• Nuclear Engineering and Technology
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• 제50권5호
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• pp.654-664
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• 2018

Self-pressurization analysis of the natural circulation integral nuclear reactor through a new dynamic model is studied. Unlike conventional pressurized water reactors, this reactor type controls the system pressure using saturated coolant water in the steam dome at the top of the pressure vessel. Self-pressurization model is developed based on conservation of mass, volume, and energy by predicting the condensation that occurs in the steam dome and the flashing inside the chimney using the partial differential equation. A simple but functional model is adopted for the steam generator. The obtained results indicate that the variable measurement is consistent with design data and that this new model is able to predict the dynamics of the reactor in different situations. It is revealed that flashing and condensation power are in direct relation with the stability of the system pressure, without which pressure convergence cannot be established.

• Nuclear Engineering and Technology
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• 제49권8호
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• pp.1733-1739
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• 2017

Molecular dynamics simulations were performed to investigate the uniaxial tensile properties of nanocrystalline $U_{0.5}Th_{0.5}O_2$ solid solution with the Born-Mayer-Huggins potential. The results indicated that the elastic modulus increased linearly with the density relative to a single crystal, but decreased with increasing temperature. The simulated nanocrystalline $U_{0.5}Th_{0.5}O_2$ exhibited a breakdown in the Halle-Petch relation with mean grain size varying from 3.0 nm to 18.0 nm. Moreover, the elastic modulus of $U_{1-y}Th_yO_2$ solid solutions with different content of thorium at 300 K was also studied and the results accorded well with the experimental data available in the literature. In addition, the fracture mode of nanocrystalline $U_{0.5}Th_{0.5}O_2$ was inclined to be ductile because the fracture behavior was preceded by some moderate amount of plastic deformation, which is different from what has been seen earlier in simulations of pure $UO_2$.

• Nuclear Engineering and Technology
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• 제54권1호
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• pp.72-83
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• 2022

This paper presents a numerical investigation of two-phase natural circulation flows established when external reactor vessel cooling is applied to a severe accident of the APR1400 reactor for the in-vessel retention of the core melt. The coolability limit due to external reactor vessel cooling is associated with the natural circulation flow rate around the lower head of the reactor vessel. For an elaborate prediction of the natural circulation flow rate using a thermal-hydraulic system code, MARS-KS1.5, a three-dimensional computational fluid dynamics (CFD) simulation is conducted to estimate the flow rate and pressure distribution of a liquid-state coolant at the brink of significant void generation. The CFD calculation results are used to determine the loss coefficient at major flow junctions, where substantial pressure losses are expected, in the nodalization scheme of the MARS-KS code such that the single-phase flow rate is the same as that predicted via CFD simulations. Subsequently, the MARS-KS analysis is performed for the two-phase natural circulation regime, and the transient behavior of the main thermal-hydraulic variables is investigated.

• 대한기계학회논문집B
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• 제37권9호
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• pp.855-862
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• 2013

개방 노심 열적여유도 해석 코드에 입력으로 제공되는 APR+ (Advanced Power Reactor Plus)의 수력학적 특징을 결정하기 위해 일련의 1/5 축소 원자로 유동분포 시험이 수행되었다. 본 연구에서는 원자로 내부 유동 계산시 다공성 모델을 사용한 전산유체역학의 적용성을 평가하기 위해 상용 전산유체역학 소프트웨어인 ANSYS CFX V.14를 사용하여 계산을 수행하였다. 결론적으로 본 연구에서 사용한 일부 원자로 내부 구조물에 대한 다공성 영역 처리방식을 통해 원자로 내부의 유동 특성을 정성적으로 적절히 파악할 수 있을 것으로 판단된다. 만일 충분한 계산 자원이 확보된 조건인 경우라면 노심 입구 상류에 위치한 원자로 내부 구조물의 실제 기하 형상을 고려함으로써 노심 입구 유량분포를 보다 정확하게 예측할 수 있을 것으로 예상된다.

• 설비공학논문집
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• 제25권5호
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• pp.269-278
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• 2013

In this study, in order to examine the applicability of computational fluid dynamics with the porous model to the analysis of APR+ (Advanced Power Reactor Plus) internal flow, simulation was conducted with the commercial multi-purpose computational fluid dynamics software, ANSYS CFX V.14. In addition, among the various reactor internals, the effect of flow skirt geometry on reactor internal flow was investigated. It was concluded that the porous model for some reactor internal structures could adequately predict the hydraulic characteristics inside the reactor in a qualitative manner. If sufficient computation resource is available, the predicted core inlet flow distribution is expected to be more accurate, by considering the real geometry of the internal structures, especially located in the upstream of the core inlet. Finally, depending on the shape of the flow skirt, the flow distribution was somewhat different locally. The standard deviation of the mass flow rate (${\sigma}$) for the original shape of flow skirt was smaller, than that for the modified shape of flow skirt. This means that the original shape of the flow skirt may give a more uniform distribution of mass flow rate at the core inlet plane, which may be more desirable for the core cooling.

• Nuclear Engineering and Technology
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• 제53권4호
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• pp.1119-1126
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• 2021

Spacer grid with mixing vane had been widely used in nuclear reactor core. One of the main feather of spacer grid with mixing vane was that strong swirl flow was formed after the spacer grid. The swirl flow not only changed the bubble generation in the near wall field, but also affected the bubble behaviors in the center region of the subchannel. The interaction between bubble and the swirl flow was one of the basic phenomena for the two phase flow modeling in fuel assembly. To obatin better understanding on the bubble behaviors in swirl flow, full three dimension numerical simulations were conducted in the present paper. The swirl flow was assumed in the cylindral calculation domain. The bubble interface was captured by Volume Of Fluid (VOF) method. The properties of saturated water and steam at different pressure were applied in the simulation. The bubble trajectory, motion, shape and force were obtained based on the bubble parameters captured by VOF. The simulation cases in the present study included single bubble with different size, at different angular velocity conditions and at different pressure conditions. The results indicated that bubble migrated to the center in swirl flow with spiral motion type. The lateral migration was mainly related to shear stress magnitude and bubble size. The bubble moved toward the center with high velocity when the swirl magnitude was high. The largest bubble had the highest lateral migration velocity in the present study range. The effect of pressure was small when bubble size was the same. The prelimenery simulation result would be beneficial for better understanding complex two phase flow phenomena in fuel assembly with spacer grid.

• Nuclear Engineering and Technology
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• 제53권1호
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• pp.93-102
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• 2021

The analysis of the fluid flow characteristics in reactor pressure vessel is an important part of the hydraulic design of nuclear power plant, which is related to the structure design of reactor internals, the flow distribution at core inlet and the safety of nuclear power plant. The flow distribution and mixing characteristics in the pressurized reactor vessel for the 1000MWe advanced pressurized water reactor is analyzed by using Computational Fluid Dynamics (CFD) method in this study. The geometry model of the full-scaled reactor vessel is built, which includes the cold and hot legs, downcomer, lower plenum, core, upper plenum, top plenum, and is verified with some parameters in DCD. Under normal condition, it is found that the flow skirt, core plate holes and outlet pipe cause pressure loss. The maximum and minimum flow coefficient is 1.028 and 0.961 respectively, and the standard deviation is 0.019. Compared with other reactor type, it shows relatively uniform of the flow distribution at the core inlet. The coolant mixing coefficient is investigated with adding additional variables, showing that mass transfer of coolant occurs near the interface. The coolant mainly distributes in the 90° area of the corresponding core inlet, and mixes at the interface with the coolant from the adjacent cold leg. 0.1% of corresponding coolant is still distributed at the inlet of the outer-ring components, indicating wide range of mixing coefficient distribution.

• Nuclear Engineering and Technology
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• 제53권4호
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• pp.1088-1099
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• 2021

Time-frequency analysis technique is an effective analysis tool for non-stationary processes. In the field of reactor neutron noise, the time-frequency analysis method has not been thoroughly researched and widely used. This work has studied the time-frequency analysis of the reactor neutron noise experimental signals under bubble disturbance and control rod vibration. First, an experimental platform was established, and it could be employed to reactor neutron noise experiment and data acquisition. Secondly, two types of reactor neutron noise experiments were performed, and valid experimental data was obtained. Finally, time-frequency analysis was conducted on the experimental data, and effective analysis results were obtained in the low-frequency part. Through this work, it can be concluded that the time-frequency analysis technique can effectively investigate the core dynamics behavior and deepen the identification of the unstable core process.

• Nuclear Engineering and Technology
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• 제47권7호
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• pp.814-826
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• 2015

The most important advantage of nanoparticles is the increased thermal conductivity coefficient and convection heat transfer coefficient so that, as a result of using a 1.5% volume concentration of nanoparticles, the thermal conductivity coefficient would increase by about twice. In this paper, the effects of a nanofluid ($TiO_2$/water) on heat transfer characteristics such as the thermal conductivity coefficient, heat transfer coefficient, fuel clad, and fuel center temperatures in a VVER-1000 nuclear reactor are investigated. To this end, the cell equivalent of a fuel rod and its surrounding coolant fluid were obtained in the hexagonal fuel assembly of a VVER-1000 reactor. Then, a fuel rod was simulated in the hot channel using Computational Fluid Dynamics (CFD) simulation codes and thermohydraulic calculations (maximum fuel temperature, fluid outlet, Minimum Departure from Nucleate Boiling Ratio (MDNBR), etc.) were performed and compared with a VVER-1000 reactor without nanoparticles. One of the most important results of the analysis was that heat transfer and the thermal conductivity coefficient increased, and usage of the nanofluid reduced MDNBR.

• Nuclear Engineering and Technology
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• 제55권9호
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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.