bluuebunny/crossref_metadata_2025 huggingface dataset:
Properties of Irradiated Additively Manufactured 316L Steel for In-Core Applications – Mechanical and Microstructural Analyses
Authors: Jedlan Štěpán , Ševeček Martin , Hodek Josef , Prantl Anotnín , Šoltés Jaroslav , Fedoriková Alica , Sovadina Michael
Published: November 2024

This paper focuses on separate aspects of the utilization of additive manufacturing (AM) for the in-core applications in the nuclear power industry. This study investigates the application of the laser direct energy deposition method (L-DED) to produce in-reactor metallic components within LWR nuclear reactors. 316L stainless steel with heat treatment was chosen as the material of interest for experimental evaluation of the material properties after the neutron irradiation. Samples for the tensile test at 300 °C and microstructural analysis (LOM, EBSD, SEM) were irradiated inside a drop-in capsule in the LVR-15 research reactor at Research Center Rez. Subsequently, the postirradiation examination (PIE) results are compared to those of non-irradiated counterparts to analyze the impact of in-reactor conditions. Furthermore, the non-irradiated samples, both conventional and AM, were tested at both the Research Center Rez and COMTES FHT a.s, to obtain independent results and validate the methodologies.

To design the irradiation campaign, the calculation of residual activity after the irradiation period is performed using a SCALE module ORIGEN to inform future handling procedures and scheduling of post-irradiation examinations. The calculation of expected displacement per atom (dpa) is also included for direct comparison with results available in the literature.
Research and Preliminary Verification of the Resonance Self-Shielding Calculation Method for the High-Fidelity Neutronics Code VITAS
Authors: Yin Han , Zhang Tengfei , Liu Xiaojing
Published: November 2024

VITAS is a high-fidelity neutronics code developed by Shanghai Jiao Tong University based on the heterogenous variational nodal method (HVNM). In previous publications, VITAS has been successfully verified for transport calculations against macroscopic benchmarks. However, advancing the research on the resonance self-shielding calculation method within VITAS remains crucial for efficient one-step neutronics calculations. This paper presents the implementation of the subgroup method in VITAS. Subgroup parameters are obtained through least squares fitting based on the resonance cross-sections table. The narrow resonance approximation elastic scattering model is employed for the scattering source term, and a subgroup transport equation is established. The subgroup transport equation is solved using the HVNM, yielding subgroup fluxes. Effective resonance self-shielding cross-sections are obtained using the subgroup fluxes to weight the subgroup cross-sections. Additionally, the Bondarenko method is applied to address multiple resonance nuclide interference effects. This paper conducts preliminary verification of resonance self-shielding calculation method through a series of benchmark problems. The numerical results suggest that the effective resonance self-shielding cross-sections and eigenvalue obtained from VITAS show good agreements with the reference solutions provided by Monte Carlo code OpenMC. This consistency demonstrates the capability of VITAS in addressing heterogeneous resonance self-shielding problems.
Enhancing Image Quality in Limited-Angle CT Reconstruction for Reactor Pressure Pipelines Detection
Authors: Fu Jintao , Chang Jiahao , Sun Yuewen , Zeng Tianchen , Tang Peng , Xu Shuo , Cong Peng
Published: November 2024

The pressure pipelines of a nuclear reactor play a pivotal role in ensuring the structural safety of the reactor, and the accurate detection of internal and surface defects constitutes a critical step in safeguarding its integrity. Despite the remarkable performance of CT technology in detecting surface and internal defects, the inspection of reactor pressure pipelines faces significant challenges due to the limitations imposed by limited-angle CT scanning. This paper is dedicated to the investigation of limited-angle CT reconstruction algorithms, conducting an in-depth analysis of traditional iterative reconstruction algorithms such as SART and ADMM-TV. The primary objective is to enhance the image quality by addressing the artifacts caused by the limited-angle acquisition in the detection of reactor pressure pipelines. Experimental results from CT detection of reactor pressure pipelines indicate conspicuous artifacts under limited-angle conditions. Although iterative reconstruction algorithms like SART and ADMM-TV contribute to alleviating some of these artifacts, our proposed DRP reconstruction algorithm demonstrates a more pronounced improvement in reconstruction quality. Through the refinement of limited-angle CT reconstruction algorithms, this paper provides a preliminary exploration of non-destructive testing and reconstruction for reactor pressure pipelines, offering valuable insights to drive technological advancements in the field. The study aims to serve as an important reference for promoting the development of techniques in reactor pressure pipelines detection.
The Effect of Liquid and Wall Properties Toward the Increase of Wetting Velocity in the Case of Using Multiple Plates in the Top-Reflood Vertical Surfaces
Authors: Akbari  , Umebayashi Hiroyuki , Okawa Tomio
Published: November 2024

Rewetting is an important mechanism in the quenching process, particularly in the case of nuclear reactor safety. During the Loss-of-coolant accident (LOCA) in the Emergency Core Cooling System (ECCS) of the nuclear reactor. In the location near the wetting front, heat transfer coefficient (HTC) increases, and the quenching rate also occurs rapidly. The mechanism of wetting velocity on the case of 2 (two) plates is assumed to be different compared to 1 (one) plate.

The test section consists of two metal plates which are used as the heat transfer surface. Each of up and down edges of the metal plates (brass or copper) was sandwiched by the brass block columns. The block columns are set to heat the plate until the temperature of 250–300 °C. Meanwhile, distilled water is sprayed through the nozzle onto one or two plates. Furthermore, wetting velocity is calculated using the video captured by infrared camera FLIR™ to observe the liquid motion on the heated wall.

Several parameters are being observed, including the effect of mass flux, initial wall temperature, thermal conductivity, wall thickness, and subcooled degree. The results show the significance of droplets impact for cooling the wall, thus contributes to higher wetting velocity in the case of multiple plates.
Design, Development and Testing of Large Diameter High Temperature Alkaline Metal Heat Pipe
Authors: Zhang Jiarui , Wang Chenglong , Tian Zhixing , Guo Kailun , Zhang Dalin , Tian Wenxi , Qiu Suizheng , Su Guanghui
Published: November 2024

Heat pipe is the most critical heat transfer element in the silent heat pipe reactor, which utilizes the phase change of the work material to carry out heat transfer work through gravity, centripetal force and capillary force, and works stably under high temperature conditions with high heat transfer efficiency. Large-diameter high-temperature heat pipe relies on a larger cross-sectional area of the vapor chamber and a larger amount of liquid filling, so that the heat transfer limit of the high-temperature heat pipe is greatly improved, increasing the heat transfer capacity of the heat pipe under the same length. In this paper, the design optimization and development test of large diameter high temperature heat pipe is carried out, the heat pipe shell material is stainless steel 310S, the length is 2 m, the diameter is 89 mm, and the wall thickness is 5 mm. The heat transfer medium is sodium and potassium. The filling of the heat pipe is carried out in the glove box using the hot state filling method to ensure the purity of the work material. The manufactured heat pipe is placed vertically for performance testing, utilizing electromagnetic induction to heat the evaporator section, and the condenser section is cooled under air-cooling conditions. It is found that the wick has a greater impact on the heat transfer performance of the heat pipe. Heat pipe #1 is a thermal siphon without the wick, the maximum heating power is 10 kW, due to the intermittent boiling of the large liquid pool at the bottom of the test process led to the existence of a violent vibration. The selection of the wick also has a great impact on the heat transfer performance of the heat pipe. Heat pipe #2 and heat pipe #3 are capillary core heat pipes with different mesh numbers of stainless steel wire mesh as the wick. The maximum electromagnetic power of heat pipe #2 is 12.3 kW, the cooling power is 3.72 kW, the maximum temperature of heat pipe is 903 °C, and there is local high temperature in the evaporator section. The maximum electromagnetic power of the heat pipe #3 is 17 kW, the cooling power is 6.11 kW, the maximum temperature is 826.1 °C, the heat transfer power is higher, and the overall temperature of the heat pipe is flattened, and no overheating phenomenon occurs. Compared with the small diameter heat pipes, the temperature fluctuation during the startup and steady state of the large diameter heat pipes is greater, and the temperature fluctuation in the condenser section reaches ± 10 °C, and the entrainment phenomenon is more obvious. However, relying on a larger cross-sectional area of the vapor chamber and a larger amount of liquid filling, the heat transfer power of large diameter heat pipe is higher.
Coalescence-Induced Irregular Particle-Droplet Removal
Authors: Li Yanzhi , Du Jiayu , Min Qi , Sun Libin
Published: November 2024

In this work, we experimentally investigated the coalescence-induced irregular particle-droplet removal phenomenon. We chose graphite and iron particles, common contaminant dust in HTR-PM, with large density and irregular (non-spherical) shapes at the sub-millimeter scale. It is found that the coalescence mechanism achieves excellent performance in droplet jumping enhancement and particle removal. The maximum transport height of graphite particles was 0.55mm, and the maximum transport height of iron particles was 0.40mm. It is revealed that since the particles always sunk at the bottom of the droplet, their maximum transport height was related to the droplet departure velocity. Besides, irregular particles enhanced the droplet departure velocity. The droplet departure velocity with all graphite particles and part of iron particles was higher than that without particles. The maximum dimensionless departure velocity is 0.35 (graphite) and 0.29 (iron), 40% and 16% higher than droplet jumping with no particles. Notably, it is illustrated that the mechanism of droplet jumping enhancement with particles was similar to that with macrostructures, which reduced the viscous dissipation and improved the velocity consistency of droplets by shortening the duration of the merge stage and prolonging the duration of the contact stage. This study provided an efficient method for irregular particle removal and opened up a new possible research direction for dust removal and aerosol removal in nuclear industry applications.
High-Fidelity Depletion-Pebble-Flow Coupling Simulation of Pebble Bed Reactor HTR-PM
Authors: Li Ruihan , Liang Jingang , She Ding , Zhang Weijian
Published: November 2024

The high-fidelity depletion and pebble-flow coupling simulation of pebble bed reactor is realized based on Monte Carlo (MC) and Discrete Element Method (DEM). An HTR-PM full scale model with 420,000 fuel pebbles is built using the open-source MC code OpenMC, and the pebble trails are obtained with the open-source DEM code LAMMPS. The OpenMC code is modified so that the model geometry can change from one depletion step to another, and thus can be coupled with pebble-flow calculation. To obtain the high-fidelity results, each fuel pebble is specified to be a single burnup-zone. The pebble-wise power distribution and nuclide composition results are obtained and agree with the physical expectations. The model is depleted for 39 steps with each step representing 2 days. The time-consumption is 2.5 hours per step with a neutron history of 50,000,000 and a parallel scale of 40 MPI processes and 28 OpenMP threads per process. The memory-consumption is 40 GB per process.
Research on Long Term Trend Prediction of Nuclear Power Plants Based on Integrated Framework
Authors: Tan Canyi , Liang Biao , Wang Bo , Li Jiangkuan , Han Rui , Tan Sichao , Tian Ruifeng
Published: November 2024

In the current field of nuclear energy, the development of the third generations nuclear power plants (NPPs) are the mainstream, and its safety and economy are the important direction of nuclear energy development. Therefore, the normal, stable and safe operation of the nuclear reactor system is particularly important. In order to improve the safety and stability of the nuclear power system, this paper proposes a life prediction model of a NPP based on a comprehensive framework. In this study, an integrated framework based on BP, LSTM, Informer and other models is built, and the model is verified by the running data of Qinshan simulator. The results show that, compared with the traditional BP and LSTM, the long-term trend prediction model of NPP based on integrated framework has lower MAE and MSE values, indicating that the long-term trend prediction model of NPP based on integrated framework has lower error, which provides strong support for the safe and stable operation of nuclear power system.
Reconstruction of 3D Nuclear Graphite Based on Generative Adversarial Neural Networks With Gradient Penalty
Authors: Peng Lei , Liu Zhiyong , Wang Pengyu , Liu Hao , Zhang Huang , Yin Huaqiang , He Xuedong , Ma Tao
Published: November 2024

Nuclear graphite is an important structural material in reactor, and it is important to study the dehumidification method to improve the economy of reactor. In order to accurately obtain various physical property parameters of nuclear graphite from the pore scale, constructing a 3D digital graphite that can accurately characterizes the pore structure of micropores and nanopores is crucial. Due to the high experimental cost or a long reconstruction time of tradigital reconstruction methods, we use the generative adversarial neural networks with gradient penalty (WGAN-GP) to reconstruct 3D digital graphite. Taking binary image of graphite as training sample for model training, the generator model that can generate 3D graphite images is obtained, and then multiple graphite images are obtained. The generated images and original images were compared and analyzed from two-point statistics, pore distribution, porosity and specific surface area. The results show the reconstructed graphites are consistent with the original graphite, which proves the reliability of this method.
Numerical Simulation of Condensation-Induced Water Hammer With Dynamic Valve Control Using a Phase Change Model
Authors: Zhao Danfeng , Zhang Zihao , Chen Xiaoxia , Tu Ruiyang , Zhao Jiaming , Xiong Feng , Chen Zhengyu , Guo Wentao , Wang Shengfei
Published: November 2024

This article focuses on the phenomenon of complex water hammer that occurred during the testing period of the steam pipeline in a nuclear power plant. Computational fluid dynamics (CFD) techniques are utilized to simulate the transient behavior of condensation-induced water hammer (CIWH). An phase change model is implemented to accurately capture the phase transition process and its impact on the system dynamics. A dynamic valve control function is implemented to investigate the impact of valves on CIWH, allowing for a comprehensive analysis of their influence. The simulation results are compared with the actual operating conditions in the steam pipeline for validation. Overall, the numerical simulation approach established in this article effectively reproduced the process of complex water hammer, laying the foundation for subsequent model optimization and triggering mechanism research.
Experimental Study of Rectangular Channel Heat Transfer With Buoyancy-Aided Flow Under Asymmetrically Heating
Authors: Ji Yongan , Guo Zehua , Ding Ming , Sun Zhongning
Published: November 2024

Many nuclear power facilities use natural air circulation as part of their non-energetic safety mechanisms to remove heat. Both forced and spontaneous convection are frequently involved in the flow during this cooling process in vertical channels. The flow is referred to as buoyancy-aided flow when the direction of forced convection and natural convection coincide. The mixed convection heat transfer coefficient is lower than the forced convection heat transfer coefficient, and this type of flow will be in a specific buoyancy range. The supercritical fluid and water as the working medium, four-sided test mode heating, asymmetric heating, and air as the work material of the experimental research are the main topics of current research on buoyancy-aided flow. Thus, the investigation of mixed convection under buoyancy-aided conditions is presented in this study, along with a comparison of the experimental findings with the current correlations and an analysis of the causes of the differences. Lastly, a modified Buoyancy number with an average deviation of +20% is offered as a new correlation. This will serve as a point of reference for the relevant mixed convection heat transfer calculation.
Methodology for Acquiring Reliability Data in Novel Nuclear Energy Systems Using a Non-Homogeneous Poisson Process
Authors: Li Zefeng , Chen Pu , Liu Tao
Published: November 2024

In nuclear power plant maintenance, Reliability-Centered Maintenance (RCM) increasingly focuses on preventing equipment failure. This is particularly important in novel reactor models where mature data for maintenance decisions is limited. Bayesian Networks (BNs) are employed to fill this gap, modeling variables and their dependencies to assess component impact on system failure rates. Innovative approaches like Bayesian updating with non-homogeneous Poisson processes (NHPP) for small samples can modeling failures more accurately. This article utilizes the Markov Chain Monte Carlo (MCMC) method for Bayesian updating of component reliability data and employs BNs to analyze the reliability parameters of first of a kind equipment. These methodologies enhance the safety and efficiency of nuclear power plants by ensuring maintenance decisions are informed by existing and comprehensive reliability data.
Key Technologies Research for the Implementation of In-Service Inspection Strategies Optimizing for the Reactor Coolant System of In-Service Nuclear Power Plants
Authors: Shen Yunhai , Deng Feng , Duan Yongqiang , Yu Xiaoquan , Ye Shuixiang
Published: November 2024

The in-service inspection code for the M310 nuclear power plant in China is RSE-M. According to the standard, a periodic hydrostatic test of reactor coolant system is required every ten years, which reduces the economic efficiency of the plant. Based on current research results, the integrity of the pressure boundary of the reactor coolant system can be verified through system leakage testing after refueling outage combined with appropriate in-service inspection strategies. This paper is based on previous research results and focuses on the application research of periodic hydrostatic test and in-service inspection strategy optimization for a nuclear power plant. Key technical research during the application implementation process has been carried out and completed, including the applicability analysis of ASME standards and related regulations, the analysis of pressure boundary differences between system leakage test and hydrostatic test, and the mechanical analysis of nuclear safety class 1 pipelines, and analysis of the effect of related system equipment. Finally, the application and implementation of the optimization of the in-service inspection strategy for the reactor coolant system of the M310 nuclear power plant was completed.
Direct Numerical Simulation Study on Turbulent Flow and Heat Transfer Behavior of Helium-Xenon Gas Mixture in a Round Pipe
Authors: Zhang Ruini , Deng Jian , Wei Zonglan , Lu Qi , Han Wenbin
Published: November 2024

Helium-Xenon (He-Xe) gas mixture is widely adopted as the working fluid for space gas cooled nuclear reactors with CBC engines. The Prandtl number of He-Xe gas mixture in the preferred mixing-ratio range is approximately at 0.2–0.3, which is significantly lower than air or water. To better reveal the flow and heat transfer mechanism of He-Xe gas mixture, 2 direct numerical simulation (DNS) cases of He-Xe gas mixture flowing through a heated pipe were performed. One is based on the experiment case (RUN696) of Taylor (1988), and another one is set with a higher temperature range. The DNS results are in good agreement with the experiment data. The Taylor Nu heat transfer correlation was evaluated by DNS results. Further analyses of other turbulent statistics distribution, such as Reynolds stress, and turbulent heat flux were conducted. Reynolds-Averaged Navier-Stokes (RANS) simulations with 6 different turbulent models of the same case were also performed. The DNS data was used to evaluate the calculation results of different RANS turbulent models. The result shows that the three k-ε and the linear pressure-strain RSM models perform barely the same, and the SST k-ω model relatively behaves the best. It is also found that the current default turbulent Prandtl number model (Prt = 0.85) deviates largely from the calculated DNS data, especially in the near wall region.
Remote Impact Acoustic Inspection of Structures Using Water Jet Impacts
Authors: Tokuomi Saeko , Mori Kazuya , Ohshima Yasutaka
Published: November 2024

We have developed an advanced method for remotely detecting defects in concrete structures using pulsed water jets, offering significant improvements over traditional continuous flow impact echo testing. Unlike the conventional approach, our method utilizes intermittent, high-pressure water jets, which enable more distant and efficient inspections while minimizing water consumption. Notably, the pulsed nature of the water jets makes the testing process less susceptible to wind interference and broadens the excitation frequency range, potentially enhancing defect detection capabilities. In our experiments, we targeted concrete specimens embedded with artificial cavities to simulate common defects, employing a 4 mm diameter nozzle at 1 MPa pressure to release water pulses for 0.01 seconds. The testing distances extended up to 5 meters, demonstrating the method’s effectiveness in detecting 300 mm diameter cavities from a significant range. Defect identification was facilitated by analyzing the acoustic impact signatures through spectrogram visualization, focusing specifically on frequency bands resonating with the defects’ natural frequencies.
Research Methods and Protective Measures for Internal Flooding Analysis of Nuclear Islands
Authors: Du Guang , Zhao Danfeng , Lin Daping , Chen Xiaoxia
Published: November 2024

Internal flooding of the nuclear island in nuclear power plants is a common mode failure of systems and equipment, and the potential hazards must be considered during design. The protective design for such hazards should ensure the realization of various safety functions of the power plant system. Based on the analysis and research on internal flooding of the nuclear island in third-generation nuclear power units, this article follows the standards and specifications of advanced pressurized water reactors, draws on engineering design experience of nuclear power plants, and studies the analysis methods and protective measures for internal flooding of the nuclear island. Firstly, it determines the background and evaluation criteria for internal flooding protection design in nuclear power plants, compares the recommended flooding hazard evaluation methods in the standards, then analyzes in detail the steps and process of screening internal flooding sources, analyzes the spreading path of water sources and water level calculation methods, and conducts risk analysis on protective equipment within the subarea, so as to determine protective measures. The analysis results are given for the safety objectives of nuclear power plants.
Effect of Different P/D Ratios on the Thermal-Hydraulic Performance of Fuel Assemblies in a Lead-Cooled Fast Reactor
Authors: Zhang Yangguang , Lu Daogang , Cao Qiong , Wang Xiaotian
Published: November 2024

The core is the area of the nuclear reactor where the fission reaction occurs, therefore, it is necessary for a complete thermal-hydraulic study to offer vital research data to assure the safe and stable functioning of the nuclear reactor. Due to the large number of fuel rods in each box and the high average heat flux density on the surface of the fuel rods, the ratio of the pitch to the diameter of the fuel rods (P/D) and the width of the side channels of the fuel assembly have a large influence on the uniformity of the coolant temperature at the outlet of the fuel tank in lead-cooled fast reactors. The thermal and hydraulic calculations of the fast reactor fuel assemblies are carried out in this paper based on the parameters of the fuel assemblies of the small and micro lead-cooled fast reactor SMILE, and the feasibility of the numerical calculation method is verified. Second, three models with P/D of 1.15, 1.20, and 1.24 were created, and sensitivity analysis was performed for the width of each model’s opposite side channel. The findings demonstrate that as P/D increases, the coolant temperature uniformity at the fuel box’s outlet improves significantly. Additionally, when P/D = 1.24 and the side channel’s width is 1.95 mm, the fuel assembly’s outlet temperature is effectively flattened, increasing the thermal safety margin.
Design Calculation of Ultra-High Temperature and High Pressure Helium-Xenon Mixed Gas Radiator
Authors: Xu Yongwang , Duan Minghui , Wei Junhan , Zhao Minfu , Zhang Dongxu , Li Qingyuan
Published: November 2024

Helium-xenon mixed gas has higher heat transfer performance under a certain mixing ratio than pure helium. There are great limitations and uncertainties in applying existing theories and empirical models of helium-xenon mixed gases to ultra-high temperature and high pressure ranges.

Therefore, it is necessary to build a loop and conduct experiments to verify the flow heat transfer characteristics of helium-xenon mixed gas under ultra-high temperature and high pressure conditions. The working conditions requires the radiator to have good heat dissipation efficiency under different pressures, temperatures and flows. Therefore, in this loop design, the radiator is a key device. So, it is necessary to perform thermal calculations before determining the structure of an ultrahigh temperature and high pressure radiator.

Two radiator structural designs are currently proposed. Method of computational fluid dynamics (CFD) were used to study two radiators with different structures. Calculations are conducted for typical stable operating conditions. The calculation results show that under the same working conditions, the radiator designed with “fin - low temperature helium-xenon mixed gas” has higher heat dissipation efficiency and is an ideal design choice for ultra-high temperature and high pressure radiators. In addition, its performance under different operating conditions is also studied.
Evaluation and Sensitivity Study of Uniform Kernel Method in Atmospheric 41Ar Dispersion Against the Belgian Field Experiment
Authors: Sun Yeshuai , Zhang Yujie , Wang Zhaoyang , Yang Li , Li Xinpeng , Chen Yixue , Dong Xinwen , Fang Sheng
Published: November 2024

The Uniform kernel is an effective kernel density estimator used in the Lagrangian Particle Dispersion Model (LPDM), estimating the dispersion trajectory of radionuclides, and determining radioactive concentration to assess environmental contamination levels. This method significantly decreases computational resources compared to the traditional box counting concentration calculation in LPDM, serving as a fast and accurate tool to ensure effective nuclear emergency response in the early stages. In this paper, the effectiveness of the Uniform kernel is evaluated systematically via the measurements from eight gamma dose rate monitoring sites deployed at the Belgian field experiment by coupling the California Meteorological Model (CALMET) and the Lagrangian Particle Model (LAPMOD). Sensitivity studies are performed on three important parameters, including bandwidths multiplication factors, horizontal grid size, and particle numbers. The results demonstrate satisfactory agreement between simulations and measurements, with acceptable statistical metrics. The sensitivity analysis highlights the significant impact of bandwidths multiplication factors on the Uniform kernel, summarizing appropriate parameters for applying the Uniform kernel in small-scale dispersion for the nuclear power plant scenarios.
Neutronics Research on Annular Axial-Non-Uniform Minor Actinides Transmutation Targets in Commercial Pressurized Water Reactor
Authors: Ye Bin , Ala Ayodeji A.
Published: November 2024

This paper introduces the thermal-fast neutron conversion material 6LiD to improve the transmutation efficiency of MAs in commercial PWR, especially the fission rate, and investigate the feasibility of using annular axial-non-uniformly transmutation targets to flatten the axial power distribution of the core and optimize the temperature reactivity effect of the moderator at high boron concentration. In this research work, four annular axial-non-uniform transmutation targets are designed and they have the same geometry structure, but axially have different MA concentration in the transmutation material which composed of MA, 6LiD and UO2. The annular transmutation targets are axially three, five and seven segments respectively, and MA concentration in the transmutation material gradually decreases from the middle to both ends. The transmutation performance of these targets is investigated while its parameters change, such as the thickness of the annular transmutation rod and the composition of the transmutation material. The results show that, compared with axially uniform rods, the annular axial non-uniform transmutation targets have high transmutation efficiency, especially the fission rate. The transmutation effect is the best while the mass ratio of 6LiD to MAs in the transmutation material is 1:9; MA fission rate of these transmutation targets can reach up to 5.05%. At the same time, these targets have good power flattening effect, especially axial power flatten results that they could reduce the axial power peak factor of the core from 1.778 to 1.388. Meanwhile, they can also optimize the temperature effect of the moderator well at high boron concentrations.
An Improved Multiphase Lattice Boltzmann Method Scheme for Large Density Ratio Simulation in Heat Pipe
Authors: Quan Lie , Ma Yugao , Liu Xiao , Huang Shanfang , Wang Kan
Published: November 2024

Heat pipe reactor is a cutting-edge energy power technology for future deep space and deep sea exploration, its current bottleneck is the development of high performance heat pipe, and the wick structure is the key factor affecting the performance of heat pipe. To promote research on flow and heat transfer mechanism inside heat pipe, an improved scheme of Multiphase LBM is introduced, which not only eliminates error caused by inconsistency of discretization scheme of pressure term when introducing non-ideal gas equation of state, but also the cubic term error of velocity field in the recovered Navier-Stokes equation through LBM. With this improved scheme introduced, results of better accuracy can be obtained, so as its application into permeability prediction of porous media. An ideal Kelvin structure is constructed according to the production process and geometric parameters. The lattice Boltzmann method was used to directly simulate the three-dimensional flow inside Kelvin structured metal foam on the pore scale, and the permeability are measured through numerical experiments based on Darcy’s Law.

When solving the multiphase flow problems such as gas-liquid flow with multiphase Lattice Boltzmann method (LBM), the gradient of the macroscopic quantity is required to calculate the interaction forcing term between the neighboring fluid particles. It is found that unphysical numerical oscillations would occur in the vicinity of the phase interface. Through analysis and test, this numerical oscillation is found to be related to the discretization scheme of gradient operator between the phase interface, and will lead to numerical instability, which limits the simulation ability of LBM in dealing with the two-phase interface problem with high density ratio. Simulations of flat interface problem and liquid droplet system are made to confirm better performance of this optimized forcing scheme, and results of higher resolution of the flow field was obtained, which are in good agreement with analytical solutions and experimental results.
TRIPOLI-4® Monte Carlo Code Verification and Validation Using T4G Tool
Authors: Lee Yi-Kang , Hugot François-Xavier
Published: November 2024

TRIPOLI-4® is a general-purpose Monte Carlo radiation transport code developed at CEA Paris-Saclay center. It uses continuous-energy nuclear data to simulate neutron, photon, electron and positron transport in fields like radiation protection and shielding, fission reactor physics, nuclear criticality safety, and nuclear instrumentation. The verification and validation assessments of TRIPOLI-4® input and output (I/O) data are one of the fundamental processes to assume the accuracy and credibility of numerical studies for its nuclear engineering applications.

T4G is an interactive tool for the TRIPOLI-4® I/O data display. It is a graphics viewer displaying 2D and 3D plots of geometry, materials of volume-cells, radiation source, collision sites projection of particle transport in space and energy, particle trajectories and leakage sites. It also provides iso-importance curves for variance reduction study, tally results display including iso-dose and iso-reaction rates curves, and volume-cell numerical values. T4G tool relies upon the same geometry module as TRIPOLI-4® code and therefore displays exactly the geometry that TRIPOLI-4® uses for neutron and photon tracking.

In this paper, we first introduce the TRIPOLI-4® code and its verification and validation (V&V) study. Secondly, we present the basic features of T4G display tool. Then we demonstrate and discuss advanced application cases summarizing the real practices of T4G tool for verifying and studying of TRIPOLI-4® I/O data in diverse fields of nuclear engineering.
Experimental Study on the Influence of Spray Characteristics and Temperature Control on the Removal Efficiency of Iodine Aerosols
Authors: Tang Jiaxuan , Yang Yang , Zhang Luteng , Li Jialong , Pan Liangming , Liu Zhuo , Gao Li , Yuan Yidan
Published: November 2024

Radioactive aerosol may be released into the gas and retained in the containment under severe nuclear accidents conditions. Spray system removal is one of the main ways to remove aerosols. Spray droplets and overpressure of primary loop in severe accidents create a complicated high temperature and high humidity environment. A significant portion of radioactive aerosols in nuclear accidents are soluble. These soluble radioactive aerosols (CsOH/CsI) in the containment of the high humidity environment will occur significant hygroscopic growth behavior and may dissolve in droplets. Moreover, if the spray solution is alkaline, changes in particle density, size, shape, and chemical composition of iodine aerosols may occur, significantly influencing the dynamic behavior of aerosol agglomeration, diffusion and deposition, and consequently, the aerosol removal efficiency of the spray system. While extensive research has been conducted on the spray removal of insoluble aerosols (TiO2\SiC), studies on typical soluble iodine aerosol spray removal experiments are largely lacking.

In this paper, the experimental study of single nozzle with different spray characteristics on the removal efficiency of iodine aerosol was carried out. This study provides insights into the effects of different spray pressures, flow rates, and pH values on the removal efficiency and influencing patterns of soluble iodine aerosols under various initial particle sizes and different aerosol compositions. Results indicate that, Alkaline spray water with different pH values had no significant effect on the removal efficiency of iodine aerosol. The aerosol removal coefficient in the 0.2μm and 0.5μm particle size ranges of iodine aerosols is the smallest, which is 0.0007s−1–0.0012s−1. Increasing the proportion of soluble aerosols in multi-component aerosols will increase the removal rate of spray aerosols.

This study can provide reference for fundamental understanding of the aerosol removal mechanism of soluble iodine aerosols by containment spray droplets and improving the containment spray removal models.
Generative Design of a Gas-Cooled Micro Reactor Based on Multi-Physics Analysis
Authors: Han Wenbin , Lu Qi , Deng Jian , Huang Shanfang
Published: November 2024

Additive manufacturing (AM) is a “game-changing” digital fabrication technology that can fundamentally alter the way products are designed and manufactured. In nuclear engineering, the application of additive manufacturing can enable advanced reactor designs with better performance, lower cost, and quicker deployment. To exploit the maximum potential of additive manufacturing, design for additive manufacturing (DfAM) is needed, DfAM can leverage the complexity freedom offered by additive manufacturing technology to break traditional design constraints and achieve enhanced performance. Generative design, one of the key technologies in DfAM, is an iterative exploration method using AI-driven algorithms to generate a series of design solutions that meet specified constraints.

In this work, a generative design platform for micro-reactors is developed based on DASSULT SIMULIA Isight. To simulate and analyze the multi-physics performance of the reactor design, various multi-physics simulation codes are integrated on this generative design platform, including RMC for neutron physics analysis, ANSYS Fluent for computational fluid dynamics analysis, and ANSYS Mechanical for structural mechanics analysis. A multi-objective optimization method is used to explore geometric parameters and fuel configurations to achieve better neutronics, thermal-fluidics, and structural mechanics performance.

The generative design solution is a 4 MW helium-cooled nuclear reactor which includes 3D-printed silicon carbide (SiC) matrix embedded with tri-structural isotropic (TRISO) particle fuel. Helium at 3 MPa is chosen as the coolant, and the cross-shaped coolant channels are distributed within the matrix to remove the heat generated by fission. According to the multiphysics analysis, the reactor has a initial keff of 1.0571, power peaking factor FQ of 1.50, and operating lifetime of 1200 EFPD (Effective Full Power Days), with a maximum temperature of 1331 K and a peak thermal stress of 141.3 MPa.

The generative design method and platform provide a systematic and efficient approach for the design of additively manufactured micro-reactors, and the generative design solution demonstrates the feasibility and effectiveness of the proposed method.
Research on Dynamic Rod Measurement Method Based on Torch Software Package for PWR
Authors: Cai Yun , Li TianYa , Guo Rui , Liu Tongxian , Li Qing , Yang LingFang , Zhao Wenbo
Published: November 2024

Dynamic rod measurement method proposed by Westinghouse and Electric Power Research Center is an effective control rod worth measurement method, which has been applied widely. The main feature of dynamic rod measurement method includes the high accuracy and the short process of the corresponding rod worth measurement test. Torch software package developed by Nuclear Power institute of China can be used to analyze the neutron physics of PWR. Corca-K in the Torch software package is the tool that is used to solve the three-dimensional space-time neutron kinetics equations. A tool based on dynamic rod measurement method and Corca-K was developed, and verified by the control rod worth measurement test of Unit #1 of FangJiaShan and Unit #4 of FuQing power plant. The results showed that the developed tool was accurate, which the relative error of the different control rods between the calculated control rod value and the measured control rod value was less than 10%. The results showed that the developed tool could be applied to the measurement of the control rod value.
Research on Multi-Scale Coupling Program Based on Lead Bismuth Fast Reactor
Authors: Zhang Yixin , Wang Chenglong , Zhang Jiaxin , Tian Wenxi , Qiu S. Z. , Su G. H.
Published: November 2024

With the increasing emphasis on the refinement of reactor thermal hydraulic calculations, research on multi-scale coupled models of nuclear reactor thermal hydraulic has also developed. This article utilizes the self-developed lead bismuth system analysis program SACLER, combined with the mature commercial CFD software FLUENT, to develop a coupled program suitable for 1D/3D calculations of lead bismuth fast reactors. In order to ensure the accuracy of the developed coupling program, the calculation function of the developed coupling program was first verified through a simple circular tube case. The coupling program matches well with the data calculated by SACLER and FLUENT respectively, with temperature and flow rate errors within 10%. Subsequently, the response of the coupling program to the thermal hydraulic phenomena of the lead bismuth reactor under transient operating conditions was explored, and the transient experiment on the NACIE facility was simulated. The calculation results showed that the coupling program was in good agreement with the experimental results and could effectively simulate the changes in transient thermal hydraulic parameters.
Numerical Simulation of Single/Two-Phase Flow in a Stratified Porous Bed
Authors: Bouloudenine Aimad , Li Liangxing , Xiang Zutao , Shi Shang , Abu Bakar Muhammad
Published: November 2024

This study investigates the coolability of debris beds formed from Fuel-Coolant Interactions (FCI) within nuclear reactors, a critical safety concern when cooling systems fail, leading to significant core melting in light-water reactors. The research focuses on understanding the flow dynamics within two-layer stratified porous beds, which develop from the mixture of corium and coolant, particularly around the interface between layers. Using ANSYS Fluent software, numerical simulations were conducted to estimate velocity profiles and pressure drops across a stratified porous structure. The Euler-Euler two-fluid model was employed alongside the k-ω-based Shear Stress Transport (SST) turbulence model to separately model turbulence for each phase. The results reveal that lateral flow dynamics are significant as fluids traverse through the porous media, affecting pressure drops across layers with varying permeabilities. Specifically, the pressure drops across the lower-permeability layer decrease while that across the higher-permeability layer increases, highlighting the influence of lateral flows between layers of different particle sizes due to variations in porosity. This lateral flow is most likely formed during the initial stages of fluid movement through the bed. The study underscores the importance of understanding these flow characteristics for both single/two-phase flow to enhance the safety and design of nuclear reactors by improving debris bed coolability predictions.
Prediction of Critical Heat Flux of Wire-Wrapped Rod Bundle Based on Artificial Neural Network
Authors: Zhang Wei , Xiao Yao , Yu Lijun , Gu Hanyang
Published: November 2024

Accurate prediction for the critical heat flux of rod bundle assemblies is important for reactor safety assessment. However, the existing prediction methods fail to unify their generalization and accuracy. The theoretical prediction models are unable to provide a satisfactory estimation. while the empirical methods that have been proposed for several decades could give results precisely, their suitable conditions are strictly limited, causing the endless development of different correlations for every specific rod bundle design. An exploration of new methods has been undertaken to provide a better method for critical heat flux prediction based on an artificial neural network since neural networks have gained recent attention for their outstanding performance in predicting complex problems. This study discusses the prediction performance of a traditional polynomial fitting method, a neural network method, and a hybrid method combined with empirical correlations and a fully connected neural network regression model. The analysis concludes that when artificial neural networks replace empirical models in prediction tasks, achieving equivalent prediction accuracy is easily accomplished; however, there is a need for improvement in error characteristics. Besides, the hybrid method was validated as a superior approach, demonstrating its effectiveness in enhancing prediction performance across various input variable regions. The ratio of predicted values to true values is well-maintained, indicating that this proposed method yields improved error characteristics, and consequently, more accurate critical heat flux estimations. The research will help improve the feasibility of the application of machine learning algorithms such as neural networks in the nuclear energy industry.
Study on Flow Fluctuation Attenuation in the L-Shaped Pipe With Spring Vibrator
Authors: Wang Sipeng , Chen Jin , Kang Zehang , Yue Zhongcheng , Zhuang Kun
Published: November 2024

In view of the flow instability phenomenon that often occurs in passive residual heat removal system (PRHRS), the idea of using spring vibrator structure to attenuate flow and pressure fluctuations is proposed. Based on the finite element method, the fluid-solid coupling analysis is carried out in a typical L-shaped pipe which contains a spring vibrator structure. The results confirm the fluctuation attenuation effect of the spring vibrator structure. The simulation results shows that flow fluctuation amplitude at the outlet reduce, and the peak pressure which is the most threatening to thin-walled structures reduces. Based on the influence of relevant parameters on the attenuation coefficient, an optimized spring vibrator structure is obtained by changing the values of relevant parameters. Comparing the outlet flow fluctuation before and after the optimization of the spring vibrator structure, the flow fluctuation amplitude at the outlet is significantly reduced. The pressure fluctuation is greatly weakened compared with that before the optimization. The simulation results prove that the optimized spring vibrator has excellent fluctuation attenuation performance.