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[en] The steam driven jet pump (SDJP) is a device without moving parts, in which steam is used as an energy source to pump cold water from a pressure much lower than the steam pressure to a pressure higher than the steam pressure. In this study, a design of a passive core injection system (PCIS) with SDJP is given for application to Boiling Water Reactor (BWR). The operation range of PCIS depends on the operation characteristics of the SDJP. Thus, the operation characteristics of the SDJP has been investigated in terms of independent parameters, such as water temperature, pressure, and steam inlet pressure which is the driving force of the SDJP to obtain a wide operation range. The water tank pressure is chosen as 320 kPa to increase the maximum operation pressure of the SDJP. The supply water temperature is chosen as 15 deg. C which is about room temperature, and the steam which comes from the reactor pressure vessel is assumed as saturated. The operation range of the SDJP in terms of the steam inlet pressure is obtained from 2 to 10.5 MPa. The study shows that when the discharge pressure increases, the discharge mass flow rate and the temperature increases as well. The probabilistic reliability of the PCIS with the SDJP is assessed as well. The reliability assessment of the PCIS with SDJP is investigated using a fault tree, and compared with the core injection system with turbine-driven pump. It is found that the PCIS with SDJP is more reliable than the core injection system with turbine-driven pump
[en] It is shown that it is possible to regulate the energy of each pulse of a powerful pulsed periodic reactor using an injector of relatively low power. The target in the reactor core generates the neutron pulses forced by the injector. The correlations for determination of the moment of injection are obtained. The equation connecting the mean reactor power, the intensity integral of the target and the scatter multiplicity of the pulse energy without injection is obtained. In addition to its function as a regulator, the injector plays the role of an auxiliary emergency unit. It is shown that using an injector provides a regime in which the reactor can generate power pulses in the form of periodic packets
[en] Highlights: •This paper reports an experimental study of air-water pool entrainment with side exit. •Total 150 sets of test data cover all sub-regimes of momentum controlled region. •The validity of existing correlation in low and intermediate gas flux regimes is proved. •A new correlation of pool entrainment with side exit in high gas flux regime is proposed. -- Abstract: Pool entrainment in upper plenum is an important safety related phenomenon in SBLOCA transient for advanced PWR plants like AP1000. Due to the unique geometric characteristics, the application of standard Kataoka model in the upper plenum/hotleg entrainment phenomena requires significant caution. Based on a careful review of recent pool entrainment studies, this research carried out an experimental study of air-water pool entrainment with and without side exit/outlet to better capture the prototypic geometry of upper plenum/hotleg arrangement. Total 150 sets of test data are obtained at various combinations of gas velocities and water levels. The test data covers several entrainment regions from low gas flux region to near saturation region. The results show that the side exit/outlet will reduce the entrainment rate in high gas flux region and near saturation region. The mechanism is analyzed using visualization and CFD simulation. Based on the experiment data, a new correlation of pool entrainment with side exit/outlet in high gas flux region is proposed.
[en] Highlights: • The choice and justification of operational software and systems are reviewed. • The standards IEC 61513:2011, ISO/IEC 23360-1:2006 and ISO/IEC 27032:2012 are analysed. • The question whether the Linux operational system conforms to the demands of the above standards is discussed. - Abstract: In this paper, we discuss problems of selecting and justifying operational software, especially operating systems. Operating systems must meet the requirements of international organisations (e.g., IEAE, IEC, ISO). The most important standards (IEC 61513:2011, ISO/IEC 23360-1:2006 and ISO/IEC 27032:2012) were analysed. Furthermore, the issue of conformance one of the most widespread operating system Linux to the requirements of the standards was analysed
[en] Common-mode/common-cause (CM/CC) failure and its prevention has been a serious concern in the nuclear safety community during the past few years. Since redundancy was first used in an attempt to achieve high reliability in systems, the CM/CC failure phenomenon has been inherent in system designs. The concern is that high-reliability systems are subject to compromise by human error and environmental factors. Potential CM/CC failures are the result of adding complexity to system designs. They are the product of a supersafe philosophy. The CM/CC failure phenomenon is reviewed. Classes of CM/CC failures are compiled, and the defenses against such failures and their weaknesses are surveyed. Some regulatory considerations, operating experiences, and reliability analysis methodology are touched upon. (author)
[en] Optimization problems involving multiple criteria are commonly found in a nuclear reactor design. For example, the focus on economical or safety aspects may lead to different reactor configurations. Solutions, which improve safety, may not lead to economical designs. Aiming to deal at same time with multiple criteria in reactor designs, we have developed a multiobjective genetic algorithm (MOGA) using concepts of Pareto optimality and niching techniques. Here, intended to show the advantages of using the MOGA, we applied it to a simplified two-criterion reactor core optimization problem. Using a simplification of a real-world problem, the computational cost associated to the reactor simulation could be reduced and exhaustive experiments could be done. In such experiments the MOGA could be compared not only with a standard genetic algorithm (SGA) but also with a brute force method in which the solutions search space was scanned. The obtained results have shown that the use of MOGA in such kind of problem enhances the quality of the optimization outcome, providing a better and more realistic support to the nuclear engineer decision
[en] Highlights: • This paper provides a comprehensive overview of dependability analysis. • Dependability evaluation taxonomy includes metrics, threats, means, and techniques. • The limitations of the dependability analysis process are analyzed. • Highlights: various gaps, challenges and needs in the context of such systems. • Direction for future research is suggested to extend the furthest scope of research. - Abstract: Safety critical systems progressively used in domains such as nuclear power, transport, medical and information systems are often concerned with a formal process of dependability certification. The intent of dependability process is to ensure that these systems will deliver the expected services to its users. In order to ensure the dependability of large safety-critical systems, the software engineer or security professional needs a thorough knowledge of the process of dependability analysis. In the past several decades, a significant amount of attention has been devoted to the dependability assessment of safety-critical control systems from some perspectives such as reliability, availability, safety, and security. However, for analysis of the critical systems, there is no any universal accepted rigorous dependability analysis process, which helps to choose the metrics, techniques and methodologies for the dependability evaluation of such critical systems. This paper provides a comprehensive detailed literature survey in order to investigate different metrics, threats, means, techniques and methodologies to ensure the dependability of computer-based critical systems. The limitations of these elements are also analyzed with respect to their applicability in SC systems. In addition to this, highlighted various issues (gap), challenges and needs in the context of such systems. The direction for future research is suggested to extend the future scope of research. The purpose of this paper is to interpret a rigorous review concept, of relevance across a wide range of affairs. Therefore, this work helps to the academicians, researchers, and practitioners to put this into practice, analyze the shortcomings of existing research and identifying the open areas that are important for the related community.
[en] International regulations for nuclear power plants strictly prescribe the design requirements for local impact loads, such as aircraft engine impact, and internal and external missile impact. However, the local impact characteristics of Steel-plate Concrete (SC) walls are not easy to evaluate precisely because the dynamic impact behavior of SC walls which include external steel plate, internal concrete, tie-bars, and studs, is so complex. In this study, dynamic impact characteristics of SC walls subjected to local missile impact load are investigated via actual high-speed impact test and numerical simulation. Three velocity checkout tests and four SC wall tests were performed at the Energetic Materials Research and Testing Center (EMRTC) site in the USA. Initial and residual velocity of the missile, strain and acceleration of the back plate, local failure mode (penetration, bulging, splitting and perforation) and deformation size, etc. were measured to study the local behavior of the specimen using high speed cameras and various other instrumentation devices. In addition, a more advanced and applicable numerical simulation method using the finite element (FE) method is proposed and verified by the experimental results. Finally, the experimental results are compared with the local failure evaluation formula for SC walls recently proposed, and future research directions for the development of a refined design method for SC walls are reviewed.
[en] In this work, we briefly describe the accelerator breeder reactors (ABR) and their possible uses both for production of energy and isotopes. This study indicates that ABRs can produce fuel, which would generate 2-15 times the electrical energy, ABRs consume. The energy gain depends on the type of ABR used. ABRs should also have several important advantages in safety over the modern breeder reactors. First, they have criticality less then 1 which makes an accident less likely. Second, they can be turned off any moment when their accelerator is turned off.
[en] Highlights: • A BBN model that estimates the number of software faults and reliability is proposed. • A model was established based on the SDLC and software-self characteristics. • Three rounds of expert elicitation were used to estimate the BBN model parameters. • A BBN model was applied to target digital protection software to assess its feasibility. - Abstract: As the instrumentation and control (I&C) systems in nuclear power plants (NPPs) have been replaced with digital-based systems, the need has emerged to not only establish a basis for incorporating software behavior into digital I&C system reliability models, but also to quantify the software reliability used in NPP digital protection systems. Therefore, a Bayesian belief network (BBN) model which estimates the number of faults in a software considering its software development life cycle (SDLC) is developed in this study. The model structure and parameters are established based on the information applicable to safety-related systems and expert elicitation. The evidence used in the model was collected from three stages of expert elicitation. To assess the feasibility of using BBN in NPP digital protection software reliability quantification, the BBN model was applied to the Integrated Digital Protection System–Reactor Protection System and estimated the number of defects at each SDLC phase and further assessed the software failure probability. The developed BBN model can be employed to estimate the reliability of deployed safety-related NPP software and such results can be used to evaluate the quality of the digital I&C systems in addition to estimating the potential reactor risk due to software failure.