bluuebunny/crossref_metadata_2025 huggingface dataset:
Study on the Weights of Influencing Factors Contribution to Background Count Rate in Liquid Scintillation Measurement
Authors: He Shuijun , Tang Lijuan , Liang Manchun , Li Ke , Wang Jihong , Du Xiaochuang
Published: November 2024

The background count rate is a key factor to MDA and accuracy of measuring gross alpha and gross beta in water sample using liquid scintillation analysis. In this paper, the weights of the measurement environment, the material of scintillation vial, the volume of scintillation cocktail, type and volume of sample solvent to the background count rate were studied in Tsinghua University Laboratory and China Jinping Underground Laboratory using liquid scintillation counter. The weight of the external measurement environment is more than 60% to the background count rate in Tsinghua University Laboratory. The wights of the volume of the liquid scintillation cocktail, pure water and 0.05M nitric acid are linearly positive correlation to the background count rate. Pure water and 0.05M nitric acid were used as sample solvent, the weights of the volume of the two kinds of sample solution to the background count rate are comparable. At the same volume, the weights of pure water and 0.05 M nitric acid to the background count rate are lower compared to weight of the liquid scintillation cocktail to the background count rate. Compared with low-potassium glass scintillation vial, plastic scintillation vial has a lower weight to the background count rate. The results can provide data support to reduction of background count rate of liquid scintillation measuring gross alpha and gross beta in water samples, optimization of measuring condition and promotion of detection capability.
Research on Nuclear Power Plant Equipment Layout Method Based on Genetic Algorithm
Authors: Su Jincheng , Jia Xiaopan , Zhao Xiaoshan , Hao Dong
Published: November 2024

Based on the process of nuclear power plant equipment layout, this paper proposes a design method for automatic layout of nuclear power plant equipment in the form of solving a two-dimensional box packing problem based on genetic algorithm. By considering factors and requirements such as the overall dimensions of the nuclear power plant equipment, the process system to which it belongs, the maintenance space of the equipment, radiation levels, etc., a numerical model of the equipment process room is established. Taking factors such as the transportation path of the nuclear power plant equipment layout plan as the main indicators, establish an evaluation system for the nuclear power plant equipment layout plan. Based on the solution to the two-dimensional box packing problem, the nuclear power plant equipment layout process is simplified and a mathematical model for nuclear power plant equipment layout optimization is constructed. The advantage of genetic algorithm in solving optimization problems is used to realize the automatic layout of nuclear power plant equipment and obtain the optimal layout plan. Programming simulation is carried out according to the above method and flow, and under the set equipment layout constraints and genetic algorithm parameters, the automatic layout scheme design of nuclear power plant equipment is carried out and the equipment layout scheme is provided. The relative position of the equipment and the direction and location of the main channels reflected in the plan can provide a design basis for subsequent plan design.
Design Management Platform Maturity Research and Improvement Application
Authors: Shi Haili , Hu Yuefei , Zhao Wenzhao
Published: November 2024

Since its implementation in 2013, the Teamcenter design information management system has served as a comprehensive design management platform for the Nuclear Engineering Research and Design Institute. Over the years, this platform has significantly enhanced the informatization and digitalization of China’s third-generation nuclear power design, equipment manufacturing, and engineering construction. It ensures the advancement, safety, and economic efficiency of the Hualong One project. Nuclear power design and information technology are deeply integrated, allowing various departments to accumulate extensive experience in adapting the platform to suit the business processes, rules, and data of engineering management. However, there is a recognized need for improvements in design management and platform optimization. This paper proposes the establishment of a “design management platform maturity evaluation system” to assess the platform’s effectiveness and visualize evaluation data. Through the definition of evaluation indicators and the establishment of an Analytic Hierarchy Process (AHP) table, this study develops a maturity evaluation system for the design management platform. It analyzes key platform elements via expert scoring to perform the maturity assessment. Subsequent improvements are applied in the CIMS3.0 system, enhancing the safety, efficiency, and convenience of data querying and statistical analysis. This includes features such as quick access to statistical reports and comprehensive data queries, thereby improving the quality and efficiency of design management.
Irradiation-Thermal-Mechanical Coupling Behaviors of a Typical TRISO/SiC Fuel Element Based on Actual Distribution
Authors: Liangjie Zhang , Ping Chen , Shichao Liu , Haoyu Wang , Chong Wei , Xiaoqiang Pan
Published: November 2024

TRISO (Tristructural isotropic)/SiC Fuel Elements is developed and considered a kind of high-temperature gas cooled reactor (HTGR) fuel, it helps to improve the safety, competitiveness, and economics of commercial nuclear power. This study focuses on a typical fuel element, which is manufactured by ceramic SiC and contains more than 10000 surrogate TRISO particles. Then the actual position information of particles is applied in finite element analysis. The finite element model, considering comprehensive coupling behavior (including irradiation-induced dimensional change, irradiation-induced creep, thermal expansion, fission gas pressure, heat generation and conduction, elasticity and plasticity, etc.) and particle to particle/particle to matrix interaction, were developed based on the COMSOL platform and accurately verified. The verification process mainly based on the IAEA benchmark of single TRISO particle, a three-layer model and the fission gas release result using BISON. Based on the single particle model, a more complicated FEM is considered. It provides detailed modeling and analysis of the performance of each layer, which may occupy a huge amount of computing resources. Therefore, representative volume element (RVE) is selected, modeling the geometric structure with symmetry. The result indicated there is a significant high stress in the densely packed areas of fuel element, which increased the failure possibility of particles. By analyzing the stress field, temperature field and deformation of the model, it is indicated that in order to reduce the failure possibility of TRISO particle, the particle clusters should be avoided during fabrication process.
Study on Sodium Extraction and Transport at Monju
Authors: Isobe Yuta , Tanigaki Takanori , Tone Kohei , Joboji Yuya , Matsui Kazuaki , Obata Ikuhito
Published: November 2024

The prototype fast breeder reactor “Monju” is a loop-type, sodium-cooled nuclear reactor in Japan. Its operator, Japan Atomic Energy Agency (JAEA) started decommissioning of Monju in 2018, and the first phase of the decommissioning was completed in FY2022 when all the fuel assemblies were removed from the reactor core and Ex-vessel storage tank (EVST) and transferred to the spent fuel pool (SFP). In FY2023, we started the second phase, and are currently working to remove neutron shields and other core elements from the reactor core and EVST, to put them in SFP. We are also dismantling water and steam system such as turbine generator, condensers and feed water heaters, etc.

MONJU has a primary sodium cooling system for the reactor core and its fuel, and a secondary sodium system that transferred heat. Since the sodium in the primary system is radioactive while the sodium in the secondary system is not, we had developed plans to drain/extract non-radioactive sodium first. According to these plans, we defined the sodium extraction area first, to set up International Organization for Standardization (ISO) tanks for non-radioactive sodium and developed a route for extracting sodium from the existing tanks to the ISO tanks. New piping for these operations will be installed, so that the sodium can be extracted using existing gas systems and electromagnetic pumps. We are ready to take appropriate measures to prevent sodium leakage and continue to develop methods, ensuring safety, to achieve efficient extraction and transport of the sodium.
Numerical Analysis of Thermal Stratification for Upper Plenum in Pool-Type Sodium Fast Reactor
Authors: Xing Mingdi , Cheng Daoxi , Ma Xiaoyao , Zhai Weiming , Zhou Ping , Yu Song , Li Ruizhi , Gao Weilong
Published: November 2024

During the scram transient of China demonstration sodium fast reactor, thermal stratification phenomenon is observed in the upper plenum. The paper simulates the thermal stratification which occurs when hotter sodium from the core outlet enters the upper plenum along with colder sodium from the DHX entering the bottom of upper plenum using ANSYS commercial code. The results illustrate three transient stages of thermal stratification: In the first stage, hotter sodium accumulate in the top of the plenum while interface continues descending; simultaneously colder sodium from the DHX enter into plenum forming another thermal stratification phenomenon with rising interface until both interfaces meet forming a stable interface near the core outlet; In the second stage, as the flow rate and temperature at the core outlet gradually decrease, the hotter sodium cannot fully reach the top of the upper plenum, and another interface is formed at the top. In the third stage, the thermal stratification interface at the top gradually disappears, and the final thermal stratification interface is formed near the core outlet. Furthermore, it is also concluded that the interface descend rate decreases with the increase of Richardson number.
Optimization Design and Application Analysis of HLW Glass Container Under High Level Liquid Waste Separation Strategy
Authors: Duo Ma , Meng Wei , Jing-Yang Wang
Published: November 2024

This paper analyzes the “high heat release” logistics source term under the strategy of spent fuel reprocessing and high-level liquid waste separation. It calculates the annual production of high-level waste glass in a reprocessing plant with an annual processing capacity of 800 tons under different limiting conditions. It is found that according to the current design scheme, the thermal power of a single glass solidified cylinder is the determining factor for the production of high-level waste glass. Based on this, this article optimizes the design of strengthened thermal conductivity for high-level waste glass containers at La Hague post-treatment plant in France, and conducts thermal calculation analysis under its application conditions using ANSYS software to optimize the design scheme. The main findings of this study are: 1) Under the high level radioactive separation strategy, by strengthening the thermal conductivity of high-level waste glass containers, the theoretical maximum production of high-level waste glass can be reduced by 47.4% (reaching a limit of about precious metal content); 2) Adding a thermal barrier in the high-level waste glass container can reduce the maximum temperature of the high-level waste glass by 79 °C under certain conditions; 3) Through preliminary quantitative analysis, the optimized high-level waste glass container design scheme can reduce the production of high-level waste glass by about 30%.
Comparative Study of Cluster Analysis Methods in Dynamic Safety Analysis of Nuclear Power Plants
Authors: Suliman Mohamedelmogtabh Omer Elfadni , Wang He , Xiong Binfu
Published: November 2024

In nuclear power plant safety analysis, employing dynamic safety analysis methods such as Dynamic Event Trees (DET) to analyze individual initiating events often generates numerous scenarios, posing challenges in extracting useful information. This study utilizes clustering analysis methods to categorize these accident scenarios and extract their main features, thereby simplifying the analysis process and enhancing efficiency. Different clustering analysis methods can yield varying results, affecting the ability to reflect accident processes and extract key features. To ensure that clustering analysis better reflects accident processes and extracts the main features of accidents, this study compares and investigates two clustering analysis methods: one that considers only extreme parameter values of accidents and another that considers parameters throughout the entire accident process. In the method focusing solely on extreme parameter values, key parameters are extracted as clustering analysis parameters from each accident scenario. In contrast, the method considering parameters throughout the entire accident process treats accident scenario parameters as time-series data and applies Dynamic Time Warping (DTW) algorithm for time-series clustering.

Using the Steam Generator Tube Rupture (SGTR) accident as an example, this study applies both methods to analyze SGTR accidents and obtains corresponding results. The characteristics and differences between the two methods are analyzed.
Investigating the Effect of Supports on Reinforced Concrete Slab Subjected to Oblique Missile Impact
Authors: Kang Zuoyi , Okuda Yukihiko , Nishida Akemi , Tsubota Haruji , Itoh Masaharu , Li Yinsheng
Published: November 2024

As an outer protective engineering structure for nuclear power plant (NPP), the reinforced concrete (RC) containment wall at NPP should be evaluated for safety requirement related to external threats. Therefore, much attention has been paid to investigate the impact resistance ability of the RC slab structure subjected to missile impact. In these studies, most of them focused on local damage of RC slab structures struck by a rigid missile along the normal direction of the impact surface, while only few studies have focused on oblique impact. So, we have conducted a series of impact tests under different impact conditions covering impact angles, stiffnesses of missiles and so on. The objective is to figure out the different impact behaviors of the RC slab structures. We also intended to put forward a numerical method based on the test results, and to validate the proposed method through comparisons between experimental and numerical results.

In the test results, the reaction forces of RC slab were measured by using the load cells installed on the four corners of RC slab’s back surface. In the oblique impact, a special support structure was manufactured to fix the RC slab. It is speculated that the different supporting structure may influence the test results of reaction force. Thus, this work concentrates on the effect of the stiffnesses of the supports for oblique impact on the reaction forces of RC slab. Until now, static loading tests were carried out to confirm the stiffnesses of the components of supporting parts. The findings obtained from the comparison between the numerical results using the experimental values as spring constants in the finite element (FE) model and the reaction force measurement results will be reported.
A Study on Analysis of Actual Data in “Fugen” Decommissioning
Authors: Kodua Yuya , Nakamura Yasuyuki , Iguchi Yukihiro , Yanagihara Satoshi
Published: November 2024

Decommissioning Engineering Center (“Fugen”), has been proceeding with decommissioning work since receiving approval for the decommissioning plan in 2008, and is currently in the second step of dismantling the reactor peripheral equipment.

’Fugen” has a structure other than the reactor that is similar to a BWR, so it will be reflected in future commercial reactors.

In decommissioning, project management is important to select strategies and optimize processes, waste, safety, costs, etc. In Japan, it is expected that decommissioning work for nuclear facilities will become more serious in the future, and the knowledge and project management data obtained from the decommissioning work carried out at Fugen to date will be useful for planning.

Since the dismantling and removal of the main turbine equipment and the peripheral equipment in preparation for the reactors have been completed so far, the actual data has been organized and analyzed and compiled into unit work coefficients.

The project management data (Man-powers, amount of waste, etc.) collected during the decommissioning work carried out at “Fugen” is organized and calculated as a unit work coefficient (Man-powers required per unit weight (person/hour/ton)).

In summary, we analyzed the differences in workability due to differences in the working environment.
Improvement of the Thermal-Hydraulic Feedback Model in Bamboo-Core From Single-Channel to Sub-Channel
Authors: Zhang Hengrui , Li Yunzhao , Wang Sicheng , Li Yisong , Liang Yiling
Published: November 2024

The thermal-hydraulic feedback is critical for PWR-core analysis. The real flow distribution at the core inlet shows some non-uniformity and varies with operating conditions. Most core analysis codes assume that the core inlet flow distribution is uniform. To minimize the power distribution variations due to different core inlet flow distributions, the sub-channel model is employed to replace the single-channel model in the thermal-hydraulic feedback module of Bamboo-Core, which is a full-core reactor core analysis code for the Pressurized Water Reactor (PWR). The new sub-channel thermal-hydraulic feedback model is validated by comparing its calculation results with the measured data from the first three operational cycles of M310. When both uniform and non-uniform core inlet flow distribution were used to calculate the CNP650. The maximum bias between two core inlet flow distribution is 8.2%. Subsequently, the bias of 3D power distribution due to two different flow distributions was calculated. For single-channel thermal feedback, the maximum bias is 0.598%, while for sub-channel feedback, the maximum is 0.205%. The Changes in core inlet flow distribution have less impact on the results with sub-channel thermal feedback than single-channel thermal feedback. Therefore, the sub-channel thermal feedback is more suitable for the case of approximate processing of coolant flow distribution at the core inlet than the single-channel thermal feedback.
Research on Energy-Saving Operation Optimization of Circulating Water System Under CHP of Nuclear Energy Comprehensive Utilization
Authors: Wu Fang , Tang Te , Liu Weili , Li Yingnan , Wu Xinzhuang
Published: November 2024

Nuclear energy heating is the most elementary and the most essential mode of comprehensive utilization of nuclear energy. Based on the principle of variable operating conditions of the turbine, the principle of energy balance and mass balance of the thermal system, and the method of calculating condenser pressure, we investigated the impacts of extracting steam for heat supply on the distribution of steam and water parameters and thermal characteristics of the turbine, and determined the heat balance diagram of the turbine after changing the rated condensing condition to heat supply and the power-backpressure characteristics of AP1000 nuclear turbine. Based on the thermal characteristics of AP1000 nuclear turbine under rated heat supply condition, we investigated the effect of changing the number of operating fixed-speed circulating water pumps on the unit economics. Based on the optimal vacuum principle, we analyzed the frequency conversion renovation and operation optimization of AP1000 nuclear turbine circulating water pumps under heat supply condition, and identified the optimal renovation scheme and economic benefits.
A Comparative Analysis of CFD Approaches to Model Wire-Wrapped Fuel Bundle and Experimental Validation
Authors: Halim Ossama , Pucciarelli Andrea , Forgione Nicola
Published: November 2024

A wire-wrapped fuel bundle of 19 rods using Lead-Bismuth Eutectic (LBE) as a coolant was chosen for this CFD study and the predictions are compared to the NACIE-UP benchmark experimental data. The k-ω SST turbulence model is used to investigate the heat transfer of the LBE together with the adoption of a constant turbulent Prandtl number of 1.5. The thermo-dynamic properties of the LBE are modelled as temperature-dependent polynomials. Three different approaches have been adopted in this work to model the FPS, namely, Bare, Detailed, and the Porous Rod Bundle model. A model-to-model comparison is performed to understand the benefits, limitations, and accuracy of using porous models for representing wrapped wire. Furthermore, the integration of NACIE-UP benchmark experimental data into the comparative analysis adds an additional layer to validate the modelling approach.
Coupled Thermal-Hydraulics and Neutron Transport Calculations of Small Modular Reactor Using Serpent, OpenFOAM and SUBCHANFLOW Codes
Authors: Kořínek Tomáš , Zavorka Jiri , Lovecky Martin , Skoda Radek
Published: November 2024

Nowadays, several Small Modular Reactors are under development and high-fidelity neutron transport calculations with thermal-hydraulic feedback are a powerful tool for design and optimization of such reactors. The present work focuses on analyses of heavy water small modular reactor using coupled CFD, Monte Carlo and subchannel code. The coupling of Monte Carlo and subchannel codes was based on Picard iterations with power relaxation using stochastic approach. CFD simulations served for a generation of inflow conditions for subchannel code. The coupled pin-by-pin subchannel analyses was compared against fuel assembly level subchannel analyses. Additionally, a case with asymmetrical inflow conditions was analyses using coupled codes to reveal influence of inflow uncertainty on criticality and maximum fuel temperatures.

Comparison of coupled calculations on pin-by-pin level and fuel assembly level showed that a lower resolution led to underprediction of maximum fuel temperatures by 108 °C and maximum cladding temperature by 9 °C. Further, the low resolution led to the underprediction of multiplication factor by 410 pcm.

The asymmetrical inflow case led to a slightly higher coolant, cladding, and fuel temperatures in order of few degrees due to a lower flow rate in half of the core. Further, the asymmetric condition did not influence the reactivity as the multiplication factor changed insignificantly.
Constrained Nonlinear Model Predictive Control for High Temperature Gas-Cooled Reactor in Load Following
Authors: Zhu Yunlong , Dong Zhe , Huang Xiaojin
Published: November 2024

As the climate issue has become the focus of global attention, new energy with less carbon emissions have attracted the attention of the world, especially nuclear energy. High temperature gas-cooled reactor is an inherently safe reactor designed by the Institute of Nuclear and New Energy Technology of Tsinghua University and has been put into commercial operation. The nuclear steam supply system (NSSS) is the key system of the modular high temperature gas-cooled reactor for generating superheated steam, including a reactor core and a once-through steam generator (OTSG). In the process of nuclear power plant operation, power-level control is the key technology to ensure the NSSS to operate in the desired state. By adjusting and controlling the control rod speed, the rotation-rate of the main helium circulator and the valve-opening, the neutron flux, helium temperature and water supply flow rate are adjusted, and then the power-level maintenance, maneuvering and load tracking are realized. Considering the strong nonlinearity of NSSS, it is necessary to develop nonlinear power-level control methods. In this paper, model predictive control (MPC) is proposed to realize load following in the NSSS of high temperature gas-cooled reactor. MPC has the advantage of obtaining control input by iterating and solving optimization problems, which has a good application scenario for power-level control of nuclear power plants. The research in this paper has an enlightening significance for the application of MPC to the power-level control of high temperature gas-cooled reactor. In the future engineering application, facing the actual situation of multi-objective and multi-constraint conditions, the MPC iterative optimization idea can be used to obtain the control input that can make the nuclear power plant stable and meet the operation requirements.
Study on Single-Phase Flow and Heat Transfer Characteristics of Vertical Narrow Rectangular Channel Under Rolling Motion with Nonuniform Heating
Authors: Di Zhu , Rulei Sun , Yang Li , Sichao Tan , Ruifeng Tian
Published: November 2024

Plate type fuel elements have wide application potential in the integrated reactors due to their high power-to-volume ratio, low core temperature, high burnup and compact structure arrangement. The flow channels between adjacent fuel element plates are small, showing typical narrow rectangular structures with large aspect ratio. Due to the influence of fuel self-shielding effect, irradiation in the reactor and component arrangement, the plate fuel element has obvious uneven heat release in the transverse direction. As a result, the thermal hydraulic characteristics in the flow channel may be different from those in the uniform heating and conventional channels. In this paper, a typical vertical narrow rectangular channel is selected as the research object, and the channel section is 65mm × 2.23mm. Based on the computational fluid dynamics method, the mathematical and physical model of flow and heat transfer is established, and refined numerical calculations are carried out. The research parameters are that the inlet temperature is 38°C, the outlet gauge pressure is 0MPa, the heat flux is 15–240kW/m2, and the Reynolds number is 1000–14000. The research results show that the effect of transverse non-uniform heating on the heat transfer characteristics is greater than that on the flow characteristics. And with the decrease of Reynolds number and the increase of heat flux, the influence of non-uniform heating is more significant. Subsequently, the flow heat transfer analysis under the rolling and pitching conditions of the ocean is carried out to analyze the influence of non-uniform heating. It is found that Nusselt number and friction coefficient are macroscopic consistent with vertical state under forced circulation ocean conditions. However, under the same rolling period and Angle, the secondary flow intensity in the rolling state is greater than that in the pitching state. In the rolling state, different nonuniform heating forms have obvious influence on the secondary flow, in which the linear distribution form has the highest secondary flow intensity.
Method Research on the Model Optimization of Whole-Core Fuel-Assembly Bowing Based on 3D Variational Algorithm and the Measurement Values
Authors: Guo Lin , Zhang Kai , Wan Chenghui , Wu Hongchun
Published: November 2024

The phenomenon of fuel-assembly bowing would inevitably occur during operation, but the fuel-assembly bowing cannot be measured directly. Therefore, the challenge lies in obtaining the actual fuel-assembly bowing, which is a prerequisite for ensuring simulation accuracy. In this paper, a newly method for the model optimization of whole-core fuel-assembly bowing was proposed. The measurement values and theoretical whole-core model are combined in this method based on the 3D variational (3DVAR) and artificial neutral network (ANN) algorithm. In 3DVAR, the cost function is constructed to characterize the difference between the actual fuel-assembly bowing and theoretical model. The 3DVAR algorithm was applied to model optimization and the ANN algorithm was applied to establish the relationship between the whole-core model of fuel-assembly bowing and corresponding simulation of measurement values. For method verifications, the power-distribution measurements of a commercial NPP in China has been applied, and the model that need to be optimized was taken as the bowing amplitude of each fuel-assembly. The numerical results indicate that the relative errors of the power distribution in entire cycle could be reduced notably to meet the industrial criteria, and the maximum value is reduced from 13.4% to 7.7%. As conclude, the proposed method based on 3DVAR algorithm and the measurement values is efficient to optimize the model of fuel-assembly bowing.
Effect of Fission Products on Equilibrium U-Pu and Th-U Cycles for 16 Selected Reactors
Authors: Sabol Josef , Borys Francis , Křepel Jiří
Published: November 2024

Three papers related to the equilibrium fuel cycle in 16 selected reactors were submitted to ICONE 31 conference. In this second part, the impact of fission products on the equilibrium is discussed.

For each of the 16 reactors in the U-Pu and Th-U cycle and in the infinite lattice, the once-through depletion was calculated with a specific share of fission products. During the calculation, the total number of actinides was kept in balance in a closed cycle, only the proportion of fission products grew. In the second step, two possible approximations for multi-batch operation were compared: linear theory of reactivity and equilibrium approximation.

It was found that the equilibrium does not differ so much with burnup, significant changes were seen only in thermal systems in the U-Pu cycle. Since actinides were kept in equilibrium, only a decrease in reactivity was observed, which was purely linear for fast systems and nonlinear for thermal systems. When comparing both approximations, it turned out that they do not fit. In all cases, the equilibrium approximation stabilizes the different concentrations of long-lived and stable fission products, which results in a decrease in parasitic absorption and in an increase in reactivity. For this reason, the paper further analyses the concentration of selected fission products between each other.

All results were obtained using the EQL0D v4-procedure and with ENDF/B-VIII.0 nuclear data library.
Numerical Study of Gas Flow and Diffusion Driven by Buoyancy in a Multi-Openings Compartment
Authors: Jiang Jinghua , Hu Peizheng , Tong Lili , Cao Xuewu
Published: November 2024

During a severe accident in a pressurized water reactor, if the coolant pipe ruptures in a multi-openings compartment, a mixture of hydrogen, vapor, and air migrates and diffuses between compartments, driven by density disparities during the extended period following the cessation of coolant rapid discharge. The same phenomenon occurs during a “wet bypass” in a fusion reactor, a mixture of hydrogen, tritium (HT/HTO), vapor, and air migrates and diffuses between the suppression system, vacuum vessel, and Tokamak gallery. In the scenarios mentioned above, the buoyancy-driven flow resulting from the mixing of light and heavy gases will influence the long-term hydrogen risk and migration of radioactive materials. Previous research has primarily focused on studying the migration behavior of multi-component gas mixtures propelled by jet flow. An experimental device (A4Q-FB) with a multi-openings container was established, to clarify the influence of buoyancy-driven flow on gas migration. The test is carried out with well-controlled boundary conditions, in which, a mixture of helium and vapor with a mass fraction of 2:15 is injected through a vertically downward nozzle into the compartment filled with air. The injection was stopped after 11200 s from the beginning of the experiment, and other boundary conditions were controlled to be unchanged, and the data of the temperature and concentration of the gases inside the container and the flow velocity of the gas mixture in the openings were obtained. In this study, a three-dimensional analysis model for this test was developed, and the predictive accuracy for gas mixture migration and diffusion behavior was assessed. The results indicate a fundamental alignment between the simulation outcomes and the trends observed in the experimental data. This consistency suggests the reliability of the model in the macro-level predictions of phenomena and supports our understanding of the gas mixture migration under buoyancy-driven flow. The simulation results indicate a maximum relative error of 12% in predicting the concentration of vapor at the top and a maximum relative error of 20% in estimating the gas flow velocity within the top openings. The simulation also provides the 3D distribution of velocity and concentration in both container and openings.
Analysis on Permeation Leakage Through Rubber Sealing Ring for Radioactive Material Transport Packages
Authors: Wang Pengyi , Zhuang Dajie , Sun Hongchao , Chen Lei
Published: November 2024

To ensure the containment of radioactive contents, IAEA SSR-6 specifies graded requirements on released radioactivity of transport packages. The O-ring sealing rubber is usually used in the radioactive material transport container to achieve the containment of the contents and air-tightness requirements. However, there are two unexpected phenomena occurring with rubber material. The first one occurs during leakage testing of transport packages containing non-radioactive gaseous contents, helium which is typical tracer gas will penetrate through the O-ring to the side with smaller pressure, thus interfering with the true leakage rate, in addition to helium leakage caused by cracks or defects between the flange and the O-ring. The second unwanted phenomenon is radioactivity release through penetration process, for transport packages containing radioactive gaseous contents. The helium permeation goes through three main processes: dissolution, diffusion and desorption. This paper designed one experimental apparatus and utilized typical rubber material as sealing rings to observe and analysis the gas permeation process and mechanism. Based on the outcome of test results, addressing the issues arisen from permeation leakage, this paper proposed recommendations on improvements of gas leakage testing method and procedures, and one calculation model of radioactivity permeation release source term.
Evaluation of Three-Dimensional Initial Dynamics of the Decompression Wave at LB LOCA in the Advanced Supercritical Reactor ECC-SMART
Authors: Filonov Vladislav , Dubyk Yaroslav , Kovalenko Oleksandr , Filonova Yuliia , Fedorov Dmitriy
Published: November 2024

The paper presents an adaptation of the modelling approach to initial dynamics during decompression for an advanced small modular supercritical reactor, which was previously developed and applied for the VVER-1000 reactors. The practice of performing such an assessment in one- and two-phase formulations indicates that three-dimensional assessment using CFD is more realistic than similar calculations using system codes (e.g. RELAP5). The main feature of modelling decompression from a supercritical state to a subcritical state is the complexity of the transition through critical pressure, where the medium can go both into the two-phase region and into superheated vapor/liquid. These features require the introduction of some abstract functions to be able to calculate a supercritical medium in a two-component approximation. The paper represents the assessment of dynamic forces during the initial dynamics of LB LOCA using a representative CFD model of a perspective supercritical reactor concept. The initial dynamic analysis is performed for the following conditions: hot zero power reactor state, break of the inlet piping with maximum reference pressure. The structural integrity analysis of some critical components based on preliminary (pre-conceptual) ECC-SMART SMR design is done using CFD results.
Advanced Radioactive Material Removal System by Silver Zeolite (9) Performance Evaluation of Silver Zeolite AgX by High Temperature and Atmospheric Pressure Test Equipment
Authors: Ishikawa Yoshihiro , Endo Koji , Narabayashi Tadashi , Kawahara Yasuhiro , Watanabe Tomonori
Published: November 2024

Severe accidents at nuclear power plants produce a variety of materials such as inorganic and organic iodine, noble gases, and aerosols. Removal filters are taken against these radioactive materials by installing filter venting systems such as FCVS. The FCVS consists of a scrubber pool, metal fiber filter, silver zeolite, etc. Radioactive materials generated during a severe accident, water-soluble materials such as inorganic iodine are removed by the scrubber pool, granular materials and solids are removed by the metal fiber, and gaseous organic iodine is removed by the silver zeolite.

We have developed silver zeolite AgX as a radioactive iodine adsorbent produced during severe accidents, and it has been used in nuclear power plants in Japan. In response to requests for use in a variety of environments, we developed a radioiodine adsorbent AgR, which is different from AgX. AgR is a silver zeolite like AgX, but the difference is that AgX has strong ability to bind hydrogen and oxygen, while AgR has a weak ability.

To evaluate these silver zeolites, we fabricated a high-temperature test facility equipped with a scrubber pool, metal fiber filter, and silver zeolite, which is the same configuration as the actual facility, and evaluated the methyl iodide removal performance of the silver zeolite. The test was conducted under the assumption that methyl iodide-containing steam flows through the silver zeolite at room temperature (cold start), simulating a severe accident.

As a result, it was confirmed that AgX and AgR shoed high methyl iodide adsorption performance even at 100% steam.
Numerical Simulation of Pool Boiling With Two-Fluid Model and Grid-Resolved Heat Transfer Mechanisms
Authors: Petrovic Milan M. , Stevanovic Vladimir , Ilic Milica , Milivojevic Sanja
Published: November 2024

Most nucleate boiling simulations are based on the Eulerian modelling of bulk liquid-vapour two-phase flow and subgrid presence of all mechanistically partitioned heat transfer modes in fluid control volumes in contact with the heated wall. Such mechanistic partitioning heat transfer approach predicts the total heat flux from the heated wall towards the boiling two-phase mixture as the sum of (i) the single-phase convection heat transfer from the heated wetted wall not covered with the rising bubbles, (ii) the transient conduction caused by the rewetting of the heated wall hot spot after the bubble detachment, known as quenching heat flux and (iii) the evaporation heat flux. Here presented research introduces a new approach to the numerical modelling of pool boiling based on the grid resolved mechanisms of boiling heat transfer. The spots of bubble growth and the wetted areas are distinguished and two grid resolved modes of heat transfer are considered: the conjugate heat transfer from the heated wall to the rising bubble at the bubble footprint until the bubble departure and the conjugate heat transfer from the heated wall to the wetting liquid. The constituents of the heat transfer model at the footprint of the bubble growth are the bubble residence time and the bubble departure diameter. The presented model is validated by experimental data from the literature. It was shown that the model can predict the wall temperature transient at bubble footprint location as well as the mean wall superheating. Appropriate modelling of vapour generation at the discrete locations of the bubble growth enables good prediction of two-phase mixture pattern in the boiling pool, the void fraction distribution along pool height and swell level position. Further analysis includes the application of both subgrid and grid resolved modelling approaches to the simulation of pool boiling under high heat fluxes. It is shown that the subgrid wall boiling model does not predict adequately the wall temperature transient behaviour and the void fraction distribution in the boiling pool under high heat fluxes, while the grid resolved model provides plausible results. Presented simulations are obtained with the in-house computer code based on the modified Semi-Implicit Method for Pressure Linked Equations (SIMPLE) for the solving of mass, momentum and energy balance equation for each phase and appropriate closure laws for the prediction of vapour-liquid interface transport phenomena.
High Temperature Gas Velocity Profile Measurement Using Fiber Optic Hot-Wire Velocimetry
Authors: Balbier Christopher , Bucks Scout , Leoschke Matthew , Scurti Federico , Lee Saya
Published: November 2024

Most advanced reactors operate at high temperatures; however, there is a lack of local velocity measurement techniques that survive in the high-temperature, corrosive, and irradiation environments present in these reactors. In this experimental study, a Fiber Optic Hot-wire Anemometer (FO-HWA) using Rayleigh Backscattering is developed and applied to high-temperature airflow in a circular tube operating at temperatures exceeding 600 °C as the first step to developing a velocity profile measurement technique for high-temperature gas-cooled reactors.

The circular tube is made of transparent borosilicate glass to allow for flow visualization measurement techniques to be applied at the desired temperature ranges. Additionally, the flow channel will be equipped with thermocouples to monitor the temperature of the flow. Also, inlet conditions were well controlled by applying a nine-to-one contraction and a flow straightener, which generates a uniform inlet velocity profile. A Time-Resolved Particle Image Velocimetry (TR-PIV) system consisting of a high-speed camera capable of recording at up to 9,000 frames per second and a high-power continuous green (532 nm) laser is applied simultaneously to the flow to generate validation data. The flow rate was controlled to cover laminar, transition, and turbulent flow regimes (Re: 1,800, 5,500, and 7,300). High-temperature air velocity profiles are obtained at those different flow rates by the FO-HWA and compared with the PIV data and existing experimental data in the literature.

Considering the relatively thick sensor diameter compared with the traditional hot-wire sensors, the velocity data is post-processed to generate a time-averaged velocity profile, and the accuracy of the time-averaged velocity profile was discussed by comparing the results with the PIV data. The time-resolved capability of the FO-HWA is assessed. Also, the second moment of the temperature fluctuation measured by the FO-HWA and the velocity fluctuation measured by PIV were compared.

The FO-HWA technique demonstrated in this study will be applied to higher temperatures, targeting over 800 °C, found in helium-cooled reactors and potentially the sodium vapor core of heat-piped cooled microreactors. In the present setup, the oxidation issue is not thoroughly resolved, although acceptable stability of the measurement is achieved. This may cause an issue for the long-term operation of the current setup. However, in the realistic target conditions mentioned above in advanced reactors, with the same stainless steel capillary tubing protection of the current setup, the oxidation issue can be mitigated compared with the present experimental condition with airflow. Eventually, with additional coating/protection techniques, the FO-HWA concept can be applied to other high-temperature energy-transporting fluids such as molten salt that can be used in nuclear fusion & fission, and solar energy applications.
Safety Evaluation and Optimization of Nuclear Power Plant Under Adaptive Sampling
Authors: Li Linfeng , Zhang Zhen , Xu Anqi , Wen Ting , Dong Xiaomeng , Yang Ming , Chen Linxian , Liu Yong
Published: November 2024

At present, the safety analysis methods used worldwide in nuclear power plants (NPPs) are divided into Deterministic Safety Analysis (DSA) and Probabilistic Safety Analysis (PSA). The traditional DSA is based on a concept of Defense-in-depth, which lacks the correct grasp of the state of NPPs and the economic consideration. The static fault tree and event tree methods used in PSA analysis are difficult to simulate the dynamic timing sequence. Previous studies on the safety margin of NPPs have mostly focused on the evaluation methods for the perspective of regulation, but there are no consensus on how to support the risk-informed decision-making of nuclear power plant (NPP). In addition, traditional DSA often only considers the epistemic uncertainty and does not include the random uncertainty such as random failure of equipment. In this way, the estimated safety margin is not accurate enough. In order to meet the comprehensive improvement of safety and economy in the Generation IV reactor, this paper is based on the Risk-informed Safety Margin characterization (RISMC) methodology, and focuses on the issues such as the large amount of calculation and long calculation time due to the requirements of high-precision under this methodology. A surrogate model training method based on Support Vector Machine (SVM), K nearest neighbor algorithm (KNN) and adaptive sampling strategy is proposed. During the process, key uncertainty parameters are screened through sensitivity analysis and the parameter dimensions are reduced. The surrogate model was then applied to RISMC methodology and combined with the Best Estimate Plus Uncertainty (BEPU) analysis to calculate the safety margin of NPPs. Finally, a Small-break Loss of Cooling Water Accident (SLOCA) is used as a case to demonstrate this method. By comparing the computation time of the surrogate model combined with adaptive sampling with a high-fidelity thermal-hydraulic model, it was found that this method can significantly reduce computation time. This provides technical support for applying RISMC method to engineering operation of nuclear power plants.