Nuclear Power Plant Applications Research Articles

Tri-Loop design and thermoeconomic analysis for the high temperature gas cooled nuclear reactor coupling energy storage system

Thermal energy storage is a proposed solution that enables nuclear power plants to adjust their output without altering power levels. This technology manages fluctuations in the power generation process by storing generated heat above demand levels until it is required to produce steam. Despite the initial investment needed for infrastructure construction, these costs are easily offset by the ultimate returns, especially when efficiency is improved to 85% to 90%. In the context of nuclear power plant applications, helium molten salt energy storage systems demonstrate significant potential as an innovative power generation configuration. In this study, three cycle systems—helium, molten salt, and water (steam)—were employed to investigate and optimize the performance of the system. A series of modeling and simulation processes were conducted using EBSILON software to explore the impact of various parameters (such as pressure, reheat stages, and temperature) on the thermal efficiency of the power plant under both design and off-design conditions. To further study the system's performance, a thermo-economic analysis method based on the second law of thermodynamics was adopted. A thermal-economic analysis model was established, and a fire use analysis was performed on key equipment under representative conditions. Through this approach, we identified the distribution of fire use losses within the system, with the steam generator being the primary contributor to fire use losses among the three systems. Through comprehensive analysis of the simulation results, we conclude that the system exhibits outstanding thermal-economic performance when high-pressure parameters are combined with high temperature. This insight is crucial for modeling and optimizing the construction of nuclear power plants in China, providing a more comprehensive understanding of factors influencing thermal efficiency and contributing to more informed decision-making for the design and operation of tri-generation nuclear power stations.

Effect of non-clustering element C and Mo on irradiation-induced clustering in reactor pressure vessel steels

Commercial reactor pressure vessel steels in nuclear power plant applications contain C and other minor non-clustering elements in addition to the clustering elements of Mn, Ni and Si. However, the effect of such non-clustering elements on irradiation-induced solute clustering in ferritic steels has seldom been addressed. Herein, we reveal that a commercial RPV steel and its Fe-Mn-Ni-Si model alloy with coarse/fine grains subjected to ion irradiation exhibits different irradiation-induced clustering, particularly in chemistry, mean size, and number density of irradiation-induced MnNiSi-clusters. The different chemical compositions of the clusters indicate different thermodynamics in driving solute clustering in RPV steel and its model alloy. The different number density and size evolutions of the clusters indicate different kinetics in governing irradiation-induced clustering. The presence of C and Mo, as non-clustering elements, is primarily responsible for the different clustering behaviors of Ni, Mn and Si in the two materials exposed to ion irradiation. The research finding is of significant implication for modelling and fundamental understanding irradiation-induced clustering in commercial RPV steels.

Advancements and environmental considerations in portland cement-based radiation shielding concrete: Materials, properties, and applications in nuclear power plants– review

Radiation Shielding Concrete plays a critical role in safeguarding against ionizing radiation, in nuclear power plants, medical facilities, and other radiation-sensitive environments. This review paper presents a comprehensive examination of radiation shielding concrete development, highlighting a novel approach that incorporates a diverse range of natural and industrial materials. Traditional radiation shielding concrete formulations rely heavily on virgin materials, contributing to environmental degradation and resource depletion. In contrast, this paper explores the innovative use of natural materials, industrial by-products, and waste materials, emphasizing the environmental sustainability and resource efficiency of such approaches. By harnessing these alternative materials, the paper outlines how radiation shielding concrete production can significantly reduce its ecological footprint while maintaining or even enhancing its radiation shielding capabilities. The paper delves into the functional aspects, composition variations, and performance characteristics of radiation shielding concrete, providing a holistic perspective on the diverse materials employed. Additionally, it thoroughly examines the factors influencing the effectiveness of radiation shielding concrete in nuclear power plants. Furthermore, this review discusses recent developments and advancements in radiation shielding concrete, offering insights into the cutting-edge research and innovations that are shaping the future of radiation shielding. The integration of natural and industrial materials presents promising avenues for enhancing the versatility, cost-effectiveness, and sustainability of radiation shielding concrete.

Feasibility Study on the Application of Boron Carbide for Long Term Reactivity Control in the LOTUS Small Fast Reactor

Abstract 200 MWth lead-cooled fast reactor (LOTUS) reactor core is a small modular lead-cooled fast reactor with designed power of 200 MWth under development at Vietnam National University (VNU) University of Science, Hanoi for a floating nuclear power plant application. For that purpose, advanced passive safety features and no refueling requirement are the priorities in the core design process. To endure the continuous operation over a long lifetime, the startup core exhibits excess reactivity to cover the reactivity loss due to burnup. The reactivity control system includes burnable poison and absorber rods and layers made of B4C which are employed in the reactor to minimize the excess reactivity of the core to about 1 $ to enhance the safety features of the core. The burnable poison is fixed inside the reactor while absorber rods/absorber layers are withdrawn or inserted in sequence to achieve the required excess reactivity of about 700 pcm. The reactivity control was arranged into ten steps to achieve the operating time of 15 effective full-power years without refueling. Good neutronic behavior of the core was observed with negative fuel temperature coefficient and coolant void reactivity and maximum radial power peaking factor of 1.32. However, a quite large residual absorption caused by fixed burnable poison inside fuel assemblies was revealed. In further study, to increase the neutron absorption efficiency of burnable poison in the fast spectrum as well as the reactor lifetime, a neutron moderator will be considered to add to the burnable poison rods.

Event diagnosis method for a nuclear power plant using meta-learning

Artificial intelligence (AI) techniques are now being considered in the nuclear field, but application faces with the lack of actual plant data. For this reason, most previous studies on AI applications in nuclear power plants (NPPs) have relied on simulators or thermal-hydraulic codes to mimic the plants. However, it remains uncertain whether an AI model trained using a simulator can properly work in an actual NPP. To address this issue, this study suggests the use of metadata, which can give information about parameter trends. Referred to here as robust AI, this concept started with the idea that although the absolute value of a plant parameter differs between a simulator and actual NPP, the parameter trend is identical under the same scenario. Based on the proposed robust AI, this study designs an event diagnosis algorithm to classify abnormal and emergency scenarios in NPPs using prototypical learning. The algorithm was trained using a simulator referencing a Westinghouse 990 MWe reactor and then tested in different environments in Advanced Power Reactor 1400 MWe simulators. The algorithm demonstrated robustness with 100 % diagnostic accuracy (117 out of 117 scenarios). This indicates the potential of the robust AI–based algorithm to be used in actual plants.

Unsupervised Causal Deep Learning-Based Anomaly Detection in Nuclear Power Plant Applications

Nuclear power generation will be key to meeting carbon free energy transition goals. However, nuclear power must provide agility and flexibility to fluctuating power demands when other sources of carbon-free energy like solar and wind may not be as flexible. This has led to small modular reactor (SMR) development where nuclear power will be generated from a distributed fleet of smaller reactors where units can be brought online or offline as needed. Due to its high operational and maintenance (O&M) costs as it is, a distributed fleet will put additional cost burden if remote monitoring and crew sharing is not enabled. This requires prognostics and health management (PHM) capabilities such as early warning, diagnostics, and prognostics to enable predictive maintenance with high accuracy. Typically, monitoring solutions are developed on component and subsystem levels targeting specific failure modes. However, it is argued that a systemwide monitoring, in addition to specific targeted analytics, would be of key importance. This paper presents a deep-causal unsupervised anomaly detector that has been successfully applied in various aerospace and renewable energy applications. In this paper we share our experience applying this method on a nuclear power plant (NPP) application. Specifically, we share how we dealt with practical challenges of data quality, ground truth labeling, performance evaluation and field validation in an unknown-unknown setting where prior knowledge of failures and failure modes were not available to begin with.

Risk-informed Emergency Response Training for Backdraft in Nuclear Power Plants

Research has been conducted for developing fire evacuation and response training programs for nuclear power plant (NPP) application. Among numerous fire scenarios that may occur in an NPP environment, three different points of origin for a fire were selected for the program based on a risk-informed approach: switchgear room, main control room, and safety injection pump room. Fire outcomes were predicted for these scenarios via numerical modeling and the results were incorporated into the newly developed fire evacuation and response training program for the APR1400, Korea’s next-generation NPP model. The switchgear room fire scenario was found to have the most potential for backdraft to occur during manual fire response following automatic gaseous fire suppression system activation. The emergency response manual does discuss this possible backdraft occurrence; however, the guidance to avoid injuries is qualitative, such as to be cautious of backdrafts and wait a sufficient amount of time after opening a door before entering the. In this study, backdraft phenomenon that may occur from a switchgear room fire was numerically examined using the recent version of the Fire Dynamics Simulator to develop an appropriate timeline to be implemented in the fire evacuation and response training program. Based on the findings, the following guidance is provided: (1) backdraft can only occur when the fire originates in the space near the door; (2) wait at least 10 minutes after opening the door before entering the room; (3) watch for rapid smoke production, as this may be an antecedent phenomenon of backdraft; and (4) when smoke production increases rapidly, leave the room as soon as possible to avoid being caught within the deflagrating flames from a backdraft.

Study on Performance of New Flexible Shielding Materials and First Demonstration Application in Chinese Pressurized Water Reactor Nuclear Power Plant

Abstract Excessive gamma rays will be emitted when a nuclear power plant is under overhaul, leading to a certain number of hotspots. To meet the shielding requirements of these hotspots, a new type of flexible shielding material with a “sandwich structure” has been developed. The new flexible shielding materials consist of two parts: one is silicone rubber with excellent physical and chemical properties by adding reasonable multicomponent additives as a matrix, and the other is lead powder as a shielding body. This kind of material, with lead powder dispersed uniformly in silicone rubber, was prepared through a mixing process. Based on this, the article has carried on the discussion with the performance of the functional layer of the new flexible shielding materials. The problem of the process stability of ultrafine powder distributed uniformly in a matrix with various densities among different regions has been solved as well. Furthermore, this kind of material has been adopted at the Chinese pressurized water reactor nuclear power plant as the first demonstration project. The site shielding effect has been recognized by the World Association of Nuclear Operators. As a result of this research, it is possible that the design of a radiation shield for a short or long time can be made quickly when a nuclear power plant is running or being overhauled. Meanwhile, this study is of great significance for the popularization and application of shielding materials.

Mechanisms and inhibition method investigation on condensation induced water hammer in an equal-height-difference passive heat removal system

The equal-height-difference passive heat removal system utilizes the ocean as an ultimate heat sink for residual heat removal, demonstrating obvious advantages for ocean nuclear power plant applications. However, the structure with a zero-height difference can trigger the reverse flow of subcooled water. This could cause serious condensation induced water hammer events. In this study, a fully visual experimental setup was established to investigate the occurrence features. First, two kinds of formation mechanisms were defined by the visual experimental results. In this regard, the subcooling and the isolated steam slug are the necessary conditions for the formation and development. In addition, the occurrence positions were found to be concentrated in the test pipe section near the water tank. Then, a butterfly check valve was designed to avoid the condensation induced water hammer events based on the formation mechanisms. The experimental results showed that the check valve usage can effectively inhibit the reverse flow of the subcooled water and thus achieve its purpose of eliminating condensation induced water hammer events. Moreover, the check valve transforms the two-phase natural circulation flow from the condensation oscillation mode to the stable boiling two-phase flow mode. As a result, it enhances the two-phase natural circulation capacity.

Preliminary Study on Alternative Reflectors for LOTUS Reactor

LOTUS reactor - a compact lead-cooled fast reactor is currently being studied to generate 200 MWth of capacity and operate for 20 years without refueling for the floating nuclear power plant (FNPP) application. Therefore, it would be a big advantage to achieve a long operation without refueling. However, a small reactor usually has a higher neutron leakage, and a good neutron reflector is essential to maintain the neutron economy. The main objective is selecting potential reflectors materials for the LOTUS reactor using Monte Carlo code, Serpent. The various candidate reflector materials, including Al2O3, BeO, MgO, PbO, SiO2, and ZrO2 are calculated from the neutronics characteristics to determine a good neutron reflector. In this work, we have investigated the parameters of neutronics characteristics, such as core neutron flux spectrum, evolution of keff due to burn-up, power distribution, and lead coolant void reactivity with each reflector. From the comparison of those parameters, MgO material was found to be a good candidate for the relector of LOTUS reactor.

Experimental assessment of impedance-based structural health monitoring in radioactive environment

This work presents an experimental study of the influence of radiation on structural health monitoring systems based on the electromechanical impedance method using low-cost piezoelectric diaphragms. For application of the method, the baseline was obtained without radiation and then compared by damage indices after application of radiation. Considering applications in nuclear power plants, the irradiation and calculation of damage indices were performed in the range of 10 kGy–60 kGy at 10 kGy intervals. Impedance measurements were performed in seven frequency bands between 0 and 65 kHz. The results show that, due to the change in the impedance signatures, the damage indices values increase according to the total applied radiation dose, which may indicate a false indication of damage. Besides, results indicate that there are frequency bands less sensitive to radiation and a threshold can be defined to distinguish radiation from structural damage.

Dissimilar welding of austenitic and ferritic steels using nickel and stainless-steel filler: Associated issues

The microstructure development and mechanical behaviour of dissimilar metal welds between ferritic and austenitic steel, as well as their application in nuclear power plants, are discussed in this review paper. Nuclear reactor components, such as steam generators and pressure vessels, consist primarily of SA508 due to their low cost and high operating temperatures and pressures. The welding of dissimilar metals is crucial due to variations in physical characteristics such as thermal conductivity, thermal expansion coefficient, mechanical properties and chemical compositions. The principal challenges of dissimilar ferritic and austenitic steel welding are the subject of this review work. Weldability issues include a sharp change in mechanical and metallurgical characterization across the fusion line, carbon migration, cyclic thermal stresses and residual stresses, which necessitate a thorough investigation of the welded joint. Generally, austenitic steel and nickel-based fillers are used to join austenitic and ferritic steel materials; however, owing to many weldability concerns, nickel-based consumables are replacing austenitic consumables. Another critical issue in the weld joint is the selection of appropriate welding consumables, and detailed explanations of the benefits of employing a buttering layer on the ferritic side are provided. The effect of heat treatment on the metallurgical and mechanical characteristics of the weld joint, as well as the formation of residual stress, has also been thoroughly explored.

Noninvasive liquid level sensing with laser generated ultrasonic waves

This article proposes a noninvasive liquid level sensing technique using laser-generated ultrasound waves for nuclear power plant applications. Liquid level sensors play an important role of managing the coolant system safely and stably in the plant structure. Current sensing techniques are mostly intrusive, performing inside the fluidic structure, which is disadvantageous in terms of the regular maintenance of the plant system. Furthermore, typical intrusive sensors do not perform stably under varying environmental conditions such as temperature and radiation. In this study, sensing units are attached to the outer surface of a liquid vessel to capture guided ultrasound waves in a nonintrusive manner. The signal intensity of the guided wave dissipates when the signal interacts with the internal liquid media. The sensing mechanism is mathematically expressed as an index value to correlate the liquid level with the sensor signal. For the acoustic wave generation, laser-generated ultrasound was adopted instead of using typical contact type transducers. Following the simulation validation of the proposed concept, the performance of the developed sensor was confirmed through experimental results under elevated liquid temperature conditions. The nonlinear multivariable regression exhibited the best-fit to the datasets measured under the variable liquid level and temperature conditions.

Investigations of tool wear of coated carbide tool while machining X-750 nickel alloy

Modern advanced structures and machinery rely heavily on advanced materials. Ni-based superalloys have applications in nuclear power plants, gas engines, airplane engines, turbines, etc. All these environments have high temperatures, thus; the material is selected such that it demonstrates sufficient strength at high temperatures. Ni-based superalloys present this characteristic due to their “high strength temperature resistant” (HSTR) nature. In the present work, an attempt has been made to machine X-750 nickel alloy on conventional machining considering rotational speed (TRS), feed (F) and depth of cut (DoC) as input machining variables. The coated carbide insert has been used at different machining cutting of input machining variables. The tool wear is assumed as an output parameter and the effect of input of parameters on the tool wear has been investigated. It has been observed that with the increase in rotational speed, feed and depth of cut the value of flank wear increases. The main mechanisms of tool wear exist in the present study are abrasion, adhesion and chipping, which are observed after the analysis of tool inserts by optical as well as scanning electron microscopy. The optimized setting for minimum TW (0.2716 mm) is TRS: 1362RPM; DoC: 01 mm and F: 0.07 mm/rev.

The impact of outgassing on the nano-oxides kinetics of ferritic oxide dispersion strengthened steel

Oxide Dispersion Strengthened (ODS) steels are studied for their application in fission and fusion nuclear power plants. These steels are processed by powder metallurgy involving a high energy milling of Fe-Cr with oxides powders, followed by outgazing and hot consolidation. The temperature increase during these last two stages triggers the precipitation of a fine dispersion of nano-oxides. The outgazing stage is generally conducted at temperature between 300 and 800 °C for several hours, which can already trigger the formation of nano-oxides, and thus may have an influence on the final nano-oxide characteristics. In this work we assess the nano-oxides size by Small Angle X-Ray Scattering (SAXS) and get insight on the nano-oxides chemical and structural properties by comparison between SAXS and Atom Probe Tomography (APT) during a thermal treatment at 700 °C for 5 h, simulating the outgazing stage. We show that despite that the nano-oxides slightly grow during the 700 °C heating, the final properties of the nano-oxides in term of size and number density are extremely close to those where no 700 °C isothermal was applied. Moreover, the growth rate of the nano-oxides is surprisingly higher at 700 °C than 1100 °C. APT and SAXS comparison shows that the nano-oxides do not stabilize in terms of stoichiometry and structure in the studied conditions, and that a small quantity of Al remain in the precipitates during this 700 °C isothermal treatment. Both of these factors could influence the nano-oxides kinetics at low temperature. • The outgazing shows no influence on the final nano-oxides size and number density. • No stabilization of the nano-oxides is observed at 700 °C after 5 h isothermal. • The nano-oxides growth rate is surprisingly higher at 700 °C than at 1100 °C. • Nano-oxides contains Al on the 700 °C isothermal, which is removed after heating at 1100°C. • TiC or Ti(C,N) is likely to form after the outgazing, between 700 and 1100 °C.

Analysis and Implementation of Sliding Mode Controller-Based Variable Frequency Drive Using the SCADA System

Vector control of an asynchronous machine is traditionally accomplished by analogizing it to a separately excited DC machine. It provides decoupled torque and flux control that is perpendicular to each other, ensuring that neither vector interferes with the other. So, despite their close interconnection, torque and speed control are accomplished separately. The rotor flux is aligned with the direct axis of the synchronously rotating reference frame to achieve this. The PI controllers are critical in achieving the variable frequency drive (VFD) desired topology. The system employs three types of controllers: flux, speed, and torque. The flux controller is easy to tune, but the speed and torque controllers are more difficult to tune because the speed controller's output is the torque controller’s reference signal. Furthermore, there is no well-defined method for tuning the controllers in a vector control system. However, perfect tuning is required for the machine’s better dynamic behavior. It is clear from the above analysis that system identification is critical for tuning PI controllers. However, as an asynchronous machine, obtaining a decoupled system transfer function is extremely difficult. To solve this problem, the proposed system combines a seven-level pulse width modulation (PWM) inverter for vector control of a three-phase asynchronous nonstandard induction machine used in critical applications in nuclear power plants with a sliding mode control technique that eliminates the complexity of PI tuning. A second-order sliding mode controller could be used in the future to reduce the chattering and parameter variation effects. This controller can be enhanced with fuzzy logic principles to make it more robust and reliable, allowing it to be used in future drive designs for high-rating motors with critical applications.

Numerical Investigation of Thermal-Hydraulic Performance of Printed Circuit Heat Exchanger with Different Fin Shape Inserts

The printed circuit heat exchanger is one of the most recent important heat exchangers, especially in the nuclear power plant and aerospace applications, due to its very compact geometry and small print foot. This paper presents a 3D numerical investigation of the thermo-hydraulic performance of PCHE with a new non-uniform channel design configuration. The new channel design consists of two different fins and shape inserts: the diamond and biconvex shapes. The influence of two design parameters on the heat exchanger performance was studied and optimized, the longitudinal and transverse pitch length (Pl) and (Pt). Air with constant properties as the working fluid with constant heat flux at the walls envelope. The Reynolds number varied from 200 to 2000. Different Pitch lengths were used (Pl=20, 30, 40, and 50) mm and (Pt=3, 4, and 5) mm. Three performance parameters were studied the Nusselt number, friction factor, and the overall performance evaluation factor. Results show that the thermal performance enhanced with decreasing the pitch lengths, and it was shown that this enhancement was found only at high Reynolds numbers above 1400. The higher enhancement factor was with NACA 0020 airfoil fins at pt=3 mm and pl=20mm of η=2.75 at Re=2000, while the worst performance was obtained with biconvex fins. The main reason behind the enhancement is the disruption of the boundary layer and the good mixing induced in the fluid flow.

Study of cracking and microstructure in Co-free valve seat hardfacing

• Through-thickness cracks were found on Co-free hardfacing with NOREM 02. • Brittle cracking propagate along the interdendritic network of carbides. • Solidification defects and lacking of ferrite phase was observed. For reducing the radiological hazard, Co-free hardfacing alternatives have been used to replace the traditional Co-based hardfacing alloys in nuclear power plant applications. However, manufacturing a defect-free Co-free hardfacing is challenging. Reduced galling resistance and mechanical properties exhibited by common replacement alloys in some conditions limit the application. This issue is important for the operating nuclear power plants with lifetime extension and also for the new builds. In this paper, a failed industrial hardfacing was studied. Through-the-thickness cracks of the Co-free hardfacing deposit with NOREM 02 on a nuclear swing check plug were detected. The potential cause of the brittle interdendritic cracking was studied and the detailed microstructure of the Co-free hardfacing analysed. The solidification defects during the manufacturing and the lacking of ferrite phase in the microstructure together with the potentially non-optimized handling procedures are concluded to be the causes of the observed cracking.

Corrosion Enhancement for FGM Coolant Pipes Subjected to High-Temperature and Hydrostatic Pressure

The objective of this paper is to enhance the corrosion resistance of coolant pipes in high temperature and lateral hydrostatic pressure in critical engineering environment, especially for circular coolant pipes under hydrostatic pressure of LBE (lead-bismuth eutectic) applications in nuclear power plants. The resistance against corrosion caused by LBE liquid is mainly formed by Fe-12Cr-2Si solid solutions coatings on the pipe. The silicon concentration in Fe-12Cr-2Si can interact with LBE as an effective oxidized compound such as SiO2 and Fe2SiO4 when the silicon concentration is higher than 1.25 wt.%. The oxide film formed on the coating can resist the LBE corroding in the Fe-12Cr-2Si structure. The primary material of a constructing coolant pipe is T91 ferritic-martensitic alloys, and the surface anti-corrosion coating is Fe-12Cr-2Si solid solution. With a high strength structure, FGC (functionally graded composite material), ensures that the pipe resists the corrosion from LBE liquid. In this study, both the steady-state stress values and silicon concentration are evaluated at 700, 1000, and 1200 °C to know the fatigue problems. The research result indicates the FGM (functionally graded material) structure performs better in promoting the margin of safety on stress distribution and reserving the silicon concentration on the inner surface higher than 1.25 wt.% over 60 years as compared to the FGC structure with 34 μm thickness of Fe-12Cr-2Si coating in a high temperature environment.

Thermal and radiation degradation effects on fluoroelastomer destined to gasket applications in nuclear power plants

Thermal and radiation degradation effects on fluoroelastomer destined to gasket applications in nuclear power plants