Circuit Of Reactor Research Articles

Evaluation of the unavailability of the primary circuit of Triga SSR reactor, importance factors and risk criteria for its components

Abstract Probabilistic Safety Assessment is a tool for improving a nuclear installation and its safety in all its function phases. PSA provides a methodical way of identifying sequences resulting from a wide range of initiating events concerning an accident, and includes systematic and realistic determination of accident frequencies and their consequences. This paper analyzes the unavailability of the TRIGA SSR reactor primary circuit, obtaining the system unavailability value, identifying main contributors to system failure, and calculating risk importance factors. Two cases were considered: one without considering Common Cause Failure (CCF) in the developed fault tree, and the second considering CCF on pumps and heat exchangers lines. The aim was to assess their influence on the primary system’s unavailability. Two importance factors, Fussell–Vesely and Risk Achievement Worth, were employed to identify components with a strong impact on the risk (probability of system failure). These factors aid in enhancing maintenance activities and testing. The analysis utilized the EDFT code, part of the PSAMAN code package developed in SCN-Pitesti.

Validation of coupled ATHLET-OpenFOAM simulation on a large-scale single- and two-phase flow experiment

Abstract System thermal-hydraulics (STH) codes are successfully used in the last decades for design and analysis of the physical behavior of nuclear power plants (NPPs). These programs provide reliable results at comparatively low computational cost, but have limited capabilities to predict complex 3D flow phenomena, which is possible with Computational Fluid Dynamics (CFD) tools. In order to combine the advantages of STH and CFD codes, algorithms to couple these are developed. At GRS, a new OpenFOAM solver, based on the volume of fluid (VOF) method, was developed together with new coupling boundary conditions to couple the CFD program OpenFOAM with the system code ATHLET. In a next step, verification and validation activities were carried out. These included the analysis of a Pressurized Thermal Shock (PTS) scenario. For this purpose, data from the OECD/NEA ROSA V Test 1.1, carried out in the Japanese Large-Scale Test Facility (LSTF) was used. The experiment investigated the temperature stratification after emergency core cooling (ECC) injection under single- and two-phase natural circulation conditions in the primary cooling circuit of a pressurized water reactor (PWR). This paper discusses the results of the coupled simulations, compares the numerical results with experimental data, and identifies areas for future improvements.

Study on tritium permeation behavior in primary and second circuits of the hydrogen production system by methane steam reforming using HTGR

Hydrogen production by methane steam reforming (MSR) using the process heat from High-temperature gas-cooled reactor (HTGR) is one of the most promising technology for hydrogen production using nuclear energy in the near future. The permeation behavior of tritium produced by the reactor is extremely important when analyzing the safety of this hydrogen production system. There are two methods of connection between a nuclear reactor and a hydrogen production system. One is direct coupling that the steam reformer of the hydrogen production system is directly connected to the first circuit of the nuclear reactor. The other is indirect coupling via an intermediate heat exchanger (IHX) to connect nuclear reactor and hydrogen production system. Hence, this work proposed a model to analyze the difference in tritium permeation between these two coupling ways and to compare the tritium permeation behavior for different IHX alloys at various operating conditions. This model includes tritium's production rate, release fraction, and decay reduction in the system. The results show that the pre-exponential factor of the IHX alloy mainly limits the tritium permeation rate. In addition, the tritium permeation rate increases about two times when the temperature rises from 750 °C to 950 °C at a steady state. Further, the decrease of tritium permeation rate is seen from the calculation results for the system with IHX, e.g., for T = 750 °C, the mean permeation rate of tritium into the product gas is 9.31 × 10−12 mol/s at the whole year for indirect coupling with an IHX, compared to 3.86 × 10−12 mol/s for direct coupling, which reduced 58.54 % (This means that with IHX the permeation rate is lower than the system without IHX). These results will contribute to designing an effective and safe hydrogen production system by MSR using HTGR process heat.

Effect of heavy ion irradiation on the electrochemical behavior of 321 stainless steel

Aiming to the environment for 321 stainless steel (SS) in primary circuit of pressurized water reactor (PWR), the effect of heavy ion irradiation to simulate neutron irradiation on the electrochemical behavior of alloy is unclear. The purpose of this work is to evaluate the corrosion resistance of 321 SS at different doses and the contribution of heavy ion irradiation in it. This work reveals that the corrosion resistance decreases in the order of 2 dpa > 0 dpa > 10 dpa > 60 dpa at room temperature, consistent with the sequence of δ-phase content near the surface. The effect of irradiation dose on the corrosion resistance of 321 SS is inversely proportional at higher dose and optimal performance of the SS irradiated to 2 dpa is attributed to the highest content of Cr-rich δ-phase. The pits occur in C enrichment zone, accompanied by increase of O content, while no pit occurs in Ti enrichment area. The results provide a reference for the irradiated SS with heavy ion to acquire optimal pitting resistence.

(Invited) COVID-19 As a Technology Accelerator for High-Speed Polymerase Chain Reaction (PCR)

Limit of detection (LOD), speed and cost for some of the most important diagnostics tools for COVID-19 i.e., lateral flow assays (LFA), enzyme-linked immunosorbent assay (ELISA) and Polymerase Chain Reaction (PCR) all benefitted from both the financial and regulatory support brought about by the pandemic. Before COVID-19, LFA was the fastest and least expensive but came with the poorest limit of detection (LOD). The slowest, most expensive but with the best (lowest) LOD was PCR. In between those two extremes, in terms of cost, LOD and speed, was an ELISA . Today, through the “COVID-19 accelerator effect” there are point of care (POC) reverse transcriptase quantitative PCR (ie., RT-qPCR) COVID-19 tests on the market that provide sample to answer in under 20 minutes, while retaining the same very low LOD but at a lower cost [1]. There are even POC PCR tests under development that come in at less than 10 minutes: beating both LFA and ELISAs in terms of time and LOD [2]. Although RT-qPCR has gained the most in overall performance, signal-amplified POC LFAs are possible now with an LOD that rivals that of an ELISA.However, implementing the next generation PCR equipment in point-of-care settings, remains very challenging because of difficulties in reaching a 10 minute or below sample to answer time , instrument complexity and cost. In this work, we discuss the engineering challenges involved in the development of high-speed microfluidic PCR equipment. To guide this review/comparison of PCR advances we introduce three parameters. The first one is the overall sample to answer time (t), the second is called lambda (λ) and is a measure of the degree to which stochastic effects interfere with obtaining analytical results and thereby setting the minimal number of copies required per volume. The third parameter gamma (γ), introduced first by Dong et al [3], is the ratio of thermal cycling time (in min) to the heat cycled sample volume (in mL). It describes the trade-off between the speed of the PCR system and the desired level of sensitivity. It can be interpreted as a system efficiency: the most efficient system heat cycles the largest amount of sample in the smallest amount of time.With these three criteria in mind we set out to suggest designs for an optimized POC reverse transcriptase (RT) quantitative qPCR or RT-qPCR system. Applications of RT-qPCR include quantifying gene expression levels, validating RNA interference (RNAi), and detecting pathogens such as viruses. The aim of this specific POC RT-qPCR system is to provide sample to answer in ten minutes or less for a respiratory pathogens panel, with an LOD of 500 copies/mL and the capability to multiplex (8 chambers) while remaining portable and low cost.Finally we introduce two of the most daunting, remaining, microfluidics challenges that both need to be tackled to implement an ideal microfluidic platform like the POC RT-qPCR system described above. One is liquid storage and the other is proportional valves on a disposable microfluidic platform. Liquid storage is essential when the assay requires repeated washes, if these washes can come from one and the same reservoir via a proportional valve, real estate is gained and system complexity is reduced. The storage issue is further complicated when the design also calls for storing lyophilized or dried reagents on the same disposable. We compare some of the most recent approaches to liquid storage and proportional valving that we believe will enable the fast POC diagnostic platforms of the future. 1. Tsang, Hin Fung, Wai Ming Stanley Leung, Lawrence Wing Chi Chan, William Chi Shing Cho, and Sze Chuen Cesar Wong. “Performance Comparison of the Cobas® Liat® and Cepheid® GeneXpert® Systems on SARS-CoV-2 Detection in Nasopharyngeal Swab and Posterior Oropharyngeal Saliva.” Expert Review of Molecular Diagnostics 21, no. 5 (May 4, 2021): 515–18. 2. BUCKLAND, Justin, Tom JELLICOE, Alex STOKOE, and Amaru ARAYA-WILLIAMS. “VARIABLE TEMPERATURE REACTOR, HEATER AND CONTROL CIRCUIT FOR THE SAME.” Patent Application, January 30, 2020. 3. Dong, Xiaobin, Luyao Liu, Yunping Tu, Jing Zhang, Guijun Miao, Lulu Zhang, Shengxiang Ge, Ningshao Xia, Duli Yu, and Xianbo Qiu. “Rapid PCR Powered by Microfluidics: A Quick Review under the Background of COVID-19 Pandemic.” Trends in Analytical Chemistry 143 (October 2021): 116377.

Numerical analysis of liquid sodium spreading on sloped floor surface under pool fire scenario in SFR systems

The accidental leaks of coolant from the heat transport circuit of Sodium-cooled Fast Reactor (SFR) can create fire events due to the reactivity of sodium with atmospheric oxygen, which can pose threat to safe operation of the reactor. The leaked sodium eventually reaches the cell floor, spreads gradually, and forms a burning pool. The thermal consequences of pool fire depend on its surface area, which is determined by the spreading process of sodium on the floor surface. In this context, numerical investigation has been carried out on the spreading behavior of burning sodium pool on the plane/sloped cell floor by developing a 3-D numerical model based on the viscous-gravity current equation of the shallow layer model. The governing equations of the model have been solved numerically by implementing the Finite Volume Method. The model developed has been validated using the experimental results published on liquid pool spreading under burning and non-burning conditions on plane and sloped floors for different leak rates. Using this model, numerical simulations have been carried out on sodium pool spreading under various leak scenarios envisaged in SFR cells based on the specific burning characteristics of sodium pool fire. The time-varying pool size and shape, pool height profile, and maximum pool area have been estimated for a wide range of sodium leak rates and typical floor slopes of interest to nuclear power plants. These predictions contribute towards the numerical evaluation of sodium pool fire scenarios pertaining to the safety of SFR.

Gas entrainment in surge tank of CFR600

The Surge Tank in the secondary sodium circuit of China Demonstration Fast Reactor(CFR600)is filled with sodium and argon. The entrainment of gas in the surge tank will be occurred if the operation conditions is not appropriate. Through theoretical analysis and experimental studying it was found that the mechanisms of gas entrainment for surge tank were focused on three points: (1)the vortex of the free surface, (2)the design of the inlet and outlet nozzle, (3)the design of the route for suction- flow. A water experiment was performed using a 1/10 scale model by geometric similarity. It was observed from the experiment that the minimum flow rate required to gas entrainment occurred was about 130% by standard running condition. The experimental critical Froude number is 4.9, when the design value is 2.5. Through the test of the water model we can see that vortex at the free surface of sodium-argon grow up along with the ever-increasing flow velocity. Entrainment of gas was occurred when the intensity of vortex was enough. The results of the experiment confirmed that the vortex is the major factor for gas entrainment of surge tank of CFR600. But anti-vortex baffle can delay the forming process of vortex effective, and raise the critical velocity substantially.

Influence of uniform magnetic field on hydrothermal characteristics and entropy production in a nanofluid filled rectangular grooved channel

The implementation of a uniform magnetic field can be beneficial for controlling thermal convective processes in magneto-thermal devices and systems. In research of productive energy employment and superior thermal efficiency, the present numerical research study aimed to investigate the flow structure, thermal behaviors, and entropy production of nanofluids with a mixture of CuO and water in a rectangular grooved channel subjected to the uniform magnetic field in the transverse direction. Convective dynamics in such rectangular grooved channels subjected to a uniform magnetic field have not been comprehensively investigated considering flow structures, thermal performances and entropy characteristics. The impact of different variables namely nanoparticle volume concentrations, Φ Hartmann numbers, Ha and Reynolds numbers, Re on the hydrothermal characteristics is numerically explored. According to acquired outcomes, the uniformly implemented magnetic field in the rectangular grooved channel causes significant variations of flow characteristics demonstrated by the streamline patterns and reduces the extent of the recirculation flow zone in rectangular grooves at Re = 250 and 1250. During the application of the uniform magnetic field, the thermal boundary layer formed in the rectangular grooves becomes thinner, while the temperature gradient in the near regions of the heated walls increases. Numerical simulations show that the average Nusselt number, Nuavg increases by approximately 9.08% for Ha = 8 and 30.42% for Ha = 24 at Re = 250, and by 0.087% for Ha = 8 and 21.13% for Ha = 24 at Re = 1250, compared to the case where no uniform magnetic field is applied, Ha = 0. It can be observed that for a given value of Re, the total entropy reduces sharply for Ha>8, due to the significant effect of high magnetic field intensity over both thermal and flow distributions. The outputs of this study could be very beneficial for designers modeling any device or system including thermal energy transportation in various industrial applications such as cooling circuits of electronic components and fast fission reactors, materials science and metallurgical processes.

An Increase in the Energy Efficiency of a New Design of Pumps for Nuclear Power Plants

The reliability of pumping units at nuclear power plants (NPPs) is critical in terms of their energy efficiency and safety. Remarkably, WWER-1000 reactors at Ukrainian NPPs are equipped with outdated pumping units that have already served their full-service life. This fact leads to an urgent need to develop a new, more efficient pump. In the article, a promising pump, ACNA 600-35, was developed. It was designed to increase the energy efficiency of pumps TX 800/70/8-K-2E, applied at the holding pool and the industrial circuit of the nuclear reactor. Since these pumps should be imported from the monopoly suppliers, this affects both the energy efficiency of pumping equipment and the energy independence of Eastern Europe. The proposed pump ACNA 600-35 is characterized by an increased efficiency of up to 0.12–0.13 compared with the TX 800/70/8-K-2E pump. In general, the life cycle cost of the proposed pump is 15–20% lower than for the analog TX 800/70/8-K-2E. The design of the developed pump ACNA 600-35 and the related pumping unit based on its production at industrial facilities allows for further development of the industrial and fuel-energy complex, increasing the state’s energy independence and employment. According to expert estimates, the average economic effect from supplying the developed pump can reach 10 mln USD/year.

Review of Inaccuracies in the Analysis of the Flow Stability Specific of the LW-SMR Modular Design

The presented analysis has been issued to express doubts about the correctness of the engineering and operational safety assessment, which arises from the modeling of transient modes and methodologically should help light water small modular nuclear reactor designers/developers to correct the model they are usually creating for studying the stability of the thermal-hydraulic circuit of the light water small modular nuclear reactor with the natural circulation of primary coolant. After the NuScale report publication and the following presentation review, additional questions arise regarding the analysis methodology, which was carried out and presented as part of the licensing package presented to NRC. After having carefully and critically read the available part of the presented report on the study of instabilities and the presentation issued by NuScale, the answer to the NRC safety committee request could be concluded that the instability analysis has not been performed correctly not only technically, but theoretically. Some clarification of the problem of analysis of instabilities is required additional explanations for further integral light water small modular nuclear reactor design developments and correct analysis implementations.

Hydrogen Sensor to Monitor the Conditions in the Primary Circuit of a Nuclear Reactor

AbstractA hydrogen concentration measurement device is developed for nuclear reactor primary circuits water. Despite the necessity to monitor hydrogen, only one producer offers a really selective hydrogen sensor on the world market because of the complexity of the device preparation when long‐term signal stability is required. Therefore, the Mass Transfer Laboratory at UCT Prague shares the experience gained during the sensor development. The results of the amperometric hydrogen sensor laboratory tests are presented, which demonstrate how to improve the sensor durability. An operational test in Dukovany nuclear power plant confirmed a significant improvement of the latest sensor version and showed that a simple principle could be realized in the form of a reliable device for industrial measurements.

Studies on In-situ regeneration of cold trap of a Bench-Top sodium loop

A cold trap (CT) is one of the crucial modules of the sodium coolant circuit of a sodium-cooled fast reactor as it controls hydrogen, oxygen and other impurities in liquid sodium within the threshold levels in dynamic mode. The major impurities are crystallized as NaH and Na2O in CT and get loaded during its continuous service. The CT at the secondary sodium circuit needs regeneration at regular intervals of time. In-situ regeneration of the CT during the reactor operation would be cost-effective while generating power. A bench-top sodium loop was employed to carry out the present study. The regeneration of CT was carried out by adopting a thermo-vacuum process in which the NaH decomposed into sodium and hydrogen. During the regeneration process, hydrogen release behavior as a function of time with varying temperature and cover gas pressure was investigated by deploying in-house developed online hydrogen sensors. An electrochemical hydrogen meter (ECHM) to measure hydrogen in liquid sodium in the ppb range, a semiconducting metal oxide-based hydrogen sensor (SMOHS) and a polymer electrolyte membrane-based hydrogen sensor (PEMHS) to measure hydrogen in the argon cover gas were used. Simultaneous measurement of hydrogen in both liquid sodium and cover gas phase played a central role in analyzing the hydrogen distribution phenomenon. The quantity of total hydrogen evolved and decomposed NaH was determined after the successful regeneration of the cold trap.

Criteria of thermochemical conditions of steam gas explosions in dynamic accident modes at nuclear power units with WWER reactors

The main lessons of the major accident at the Fukushima-Daiichi NPP in 2011 for the nuclear power industry identify the need to model, analyze and develop emergency measures for relatively unlikely events with catastrophic environmental consequences, taking into account multiple failures of safety systems. Steam-gas explosions became one of the main causes of the catastrophic environmental consequences of the Chernobyl and Fukushima accidents. Criteria and conditions for the occurrence of steam-gas explosions in dynamic emergency modes in the "tight" reactor circuit of nuclear power reactor units (NPP) with water-water reactors (WWER) with failures of safety systems valves and emergency steam gas removal are determined by the rate of change of thermodynamic and physico-chemical parameters. A method for determining the criteria and conditions for the occurrence of steam-gas explosions in dynamic accident modes with a "tight" reactor circuit and failure of safety valves for modeling the initial emergency events – seismic effects, falling of massive objects, etc. is presented. The conditions for the occurrence of hydrogen explosions are determined by the maximum rate of increase in the temperature of fuel oil shells, and the conditions for steam explosions are determined by the maximum rate of pressure increase as a result of the intensification of vaporization processes. The criteria for the occurrence of steam explosions in dynamic emergency modes are determined by the maximum pressure amplitude and the propagation speed of acoustic disturbances in the steam volume. And the criteria for hydrogen detonation in dynamic emergency modes are determined by the maximum amplitude of the increase in the temperature of the fuel shells and the average flow rate of the coolant in the active zone of the reactor.

The Influence of High-Temperature Helium and the Amount of Revert Material on the Material Properties of Inconel 738

Nickel-based alloys are considered promising materials for primary circuits of high-temperature gas reactors (HTGRs), specifically for gas turbines. The primary helium (He) coolant in the gas-turbine-based HTGRs is expected to reach temperatures of up to 900 °C; therefore, the selected materials should adequately perform over a long service life at such an environment. A promising manufacturing method in the production of reactor components is precision casting, where the content of revert (recyclate) material in the alloy differs and can influence the material behavior. In our study, Inconel alloy 738 was manufactured by casting 50% and 100% of revert material and tested in HTGR conditions to examine the influence of helium coolant on the material’s properties. Tensile specimens were exposed at 900 °C for 1000 h in helium containing a specified amount of gaseous impurities. Scanning electron microscopy (SEM) with energy dispersive spectroscopy (EDS), in combination with X-ray diffraction (XRD) and nano-, microhardness methods, were used for material characterization after performing the tensile tests at room temperature. The presence of three types of layers was observed: a thin layer formed by aluminum and chromium oxides on the surface; non-uniform surface oxides Ti3O5 with inner (Al,Cr)2O3; and the inner fine-grained Inconel Cr-enriched phase (approx. 10–20 µm below the surface), which can act as a protective surface layer. Mechanical properties of both revert materials decreased after exposure to HTGR conditions but did not show a significant difference as a result of the content of the revert material. The increase of nano-hardness in line profiles throughout the specimen’s cross-section was observed locally at the surface oxides and in the precipitates and grain boundaries. After exposure, Rp0.2 values decreased by 20% and 17.7%, and Rm values by 12.3% and 20.8% in samples with 50 and 100% revert content, respectively. Furthermore, a decrease in microhardness values (HV0.1) was detected by 4.98% in longitude and 5.80% in cross-section for samples with 50% revert material and by 3.85% in longitude and 7.86% in cross-section for samples with 100% revert material. It can be concluded that both revert materials have similar corrosion resistance in HTGR conditions. The presented results complement the knowledge about the degradation of alloys in the coolant environment of advanced gas-cooled reactors.

High-Temperature Corrosion Behavior of Incoloy 800H Alloy in the Impure Helium Environment

The helium coolant in the primary circuit of the high-temperature gas-cooled reactor (HTGR) contains traces of impurities, which can induce the corrosion of superalloys when exposed to elevated temperatures. The superalloy damage caused by the corrosion could threaten the safe operation of the reactor. In this work, the corrosion behavior of a representative superalloy (chromium-rich iron base alloy Incoloy 800H) was investigated under the impure helium at different typical temperatures of HTGR. An experimental setup developed for studying the high-temperature corrosion of superalloys was used to investigate the chemical reactions and corrosion behaviors of Incoloy 800H. It was found that CO2 is an important oxygen source in the reaction with chromium, and CO is released as the product. In addition, the observation and computation of the critical temperature (TC) of the reaction between CO2 and carbon in the alloy show that TC is much lower than that (TA) of the microclimate reaction, which indicates that CO2 can protect the scale from destruction. Furthermore, the slight decarbonization of the alloy was found above TC. Also, a model developed by the thermodynamic analysis was proposed to explain the mechanism of slight decarbonization and predict the critical temperature when the CO2-C reaction occurs. This work presents a guideline for protecting the oxide scale of superalloys used in HTGR.

An Analysis of Water Characteristics of Shiddhirganj 210MW Steam Power Plant, Sirajganj 225MW Combined Cycle Power Plants and Rooppur 2400MW Nuclear Power Plant in Bangladesh: A Comparative Study of Power Plant Grade Water

Water quality has long been an important part of the operation of nuclear and thermal power plants. Water is used as a working and cooling fluid in power plants. The quality of source water to be used in the power plants after treatment should conform to the prescribed values of Physico-Chemical properties like pH, EC, TDS, Alkalinity, Hardness, presence of Chloride content, Silica, and Heavy metals as recommended by technical guidelines. To analyze the water characteristics, water samples were collected from various sources. Two river water samples and a groundwater sample (marked as PP1, PP2, and NPP site) were tested to determine the water properties. The Physico-chemical properties of the water samples were determined using various analytical tests. Heavy metals were also investigated using the Atomic Absorption Spectrophotometric (AAS) process. The Physico-chemical properties of investigated samples were so good as to the recommended values of the World Health Organization (WHO), United States of Public Health (USPH), and Bangladesh Environmental Department. But the values were very far from the recommended values for thermal power plants operation. So, it needs to treat to use in the boilers. Gravitation, Carbon filtration, Ion exchange method, and Reverse Osmosis (RO) are good ways to treat the water before use in power plants. The Padma River water is better than the other two samples when the Physico-chemical characteristics of the sample waters were analyzed to the prescribed levels for the boiler water of a thermal power plant and the secondary water circuit of a third-generation nuclear PWR reactor. The aim of this work is to explore the difference between the source water quality parameters values and those of the recommended values of technical guidelines. To minimize the water treatment cost, a new site for the installation of a power plant project in Bangladesh can be advocated depending on the availability of good quality water.

Online intelligent fault diagnosis of redundant sensors in PWR based on artificial neural network

Sensors in the primary circuit of a pressurized water reactor (PWR) are normally designed with redundant structures to improve system safety and reliability. However, reliability of the actual system is often lower than that obtained by theoretical calculation due to the inevitable occurrence of common mode fault (CMF), which is a dependent failure event that can cause multiple failures in redundant channels. CMF may increase the reliability deviation of the system by orders of magnitude and, hence, seriously affects the reliability of the system. To mitigate the CMF of redundant sensors in nuclear power plants, an artificial neural network (ANN) can serve as a data-driven analytic model to monitor sensor parameters, to identify any possible abnormal status of the sensors, and provide an early warning. In this study, by using the high-fidelity dataset obtained in a full-scope PWR simulator as training, validation, and test data, a relevant parameter-based ANN black-box model (RPANN) was established by employing the back-propagation (BP) learning algorithm, which was then defined as an analytic redundancy. Time series-based ANN checking models (TSANNs) were also established for each of the input and output parameters of the RPANN in order to identify its abnormal state based on historical data in the past. When combined with the existing hardware redundancy, the ANN-based analytic redundancy can serve as an online monitoring tool of the hardware status and an online diagnosis strategy for sensor faults. Furthermore, ANN-based analytic redundancy can replace faulty hardware sensors to analytically reconstruct the reading of the monitored sensor parameter without having to reduce the reactor output power or even shut down the reactor for emergency maintenance so that the on-site calibration frequency of hardware sensors in redundant channels can be effectively reduced. This is not only of vital importance in reducing operation and maintenance costs of existing PWR power plants but also plays an important role in building reactor operation schemes with rapid and frequent changes in power output in the future. Simultaneously, the diverse redundancy combining analytic software redundancy and physical hardware redundancy can effectively reduce the threat of CMF of hardware sensors on the operation safety of reactor systems.

Analysis of coasting-down hydrodynamic characteristics of sodium pump in the primary circuit of fast reactor

The variation of coasting-down hydrodynamic characteristics of the sodium pump in the primary circuit of fast reactor is related to the safe and stable operation of nuclear power plant. In order to explore the variation rule of performance parameters during the coasting-down process, the prototype of the sodium pump (Vertical double suction coaxial in and out submersible pump) in the primary circuit of fast reactor was taken as the research object in this paper. Based on the N-S equation and the RNG k-ε turbulence model, the Fluent software was used. The UDF function was written to carry out unsteady numerical calculation on the coasting-down process of sodium pump, and the variation rule of rotating speed, flow rate, head, torque and pressure with time in the coasting-down process was obtained. The results show that during the coasting-down process，the time for the model pump rotating speed to drop to half (296.5 r/min) is more than 15 s, and the time to drop to the minimum (112 r/min) is more than 74 s, which meets the nuclear safety standards. The flow rate ratio and the rotating speed ratio have the same change law, the head ratio and the torque ratio have the same change law, and the decline speed of the head ratio and torque ratio is greater than the rotating speed ratio and the flow rate ratio, which conforms to the similarity law of the pump. During the coasting-down process, the energy of the model pump pressure pulsation is mainly concentrated at middle and low frequencies. With the increase of the coasting-down time, the amplitude of pressure pulsation decreases gradually, and the uniformity of pressure distribution at circumferential direction decreases gradually. In addition, under the influence of pressure pulsation and the complexity of flow passage of the flow components, the head has a pulsating downward trend.

Study on Fault Characteristics of Layer Short Circuit of Dry-Type Air-Core Series Reactor

Aiming at the layer short circuit fault of dry-type air-core series reactor winding, in order to reveal the characteristics of electrical parameters when the fault occurs and improve the accuracy of online detection, this paper uses ANSYS Maxwell software to study the changes of electrical parameters when the layer short circuit fault occurs in dry-type air-core series reactor. The changes of equivalent resistance, equivalent reactance, and power factor are simulated and analyzed when the layer short circuit fault occurs in normal operation and different positions of winding. The simulation results show that with the different locations of layer short circuit fault points, the variation laws of electrical parameters of reactors are also different, in which the variation amplitude of power factor is obvious and the variation trend is consistent, which can provide important theoretical support for the online detection and protection of layer short circuit fault of dry-type air-core series reactors.

Synthesis of Submicron Boron Carbide by the Non-Vacuum Method with Indirect Supply of the Thermal Energy of a DC Arc Discharge

The results of experimental studies demonstrating the possibility of obtaining boron-carbide powder by treating a mixture of carbon and amorphous boron containing a small amount of boron oxide with DC arc discharge plasma are presented. A feature of the applied nonvacuum electric-arc method for the synthesis of boron carbide is its implementation in an open air environment due to achievement of the effect of screening the reaction zone with gaseous oxide and carbon dioxide. In this case, the initial reagents are heated to the required temperatures by a DC arc discharge plasma burning in the immediate vicinity of the initial-material mixture in the cavity of the graphite crucible. The resulting material contains particles of boron carbide with a size in the submicron range, which are characterized by the “shell–core” type structure. The scheme of the discharge circuit of a laboratory reactor with horizontal arrangement of electrodes and an indirect supply of thermal energy to the initial reagents described in the paper makes it possible to significantly increase the content of the boron-carbide phase and reduce the content of the graphite impurity phase.