‐thinning nanofluid. Furthermore, oxytactic microorganisms were used as mixers to actively stabilize the nanoparticles. The movement of these microorganisms within the nanofluid gives rise to a major phenomenon termed bioconvection. The flow of nanofluid past a rotating cone finds applications in the field of nuclear reactors, biomedical applications, solar power collectors, steam generators, and so on. The mathematical model is designed using Buongiorno's model that describes the two major slip mechanisms experienced by the nanoparticles moving within a fluid namely thermophoretic force and Brownian motion. The model thus formed is nondimensionalized using the apt similarity transformation. The resulting system is solved by the technique by adapting the shooting method. The velocity, temperature, concentration, and motile density profiles are graphically interpreted for different flow parameters involved in the study. It was observed that thermophoresis reduces concentration and enhances the temperature whereas Brownian motion enhanced both temperature and concentration profiles. Also, the increase in the mixed convection parameter effectively decreased the temperature of the nanofluid.

Socio-Economic and Environmental Impact of Nuclear Desalination

Nuclear desalination concept and implementation spanning 50 years are recognized as an economical viable option for water and electricity production but could not receive wider applications. This is due to various factors, in addition to technical design parameters, other factors, such as social, economic, and environmental issues, need to be considered. For this purpose, the current studies start with performing a critical and up-to-date literature review on previous investigations in the field of nuclear reactors and integrated nuclear power with desalination plants with a specific focus on performance criteria, technical specifications, etc. Reviewing and compiling the most updated technical specifications, cost estimations, and environmental data related to nuclear power and desalination plants are also important steps. Previous studies show a special focus on other important issues on nuclear desalination characteristics in countries including Saudi Arabia, Egypt, United Arab Emirates, Pakistan, India, and Kuwait. This work presents a concise review of previous works on the relevancy of other issues, such as economic, environmental, and social, associated with the use of nuclear energy in power generation and fresh water production. Preliminary assessment of possible hybrid configurations of nuclear and desalination technologies is developed and assessed by a computational program. Both operating and capital cost of the integrated plants are calculated. The simulation model is then extended to compare with other heating reactors as well for the verification analysis. The results obtained from comparative assessment depicts the accuracy of the simulation model used for preliminary assessment of the integrated nuclear desalination option. The main objective of the research is to assess the nuclear desalination plant development in terms of social, economic and environmental aspects. The results will pave the way for countries interested in developing nuclear desalination plants.

The Development of a Thermal-Hydraulic and Nonlinear Dynamic System for Molten Salt Reactors

The Molten Salt Reactor (MSR) is the most important system suggested by Generation IV for the future direction in the nuclear reactor field. For more development of the MSR reactor, the core system inside the tube is proposed by naturally circulating molten fuel salt. The nonlinear kinetic equations form a linearized function and are obtained in state-space form. Reactivity feedback and delayed neutrons are extremely important for reactor control. In this paper, a thermal-hydraulic system for the commercial computation dynamic model is proposed. Currently, there is no commercial software to simulate the natural circulation flow. The proposed method can be easily employed to detect faults and can provide a feasible overall system performance.

Thermo-Physical Properties of CaO-Fe2O3 Binary Mixture and its Application in the Field of Nuclear Reactor as Simulant Material

The simulant materials play important role in the melt coolability experiments to understand the actual scenarios of core melt accidents in the field of nuclear reactor. Simulant materials are generally oxide/ceramics materials which have the properties similar to the properties of corium (mixture of UO2 , ZrO2 , Zr alloy, Fe, Ni and Cr etc.). This work was carried out to determine the thermo-physical properties of CaO-Fe2O3 binary mixtures of different ratio of CaO and Fe2O3 (23C77F, 26C74F, 29C71F, 32CF68 and 35C65F; here the ratio is in the wt% and C for CaO and F for Fe2O3 ) and compare the properties of CaO-Fe2O3 with the properties of corium. It was observed that the thermo-physical properties of CaO-Fe2O3 binary mixture are close to the properties of corium, so it can be said that this material is simulant material. In this research work, thermal behavior of CaO-Fe2O3 was also carried out using differential scanning calorimetry (DSC). The simulant material CaO-Fe2O3 can be used in the melt coolability experiment to understand the phenomena that happened during the core melt accidents in the nuclear reactor.

Application of causality analysis on nuclear reactor systems

Causality analysis is a substantial tool for identifying cause-and-effect links between different components of a system and has been extensively used in various areas of science such as neuroscience, climatology, and econometrics. This analysis is carried out in terms of the renormalized partial directed coherence and the directed transfer function connectivity measures. Applying such analysis in the nuclear reactor field is of paramount importance since it can help in inferring cause-and-effect relationships between highly coupled processes, and consequently, it can assist on the safe and reliable operation of a nuclear power plant during the occurrence of possible disturbances or malfunctions. The effectiveness of the connectivity analysis is demonstrated through several simulated and measured test cases. Results show that the connectivity analysis is able to identify accurately the importance and central role of the activation signal when it is applied on a simple analytical model and a simulated nuclear reactor system. In addition, the application on more realistic and complex measured data sets of a Swiss boiling water reactor illustrates the capability of this analysis to indicate possible causes behind the observed anomalies or trends observed at certain conditions and, more importantly, allows a better understanding of the underlying interactions among different neutronic and thermal-hydraulic processes.

Experience in nuclear reactor physics laboratory exercises using Kartini research reactor

The experience in using Kartini nuclear research reactor for reactor physics laboratory exercises is presented. The aim of this paper is to discuss the effectiveness of nuclear reactor physics laboratory exercises using Kartini reactor. The method used is through a comparative reference study with the similar activities at other institutions. Software packages and procedural guides have been developed and used for reactor power and control rod worth calibrations, criticality experiment, experiments to study the xenon stability, neutron flux measurement, reactor start-up / shutdown operations, etc. For illustration, a computer simulation for reactor criticality experiment is described and result from practical experiment is presented. In conclusion, it has been demonstrated that the implementation of reactor physics laboratory exercises using Kartini reactor has a good performance and it was an important role in developing human resources in nuclear reactor field for the country. It is hoped that in the future, Kartini reactor will contribute to regional nuclear education and training programs.

Global parameter optimization and criterion formula of supercritical carbon dioxide Brayton cycle with recompression

Supercritical carbon dioxide (S-CO2) Brayton cycle has caught the attention in many fields, especially the field of nuclear reactors. The recompression cycle is considered to be one of the most feasible cycles to solve the pinch problem of heat exchanger. Aiming at the disadvantages of conventional cycle analysis, this paper proposed sensitivity analysis of several parameters and global parameter optimization. The result of sensitivity analysis showed that maximum pressure and efficiency of turbine had the most influential power upon relevant parameters that may determine the final cycle efficiency, followed by re-compressor efficiency, split ratio and main compressor efficiency, with total pressure loss the least. Based on the sensitivity analysis, the thermodynamic parameters optimizing with cycle efficiency as its objective function was presented in this paper. The global parameter optimization results indicated that, for the recompression cycle, the optimal pressure and split ratio could maximize the cycle efficiency. The cycle efficiency could be maximized to 0.447 by adjusting the split ratio and the maximum pressure to 0.687 and 27.75MPa, while the heat source temperature was 600°C, the efficiency of the compressor and re-compressor was 0.8 and the efficiency of turbine was 0.9. Depending on the simulation dates of optimization, criterion formulas of cycle efficiency, maximum pressure and split ratio under the best operating conditions were established. With the input values of given system parameters, the key cycle parameters under the best operating conditions could be received easily and accurately by the criterion formulas.

Qirâ’ah Siyâqîyah Nasr Hâmid Abû Zayd tentang Hak-hak Wanita dalam al-Qur’an

Starting from the problems that faced by an Egyptian female experts in field of nuclear reactors as well as a lecture at an university who divorced her husband because she refused her husband’s request for pregnancy/childbirth specially the question of divorce judgment “suddenly” assigned by religious judge based on “capping” women role only in domestic region of the family or motherhood (<em>al-ummah</em>). Abû Zayd then analyze the concept of women’s right in the Koran including its relationship with the polygamy case. At the end of the discussion paper, the authors citing the opinion of Abû Zayd that regard bad argument of supporting of polygamy parties whe base themselves on interpretations and interpretations of <em>fiqh</em> and argued on the basis of the conditions of contemporary morality. That polygamy is the best way, in the sense of it’s smaller dangers than the adultery especially when it is figured on the social-moral in the west

Characterization of Vacuum Brazing of SS 304 and Alumina Ceramic with Active Brazing Alloy

Brazing of ceramic and metal components is an emerging manufacturing technology in which a furnace or a heat source is used to join the substrates using a brazing filler material. Brazing of ceramic and metal has received considerable attention in the field of nuclear reactor, aerospace, automobile, medical and electrical engineering. A brazing alloy can either be a conventional brazing alloy which requires the ceramic surface to be metalized or an active brazing alloy which does not require metallization. A sound interfacial adhesive bond is formed during the process. Vacuum brazing of alumina and steel (SS304) with Cusil active brazing alloy has been carried out for the analysis of behavior of joint in different loading condition. Uniformity and presence of defects of the joints is confirmed by observing the Scanning electron microscope (SEM) images. The elements and phases present in the joint interface are observed by energy dispersive spectroscopy (EDX) and X-ray diffraction (XRD) analyses. Average shear stress was found to be 88 ±8MPa. The micro- Vickers hardness test showed that the micro hardness of the interface was lower than the substrates.

Whither the nuclear fuel cycle?

The government context that serves as a backdrop to energy policy-making at the turn of the millennium is characterised by trends towards deregulation of the electricity markets and increasing awareness of'sustainable development goals. Both trends are affecting nuclear energy and its fuel cycle, offering new opportunities and raising new challenges for the future. Industrial practices and research and development programmes supported by governments and/or industry are responding to evolving market and social requirements. Ongoing R&D programmes in the field of nuclear reactors and the fuel cycle are contributing to the enhancement of the technical, safety and economic performance of nuclear systems in operation and under development. Scientific and technological progress follow either evolutionary or innovative routes, but always aim at improving overall efficiency throughout the nuclear energy chain.

99/00422 Feynman-alpha measurement in a 100 kW research reactor

The heuristically based generalized perturbation theory (HGPT) methodology has been extensively used in the last decades for analysis studies in the nuclear reactor field. As well known, its use leads to fundamental reciprocity relationships from which perturbation, or sensitivity expressions can be derived, to first and higher order, in terms of simple integration operation of quantities calculated at unperturbed system conditions. Its application to subcritical, source-driven systems, now considered with increasing interest in many laboratories for their potential use as nuclear waste burners and/or safer energy producers, is here commented, with particular emphasis to problems implying an intensive system control variable.

Sensitivity analysis of source driven subcritical systems by the HGPT methodology

The heuristically based generalized perturbation theory (HGPT) methodology has been extensively used in the last decades for analysis studies in the nuclear reactor field. As well known, its use leads to fundamental reciprocity relationships from which perturbation, or sensitivity expressions can be derived, to first and higher order, in terms of simple integration operation of quantities calculated at unperturbed system conditions. Its application to subcritical, source-driven systems, now considered with increasing interest in many laboratories for their potential use as nuclear waste burners and/or safer energy producers, is here commented, with particular emphasis to problems implying an intensive system control variable.

Nuclear district heating plants AST-500. Present status and prospects for future in Russia

Development activity in the field of nuclear reactors for district heating purposes was initiated in the ex-USSR in the middle 1970s. It was stimulated by a growing deficiency of fossil fuels needed to provide for the heating of large cities in the European regions of the country. Construction of two pilot twin-reactor nuclear district stations was started in the early 1980s in the cities of Gorky and Voronezh. Nuclear heating reactor AST-500, developed by OKB Mechanical Engineering, has a number of unique engineering features, e.g. integral lay-out of primary system's components, natural coolant circulation, steam-gas self-pressurizing system, additional safety vessel around the reactor unit, etc. All these specific features, which were adopted in order to assure enhanced safety of the reactor plant, required fulfilment of an extensive R&D program. Description of the reference AST-500 design features is given in this paper, as well as respective R&D activities carried out for the design solutions and safety validation. Currently, activity in the field is directed to AST-500 design upgrading and development and also to a whole series of heating-only and co-generation reactor plants ranging from 30 to 600 MW(th) for the new generation nuclear stations.

Fault Diagnosis Via Neural Networks: The Boltzmann Machine

The Boltzmann machine is a general-purpose artificial neural network that can be used as an associative memory as well as a mapping tool. The usual information entropy is introduced, and a network energy function is suitably defined. The network’s training procedure is based on the simulated annealing during which a combination of energy minimization and entropy maximization is achieved.,An application in the nuclear reactor field is presented in which the Boltzmann input-output machine is used to detect and diagnose a pipe break in a simulated auxiliary feedwater system feeding two coupled steam generators. The break may occur on either the hot or the cold leg of any of the two steam generators. The binary input data to the network encode only the trends of the thermohydraulic signals so that the network is actually a polarity device. The results indicate that the trained neural network is actually capable of performing its task. The method appears to be robust enough so that it may also be applied with success in the presence of substantial amounts of noise that cause the network to be fed with wrong signals.

Identification of a line break by a neural network methodology

Early failure detection and identification play a fundamental role in the operation of plants for which safety and reliability are of paramount importance. Therefore many efforts have always been, and still are, devoted to the attainment of this goal. An emerging technique, nowadays widely applied in many different scientific and industrial fields, which seems worthwhile exploring is the artificial neural network methodology. In this paper we consider its application in the nuclear reactor field. We explore the feasibility of using an artificial neural network for the quantitative identification of a line break in a simulated auxiliary feedwater system. The system is coupled to four steam generators and the signals consist of thermal-hydraulic data. Here we employ a supervised, feedforward neural network trained by the back-propagation algorithm. The results confirm the validity of the approach. Indeed, not only the method allows to correctly forecast the location of the break, but also it is capable of performing the much more difficult task of estimating the break dimension, with very good accuracy. Moreover the technique appears to be robust against noisy signals.

Dynamics of Nuclear Reactors

"Dynamics of Nuclear Reactors." Nuclear Technology, 12(4), p. 415 Additional informationNotes on contributorsM. A. SchultzM. A. Schultz (BS, electrical engineering, MIT, 1939) is presently a professor of nuclear engineering at The Pennsylvania State University, after 30 years of industrial experience, 20 of which were in the nuclear reactor field. He is the author of one of the first books on reactor dynamics and control and together with his friend, Joe Harrer, succeeded in maintaining a monopoly in this field for many years.

A digital nuclear reactor control system

THE GREAT EXPANSION in the nuclear reactor field places increased emphasis on the problem of reactor control. The complexity of a nuclear reactor installation suggests the use of digital computation in the over-all power-plant control system. The chief advantages to be gained from the use of a digital system would be a sizable increase in the flexibility of the programing of the control system and an increase in the accuracy of the computations.