Abstract: Zirconium alloy components used in nuclear reactors maybe occur to fail due to hydrogen induced delayed cracking (HIDC) during service. Whether the microdefects on the surface of the components will cause HIDC is worth studying. In this paper, samples with microcrack defects on the surface of zircon...
Abstract: The research, development and application of health monitoring and fault diagnosis technology for reactor critical equipment are important guarantee for the safe, reliable and economic operation of nuclear reactors, which also contribute to the improvement of operation and maintenance of nuclear rea...
Abstract: The assembly code based on the Monte Carlo method can handle problems with complicated geometry and avoid the resonance self-shielding calculation in the deterministic assembly code. However, the Monte Carlo assembly code has certain difficulties when generating diffusion coefficient and assembly di...
Abstract: In order to improve the modeling and calculation accuracy of the traditional Consistent Adjoint Driven Importance Sampling (CADIS) method that relies on structural-mesh finite-difference discrete ordinate (SN) code to determine the importance distribution of particles to further enhance its capabili...
Abstract: In order to evaluate the radiation damage of fast reactor structural materials, a set of evaluation methods for radiation damage of fast reactor structural materials is proposed in this paper. According to the characteristics of the fast reactor energy spectrum, the irradiation scheme of the neutron...
Abstract: The generalized equivalence theory based CMFD (GET-CMFD) has great convergence behavior and has been successfully applied to the axial NEM based 2D/1D coupling method. However, when applying the high precision SN as the axial solver, the 2D/1D coupling method faces the convergence problem. To solve ...
Abstract: In the calculation of neutron transport problems with spatially localized source or largely void region, the discrete ordinate method (SN) suffers from ray effects, which cause computational inaccuracies. The first collision source method is often used to alleviate the ray effect to improve the reli...
Abstract: Based on the system code RELAP5 and the sub-channel code CTF, a multi-scale coupling domain decomposition approach for pool pulsed reactor with natural circulation and a grid mapping method that meets the precise conversion between different scale parameters are proposed in this paper. The correctne...
Abstract: Subcooled boiling is widely used in cooling applications with high heat fluxes represented by International Thermonuclear Experimental Reactor (ITER). In this paper, predictions on the critical heat flux (CHF) of subcooled water boiling under high heat flux and swirl flow conditions are focused and ...
Abstract: The coolant leakage and bypass flow channel in PWR is mostly a narrow gap. The size of narrow gap is sensitive to be easily affected by system pressure, differential pressure, temperature, vibration and other factors. Small size changes will cause significant changes in resistance coefficient, resul...
Abstract: In the helical coiled tube steam generator of liquid metal fast reactor, the temperature difference between inlet and outlet of the primary side increases significantly, and the outlet steam superheat degree of the secondary side also increases significantly. This is a common problem and brings new ...
Abstract: In the severe accident of lead-based reactor, the fuel particles may lead to core blockage during the migration process in the reactor, and there is a risk of re-criticality. In this study, in order to obtain the flow and migration characteristics of metal particles entrained in liquid lead-bismuth ...
Abstract: In order to study the oxidation corrosion characteristics of horizontal lead-bismuth reactor core, a liquid lead-bismuth oxidation corrosion model was established in this study. Based on computational fluid dynamics method, the self-defined source term method of transport equation was used to realiz...
Abstract: In this paper, a physical model of 90Sr radioisotope thermoelectric generator was designed, including the heat source, thermoelectric conversion module and heat dissipation module. On this basis, the finite element analysis of generator was carried out by using the COMSOL, and the temperature distri...
Abstract: Accurate flow regime identification of the boiling flow is of great significance for using closure correlations in thermal-hydraulic system codes. The flow regime identification is achieved by the unsupervised machine learning (ML) approach, and the current work integrates the data-driven ML model w...
Abstract: The TRISO microspheres were prepared by fluidized-bed chemical vapor deposition (FBCVD) process with ZrO2 microspheres as simulated cores. The microstructures of the coating layers of the microspheres were inspected by SEM and TEM, and the elastic modulus as well as Vickers hardness of the coating l...
Abstract: At present, there is no reference standard for hoop tensile testing of small-diameter tubes and there is also no uniform research method for hoop tensile testing of zirconium alloy cladding tubes in the worldwide research field. As to home-made Zr-Sn-Nb alloy cladding tubes, there is no data in hoop...
Abstract: In this paper, the seismic analysis and seismic test of a nuclear vortex process fan are carried out. Based on the test, the seismic analysis method of this kind of fan is obtained. Firstly, the finite element model of the fan is established, and the main modal frequencies of the fan are obtained th...
Abstract: In this paper, the impact of magnetic saturation on the electromagnetic parameters and magnetic circuit parameters of the control rod drive mechanism (CRDM) is studied by combining theoretical and finite element analysis (FEA). The method of obtaining magnetic circuit parameters is derived by establ...
Abstract: The application of the heat pipe reactor (HPR) urgently requires unmanned autonomous operation technology. Applying autonomous operation technology to HPR can realize state sensing, trend prediction and strategy optimization, which can effectively avoid human errors, improve the technical performanc...
Abstract: To study the influence of parameter uncertainties of helical coil once-through tube steam generator (HCOTSG) on the operation of small pressurized water reactor, based on the establishment of the core power control system, the feed water control system of steam generator and the pressure and water l...
Abstract: In applying the design of secondary-side passive residual heat removal system (PRS) of Generation 3 PWR to Generation 2+PWR, there are limitations due to the long distance between the cooling water tank of the PRS with the shell of the PWR, which complicates the arrangement of the relatively long st...
Abstract: In the power test phase of China Experimental Fast Reactor, the power reactivity measurement test was carried out. The results showed that there was a large deviation between the measured values and the theoretical calculation values, and the calculation method of power reactivity needed to be impro...
Abstract: In order to solve the assembly problem of insulation materials for steam generator of high temperature gas cooled reactor demonstration project (HTR-PM) and ensure that the insulation materials meets meet the design requirements of filling density and pressure, the assembly process test was carried ...
Abstract: Based on the needs of passive heat removal system using medium-temperature heat pipes (MTHPs) and heat pipe heat exchangers (HPHEs) in the high temperature circuit of large nuclear power plant and the novel nuclear reactor, and promoting their applications in industrial heat transfer and energy cons...
Abstract: The fuel cladding tubes damage often occurs during the operation of the reactor. When a fuel assembly is damaged, the power plant usually performs a sipping test on each fuel assembly using an online sipping device during the re-fueling process, and determines whether the tested assembly is damaged....
Abstract: In the logic function verification test (T2 test) of the Diversity Actuation System (DAS), the verification of the timer in the logical link is often carried out by waiting, making the test last for more than 2 hours, causing human error in the waiting process. In order to reduce T2 test time and hu...
Abstract: The operating experience feedback can significantly improve the safety level and operation level of nuclear power plants. China’s nuclear industry regulatory agencies and various operating units have established different levels of experience feedback systems that are operating well, collecting larg...
Abstract: In order to further explore the basic law that the environmental conditions in the nuclear power plant containment affect the leakage rate measurement, a data analysis program for measuring the containment leakage rate by the pressure drop method was developed in this investigation. The measurement ...
Abstract: The control of radioactive source of the activated corrosion products during the shutdown of PWRs is one of the most effective means to reduce the CRE. The effect of coolant radiolysis should be taken into account. In this paper, a coolant radiolysis kinetic model was developed based on water radiol...
Abstract: Self-powered neutron detector (SPND) is an important nuclear sensing device in the core, whose health status affects the safe operation of the reactor directly. Considering the measurement correlation between SPNDs at different positions in the reactor, a twin model-based anomaly detection method fo...
Abstract: Ammonia is applied in the coolant to eliminate the O2 and H2O2 in the PWR, thus mitigating the corrosion of structural materials. The present work studied the γ-radiolysis of ammonia solution under different conditions including ammonia concentration, absorbed dose, absorbed dose rate, gas-liquid vo...
Abstract: The complex working environment of the fuel elements in the reactor may cause changes in the performance of fuel elements and deviations from their initial geometric state, which can affect flow and heat transfer characteristics and threaten the safety of the reactor core. In this paper, the ANSYS W...
Abstract: Refined physical-thermal coupling calculations of reactors can simulate the core behavior more accurately. However, existing analysis programs adopt different discrete formats and mesh divisions when calculating different physical fields, resulting in a complex mesh mapping relationship for the tran...
Abstract: The core outlet temperature measurement is of great significance for mastering the reactor operation status. This paper analyzed the characterization of the core outlet temperature measurement by computational fluid dynamics (CFD) method. By simulating the structure of the fuel assembly and instrume...
Abstract: The design of reactor primary loop support should meet the vibration reduction and anti-shock requirements. Accordingly, mechanical models with different fidelity should be established. This paper first adopted correlation analysis to quantify the compatibility of static and dynamical characters of ...
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