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2020 International Conference on Nuclear Engineering, ICONE 2020, collocated with the ASME 2020 Power Conference

作者： Ma, Yugao Liu, Minyun Han, Wenbin Xie, Biheng Chai, Xiaoming Huang, Shanfang Yu, Hongxing Department of Engineering Physics Tsinghua University Beijing China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9780791883785

Space fission power systems can enable ambitious solarsystem and deep-space science missions. The heat pipe cooled reactor is one of the most potential candidates for near-term space power supply, featured with safety, simplicity, reliability, and modularity. Heat pipe cooled reactors are solid-state and high temperature (up to 1500 K) reactors, where the thermal expansion is remarkable and the mechanical response significantly influences the neutronics and thermal analyses. Due to the considerable difference between heat pipe cooled reactors and traditional water reactors in the structure and design concept, the coupling solutions for light water reactors cannot be directly applied to heat pipe cooled reactor analyses. Therefore, new coupling framework and program need to consider the coupling effects among neutronics, heat transfer as well as mechanics. Based on the Monte Carlo program RMC and commercial finite element program ANSYS Mechanical APDL, this work introduces the three coupling fields of neutronics (N), thermal (T), and mechanics (M) for heat pipe cooled reactors. The neutronic and thermal-mechanical (N/T-M) coupling strategy is developed theoretically, focusing on the formulation of the nonlinear problem, iteration schemes, and relaxation methods. Besides, the finite element method and the Monte Carlo program use different meshes and geometry construction methods. The spatial mapping and geometry reconstruction are also essential for the N/T-M coupling, which is discussed and established in detail. Furthermore, the N/T-M coupling methods are applied to the preliminary self-designed 10 kWe space heat pipe cooled reactor. Coupling shows that the thermal-mechanical feedback in the solid-state reactor has negative reactivity feedback (-2007 pcm) while it has a deterioration in heat transfer due to the expansion in the gas gap. © 2020 American Society of Mechanical Engineers (ASME). All rights reserved.

关键词： Heat pipes

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Chemical Speciation Distribution and Thermal Stability of Heavy Metals Along Flue Gas Cleaning systems in a Hazardous Waste Incinerator

SSRN

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SSRN 2022年

作者： Li, Wenhan Ma, Zengyi Deng, Lilin Huang, Qianming Zhai, Zian Huang, Bochen Liu, Ting Wang, Shuang Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China State Key Laboratory of Clean Energy Utilization Institute for Thermal Power Engineering Zhejiang University Hangzhou310027 China

In this study, chemical speciation distribution and thermal stability of heavy metals (As, Cd, Cr, Cu, Pb, Se, Zn, and Hg) along flue gas cleaning systems were investigated in a hazardous waste incinerator located in Zhejiang, China. The chemical speciation and thermal stability of heavy metals, environmental risk evaluation, and morphological and mineralogical characteristics of incineration residues were assessed before and after heat treatments at different temperatures. The results showed that As and Cr in the raw incineration residues mainly fractionated as the residual form with the proportion of over 48.9% and 74.5%, respectively, whereas most of Se was associated with the oxidizable and residual fraction. The chemical speciation of Cu and Pb in bag filter ashes was related to the chlorine content in the waste, and high chlorine availability favored the formation of acid soluble Cu and Pb. The residual fraction of Hg (over 68.0%) was the most dominant in bag filter ashes, which might exist in the form of HgS. During heat treatments at different temperatures, the vaporization of heavy metals increased with increasing temperature. As, Cr and Se showed high thermal stability due to the fact that As, Cr and Se were involved in the crystallization reaction, and heat treatment would promote the transformation from the mobile fractions into residual fraction. Hg and Cd displayed the worst thermal stability, and high temperature heat treatment would contribute to secondary volatilization from the incineration residues. Cu, Pb and Zn had moderate thermal stability, and heat treatment would cause the volatilization of the mobile fractions or transformation from the mobile fractions into residual fraction. With the increase of heat treatment temperature, the environmental risk of incineration residues displayed a decreasing trend and reached the medium risk level or below after heat treatment at 1200 °C without any additives or pretreatments. © 2022, The Authors. All ri

关键词： Heavy metals

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Experimental Study on Surface Characteristics and Thermal Conductivity of New Zirconium Alloy with Oxide Layer

SSRN

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SSRN 2022年

作者： Zhong, Lei Chen, Deqi Yu, Hongxing Liu, Haidong Chen, Mingjing Liu, Hanzhou Deng, Jian Ding, Shuhua Wu, Dan Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China Key Laboratory of Low-grade Energy Utilization Technologies and Systems Chongqing University Ministry of Education Chongqing 400044 China

The surface properties and thermal conductivity of oxide layer formed on the new zirconium alloy due to high temperature steam oxidation are studied in this paper. The thickness of oxide layer, surface wettability and thermal conductivity are measured. The effects of oxidation temperature and time on the wettability and thermal conductivity of zirconium alloy are analyzed. The results show that zirconium on the surface is fully oxidized to ZrO2 , and the difference of surface wettability is caused by different surface morphology. The characteristics of oxide layer surface cracks will affect the surface wettability. At the same oxidation temperature, with increasing time duration, the size and number of oxide layer cracks increase. The change of thermal conductivity is the result of competition among factors such as the development of recrystallization process, the increase of oxide layer thickness and the influence of pores and cracks. © 2022, The Authors. All rights reserved.

关键词： Wetting

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Verification of the 2D/1D coupling 3D transport code TIGER with C5G7 benchmarks

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International Journal of Advanced Nuclear reactor design and technology 2019年 1卷 43-50页

作者： Wu, Wenbin Yu, Yingrui Li, Qing Yao, Dong Luo, Qi Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

TIGER is a 2D/1D coupling 3D transport code developed by Nuclear Power Institute of China based on large-scale parallel computation. It employs 2D domain-decomposed parallel Matrix MOC in the radial direction and 1D finite difference diffusion in the axial direction. 2D domain-decomposed parallel Matrix MOC is accelerated with multi-group and multi-domain coupled PGMRES algorithm. Generally, 2D planar problem are solved independently and parallel degree is natural in the axial direction. Combination of the parallel degrees of both radial and axial directions is adopted in TIGER to achieve the large-scale parallel computation. In this paper, a few variations of C5G7 benchmarks were created based on the published benchmarks. Reference results of these variations were provided by the Monte Carlo code RMC. To demonstrate the accuracy and efficiency of TIGER, C5G7 benchmarks and these relevant variations were calculated by TIGER, and the comparisons were made with the published results and Monte Carlo results. © 2019 Xi'an Jiaotong University

关键词： Matrix algebra

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Experimental study of thermal and hydraulic characteristics of pressurizertype heating container during blowdown process

Experimental study of thermal and hydraulic characteristics ...

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2020 International Conference on Nuclear Engineering, ICONE 2020, collocated with the ASME 2020 Power Conference

作者： Ke, Bingzheng Gao, Puzhen Wang, Bo Chen, Bowen Li, Ru Tian, Ruifeng Cheng, Kun Lu, Chuan Fundamental Science on Nuclear Safety and Simulation Technology Laboratory Harbin Engineering University Harbin China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9780791883785

Blowdown process is an important process in LOCA. Studying the internal thermal hydraulic characteristics of the pressurizer during LOCA blowdown process can help to accurately analyze the process of the whole LOCA, which is of great significance to reduce the hazard of accidents and improve the safety of nuclear power plants. An experiment system of blowdown process under low pressure conditions is built, and it uses a pressure gauge and a differential pressure level gauge to measure the pressure and liquid level in the pressurizer-type heating container, and uses a thermocouple to measure the temperature in the pressurizer-type heating container at different heights and distances from the centerline. The changes of pressure, liquid level and temperature characteristics during the blowdown process were obtained. The research results show that the higher the heating power, the greater the slope of pressure change in the heating container, and the pressure changes from negative to positive growth;The greater the heating power is, the faster the liquid level decreases, and the liquid level decreases sharply after the blowdown inflection point. The greater the heating power is, the greater the rate of temperature change, from negative growth to positive growth. The temperature change rate after the blowdown inflection point is reduced. As the height increases, the temperature inside the container first increases and then decreases;the further away from the container's centerline, the higher the measured temperature. © 2020 American Society of Mechanical Engineers (ASME). All rights reserved.

关键词： Containers

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Synthetic subnanochannels in porous aromatic frameworks accelerate selective water permeation in membrane desalination

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science China Materials 2022年第7期65卷 1920-1928页

作者： Lei Wang Panpan Zhang Kaikai Chen Junchao Dong Feng Luo Qinglin Huang Zhixia Sun Xiaoqin Zou Guangshan Zhu Key Laboratory of Polyoxometalate and Reticular Material Chemistry of Ministry of Education Faculty of ChemistryNortheast Normal UniversityChangchun 130024China Institute of Chemical and Industrial Bioengineering Jilin Engineering Normal UniversityChangchun 130052China State Key Laboratory of Separation Membranes and Membrane Processes Tiangong UniversityTianjin 300387China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of ChinaChengdu 610213China

Creation of freeways in a membrane accelerates water transport for efficient desalination;however,the synthesis of target membrane materials is very *** the present study,the molecular-design strategy is developed for the synthesis of porous aromatic framework(PAF)membranes with rigid and open networks to regulate water ***-based monomers in triangle shape are designed for the construction of PAF-98 and its sister frameworks,and these PAF membranes are formed through trifluoromethanesulfonic acid *** rigidity and geometry of the monomers endow PAFs with unique structures of connected pores with edging size(4.5–6.7Å)between water(2.8Å)and hydrated sodium molecules(7.2Å),enabling effective differentiation of water and *** fabricated membranes with high flexibility and continuity are ideal for water desalination,and reverse osmosis experiments demonstrate that the three PAF membranes are selective for permeation of water molecules over hydrated sodium *** water flux of 547.1 L m^(-2) h^(-1) and high rejection degree of 90%are obtained on a representative membrane of PAF-98-E;and the desalination performance is superior to currently available membrane ***,PAF-98-E membrane is capable of rejecting various cations with high efficiency in real seawater,holding great promise in practical water desalination application.

关键词： porous aromatic framework water sieving reverse osmosis molecular design membrane desalination

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Experimental investigation on temperature characteristics of direct contact condensation in a natural circulation system

Experimental investigation on temperature characteristics of...

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2020 International Conference on Nuclear Engineering, ICONE 2020, collocated with the ASME 2020 Power Conference

作者： Sun, Jianchuang Ding, Ming Mi, Zhengpeng Zhang, Zhuohua Fundamental Science on Nuclear Safety and Simulation Technology Laboratory Harbin Engineering University Harbin China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9780791883785

Direct contact condensation (DCC) is a common physical phenomenon appearing in the nuclear power plants and other industrial applications. The current research on DCC focuses on steam-water counterflow or jet flow under forced flow conditions. For some natural-circulation passive safety systems in floating nuclear power plants, the heated section is connected with the heat sink by the horizontal pipes. The heat sink is usually the ocean. In such natural circulation systems (NCSs), the steam produced in the heated section will inevitably contact the subcooled water in the ocean and result in DCC event. In this paper, the fundamental characteristics of two-phase flow were described in detail. In addition, the influences of the subcooled water temperature on the flow rate, outlet temperature, steam bubble behaviors, and pressure surge were emphatically discussed. The experimental results revealed that the subcooled water in the water tank could reversely flow into the pipe, which was able to cause condensation induced water hammer (CIWH) event and flow oscillation. Further research showed that the subcooled water was evidently prevented from reversely flowing into the pipe with the increase in the subcooled water temperature. The position where the DCC event occurs transferred from the pipe to the water tank, and the pressure peak obviously reduced, even disappear when the subcooled water temperature is larger than 61 oC. © 2020 American Society of Mechanical Engineers (ASME). All rights reserved.

关键词： Water temperature

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Experimental research on the startup and shutdown heat transfer behavior of the coupling system of multi-stage heat pipes and radiators

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Annals of Nuclear Energy 2025年 222卷

作者： Liu, Limin Du, Zhengyu Guo, Ziang Liu, Ziyin He, Xiaoqiang Gu, Hanyang School of Nuclear Science and Engineering Shanghai Jiao Tong University China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China China

The heat pipe cooled reactor (HPR) offers high safety and compactness for space applications. This study experimentally investigates the coupled transient thermal behavior of the primary sodium heat pipe, thermoelectric generator (TEG), and secondary potassium heat pipe radiators during startup and shutdown. The primary heat pipe reached ∼700 °C at the evaporator within 3 h, while the secondary radiators exhibited delayed activation, with most condensing sections exceeding 299 °C after 21 h. A sonic limit effect caused local thermal and pressure drops at 800 mm. 75–80 % of input power was dissipated radiatively, exhibiting a thermal lag due to system inertia. During shutdown, the primary heat pipe responded rapidly to power reduction, while secondary heat pipe radiators showed accelerated cooling after passing their transition temperature. Stepwise power reductions led to faster temperature drops in the primary pipe than constant-rate reduction conditions. These results offer insights into HPR thermal dynamics for space system design. © 2025 Elsevier Ltd

关键词： Space applications

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Research on probabilistic safety assessment (PSA) method of safety class distributed control system (DCS) in nuclear power plant based on fuzzy Markov process

Research on probabilistic safety assessment (PSA) method of ...

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2020 International Conference on Nuclear Engineering, ICONE 2020, collocated with the ASME 2020 Power Conference

作者： Tian, Xufeng Hao, Yan Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China HuaFu avenue 299# section 3 Chengdu city610213 China

ISBN: (纸本)9780791883778

Safety class DCS(Distributed Control system) in nuclear power plant is characterized by multi-state interaction, complex function and unknown fault mechanism, which limits the use of tradition PSA(Probabilistic Safety Assessment) method such as event tree and fault tree. An improved dynamic probabilistic security simulation method is proposed based on fuzzy Markov process to quantitative evaluation the risk of Safety class DCS in nuclear power plant, which resolved the failure uncertainty problem without in-depth understanding of failure mechanism. Firstly, the fuzzy fault tree reliability evaluation model is established based on fuzzy theory and FTA(Fault Tree Analysis), which describes the probability and degree of failure with fuzzy number. Then, the fuzzy fault tree is transformed into fuzzy Markov chain, this model is described mathematically, and the reliability indexes are simulated and analyzed with Monte Carlo simulation method, the change of system safety performance in long-term operation is obtained. Finally, the scientificity and effectiveness is verified by the example of redundant analog output module of safety class DCS in nuclear plant. Copyright © 2020 ASME

关键词： Nuclear power plants

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Experimental Study on Flow Patterns in Narrow Rectangular Channels

SSRN

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SSRN 2022年

作者： Lv, Xindi Zhang, Dalin Song, Gongle Su, Guanghui Tian, Wenxi Qiu, Suizheng State Key Laboratory of Multiphase Flow in Power Engineering Shaanxi Key Laboratory of Advanced Nuclear Energy and Technology School of Nuclear Science and Technology Xi’an Jiaotong University Xi’an710049 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China

Plate fuel elements are commonly used in compact nuclear reactors due to the advantages of high fuel consumption. In this paper, the two-phase flow pattern in the coolant channel of plate fuel elements, i.e. the rectangular narrow channel was studied. Visualization flow pattern experiments of upward steam-water two-phase flow in narrow rectangular channels, 745mm long and 67mm wide with gap widths of 1.3mm and 1.9mm, were investigated. The experiments were performed in the conditions of pressure ranging from 1 to 2MPa, mass flow rate ranging from 200 to 1500kg/(m2 ⋅s), inlet subcooling ranging from 20~100°C and tilt angle ranging from 0~ 45°. Three flow patterns, consisting of bubble flow, churn flow and annular flow, were obtained in a wide range of thermal parameters. Flow patterns maps were plotted and the influences of pressure, mass flow rate, inlet subcooling, channel gap and tilt angle on the flow pattern transformation of narrow rectangular channel were studied. This study is of great significance for investigating the subsequent heat transfer process of two-phase flow in compact nuclear reactors. © 2022, The Authors. All rights reserved.

关键词： Two phase flow

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