Safety Of Nuclear Power Plants Research Articles

Deep learning for predicting the residual concentration of sodium hypochlorite in the cooling water OF nuclear power plants

Sea water is used as cooling water in several operating nuclear power plants, as well as in other industries such as petrochemicals. Biofouling is a common problem in systems that use sea water, causing corrosion in pipes and equipment, blockages, and loss of efficiency in heat exchangers. The use of sodium hypochlorite has proven to be effective in minimizing the damages caused by biofouling, provided that the residual chlorine concentration remains within a specific range, as low dosages will not protect the equipment and high dosages can cause environmental damage. With the aim of reducing maintenance and operation costs of equipment, and particularly ensuring the operability of systems related to the safety of nuclear power plants that use sea water, this study developed deep artificial neural networks (ANN) models capable of predicting the residual chlorine concentration in the main cooling and safety systems of a nuclear power plant. This prediction is based on environmental and process parameters that influence the residual chlorine concentration. Proposed model achieved average mean squared error of 0.002 and correlation coefficient (R2-Score) of 0.934, demonstrating to be a promising tool for predicting residual chlorine concentration.

Dimensional synthesis of an Inspection Robot for SG tube-sheet

To ensure the operational safety of nuclear power plants, we present a Quadruped Inspection Robot that can be used for many types of steam generators. Since the Inspection Robot relies on the Holding Modules to grip the tube-sheet, it can be regarded as a hybrid robot with variable configurations, switching between 4-RRR-RR, 3-RRR-RR, and two types of 2-RRR-RR, and the variable configurations bring a great challenge to dimensional synthesis. In this paper, the kinematic model of the Inspection Robot in multiple configurations is established, and the analytical solution is given. The workspace mapping is analyzed by the solution-space, and the workspace of multiple configurations is decomposed into the workspace of 2-RRR to reduce the analysis complexity, and the workspace calculation is simplified by using the envelope rings. The optimization problem of the manipulator is transformed into the calculation of the shortest contraction length of the swing leg. The switching performance of the Inspection Robot is evaluated by stride-length, turning-angle, and workspace overlap-ratio. The performance indexes are classified and transformed based on the proportions and variation trends of dimensional parameters to reduce the number of optimization objective functions, and Pareto optimal solutions are obtained using an intelligent optimization algorithm.

Responses of microeukaryotic community structure to a Phaeocystis globosa bloom in a semi-enclosed subtropical bay

The occurrence of Phaeocystis globosa, a harmful algal bloom species in Chinese coastal waters, has significant impacts on marine organisms and poses a threat to the safety of coastal nuclear power plants. Although previous studies have established a close association between P. globosa blooms and the bacterial community, the relationship between the microeukaryotic community and P. globosa blooms remains poorly understood. In this study, the variations in the microeukaryotic community resulting from a P. globosa bloom were analyzed using 18S rRNA gene amplicon sequencing. The results indicated that the diversity of the microeukaryotic community during the bloom phase was significantly higher than that during the dissipation phase. The microeukaryotic community compositions varied significantly between the two phases of the P. globosa bloom. During the bloom phase, the dominant microeukaryotic was Viridiplantae, which was then replaced by Dinoflagellata during the dissipation phase. Co-occurrence network analysis showed that the relationship among the microeukaryotic community during the bloom phase was more complex than that during the dissipation phase, and the keystone taxa varied as the bloom progressed. Additionally, microeukaryotic community assembly was primarily driven by stochastic processes during the bloom phase based on the β-nearest taxon distance, whereas it was driven by both deterministic processes and stochastic processes during the dissipation phase. Overall, our findings provide novel insight into the mechanisms and interactions involved in microeukaryotic community dynamics in environments disturbed by P. globosa blooms.

Atomic blackmail – Russia-Ukraine war and Ramberg’s theory of vulnerability

ABSTRACT The Russia-Ukraine War is being fought between states committed to nuclear power. This paper references Ramberg’s nuclear powerplant (NPP) vulnerability treatise, the IAEA’s Seven Pillars of Safety metric and Pidgeon and O’Leary’s safety imagination approach to assess the safety of Ukrainian and Russian NPPs. It suggests governments should reflect on events in Ukraine, and, with reference to the above-mentioned approaches, decide whether the upside of nuclear power – the largely carbon-neutral production of electricity – outweighs the downside – giving a blackmail opportunity to hostile states. Although a cost-benefit analysis to inform decision-making is effortful, the paper suggests that failure to consider every advantage and disadvantage of nuclear power could prove costly in human and environmental terms. Drawing on Ukraine’s experience, the paper suggests the national interest is best served in time of war by generating electricity in multiple small plants rather than a few large plants.

Investigating Natural Disaster‐Related External Events at Nuclear Power Plants: Towards Climate Change Resilience

Climate change is causing rapid shifts in the intensity and frequency of natural disasters worldwide, which may have implications for the safety of nuclear power plants. As these natural disasters become more frequent and severe, their potential impact on nuclear facilities is expected to increase. Several instances of nuclear power plants being forced to shut down due to climate change have already been observed. The changing patterns of natural disasters can lead to alterations in external events affecting nuclear power plants. Therefore, it is crucial to investigate how climate change could affect the safety and operations of nuclear power plants in order to establish protective measures and mitigation strategies. This study examined external events associated with climate change that could potentially impact nuclear power plants, and both Korean and international cases were summarized. Recommendations for future tasks were proposed based on these findings. Currently, there have been no significant reported cases where climate change has severely impacted the operations and safety of nuclear power plants. However, it is essential to undertake periodic evaluations, long‐term management, and further research to ensure the safety and continued operation of nuclear power plants in the face of climate change.

Zero-suppressed ternary decision diagram algorithm for solving noncoherent fault trees in probabilistic safety assessment of nuclear power plants

Probabilistic safety assessment (PSA) plays a critical role in ensuring the safe operation of nuclear power plants. In PSA, event trees are developed to identify accident sequences that could lead to core damage. These event trees are then transformed into a core-damage fault tree, wherein the accident sequences are represented by usual and complemented logic gates representing failed and successful operations of safety systems, respectively. The core damage frequency (CDF) is estimated by calculating the minimal cut sets (MCSs) of the core-damage fault tree.Delete-term approximation (DTA) is commonly employed to approximately solve MCSs representing accident sequence logics from noncoherent core-damage fault trees. However, DTA can lead to an overestimation of CDF, particularly when fault trees contain many nonrare events. To address this issue, the present study introduces a new zero-suppressed ternary decision diagram (ZTDD) algorithm that averts the CDF overestimation caused by DTA.This ZTDD algorithm can optionally calculate MCSs with DTA or prime implicants (PIs) without any approximation from the core-damage fault tree. By calculating PIs, accurate CDF can be calculated.The present study provides a comprehensive explanation of the ZTDD structure, formula of the ZTDD algorithm, ZTDD minimization, probability calculation from ZTDD, strength of the ZTDD algorithm, and ZTDD application results. Results reveal that the ZTDD algorithm is a powerful tool that can quickly and accurately calculate CDF and drastically improve the safety of nuclear power plants.

Proposal of a novel AI-based plant operator support system for the safety of nuclear power plants

Proposal of a novel AI-based plant operator support system for the safety of nuclear power plants

Relative humidity prediction of a leakage area for small RCS leakage quantification by applying the Bi-LSTM neural networks

In nuclear power plants, reactor coolant leakage can occur due to various reasons. Early detection of leaks is crucial for maintaining the safety of nuclear power plants. Currently, a detection system is being developed in Korea to identify reactor coolant system (RCS) leakage of less than 0.5 gpm. Typically, RCS leaks are detected by monitoring temperature, humidity, and radioactivity in the containment, and a water level in the sump. However, detecting small leaks proves challenging because the resulting changes in the containment humidity and temperature, and the sump water level are minimal. To address these issues and improve leak detection speed, it is necessary to quantify the leaks and develop an artificial intelligence-based leak detection system. In this study, we employed bidirectional long short-term memory, which are types of neural networks used in artificial intelligence, to predict the relative humidity in the leakage area for leak quantification. Additionally, an optimization technique was implemented to reduce learning time and enhance prediction performance. Through evaluation of the developed artificial intelligence model's prediction accuracy, we expect it to be valuable for future leak detection systems by accurately predicting the relative humidity in a leakage area.

Optimization of prevention and mitigation of severe accidents through Research: Major outcomes and outlook of the Spanish programme

Severe accidents research activities in Spain have grown substantially since the Fukushima-Daiichi accident. Most of these research activities have been conducting by a few organizations (national research center and academia) with the support of the Regulatory Authority, utilities and engineering companies. A great number of these activities were conducted in the analytical field and their main outcomes are contributing to a better understanding of severe accident phenomena (i.e., pool scrubbing; combustible gas distribution and combustion; fission products transport; etc.) and the efficiency of severe accident management actions (i.e., FLEX; PARs; activation of safety systems, long-term measures; etc.). The national framework set between researchers and end-users (regulators and utilities) is improving the safety of the Spanish nuclear power plants and allows for the assimilation of the research results by the nuclear industry.

Interplay of human factors and safety culture in nuclear safety for enhanced organisational and individual Performance: A comprehensive review

The incidents at Three Mile Island and Chernobyl have poignantly spotlighted the indispensable role of human factors and organisational dimensions in ensuring nuclear safety. These events illuminated that beyond the intricate technicalities of nuclear operations, lie the equally significant non-technical facets that warrant utmost attention. While human factors delve into the cognitive and ergonomic challenges faced by operators, safety culture represents the ingrained organisational ethos that promotes or impedes safety standards. While a myriad of studies has dissected the impact of these elements on nuclear safety, the quest for an encompassing methodology that holistically integrates them is still ongoing. This paper embarked on an exhaustive review that probed the symbiotic relationship between human factors, safety culture, organisational performance, and individual performance within nuclear power plants, all the while bearing the indelible imprints of the Three Mile Island and Chernobyl disasters. Beginning with an elucidation of the core tenets of human factors and safety culture and their intrinsic ties to nuclear safety, the review delved deeper into empirical research that illustrated their influence on the overarching safety climate of nuclear establishments. It shines a light on recurrent challenges like communication lapses, leadership oversights, and the ever-present spectre of human error. In its culmination, the review underscored the intertwined nature of these pivotal elements and expounds on their collective import in sculpting a resilient safety culture in nuclear setups. Recommendations for bolstering human factors and safety culture practices, gleaned from the bitter lessons of Three Mile Island and Chernobyl, are proffered. At its core, this paper advocated for an integrative paradigm that melds organisational and individual performance metrics with human factors and safety culture insights, perpetually underscored by rigorous research and evaluation, to ensure the unerring safety and dependability of nuclear power plants.

Study of the impact of hydrides on the fracture toughness of Zr-2.5Nb alloy

Hydrogen absorption and hydride formation in zirconium (Zr) alloys is a well-known issue and has been extensively studied in the past. Hydrogen has an impact not only on the strength but also on the fracture toughness of Zr alloys. As various Zr alloys are used for manufacturing the core components of nuclear reactors, such as nuclear fuel claddings, fuel channels, fuel assemblies, fixtures, and etc., their structural integrity has a great impact on the safety and reliability of the whole nuclear power plant. Furthermore, the hydrogen absorption ageing mechanism retains its significance not just during reactor operation but also throughout decommissioning and storage of spent nuclear fuel, especially when sufficient levels of absorbed hydrogen concentrations are reached, potentially leading to a delayed hydride cracking mechanism. Assessing the impact of hydrogen absorption on the strength of zirconium alloys typically involves time-consuming, costly, and sample-intensive experimental studies. Consequently, there is a demand for alternative research approaches. This article introduces the methodology to evaluate the fracture toughness of the Zr-2.5Nb alloy with varying hydrogen concentrations.

Analysis of main steam line break accident on a BWR test facility using TRACE

PUMA-E is an integral test facility, which is used for extensive experimental investigations to study the behavior of Boiling Water Reactor (BWR) plants under accident conditions. The TRACE code is the best estimate thermal hydraulic system code developed by the U.S. Nuclear Regulatory Commission for performing nuclear power plant safety analysis. The aim of the present work is to conduct a short validation study of the TRACE code applied to the PUMA-E facility. Firstly, a TRACE model of the PUMA-E facility was developed based on the existing RELAP5 input deck. Then, an additional facility TRACE model with reactor pressure vessel modeled by three dimensional “VESSEL” component was also established. After that, a Main Steam Line Break accident test was simulated, and the calculation results were compared with experimental data and RELAP5 simulation results. The TRACE simulation capability to predict the accident phenomenon of BWR test facility was evaluated using both qualitative and quantitative methods. On the basis of the above research, the base case accompanying with one Gravity Driven Cooling System tank failure accident was carried out based on the three facility models. The validity and reliability of the simulation results of the different facility numerical models were mutually verified through a qualitative comparison of the simulation results.

Event Sequence Based Fault Tree Analysis to Evaluate Minimal Combination of Event Sequences Leading to the Reactor Core Damage

Probabilistic safety assessment has been widely used to evaluate nuclear power plant safety systems. It couples fault tree (FT) analysis and event tree (ET) analysis. FT analysis is to develop failure scenarios, to quantify the failure probability and to identify critical components. ET analysis is to develop event sequences (ESs) leading to reactor core damage and quantify core damage frequency. Currently, there is no approach in probabilistic safety assessment that can be used to identify minimal combinations of safety system failures leading to reactor core damage. This study proposes an event sequence based fault tree analysis, which integrates FT model with ET model. The ET model is to identify the ESs leading to the reactor core damage. Meanwhile, the FT model is to identify minimal combinations of those ESs found in the ET model. The motivation of this study is on how to identify critical safety systems in nuclear power plants to mitigate disturbances caused by a group of postulated initiating events to avoid core damage. To confirm the feasibility and the applicability of the proposed approach, the performance of the AP1000 passive core cooling system is evaluated. It is found that if in-containment refueling water storage tank, automatic depressurization system - full, accumulator, core make-up tank, and passive residual heat removal work properly, the reactor core will intact. These results confirmed that the proposed approach can be applicable to identify minimal combinations of safety systems to keep the reactor core intact.

Rocking induced axial response and performance of lead‐rubber bearings in seismically isolated nuclear plants

AbstractSeismic isolation using lead‐rubber (LR) bearings can be an effective earthquake mitigation technique to ensure the safety of Nuclear Power Plants (NPPs) against extreme earthquake events. Accurate estimation of the axial and rocking response of NPPs is critical to the design of the safety‐related secondary components, piping, and auxiliary system crossing the interface of the isolation system. The isolators along the periphery of the isolation interface are especially susceptible to buckling in compression and cavitation in tension due to the rocking motion of the structure. No study so far has looked into rocking response of base‐isolated NPP and its consequence on isolator's performance. This paper investigates the axial and rocking behavior of base‐isolated NPPs at a site of high seismic hazard. A numerical study is conducted on the different representations of NPP models with the varying height of the center of mass isolated using a range of isolation systems of different periods and characteristic strengths. The rocking responses are obtained using simplified and advanced isolator models, which consider the interaction between the axial and shear response. The results provide estimates of the rocking responses and their influence on the axial response of isolators. It was observed that the rocking response of a base‐isolated NPP increases with an increase in the isolation period and the characteristics strength of the isolation system. Rocking effects may amplify the axial forces in the lead‐rubber isolators by more than 100% and need to be considered for true performance assessment of isolators when the horizontal to vertical shaking intensity exceeds a ratio of 0.5.

Condition monitoring for nuclear turbines with improved dynamic partial least squares and local information increment

Performing online condition monitoring for nuclear turbines in the rapidly changing environment is a challenging but imperative task to enhance the safety and reliability of nuclear power plants. Given the nonlinear and dynamic properties of nuclear turbine operation, this paper proposes an innovative method for condition monitoring. Specifically, the paper first redesigns time augmented matrices based on lagged data to reflect the process dynamics. Subsequently, a dynamic auto-regressive model, integrated with the variant of kernel partial least squares, is built between input and output variables, which represents auto-correlations and cross-correlations of operation data simultaneously. The prediction of the model serves as the baseline for the monitoring indicator. Additionally, since the operation process involves variable working excitation and random noise, making static control limits insufficient to satisfy the requirements of condition monitoring, the proposed method utilizes a novel monitoring indicator based on local information increment. The indicator comprehensively incorporates the prediction value and past measurement for monitoring statistics and control limits. Finally, the proposed method is applied to a real nuclear turbine operation process, and the results are compared with three other methods to demonstrate its superiority.

Optimization strategy for SAM in nuclear power plants based on NSGA-II

Abstract The Severe Accident Management Guide (SAMG) is an important component of nuclear safety regulations. Many studies are being conducted to optimize severe accident management (SAM) strategies. To ensure the safety of nuclear power plants, decision makers need to monitor multiple parameters with security threats. Therefore, it is particularly important to search optimal SAM strategies under different numbers of mitigation targets. The Non-dominated Sorting Genetic Algorithm-II (NSGA-II) is an evolutionary algorithm that does not require derivative differentiation and is capable of population search. In this study, a nuclear power plant accident optimization strategy is developed using the Modular Accident Analysis Program (MAAP) in conjunction with NSGA-II. The strategy enables decision makers to consider multiple mitigation objectives in a complex decision environment. Focusing on the CPR1000, this study applies the optimization strategy to automatically search for optimal mitigation strategies for small break loss of coolant accident (SBLOCA) and station blackout hot leg creep rupture accidents (SBOHLCR). Comparing the optimization results with the basic accident sequence, it is found that the reactor pressure vessel (RPV) failure time is delayed from 72,702 s to 128,730 s under SBLOCA and from 23,828 s to 28,363 s under SBOHLCR. This study has also verified that the optimal SAM strategy obtained by the strategy through dual objective optimization has better mitigation effects than a strategy that only considers one objective. This optimization strategy has the potential to be applied to other types of severe accident management studies in the future.

Comparison of seismic analysis methods within analytical approach

Introduction. In the nuclear power industry, ensuring nuclear and radiation safety is of paramount importance, therefore all aspects that may affect safety are considered in a deterministic and probabilistic approaches.
 As part of the safety justification, consideration of earthquakes of various intensity levels is regulated by safety standards. At the same time, the deterministic analysis of the earthquake and the models and calculation methods used are quite well-established, and in the probabilistic way, a relatively large number of different approaches have been developed and used as part of PSA level 1. The results of the evaluation of the conditional probability of failure of system elements, buildings and structures depending on the intensity of seismic impact on the free ground surface of the site depends on the approach used for the calculations.
 Aims. To perform a comparative analysis of various methods for constructing a failure probability curve under seismic impact.
 Materials and methods. A comparative analysis of the use of various approaches to the construction of a failure curve under seismic impact for a relatively simple case is performed. The results of the calculations showed that the scaling method leads to the most conservative results. The use of linearization methods and the method of moments give more realistic estimates, which can have a positive impact on the results of PSA level 1.
 Conclusions. A comparative analysis of some approaches has practical interest and is relevant, since various methods can contribute more than 50% to the core damage frequency, which creates a significant imbalance in the safety of nuclear power plants.

Determination of heat transfer process in vertical cable tunnels of nuclear power plants under real fire conditions

The object of the study was heat and mass transfer processes occurring in vertical cable tunnels. The problem to be solved was the definition of process mechanisms in the inner space of the tunnel. For this, field tests were conducted, mathematical models were created, and computational experiments were conducted to establish specific parameters that affect the temperature regime of a fire in a vertical cable tunnel of a nuclear power plant. The dynamics of temperature changes with known geometric parameters and fire load were determined, and the adequacy of mathematical models built in the Fire Dynamics Simulator software was investigated and computational experiments were carried out. It has been proven that they consist in determining the temperature regime in a vertical cable tunnel of a nuclear power plant with known technical and geometric parameters. Such studies have practical applications in the field of safety of nuclear power plants and the development of new technologies in this field. An important conclusion of these studies is the possibility of determining the fire resistance of building structures of vertical cable tunnels of nuclear power plants with the selection of the most severe temperature regime, according to the conducted field test. This means that research results can be used in practice in designing and evaluating the safety of such objects. The conducted research established that the temperature in the inner space of the tunnel can reach values from 1200 to 1400 ℃. The following factors influence the maximum temperature value and the maximum time to reach the maximum temperature in the fire cell: fire load, height and area of the tunnel. With a lower fire load, the maximum temperature in the vertical cable tunnel of the nuclear power plant was 75 % lower. Therefore, the results of these studies have a direct practical application in the field of safety of nuclear power plants and can be used to improve and develop new technologies in this field

A framework to determine the holistic multiplier of performance shaping factors in human reliability analysis – An explanatory study

The safety of nuclear power plants is the uppermost goal to be emphasized for their sustainability. To this end, it is crucial to evaluate their anticipated risk by visualizing accident scenarios resulting in undesirable consequences. In quantifying the likelihood of an accident scenario occurrence, it is necessary to estimate values of human error probabilities with respect to the performance of safety-critical tasks following relevant human reliability analysis (HRA) methods. One of the technical issues pertaining to the quality of HRA results is to determine the multipliers of performance shaping factors (PSFs) that specify the contexts in which human operators have to accomplish the required safety-critical tasks. Unfortunately, since PSFs are entangled with complicated interrelations, it is hard to properly determine their multipliers. To address this problem, in this study, a framework that can quantify the holistic multiplier of a specific PSF is proposed that can encompass the effect of these complicated interrelations. In order to corroborate the applicability of the proposed framework, an explanatory case study is conducted by using data available from the literature. Based on the result of this case study, it is expected that the proposed framework would be effective for determining PSF multipliers.

Autonomous Thermosiphon System of Passive Residual Heat Removal from the Primary Circuit of the Reactor Plant: Features of Operation, Characteristics and Basic Advantages

An autonomous system of passive removal of residual heat (PRRHS) of a reactor installation with VVER designed to ensure the safety of nuclear power plants in an accident with complete long-term blackout is considered. The system provides for the removal of heat directly from the first circuit of the reactor plant (PRRHS R). In order to increase the reliability and safety of the emergency heat sink, heat exchange equipment based on closed-type evaporation and condensation devices – two-phase thermosyphons – has been used in the system. The main feature of such heat exchangers is that their thermosiphon assemblies structurally separate the primary circuit and the auxiliary circuit of the PRRHS, which is removed outside the reactor compartment, and provide safe and efficient heat removal, reduce the risk of radioactive contamination spreading beyond safety barriers. Such autonomous passive systems will provide effective heat removal directly from the primary circuit by changing the chain of successive heat transfer sites from nuclear fuel to the final absorber and excluding from it such elements, as for example steam generators, the condition and operability of which in the emergency process of heat removal have a major impact on the safety of the reactor core. The article presents a diagram of an autonomous heat sink system; also, a description of its operation is given. The main characteristics of the course of the emergency process of removal of residual heat by the autonomous thermosiphon PRRHS R obtained by computational modeling have been considered. The advantages of an autonomous thermosiphon passive system in comparison with a passive heat removal system of a reactor installation with VVER through the second circuit are analyzed. The obtained results are proposed to solve the problems of diversification of passive safety systems of evolutionary reactor plants of nuclear power plants with VVER type reactors.