bluuebunny/crossref_metadata_2025_split huggingface dataset:
Study on Multi-Objective Optimization Design of LP Section for 1000MW Class Nuclear Turbines
Authors: Hu Yifeng , Hu Zhiqiang 
Published: November 2022

Abstract
With the development of nuclear reactors, the domestic market demand for high-capacity nuclear power turbines is also growing. Low pressure (LP) section (include LP inner casing and LP outer casing) is one of the important components of nuclear steam turbine, which are mainly welded by plates and rods. The increasing size of LP section leads more vacuum load, more self-weight, more stress, more deformation and so on. Therefore, the traditional design methods can no longer adapt to this situation very well.
In this paper, the stress, deformation, buckling behaviors and seismic loadcase of the LP section of steam turbine are evaluated by the numerical finite element analysis. Then, the multi-objective optimization (include size optimization and topology optimization) design of the LP section is done, so that the structural efficiency and performance of LP section is achieved. The performance of all aspects of the optimized LP section has been greatly improved. On the basis of a small improvement in stress and deformation, the self-weight of the LP section is greatly reduced. At the same time, the optimization boundaries, the choice of the model and the process are discussed. This study proposes a more optimized design method of LP section of steam turbines, and it offers considerable economic benefits.
Application of Risk-Informed Approach in Emergency Compressor Emergency Backup Function Test
Authors: Liu Deyi , Zha Sixing , Wu Jian , Cao Yong , Wang Zilong 
Published: November 2022

Abstract
The emergency compressed air production system is a part of the compressed air production system of the nuclear power plant. Its function is to provide the necessary compressed air to the nuclear island when the main air compressor fails and cannot provide compressed air. Each unit of the nuclear power plant has two emergency air compressors to provide backup air source for the compressed air distribution system for instrumentation. Under normal working conditions, one emergency air compressor is in the basic load state and the other is in the standby state. The two emergency air compressors will automatically start according to the degree of pressure drop of the compressed air production system. The purpose of the emergency backup function of the emergency air compressor is to verify that the two air compressors will start up as the system pressure drops in different states, and to verify the operating parameters of the two emergency air compressors. The test period is one week. Periodic testing is an important activity for nuclear power plants to ensure the availability of equipment that performs important functions. However, periodic testing exceeding a certain frequency may reduce the safety of the power plant and increase the unnecessary burden of the power plant. At present, PSA has been widely used in nuclear power plant operation guidance, maintenance rules and many other fields, PSA has developed into an important tool for safety assessment and decision-making. With the development of PSA technology, the risk-oriented model combining probability theory and determinism has also been applied in the optimization of the periodic test cycle of nuclear power plants, so that the resource allocation of power plants can be optimized under the premise of ensuring sufficient safety of nuclear power plants. Improve regulatory and operational efficiency. This paper will take the emergency compressed air production system of Fangjiashan Nuclear Power Plant (FSJ NPP) as an example to introduce the application method of the risk-oriented periodic test optimization method.
Comparative Machine Learning Study for Estimating Peak Cladding Temperature in AP1000 Under LOFW
Authors: Najar Merouane , Wang He 
Published: November 2022

Abstract
For a more realistic estimation of safety margins, the conservative approach is replaced by integrating the best estimate approach (BE) with uncertainty quantification, the integration which knows as best estimate plus uncertainty (BEPU), which can predict the key safety parameters such as peak cladding temperature (PCT) and departure from nucleate boiling ratio (DNBR), etc. In this sense, a fast and cost-effective tool for uncertainty quantification is developed through a data-driven approach to predict PCT under loss of feedwater accident (LOFW) in AP1000 reactor.
This paper includes performing a comparative study between different regression ML algorithms to find the best algorithm which can predict the PCT with higher accuracy.
Intent to generate the required data for training and testing the ML algorithm, an uncertainty quantification framework is developed by coupling a best estimate code (RELAP5) with a statistical tool (RAVEN). RELAP5 is used to simulate the thermal-hydraulic response under LOFW accident while a set of uncertainty parameters are propagated through the RELAP5 model using RAVEN. These distributions were sampled using a Latin Hypercube Sampling (LHS) technique to generate sets of sample cases to simulate using the RELAP5 code. 5,000 runs were generated in order to acquire a large database for training purposes. The examined algorithms are linear regression, supported vector machine, k-nearest neighbors (KNN), and random forest. The evaluation of algorithms depends mainly on mean absolute error (MAE) and determination coefficient R2.
The result shows that the random forest provides high accuracy in predicting PCT within four algorithms, which reaches 98.96%.
Heating Start-Up Method and Demonstration for Full Natural Circulation Reactor
Authors: Yu Jianhui , Tan Wenji , Zhang Xingliang 
Published: November 2022

Abstract
Since there is no main coolant pump (MCP), the full natural circulation reactor can not use the MCP to heat the reactor coolant system to normal operating temperature prior to reactor criticality. This paper presents an external starting heating method. The principle and brief system description of the external start-up heating system is presented. By reasonably arranging the inlet and outlet nozzles of start-up heating in the reactor, the external start-up heating source is used to heat the reactor coolant system, by establishing natural circulation in the reactor.
In order to demonstrate the feasibility of external start-up heating system, the heating effect of start-up is analyzed, and the analysis results are verified by experiments. On the other hand, the impact of external start-up heating system on reactor operation and safety is evaluated. In order to ensure that the external start-up heating system does not affect the operation and safety of the reactor, the suggestions on the configuration and operation of external start-up heating system are given.
By the external start-up heating system, the problem of startup heating without main pump of full natural circulation reactor is successfully solved. Therefore, it provides an option of hot start for the full natural circulation reactor.
Assessment on Accident and Mitigation of Spent Fuel Pool in a Typical PWR
Authors: Guo Chao , Ge Zhenzhen , Liao Feiye , Chen Peng , He Dongyu 
Published: November 2022

Abstract
In consideration of low power rate and large water inventory in spent fuel pool (SFP), the accident progression may be very slow and the fuel melt accident was considered to be very unlikely in SFP. However, the accident in Fukushima Dai-ichi Nuclear Power Plant implies the possibility of severe accident in SFP and highlights the importance of accident management of SFP. As the fuel building has no three-barriers containing radioactive products like the reactor building and SFP has significantly larger fuel inventory, the potential consequences of severe accident (radioactive materials release and hydrogen production) could be far beyond the capability of practical mitigation measures in design. Since Fukushima accident, lots of efforts related to SFP simulation are made by developers of several severe accident simulation codes. For ASTEC code, models about radiative heat transfer and Zircaloy oxidation are added in V2.1 which makes it possible to have a better prediction of severe accident in SFP. In this paper, representative scenarios (loss of coolant, loss of cooling) in SFP of a typical pressurized water reactor (PWR) design are simulated with ASTEC code. Accident progression and phenomena are presented. The influence of potential mitigation measures is studied. Suggestions to the design of SFP and accident mitigation strategies are provided according to the analysis results.
Research on Fretting Wear of Grid and Cladding Tube Under Simulated Primary Circuit Environment of Pressurized Water Reactor
Authors: Liu Tingting , Li Kun , Ma Wenhui , Zhang Yuxiang , Chen Minli , Wu Xiaohang 
Published: November 2022

Abstract
With the deepening of the fretting wear process, the wear behavior and wear mechanism of the contact surface between the grid and the fuel rod cladding will gradually change, but there are relatively few research reports in this area at home and abroad. In this paper, the chemical environment of high temperature and high pressure water in the primary circuit of the PWR is simulated. In this environment, the friction pairs formed by the grid and the cladding tube are subjected to fretting wear tests of different durations (192h, 288h, 384h). The cladding samples is subjected to wear scar detection, and the wear topography data (including the wear volume and maximum wear depth) of the contact point between the cladding tube and the grid clamping system are obtained, and the wear coefficient is finally calculated using the ARCHARD equation. The results show that with the increase of the test time, the spring wear scar data (including the wear volume and the maximum wear depth) formed by the grid spring and the cladding tube tend to increase, the wear rate is basically stable at 0.8μm3*s−1, and the wear coefficient is basically stable at 0.4 × 10−17 Pa−1, there is no obvious change. For the dimple wear scar formed by the dimple of grid and the cladding tube, the wear scar data (including the wear volume, the maximum wear depth and the wear coefficient) are significantly greater than those obtained under the test conditions of 192h and 288h at 384h. However, the data obtained at192h is slightly larger than that at 288h, combining with the Archard equation, it can be seen that this may be caused by the decrease in the wear rate and wear coefficient in the middle of the fretting wear. The results of this paper show that with the deepening of the fretting wear process, the wear characteristics of the grid and cladding samples are constantly changing, it has a good reference significance for the existing fretting wear research between the grid and the cladding tube and the related nuclear fuel structure.
Study on High Frequency Flow Induced Vibration Mechanism and Evaluation Method of PWR Fuel Assembly Grid
Authors: Ding Jie , Zhang Chaozhu , Sun Jingya , Gan Fujun , Gong Shuhe 
Published: November 2022

Abstract
By the analysis of the three-dimensional flow field of the fuel assembly grid, the high-frequency flow-induced vibration test of small-scale test assembly, the preliminary research on the overall and local flow-induced vibration mechanism and influencing factors of the fuel assembly grid is carried out. A system of evaluation index based on the three-dimensional flow field analysis of the grid and the flow-induced vibration test of 5 × 5 small test assembly is preliminarily established, which can more accurately evaluate the flow-induced vibration characteristics of the PWR fuel assembly grid.
Working Fluids Analysis and Discussions of the Stirling Cycle for Space Nuclear Reactors
Authors: Yang Chenhao , Zhuang Nailiang , Zhao Hangbin , Tang Xiaobin 
Published: November 2022

Abstract
Space nuclear power is the most potential power source for future deep space exploration, interstellar navigation, and planetary surface base station. For medium and high-power space missions, an efficient, compact and reliable energy conversion system that converts nuclear reactor thermal energy into mechanical energy or electrical energy is critical for the entire power system. The Stirling cycle that converts thermal energy into mechanical energy through expansion and compression of working fluids has the advantages of strong load adaptation, high conversion efficiency, modular combination, robust reliability, redundancy, etc. Therefore, it is very suitable for future high-power deep space missions. The technical route of space nuclear power proposed by Nanjing University of Aeronautics and Astronautics is based on the modular Stirling thermoelectric conversion and liquid molten salt core scheme. The present study is the simulation prediction of the thermodynamic performance of the Stirling engine. The working fluid’s composition, physical properties, and leakage properties have great influences on the thermal efficiency, operating lifetime, and reliability of the Stirling cycle. In this regard, a comprehensive study on the properties and effects of H2, He, He-Xe mixture, N2 and air as working fluids was carried out in the present study. The modified Stirling thermodynamic model IPD-MSM was used to simulate the Stirling cycle, and the heat and power losses caused by various irreversible factors were analyzed. Moreover, impacts of operating pressure, heat source temperature, and piston frequency on every heat and power loss were discussed. The results show that the pressure loss and the non-ideal heat transfer loss are the dominant losses. However, different working fluids and operating conditions have different performances of power loss. The effect of operating frequency and working pressure are significant, while the effect of operating temperature is relatively small. The present study provides theoretical support for selecting thermoelectric conversion methods for future medium and high-power space missions.
Evaluation of Detectability of Pump/Diagrid Link Rupture in Pool-Type Sodium-Cooled Fast Reactor
Authors: Onoda Yuichi , Uchita Masato , Tokizaki Minako , Okazaki Hitoshi 
Published: November 2022

Abstract
The safety analyses were carried out to confirm the sufficiency of the function of the plant protection system against the pump/diagrid link rupture. The target plant is a pool-type SFR of about 600 MWe class equipped with an axially homogeneous core currently under development in Japan. In the pool-type SFR, the primary system piping connects primary pump and the high-pressure sodium plenum located at the inlet of fuel sub-assemblies and called “pump/diagrid link”. Because this piping is submerged in the reactor vessel, it is difficult to detect small scale sodium leakage in this piping, and thus a certain large pipe break like guillotine should be assumed and evaluated as a design basis event. In order to confirm the detectability of pump/diagrid link rupture by safety protection system signals, a series of analyses of the guillotine break for a pump/diagrid link were carried out. Sensitivity study had also been performed to consider the uncertainty of the reactivity coefficient in the analyses. The sufficiency of the function of the plant protection system against the pump/diagrid link rupture was confirmed by the analysis results that at least two signals are transmitted for the detection of the event, which is the development target of the plant protection system in pool-type SFR.
Investigation on the Active Learning Optimization on Rotor Dynamics of SCO2 Turbine
Authors: Zheng Zhaoli , Wu Jun , Ke Zhiwu , Zhao Zhenxing , Yang Xiaohu , Dai Lu 
Published: November 2022

Abstract
The supercritical carbon dioxide (SCO2) Brayton cycle has the characteristics of high power density and high thermal efficiency, which is an important development direction of the micro power plant. SCO2 turbine is the core component of the SCO2 Brayton cycle of which the dynamics have important influences on operational reliability of the entire system. With regard to the rotor of SCO2 turbine, the equations of motion is established by adopting finite element method. The complex eigenvalues of the rotor are solved in the state space, and the campbell diagram has been drawn to obtain the critical speeds. The steady state response of the rotor is obtained by the harmonic balance method, and the safety of the system is estimated based on API684. Results show that the rotor is not safe with its original geometric paramters. Aiming to improve the operational safety, an optimization method based on active learning is developed to maximize the separation margin. Results show that after the optimization, the separation margin has been greatly increased. Comparing with the genetic algorithm (GA) and the parttern search (PS) method, the iteration number of the active learning optimization method has been greatly reduced. The effectiveness of the developed optimization method is proved, and the study method and conclusions can serve as a reference to the optimization of SCO2 turbine rotors in the industry.
Characteristic Behavior of Flow Field in Blocked Fuel Assembly Flow Channel under Steady Flow Condition
Authors: Jayeola Mathew , Tan Sichao , Qiao Shouxu , Ala Ayodeji , Dione Fulgence 
Published: November 2022

Abstract
Past histories of fuel assembly blockage incidents had made investigation of flow channel blockage in rod bundle a compelling field of thermal-hydraulic safety. This study experimentally investigate flow field structural characteristics in a 5 × 5 rod bundle under two sub-channel blockages with geometries; 2 × 5 and 2 × 3, in a non-mixing spacer grid. The blockage effects on velocity distribution and turbulence characteristics is examined under laminar, transitional and turbulent flow regimes across a range of Reynolds number between 1108 ∼ 8103. Under laminar flow, result of spatiotemporal analysis for flow downstream of the blocked spacer grid-SP2, showed formation of a large symmetric vortex pair shedding in wake formed at both edges of the blockage. For transitional and turbulent flow, the vortex gradually dissipate especially towards the flow channel’s wall due to interference of friction effect. Damping effect of blockage on upward velocity flow and pressure in the vortex dominated region creates a backflow zones accompanied by flow reversal. Analysis of turbulence characteristics examined for 2 × 5 blockage across adjacent bundle gaps and sub-channels showed immense variations with Reynolds number and channel wall impact. The study aimed to provide relevant reference basis to compare and analyze similar scenario under unsteady flow condition.
Experimental Study on the Relationship Between Performance and Structural Parameters of Metal Fiber
Authors: Li Lintai , Zhou Yanmin , Li Yuanyuan , Sun Zhongning , Wang Yin , Gu Haifeng , Ma Song 
Published: November 2022

Abstract
As a new filter material, metal fiber has good high-temperature resistance, moisture resistance, and radiation resistance. To explore the influence of the fiber’s primary structural parameters on the fiber’s performance, this paper experimentally studied the relationship between the flow resistance, filtration efficiency, and dust holding capacity of the metal fiber filter material and the thickness and solid volume fraction (SVF). Under the condition of ensuring uniform deposition of NaCl aerosol, we found that the deposition law of NaCl polydisperse aerosol is the same as that of monodisperse aerosol. The results show that for the same fiber diameter, increasing the fiber’s thickness and SVF will increase the flow resistance of the fiber under the same dust holding capacity; that is, reducing the dust holding capacity of the fiber. In the depth filtration, the efficiency increases rapidly due to the aerosol deposition in the fiber, while the efficiency changes gently when the filter cake is covered on the filtration surface. At the same time, for each high filtration efficiency fiber, the variation characteristics of resistance with dust holding capacity in the deep filtration stage are the same. After the filter cake is completely covered on the filter surface, the growth trend of resistance is the same, independent of the fiber’s structural parameters.
Study on Frequency and Time Domain Analysis of Flow-Induced Vibration Response of Core Barrel
Authors: Cao Guitao , Zhu He , Liu Guangdong 
Published: November 2022

Abstract
Different calculation methods of flow-induced vibration of the reactor vessel internals (RVI) are studied in this paper, that is frequency domain analysis and time domain analysis. Firstly, considering the fluid-structure interaction between the core barrel (CB) and the reactor pressure vessel (RPV), the CB and the heavy reflector, the finite element model of the CB in water is established and the results of its vibration characteristics are obtained, which is consistent with the test results. Based on the flow-induced vibration test of the RVI scale model, the fluctuating pressure data of the CB is obtained. The parameters such as power spectral density (PSD), correlation length and coherence function are obtained by processing the test data. Thus, the cross-power spectral density between any two points is got through these parameters. The root mean square (RMS) response of the CB is obtained by random vibration analysis (frequency domain). Then, the CFD model is established and large eddy simulation (LES) is used to obtain the time history of the fluctuating pressure of the CB. The response of the CB is obtained by time integration method (time domain). The calculation results by using these two methods are in good agreement with the experimental results.
ISO 19443: Fundamental Requirements of the New Quality Management Standard for the International Nuclear Supply Chain
Authors: Bargsten Jeanne , Treier Clemens , Yilmaz Özlem , Pades Amandine , Kim Andrea , Lee Jung-Kyu 
Published: November 2022

Abstract
The new quality management standard ISO 19443 [2] defines specific requirements for supply chain organizations in the nuclear industry. The standard became increasingly popular in the last years and a certification is now often a prerequisite for equipment manufacturers and service providers to participate in the bidding process of new build projects.
These companies face the challenge, to adapt their existing quality management system to the new requirements of ISO 19443 [2], since this is necessary to receive a globally accepted accredited certification.
A conducted analysis of the ISO 19443 standard [2] identified major differences to the established quality management system standard ISO 9001 [3], the results are presented. Furthermore, the key requirement areas of the standard with particular significance to the nuclear industry, are evaluated.
ISO 19443 [2] enshrines nuclear safety culture as the fundamental element throughout the entire supply chain and enables a graded approach to the application of quality requirements. The standard aims at the global harmonization of management standards and facilitates the regulation of the international nuclear supply chain.
Best practice approaches are highlighted, on how to successfully adapt a quality management system. It is based on experience of one of the first accredited inspection bodies, already supporting several manufacturing companies on their way to certification.