bluuebunny/crossref_metadata_2025 huggingface dataset:
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

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

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

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

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

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

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

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

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

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.
Study on Reactor Cavity Cooling System of HTR-PM in DLOFC Accident With TIN-CAVCO Code and SPECTRA Code
Authors: Xu Xinsheng , Ye Yiyang , Zheng Yanhua , Marek M. Stempniewicz
Published: November 2024

High Temperature Gas-cooled Reactor (HTGR) is recognized by the international nuclear power industry as a safe type of reactor. The modular pebble-bed high temperature gas-cooled reactor uses reactor modules with relatively small thermal power between 200–600MWt, utilizing the high-temperature resistance that can be achieved by coated fuel particles. Institute of Nuclear and New Energy Technology (INET) of Tsinghua University has developed a commercial-scale 200 MWe High Temperature gas-cooled Reactor Pebble-bed Module project (HTR-PM). The decay heat can be removed by natural mechanism, such as thermal radiation, heat conductivity, natural convection, thereby guaranteeing the safety of reactor under accident and eliminating the possibility of core melting. A passive Reactor Cavity Cooling System (RCCS) has been designed for HTR-PM to ensure the safety of the Reactor Pressure Vessel (RPV) and reactor cavity under normal operation and accident conditions. RCCS transfers heat from pressure vessels to the atmosphere through thermal radiation, heat conduction, and natural convection, and the system utilizes passive mechanism with high safety and reliability. The RCCS of HTR-PM has been designed as three independent sets. According to the preliminary design, two sets work can guarantee the safety of the PRV and reactor cavity.

In this paper, the working performance of RCCS under depressurized loss of forced cooling (DLOFC) accident has been analyzed by INET with TIN-CAVCO code, and also by NRG with SPECTRA code. The analysis results of these two codes are in good agreement, proving the applicability of both codes in analysis of the HTGR as well as its RCCS. During the accidents, the maximal fuel temperature keeps below the limitation of 1620°C. The results also prove the excellent performance of the RCCS. The RCCS meets the design requirements, that is the normal operation of two sets can effectively ensure the safety of the pressure vessel and the reactor cavity during the accident. Even if only one set works normally, the decay heat still can be carried out effectively and the RPV temperature will not exceed the limitation of 425 °C.
A Preliminary Study on the Transport of Critical Radionuclide Polonium in the Marine Environment After Source Term Release
Authors: Zeng Junyang , Wang Yuqing , Zhang Minghao , Ni Muyi
Published: November 2024

Regional Ocean Modeling System (ROMS) was employed to model the transport of radionuclide Po-210 in the ocean over a one-week period, simulating the scenario of a hypothetical nuclear accident. The simulation incorporated considerations for the radioactive decay of Po-210, various release modes, and the combined impact of fifteen tidal constituents. The initial release concentration of Po-210 was set at 10000 unit/m3, with the release point situated near a coastal nuclear power plant in Guangdong Province, China.

The concentration distribution analysis revealed that, driven by ocean currents, Po-210 initially moved towards the northwest coast from the release point. Subsequently, it followed a southwesterly trajectory along the coast, impacting the entire coastal region of western Guangdong Province. Five days postrelease, the influence of Po-210 extended to the South China Sea.

Analyzing the concentration distribution and results from three coastal concentration detection stations revealed consistently high concentration values in coastal waters, indicating a significant impact of polonium on the coastal environment. Notably, high Po-210 concentrations in the waters surrounding Port X (the estuary) in Guangdong Province suggested a potential impact on inland rivers, directly affecting nearby residents. Cross-sectional analysis unveiled that Po-210 exhibited rapid vertical mixing, a phenomenon closely associated with tidal influences and water depth. Comparing simulation results with and without tidal effects demonstrates that that the rising tides exerted an inhibitory effect on Po-210 transport, while ebbing tides had a promoting effect.
Steam-Water Two-Phase Flow Testing Under Nominal Pressure and Temperature Conditions for Natural Circulation Performance of Chimney in BWRX-300 SMR
Authors: Furuichi Hajime , Povolny Antonin , Katono Kenichi , Yasuda Kenichi , Kito Kazuaki , Heck Charles
Published: November 2024

The BWRX-300 is a water-cooled, natural circulation, small modular reactor developed by GE Hitachi Nuclear Energy in cooperation with Hitachi-GE Nuclear Energy. BWRX-300 is equipped with passive safety systems and leverages the design and licensing of USNRC-certified ESBWR. The BWRX-300 uses a tall chimney above the core to provide natural circulation head to drive sufficient core flow and ensure fuel is operated within its thermal limits. Thus, calculated void fraction in the chimney is a key quantity for estimation of the natural circulation core flowrate. BWR transient analyses are performed with Transient Reactor Analysis Code, GE version (TRACG) which has been qualified for calculating thermal limits in BWRs. This extent of qualification includes the evaluation of core flow rate and evaluation of void fraction in chimney or upper plenum of BWRs. To support extending the TRACG qualification of void fraction evaluation to the large-diameter chimney of BWRX-300, additional test data were obtained with steam-water two-phase flow test and the void fraction modeling and calculation in TRACG were validated. The test was conducted in Hitachi Utility Steam Test LEading facility (HUSTLE) using a chimney test section, which was down-scaled to around 0.5 m in diameter. Distributions of the void fraction and the gas-liquid interfacial velocities in the chimney were measured by differential pressure gauges and the Wire-mesh Sensor (WS) for nominal BWR operating conditions of 7 MPa and 290 °C and flow conditions scaled from conditions of BWR-300 chimney at a range of reactor power levels. The average void fraction in the chimney estimated by the TRACG model and compared with the HUSTLE data. Consequently, the qualification of the void fraction estimation of the TRACG model was extended to the large-diameter chimney of BWRX-300.
Simulation Study on the Transient Operating Characteristics of Natural Circulation Coupled With PCM Energy Storage Module for the Passive Containment Cooling System
Authors: Huang Yufan , Wang Han , Chen Weixiong , Yang Xiaohu , Li Shaodan , Yan Junjie
Published: November 2024

To achieve a further reduction in pressure peak during accidents, phase change materials (PCM) are utilized in conjunction with natural circulation in the passive containment cooling system (PCCS). The dynamic simulation model of PCCS has been developed using Apros software. This article focuses on a comparative analysis of cooling capacity, considering various thicknesses and types of PCM modules. The results indicate that employing a 4 mm PCM module decreases the pressure peak in the containment from 1.85 MPa to 1.68 MPa. Additionally, the impact of thickness on the maximum power of the heat exchanger is constrained by the lower thermal conductivity of the PCM. Another type of PCM with the same latent heat capacity but lower thermal conductivity maintains pressure peak suppression performance. However, it results in a diminished pressure drop performance during the continuous condensation of steam. As a consequence of increased accumulation of non-condensable gases at the bottom, the heat transfer for steam condensation at the corresponding position is impeded.
Neutronics Study and Conceptual Design for Small Sodium-Cooled-Fast-Reactor With Annular Fuel
Authors: Ye Bin , Ala Ayodeji A.
Published: November 2024

This paper presents a conceptual design for a small Sodium-Cooled-Fast-Reactor core with annular fuel through the analysis of neutronic character, various power strategies and control rod control system. The design work of the conceptual core is based on BN800. Take the universality of fuel assembly into account, the hexagonal fuel assembly cladding of the conceptual core is consistent with BN800. The inner and outer diameter of the annular fuel are 4.0mm and 11.8mm separately, and the mixed oxide (MOX) fuel enrichment from the inside to the outside of the core are 19.5%, 22.1%, and 24.7%, respectively. The number of fuel assemblies in the inner/middle/outer region of the core is 109/102/156. The finite multiplication factor (keff), Neutron flux (Φ), neutron energy spectrum, effective delayed neutron fraction (βeff), sodium void reactivity effect (αv), and fuel burnup performance parameters of the reactor are all studied, furthermore, the control rod performance with different absorbing materials (boron carbide (B4C), hafnium diboride (HfB2), and hafnium hydride (HfH1.62)) are also compared and evaluated. This research work provides a reference for the optimization design of dual-cooled annular fuel for small sodium-cooled fast reactor.
Study on Heat Transfer Behavior of a Rectangular Particle Bed With Volumetric Heating
Authors: Junlang Wen , Yuto Kamada , Kosei Yokoyama , Tatsuya Matsumoto , Wei Liu , Koji Morita , Yuya Imaizumi , Hirotaka Tagami , Kenichi Matsuba , Kenji Kamiyama
Published: November 2024

During core disruptive accidents (CDA) in sodium-cooled fast reactors, the core debris bed could be formed due to the interaction between the molten fuel and coolant. At the late stage of CDA, the heat removal of decay heat will be achieved mainly by the natural circulation of the molten stainless steel to the surrounding environment. To investigate the coolability of fuel debris bed, a rectangular experimental system was built in which the particle bed was volumetrically heated via direct current heating. The experimental apparatus consists of a Nd-Fe-B magnetic particle bed immersed in water and a water pool above it, which simulate disrupted solid fuel and molten steel, respectively. Magnetism ensures the good contact between particles; therefore, the stable volumetric heating is truly achieved, which is similar to the characteristic of fuel debris with internal decay heating. Computer code simulations with the reactor safety analysis code SIMMER-IV were performed to help understand the heat transfer characteristics and to validate the applicability of the newly embedded momentum exchange function (MXF) models. The time variation of average temperatures of particle bed and water pool, and their difference were evaluated to compare with the experimental results. The comparison results show the most applicable MXF model under current experimental conditions and well reproduce the influence of water pool height in the experiments.
Inventory Control Strategy and Dynamic Modelling of a 1 MWe HeXe Brayton Cycle Reactor System
Authors: Li Xinyu , Guan Chaoran , Chai Xiang , Liu Xiaojing
Published: November 2024

The small mobile nuclear power system using the HeXe Brayton cycle as its energy conversion system has promising application potential. Among the research, control strategies get a lot of attention, which makes variable operation of the system possible. Among various control strategies, inventory control is widely used in the load-following process as it enables the system to remain high efficiency during load regulation. However, compared to other control strategies, it has a slower response time, which needs to be improved. In this study, a one-dimensional dynamic model of a simple HeXe close Brayton cycle reactor power conversion system, as well as its main equipment, has been proposed. Focusing on inventory control, the effects of initial tank volume and pressure on load range that the control strategy could meet were investigated. The open loop simulation of coolant mass in the loop was carried out. After that, to overcome its disadvantage, inventory control has been modified for rapid load regulation by combining with valves, so that the optimized inventory control method could meet the rapid load change process while maintaining high efficiency and safe operation of the system.
A Study on the Burnup Effect of Silver Self-Powered Neutron Detector
Authors: Sang Yaodong , Deng Bangjie , Liu Shiyu , Zhang Qingmin
Published: November 2024

Self-Powered Neutron Detector (SPND) has been widely used in reactor core for in-core neutron flux monitoring. As a typical delayed type of SPND, the burnup behavior of Ag SPND is more complicated than Rh SPND and VSPND, because Ag is composed of two natural isotopes 107Ag (51.839%) and 109Ag (48.161%) with distinct neutron capture cross-sections. This will lead to change in neutron sensitivity of Ag SPND during operation, which in turn affects the monitoring of neutron flux. In this work, the burnup effect of Ag SPND utilized in CE-type assembly was studied by using SPNDSignal toolkit. By solving the burnup equations, we calculated the composition change of the emitter material over 10.5 operation years. Ag SPNDs with different burnup rates were modeled in SPNDSignal, and the components contributing to neutron sensitivity were calculated and analyzed. After 10.5 years of use, 7.84% of 107Ag and 21.08% of 109Ag are burned up, and correspondingly it leads to neutron sensitivity decrement by 16.68%; however, the sensitivity to external gamma is negatively small and remains almost unchanged during the whole operation period. On the other hand, signals’ timing analysis shows the rising time of the total signal current to a step neutron flux increases from 115 s to 125 s. All of these suggest that the neutron sensitivity and delay compensation coefficients need to be recalibrated during the use of Ag SPND.
Preliminary Investigation on Burst Behaviors of Zr-Based Cladding Under Simulated LOCA Conditions
Authors: Duan Zhengang , Yuan Bo , Wen Qinglong , Chen Kang
Published: November 2024

The Fukushima accident exposed the shortcomings of zirconium-based fuel cladding in withstanding severe loss of coolant accidents (LOCA), thereby accelerating the development of accident-tolerant fuel (ATF). Over the past decade, extensive technical exploration has led to chromium (Cr) coated zirconium cladding emerging as the primary technological route slated for commercial application soon. In current study, the burst behaviors of zirconium-based claddings were systematically investigated under simulated LOCA conditions. The initial internal pressures ranged from 2–10MPa, and the heating rates were 14 °C/s and 28 °C/s. Although research on Cr-coated cladding is still ongoing, preliminary findings from the zirconium-based cladding analysis indicate that at a consistent heating rate, the burst time decreases with increasing internal pressure, and the burst temperature decreases accordingly. Conversely, at a constant initial internal pressure, the burst time decreases as the heating rate increases, but the burst temperature elevates.
Effect of Heat Treatment on Microstructure Characteristics of Hot Worked Zr-2.5Nb Alloy Tubes
Authors: Yin Jianming , Gao Bo , Luan Baifeng , Zhang Yan
Published: November 2024

Characterization and analysis techniques including XRD, FEG-SEM with attached EBSD detector, TEM with attached EDS detector were employed to investigate the microstructure characteristics of a Zr-2.5Nb alloy tube. The samples as received from the hot worked tube were treated under 550°C/5.5h annealing, and 850°C/1h solution heat treatment followed by 550°C/5.5h aging (SAHT) respectively. Hardness test was also performed for predicting of the strength changes. Results show that there was residual β-Zr existed in the hot worked tube, and after annealing at 550°C, decomposition of the residual β-Zr in the hot worked tube happened and the low angle grain boundaries reduced. There is no obvious texture memory effect after solution at 850°C flowed by water quenching, however the effect can be well observed after aging at 550°C. Nb-rich second phase particles appeared during aging process, with their locations near the boundaries inside primary a grains. Hardness of the hot worked tube didn’t show obvious change after annealing, while after solid solution and aging treatment (SAHT), the hardness increased significantly both in LD and TD directions.
Machine Learning Approach to Analyzing Data in Japanese BWR Chemistry Database
Authors: Yamazaki Gaku
Published: November 2024

Understanding the relationship between the dose rate of the primary loop recirculation (PLR) piping and chemistry parameters is the beginning of developing the techniques of reducing the radiation exposure in boiling water reactors. In this study, a regression model was built using a machine learning method, gradient boosting, to identify the relationship between the dose rate of the PLR piping and several chemistry parameters.

Categorical Boosting, EXtreme Gradient Boosting, and Light Gradient Boosting Machine were used to build a regression model, with the dose rate of the PLR piping as the target variable and chemistry parameters as the explanatory variables. The coefficient of determination and root mean squared error across these three models are nearly identical, demonstrating the similar accuracy of different gradient boosting methods.

SHapley Additive exPlanations values were used to interpret the internal structure of these models and rank the importance of parameters associated with the dose rate of the PLR piping. All three models highlighted the same top-tier parameters, indicating that comparing regression models could be beneficial for enhancing the robustness of parameter extraction. Only within certain ranges was a strong relationship found between the dose rate of the PLR piping and these chemistry parameters, implying non-linear relationships and optimal ranges corresponding to lower values of the dose rate of the PLR piping.
Research on Fluid-Structure Coupling of Spent Fuel Storage and Transport Basket Based on Immersed Boundary Element
Authors: Zhang Feifan , Liu Yu , Lu Daogang , Xie Fei
Published: November 2024

Nuclear power plants are intricate systems housing a multitude of equipment crucial for their functioning. Among these components, the spent fuel storage and transport basket plays a pivotal role in securely containing highly radioactive spent fuel. Particularly under seismic conditions, the spent fuel storage and transport basket encounters challenges associated with fluid dynamics. Fluid permeates its interior, and external fluid dynamically envelops it, introducing complex fluid-structure coupling effects. Traditional fluid-structure coupling models, characterized by closed structures, tend to adopt overly conservative approaches. Additionally, transient Computational Fluid Dynamics (CFD) simulations, relying on intricate modeling, pose computational challenges due to their intensity. In response to these challenges, this research introduces an innovative approach by leveraging the immersed boundary method. The research not only proposes a new methodology for studying fluid-structure coupling phenomena but also utilizes the OpenFOAM software for numerical simulations. The study provides comprehensive insights into the variation of lift force, contributing to a deeper understanding of the fluid-structure interactions. To validate the proposed approach, the numerical results undergo a rigorous comparative analysis with both experimental data and previously modeled outcomes. In conclusion, through a meticulous and comprehensive numerical investigation, the findings not only offer a foundation for structural dynamic analysis but also hold implications for enhancing the seismic resilience of crucial components within nuclear power facilities.
Accident Progression Analysis of an Offshore Floating Nuclear Power Plant Under Heaving Motion
Authors: Ui Atsushi , Ikeda Yutaka , Arai Takahiro
Published: November 2024

Offshore Floating Nuclear Power (OFNP), in which a proven commercial light water reactor (LWR) is mounted on a floating structure such as an oil drilling rig has been proposed by the Massachusetts Institute of Technology. Based on this concept, the Japanese industry is conducting research and development on the feasibility of an OFNP with a boiling water reactor (BWR) on a floating structure and operating 30 km off the coast of Japan. As part of this study, safety analysis was conducted to confirm whether the heaving oscillation affects accident progression in design extension conditions during the oscillation of OFNP, which is assumed to occur in stormy weather and other high-wave conditions. Prior to the analysis, the terms including gravitational acceleration in the equations of motion of the reactor system analysis code RELAP5/MOD3.3 were modified to handle acceleration during oscillation. In addition, experiments corresponding to the oscillation were conducted under the same conditions of approximately 290°C and 7.2 MPa as the BWR core, and the experimental results were used to validate the code. After confirming that the code was able to reproduce the void fraction variations of the experiment, uncertainty analysis of accident progression, such as TQUV and ATWS, including various heaving conditions, was performed to analyze the effects of the oscillations.
A Brief Introduction to Several Important Design Optimizations of Passive Nuclear Power Plants in China
Authors: Ma Baisong , Miao Zhengqiang , Ma Yuanhua
Published: November 2024

Passive nuclear power technology was developed by Westinghouse in the United States and introduced to China around 2005. Currently, four passive nuclear power plants have been operating safely and stably in China for about 20 reactor years. However, some design defects have also been exposed. Therefore, a series of design optimizations have been carried out to improve the reliability and ease of operation and maintenance in passive nuclear power plants. This paper briefly introduces the several important design optimizations implemented and planed in Chinese passive nuclear power plants. We hope it can provide useful reference for other countries in the construction and operation of the passive nuclear power plants. (1) A gate has been added in refueling cavity. It creates conditions for equipment maintenance during refueling operations, which require the water level in the Reactor Pool to be lower than the level of flange of the Reactor Vessel. (2) A spare Variable Frequency Drive (VFD) has been added to solve the problem of unreliable long-term operation of the VFD for Reactor Coolant Pump (RCP). A 50Hz RCP has been selected for the plants under construction. (3) The anchor point of the support for Main Stream Pipelines has been adjusted from the Auxiliary Building in Nuclear Island to the adjacent wall of the First Bay so as to control Main Control Room (MCR) noise. The noise source has been also reduced by reducing the main stream flow rate. (4) A Spare Glove Box has been added, which is standby with the Grab Sample Panel. According to the characteristics of the sampling object, the sampling pipeline shall be reasonably separated to reduce or avoid the mixing of the sampling liquid, so as to improve the sampling representativeness.
Preliminary Study on Laser Inertial Confinement Fusion Reactor Energy Transfer
Authors: Li Xinze , Chen Ronghua , Zhang Bingqian , Zhang Kui , Zhang Dalin , Tian Wenxi , Qiu Suizheng , Su Guanghui
Published: November 2024

This study focuses on energy transfer within the implosion target of a Laser Inertial Confinement Fusion Reactor (LICFR). The study involves an analysis of heat transfer and tritium production within the Tritium Breeding Blanket (TBM) and the energy cycle of the secondary loop.

The research is based on an operating condition of 200MW. Super critical carbon dioxide (S-CO2) is utilized to cool the first wall of the TBM. Lead-Lithium (PbLi) is employed as both a coolant and tritium breeder material. With the characteristics of this dual-cooling setup, a multi-stage Brayton cycle is established and analyzed. The calculations couple the Monte Carlo code OpenMC with a self-developed heat transfer model. The study investigates the S-CO2 and PbLi dual-cooling blanket models under the 200MW operating condition. Based on the results of these calculations, optimizations are made to the structure and coolant parameters, and a multi-stage Brayton cycle is designed accordingly.

The research results show that the nuclear-thermal coupling model is capable of providing preliminary calculations. Tritium production and energy extraction from the blanket are influenced by both the dimensions of the reactor cavity and the breeding zone.
Three-Dimensional Temperature Field Monitoring of Heat Pipe Reactor Core
Authors: Wang Shidi , Zheng Youqi , Wang Xiayu , Fan Lunhe , Xiao Bowen
Published: November 2024

The heat pipe reactor is a new type of nuclear reactor based on the principle of heat pipe. It adopts the concept of solid core design, and transfers the fission heat through the matrix to the heat pipe by heat conduction. This type of reactor exhibits good inherent safety and single point failure resistance, and is an ideal candidate for space nuclear power systems and transportable nuclear power supply. This paper studies the core temperature monitoring method of heat pipe reactor based on three-dimensional temperature field reconstruction method. The reconstruction method is based on the value of thermocouple arranged inside the matrix. In each assembly, there are 6 thermocouples. They are used to monitor the temperature of matrix and to predict the temperature of cladding and fuel pins. The temperature field reconstruction method reconstructs the whole temperature field through a small number of discrete measuring points of thermocouple, and then expand the measurement data in large quantities. Using this method, the heat pipe failure and cladding failure were predicted, the specific location of the failed heat pipe is determined by monitoring the relative temperature change of the thermocouples at different positions.
The Variation Law of Aerosol Distribution Coefficient at Different Condensation Rates and its Impact on Heat Transfer of Tube Bundles
Authors: E. Xinnuo , Gu Haifeng , Chen Junyan , Wang Hui , Zhou Yanmin , Sun Qingyang , Yu Jianqun , Wang Xiao
Published: November 2024

When the PCCS (passive containment cooling system) operation, the aerosols in the containment along with the steam condensation migration to the condensate, will remain in the heat exchanger tube or by the condensate carried to the ground inside the containment or pit. In order to assess the change rule of aerosol distribution coefficient and the degree of influence on the PCCS heat transfer, this paper carries out an experimental study on the phenomenon of removing aerosols from the PCCS, and the experimental results show that the higher the rate of vapor condensation, the larger the aerosol distribution coefficient; lateral comparison of the aerosol distribution coefficient, in the experimental process of the aerosol distribution coefficient per unit of time is more stable, which indicates that the behavior of aerosols in the liquid phase and the wall of the tube between the liquid phase and the behavior is a stable process; the heat exchanger tube cooling tube is a stable process. A kind of stable process; heat exchanger tube condensation heat exchanger quantity change fluctuation is very small, there is no significant trend of change, the effect of aerosol on the condensation heat exchanger characteristics is not significant.
Research on Heat Transfer Calculation of Ellipsoidal Melt Pool
Authors: Liu Peng , Zhou Tao , Zhang Haolei , Liu Xiaofang
Published: November 2024

The in-vessel melt retention (IVR) technology, as one of the typical technologies to cope with severe accidents in third-generation nuclear power, has been widely used in pressurized water reactor (PWR) nuclear power plants. Based on the computational study of melt pool heat transfer in VVER-1000 under severe accidents, a steady-state heat transfer model of melt retention inside the elliptical lower head pressure vessel was established, and the wall heat flux at each place of the ellipsoidal lower head wall, the thickness of the oxide crust formed by the liquid-phase oxidic layer being cooled and solidified, the residual wall thickness of the pressure vessel, and the failure of the pressure vessel were computed and analyzed. The calculation results show that: the heat flux at the bottom of ellipsoidal pressure vessel lower head is the smallest, the heat flow distribution on the wall surface from the ellipsoidal part to the cylindrical part is discontinuous, and obvious heat focusing effect is observed in the metal layer; the thickness of oxidic crust decreases with the increase of axial height of ellipsoidal melt pool, and the higher the volumetric heat release rate of oxidic pool is, the smaller the thickness of oxidic crust is; the wall of the pressure vessel starts to melt at a certain place of the lower head, and the thickness of wall thins with the increase of heat flux. The wall thickness decreases with the increase of heat flux, and the ablation of the pressure vessel wall increases to the maximum value due to the drastic increase of heat flux in the metal layer area, and the thinner the metal layer is, the smaller the wall thickness of the pressure vessel in contact with the metal layer is, and the greater the probability of failure of the pressure vessel is; therefore, the wall surface of the pressure vessel in contact with the metal layer is the most dangerous position in the case of a serious accident.
Investigation About Classification of Stresses due to Radial Deformation Induced by Design Pressure Applied to the Tubes of Square Pitch Multi-Perforated Portion of Thick Plates
Authors: Duranton Pascal , Bernion Marie , Amzil Aboubakr , Jolly Divjot , Benrabia Abdelhak , Da Silva Tony
Published: November 2024

The design of thick Tube Sheets (TS) of a level 1 heat exchangers in the nuclear industry is a precise and complicated process, and one needs to consider the effects of numerous perforations. It must be noted that the residual stresses from the tube welding and expansion processes are not considered in this study.

The present paper focuses on the stresses induced by the radial deformation due to internal design pressure applied on the inner skin of the tubes expanded into the multi-perforated portion of plates. It is assumed that the stresses due to welding, hydraulic expansion and/or roller expansion have no impact. Indeed, pressure in the perforations tends to expand the multi-perforated portion of the plate. This radial deformation is resisted by the unperforated rim, the surrounding components and by the ligaments themselves.

The present paper studies the radial effect of the pressure in the tube expanded in the Tube Sheet in order to suggest a classification of the induced stress as either primary or secondary.

The multi-perforated portion of plate is often represented by a homogenous equivalent material model, developed to replicate the mechanical behavior of the multi-perforated plate while being far less computationally expensive. These methods consider only the primary to secondary pressure differential across the Tube Sheet in the numerical calculation and not the internal pressure in the tubes expanded in the multi-perforated portion of plate. It can be introduced in the stress results by additional post-processing, usually conservatively or excluded if considered as secondary stresses.

During the Tube Sheet service life, primary stresses, together with, to a lesser extent, secondary stresses, should be limited to avoid the risks of excessive deformation. To assess the risk of excessive deformation, a limit load analyses is performed.

The Finite Element Method (FEM) used is an explicitly modeled perforated zone of a square pitch Tube Sheet with its surrounding components. The tubes having not been explicitly modeled, two approaches are considered to represent the radial load applied to the holes of the Tube Sheet without considering the stiffness of the expanded tubes.