bluuebunny/crossref_metadata_2025_split huggingface dataset:
Development and Verification of New Drum Filter Equipment for Nuclear Power Plants
Authors: Xiaojie Huang , Yishan Lin , Honghu Xie , Yanqing Li , Jinchun Yang 
Published: November 2022

Abstract
Drum filter is a large-scale filtration equipment in nuclear power plants, providing clean cooling water for nuclear power plants. At present, there are problems of easy blockage, short response time of the unit after blockage, and shutdown of the unit in severe cases during the operation of the drum filter in the in-service nuclear power plants. The analysis shows that the drum filter adopts the structure of “internal water intake”, and the garbage such as marine organisms remain on “inside” of the drum filter during operation, which makes it difficult to clean up and operate the inner garbage. When the marine organisms suddenly break out, if the amount of garbage cleaned by the drum filter is not enough to offset the amount of garbage entering, the garbage will accumulate continuously inside, which may lead the pressure difference between the inside and outside of the drum filter to suddenly increase, and then cause the circulating pump to trip or even the reactor to shut down. The inherent safety of the drum filter of this water intake method is low. To solve this problem, this paper puts forward a new concept of water intake of drum filter, that is, the concept of “external water intake”, which changes the water intake method of the drum filter, intercepts the garbage on the “outside” of the drum filter. Under the action of water flow, the garbage outside is not easy to remain on the net and block the drum filter, thereby fundamentally improving the inherent safety of the drum filter. A new type of “external water intake” drum filter was developed by using parametric and modular design method. The reliability of the drum filter was analyzed by using the FMEA failure mode and impact analysis method, and the weak links of the drum filter were found out, and optimization and improvement measures have been taken for the weak links. The ANSYS finite element analysis software is used to conduct a seismic analysis of the drum filter, which proves that the drum filter meets the requirements of the RCC-M specification and the seismic requirements of the third-generation nuclear power plants, which demonstrates the safety of the drum filter. Finally, through the physical prototype test, the design concept of “external water intake” was verified, and garbage cleaning capacity of the new drum filter meets the requirements of nuclear power plants. This type of drum filter has been popularized and applied to many HPR1000 projects under construction and newly built ones.
Intelligent Monitoring and Diagnosis of CCWS Heat Exchanger For Nuclear Power Plant
Authors: Ming Qian , Hu Jian , Wang Wu 
Published: November 2022

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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.