bluuebunny/crossref_metadata_2025_split huggingface dataset:
Analysis of Flow Resistance Influence on Step-Down Process of the Control Rod Hydraulic Drive System
Authors: Yang Linqing , Qin Benke , Bo Hanliang 
Published: October 2021

Abstract
Control rod hydraulic drive system (CRHDS) is a new type of built-in control rod drive technology which is invented by INET, Tsinghua University. The integrated valve (IV) is the main flow control component of the CRHDS. Flow resistance of IV has a great influence on the control rod dynamic step-down process. The step-down performance experiments of CRHDS with different flow resistance of IV were conducted under room temperature conditions. Meanwhile, the theoretical model of hydraulic cylinder step-down process was established and combined with the relationship of the flow resistance of IV under the experimental conditions to get the dynamic response of the hydraulic cylinder. The calculation results of theoretical model agree well with the experimental data. On this basis, the theoretical model of hydraulic cylinder step-down process was applied to the high temperature working conditions with different flow resistance of IV. The analysis results show that at higher working temperature, with the increase of the flow resistance of IV control rod step-down average velocity decreases and step-down time increases correspondingly. There is an inflection point in the transient pressure curve and the pressure of the inflection point decreases gradually with the increase of the flow resistance. The pressure lag time after step-down also decreases. The research results lay the base for the design and optimization of the flow resistance of the IV for the CRHDS.
Dynamic Modeling of Nuclear Hydrogen Production Using Methane Steam Reforming
Authors: Li Junyi , Dong Zhe , Li Bowen 
Published: October 2021

Abstract
Methane steam reforming (MSR) technology is one of the promising methods of hydrogen production and already available at an industrial scale, in which steam is added to methane to generate hydrogen. MSR carries out at a temperature of 500°C when catalysts and Pd-based membrane reactors are used. The nuclear steam supply system (NSSS) of a modular high-temperature gas-cooled reactor (MHTGR) can provide high-quality steam of around 570°C, which is an excellent heat source for MSR. MHTGR is a typical small modular reactor (SMR), of which the coolant is helium, and the moderator and structural material are graphite. The number of the MHTGR can be decided based on the thermal power required for MSR and electricity generation. In this paper, a six-modular MHTGR nuclear power plant with 1500MW thermal power coupled with the MSR process is designed. The hydrogen production rate is 9.72 tons per hour. The dynamic modeling is based on conservation laws of mass and energy. To examine the dynamic characteristics of the nuclear hydrogen production plant, open-loop responses of the model under different disturbances are presented.
System Design for Ammonia Nuclear Thermal Propulsion
Authors: Mao Chenrui , Ji Yu , Sun Jun , Liang Zhaoyu , Shi Lei 
Published: October 2021

Abstract
Nuclear Thermal Propulsion (NTP) for deep space exploration is promising. Hydrogen is employed as the propellant to enhance the maximized specific impulse of NTP. Whereas, the low boiling point and density of hydrogen lead to several issues related to safety and storage. From the perspective of reactor physics, thermal hydraulics, material compatibility, and experience in the aerospace field, ammonia (NH3) is the alternative propellant to replace hydrogen.
To analyze the specific impulse and system parameters of ammonia NTP, the hydrogen expander cycle NTP system analysis program PANES (Program for Analyzing Nuclear Engine Systems) is improved. The thermophysical properties library of ammonia has been increased in PANES. Based on the same thrust with hydrogen NTP, the design and analysis of ammonia NTP systems are carried out.
This paper improves the PANES program and calculates the NTP performance parameters for potential alternative propellant ammonia, which provides a reference for the design of ammonia NTP system.
Study on Potential Leakage and Electrical Performance for Electrical Penetration Assemblies Under Severe Accident Conditions
Authors: Liu Yu , Liu Jing , Wang Cong , Gao Heng 
Published: October 2021

Abstract
EPAs are the penetrations used to minimize leakage and supply instrumentation or power cables path. Hence, sealing performance of EPA is important to contain radioactive materials and minimize their release without any filtration. Leakage behavior of containment under SAC is also required for the evaluation of mitigation strategies, risk analysis, PSA and emergency planning. In addition, electrical continuity for EPA cables is important for the operation of mitigation systems under different conditions, especially the SAC.
In this paper, it is focused on the review and study on leakage behavior and electrical continuity for EPA under SAC, which consists of the following key issues:
(1) Requirements, criteria and introduction of EPAs.
(2) Study on sealing and electrical performance of EPAs, especially the leakage mechanism, potential leak paths and sealing material performance, electrical degradation under different conditions.
(3) Preparations and plans for the EPA sealing and electrical performance test, introductions of our expectations and methods.
(4) Summary and outlook
Analysis and Research on Sodium Single Droplet Combustion
Authors: Zhao Lei 
Published: October 2021

Abstract
The sodium-cooled fast reactor is one of the preferred reactor types for the fourth-generation nuclear energy system. The sodium-cooled fast reactor uses metallic sodium as the coolant, which has a lot of advantages such as good thermal conductivity, larger heat capacity and lower neutron absorption rate. However, sodium is relatively active in chemical properties, and is prone to chemical reactions with oxygen and cause fires. The sodium fire caused by the sodium leakage of the sodium-cooled fast reactor seriously affected the safe operation of the reactor. The combustion of a single sodium droplet is the basis of sodium fire calculation and analysis. In this paper, by modeling the combustion of a single sodium droplet, a calculation and analysis program for the combustion of a single sodium droplet is developed, and compared and verified with related experiments. Through this program, the factors affecting the combustion of sodium droplets were analyzed and studied. The study showed that the diameter of sodium droplets, the temperature of sodium droplets, and the oxygen concentration have important effects on the combustion of sodium droplets. This work lays the foundation for the analysis of spray sodium fire accidents and is of great significance to the safety of fast reactors.
Analysis of Temperature Field in Hot Leg Piping of Space Nuclear Closed Brayton Cycle
Authors: Ma Wenkui , Ye Ping , Gao Yue , Zhao Gang , Yang Xiaoyong , Wang Jie 
Published: October 2021

Abstract
Direct gas cooled closed Brayton cycle can efficiently convert the fission energy of the reactor into electricity, which is an ideal energy conversion system for space reactor. The pipe between the reactor outlet and turbine inlet is designated as the hot leg piping. To reduce effect of radiation heat dissipation on the cycle performance, the hot leg piping requires unique design features. In this paper, the mathematical models of the space Brayton cycle and the radiation heat dissipation of the hot leg piping are established, and the influence of the heat dissipation on the cycle performance is analyzed. The results show that the heat dissipation of hot leg piping decreases the cycle performance. The turbine inlet temperature decreases by 10K and the cycle efficiency decreases by 1.35%. Based on the numerical simulation method, insulation performance of hot leg piping is studied. The results show that the insulation performance of internal insulation is better than the external insulation. Internal insulation with low thermal conductivity and large thickness of insulation layer decrease the maximum temperature of the hot leg piping, reduce the radiation heat dissipation and improve the cycle performance. The higher the space environment temperature and the lower the emissivity of the coating material reduce the radiation heat dissipation. However, the maximum temperature is easy to exceed the limit of pipe material. The analysis provides a reference for the hot leg piping insulation design of the space closed Brayton cycle in the future.
Numerical Simulation of Added Mass in Narrow Gaps of Multi-Layer Thin-Walled Shell of Fast Reactor
Authors: Duan Dexuan , Lu Daogang , Liu Yu , Li Donghao 
Published: October 2021

Abstract
Sodium-cooled fast reactor is one of the important reactors of the fourth-generation advanced nuclear power technology. Since the pool-type fast reactor has large size and thin wall that its rigidity is usually low. Therefore, the seismic design, analysis and verification of fast reactors are the key concerns of fast reactor safety evaluation. The fluid structure interaction effects of many special structures with narrow gaps under seismic loading need to be more accurately evaluated. At present, the method of added mass in the narrow gap fluid domain and the added mass formula of the infinitely long flexible coaxial cylinder given by the pressure field wave equation is not suitable for direct use. Based on the finite element analysis method, the frequency of different narrow gap annular cylinders such as coaxial three-layer cylinders in different modes is studied. According to the vibration formula, the added mass of narrow gap fluid domain is obtained, and the main pump support in the fast reactor is obtained. And explore the effect of the height of the fluid in the shell gap on the added mass of the shell. The design and inspection of narrow gap watershed structures such as cylinders and stack bodies have reference value.
Hydrodynamic Analysis of Steam Generator Under LOCA Conditions
Authors: He Xiaoqiang , Gao Puzhen , Yuan Weichao 
Published: October 2021

Abstract
Loss of coolant accident (LOCA) is one of most severe accidents of nuclear power plant. The coolant will discharge from the break when this accident occurs, consequently exert large hydrodynamic load on the primary loop system because of the fluid acting force as well as the propagation of pressure wave, then may further influence the pressure boundary of the primary loop. Steam generator (SG) is a crucial device with complex structure, hydrodynamic forces of its U-type tube and plate have been analyzed under various cases in this paper. Firstly, the reactor core and primary loop system of pressurized water reactor nuclear power plant (NPP) was modeled with RELAP5. Then a LOCA was simulated and the thermal hydraulic characteristics such as the pressure and mass flow rate were obtained. Finally, the forces at key positions of SG were calculated with the control volume integral method. The sensitivity analysis was also performed, including the break opening time, number of control volumes as well as the time step chosen in the hydrodynamic force calculation. The results have shown that the break opening time has a great influence on the numerical calculation that should be carefully set before the simulation, and the maximum hydrodynamic force is obtained under the condition that double-ended offset shear break at outlet nozzle of SG, and the pressure difference between the two sides of SG divider plate obviously increases compared with the normal operation condition. Thus, some meaningful suggestions have been provided for the design and safety analysis of SG.
Loss of Main Feedwater ATWS Accident Analysis for Ship Nuclear Power Platform
Authors: Qiu Jinrong , Feifei Song , Longze Li , Xiaofan Hou , Fei Chao 
Published: October 2021

Abstract
In this paper, the complete accident analysis calculation model of the ship nuclear power platform is established by using the developed accident analysis system code SACO (System Analysis Code under Ocean conditons), which is coupled with the ocean condition. Combined with the accident progress of loss of main feedwater ATWS, and on the basis of considering the reasonable assumption of the accident boundary, the main feedwater loss ATWS accident is analyzed. The accident sequence of loss of main feedwater ATWS is given, and the changes of the key thermal safety parameters such as the reactor power, heat removal from the active residual heat removal system, primary circuit pressure, pressurizer water volume, average reactor coolant temperature, peak temperatures of the fuel pellet and cladding are analyzed in detail. The results show the active residual heat removal system, reactor core reactivity feedback and design of the pressurizer safety valve could limit the primary circuit pressure and other key parameters without exceeding the safety limit value in the case of loss of main feedwater ATWS accident for ship nuclear power platform.
Study on Flow Boiling of Refrigerants in Micro/Mini-Channels
Authors: Wen He , Chenru Zhao , Hanliang Bo 
Published: October 2021

Abstract
The flow boiling heat transfer in micro/mini-channels is encountered in many compact heat exchangers such as the once-through steam generator in nuclear reactors due to its high heat transfer efficiency and thus accurately predicting the heat transfer coefficients is of great importance. Nowadays, for the high saturation temperature of water, refrigerants are more widely used in the flow boiling heat transfer experiments and a lot of heat transfer correlations are raised based on the data of refrigerants. Therefore, the accuracy of heat transfer correlations for refrigerants in flow boiling micro/mini-channels are analyzed in this paper. Firstly, five commonly used heat transfer correlations are summarized which can be simply divided into Chen type and Nusselt type and nearly 1000 experimental data of refrigerants conducted in the micro/minichannels are collected from 5 separate experiments. Then, due to the relatively low predicting accuracy of current heat transfer correlations, a new non-dimensional heat transfer correlation has been proposed by analyzing the experimental data with the application of Bond number in the range of 0.42–9.61, liquid Reynold number in the range of 34–8793, Boiling number in the range of 0.55 × 10−4 – 20.66 × 10−4. Finally, this new correlation is applied to water in different working conditions and the predicting accuracy has been analyzed.
The Effect of Flow Channel Geometry on Thermomechanical Performance of Printed Circuit Heat Exchanger (PCHE)
Authors: Jiragoontansiri Witiwat , Woravisuttsarakul Teerapat , Sae-pueng Rinrada , Sukjai Yanin , Shirvan Koroush 
Published: October 2021

Abstract
Climate change has garnered attention of communities world-wide. The carbon-free thermal energy systems such as nuclear benefit from compact and highly efficient heat exchanger technologies. The plate-type compact heat exchangers such as the Printed Circuit Heat Exchanger (PCHE) holds promise to fulfil these requirements. The chemical etching and diffusion bonding technology expands PCHE application for several different nuclear technologies. This work presents the thermal-hydraulic and structural analysis of PCHE for molten salt application with thermal energy storage. In this study, three different types of geometry are chosen for the analysis i.e., the zigzag channel type, the airfoil fin type, and the slotted fin type. For the working fluid, FLiBe (Li2BeF4) and Solar Salt (60% NaNO3 and 40% KNO3) are chosen for hot side and cold side respectively. Titanium grade 5 is chosen as the structural material. The study is conducted by Computational Fluid Dynamics (CFD) and Finite Element Method (FEM) analysis. The thermomechanical behavior including pressure drop, fluid temperature, velocity profile, stress, and deformation of the flow channel were considered in this work. From the results, the zigzag channel geometry gives the best thermal hydraulic performance in terms of heat transfer and pressure drop. The structural analysis shows that the stress intensity of all three geometries can exceed the maximum allowable stress (MAS) under certain conditions. Overall, the zigzag channel PCHE is still the most suitable geometry for this application.
Improvement of Conversion Ratio of Thorium Fuel in LWR by Adding Neutron Absorber
Authors: Takeishi Taichi , Takeda Satoshi , Kitada Takanori 
Published: October 2021

Abstract
The reproduction factor of Th232 is high in the thermal energy range and there is a possibility to achieve the breeding in LWRs. However, it is necessary to improve the conversion ratio since the breeding is difficult in LWRs. The conversion ratio can be improved by suppressing capture rate of Pa233 and by promoting capture rate of Th232. In addition, these capture rates can be modified by adding neutron absorber. Therefore, the neutron absorber is focused for improving the conversion ratio in this study.
The high resonance peaks of Pa233 capture cross section exist around 1∼100 eV. The resonance peaks of Th232 are higher than 10 eV. Thus, when the 1∼10 eV neutrons are suppressed in the fuel, the Pa233 resonance capture reaction is suppressed and the Th232 resonance capture reaction is promoted by neutron spectrum hardening. Therefore, six neutron absorbers that have high capture cross section peaks at 1∼10 eV were selected. The PWR pin cell calculations were carried out by Monte Carlo code MVP. The fuels are composed of a base material and an absorber. The base material is an oxidized fuel composed of U233(10 wt%), Th232(89.95 wt%), and Pa233(0.05 wt%). The amount of neutron absorber was adjusted so that the infinite multiplication factor becomes 1.33.
The impact of adding neutron absorber on the reaction rate was evaluated. As the result, the hardening of the neutron spectrum leads increase of the capture rate of Pa233, and the capture rate of Th232 in the epithermal energy range is increased. The change of capture rate of Th232 is greater than that of Pa232. Therefore, the conversion ratio is found to be improved by adding neutron absorber.
Nonlinear Low Bias Current Control for Magnetic Bearing System Using Active Disturbance Rejection Technology
Authors: Yao Yichen , Su Yixin , Yu Suyuan 
Published: October 2021

Abstract
Magnetic bearing is widely used in helium-turbine circle of the high temperature gas-cooled reactor and many other highspeed rotating machinery because of its unique advantages in vibration and noise reduction. However, the power consumption of magnetic bearing increases its cost of use. Moreover, the design of magnetic bearing controller relies on accurate system modeling. All these restrict the industrial application of magnetic bearings.
Based on the structure of the eight-pole magnetic bearing and its corresponding traditional decentralized differential PID control strategy, this paper proposes a magnetic bearing control framework including expected bearing force realization control strategy and centralized control strategy. Under this framework, a nonlinear low bias current control method for magnetic bearing system is given. Afterwards, an active disturbance rejection controller based on low(zero) bias current is proposed to compensate the gyroscopic disturbance and modeling uncertainty of the system. The controller can keep small loss of magnetic bearing and have good stability. It has a frame of active disturbance rejection control (ADRC) and its compensation performance is analyzed.
In order to verify the effectiveness of the controller, a corresponding experimental verification is carried out on the test rig. The results show that the control strategy is effective.
Evaluation of Serpent Capabilities for Hyperfidelity Depletion of Pebble Bed Cores
Authors: Robert Yves , Fratoni Massimiliano 
Published: October 2021

Abstract
Accurate burnup calculation in pebble bed reactor cores is today necessary but challenging. The continuous advancement in computing capabilities make the use of Monte Carlo transport codes possible to efficiently study individual pebbles depletion without making strong assumptions. The purpose is to eliminate unnecessary typical assumptions made in existing codes, while being flexible and suitable for commonly available computing machines.
Among the available codes, Serpent 2 provides extremely useful tools to make pebble beds modeling and simulation efficient. The explicit stochastic geometry definition handles irregular pebble beds with comparable performances to regular lattices. Optimization modes controlling the use of unionized energy grids, cross-sections pre-calculation and flux calculation through spectrum collapse or direct tally lead to high flexibility and optimal memory usage while limiting calculation time. Automated burnable materials division is a useful tool to lower the memory requirements while quickly generating the geometry and materials. Finally, parallelization and domain decomposition allow for decreasing unreasonable memory constraints for large cores.
This work thus explores the possibilities of Serpent 2 when applying depletion in pebble beds, compares the optimization modes and quantifies the simulation time and memory usage depending on the conditions of the calculation. Overall, the results show that Serpent 2 is well adapted to the use of small to large cores calculations with commonly available resources.
Evaluation of Interface Stresses and Cladding-Cladding Delamination Failures In U-Mo Fuel Plates
Authors: Ozaltun Hakan 
Published: October 2021

Abstract
Monolithic fuel is a candidate fuel form for the conversion of high-performance research reactors. This plate-type fuel consists of a high-density, U-Mo alloy-based fuel in a monolithic form that is sandwiched between zirconium diffusion barriers and encapsulated in an aluminum cladding. To qualify this fuel system, the program must demonstrate that the fuel meets the safety standards, has a predictable behavior, and performs well in a reactor. Although many plates have shown satisfactory thermo-mechanical performance in reactor, a small number of experimental plates exhibited failures, rendering performance of these plates unacceptable. Post-irradiation examinations of failed plates have revealed that various failures modes are present, including pillowing, blistering, buckling, warping, delamination, fracture, rupture, oxide spallation, etc. Previous studies have implied that such various operational and loading conditions could generate a complex stress-strain state, facilitating distinct failure modes that could impact the structural integrity. To understand such various failure modes, a comprehensive set of studies focusing on failure modes is being performed. As a part of these studies, this work studies the cladding-cladding separation failures that were observed in some of previous irradiation tests. To understand the root cause of cladding separation failures, selected plates from previous irradiation experiments were simulated under comparable operational conditions. Stress-strain state of the plates at the interface were examined. The results were comparatively evaluated to understand possible effects of stress state of the bond-line on delamination failures. This paper presents these findings, discusses such failure modes, and examines the influence of various factors on previously observed cladding delamination failures.