bluuebunny/crossref_metadata_2025 huggingface dataset:
Conceptual Design and Feasibility Analysis of a Modular Supercritical CO2 Fast Reactor Core
Authors: Lai Guihua , Wang Tian , She Ding , Wang Hong
Published: November 2024

Small Modular Reactor (SMR) is a type of prospective nuclear reactor which has gained extensive attention due to its flexibility, economics, and safety. This paper presents an innovative design of a fast reactor core cooled by supercritical CO2 (SCO2), which can operate at 50 MWth for at least 20 years without refueling. A tube-in-duct structure using beryllium oxide (BeO) as matrix material is proposed. Specifically, the heat generated by the fuel rods is conducted by BeO matrix to adjacent coolant channels, and the SCO2 removes the heat out of the core. Reflectors, shieling, and control rods are applied for operation modification and safety considerations. To enhance the inherent safety, the reactor used double-layer SiC particles with uranium nitride (UN) as fuel. These particles are randomly dispersed in lead contained in the middle of the fuel rods, while the lead offers a better conduction performance. Partial 238U is substituted by minor actinides (MA) to harden the neutron spectra. In the feasibility analysis, both neutronic and thermal-hydraulic calculations are performed under normal condition. Results show that when the reactor is operated at 50 MWth for 30 years, the burnup is about 98.123 MWD/kgHM and the maximum power peaking factor is 1.57 in the radial direction and 1.2 in the axial direction. The inlet/outlet temperature of SCO2 is 660K/873K, while the maximum temperature inside the core holds below 1600 K, which is under the temperature limits. This work may provide a reference for the design of small modular reactor in the future.
Study on Radial Thermal-Hydraulic Characteristics of Once-Through Steam Generator of Sodium-Cooled Fast Reactor
Authors: Liu Yapeng , Wang Bo , Zhang Dalin , Feng Zhenyu , Li Xinyu , Tian Wenxi , Qiu Suizheng , Su Guanghui
Published: November 2024

In sodium-cooled fast reactor, once-through steam generator (OTSG) is an important heat transfer component, and it is also one of components that is prone to failure. The safe and stable operation of this equipment is crucial to the safety of sodium-cooled fast reactors. The steam generator has a sizable number of heat exchange tubes. Variations of mass flow rates in distinct radial zones on the primary side of the steam generator are caused by structural variations in the steam generator, which in turn cause variations in flow and heat transfer along the radial direction. Further thermal stresses are produced as a result of these variations in the fluid and structure’s temperatures. As a result, more detailed modeling of the primary side of the steam generator is required. An essential technique for a detailed simulation of the primary side of the once-through steam generator is computational fluid dynamics (CFD). However, CFD method is time-consuming and difficult to apply to long-term transient analysis. Therefore, in order to calculate the long-term transient operating conditions, and considering the radial thermal-hydraulic difference on the primary side of the steam generator, a radial parallel multi-channel model is developed to analyze the radial thermal-hydraulic difference of the once-through steam generator. This model can use multiple parallel channels to model the primary and secondary sides of the heat exchange tubes in different areas of the steam generator. The lateral momentum exchange between the heat exchange tubes is not considered in the model, but the energy exchange between different radial regions can be considered in the form of heat conduction. First, the model was verified using the transient test conditions carried out on the once-through steam generator test facility (SGTF) in India. The calculation results were consistent with the test results, and the deviation was within 10 K. The test results and calculation results in different radial areas were compared, and the program’s ability to calculate radial thermal-hydraulic differences was verified. Finally, the test conditions of the steam generator prototype of the China Fast Reactor 600 (CFR600) were analyzed. By considering the difference in radial flow distribution, the results of single channel model and radial multi-channel model are compared, and the impact of the radial thermal-hydraulic difference of the once-through steam generator on the simulation results was analyzed. The calculation results show that the radial multichannel model can effectively analyze the flow/heat transfer differences of the steam generator in different radial areas. It is verified that the model can analyze the radial thermal-hydraulic difference of the steam generator, and can provide reference data for the rapid thermal-mechanical coupling analysis of the steam generator.
Containment Depressurization Rate Requirements for Design Basis Loss-of-Coolant Accidents
Authors: Zheng Hua
Published: November 2024

Currently passive containment cooling system (PCCS) of HPR1000 is used in design extension conditions. PCCS uses loop-type heat pipes with evaporators inside the containment and condensers and a large water storage pool outside the containment. One of HPR1000 potential optimizations is using PCCS to mitigate design basis accidents, but there is an issue of containment depressurization rate. Section 6.2.1.1.A of Standard Review Plan (SRP) requires the containment pressure should be reduced to less than 50% of the peak calculated pressure for the design basis loss-of-coolant accident within 24 hours after the postulated accident. Due to the working principles of PCCS and available layout space inside the containment, it ‘s difficult for PCCS to meet the containment depressurization rate requirement. Safety requirements of nuclear safety regulations, guides, standards and utility requirements documents are analyzed to trace back to the provenance and its technical basis. Design features for meeting containment depressurization rate requirements of typical light water cooled reactors (LWR) are analyzed and compared. A combination of increasing the containment design pressure and increasing heat transfer area of PCCS is proposed for HPR1000. Loss of coolant accident is analyzed for HPR1000, which demonstrates that after increasing the heat transfer area and the design pressure of the containment, HPR1000 can meet the containment depressurization rate requirement.
Droplet Evaporation Characteristics on High-Temperature Porous Surfaces for Cooling Fuel Debris
Authors: Yuki Kohei , Horiguchi Naoki , Yoshida Hiroyuki , Yuki Kazuhisa
Published: November 2024

Fuel debris at the Fukushima nuclear power station is typically cooled under immersion. However, significant drop in water level results in coolant contact with high-temperature, porous fuel debris. In such scenarios, rapid cooling is essential, yet the thermal behavior at the liquid-solid interface, including capillary phenomena, is not well understood. This paper presents basic research evaluating the evaporation characteristics of droplets upon contact with metallic porous media featuring small pores under 1 mm. We conducted experiments using bronze or stainless steel porous media with pore diameters of 1, 40, or 100 μm to derive lifetime curves for droplets. Our findings indicate that Leidenfrost effect is mitigated on porous surfaces as the vapor can escape through the pores. Moreover, in bronze porous media, as the temperature increases, oxide film with a fine structure facilitates capillary action. In contrast, the low wettability of stainless steel porous media prevents capillary action, inhibiting droplet absorption and dispersion into the pores. Consequently, rapid cooling via the capillary action is unlikely if the fuel debris shares similar characteristics with steel porous media.
Research on CFD Calculation and Analysis Technology of Porous Media in Sodium Cooled Fast Reactor Core
Authors: Sun Yuchen , Chen Guangliang , Yin Xinli , Qian Hao , Tian Yizhi , Zhang Senyong , Li Menglai
Published: November 2024

The composition of sodium cooled fast reactor fuel components is relatively simple, without too many complex structures. Therefore, it is necessary to carry out thermal hydraulic simulation of the entire core and large watershed. However, Due to the complexity of the existing CFD refined methods in modeling and grid partitioning, the long calculation time lead to the inconvenience of the real time analysis. Therefore, the porous medium method is used for CFD analysis in the entire core and inter-wrapper flow of sodium cooled fast reactors. At the same time, the conjugate heat transfer method is used to simulate the fuel assembly box wall and separate the flow inside and between the boxes for research. The research results indicate that the porous media method can greatly simplify the difficulty of modeling and the number of grid divisions. Through comparative analysis and verification, it is found that the influence of inter-wrapper flow on the flow inside the core is limited. Therefore, in actual modeling, the inter-wrapper flow, as well as components with extremely low power flow rates such as neutron sources and reflector layers, are ignored. Only fuel components are analyzed, while other components are ignored. On this basis, flow partitioning was applied to the core, and the results were compared with those without partitioning. The results showed that flow partitioning promoted the flattening of the core outlet temperature. Finally, the calculation results will be compared with experimental data to demonstrate the effectiveness of the method using porous media. The calculation has achieved sub real-time and has good engineering application value.
Transition Mechanism of Supercritical Carbon Dioxide Between Gas-Like and Liquid-Like States
Authors: Liu Minyun , Han Yansong , Huang Yanping
Published: November 2024

In recent years, the supercritical carbon dioxide power cycle has made great breakthroughs and has been applied to new energy fields such as solar heat and waste heat recovery. In the field of nuclear energy, the supercritical power cycle has the advantages of high efficiency, compact structure, and good compatibility with active media, so it has broad application prospects in the new generation of nuclear energy systems combined with high-temperature gas cooled reactors, sodium-cooled fast reactors, and even fusion reactors. Supercritical fluids are typically regarded as homogeneous, continuous, indistinguishable, single-phase substances without undergoing phase transitions. However, recent research has demonstrated the existence of a transition mechanism between gas-like and liquid-like structures within supercritical fluids. Considering that this structural transition may lead to the mixing of nitrogen and supercritical carbon dioxide within the pressurizer, this paper devised a straightforward visual experiment. The experiment investigated the characteristics of the interface between supercritical carbon dioxide and nitrogen under different pressure and temperature conditions. The results revealed that, under the same pressure, as the temperature rises, the interface between supercritical carbon dioxide and nitrogen evolves from being clear to gradually becoming blurry and ultimately vanishing. Under a pressure of approximately 7.7 MPa, nitrogen and supercritical carbon dioxide will completely mix at temperatures above 31 °C. To address the significant interference caused by this issue in thermal-hydraulic experiments, this paper also introduces an effective pressurizer design.
Numerical Analysis of Three-Dimensional Flow Field Impact Under Strong Transient Conditions in Advanced Reactors
Authors: Zhao Jian , Zhang Jing , Wu Yingwei , Su Guanghui , Tian Wenxi , Qiu Suizheng
Published: November 2024

As a representative of Generation IV nuclear energy systems, sodium-cooled fast reactors (SFR) may cause the liquid sodium flowing through the core to boil after a Core Disruptive Accident (CDA), which then forms a high-energy vapor cavity expansion to push the liquid sodium above, causing an impact failure to the main vessel structure in a very short period of time, which is a safety hazard. In this study, the computational analysis of the process of high-energy vapor cavity expansion pushing liquid sodium impacting on the top of the reactor vessel was carried out from the full scale to derive the effects of high-energy impact on the inner wall and upper components of the reactor vessel after the accident. The results show that after the CDA, the sodium-cooled fast reactor shielding lid is subjected to a great impact pressure at the center of the cork, while the covering argon gas at the top is compressed, resulting in a rapid increase in the internal pressure of the reactor vessel within a short period of time, and extrusion of the vessel inner wall. Through this study, the flow field distribution characteristics inside the reactor vessel in the late stage of the core disruptive accident are grasped, and the distribution of the impact pressure on the top shielding cover at the full reactor scale in the core disruptive accident is obtained, which can provide technical and result references for the studies related to the CDA of pool-type sodium-cooled fast reactors (PSFR).
Research on the Criteria for Single Protection Channel Power Failure in Nuclear Power Plants
Authors: Zhang Xiuchun , Xia Hong , Liu Yongkang , Zhu Shaomin , Zhang Jiyu , Xu Junjun , Jiang Yingying , Yin Wenzhe , Zhang Jie , Liu Jie
Published: November 2024

In order to ensure the safe operation of a single protection channel in a nuclear power plant in case of power failure, it is necessary to carry out the consequence analysis of power failure in the construction drawing design stage. Moreover, in order to ensure the effective application of engineering design experience in a new power plant, it is necessary to establish design criteria and operation management criteria for the power failure of a single protection channel. Based on the principle of digital design of protection channel in a typical nuclear power plant, this paper analyzes the behavior, main consequences, possible safety injection risks, precautions for power transmission restoration after power failure of a single protection channel, and explains the verification process of power failure analysis results of a single protection channel in the design process, the design idea of corresponding operation documents and the logic of the operator’s execution of operation documents. Through the analysis and research process, the design criteria and operation criteria applicable to the power failure of single protection channel of nuclear power plant are obtained. The research results of this article can provide a basis for the design and operation management of single protection channel power failure in newly built nuclear power plants in China. Furthermore, it ensures the operational safety of nuclear power plants in the event of a single protection channel failure.
Research and Application of Containment Structural Health Monitoring System During Operation of Nuclear Power Plant
Authors: Mengyan Song , Di Yao , Chaoqi Zhang , Di Jiang , Peiyao Zhang
Published: November 2024

The containment of the nuclear power plant is an important item of the tolerance nuclear steam supply system. It is the third barrier to prevent radioactive waste from leakage after an accident. It is the most important structure of nuclear power plant ensuring the safety of nuclear power plants. To ensure that the containment meets the design requirements during the service period during service, Structure integrity test is done every ten years to verify whether the performance of the containment meets the design requirements. However, through the cycle Structure integrity test, the structural state of the containment is obtained every ten years. It is unable to obtained in real time. When the accident conditions occur, the structural state of the containment cannot be evaluated quickly and unable to assist the containment related decision -making. In order to ensure the safe operation of the containment during the service time, at the same time improve the intelligent level of the safety monitoring of the containment, developing a set of intelligent monitoring and evaluation and early warning systems to achieve daily monitoring, abnormal state warning, healthy real-time real-time monitoring, abnormal state warning, abnormal state warning, abnormal state early warning, healthy Status assessment are very necessary.
Study on Influence of Braking Circuit and Stator-Rotor Structure on the Damping Characteristics of Permanent Magnet Motor
Authors: Ren Zhiyuan , Wu Hongyu , Yan He , Li Tianjin , Diao Xingzhong
Published: November 2024

Permanent magnet motor is used as the power source to provide driving force for the lowering, lifting and holding action of the control rod in high-temperature gas-cooled reactor. When reactor scram and the rods falling by gravity, the permanent magnet motor also can act as a damping system to convert the gravitational potential energy of the control rods into the thermal energy consumed in the braking resistor, that enables a damping function of limiting the speed along the long dropping stroke and effectively buffering the impact of the control rods on the reactor internals. In order to further investigate the damping capability of the motor, a finite element model was established for the permanent magnet motor, as well as its external circuit. The braking and damping performance of the motor under different rod drop speeds was analyzed by simulation methods. An important performance parameter for evaluating the damping capability of motor is the damping torque, which is mainly affected by the braking resistance in the external circuit model and the stator-rotor structure of the motor. First, a full-size motor with external circuit model was established based on Ansys Maxwell software, and reasonable meshing and irrelevance verification of the motor body model was carried out to improve the solution accuracy and modeling efficiency of finite element simulation. Then the influence of the motor structural parameters (such as air gap, stator-rotor tooth shape, permanent magnet structure, and braking resistance setting of the external circuit) on the damping torque was studied through parameter-sensitive simulation and analysis, especially through the reasonable setting of the external circuit the adjustment of the damping torque was realized. The effect of motor structural parameters on damping performance was also obtained. According to the required falling speed of the control rod, the appropriate resistance value could be selected to regulate the falling rod speed, which provided a strong support for the realization of the falling rod buffer of the control rod drive mechanism with the permanent magnet motor serving as the second damping system. At present, the research work had been carried out based on the finite element model of the permanent magnet motor, which would provide a validation basis for the subsequent experimental research.
Simulation Study of Nuclear Heating Reactor Control Method in Slight Boiling Operation Mode
Authors: Huang Tianyi , Sun Jun , Xie Heng
Published: November 2024

Nuclear Heating Reactors (NHRs) could achieve multi-purpose target including heating and seawater desalination without emission of greenhouse gases. Institute of Nuclear and new Energy Technology (INET) have designed two operation modes for NHRs: pressurized mode and slight boiling mode. The former shows a higher level of maturity, while the latter guarantees better economic outcomes but still lacks comprehensive testing and analysis. Previous researches on operation of NHR slight boiling mode mainly concentrate on avoiding instabilities during reactor start-up process, while the power adjustment process is also important but seldom researched. NHR-200II/S is designed to obtain the slight boiling operation capability based on the parameters of NHR-200II, and RELAP5 code is employed for the simulation of operation processes. A proportional-integral-derivative (PID) control scheme is introduced to maintain the primary loop pressure by adjusting the flow rate of intermediate loop. The PID coefficients are determined and the effect of the control scheme is verified by power adjustment processes. The ability of proposed control scheme to maintain stable operation of NHRs in slight boiling mode is shown.
Concept Study of a Fission Fragment Propulsion System
Authors: Zhang Dacai , Yu Ganglin
Published: November 2024

Deep space exploration holds immense significance for the future of space exploration, and it requires propulsion systems with high specific impulse characteristics to accomplish long-duration space missions. The performance of prevalent nuclear thermal and nuclear electric propulsion systems is constrained by fuel element cladding temperatures and energy conversion systems efficiencies, precluding further enhancement of specific impulses. This paper focuses on a space nuclear propulsion system with a magnetic collimation structure based on the direct propulsion of fission fragments of Am-242m. The system employs eight control rods for reactive adjustments, the system’s critical and propulsion characteristics are analyzed by the combination of RMC and MATLAB. The calculation results demonstrate that the system exhibits superior comprehensive performance, with a specific impulse of approximately 4 × 104s while the keff of the system is optimized to reach about 1.06. The results also indicate that the system possesses excellent safety features and can play a crucial role in deep space exploration. The exploration of propulsion systems utilizing Pu-239 and U-235 as well as the investigation of enhancing the fission fragment escape rate and energy utilization rate will be continued in the future.
Numerical Study of Flow-Induced Vibration of Square Tube Bundle Structures Under Ocean Conditions Based on Large Eddy Simulation
Authors: Sun Junshuai , Sun Rulei , He Xiaoqiang , Wen Jiming , Tan Sichao , Tian Ruifeng
Published: November 2024

Marine nuclear power equipment is an important support to solve the problems of long-lasting power energy supply in the development of far-reaching marine resources and “carbon emission reduction” in the marine field. The additional force under the ocean environment will lead to the change of the flow field characteristics of the tube bundle structures of the steam generator in the nuclear power equipment, which in consequence will lead to the change of the flow-induced vibration characteristics, and the safety of the system will be tested. In order to study the flow-induced vibration characteristics of the tube bundle structures under ocean conditions, taking the square tube bundle structure as the research object, based on the large vortex simulation method, using the secondary development of customized functions, combined with the local dynamic mesh deformation technology, the fluid-structure interaction computational model of the tube bundle structures under ocean conditions was established. Firstly, the effect of grid discretization on the predictive ability of the flow field characteristics of the tube bundle structures was studied by the calculation of the flow field of the square tube bundle structures under the static conditions with typical parameters. Secondly, the accuracy of the fluid-structure interaction computational model was verified by the calculation of the flow-induced vibration of the square tube bundle structures under the non-ocean conditions. Thirdly, the flow field characteristics of the square tube bundle structures under the ocean conditions was researched and compared with the results of the non-ocean conditions, the effect of oceanic conditions on the flow field characteristics of the tube bundle structure was obtained. Finally, the flow vibration characteristics of the square tube bundle structures under ocean conditions were investigated and compared with the results of the non-oceanic conditions, the effect of ocean conditions on the flow-induced vibration characteristics of the tube bundle structures was obtained. The results show that the established fluid-structure interaction computational model can accurately predict the flow-induced vibration behavior of the tube bundle structures under oceanic conditions; the existence of ocean conditions will lead to the generation of transverse flow in the flow field of the tube bundle structures, which will enhance the fluid mixing between the tube bundles and make the flow-induced vibration more severe. This study is of great significance for the design and optimization of steam generator tube bundle structures in the marine nuclear power equipment.
Pool Boiling Heat Transfer Enhancement Using Bi-Conductive Surfaces
Authors: Tang Longchang , Liu Xiaojing , Xu Wei , Tang Luyao
Published: November 2024

Due to its superior heat dissipation capability, pool boiling has been a research hotspot over the decades. This paper provides insights into the underlying pool boiling enhancement mechanism of bi-conductive surface which is an effective modification method. In this study, a new type of bi-conductive surface is created by embedding a reticular structure made of low-conductive epoxy into a high-conductive copper substrate. The influence of the width of epoxy is studied by conducting experiment as well as CFD simulation. Due to extreme differences in thermal conductivity, the heat flux almost exclusively flows throw copper. Moreover, epoxy inhibits nucleation because of its low surface energy. Bubbles are almost only generated on copper and the colder epoxy remains moist. The existence of epoxy creates periodic spatial surface temperature variations and induces separated liquid and vapor paths. As a result, compared with bare copper surface, bi-conductive surfaces have better heat transfer performance. For the bi-conductive surface with 0.75mm width of epoxy, replacing 44% of the high-conductive copper with low-conductive epoxy results in 270% increase in CHF value and 108% increase in the total effective thermal power at CHF.
Experimental Study on Characteristics of Entrained Droplets and Liquid Film at Swirler Outlet in Swirl-Vane Separator
Authors: Kang Ruiqi , Xiong Zhenqin
Published: November 2024

The swirl-vane separator, recognized for its compact design and high separation efficiency, experiences a gas-liquid flow transition from non-swirling to swirling after passing through the swirler. Driven by centrifugal force due to gas-liquid density disparities, efficient separation is achieved as liquid is propelled towards the wall. However, complexities arise in the flow filed within the separator due to significant variations in cross-sectional area from upstream to downstream of the swirler. In conditions of insufficient gas flow or low liquid phase flow, a swirling churn flow emerges at the swirler outlet, causing fluctuations in the liquid film and entrained droplet generation. The limited separator riser height impedes timely droplet-wall contact, leading to compromised separation efficiency. To delve into this complexity, an air-water experimental facility was established, investigating the two-phase flow at the swirler outlet. Wire mesh sensors gauged liquid film and droplet characteristics under varying conditions (air superficial velocity: 12.4–20.4 m/s, liquid superficial velocity: 0.1–1.3 m/s). Results revealed that liquid film thickness increases with decreasing gas velocity and rising liquid velocity. Entrained droplet diameter correlates positively with liquid velocity and swirl vane inclination but inversely with gas velocity. Notably, a 45° swirl vane inclination induced dispersed liquid droplets within the swirling annular gas core. Understanding these distribution characteristics is pivotal for setting separator working conditions to enhance separation efficiency and optimize swirler structure.
Experimental Investigation of Early Response of Hot Water Rapid Blowdown in Separated Pipeline
Authors: He Xiaoqiang , Gao Puzhen , Zhang Ze , Wang Jianjun
Published: November 2024

In this study, a series of separate effects tests are carried out to investigate the phenomena of the early response of the blowdown process in the ruptured pipe. The experimental section is a 50 mm nominal stainless steel pipe about 1600 mm long. Several high-precision piezoelectric sensors are placed along the pipe and the outside space near the rupture point with a data acquisition frequency of up to 20 kHz. Two methods are used for the postulated pipe rupture simulation: one using a quick-opening solenoid valve, and the other is realized by a thin glass cap and shatter trigger. Based on the experiments, pressure distribution and its time trends are obtained during the rapid blowdown process with various initial conditions, including different pressures and subcooling degrees. The characteristics of different stages in depressurization are identified, and some other vital parameters, such as the rate of depressurization and the initial discharge speed are also analyzed. In addition, the time response curve of back pressure in the outside space is recorded, and the flashing of superheated water during the blowdown is filmed. Moreover, the initial speed of the discharge flow is calculated after the data processing. The results of the tests help to understand the transient blowdown process and provide verified data for the simulations. It also provides a foundation for future experiments under higher working conditions and for developing boundary models used in analysis code.
Study on Heat Transfer Characteristics of 2×2 Helical Cruciform Fuel Assembly Under Non-Uniform Heat Source
Authors: Luo Menghao , Lu Zhiwei , He Yanan , Wu Yingwei , Tian Wenxi , Su Guanghui , Qiu SuiZheng
Published: November 2024

Helical cruciform fuel, due to its unique geometric shape, affects heat transfer, thereby improving the power level and safety margin of existing pressurized water reactors. This article establishes a unidirectional neutron thermal coupling calculation method for the complex geometric cross-section of helical cruciform fuel to study its flow and heat transfer characteristics under non-uniform heat sources. Based on the Monte-Carlo particle transport encoding OpenMC, the fission power distribution of the two-dimensional cross-section of a helical cruciform fuel rod were studied, and we used this as a non-uniform heat source for calculation of flow and heat transfer of 2 × 2 fuel assembly. The neutron calculation results show that the fission energy is concentrated at the fuel petal convexity, and the radial power factor peak fluctuates around 1.3. The flow heat transfer calculation results show that the temperature and heat flux at the petal convexity of the fuel rod cladding outer are lower than those at the petal concavity due to the longer heat conduction path; the swirling flow in the coolant channel enhances the heat transfer on both sides of the petal convexity, and flattens the heat flow distribution at the petal convexity, resulting in a smaller heat flow peak here. The trend about distribution of heat flow and wall temperature on the outer surface of the cladding under non-uniform heat source is basically the same as that under uniform heat source, with a relative deviation of 2.38% for temperature and 18.32% for heat flow. The influence of large heat flux deviation on the local thermal design of fuel cannot be ignored, and it is necessary to further optimize the coupling method and accurately the actual deviation value.
Analysis of Turbulent Structures Based on Large Eddy Simulation in Rod Bundle Channels With Mixing Vanes
Authors: Zhang Lixuan , Chen Guangliang , Tian Zhaofei , Qian Hao , Zhang Yuhang , Li Rui
Published: November 2024

The refinement of thermal-hydraulic calculations for reactors is crucial for both the economy and safety of reactors. However, the spacer grids and mixing vanes will affect the turbulence and vortices. In particular, the prediction of near-wall turbulence using computational fluid dynamics (CFD) is more difficult due to the strong shear stress and anisotropic properties near the coolant channel walls. By studying the turbulent coherent structures in the core coolant channel, it is possible to better predict the turbulent flow and obtain more accurate core thermal-hydraulic parameters. This paper focuses on the experimental objects of AFA-2G spacer grids and mixing vanes and establishes a model that meets the requirements of Large Eddy Simulation (LES) to study turbulent structures. The parameters of the flow field acquired from LES were in good agreement with the experimental results. By comparing with the Reynolds time-averaged model, it was found that its mean field is more accurate. This paper quantitatively analyzed the anisotropic characteristics of coherent structures and visualizes their spatial distribution using barycentric map. The visualization of vortices was performed using Ω criterion, and the dynamic analysis of vortices was conducted to investigate the effects of spacer girds and mixing vanes on the generation and evolution of vortices, as well as their mutual interactions.
Study on the Spent Fuel Management in Different Developed Countries
Authors: Shi Lei , Su Jiqiang , Chen Ding , An Yan
Published: November 2024

Spent fuel is nuclear fuel that has been exposed to radiation, usually produced by nuclear reactors in nuclear power plants, and is an inevitable product of the development of nuclear energy. Spent fuel contains a large amount of radioactive elements, which are characterized by high levels of radioactivity, difficult processing techniques, and strict safety management requirements. It is related to the sustainable development of nuclear energy, as well as the safety of the environment and human life. Ensuring long-term safety management of spent fuel has always been an important prerequisite for countries around the world to develop nuclear energy. To this end, major nuclear power countries around the world have established complete management systems and mechanisms, formulated sound policies and regulations, and ensured the systematization, standardization, and effectiveness of the entire process of spent fuel management. This study focus on national management policies and regulatory systems, investigates the practices of spent fuel management in major nuclear power countries such as France, Japan, and South Korea, and summarizes experience to provide relevant inspiration and suggestions for the management of spent fuel in China. Study has shown that it is necessary to strengthen top-level design, clarify the responsibility and authority boundaries of all parties involved, and to formulate policy systems such as the long-term development plans for the industry.
New Turbulent Prandtl Number Model for Liquid Metal Based on DNS Results
Authors: Fu Hao , Zhao Houjian , Li Xiaowei , Wu Xinxin , Lyu Xuefeng , Liu Fang , Yu Yu , Xu Wei
Published: November 2024

Liquid metal is widely used as the primary coolant in many advanced nuclear energy systems. Prandtl number of liquid metal is much lower than the conventional coolant of water. Due to the high molecular thermal diffusion coefficients of liquid metal, the thermal conduction dominated region of liquid metal is much thicker than that of water. Logarithmic region for dimensionless temperature in the cross section is diminished due to the low Prandtl number effects. Conventional Nusselt number correlations derived from the logarithmic temperature distribution can not be used for liquid metal. For conventional coolant of water or air, the turbulent momentum field is similar with turbulent scalar field. Based on Reynolds analogy, the turbulent Prandtl number is assumed as one for the Reynolds Averaged Navier Stokes method. For turbulent convection of liquid metal, dissipations of turbulent scalar energies are larger than that of turbulent kinetic energies. Dissimilarity between the thermal and momentum field increases with the decreasing of Prandtl number. For turbulent convection of liquid metal, the turbulent Prandtl number is larger than one.

In the current investigation, turbulent convection of liquid metal in channel is directly simulated with OpenFOAM. Turbulent statistics of the momentum and the thermal field are compared with the existed database to validate the numerical model. Power law for dimensionless temperature distribution with different Prandtl number is obtained by regression analysis of DNS results. New Nusselt number correlation is derived based on the power law of temperature distribution. The new Nusselt number correlation agrees well with the DNS results in the literature. Using the method of order magnitude analysis, relationships between the turbulent heat flux and turbulent scalar energy is analysed. Turbulent scalar energy is used to calculate the turbulent thermal diffusion coefficients. New correlation is proposed to consider Prandtl number effects on dissimilarity between the turbulent kinetic energy and the turbulent scalar energy. Then, new turbulent Prandtl number model for liquid metal is obtained. Combined with SST turbulence model, numerical results with the new turbulent Prandtl number model agree well with DNS results. The new Prandtl number model can be used for turbulent convection with different Prandtl number and with different Reynolds number.
Experimental Study of the Effect of Pulsatile Flow on Corner Vortices at the Sudden Contraction Before the Inlet of Two Parallel Narrow Rectangular Channels
Authors: Ala Ayodeji , Ye Bin
Published: November 2024

Corner vortices can be found in regions where there is considerable flow expansion or contraction. Experimental measurement was used to investigate the changing vortices at the contraction region before a dual parallel channel arrangement. The flow in the loop was pulsatile with a flow period of 5 s, amplitude between 0.2 and 0.4, and flow rate in the range of 0.4 m3/h to 0.9 m3/h. The mean transient flow rate in the loop was within ± 6% of the ideal sinusoidal fluctuation at lower amplitude. The corner vortices were analyzed by their change in size during pulsation and the turbulence structure in the contraction area. The corner vortexes are quasi-symmetrical at the channel corners below the contraction plane. The corner vortices were larger when the flow velocity was off-peak. Regions of slower flow increased during the deceleration half-cycle of the pulsatile flow. The results provide more explanation of flow behavior at channel corners and contraction planes.
Exploration of Ultra-High Performance Concrete (UHPC) Application in Nuclear Power Plant Structures
Authors: Peiyao Zhang , Mengyan Song , Di Jiang , Jingwei Gao
Published: November 2024

Ultra-High Performance Concrete (UHPC), as a rapidly developing high-performance cement-based material in recent years, has three major characteristics: high strength, high ductility, and high durability. Based on these characteristics, UHPC offers the potential for conducting “ultra-long life” highways or bridges. Therefore it was initially applied in bridge structures. Due to the safety requirements of nuclear power plants and the fact that most nuclear power plants are distributed along the coast, the material properties of UHPC are very suitable for the structures of nuclear power plants. Currently, UHPC has been preliminarily applied in Nuclear Power Plants. In this paper, a new type of containment structure called UHPC containment is proposed and studied. The finite element analysis is carried out on the UHPC containment by varying the different tensile strength and reinforcement ratios. The results obtained from finite analysis are compared with those from the current pre-stressed concrete containment structures to discuss the feasibility of applying UHPC in such structures. The findings suggest that reasonable use of UHPC can contribute to the enhanced safety of containment structures. Finally, this study proposed required material characteristics and design methods for this new type of UHPC containment structure.
CFD-DEM Simulation on Gas-Solid Separation and Sphere Packing Behaviors in a Storage Bin of the Small Absorber Sphere System
Authors: Li Tianjin , Hu Haitao , Yan He , Diao Xingzhong
Published: November 2024

The small absorber spheres system is one kind of shutdown system in the pebble-bed High Temperature Gas-cooled Reactor (HTGR or HTR). Vertical pneumatic conveying of small absorber spheres in pebble-bed HTGR is a special application of pneumatic conveying technique. The whole process consists of several sub-processes, e.g., the granular discharge from the sphere storage bin into the boring, granular discharge from the side reflector boring into the feeder, particle entrainment in the feeder, sphere conveying in the transport pipeline, and gas-solid separation in the storage bin. The Computational Fluid Dynamics and Discrete Element Method (CFD-DEM) coupling simulation has recently been extended and used to investigate particle vertical conveying in the small absorber sphere system in our previous work. In this study, we further investigate the gas-sphere separation and sphere packing behaviors in a storage bin of the small absorber sphere system by CFD-DEM coupling simulation, to make a better understanding of the gas-solid flow and particle flow behaviors in the small absorber spheres system. The influence of operation parameters is considered, including gas velocity, particle velocity and solid flow rate. The effect of gas and solid properties are also considered, including media type (high pressure Helium and ambient air) and friction coefficient of the sphere and wall. Attention is paid to the sphere motion behaviors and heap up characteristics of the spheres in the storage bin, to gain more knowledge of the evolution of heap up surface for the design of sphere level indicator.
Application of Plasma Arc Cutting Technology in Dismantling of Large Metal Components in Containment Building
Authors: Zhang Naizhe , Wang Weishuai , Gao Zhen , Liu Hai , Liu Xuegang
Published: November 2024

The decommissioning of nuclear reactor is a challenge to the sustainable development of nuclear energy. One of the most challenging tasks in reactor decommissioning is the dismantling and segmentation of radioactive large metal components in the containment building. A variety of cutting technologies have been developed for the special purpose. This paper is focused on the underwater plasma arc cutting technology and its potential applications.

The characteristics of large metal components in nuclear containment building were summarized. The matters needing attention in the process of dismantling and segmentation were analyzed as well. The technical principle of plasma arc cutting was briefly introduced, and the selection of the key operational parameters are discussed, such as the types of gas and electrode materials.

Cutting noise, arc light, and smoke will be efficiently reduced by underwater cutting. More importantly, underwater cutting can reduce the radiation dose to the occupational staff and optimize the radioactive waste management. A comparative evaluation was conducted among underwater plasma arc, arc saw and wire saw cutting. In the last part, a hypothetical scenario taking a steam generator as an example for dismantling is proposed and analyzed. The differences between the plasma arc cutting and the arc saw cutting were compared in terms of cutting time and the amount of secondary waste generation.

In this paper, the underwater plasma arc cutting was comprehensively analyzed and discussed to validate the technology in the decommissioning of large metal components in the containment building. Some concerns for industrial application were revealed.
Thermohydraulic Simulation of Energy Well Micro Modular Reactor Using System Codes
Authors: Mazzini Guido , Benčík Marek , Reungoat Mathieu
Published: November 2024

Due to the actual energy need worldwide, which should be carbon oxide emission free, several initiatives to prepare advanced Small Modular Reactor (SMR) and Micro Modular Reactors (MMR) are ongoing in order to offer new energy solutions (not only for electrical production) with high standard in safety. The goal of these technologies is to obtain clean energy with low carbon emission, which will be located nearby industries or cities. In particular, several countries such as United States of America, Great Britain and Canada plan to build a fleet of SMRs in order to find a compromise between, the actual energy needs, the construction period and investment cost. Particular attention is given to the Gen. IV reactors, which are bringing new solutions and aspects in the nuclear panorama. One of the goals is the practically elimination of severe accident addressed by approaches and phenomena, which are considered during the design of SMRs and MMRs. For this reason, the SMRs development is conditioned by a portfolio of technological solutions, which are currently in preparation, and by the readiness of the supervisory authority to legislatively regulate the construction, operation and supervision.

One of these concepts is Energy Well (EW), an MMR developed in Centrum Výzkumu Rez (CVR) located in the Czech Republic with support of a national consortium, which includes universities and other Czech companies. Especially, CVR uses its experience of more than 60 years in running research reactors, which has several similarities with MMRs, to design EW and to build several facilities useful to assess the molten salt technology. Several activities are ongoing in a framework of a project financed by the Technologická agentura České republiky (TAČR).

Part of these activities focuses on thermohydraulic assessment and safety analyses, which are performed by CVR and Nuclear Research Institute (UJV) in a parallel benchmark activity. The codes used in these activities are TRACE 5 patch 7 developed by Nuclear Regulatory Commission (US NRC) and RELAP5-3D developed at Idaho National Laboratory, which offer several working fluids represented in different equation of states. The aim of the analyses is to provide the results of a benchmark, particularly, assessing the capability of TRACE and RELAP to analyze molten salt technology. Additionally, the EW thermohydraulic will be investigated starting from the steady state and reaching a transient analysis. Particular attention will be given to the point kinetic module, which is introduced and implemented into the models to study the core behavior under extreme accident conditions.