Safety Of Nuclear Power Plants Research Articles

A High-Performance Gamma Spectrometer for Unmanned Systems Based on Off-the-Shelf Components

Since the Fukushima Daiichi Nuclear Power Plant accident in March 2011, the technology available for unmanned aerial vehicles (UAVs) for radiation monitoring has improved greatly. Remote access to radiation-contaminated areas not only eliminates unnecessary exposure of civilians or military personnel, but also allows workers to explore inaccessible places. Hazardous levels of radioactive contamination can be expected as a result of accidents in the nuclear power industry or as a result of the intentional release of radioactive materials for terrorist purposes (dirty bombs, building contamination, etc.). The possibility to detect, identify, and characterize radiation and nuclear material using mobile and remote sensing platforms is a common requirement in the radiation sensing community. The technology has applications in homeland security and law enforcement, customs and border protection, nuclear power plant safety and security, nuclear waste monitoring, environmental recovery, and the military. In this work, the authors have developed, implemented, and characterized a gamma-ray detection and spectroscopy system capable of operating on a UAV. The system was mainly developed using open-source software and affordable hardware components to reduce development and maintenance costs and provide satisfactory performance as a detection instrument. The designed platform can be used to perform mapping or localization tasks to improve the risk assessment process for first responders during the management of radiological and nuclear incidents. First, the design process of the system is described; the result of the characterization of the platform is then presented together with the use of the prototype installed on a UAV in an exercise simulating a radiological and nuclear contamination scenario.

Safety Analysis in VVER-1000 Due to Large-Break Loss-of-Coolant Accident and Station Blackout Transient Using RELAP5/SCDAPSIM/MOD3.5

It is crucial to do safety evaluation of different postulated transient scenarios in actual nuclear power plants (NPPs). Some of the common analyzed scenarios are primary coolant tube rupture and station blackout (SBO). In this research, it was supposed that after establishment of a steady-state condition, an instantaneous guillotine large-break loss-of-coolant accident (LB-LOCA) of 850-mm inside diameter in one of the reactor vessel cold legs occurred, accompanied with SBO. The event progression and the variation of different reactor parameters like loop pressures, mass flow rates, fuel and clad temperature, injection rate of accumulators (ACCs), decay, and reactor power were investigated using the RELAP5/SCDAPSIM/MOD3.5 thermal-hydraulic program. The reactor consequences due to availability and unavailability of passive ACCs were compared. These kinds of analyses assist in estimating the time available to perform operator safety actions. This in turn aids in emergency planning and severe accident management. The results reveal that fuel damage decreased after the introduction of ACCs. Actuation of ACCs at their actuation setpoints provided core cooling by injecting water into the reactor core. However, ACCs alone are inadequate to contain long-term core cooling during a persistent LB-LOCA. The results obtained in the research were compared with MELCOR 2.1 and ASTEC V1.3, and a cohesive agreement was obtained. Therefore, RELAP5/SCDAPSIM/MOD3.5 is capable of modeling a LB-LOCA and SBO in VVER-1000, and it provides a significant analytical capability of safety systems for specialists in the field in NPP safety.

Status and Perspective of China’s Nuclear Safety Philosophy and Requirements in the Post-Fukushima Era

The significant impact brought by a severe nuclear accident at the Fukushima Daiichi Nuclear Power Plant (NPP) in Japan in March 2011 has made global regulators to review the requirements against severe accidents. In China, comprehensive safety inspection and external hazard safety margin assessment on NPPs were carried out, regulatory requirements on improvement measures for NPPs based on the inspection were given, the nuclear safety 5-year plan was made and executed, and the safety requirements on the new NPP design were drafted. The Nuclear Safety Law came into effect in 2018. The “Code on the Safety of Nuclear Power Plant Design” (HAF102) was revised in 2016, and relevant safety guides were developed. In this article, improvement actions and requirements about NPP safety in China over the past 10 years were reviewed, and the nuclear safety philosophy and requirements including practical elimination, classification of accident conditions, and defense in depth portable equipment were elaborated. In summary, some suggestions of NPP safety in China in the future were provided.

Study on response evaluation method for low-seismic-class piping with U-bolts

Simple support structures such as U-bolts and clamps are used for the low-seismic-class piping designed using the constant pitch span method. Recently, the safety of the piping is required during the earthquake, and the demand for safety in nuclear power plants has increased. As a result, the high seismic resistance has also been required for the low-seismic-class piping. In the response analysis, the support structure constrains the displacement in two translational directions, but not the axial displacement of the pipe. Depending on the routing of the piping, a large axial response displacement may occur, and the structural integrity of the piping and the support structure may not be ensured. Therefore, measures have been taken to add or reinforce support structures. In practice, pipes are subjected to the axial friction resistance by the support structure, which reduces the response displacement. A static element test was conducted to understand this axial resistance. In this test, the friction resistance force was measured with no tightening force. Specimens were G28 (about 3 kg) and G54 (about 6.5 kg) pipe elements. The resistance force about 3.5 to 7 N for G28 and about 7 to 15 N for G54, and the friction coefficient about 0.18 were observed. By taking this into consideration, it is expected that the response displacement during the earthquake will decrease. The development of a rational seismic evaluation method considering the frictional resistance in the axial direction will be continued.

PREVENTION OF HYDRODYNAMIC INSTABILITY CONDITIONS IN SAFETY SYSTEMS WITH PUMPS OF NUCLEAR POWER PLANTS

The study of hydrodynamic instability in the safety systems of nuclear power plants is relevant. In the deterministic analysis of the safety of nuclear power plants based on accident simulation, it is necessary to take into account the possibility of hydrodynamic instability in the operational and transient modes of safety systems. The consequences of the emergence of hydrodynamic instability in safety systems can be following: a significant deterioration of the heat and mass exchange conditions in the reactor and steam generators during the heating process, an increased power of thermo-hydro-shock on the equipment of the nuclear installation and other negative effects. The negative consequences of the hydrodynamic instability in the safety systems of nuclear power plants can be a significant deterioration in the conditions of heat-mass exchange and the thermal water hammers with increased power. The main reasons for the hydrodynamic instability in safety systems are inertial lag in the response of control valves and head-flow characteristic of pumps to “fast” changes in hydrodynamic parameters in nuclear power plant systems. The purpose of this work is to determine methods for minimizing the impact of the causes of hydrodynamic instability in security systems. The methods of substantiating effective structural and technical parameters of damping devices to prevent conditions of hydrodynamic instability in stationary working and transient modes of safety systems with pumps are given. A method for substantiating effective design and technical parameters of damping devices to prevent conditions of hydrodynamic instability in transient modes of starting pumps of safety systems is presented. Stability conditions in stationary operating modes of the initial steam-gas volume of damping devices are determined. The minimum permissible dimensions of damping devices that meet the conditions of hydrodynamic stability in the transient modes of SB pumps are determined.

State identification of large diameter wet steam pipeline in nuclear power conventional island

The safety of nuclear power plant is an inevitable condition for maintaining the long-term and stable development of nuclear power. Pipeline vibration is one of the common causes of power plant operation faults. Therefore, the state identification of pipeline vibration is very important. Taking a nuclear power heating pipeline as an example, the original vibration signal is decomposed and reconstructed by Wavelet Packet Transform, and the energy eigenvectors in X, Y and Z directions are established. In order to analyze the state of pipeline more accurately, two recognition algorithms of support vector machine (SVM) and SOM neural network are proposed. By comparing the classification effect and classification accuracy of the SVM before and after optimization, the optimized SVM and SOM neural network, it is concluded that the optimized SVM has better effect on pipeline state recognition, and the effectiveness of this method on pipeline state recognition is verified.

Design and Features of Advanced Decision Support Systems

The design of systems important to the safety of nuclear power plants (NPPs) is based on compliance with regulatory documents’ requirements and ensuring the object's reliable operation throughout its entire life cycle. One of the aspects of increased requirements for the reliable and safe operation of NPP process control systems is perfecting human-operator information decision support systems. In the paper, an approach is considered in which the operator support information system kernel is a flexible modeling program complex, which, depending on the tasks being solved, can be used as a digital twin or a digital shadow. The digital twin furnishes the past state of the product, the current state, and probable future states, which is especially significant for its use. The paper also discusses methods for assessing the proximity of a physical object and its digital twin based on a consistent and normalized measure of dependence of random vectors.

A Lack of Knowledge Acquisition Can Impair Nuclear Power Plant Safety

Abstract A lack of knowledge acquisition (LOKA), among engineering staff members in supply firms (e.g., vendors), owner/operator utilities (e.g., licensees), and in regulatory agencies (e.g., the NRC), can impair nuclear power plant (NPP) safety in ways that can persist throughout the operating lifetime of an NPP. A LOKA occurs when experienced technical reviewers fail to pass enough information, or technology, to less-experienced technical reviewers. The existence of a LOKA, among technical reviewers, can lead to errors and omissions that can result in misleading or incomplete licensing bases. Eight examples of errors and omissions are presented, each of which is evaluated in the context of physical phenomena, logic, licensing strategy, and effects upon regulation. These errors and omissions could be attributed to several causes, one of which could be a LOKA. Reliable attribution to a LOKA or its causes is not directly addressed, since attention is focused principally upon the safety implications of errors and omissions that may possibly, but not exclusively, be due to a LOKA. The epistemology of a LOKA, which may consider training or human relations, is generally addressed in other studies, which apply to issues that affect more than the nuclear power industry. If those who design, analyze, license, operate, and regulate NPPs do not adequately understand and apply proven engineering principles, standards, and established regulations, critical thinking, and sound logical reasoning, then it could be said that a LOKA exists. A LOKA could hamper the development of defensible conclusions in safety analyses, viable licensing strategies, and fair regulatory judgments. Eight examples of errors and omissions are presented, each of which leads to a conclusion that seems to conflict with an industry standard, a federal regulation, an engineering principle or physical phenomenon, or just plain logic. The examples are generally evaluated in accordance with the requirements of a well-known, oft-cited nuclear industry standard, which is now almost half a century old. This standard was published in 1973 by the American Nuclear Society (ANS) (1973, Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants, La Grange Park, IL, ANS-N18.2-1973). It expresses the fundamental principle of nuclear safety and licensing, which is applied by vendors, licensees, and regulators alike. This Standard defines nuclear safety criteria and plant design requirements for plant operating situations or events according to their expected frequencies of occurrence. Those events that have high frequencies of occurrence must not pose a danger to the public. Events that could pose the greatest danger to the public must be limited, by design, to extremely low expected frequencies of occurrence. This concept is implemented by grouping postulated plant situations (or events) into categories that are defined according to their expected frequencies of occurrence. Licensees are required to present analyses and evaluations of the events, in each category, to demonstrate that the events’ consequences do not exceed the category’s specified acceptance limits. Furthermore, licensees are required to demonstrate that certain events would not develop into more serious events (e.g., events that would be grouped into more serious, higher-consequence categories), without the concurrent occurrence of independent faults. That is, the Standard (American Nuclear Society, 1973, Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants, La Grange Park, IL, ANS-N18.2-1973) requires that NPPs must be designed in a way that does not allow high-frequency, low-consequence events to degrade into high-frequency, high-consequence events. The errors and omissions that are considered in the example evaluations could be due to a LOKA, among other possible causes. In each example, a LOKA is sufficient; but not necessary, to produce the noted errors and omissions. Attention is focused upon the safety implications of a LOKA, not its epistemology. The LOKA in the nuclear power industry, including its regulators, is ongoing, since it is not recognized and remedied.

Development of a Buoy-Borne Underwater Imaging System for In Situ Mesoplankton Monitoring of Coastal Waters

This article reports the development of an underwater imaging system and its trial on a moored surface buoy for <i>in situ</i> plankton monitoring of coastal waters. The imager features shadowless white light illumination by an orthogonal lamellar lighting design, resulting in high-quality underwater darkfield color imaging of planktonic particles in the size range of &#x223C;200 <i>&#x03BC;</i>m to 40 mm and effective reduction of zooplankton phototaxis. Through raft and buoy trials, 46&#x2009;804 plankton and suspending particle images have been annotated through a human&#x2013;machine mutually assisted effort into a data set with 90 categories. In the meanwhile, a deep learning model based on a triclassification VGGNet-11 and multiclassification ResNet-18 convolutional neuron networks in a two-staged hierarchy has also been trained and developed. The model has been applied with human supervision to semiautomatically analyze a total of 1&#x2009;545&#x2009;187 images obtained from a buoy trial for six months from late spring to early winter of 2020. The high temporal resolution results well documented the variation of the mesoplankton community structure in two time series of 38 days in summer and 54 days in autumn of the target sea region. In addition, the dominant species in the trial period and a zooplankton outbreak that had threatened the safety of the nearby nuclear power plants were quantitatively analyzed. The system is expected to become a new entry into the toolkit for marine surface buoy platforms to upscale their capabilities for more comprehensive <i>in situ</i> plankton monitoring.

Some Aspects of Intelligent Human-Operators Decision Support Systems for NPP

The design of systems important to the nuclear power plants (NPPs) safety must base on compliance with regulatory documents' requirements and ensuring the object's reliable operation throughout its entire life cycle. One of the aspects of increased requirements for the safe and dependable operation of NPP instrumentation and control systems is perfecting human-operator information decision support systems. In the paper, an approach is considered in which the operator support information system kernel is a flexible modeling complex, which can be used as a digital twin. The paper also discusses methods for assessing the proximity of a physical object and its digital twin based on a consistent and normalized measure of dependence of random vectors.

NSGA-II algorithm-based LQG controller design for nuclear reactor power control

The design of reactor power control system is closely related to the safety and economy of nuclear power plants. To improve its power control performance, this study takes a pressurized water reactor as the research object. The nonlinear dynamic mathematical model and its state space model were first derived; Then the Linear Quadratic Gaussian (LQG) optimal control method was adopted to design the reactor power controller; Furthermore, the multi-objective optimization of the linear quadratic regulator (LQR) weighting coefficient is carried out based on the elitist nondominated sorting genetic algorithm version II (NSGA-II). The results show that the LQG controller optimized based on NSGA-II method not only has fast response, high control accuracy and good stability, but also has good robustness and can obtain superior reactor power control performance.

A Dialogic Perspective on Managing Knowledge Differences: Problem-solving while building a nuclear power plant safety system

To manage knowledge differences, existing research has documented two sets of practices: traversing and transcending knowledge boundaries. What research has yet to explore, however, is the dynamics through which traversing or transcending practices emerge in response to a particular problem situation. Using a qualitative, inductive study of the problem episodes encountered by groups of experts working on a large-scale project to build the safety system for a nuclear power plant, we observed that the emergence of traversing or transcending depended on how experts interpreted problems and initiated dialogues around specific problems. Our work provides insight into the condition through which knowledge integration trajectories may emerge.

Experimental study on the floor responses of a base-isolated frame structure via shaking table tests

Seismic isolation represents one of the innovative means to improve the seismic safety of nuclear power plants (NPPs) against strong earthquakes. Many difficulties remain, though, in applying seismic isolation systems to real NPP structures due to the variability of the material properties of seismic isolators and the nonlinear behavior characteristics of the base isolation system in spite of the elastic behavior of the superstructure. Specifically, there is a limit to estimate the dynamic properties of base isolation systems for numerical analysis. The in-structure response of a base isolated-structure, which is related to equipment safety installed on the floors, is more sensitive than the base shear of the superstructure to the nonlinear properties of the isolation system. In this study, the complex behavioral characteristics of a base-isolated elastic frame structure are experimentally investigated through shaking table tests. Variation in response according to the frequency contents of the input earthquake is investigated, and the earthquake responses of the base-isolated structure for each frequency range are analyzed. In addition, the nonlinear increment of the in-structure response spectra to input ground motion intensity is evaluated, along with the nonlinear behavior of the seismic isolators. The experimental results obtained in this study can help identify various behavioral characteristics of base-isolated structures, with the nonlinear properties examined here acting as important references for evaluating the seismic safety of base-isolated NPPs.

Crack fault diagnosis of rotating machine in nuclear power plant based on ensemble learning

Crack faults in rotating machines can cause machine shutdown or scrapping, endangering the normal operation and safety of nuclear power plants. Intelligent diagnostic techniques based on machine learning have the potential to diagnose crack faults. However, problems such as scarcity of field fault data and high noise of plant measurements pose challenges to the application of machine learning. This study proposes an ensemble learning approach to mitigate the negative impacts of the problems. Ensemble learning is a strategy for combining multiple machine learning models into a composite model. The basic idea of ensemble learning is that even if one model makes a mistake, other models can correct it. Case studies based on bearing and gear system fault experiments show that the proposed ensemble learning models have better diagnostic results than the single model in the presence of noise and small data.

Analysis of numerical studies on the thermal-hydraulic and neutronic thermal-hydraulic stability of supercritical water reactors

The paper presents a review of modern studies on the potential types of coolant flow instabilities in the supercritical water reactor core. These instabilities have a negative impact on the operational safety of nuclear power plants. Despite the impressive number of computational works devoted to this topic, there still remain unresolved problems. The main disadvantages of the models are associated with the use of one simulated channel instead of a system of two or more parallel channels, the lack consideration for neutronic feedbacks, and the problem of choosing the design ratios for the heat transfer coefficient and hydraulic resistance coefficient under conditions of supercritical water flow. For this reason, it was decided to conduct an analysis that will make it possible to highlight the indicated problems and, on their basis, to formulate general requirements for a model of a nuclear reactor with a light-water supercritical pressure coolant. Consideration is also given to the features of the coolant flow stability in the supercritical water reactor core. In conclusion, the authors note the importance of further computational work using complex models of neutronic thermal-hydraulic stability built on the basis of modern achievements in the field of neutron physics and thermal physics.

Knowledge, belief, and attitude of Bangladeshi youth toward the development of nuclear power

Public support is one of the major challenges for maintaining a sustainable nuclear power program for any country. Although Bangladesh’s decision to build its maiden nuclear power plant (NPP) has received significant scholarly attention, the study regarding public perception is largely ignored. To fill this gap, this study investigated how Bangladeshis view the government’s decision to build NPPs. The study evaluated Bangladeshi youth’s knowledge, belief, and attitude toward the development of nuclear power. Applying the survey research technique, we found that 49% of the 450 respondents were positive toward NPPs, while 12% had no idea about NPPs. Women, in comparison to men, were found to have less knowledge of nuclear energy. Additionally, 41% of the respondents believed that the decision for establishing the country’s first NPP at Rooppur was right. However, 54% of the respondents expressed concerns over the safety, security, and sustainability of NPPs. Most importantly, a significant distrust among the respondents was observed in the capability of the regulator and operator for maintaining the NPP in a safe, secured, and sustainable manner. The study concludes that more policy actions are necessary to increase public support for nuclear power.

Feasibility of producing cement-based sacrificial materials with strontium ferrite: A preliminary study

The safety of nuclear power plant can be improved via core catcher, and cement-based sacrificial materials are widely used in most core catchers because of their simple construction process and the low manufacturing cost. This paper carried out a preliminary investigation to explore the feasibility of using strontium ferrite to prepare cement-based sacrificial materials. The effects of strontium ferrite on the room temperature and high temperature properties of cement mortar were systematically studied in the work. In addition, the damage evolution of cement mortar containing strontium ferrite subjected to elevated temperatures was also identified according to the damage mechanics theory. It was found that, (1) the addition of strontium ferrite led to the deterioration of the mechanical properties, microstructure and pore structure of cement mortar; (2) it was feasible to prepare cement-based sacrificial materials containing strontium ferrite that met the requirements on basic properties like mechanical properties, free water content, volume stability, and the high temperature integrity by controlling the content of strontium ferrite; (3) the correlation between the internal damage of cement mortar and the temperature could be described by Weibull Distribution Model, which provided a basis for the identification and evaluation of serious nuclear power accidents.

Dynamic Reliability Evaluation of Diesel Generator System of One Chinese 1000MWe NPP Considering Temporal Failure Effects

Loss of power supply from the diesel generator system (DGS) after loss of offsite power (LOOP) will pose great threat to the safety of GEN-II pressurized water reactors (PWR). Therefore, it is very desirable to evaluate the DGS’s reliability. The traditional analyzing tools are limited to static approaches neglecting the dynamic sequence failure behaviors, such as reliability block diagram (RBD), static fault tree (SFT). Static reliability modeling techniques are not capable of capturing the dynamic sequence-dependent failure behaviors typically existing in NPP safety systems such as DGS, and thus often overestimate the unreliability of systems. In this paper, motivated to study the effects of sequence failure behaviors, dynamic fault tree (DFT) is applied to evaluate the reliability of the DGS of one Chinese 1000MWe Nuclear Power Plant (NPP), and an integrated two-phased Markov Chain model is also developed, which can be considered as a contribution of this article. Comparative study of DGS reliability between DFT and SFT is carried out. The results indicate that compared with the result derived from the DFT model, the unreliability of DGS calculated by SFT is greatly overestimated by about one to two orders of magnitude. Therefore, DFT has a potential to improve the economy of NPP by relaxing the overestimated unreliability of nuclear power systems.

Study on Micro-Structure and Tensile Mechanical Properties of Dissimilar Metal Weld Joint Connecting Steam Generator Nozzle and Safe-End

The safe-end of a steam generator (SG) nozzle dissimilar metal weld (DMW) for pressurized water reactors (PWRs) is the weakest point of failure which is crucial for the safe operation of a nuclear power station. Related to materials micro-structures, a uniaxial stress–strain relationship is the basic input parameter for nuclear power plant design, safety evaluation, and life management. In this paper, the micro-structure and tensile mechanical properties of a DMW of a European pressurized water reactor (EPR) were studied. Vickers hardness tester, optical microscope, and electron back scatter diffraction were used to analyze the micro-structure of the DMW joint. In addition, the residual strain of the DMW joint base material, heat-affected zone, weld metal, and fusion boundary region were studied. Based on digital image correlation (DIC) technology, tensile mechanical properties of the DMW joint were obtained. The results show that an accurate tensile stress–strain relationship of dissimilar metal welded joints can be obtained by using the DIC technique, the weld is the relatively weak link, and the residual strain is concentrated in the heat-affected zone. This study provides valuable engineering information regarding nuclear power plant design, in-service performance testing, and structural analysis and evaluation of welds containing defects.

Study of a High Temperature–Resistant Shielding Material for the Shielding Doors of Nuclear Power Plants

An optimization design and application of high temperature–resistant shielding material was carried out according to the nuclear power plant source characteristics and special protection requirements such as loss-of-coolant accident (LOCA). The composition of lead–boron polyethylene shielding composite was optimized based on the genetic algorithm and Monte Carlo methods and then realized by blending modification and graft copolymerization to improve its high temperature–resistant, shielding, and mechanical properties. Then comprehensive properties such as mechanical, neutron shielding, damp heat aging, irradiation resistance, and high temperature resistance were tested. These experiments proved that the high temperature–resistant lead–boron polyethylene shielding composite has excellent performance; especially, as it is able to keep a complete structure in a high-temperature environment of up to 190°C for 48 h. Finally, the shielding composite was applied to the shielding door design of a reactor pit chamber. When the shield thickness is 60 mm, the level of the neutron dose rate was reduced by 10 times, and that of the γ dose rate was reduced by 5 times, which meets all the requirements of radiation protection safety for nuclear power plants.