bluuebunny/crossref_metadata_2025_split huggingface dataset:
Development of an Extra-Trees-Based Program to Calculate Gamma-Ray Exposure and Energy Absorption Buildup Factors
Authors: Chen Zhitao , Liu Yongkuo , Hu Jifeng 
Published: November 2024

Abstract
The calculation of gamma-ray energy absorption and exposure buildup factors is studied in this paper. The GP formula requires logarithmic interpolation calculation about the photon energy and the equivalent atomic number of the material in the interpolation calculation of the fitting parameters, which leads to a complicated process of calculation of the buildup factor. In order to solve this problem, a program for buildup factor calculation based on the Extra-trees regression algorithm has been developed in this work, which can calculate the energy absorption and exposure buildup factors by inputting the photon energy and the elemental composition of the material. The code was validated across a wide range of materials by calculating the buildup factors for a variety of mixtures/compounds, including water, air, concrete, human tissues, and bricks. The tests show that the results of the program are in good agreement with the values from the ANS, GP formulas, and MCNP program. Moreover, the accuracy of the Extra-trees regression model is improved in this paper by using the GP formula to generate the buildup factors of elements 4–92 as a training set compared to using ANS data.
Multi-Mesh Approach for Geometry-Independent Neutronic/Thermal-Hydraulics Coupling Analyses Using RMC and Ansys Fluent
Authors: Zhao Xingyu , Liu Guodong , Huang Shanfang , Chen Qiaoyan , Luo Hao , He Ying , Hou Junren , Wang Kan 
Published: November 2024

Abstract
This paper presented a multi-mesh approach for high-fidelity and geometry-independent neutronic/thermal-hydraulics (N-TH) coupling calculation capable of various nuclear reactor analyses using Reactor Monte Carlo code (RMC) and Computational Fluid Dynamics (CFD) code Ansys Fluent. Verifications and coupling analyses were performed on various models.
Customized RMC/Fluent interfaces were developed for direct data transfer through hdf5 files without text files to achieve higher transfer efficiencies and precisions. This study presented a multi-mesh approach to realize mesh mapping between solvers by separating the data statistical Q-mesh and T-mesh from the computational cells. Q-mesh tallied the power distributions in 3D structured forms using mesh tallies of RMC, while T-mesh tallied the temperatures and densities in Fluent model by the UDFs. Both solvers read data through Q-mesh and T-mesh without directly interacting with each other, and thus the geometry-independent coupling was realized. This study also considered relaxation method during iterations to stabilize coupling solutions and improve convergence.
Verifications on a unit cell model showed great agreement to previous studies with a maximum fuel temperature deviation less than 0.6%. Coupling calculations performed on a light water-cooled unit cell model and a helium-cooled assembly model both converged in 3 iterations. The results showed that the coupling effects were dominated by the moderator density variation. The maximum fuel temperature of coupled results was 17.6% higher than that of decoupled results, and 7% higher than that of the coarse model, showing the conservative nature of high-fidelity coupling, and thus the high-fidelity coupling was beneficial for reactor safety.
The Effective Thermal Conductivity Model of Dispersion Fuel Elements Based on Effective Matrix Approximation
Authors: Qiu Tianchen , Sun Jun 
Published: November 2024

Abstract
The modular high temperature gas-cooled reactors use fuel elements with huge number of Tri-structural isotropic (TRISO) fuel particles dispersed in graphite matrix, which contributes to the inherent safety and enables high temperature process heat applications. Featured by complex geometry and material combinations, a lumped model of effective thermal conductivity (ETC) is practically used to describe the heat transfer performance and temperature distribution of the fuel element the in thermal hydraulic design and safety analysis of HTGRs. Since the fuel particles release fission or decay heat, classical analytic ETC models fail in predicting the temperature distribution inside the fuel elements, without denoting the inner heat sources. Previous researches have explained the definition of ETC model with inner heat sources. Some new ETC models that indicate the function of distributed inner heat sources in the dispersion fuel element were also proposed.
In this paper, an ETC model to predict the volumetric average temperature of dispersion nuclear fuel elements is proposed. The proposed model considered the influence of distributed inner heat sources in the composite media by introducing effective matrix approximation. The ideal dispersion fuel pebble, which is simplified from the element currently used in pebble bed HTGRs, is taken as an example to show the application of the proposed ETC model. And the validity and conservativeness of the proposed ETC model are preliminarily revealed by finite element numerical calculation results.
A Preconditioning Method Based on Strongly Implicit Procedure for the Multi- Diagonal Equation
Authors: Tan Bo , Zhang Haojie , Wen Yutong , She Ding 
Published: November 2024

Abstract
Ellipsoidal partial differential equation is usually discretized into a multi-diagonal equation when using the finite-difference or finite-volume numerical method. The neutron diffusion equation in reactor physics and the thermal conduction convection equation in thermal fluid calculations are elliptical partial differential equations. Therefore, the efficient solution of multi-diagonal equation is of great significance to nuclear engineering calculations. GMRES algorithm has been widely used to solve multi-diagonal equations, but the computational efficiency of this method depends on the effect of preconditioning method. This paper combines the SIP preconditioning method with GMRES to solve steady-state thermal conductivity problems for flat cartesian and cylinder. The calculation results show that the SIP preconditioning method can reduce the iterations of multi-diagonal equations, without increasing the additional preconditioner’s memory compared with the traditional ILU(0) preconditioning method.
The Experimental Research of Integrative SMR Using TEG for Power Generation
Authors: Wang Suhao , Li Yong , Xiao Hui , Guo Liang , Li Ying , Lou Ruifan , Tang Hairong 
Published: November 2024

Abstract
A Thermoelectric Generator (TEG) is a device that directly converts heat into electricity based on the Seebeck effect. An Integrative Small Modular Reactor (SMR) utilizing TEG as a power generator is characterized by the direct conversion of thermal energy into electric energy, absence of moving parts, an extended lifespan, and a reduction in reactor space. This paper presents an experimental research study on an SMR-TEG test facility using high-temperature Liquid Metal as the TEG’s hot side and normal-temperature water as the TEG’s cold side. The experiment yielded electric output power and thermo-electric conversion efficiency data for the TEG testing module. These results can be conventionally utilized for TEG material evaluation and reactor design optimization in subsequent stages of research and development.
Model Evaluation and Parallelization of Subchannel Analysis Code for Full-Core Pin-by-Pin Analysis
Authors: Yu Zhonghao , Xiong Jinbiao 
Published: November 2024

Abstract
Subchannel analysis, as a common tool utilized in nuclear reactor core thermal-hydraulic calculation, is essential to establish high-fidelity high-resolution reactor core simulator which shows growing potential in nuclear reactor operation and monitoring. However, thermal-hydraulic models and correlations concerning the effect of spacer grids with mixing-vane currently employed in subchannel analysis codes are imprecise due to narrow applicability and insufficient accuracy, resulting in large calculation errors. Meanwhile, vast amount of data generated from pin-by-pin solution analysis brings challenge to current non-parallel calculation process which leads to low calculation efficiency. In this paper, an optimization for subchannel analysis program on the capability to run full-core scale subchannel analysis on existing PWRs with pin-by-pin resolution is carried out. Both OpenMP an MPI method is adopted to optimize the subchannel program and the result of adopting OpenMP is discussed. The correlation of single-phase turbulent mixing factor and transverse crossflow friction factor considering the effect of spacer grids are coded into the program. The evaluation and validation of the models are carried out through sensitivity analysis and results comparison with the reference database of subchannel benchmark PSBT. The new models’ effects on the flow field in the axial direction are also studied.
Development of Low-Pressure Loss Separator for Advanced BWRs
Authors: Povolny Antonin , Katono Kenichi , Ishida Naoyuki , Fujimoto Kiyoshi , Yasuda Kenichi , Kito Kazuaki 
Published: November 2024

Abstract
In Boiling Water Reactors, steam water mixture is produced by the reactor core. Steam separators separate the steam from the water downstream the core. The steam separators are designed to provide good steam-water separation ratio (limit the carry-over ratio of liquid water in steam and carry-under ratio of steam in water) without inducing too large pressure loss on the recirculation flow (affecting core flow rate or pump power). The steam separators use flow rotation and centrifugal force to separate water from the steam. Good steam-water separation requires stronger rotation within the barrel, which usually increases the pressure loss presenting a trade-off. Many studies have aimed at both improving the separation performance and lowering the pressure loss. This work introduces recent activities of Hitachi-GE Nuclear Energy. A previously developed prototype was optimized to further improve performance as well as manufacturability by taking the advantage of quick prototyping with CFD modelling, 3D printer prototyping and half-scale air-water testing at room temperature and atmospheric pressure. The final design prototype will be confirmed with a test that is being planned in Hitachi Utility Steam Test LEading facility (HUSTLE) under nominal BWR operating conditions. Confirming the separator performance will allow adopting the separator in future advanced BWR plants or potential replacement of older separators in operating BWR plants. The adoption of the low-pressure loss separator would lead to improved plant performance and safety.
Practice and Validation of Harmonic Method for HTR-PM Power Distribution Monitoring
Authors: Li Jinpeng , Guo Jiong , Li Fu , Wei Chunlin 
Published: November 2024

Abstract
Due to the pebble-bed reactor design, core of the HTR-PM consists of a stack of pebble fuel elements, which, together with the high temperature environment inside the reactor, prevents the installation of power detectors in the active zone of the core. Power detectors of HTR-PM are in a limited number within the concrete shielding wall, which is a certain distance from the active zone. As a result, detectors are not capable to give sufficient information and it is challenging to monitor power distribution in HTR-PM.
In this paper we reconstruct power distribution of various operating states of the reactor using harmonic synthesis method based on actual operating data of HTR-PM and readings of detectors at different positions. Afterwards, the VSOP program is used to track the reactor operation process and calculate the power distribution of various states. Comparisons are made between reconstructed results and VSOP’s results for verification. We notice that harmonic synthesis method is capable to realize power distribution reconstruction of high-temperature gas-cooled reactors, but reconstructive ability of the harmonic set is still a little bit insufficient for some special operating states, and it’s necessary to improve the ability and accuracy of the harmonic set and spatial response function for certain special operating states.
Reactivity Feedback Effects of Fuel and Grid Plate Expansion in Small Gas-Cooled Fast Reactors
Authors: Ma Yugao , Zhang Suyi , Deng Jian , Wang Jinyu , Liu Yue , Ding Shuhua , Zhang Muhao 
Published: November 2024

Abstract
Small gas-cooled fast reactor is considered as an important candidate for micro nuclear power, with the lightweight, compact structure, and high thermal conversion efficiency. To achieve an ultra power-to-mass ratio, the reactor typically employs high-enriched uranium fuel designs. However, as the relative enrichment of 238U in the fuel decreases, the fuel temperature Doppler reactivity feedback effect in reactors may transport from negative to zero or positive. Furthermore, the fuel expansion reactivity feedback coefficient is small and instable, because of the low thermal expansion coefficient of UO2 and the fragmentation of fuel pellets after high burnup. Discussing the negative reactivity feedback by changes in the core structure is crucial to ensure the negative reactivity feedback for the entire core and system operation stability.
The large temperature difference (∼300°C) at the inlet and outlet of small gas-cooled fast reactor leads to increased differential thermal expansion between the fuel restraining holes in the top and bottom grid plates, which result in the expansion named as free-flowering of fuel rod bundles. This phenomenon will lead to a decrease in the effective fuel density and an increase in the neutron leakage. Ultimately, the expansion reduces the excess reactivity of the core and manifests as a grid plate expansion negative reactivity feedback effect.
In this paper, the reactor Monte Carlo neutron transport code RMC and the finite element analysis software ANSYS Mechanical are used to model and analyze the Prometheus space reactor. The focus was primarily on discussing the effect of grid plates expansion on reactivity feedback and power distribution. The results showed that grid plates expansion had a minimal impact on the axial power distribution in the core, as the difference of axial distribution between before and after free-flowering expansion is 0.05% in average, and the difference of radial distribution is less than 0.81%. Furthermore, the reactivity feedback induced by the free-flowering expansion of the grid plates is at the same magnitude of that caused by the fuel expansion, which ranges between −0.3 ∼ −0.4 pcm/K. Thus, the grid plates expansion reactivity feedback coefficient should not be neglected in small gas-cooled fast reactors.
Study on Analytical Model of Permanent Magnet Eddy Current Retarder for High Temperature Gas-Cooled Reactor
Authors: Wu Hongyu , Yan He , Dong Yujie , Diao Xingzhong 
Published: November 2024

Abstract
The permanent magnet eddy current retarder serves as the main rod drop speed-limiting device in the control rod drive mechanism of the pebble-bed modular high temperature gas-cooled reactor. It can dissipate the kinetic energy of the control rod drop by generating eddy currents in the rotating conductor disk. To achieve fast calculation and analysis of the damping torque, an analytical model was established for a typical simplified structure of the permanent magnet eddy current retarder. Firstly, the three-dimensional structure was simplified into a two-dimensional model, and a two-dimensional magnetic circuit was established based on the magnetic equivalent circuit method, then the magnetic reluctances and fluxes of the circuit were calculated. Secondly, based on Faraday law of electromagnetic induction law and Ampere’s circuital theorem, the eddy current-induced magnetic field was solved, and the damping torque was calculated based on energy conservation. Furthermore, a three-dimensional correction was made. Finally, the analytical model was validated through three-dimensional finite element simulation. The results indicate that the analytical model can effectively calculate the magnetic flux density distribution in the air-gap and the torque speed characteristics. Within the general parameter range, the variation characteristics of torque with structural dimensions can be accurately analyzed. In addition, the temperature influence can be effectively analyzed within the range of 20∼200 °C.
The General Quadrant Meshing Approach for Constructive Solid Geometry Based Method of Characteristics
Authors: Guo Jian , Zhu Guifeng , Yan Rui , Zou Yang 
Published: November 2024

Abstract
The constructive solid geometry (CSG) method is widespread in codes based on Monte Carlo and the method of characteristics (MOC). Unlike just modelling the material regions in the Monte Carlo method, the MOC needs to split the material regions into small flat source region (FSR) meshes. The main obstacle standing ahead in using CSG methods for geometric modeling of MOC is the meshing of complex regions. Unstructured triangular mesh is introduced in MOC codes such as DIAMOND and ThorMOC to mesh on complex geometry. The geometry modeling ability is indeed greatly enhanced by using unstructured triangular mesh. However, the unstructured triangular mesh requires a small amount of initial overhead to generate and refine the mesh. Furthermore, small regions, like fuel rod cladding in PWR, need a lot of triangle meshes, which lead to very small track spacing and huge memory requirements. As a result, a general quadrant meshing (GQM) technique for complex geometry modeling with CSG is proposed in this paper and has been implemented in the 2D MOC. The GQM bounds the target CSG grid with a rectangle. By setting number of grids on the horizontal and vertical edge, the rectangle is then converted to a quadrant mesh. The grid that is in the CSG grid will be added as the sub-cells. The grid that is out of the CSG grid will be ignored. And the grid that crossed by the CSG grid boundaries will be added as the sub-cells after merging all the CSG grid surfaces. The volume of grid that crossed by the CSG grid boundaries is recomputed by the Monte Carlo method or the integration method. The point coordinates are random generated in the Monte Carlo method and the coordinates are the center of each quadrant mesh in the integration method. If the number of points is greater than 106 in the bounding rectangular cell, the volume relative error is less than 10−2 under most circumstances. A performance comparison between triangular unstructured meshes and the GQM is presented in this paper. The results of CANDU-6 bundle problem demonstrates the superiority of the CSG quadrant meshing method with respect to the unstructured triangular mesh. In the CANDU-6 bundle problem, the scalar flux and keff of GQM are nearly the same as triangular unstructured mesh while the grid number is only slightly more than half of the triangular unstructured mesh. To verify the accuracy of the 2D MOC code developed in this paper, many other transport benchmarks with different geometries, including BWR and PWR lattice, etc. were performed. Results of all the tested problems show good agreement with the references.
A Text Intelligence-Based Approach for Automatic Generation of Fault Trees in Nuclear Power Plants
Authors: Xiao Xingyu , Liu Songlin , Zuo Zhiyong , Chen Peng , Qi Ben , Liang Jingang , Tong Jiejuan 
Published: November 2024

Abstract
Fault Tree Analysis (FTA) is an indispensable tool in high-stakes industries like nuclear power for conducting thorough risk assessments. However, the development of fault trees for Nuclear Power Plants (NPPs) is often marred by the necessity of interdisciplinary and intricate knowledge, posing a significant hurdle for non-experts. This necessity for specialized knowledge limits the wider adoption of FTA across various sectors. In response to these challenges, this study introduces the Nuclear Large Language Model Fault Tree Generator (NuLLM-FTG), an innovative solution aims at streamlining and augmenting the FTA process. Central to our approach is the implementation of large language models (LLMs) supervised fine-tuning (SFT) techniques. Specifically, we have developed a uniquely textual data structure, meticulously crafted to encapsulate the distinct characteristics of fault trees. To enhance the precision of our methodology, a comprehensive dataset containing several thousand examples is constructed for SFT. NuLLM-FTG’s performance is subject to a thorough evaluation. This process involved unraveling the “black box” nature of the model, allowing for an in-depth examination of performance enhancements, particularly in terms of horizontal conversational pattern alignment and vertical fault tree knowledge evolution. The practicality and effectiveness of NuLLM-FTG are corroborated through online experiments involving real operators, coupled with evaluations conducted by domain experts. Additionally, the applicability of our method in real-world scenarios is demonstrated through its integration with the Risk Spectrum (Version 14.0), thereby confirming its effectiveness and robustness. Crucially, our findings indicate that the finely-tuned NuLLM-FTG attains a level of performance comparable to experienced fault tree professionals across various metrics, including professionalism, completeness, and satisfaction. Notably, under specific conditions, our model outperformed GPT-4, and the utilization of an English-language corpus within our model proved to be more effective. Ultimately, our proposed method facilitates novices’ involvement in FTA.
Finite Element Analysis of Rupture Pressure of Alloy 690 Stream Generator Tubes With Defects
Authors: Zhang Bin , Jiang Nai-Bin 
Published: November 2024

Abstract
In order to predict the load-bearing capacity of stream generator tubes with defects, this paper proposes a finite element method for analyzing the rupture pressure of SG tube. The simulation is based on the mechanical performance test of 690 alloy, and establishes a plastic constitutive and ductile damage model of 690 alloy. The finite element analysis software ABAQUS is used to establish a finite element model of tubes with defects, and the rupture pressure of tubes with different defect types and crack sizes is analyzed. Comparing the finite element prediction with the model formula calculation, the results match well. Therefore, this method can be used to analyze the rupture pressure of tubes under more complex geometric conditions such as U-bend tube or multiple defects.