bluuebunny/crossref_metadata_2025 huggingface dataset:
Research of the Standards Compilation of Welding Material for Carbon Steel in Pressurized Water Reactor Nuclear Power Plants
Authors: Lin Bai , Xiaoru Liang
Published: November 2022

Carbon steel welding is a major manufacturing method for nuclear power construction. For the moment, there are two sets of standards of welding material for carbon steel which used in pressurized water reactor nuclear power plants of China. The coexistence of standards of two different technical routes brings unnecessary cost and waste to pressurized water reactor nuclear power plant manufacturing. With the rapid development of pressurized water reactor nuclear power plants in China, the research of developing a set of unified standards for carbon steel electrodes and rods for gas shielded arc welding and carbon steel covered electrodes is necessary. At present, two main standards will be revised and published. Through the interpretation of “Welding material for pressurized water reactor nuclear power plants — Part 1 Carbon steel covered electrodes” and “Welding material for pressurized water reactor nuclear power plants — Part 6 Carbon steel electrodes and rods for gas shielded arc welding” and the introduction to their compilation thoughts and technical points, the article guides engineers to understand the compilation thoughts of these two standards quickly. The standards specify classification, technical requirements, test methods and testing of carbon steel electrodes and rods for carbon steel in pressurized water reactor nuclear power plants.
The Dose Assessment Method Based on the Local Method of Characteristics and Skeletal Animation
Authors: Cao Lun-xiu , Chao Nan , Liu Yong-kuo , Chen Zhi-tao
Published: November 2022

Flexible and accurate estimation of dose absorbed by critical human organs is important to ensure and improve workers’ radiation safety. Absorbed doses for different organs and tissues are affected by their positions and the orientation of the body in the radiation field. This paper presents a skeletal animated local method of characteristic (SALMOC) dose assessment method that considers the working dynamics and organ-level dose for occupational workers during nuclear facility decommissioning. The proposed method combines the verisimilitude of skeletal animation technology (SA) with the flexibility of the local method of characteristics (LMOC). To account for the working posture, the change in different working postures are controlled by skeletal animation, and then the voxel model in a certain pose is generated according to the changed human model. Finally, the LMOC is used to perform the dose assessment. The proposed SALMOC method not only accounts for different working postures during the decommissioning of nuclear facilities but also considers the occlusion and scattering effect of the human model to achieve accurate organ-level dose assessment. The effectiveness of the proposed method is verified by comparing the results with those derived from the Monte-Carlo method and Point-Kernel method evaluated on two different organs. The result shows that the proposed method can generate voxel models for irregular organs, and the tests show better consistency with the Monte-Carlo method compared with the Point-Kernel method.
Application of Probabilistic Safety Assessment in the Risk Analysis and Management of Interface Systems LOCA in a Nuclear Power Plant
Authors: Liu Leilei , Liu Weidong , Wang Zhao
Published: November 2022

Interfacing systems LOCA is a kind of special loss of coolant accident in nuclear power plants, which has an important impact on the nuclear power plants safety. Because the break of this kind of accident is located outside the containment, the interfacing systems LOCA will cause a large amount of radioactivity to be directly released from the core to the atmospheric environment without the detention of the containment, so which is an important contribution to the large early release of nuclear power plants. The Probabilistic safety assessment have been applied to a nuclear power plant in Chinese mainland in this paper. The risk of interfacing systems LOCA in the nuclear power plant is comprehensively evaluated from the occurrence scenarios and frequency of initial events, accident mitigation strategy and event tree modeling, accident consequence and impact analysis. Through the application of probabilistic safety assessment, the key factors that may lead to a core damage and large release of the interface LOCA in the nuclear power plant has been find out, and the weakness existing in the power plant in dealing with such accidents has been discussed at the same time. According to the results of the analysis, how can the nuclear power plant deal with such accidents much more better is discussed, the improve suggestions are given.
Investigation of Swelling Behaviors of U-10Zr Metallic Fuel in the Low Temperature Regime via a Combined Thermo-Mechanical and Rate Theory Analysis
Authors: Qian Zhengyu , Yun Di
Published: November 2022

Uranium zirconium (U-Zr) metallic nuclear fuel has long been considered as a promising candidate for fast reactor fuel in the past few decades duo to its excellent performance. One of the key performance problems of U-10Zr metallic fuel is its strong swelling which would lead to significant fuel cladding mechanical interaction (FCMI) at high burn-ups. Therefore, the mechanism of fuel swelling has long been a concern of great interest in a wide range of this fields. Recently published research conducted by the Institute of Nuclear Physics and Chemistry in China got the experimental results of U-10Zr fuel swelling in a low temperature regime (400–600K) which showed perceivable swelling of the in-pile irradiated fuel. The previous exploration on the swelling mechanism of uranium zirconium metal fuel at low temperature can’t be carried out due to the lack of data and this research filled the gap for the swelling data of uranium zirconium metallic fuel at low temperature. In our work, combined with the input parameters given by finite element analysis, we utilized a rate theory code based on the cavitational void swelling model to catch the fuel swelling behaviors of the U-10Zr fuel in the low temperature regime building upon these data. And a new set of key parameters for the rate theory model were also determined.
Manufacturing Technology of Thick-Walled Integrated Cylinder With Large-Sized Side Nozzles for Nuclear Pressure Vessel
Authors: Li Shaofei
Published: November 2022

This paper introduces a manufacturing technology of thick-walled integrated cylinder with large-sized side nozzles for nuclear reactor pressure vessel. This forging integrates the inlet nozzle and the outlet nozzle on the nozzle shell with integral flange to form an integrated forging. China First Heavy Industry (CFHI) has successfully realized the manufacturing process of this forging. During the forming stage of the forging, with the help of numerical simulation method, the size of the forging blank is reduced to the greatest extent, and a thick-walled cylinder with a convex structure is designed and forged, and then the punching guide hole is machined at the convex structure of the cylinder, which plays the role of positioning and guiding in the forging process of the side nozzle. We use special-structured forging aids to transfer the load of the forging press to the punch to complete the hot extrusion of the side nozzles, and finally realize the integrated near-net shape of the forging. The forging flow line of the forging is preserved to the greatest extent, and the best forging effect of each part of the forging is guaranteed. After the final heat treatment process of the forging, all parts of the forging obtained excellent mechanical properties.
Research and Development of SA508Gr.4N Steel Ultra-Thick Forgings With High Strength and Toughness for Nuclear Engineering
Authors: Du Dongxu , Zhao Deli , Ren Liguo , Zhang Wenhui , Li Jiaju , He Xikou
Published: November 2022

Due to low hardenability, the conventional SA508Gr.3 steel cannot meet the design requirements of increasing wall thickness of forgings for nuclear power pressure equipment. In this paper, the hot processing and manufacturing technology of SA508Gr.4N steel ultra-thick forgings for the new generation of nuclear reactor pressure vessel was introduced. The engineering development of nuclear SA508Gr.4N steel forgings with wall thickness from about 700mm to 1000mm was realized by China First Heavy Industry (CFHI). The largest forging was made of 228 tons ingot and the final maximum dimensions reached about ϕ4000mm × 1000mm. The samples were taken anatomically from the forging at various thickness and radial positions, and subjected to the tensile, Charpy impact and Drop-weight tests. As a result, the full-section property of the forging met the requirements of the ASME code. Compared with nuclear SA508Gr.3 steel forgings, the strength and toughness of SA508Gr.4N forgings were greatly improved. Meanwhile, based on the fact that the actual property value is much higher than the requirements of ASME code, higher design property requirements of SA508Gr.4N steel forging was proposed. In addition, the nuclear neutron-irradiation-embrittlement data of SA508Gr.4N steel, which were cut from the actual engineering forgings and experimented in a high flux engineering test reactor (HFETR), was obtained. The main research and development goals for engineering application of SA508Gr.4N steel in the future were also discussed.
Finite Element Analysis and Prediction of the Cross-Section Morphology of Grid Strips
Authors: Mao Jianzhong , Cheng Zhi , Su Jun , Wang Ben , Zhang Pengfei , Mei Zezhi , Yuan Jiajian , Gao Bo
Published: November 2022

To investigate the relationships between the quality of the section morphology of the zirconium alloy strips and the blanking parameters as well as the design of the blanking mould, 64 processing parameter combinations are generated by using Latin Hypercube Sampling and the finite element models with different boundary conditions have been constructed to acquire the corresponding section qualities. The calculation results of the finite element model are verified by the experiment data. Furthermore, the random forest algorithm is employed to build the prediction model between the processing parameters and the lengths of the section bright strip. The sensitive analysis of the processing parameters influenced on the section quality have been conducted through the prediction model. The results indicate that the random forest prediction model have a satisfied accuracy, which could effectively forecast the length of the bright band in the section, and then provides guidance on mold design as well as process parameter selections. The blanking clearance included in the design factor has a decisive influence on the length of the bright strip; The friction coefficient has the weakest effect on the results. The effect of die fillet radius and blanking speed on the length of bright band is not significant.
Measurement of the Two-Phase Flow Void Fraction Downstream Of Spacer Grids in Tight Lattice Bundles Using Wire-Mesh Sensor
Authors: Yan Xu , Xiao Yao , Zhang Hengwei
Published: November 2022

The tight lattice fuel assemblies (P/D < 1.1) have huge application potential in the new generation of the nuclear reactor systems, which possesses the advantage of higher power density and higher fuel conversion ratios, etc. Spacer grids can fix the fuel rod position and enhance the heat transfer performance. However, the two-phase flow characteristics downstream of spacer grids in the tight lattice are still not clarified. In this paper, a gas-liquid two-phase flow experiment study measuring the void fraction distribution downstream of spacer grids without mixing vanes (SGWMVs) in a double subchannels tight lattice bundle was conducted using a double-layer wire-mesh sensor. The experiment channel and the SGWMVs were both up-scaled (1:2.7) considering the difference between the experimental flow conditions and the prototype flow conditions. The phase distributions for a series of flow conditions (0.07 m/s < jg < 1.04 m/s, 0.93 m/s < jl < 1.86 m/s) were measured at Z/Dh = 115.81 of the bare rod channel as the flow data upstream of the spacer grid. Whereafter, the SGWMVs was installed at the position of Z/Dh = 145.92. The phase distributions 45mm downstream of SGWMVs were measured successively for the same flow conditions. Three types of flow patterns were obtained, including the bubbly flow, the cap-bubbly flow, and the slug flow. And the effects of SGWMVs on different flow patterns were described and analyzed. The newly obtained data established the reliability database contributing to the development of the computational fluid dynamics codes and the interfacial area transport equation.
A New Correlation for Predictions of Heat Transfer of Lead-Bismuth Eutectic in Circular Tubes
Authors: Wang Zhengquan , Wen Qinglong , Xu Shijia
Published: November 2022

Liquid lead or lead-bismuth eutectic has good thermal physical properties, neutron performances, and natural circulation capacity. It is the first choice of coolant material for the lead-based fast reactor (LFR). However, there is a fundamental problem, heat-transfer characteristics and quantitative relationship of liquid lead and lead-bismuth eutectic, to be solved in the process of thermal-hydraulic design and safety analysis code development of lead-based fast reactor. The existing correlations are mostly fitted by the experimental data under specific conditions and the applicability is poor in a wide range of parameters, which affects the accuracy of thermal-hydraulic design and safety analysis results. Therefore, it is a necessity to establish experimental databases and correlation databases to obtain a more accurate heat transfer correlation of liquid lead and bismuth eutectics within a wider range of parameters.

This paper reviews the typical theoretical and experimental studies on the heat transfer characteristics of liquid lead and lead-bismuth eutectics in the recent 70 years, and an experimental database based on the heat transfer experimental data in the published literature is established, which data points include a wide range of thermodynamic parameters — inner diameter ranging from 9.00 to 16.56 mm, temperature from 250.0 to 412.80 °C, heat fluxes from 20.0 to 12900.0 kW/m2, and mass fluxes from 0 to 12.0 kg/m2s. Based on the database, the typical heat transfer correlations of liquid lead and lead-bismuth eutectics are evaluated and analyzed, and a new heat transfer correlation is developed, its suitable range is 100 < Pe < 5800. The results show that the mean absolute relative error of the new correlation is 16.64%, the average relative error is 0.74%, the standard deviation is 4.37%, and the error of 65.27% of the total data points is within ± 20%. Therefore, the heat-transfer correlation proposed in this study can well predict the heat-transfer coefficient of high-temperature liquid lead or lead-bismuth eutectic in the tube.
Thermal-Hydraulic Safety Analysis of Natural Circulation Lead-Cooled Fast Reactor SNCLFR-100 Core Based on Porous Medium Approach
Authors: Feng Wenpei , Yang Guangliang , Zhang Kefan , Qin Chong , Luo Xiao , Ding Tao , Chen Hongli
Published: November 2022

In this study, a thermal-hydraulic analysis is carried out for the core of natural circulation lead-cooled fast reactor SNCLFR-100. Steady-state and transient analysis are performed with porous medium approach-based code TWOPORFLOW. In the steady-state analysis, mass flow distribution and temperature distributions of the assemblies are analyzed in assembly-wise mode. At the same time, the hottest assembly is analyzed in pin level and the safety performance is investigated. In the transient analysis, a typical Design Extension Conditions unprotected over-power transient is simulated with one-way coupling method.
Investigation of Multifarious Zircaloy-Water Oxidation Models on Core Degadation in a Typical SBO Accident
Authors: Sun Hongping , Luo Yuejian , Deng Jian , Zhang Ming , Wu Xiaoli , Liu Lili , Xu Youyou
Published: November 2022

Cladding oxidation is a particularly important process which can affect the process of core melting. Synchronously, hydrogen released from zircaloy-water reaction is a threat to the integrity of containment. Up to now, multifarious zircaloy-water oxidation models are applied in different severe accident analysis programs. In order to investigate the influence of different cladding oxidation models on the core degradation process, a station blackout (SBO) accident model of CPR1000 nuclear power plant (NPP) was established and simulated by modifying the cladding oxidation model in this work. Five oxidation relations were chosen for evaluating thermal hydraulic, metal oxidation and core degradation analysis. Thermal hydraulic characteristics of five types remain consistent before core oxidation. For more concerned hydrogen production, the results show that the ratio of the maximum hydrogen production to the minimum is approximately 2.5 times in comparison of the different oxidation models. Four core nodes temperature distributions at different time points are selected to analyze the core node degradation process, which can help readers understand the effect of different models on core degradation more intuitively. The oxidation heat generated is consistent with that of hydrogen production, indicating that the core degradation time in the case of less hydrogen produced is delayed than that of more hydrogen produced. It follows that different oxidation models have great influence on the core oxidation behavior, which is of great significance to understand different oxidation models of zircaloy and the core degradation behavior of severe accident. It is critical to choose the appropriate model to simulate the whole severe accident.
Analysis of Water Consumption in Nuclear Power Plants and National Standard Formulation for Norm of Water Intake
Authors: Wang Rui , Xu Kechen , Zeng Kehao , Yan Hongmei , Bai Wei
Published: November 2022

The nuclear power industry consumes a large amount of water. Its water process systems are complex, and require high water guarantee rate. In general, water process and water consumption indexes for different nuclear power plants are still in the exploratory stage. For a long time, there are huge differences between the design values and actual values of water intake, water consumption, and water withdrawal in China nuclear power industry, which brings difficulties to the management. At present, the construction of nuclear power plants in China is in a period of rapid development, but there is no national standard for norm of water intake, yet. Nearly all the design water consumption indexes of nuclear power plants are based on the design values of nuclear power plants that have been built and part water consumption indexes required by few relevant design specifications. With the implementation of the strictest water resources management system, total volume control and norm management have become important tools of water resources management. The national standard formulation for norm of water intake in China’s nuclear power industry can provide a basis for the approval of water intake permits for design and construction of the nuclear power plants, to strengthen the management of water resource and improve the water utilization efficiency of the nuclear power plants. This paper introduces the current situation of water consumption in each nuclear power plant and the basis for national standard formulating for norm of water intake.
Development and Regulatory Reform on Radiation Safety Training for Nuclear Technology Utilization in China
Authors: Zhang Yang , Chen Fangqiang , Chen Baolong , Jia Wei , Li Zhaotong
Published: November 2022

Radiation safety training for the use of nuclear technology is an essential requirement of the regulations, and an important means to ensure the health of radiation practitioners and the safety of radiation environment. After more than ten years of development, the scale of the previous off-line training model basically met the needs of the industry, but some drawbacks also gradually appeared, such as lack of regulatory basis and supervision, high pass rate and poor training effect, increased corporate burden, etc. Since 2020, the National Nuclear Security Administration (NNSA) has officially implemented the reformation of radiation safety training. A unified radiation safety training and assessment platform has been established, covering the training and assessment needs of 11 nuclear technology utilization disciplines, and the training and assessment data are connected with the regulatory platform. The training assessment has changed from “accredited training institutions and paid off-line training and examination” to “free online self-learning and on-site online examination”.

This paper introduced the basic situation of the radiation safety training surveys, carried out multi-dimensional data analyses on the reformation of radiation safety training and assessment, discussed the existing issues in rules and the division of responsibilities between the companies and agencies. Some suggestions, such as adjusting the requirements of rules and regulations, increasing the coverage of nearby assessments, and adding the practical operation assessment, were proposed.
Specifics of Calculating Thermophysical Properties of CO2 and R134a in Critical Point Using NIST REFPROP
Authors: Xie Hanqing , Pioro Igor
Published: November 2022

Thermophysical properties of various fluids (liquids, vapours, and gases) are the fundamental knowledge for the application of these fluids. A computer program can be considered as a very useful tool if it is able to calculate various thermophysical properties of various fluids within a wide range of pressures and temperatures from lower ones and up to critical and supercritical ones. NIST REFPROP is such a program. For tens of years, it can be possibly considered as the best one in the world. However, it is not absolutely perfect. In the previous versions(s) of the NIST REFPROP 9.1 and lower, three basic thermophysical properties, specific heat, thermal conductivity, and volumetric expansivity, were subjected to very significant variations within the critical region. They had almost infinite peak values in the critical point, which was a theoretical approach.

In 2018, the latest version was released, Ver. 10.0 (https://www.nist.gov/srd/refprop). It was updated with new fluids, a wider range of pressures and temperatures, and with improved equations / correlations. The objective of this paper is to check if the deficiencies occurred in the previous version(s) have been fixed in NIST REFPROP Ver. 10.0. In this paper, various thermophysical properties of CO2 and R134a have been calculated by using NIST REFPROP Ver 10.0. The results around the critical point have been analyzed.
Inadvertent Reactor Pit Flooding Risk Research on the Generation III Nuclear Power Plant
Authors: Xiao Lingmei , Zhang Bing , Wang Ming , Wang Zhiwen
Published: November 2022

In-vessel Melt Retention (IVR) system is a typical design feature of the the generation III nuclear power plant. In case of severe accident, enough water is injected into the reactor pit to cool the outside of the Reactor Pressure Vessel (RPV) to maintain the integrity of the reactor pressure vessel, so that the core melt is retained in RPV. The design concept of IVR is very meaningful for maintaining the integrity of RPV and reducing the risk of large radioactivity release, but on the other hand, it may also introduce the possibility of inadvertent reactor pit flooding during normal operation, which may threaten the integrity of RPV. At home and abroad, the research on IVR mainly focuses on whether its safety function can be realized, but less attention is paid to its potential negative impact. This paper takes the China generation III reactor with the IVR design feature for the study, carry out the possible inadvertent reactor pit flooding path analysis and frequency analysis. Based on the risk study, put forward the concerns to prevent inadvertent reactor pit flooding, which will provide guide for the subsequent IVR design and optimization.
Analysis of Behavior of Prestressed Containment Under Close-in Explosion Based on CEL Method
Authors: Li Rongpeng , Yao Qinqin , Liu Mengsha , Jiang Di
Published: November 2022

The prestressed containment structure is the last physical barrier against the spread of radioactive material. In addition to being able to resist the pressure caused by internal accidents, it should also have a certain ability to resist external projectiles or explosive shocks. Under the background of frequent terrorist incidents and intensified local conflicts, evaluating the performance of the containment structure under the close-in explosion of explosives carried by drones or light weapons is of great significance for maintaining the safety of nuclear power plants. In this paper, the CEL (Coupled Eulerian and Lagrangian) method is used to calculate the overpressure of the air blast, which is compared with the empirical formula. The test of the damage to the concrete slab under TNT near-in explosion reported in the literature is simulated, and the test results are compared. A numerical model of containment was established and calculated to predict damage behavior. The results show that the air overpressure simulated by the CEL method is in good agreement with the empirical formula, and the simulated results are consistent with the experimental results reported in the literature. Finally, predictions are made for the performance of the containment model under the close-in explosion. This method can be used to design or evaluate the performance of containment against close-in explosions.
Measurement of Two-Phase Flow: Static CT System Based on Carbon Nanotubes
Authors: Zhang Yucheng , Xu Shuo , Xiang Xincheng
Published: November 2022

Two-phase flow measurement with high precision plays a significant role in safe and efficient operation of nuclear reactor. This paper focuses on the void fraction of gas-liquid flow in a non-transparent tube with the inner diameter of 13mm. X-ray computer tomography (CT) is considered to be effective for two-phase flow measurement because of its good qualities of non-invasion. However, due to the mechanical limitation of rotating speed, solutions for fast CT system to reduce the scanning time have been suggested relied on multiple sources and detectors. Recently, carbon nanotube (CNT) based X-ray source has been applied to CT imaging, significantly improves the temporal resolution by increasing the number of sources and avoid the gantry rotation. This paper proposes a potential static CT system design for the imaging of two-phase flow in straight steel tube. The setup of this system employed 90 couples of CNT X-ray sources and detector arrays arranged in a circle. Gas-liquid flow was simulated with different sizes of spheroidic bubbles randomly placed in the water inside the tube. To substitute for the flow moving, the z-axis of the simulator was added according to the flow velocity and exposure duration. Iterative image reconstruction was applied for the inverse problem of density distribution, and the reconstruction result of the experiment indicates the static CT system is useful to distinguish the flow pattern and measure the void faction distribution of gas-liquid flow.
Research on Precision Water Reactor Fuel Servo Operation System
Authors: Ping Li , Deng Peiqiang , Qian Wei , Yang Kaiyu , Zhai Jiansong , Huang Shuiyang
Published: November 2022

The PMC system of PWR nuclear power plant is a servo control system based on PMSM permanent magnet synchronous motor. The traditional Clarke and park coordinate transformation theory algorithm is used for precise control. This paper analyzes the composition of the servo system of PMC loading and offloading equipment, the principle and control algorithm of PMSM permanent magnet synchronous motor, and studies the double closed-loop control model based on current sampling and position sampling, the protection function setting of servo system and the self-learning method of PMSM. The research results show that the PMC system adopts the double closed-loop control theory of field oriented control (FOC) with DSP as the control core. Through the close-loop cooperation of PLC, servo driver, PMSM permanent magnet synchronous motor and resolver, the precise control of speed loop and current loop is realized, and the three closed-loop is realized by the position loop of PLC Control. This study provides a reference for the realization of independent deep maintenance of PMC equipment, and provides theoretical support for the future transformation of PMC servo control system and fault prediction of servo system.
Optimal Design of Thermal Scheme for Liquid Metal Cooled High Flux Reactor Fuel Assembly
Authors: Pan Rui , Zhang Kefan , Zhang Xilin , Deng Jian , Zhang Yong , Zhu Dahuan , Chen Hongli
Published: November 2022

This paper studies the structural parameters of fuel assemblies suitable for high neutron flux density environments.

Due to the high neutron flux density in the core, high-flux reactors provide an experimental environment for neutron irradiation. However, high neutron flux leads to high heat flux density on the fuel assembly surface, which brings challenges to the design of fuel assembly. Therefore, it is very important to study the structural design of fuel assemblies suitable for high neutron flux density environments.

Through theoretical derivation of the thermal model of the fuel assembly and sensitivity analysis of the design parameters of the fuel assembly using a single-channel program, the results show that the liquid metal cooling plate fuel element can be well adapted to the high neutron flux Density environment; at relatively low neutron flux densities, bundle fuel elements can also meet reactor design requirements.
Comparison Of Radiation Embrittlement Prediction Models for Reactor Pressure Vessel
Authors: Cai Kexin , Huang Ping , Qian Wangjie , Fan Minyu , Qi Shuang , Zhang Guodong , Zhang Yanwei
Published: November 2022

In order to promote the establishment of the material prediction model of domestic reactor pressure vessel (RPV) under high neutron flux irradiation, the prediction model of reference toughness and brittle-transition temperature variation (ΔTR,NDT) in foreign relevant guidelines and standards were compared and analyzed from the aspects of data basis, scope of applicable neutron flux and reliability of results. The results show that all prediction models are based on a large number of experimental data, and their reliability varies greatly. RG1.99 (Rev.2) has the worst reliability, while FIS has the highest reliability. When the neutron flux is less than or equal to 7.2 × 1019 cm−2 (E ≥ 1.0 MeV), the results of JEAC 4201 model are more conservative than NUREG/CR-6551, but when the neutron flux is larger than 7.2 × 1019 cm−2 (E ≥ 1.0 MeV), the results are opposite. The range of neutron flux applicable to each model is different, among which RG1.99 (Rev.2) and JEAC-4201 prediction models are the highest, reaching 10.0 × 1019cm−2 (neutron energy E ≥ 1.0 MeV). The establishment of the prediction model of domestic RPV materials under the condition of high neutron flux irradiation requires a large number of irradiation tests and in-depth research on the irradiation embrittlement mechanism, which lays the data and theoretical foundation for the establishment of the prediction model of domestic RPV materials under the condition of high neutron flux irradiation.
Study on Flow and Heat Transfer Performance of Space Reactor Radiator Water Heat Pipe
Authors: Jiao Guanghui , Xia Genglei , Tao Zhou , Bi Yuepeng
Published: November 2022

To ensure that the heat of the space reactor can be exported in time, the parameters and operating characteristics of the radiator heat pipe must be clear, and the heat pipe must work under various heat transfer limits and meet the safety requirements of avoiding single-point failure. Based on the basic principle of the heat pipe and fully considering the phase change characteristics of the gas-liquid interface in the heat pipe, this paper establishes the numerical calculation model of the operation characteristics of the space reactor radiator water heat pipe. The influence of phase transition on flow and heat transfer is considered. Through the calculation of the heat pipe, the operation parameters such as temperature, pressure and speed of heat pipe are obtained. On this basis, according to the requirements of safety characteristics analysis in the heat pipe, four heat transfer limit calculation methods of heat pipe viscosity limit, sonic limit, entrainment limit, and capillary limit are established. The heat transfer limits of the heat pipe of the space reactor radiator are calculated by using the constructed model, and the main heat transfer limits affecting the heat transfer performance of the heat pipe are determined. By coupling the heat pipe calculation program with the heat sink calculation model, the temperature distribution of the radiator panel is obtained, and the influence of the length of the radiator panel on the heat transfer characteristics is analyzed.
Research on the Reactor and Turbine Matching Key Technology of CPR1000 Nuclear Power Plant for Industrial Steam Supply
Authors: Mingjia Zhou , Kai Wang , Wei Zhang , Jianmin Zhu , Yulong Mao
Published: November 2022

The main steam extracted from the secondary side of nuclear power plant for steam supply in industry will affect the power matching between reactor and steam turbine, and the adaptability of reactor control system is the key of reactor-turbine matching research. By analyzing the influence of main steam extraction on reactor control system in CPR1000 nuclear power plant, a new reactor-turbine matching signal is designed and verified by modeling and maximum extraction steam flow rate is discussed. This study can provide important reference for the development of other nuclear energy comprehensive utilization.
Simulation of One-Dimensional Nuclear-Pumped Laser Generation System Based on Modelica Language
Authors: Tang Xiang , Wang Xiang
Published: November 2022

Nuclear-pumped lasers (NPL) utilize nuclear energy pumping to generate laser light and represent a viable approach to directly converting nuclear energy into useful products (coherent photons). Most nuclear-pumped lasers use thermal neutrons provided by pulsed nuclear reactors to initiate nuclear reactions. Nuclear-pumped lasers that use nuclear reactors as thermal neutron sources are also known as reactor-pumped lasers (RPL). Gas lasers mainly use inert gas as the laser medium, which has the characteristics of high output energy and long service life. This paper mainly establishes a one-dimensional He-Xe laser system, which can excite a laser with a wavelength of 2.03 μm, and the efficiency can reach 1% to 4%. According to the different physical processes of the laser generation system, we divided the system into three main components: nuclear reaction component, pump reaction component, and laser generation component. Based on the modeling method of Modelica for physical process componentization, it is used to simulate and verify each component, and the relevant parameters of the component are determined through theoretical calculation and numerical simulation. The componentized modeling method increases the universality and ease of editing of the model, and facilitates the maintenance and later expansion of the model. When the neutron fluence rate is 6.0 × 1014 cm−2·s−1, the laser with a wavelength of 2.03 μm and a power of 38.55 W can be output.
Structural Design and Dynamic Analysis of the Heat Exchanger Used in Nuclear Fuel Test Loop
Authors: Gang Huang , Sheng Sun , Liangming Pan , Junping Si
Published: November 2022

The heat exchanger was required to cool the high temperature and high pressure primary water from the test chamber of the nuclear fuel test loop generated by test fuel assembly. The heat exchanger was composed of shell-tube heat exchanger group and double-pipe heat exchanger group, the regenerative structure was adopted to greatly reduce the thermal stress caused by the large temperature difference between primary water and secondary water, and the long-term safe and stable operation of the device was ensured, the establishment of 3D analysis model and dynamic analysis of the heat exchanger were carried out in the paper, stress strength of the heat exchanger by different loads under design condition, normal condition, abnormal condition, accident condition and test condition were obtained, stress strength assessment for the heat exchanger structure based on RCC-M specification design and construction rules for mechanical components of PWR nuclear island was also carried out, the results indicate that the strength of the heat exchanger satisfied requirements of every RCC-M level.
Neutronic and Thermal-Hydraulic Analyses for an Additive Manufacturing Reactor Core With SiC Matrix and TRISO Particle Fuel
Authors: Han Wenbin , Deng Jian , Lu Qi , Chen Chong , Xu Youyou , Zhang Tao , Huang Shanfang
Published: November 2022

Additive manufacturing (AM) is a transformational digital manufacturing technology featured with rapidity, customizability, precision, and economy, which is fundamentally altering the way components are designed and manufactured. AM enables the freedom of design, and makes full use of complexity of geometry which “comes for free”. Applying AM technology to nuclear industry can yield advanced reactor designs with function and structure matched for the best thermal, hydraulic and mechanical performance. In this work, an AM-informed reactor core design with silicon carbide (SiC) matrix and tri-structural isotropic (TRISO) particle fuel is proposed and analyzed. The core is an integrated 3D-printing SiC bulk with helical cruciform coolant channels, and the UO2-bearing TRISO fuel particles are dispersed in the bulk. A multi-physics analysis framework for irregular geometry is developed to analyze and further optimize the reactor design. The TRISO particle positions are generated with discrete element method (DEM). The Reactor Monte Carlo code (RMC) and the commercial CFD software STAR-CCM+ are used for the neutronic and thermal-hydraulic analyses, respectively. RMC simulates the neutron transport to predict the effective multiplication factor and power distribution. STAR-CCM+ calculates the flow and heat transfer in coolant channels and heat conduction in solid matrix with the power distribution as the heat source. Preliminary results show that the power peaking factor FQ decreases below 1.65, the heat transfer area increases by 30.3% and the fuel peaking temperature decreases by 25 K. The optimized AM-informed design enjoys better neutronic and thermal-hydraulic performance than those with regular geometry.
Selection and Economic Benefit Analysis of Current Limiting Circuit Breaker in a Project of a Newly-Built Large Nuclear Chemical Plant
Authors: Gao Shuo , Tian Shizhong , Huang Zhi
Published: November 2022

With the continuous development of science and technology and the iteration of various types of low-voltage protection products, electrical designers are focusing not only on the reliability and security of the power supply system but also its economic benefits. From the calculation of electricity load in the early stage, to the selection of transformers, and then to the determination of the number of low-voltage distribution cabinets, as well as the selection of each feeder circuit protection devices, designers strive to come up with the most reliable and economical choice in each link through rigorous calculation and analysis. In the design and installation of large communities or factories, a small technical improvement can bring substantial economic benefits. Combined with engineering practice, this paper briefly introduces several protective appliances and their functions in low-voltage distribution circuits, the main differences between current limiting circuit breakers and ordinary circuit breakers are also revealed. The selection principle and basis of current limiting circuit breaker in the electrical design of a nuclear chemical engineering project are described in detail. The economic benefits and significance of selecting current limiting circuit breaker are compared and analyzed in this paper. The development prospects of current limiting circuit breaker selection in power supply and distribution system in large nuclear chemical plants and its key role for the whole plant in the future are discussed.
Implementation and Analysis of Thermal Hydraulics Feedback in High-Fidelity Neutron Transport Program HNET
Authors: Zhu Yanling , Hao Chen , Li Peijun , Zhou Xiaoyu
Published: November 2022

With the growing requirement of predicting reactor behavior in high-fidelity detail at practical conditions, it is urgent to accomplish thermal hydraulics (T-H) feedback in the high-fidelity neutron transport program HNET. For better convergence behaviors than Picard iteration, the Matrix Free Newton/Krylov (MFNK) method was employed to resolve neutronics and thermal-hydraulics coupling system. MFNK treats each subsystem as a black box within the Newton method framework, so it can facilitate the coupling procedure without surrendering efficiency or robustness. For the T-H feedback effects, a simplified internal thermal hydraulics model was adopted to provide T-H conditions for neutronics. The convergence behaviors of MFNK and Picard iteration were assessed through simple typical cases. Finally, the effectiveness of the coupling system was verified by the VERA problem #6. Numerical results demonstrate the efficiency and stability of MFNK compared with Picard iteration. Moreover, it turns out that the coupling system has a good performance in realistic cases.
Research and Treatment of Pressure Regulation Defects in Fuel Circuit of Emergency Diesel Engine of One Nuclear Power Plant
Authors: Feng Guo
Published: November 2022

For one nuclear power plant, the pressure of the diesel engine fuel oil circuit happened to fail to fit the design because of low fuel oil inlet pressure and fall-off of the O-ring of the pressure regulating valve. A lot of data has been collected from tests performed in different conditions during the commissioning phase of this type of diesel, and after a lot of work the wrong diameter of the orifice plate and the improper inner structure of the pressure regulating valve in the fuel oil returning circuit were proved to be the main reasons. This project tried to solve the problems in a systematic way. Single-variable method was used to find out that the inner hole diameter of the orifice plate was too large to maintain the fuel oil pressure at a high level even if the pressure regulating valve was blocked and the right diameter was found and tested on site. The inner part (valve disc) of the pressure regulating valve was found fail to keep the O-ring under the alterable pressure, then fail to regulate the fuel oil pressure. The structure of the valve disc was optimized to fit the alterable pressure working situation. Finally, the defect in fuel circuit of emergency diesel engine was solved by adjusting the bore diameter of the orifice plate and changing the structure of the pressure regulating valve disc. The relevant tests were successfully done after the modification implementation.
Shutdown Seal Improvement Feasibility Analysis for CPR1000 Reactor Coolant Pump
Authors: Hengrui Zhang , Qian Huang , Peiren Fu , Daoke Zhang , Rui Cai , Lei Sun , Qiuyang Chen , Feng Xue
Published: November 2022

After the Fukushima accident, there were two important feedbacks. One was the simultaneous occurrence of serious accidents of multiple units at the same plant site, and the other was the emergency diesel generator was unavailability under a serious accident. The events at Fukushima highlight the possibility that extreme natural phenomena could challenge the prevention, mitigation and emergency preparedness defense-in-depth layers. In the design of CPR1000 Nuclear power unit, the important contributor is the integrity of the reactor coolant system considering potential leaks from the reactor coolant pump seals following a station blackout (SBO) or loss of heat sink accident. After the SBO or loss of heat sink of whole plant occurs, normal main pump shaft seal injection will be lost. The hydrostatic test pump will start automatically to ensure emergency shaft seal injection. But the flow of hydrostatic test pump can only meet the requirements of one unit shaft seal injection, and the other unit shaft seal injection cannot be guaranteed, which may lead to loss of coolant accident (LOCA). If the LOCA is not controlled, it will lead to reactor core damage. At present, the improvement of shutdown seal used in the main pump of the new power plant can effectively solve this problem and reduce the frequency of reactor core damage. This paper mainly analyzes the technical characteristics of shutdown seal and the structure and function of main pump, so as to determine the feasibility and necessity of the improvement of shutdown seal of CPR1000 main pump.
State Monitoring of Electric Gate Valve of Nuclear Power Plant Based on Kernel Principal Component Analysis and Jensen-Shannon Divergence
Authors: Wang Xiaokun , Wang Hang , Deng Qiang , Xu Renyi
Published: November 2022

With the rapid development of the global nuclear power industry, the safety of nuclear power equipment has received increasing attention. In the event of an accident in its equipment, it is likely to cause significant losses and harm to personnel and the environment. Therefore, real-time monitoring of nuclear power plants has very important practical significance. Based on Kernel Principal Component Analysis (KPCA) and Jensen-Shannon (JS) divergence, equipment-level state monitoring of electric gate valves of nuclear power plants will be conducted. The traditional Principal Component Analysis (PCA) method, which can only reduce the dimensionality of nonlinear data, is improved by KPCA. The traditional statistics T2 and Q, which have the problem of low recognition rate of small faults and poor timeliness, can be greatly improved by the JS divergence. The main model process is to first perform KPCA dimensionality reduction on the data, calculate the JS divergence value of the data after dimensionality reduction, and compare the change trend of the JS divergence value to reflect the operating status of the electric gate valve, and perform real-time status monitoring. The experimental results show that the state detection method combining KPCA data dimensionality reduction and JS divergence has improved the accuracy of monitoring compared with other methods, and has better timeliness.
Preparation and Properties of Ceramic Solidified Product Containing Cs and Sr
Authors: Sang Hongji , Gu Zhengcheng , Cui Zheng , Zou Ruoxue , Wu Yan
Published: November 2022

137Cs and 90Sr in high level liquid waste, with high radioactivity, large heat-generating and relatively long half-life. In order to solve the problem of final disposal of 137Cs and 90Sr, natural mineral allophane was chosen as the base materials to synthesize the silicate ceramic solidified products with cold pressing/sintering method. The microstructure, phase composition and surface element distribution of the solidified products were analyzed. The solidification mechanism of the solidified products was also discussed. The surface of the solidified products appeared melting phenomenon after sintering, and the structure was more compact. CsAlSiO4, Sr2Al2SiO7 and SrAl2Si2O8, which can stably solidify Cs and Sr, were formed in the solidified products. The content of allophane in the solidified product has an important influence on the immobilization ratio of Cs and Sr. And at the sintering condition of 1 h duration time at 1200 °C, the immobilization ratios of Cs and Sr can reach 100 %. By increasing the content of cured substrate, the surface characteristics and crystallization properties of the solidified product can be improved, and the volume reduction effect was more obvious.