Operation Of Nuclear Power Plants Research Articles

Study on irradiation embrittlement behavior of reactor pressure vessels by machine learning methods

A reliable irradiation embrittlement model of reactor pressure vessel (RPV) is critical to the safe operation and life extension of nuclear power plants (NPP). In this study, 474 datasets from NUREG/CR-6551 and 6 machine learning (ML) methods were adopted to accurately predict the irradiation embrittlement behavior (ΔRTNDT) of RPV with comprehensive input factors involving irradiation conditions and metal composition. The GBDT model has the best prediction performance. The calculated result by GBDT model is close to those of JEAC-4201 and ASTM E900-15. Cu content is the most sensitive factor for ΔRTNDT, followed by neutron fluence. This work demonstrates a successful application of ML in promoting NPP safety and RPV materials design.

IMPACT OF URBAN LAND ON THE RISKS OF KARST PROCESSES IN THE NPP LOCATION AREA (ON THE EXAMPLE OF RIVNE NPP)

The aim of the article is to assess the impact of urbanized areas on the risk of developing karst processes in the area of nuclear power plant location and on the safety of the plant (using the example of the Rivne nuclear power plant). Karst features in the area around the Rivne nuclear power plant were visualized using GIS methods. Systematization of karst hazards was carried out with regard to their impact on the structural integrity and safety of the long-term operation of the Rivne nuclear power plant. The impact of urbanized areas on the geological environment was assessed, and a correlation was established between urbanization factors and the activation of karst-suffusion processes. It is shown that urbanization can contribute to the emergence of karst manifestations and, as a result, to threaten to the safety of NNP. Approaches aimed at improving the assessment of karst hazards, monitoring of the main factors of their activation, and managing urbanized areas to increase the safety of nuclear power plants were proposed.

Monitoring of tritium in surface waters in PC, Nuclear facilities of Serbia

The isotope of hydrogen, tritium, is a radioactive isotope, in most cases it is found bonded to molecules that replace hydrogen. It can be found in nature as a product interactions of cosmic radiation with air or atmospheric molecules nitrogen, such tritium is natural. There is also anthropogenic tritium that contributes increased levels of natural tritium. Artificially produced tritium is produced in testing of nuclear weapons, in the operation of nuclear power plants and processing nuclear fuel. This paper will describe the methods of tritium detection liquid scintillation detector Quantulus 1220 by direct method in duration of 300 minutes at the Department of Nuclear Physics. Surface water samples with tritium was taken from the location of PC Nuclear facilities of Serbia from the Mlaka stream. The monitoring recorded an increase in tritium content at two locations, which indicates failures of nuclear waste storage.

Modern approaches to strength design substantiation for critical elements of WWER NPP reactor internals

This study presents the theoretical and applied results on the development and application of refined models and methods for solving actual problems of mechanics related to the substantiation of strength and forecasting the service life of elements of the primary circuit equipment of WWER nuclear power plants (NPPs). The scientific and applied developments take into account current trends in global practice and contain new conceptual approaches to solving nonlinear boundary value problems in the mechanics of deformed structures. Refined mathematical models have been proposed and theoretically substantiated to describe non-isothermal inelastic deformation processes, taking into account the deformation history of thermal and power loading, radiation effects of hardening, swelling and creep of irradiated metal, and accumulated damage of brittle-ductile fracture. The obtained scientific results are the basis for the methodology of refined design strength analysis and residual life assessment of critical elements of the equipment of the first circuit of WWER reactor units to solve urgent problems of ensuring the conditions for safe operation of Ukrainian NPPs and justifying the terms of their service life extension

Deep reinforcement learning for a multi-objective operation in a nuclear power plant

Nuclear power plant (NPP) operations with multiple objectives and devices are still performed manually by operators despite the potential for human error. These operations could be automated to reduce the burden on operators; however, classical approaches may not be suitable for these multi-objective tasks. An alternative approach is deep reinforcement learning (DRL), which has been successful in automating various complex tasks and has been applied in automation of certain operations in NPPs. But despite the recent progress, previous studies using DRL for NPP operations have limitations to handle complex multi-objective operations with multiple devices efficiently. This study proposes a novel DRL-based approach that addresses these limitations by employing a continuous action space and straightforward binary rewards supported by the adoption of a soft actor-critic and hindsight experience replay. The feasibility of the proposed approach was evaluated for controlling the pressure and volume of the reactor coolant while heating the coolant during NPP startup. The results show that the proposed approach can train the agent with a proper strategy for effectively achieving multiple objectives through the control of multiple devices. Moreover, hands-on testing results demonstrate that the trained agent is capable of handling untrained objectives, such as cooldown, with substantial success.

Far-Field Influences Shadow the Effects of a Nuclear Power Plant’s Discharges in a Semi-Enclosed Bay

The sustainable development of society depends on the reliable supply of electricity while keeping impacts on the environment to a minimum. A 951 MWe nuclear power plant in the semi-enclosed Nanwan Bay at the southern tip of Taiwan began operating in May 1984. Part of the bay is in Kenting National Park, which is known for its coral reefs and abundant marine life; thus, thermal pollution from the cooling water discharge is a great concern. Fortunately, the bay opens south to face the Luzon Strait, where the world’s strongest internal tides are generated. Because the bay is deep enough, internal waves bring up cold deep water and reduce the surface temperature by as much as 10 °C for a few hours every day. These internal waves and topographically generated upwelling also bring nutrients to the euphotic layer from the depths, but the upwelled waters quickly leave the bay along with the cooling water. As a result, a thermal plume with a temperature of 1 °C or higher than the ambient temperature only covers 1 km. By way of comparison, El Niño—Southern Oscillation- or Pacific Decadal Oscillation-related interannual variations in temperature are as high as 5 °C. The rapid turnover of the upwelled waters also helps to prevent heat released by the power plant from accumulating and diminishes the thermal stress, thus sustaining corals and other marine life forms. Typhoons, even hundreds of kilometers away, could also induce the upwelling of cold subsurface water. Consecutive typhoons have been observed to reduce the water surface temperature by up to 10 °C for two weeks or longer. Furthermore, the currents are such that the thermal plume flows out of the bay most of the time. All of these factors make the surface waters in the bay about 0.5 °C cooler than the waters outside of the bay, despite the operation of a nearby nuclear power plant.

Resistance of Gable Structure of Nuclear Island to Progressive Collapse in Conventional Island Shield Building of Nuclear Power Plants

The gable wall of the conventional island of a nuclear power plant carries various pipeline loads connecting the nuclear island and the conventional island. If the gable wall collapses and is damaged due to the progressive collapse of the main plant structure of the conventional island, it will directly threaten the safety of the nuclear island. Therefore, it is of great significance to systematically study the resistance of the gable structure of the nuclear island to progressive collapse for the safe operation of nuclear power plants. Based on the SAP2000 finite element analysis platform, this paper established a model of the reinforced concrete frame structure of the first span of the shield building on the conventional island of a nuclear power plant, including the gable. According to the alternate path method, the node displacement and bearing capacity responses were used to determine the critical components in the gable, and single column failures of the top four in importance were selected as different analysis conditions. Nonlinear static and dynamic methods were used to analyze the displacement, internal force change, hinge sequence of the gable structure on the side of the nuclear island, and the ability of the gable to resist progressive collapse. The analysis results showed that the gable structure of the nuclear island had good resistance to progressive collapse under four single column failure conditions, whether by nonlinear static or dynamic analysis, and the overall vertical displacement obtained by the nonlinear static analysis was larger than that obtained by the dynamic analysis.

Adapted FLASHCAT methodology to model horizontal cable tray fires using computational fluid dynamics

Nuclear power plants (NPPs) have an extensive network of electrical cables and associated trays to support daily operations. As electrical cables have combustible mass, their presence adds to the fire load in such plants. Thermal stress from cable fires can cause cable failures and spread fire to other redundant cables or other equipment, jeopardizing the safe operation of NPPs. Assessing the heat release rates (HRR) of cable fires is critical for the nuclear industry as HRR regulates the thermal stress in the surrounding of a fire and other fire products that can damage the facility. This paper presents a computational fluid dynamics-based method to model the HRR of such cable fire. The method is partly based on the FLASHCAT model (modified and adapted here) and the surface temperature ignition model of the Fire Dynamics Simulator (FDS) software. The simulation set up with validation objective replicates an experiment of OECD PRISME 3 program in which cables arranged over two horizontal trays are burnt. HRR obtained from the simulation had peak HRR underestimated by 4%, and the time to reach peak HRR overestimated by 5%. The result is encouraging as it provides confidence in using the method outlined in the paper.

Progress of the Analytical Methods and Application of Plutonium Isotopes in the Environment

Since the beginning of the use of nuclear energy in the twentieth century, atmospheric nuclear weapons testing, nuclear accidents, and spent fuel reprocessing have released large amounts of radioisotopes into the environment, especially plutonium isotopes. As an important anthropogenic radionuclide, plutonium is mainly produced by neutron irradiation of uranium. Plutonium isotopes and their ratios from different sources have a specific ‘fingerprint’, which is determined by irradiation time and intensity. Accordingly, several methods have been developed for the analysis of plutonium isotopes. As a popular new tracer in recent years, plutonium isotopes could be used for environmental tracer applications and source identification of regional environmental contamination sources. This paper summarizes and compares the pretreatment of plutonium isotopes in various samples and the detection and analysis methods in recent years. Plutonium isotopes in various environmental samples were enriched by chemical treatments, such as ashing, acid digestion, and alkali fusion. Then it was purified and separated by extraction resins to remove interfering nuclides for the subsequent mass spectrometry. The practical applications of plutonium isotopes and their related ratios in the environment are summarized, such as the determination of plutonium isotopes and their related ratios in the environment near two representative nuclear power plant accidents and the monitoring of the safe operation of nuclear power plants; the establishment of a plutonium isotope database in the vicinity of Chinese waters; and the traceability of plutonium isotopes in environmental regions through the global atmosphere and sand and dust. Finally, the outlook for subsequent research in plutonium isotope detection and applications is presented.

The geochemical role of B-10 enriched boric acid in cemented liquid radioactive wastes

Boric acid is a significant radioactive waste generated during the operation of nuclear power plants. Cementitious materials have been widely studied for the immobilization of boric acid. The generally used natural boric acid has been replaced by enriched boric acid for geochemical reasons and are expected to have varied behaviors in cementitious matrices. Results showed that simulated enriched/natural boric acid liquid wastes mostly contain boron in {mathrm{B}(mathrm{OH})}_{4}^{-} and {{mathrm{B}}_{5}{mathrm{O}}_{6}(mathrm{OH})}_{4}^{-} ionic forms, but the mass ratio of these species is higher in enriched boric acid solutions. In function with the concentration of enriched/natural boric acid, the solidified cementitious materials show different mineralogy.

Surveillance of the seashore using the KATERINA II geo-referenced detection system

In this work, the design and initial demonstration of the KATERINA II detection system for rapid mapping of radionuclides in areas near to seashore is presented. A new development has been realized by integrating a GPS module in KATERINA II detection system and synchronizing its data with the acquired spectra in real-time. The new system may be used in a backpack, for areas with low activity concentration, or can be installed in an unmanned vehicle, for observing and mapping the source(s) of radioactivity, e.g. at the seashore, in areas with high contamination. A quantitative solution is provided for natural and artificial radionuclides, taking into account the characteristics of the detector, the parameters of measurement geometry and a mean beach sand/sediment composition. This paper reports field results for site characterization issues through automated analysis of gamma-ray spectra including low-level and low-energy γ-emitters. Perspectives of the future application of the system in a worldwide basis are related to radionuclides mapping and the assessment of dose rates in seashore areas that may be contaminated due to the operation of nuclear power plants, desalination plants and NORM industries, and/or due to the decommissioning of nuclear installations.

Study of the radioactivity inventory deposited in the reactor coolant system during the decommissioning of Kori Unit 1

The Kori Unit 1, a first nuclear power plant (NPP) in South Korea stopped operation permanently on June 18, 2017 after 40 years of operation. Before the main decommissioning activities begin, full-system chemical decontamination operation (FSCDO) of the primary systems is planned to secure the safety of workers and minimize the quantity of radioactive waste. In this study, the radioactive inventory at the time of decommissioning of Kori Unit 1 was investigated. The CRUDTRAN code was used to make predictions by using the actual operation and maintenance history and design data of Kori Unit 1. Regarding the steam generator (SG), the prediction results obtained using the CRUDTRAN code were 17–30% higher or more conservative than the direct measurement results obtained using survey meters. By applying the oxide layer removal ratio (OLRR; obtained through a chemical decontamination field test using the contaminated specimen) to the oxidation/deoxidation process during FSCDO, the removal rate of radioactive inventory could be predicted. As an actual decommissioning work proceeds, the detailed data of radioactive elements in RCS systems can be obtained through sampling and chemical analysis. Those data will be much valuable for both the operation and future decommissioning of current NPPs.

Thermodynamics and vapourization of Cs-, Sr-, Ba-containing oxide systems valid for nuclear safety problems

The review provides a systematic analysis of the studies carried out mainly in the last decade in which the thermodynamic properties and vapourization of the systems containing cesium, strontium, and barium were studied by high-temperature mass spectrometry. Such systems are of particular interest for considering the problems of environmental safety in the nuclear power industry. Particular attention is paid to the issues of reliable identification of the content of the gaseous phase over oxide systems, which are important in various high-temperature technologies, including the disposal of radioactive waste, reprocessing of nuclear fuel, as well as ensuring the safe operation of nuclear power plants. A discussion and comparison of thermodynamic data found in the literature for the systems under consideration in a wide temperature range were also carried out featuring the main advantages of the Knudsen mass spectrometric effusion method.<br> The bibliography includes 117 references.

DESIGN AND DEVELOPMENT OF HMI IN SIMULATOR FOR FRESH FUEL HANDLING SYSTEM

KALBRSIM operator training simulator is used for training Prototype Fast Breeder Reactor (PFBR) nuclear power plant operators. In simulator, using virtual panel HMI which is an exact replica of actual hardware panels, all the essential fresh fuel operations can be done. Visualization power is also provided by 3D models which aid in quicker understanding of the complex process involved in Fresh Fuel Handling System (FFHS). Operation using Virtual Panel is an excellent tool for system developers to test the logic and process models of FFHS in KALBRSIM before integration with hardware panels. Also, operator training through virtual panel HMI can enable operator to monitor and control FFHS by running multiple screens at once at a central location. Teaching operators their effect of actions using 3D models is an add on with virtual panels which help the operator visualize the system accurately. This paper discuss the design and development of virtual panel HMI and 3D models of fresh fuel handling system used in full scope replica PFBR operator training simulator. Advantage of using HMI in simulator is also covered.

Deep learning health management diagnostics applied to the NIST smoke experiments

Fire is one of the most important hazards that must be considered in advanced nuclear power plant safety assessments. The Nuclear Regulatory Commission (NRC) has developed a large collection of experimental data and associated analyses related to the study of fire safety. In fact, computational fire models are based on quantitative comparisons to those experimental data. During the modeling process, it is important to develop diagnostic health management systems to check the equipment status in fire processes. For example, a fire sensor does not directly provide accurate and complex information that nuclear power plants (NPPs) require. With the assistance of the machine learning method, NPP operators can directly get information on local, ignition, fire material of an NPP fire, instead of temperature, smoke obscuration, gas concentration, and alarm signals. In order to improve the predictive capabilities, this work demonstrates how the deep learning classification method can be used as a diagnostic tool in a specific set of fire experiments. Through a single input from a sensor, the deep learning tool can predict the location and type of fire. This tool also has the capability to provide automatic signals to potential passive fire safety systems. In this work, test data are taken from a specific set of the National Institute of Standards and Technology (NIST) fire experiments in a residential home and analyzed by using the machine learning classification models. The networks chosen for comparison and evaluation are the dense neural networks, convolutional neural networks, long short-term memory networks, and decision trees. The dense neural network and long short-term memory network produce similar levels of accuracy, but the convolutional neural network produces the highest accuracy.

Effects of throttling windows on cavitation flow of sleeve control valve

The control valve is the key equipment in nuclear power plants, which is mainly used to control the fluid flow, and essential for the safe operation of nuclear power plants. Cavitation is one of the most common phenomenon inside control valves, which will not only generate noise and vibration, but also reduce the service life of control valves. In this paper, the effects of throttling windows numbers and flow directions on the cavitation characteristics of a typical sleeve control valve are analyzed by numerical simulation. Results show that the vapor volume of 2, 4, 6, 8 throttling windows when fluid flows in the forward direction are 33.83, 39.25, 71.56, 59.34 times of the vapor volume in the backward direction respectively, which means increasing the throttling windows number can greatly reduce the vapor volume. Meanwhile, the vapor distribution is significantly different under different flow directions. The energy loss is more serious when the fluid flows in backward direction. Therefore, the erosion and abrasion of the valve body caused by the fluid impact can be reduced by changing the flow direction of the fluid from forward to backward and increasing the throttling windows number. This paper analyzes the cavitation characteristics of the control valve based on the switching conditions and discusses the mechanism of the working conditions and structural parameters on the cavitation characteristics, which is of significance for the anti-cavitation design of control valves.

Unique safety features and licensing requirements of the NuScale small modular reactor

Small modular reactors (SMR) offer a novel approach to the construction and operation of nuclear power plants. The NuScale VOYGR™ plant uses a simplified SMR design that is based on proven light-water reactor technology with substantial improvements in nuclear safety. It consists of a 250 MWt reactor core housed with other primary system components in an integral reactor pressure vessel surrounded by a steel containment vessel, all of which is immersed in a large pool of water that also serves as the ultimate heat sink. At the core of the NuScale safety case are three primary safety systems: the decay heat removal system, the emergency core cooling system, and the containment. The ability of the NuScale Power Module (NPM) passive safety systems to remove core decay heat for an unlimited duration is demonstrated through analysis of a beyond-design-basis extended loss of AC power with no replenishment of water to the ultimate heat sink or operator actions. The NuScale methodology to evaluate an indefinite loss of AC power uses the proprietary NRELAP5 systems analysis computer code. Analysis results show that the reactor coolant system liquid level above the core is maintained and that containment pressure remains below the vessel design pressure. Once full passive air cooling is established, containment pressure and temperature will decrease over time with decreasing core decay heat. NuScale received standard design approval in September 2020 and design certification in January 2023 for its 50 MWe NPM configured as a 12 module plant. NuScale is currently seeking standard design approval to increase its core power to 250 MWt, nominally 77 MWe per module, in a 6-module plant (VOYGR™-6) configuration. The high-level safety of NuScale’s SMR technology is foundational to a new standard of nuclear power plant resilience.

Analysis of time-varying mechanical properties of prestressed concrete containment during the tensioning process and service considering the influence of creep

The containment structure is an important barrier that ensures the safe operation of nuclear power plants. Currently, most studies on concrete creep in containment structures assumed constant stresses in the calculation of creep, which is not consistent with realistic cases or circumstances. In this study, a combination of theoretical analysis and numerical simulation was used to investigate the prestressed concrete containment (PCC) of nuclear power plants. Based on the age-adjusted effective modulus method, three widely accepted concrete creep models were improved to obtain a variable stress creep model. Using the improved creep model, the effect of concrete creep on the PCC structure during prestressed tensioning and long-term service was systematically analyzed. Then, the internal pressure bearing capacity of the PCC in long-term service was assessed, and the effects of long-term concrete creep on the mechanical properties of the PCC were evaluated.

An Operation Scheme Generation Method for Nuclear Power Plant Operation under the Condition of No Operating Procedures Guided

Large-scale, complex, and high-risk industrial systems such as nuclear power plants have developed detailed operating procedures. Under the expected conditions, the operators operate the system according to the operating procedures to avoid human error. However, under complex and unfamiliar conditions, once the guidance of operating procedures is lost, serious consequences may be caused. This paper proposes a No-Procedure guided Operation Supervision Technology (NoP-OST). The key idea is to identify the success paths by a forward-searching approach from source to sink functions. On this basis, the success paths are combined to generate various operation schemes to achieve the main objectives of the system. The paper illustrates the application of NoP-OST through the case of loss of the Residual Heat Removal System (RHRS) in a Pressurized Water Reactor (PWR) nuclear power plant. The research results show that the method proposed in this paper can provide a valuable operational reference for complex systems by making use of the available functions of the system.

Modeling of deposition and erosion of CRUD on fuel surfaces under sub-cooled nucleate boiling in PWR

Simulating the Corrosion-Related Unidentified Deposit (CRUD) on the surface of fuel assemblies is necessary to predict the axial offset anomaly and the localized corrosion induced by the CRUD during the operation of nuclear power plants. A new CRUD model was developed to predict the formation of the CRUD deposits, considering the deposition and erosion mechanisms. The heat transfer and capillary flow within the CRUD were also considered to evaluate the boiling amount within the CRUD layer. This model predicted a CRUD deposit thickness of 44 μm during a one-cycle operation of the Seabrook nuclear power plant. The CRUD deposition tended to accelerate and decelerate during the simulation, by being related to boiling mechanism on the deposits surface. Additionally, during a three-cycle operation corresponding to the refueling period, the CRUD deposition was saturated at a thickness of 80 μm, which was in good agreement with the suggested thickness for CRUD buildup in pressurized water reactors. Surface boiling on the thin CRUD deposits enhanced the acceleration of the deposition, even when the wick boiling properties were not favorable for CRUD deposition. To ensure the certainty of the simulation results, sensitivity analyses were conducted for the porosity, chimney density, and the constants employed in the proposed model of the CRUD.