bluuebunny/crossref_metadata_2025 huggingface dataset:
Understanding the Flow Anisotropic Turbulent Flow in the Subchannel of Fuel Assembly Under Effect of the Bare Rod and Mixing Vane Grid by LES
Authors: Qi Siwei , Han Bin , Xing Tianyang , Zhu Xiaoliang , Yang Bao-Wen , Yin Yuanyuan , Liu Aiguo , Liu Shenghui
Published: November 2024

The grid has the functions of supporting fuel rod bundles and mixing the coolant. The flow of the coolant through the mixing vanes increases its perturbation, which directly affects the Critical Heat Flux Density (CHF) of the reactor core, enhances the wall heat transfer performance, and improves the safety and economy of the reactor. The geometric asymmetry of the sub-channels in the Pressurized Water Reactor (PWR) fuel assemblies leads to anisotropy in the flow, and accurate prediction of the flow mixing characteristics in the rod bundle channels plays a significant role in the prediction of the onset of nucleate boiling and CHF. Traditional RANS turbulence models struggle to accurately predict the anisotropic turbulent flow in rod bundle channels. This paper, based on LES simulation calculations, investigates the anisotropic flow characteristics in 2 × 1 and 3 × 3 bare rod channels, comparing the effects of different sizes of rod bundle geometries on flow anisotropy. On this basis, the study examines the anisotropic turbulent flow in rod bundle channels under the influence of a 3 × 3 grid. By conducting Large Eddy Simulations (LES) on the flow field of a simple 3 × 3 rod bundle sub-channel, this paper analyzes the impact of grid mixing vane angle variations on the flow anisotropy between nuclear fuel rods. The results show that changes in the grid mixing vane angles can produce varying degrees of flow field perturbations. The study of the evolution of turbulent anisotropy between rods under the coupled effect of rod bundle channels and grids, and flow field characteristics, lays the foundation for the construction of URANS models that can accurately reflect anisotropic effects, and also establishes a basis for the design of high-performance fuel assembly grids.
Impact of Tube Fouling on Steam Generator Performance Using a Thermal-Hydraulic Code
Authors: Jing Liu , Zhen Li , Zhenqin Xiong , Yuan Gao , Yuchen Lin , Linpeng Shi
Published: November 2024

In order to study the deposition and distribution of Fe3O4 fouling and its impact on the thermal performance of steam generator (SG), numerical simulation model of SG has been developed based on porous medium model. The impact of Fe concentration of feed water on the fouling deposition characteristics and the thermal performance of steam generator has been quantitatively analyzed. The fouling deposition mass after ten-year operation is 65.0 and 166.5 kg for Fe impurity concentration in the SG feedwater 1ug/kg and 3ug/kg, respectively. The thermal power is decreased by 1.2% and 3.1% compared to the situation without fouling accumulation, respectively. The heat flux degradation and the thermal resistance due to fouling of tubes in different locations of SG are compared. The mean heat flux of the shortest tube in the central region is most serious affected by fouling deposition. The fouling thermal resistance contributes to 12%∼25% of the thermal resistance along the bend for Fe concentration 2ug/kg and ten-year operation, while it is in the range of 3–21% and 0–6% for the straight pipe section on the hot side and on the cold side, respectively.
Optimization Design Study of a Radiant Heat Exchanger for a Small-Scale Underwater Nuclear Power Plant’s Residual Heat Removal System
Authors: Guo Yinan , Li Lei , Wen Yongsheng
Published: November 2024

With the continuous development of small-scale underwater nuclear power plants, the need to develop residual heat removal systems for underwater nuclear power plants has become urgent. The efficient and lightweight residual heat removal system is one of the key research directions that need to be developed. The mass and volume of the heat exchanger of a small-scale underwater nuclear power plant account for a large proportion of the residual heat removal system. The heat exchanger in the residual heat removal system in the liquid metal-cooled nuclear reactor of a small-scale underwater nuclear power plant adopts radiant heat exchange, and the fins of the radiant heat exchanger are made of graphite carbon fiber composite material with a rectangular structure. The heat exchanger and heat trap are filled with nitrogen gas. This paper proposes to replace the rectangular cross-section structure of the radiant fins with the triangular structure of the heat exchanger fins. The heat transfer performance of the optimized structure of the radiant heat exchanger is investigated using Fluent. The results show that the optimized triangular fin structure of the radiant heat exchanger makes the residual heat removal system have a better performance of the radiant heat exchanger. The weight of the optimized system is reduced by 16.53%.
Calculation of Noble Gas Ion Mobility Based on SRK EOS Under High-Pressure
Authors: Chang Jiahao , Fu Jintao , Tang Peng , Liu Haoyu , Wang Zhentao , Wu Zhifang
Published: November 2024

The time response characteristic of detectors in radiation imaging systems holds significant importance, and the response time of the detector will influence the image quality of the imaging system. However, the extended response time of parallel plate ionization chamber detectors currently in use contributes to image blurring. Enhancing the imaging quality necessitates a reduction in detector response time. There are many ways to reduce the response time of ionization chamber detectors. Modifying the working gas within the detector and opting for a working gas component with high ion mobility, while keeping the detector structure unchanged, has proven effective. Typically, gas detectors in radiation imaging systems contain noble gases at high pressure. To select the appropriate working gas components, we need to measure the mobility of ions of different gas components at high atmospheric pressure. Measuring the mobility of ions at high atmospheric pressure requires an accurate calculation of the molecular number density, and it is vital to choose an appropriate equation of state (EOS) for the gas. Hence, this paper suggests a method for calculating molecular number density grounded in the Soave-Redlich-Kwong (SRK) gas equation of state. It involves deriving the explicit form of the SRK equation and integrating an iterative approach to compute molecular number density in high-pressure gases. In order to compare the calculated results of different gas equations, various gas equations of state such as the Van Der Waals (VDW), Peng Robinson (PR) equation, ideal gas equation of state, etc. are utilized in this paper. The calculated results of the above equations are compared with the National Institute of Standards and Technology (NIST) data. The calculated results show that the SRK EOS is closer to the real value. The calculated results of the ideal gas equation of state deviate the most from the actual value. For some gases (e.g., krypton), the van der Waals equation has no real-number solution when the gas pressure is more than 2900KPa. Thus, in the subsequent measurements of the ion mobility of noble gases under high-pressure conditions, the ion mobility can be calculated with the assistance of the SRK equation and more accurate data can be obtained.
Investigation of Reverse Flow in the Passive Residual Heat Removal System of NHR-200II
Authors: Geng Yiwa , Li Ziyi , Liu Xiongbin , Huang Shuliang , Qin Haiqi
Published: November 2024

The NHR-200II is a multi-purpose small integrated pressurized water reactor (PWR). Its passive residual heat removal system (PRHRS) to remove core decay heat. During a series of PRHRS experiments conducted in a scaled integral test facility for the NHR-200II, a reverse flow phenomenon was observed. To study this phenomenon, a system model using the RELAP5 code was developed for the integral test facility, and the reverse flow phenomenon was numerically simulated. The system code was used to calculate the flow rate and heat transfer power of each RHE and PHE branch under the test conditions. The calculated flow rates in the PHE branches showed good agreement with the observed reverse flow phenomenon in the test facility. Similarly, the flow rate and heat transfer power of the RHE branch matched well with the experimental data. The cause of the reverse flow phenomenon is attributed to the asymmetric location of the RHE branch on the hot and cold headers relative to the PHE branches. The non-uniformity flow rates in the PHE branches are caused by the different distances between the RHE and each PHE branch. Although small initially, this non-uniformity is amplified by the positive feedback effect of pressure drop and flow rate between the PHE branches, eventually resulting in reverse flow. To address the reverse flow phenomenon in the PRHRS, two methods are proposed. The first method involves reducing the length of the ring header, while the second method focuses on increasing the buoyancy force of the natural circulation system. The effectiveness of these methods is calculated and discussed. This paper highlights the NHR-200II reactor PRHR system, discusses the observation of reverse flow in experiments, describes the numerical simulation using a system model, and proposes two potential methods to mitigate the reverse flow.
Intelligent Fault Diagnosis of Rotating Machinery Based on Deep Neural Network
Authors: Zhang Xiuchun , Xia Hong , Liu Yongkang , Zhu Shaomin , Jiang Yingying , Zhang Jiyu , Liu Jie , Yin Wenzhe
Published: November 2024

In order to improve the fault diagnosis for rotating machinery in nuclear power plants, this paper proposes a fault diagnosis model based on short-time Fourier transform (STFT) and Residual Network (ResNet), using information from multiple sensors. Firstly, acceleration sensors are installed on the different positions of the motor to measure vibration signals, and the normal condition and eight typical faults of the motor are simulated by the mechanical fault simulation test platform. Then, sensor signals from different positions of the motor are obtained as inputs to the fault diagnosis model to reduce the impact of the fault feature attenuation during mechanical propagation. In data preprocessing, STFT is used to convert the multi-source signals from time domain to time-frequency domain to extract the fault features of the signals. Finally, ResNet is used to further extract advanced features from time-frequency feature signals and identify fault patterns. Results obtained show that the proposed model based on STFT, ResNet and multi-sensor information strategy can accurately classify the fault modes, and comparisons with the traditional convolutional network model and single sensor strategy model highlight the advantages of the proposed method in rate of convergence, accuracy and noise robustness.
Application of Sensitivity Analysis Techniques to a Low Dimensional Problem in the Frame of Severe Accidents
Authors: Angelucci Michela , Paci Sandro , Cancemi Salvatore Angelo , Lo Frano Rosa
Published: November 2024

The quantification of uncertainties in Severe Accident (SA) simulations is crucial for ensuring the robustness and enhancing the reliability of nuclear reactor safety assessments. However, traditional UQ methods often lack the ability to discern the individual contributions of various input parameters to the overall uncertainty in simulation results. To this end, a sensitivity analysis complementary to the uncertainty one seems indispensable.

In this framework, the present paper presents a twofold approach to address this challenge: on one side, by applying the so called single variable approach, in which the relationship between the selected output response and each input parameter taken individually is studied; on the other side, by integrating supervised Machine Learning (ML) techniques into the sensitivity analysis of SA simulations.

The selected sensitivity analysis techniques are applied to the outcomes of SA simulations conducted as part of the recently concluded “Management and Uncertainty of Severe Accidents” (MUSA) EURATOM project. In particular, results from a training exercise, involving an exploratory UQ application with a simplified SA scenario (namely the Phebus FPT1 test) and a limited number of input parameters, are employed as a basis for the performed sensitivity analysis.

As a result of the application of the different sensitivity methods, the parameters, or “features”, selected by each one of them are reported and documented. Additionally, a thorough cross-comparison of the obtained results is conducted in the attempt to work out an optimization of the sensitivity analysis process.
Study on Radionuclide Desorption in Matrix Graphite Under Accident Conditions
Authors: Xu Wei , Li Jian
Published: November 2024

High temperature gas cooled reactor (HTGR) is one of the promising advanced nuclear reactor for future electricity generation. The spherical fuel element with thousands of TRISO-coated particles dispersed in the graphite matrix are adopted by the HTGR, which is an important guarantee for the inherent safety of this type of reactor. The TRISO fuel particle typically consists of a very small fuel kernel in the central region, and surrounded by four coating layers. These coating layers constitute the principal containment barrier for radioactivity release under normal and accident conditions. In the spherical fuel element, the radionuclide release is dominated by multiple physical and chemical processes, such as fission product transport, recoil, adsorption, desorption, and etc. The matrix graphite would be oxidized by water or air under water ingress or air ingress accident, which causing an increase of the porosity of graphite. The increase of graphite porosity affects the diffusion and adsorption of radionuclides in the matrix graphite, thereby impacting the release of radioactive nuclides. This paper studies the adsorption process of radionuclides and analyzes the influence of graphite oxidation on the adsorption rate under water ingress and air ingress accident. The adsorption rate of several safety related radionuclides (such as 131I, 110Ag, 137Cs, 90Sr) in the matrix graphite were analyzed in detail. The influence of temperature, adsorption energy and density of radionuclide in the matrix graphite on the adsorption rate is discussed. Among these relevant factors, temperature has the most significant influence. The density of radionuclide in the matrix graphite increases with the increase of graphite porosity and surface area due to graphite oxidation. Although the adsorption rate of nuclides in the matrix graphite will increase due to oxidation, the adsorption rate of nuclides is almost zero under water ingress and air ingress accident, and the increase of the adsorption rate caused by oxidation can be negligible.
Study on NUREG-2215 Regulatory Condition for Reflooding of the Previously Dried High Burnup Fuel Is Not Allowable During Dry Storage
Authors: Liao Ting-Yi , Tseng Yi-Fan , Lan Kuan-Che , Chiu Wan-June , Tung Hsiao-Ming
Published: November 2024

This paper studies the regulatory condition for reflooding of the previously dried high burnup fuel is not allowable in the NUREG-2215 review guidance issued by the NRC, to evaluate the regulatory applicability of this condition to the retrieval operations in Taiwan’s indoor dry storage facilities.

The discussion content firstly explains the research background and purpose of this paper, and then introduces NUREG-2215. And refer to NUREG CR-7198, Revision 1, NUREG-2224 and other research reports, as well as NRC’s response to public comments, to discuss the regulatory basis for reflooding of the previously dried high burnup fuel is not allowable. And from the NUREG-2224 evaluation results, it is known that intact high burnup fuel can withstand at least one thermal cycle exceeding 65°C without affecting the safety analysis of transportation or storage based on design basis accidents. The latest NRC review operation has also eliminated the restriction of reflooding, but high burnup fuel stored for more than 20 years needs to undergo additional supplementary analysis according to the recommended method.

Subsequently, the feasibility of applying this regulatory condition to Taiwan is evaluated based on items such as the regulatory situation, characteristics of spent nuclear fuel cladding, retrieval facilities, and retrieval operations.
A Benchmark Test of CHF Test Facility in CIAE to Columbia University HTRF
Authors: Duan Minghui , Xu Yongwang , Zhang Dongxu , Zhao Minfu , Wei Junhan , Yang Bing , Li Qingyuan , Wang Wei
Published: November 2024

Critical heat flux (CHF) is one of the most concerned thermal hydraulic phenomena in reactor design and safety analysis. As CHF involves complex two-phase flow heat transfer mechanism, and hasn’t been fully understood, the prediction of CHF mainly relies on CHF correlations obtained under limited experimental conditions. At present, CHF tests of a new type of PWR fuel assembly are being carried out on the Thermal-hydraulic Comprehensive Test Facility (TCTF) in China Institute of Atomic Energy (CIAE). In order to verify the reliability of the facility, testing method and measurements, a benchmark test has been performed on TCTF, which repeated the CHF test performed on the Heat Transfer Research Facility (HTRF) at Columbia University in the United States. The test section on TCTF was a 5 × 5 uniformly-heated rod bundle with spacer grids, and its sizes were the same as prototype. The range of test conditions covered the local pressure of 7.4∼16.8 MPa, the inlet mass flux of 1.0∼4.0 t/m2s, and the inlet temperature of 120∼266 °C. Verifications including the redundant measurements comparison, the repeatability of test data under the same thermal hydraulic parameters, the thermal balance verification and linearity of data were carried out to prove the reliability of the test data. And CHF test data comparison between TCTF test and HTRF test has been carried out. The comparation results showed that the deviation of the CHF values in TCTF test from HTRF test was within 6.6%, which indicated that the CHF tests in HTRF have been well repeated on TCTF. The benchmark test verified the accuracy and reliability of TCTF, testing methods and measurements, and TCTF can perform CHF tests to obtain reliable CHF data to evaluate the thermal-hydraulics performance of new PWR fuel assemblies.
Research on Pulse Discrimination for Alpha and Beta Measurement Based on Auto-Encoder
Authors: He Shuijun , Tang Lijuan , Liang Manchun , Li Ke , Wang Jihong , Du Xiaochuang
Published: November 2024

Pulse shape discrimination is an important method for measuring both gross alpha and gross beta in a single measurement. When the quenching level of the sample changes significantly, the percent spillover of the alpha events into the beta count and the beta events into the alpha count will change greatly. In order to ensure the accuracy of the measurement results, it is necessary to re-optimize the pulse shape discrimination parameter. For each sample measurement, a spiked sample with the same quenching level needs to be established for pulse shape discrimination parameter optimization. In this paper, we proposed a pulse discrimination method for alpha and beta decay based on auto-encoder, which only needs to be optimized once, and can adapt to the significant changes in the quenching level of the sample. The pulse shape discrimination model was trained on the measurement data of Am241 spiked samples with different quenching levels, and verified by the measurement data of Sr90 spiked samples with different quenching levels. Finally, the optimal parameter for pulse shape discrimination was determined. The model was tested using the measurement data of Am241 and Sr90 spiked samples with different quenching levels. These samples quenching level range from 306 to 433 which represented by transformed spectral index of the external standard. The percentage of alpha pulse misclassification is 2.21%–3.35%, and the percentage of beta pulse misclassification is 2.37%–2.82%.
Enhancing Nuclear Power Plant Operational Forecasting With Transformer Neural Networks: a Time-Series Data Approach
Authors: Tuo Yanjie , Liu Xiaojing
Published: November 2024

The forecasting of critical operational parameters within nuclear power plant is paramount for ensuring their operational safety and efficiency. In the realm of predictive analytics, the Transformer neural network architecture has distinguished itself by delivering superior performance across a diverse array of fields. Despite its proven efficacy, the exploration of Transformer models within the nuclear energy sector is notably sparse. This investigation seeks to bridge this gap by deploying Transformer-based predictive methodologies on a dataset characterizing the loss of offsite power scenario at a nuclear power plant. Through the application of the Tree-structured Parzen Estimator optimization algorithm, the study meticulously optimizes the neural network architecture to refine its predictive precision. The research effectively forecasts critical physical metrics such as core pressure vessel water level, core outlet temperature (average), hotleg temperature, coldleg temperature, steam line pressure, steam generator secondary feedwater flow rate. The empirical results of this study are promising, demonstrating the Transformer model’s good capability in predictions, with the normalized Root Mean Square Error for all evaluated physical parameters on the test dataset consistently remaining below 0.056. This outcome underscores the Transformer’s potential as a robust tool for predictive analytics in nuclear power plant operations.
Effect of Temperature and Lithium Concentration on the Out of Reactor Corrosion Property of SZA Alloys and Model Study
Authors: Zeng Qifeng , Wang Chen , Wang Ben , Lu Junqiang , Li Cong , Zhang Lefu , Si Shengyi
Published: November 2024

Zirconium alloys have been widely used as fuel cladding and structural materials for pressurized water reactor (PWR) fuel assemblies because of their stable dimensional and physical properties under neutron irradiation environment, low thermal neutron absorption cross section, good corrosion resistance and mechanical properties at operating temperature. During the operation of zirconium alloy cladding, corrosion resistance is a critical life-limiting degradation mechanism, temperature and lithium concentration are the key factors for corrosion resistance. In order to study the effects of temperature and lithium concentration on the corrosion resistance, the microstructural analysis of two new zirconium alloys named SZA and static autoclave corrosion tests under various conditions were carried out. The microstructure of two SZA alloys were analyzed by scanning electron microscope (SEM) and transmission electron microscope (TEM), including the microstructure of the zirconium alloys matrix, the size, distribution and structure of the second phase particles (SPPs). The corrosion weight gain curves of SZA alloys and Zr-4 alloy under different temperatures and lithium concentrations in high temperature and high pressure water were obtained through out of reactor static autoclave tests. The corrosion rate-temperature and corrosion rate-lithium concentration models of SZA alloys were obtained through data analysis and model fitting.

The experimental results show that the two SZA alloys are both partially recrystallized structure, and a large number of fine SPPs are dispersed in the zirconium alloys matrix. The average diameters of the SPPs in SZA-4 and SZA-6 alloys are 98.6nm and 100.7nm, respectively. There are mainly two kinds of SPPs containing trace Ge and small amount of Nb in SZA-4 alloy, including Zr(Fe,Nb,Cr,Ge)2 with face centered cubic (FCC) crystal structure and Zr(Fe,Nb,Cr,Ge)2 with hexagonal close packed (HCP) crystal structure. There are three kinds of SPPs containing trace Cr in SZA-6 alloy, including a large amount of Zr(Fe,Nb,Cr)2 with HCP crystal structure and Zr2(Fe,Nb,Cr) with FCC crystal structure, and a small amount of Zr(Fe,Nb,Cr)2 with FCC crystal structure. SZA alloys have better corrosion resistance than Zr-4 alloy under a variety of temperature and lithium concentration test conditions, except that there is no obvious difference in the corrosion resistance of the three alloys at 315 °C due to the low test temperature. At the same lithium concentration, increased corrosion rate of the three zirconium alloys were observed with an increase in test temperature. At the same temperature, increased corrosion rate of the three zirconium alloys were observed with an increase in lithium concentration. The model fitting results indicate that the corrosion rate of SZA alloys are exponentially related to temperature, and a quadratic or linear relationship with lithium concentration.
Improving the Accuracy of RCCS Simulator Based on Ensemble Kalman Filter Algorithm
Authors: Liu Songsong , Zhang Han , Wu Yingjie , Lang Minggang , Dong Yujie , Li Fu
Published: November 2024

Data assimilation, an effective approach that utilizes measured data and suitable assimilation algorithms to reduce existing model calculation errors, has its origins in weather forecasting research. In this study, a framework of data assimilation based on Kalman filter (EnKF) algorithm is established for the computational program of the reactor cavity cooling system (RCCS), an important component of high temperature gas-cooled reactor-pebble-bed module (HTR-PM). Original model with uncertain parameters and conservative assumptions gives a result deviating from the measured data. After assimilation, parameters including the thermal conductivity of the water cooling wall, blackness coefficient, material reflectivity, resistance coefficients of various components, and the frontal area of the air cooler, are recalibrated. The modified model significantly improves the accuracy of predicting the residual heat removal power, reducing the calculation error from 16.89% to 1.22% compared to the original model. By applying the EnKF algorithm, this paper expands its application scope within nuclear engineering, providing inspiration for future relative studies.
Optimal Sensor Placement in Nuclear Power Steam Water System Based on Signed Directed Graph
Authors: Xing Tianyang , Zeng Chunyang , Han Bin , Jiang Mudi , He Yunze , Liu Shenghui , Huang Junling , Zhu Xiaoliang
Published: November 2024

Nuclear power plant is expected to achieve the comprehensive fault monitoring and ensure the secure and stable operation. Enhancing the reliability of the digital measurement system will effectively mitigate the potential risks associated with compromised operation of the nuclear power system, confine the extent of fault impact, and bolster the operator’s situational awareness proficiency. This paper presents a comprehensive modeling strategy for obtaining the fault-sensor correlation matrix in the deaerator system of a nuclear power plant based on SDG theory. Based on the utilization of Advanced Process Simulation Software (APROS), we successfully developed a dynamic numerical simulation model for the deaeration system in nuclear power plants. By conducting a comprehensive analysis of the correlation between fault propagation path and sensor, we have successfully established a probability model that accurately assesses the performance of sensor fault detection. A multi-objective optimization model for sensor layout, considering the cost of sensor placement, detection ability, and fault correlation, was developed based on constraint relationships. Compared to conventional sensor placement schemes, the optimized sensor placement scheme exhibits superior performance in terms of cost-effectiveness in arranging sensors, reliability in fault detection, and correlation between faults and sensor points.
Investigation on Oxidation Behaviors of FeCrAl Alloy Cladding Under the Simulated LOCA Conditions
Authors: Chen Kang , Duan Zhengang , Wen Qinglong , Zhang Ruiqian
Published: November 2024

During a LOCA (Loss of Coolant Accident), the zirconium alloy cladding undergoes oxidation, generating significant heat and hydrogen gas. In extreme circumstances, this can result in hydrogen explosions compromising reactor safety. Recently, the ATF (Accident Tolerant Fuel) cladding has been introduced as a new generation of fuel cladding. Among the ATF claddings, FeCrAl alloy cladding exhibits a low oxidation rate during LOCA due to the formation of a dense alumina layer during high-temperature oxidation. However, current research on this FeCrAl alloy cladding remains limited in scope and depth. The present study focuses on the oxidation behavior of FeCrAl alloy cladding at 1200 °C in steam. Key findings include: (1) Spots on the sample were identified as oxides of Fe, Cr, and Al through microscopic analysis. (2) An oxide layer of Al was observed with a thickness ranging from 0.86 μm to 1.98 μm. (3) The ring compression test revealed that the FeCrAl alloy cladding neither fractured nor underwent a ductile-brittle transition.
Suppression of Harmonic Vibration in AMB System Using Nonlinear Adaptive Resonant Controllers
Authors: Bian Xiaoyu , Shi Zhengang , Mo Ni , Sun Zhe
Published: November 2024

The primary helium circulator (PHC) serves as the core of the high-temperature gas-cooled reactor. Due to the non-contact nature of active magnetic bearing (AMB), it is employed as the supporting apparatus for the rotating machinery of the PHC to ensure the purity of the helium environment within the high-temperature gas-cooled reactor. Nevertheless, the rotor mass imbalance and sensor runout may lead to synchronous vibrations and harmonic vibrations within the AMB system. These vibrations are subsequently transmitted to the shell, resulting in foundational vibrations that pose a significant threat to the stability and safety of the system. This paper introduces a novel nonlinear adaptive resonance controller to suppress both synchronous vibrations and harmonic vibrations in the system. The algorithm features a straightforward structure with minimal computational complexity and operates without reliance on the mathematical model of the system. The algorithm proposed in this paper consists of three modules: the rotor phase calculation module, the resonance signal extraction module, and the adaptive iteration module. The rotational speed sensor is utilized by the rotor phase calculation module to calculate the rotor phase within one rotation period. The resonance signal extraction module can calculate the synchronous and harmonic energies within one iteration period. The adaptive iteration module can output the feed-forward compensation signals with proper amplitudes and phases. Finally, the simulation results validated the effectiveness of the algorithm.
A Control-Oriented Hybrid Model for Nuclear Reactors Based on Neural Network
Authors: Zhu Ze , Liang Wenlong , Ye Baiqing , Jiang Qingfeng , Wang Pengfei
Published: November 2024

Accurate modeling is the basis for analyzing the dynamic response characteristics of the model. However, due to the complexity and time-varying nature of the internal mechanisms of the reactor, it is inevitable that inaccurate model parameters will be used in the modeling process. These will lead to discrepancies between the modeled mechanisms and the actual reactor. In this paper, the difference is evaluated and shortened by means of neural network hybrid modeling. Based on the MATLAB/Simulink simulation platform, this paper firstly obtains the parameters that have the greatest influence on the linear model through sensitivity analysis and takes them as the object of neural network correction, then obtains the data required for offline training of neural network according to the mechanism model, linear model and the deviation of the two under different working condition levels, retains the neural network weights and thresholds obtained from the offline training, and finally utilizes the gradient descent algorithm to update the neural network weights and thresholds in real time in order to achieve the online calibration of the linear model. The final results show that the hybrid model can effectively reduce the steady-state deviation between the two models, which indicates that the hybrid modeling can effectively improve the accuracy of the established model and provide a solid foundation for the subsequent design of the control system based on the linear model.
Study on the Influence of Internal Heat Source Characteristics and Distribution on Effective Thermal Conductivity of Cylindrical Particle Dispersed Fuel
Authors: Liu Yuhao , Sun Jun
Published: November 2024

Dispersing fuel particles within a matrix is a pivotal strategy for next-generation nuclear fuels. The internal temperature distribution of dispersion fuels is generally determined by the lumped parameter method, where Effective Thermal Conductivity (ETC) plays a critical role. Most computations yet use classical ETC models that lack consideration for internal heat sources, such as Maxwell-Eucken model, Chiew-Glandt model, and Effective Medium Theory. Some researchers have noted that classical models tend to underestimate the temperature when internal heat sources exist. Additionally, the stochastic distribution of heat sources introduces uncertainty in ETC values and, consequently, temperature predictions. This study focuses on cylindrical dispersions with randomly distributed particles, and employs the three-dimensional finite element method to calculate their ETCs. Various parameters, including component thermal conductivity, particle radius, volume fraction, and minimum particle spacing, are adjusted in different cases. For each parameter configuration, a large number of cases are randomly generated and simulated. Statistical characteristics of ETC results are analyzed. It reveals that, when internal heat sources are considered, the value and uncertainty of cylindrical dispersion ETC are influenced not only by volume fraction and component thermal conductivity, but also by the size of heat-generating particles and the minimum particle spacing. It underscores the necessity of characterizing and quantifying the distribution of internal heat sources when proposing novel ETC models for dispersion fuel.
PID Parameter Tuning Method for Steam Generator Level Control Based on Ant Colony Algorithm
Authors: Wang Yulong , Wei Xinyu , Sun PeiWei
Published: November 2024

Steam generator levels in pressurized water reactor nuclear power plants at home and abroad use the classical serial PID control scheme. At present for the steam generator level control system control strategy method research has been fruitful, but when the operating conditions change, the original PID control effect may no longer meet the demand, so the need for PID control parameter calibration of the problem is still very few relevant methods of research. For this reason, this journal proposes a method to train the RBF neural network model based on the historical data of the system, to improve the optimization data, and establish an accurate mathematical model of the system based on the optimized data by the least squares method for the PID parameter tuning method of the subsequent ACO algorithm. The RBF neural network is trained through historical data to obtain the feature model. The optimized data is modelled using F-test to determine the order of the system data, and the corresponding differential equations are used for identification, and the identified differential equations are transformed into transfer functions, through which an accurate mathematical model can be identified in the case that the order of the system structure is not clear. Finally, the ACO algorithm is used to adjust the PID parameters according to the identified mathematical model of the system, and the adjusted PID parameters have a good control effect in different working conditions, and the results show that the optimized data identification transfer function is basically the same as the actual system, and the parameter-optimized control system of the steam generator level control system is better than the original control system in terms of the overshooting control indexes, and it achieves the control of the liquid level inside the steam generator under different working conditions. Steam generator level control under different operating conditions. It is concluded that when the model parameters are changed, the PID parameter control effect of the identification model adjusted by the ACO algorithm is better than the original PID control of the steam generator.
Experimental Study of Heat Transfer Characteristics During Reflooding Process Under Partially Exposed Conditions in Narrow Rectangular Channels
Authors: Wu Junchen , Liu Haidong , Chen Deqi , Zuo Qianlong , Liu Hanzhou , Deng Jian , Wu Dan , Chen Mingjing
Published: November 2024

In the event of a small and medium-sized breach accident in the primary loop of a nuclear reactor, the pressure relief and discharge causes a loss of coolant flow in the core, exposing the top of the fuel elements to overheated steam while the bottom remains submerged in residual cooling water. At this point, the fuel elements experience significant axial temperature differences, making the boiling two-phase flow behavior during the reflooding process of the fuel elements more complex compared to complete exposure. This experiment conducted reflooding tests with varying initial water levels for localized exposure, and the results showed that the minimum wall film boiling temperature in the completely exposed state gradually decreased downstream along the axial direction, while it increased in the case of localized exposure. The precursor cooling duration experienced during the reflooding process was longer for complete exposure than for localized exposure, and at the same wall superheat, the heat transfer coefficient for localized exposure was larger. The average heat transfer coefficient during the reflooding process of localized exposure was higher than for complete exposure, and as the degree of exposure decreased, the average heat transfer coefficient during the precursor cooling phase increased.
Total Monte Carlo Uncertainty Analysis of VVER-440 Spent Nuclear Fuel in LWR and HWR Reactor Environment
Authors: Lovecký Martin , Závorka Jiří , Kořínek Tomáš , Škoda Radek
Published: November 2024

Total Monte Carlo (TMC) extends the Monte Carlo method, using stochastic techniques and random sampling to solve the Boltzmann transport equation in spent nuclear fuel (SNF) depletion analysis. TMC evaluates the impact of uncertainties in nuclear data, such as cross-sections and fission product yields, on SNF characteristics, focusing on decay heat, which is crucial for SNF handling and management. TMC generates random nuclear data variations within uncertainty ranges for Monte Carlo simulations, resulting in outcome distributions (e.g., decay heat) that reflect real-world behavior uncertainties.

The uncertainty analysis examined two reactor environments: a standard VVER-440 reactor using a light water coolant at 1375 MW thermal power and a Teplator district heating reactor VVER-440 fuel in a HWR environment at 50 MW thermal power. This comparison offers insights into VVER-440 SNF behavior under different reactor conditions and coolants.

Using the Serpent 2 code and referencing the ENDF/B-VIII.0 and TENDL nuclear data libraries, the study focused on the impact of variable cross-sections on critical nuclides and fission product yield variations. It analyzed decay heat generation immediately post-reactor shutdown and long-term decay heat for spent fuel cask loading, providing a comprehensive view of the nuclear fuel cycle.
Preliminary Assessment of CATHARE-MODELICA Codes Coupling for Safety Analyses in Nuclear Cogeneration
Authors: De Angelis Alessandro , Olita Paolo , Ambrosini Walter
Published: November 2024

To assure a viable energy transition towards a low-carbon future energy mix, Small Modular Reactors have gained attention in the last years due to their peculiarities, e.g., potential enhanced safety, economic viability due to series production, ease of installation and modularity. Moreover, their implementation in Hybrid Energy Systems in cogeneration operation could bring to the decarbonization of other carbon-emitter industrial sectors. However, the introduction of Small Modular Reactors in Hybrid Energy Systems for cogeneration can result in different transients induced by the thermal users on the nuclear power plant, which must be carefully considered when performing a safety assessment. To this purpose, the paper describes the methodology adopted to perform a code coupling useful for future safety analyses of these hybrid energy systems. In particular, the coupling is performed between a well-assessed thermal hydraulic system code used to perform nuclear safety analyses, namely CATHARE 3, and a widely-adopted a-causal dynamic modelling language, namely MODELICA, exploiting the features of both modelling codes. The results, although preliminary, show good coupling capabilities between the two codes, providing an excellent starting point to perform future analyses of some transients induced on nuclear systems in cogeneration mode.
Study on the Development Path of AFR Storage for Spent Fuel in China
Authors: Shi Lei , Lu Haoran , Wang Yihan , Hu Jian
Published: November 2024

Away-from-reactor (AFR) storage of spent fuel is the “intermediate storage” after transported from a nuclear power plant site before final treatment and disposal. AFR storage not only ensures the safe operation of nuclear power plants, but also provides a buffer for spent fuel reprocessing or geological disposal. It is considered an important industry sector in spent fuel management in various nuclear power countries around the world. Under the background of the “carbon peak and carbon neutrality” goal, China nuclear energy will maintain a steady and rapid pace of development. The annual production of spent fuel will also increase year by year, and the demand for AFR storage will become increasingly large. However, there are various types of pressurized water reactor nuclear power units in China, resulting in different requirements for spent fuel management. Therefore, it is necessary to strengthen top-level design, taking into account many factors such as safety, technological maturity, and economy, and combine with the demand forecast of China’s nuclear power development to develop an AFR storage capacity construction plan. This study shows that it is necessary to adhere to the principle of “more concentration and less dispersion”, and start the construction of China’s spent fuel AFR storage capacity as soon as possible.
Experimental Investigation of Fluidelastic Instability in In-Line Tube Arrays
Authors: Wang Yuqi , Wang Xiaoxin , Shi Li , Wu Xinxin
Published: November 2024

The Helical Tube intermediate heat exchanger (Helical Tube IHX) is a crucial component for process heat applications and hydrogen production in Very High-Temperature Gas-Cooled Reactors (VHTRs). Due to the coil design with a large spiral diameter in the Helical Tube IHX, the fluidelastic instability needs to be given special attention. The simplified straight tube bundle model and inclined tube bundle model were designed and constructed to inclined study the fluidelastic instability of single flexible tube in the center of the bundle. A variable magnetic eddy current damping system has been constructed and utilized to change the tube’s damping ratio. The critical flow velocities for fluidelastic instability in tube bundle model were obtained under different mass damping parameters. By comparing the instability maps of the straight tube bundle and the inclined tube bundle, the results indicate that the inclined tube bundle is more prone to instability compared to the straight tube bundle.
Technology Review and Safety Assessment of Nuclear-Renewable Hybrid Energy Systems With Light-Water Small Modular Reactors
Authors: De Angelis Alessandro , Frignani Michele , Pucciarelli Andrea , Sevbo Oleksandr , Lapuente Miriam Minchole , Schneidesch Christophe , Vaglio-Gaudard Claire , Miss Joachim , Hollands Thorsten , Ambrosini Walter
Published: November 2024

An increase of renewable energy share is expected in the next future to drive the energy transition towards a low-carbon future energy mix. However, renewable energy sources are intermittent and non-dispatchable by nature and could lead to grid stability issues. For this reason, the idea of coupling renewable energy sources with Nuclear Energy (in particular, Small Modular Reactors) in the so-called Nuclear-Renewable Hybrid Energy Systems has gained more attention during the past years. During the second half of 2022 a European project belonging to the Horizon Europe program, namely TANDEM, has started to investigate the safety aspects of Light-Water Small Modular Reactors when introduced in a Hybrid System scenario. This paper aims at summarizing some of the first outcomes of the TANDEM project regarding the technology assessment of a Nuclear-Renewable Hybrid Energy System with Light-Water Small Modular Reactors, with preliminary considerations about safety of nuclear cogeneration. To this purpose, the main components of a general Hybrid system are considered, i.e., Light-Water Small Modular Reactor, low and high temperature hydrogen production, thermal and electrical storages and seawater desalination, aiming at providing general rules to perform the future modelling of the system in the frame of the project.
Experiment and Simulation Analysis for Performance of Squat Shear Wall in Nuclear Island
Authors: Wang Shaojie , Wang Dongmei , Li Yumin
Published: November 2024

In nuclear island, shear wall has the characteristic of low shear span ratio (low aspect ratio) and high reinforcement ratio, which can be called as the squat shear wall. The performance of squat shear wall determines the seismic PSA analysis conclusion of nuclear island structures. According to the typical wall in nuclear island, three large-scale specimens of squat shear wall are designed. Considering different axial compressions, quasi-static experiments are conducted. The load, displacement, cracks and final failure mode are obtained. Hysteresis curve, skeleton curve, energy dissipation capacity, ductility coefficient, stiffness degradation and other performance data are calculated and analyzed. A proper nonlinear simulation method is presented where the key parameters in simulation are calibrated by experimental data. The FEM results and experiment results are in good agreements. The result shows that the squat shear wall in nuclear island has high loading bearing capacity, but the ability to resist deformation is limited, so the wall tends to be brittle. The reduction of axial compression ratio can enhance the energy dissipation capacity of shear walls. The variation of axial compression ratio has little impact on the stiffness degradation.