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检索条件"机构=Science and Technology on Reactor System Design Technology Laboratory"

共 1276 条记录，以下是101-110 订阅

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Simulation of esprit of hualong secondary passive residual heat removal system by using relap5 16

Simulation of esprit of hualong secondary passive residual h...

引用

16th International Topical Meeting on Nuclear reactor Thermal Hydraulics, NURETH 2015

作者： Feng, Li Science and Technology on Reactor System Design Technology Laboratory ChangShun Avenue 328 ShuangLiuCounty ChengDu China

ISBN: (纸本)9781510811843

A 3rd generation reactor called HuaLong is developed by China National Nuclear Corporation (CNNC). S econdary passive system (PRS) is designed to remove the residual heat during 72 hours after the reactor trips. To verify the ability of PRS system, the test facility ESPRIT is built in Chengdu which is 1:1 in heigh t and 1/62.5 in power. Tests are consisted by three Steady-states and one transient. One model of RELAP5 mod3.2 is developed and used to simulate the steady and transient tests. The results of RJELAP5 show con sistent with facility results, but the early phenomenon during transients is different between RELAP5 and tests as RELAP5 underestimates the direct-contact condensation power.

关键词： Condensation

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Atomic Scale Simulation on Liquid Metal Embrittlement Induced by Segregation of Lead Element in Lead-Cooled Fast reactor 23rd

Atomic Scale Simulation on Liquid Metal Embrittlement Induce...

引用

23rd Pacific Basin Nuclear Conference, PBNC 2022

作者： Bida, Zhu Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9789811988981

Lead-Cooled fast reactor, which uses liquid lead or lead-bismuth alloy as coolants, is one of the six main reactors of the fourth generation reactor system. Most recently, the Lead-Cooled fast reactor received a lot of attention due to its safety, economy, and sustainability advantages. However, the liquid metal embrittlement (LME) could be a potential threat to the reliability of structure materials used in the Lead-Cooled fast reactor. Several experimental results show that the segregation of alloy element in grain boundary has a close association with liquid metal embrittlement. In this paper, by applying atomic simulation techniques, including molecular dynamics and Monte Carlo method, the segregation of lead element on poly-crystal ferritic and the resulting mechanical properties variations are simulated with the newest interatomic potential. The simulation results show that the lead element segregates on grain boundary. Moreover, such segregation of lead on grain boundary induced hardening and embrittlement of poly-crystal systems in the following uniaxial tensile testes. The atomic-scale mechanisms for such mechanical properties variation induced by lead element segregation are discussed carefully. In conclusion, these results verify that the segregation of lead element could be one of the critical mechanisms for LME, which well agree with previous experimental evidences. The present study may allow us to understand the mechanism of LME better. It can serve as a good reference for selecting more suitable materials for the Lead-Cooled fast reactor. © 2023, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Grain boundaries

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Effect of clamping failure on flow induced vibration and fretting wear of fuel rods

Effect of clamping failure on flow induced vibration and fre...

引用

ASME 2018 Pressure Vessels and Piping Conference, PVP 2018

作者： Huan-Huan, Qi Zhi-Peng, Feng Fu-Rui, Xiong Nai-Bin, Jiang Qian, Huang Xuan, Huang Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9780791851654

Fuel rods are subjected to both axial and lateral flow in the reactor core. In this study, we present a study on the flow induced vibration (FIV) and fretting wear of fuel rod with failed clamping at grids. First, according to the flow distribution around a type of pressurized water reactor (PWR) fuel rod, the power spectral density (PSD) is obtained to characterize the turbulence excitation. Next, by combining the correlation PSD test parameters, the mean square value of the vibration displacement of each rod mode is found, and then the wear depth of dimple position is calculated based on the ARCHARD wear formula. The clamping of fuel rod at various grids may fail due to inaccurate manufacturing, fuel transportation and in-core irradiation. The absence of clamping force would significantly influence the rod mode and thereby changes its FIV responses. Simulation results show that the failure of the leaf spring has negligible effect on the rod natural frequency whereas the dimple failure near the location with larger FIV amplitude has a much significant effect. The lateral flow velocities at the inlet and outlet of the core are larger. For the fully clamped fuel rod, the response amplitude of turbulent excitation at the bottom and top of the fuel rod is larger. This is even more obvious with a failed dimple at these locations. Comparatively, the effect of dimple support failure in the middle is less influential. The influence of dimple support failure over the rod wear depth depicts basically the same trend as on the maximum FIV amplitude. In addition to the FIV amplitude, wear depth is also related to rod natural frequency. By examining the multiplication of amplitude and frequency at the top and bottom grid, we found that the dimple failure has greater impact at the top grid. Copyright © 2018 ASME.

关键词： Natural frequencies

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STUDY ON STEADY OPERATIONAL CHARACTERISTICS OF FLOATING NUCLEAR reactor UNDER OCEAN CONDITION 29

STUDY ON STEADY OPERATIONAL CHARACTERISTICS OF FLOATING NUCL...

引用

2022 29th International Conference on Nuclear Engineering, ICONE 2022

作者： Cheng, Kun Chu, Xiao Qiu, Zhifang Xian, Lin Deng, Jian Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9780791886380

Different from land-based reactors, floating reactors are affected by the ocean condition. Numerical simulation is conducted to study the effects of heaving and rolling motion on typical floating nuclear reactor (FNR) under rated power operation by ARSAC code with ocean condition calculation module. The calculation results have shown that heaving motion can cause the periodic fluctuations in the flow rate of both the primary and secondary loops in FNR. The amplitude of heaving motion is the main influence factor on the flow rate fluctuation of the primary loop while the effect of heaving motion period is limited. Rolling motion can cause the out-phase flow fluctuation in the two symmetric loops of the primary system. However, the rolling motion has limited impact on the core flow rate and cannot cause dramatic changes of system parameters at the core. Within the range of motion parameters used in this paper, heaving and rolling motion cannot have a substantial impact on the safe operation of FNR under rated power operation. Copyright © 2022 by ASME.

关键词： Oceanography

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EXPLORATION AND APPLICATION OF ENSEMBLE LEARNING METHOD ON PREDICTING TWO-PHASE FLOW KEY PARAMETERS OF RECTANGULAR CHANNEL 29

EXPLORATION AND APPLICATION OF ENSEMBLE LEARNING METHOD ON P...

引用

2022 29th International Conference on Nuclear Engineering, ICONE 2022

作者： Huang, Qingyu Yu, Yang Zeng, Hui Yang, Hui Zhang, Siyuan Yang, Kunlin Lin, Yuanfeng Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9780791886410

Among the current mainstream nuclear reactor thermal-hydraulic calculation system analysis softwares (such as RELAP5, TRACE, etc.), different empirical or semi-empirical relationships are employed to calculate two-phase interfacial flow parameters. However, due to the lack of understanding of the physical laws of reactor fluids and the limited experimental data currently, the application scope and accuracy of these methods are relatively limited. As an important paradigm in machine learning methods, ensemble learning has received great attention in numerical prediction tasks in many fields due to its excellent robustness, nonlinear fitting ability and strong interpretability. In this study, three typical ensemble learning methods including random forest, gradient boosting regression tree, and extreme gradient boosting tree are used to build a two-phase interfacial parameter prediction model for rectangular channels. The investigated interfacial parameters include interfacial area concentration, void fraction, bubble velocity, bubble chord length, and bubble frequency. The result reveals that XGBoosting has the best performance in forecasting. In order to further improve the performance of the ensemble learning method and solve the problem of poor generalization ability of the homogeneous-ensemble model, the heterogeneous-ensemble strategy are adopted to combine different types of algorithms to provide a model with stronger scalability and generalization ability for two-phase flow interfacial parameter prediction. The results show that the usage of the heterogeneous-ensemble method can improve the prediction ability of the model to a certain extent, and further ensure the reliability of the data-driven model for the prediction of the two-phase interfacial flow key parameters in the rectangular channel. Copyright © 2022 by ASME.

关键词： Two phase flow

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NUMERICAL INVESTIGATION ON THE THERMAL-HYDRAULIC AND FIV CHARACTERISTICS IN LBE-COOLED HELICAL CRUCIFORM FUEL AND WIRE-WRAPPED FUEL 31

NUMERICAL INVESTIGATION ON THE THERMAL-HYDRAULIC AND FIV CHA...

引用

2024 31st International Conference on Nuclear Engineering, ICONE 2024

作者： Qi, Zhang Haoyu, Wang Cao, Junxian Yuanming, Li Chenxi, Li Zhenhai, Liu Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9780791888261

The lead bismuth eutectic (LBE) cooled fast reactor in one of the General 4 advanced reactors, which plays significant role in the nuclear fuel cycle. In the recent years, the wire-wrapped fuel with four wires surrounds the cylindrical fuel is developed and promoted, which may be accept by the LBE-cooled fast reactor. The helical cruciform fuel of the Lightbridge corporation has similar geometrical features with the wire-wrapped fuel. The cross-section of helical cruciform fuel consists of the convex blades and the concave elbows, and the cross-section is rotated around the axis. The blade and elbow of helical cruciform fuel function as the wires while they are part of the nuclear fuel. In the present work, the comparison on the thermal-hydraulic and flow-induced vibration (FIV) characteristics of the helical cruciform fuel and wire-wrapped fuel is carried out by numerical simulation. The fluid-structure interaction is solved by the coupling between the fluid and solid domains, and the transient turbulence fluctuation is solved by large eddy simulation (LES). The results show that the convective heat transfer is enhanced by the spiral structures even though the hot spot phenomenon is caused simultaneously. The hot spot temperature of the helical cruciform fuel is larger than the wire-wrapped fuel duo to the non-uniform distributed heat flux. As the coolant in the flow channel is blocked by the blades and wires, the directional transverse flow is generated and thus the transverse mixing is enhanced. The one-way mixing with opposite direction occurs in the two sides of the single rod. There are similarities in the FIV characteristics for the spiral shaped fuels. The maximum vibration displacement and dominant vibration frequency of the helical cruciform fuel and wire-wrapped fuel are almost the same. © 2024 by ASME.

关键词： Fluid structure interaction

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Comparative study on thermal stress analysis and fatigue curve in stress and fatigue calculation of nuclear equipment 28

Comparative study on thermal stress analysis and fatigue cur...

引用

2021 28th International Conference on Nuclear Engineering, ICONE 2021

作者： Shao, Xuejiao Xie, Hai Yixiong, Y. Zhang, Liping Fu, Xiaolong Mi, Xue Li, Hui Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9780791885246

In the stress and fatigue evaluation of primary nuclear equipment, thermal stress calculation and fatigue curve have a great impact on the calculation results. There is a problem of stability in the step integral when calculating the thermal stress. The transient heat conduction changes rapidly near the initial temperature time point. It is critical to choose the right sub steps. In the nuclear code, the design fatigue curve in the air environment is given by applying a factor of 2 on stress or a factor of 20 on cycles. The factors are used to consider various influences such as data scatter, structure size, surface roughness and load. The factors of fatigue curves in ASME BPVC 2007 edition + 2009 addendum and French AFCEN's RCC-M 2017 edition RPP 2 have changed. Due to the changes of the original test data and modificative factors, the new fatigue curve and the original fatigue curve are crossed. In code case N-761 of ASME, fatigue curves considering the influence of primary coolant environment are given under different strain rates. In the calculation of thermal transient using commercial ANSYS software, the setting of time step has a great influence on the thermal stress. This paper compares and analyzes the influence of different calculation methods on the thermal stress at different positions of the model. The effect of different fatigue curves on fatigue life is also discussed considering different air design curve and different strain rate fatigue curve in water. The results show that the fatigue results obtained by using the modified ASME or RCC-M air fatigue curve are smaller than those before the change when the usage factor is large to a certain extent. After considering the environmental fatigue curve, the fatigue results increase significantly. The accuracy of thermal stress calculation can be significantly improved by adding time points and refining time steps near the time points after the transient sharply changes, otherwise the influence of thermal

关键词： Thermal stress

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Research on the effect of grid clamping failure on flow elastic stability and vortex shedding of fuel rod 26

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2018 26th International Conference on Nuclear Engineering, ICONE 2018

作者： Huan-Huan, Qi Zhi-Peng, Feng Nai-Bin, Jiang Qian, Huang Xuan, Huang Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9780791851456

Flow elastic stability and vortex shedding were two important mechanisms of the flow induced vibration analysis. Due to the influence of manufacturing process, transportation and irradiation, the clamping action of grid on fuel rods may be invalid. Taking one fuel assemblies as an example, the effects of clamping failure on the natural frequencies, mode shapes, flow elastic stability and vortex shedding were studied. The results show that the effect of the rigid convex support failure on the natural frequency was directly related to the mode shape. The effect of the grid rigid convex failure near the nodes with larger amplitude on the natural frequency was obvious. The velocity of flow at the top and bottom of the fuel rods were larger and the size was comparable, this induced that the rigid convex failure of the top and bottom grids had a significant effect on the ratio of flow elastic stability and the vortex shedding. Copyright © 2018 ASME.

关键词： Vortex shedding

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POST-HOC INTERPRETABILITY BASED PARAMETER SELECTION FOR DATA ORIENTED NUCLEAR reactor ACCIDENT DIAGNOSIS system 29

POST-HOC INTERPRETABILITY BASED PARAMETER SELECTION FOR DATA...

引用

2022 29th International Conference on Nuclear Engineering, ICONE 2022

作者： Li, Chengyuan Li, Meifu Qiu, Zhifang Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9780791886427

During applying data-oriented diagnosis systems to distinguishing the type of and evaluating the severity of nuclear power plant initial events, it is of vital importance to decide which parameters to be used as the system input. However, although several diagnosis systems have already achieved acceptable performance in diagnosis precision and speed, hardly have the researchers discussed the method of monitoring point choosing and its layout. For this reason, redundant measuring data are used to train the diagnostic model, leading to high uncertainty of the classification, extra training time consumption, and higher probability of overfitting while training. In this study, a method of choosing thermal hydraulics parameters of a nuclear power plant is proposed, using the theory of post-hoc interpretability theory in deep learning. At the start, a novel Time-sequential Residual Convolutional Neural Network (TRES-CNN) diagnosis model is introduced to identify the position and hydrodynamic diameter of breaks in LOCA, using 38 parameters manually chosen on HPR1000 empirically. Afterwards, post-hoc interpretability methods are applied to evaluate the attributions of diagnosis model’s outputs, deciding which 15 parameters to be more decisive in diagnosing LOCA details. The results show that the TRES-CNN based diagnostic model successfully predicts the position and size of breaks in LOCA via selected 15 parameters of HPR1000, with 25% of time consumption while training the model compared the process using total 38 parameters. In addition, the relative diagnostic accuracy error is within 1.5% compared with the model using parameters chosen empirically, which can be regarded as the same amount of diagnostic reliability. Copyright © 2022 by ASME.

关键词： Nuclear power plants

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Flow-induced vibration analysis of the check valve 25

Flow-induced vibration analysis of the check valve

引用

25th International Congress on Sound and Vibration 2018: Hiroshima Calling, ICSV 2018

作者： Shuai, Liu Peng, Feng Zhi Zhou, Jiang Xiao Xi, Lv Shou, Zhang Feng Xuan, Huang Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9781510868458

As an important over-current adjustment unit in the nuclear power plant piping system, check valve often causes some vibration and noise problems. The study of the existing valves mainly focuses on the simple or fixed opening two-dimensional steady-state flow field, which can't be used to explain the vibration and noise problems. In this paper, a complex passive check valve which has the asymmetric distribution with one import and two exports is proposed. In order to investigate the coupling mechanism between the internal pressure fluctuations and the vibration of system with the automatic adjusting process, the three-dimensional fluid simulation model is constructed. In addition, the vibration characteristic of internal flow field and the irregular fluctuating of the flow field are also calculated when the spool is moving. The coupling vibration characteristics of the flow field is obtained, which provides a meaningful guidance for the vibration reduction of the valve. Copyright © (2018) by International Institute of Acoustics & Vibration. All rights reserved.

关键词： Flow fields

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