bluuebunny/crossref_metadata_2025 huggingface dataset:
Simulation and Analysis for Hualong Reactor Advanced Integrated Test Facility Based on RELAP5
Authors: Geng Yiwa , Zhan Jingxiang , Zheng Yuntao , Liu Jiatai , Huang Shuliang , Ni Si , Yang Changjiang
Published: November 2024

Hualong One and its subsequent rectors are the third-generation pressurized water reactor (PWR) and constructed by the China National Nuclear Corporation (CNNC) which possess completely independent intellectual property rights. China Nuclear Power Engineering Co., Ltd. (CNPE) designed and built an integrated test facility for the Hualong’s subsequent rectors emergency core cooling system (referred as Integrated Test Facility). The Integrated Test Facility is a large-scaled facility and designed based on the Hualong’s subsequent rectors’ safety systems, which adopts a passive safety design. The Integrated Test Facility includes main loop system, High-Pressure Makeup Tank (HPMT), Accumulator (ACC), Automatic Depressurization System (ADS), Outside Reactor Water Storage Tank (ORWST), Direct Vessel Injection (DVI) pipeline, Sump, Passive Residual Heat Removal System (PRS), Passive Containment Cooling System (PCS), etc. These systems and equipment work together to achieve passive emergency core cooling function. In this paper, a RELAP5 model is establish to analyze a small LOCA accident sequence. Simulations are conducted for five stages after the accident, including natural circulation, automatic depressurization, safety injection, and long-term sump recirculation. A 4MW core power (with operation pressure 15.5MPa, core inlet temperature 292.4 Celsius degrees, core outlet temperature 327.7 Celsius degrees) and medium pressure model was established. The emergency core cooling system operational characteristics were investigated under small break loss of coolant accident (SBLOCA). The simulation results demonstrated that the RELAP5 model can simulate the transient key system parameters accurately. This paper investigates the operational characteristics of the emergency core cooling system under different LOCA conditions, elucidates the interaction mechanisms of passive safety systems, and explores the physical mechanisms of thermal-hydraulic phenomena and processes in the integrated test facility. Based on numerical simulation analysis, the experimental data under accident conditions are obtained, thereby validating the rationality of the safety system design for the integrated test facility and the effectiveness of the accident analysis modeling method.
Research on Computation Method of Bubble Volume Based on Machine Learning
Authors: Ding Yongwang , Liu Xinxing , Meng Zhaoming , Zhang Han , Li Fu
Published: November 2024

In the case of low-speed jet flow and high undercooling degree, the condensation process in the reactor suppression system is often accompanied by low-frequency oscillations, namely the chugging phenomenon. The mechanism of oscillation is related to the formation, condensation, and collapse of bubbles. The accuracy of extracting bubble geometry parameters directly impacts the analysis of condensation oscillation mechanisms. The calculation of geometric parameters of bubbles using traditional image processing techniques often involves complex pre-processing, and the feature parameters need to be defined manually. There are some issues, such as significant user effects and long computation time. Therefore, based on traditional methods, this paper proposes a more efficient algorithm for extracting image feature parameters using convolutional neural networks (CNN). The results show that the performance of the newly proposed method under a single condition is satisfactory, and its generalization performance across different backgrounds is also good. The average additional error of extracted feature parameters is less than 2%, but the extraction cost can be significantly reduced.
Feasibility and Safety Analysis of Solid-State Reactor for Lunar Surface Exploration
Authors: Ma Yugao , Lu Jiahao , Liu Yue , Zhang Suyi , Zhang Muhao
Published: November 2024

The Fission Surface Power System is a feasible solution for future lunar surface exploration energy supply, with the solid-state reactor being considered a primary candidate. This paper presents a modeling and analysis of a solid-state reactor based on publicly available information.

Using the reactor Monte Carlo program RMC and the commercial finite element analysis software ANSYS, the reactor is modeled and analyzed. The power distribution of the active zone and reactivity feedback coefficients of the entire core are obtained. The result indicates a negative reactivity feedback coefficient, confirming this reactor design’s inherent safety. Furthermore, a two-dimensional reactor model is constructed to obtain the temperature distribution during full-power operation, and non-passive residual heat removal analysis and radiative heat loss failure analysis are performed. At the failure condition of all heat pipes, the maximum core temperature is 742 K, which is below the reactor safety temperature limit, confirming the residual heat removal capability of the solid-state reactor in the loss of heat sink. Additionally, the self-developed transient reactor system analysis program is used for thermal safety analysis. The thermal parameters and heat transfer limits of high-temperature alkali metal heat pipes within the reactor are obtained, with the ratio of heat transfer limit to the heat load of 2.05 demonstrating sufficient safety margins of the heat pipe operation in the reactor. The results manifest that the core is satisfied with the existing criteria for anticipated operating occurrences of solid-state reactors, verifying the excellent safety performance of this reactor design.

Based on the analysis and calculations in this study, it is concluded that this reactor design exhibits superior safety performance, which is attributed to the neutron physics design concept of the intermediate neutron spectrum and the thermal-hydraulic design concept of non-passive residual heat removal. This study contributes to the field of advanced nuclear reactor engineering by providing a comprehensive analysis of the solid-state reactor design for the Fission Surface Power System. The findings highlight the superior safety performance and potential application of this solid-state reactor design for future lunar surface exploration, which has important implications for the development of lunar surface reactors.
Seismic Response of NPP Considering the Non-Linear Soil-Structure Interaction Effects Using DRM
Authors: Jang Youngsun , Lee Youngoh
Published: November 2024

For improvement of seismic capacity of NPP, the Non-linear SSI (NLSSI) analysis concepts are studied and applied to practical NPP (Nuclear Power Plant) Structural Analysis. Nonlinear SSI analysis requires a different approach from existing linear and frequency domain SSI analysis. In order to consider the non-linear behavior of the ground and structural materials, the analysis must be performed in the time domain, but this is difficult due to a large analysis model and many repetitive analyzes. Accordingly, many researches are being actively conducted to introduce a hybrid method that sets a certain region as the near field, adds boundary elements, and conducts analysis in the frequency domain or time domain. DRM (Domain Reduction Method) is a method proposed by Bielak et al. by developing several theories for calculating effective seismic force, and is the most widely used technique when analyzing NLSSI using the sub-structure method. This method is also provided in LS-DYNA. In this study, the Korean NPP Structures are analyzed with ABAQUS program in the time domain for the NLSSI analysis. The Linear SSI analysis of the NPP Structure is performed with ACS SASSI program in the frequency domain at same time, and LSSI analysis result is compared to NLSSI analysis result. The final produce of NLSSI analysis is the FRS (Floor Response Spectrum) at each story of building. After NLSSI analysis, FRS will be provided for re-calculation of the HCLPF (High Confidence of Low Probability of Failure) of some critical equipment that contribute significantly to the CDF (Core Damage Frequency).
Experimental Study on Startup and Heat Transfer Performance of High Temperature Sodium Heat Pipe
Authors: Jiang Shunli , Zhou Huihui , Zhang Youjia , Yuan Dewen , Huang Shanfang
Published: November 2024

The screen-wick sodium heat pipe is a highly effective passive device for heat transfer that can operate under complex operating conditions. In this work, the startup process, steady-state heat transfer, and heat transfer limits of sodium heat pipe were experimentally studied and the influence law and mechanism of the heating power, inclination angle, and cooling condition on the heat transfer of the heat pipe have been discussed. The experimental results show that high heating power will cause overheating, and even local dry out, leading to the failure of heat pipe. The increase of inclination angle will lead to irregular change of heat transfer. Geyser boiling will occur within a heating power range under the positive inclination angle, resulting in the significant temperature oscillation. The increase of cooling capacity will improve heat transfer power, and the critical startup power of the heat pipe is basically in a linear relationship with the length of the condenser. However, the excessive cooling may cause the heat pipe to fail to start and the operation of heat pipe at low temperatures is restricted by the sonic limit. This research may provide valid experimental data support and references for the design optimization and operation of high temperature heat pipes.
Study on Laser-Controlled Surface Preparation and Condensation Heat Transfer Characteristics of Copper-Based Micro-Nano Structure Surface
Authors: Wang Zhikai , Chen Feng , Chen Jing , Huang Zhiyong , Yan He , Bo Hanliang , Dong Yujie
Published: November 2024

The micro-nano structure surface has significant influence on the condensation heat transfer. Studying and designing micro-nano structure parameters help to better explore methods for improving surface condensation heat transfer performance. In this paper, we firstly processed hill shaped microstructure with customizable structural parameters on the copper surface using femtosecond laser technology. Subsequently, copper oxide nanowire structures were prepared on copper surfaces with hill shaped microstructures utilizing thermal oxidation method by tube furnace. We have been conducted a series of condensation heat transfer experiments on the copper-based micro-nano structure surface. Further the condensation heat transfer coefficient on the surface of micro-nano structures were analyzed. Additionally, the dynamic behavior of condensation droplets was also investigated. The surface processing results show that the surface structure parameters can be well customized using femtosecond laser technology by adjusting the laser processing parameters and designing patterns. The experimental results of condensation heat transfer indicate that the surface heat transfer coefficient of the copper-based micro-nano structure has been significantly improved compared with smooth copper surface. It has been found that there is droplet jumping phenomenon at low subcooling, while flooding phenomenon occurs at high subcooling.
Thermo-Neutronics Coupled Simulation of a Heat Pipe Reactor Based on OpenMC/COMSOL
Authors: Nie Jingyu , Li Binqian , Wu Yingwei , Zhang Jing , Zhang Guoliang , Ren Qisen , He Yanan , Su Guanghui
Published: November 2024

As an advanced small nuclear reactor, the heat pipe reactor possesses several advantages, including high energy density, long operational lifetime, compact size, and strong adaptability to various environments, making it an optimal choice for specialized energy needs in future applications, such as deep-sea and deep-space domains. In this study, we developed a code system using OpenMC/COMSOL for neutron and thermodynamic simulations. The continuous-energy Monte Carlo code, OpenMC, was employed to generate homogenized cross-section databases, offering significant modeling flexibility compared to traditional deterministic lattice transport codes. The generated multi-group cross-sections from OpenMC were utilized in COMSOL for the coupled neutron and thermodynamic simulations of the entire core. To validate the OpenMC/COMSOL code system, benchmark problems for pressurized water reactors were computed, and the results of the “two-step” scheme for neutron physics were compared with full-core Monte Carlo neutron results. Furthermore, to investigate the applicability of the thermal-hydraulic coupling in the heat pipe reactor neutron physics model, typical heat pipe reactor assemblymodels were established and verified under various energy group numbers and homogenized regions. The results demonstrated good agreement for multiplication factors and power distributions. The research results indicate that the utilization of the cross-section library generated by OpenMC enables the capability for steady-state analysis and core design. The results obtained from COMSOL exhibit good overall agreement with respect to multiplication factors and power distribution.
Ultrasonic Doppler Velocimetry (UDV) for Cold Shock Transients in a Scaled Liquid Metal Cooled Reactor Plenum
Authors: Sieh Broderick M. , Bindra Hitesh
Published: November 2024

Liquid metal reactors experiencing low convective flux face a handful of issues that may deteriorate reactor safety. These long-standing issues, such as thermal stratification or thermal striping, can negatively impact the thermal behavior, mechanical systems, and neutronics of the nuclear reactor. To overcome these issues, safety analysis codes can be established. However, high-quality, validation grade experimental data are required for the development of such computational models, yet, such high-quality data are limited for low Prandtl number fluids such as liquid metals. To provide validation grade data for liquid metal, researchers at Purdue University have constructed the Gallium Thermal-hydraulic Experiment (GaTE).

The GaTE is an experimental facility designed to study liquid metal cooled reactor transient conditions within a 1/20th scale model of the Advanced Burner Test Reactor (ABTR) outlet plenum. Gallium serves as a surrogate for sodium in the facility as the metal also exhibits a low Prandtl number in liquid form. Additionally, gallium is not cost prohibitive, nontoxic, and non-reactive with air and water, unlike sodium. To produce the desired transient conditions within the scaled plenum, the GaTE is capable of controlling the plenum inlet jet velocity and pool temperature as well as the system circulation. Control of the flow throughout the system is permissible through use of a moving magnet pump. An electromagnetic flow meter (EMFM) is employed to provide flow rate feedback data. The temperature of the gallium entering the plenum is regulated via a heat exchanger utilizing hot oil on the secondary side. Similarly, the fluid exiting the plenum passes through a heat exchanger using cool water on the secondary to reduce the temperature of the liquid metal. Thermocouples monitor the temperature of the gallium throughout the loop. A fiber optic distributed temperature sensing system is positioned within the plenum to provide temperature measurements of the liquid metal pool.

Previously, the GaTE has been used to examine thermal stratification in a scaled liquid metal cooled reactor plenum based on flux Richardson number (Rif). These tests utilized the distributed temperature sensing system to characterize thermal stratification and the fluctuations associated with the stratification thermocline based on the proximity to the critical Rif. To relate the flux parameters to global parameters, flow velocity data within the plenum must be collected in addition to the temperature data. Ultrasonic Doppler velocimetry (UDV) is employed to obtain the desired data under cold shock transients in the scaled liquid metal cooled reactor plenum.
Load-Following Control of Nuclear Reactor Using Variable-Order Fuzzy Fractional PID Controller Based on the Fractional-Space Neutron Point Kinetics Model
Authors: Sun Hetao , Mengsitu Hailemichael Guadie , Luan Xiuchun , Zhou Jie , Wang Junling , Yang Zhida
Published: November 2024

With the increase in the proportion of nuclear power, the load-following capability of NPPs has increasingly become an important factor affecting the development of a system security economy. In this paper, a variable-order fuzzy fractional PID controller is applied for load-following control of a nuclear reactor based on the fractional-space neutron point kinetics (F-SNPK) model, which can show the dynamics of the nuclear reactor in a better way than the classical neutron point kinetics (CNPK) model. Three-group of delayed neutrons and reactivity feedback effects (temperature and poison feedback effects) are considered. The simulation results are compared with PID and PIλDμ controller and it is shown that the variable-order fuzzy fractional PID controller exhibits a better satisfaction and capacity during the load-following operations compared to the PID and PIλDμ controllers.
Quantifying Software Safety in Nuclear Power Plants: A Framework for Requirements Phase Assessment
Authors: Li Boyuan , Li Duo , Li Jianghai , Guo Chao , Xiong Huasheng , Zhou Shuqiao , Huang Xiaojin
Published: November 2024

With the increasing dependence on digital instrumentation and control (I&C) systems in nuclear power plants, software has become a significant determinant of system safety assurance. To expand the use of digital technology in the nuclear industry, systematic methods are required for quantifying the safety of software-based I&C systems in safety critical applications. A software safety quantification model limited to the requirements phase is built in this paper based on the causal mechanisms that challenge safety. A preliminary mathematical method was developed to assess the number of requirements faults, their sub-types, their hazard rates and a control gate that allows determination of when to move beyond the requirements phase. A case study is conducted on a function of a reactor protection system to verify the validity of the quantification model. The framework allows us to assess software safety quantitatively.
Dynamic Entropy Investigation of Feed Water Heater in Nuclear Power Plant During Start-Up Process
Authors: Zhang Zhijiang , Tian Zhaofei , Ma Xiaoyu
Published: November 2024

A dynamic simulation model has been established to study the entropy characteristics of the feed water heater start-up process, and the dynamic entropy analysis was conducted. The effects of step and continuous changes in extraction steam flow on the entropy characteristics of the start-up process were investigated. With the increase of the extraction mass flow rate, the entropy generation rate at steady state increased gradually. To study the effects of the continuous change in extraction steam flow, the entropy was characterized and analyzed with three types changes: the quick opening (QO), the linear and the equal percentage (EP). The optimal start-up method should have the EP extraction steam flow change characteristics with the smallest accumulated entropy generation (28.16 MJ/K), which was 32.01% less than that of the linear flow characteristic and 47.548% less than that of the QO flow characteristic. The proportion of entropy generation rate of each control volume under the design condition was obtained, and the proportion of steam condensing zone was the largest (46.08%). This study could provide important guiding significance for the start-up process of the feed water heater.
Investigating the Impact of Various Molten Salt Combinations on Reactor Criticality and Thermal Neutron Flux Distribution in SD-TMSR
Authors: Yilmaz Seda , Revankar Shripad , Xu Yunlin
Published: November 2024

Molten salt reactors represent a groundbreaking advancement in both advanced and conventional reactor technologies, leveraging fluid fuel/coolant salt for enhanced temperature control and thermal efficiency. In the pursuit of optimizing molten salt reactor efficiency, our research focuses on exploring diverse salt compositions within the structured framework of the SD-TMSR core geometry. This investigation maintains a fixed hexagonal lattice structure, distinguishing two core zones, including a crucial blanket region for thorium-based fuel breeding. The study systematically alters salt composition with various isotopes, specifically fissile (U-233) and fertile (Th-232) materials, providing critical insights into reactor criticality and thermal neutron flux distribution. The molten salts examined here include different combinations of materials listed as lithium (Li), fluoride (F), beryllium (Be), sodium (Na), zirconium (Zr), and rubidium (Rb). Nine molten salts were tested for their criticality performance, and the reactor reached criticality with only four of them. Therefore, the neutron flux distribution simulations were performed only with those salts. The primary purpose of the research is to compare the performance of the core reactivity with different molten salts and identify the most suitable salt for SD-TMSR. Through the broadening of our knowledge and understanding of the complexities in MSR design, this study emphasizes the crucial position of molten salt reactors in advanced next-generation nuclear technology, establishing them as robust and sustainable contributors to the global energy perspective.
Methodology Development for Explosion Hazard Evaluation in Hydrogen Production System Using High Temperature Gas-Cooled Reactor
Authors: Morita Keisuke , Aoki Takeshi , Shimizu Atsushi , Sato Hiroyuki
Published: November 2024

High temperature gas-cooled reactor (HTGR) is expected to use nuclear heat to wide range of industrial applications such as hydrogen production due to its high temperature heat supply capability and inherent safe characteristics. One of the key technologies is a safety evaluation method which can simulate an impact of explosion hazards occurred in the hydrogen production plant to safety functions in the nuclear plant because HTGR hydrogen production system contains large amount of combustible gases such as hydrogen.

A computational fluid dynamics code FLACS has been sufficiently validated for dispersion and explosion of combustible gases such as hydrogen and methane gas worldwide, however, only few attempts have been made for validation of analysis in closed area with small space. A leak of combustible gases to the piping in HTGR hydrogen production system may occur in case of abnormal condition in hydrogen production plant and therefore an explosion in the piping should be considered.

Two experiments on combustible gas explosion in piping are selected to compare with calculated results obtained by FLACS code. One is the deflagration of natural gas in a simple closed system consisting of two spherical vessels of different sizes and a thin straight piping (Experiment 1), and the other is the deflagration of mixture of hydrogen and methane gas in an opened complex piping including multiple branches (Experiment 2).

The comparison of calculated and experimental results showed good agreement for maximum pressure and temperature in Experiment 1. In Experiment 2, the maximum pressure and flame spread velocity in a section of straight piping from the ignition point to the first branch also agreed well in calculated and experimental results.
Remote Tapping Inspection Method Employing a String Shooter
Authors: Tokuomi Saeko , Mori Kazuya , Ohshima Yasutaka
Published: November 2024

Tapping inspections of structures are commonly performed through manual efforts by inspectors. However, in the case of nuclear structures, there may be situations where it is preferable to avoid close proximity to the objects under inspection. Moreover, there is a desire for the automation of these inspections. This study proposes a completely new tapping inspection method that enables inspections without the need to approach the object.

Our specific approach involves the use of a string shooter for tapping inspections. Known commonly as a toy that launches a looped string into the air using two rollers, we hypothesized that attaching a steel ball for impact to this string could enable remote tapping of the inspection object. Consequently, we attached a 4mm diameter stainless steel ball to the string of a string shooter and attempted to detect a circular void defect with a diameter of 300mm within a concrete test specimen measuring 600mm × 600mm × 175mm from a distance. When the specimen was randomly tapped from a 2-meter distance, a unique acoustic signal was generated above the void defect. By assigning colors to the impact points based on the intensity of the acoustic frequency spectrum, we were able to differentiate the area corresponding to the void defect from healthy areas by color, thus facilitating the identification of void defect areas. This method allows for tapping inspections to be conducted remotely from several centimeters to several meters away, and it is capable of tapping inspection surfaces in any direction, from horizontal to vertical. By mounting the inspection device on a drone, it also becomes possible to inspect high places.
Numerical Investigation of Convective Heat Transfer Performance of Hybrid Nanofluid Flowing Through Square Channel in PWR Fuel Rod Assembly
Authors: Oke Samuel , Gao Puzhen , Fadodun Olatomide , Jayeola Mathew A.
Published: November 2024

The invention of hybrid nanofluid provides a better cooling ability than the ordinary nanofluid. Its convective heat transfer performance and thermal conductivity foster a new opportunity for reducing thermal hydraulic challenges of nuclear reactor. This work investigates the effect of Reynold’s number (1.5 × 104 ≤ Re ≤ 2.8 × 104) and concentration of nanoparticle (0 ≤ VR ≤ 2%) on cladding wall temperature, average Nusselt number, pressure drop and performance evaluation criterion of copper-alumina hybrid water based nanofluid flowing through square sub channel of nuclear fuel rod assemblies in typical pressurized water reactor. The flow of turbulent hybrid nanofluid in square channel PWR nuclear fuel rod assembly was modeled using single phase and the k-ϵ turbulent models. The results are compared with existing correlations and give good agreement. The result shows that the addition of very moderate copper-alumina hybrid nanoparticles into water causes significant augmentation in heat transfer performance of about 40% than ordinary water. Also, copper-alumina hybrid nanofluid gives better heat enhancement ranging from 10% to 30 % compared to ordinary alumina nanofluid. with higher volume fraction. This implies that usage of hybrid nanofluid should be considered instead of mono nanofluid. since hybrid nanofluid has capability to increase the performance evaluation criteria (PEC) and lower the cladding wall temperature than water and mono nanofluid.
Numerical Investigation of Vapor-Liquid Direct Contact Condensation Heat Transfer Behavior Under the Spray Condition
Authors: Wang Yiwei , Wei Hongyang , Gao Puzhen , Tian Ruifeng , Tan Sichao
Published: November 2024

The direct contact condensation heat transfer between the liquid film formed by the nozzle and the saturated steam is a complex process, and many experiments and simulations have been done by predecessors. In this study, CFD was used to calculate the heat and mass transfer process of direct contact condensation between liquid film and saturated steam. The VOF model was used to obtain a clear two-phase interface. The fixed cone angle formed by the liquid film was used to improve the phase interface area density of contact condensation heat transfer between liquid film and steam in the heat and mass transfer model. The results are in good agreement with the previous experimental and simulation results. Comparing the numerical results obtained before and after the treatment of the interfacial area density, the variation of the temperature and thickness of the liquid film is basically the same, but there are some differences in the local heat transfer coefficient of the liquid film. It can be seen that the treatment of the interfacial area density has a great influence on the heat transfer coefficient of the liquid film.
A Novel Framework for Spatiotemporally Decoupled Source Reconstruction of Radionuclides Released Into the Atmosphere
Authors: Xu Yuhan , Dong Xinwen , Zhuang Shuhan , Hu Hao , Fang Sheng
Published: November 2024

In recent years, Europe has been confronted with a series of radionuclide leak incidents, emanating from sources that remain unidentified. Such incidents are typically assumed to originate from a single, static point source, allowing the source location to be pinpointed to a specific, time-unvarying coordinate. However, the unavailability of detailed information about these leaks, including crucial data like the operating status of involved nuclear plant, renders direct estimation of release details using a top-down approach unfeasible. Consequently, a bottom-up approach, relying on environmental measurements and atmospheric dispersion modeling, becomes the only feasible choice. This approach focuses on the spatiotemporal source reconstruction of atmospheric radionuclide release, aiming to optimize the source location and release rate by minimizing the discrepancy between simulations and measurements. Owing to the interdependence of these spatiotemporal source parameters, traditional source reconstruction methods often attempt to resolve both temporal and spatial aspects simultaneously. The resulting tremendous solution space may cause substantial estimation errors in both the source location and release rate.

To tackle this issue, this paper proposes a novel framework for spatiotemporally decoupled source reconstruction. This framework splits the reconstruction into the source localization and release rate estimation stage, reducing the solution space at each stage. Initially, an artificial neural network model is employed to map a relationship between the source location and several simulation-measurement error metrics. The error metrics for each source location are obtained based on the release rates calculating by the Tikhonov regularization method. Subsequently, a likelihood function is constructed based on these error metrics, followed by assessing the posterior distribution of the source location using the Bayesian statistics. Finally, based on the estimated source location, the release rate is calculated using the Tikhonov regularization method.

The proposed framework has been first validated using a European Tracer Experiment (ETEX) whose source is previously known. The source location error was less than 100 km and the estimated release rate profile match the truth well. Furthermore, this framework was applied to reconstruct the source of the 2017 106Ru leak incident, a case where the source remains unknown. The estimated source location and release amount are consistent with public literature. The successful application of this framework to both controlled experiment and real-world radionuclide leak incident underscores its potential as a key component of the nuclear emergency response system.
Remote Measurement by Robot Arm Equipped With PAUVP and LIBS
Authors: Ji Chengzuo , Chen Yuan , Brice Christian , Kikura Hiroshige , Takahashi Hideharu , Endo Gen
Published: November 2024

This paper presents a prototype of a system for remotely detecting leaks and determining the shape and elements of the sediment in the primary containment vessels (PCVs) of the Fukushima Daiichi Nuclear Power Plant, which is currently undergoing decommissioning. The accident resulted in breaches of the PCVs, causing the cooling water injected to cool the fuel debris to leak, along with relatively large sediments present in the leakage field in addition to small fuel debris. Up to now, the primary focus of the investigation is the internal structure of the vessel, leakage position, and the distribution of nuclear fuel debris within the reactor.

The paper applies the Phased Array Ultrasound Velocity Profile (PAUVP) and Laser-Induced Breakdown Spectroscopy (LIBS) is employed in this study to identify the elements present in the sediment above the water surface part. Additionally, an arm robot capable of carrying both PAUVP and LIBS was developed as a result of the radioactivity.

The experiments conducted in this study confirmed that the location of the leakage and the sediment characteristics were estimated from the visualization flow field map and the elements of sediment above water were identified, enhancing the efficiency of the detection process. The results showed that the PAUVP technique could accurately capture the velocity distribution and the leakage location in the simple flow field and that the LIBS technique could effectively determine the elemental composition of the above-water part. The arm robot was able to perform the measurements remotely and safely, avoiding human radiation exposure.

This study demonstrates the feasibility and potential of using the PAUVP and LIBS techniques combined with an arm robot for investigating leak locations and sediment properties in the PCVs of the Fukushima Daiichi Nuclear Power Station. The system developed in this paper could provide valuable information for the decommissioning process and contribute to the safety and environmental protection of the nuclear power plant.
Numeral Research on Thermal-Hydraulic Characteristics of Pebble Bed in WCCB Blanket Under In-Box LOCA
Authors: Bai Hanlin , Cheng Jie , Han Shuaiyu , Wu Di , Wang Jianjun
Published: November 2024

After the breakage of the coolant pipeline in Water Cooled ceramic Breeder blanket (WCCB), the coolant at high pressure inside the pipeline will be sprayed into the pebble bed environment, and the specific behavioral characteristics of this process are not clear. In order to investigate the flow characteristics of the coolant in the pebble bed after the rupture occurs, numerical simulation of this physical process is carried out based on the three-dimensional computational fluid dynamics (CFD) method. The flash phase transition process is described by the VOF model and the buckling flash model. The simulation results show that, after the In-box LOCA occurs, the coolant enters the proliferating spherical bed and rapidly vaporizes into water vapor by flash evaporation. Due to the geometrical characteristics of the blanket structure, the water vapor diffuses inside the pebble bed and then gradually diffuses to both sides and contacts the underside of the pebble bed; in addition to this, helium will flow out of the cladding along the helium pipeline during the In-box LOCA process; The In-box LOCA process can be initially divided into 2 stages. Stage 1: water vapor rapidly diffuses to contact the closer wall of the pebble bed at the beginning of the breach, which has a longer duration and a faster rate of pressure rise. Stage 2: water vapor diffuses mainly to the farther side of the wall, and because the gas flows out along the helium pipeline, this stage will last longer and the pressure rises slowly and is relatively stable.
Investigation on Applicability of Dynamic Mode Decomposition in Burnup Analysis
Authors: Xue Pengchao , Zhang Qian , Zhao Qiang , Zhang Yunfei
Published: November 2024

In burnup analysis of a reactor, neutron transport calculation and depletion calculation are carried out alternately, bringing a serious computational burden to large-scale problem calculation. In this study, we used a data-driven strategy based on dynamic mode decomposition to extract the modes and characteristics of nuclide density to construct a prediction expression for the evolution of nuclide density. Reducing the burden of burnup calculation can be achieved through direct prediction of the nuclide density in a certain time point in the future based on the dynamic mode decomposition algorithm[1]. Firstly, six snapshots of different size are selected to test the relationship between the size of snapshot and prediction accuracy. The results indicate that when the time scale of the snapshots is the same as the time scale of the period to be predicted, the dynamic mode decomposition algorithm can extract a more comprehensive spatiotemporal evolution model. Secondly, we selected appropriate-sized snapshots to test the impact of different truncated rank on prediction accuracy. The results show that the optimal truncated rank for the assembly-level problems is 11. Thirdly, we explored the effect of the spatial field information on the results of prediction. The results indicate that using snapshots containing spatial information for prediction does not effectively improve accuracy and still incurs a high computational burden. Finally, the applicability of the dynamic mode decomposition algorithm to different fuel assemblies and the computational efficiency of the DMD algorithm are tested. The results show that the prediction deviation of nuclide density for different assemblies is around 1%, while saving 99.87% of the burnup calculation time. The result proves that the dynamic mode decomposition algorithm can efficiently discover the hidden dynamics in burnup analysis and has a considerable application prospect in large-scale burnup analysis.
Molecular Simulation of I2-H2O Diffusion in Different Pore Structures of Activated Carbon Fiber
Authors: Li Xin , Liu Jingguo , Shuang Yingzhong , Liu Zhixin , Li Jian , Yu Jie , Li Yongguo , Liu Liye
Published: November 2024

Activated carbon fiber (ACF) is a new, efficient, and multifunctional adsorption material developed in the 1970s. In order to develop an iodine adsorbent of ACF, the diffusion effect of iodine in the pores of ACF was studied. Based on molecular dynamics, and Monte Carlo methods, a pore structure model of ACF was constructed based on carbon nanotubes. The accuracy of the model was verified through diffusion experiments and comparison with simulation results. And then, according to the basic structural unit of the constructed ACF pores, the mean square displacement and radial distribution functions of I2 and H2O under different pore sizes, functional groups and temperatures were studied, and the interaction between I2-H2O was obtained. The results show that for different diffusion components at the same temperature, the diffusion rate shows I2 > CH3I > H2O with the increase of time step. The diffusion rate of I2 is positively correlated with temperature and pore size. For different relative humidity, the diffusion rate is 40% > 80% > 10% with the increase of time steps. This study provides a theoretical basis for achieving the goal of iodine removal of activated carbon fiber.
Study on the Effective Diffusion Coefficient of Inert Gases in Porous Graphite Materials
Authors: Nie Rui , Wang Yu , Zhou Ziling , Zhang Weihua , Xie Feng , Li Hong , Wei Liqiang , Fan Zhixiang , Fu Jia , Fan Qunchao
Published: November 2024

The high-temperature gas-cooled reactor (HTGR), is globally recognized as one of the reactor types embodying the safety features of the fourth generation, with the advantages of safety, high thermal efficiency in power generation, a small emergency planning zone, and the ability to use high-temperature helium for hydrogen production. HTGRs utilize a significant amount of graphite materials, including fuel graphite matrices, graphite moderator material, and graphite reflector layers and so on. Fuel containing fissile materials produce inert gas fission products such as Kr and Xe, while isotopes like Ar-40 in the pores of graphite materials can become activated, forming inert gases such as Ar-41. The fuel element matrix graphite is the necessary pathway for these inert gas nuclides to enter the primary coolant circuit. However, current research rarely provides a comprehensive analysis of the diffusion behavior of inert gases in porous graphite materials. This study establishes a computational model and methodology for determining the effective diffusion coefficients of inert gas nuclides in porous material graphite. It quantitatively calculated the cumulative release fractions and release ratios of Ar, Kr, Xe, in both open and closed pores of the fuel element matrix graphite and elucidated the impact and patterns of factors such as pressure, temperature, and half-life on the diffusion behavior of inert gases in porous material graphite. This research provides valuable insights for future studies on the diffusion behavior of Ar, Kr, and Xe in graphite.
Conceptual Design of High-Power Nuclear Power System on the Surface of Mars
Authors: Han Yansong
Published: November 2024

Mars has always been an important target of deep space exploration, and it is expected to achieve the goal of manned landing around 2050. For manned Mars missions, a reliable and powerful power system is essential. Because of its simple, reliable, safe and high energy conversion efficiency, sCO2 Brayton cycle with heat pipe reactor has become a promising high-power nuclear power scheme on the surface of Mars. In this paper, the concept design of Mars surface megawatt high-power nuclear power supply system based on sCO2 Brayton cycle with heat pipe reactor was carried out. In order to explore the native application potential of Martian surface atmosphere, two working fluids (pure CO2 and the CO2+N2+Ar mixture of Martian surface atmosphere) were compared. The thermodynamic properties of working fluid in supercritical state are complex. Therefore, it is very important to use appropriate thermodynamic property calculation tools. In this paper, the commonly used calculation tools for the thermodynamic properties of working fluid were verified to select the appropriate tools. Then, the concept of Mars surface megawatt high-power nuclear power system was constructed, and the thermodynamics and mass models were completed. On this basis, the operation characteristics of the system were analyzed, and the application potential of Martian atmospheric mixture was preliminarily evaluated. Finally, the system was comprehensively optimized to determine the optimal parameter setting, finishing the conceptual design of the system and the exploration of application value of Martian surface atmosphere.
Study on Long-Life Minor Actinide Temporary Storage Scheme Based on High-Level Liquid Waste Partitioning Strategy
Authors: Meng Xiantao , Wei Meng
Published: November 2024

At present, transmutation and fast reactor fuel cycle cannot be implemented in engineering, minor actinides obtained by high-level liquid waste partitioning may be managed by the concept of “conditioning - temporary storage - retrieval - reuse”. In this paper, Am/Cm stripping stream of TRPO process is analyzed as example of this concepts. The physical properties (composition, mass, radioactivity, heat power) of the americium-curium mixed oxide, and the variation of these properties over time have been investigated. Although the mass ratio of Am and Cm in the Am/Cm stripping stream obtained by TRPO process is very low (less than 5%), the Ln/An separation may not be necessary under the concept of “conditioning - temporary storage - retrieval - reuse”, because during the temporary storage of americium-curium oxides, U, Pu, Np will be produced through alpha decay of Am and Cm. The rare earth elements and transition metals contained in the Am/Cm stream are the main contributors of β/γ radioactivity and short-term heat in the mixed oxides, but the β/γ radioactivity and short-term heat release will decay rapidly, leaving only a radioactivity and long-term heat of Am and Cm.
Dimensionless Analysis of Aerosol Dynamic Behaviors in the Light Water NPP Severe Accidents
Authors: Wang Jinghong , Peng Wei , Yu Suyuan
Published: November 2024

The aerosols dispersed inside the containment space undergo a variety of complex dynamic behaviors such as convection, sedimentation, thermophoresis, diffusiophoresis, and Brownian coagulation after a severe accident in a modern light water nuclear power plant (NPP). These aerosol behaviors occur simultaneously and compete with each other. The relative significance of the aerosol behaviors varies greatly under different environmental conditions. In order to study the significance of convective transport, gravitational sedimentation, thermophoresis, diffusiophoresis, brownian coagulation under different environmental conditions after a severe accident, a population balance equation for aerosol dynamic behaviors after a severe accident was constructed in the present work. A dimensionless analysis was carried out on the population balance equation, and the Peclet number and the Damköhler number, were introduced to describe the relationship between the aerosol behaviors after a severe accident. The variation rules of the dimensionless parameters were obtained under various environmental conditions such as temperature, pressure, particle sizes, and particle number concentration. The results show that, under typical severe accident conditions, the changes of temperature and pressure have significant effects on the diffusiophoresis process. The change of aerosol number concentration mainly affects the aerosol diffusion and coagulation rate. While the aerosol particle size affects most aerosol dynamic processes, and the gravity sedimentation process has the highest sensitivity to the change of particle size. The results of this study are helpful to understand the dominant mechanisms of aerosol behavior after severe accidents and to understand the spatial-temporal evolution of aerosols after severe accidents.
Surrogate Modeling Advancements in Nuclear Fuel Engineering: Bridging the Gap Between Accuracy and Efficiency
Authors: Cancemi Salvatore Angelo , Angelucci Michela , Lo Frano Rosa , Paci Sandro
Published: November 2024

In nuclear reactor fuel engineering, the phenomena of Pellet-Cladding Interaction (PCI) and Pellet-Cladding Mechanical Interaction (PCMI) present significant challenges. These issues affect the design and safety of NPPs, mainly due to factors like fission gas release and fuel swelling. Accurately modeling these interactions is complex, as they are complexly linked with the fuel’s neutronic and thermal responses. Traditionally, the simulation of these interactions, essential for nuclear safety and risk assessment, has been complex and resource intensive. This study focuses on developing innovative surrogate models to enhance computational efficiency in nuclear fuel code analysis. A surrogate model based on a neural network approach is a data-driven computational model that approximates the behavior or output of complex, time-consuming, or resource-intensive simulations. This study specifically employs Artificial Neural Networks (ANNs) and statistical algorithms, aiming to reduce high computational cost of traditional approaches. A validated synthetic dataset, representing thermal analysis under steady-state conditions, is used to train the machine learning model. The dataset specifically focuses on Cladding Temperature. The study investigates eight different Test-Case. The surrogate model, trained on only 20% of the dataset, can predicts the entire time series of temperature using ARIMA, LSTM, and Prophet algorithms. The maximum error achieved by the surrogate model is 3.19°C compared to the validated temperature. The study demonstrates that surrogate models offer a time-efficient alternative for simulating complex physical phenomena, achieving a balance between accuracy and efficiency. This approach is particularly beneficial in scenarios where full-scale 2D and 3D simulations are excessively time-consuming, providing quicker results and significantly reducing computational resources.