bluuebunny/crossref_metadata_2025_split huggingface dataset:
A Novel Numerical Model of a 7-Pin Sodium-Cooled Fast Reactor Wire-Wrapped Fuel Bundle
Authors: Lyu Daizhen , Wang Naihua 
Published: November 2022

Abstract
As a low Prandtl number fluid, the numerical simulation of liquid sodium is different from the typical numerical simulation, because the turbulent Prandtl number is no longer a constant and the turbulent Prandtl number model must be developed. In this study, a novel numerical model is used to carry out the steady-state numerical simulation of a 7-pin sodium-cooled fast reactor wire-wrapped fuel bundle. The turbulent Prandtl number of liquid sodium is no longer approximate to the default value of 0.85. Kays model is used as a UDF of Fluent to modify the turbulent Prandtl number. The novel numerical model is computationally economical and can ensure the accuracy of the calculation. Numerical simulations in five different working conditions are performed (104 &lt; Re &lt; 105).
Comparing the CFD results with the experimental correlation, it is found that the numerical model with Prt = 0.85 which used the same geometry would overestimate the heat transfer performance by 9% compared to the novel numerical model, and the novel numerical model is closer to the experimental correlation. In addition, the analysis of the flow field and temperature field shows that the axial velocity changes periodically along the axial direction due to the influence of wire wraps. The period equals one-quarter of the wire pitch. In the different cross sections, the cross flow which is induced by the wire wraps forms vortices in the subchannels. The surface temperature of fuel rods on the windward side is different from that on the leeward side due to the blockage of wire wraps. And the maximum surface temperature occurs at the location of the gap between fuel rod and wire wrap. The conclusions of this study provide effective suggestions for the optimization of numerical simulation and the prediction of the hot spot of the sodium-cooled fast reactor wire-wrapped fuel bundle.
Numerical Investigation of Large-Scale Pool Boiling Outside the Condenser Tube of Passive Containment Cooling System
Authors: Pan Adong , Wang Naihua 
Published: November 2022

Abstract
The passive containment cooling system (PCCS) based on gravity heat tube is introduced to the new generation of nuclear technology. Large-scale pool boiling outside the condenser tube plays an important role in the operation of PCCS due to its excellent heat transfer capacity. Nonetheless, there are few studies on pool boiling outside the condensate tube of passive containment cooling system. In this paper, the pool boiling phenomenon outside the condenser tube of PCCS was numerically investigated. The effects of inclined angle and tube diameter on the heat transfer characteristics of the condenser tube were studied. The VOF method combined with the Lee mass transfer model was employed to model the pool boiling outside the condenser tube of PCCS. The influence of inclined angles and diameters on the boiling heat transfer has been analyzed. The results indicate that in the case of vertical tube, the average boiling heat transfer coefficient increases with the increase of tube side height. When the condenser tube is inclined, the bubbles on the upper wall of the condenser tube will leave the heating wall earlier, while the bubbles on the lower wall tend to merge into large bubbles. Combining the influences of the two aspects, the tube inclination has little effect on the heat transfer effect. In the case of vertical tube, the boiling heat transfer coefficient decreases with the increase of diameter.
Study on Social Stability Risk Assessment of Nuclear Facilities
Authors: Zhu Rongxu , Zhao Feng , Zhang Xiaofeng , Zhang Meng 
Published: November 2022

Abstract
This paper studies the social stability risk assessment of nuclear facilities, formulates the assessment procedure, determines the assessment content and scope, and establishes the assessment risk index database and risk evaluation criteria. The risk matrix method is introduced to judge the risk degree. The expert evaluation method is used to calculate the weight of social stability risk indicators. Finally, the project risk quantification and quantification are combined to carry out the work, which has high scientific and credibility, and is operable and implementable for the social stability risk assessment of nuclear facilities. Finally, a nuclear facility is taken as an example.
Drop Analyses and Verification of the Spent Nuclear Fuel Transportation Cask Under Hypothetical Accident Conditions
Authors: Liu Guangdong , Wu Weiliang , Zhu He , Ma Qingjun , Li Qipeng 
Published: November 2022

Abstract
According to IAEA regulations, the spent nuclear fuel transportation cask must be able to withstand a free drop from the height of 9 meters on an unyielding surface. The impact limiter of the cask absorbs most of the impact energy and reduces the deceleration loads on the cask and contents. In order to accurately estimate the safety of the cask and the spent nuclear fuel, the calculation of drop acceleration is very important. In this study, based on the absorption of energy by an impact limiter is equal to the work done in crushing a volume, the maximum deceleration is estimated during collision process between cask and the ground. Then using the LS-DYNA computer program to simulate the cask drop behavior of 9 meters free drop. By comparing the results of theoretical analysis, numerical simulation and actual drop tests, the cask maximum acceleration and the impact limiter deformation are in good agreement, which prove the validity of the spent nuclear fuel transportation cask drop analysis. The stress intensity results for 9 meters free drop loading cases satisfy the allowable stress limit criteria as presented in “ASME Boiler and Pressure Vessel Code”. The spent nuclear fuel transportation cask meets the requirements of IAEA in the 9m free drop hypothetical accident conditions.
Research and Application on Simulation Test Between Turbine Governing System of Half-Speed Turbine and Rod Position Control System in Nuclear Power Plant
Authors: Yang Liping 
Published: November 2022

Abstract
Coordinated control between reactor and turbine is taken by Co-commissioning test between turbine governing system (GRE) and reactor rod position control system (RGL). Matching and completeness of interfaces between reactor and turbine could not be verified by static verification methodology before the unit starts up. Characteristics of coordinated control are analyzed in pressurized water reactor plant. Simulation Co-commissioning test program by static and dynamic verification methodologies between turbine governing system and rod position control system is put forward based on functional classification of interfaces signals. Static characteristics of interface signals between GRE and RGL are checked to verify static matching and completeness of interface signals by simulating the steam turbine to run under various operating conditions in verification method of static simulation. Dynamic response characteristic parameters of interface between GRE and RGL are analyzied and computed to verify dynamic matching and completeness of interface functions by simulating GRE and RGL to run under 4 typical coordinated control operating conditions in verification method of dynamic simulation. The case results have shown that matching and completeness of coordinated control interfaces between reactor and turbine can be verified efficiently by applying the test program into nuclear power plants successfully. And dynamic logic problems of interface are discovered and solved before the unit starts up.
Spent Fuel Pool Risk Analysis for a CPR1000 Unit
Authors: Weidong Liu , Leilei Liu 
Published: November 2022

Abstract
After the Fukushima nuclear accident, the safety level of the spent fuel pool has gradually attracted increasing attention. This paper develops a probabilistic safety assessment (PSA) model for the spent fuel pool of a CPR1000 unit, and quantitatively evaluates the frequency of fuel damage. Possible weaknesses of the spent fuel pool are identified, and suggestions are provided to optimize the operation and management of nuclear power plants, thereby further improving the accident prevention and mitigation capabilities of the spent fuel pool and enhancing the overall safety of nuclear power plants. Through the quantitative analysis of spent fuel pool PSA, SFP fuel damage frequency of 6.99E−09/reactor year is finally developed. Compared with the core damage risk, SFP fuel damage frequency is very low, less than 1% of the core damage frequency, and the risk level is far lower than the reactor core. From the quantitative results of the initiating events, PTR discharge piping breakup and loss of offsite power are the dominant contribution to fuel damage frequency, mainly due to 1) After PTR discharge piping breaks, the ruptures can’t be isolated, and the emergency make-up is not available, finally leading to fuel damage; 2) Under refueling state, the loss of offsite power, diesel generator failure, while the main transformer, RRI/SEC, LHP/LHQ is to carry out planned maintenance, and also the emergency water supply failure jointly lead to fuel damage.
Prediction and Uncertainty Analysis of Stress Corrosion Fraction Failure Probability for Pressurized Water Reactor Materials
Authors: Kong Deyan , Liu Xu , Cheng Jie , Wang Jianjun , Zhang Jinqiang , Chen Dingyong , Li Qingzhao , Liu Yanting 
Published: November 2022

Abstract
The failure of stress corrosion fraction (SCF) will lead to the ejection of nuclear power plant related equipment, which will affect the safety and economy of nuclear power plant. In this paper, a SCF failure analysis method for primary coolant circuits materials was established based on Paris model, and the uncertainty of fracture toughness was transformed into an integral form to improve the calculation efficiency. Taking the weld of thermowell as an example, the probability of SCF failure was calculated, and the uncertainty was analyzed by Wilks’ method. The influences of simple random sampling (SRS), Latin hypercube sampling (LHS) and Halton low discrepancy sequence on the uncertainty quantification of the calculated results were studied. At the same time, a surrogate model was established based on polynomial chaos expansion (PCE) method to study whether this method was suitable for SCF probability calculation. The results show that Halton sequence with 1000 samples can make the mean and variance convergence of failure probability better than SRS and LHS. When calculating the upper limit of tolerance interval, the mean and median of results corresponding to LHS are similar to those of SRS, but the dispersion degree of LHS is lower than SRS, while the results corresponding to Halton sequence are smaller than those corresponding to SRS and LHS. The increase of Wilks’ order can reduce the conservatism. When the order is 4, both the computational efficiency and the computational accuracy are considered. The results of the surrogate model based on PCE are basically consistent with those of the original program, but the amount of calculation is greatly reduced. This method is suitable for SCF probability analysis.
Multiphysics Analysis of UN Fuel Performance With a Two-Layered SiC Cladding Under Multiple Operating Conditions in a Light Water Reactor
Authors: Liu Rong , Han Huiyun , Liu Shengyu 
Published: November 2022

Abstract
In recent years, UN fuels have emerged as potential alternatives to UO2 fuels, due to its superior thermophysical properties, such as high thermal conductivity, melting point and fission density. And SiC is considered as one of the new generations of accident-tolerant cladding materials because of its excellent high temperature strength, creep resistance, lower thermal expansion and better irradiation performance. In this paper, the thermal and mechanical properties of the two-layered SiC cladding and UN fuel are reviewed. Then, based on the developed fuel performance analysis code CAMPUS, the multiphysics models of two-layered SiC cladding and UN fuel are implemented into the CAMPUS code. After that, the fuel performance of three fuel cladding systems, UN-two-layered SiC cladding, UN-Zircaloy cladding and UO2-Zircaloy cladding, are simulated and analyzed under both normal operating and accident conditions of PWR.
Our calculation results show that using UN fuel can significantly reduce the maximum temperature of the fuel under LOCA condition, resulting in a significant delay in cladding failure time. Thus, the safety margin of the reactor can be improved. After the two-layered SiC cladding is further assembled, the failure time of fuel rod under accident condition is found to be effectively delayed due to better mechanical properties of SiC.
3D-Printed SAW Sensor Development for Enabling Remote Monitoring of Advanced Reactors
Authors: Gkouliaras Konstantinos , Marr Jake , Theos Vasileios , Chatzidakis Stylianos 
Published: November 2022

Abstract
Recent and future advancements in the upcoming digital era of Nuclear Industry would rely heavily on real-time, remote, and unattended monitoring, as well as continuous data collection. Advanced reactor designs (e.g., Microreactors, SMR, etc.) intended for installation in remote and possibly inaccessible locations would require robust, compact, passive, low-cost, radiation-tolerant, and energy efficient sensor systems that could be remotely interrogated and provide real-time information of operating and safety-related parameters (temperature, flux, etc.). In addition, to ensure economic competitiveness and minimize sensor costs, it is equally important to guarantee a simple, streamlined, and cost-efficient manufacturing procedure. To achieve this, we explore the development of low-cost 3D-printed Surface Acoustic Wave (SAW) sensors optimized for remote monitoring of advanced reactors.
SAW sensors are passive devices that can convert an electromagnetic wave into a surface acoustic wave propagating along a piezoelectric substrate. The mechanical wave is then reflected and eventually converted back into an electrical signal which can be broadcasted. These characteristics offer unique advantages including sensor simplicity, power-free operation, and radiation tolerance since no integrated circuits are needed. To develop customized SAW sensors, we used a state-of-the-art aerosol jet printer that enables precise fabrication of sensors with features as small as 10 μm. In this work, LiNbO3, LiTaO3 and Quartz substrates are used to demonstrate the feasibility of the proposed approach by 3D-printing two-port SAW structures using conductive silver nanoink. A parametric analysis is performed in order to optimize the printing procedure and investigate optimization of printing line resolution using optical and Scanning Electron Microscopy.
Research on the Influence of Water Tank Liquid Level on Low Pressure Two-Phase Open Natural Circulation Flow and Heat Exchange
Authors: Zhang Shengnan , Sun Zhongning , Zhang Nan 
Published: November 2022

Abstract
In order to study the influence of the water level of the water tank on the flow instability and heat transfer character istics,which is of the low-pressure two-phase open natural circulation system. The open natural circulation system was numerically simulated and experimentally verified. Numerical simulations are calculated by using a homogeneous flow model in this study, The flow rate and heat removal power of the system under different water level conditions were calculated and recorded, and the trend of changing with the water level was obtained. The results were then compared with the experimental results. It shows that the calculated open natural circulation flow and heat removal power results are consistent with the experimental. As the liquid level decreases, the flow state of the system gradually changes from the periodic flashing state inside the adiabatic section to the stable flashing operating state, and the heat exchange power also increases gradually. times. Therefore, in the low-pressure system, reducing the liquid level of the water tank can achieve a good effect of enhancing the flow stability of the low-pressure two-phase open natural circulation, and also has a good effect on the improvement of the heat removal power of the system.
Numerical Study of the Gas-Liquid Mass Transfer in the Volume Control Tank
Authors: Hu Jian , Chi Xiangyu , Wang Naihua , Su Qian 
Published: November 2022

Abstract
In this paper, a novel numerical model is established to conduct transient simulations of gas-liquid mass transfer in volume control tank (VCT). The surface divergence model was coupled with the CFD model to model the gas-liquid mass transfer near the interface. And the species transport model was used to model the homogeneous mass transfer inside the liquid. The simulated data is verified by experimental results. And the deviation between simulation results and experimental results is less than 6.67%. This model can accurately predict the gas-liquid mass transfer in VCT. Moreover, the flow field and concentration field in the VCT were analyzed. And the results show that there are three dominant mechanisms of gas-liquid mass transfer in VCT. The first mechanism is the entrainment of high-concentration fluids by surface jets at the gas-liquid interface. And the second is a double vortex structure above the jet inlet to pump the high-nitrogen-concentration fluid to a low-nitrogen-concentration area. The third is the mass exchange between the vortex at the end of the jet and the jet. Furthermore, the effects of thermal parameters (the temperature, operating pressure and inflow flow rate) on the VCT interface mass transfer process were also studied.
Optimized Moderator Design and Analysis of a Pin-Type Supercritical Carbon Dioxide Reactor Based on RMC
Authors: Zhao Xingyu , Liu Minyun , Cui Rongyi , Huang Shanfang , Wang Kan , Lu Chuan 
Published: November 2022

Abstract
This study analyzed an yttrium hydride (YH2) moderated supercritical carbon dioxide cooled reactor loaded with pin-type, beryllium oxide diluted oxide fuel elements to reduce the critical enrichment. The impact of the YH2 on the coolant void reactivity was studied along with a moderator zoning scheme to flatten the radial power distribution. The YH2 was added as hexagonal moderating rods at the center of the fuel assemblies. The core was modeled using the continuous-energy Reactor Monte Carlo code, RMC, with the On-The-Fly cross sections treatment. The results showed that the YH2 moderator increased the thermal fission and reduced the critical enrichment of the core with the same diluent volume fraction by more than 30%. The YH2 moderator significantly softened the neutron energy spectrum and reduced the neutron leakage upon core voiding, resulting in both a weaker positive spectral reactivity feedback and a weaker negative leakage reactivity feedback during core depressurization. For an UO2-loaded core, the YH2 gave a lower negative coolant void reactivity, while for a MOX-loaded core with diluent volume fractions smaller than 35%, the spectral feedback was more important and the YH2 strongly reduced the positive coolant void reactivity to less than $1. Arranging the YH2 in the peripheral assemblies reduced the radial power peaking factor to 1.319. The study shows that the YH2 moderator can reduce the critical enrichment, make the core less sensitive to voiding, and can flatten the radial power distribution of a single-enrichment core through moderator zoning.
Numerical Investigations on Structure Effect for Steam Condensation in the Presence of Air Under Vertical Condition
Authors: Zhou Shuhang , Meng Xianke , Bian Haozhi , Ding Ming 
Published: November 2022

Abstract
Steam condensation in the presence of non-condensable gas is a general phenomenon in various industrial applications. It has been broadly investigated by experimental studies and numerical analyses. Previous experimental studies and numerical analyses have focused on single structure. To evaluate the condensation heat transfer characteristics of steam for various condensation structure, the present work conducted numerical simulations based on various structure at a wide parameter range (pressure from 0.2 to 1.6 MPa and air mass fraction from 0.16 to 0.71). In the assessments, the condensation heat transfer coefficient with various condensation structure (single tube, plate, cylinder, spherical) were investigated. The results indicate that the structure of the condensation surface has a significant effect on the heat transfer characteristics of steam condensation. There is almost no difference in the heat transfer performance of the four condensation structures under the low pressure and high air mass fraction condition. With the increase of pressure and the decrease of air mass fraction, the heat transfer performance of the cylinder is the best, followed by the spherical and plate, and the worst is the single tube. In addition, the condensation heat transfer coefficient (CHTC) of the cylinder can be 40% greater than that of the single tube at a pressure of 1.6MPa and an air mass fraction of 0.16.
Performance Evaluation of Typical 3-Loop PWR With Annular Fuel Assemblies Under Typical Accidents
Authors: Yang Bowen , Shan Jianqiang , Ge Li , Chen Kaidong 
Published: November 2022

Abstract
The externally and internally cooled annular fuel is considered to have a good prospect with the shorter heat transfer path, larger heat transfer surface and higher burnup depth, which can effectively improve the safety of the reactor. This paper takes the nuclear power plant CPR1000, which is representative in China, as the reference power plant. The system analysis code NUSOL-SYS was used to estimate the thermal and hydraulic parameters of the reactors during the transient period, and the sub-channel analysis code NACAF was used to estimate the MDNBR of the core. The results showed that in the large break loss of coolant accident, since the annular fuel has less heat storage and a larger heat transfer surface during steady-state operation, the peak temperature of its cladding surface was only 861K, which was much lower than the corresponding temperature 1233K of the solid fuel. In addition, because the thickness of the annular fuel is small, the temperature of the inner and outer cladding was basically the same, but because the inner channel does not have the spacer grid and other effects, the level of the inner channel was slightly higher than the outer channel, and the inner wall surface was completely quenched slightly earlier than the outer wall surface. In the main steam line break accident and the reactor coolant pump rotor seizure accident, the annular fuel had a larger MDNBR, which caused a smaller risk of fuel damage. During the rod ejection accident, the average temperature of the annular fuel was lower than 900K, and the average enthalpy was lower than 200kJ/kg, which is far lower than the corresponding value of the solid fuel. In addition, the results of the annular fuel calculation model at 150% power showed that in the above four accidents, it was feasible to increase the power of the core loaded with annular fuel by 50%.