bluuebunny/crossref_metadata_2025_split huggingface dataset:
Seismic Analysis of the Scale Reactor Structure Model of a Lead-Bismuth-Cooled ADS Reactor
Authors: Rong Gang , Zhu He , Yang Yi 
Published: November 2022

Abstract
This paper is aimed to analyze the structural responses of a scale reactor structure model of a lead-bismuth-cooled ADS reactor under the design seismic time history loads. As a heavy-liquid-metal-cooled reactor, the fluid-structure interaction due to seismic loading is a key issue that may significantly affect the responses of the structures. Two three-dimensional finite element analysis models based mainly on shell element have been built, considering the gap and contact effect between components by nonlinear contact elements with gap specification. The fluid-structure interaction of the structure with an annular cavity in the scale reactor structure was included in the analysis model through a hydrodynamic mass matrix element considering no leakage effect on the ends of the annulus for the first model, and assuming some mass loss due to leakage at the ends for the second model, while additional mass elements due to fluid effect were used for all other structures in both models. Simulation analyses for the scale reactor structure were carried out to obtain the dynamic acceleration response data of the main components, and the comparison with measured data of vibration characteristic tests was conducted, the result showed that considering some mass loss at the ends of annulus would obtain a more precise result. The results of this paper will provide a theoretical recommendation for the design scheme of the subsequent seismic experiment of the scale reactor structure model.
Development and Validation of a New Gas Conductance Model
Authors: Yang Shengzhi , Wei Xiaoyan , Zhu Yanan 
Published: November 2022

Abstract
The analysis of fuel rod behavior has become increasingly important and reliable assessment of temperature distribution is vital. Gap conductance is a key parameter to calculate the heat transfer between fuel pellet and cladding, then fuel temperature and thermo-mechanical of fuel rods will be influenced. Gas conductance often plays the most important role comparing convective heat transfer, thermal radiation and solid contact conductance in fuel rod analysis. A new gas conductance model has been proposed, considering the effects of mixture gaseous component, the temperature discontinuity on the surfaces of pellet and cladding, and the roughness of solid surface, since the irregularity will become significant when the gap is closed at the extended burn-up. Moreover, the model has been calibrated with Modified Pulse Design (MPD) and Modified Longitudinal Design (MLD) experiment gas conductance data performed by John. E. Garnier and Stefan. Begej, and shown a good prediction accuracy at various experiment conditions. Furthermore, the model has been successfully implemented in fuel rod performance analysis code JASMINE to improving the temperature distribution prediction. Based on the comparison of code prediction of centerline temperature with HALDEN in-core irradiation tests data, the results show that the new gas conductance model has a decent suitability to JASMINE.
Numerical Investigation of Melt Fragmentation in Sodium-Cooled Fast Reactor Based on Moving Particle Method
Authors: Wang Wenpeng , Zhang Bin , Cao Sheng , Hu Minghao , Shan Jianqiang 
Published: November 2022

Abstract
The Melt-Fuel-Coolant Interaction (MFCI) is one of the important phenomena during reactor severe accidents. In particular, the fragmentation process of the molten fuel in the metal coolant is a complicated multi-phase and multi-component heat transfer phenomenon accompanied by solid-liquid coupling and phase change, which is of great significance to the formation of debris bed on core catcher as well as the coolablity of debris bed after severe accident. It is of great significance to conduct research on MFCI of sodium-cooled fast reactor. A moving-particle-method based program is developed for the melt fragmentation analysis of sodium-cooled fast reactor in the present study. The program is based on the finite volume particle method and improved free surface pressure model, smoothing model, heat transfer and phase change model, interface surface tension model and solid-liquid coupling model are incorporated. Afterwards, the process of fragmentation is compared against the experimental data. The effects of initial temperature of melt, velocity of melt entering coolant, and jet diameter on the fragmentation are analyzed. The results show that the initial temperature of the melt has a great influence on the melt fragmentation, and a higher initial temperature makes the melt fragmentation more complete. Secondly, in the simulation of melt jet fragmentation, the process can be divided into six phases by analyzing the variation of the amount, the maximum mass, and the maximum size of fragments with time. Thirdly, by changing the initial velocity and diameter of the melt jet, it is concluded that the higher initial velocity and the larger melt jet diameter will produce a higher peak amount of fragments.
Probability Safety Analysis for Nuclear Power Plant Long-Term Temporary Shutdown State
Authors: Chen Yanzhu , Yang Zhi 
Published: November 2022

Abstract
In recent years, due to the influence of power grid demand and other factors, some nuclear power plants have long-term temporary SHUTDOWN. At present, there is no mature quantitative analysis method to measure which Plant operational state is safer under long-term temporary shutdown. Taking a second-generation nuclear power plant as an example, this paper first briefly introduces the PSA analysis method; Secondly, the difference is analyzed qualitatively, which between standard long-term temporary shutdown condition and refueling shutdown condition. Thirdly, from the quantitative point of view, the model processing methods and analysis assumptions are introduced, the risk views are given, and the quantitative calculation and sensitivity analysis are carried out; Finally, the analysis conclusion is obtained.
Stay in the standard long-term temporary shutdown state for 15/30/45/60 days, after supplementing the calculation of the thermal shutdown condition,the risk per unit time of long-term temporary staying Maintenance Cold Shutdown mode is lower. The risk sequence is: Cooldown with the residual Heat Removal mode &gt; Hot shutdown condition &gt; Maintenance Cold Shutdown mode with closed primary loop. From the perspective of the risk of core damage, it is suggested that the Maintenance cold shut mode in which the primary loop is kept closed is recommended for long-term temporary shutdown of the power plant.
Layout Optimization of Self-Powered Neutron Detector for HPR1000
Authors: Zhou Yao , Cao Liangzhi , Zhao Changyou , Li Zhifeng , Zhu Yuxiang 
Published: November 2022

Abstract
Self-Powered Neutron Detector (SPND) has been widely applied for the on-line monitoring of the key core parameters, including the power distribution, power-peak factor and so on. For the SPND layout design, the vital parameters include the SPND amount and loading position in the radial direction and segment amount in the axial direction. Different SPND layout reflects different ability of power distribution reconstruction, which is directly related to the safety and economy of nuclear power plant. For HPR1000, the neutronics sensitivity of SPND layout design is analyzed based on a 3D on-line power distribution monitoring system NECP-ONION. Through the comparison and analysis of the reconstruction errors of power distribution, the design rationality of SPND is evaluated and the optimization design is recommended. For the recommended SPND layout design, the loading positions in the radial direction are arranged at intervals and segment amount in the axial direction is 15. The recommended SPND layout design shows stable performance in the whole life for HPR1000 reactor, and the RMS errors of the power distribution reconstruction has been in the range of [0.32%, 0.48%], meeting the safety limit and accuracy requirements of power distribution reconstruction. This work verifies the on-line monitoring ability of NECP-ONION system. Additionally, it provides a technical reference for the SPND experiment in HPR1000.
Research on Martensitic Stainless Steel Used on the Latch Housing of European Pressurized Reactor (EPR) Control Rod Drive Mechanism (CRDM)
Authors: He Cuizhu , Lan Yinhui 
Published: November 2022

Abstract
As one of the most important equipment for reactivity control, Control Rod Drive Mechanism (CRDM), which is widely used in Pressurized Water Reactor (PWR) nuclear power plant, has a series of important security functions. As part of the reactor pressure boundary, materials for European Pressurized Reactor (EPR) CRDM pressure housing needs to have relatively high strength and sufficient toughness (i.e., appropriate yield ratio). As an important part of the magnetic circuit, martensitic stainless steel used for latch housing (a part of pressure housing) needs to have good magnetic properties. It is difficult for the yield ratio and the magnetic properties of the materials to meet the specification requirements simultaneously. Based on the design specification, this research refers to the relevant international manufacturing experience and combines the current manufacture with engineering practice. The manufacturing process and key parameters of the material for latching housing of CRDM are established. The martensitic stainless steel which meets the design requirements has been successfully prepared. The comprehensive properties (e.g. yield ratio and magnetic properties) of the material are improved as well. The performance indicators have reached international advanced level. The successful development of the martensitic steel for EPR CRDM latch housing provides strong technical support for the construction of nuclear power projects at home and abroad.
Research and Practice on Operational Risk Management of Nuclear Power Plants
Authors: Wei Haifeng 
Published: November 2022

Abstract
Risk management is a very important work for nuclear power plants, which is related to the sustainable, healthy and stable operation of nuclear power plants. Almost all the work of nuclear power plants requires risk identification and risk control. Based on the risk management theory, advanced concepts and good practices, this paper defines the operational risk management of nuclear power plants, puts forward five modes of operational risk management of nuclear power plants and corresponding control measures for the first time. Combined with the practical feedback of nuclear power plant operation, it clearly puts forward the principles and suggestions for formulating operational risk control and response measures of nuclear power plants. Referring to the analysis mode and corresponding control measures of nuclear power plant operational risk management proposed in this paper, we can accurately and comprehensively analyze the operational risk of nuclear power plants, formulate the risk control measures of validity, the effectiveness of operational risk management of nuclear power plants in guaranty, and prevent the occurrence of accidents/events. At the same time, the research results of operational risk management of nuclear power plants have important reference and guiding significance for the promotion and optimization of risk management of nuclear power plants.
Study on the Effect of Random Distribution of the Fuel in Dissolver
Authors: Yu Miao , Yi Xuan , Shao Zeng , Yang Haifeng 
Published: November 2022

Abstract
The main role of the dissolver is to dissolve sheared short sections of the spent fuel assembly. During the dissolution process, there is the situation where solid fuel coexists with and randomly distributed in the solution. Generally, the optimal moderation grid array is used as the bounding model in the criticality safety analysis, which has a considerable conservative margin. Based on a hypothetic dissolver equipment, this paper studies the effect of the random distribution of the solid fuel in the dissolver. Sensitivity analysis is carried out on the random distribution states, fuel composition, solid fuel volume proportion and radii of fuel particles. According to the results of sensitivity analysis, the suggestion for criticality safety analysis model of dissolver is given. The research on the effect of the random distribution of the solid fuel in the dissolver provides a reference for the optimization of the criticality safety analysis model of the dissolver, which can be referred in the subsequent criticality safety analysis of the dissolver to increase the processing capacity of the dissolver and improve the dissolver economy on the premise of ensuring safety.
Multiphysics Thermo-Mechanical Behavior Modeling for Annular Uranium-Plutonium Mixed Oxide Fuels in a Lead-Cooled Fast Reactor
Authors: Cai Yutai , Zhou Wenzhong , Gu Long 
Published: November 2022

Abstract
A model for fuel performance of annular uranium-plutonium mixed oxide (MOX) fuels in lead-cooled fast reactors is developed. This model has several modules working in fully coupled approach based on COMSOL Multiphysics. These modules describe (1) Fission gas release, (2) Plutonium redistribution, (3) Oxygen redistribution, (4)Heat transfer, (5) Gap width evolution. The modeling results were benchmarked with existing codes and experiment data, the results were in satisfactory agreement. The parameters analysis was carried out in this work. Consistent with the solid MOX fuel, the O/M ratio (or deviation from stoichiometry of oxygen) has a great effect on temperature evolution, fission gas migration and release behavior and plutonium redistribution. The effect of linear power on fuel performance are also analyzed.
Assessment of Passive Cooling Capability of the Westinghouse Lead Fast Reactor Under Station Blackout Conditions
Authors: Tatli Emre , Lee Sung Jin , Liao Jun , Ferroni Paolo 
Published: November 2022

Abstract
Westinghouse is developing its next generation of high-capacity nuclear power plants with the Westinghouse Lead-cooled Fast Reactor (LFR) (Ref. 1). The Westinghouse LFR is a pool-type 950 MWt (∼450 MWe Net) passively safe modular construction plant that contains all the primary system components within the reactor vessel, including the reactor coolant pumps (RCPs) and the primary heat exchangers (PHEs). Similar to the Westinghouse AP1000® Pressurized Water Reactor design, the Westinghouse LFR design incorporates passive cooling capability primarily through its Passive Heat Removal System (PHRS). During accidents, the PHRS allows decay heat to be transferred from the core to a pool of water surrounding the guard vessel, which transitions to long-term air cooling upon water depletion. All of this takes place without operator action or need for moving parts or power supply. Overall, when combined with natural circulation of liquid lead inside of the reactor vessel, this passive system fully relies on natural circulation, thermal radiation and boiling mechanisms to transport decay heat from the core to the atmosphere. In this study, complementary Computational Fluid Dynamics (CFD) and system code analyses were performed to assess the passive cooling ability of the Westinghouse LFR design during station blackout (SBO) conditions. The postulated scenario examined in this study is the longer-term phase of the SBO event, with the RCPs and PHEs assumed to be inactive and the water inventory in the pool exterior to the guard vessel fully depleted, thus relying purely on air cooling as the ultimate heat sink. The CFD analysis was conducted with Siemens STAR-CCM+ code using, as domain, the primary lead pool and with boundary conditions provided by the containment code GOTHIC. The CFD analyses provided high fidelity resolution of the three-dimensional natural circulation flow patterns formed throughout the reactor vessel. The results of the study not only provided confirmation of the passive cooling capability of the Westinghouse LFR design during an SBO event, but also provided granular input for vessel structural analysis activities accounting for thermal gradients developed during an SBO event.
Study on Hydrogen Risk Venting Mitigation Measure of Small Steel Containment During Severe Accident
Authors: Zhiqiang Zou , Hang Zhang , Ming Zhang , Huanhuan Peng , Youyou Xu , Jian Deng 
Published: November 2022

Abstract
The hydrogen risk in small steel containment is one of the most important safety problems of small modular reactor (SMR), effective measures need to be taken to mitigate the risk of high concentration hydrogen in a small free volume during severe accidents. In the early stage of a severe accident caused by a loss of coolant accident, the pressure in the containment may rise rapidly, probably triggering the immediate venting of the containment, which can protect the integrity of the containment in the stage of pressure rise in the short term. In the later stage of a severe accident, the generation of hydrogen from reactor core may cause hydrogen combustion or even explosion, which may cause the possibility of overpressure in the containment. In this paper, Computational Fluid Dynamics (CFD) GASFLOW-MPI code is used to analyze the effect of containment venting measures on hydrogen risk in small steel containment of SMR during severe accident initiated by direct vessel injection line break loss-of-coolant. The analysis results show that the venting action can effectively depressurize the containment, the larger the area of the venting pipe is, the more obvious the depressurization effect is. However, hydrogen cannot be effectively discharged in the middle and later stage of the accident, so the risk of hydrogen cannot be completely eliminated, and the possibility of flame acceleration in containment cannot be ruled out for the reason that the negative pressure generated by water vapor condensation will suck in the air outside the containment and form a combustible gas cloud again.
Prediction of Critical Heat Flux Based on Random Forest Algorithm
Authors: Sun Zhuo , Han Mingyang , Yuan Peng , Gong Lei , Chen Di , Yuan Guanghui , Zhang Shuang 
Published: November 2022

Abstract
The accurate prediction of critical heat flux (CHF) is of great significance to the safe operation of the reactor. At present, most of the traditional methods for predicting CHF are based on relational fitting in mathematical form, which are not accurate enough and may can not taking full advantage of experimental data. Therefore, a prediction model based on random forest algorithm (RF) is proposed, the model is based on CHF 2006 look-up table, and the accuracy and smoothness of the model have been verified when predicting CHF. In this paper, the variation trend of CHF predicted value and actual value with pressure, mass flow rate and thermodynamic quality is given. It has been verified that the error of this method is acceptable, the predicted value is in good agreement with the experimental value, and conforms to the corresponding physical laws, which can provide a reference for the development of high-precision CHF prediction models or CHF-related experiments.
Optimization in Heat Transfer Scheme for Load Rejection and Emergency Shutdown Conditions of Sodium Fast Reactor
Authors: Yu Yang , Yang Jun , Wang Jin , Wang Xiaokun , Hu Wenjun , Zhang Donghui 
Published: November 2022

Abstract
For a demonstration pool-type sodium fast reactor (CFR600), load rejection and emergency shutdown conditions put forward sharp requirements for the reactor control system by introducing abrupt changes in power and flowrates. While the sodium pool has a slow temperature response resulting from the heat capacity, the change in the circuit flowrates at the initial stage of the transients could lead to fluctuations in the intermediate heat exchanger, steam generator, and other essential devices. This research firstly clarified the constraints and objectives of the working condition design, and adopted an envelope calculation method to optimize the scheme under load rejection and emergency shutdown conditions. The research proposed several optimizations of secondary pump flowrate and the BOP water supply scheme, which enhanced the stability of devices in transient conditions. The validity of the scheme is finally re-checked from the perspective of energy transportation. It is concluded that by targeting at increasing heat transfer and reducing equipment fluctuation in the initial stage of transient, the optimized scheme realizes the balance between safety and economy to a certain extent. This research may provide the guidance for experiment configuration and the further study.
Rod Ejection Accident Simulation of a Small Pressurized Water Reactor
Authors: Fan Ziqi , Sun Peiwei , Wei Xinyu 
Published: November 2022

Abstract
Rod Ejection Accident (REA) is one of the major accidents which may result in the serious consequences during nuclear reactor operation. The study on REA is helpful to understand the physical characteristics and to verify the safety design of the core. The small pressurized water reactor (SPWR) adopts a relatively small amount of control rods. Therefore, when REA occurs, the positive reactivity introduced by the accident may be larger, resulting in more serious consequences. The integral design of SPWR results in different responses to the REA. Thus, it is necessary to investigate the REA of SPWR.
In this paper, 3KeyMaster simulation platform, which is coupled with a thermo-hydraulic model and a three-dimensional space-time neutron dynamics model, is used to establish the reactor core model. Four single-beam control rods are selected to carry out REA, respectively. Simulations are performed, and results are collected. The responses of the SPWR are analyzed and accident procedures are obtained.
Through analysis, it can be concluded that the serious consequences are caused when the R rod closest to the core, resulting in the maximum nuclear power peak. But the fuel center temperature and cladding temperature do not exceed the limit value, which proves the safety of SPWR and the rationality of core design.
Analysis of Hidden Fault in Digital Instrumentation and Control System of Nuclear Power Plant Based on Environmental Model
Authors: Guo Chao , Zhou Shuqiao , Li Duo , Huang Xiaojin 
Published: November 2022

Abstract
Nuclear safety is one of the key issues for a nuclear power plant (NPP). The instrumentation and control system (I&C system) plays critical roles for the safe and efficient operation of an NPP. Due to the large scale and complex logic of the I&C system, some hidden faults cannot be found by factory testing or on-site debugging and may cause system degradation or even system failure during NPP operation. These hidden faults may come from system design omissions, software programming errors, personnel operation failures, natural environment, inappropriate preventive maintenance, etc. Timely identification of these hidden faults is of great significance for improving the reliability of the I&C system. This paper takes the hidden faults of the I&C system as the research object, and carries out the analysis and investigation of the hidden faults based on the environmental model. After determining the key and important characteristics of the I&C system to be analyzed, this paper analyzes and classifies the triggering causes of the faults from three dimensions of specific application environment, specific application method and long-term dynamic operation. This paper takes the hidden fault “Failure of Power Module of the DCS Cabinet”, “Operation failure of KCS troubleshooting personnel”, and “Communication abnormal of DCS internal network” actually found in the I&C system of an NPP as cases to establish the mapping relationship between the hidden fault and the environmental model to verify the correctness of the built model. The work showed in this paper can contribute to improve the identification capability of hidden faults in I&C system of the NPPs, which has important engineering significance for improving the reliability of the I&C system.