bluuebunny/crossref_metadata_2025 huggingface dataset:
Intelligent Monitoring and Diagnosis of CCWS Heat Exchanger For Nuclear Power Plant
Authors: Ming Qian , Hu Jian , Wang Wu
Published: November 2022

Component Cooling Water System (CCWS) heat exchanger in nuclear power plant plays an important role in normal operation and accident condition. Because of the different requirements of heat conduction performance under normal operation and accident conditions, it is necessary to monitor the heat conduction performance under normal conditions to valuation the performance under accident condition. At present, the heat transfer coefficient is calculated by collecting the temperature and flow parameters of the cold and hot sides of CCWS heat exchanger. This way of valuation does not consider the influence of the change of plant operating conditions on the heat transfer coefficient of the heat exchanger, and it is also impossible to predict the failure time.The intelligent monitoring and diagnosis system using big data intelligent algorithm and failure mechanism model can automatically calculate the evaluation parameters that can characterize the trend changes of thermal and hydraulic performance of heat exchanger in real time, and realize the purpose of predicting the faults of heat exchanger in advance, thus guiding on-site maintenance personnel to accurately arrange maintenance activities.Based on the research of intelligent monitoring and diagnosis system of CCWS heat exchanger, this paper introduces the development direction and suggested realization method of CCWS heat exchanger in the intelligent process of nuclear power plant, and puts forward the conception of developing intelligent monitoring and diagnosis system of CCWS heat exchanger.
Simulation and Analysis of 4H-SiC-Based PiN Fast Neutron Detector With Polyethylene Converter Using Geant4
Authors: Yanli Xiong , Xiaopeng Qiu , Shuhuan Liu , Zhuoqi Li , Shuangying Liu , Yong Ma , Ci Song , Ning Han
Published: November 2022

When detecting fast neutrons, generally select materials with high hydrogen content, such as polyethylene and lithium hydride. It’s known that polyethylene is widely used and has high detect efficiency. Therefore, this paper uses Geant4 software to simulate 4H-SiC-based PiN fast neutron detector with polyethylene converter of varies thickness. With the increase of converter thickness, the conversion efficiency of 1–14MeV monoenergetic fast neutron will first increase and then decrease. By analysing the trend of conversion efficiency, the thickness is determined as 1000um. Then, the 1MeV, 2MeV, 4MeV and 8MeV monoenergetic neutrons and typical fats neutron source: Am-Be neutron source are simulated under the thickness of 1000um convertor, the particle types and energy distribution to the sensitive volume and the energy deposition of particles in the sensitive region are also obtained. Through the analysis of recoil proton energy spectrum, the relationship between incident neutron energy and recoil proton energy are studied. Also, through the analysis of PKA (primary knocked-out atoms) energy spectrum, we found that the main reactions in epi-layer are 28Si(n,n’) 28Si, 12C(n,n’) 12C, 28Si(n,α) 25Mg, 28Si(n,p) 28Al, etc. It is found that heavy ions such as C and Si produced by neutrons in sensitive volume account for a large proportion. The above research finally helps to investigate the deposit energy in the epi-layer.
Modal analysis and additional fluid mass of coaxial three-layer thin-wall member in annular flow gap
Authors: Zhang Chaofan , Lu Daogang , Liu Yu , Qiu Haotian
Published: November 2022

The sodium-cold pool-type fast reactor is the preferred reactor type for the fourth generation advanced reactor research. The reactor structure of a pool-type fast reactor has the characteristics of large size, thin wall, and relatively low stiffness. Therefore, the seismic design, analysis, and verification of the pool-type fast reactor are the key concerns of the fast reactor safety evaluation. In the fast reactor, many important reactor structures, such as the main pump support cylinder, are immersed in the ring basin of the reactor body. These devices are equipped with thermal shields and can therefore be regarded as coaxial housing systems with water gaps between the cylinders. Under the seismic condition, the fluid-structure coupling effect between the shell and the water gap will change the natural frequency of the shell itself, and the change of structural vibration characteristics is very important to study the seismic performance of the equipment structure in the fast reactor. However, the calculation method based on the dry/wet modal frequency ratio is not suitable for the calculation of additional mass under non-full water conditions. In this paper, based on ANSYS finite element analysis method, a coaxial three-layer shell model with fluid clearance is established to simulate the main pump supporting cylinder and other equipment, and the modes under different frequencies are studied. The additional mass of fluid clearance under different modes is calculated by the relationship between fluid pressure field and shell areal-displacement during vibration. The results are of reference value to the structural design and test of coaxial multi-layer shells such as the support cylinder of the main pump of the fast reactor.
Monte Carlo Simulation of In-Situ Multi-Scale Damage Detection of AGR Graphite
Authors: Zeng Tianchen , Sun Yuewen , Cong Peng
Published: November 2022

The Advanced Gas-Cooled Reactor (AGR) is the most common design of nuclear reactor in the UK. These reactors were commissioned in the 1980s and have now reached the end of their design life. As a type of thermal reactor, it uses graphite as a moderator to control the reaction. The cracking of graphite blocks in the core is an important factor limiting the lifetime of AGR stations. The damage of the AGR graphite will affect the power of the reactor and also become a hidden danger threatening the safety of the reactor. As a thermal reactor, it uses graphite as a moderator. Since graphite bricks cannot be replaced or repaired, in order to ensure the continued operation of the reactor, it is necessary to obtain timely and accurate damage information of graphite bricks.

In order to carry out the research on in-situ multi-scale damage detection of AGR graphite, a linear CT imaging detection scheme was designed in combination with the core geometries of the AGR stations. Based on the GEANT4 Monte Carlo simulation software, an experiment is carried out to simulate a series of processes from signal acquisition to image reconstruction. Through the simulation experiment, the detection ability of the linear CT detection system under the interference of strong background radiation field signal in the AGR reactor and the geometric error of the detection system is verified, so as to evaluate the feasibility of the linear CT imaging detection scheme and guide the design of the detection system.
Root Cause Analysis of Nonrelevant Indication of Magnetic Particle Test on 18MND5 Steel Welded Joints in the Steam Generator of a Nuclear Power Project
Authors: Guo Yanbing , Lu Yanhong , Zhang Wang , Yang Miaosen , Liu Xiyang , Lu Binfeng
Published: November 2022

The 18MND5 steel is rolled and welded by the submerged arc welding process to fabricate the upper cylinder shell of the steam generator of a nuclear power project. After the shell is welded, Heat treatment is performed. Then, magnetic particle test (MT), ultrasonic testing (UT), and radiographic testing (RT) were performed on the longitudinal Joint, and no defects were found. However, there are MT nonrelevant indications characterized continuous linear at the position of the weld fusion line on both sides of the longitudinal Joint. Through the peeling observation of the Product Weld Test Coupon with the same linear indication, no defects were found. It can be determined that continuous linear indications are nonrelevant. Optical microstructure observation (OM), scanning electron microscopy (SEM), X-ray diffraction (XRD), and electron backscattered diffraction (EBSD) analysis was carried out on the 18MND5 welded joints to determine the cause of MT nonrelevant indication. The result showed that there is a noticeable growth texture in the coarse-grained lath martensite (M) region of the weld with MT nonrelevant indication, which is closely related to the ferrite (F) and parent metal of the weld metal on both sides. The ferrite (growth direction and morphology of the material structure are different. This growth texture leads to a significant difference in magnetic properties at the fusion line position of the welded joint, resulting in a mutation in the magnetic properties of the coarse-grained region, resulting in an uncorrelated MT indication.
Numerical Analysis on Pressurizer Liquid Level Characteristic Under Marine Conditions
Authors: Xu Sijie , Wei Tianyi , Yue Jinyu , Li Dongyang , Tan Sichao
Published: November 2022

The pressurizer of floating reactor moves dramatically with the of ocean waves, and liquid in pressurizer will slosh when subjected to additional inertial force. To study the hydro characteristics of the liquid in pressurizer, this paper analyzed the flow field and pressure variation of pressurizer under typical moving conditions. Combined with the influence factors of typical motion on the transient pressure variation characteristics at monitoring position under different working conditions, optimize the differential pressure measuring method of liquid level through numerical simulation, result in reduce the liquid level measuring deviation. Traditional numerical approach of liquid sloshing is improved based on the modified motion equation and coordinate transformation matrix approach. Combined with the moving grid and modified algorithm. Through comparison. The results show that the liquid level measuring deviation of heave motion with vertical additional acceleration and the pressure variation of rolling motion is mainly caused by the spatial position variation of the model and the liquid free surface. The mechanism of coupling of heave motion and rolling motion can be comprehensively considered in combination with the influences of every individual motion. In addition, through verification, the liquid level measuring can be corrected according to the attitude data and geometric information of pressurizer.
New Gamma Scanning Technology for 200L and 400L Low and Intermediate Level Solid Waste
Authors: Wang Dezhong , Yang Hui , Gu Weiguo , Qiu Jie , Zhou Wentao
Published: November 2022

Segmented gamma scanning (SGS) and tomographic gamma scanning (TGS) are two traditional detection techniques for low and intermediate level radioactive waste drum. However, detection accuracy of SGS is low and detection time of TGS is very long. This paper summaries three types of new gamma scanning techniques to assay the radioactive waste drum. Compared with SGS and TGS, these techniques exhibit the well-balanced performance which can improve the detection accuracy and shorten the detection time.
Study on Methodology for Quantification of Radioactive Discharges and Limits for Pressurized Water Reactor HPR1000 Based on Operating Experience
Authors: Chen Yujia , Lv Weifeng , Jiang Zhenyu , Zhou Yonghai
Published: November 2022

For a nuclear power plant, the radioactive gaseous and liquid discharges are the main contributor to radiation exposure to the member of public and non-human biota during normal operation, which hence need to be quantified and to support the environmental impact assessment.

When applying the traditional theoretical methodology, due to various and complex mechanisms involved in radioactive gaseous and liquid effluent streams, a number of assumptions need to be made to support the theoretical modeling. The combination of these assumptions can easily lead to overestimate or underestimate of the radioactive discharges and limits and may not represent the actual performance of the plants. As such, to obtain predicted discharges and limits closer to the future actual performance of the plant, it is meaningful and necessary to develop a methodology based on operating experience.

This paper has studied and developed a systematic methodology based on operating experience for quantification of radioactive discharges and limits for the 3rd generation pressurized water reactor HPR1000 during normal operation, taking into account the differences on design features and operation management between the HPR1000 and the operating units, the fluctuations due to the variations of plant and system operation parameters and the potential influences from expected events within the normal operation range. This methodology has been successfully applied to HPR1000 and the results have been verified reasonable and appropriate by comparing with the operating experience data from comparable international PWRs.

This methodology has been applied to HPR1000 successfully for Generic Design Assessment (GDA) in the UK and the European Utility Requirements for LWR Nuclear Power Plants (EUR) and can also be widely applied for other PWRs with slight adjustment.
EMC Performance of Different Cable Routing Styles for I&C Systems of NPPs
Authors: Yang Youwei
Published: November 2022

In the nuclear power plants, cables are signal transport links between separated signal processes that must be allowed to communicate with one another in a specified manner. Nowadays, for the instrumentation and control system of nuclear power reactor, there have always two ways to place the cables, i.e. above the cabinet or underneath the cabinet. One of the two ways is often preferred on account of the local installation environment or the human factor, while the electromagnetic compatibility (EMC) performance of these two ways is seldom reported. The pitfall may compromise the electromagnetic compatibility performance of the cabinet, or the system or installation it is used in, even when all the components used in the cabinet have excellent EMC compliance individually. In this paper, based on the numerical simulation technology, numerical models were built for the instrumentation and control system with the two different cable routing styles. Radiated emission, radiated susceptibility and crosstalk of cables connected to the cabinets are mainly researched. Results show that the different cable routing styles mainly affect the radiated susceptibility performance and the way that underneath the cabinet is preferred.
A Chinese 300MWe Two-Loop PWR NPP LBLOCA Analysis Based on the Deterministic Realistic Hybrid Methodology
Authors: Li Yuhan , Kuang Bo
Published: November 2022

Loss of coolant accident (LOCA) is among the important limiting design basis accidents for a PWR nuclear power plant (NPP). In China, a 300MWe two-loop PWR NPP, although facing the challenge of life extension, still adopted rather conservative tools and methods for safety analysis. This is supposed to have guaranteed sufficient margin for safe operation of the plant during the past years, yet, at the expense of plant economy and operation flexibility. To evaluate the safety margin more reasonably and realistically, the mixed methodology of DRHM (deterministic realistic hybrid methodology) is introduced for LBLOCA analysis of the Chinese 300MWe two-loop PWR NPP in the paper, with which conservative evaluation model plus best estimation analysis tool is applied, and effects of uncertainty of important plant state parameters are quantified.

In the DRHM analysis of postulated LBLOCA caused by double ended-guillotine-cold-leg break for the 300MWe two-loop PWR NPP in this paper, the evaluation model RELAP5-APK (the conservative Appendix K physical models plus best-estimate system analysis code RELAP5/MOD3) is developed and verified. And during the transient analysis of the LBLOCA scenario, uncertainty of the effects of important plant state parameters are quantified through statistical sampling and corresponding calculation. Taking the cladding peak temperature (PCT) index for demonstration to measure the safety margin, the single-sided confidence upper limit including 95% PCT of the sampling population with 95% confidence level is acquired. The resultant shows that a greater PCT margin is achieved compared with that in the original FSAR. This provide a further confidence for life extension or power uprate of the plant.
A Steady-State Analysis of U-Pu-Zr Metallic Fuel Performance on Constituent Redistribution by Multiphysics Modeling
Authors: Liu Rong , Liu Shengyu , Han Huiyun , Yang Liwen
Published: November 2022

A Multiphysics approach to calculate the constituent redistribution in U-Pu-Zr metallic fuel is developed in this work and the fuel performance of U-Pu-Zr metallic fuel under normal operating condition is investigated by CAMPUS code where the model of constituent redistribution is applied. The metallic nuclear fuel is proposed to be used as a candidate of Sodium cooled Fast Neutron Reactor fuel due to its superior thermodynamic properties, such as thermal conductivity and heat capacity. Recently, the constituent redistribution is found to be an important phenomenon in the metallic fuel pellet, which influences the thermodynamic properties of metallic fuel greatly. In this work, the properties of metallic fuel and the calculation method of constituent redistribution are introduced firstly. And the model of constituent redistribution in U-Pu-Zr metallic fuel is developed and implemented into CAMPUS code, and then the metallic fuel performance in the sodium-cooled fast reactor is further studied by CAMPUS code under normal operating condition. Thirdly, the model of constituent redistribution is verified and the fuel performance of U-Pu-Zr metallic fuel is presented and discussed. The Zirconium element is found to migrate to the fuel centerline and fuel surface and the performance of the U-Pu-Zr fuel is found to be influenced by the constituent redistribution dramatically.
Research of Probabilistic Safety Assessment of CPR1000 Considering Aging Effects
Authors: Zhang Xiaoming , Wang Lijuan
Published: November 2022

In order to quantify aging impact to core damage risk of CPR1000 units which are widely distributed in China, it is necessary to develop a technical framework of probabilistic safety assessment (PSA) considering aging effects. Several basic technical elements affected were identified. At first, general rules were established to identify and screen structures, systems and components (SSCs) based on their risk contributions and aging sensitivities. No further work was required for SSCs screened out. Some SSCs which are not included in current PSA model should be added if aging effects are considered, e.g. pipes and cables. Secondly, a formula to calculate the aging failure rate was developed, considering aging effects and aging management actions. Thirdly, a method of integrating aging effects into current PSA model was developed, including the development of aging basic events and handling of different failure rates with different operating time. The accident of large break loss of coolant was selected for case analysis. A brief PSA model was developed and a quantitative result was generated for the selected accident. The result of the case analysis showed that aging effect could have a significant influence to the core damage frequency and relevant risk insights of PSA for a CPR1000 nuclear power unit.
Dynamic Precision Evaluation of Compensated Ion-Chambers for Reactor Power Monitoring
Authors: Huang Qichang , Liu Mingxing
Published: November 2022

During reactor power monitoring, compensated ion-chambers (CICs) are assembled by two similar gas-filled detectors. The gamma compensation of CICs is essential for detecting precision of thermal neutrons, and requires a pairs of fully matched voltages. Descriptions of detecting precision include current formation model, in which applied voltages and radiation strength act in the process of collecting charged particles. However, the model of CICs are seldom applied in design work to evaluate the monitoring precision. Here we show that imbalance degree of collecting charged particles leads to mismatch of saturation currents depending on difference of recombination and diffusion in each sub-chambers. We analyzed a coaxial type compensated ion-chamber (DL-126), and found that applied voltages and radiation strength have significant impact for ion recombination and electron diffusion. For increasing precision of CICs, this result suggests a dynamic way to evaluate the output deviation using radiation strength. Furthermore, we found the compensation factor of two sub-chambers is only related to the applied voltages and remain unchanged in various radiation. Our results demonstrate how collecting regulars of charged particles are likely to function in setting applied voltages, calculating the charge loss as well as quantifying the saturation currents. We anticipate our analysis to be a starting point for revising detection precision of compensated ion-chambers. For example, evaluating the neutron fluence in the reactor shutdown and restart process, including monitoring task after accident. Furthermore, dynamic precision evaluation of these detectors is a major target of prognostics and health management for Nuclear Instrument.
Corrosion Behavior of a Novel Alumina Forming Austenitic Steel Exposed to Supercritical Water
Authors: Sun Dayun , Gao Yang , Cong Shuo , Guo Xianglong
Published: November 2022

The corrosion behavior of novel alumina-forming austenitic steel Fe-26Ni-19Cr-2.5Al-1Nb-0.5Si and 310S steel in aerated supercritical water (SCW) at 550 °C/25 MPa was investigated. The AFA and 310S steels has been exposed in supercritical water for up to 1000h. The results show that both the weight gain curves of AFA and 310S steels follow near-parabolic law. Although the weight gain of 310S steel exposed in supercritical water after 1000h was up to 37 mg/dm2, the weight gain of AFA steel exposed in supercritical water after 1000h was near 18 mg/dm2, which was only half of that of 310S steel. The weight gain curves indicating that the AFA steel has better corrosion resistance than 310S steel. Besides, microstructure characterization of two steels has been conducted by SEM, EDS and XRD. SEM images shows that there are some differences between surface morphology of 310S steel and AFA steel. The microstructure results show that 310S steel has a double oxide layer: a Fe-riched outer layer and a Cr-riched inner layer, while a multilayer structure mainly composed of Fe-riched oxide layer, Cr-riched oxide layer and Al-riched oxide layer was formed on AFA steel, indicating a different corrosion process from 310S steel. The corrosion mechanism of two steels based on the microstructure is discussed in detail.
Bubble Behaviors in Saturated Flow Boiling Based on the VOF Model
Authors: Gao Deyang , He Wen , Bo Hanliang
Published: November 2022

Flow boiling in vertical pipes widely exists in nuclear reactors. The bubble motion and behavior attract more and more attention due to their influence on the heat transfer capacity. However, the most used two-phase flow model in numerical simulation, called Eulerian two-fluid model, is not able to obtain parameters of phase interfaces and provide an observed image of bubble movements. Therefore, the present study aims to simulate the bubble behavior and trajectory in the saturated flow boiling in the vertical pipe using VOF model which can treat the deformation of interfaces. Firstly, the bubble parameters were calculated based on existing bubble dynamic models. Then, the random generation of bubbles was simulated by directly adding mass and momentum sources into the flow region by the User Defined Function (UDF). Finally, the deformation, collapse, and coalescence of bubbles were simulated and predicted in bubbly flow. The feasibility of the newly proposed model is proved by comparing with experimental data and other simulation results. The present study provides a new idea for simulating the movement of a large number of bubbles in the saturated flow boiling in vertical pipes, which adopts bubble parameters as boundary conditions directly, instead of heat and mass transfer models. In comparison, the present model can greatly improve the calculation efficiency.
Analysis of Contribution Degree of Vibration Transmission of Bolted Structure
Authors: Liu Shuai , Li Wanyou , Huang Xuan , Xiong Furui , Jiang Xiaozhou , Deng Liwei
Published: November 2022

In practical engineering, the vibration transmission between bolted structures supported by cantilever beams involves many factors, including the friction between plates and the bolt connection relationship. The vibration transmission effect is the embodiment of the coupling of various factors, so it is impossible to accurately and quantitatively evaluate the influence degree of various factors. In this paper, the influence factors of vibration between bolted plates in the form of cantilever beam are studied by numerical calculation method. The effects of friction coefficient, bolt preload, temperature and internal and external pressure on vibration characteristics are studied. Finally, the contribution of different influencing factors to vibration transmission is given. The preliminary research shows that the bolt connects the two plates into a whole structure, and its stiffness changes little with the friction coefficient, bolt preload, temperature and internal and external pressure. The natural frequency of each order increases with the increase of friction coefficient, bolt preload and internal and external pressure, decreases with the increase of temperature, and the overall trend changes little. At the same time, the bolt preload is verified by test, the difference of natural frequency under the same preload is compared and analyzed, the numerical calculation is compared with the test value, and the vibration contribution of preload is verified.
Experimental Studies and Mechanical Properties Analysis of Three-Dimensional Oblique Rotating Friction Seismic Isolators
Authors: Wu Saifeng , Liu Wenguang , He Wenfu
Published: November 2022

Traditional isolation devices widely provide protection only against the horizontal components of earthquake motions. The vertical ground motions are transmitted through the isolation system unchanged or even magnified. Thus, in this paper, a mechanical model of a novel three-dimensional oblique rotating friction seismic isolator is proposed. The device with oblique-torsion coupling deformation is designed by combining three inclined rubber bearings. The test results show that the device have asymmetric behavior and a plump hysteretic loop in a vertical direction. After three times cyclic loading, the slope of the curve for the device hardly changes, and there is no damage in the loading process. The error of vertical post-yield stiffness of the device between the experimental result and theoretical prediction is within 10%, and it is in good agreement with the theoretical model prediction. In addition, the effects of different parameters are highlighted based on the theoretical mechanical model. The results show that the inclination angle, the friction coefficient, radius of PTFE, distance from the center of the inclined LRB to the center of the rotating block, and the performance of the inclined LRBs strongly influence the vertical behavior of the proposed device and should be worthy of attention in the design of devices.
An Experimental Study on the Retention of Graphite Dust Aerosol in Pool Scrubbing and the Possible Application in HTGRs
Authors: Wang Yating , Zhang Yiyang , Fang Zhu , Wu Xinxin
Published: November 2022

Owing to the inherent safety, the modular high-temperature gas-cooled reactor (HTGR) eliminates the possibility of severe accidents in principle. Therefore, the source term is the main concern for HTGR accidents. Especially, the graphite dust problem, arising from the abrasion of fuel pebbles, complicates the evaluation of source terms. In the water ingress accidents, the graphite dust may be released to the containment during the overpressure discharge, which not only contributes to the source term but also makes trouble for post-accident measures. In this work, we consider the possibility to develop a pool scrubbing measure to retain graphite dust and fission products for the overpressure discharge of HTGR. A scrubber is set up with aerosol particle concentration measurement at both the inlet and outlet, which enables us to determine the decontamination factor (DF). By conducting a series of experiments for different submergence depth and particle size, it is demonstrated the DF value increases with the increase of submergence depth. When the submergence depth is 105 cm, the retention efficiency of graphite particles larger than 2 μm exceeds 96%. Our previous study has shown due to a fragmentation effect, most of the graphite particles released during the overpressure discharge are below 2 μm, while the DFs for these small particles fall below 20. The experimental results show that the pool scrubbing technology can efficiently retain the graphite particles for the overpressure discharge in HTRGs. More efforts are still needed to effectively improve the DF for graphite particles below 2 μm.
Simulation of Plasma Magnetic Confinement System Based on Modelica Language: Starting with a Single Particle
Authors: Lyu Hanyang , Wang Xiang
Published: November 2022

This article uses Modelica language to create a plasma magnetic confinement system with a single charged particle. The physics used in the article is basically the equations of a charged particle moving in a magnetic field. The Modelica language is an object-oriented component-based programming language that can easily modularize the system. The geometric part represented by the program can be processed automatically after the input of parameters. At last, the program simulates three different operating conditions for a magnetic confinement system. The initial velocities and initial positions that are different from the standard case, and the combination of the two are selected as the experimental objects. By comparing with the simulation with the case without magnetic confinement, the results show that the confinement system is effective.
Seismic Isolation Reliability Analysis of a Lead-Bismuth Small Modular Reactor
Authors: Huang Yi , Qi Min
Published: November 2022

Seismic isolation technology has been proven to significantly reduce the seismic response of the structures, systems and components (SSCs) in a nuclear power plant (NPP), improve the seismic reliability and safety of nuclear power plants, reduce construction costs, and shorten construction periods. Reliability analysis of a lead-bismuth small modular reactor (SMR) with a seismic isolation design was conducted: ANSYS APDL was used to establish a finite element model of the SMR, modal analysis was performed on the isolated SMR, and the dynamic characteristics of the SMR was obtained; seismic design spectrum proposed by Newmark-Hall was used as input for seismic spectrum analysis, and the seismic response evaluation of SMR was carried out; ANSYS PDS (Probability design system) module was used to carry out probabilistic calculation of the above process, and the calculation considers the uncertainties in the physical parameters of structure materials and isolation bearings, geometric dimensions and seismic response spectrum, the random variables were sampled by the Latin hypercube sampling method in the Monte Carlo method, and failure probability of isolated SMR and non-isolated SMR were obtained through mathematical statistics. The calculation results show that the seismic isolation design can significantly reduce the failure probability under earthquakes and improve the safety of lead-bismuth SMR.
The Reconstruction of 3d Radiation Field Based on Sparse Measurement Data
Authors: Xie Xingwen , Cai Jiejin , Tang Zhihong
Published: November 2022

With the development of nuclear technology, more and more occupational workers work in radioactive environments. The three-dimensional (3D) dose rate field is of great significance in the radiation protection field. Many researchers focus on the reconstruction of 2D radiation fields but few of them paid attention to the reconstruction of the 3D radiation field based on sparse measurement data. In this work, an interpolation method is introduced to reconstruct 3D radiation field using sparse sampling nodes. The proposed method includes interpolation on sampling grid lines, interpolation on sampling grid plane and interpolation in the three-dimensional space. Additionally, two simulation cases are conducted in MATLAB to prove the effectiveness and accuracy of the proposed method. Case1 and Case 2 are 3D radiation fields with a single radioactive source and two radioactive sources respectively. The reconstruction radiation field is compared to the dose rate field simulated by the Monte Carlo method. By using less than 1% of measurement data, the proposed method precisely reconstructs radiation fields with an average relative error of 0.82% and 1.71% respectively. What’s more, the accuracy of the proposed method was studied at different sampling node ratios. As the ratio of sampling nodes decreases, the average error and maximum of the relative error increase obviously. The proposed method is an interpolation method suitable for the radiation field and it can reconstruct the 3D radiation field accurately with sparse data.
Research on Cold End Optimization of Water Supply Mode of Expanded Unit System of Circulating Water System in Coastal Nuclear Power Plant
Authors: Lin Fang , Jian Jiale , Wang Fang , Yang Jia
Published: November 2022

A typical northern nuclear power plant and a typical southern nuclear power plant are selected to optimize the cold end of the circulating water system with one machine and two pumps in an expanded unit system. And study the optimal operation mode of the circulating pump and the cold end configuration scheme with the minimum annual cost in this water supply mode. It is discussed how to combine the site conditions of the nuclear power plant, the price of raw materials and the main cold-end equipment of the steam turbine, the cost of the main structures of the water supply and drainage system and the related annual operating costs, power generation income and other factors, to seek the optimal solution through economic and technical comparison of multiple schemes. That is to select the optimal matching combination scheme of the configuration of the cooling water system such as the configuration scheme of the low-pressure cylinder of the steam turbine, the area of the condenser, the flow rate of the cooling water pump, and the cooling water pipe trench, and determine the economical and reasonable design parameters of the equipment at the cold end. By comparing the cold end optimization results of the constant speed scheme of 1 machine with 2 pumps, the frequency conversion scheme of 1 machine with 2 pumps and the expansion unit system scheme of 1 machine with 2 pumps, the frequency conversion scheme of 1 machine with 2 pumps is superior in terms of technology and economy and has a great advantage for the northern site. For the southern site, due to the higher water temperature and smaller variation, the advantage is less.
Study on Technology and Application for Health Assessment of Electronic Equipment
Authors: Guilian Shi , Xiaojin Huang , Xiaopeng Zhao
Published: November 2022

Electronic equipment is an indispensable part of high reliability equipment such as aerospace and nuclear power plant (NPP). It is difficult to evaluate the health condition of the electronic equipment because of its complex structure and uncertain fault behavior. This paper firstly classifies the health assessment methods of electronic equipment, introduces in detail the main technical methods under the three categories of classical assessment methods, data-based assessment methods and physical model-based assessment methods, and analyzes the advantages and disadvantages of each method. In addition, considering the increasingly complex functional structure of electronic equipment, its physical failure mechanism is more difficult to grasp. From the perspective of application potential, this paper focuses on the research of data-based health assessment strategy, proposes a multi-model health assessment method based on working conditions, gives a general model of electronic equipment health assessment, and gives a way to obtain model parameters through the local optimal least square method. Finally, the above strategy is applied to study the degradation condition of DCDC power module aging test data of nuclear power board, and the validity of the proposed algorithm is verified based on the degradation data of board aging test, which provides a reference for electronic equipment health evaluation.
Seismic Behavior Study of a new Isolated Structure System With Damping Supporter for Nuclear Power Plants in a Cave or Underground
Authors: Feng Yixin , Xu Hao , Liu Wenguang , Zhang Qiang
Published: November 2022

There have been heated discussions on how to reduce the earthquake response of NPPs. Generally, isolation technology is one of the most efficient methods but researches on traditional isolated structure and 3D isolated structure for NPPs, to some extent, have deficiencies. Therefore, in this paper, a new structural system which included base isolation system and lateral damping supporters for NPPs (hereinafter referred to as DISO) in a cave or underground was proposed to obtain a positive seismic response, aiming to protect the nuclear island. The motion characteristics of the system have been studied and a simplified mathematical model has been set up. The dynamic equation was deduced and solved. The theoretical analysis results that there is an optimal range of parameter β, the ratio of the damping coefficient of the additional damping support to the equivalent linear damping coefficient of isolation layer, to get a positive response of DISO system. The seismic response of the DISO structural system was analyzed and the correlation that the parameters including damping arrangement coefficient, damping index and damping coefficient have with the seismic response of the structure were discussed. The results show that the acceleration response of the DISO system decreases to minimum value with the increasing of damping coefficient, along with whiplash effect, and then increases regularly. There is a positive correlation between the damping coefficient and displacement decline rate. A case study has been completed. The numerical results show that the DISO system performs more effectively than the ISO system (refer to the traditional isolated structural system). Given that the equipment and pipelines in NPPs are sensitive to large isolation displacement, the displacement response of DISO system is less than that of the traditional isolated structural system without acceleration amplifying. Moreover, both acceleration of superstructure and displacement get decreased. And IDA-based (Incremental Dynamic Analysis) seismic fragility analysis of ISO and DISO system has been performed. DISO system has the smaller probability of failure compared with the ISO system.
Measurement of Unbalance Vibration Phase for Active Magnetic Bearing in High Temperature Gas-Cooled Reactor
Authors: Bian Xiaoyu , Shi Zhengang , Zheng Yangbo , Mo Ni , Shi Lei
Published: November 2022

In high temperature gas-cooled reactor (HTGR), the primary helium circulator (PHC) is the key component of the reactor. The active magnetic bearing (AMB) has some outstanding advantages such as micro-friction, no need of lubrication, high temperature resistance, extended life and so on. It is the ideal support device for the rotor of the PHC because it can ensure a clean helium environment. Generally, due to the existence of mass imbalance of the rotor, the AMB system will generate unbalance vibrations during operation. Accurate measurement of the unbalance vibration phase is of great significance to the analyses of dynamic balance and speed up and down of the AMB system. In this work, the principle of the unbalance vibrations and the incidental problems are analyzed in great detail. And a simple and practical unbalance vibration phase measurement algorithm is proposed by combining hardware structure with software programing. This algorithm consists of real-time peak detection and phase calculation, which can realize real-time and accurate measurement of the unbalance vibration phase. In addition, the proposed method can reduce the effect of the noises of the AMB system. The experiment results demonstrate that the proposed algorithm effectively revises the conventional phase measurement methods, improves the accuracy and reduces the inherent error of the system.
Study on the Fallback Time Analysis Method for Loss of Ultimate Heat Sink Under Normal Power Operation of Nuclear Power Unit
Authors: Zhang Zhao , Yan Jun-Jie , Li Gen , Chong Dao-Tong , Liu Ming , Wang Zi-Chun , Shen Rong-Fa
Published: November 2022

For the beyond design basis accident, loss of ultimate heat sink under normal power operation of CPR1000 nuclear power unit, an analytical model is proposed to evaluate the process during which the component cooling system is cooled by the Refueling Water Storage Tank, which is very important for the accident treatment. The results of the analytical model agree well with that provided by EDF. Based on the according strategies of State-oriented Procedure, a fallback time analysis method is carried out, the result is in accordance with the simulation time on the full-scope simulator. These indicate that the analytical model for the cooling process and the fallback time analysis method are correct.
A Novel PID Self-Tuning Method Based on Closed-Loop Identification Used in Nuclear Power Plants
Authors: Zhang Pengbin , Xing Kunpeng , Zou Tao , Yang Zhijia , Wang Jingyang , Wang Hongrui
Published: November 2022

Nuclear power is one of several significant clean energy and its production capacity is increasing rapidly. This paper proposed a novel PID self-tuning method based on closed-loop identification used in nuclear power plants to deal with the turbine’s load tracking problem. A setpoint-shifting method is adopted to achieve the informativity of the closed-loop system, and the PID parameters are designed through an optimization method. For validating the proposed method, this paper modeled a pressurized-water reactor (PWR) nuclear power plant, a complex system consisting of three circuits, through the first principle. The primary circuit contains three main devices: a nuclear reactor, a pressurizer, and a steam generator. The models of these three devices are provided by previous literature separately. The whole model is generated through mass and energy balance. This mechanism model is used as the plant model, while the identified model is used as the control model. The initial PID parameters are set through open-loop identification using an external excitation. Then the proposed method works online to track the turbine’s load in time. The effectiveness of the proposed method is illustrated by simulations.
Investigation on the Inspection Technology of GH3535 Alloy Butt Weld With Linear Array Probe
Authors: Yuan Xiaodan , Li Zhijun
Published: November 2022

The critical problem to accurately characterizing defects in GH3535 alloy weld is the deflection of the acoustic beam and scattering due to the coarse columnar grains. Signal-to-noise ratio is an important index to indicate whether the grain scattering is severe in the ultrasonic inspection. In this paper, the phased array ultrasonic testing of GH3535 alloy butt weld with the thickness of 46mm was studied using sector scan mode with linear array (LA) probe with 64 channels. The center frequency is 2.25MHz, and the incident angle is 19.87°. The sound field characteristics of the linear array probe with different focusing parameters were analyzed, and the signal-to-noise ratio (SNR) in the detection was calculated. The results show that: the grain size of butt weld ranges from 40um to 500um. The acoustic beam of the linear array probe can cover the weld thickness range of 5–39mm, based on the removal of weld reinforcement. The signal-to-noise ratio of transverse hole with φ3mm located at the weld-fusion line is more than 15dB, when the front end of the probe is located directly above the transverse hole of weld-fusion line.
A Modified Simulation Model of Blocked Regions After Rod Cladding Deformation During LOCA Transient
Authors: Cao Zhiwei , Lin Zhikang , Xu Miaomiao , He Jimin , Song Jianyang
Published: November 2022

During a loss-of-coolant-accident (LOCA) transient, the ballooning and burst of the cladding can lead to blockage of flow path and worsen core coolability. Studies have shown that the core coolability is not changed much up to about 60% blockage ratio, while can be threatened by a severe blockage with a high blockage ratio and long blockage length. A blockage simulation model based on the non-coplanar assumption is widely used in LOCA transient analysis. However, regarding the coplanar deformation observed in experiments, validity of this assumption is questionable. Furthermore, it is found that this model cannot predict reasonable results for severe blockage conditions.

In this paper, the blockage model is modified for the thermal hydraulic simulation of LOCA transients with the occurrence of severe blockages. Pre-test simulations of a test facility studying the influence of blockage on core coolability have been conducted with the modified model. The results show that, the modified model shows better agreement with the observations in the experiments.
Study on Reactor Vessel Water Level Measurement Method for Emergency Decision-Making of Nuclear Power Plant
Authors: Shi Yanming , Wang Zhenying
Published: November 2022

After a nuclear accident, the nuclear power plant will monitor and analyze the unit status, environmental characteristics and accident process, determine the emergency state classification under the nuclear accident conditions, and respond to the emergency state hierarchically, to reduce the impact of the accident and ensure the safety of personnel. For determining emergency state classification of nuclear power plant cold shutdown and refueling shutdown operation mode, reactor vessel water level is an important basis for judging. The Three Mile Island nuclear accident made the industry realize the necessity of monitoring the primary water load, especially the water level in the Reactor vessel. Reactor vessel water level provides an essential basis for monitoring the core cooling state after an accident, to ensure that the core cooling state can be diagnosed correctly in time, accurately and conveniently, thus providing a criterion for determining the emergency state level, and then selecting the appropriate accident operation strategy.

Based on the demand analysis of emergency state classification based on reactor vessel water level, combined with the current situation of reactor vessel measurement in Reactor vessel and the setting of peripheral dose rate monitoring channels, this paper supplements the water level measurement method under potential water loss conditions, realizes the full-range measurement of Validity water level in Reactor vessel, evaluates Validity water level Validity in Validity, and then efficiently determines the emergency state classification, providing Validity basis for emergency response.