bluuebunny/crossref_metadata_2025 huggingface dataset:
Feasibility Verification of One Operator Monitoring Two NSSS Modules in Power Plant With Six High Temperature Gas Cooled Reactors Under Accident Conditions
Authors: Miao Runfa , Jia Qianqian , Li Duo
Published: November 2022

The multi-module high temperature gas cooled reactor (MMHTGR) adopts the design of multiple reactor modules coupled to one steam turbine. The number of modules can be as much as six. According to staffing model of the preliminary stage, each reactor operator needs to monitor and control two modules, and each NSSS (nuclear steam supply system) module has the similar displays and operating procedures. The operation concept of the multi-module plant is different from single-module HTGR. Therefore, the human performance of operators in the multi-module HTGR need to be studied.

In order to verify the feasibility of one operator monitoring and controlling two NSSS modules under accident conditions, experiments are designed. During the experiment, a variety of accident scenarios are selected, including accidents in two or more modules. During the experiment, the operator will record the completion of the operator’s tasks. At the end of each experimental scenario, each operator is invited to fill in SART scale and NASA-TLX scale for measuring situation awareness and workload. The experimental results show that because the inherent safety of the HTGRs and a high degree of automation is conducive to the operation of the reactor, it is feasible for one operator to control and monitor the two modules under the accident condition. The correlation analysis shows that there is a strong negative correlation between the operator’s situation awareness and Demand use the SART scale. The workload of operators has a strong positive correlation between Mental and Temporal in the NASA-TLX scale. By improving the components of Demand, Temporal and Mental, operators’ workload can be reduced and situation awareness can be improved.
Study on Radionuclide Migration Under Accident Conditions of Mobile Heat Pipe Small Module Reactor
Authors: Tang Jianyu , Zhou Tao , Xu Peng , Lu Huaichang , Gao Tianyu
Published: November 2022

Mobile heat pipe reactor has a wide application prospect in many fields. In the event of a nuclear leak, the reactor will have adverse effects on the environment. So, mastering the circumstances of nuclide migration has great value to enhance the capacity and efficiency of nuclear accident emergency response. This research uses a programming method to describe the diffusion and deposition of nuclides in the atmosphere after a leakage accident. Subsequently, further study the deposition of nuclide diffusion in the unsaturated zone. Through these studies, we can make predictions and evaluate the consequences of accidents. The result shows that there is a certain relationship between the concentration of nuclide diffusion in the atmosphere and the distance downwind direction and there is a maximum value. The maximum diffusion distance of a nuclide is positively correlated with the amount of leakage nuclide when it diffuses over a short distance and near the ground. The deposition conditions of nuclides are different under different conditions. Dry deposition flux can be neglected in short-term calculations. Wet deposition flux is the main source of the diffusion of nuclides into the unsaturated zone. Soil has a good retention effect on radionuclide diffusion and the unsaturated zone can slow the migration velocity. However, rainfall will accelerate the migration of nuclides.
Evaluation of the Parameter Importance Changes Over Time for Nuclear Reactor Transients
Authors: Xiong Qingwen , Song Gongle , Du Peng , Huang Tao , Deng Jian
Published: November 2022

The uncertainty analysis and sensitivity analysis are two important parts in the safety analysis of nuclear reactor accidents, and the main purpose of the sensitivity analysis is to quantify the importance of the input on the target output. In typical sensitivity analysis for the nuclear reactor transients, specific techniques such as the regression analysis or correlation coefficient analysis are utilized to calculate the sensitivity measures of the inputs on the target output at scalar points, for instance the peak cladding temperature (PCT). However, the phenomena and parameters vary significantly over time during the transients, thus the influence of each input also changes. In order to evaluate the importance changes of the inputs on the target output over time, a time-dependent global sensitivity analysis technique was developed base on the B-spline function system of the functional data analysis, the landmark registration strategy, and an efficient moment-independent sensitivity analysis method. The developed technique was adopted in the sensitivity analysis of a large break loss of coolant accident (LBLOCA) of the nuclear reactor, and the time-dependent sensitivity measures of the inputs on the maximum cladding temperature were acquired and analyzed. Results showed that the proposed technique is capable of evaluating the importance changes of the key parameters in nuclear reactor accident transients.
Creep Properties of FeCrAl Alloy at High Temperature Under Neutron Irradiation
Authors: Yao Huan , Ye Tianzhou , Wu Junmei , Wu Yingwei , Yin Chunyu , Chen Ping
Published: November 2022

Nuclear fuel cladding is subjected to neutron irradiation in a high-temperature stress environment, and the structural integrity of the cladding is very important for the safe operation of nuclear reactors. FeCrAl alloy has become a promising candidate cladding material for the accident tolerance fuel development in view of its excellent irradiation resistance and high temperature strength. This work aims to study the creep properties of FeCrAl alloy at high temperatures under neutron irradiation. Thermal and irradiation creep behavior in nanocrystalline FeCrAl samples is examined using molecular dynamics simulation method. And the effects of temperature, stress, irradiation dose rate on the creep rate and parameters of the creep constitutive equations are discussed. The results show that the thermal creep rate is greater than irradiation creep rate. The effect of temperature on the thermal creep stress exponent is relatively small at low stress, but is obvious when stress exceeds 0.8 GPa. The higher the temperature, the larger the thermal creep stress exponent. The irradiation creep rate increases almost linearly with the dose rate, that is, the exponent of dose rate for irradiation creep approach 1.0. Irradiation creep stress exponent fluctuates very little around 1.1 within the scope of the present research. Besides, higher temperature accelerates the linear increase of irradiation creep rate with dose rate, and the irradiation creep pre-factor becomes higher.
Research on Scenarios and Development Paths of China’s Commercial Closed Nuclear Fuel Cycle
Authors: Xiao Min , Xiao Xiaojun
Published: November 2022

China implements the established policy of closed nuclear fuel cycle for the sustainable development of nuclear power. However, there seems no feasible development plan and roadmap to initiate and deploy a commercial closed fuel cycle in China up to now.

The industrialization of the nuclear fuel cycle requires gradual and phased progresses. Since most of the operating nuclear power plants in China are PWR units, and China implements and promotes a commercial closed nuclear fuel cycle, how to initiate a closed nuclear fuel cycle from the mature commercial pressurized water reactors is an unavoidable reality.

Different from the implementation of closed nuclear fuel cycle reactors in countries such as France and Russia, the operating status and modes of PWRs in China are varied significantly. Most PWRs in China have implemented different plant modifications such as reactor power upgrading, core design and fuel management improvements with different new fuel types, different burnups and different cycle lengths, which have consumed certain degree of safety margins. These characteristics and differences bring challenges and difficulties to the implementation of a closed nuclear fuel cycle in China.

Based on international experiences and China’s situation, this paper discusses the necessity of initiating a closed nuclear fuel cycle from mature commercial nuclear power plants in China as the initial stage of the closed fuel cycle to lay the foundation for the future advanced nuclear fuel cycle, analyze and discuss the initiating mode of China’s commercial closed nuclear fuel cycle, review the nuclear fuel types to be utilized in the closed nuclear fuel cycle, and discuss the possible configuration and development path of China’s closed fuel cycle in the future.
Research on Signal Reconstruction Technology for Nuclear Power Plant Based on Self-Associative Kernel Regression
Authors: Miao Zhuang , Yu Fangxiaozhi , Xu Zhao
Published: November 2022

The instrument control system of the nuclear power plant is an important part of ensuring its safe and reliable operation. The sensor is the core component of the instrument control system to monitor the operation status of the nuclear power plant and trigger the operation of the instrument. Therefore, the availability of the sensor plays an important role of the safe operation of the nuclear power plant. The virtual sensor technology based on self-associative kernel regression can accurately determine the fault of the sensor signal, such as instrument drift, missing instrument values, etc. And it can replace the fault sensor with reliable and accurate values in a short period of time, ensuring normal operation of the system in a short period of time. In addition, the output value of the virtual sensor can also be used to calculate the measured value of the unmounted sensor part. This paper introduces a method to replace sensor values with right values when measured sensor values has something unusual. This paper only concerns about replacing sensor values and the paper assumes that it can find sensor fault correctly and timely. The paper selects the main steam system of the pressurized water reactor of the nuclear power plant as an example to verify the feasibility and reliability of the virtual sensor technology based on the self-associative kernel regression. The research results show that it is basically feasible to implement the virtual sensor technology by using the related technical solutions proposed in this paper.
Relay Chatter Analysis With Practical Application in Seismic PSA in Nuclear Power Station
Authors: Yang Jianfeng , Gu Xiaohui , Feng Bingchen , Wang Hangding
Published: November 2022

In the seismic PSA analysis of a nuclear power plant, the relay chatter analysis is a unique and important research content. An earthquake may cause inadvertent vibration of the relay contacts with an undesired output due to the low seismic ruggedness, and if the chatter time sustained more than 2 milliseconds, it will give an error signal. The error signal may lead to a refusing action or a mis-operation of relevant equipment, which will cause trouble to the mitigation of the earthquake accident of the nuclear power plant. Relay chatter could also actuate the lockout devices or seal-in devices associated with the supply breaker for essential buses, thus operator manual recovery is needed to re-close these breakers. Considering the accessibility of these equipment locations after an earthquake, it is difficult for operators to manually restore them if the number of these equipment is large. So relay chatter may cause an adverse impact on the accident mitigation functions by rendering selected accident mitigation equipment unavailable. This paper briefly introduces the method and procedure of Relay Chatter Analysis for the nuclear power plant PSA or SMA, then, illustrate it with an example of 380 VAC contactor control circuit. Finally, the modeling and quantification process for relay chatter analysis results in seismic PSA is briefly introduced, and a reference for solving similar problems is given.
An Evaluation of Critical Heat Flux Prediction in Single-Side Heated Narrow Rectangular Channel Under Upflow Condition
Authors: Yan Meiyue , Ma Zaiyong , Pan Liangming , He Qingche , Xv Wangtao , Li Xiang
Published: November 2022

Safety is the biggest concern of nuclear power plant and research reactors.As one of the three primary thermal design criteria for rectors, the critical heat flux (CHF) guarantee the reactor’s safety and economics. The bubble dynamics in narrow rectangular channel is didderent from that in conventional channels. Therefore the CHF characteristics obtained in the conventional channel may not be suitable in the narrow rectangular channel. To analyze CHF characteristics in the narrow rectangular channel, a visual experiment study on CHF was carried out in one side heated narrow rectangular channel under upflow condition condition. The experiments were performed at pressures ranging from 1 to 4 MPa, with inlet subcooling ranging from 65 to 120 K and mass flux ranging from 350–200 kg/(m2s). The deionized water was used as the working medium. The relevant thermal-hydraulic parameters and visualization results were collected synchronously. Then compare the CHF experimental values with the predicted values of the W-3 correlation,Mishima correlation, and Chang H correlation. The result shows the predicted values of W-3 correlation and Mishima correlation are always larger than the experimental values. Among these corrrelations, we found that the errors of Chang H correlation is within 30% in 2mm gap size.
Experimental Research of Bundles Radial Power Distribution Influencing Factor on Fuel Assembly Mixing Characteristics
Authors: Cheng Cheng , Donghua Lu , Qianhua Su , Chengyue Wang , Yanyan Xi
Published: November 2022

It can be seen that the studies on fuel assembly mixing performance in the reports are all focused on the influence factors of thermal parameters, and few on the structural parameters. In order to systematically investigate the influence of radial power distribution of rod bundles on the mixing characteristics of fuel assembly, the experimental study was carried out by a 14 ft 5 × 5 full-length rod bundle fuel assembly with both left-right symmetry and center-symmetric arrangement of cold and hot rod bundles. CFD numerical simulation method was used to calculate the same condition. The results show that the radial power ratio of the rod bundles has little effect on the mixing coefficient kT of the fuel assembly when the rod bundle is centrally symmetrical. When the rod bundle is left-right symmetrical, the temperature matrix of the sub-channel section presents an angular symmetry phenomenon near 45°. The experimental results and CFD simulation both show that the configuration of the mixing vanes forms a macroscopic diagonal circulation of the fluid in the sub-channel. FLLICA-IIIF subchannel software could not accurately simulate the diagonal circulation phenomenon of the fluid in the subchannel.
U Influence of Non-Condensable Gas in Mobile Heat Pipe Small Reactor on Heat Pipe Heat Transfer Performance
Authors: Tianyu Gao , Tao Zhou , Wenbin Liu , Jianyu Tang , Huaichang Lu
Published: November 2022

Heat pipe cooled reactors have a promising future in the next generation of reactor design. The heat pipe, which is the core device for heat transfer from the core to the outside, determines the safety of the heat pipe reactor and the efficiency of power generation. In the high-temperature, high-pressure, high-irradiation environment of a nuclear reactor, the non-condensable gas produced by the lithium heat pipe - helium - affects the heat transfer efficiency of the heat pipe. Therefore, by modeling the heat transfer process of the heat pipe, the effect of the content of the non-condensable gas on the heat transfer performance of the heat pipe is calculated. It can be concluded that: at a certain input power, the temperature of both the hot end and the cold end of the heat pipe increases with the increase of the content of non-condensable gas; the distribution of non-condensable gas in the heat pipe is irregular, more at the hot end and less at the cold end; at a certain input power, the equivalent thermal resistance of the heat pipe increases with the increase of the content of non-condensable gas, and the higher the content of non-condensable gas, the greater the effect on the heat transfer effect of the heat pipe; in a certain range The effect of power on the equivalent thermal resistance of heat pipe is almost none.
Application of Recurrent Neural Network in Improving Software Reliability Model of Reactor Protection System
Authors: Li Duo , Zhang Wenji , Guo Chao , Zhou Shuqiao
Published: November 2022

The safety-critical software of Reactor Protection System (RPS) plays a significant role for the safe operation of the Nuclear Power Plant (NPP). The reliability analysis of safety-critical software is also expected by the nuclear regulation agencies and the software development groups for test evaluation and optimization.

Software Reliability Growth Model (SRGM) is a commonly used reliability model. SRGMs can be divided into Markov model, non-homogeneous Poisson process (NHPP) model and so on according to the stochastic process method of its application. The NHPP model occupies a significant proportion in the software reliability models, which considers the failure rate as a function of time and, in the ideal case, the failure rate decreases with the removal of faults; after the software is released, the failure rate no longer changes unless the software is debugged again.

SRGM researches are based on the recorded software error data which are a set of time series data and collected during software development. There are lots of extensive and in-depth researches to improve the software reliability growth model, such as considering more hypothesis in software development, analyzing software error severity, and so on. The rapidly developed neural network technology provides a new kind of research tool to analyze these recorded data. This paper uses software error data to train a Recurrent Neural Network (RNN), the trained RNN can give more correct software reliability growth prediction and improve SRGM results.
Scaling of Thermal and Flow Behavior in Dry Storage Casks for Nuclear Spent Fuel
Authors: Wen Qinglong , Yang Ting , Xu Shijia , Wang Zhengquan , Yuan Bo
Published: November 2022

In China, the demands for nuclear power have increased with increasing energy consumption. Thus, requirements in the treatment of spent fuel have also increased. As spent fuel is radioactive, cooling is required to prevent the spent fuel from reaching high temperatures. From the perspective of long-term development, dry storage casks have become a priority intermediate storage method in China. Before this storage method is put into application, the nuclear regulatory authority requires a thorough safety evaluation by experiments or analyses to ensure that it maintains its thermal safety function under normal, abnormal, and accident conditions. It is pointed out that experiments with the full-scale model are expensive and are typically very large. However, the safety of such structures may be examined through the use of a dimensionally scaled-down model. To ensure an accurate analysis of the scaled-down structures, the scaling of thermal-hydraulic behaviors from the prototype dry storage casks for spent fuel is required.

Firstly, a general scaling method of a conventional testing system is introduced in this paper. Secondly, the principle used to perform the scaling of a dry storage cask simulator is carefully explained. Then a Phenomena Identification and Ranking Table (PIRT) of thermal and flow behaviors involved in a dry storage cask are presented. After that, dimensionless criteria which are used to characterize thermal-hydraulic behaviors in the dry storage cask are derived. Finally, the distortion is numerically analyzed.
Rejection of Synchronous Vibration Based on Generalized Adaptive Period Iteration Algorithm for Active Magnetic Bearing
Authors: Bian Xiaoyu , Shi Zhengang , Mo Ni , Zheng Yangbo , Shi Lei
Published: November 2022

High temperature gas-cooled reactor (HTGR) is the fourth-generation advanced nuclear reactor developed by Institute of Nuclear and New Energy Technology, Tsinghua University, which has advantages of inherent safety and modular design. Since the rotor is suspended by electromagnetic forces, active magnetic bearings (AMBs) possess the advantages of on friction and no need of lubrication. Hence the AMBs are widely applied in HTGR, which can ensure the pure helium environment for rotational machineries. However, in general, the imbalance of the rotor will result in the vibrations of the rotor and the shell of the machine during operation, which seriously affects the stability and security of the AMB system. This paper proposes a generalized adaptive period iteration (GAPI) algorithm. This method can estimate the rotating speed based on the displacement vibration signals and realize the rejection of the synchronous vibrations, which keeps the rotor rotating around its geometric axis. Unlike conventional unbalance control methods, the proposed algorithm does not require the rotating speed sensor, which simplifies the structure of the AMB system. Moreover, this method is independent of mathematical model of the system and can adoptively adjust the amplitude and phase of the feed-forward compensation signals to suppress the synchronous vibrations. Simulations and experiments show that this algorithm can quickly and accurately estimate the rotating speed, and has good stability and robustness.
Seismic Analysis of Fuel Transfer Tube Gate Valve
Authors: Ziwei Su , Xin Zhang
Published: November 2022

The Gate valve is an important functional part of the Fuel Transfer system, in order to achieve complete independence of Fuel Transfer unit, the scientific research and manufacture of the gate valve is imperative. According to the objectives, the overall structure design, component function design and material selection are completed, and then manufacture scientific research prototype, carry out a series of performance tests such as strength test, sealing test and whole machine life test, etc. Among them, it is an important step to verify the integrity of the Gate valve under seismic conditions, so in accordance with the design structure, establish the 3D geometric model, divide mesh, creat the finite element model, and apply the boundary conditions, the numerical simulation for Gate valve was presented by using ANSYS software.

The results show that under a combination of loads of various working conditions, including SL-1 abnormal condition or SL-2 accident condition, the stress of each component of Gate valve body, Support, Operating mechanism and Driving rod is less than the specified limit and meets the requirements of the judgement criterion; so the equipment achieve complete independence of Fuel Transfer unit, and be with reliable and stable performance, which can effectively guarantee the safe operation of nuclear power plant.
Design of Loading Elevator Control System Based on Servo Positioning Control
Authors: Zhao Lihui , Dong Shengguo , Zhu Hao
Published: November 2022

Loading elevator is one of the important equipment for refueling system. The refueling process of the sodium-cooled fast reactor is sealed and the equipment for refueling process is complicated and in large quantity. It is necessary to research and develop a set of high-accuracy and high-reliability control system to ensure loading elevator for the refueling process can operate safely and reliably. This article illustrates and describes loading elevator’s control system design detail. For meet the requirement that the positioning deviation is limited to ±5mm, this control system hardware for loading elevator adopts the advanced digital AC server to position the equipment. Compared with the traditional PLC control method, it has the advantages of high positioning accuracy, simple commissioning process and stable operation. Meanwhile, the multi-fault detection logic is used for fault detection. Though acquire real-time data, the fault logic transfers protection signals and interlocking signals to DCS. This fault logic can ensure that when the equipment operates abnormally, it can detect the fault in real time and stop the motor to ensure the safety of the equipment and components. The control system design plan has been applied in the loading elevator test bench safely and reliably for many times, which proves that it can realize the function of equipment monitoring and control, and also the positioning accuracy can meet all requirements.
Application of Qualitative and Quantitative Configuration Risk Assessment in AP1000 Unit Outage Scheduling Optimization
Authors: Wang Zhenqi , Huang Weibing , Zhou Jian , Lin Huan , Tao Yinyong , Zhang Jun
Published: November 2022

Configuration risk management refers to the use of probabilistic safety assessment technology to calculate risk indicators and perform risk management based on the actual operation configuration of nuclear power plant. Since its development in 1990, this method has been widely used in nuclear power plants. It can help power plant personnel optimize maintenance plan scheduling, control the nuclear risk level of units and improve the safety of power plants. During the shutdown and outage of the nuclear power plant, a lot of work has to be performed, resulting in the centralized shutdown of system equipment, and the nuclear safety during this period has attracted the attention of the power plant. Therefore, in order to ensure that the nuclear risk of the unit shutdown and outage can be controlled, Sanmen Nuclear Power, as the world’s first AP1000 unit, actively performs configuration risk management and control during the shutdown, It is the first time in China to optimize the outage plan scheduling through the combination of qualitative and quantitative configuration risk assessment method, so as to comprehensively investigate and eliminate the potential risk sources. This paper will mainly introduce the development background of configuration risk management, and take Sanmen Nuclear Power AP1000 unit as the object to introduce the application and achievements of qualitative and quantitative configuration risk assessment in outage scheduling optimization. In addition, it will also make a comparative analysis of qualitative and quantitative configuration risk assessment techniques.
Design of Radiation Damage Monitoring Scheme for Materials in China Experimental Fast Reactor
Authors: Ziao Hu , Xiao-xian Chen , Xiao-liang Chen
Published: November 2022

In order to monitor the radiation effects of advanced fuel element cladding materials, it is necessary to install irradiation supervision tanks in the reactor, which contain kinds of detector foils. And then the detector foils are used to determine the neutron energy spectrum and neutron flux, so as to evaluate the radiation damage of the material. Taking China Experimental Fast Reactor (CEFR) as the research object, seven neutron detector foils were selected and five supervision schemes were designed in 316 stainless steel. The size of the neutron detector, the irradiation time and the corresponding relationship with the radiation damage are calculated by the Monte Carlo code and SPECTER program, and finally the sound monitoring scheme is given. It has guiding significance for irradiation experiments of structural materials, and provides a basis for monitoring the irradiation damage of large commercial fast reactor structures.
Multiplication Factor Prediction Feasibility Study for the BEAVRS Core Based on the LSTM Algorithm
Authors: Chang-an Ren , Ji-chong Lei , Xiao-hua Yang
Published: November 2022

This paper focuses on exploring the feasibility of the LSTM (Long Short-Term Memory) algorithm in deep learning for effective multiplication factor keff prediction at the core level, modeled by BEAVRS (Benchmark for Evaluation And Validation of Reactor Simulations) core first cycle loading with keff of operating at full power for 0–300 days was used as the study subject. The first 65% of the dataset is the training and validation set, and the last 35% of the dataset is the prediction target. The training and alignment results of the physical parameters of the components were obtained using the DRAGON4.1 and DONJON4.1 codes, and the LSTM algorithm in deep learning was applied. By adjusting the number of LSTM cells, L2 regularization parameters, optimizer type, and other parameter coefficients in the algorithm, the results showed that the absolute error of the predicted core effective multiplication factor keff could be made within 2 pcm by adjusting the appropriate parameters, which validated the successful application of machine learning to transport equations.
On Multiaxial Strain Characteristics of Piping Systems Subjected to Earthquake Loads
Authors: Morishita Masaki , Otani Akihito
Published: November 2022

The JSME Code Case N-CC-008 provides design rules for piping seismic design by inelastic response time history analysis and strain-based fatigue criteria. Since the strain time histories obtained from the analysis are multi-axial and random, the Code Case uses the Rainflow method for cycle counting in its procedure of evaluating the ranges of equivalent strain and the fatigue usage factor. The Rainflow method is used for identifying pairs of peaks/valleys of the representative strain and their occurrence times. Then, with the other components of strain and stress at these times, the range of equivalent strain is calculated sequentially. On the other hand, ASME Boiler and Pressure Vessel Code Section VIII, Division 2 reads “The Rainflow cycle counting method is recommended but not applicable for non-proportional loading.” The authors, then, analyzed the behavior of multi-axial strain from the inelastic response analyses of some dynamic tests of piping systems to test if the piping strains due to seismic excitations are proportional or non-proportional, or somewhere in between. The IS method, which was proposed by Itoh et.al for quantifying the degree of non-proportionality of multi-axial fatigue cycles, was used for the analysis. The analysis results showed that the multi-axial strain in piping systems induced by seismic excitation is almost proportional. Based on this result, the authors concluded that the Rainflow cycle counting method is well applicable for the piping seismic fatigue evaluation. In addition to these numerical analyses with the IS method, some discussions are also made in this paper from the viewpoint of the relation between vibration mode and strain components (compatibility condition between deformation and strain) to support the conclusion that the piping strain by seismic loads is proportional.
TCR Central Shutdown Rod Fine Motion Control
Authors: Varma Venugopal K. , Fountain Eliott , Burkhardt Earle , Cetiner Sacit , Ottinger Dustin
Published: November 2022

The Transformational Challenge Reactor (TCR) is a helium-cooled 3 MWt test reactor with a design that leverages advances in materials and manufacturing, computing, and artificial intelligence. Gravity, pneumatics, or springs are typically used to quickly release the central shutdown rod — which contains the neutron absorber — into the reactor core, thus stopping the nuclear reaction. The rod’s motor/actuator typically resides outside the reactor, and the motion is transmitted through a penetration into the pressure vessel. Some control rod designs incorporate magnetic latches with a coil outside the pressure boundary to ensure precise position control.

The proposed design comprises a magnetic coupling to position the shutdown rod, with no penetration into the pressure vessel for the entire drive length of TCR shutdown rod. Electromagnets are used in nonpower nuclear reactors for shutdown rod control, but they reside inside the reactor pressure vessel. This paper describes the use of an electromagnet outside the pressure vessel to position and release the shutdown rod. A prototype was developed at Oak Ridge National Laboratory to demonstrate the concept, and the ferritic core and material design was optimized to maximize the electromagnet’s lift force. An elevated temperature test was also conducted to ensure that the system will perform effectively inside an operating reactor.
Study on Preparation and Radioactive Methyl-Iodine Adsorption Performance of TEDA Ionic Liquids
Authors: Li Xin , Li Yongguo , Yu Jie , Chen Xin , Hou Jianrong , Li Jian , Qiu Dangui , Shi Xu
Published: November 2022

As the most important impregnation component of iodine adsorption materials, Triethylenediamine (TEDA) has a disadvantage of being volatile. This disadvantage has a negative impact on the adsorption performance of iodine adsorber, resulting in an increase in the operation cost of nuclear facilities and the amount of solid waste. Based on this problem, TEDA ionic liquids (TEDAILs) was prepared by quaternization of TEDA, which is a method to effectively solve the volatile problem of TEDA. The changes of the yield of TEDAILs and the proportion of single-cationic ionic liquids (SCILs) were investigated under different reactant ratio, reaction temperature and reaction time. According to the law of change, the optimal preparation condition of TEDAILs was revealed. In addition, the composition of the TEDAILs was analyzed by infrared spectrum and nuclear magnetic spectrum, and the thermal stability was also verified by thermogravimetric analysis. On this basis, TEDAILs loaded activated carbon fibers (TEDAILs-ACF) was designed and prepared, and the iodine adsorption performance was tested. Verification test results show that the complete decomposition temperature of TEDAILs is about 362 °C, which is about 187% higher than that of TEDA. Moreover, the adsorption efficiency of radioactive methyl iodine by TEDAILs-ACF under normal temperature and humidity is 99.94%. In summary, TEDAILs-ACF as radioactive methyl-iodine adsorption material has potential application prospect.
DFMEA Application in Rotor Design of Million-Kilowatt Nuclear Power Stream Turbine
Authors: Binghan Xue
Published: November 2022

DFMEA is a systematic and scientific method for new product design and development, which is applicable throughout the entire process of new product design and development. In this paper, DFMEA method is adopted in the systematic analyzing of the million-kilowatt nuclear power steam turbine rotor new product design and development process. On the basis of the introduction of product characteristics and current development requirement, DFMEA is introduced to the design and development process, applying the block diagram and structure-function matrix to locate the possible failure factors. Subsequently, the evaluation criteria of Severity (S), Occurrence (O), and Detection (D) are introduced and defined. According to the criteria, the Risk Priority Numbers (RPN) and Action Priority (AP) of key risks are sorted, therefore indicating the terms critical to product quality improvement.

Through the application of DFMEA, an effective risk identification and analysis of the million-kilowatt nuclear power steam turbine rotor is carried out, achieving the results that operational failures are advanced to the design stage to implement countermeasures, therefore realizing the improvement of product and system operation reliability, optimization of project cycle and manufacturing cost reduction. DFMEA promotes the optimization of overall new product development process. Meanwhile it provides ideas for the subsequent design optimization methods of similar kinds.
Volume Change Analysis of Primary Loop in a Small Modular Thorium-Based Molten Salt Reactor
Authors: Zhu Guifeng , Zou Yang , Yan Rui , Guo Wei , Kang Xuzhong , Xu Tiangui , Yang Pu , Zhou Bo
Published: November 2022

In a molten salt reactor, the volume change in the primary loop may lead to an overflow of fuel salt, which will challenge the radioactive containment. Several factors, such as the thermal expansion, graphite dimensional change caused by fast neutron irradiation, fuel feed online and fuel salt reconstitution that affect volume change are analyzed based on a designed small modular thorium-based molten salt reactor (SM-TMSR) in this paper. The results show that the thermal expansion of nickel alloy is comparable to that of fuel salt, the net volume change due to isothermal expansion is not obvious, about 0.04 m3 per 100 °C in SM-TMSR, which can be easily contained in a pump bowl without an additional overflow tank. Furthermore, the volume decrement due to graphite shrink by irradiation (based on the IG-110 data) is more obvious than the volume increment by fuel feed online, which will lead to a decline of liquid level during the long-term operation. 0.13 m3 excess volume of fuel salt should be provided in the pump of SM-TMSR. Finally, as we analyzed, fuel salt reconstitution will add to the difficulties in volume control, reactivity control and the control of heavy metal mole ratio in fuel salt. It is recommended that the used fuel salt can be recycled into a new power module without any treatment, and the increased volume of fuel salt caused by feed online in last power module can be accommodated in a larger downcomer.
Study on Application of Hollow Springs in Seismic Isolation in Nuclear Power Plants
Authors: Hou Nannan , Dong Jianling
Published: November 2022

Earthquake is always an important consideration when constructing nuclear power plants. There are many studies on how to resist or avoid earthquake damage. And recently more and more seismic isolation technology has been applied to nuclear structure in nuclear power plants using isolators mainly made of rubber which can not only bear the vertical load, but also reduce the response of seismic on the upper structure due to their low lateral stiffness so as to improve the safety of the nuclear structure in the earthquake. Compared with traditional anti-seismic technology, seismic isolation can make the design of the main structure of nuclear power plants achieve standardization unrestricted by the magnitude of design ground motion.

The two main categories of seismic isolators are elastomeric bearings and spherical sliding bearings. And low damping bearings (LDR), lead rubber bearings (LR) and spherical sliding bearings (FP) are currently considered sufficiently well characterized for use in nuclear facilities. However, the rubber which is the main material of elastomeric bearings cannot withstand high temperature, so the properties of the bearings might be affected and even lost under the complex and severe environment of nuclear power plants, which could compromise the safety of nuclear reactor. In addition, both rubber bearings and spherical sliding bearings can only apply to horizontal seismic isolation.

The metal springs have more stable performance than rubber under high temperature condition and also has better isolation effect. Springs are usually made of solid spring wire by metal forming. In production practice, springs are used for vertical isolation because of their low vertical stiffness, and calculations show that they also work in the lateral direction. Springs have been widely used for vibration isolation in many light load and high requirements situations like precision instruments at this stage, but need more research to be applied in nuclear structure of nuclear facilities.

Compared with air spring, under the same load-bearing capacity, the stiffness of solid spring is higher, which leads to the higher natural frequency of the isolation system, so it is not suitable for the situation that requires low frequency and heavy load. This paper tries to use springs made of hollow wire to solve the problem. In this paper, some research results of hollow springs are given, including frequency and damping property when filled with damping materials. In addition, a calculation formula of hollow springs’ mechanical property, including the vertical and lateral stiffness and carrying capacity, is derived, and the finite element calculation are used to validate the formula. And the paper also gives a finite element calculation case on how the hollow spring reduce the natural frequency of an isolation system.
Application of Nuclear Insurance Risk Survey Results in Nuclear Power Plant Safety Improvement Program
Authors: Ping Jiang , Jiangtao An , Changliang Liu
Published: November 2022

From the prospective of loss prevention, economic risk reduction and catastrophe risk reduction, nuclear insurance risk survey provides professional and foresighted safety improvement recommendations based on the international nuclear and insurance experience. As a common international risk management practice, nuclear insurers provide loss prevention evaluation services, which is considered as peer review by nuclear operators. With the development of Chinese insurance industry and increasing loss prevention demand from Chinese nuclear power plant, Chinese insurance industry gradually integrates international practice, and establishes a systematic, Chinese characterized nuclear insurance survey methodology. This thesis introduces insurance survey background and importance, analyses the areas in which insurance survey could support nuclear power risk management. This paper focuses on risk survey methodology and implementation plan based on insurance service practices, and analyses survey result trends. The methodology has been implemented in nuclear power risk management regime. Recent cases have verified the insurance survey could support nuclear power plant safety improvement. Besides, based on the current practice, this paper suggest the potential ways to improve the interface between insurance risk survey and nuclear power risk management, improve common risk and enhance survey effectiveness.
Effect of Water Jet Behavior on the Operating Range in Supersonic Steam Injector
Authors: Minamizono Yukiya , Kaneko Akiko
Published: November 2022

In the case of a severe accident, such as a loss of power supply, it is necessary to keep the nuclear power plant in a safe condition. Therefore, steam-water injectors (SI) have been attracting attention as a device that can cool the plant without power supply in the event of a severe accident. In a SI, the steam condenses by direct contact between steam and water interface. As the steam condenses, negative pressure is generated, and a force that attracts water is generated. The SI can process a large amount of cooling water with this force. It can also function as a reactor condenser for nuclear reactors due to its high condensation performance. There are some characteristics of SI that the discharge pressure is expected to exceed the inlet pressure after being accelerated at the throat, and it has a very simple structure with converging-diverging part, which is expected to reduce the cost of installation and maintenance. Previous studies have indicated that GROLMES et al. (1968) inferred the flow in the reduced test section by measuring the void fraction and MIWA et al. (2018) suggested a relationship between internal flow and the operating range of SI, referring to the results inferred by GROLMES et al. (1968). Although these existing studies have suggested some relationship between the internal flow and the operating range, the relationship between the stability of the water jet and the behavior of the SI has not been clarified.

Therefore, the aim of this study is to clarify the effect of water jet behavior on the operating range of SI. The taper angle of the water jet outlet was varied and the relationship between the water jet behavior and the operating range was investigated. In this study, experiments were conducted using two types of taper angles at the water jet outlet: 20 and 7 degrees. As a result of detailed observation inside the SI, it was confirmed that the water jet with the taper angle of 7 degrees was less spreading than that with the taper angle of 20 degrees. In the operating range of this experiment, no significant difference was observed when the taper angle of the water jet was changed. However, it is suggested that the water jet with a certain degree of dispersion might be better in terms of heat exchange due to large surface area. These results suggest that the stability of the water jet is a necessary but not a sufficient condition for operation.
Effect of Brownian Diffusion on Film Condensation With Non-Condensable Gas
Authors: Peng Cheng , Song Youya , Meng Xianpin
Published: November 2022

The innovative design of passive containment cooling system (PCCS) is one of the most significant safety improvements for the third generation of advanced pressurized water reactors (APWRs), however, its impact on the hydrogen risk in severe accidents for nuclear power plants (NPPs) has not been determined yet. On the one hand, natural convection through liquid film condensation on the inner wall of the containment may mitigate the hydrogen risk, but on the other hand, rapid steam condensation may also ruin the existing inerting environment within the containment. Brownian diffusion is one of the essential mechanisms which drives the steam condensation near the wall. In this paper, the distributions of the velocity field, temperature field, and concentration field in the gas mixture boundary layer are firstly obtained through the build-up of a two-dimensional mathematical model allowing for the Brownian diffusion behavior of the mixing gas under the presence of non-condensable gas. Compared with the COPAIN test data and simulation results from the GASFLOW code obtained from previous literature, the applicability and prediction error of the present model are evaluated. Besides, the comparison with the prediction results based on Nusselt’s classical liquid film condensation equation demonstrates that the Brownian diffusion of the mixing gas, including steam and air, will reduce the heat transfer rate through the liquid film by 1–2 orders of magnitude. The theoretical model from the present work can support the development of new CFD codes and help get new understandings of complex thermal-hydraulic phenomena in hydrogen risk under various containment cooling conditions.
Fuel Management Strategy of LBE-Cooled Fast Reactor BLESS-D
Authors: Luo Zhen , Yeoh Eing Yee , Chen Xiaosong , Li Linsen , Xing Mian
Published: November 2022

As one of the six nuclear reactor candidates selected by the Generation IV International Forum (GIF), lead-cooled fast reactor (LFR) has become one of the most promising concepts and attracted more attention from the institutes. A lead-bismuth eutectic (LBE) cooled fast reactor called BLESS-D has been proposed by State Power Investment Corporation Research Institute of China. The fuel management is critical to reactor design because it affects reactor economics. Although the research on fuel management of PWR has matured, there are few studies on fuel management for LFR. Fuel management design includes the entire process from initial cycle to transition cycle, and to equilibrium cycle. To obtain a better refueling scheme and optimize the key parameters of equilibrium cycle, a new refueling scheme is first proposed, and then traditionally, the simulation covering the entire refueling operation is required cycle by cycle, resulting in consuming a massive computing resource. If the scheme fails to meet expectations, expenditure will multiply until the modified scheme meets the safety design criteria. Fuel management design using “pseudo-equilibrium cycle” instead of traditional method is carried out in this work. The “pseudo-equilibrium cycle” method can directly solve the core key parameters of equilibrium cycle by replacing fuel model with approximate nuclides densities estimated from initial core arrangement and refueling scheme. In this paper, a two-batch refueling scheme is proposed with “pseudo-equilibrium cycle” method and then transition cycle are designed to ensure the feasibility that the fresh core successfully transition to the “pseudo-equilibrium cycle” state. Afterwards, neutronics parameters are solved for each cycle from the fresh core and results show that when the burnup calculation comes to the 5th cycle, the reactor key parameters, including assembly peaking factor, linear power density, delayed neutron fraction, and prompt neutron lifetime are in good agreement with “pseudo-equilibrium cycle”, which proves that the “pseudo-equilibrium cycle” method can be used accurately and efficiently to design the refueling scheme.