Reactor Power Plant Research Articles

Numerical study on the effect of CO2 on the heat transfer characteristics of steam condensation outside a vertical tube

Under severe accident conditions of a pressurized water reactor nuclear power plant, the reactor core may melt, and the molten corium even reacts with concrete in the containment, producing a large amount of CO2 into the entire containment gas space. This may have an obvious impact on the heat transfer characteristics of steam condensation in the containment. However, there are few studies focused on this phenomenon. In this paper, a commercial CFD software, STAR-CCM+, is used to study the effect of CO2 on steam condensation heat transfer. The calculations are based on 0.4 MPa gas pressure, 40 K wall sub-cooling, 0.66 or 0.2 steam mole fraction, and the CO2 mole fraction varies from 0 to 0.34 or 0.8. The results show that the CO2 demonstrates different influence laws at high and low steam concentrations, respectively. When the steam mole fraction is 0.66, as the proportion of CO2 in the non-condensable gas increases, the condensation heat transfer coefficient (HTC) will increase, and the enhancement can reach 30%. When the steam mole fraction is 0.2, the condensation HTC will first decrease and then increase, as the proportion of CO2 in the non-condensable gas increases. The effects of gas pressure and wall sub-cooling on the condensation HTC are also analyzed. The results show that the condensation HTC increases with the increase of the gas pressure and decreases with the increase of wall sub-cooling.

Fouling on the secondary side of nuclear steam generator tube: Experimental and simulated study

The fouling in the steam generator (SG), seriously affecting the internal integrity of SG tube, heat transfer efficiency and control of water chemistry, is commonly caused by the growth of thick oxides and deposition of soluble particles in the secondary circuit of pressure water reactor (PWR) nuclear power plant (NPP). Herein we describe the nucleation, aggregation and growth mechanism of fouling containing different metallic cations. Three fouling layers were observed at the crevice position of SG tube and the large Fe3O4 fouling particles exhibited a double layer structure. The growth process and direction of fouling particles were found to be related with the oxide layer. Meanwhile, the deposition process of fouling is affected by the matrix shape and distribution of the massive fouling particles, and large fouling particle grows preferentially in the direction of primary crystal plane on the surface of the oxide layer. The nucleation, agglomeration, adsorption, growth of the oxide was identified to be affected significantly by fouling deposition, however, it exhibited possibly opposing impacts on the reliability of SG tubes.

Simulation on influences of Al2O3 nanofluids as coolant for nuclear power plant based on modified empirical formula of nanofluids thermal properties at high temperature and high pressure

ABSTRACT The thermal properties of nanofluids are variable with working temperature and pressure and ignored by the existing empirical formula. In the present work, based on the application of nanofluids in pressurized water reactor (PWR) nuclear power plants, different concentrations of Al2O3/H2O nanofluids are prepared for measuring its thermal properties and then modifying empirical formulas at high temperature and high pressure (HTHP). In addition, the primary heat transfer process of PWR nuclear power plants is simulated by numerical method. The results show that, when volume fraction φ within 2.0%, the heat transfer performance of Al2O3 nanofluid increases with the increase of concentrations; however, it is on the contrary when the volume fraction of nanoparticles is larger than 2%. Comparing with water as the coolant, Al2O3 nanofluids with 2% volume fraction give the largest heat transfer enhancement of 103%. Therefore, Al2O3 nanofluids as the coolant can achieve the goals of heat transfer enhancement and safety limit improvement.

Operation optimization of auxiliary electric boiler system in HTR-PM nuclear power plant

Electric boilers (EBs) are the backup steam source for the auxiliary steam system of high-temperature gas-cooled reactor nuclear power plants. When the plant is in normal operations, the EB is always in hot standby status. However, the current hot standby operation strategy has problems of slow response, high power consumption, and long operation time. To solve these problems, this study focuses on the optimization of hot standby operations for the EB system. First, mathematical models of an electrode immersion EB and its accompanying deaerator were established. Then, a control simulation platform of the EB system was developed in MATLAB/Simulink implementing the established mathematical models and corresponding control systems. Finally, two optimization strategies for the EB hot standby operation were proposed, followed by dynamic simulations of the EB system transient from hot standby to normal operations. The results indicate that the proposed optimization strategies can significantly speed up the transient response of the EB system from hot standby to normal operations and reduce the power consumption in hot standby operations, improving the dynamic performance and economy of the system.

Seismic dynamic analysis and structural integrity assessment of the PWR spent fuel pool under beyond design basis earthquake

A series of numerical analyses were carried out to assess the safety of spent fuel pool (SFP) of a pressurized water reactor (PWR) power plant in Taiwan. These analyses performed seismic force evaluation and 3D nonlinear finite element analysis (FEA) for the SFP under beyond design basis earthquake (BDBE). In addition, this study considered various in-structural response spectra (ISRS) with various damping ratios and the reduced factors to account for the partial cracking effects from BDBE. In the seismic force evaluation, the results show that the pool base takes the greatest uniform equivalent loading resulted from the cooling water, racks, and spent fuel components. For the walls, the south wall suffers the greatest equivalent force. The results of 3D FEA indicated that the largest values of stress and strain appear at the bottom corner near south wall of the channel between the main pool and cask loading pit for the overall SFP structure. It is also found that the influences of cracking effects on the seismic force calculation and structural analysis are significant. But comparing the results of numerical analysis by 5% damping ratio with 0.7 reduced factor and 10% damping ratio with 0.5 reduced factor, the impacts on the liner of the SFP are insignificant. Because the cooling water leakage depends on the liner failure, the influence on the leakage is small.

Research on the inlet preswirl effect of clearance flow in canned motor reactor coolant pump

For a pressurized water reactor power plant, the reactor coolant pump (RCP) is a kernel component. And for a canned motor RCP, the rotor system's properties determines its safety. The liquid coolant inside the canned motor RCP fills clearance between the metal shields of rotor and stator, forming a lengthy clearance flow. The influence of inlet preswirl on rotordynamic coefficients of clearance flow in canned motor RCP and their effects on the rotordynamic characteristics of the pump are numerically and experimentally investigated in this work. A quasi-steady state computational fluid dynamics (CFD) method has been used to investigate the influence of inlet preswirl. A vertical experiment rig has also been established for this purpose. Rotordynamic coefficients on different inlet preswirl ratios (IR) are obtained through CFD and experiment. Results show that the cross-coupled stiffness of the clearance flow would change significantly with inlet preswirl, but other rotordynamic coefficients would not change significantly with inlet preswirl. For the case of clearance flow between the stator and rotor cans, influence of inlet preswirl is not so significant as the IR is not large enough.

Development of an Oxide Dispersion Strengthened Austenitic Steel Powder with 4% Aluminum Addition

Development of accident-tolerant fuel (ATF) cladding materials for LWR (Light Water Reactors) is being intensively carried out in many research centers in the world as the lesson-learned from the Fukushima Daiichi nuclear reactor power plant accident. One of the advanced high temperature materials which is intensively developed is Oxide Dispersion strengthened (ODS) steel. The strengthened mechanism is generated by homogeneous dispersion of nanometer sized ceramic oxide particles in the matrix of the steel which is done by mechanical alloying technique so called mechanosynthesis. Synthesis of an austenitic ODS steel powder of SS316L with additional element of 4% aluminum and dispersion of nanoparticles of 0.5% yttrium oxide (yttria: Y2O3) has been done. The aim of the research is to achieve a better performance of SS316L steel as the established material for power plant at elevated temperature. Mechanosynthesis was done using High Energy ball Milling (HEM) for 30 hours with BPR (Ball to Powder Ratio) of 10:1. Afterward, the sample was characterized using Scanning Electron Microscope-Energy Dispersive Spectroscope (SEM-EDS) and X-ray diffraction (XRD) to analyze the microstructure characteristics. The results showed that the alloying between SS316L and aluminum powders was formed. Furthermore, crystal grains and the particles of SS316L-4Al-0.5Y2O3 powder was smaller than as received SS316L and aluminum powders.

A New Neuro-Optimal Nonlinear Tracking Control Method via Integral Reinforcement Learning with Applications to Nuclear Systems

In this paper, a new infinite horizon optimal tracking control method for continuous-time nonlinear systems is given using an actor-critic structure. This present integral reinforcement learning (IRL) method is a novelty method in adaptive dynamic programming (ADP) algorithms and an online policy iteration algorithm. For the optimal tracking problem, the cost function is defined by tracking errors. Consequently, the goal is to minimize tracking errors toward desired trajectories. Since it is hard to solve the Hamilton-Jacobi-Bellman (HJB) equation for continuous-time nonlinear systems control problems, leveraging the actor-critic architecture with neural networks (NNs) to approximate the tracking error performance index and error control law is necessary. Instead of using conventional neural networks, we employ higher-order polynomials in the whole actor-critic architecture. Finally, we apply this new neuro-optimal tracking method to the 2500MW pressurized water reactor (PWR) nuclear power plant, and simulation results are given to demonstrate the effectiveness of the developed method.

Measurements of the 107 Ag neutron capture cross sections with pulse height weighting technique at the CSNS Back-n facility

Silver indium cadmium (Ag–In–Cd) control rod is widely used in pressurized water reactor nuclear power plants, and it is continuously consumed in a high neutron flux environment. The mass ratio of 107Ag in the Ag–In–Cd control rod is 41.44%. To accurately calculate the consumption value of the control rod, a reliable neutron reaction cross section of the 107Ag is required. Meanwhile, 107Ag is also an important weak r nucleus. Thus, the cross sections for neutron induced interactions with 107Ag are very important both in nuclear energy and nuclear astrophysics. The (n, γ) cross section of 107Ag has been measured in the energy range of 1–60 eV using a back streaming white neutron beam line at China spallation neutron source. The resonance parameters are extracted by an R-matrix code. All the cross section of 107Ag and resonance parameters are given in this paper as datasets. The datasets are openly available at http://www.doi.org/10.11922/sciencedb.j00113.00010.

Mathematical Model of the Nuclear Fuel Refueling Mechanism of the BN-800 Reactor and Optimization of its operation

In connection with the burning of nuclear fuel in the reactors of nuclear power plants, the nuclear fuel refueling procedure is periodically carried out. Part of the spent fuel assemblies is removed to the spent fuel pool. Some of them are rearranged inside the reactor to other places, and instead of the unloaded fuel assemblies, fresh fuel assemblies are loaded into the vacated places. The problem under consideration can be considered as a route problem with specific restrictions Beltyukov (2013); Korobkin (2012). The goal of the optimization route problem is to minimize the time spent on reloading fuel assemblies. This is motivated by the fact that the reactor stops during the refueling period and the nuclear power plant incurs material losses. Due to the fact that the optimality criterion is the total operating time of the mechanism, it is necessary to learn how to find the time spent on an elementary operation - moving the grip from one position to another. Knowing the time spent on elementary operations, we can calculate the time spent on the rearrangement of fuel assemblies for a given order of fuel assemblies movement. This work is devoted to the time-optimal algorithm for performing an elementary operation - transferring a grip from one position to another. Knowledge of such an algorithm will make it possible to construct an objective function for the route problem of permutation of fuel assemblies.

Analysis of Efficiency of Technical Condition of Heat Exchanger of Spent Fuel Pool Cooling System with WWER-1000 Reactor of Nuclear Power Plant

The main factors affecting the elements of the heat exchanger of the spent fuel pool cooling system during operation are considered. The analysis of design, technical and operational documentation is carried out for the purpose of preliminary assessment of the technical condition of the heat exchange equipment of nuclear power plants. The data on the installed capacity and service life of power units of the Ukrainian NPPs are given. The cooling system of the spent fuel pool with the WWER-1000 reactor, its purpose, as well as the technical characteristics of the system heat exchanger and its schematic representation are described. The potential mechanisms of aging of heat exchanger elements, aging effects, parameters of technical condition and methods of their control are determined. The efficiency of the heat exchanger of the spent fuel pool cooling system is calculated by determining the rating and design heat transfer areas. The influence of the service life of the heat exchanger on the change in the heat transfer area is shown and the assessment of the effective use of the heat transfer area is determined. The main rating parameters are summarized in the table and the analysis of these values is carried out.

Nondestructive Testing and Condition Monitoring of Pressurized Water Reactor Nuclear Power Plants: A Review

Nondestructive Testing and Condition Monitoring of Pressurized Water Reactor Nuclear Power Plants: A Review

Predictive Execution of Parallel Simulations in Hard Real-Time Systems

Hardware-in-the-loop systems are a type of Hard real-time system that require a faster than real-time simulation. A method of running hard real-time systems with slower than real-time simulations is devised. Instead of waiting for the real-time subsystem output to a simulation, multiple simulation input predictions are generated and used to run several parallel simulation paths independent of the real-time subsystem. The simulation path results are stored until needed by the real-time subsystem at which time one is selected and returned to it to continue the execution of the system. A test case using a small modular reactor nuclear power plant hardware-in-the-loop system was used to demonstrate the effects of slower than real-time simulations on a hard real-time system and to show that this method allows such a simulation to be used.

First-principles computational tensile tests of grain boundaries of zirconia with Cu segregation

Zirconium alloys are widely used as nuclear cladding in pressured water reactor of nuclear power plant, and its corrosion resistance determines the service life of Zr alloys. Alloying can improve the corrosion resistance of Zr alloys, however, the mechanism of alloying element changing the corrosion resistance is still unclear. Since the properties of oxide of Zr alloys are critical to corrosion resistance, we use first-principles calculation to investigate the role of a typical alloying element Cu on the tensile strength of monoclinic zirconia grain boundary (GB) in oxide of Zr alloys. It is found that Cu atom can segregate at GB, and enhance the tensile strength of GB, leading to the improvement of corrosion resistance of Zr alloys. Moreover, we also check the results from calculations with various exchange-correlation functionals, we confirm that it is necessary to use generalized gradient approximation (GGA) plus a Hubbard U correction or hybrid functional instead of standard GGA to calculate the tensile strength of GB with Cu.

Study on the relationship between Fe3O4 fouling and NiFe2O4 oxide layer in the secondary circuit of nuclear steam generator

Fouling is one of the most significant problem in the steam generator (SG) of pressurized water reactor (PWR) nuclear power plant (NPP). To figure out the deposition process, through changing the Fe2+concentration and temperature in the simulated configurations (523 K–603 K), adsorption mechanism of fouling Fe3O4 on steam generator (SG) tube in secondary circuit of pressurized water reactor (PWR) nuclear power plant was investigated by molecular dynamic simulation. The ion concentration distribution, the mean square displacement (MSD), the radial distribution function (RDF), and the energy change before and after adsorption were compared. The influence of the change of temperature and Fe2+ concentration on the corrosion of the SG tube was discussed. Due to the negative surface charge of NiFe2O4, the Fe2+ in the solution did not bond with OH−, but agglomerated with high-pressure high-temperature H2O in the form of triangular bipyramid and octahedron. The diffusion coefficient of OH− increases linearly with the increase of temperature (from 15.28 × 10−5 cm2/s to 18.59 × 10−5 cm2/s) and concentration (from 14.40 × 10−5 cm2/s to 16.06 × 10−5 cm2/s). While the diffusion coefficient of Fe2+reduces, the diffusion coefficient of H2O changes. Simultaneously, Fe3O4 fouling formed on the surface of NiFe2O4 dense oxide layer due to the ion adsorption and interface charges. This work can provide an atomic-scale basis for fouling inhibition in the SG of PWR nuclear power plant.

Demand and Electricity Energy Mix in Indonesia 2030 with Small Modular Reactor Nuclear Power Plant and Renewable Energy Scenario

The interest in small modular reactors worldwide has been increasing due to flexibility in the power generation for more comprehensive users and applications. Small Modular Reactors or SMRs can be the primary choice for Indonesia provided with the geographical condition, which consists of many islands and is more flexible in construction compared to the conventional nuclear power plant. The main objective of this paper is to provide an overview projection of demand and energy mix of electrical in Indonesia 2030 with SMRs NPP in the energy mix referring to RUPTL or General Plan of Electricity Supply Indonesia. Using the end-use model, which is total electricity consumption for each electricity sector, it can be calculated how much electricity demand is from these sectors. The scenario uses RUPTL, roadmap from Energy and Mineral Resources Ministry references, and policy of no coal power plant added from 2020 onwards. The results show in 2030, Indonesia needs 577,016.2 GWh of electricity, where the household and industry sectors have the highest electricity needs, which is 44% for the household sector and 31% for the industry. The transformation projection in PLTGU or Combined Cycle Power Plants scenario also shows that without replacing the power plant, renewable along without SMRs only had ±7.49% of the total capacity mix, and the second scenario with SMRs shows that renewable energy share had 16.07%.

Steam condensation separate effect tests in core make-up tank

The core make-up tank (CMT) is a crucial device of the passive core cooling system in the Westinghouse AP-series pressurized water reactor nuclear power plants. In the prototype plant, the CMT is a vertical cylindrical vessel with hemispherical top and bottom end sockets. Although the structure of the CMT system is simple, it involves many thermal-hydraulic phenomena, most of which significantly impact the peak temperature of the fuel cladding and the liquid level of the core. Since the CMT not only provides full-pressure injection but also triggers the auto depressurization system by its liquid level signal during small-break loss of coolant accident (SBLOCA), its thermal hydraulic behaviors affect the performance of the passive core cooling system (PXS). Therefore, it is necessary to carry out separate effect tests to verify the feasibility of CMT applied to passive safety and injection systems and to study the operating characteristics of CMT. This article first presents the design of a CMT separate effect test facility and analysis of the operating characteristics of the CMT with the experimental data. The influence of pressure, non-condensable gas, wall condensation, and thermal resistance on the CMT behavior were investigated. The study found that the non-condensable gas would stratify in the CMT, which had negligible effect on the wall condensation, but could greatly reduce the gas-liquid interface condensation.

Study on properties of 18MND5 steel forgings for PWR steam generator

In the pressurized water reactor (PWR) nuclear power plant, the shell material of steam generator is required to have good strength-toughness matching, anti-fatigue performance, and neutron radiation resistance to ensure the long-term safe and reliable service. The manufacturing requirements of Manganese-Nickel-Molybdenum alloy steel forgings of steam generator channel head for Hua-long Pressurized Reactor (HPR1000) and Advanced Passive PWR(AP1000) nuclear power plant were compared. The heat treatment process, chemical composition, mechanical properties, metallographic structure of 18MND5 steel forging and SA 508-III steel forging were analysed and studied. The results show that the performance heat treatment temperature of HPR1000 forging has more stringent controls and with lower element content of C, Mo, P, Si, V, Co than the AP1000 forging. By reducing the C content, HPR1000 forging got better toughness while the strength was ensured.

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.

Motion planning of a steam generator mobile tube-inspection robot

Under the influence of nuclear radiation, the reliability of steam generators (SGs) is an important factor in the efficiency and safety of nuclear power plant (NPP) reactors. Motion planning that remotely manipulates an SG mobile tube-inspection robot to inspect SG heat transfer tubes is the mainstream trend of NPP robot development. To achieve motion planning, conditional traversal is usually used for base position optimization, and then the A∗ algorithm is used for path planning. However, the proposed approach requires considerable processing time and has a single expansion during path planning and plan paths with many turns, which decreases the working speed of the robot. Therefore, to reduce the calculation time and improve the efficiency of motion planning, modifications such as the matrix method, improved parent node, turning cost, and improved expanded node were proposed in this study. We also present a comprehensive evaluation index to evaluate the performance of the improved algorithm. We validated the efficiency of the proposed method by planning on a tube sheet with square-type tube arrays and experimenting with Model SG.