bluuebunny/crossref_metadata_2025_split huggingface dataset:
Flow Dynamics in a Fuel Bundle With a Particle Bed
Authors: Franz Scott C. , Wissinger Gordon J. 
Published: November 2024

Abstract
The pressure reduction in the RCS and temperature excursion of the cladding during a postulated loss-of-coolant accident (LOCA) can lead to cladding rupture. Fuel fragmentation can occur at any burnup, but the fuel fragment sizes decrease with increasing burnup, thereby increasing the potential for relocation and dispersal via the rupture opening. Research is ongoing to better understand fuel fragmentation, relocation, and dispersal (FFRD). However, the research to date is focused largely on understanding the mechanisms generating the fragmentation and quantifying the resulting fragment sizes and amounts within the fuel rod. There is a need to understand the consequences of the dispersal from the fuel rods. Some of the dispersed particles can settle onto a spacer grid below the rupture location forming a particle bed. Framatome is researching the behavior of fluid flow through and around a particle bed that has the characteristics of fuel particles that have been dispersed from rupture locations within a fuel bundle. Understanding the behaviors and changes in the flow patterns caused by the presence of the particle bed are of importance as they relate to the cooling behavior of the fuel pins during a LOCA transient. The presence of the particle bed can either divert steam and droplet flow away from the location near the particle bed region or may slow the flow of steam and droplets through the particle bed. A reduction of flow in the local region of the particle bed may adversely impact the cladding temperature during the transient. Tests for the benchmarking of flow patterns and cooling dynamics for these kinds of conditions are limited. Consequently, Framatome has performed full scale fuel bundle flow testing to provide needed benchmarking data. The testing includes bundle geometrical aspects under scaled LOCA conditions which include clad swelling, rupture locations, spacer grids, mixing grids, the presence of a particle bed, and entrance and exit effects. Multiple sizes of particles including mixtures of sizes are taken into account as well as multiple bed depths and different bed formations in the testing. Additional testing has been performed to establish baseline characteristics of fuel particle beds in general (fluidization behavior, packing fractions, etc.). The data collected has provided results leading to in-depth understanding of the flow patterns and pressure drops in and around the particle bed with a high level of detail. This data can then be used to develop and benchmark tools capable of simulating the flow patterns within a nuclear fuel bundle with a particle bed. An increased understanding of how steam and droplets interact with a particle bed in the geometrical conditions found in a nuclear fuel bundle is a requirement for developing new safety methods.
Manufacturing Challenges for Hot Isostatic Pressing Large-Scale Vessels
Authors: Sulley John , Stewart David 
Published: November 2024

Abstract
Hot Isostatic Pressing (HIPing) has been used by Rolls-Royce to successfully manufacture nuclear plant components such as: valves, piping, and pump casings, the majority of these components being manufactured from stainless steels, typically 316L. There are potential benefits to be gained by manufacturing large nuclear plant pressure vessels via the HIP process, such vessels commonly being manufactured from Low Alloy Steel (LAS) materials such as ASME SA-508. HIPing of large-scale LAS structures present specific challenges. This paper presents the challenges Rolls-Royce has experienced in their HIP vessel development work and means of addressing those challenges. It covers key steps of the HIP manufacturing process, from powder production, through can manufacture including leak detection and HIP consolidation. Of particular note are the risks of powder oxidisation and contamination, achieving the necessary material toughness, and can failure during the HIP cycle as a consequence of poor can design and fabrication. The paper shows examples of the results of failures, e.g. contaminated and unconsolidated powder, and how the risk of such failures can be reduced.
Thermal Performance Analysis of Radioactive Material Transport Containers
Authors: Wang Changwu , Zhang Yuhang , Lian Yiren , Chen Lei , Wang Zhipeng , Sun Qian , Zhuang Dajie 
Published: November 2024

Abstract
The safe transportation of radioactive materials is a critical component of nuclear safety, with transport containers acting as the primary safety barrier during these operations. Thermal performance analysis plays a critical role in evaluating the safety performance of these containers. In our study, we conducted thermal tests and simulation analysis on AF transport containers. The focus was primarily on comparing the simulated maximum temperature results of key components with the actual test outcomes, after simulating transport accident conditions. The comparison revealed a strong correlation between the simulation analysis results and the experimental data, with an error margin of less than 14.4%. These findings demonstrate that the combined application of simulation analysis and experimental methods for collecting temperature data of critical components within transport containers can effectively evaluate the thermal safety performance of radioactive material transport containers, thereby enhancing nuclear safety standards.
Effect of the Heat Transfer Correlation Choice on the Coupled Assessment of Energy Release and Stress-Strain State of an Elementary Supercritical Fuel Rod Under Nuclear Heating Conditions
Authors: Filonov Vladislav , Kovalenko Oleksandr , Dubyk Yaroslav , Filonova Yuliia , Fedorov Dmitriy 
Published: November 2024

Abstract
The paper presents a multiphysics procedure for estimating the parameters of a fuel channel under nuclear heating conditions. The idea of the approach is to determine the multiplying properties of an elementary fuel rod in the 4-group diffusion approximation, together with the estimation of the temperature field of both the supercritical coolant and the structural elements of the fuel rod. Group constants are prepared on the basis of a precision code that takes into account the design features of the reference fuel assembly. Several dozen different correlations are implemented into the procedure to calculate heat transfer. A concept of parametric energy release functions is proposed, which can be useful, for example, when modeling the core using CFD. The concept is based on the analysis of the influence of parametric temperature profiles (according to the classification of heat transfer regimes) on the effective multiplication factor. The paper presents the results of assessments of the fuel rods temperature state and thermal power depending on the selected reactivity margin and the selected HTC correlation. The stress-strain state of fuel rods was analyzed for each calculation option. The developed approaches are fast and can be used to optimize design decisions.
Analysis of the Effect of Random Bowing of the Fuel Pin on the Power Distribution in the Fuel Assembly
Authors: Závorka Jiří , Lovecký Martin , Škoda Radek , Monhartová Petra 
Published: November 2024

Abstract
One of the main approaches to ensuring nuclear safety in neutronic calculations is the consideration of engineering factors. These factors define the margin of the individual parameters of the power distribution in the reactor core up to their safety limit, which must be assumed in the design of the fuel loading and which is controlled during the operation of the nuclear reactor.
The resulting engineering factor consists of many components, and one component of this set is the subject of investigation in this paper, namely the evaluation of the fuel pin bow on the power distribution.
This study provides a fundamental investigation of the effect of the fuel pin bow on the power distribution in the fuel assembly without knowing the maximum bow specified by the fuel supplier or determined by measurement.
The methodological approach is based on randomly generated fuel pin deflections that simulate the fuel pin deflection/bow using a specially developed program, including a few conservative assumptions.
Development of an Onset of Nucleate Boiling Model and a Nucleate Boiling Heat Transfer Model Within the Fouling Layer
Authors: Wang Xiaowen , Liu Maolong , Cong Tenlong , Gu Hanyang 
Published: November 2024

Abstract
The fouling on the reactor fuel cladding surface affects the boiling heat transfer, and the boiling within the fouling involves both nucleate boiling and film boiling regimes. It is crucial to establish the boiling onset model to determine the boiling regime within the fouling layer. By simplifying the fouling layer as the porous media with steam chimneys, it is assumed that the cladding temperature is exactly the saturation temperature at the onset of nucleate boiling, and the energy still relies on convective heat transfer when nucleate boiling occurs. A new model is proposed to determine the onset of the nucleate boiling within the fouling. And the wall temperature at nucleate boiling is obtained by simple linear interpolation. The errors of the onset of the nucleate boiling and nucleate boiling model from the experimental results is within 15% and 20%, respectively.
High-Order Harmonics Calculation for the Power Reconstruction of 3D Pin-Scale Whole-Core
Authors: Xie Zhinan , Hao Chen , Yin Wen 
Published: November 2024

Abstract
The high-order harmonics of neutron diffusion equation is used in the reconstruction method to reconstruct the core power distribution. There are two methods available for the high-order harmonics solving: power iteration method with source correction (PIMSC) and Krylov subspace method. PIMSC was commonly used in the high-order harmonics calculation of the homogeneous assembly-scale neutron diffusion equation. However, this method cannot use the discontinuity factor theory, and the diffusion system is not equivalent to the transport system. Meanwhile, the convergence rate is influenced by the dominance ratio. In order to calculate the high-order harmonics of the 3D multigroup pin-scale whole-core neutron diffusion equation and guarantee equivalence between the neutron diffusion equation and the neutron transport equation, the 3D pin-scale generalized equivalence theory based coarse mesh finite difference (gCMFD) system is used in the high-order harmonic calculation, which is equivalent to the transport system using the discontinuity factor theory. The implicitly restart Arnoldi method (IRAM) is used to solve the 3D pin-scale gCMFD system. IRAM is a kind of Krylov subspace method which utilizes the implicit restart technology to improve the solving speed. To verify IRAM in high-order harmonics calculation, the results of C5G7 benchmark calculated by the PIMSC and IRAM are compared. The whole-core model of low-temperature heating reactor (LTHR) is used to test the calculating ability of pin-scale high-order harmonics and analyze the express ability of power. The numerical results indicate that IRAM has superior computational efficiency compared to PIMSC, being approximately 431 times faster. Additionally, for the low temperature heating reactor, the high-order harmonics solved by IRAM have good accuracy in the pin-scale power reconstruction.
Analysis of Shielding Structures of Radioactive Material Transport Package Based on Isotope Γ Radiation Properties in Radioactive Materials
Authors: Wang Zhi-peng , Wang Chang-wu , Sun Qian , Zhang Yu-hang 
Published: November 2024

Abstract
The transportation package is an important equipment for the safe transportation of radioactive materials. Many countries and IAEA have restricted the shielding performance of the transportation container of radioactive materials. The radiation shielding performance provided by different materials, and different structural designs is also an important research area in the field of radiation protection. Uranium and its isotopes are taken as analysis objects, and the overall radiation characteristics (γ spectra) of the contents are obtained by calculating method based on the γ spectra of different nuclides provided by EDNF database. This paper forms a design method for the shielding structure design process of packages, and provides reference and guidance for shielding design of radioactive goods transport cargo package.
Comparison of K-Solutions of Rectangular Flaw With Semi-Elliptical Flaw for PWSCC Growth Analysis
Authors: Hojo Kiminobu , Rudland David 
Published: November 2024

Abstract
The JSME Codes and Standards has a Code Case of K-solutions of a rectangular flaw for primary water stress corrosion cracking (PWSCC) growth analysis. The shape of axial PWSCC flaws in a dissimilar material weld (DMW) joint is relatively rectangular rather than semi-elliptical. This flaw will propagate selectively in the weld metal, which makes it a deep, short flaw. The K-solutions of the JSME Code Case have been provided considering the features of the PWSCC flaw conservatively. The same K-solutions for a Code Case of ASME Section XI were proposed and have been discussed in a Working Group of the ASME Sec. XI Standard Committee. In the Working Group, sample problems were proposed to confirm necessity of introduction of K-solutions for a rectangular flaw into the ASME Code, and the appropriateness of the code description of a draft Code Case based on the JSME Code Case. Especially, there was a discussion to avoid over-conservatism of the K-solution. In the sample problems, the K-solutions, that is Gi coefficients, of a rectangular flaw and a semi-elliptical flaw were compared for various flaws with a different flaw depth and length. The deeper and the narrower the flaw, the larger the difference of the Gi coefficients between a rectangular flaw and a semielliptical flaw, up to the ratio of 0.5 to 0.6. Also, PWSCC propagation calculations of both types of the flaws assumed in a DMW joint of a primary coolant pipe with double V groove, a hot leg with single V groove and a surge line with single V groove were performed by using the K-solutions of a rectangular flaw and a semi-elliptical flaw. These Code basis calculation results were compared with the result of the advanced FE (AFEA) analyses which simulated actual axial flaw growth behavior. In the sample problems, the initial flaw was a semi-circular flaw within a weld width or a rectangular flaw with a length of the weld width. The initial semi-circular flaw grew with a semi-elliptical shape and the initial rectangular flaw grew with a rectangular shape. As a result, the calculation results of semi-elliptical flaws were quite similar to those of AFEA. The condition of a rectangular flaw with the initial flaw length of the weld width was too conservative, which was in the code description of the first draft Code Case. If the initial semi-circular flaw reached the weld boundary at the surface point and changed to a rectangular flaw, the PWSCC propagation behaviors were reasonably conservative. Based on the results of the sample problems, the code description of the Code Case was revised.
Research on the Application of Risk-Managed Technical Specifications in Passive Nuclear Power Plants
Authors: Yan Zhen , Wang Jiajian , Ma Gang , Sun Jianwen , Wang Zhenqi , Liu Dalin , Wang Shuting , Zhu Sheng 
Published: November 2024

Abstract
The Technical Specification is the top-level document for the operation of a nuclear power plant and is an important basis for action to safeguard and realize the safety of a nuclear power plant. Risk-managed technical specifications use risk-informed technology to optimize traditional technical specifications. By introducing risk-informed completion times instead of completion times for limiting conditions for operation, it enhances the flexibility of nuclear power plant operation and maintenance, avoids potential risks caused by unnecessary unplanned shutdowns and mode switching, and provides an effective way for power plants to control and reduce the risks of maintenance activities. Based on the design characteristics of passive nuclear power plants, this paper takes a passive nuclear power plant as the research object to carry out the research on the application of risk-managed technical specifications in passive nuclear power plants, which includes: determining the applicable plant operation modes by considering the purpose of the application of risk-managed technical specifications and the operation modes characteristics of the units; determining the risk management thresholds in terms of balancing the risks and the flexibility of operation and maintenance; and formulating the LCO actions screening criteria applicable to risk-managed technical specifications in light of the characteristics of passive nuclear power plants; and conducting a comprehensive screening of actions related to the limiting conditions for operation in the technical specification of the plant in light of the feedback from the operation of the plant; and finally in order to support the application of risk-managed technical specifications in operating power plants, providing development recommendations for the management and implementation process of the nuclear power plant.
Molten Salt Reactor Simulation by a MOOSE-Based Application Suanni
Authors: Zhang Ao , Dai Ming , Xia Shaopeng , Chen Jingen , Cai Xiangzhou 
Published: November 2024

Abstract
Liquid fueled Molten Salt Reactors use homogeneously mixed fuel and have less heterogeneity than traditional solid fueled reactors. The two-step method successfully applied for Light Water Reactors is expected to obtain good accuracy and efficiency in MSR simulations. This paper implements a two-step procedure into a newly developed code Suanni based on the Multi-physcis Object Oriented Simulation Environment (MOOSE) framework. Suanni utilizes an open-source Monte Carlo code OpenMC for producing few-group cross sections and providing high-fidelity solutions to calculate super homogenization (SPH) factors. A diffusion solver with continuous finite element method is implemented in Suanni for the second step core simulation with SPH corrected cross sections. A Molten Salt Thermal Reactor (MSTR) core is used for the verification of the newly developed code, and the reference results come from continuous-energy OpenMC calculations. The keff and power errors are relatively large when no SPH correction is used. These errors can be significantly reduced with 4-group SPH corrected cross sections, where kes error is less than 300 pcm and relative power errors range from −2.4% to −0.35% for 3D rod-in MSTR simulation.
Numerical Investigation of High-Temperature Heat Pipes With Different Wick Structures Under Non-Uniform Heating Boundary Conditions
Authors: Guo Kailun , Sun Qishi , Sun Hao , Wang Chenlong , Chen Kang , Tian Wenxi , Su Guanghui , Qiu Suizheng 
Published: November 2024

Abstract
High-temperature heat pipe is an ideal technical solution for microreactors applied for space missions, planet surface, and the small electricity markets on the ground. However, the internal fluid flow state and the effects of key parameters such as permeability, porosity, and capillary force of wick structure on the distribution of working fluid inside the heat pipe are difficult to measure experimentally. In this study, a general CFD model for heat pipe with wick structure that can predict the thermal behavior at different steady-state conditions was carried out. The thermal conduction through heat pipe shell, liquid film on the inner surface, and the formation of liquid reservoir at the bottom of heat pipe were considered in this model. A high-temperature heat pipe model considering the flow of liquid working fluid inside the wick structure was developed based on the Star-CCM+ numerical simulation software. The model used a combination of the liquid film model and the Volume of Fluid (VOF) model to simulate the effects of fluid flow and collection behavior inside the wick structure on the heat transfer characteristics of the heat pipe. The method included models for flow resistance and capillary driving force under steady-state conditions, centrifugal force and time-varying gravitational field under motion conditions, and mixed convective heat transfer caused by motion conditions. The accuracy of the numerical simulation was validated using typical potassium high-temperature heat pipe experiments. The effects of different permeability and capillary force characteristics of the wick structure on heat transfer and internal fluid distribution in the heat pipe were studied. Based on the current method, high-temperature heat pipes with different inclinations (0°, 90 °and 45°) were investigated. Additionally, non-uniform heating boundary conditions were added to the high-temperature heat pipe evaporator section to investigate its heat transfer performance under real reactor axial power input. This investigation lays the foundation for the design and analysis of the heat transfer characteristics of high-temperature heat pipes.
Fast Solution of Schrödinger Equation Based on Complex One-Dimensional Potential Wells
Authors: Liu Zihao , Wang Xiang 
Published: November 2024

Abstract
The Schrödinger equation (SE) is a fundamental equation in quantum mechanics that describes the behavior of quantum systems. Solving the SE is crucial for understanding quantum phenomena, but it is a challenging task, particularly for complex potential wells, which require significant computational resources to obtain high accuracy by using traditional numerical methods. Recently, deep neural networks, such as Fourier neural operator (FNO), have made significant breakthroughs in solving partial differential equations. This paper explores solving the one-dimensional SE with diversified but classified potential wells using FNO. Gaussian random fields are used to generate the potential function, and then FNO is used to learn the mapping relationship between different potential functions and the solutions to the SE. The results show that the trained FNO can accurately output the solution to the SE under the corresponding potential function in a very short time, and the output results are completely consistent with expectations. Moreover, using FNO to solve the SE is significantly faster than traditional solvers, especially when the grid size is large. Therefore, in the future, FNO will provide a new solving method in complex, high-dimensional cases, such as two-dimensional, three-dimensional, and time-dependent SEs.
Research on Diffusion and Mitigation of Severe Accident Source Terms in Floating Nuclear Power Plant
Authors: Xu Tao , Wang Junlong , Zhang Bin , Li Jishen , Liu Jiajia , Xia Mingming , Ma Haifu , Wu Yirui 
Published: November 2024

Abstract
ACP100S is a modular floating nuclear power plant independently developed by China National Nuclear Corporation (CNNC) with broad application prospects. At present, research on the source terms of severe accidents in floating nuclear power plants is not sufficient, and research is needed on the migration and distribution of source terms under severe accident conditions, as well as mitigation measures for source terms. Focusing on ACP100S, a severe accident source term analysis model for floating nuclear power plant reactor is established using the ISAA code in this paper. The accident process, release and migration of the source terms after the accident are studied. At the same time, the mitigation effect of containment spray on severe accident source terms is studied, as well. The results show that after the accident, the total amount of noble gases is about 99.7% in the containment and 0.254% in the primary circuit. The fraction of Cs is 80.1% in the containment, 19.1% in the primary circuit and 0.637% in the reactor cavity. And the share of CsI is 85.8% in the containment, 2.73% in primary circuit and 9.13% in reactor cavity. In addition, spray can reduce the concentration of Cs and CsI in the containment by about 52.48% and 73.44%, which has a significant effect on the mitigation of source terms.