bluuebunny/crossref_metadata_2025_split huggingface dataset:
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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.
Development of a Qualified Plant Nodalization for Steady State and Transient Analysis of Table Top Facility
Authors: Rozzia Davide , Makris Loukas Dikonimos , Cools Jan , Van Loy Tom , Verpoorten Tim , Van Tichelen Katrien 
Published: November 2024

Abstract
MYRRHA is a flexible, fast spectrum, pool type research reactor cooled by lead bismuth eutectic (LBE), under development at SCK CEN. The Primary Heat eXchangers (PHX) of MYRRHA are designed to operate with pressurized water as secondary coolant and are located in the reactor pool. The PHX design is based on a double-walled bayonet tube bundle, with intermediate Helium leak detection system. Three experimental facilities have been designed and constructed at SCK CEN with the aim to assess the performance of this innovative component:
The Leak Detection System facility has been designed, constructed and operated during the period 2017–2019 with the goal to perform a standalone investigation of the PHX Helium leak detection system.
The HEXACOM facility has been designed and constructed during the period 2018–2023 with the aim to investigate the performance of a full scale single PHX tube including the use of LBE from the primary coolant loop of the COMPLOT facility, which is already operational, as heat source.
The Table Top facility has been designed and constructed during the period 2019–2021 and has been operational since 2022. This latter facility is discussed further in the current paper.
Table Top is a 2.5 kW heated water/steam pressurized loop in open configuration that works with natural convective flow in the pressure range of [8–16] bara. The objective of the facility is to obtain a first estimation of the thermal resistance of the double walled PHX tube as well as provide experimental data for thermal hydraulic system code assessment and model development. The facility represents a single scaled down (in length) PHX tube and does not represent the primary LBE coolant, as heating is provided by electric heating elements. Three large experimental campaigns have been carried out including more than twenty-nine experiments.
A procedure to develop a qualified thermal hydraulic system code nodalization for steady state and transient analyses is introduced in the paper. The procedure is applied to two experimental tests on Table Top to develop and refine a “immutable and qualified” nodalization of the facility. The activity investigates the capabilities of the thermal hydraulic system code RELAP 5 mod 3.3 of representing the main phenomena which occur during a typical transient as well as a typical steady state operational mode of Table Top. Emphasis will be given to the need of representing the plant layout to within an acceptable range of accuracy as defined in the procedure. This is necessary to capture the physical behavior of the system numerically.
The SPIZWURZ Project - Experimental and Modeling Approaches of Hydrogen in Cladding Tubes Under Dry Storage Conditions
Authors: Weick Sarah , Herm Michel , Grosse Mirco , Kilger Robert , Rössger Conrado , Steinbrueck Martin , Stuckert Juri 
Published: November 2024

Abstract
When hydrogen penetrates into cladding tubes during operation of nuclear power plants, it may affect their mechanical stability. During the dry storage of spent nuclear fuel, the initially dissolved hydrogen precipitates as hydrides in the metallic matrix. This precipitation process induces a certain stress in the zirconium structure of the cladding tube. In the framework of the SPIZWURZ project, cladding tubes are charged with hydrogen and exposed to a certain pressure — either internal or external, in order to investigate the stress influence on the hydrogen diffusion and solubility under dry storage conditions. The relevant long-term bundle test that simulated dry storage conditions with a constant cooling period of 250 days has recently ended. As described in the paper, the preliminary calculations for the temperature development and the hydride state shall be compared to the real state of the bundle after the test. Therefore, various analytical methods will help to determine relevant parameters that will be used for the fuel rod performance code TESPA-ROD from the Gesellschaft für Anlagen- und Reaktorsicherheit (GRS), which shall be revalidated in this framework.
Three-Dimensional Numerical Simulation on DRACS and PRACS Heat Removal Ways of Natural Circulation Based on the PLANDTL-DHX Experimental Device
Authors: Song Haijie , Ma Xiangfeng , Zhao Haiqi , Zhang Yuhao , Lu Daogang , Zhang Xueyuan 
Published: November 2024

Abstract
The sodium-cooled fast reactor (SFR) experimental device PLANDTL-DHX adopts a loop-type arrangement. In the event of station blackout, a natural circulation is required to remove the decay heat of core. There are two main ways to remove the decay heat from the reactor core, one is the direct reactor auxiliary cooling system (DRACS), the other is the primary reactor auxiliary cooling system (PRACS). In the present work, the modular modeling and integrated calculation approach was employed to build the device model and simulate the thermal hydraulic phenomena in the whole primary loop under the conditions of DRACS and PRACS respectively. The feasibility of the modular modeling and integrated calculation method is verified, and the differences between DRACS and PRACS decay heat removal methods were also analyzed. By means of PRACS or DRACS for decay heat removal, both inter-wrapper flow (IWF) and intra-subassembly flow are evident at the core locations. The temperature at the outlet of the core and the upper plenum is significantly higher with PRACS than with DRACS. Compared with the PRACS decay heat removal method, a return flow phenomenon at the core location emerges the DRACS decay heat removal, which is more conducive to the discharge of core decay heat. The results can provide important numerical reference for the design and operation of different forms of decay heat discharge ways.
Impact of Narrow Gaps on Graphite Component Lifespan in Small Modular Molten Salt Reactors
Authors: Wang Qi , Zhong Yu , Yu Chenggang , Guo Wei , Chen Jingen , Wang Zhichao , Sun Qiang 
Published: November 2024

Abstract
In molten salt thermal reactors, graphite serves as both the moderator and the support structure, while also forming the flow paths for fuel and coolant, as well as the channels for control rods. The molten salt narrow gap flow channel consists of several small gaps that exist between graphite assemblies in the reactor core. And the fuel in the narrow gap channel will cause several modifications to the properties of the graphite and the primary channel, which will affect the life of the graphite assembly. Therefore, thoroughly examining the performance variations of graphite caused by the narrow gap as the reactor operates is critical. In order to investigate the influence of the narrow gaps on the graphite assembly lifetime in small modular molten salt reactors (SM-MSR), a high-precision neutronics physics and thermal hydraulics coupling method for the analysis of the narrow gap is established in this study. The effect of the presence of the narrow gap on the lifetime of graphite components was investigated by comparing the changes in performance of RCA components with different narrow gap widths. The results indicate that the presence of the narrow gap increases the asymmetry of temperature distribution at both ends of the graphite, which in turn aggravates the inhomogeneity of graphite axial deformation distribution. In addition, when the narrow gap channel width is sufficiently small, the molten salt fails to cool the graphite as effectively as it theoretically should. Such a situation is alleviated by increasing the width of the narrow gap.
Development of a Spatial Dynamics Model Based on Semi-Analytic Nodal Method
Authors: Cui Yong , Chen Jin-Gen , Wu Jian-hui , Guo Wei , Cai Xiang-Zhou 
Published: November 2024

Abstract
Dynamic calculation based on neutronics and thermal-hydraulics (TH) coupling model is one of the crucial aspects in reactor safety analysis and system simulation. With the continual development of the numerical model of reactor physics and the high-performance computers, now 3D spatial dependent kinetics calculation is feasible and has been already widely applied in core simulator. In this work, a neutronics/TH coupled spatial dynamics model has been developed based on the multi-group neutron diffusion theory in 3D Cartesian geometry for next-generation advanced reactor steady state and transient analysis. The spatial variables are handled by employing a non-linear iterative semi-analytic nodal method (SANM) while for the neutron kinetics solver the fully implicit scheme and exponential transformation are adopted. In the TH module, a single channel model and its steam table is implemented. To test the performance of the coupled model, both static and transient characteristics of the three benchmarks including 3D-LMW, NEACRP 3D and PWR MOX/UO2 core are obtained and analyzed in detail. Numerical results show that accuracy of the developed spatial dynamics model is similar to that of other high performance nodal codes.
Adsorption and Diffusion of Silver on SiC(111) and ZrC(111) Surfaces: A First Principles Study
Authors: Li Jiaxuan , Zhu Hongwei , Wang Taowei , Gao Zelin , Chen Xiaotong , Xu Gang , Liu Bing , Tang Yaping 
Published: November 2024

Abstract
First-principles calculation have been implemented to investigate the adsorption and diffusion behaviors of Ag atom on β-SiC(111) and ZrC(111) surfaces. The configurations of Ag adsorbed on FCC site of β-SiC(111) surface and ZrC(111) surface have strongest stability and the adsorption energies of these two configurations were −2.775 eV and −3.192 eV, respectively. The electronic structures of these two most stable configurations were analyzed by charge density difference (CDD) plots and particle density of states (PDOS), and the results shows that there are strong interactions between the Ag atom and surface atoms, indicating chemical bonds are formed between the Ag atom and surface Si atoms or Zr atoms. The surface diffusion barriers and minimum energy paths (MEP) were determined by climbing image nudged elastic band (CI-NEB) method. The results suggest that the energy barriers of Ag atom diffused from the FCC site to the neighboring FCC site are 0.166 eV for β-SiC(111) surface and 0.150 eV for ZrC(111) surface. The surface diffusion coefficients of Ag atom on β-SiC(111) and ZrC(111) surfaces at 1273 K are 4.035 × 10−8 m2/s and 2.503 × 10−8 m2/s, respectively. The surface diffusion coefficient results indicate that it is easier for Ag atom diffuses on β-SiC(111) than on ZrC(111) surface.