• Nuclear Engineering and Technology
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• v.53 no.11
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• pp.3653-3664
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• 2021

In PWR three-dimensional transient coupling calculation software CORCA-K, the nodal Green's function method and diagonal implicit Runge Kutta method are used to solve the spatiotemporal neutron dynamic diffusion equation, and the single-phase closed channel model and one-dimensional cylindrical heat conduction transient model are used to calculate the coolant temperature and fuel temperature. The LMW, NEACRP and PWR MOX/UO₂ benchmarks and FangJiaShan (FJS) nuclear power plant (NPP) transient control rod move cases are used to verify the CORCA-K. The effects of burnup, fuel effective temperature and ejection rate on the control rod ejection process of PWR are analyzed. The conclusions are as follows: (1) core relative power and fuel Doppler temperature are in good agreement with the results of benchmark and ADPRES, and the deviation between with the reference results is within 3.0% in LMW and NEACRP benchmarks; 2) the variation trend of FJS NPP core transient parameters is consistent with the results of SMART and ADPRES. And the core relative power is in better agreement with the SMART when weighting coefficient is 0.7. Compared with SMART, the maximum deviation is -5.08% in the rod ejection condition and while -5.09% in the control rod complex movement condition.

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.97-109
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• 2022

Multi-dimensional coupling analysis is a research hot spot in nuclear reactor thermal hydraulic study and both the full-scale system transient response and local key three-dimensional thermal hydraulic phenomenon could be obtained simultaneously, which can achieve the balance between efficiency and accuracy in the numerical simulation of nuclear reactor. A one-dimensional to three-dimensional (1D-3D) coupling platform for the nuclear reactor multi-dimensional analysis is developed by XJTU-NuTheL (Nuclear Thermal-hydraulic Laboratory at Xi'an Jiaotong University) based on the CFD code Fluent and system code RELAP5 through the Dynamic Link Library (DLL) technology and Fluent user-defined functions (UDF). In this paper, the International Standard Problem (ISP) No. 43 is selected as the benchmark and the rapid boron dilution transient in the nuclear reactor is studied with the coupling code. The code validation is conducted first and the numerical simulation results show good agreement with the experimental data. The three-dimensional flow and temperature fields in the downcomer are analyzed in detail during the transient scenarios. The strong reverse flow is observed beneath the inlet cold leg, causing the de-borated water slug to mainly diffuse in the circumferential direction. The deviations between the experimental data and the transients predicted by the coupling code are also discussed.

• Nuclear Engineering and Technology
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• v.55 no.9
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• pp.3213-3228
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• 2023

As an important part of the digital reactor, the pin-by-pin wise fine coupling calculation is a research hotspot in the field of nuclear engineering in recent years. It provides more precise and realistic simulation results for reactor design, operation and safety evaluation. CORCA-K a nodal code is redeveloped as a robust pin-by-pin wise neutronics and thermal-hydraulic coupled calculation code for pressurized water reactor (PWR) core. The nodal green's function method (NGFM) is used to solve the three-dimensional space-time neutron dynamics equation, and the single-phase single channel model and one-dimensional heat conduction model are used to solve the fluid field and fuel temperature field. The mesh scale of reactor core simulation is raised from the nodal-wise to the pin-wise. It is verified by two benchmarks: NEACRP 3D PWR and PWR MOX/UO₂. The results show that: 1) the pin-by-pin wise coupling calculation system has good accuracy and can accurately simulate the key parameters in steady-state and transient coupling conditions, which is in good agreement with the reference results; 2) Compared with the nodal-wise coupling calculation, the pin-by-pin wise coupling calculation improves the fuel peak temperature, the range of power distribution is expanded, and the lower limit is reduced more.

• Nuclear Engineering and Technology
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• v.49 no.6
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• pp.1339-1345
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• 2017

In the framework of the EU funded project NURESAFE, the subchannel code CTF and the neutronics code DYN3D were integrated and coupled on the NURESIM platform. The developments achieved during this 3-year project include assembly-level and pin-by-pin multiphysics thermal hydraulics/neutron kinetics coupling. In order to test this coupling, a PWR rod ejection transient was simulated on a MOX/UOX minicore. The transient is simulated using two different models of the minicore. In the first simulation, both codes model the core with an assembly-wise resolution. In the second simulation, a pin-by-pin fuel-centered model is used in CTF for the central assembly, and a pin power reconstruction method is applied in DYN3D. The analysis shows the influence of the different models on global parameters, such as the power and the average fuel temperature, but also on local parameters such as the maximum fuel temperature.

• Nuclear Engineering and Technology
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• v.53 no.3
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• pp.763-775
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• 2021

The thermal hydraulic performances of Steam Generator (SG) under both steady and transient operation conditions are of great importance for the safety and economy in nuclear power plants. In this paper, based on our self-developed SG thermal hydraulic analysis code STAF (Steam-generator Thermalhydraulic Analysis code based on Fluent), an improved new version STAF-CT (fully Coupling and Transient) is developed and introduced. Compared with original STAF, the new version code STAF-CT has two main functional improvements including "Transient" and "Fully Three Dimensional Coupling" features. In STAF-CT, a three dimensional energy transferring module is established which can achieve energy exchange computing function at the corresponding position between two sides of SG. The STAF-CT is validated against the international benchmark experiment data and the results show great agreement. Then the U-shaped SG in AP1000 nuclear power plant is modeled and simulated using STAF-CT. The results show that three dimensional flow fields in the primary side make significant effect on the energy source distribution between two sides. The development of code STAF-CT in this paper can provide an effective method for further SG high fidelity research in the nuclear reactor system.

• Journal of electromagnetic engineering and science
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• v.16 no.4
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• pp.219-224
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• 2016

In this paper, coupled-mode theory (CMT) is used to obtain a transient solution analytically for a wireless power transfer system (WPTS) when unit energy is applied to one of two resonators. The solutions are compared with those obtained using equivalent circuit-based analysis. The time-domain CMT is accurate only when resonant coils are weakly coupled and have large quality factors, and the reason for this inaccuracy is outlined. Even though the time-domain CMT solution does not describe the WPTS behavior precisely, it is accurate enough to allow for an understanding of the mechanism of energy exchange between two resonators qualitatively. Based on the time-domain CMT solution, the critical coupling coefficient is derived and a criterion is suggested for distinguishing inductive coupling and magnetic resonance coupling of the WPTS.

• Journal of Advanced Marine Engineering and Technology
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• v.31 no.8
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• pp.998-1004
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• 2007

Connecting a synchronous generator to a power system is a dynamic process, requiring the coordinated operation of many components and systems. The goal is to connect the oncoming generator to the system smoothly i.e without causing any significant bumps, surges, or power swings, by closing the ACB when the oncoming generator matches the power system in voltage magnitude, phase angle, and frequency. If oncoming generator voltage is not matched to the power system voltage, reactive power will flow either into or out of the system at the instant of ACB closure. If this voltage difference is too great, the reactive power flow may result in high transient stresses that could damage the windings of the generator. Also, if oncoming generator frequency is not matched to the power system frequency, transient power will flow between generator and power system. If the frequency difference is too great, the transient power flow is reflected into the prime mover shaft, and this may result in excessive shaft or coupling stress. This paper tries to prove the necessity of correct synchronization for ship generators through a transient phenomenon analysis.

• Proceedings of the KIEE Conference
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• 1995.07a
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• pp.24-26
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• 1995

In this paper, we investigated transient fault currents in a magnetic coupling High-Tc superconducting current limiter(HCL). It has an important effect on the reliability and stability of the power system. In order to analyze transient fault characteristics of HCL, we fabricated a magnetic coupling HCL and tested it in different fault conditions. An important parameter of design and manufacture which makes HCL inherently reliable is reduction of inrush fault currents. Without inrush fault currents, the currents flowing under such conditions can be limited to a desired-value within one cycle. Inrush fault current depends on saturation, normal spot propagation velocity, turns ratio and the fault angle.

• Steel and Composite Structures
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• v.25 no.2
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• pp.187-196
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• 2017

In this work, transient thermo-piezo-elastic responses of an infinite functionally graded piezoelectric (FGPE) plate whose upper surface suffers time-dependent thermal shock are investigated in the context of different thermo-piezo-elastic theories. The thermal and mechanical properties of functionally graded piezoelectric plate under consideration are expressed as power functions of plate thickness variable. The solution of problem is obtained by solving the corresponding finite element governing equations in time domain directly. Transient thermo-piezo-elastic responses of the FGPE plate, including temperature, stress, displacement, electric intensity and electric potential are presented graphically and analyzed carefully to show multi-field coupling behaviors between them. In addition, the effects of functionally graded parameters on transient thermo-piezo-elastic responses are also investigated to provide a theoretical basis for the application of the FGPE materials.

• Nuclear Engineering and Technology
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• v.54 no.10
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• pp.3732-3744
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• 2022

To investigate the effects of ocean conditions on the thermal stress and deformation caused by thermal stratification of a pressurizer surge line in a floating nuclear power plant (FNPP), the finite element simulation platform ANSYS Workbench is utilized to conduct the fluid-solid-thermal coupling transient analysis of the surge line under normal "wave-out" condition (no motion) and under ocean conditions (rolling and pitching), generating the transient response characteristics of temperature distribution, thermal stress and thermal deformation inside the surge line. By comparing the calculated results for the three motion conditions, it is found that ocean conditions can significantly improve the thermal stratification phenomenon within the surge line, but may also result in periodic oscillations in the temperature, thermal stress, and thermal deformation of the surge line. Parts of the surge line that are more susceptible to thermal fatigue damage or failure are determined. According to calculation results, the improvements are recommended for pipeline structure to reduce the effects of thermal oscillation caused by ocean conditions. The analysis method used in this study is beneficial for designing and optimizing the pipeline structure of a floating nuclear power plant, as well as for increasing its safety.