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[Introduction] Under the scouring impact of secondary side fluid, the steam generator tube is prone to flow-induced vibration. One of the main mechanisms leading to tube vibration is the random turbulent force. When the fluctuating pressure frequency of the fluid is close to the natural frequency of the tube, structural resonance will be caused and long-term vibration will lead to the failure of the tube. Therefore, it is necessary to study the dynamic response characteristics of steam generator tubes under fluid excitation. [Method] In this paper, a flow-induced vibration test device for the bend zone of tube bundle with a pitch-diameter ratio of 1.47 was designed. The secondary side fluid condition was simulated by air-water two-phase flow. The fluid fluctuating pressure and the vibration acceleration of tube bundle with a void fraction of 0.7～0.98 and a flow velocity between tubes of 5～13 m/s was measured. [Result] The results show that: the main frequency of fluctuating pressure is close to the natural frequency of tube at low flow velocity, which is easy to causes resonance; when resonance occurs, the amplitude of the tube bundle increases; with the increase of velocity between tubes, the fluctuating pressure on tube bundle increases correspondingly. With the increase of the void fraction, the fluctuation pressure first increases and then decreases. When calculating the main frequency of fluctuating pressure under the condition of air-water two-phase flow, the coefficient in the empirical formula can be adjusted appropriately. [Conclusion] This test simulated the working conditions of the secondary side fluid operation in the bend zone of steam generator tube bundle and considered the geometric similarity with the prototype and the similarity of support and constraint in the design of the model. This test was closer to the actual situation than the previous experimental research and can provide design references for engineering applications.

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14C has become the nuclide that contributes the most to the annual effective dose to the surrounding public among the radioactive effluents during normal operation of nuclear power plants, but for a long time, domestic nuclear power units have not been equipped with airborne 14C treatment facilities, it is important and urgent to seek technical methods to reduce 14C emissions based on potential and actual needs under the requirements of the existing emission limits. Domestic nuclear power units under construction and in operation are mainly pressurized water reactors (more than 95% of the total), and the airborne 14C in this type of reactors mainly exists in the form of alkane compounds. Therefore, in this study, 14CH4, which accounts for the largest proportion of 14C alkane compounds and has the most stable chemical properties, was taken as the treatment target, and low-temperature plasma technology was introduced to investigate its discharge behavior and CH4 treatment performance. The results show that under the optimal conditions of normal temperature and pressure, output voltage of 17.89 kV, and gas flow rate of 0.83 cm/s, the plasma’s CH4 treatment efficiency can reach 99.37%, and the CO2 selectivity can reach 46.99%. The plasma’s CH4 treatment performance can be improved by increasing the output voltage, reaction temperature, and decreasing the gas flow rate. However, since increasing the reaction temperature would bring problems about energy consumption, safety, and equipment complexity, increasing the reaction temperature is not the first choice. In addition to CO2, there are more than thirty kinds of by-products produced in the process of plasma treatment of CH4, which are dominated by organic substances. The kinetics process of plasma treatment of CH4 is in accordance with the quasi-primary reaction kinetics model, and the corresponding rate constants are 1.104 8 m3/(kW·h). The above results indicate that plasma technology has a broad development prospect in the field of airborne 14C treatment and monitoring, especially in the treatment of 14C alkane compounds. The focus of subsequent research should be focused on optimizing the reaction pathway, lowering the reaction barriers, further increasing the proportion of CH4 directed oxidation to CO2, and significantly reducing the formation of by-products.

Copyright of Atomic Energy Science & Technology is the property of Editorial Board of Atomic Energy Science & Technology and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Copyright of Atomic Energy Science & Technology is the property of Editorial Board of Atomic Energy Science & Technology and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

321 stainless steel (SS) is one kind of austenitic stainless steel, in which Ti is added as stabilizing element. Due to its excellent corrosion resistance and comprehensive mechanical properties under high temperature, 321SS is an important construction material in light water reactor (LWR). It is prone to cold work due to various factors in the whole cycle, which changes its working performance, especially corrosion resistance and stress corrosion cracking (SCC) property, while corrosion and SCC are main failure forms of 321SS in LWR. The mainstream view is that the transformation of grain boundary type, the increase of slip lines, the formation of voids and the phase change from austenite to martensite caused by cold work will increase the probability of failure. However, the dominant factors and main mechanisms are still controversial. The purpose of this study is to investigate the effect of cold work on corrosion and SCC of 321SS and preliminarily explain the mechanism. Samples with different work rate and martensite fraction were acquired through different extent of cold work and hot work. The as-received specimen was remarked as SA, the cold worked specimen was remarked as CW20 and the hot worked HW20. The corrosion behavior was studied by high temperature electrochemical tests. The electrochemical impedance spectroscopy (EIS) of SA and CW20 was measured in a simulated pressurized water reactor (PWR) first circuit environment, and HW20 was used as a comparison. The SCC performance of specimens was tested by stress corrosion test under slow strain rate tensile (SSRT) loading mode. After tensile tests, micro cracks initiated on the surface were counted and the average crack length was calculated to evaluate SCC sensitivity. The microscopic characteristics of the samples were analyzed by X-ray diffraction (XRD), scanning electron microscopy (SEM) and energy dispersive spectrometer (EDS). XRD analysis shows that barely no martensite exists in SA, the martensite volume fraction in CW20 is 22% and 11% in HW20. In EIS tests, the charge transfer resistance rises both in CW20 and HW20 compared with SA. SA has the highest film resistance, HW20 the second, and CW20 has the lowest. After SCC test, the most cracks initiated on the surface of HW, and the least on SA. The conclusions can be drawn as following: The cold work causes the transformation from austenite to martensite in the matrix, and the high temperature inhibits this process. With the increase of strain degree, the charge transfer resistance increases, and the film resistance decreases with the increase of martensite content. A Cr-depleted zone is caused by martensite at the interface between the oxidize layer and the matrix, thus leads to the reduction in protective effect of the passivation film. With the increase of martensite content, the film resistance decreases, and the corrosion resistance of the passivation film decreases. Micro cracks tend to initiate more easily after either cold work or hot work compared with primary solution annealed 321SS. The martensite phase distributed in the matrix is oxidized preferentially and suppresses the corrosion of austenitic phase. Under the conditions of this research, strain-induced martensite inhibits SCC crack initiation.

BackgroundIn thermal pipelines of nuclear power systems, thermal stratification is a common phenomenon that can cause stress concentration and deformation of pipeline structures, thereby leading to safety hazards. A stagnant branch pipe is connected to the main coolant pipe, and a large temperature difference exists between the fluid in the pipe and the coolant in the main pipe of the primary circuit. Due to factors such as turbulent flow penetration and valve leakage, thermal stratification is prone to occur in the branch pipe.PurposeThis study aims to analyze the temperature change characteristics and flow characteristics of thermal stratification in stagnant branch pipes and provide a theoretical basis for subsequent experimental research and stress analysis.MethodsFirstly, a stagnant branch pipe model was established, and numerical simulation of thermal stratification phenomenon in stagnant branch pipes was conducted using FLUENT 2022 to analyze the temperature variation characteristics of the pipe wall and the distribution characteristics of the flow field inside the pipe. Then, the SST k-ω model was used to perform three-dimensional numerical simulation of the thermal stratification of stagnant branch pipes, with a leakage flow rate of 0.062 kg·s-1, leakage temperature of 488.15 K, and leakage pressure of 6 MPa.ResultsThermal stratification is prone to occur in horizontal pipe sections. Without insulation measures and a large pipe diameter, thermal stratification can be exacerbated, while the curved section can effectively reduce the temperature difference of the cross-section. A backflow phenomenon occurs in the horizontal section of the stagnant branch pipe, while the structure of the large and small end pipe sections causes secondary backflow in the flow field inside the pipe. The backflow phenomenon is not conducive to the mixing of cold and hot fluids in the pipe; consequently, the influence time of thermal stratification is longer.ConclusionsA significant difference in the thermal stratification phenomenon exists between the stagnant branch pipe and equal cross-section pipes.

With the increase of fuel cycle period of the new generation pressurized water reactor, the influence of corrosion product deposition in the core is becoming more and more serious. The cladding surface temperature and species concentration changes caused by deposits (also called Chalk River Unidentified Deposit (CRUD)) increase the risks of core power shift induced by boron accumulation, radiation dose increase and cladding corrosion. In order to predict the risks induced by the CRUD on the cladding in the pressurized water reactor, a multi-physics process model was established and optimized for the heat transfer, mass transfer, fluid flow and chemical process of the cladding surface with CRUD. The heat transfer parameters and mass transfer parameters on the cladding surface were carried out under different operating conditions, hydrochemical conditions and CRUD structures. Combined with the relationship of multi-physics processes, the effects of the heat flux, the coolant temperature, the boron and lithium concentrations in coolant and the structural parameters of CRUD on the performance of the cladding surface were discussed, and the essential reasons of CRUD induced boron accumulation were discussed. The risk equation to evaluate the CRUD induced boron accumulation under different operating parameters, hydrochemical parameters and CRUD structural parameters was proposed. The results show that when boiling occurs in CRUD, the temperature distribution can be obtained more accurately by considering the volatilization of boric acid. The presence of CRUD leads to a dramatic increase of the outer surface temperature of the cladding. The increase degree of cladding temperature increases with the CRUD thickness, the CRUD porosity and the boron concentration, and decreases with the increase of heat flux, coolant temperature and CRUD stack density. The thermal driving force from cladding to coolant increases when the CRUD thickness is less than 50 μm, the heat flux is 0.4-1 MW/m2 and the coolant temperature is less than 300 ℃. Beyond the above ranges, the presence of CRUD reduces the heat transfer from cladding to coolant. Under the classical pressurized water reactor condition, the boron accumulation in the void structure of 40 μm CRUD mainly derives from boric acid (0.057 4 g/m2) and the adsorbed boron (4.61×10-3 g/m2). The deposited boron in the void structure of CRUD in the form of Li2B4O7 is 8.34×10-5 g/m2, while LiBO2 can not deposit in 40 μm CRUD. At 1 MW/m2, a power shift of 1.5% occurs when 1/8 of the core attaching to 80 μm thick cladding CRUD. The effects on the CRUD induced boron accumulation risk from large to small are as follows: CRUD thickness, coolant temperature, pore diameter, coolant boron concentration, heat flux, porosity, while the coolant lithium concentration and chimney density of CRUD have little effect.

BackgroundSilicon carbide (SiC) composite claddings are candidate solutions for accident resistant fuel claddings in light water reactors.PurposeThis study aims to estimate the failure probability of a double-layer structured SiC cladding under a loss-of-coolant accident (LOCA).MethodsBased on a failure probability calculation method for SiC composite cladding, a quasi-steady state method was used to simulate and calculate the SiC composite cladding failure probability under transient conditions. Sensitivity analysis of the two characteristic parameters of Weibull distribution was performed by analyzing the proportion of various stresses under accident conditions. The effects of different burn-up conditions on the failure probability were investigated, and the failure probability of the cladding under different layer thickness ratios was simulated.Results & ConclusionsSimulation results indicate that the transient failure probability of SiC composite claddings is significantly affected by changes in the proportion of the composite layer and Weibull parameter, as well as the occurrence of LOCAs under different burn-up conditions. This study makes contribution to the development and design of accident resistant fuel claddings, providing reference for further investigations on the failure probability of SiC composite claddings.

BackgroundUnder the condition of sub-cooled nucleate boiling (SNB), corrosion products in primary coolant of nuclear reactor will deposit on the outer surface of fuel cladding, which is commonly called fuel crud. Previous literature shown, zinc injection in primary coolant is an important method to inhibit the fuel crud deposition on the fuel cladding surface.PurposeThis study aims to investigate the influence of zinc concentration on the behavior of fuel crud deposition, and eventually provide guidance for zinc injection in primary coolant of nuclear power plant.MethodsThe fuel crud deposition tests of domestic zirconium alloy fuel cladding in different zinc concentrations were carried out by using a self-made fuel crud deposition device. Tubular crud deposition test specimen with built-in heating unit was designed and prepared for simulation study. After the tests, stereo microscope (SM) and scanning electron microscope (SEM) were employed to observe the macro and micro morphology of fuel crud whilst the composition of of fuel crud was observed and analyzed by the energy dispersive spectroscopy (EDS) with SEM, and X-ray photoelectron spectroscopy (XPS) was used to analyze the contents of Zn and B elements in the crud phase and inside the crud.ResultsObservation results show that the chimney-like crud formed on the fuel cladding surface becomes less obvious with increasing the zinc concentration in the coolant and the crud surface becomes flatter. Simutabeously, the crud thickness, the ratio of Ni/Fe and the boron precipitation mass within the crud are decreasing with increase of the zinc concentration. When the zinc concentration increases to 100 μg‧L-1, new Zn-containing phases precipitate within the crud.ConclusionsWithin the zinc concentration of 0~100 μg‧L-1, zinc injection in primary coolant of reactor can significantly inhibit the crud deposition on the fuel cladding surface.

BackgroundCompared with the traditional pressurized water reactor (PWR), the core design of large advanced PWR CAP1400 has significant changes, such as the increase in the number of fuel assemblies, the increase in reactor power, the increase in the average temperature of core coolant, etc. These changes have an important impact on the results of rod ejection accident, and then affect the safety of reactor.PurposeThis study aims to verify the safety of large advanced PWR under rod ejection accident condition and the influence of key input parameters on accident analysis results.MethodsBased on the neutron dynamics software TWINKLE and fuel performance analysis program FACTRAN, the typical four types of operating conditions, including the beginning of life the hot full power and the hot zero power, the end of life the hot full power and the hot zero power, were selected to carry out the simulation calculation of the control rod ejection accident analysis for large advanced PWR, and the sensitivity analysis of key input parameters of rod ejection accident conditions was performed by using the direct numerical perturbation method.Results & ConclusionsSimulation results show that the power peak is the most sensitive to the worth of rod ejection, but less sensitive to shutdown reactivity. The consequences of the control rod ejection accident designed for CAP1400 can meet the requirements of acceptance criteria and the reactor is in the safe and controllable state.

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BackgroundFor a pressurized water reactor (PWR), under the synthesis effects of primary water chemistry, reactor operation mode, system facilities replacement, fuel design and management, the corrosion products in primary coolant will deposit on the outer surface of fuel cladding, which is commonly called fuel crud. Fuel crud formed on the cladding surface may result in crud induced localized corrosion (CILC), crud induced power shift (CIPS) and raise the primary source term. The deposition behavior of fuel crud is affected by both the corrosion products concentration and the core boiling rate.PurposeThis study aims to investigate the influences of core power variation on the corrosion products concentration and the core boiling rate, and eventually evaluate the fuel crud deposition states under different core powers.MethodsThe core crud mass and the maximum crud thickness on fuel cladding surface were estimated by the CAMPSIS, an independent crud behavior analysis software developed by China Nuclear Power Technology Research Institute Co., Ltd (CNPRI). Influence of core power variation and the combined effect of core boiling rate and corrosion product concentration were analyzed.ResultsComputational results that, within the range of 75%⁓100% full power (FP), the lower the core power is, the smaller the core crud mass will be, but the maximum crud thickness will increase. When the core power decreases to 50% FP, the maximum boiling rate is close to zero, and both of the core crud mass and the maximum crud thickness are minimum.ConclusionsThe core power reduction of a nuclear reactor may lead to the increase of local fuel crud thickness, and adversely affect the heat transfer between the fuel cladding and the coolant. It is suggested that the influence of core power reduction on fuel crud deposition should be considered if a nuclear reactor tends to operate at a lower core power.

To investigate the exergy loss of fuel rod during converting nuclear energy into thermal energy, the partial differential equations of steady heat transfer of pressurized water reactor fuel rods and the first and second laws of thermodynamics were used. The exergy analysis method was innovatively combined with the numerical calculation of temperature field. The numerical calculation program was compiled to simulate the fuel rods and heat transfer channels and analyze the temperature distribution, the exergy loss distribution and the energy utilization efficiency during converting nuclear energy into heat energy and during coolant heat transfer. The results show that the fuel rod exergy loss is increased with latter decreasing in the axial direction and increased in the radial direction with the maximum exergy loss coefficient of 0.207 at the edge of fuel core. The exergy loss in the convective heat transfer process is mainly related to the heat transfer temperature difference and increased with latter decreasing along the thermal channel with the total exergy loss coefficient of 0.304. [ABSTRACT FROM AUTHOR]

In order to reduce the false alarm or omission phenomenon existing in the loose parts monitoring system, reduce the possibility of that the loose parts bring great risks to the safe operation of the nuclear reactors under the impact of high speed water flow, the collision of the loose parts and the inner wall of the equipment was simulated by the collision of the six different quality steel balls and the steel plate, and then the corresponding impact signal was obtained by the acceleration sensor. The wavelet transform was used to decompose the original shock signal in the multi-layer frequency domain, and the energy ratio was used as the mass estimation feature vector. The selected eigenvector was input as a multi-layer back propagation( BP) neural network, and a suitable steel ball mass estimation model was established after training with different samples. The results showed that the estimated errors of different quality steel balls were 0.04%, 1.14%, 6.76%, 0.01%, 0.05% and 0.07%, respectively, the errors were all within 8%. By using the scientific combination of wavelet packet transform and BP neural network, the minimization of quality estimation error is realized under the control of system cost. [ABSTRACT FROM AUTHOR]

To improve the efficiency and accuracy of numerical simulation of single-phase flow and heat transfer in Pressurized Water Reactor (PWR) rod bundle, simulations were performed in two computational models of four-subchannel without spacer and 5 x 5 rod bundle with a mixing vane spacer under different working conditions. The different turbulence models, near-wall treatments and near-wall mesh generation strategies were comprehensively evaluated from two aspects of heat transfer ( Nusseh number) and fluid flow (Secondary flow, turbulence intensity) by using high quality wall-function mesh and low-Reynolds-number mesh. Analysis shows that the numerical results of Nusselt number were seriously affected by both of the turbulence model and near-wall treatment, while secondary flow and turbulence intensity were mainly influenced by turbulence model. The Reynolds Stress Model (RSM) could predict the secondary flow field in bare rod bundle better than eddy viscosity models, and there was no significant difference for calculating the secondary flow field downstream of the mixing vane spacer grid. Among all turbulence models, only the SST k - ω model behaved well even combined with different type of near-wall meshes. Using the SST model with wall-function mesh for engineering computation can provide both the efficiency and accuracy of simulation, and the SST model combined with low-Reynolds-number mesh is recommended for numerical research. [ABSTRACT FROM AUTHOR]