bluuebunny/crossref_metadata_2025 huggingface dataset:
A Knowledge Retrieval Method for Layout Design of Nuclear Power Reactor Coolant System Based on Hypernetwork and Deep Structured Semantic Model
Authors: Wang Bin , Hu Bingtao , Zhang Zhifeng , Feng Yixiong , Tan Jianrong
Published: November 2022

The layout design of nuclear power reactor coolant system requires a large amount of knowledge that satisfies many disciplines, which will waste designers a lot of time retrieving relevant knowledge in the design process. In order to obtain the knowledge efficiently and accurately, a knowledge retrieval method based on hypernetwork and deep structured semantic model (DSSM) was proposed. The knowledge hypernetwork model consisted of a designer sub-network, a design task subnetwork, and a design knowledge resource sub-network. Nodes in each sub-network and between different sub-networks were connected through special edges, which represented correlation degree information. Then an improved DSSM model was used to evaluate relevance at the semantic level between user query information and knowledge elements in hypernetwork. Correlation scores will be obtained based on relevance at the semantic level, and knowledge elements with lower scores will be removed during the process. Finally, the Bayesian method was used to calculate the knowledge recommendation probability to obtain the most relevant knowledge retrieval results. The knowledge retrieval results were sorted from high to low according to the calculated probability. A case study conducted in this work showed that the proposed approach was effective in capturing relevance at the semantic level and supporting efficient and accurate knowledge retrieval services.
Construction and Application of Mitigation System Performance Index (MSPI) Platform for a China Nuclear Power Plant
Authors: Zhang Jie , Wang Jing , Zhang Peiying , Tan Kun , Yang Minxue
Published: November 2022

The mitigation system performance index is used to evaluate the ability of the safety-related system to be put into operation to mitigate the accident when an accident occurs in the nuclear power plant (NPP). This index can effectively monitor the status of the safety system and reasonably reflect the impact of equipment performance on the safety level of the nuclear power plant. Typical pressurized water reactor (PWR) nuclear power plant mitigation systems include emergency alternating current (AC) power system, high pressure safety injection system, auxiliary feedwater system, residual heat removal system and cooling water support system. Based on the probabilistic safety assessment (PSA) model, this paper develops a nuclear power plant MSPI computing platform to monitor and judge the risk level of the mitigation system, and guide the nuclear power plant to carry out production activities such as risk-informed preventive maintenance, corrective maintenance and production planning. This paper takes the Essential Service Water System (SEC) of a nuclear power plant in China as an example to verify the proposed method, and conducts the trend analysis of MSPI in the 8 quarters of 2020–2021. The results show that MSPI can timely reflect the influence of the unavailability and unreliability of system equipment on the overall safety status of the power plant from the perspective of risk.
Research and Development of Rock Cavity Disposal Technology in China
Authors: Yin Wencheng , Pan Yuelong , Zhang Xueling , Liu Yu , Ma Jian , Li Tongtong
Published: November 2022

Rock cavity disposal is regarded as a preferred option of radioactive waste final disposal solutions with less disturbance to surface and better isolation from people and surface environment which facilitates the long term surveillance and control of facility both during operation and after closure. This paper provides the research progress of key aspects of rock cavity disposal technology, such as site selection criteria, engineering design scheme of disposal vaults, technique process design of the handling of waste packages, research & development of intelligent control system and equipment of waste transportation, documentation of safety case and supporting safety assessment, numerical simulation of radionuclides migration process, and etc. The first practice of rock cavity disposal in China is also discussed. It shows that developing the rock cavity technology is reasonable, feasible and has significant potential in China. Finally, this paper looks forward to the research focus in the next stage. It is suggested to study the compatibility between the complexity level of engineering barrier measures and the waste hazards and to improve the reliability of future evolution and scenario assumption. The nuclides migration behavior study and migration parameter determination involving more key nuclides in waste from other origins are also of importance.
Research on Severe Accident Management Guidelines for HPR1000 NPP
Authors: Pingting Jiang , Peng Chen
Published: November 2022

The Severe Accident Management Guideline (SAMG) is an integrated document for a nuclear power plant (NPP) to deal with severe accidents effectively, with the basic objective to maintain the integrity of the containment in both short term and long term as far as possible. After Fukushima Daiichi nuclear accident, insights on severe accident prevention and mitigation are put forward in the process of a newly designed nuclear power plant. Hua-long Pressurised Reactor (HPR1000) is a generation-III NPP in China, with necessary dedicated system and equipment that designed at the beginning during severe accidents, and is also planned to have associated procedures after accidents. The paper focuses on development process of SAMG for HPR1000 nuclear power plant. Firstly, the characteristics of the plant are sorted, with both dedicated equipment required for severe accident mitigation and other equipment. Then the main objective and principles in developing SAMG are defined. After that, the structure of SAMG is discussed and finally determined referred to the latest insights all over the world and also considering the good practice in China. Additionally, diagnosis process and severe accident strategy are described. In relation to this, flexible strategy in HPR1000 NPP is discussed and included in the integrated SAMG.
Design of Liquid Level Control System of Steam Generator Based on Neural Network PID Controller
Authors: Xiao Longhao , Sun Peiwei , Wei Xinyu
Published: November 2022

Steam generator (SG) is an important equipment of the nuclear power plant, and the stability of its liquid level affects the safe operation of the nuclear power plant. SG is a complex system with nonlinear, time-varying, nonminimum-phase, small stability margin and large time delay. In actual operation, it is difficult for classical PID control to ensure a satisfactory control performance.

In this paper, the neural network methods are used to optimize the parameters of the PID controller, and a neural network controller is designed. The controller of the system consists of two components: a classical PID controller, which realizes control through a closed loop; a single-hidden-layer neural network based on the BP (back propagation) model. The neural network calculates the coefficients of the classic PID controller through matrix operations. Two weighting matrices are adjusted according to the gradient descent method to reduce the loss function and realize the training process. The control system is deployed to a SG simulation model through Simulink. The typical working conditions are simulated and investigated. The control performance is compared with that of the classical PID controller.

Through analysis, it is confirmed that the neural network PID control system can meet the control requirements with fast response speed, short settling time, stable control effect under various working conditions, and strong anti-interference ability. The results prove that the neural network control has greater advantages and better application value than the classical PID controller.
Extension of the IVR Analysis Module in Modeling the Ex-Vessel Water Cooling
Authors: Yu Peng , Zhao Yuer , Wang Chen , Yang Xiaoming , Ma Rubing , Yuan Yidan
Published: November 2022

In-vessel retention (IVR) through external reactor vessel cooling (ERVC) is an important severe accident mitigation strategy widely adopted in some advanced light water reactors, e.g. AP1000 and HPR1000. The reactor pressure vessel is supposed to be flooded externally by water, such that the decay heat could be efficiently removed and the radioactive corium materials retained inside the vessel. Lumped parameter models are used in engineering design to calculate the melt pool heat transfer and thus assess the effectiveness of IVR. In this methodology, a simple constant temperature is normally assumed for the ex-vessel surface accounting for the external water cooling. Since the RPV and cooling water are coupled in thermal behavior, modelling of the related heat transfer between them may be of interest. Under the PISAA framework, we developed the IVR module for the IVR analysis. In this paper, we further extend the code capability by considering the ex-vessel heat transfer. Possible conditions of the external cooling water e.g. pure liquid phase convection and boiling as well as the corresponding heat transfer correlations would be considered. The newly implemented model is then verified with acceptable results achieved. With this new feature, the coupled heat transfer analysis of the RPV and external cooling water is investigated. Compared with the original treatment with constant temperature boundary conditions, the new treatment with convective heat transfer affects the predictions of outer vessel temperature and residual wall thickness while the influence on the inner vessel temperature is negligible.
Heat Transfer for Forced Flow of Helium Gas in a Small Diameter Tube With Different Heated Length
Authors: Honjo Yushi , Xu Feng , Liu Qiusheng , Shibahara Makoto
Published: November 2022

As one of the cooling methods for the blanket in a fusion reactor, forced convection using helium gas has been considered as one of the promising candidates. In this study, in order to clarify the effect of tube length on heat transfer for forced flow of helium gas under various velocities and pressures, experiments on forced convection heat transfer for helium gas in a small diameter tube were conducted. A circular tube made of platinum with an inner diameter of 1.8 mm and a heated length of 50 mm was used in this experiment. The tube was heated with exponentially increasing heat inputs. As a result of the experiment, the heat transfer process can be considered in the quasi-steady state when the e-folding time is larger than about 1.5 s. In addition, the heat transfer coefficient increased with the increases in velocity and pressure. The heat transfer was also higher than that of conventional turbulent heat transfer correlation. By comparing with the experimental results of tube with different heated lengths, it was found that the heat transfer coefficients for the heated length of 50 mm were higher than those of the one with a length of 90 mm.
A Novel Numerical Model of a 7-Pin Sodium-Cooled Fast Reactor Wire-Wrapped Fuel Bundle
Authors: Lyu Daizhen , Wang Naihua
Published: November 2022

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 < Re < 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

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

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

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

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

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

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

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

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

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

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

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

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

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%.
Experimental Research on the Retention Characteristics of Hydrophilic and Hydrophobic Aerosols in Jet Regime
Authors: Ma Qianchao , Zhou Yanmin , Gu Haifeng , Sun Zhongning , Yang Jinghao
Published: November 2022

Aerosol is one of the main forms of fission products in nuclear reactor accidents, which are derived from direct leakage of solid fission products and homogeneous and heterogeneous nucleation of gaseous fission products. Pool scrubbing may reduce the leakage of radioactive aerosol, and it is vital for the prediction of radioactivity to accurately grasp the retention efficiency. In this paper, experimental research on aerosol retention characteristics in jet regime was carried out. Titanium dioxide and cesium iodide were used to simulate radioactive aerosols with different hydrophilicity, and the retention efficiency under different inlet pressure and aerosol size was obtained. The results show that the retention efficiency of aerosol increases significantly while the inlet pressure increases from 43 kPa to 110 kPa. While the inlet pressure is higher than 110 kPa, the retention efficiency still increases with the increase of inlet pressure, although the increasing trend is not obvious. The most easily permeable range of cesium iodide aerosol is smaller than that of titania aerosol, and the retention efficiency of cesium iodide aerosol is always higher than that of titania aerosol, which may be caused by hydrophilicity. However, with the increase of stokes number, the difference of retention efficiency between the two aerosols gradually decreases. The difference between different aerosols was analyzed in this paper and it was found that the influence of hydrophilicity was negatively correlated with stokes number. The results of this research can be used to assist the validation of pool scrubbing model.
Reactivity Introduction Accident and Control Strategy Analysis of Small Modular Supercritical CO2 Brayton-Cycle Reactor System
Authors: Ming Yang , Jin Yang , Qiao Yanlong , Zhao Fulong , Tian Ruifeng
Published: November 2022

In this paper, a dynamic model of a 5MW small modular supercritical CO2 Brayton-cycle reactor system is established based on Modelica language, and the control rod ejection accident simulation with different reactivity is carried out. This paper takes a 5MW small modular supercritical CO2 Brayton cycle reactor system as the research object. Based on the modular modeling idea, the dynamic model of the system including reactor, turbine, compressor, regenerator, cooler, and other main equipment is built, and the reactor power control system is based on the PID control principle is designed. The results show that the system has certain inherent safety, and the designed control system can coordinate the main equipment under accident conditions, maintain the important parameters such as reactor core fuel temperature, coolant temperature, and reactor pressure within the limit, and control the system flow, permeability and compressor rotor speed within a reasonable range. The work of this paper provides a certain reference for the selection of control strategy and safety system design of the new supercritical CO2 reactor system.
Molecular Dynamics Study on Thermal Conductivity of Unirradiated and Irradiated Symmetrical Tilt Grain Boundary 3C-SiC
Authors: Cai Ziqi , Zhang Qingmin , Shao Zhuang , Li Yuanming
Published: November 2022

The cubic silicon carbide (3C-SiC) has been considered as a candidate structural material for several types of advanced nuclear reactors. The effects of cascade collision on thermal conductivity in symmetrical tilt grain boundary (GB) were studied by Molecular dynamics (MD) simulations. The thermal conductivity of 3C-SiC at Σ5(210)[001] GB was calculated using non-equilibrium molecular dynamics (NEMD) methods. A relatively small simulation unit was used to analyze the effect of different energies of incident PKA (primary knock-on atoms) on the thermal conductivity of 3C-SiC and to compare the results with perfect structure GB system. Finally, the vibrational density of states (VDOS) of atoms in the GB region was calculated to analyze the phonon mismatch at the interface. Calculations show that cascade collisions generated by energetic atoms will result in a decrease in thermal conductivity of the Σ5(210) GB system, but the effect varies in different regions, with a sharp decrease in thermal conductivity and an increase in thermal resistance for the intracrystalline region, while the magnitude of change in either thermal resistance or thermal conductivity is not significant in the GB region. Irradiated model shows a higher GB energy compared to the unirradiated model. For all irradiated models, lattice defects have a significant effect on the thermal conductivity of the GB system, depending on the spatial structure of the GBs. the results of the VDOS analysis suggest that an increase in the degree of atomic lattice mismatch near the interface is responsible for a further increase in the thermal resistance of the irradiated GB system.
Development of the Digital Human-Computer Interaction Platform of the M2LFR-1000 Reactor Based on QT and ICoCo
Authors: Liu Tianyi , Ding Tao , Shen Yanyi , Chen Hongli
Published: November 2022

Human-computer interaction interface refers to a human-oriented, humanized and comfortable information display system. Building an interactive relationship with users in design is a bridge between users and information. The purpose of the human-computer interaction interface is to exert mutual influence between humans and machines, so that both parties understand each other’s information and keep the information exchange channel smooth. However, at present, there is a lack of visual human-computer interaction interface for lead reactor, and the results calculated by the thermal hydraulic calculation program cannot be presented directly to users. Moreover, it is very inconvenient for users to modify some parameters. Therefore, for the M2LFR-1000 lead-cooled fast reactor, this paper will introduce the development of a digital human-computer interaction platform based on the ICoCo concept and QT framework. The system code RELAP5 has been selected as the underlying thermo-hydraulic program. The platform can visually display the overall structure of the primary circuit of the M2LFR-1000, and can display parameters during calculation such as the inlet and outlet temperature of the primary circuit coolant in the core, the inlet and outlet temperature of the secondary circuit coolant in the main heat exchanger, and the mass flow of the core. The user can also adjust the parameters of the time-related components such as the inlet and outlet temperature of the secondary circuit coolant in the main heat exchanger when the calculation is being performed. In this way, the calculation of the thermal-hydraulic program can be synchronized with the display of the calculation results. At the same time, the user’s needs can be fed back to the thermal-hydraulic program in time. In addition, since QT is a cross-platform interface development software, the visual human-computer interaction interface developed by QT also has a good code compatibility, which is convenient for subsequent development and upgrade.
Experimental Investigation of Cable Arrangement on the Cable Burning Behavior in a Cone Calorimeter
Authors: Yang Ping , Huang Xianjia , Peng Lan , Xing Chaoliang , Zhao Chunyang
Published: November 2022

Cables are one of the main fire hazards in nuclear power plants (NPPs). Cable arrangement, especially cable spacing in an electrical cable tray has a significant effect on cable tray fire. Predicting cable fire behavior in NPPs is necessary for the nuclear safety analysis. In this work, cable fire experiments with halogen-free flame retardant cable (XLPE cable with a diameter of 8.55mm) were carried out in a cone calorimeter. Cable spacing was changed from 0 mm to triple of cable diameter, including 7 different spacing arrangements under 3 different external heat fluxes. Besides, the experimental data of two other halogen-free flame retardant cables in the literature available are used. With the increase of cable spacing, the heat release rate per unit area (HRRPUA) was significantly increased and the fire duration was considerably decreased, while the time to ignition (TTI) of the cable was basically invariable with different cable spacings. Furthermore, the effects of the cable spacing on the cable burning become constant when the cable spacing was larger than a critical cable spacing. This work provides experimental support for industrial cable fire risk assessment, and the above findings are beneficial for fire prevention in the design stage of NPPs.
Research and Application of Pre-Development Software Qualification Technology for Nuclear Safety I&C Systems
Authors: Wang Shengchao
Published: November 2022

With the continuous advance of the mass construction of nuclear power in China, it has become an irresistible trend to use mature commercial-off-the-shelf (COTS) to perform safety functions in the safety important instrumentation and control (I&C) systems to reduce the construction cost of nuclear power plants (NPPs). The quality of pre-developed items especially the pre-developed software (PDS) has been paid more and more attention by stakeholders. By combining with the specific requirements of international mainstream guidelines and standards related to nuclear safety I&C systems or equipment PDS qualification, a set of strong commonality, operational PDS qualification scheme has been proposed. The scheme includes qualification process, activities, tasks and methods. The main qualification activities are suitability evaluation, quality evaluation, operating experience evaluation, and additional test. According to the software requirements of implementing different safety function categories, the corresponding software integrity level qualification tasks are matched to the activities. In addition, according to the characteristics of each task, the recommended execution methods are given, which mainly include document review, traceability analysis, FMEA, code review, code static analysis, code dynamic test and so on. Finally, the application effect of the PDS qualification scheme for safety I&C systems of NPP is introduced, and the concerns in the qualification process are expounded, so as to provide reference for the safety items software qualification of NPPs.
Research on GIL Fault Diagnosis and Location Method Based on Partial Discharge and Vibration Monitoring
Authors: Liu Sen
Published: November 2022

At present, the ultra-high voltage transmission lines in nuclear power plants mostly use Gas Insulated transmission lines (GIL). This type of transmission line transmits large natural power, and GILs are mostly laid through corridors and themselves are metal-enclosed gas -insulated structures, which are affected by the natural environment and have low reliability. high. However, due to the long length of each independent air chamber of the GIL, when it fails, it is accurately difficult to monitor the fault and locate the fault, which reduces the availability of the unit and brings about greater economic losses. In this paper, through the simulation test of the vibration signal propagation when the GIL fails, the signal attenuation law of the vibration signal passing through the insulator is tested, and the fault cell is located through the signal analysis of the vibration sensors in different directions. At the same time, through the simulation test research on the discarge faults of metal spikes, floating potential and metal particles in the GIL, after obtaining the partial discharge signal through the sensor, the corresponding partial discharge eigenvalues are extracted, and the eigenvalues are classified based on the SVM method to realize the GIL fault. Which can provide support for engineering practice.
Study on the Applicability of Flat Ceramic Membrane in Simulated Primary Coolant Water Environment of PWR Nuclear Power Plant
Authors: Guan Yuxin , Chen Jiguo , Zhang Guodong , Ren Xuqiu
Published: November 2022

A flat ceramic membrane has a separating function and is a novel type of material membrane. High temperature resistance, superior mechanical strength, huge permeation flow, and reusability are all advantages. It’s been employed in water treatment industries like food, medicine, and fine chemicals with great success. However, there has been no mention of application research in nuclear power plants so far. This paper evaluates the feasibility of ceramic membranes in a primary coolant water environment in terms of filtration materials and fluid compatibility. The researchers tested the chemical stability of the flat ceramic membranes and observed the variation of dissolved elements concentration with time, mass and temperature. The dissolved elements are mostly Al, Si, Mg, and Ti, according to the results. The leachable Si element is the most abundant, followed by Al and Mg, with a very low eluted Ti percentage. With increasing duration, mass, and temperature, the dissolved elements of the flat ceramic membrane in the simulated primary coolant water environment rise. With time, the rate of dissolved element release diminishes. The dissolving rate tends to a constant value until a particular concentration (about 10 mg/L in this experiment) is reached. The existing flat ceramic membrane is not suited for the primary coolant water environment of PWR nuclear power plants due to the high dissolving rate of each element.