Future Nuclear Power Plants Research Articles

Digitalization, Management, and Synthesis of Thermal-Hydraulic Legacy Data Using the Dynamical System Scaling

A unique attribute of the nuclear industry is the extent by which designers, developers, operators, and regulators pay attention to demonstrating the safety case. Nuclear installation resiliency or safety takes overriding priority over function and performance. The preparation of the safety case heavily relies on extensive experimental campaigns that challenge the target design under a spectrum of postulated accident scenarios. For safety reasons, these experiments, typically conducted in subscale test apparatuses intended to reproduce the postulated accident scenarios and event sequences realistically, but subject to proven similarity criteria, are often referred to as scaling analyses. Since the dawn of the nuclear industry, significant resources have been committed to the construction and operation of such test facilities, and a massive amount of data has been generated. Such test data have been and are still used to validate the fundamental assumptions of nuclear power plant phenomena. As we move into the digital world, capturing this multiple decades worth of information in a transparent, easily accessible, and usable manner for the public and industry stakeholders is of paramount importance. Over the last few years, FPoliSolutions has been actively developing an enterprise digital data ontology platform (OGMA) intended to do just that. The goal is to help designers use experimental data to assess their safety case evaluation models. The platform or application programming interface (API) was designed to ontologically organize the experimental data, as well guide the user in the complex analytics associated with the interpretation and use of such test results. The OGMA API supports the user in accessing a library of sophisticated mathematical procedures for performing scaling analyses. For this purpose, the dynamical system scaling (DSS) procedure invented by Dr. Jose’ Reyes was formulated as one of the services in the digital platform. These methods have been demonstrated to provide an excelled mathematical apparatus to identify similarity criteria or quantify distortions between measured data and the target prototypical conditions. Moreover, DSS can provide a level of synthesis across separate effect tests and integral effect tests that has the promise of yielding efficiency in the evaluation code assessment activities, as setting up evaluation models that support the safety case of current and future nuclear power plants is often a daunting and expensive task.

New developments of hydrogen impurity online-monitoring in liquid lithium of IFMIF-DONES

On the way to future nuclear fusion power plants, the International Fusion Materials Irradiation Facility (IFMIF) – DEMO Oriented Neutron Source (DONES), is an important key element between ITER and DEMO to study before the influence of DEMO-neutron irradiation on foreseen fusion materials. The core of DONES, currently under construction near Granada, Spain, is based on a circuit containing liquid lithium at elevated temperatures acting as a functional material for the (d,n)-Li reaction in a materials test cell. In addition to this actual function of DONES, there are important aspects of maintenance in which hydrogen isotopes are generated under operating conditions and dissolved in this aggressive and very reactive alkali metal. This implies strong unfavourable effects on the applied structural materials, e.g. hydrogen embrittlement and others.To counteract these unfavourable effects, endangering the safe operation, an Impurity Control System (ICS) is an integral part of the DONES instrumentation. As part of the tasks of the European Neutron Source (ENS) to develop redundant systems for monitoring impurities, a special sub-task was defined for the development of an electrochemical H-sensor for concentrations in liquid lithium, ECHSLL. It is determined by detecting the electrical potentials of the lithium melts compared to a standard Li-based chemical reference system. This allows an inherent material property to be directly correlated (i.e. through appropriate electrochemical instrumentation) with chemical concentration values. This article presents important advancements in the applied ECHSLL technique, such as improving laboratory measurements from stagnant conditions to dynamic and realistic flow conditions of the liquid lithium material, as well as appropriate approaches to overcome the challenges in distinguishing the hydrogen isotopes by ECHSLL system (protium and deuterium among given laboratory conditions).

Scaling in nuclear thermal hydraulics: Advances, applications and future perspectives in Spain

The demonstration of the safety, high performance and reliability of any man-made device relies heavily in experimental programs. For nuclear reactor safety demonstration, several experimental facilities are used for both testing the design and the validation of computational codes. In the field of thermal hydraulics, most experimental facilities are built at a smaller scale, thus the scaling analysis is a requirement in the licensing process of the nuclear power plant. This includes the analysis of the scaling capabilities of the computer codes, the scale and design differences with the licensed reactor and the methodologies applied for licensing.In collaboration with the Spanish regulatory Body, several Spanish university groups have contributed considerably to the development and application of scaling methodologies. This article provides an overview of the most relevant contributions presented in the last decades. The activities deal mainly with the use of system codes in scaling analysis. Specifically the works can be classified in the use of Kv scaling analyses, analysis of counterpart experiments, plant application and the scaling of uncertainties. Finally, the future challenges that will be addressed in relation to scaling are highlighted with emphasis to the contribution to the safe operation of current and future nuclear power plants.

Transient simulation and analysis of a supercritical CO2 heat removal system under different abnormal operation conditions

The supercritical carbon dioxide (sCO2) heat removal system, which is based on multiple closed Brayton cycles with sCO2 as the working fluid, is an innovative, self-propelling and modular heat removal system for existing and future nuclear power plants. Previous studies analysed its design, layout, control and operation. In addition, this study considers different sudden failures during the accident progress, e.g. failure of single sCO2 cycles, control systems and valves. These abnormal conditions were investigated with the thermal-hydraulic system code ATHLET for a generic Konvoi pressurized water reactor. In most cases, the failure of a single sCO2 cycle can be compensated. On the one hand, failure of the fans of the gas cooler leads to a pressure increase which may be mitigated by an inventory control system or cycle shutdown. On the other hand, unintended fan speed-up can cause compressor surge without adequate countermeasures. Furthermore, the system can operate under the cyclic blow-off from the steam generator safety valves when the relief valves are not available. Finally, the unintended closure of the valve which controls the steam flow through the compact heat exchanger triggers a fast cycle shutdown but a subsequent restart might be possible.

Development of a Decommissioning Robot with a Simple Structure Capable of Traversing Steps Using Two Different Drive Systems

In response to the imperative need for robotic involvement in the decommissioning of the Fukushima Daiichi Nuclear Power Plant, a proactive initiative was undertaken. This initiative led to the organization of the “7th Decommissioning Creative Robot Contest,” in which pivotal challenges related to nuclear plant decommissioning were addressed. Recognizing inaccessible areas within a decommissioned plant, the primary objective of this study was to develop a remotely operated robot capable of executing decontamination tasks in such environments. Two distinct locomotion systems were designed: mecanum wheels and crawlers. The mecanum wheels, characterized by their ability to execute approximately 100 mm vertical movements, facilitated the robot to surmount obstacles, particularly by negotiating 100 mm steps by engaging the tracks. Simultaneously, to address decontamination tasks at elevated locations, a threefold expandable ladder truck system was integrated, augmenting the operational range of the robot. Participation in the contest provided a platform for showcasing the developed technologies and validating their practicality. Despite encountering challenges during competition, such as damage to one locomotion system, the other locomotion system can be used to reach the destination. Experience serves as a crucial means of verifying technological advancements. Issues such as malfunctions in the range of motion control due to faulty limit switch contacts and unexpected loads on the tire elevation motors were noted, which hindered the achievement of the anticipated results. Despite these drawbacks, the adopted structural design demonstrates notable advantages, particularly in terms of maneuverability. The combination of mecanum wheels and crawlers along with a ladder truck system demonstrated adaptability to various scenarios. This adaptability holds significant promise for future nuclear power plant decontamination applications. While the competition may not have yielded the desired outcome, the lessons learned and the verified practicality of the developed technologies reinforce their potential utility in addressing the future challenges of nuclear plant decommissioning.

Development of a Diffusion Bonding Route for Joining Oxide-Dispersion-Strengthened (ODS) Steels for Nuclear Applications

Oxide-dispersion-strengthened (ODS) steels are candidate materials for components in current and future nuclear power plants. One of the issues with using ODS steels is the difficulty of joining them without loss of mechanical performance. In this study, austenitic ODS 316L stainless steel was diffusion-bonded to Inconel 718 superalloy. Having optimized the bonding conditions, a number of samples were made at 1200 °C with a bonding time of 1 hour and pressure of 10 MPa. Preliminary mechanical and microstructural analyses indicated the formation of a sound joint interface, despite slight grain growth in the ODS 316L. A post-bonding thermo-mechanical treatment (TMT) was conducted to refine and restore the microstructure of the ODS 316L. Comparative TEM investigations of the parent ODS alloy and the bonded samples (with and without TMT) along with statistical analyses showed that the Y–Ti–O oxide size distribution remains unaffected by the bonding and complementary TMT cycles, indicating stability of such particles even at very high temperatures and suitability of the devised route for joining the ODS 316L steel.

ON RADON ACTIVITY OF TECTONIC FAULTS IN THE AREA OF SITING THE SINOP NPP IN THE REPUBLIC OF TŰRKIYE

Problem statement. The paper provides an additional expanded justification for applying an innovative methodology in the area of the future Sinop NPP, planned to be deployed in the Republic of Tűrkiye, to identify the activity rate of tectonic faults in soil foundation based on comprehensive radon measurements using the PSACEA universal estimation scale. The purpose of the article. Current radon activity assessment of existing tectonic faults in Earth’s crust based on archival measurements of soil radon carried out earlier in Sinop Province. Conclusions and results. Current analysis of radon measurement results obtained previously by the track method made it possible to further assess the radon activity rate of tectonic faults existing in the area of the future nuclear power plant. The proposed methodology to assess radon activity of tectonic faults in Earth’s crust involves increasing safety during the construction and operation of the future nuclear power plant. The regulation on using this methodology as part of the subsurface monitoring at the Sinop NPP can be enshrined in job descriptions. This methodology can also be used as part of the seismic monitoring of other sites at the nuclear power plant, especially those located in high seismic risk zones.

The rise of the Argentine Nuclear Program and the thwarted dream of industrialization (1950-1984)

The rise of the Nuclear Program in Argentina took place during the second stage of the process of Import Substitution Industrialization, which encourage local industry and the development of metalworking and electronics. In this context, one of the central aims of the National Atomic Energy Commission (CNEA) was to improve the domestic industry of goods and services in order to supply future nuclear power plants and promote technological autonomy. However, this philosophy began to struggle with the neoliberal reforms that were implemented after 1976. The primary objective of this work is to draw and analyze the main strategies implemented by CNEA to promote local industry. In this sense, this paper will be focus in the creation of the Metallurgy Department (1955), the Technical Assistance Service to Industry (1962), the National Industry Group (1965) and the creation of an Industrial Architect, ENACE SA (1980).

Preliminary study of the sub-medium São Francisco region overlooking the installation of future Brazilian nuclear power plants

Preliminary studies of site selection for the installation of nuclear plants in the northeast of Brazil have shown that the São Francisco river basin is a possible region for the implementation of these plants. For their operation, however, nuclear power plants require a reliable source of water for the cooling of heat rejected from the condenser that is sufficient for normal operation, shutdown, postulated accident conditions, and fire protection. The objective of this paper is to discuss safety issues and environmental impacts arising from the placement of nuclear power plants at the São Francisco river basin. The approach taken considered water availability throughout the year, the maintenance of thermal balance and climate changes. The present discussion indicates the necessity of a long-term management plan for the various uses of the São Francisco basin. An economic and social approach to the impacts of nuclear power plant operation and other agricultural and industrial activities in the region is also necessary. Finally, the study suggests the adoption of a variable-cycle cooling system, which will enable operation in all types of hydrological conditions expected in the São Francisco basin.

Modelling Irradiation Effects in Metallic Materials Using the Crystal Plasticity Theory—A Review

The review starts by highlighting the significance of nuclear power plants in the contemporary world, especially its indispensable role in the global efforts to reduce CO2 emissions. Then, it describes the impact of irradiation on the microstructure and mechanical properties of reactor structural materials. The main part provides the reader with a thorough overview of crystal plasticity models developed to address the irradiation effects so far. All three groups of the most important materials are included. Namely, the Zr alloys used for fuel cladding, austenitic stainless steels used for reactor internals, and ferritic steels used for reactor pressure vessels. Other materials, especially those considered for construction of future fission and fusion nuclear power plants, are also mentioned. The review also pays special attention to ion implantation and instrumented nanoindentation which are common ways to substitute costly and time-consuming neutron irradiation campaigns.

Graded Approach Establishment for the HTGR Maintenance Activities Using Modified Fuzzy FMEA & Expert Judgement Methodology

Grading is an important step of the Nuclear Power Plant (NPP) operation & maintenance activities. However, there are several grading difficulties for the High Temperature Gas Reactor (HTGR) as well as the other type of NPPs causing by the lack of operational experiences and availability of the reliability data. Failure Mode & Effects Analysis (FMEA) is one of the mature techniques that are commonly used to solve such kinds of difficulties. Nevertheless, traditional FMEA has several issues and possibly become an obstacle in the grading process. The modified FMEA by utilizing expert judgment elicitation techniques combined with the fuzzy logic theory is proposed to solve those issues. As a study practice, the proposed methodology is applied by examining Japanese’s HTGR, Gas Turbine High Temperature Reactor 300 for Cogeneration (GTHTR300C) design carefully. This study establishing good practice especially for the future advanced NPP maintenance activities development.

Mathematical model of the NPP Krško PCFV system for the RELAP5 computer code

Containment building is the final barrier for radioactive releases from a nuclear power plant (NPP). Preserving its integrity will minimize these releases even in a case of a severe accident with core degradation and melt discharge in the containment accompanied with the pressure and temperature increase. Installation of a venting system with ability to filter radioactive fission products is a preferred way to deal with the issue in present and future NPPs, especially after the Fukushima accident. Such system, called passive containment filtered venting system (PCFV), was installed in 2013 in the NPP Krško. Thermal hydraulic model of the PCFV system which included aerosol and iodine filters, associated pipings and valves was developed for the best-estimate computer code RELAP5. Main results and discussion are presented and compared with relevant plant documentation.

Experimentally validated multiphysics modeling of fracture induced by thermal shocks in sintered UO[formula omitted] pellets

Uranium Dioxide (UO2) fuel powers almost all commercial Nuclear Power Plants (NPPs) worldwide, generating carbon-free energy and contributing to the fight against climate change. UO2 fuel incurs damage and fractures due to large thermal gradients that develop across the fuel pellet during normal and transient operating conditions. A comprehensive understanding of the underlying mechanisms by which these processes take place is still lacking. A combined experimental and computational approach is utilized here to quantify the behavior of UO2 fuel fracture induced by thermal shock. This work introduces both (1) an experimental study to understand the fuel fracturing behavior of sintered UO2 pellets when exposed to thermal shock, and (2) a Multiphysics phase-field fracture model capable of simulating this process. Parametric studies were conducted to evaluate the effects of uncertainties in fracture properties on the fracture behavior of UO2 due to thermal shocking. A set of energy release rate (or equivalently fracture toughness) and contract area (the part of the fuel pellet in direct contact with the cold bath) were able to capture the overall fracture trends of the corresponding experimental data. Our combined approach presents a new method for accounting for the effects of microstructure and sample size on the energy release rate/fracture toughness. The experimental data were collected from multiple experiments that exposed UO2 pellets to high-temperature conditions (589–676 ∘C) followed by a quench in sub-zero water. This work demonstrates that joint experimental and computational efforts are able to advance the understanding of thermal fracture in the primary fuel source for existing and future NPPs.

Modeling and simulation of radiation effects in nanometric volumes of silicon and water

Understanding the effects of particle radiation in nanoscale electronic materials and biological materials is particularly important for future long-term space flights, nuclear medicine and nuclear power plants. Heavy charged particles arriving among galactic cosmic rays have powerful damaging actions when they pass through the electronic devices and biological objects. The present study is focused on the Monte Carlo modeling of stochastic energy depositions in charged particle tracks and calculation of absorbed energy within small sample (200 nm) of silicon and liquid water. The silicon is widely used material for memory chips, computer processors, transistors, and all other electronics. The most common biomaterial is liquid water, which accounts for 60-90% of all living organisms. The applied simulation technique is based on the Geant4 Monte Carlo transport code, which shows good agreement with experimental cross sections for dosimetry and nano-dosimetry measurements. The calculations were made for protons and carbon ions with different kinetic energies within a relatively wide range of the Bragg curve and linear energy transfer (LET) values from a few to hundreds of keV/μm. We estimated and compared the distribution of energy deposition events and production of reactive chemical species within nanometric volumes of silicon and water after irradiation. We also performed a microdosimetric calculation to estimate LET distribution and nanodosimetric calculation to analyze cluster distributions of ionizations, excitations, elastic collisions and dissociative attachment for electrons corresponds to an interaction with a nonzero energy deposition.

Simulation, analysis and control of a self-propelling heat removal system using supercritical CO2 under varying boundary conditions

The supercritical carbon dioxide (sCO21) heat removal system, which is based on a closed Brayton cycle with sCO2 as a working fluid, is an innovative heat removal system for existing and future nuclear power plants. This paper provides the design, layout and control of the system based on assumptions developed in the project sCO2-4-NPP. A self-propelling operational readiness state enables a fast start-up and consumes only 12% of the design thermal power input. The system is analysed over a wide range of ambient and steam-side conditions in ATHLET, using performance maps for the turbomachinery, which were designed recently. The performance analysis suggests that it is a good option to operate the system at the design compressor inlet temperature of 55 °C at any boundary condition. With decreasing thermal power input, the rotational speed of the turbomachinery must be decreased to keep the system self-propelling. Moreover, the turbomachinery design with a higher surge margin is preferred. By controlling the compressor inlet temperature via the air mass flow rate and turbine inlet temperature via the turbomachinery speed, the heat removal system is successfully operated in interaction with a pressurized water reactor.

Aging management experience to support reliability of RSG-GAS operation and future NPP in Indonesia

The aging process will cause a decrease in the reliability and safety performance of the nuclear reactor operation. As a consequence, the quality of service to stakeholders will decline. In order to keep nuclear reactor reliable, systematic activities are needed to manage the aging process that occurs at the GA Siwabessy multi-purpose reactor (RSG-GAS), in Serpong. The activity includes 4 main elements, i.e.: understanding and evaluating the aging process, using aging prevention methods, detection and surveillance methods, and aging mitigation methods. Aging management is carried out at RSG-GAS through routine and periodic maintenance, repair, replacement and monitoring mechanisms. Aging management must be planned from design, construction and commissioning activities to reactor operation. RSG-GAS aging management is carried out in accordance with safety guidelines from the international atomic energy agency (IAEA) and nuclear power regulatory body (BAPETEN) regulations, as well as various vendor documents. The success of aging management is demonstrated by operating parameters that meet operating limit conditions (OLC), and is evaluated through periodic safety assessments. This paper aims to convey the experience of RSG-GAS in aging management, in order to support the reliable operation of RSG-GAS. Experience in the aging management program of the RSG-GAS as long as 34 years, is an important asset in developing the aging management program of future Indonesian nuclear power plants.

Start-up, operation and thermal-hydraulic analysis of a self-propelling supercritical CO2 heat removal system coupled to a pressurized water reactor

The supercritical carbon dioxide (sCO2) heat removal system, which is based on a closed Brayton cycle with sCO2 as a working fluid, is an innovative, self-propelling and modular heat removal system for existing and future nuclear power plants. By changing the number of CO2 cycles, the heat removal capacity can be adapted. In this paper, up to four sCO2 cycles are analyzed in interaction with a pressurized water reactor, using the thermal-hydraulic system code ATHLET and considering a long-term station blackout and loss of ultimate heat sink scenario with conservatively high and low decay heat curves. The presented start-up procedure for the heat removal system might require further optimization due to the non-linear thermal gradients. Independent from the start-up, a heat removal system with three or four CO2 cycles keeps the primary loop temperatures sufficiently low. However, with only three cycles, the core is almost uncovered, and the danger of recriticality may occur due to cold leg deboration. Controlling the turbine inlet temperature via the turbomachinery speed and subsequent shutdown of single cycles successfully adapts the operation of the heat removal system to the declining decay heat. This enables reliable decay heat removal for more than 72 h.

A Critical Look at the Need for Performing Multi-Hazard Probabilistic Risk Assessment for Nuclear Power Plants

Probabilistic Risk Assessment (PRA) is one of the technologies that is used to inform the design, licensing, operation, and maintenance activities of nuclear power plants (NPPs). A PRA can be performed by considering the single hazard (e.g., earthquake, flood, high wind, landslide) or by considering multi-hazards (e.g., earthquake and tsunami, high wind and internal fire). Single hazard PRA was thought sufficient to cover the analysis of a severe accident until the Fukushima Daiichi NPP accident in 2011. Since then, efforts were made to consider multi-hazards as well; thus, multi-hazard PRAs are starting to be seen as being indispensable for NPPs. In addition to the changing frequency of global and local natural hazards, other reasons to be highlighted are that the number and diversity of NPPs will probably increase. Moreover, advanced reactors are close to becoming a reality by designing them with passive safety systems, smaller, standardized, and even transportable to make them cheaper across the design, licensing construction, and operation stages. Thus, multi-hazards should be addressed in any future full-scope PRA. Although we found a few studies discussing multi-hazards, a general framework for multi-hazard PRA is still missing. In this paper, we argue that the starting point for any multi-hazard PRA general framework should be the Advanced Non-LWR Licensing Basis Event Selection (LBE) Approach and Probabilistic Risk Assessment Standard for Non-Light Water Reactor (non-LWR) Nuclear Power Plants. For Probabilistic Risk Assessment (PRA), history has shown us the path forward before, with Three Mile Accident being seen as one milestone to understand the necessity of PRA. The Fukushima Daiichi NPP Accident is another milestone in the development of PRA, showing the need for performing multi-hazard PRA for the current and future NPPs.

Assessment of Radioactivity Level in the Terrestrial and Marine Organisms in Yangjiang and Its Adjacent Areas (China).

In order to assess the radioactive level in the terrestrial and marine organisms in Yangjiang and the adjacent areas, 90Sr, gross beta and gamma-emitting radionuclides (238U, 226Ra, 228Th, 226Ra, 40K, 137Cs, 51Cr, 55Fe, 54Mn, 58Co, 60Co and 65Zn) were analyzed from 2011 to 2012. The annual effective doses were estimated in the high natural radioactive background areas in Yangjiang (HBRAYJ). The specific activities of 238U, 228Th, 226Ra, 40K and 137Cs in all organisms were <0.05–5.20, 0.30–14.50, 0.11–3.58, 11.1–148.0 and <0.003–0.088 Bq/kg, whilst 51Cr, 55Fe, 54Mn, 58Co, 60Co, 65Zn and 110mAg were below the minimum detectable activity. 90Sr and gross beta specific activities were 20.0–143.0 and 0.021–0.316 Bq/kg. Results show that 228Th was significantly higher than 238U and 226Ra of natural U series in organisms due to the rich-Th soils in the HBRAYJ; 228Th, 226Ra, 40K, 137Cs and 90Sr were greatly lower than the limited concentrations in Chinese foods. The internal dose mainly contributes to natural 40K, 226Ra and 228Th. It is useful to provide some basic data and assess the radiological risk from the HBRAYJ and Yangjiang nuclear power plants in future.

The Study of Main Component of the High Temperature Gas-cooled Reactor Balance of Plant using Local Component

National Nuclear Energy Agency (BATAN) as a research and development institution related to nuclear energy has designing Nuclear Power Plant (NPP) as one of the solutions to meet energy needs in Indonesia. This alternative electric power plant can be used as back up of fossil base power plant and maintaining the basic load of electric consumption. This research purpose is to simulate the balance of plant of the designed NPP and comparing the main component need to be produce by local industry. The simulation was created using Chemcad software to describe the process and component analysis of the NPP. The simulations done also can be increase the economic efficiency. Simulation results show the component data which can be used as a reference for the selection of the main components and pipes would be used in the construction of the reactor. The simulation results data of this simulation also can be uses to decide the local component that can apply in future NPP.