bluuebunny/crossref_metadata_2025 huggingface dataset:
Temperature Dependent Removal Efficiency of Crankcase Control Devices
Authors: Hicks Myles D. , Northrop William F. , Potratz Daniel , Kittelson David B.
Published: November 2022

Oil aerosols formed in the crankcase of reciprocating engines are a significant source of particulate matter and may lead to engine component degradation when used in closed crankcase ventilation (CV) systems or constitute a significant emission source in open CV systems. In modern CV systems, filtration or inertial separation is used to collect oil particles. These devices are highly efficient over a range of flow rates and temperatures; however, some system conditions result in a decrease in efficiency with increasing temperature in a manner not predicted by theory.

Experiments were performed on a bench system that introduced atomized oil particles into a clean, pre-heated, air stream that was conveyed isothermally through a heated duct to an oven containing the removal device, either a coalescing filter or an inertial separator, followed by another heated section and an exit duct. The aerosol was sampled using identical upstream and downstream isokinetic sampling systems and characterized using a Dekati Electrical Low-Pressure Impactor (ELPI) that measured particle concentration and size in the range from 0.043 to 8.46μm aerodynamic diameter.

Neither single fiber efficiency (SFE) filtration theory nor physical arguments describing the performance of the inertial separator predict a significant dependence total and fractional efficiency and most penetrating particle size (MPPS) on temperature in the range explored here, 25 to 115°C, but in many cases our measurements show MPPS shifting to smaller size and total efficiency decreasing with increasing temperature. These observations are consistent with the hypothesis that, at device temperature, the particles being processed are smaller than the particles being measured, at room temperature. Particles shrink by evaporation as they pass through the heated sections, but cool and grow by condensation as the sample cools in the sampling lines or in the exit duct.

Reported carbon number distributions of typical lubricating oils and evaporation kinetics indicate that droplets in the size range investigated may lose more than half of their mass under temperature conditions, up to 115°C, found in typical crankcase removal devices under engine high-load conditions. These droplets are in a dynamic balance with their vapors and may change during sampling and measurement. Great care must be taken in design of systems used to characterize devices intended to remove volatile droplets and interpretation of measurements made, as the size of the particles measured may not be the same as the size of particles passing through the device.
Complementary Simulations to Determine Heat Transfer Coefficients and the Maximum Heat Flux in Multi-Nozzle Spray Cooling Experiments
Authors: Du Yuxuan , Bandaru Satya V. Ravikumar , Villanueva Walter
Published: November 2022

For Light Water Reactor (LWR) safety, spray cooling during severe accidents is one of the promising approaches to achieve In-Vessel Retention of corium by External Reactor Vessel Cooling (IVR-ERVC). To study the efficiency of multi-nozzle spray cooling (nozzles of 2 × 3 matrix) on a downward-facing FeCrAl heated surface, a lab-scale experimental facility was built. It should be emphasized, however, that a direct measurement of Heat Transfer Coefficient (HTC) on the sprayed side is challenging due to the strong interference of water flow and intrusiveness of standard instrumentation methods. In this paper, a 3D numerical model has been established with the same geometric and material parameters as the foil sample in a multi-nozzle upward spray cooling. Given the experimental temperature profiles on the sample’s dry side measured by an IR camera, the complementary numerical simulations have revealed the HTCs and corresponding temperature profiles on the sprayed side, which enabled the prediction of the maximum heat fluxes (MHFs). The maximum heat fluxes for the given spray cooling conditions can reach up to 3.25MW/m2, which is more than adequate for what is required for a successful IVR-ERVC for high-power reactors. At the same time, the maximum temperature on the dry side at the highest input power is still much lower than the expected failure temperature of the sample material.
Effect of the Plasticity of Piping and Support Structures on the Seismic Response of Piping Systems
Authors: Okuda Takahiro , Takahashi Hideki , Watakabe Tomoyoshi , Morishita Masaki
Published: November 2022

This study describes the results of seismic analyses conducted to clarify the effect of the plastic deformation of a pipe and its support structures on the seismic response of the entire piping system. The first type of analyses includes the investigation of plasticity with regard to the following parameters: (1) none, (2) piping, (3) support structure, and (4) piping and support structure. The second type of analyses includes the parametric survey of the yield load and post-yield stiffness of the restoring force characteristics of support structures.

The results of the first type of analyses have demonstrated significant implications on the seismic design of piping systems based on the elastic-plastic response analysis. First, the piping response, i.e., elbow strain, is significantly reduced when plasticity is considered in the piping support. On the contrary, the piping support response is not so much affected by the piping plasticity.

The results of the second type of analyses showed that the yield load of the support structure significantly affects the response of the piping system, whereas the post-yield stiffness does not affect much compared with the yield load.
Numerical Simulation of Aerosol Generation and Aerosol Spray Scavenging During Fukushima Daiichi Decommissioning
Authors: Liang Hui , Nejdet Erkan , Wang Kai , Sharma Avadhesh Kumar , Suzuki Shunichi
Published: November 2022

In the future decommissioning plan of the damaged Fukushima Daiichi nuclear reactors, the melted and re-solidified fuel debris at the bottom of the reactor pressure vessel and primary containment vessel will be cut into small pieces before retrieving them from reactor buildings. Submicron aerosol particles that generated during cutting operations must be removed inside the containment vessel before they escape to the environment. The existed spray system inside the containment vessel can be used to remove these aerosol particles under different collection mechanisms. In this study, a new numerical model is developed to simulate both the aerosol generation by cutting fuel debris and aerosol removal by spray injection in the open-source CFD code OpenFOAM using the geometry of our model facility (UTARTS facility). Simulation results provide detailed information about the time evolution of aerosol distribution, and also show that aerosol generation rate has no influence on aerosol removal efficiency, and larger particles can be removed faster because of stronger inertial impaction. The newly developed numerical model with considering both aerosol generation due to cut fuel debris and aerosol scavenging by spray droplets can be used in the real-size containment vessel of Fukushima Daiichi nuclear reactors to optimize the design of future spray systems.
Sensitivity Calculation of Lower Shell Pressure Parameters in Advanced Nuclear Power Plant Accidents
Authors: Lanyu Zhou , Yanyu Sun , Yuntao Zheng , Changjiang Yang
Published: November 2022

As the main coolant piping from the steam generator to the reactor pressure vessel, cold leg is an important part of the primary circuit. During a LOCA accident in the cold leg, the parameter sensitivity has a crucial influence for the containment response. PAREO program is used to analyze the sensitivity of the pressure response of HPR1000 containment. Quantitatively evaluated the influence of free volume, area of thermal components, spray flow rate and internal atmospheric temperature on the pressure and temperature of containment under the condition of LOCA. The results show that the area of thermal components is the most sensitive parameter to the pressure response process of containment. The research results could provide technical support for containment design, safety analysis and assessment.
Information Security Design of Nuclear Power Embedded Real-Time Operating System
Authors: Li Meng , Shi Guilian , Ma Zhonggang , Wang Xiaowei , Fan Lichen , Dou Weiwei
Published: November 2022

With the development of modern automatic control technology, digital control technology has advanced technology and inherent advantages compared with traditional analog control technology. Therefore, the use of digital instrument and control system has become an inevitable trend of global nuclear power construction. Nuclear power embedded real-time operating system is an important part of digital instrument and control system in nuclear power plant. It is responsible for managing the hardware resources of the whole system and providing operation environment for application software. Its safety directly affects the safety of instrument and control system. This paper studies the application of embedded information technology in the safety of nuclear power plant, and puts forward a safety related scheme of embedded information system. The scheme fully considers the confidentiality, integrity and availability of information in the digital instrument and control system of nuclear power plant. The application results show that the scheme can meet the information security requirements of nuclear power instrument and control system, and has reference and guiding significance for the research and development of nuclear power embedded real-time operating system.
Classification Method and Response Strategy of Public Opinion Risk Related to Nuclear-Related Event Occurred in Countries Around China
Authors: Zhao Hongru , Tong Jiejuan
Published: November 2022

Due to the particularity of the nuclear-related event, even if it does not have any serious consequences, or even does not occur in China, it may also bring a greater risk of public opinion, that is, the technical risk is low but the public opinion risk is high. In order to provide a scientific basis to identify the impact of Nuclear-related Event occurred in countries Around China (NEAC) on China, we established an integrated classification framework of public opinion risks related to these events, and the indicators of each evaluation dimension have been identified. In another aspect, based on the theory of crisis life cycle and the theory of crisis management 4R, some suggestions on the response strategy of the public opinion risk caused by NEAC were proposed in this paper.
Microstructure and Hardness Investigation on 25Cr2Ni2MoV Steel Welding Joint of Nuclear Power Steam Turbine Low Pressure Welded Rotor and Engineering Applications
Authors: Huo Xin , Liu Xia
Published: November 2022

Large Nuclear power steam turbine Low pressure (LP) rotor composed of several shafts and discs with 25Cr2Ni2MoV steel was manufactured by gas tungsten-arc welding (GTAW) for backing welding and submerged-arc welding (SAW) for filling welding. In this study, the welding joint is investigated and it is found that the main microstructure of weld metal and base metal is typically bainite, while tempered martensite is observed in heat affected zone (HAZ) by over-quenching imposed by welding cycling and post weld heat treatment (PWHT). Compared with the weld metal of SAW, the microstructure of GTAW is relatively fine and homogeneous, due to its lower heat input. The hardness test indicates that the maximum hardness of HV400 occurs in HAZ because of tempered martensite. Weld metal of SAW has the hardness of HV270, which is a little bit lower than GTAW’s on account of fine grain. Based on the researches, the trial welded rotor and production rotor have been successfully produced, relevant inspections and tests were conducted, the results are well consistently with original research.
Neutronics Analysis of RBMK-1000 Reactor Using OpenMC: Criticality and Depletion Analysis at 3D Assembly Level
Authors: Hashemi Mohammad , Macián-Juan Rafael , Ohlerich Martin
Published: November 2022

This article presents the first results of the neutronics analysis of RBMK-1000 at the 3D assembly level using the OpenMC code as an early stage of the long-term large academic research project. The goal of the large academic research project is to develop a Multi-group Monte Carlo computational module around the OpenMC code using Python programming language. The results of this paper include the criticality calculation and depletion analysis of RBMK-1000. The effective multiplication factor value (keff) has been determined. In depletion calculation, the fission reaction rates of U-235, U-236, U-238, production of fissionable materials, and two fission product poisons Xe-135 and Sm-149 have been investigated. There is ongoing work for benchmarking the first results from OpenMC against Serpent 2 Monte Carlo code. The parametric design data of RBMK-1000 have been taken from the OECD - Nuclear Energy Agency (NEA) SFCOMPO 2.0 (Spent Fuel Isotopic Composition) database that includes measured isotopic concentrations of spent nuclear fuel, with operational histories and design data. The case study was the Leningrad-1 NPP. The ENDF/B-VII.1 data library has been used in this analysis. Moreover, this is the first time to date that the OpenMC code has been used in neutronic studies of the RBMK reactors.
Numerical Study on Welding Deformation of Upper Support Structure in 600MW Sodium-Cooled Fast Reactor
Authors: Dong Wenchao , Lu Shanping
Published: November 2022

The upper support structure of 600 MW sodium-cooled fast reactor is welded by inner and outer cylinders, upper and lower tube-sheets and multiple rib plates. The structure has a large size and many welding passes, which poses a great challenge to the control of welding deformation. It is an effective method to control welding deformation of structural parts by using numerical simulation method to predict welding deformation and optimize welding process. In this work, the welding process of typical T-joint is numerically simulated and tested firstly. The reliability of the simulated results is verified by comparing the temperature field, and the inherent strain of typical joint is obtained. Then, the inherent strain method is used to calculate the welding deformation of typical T-joint. The welding deformation accuracy is compared and analyzed by using thermoelastic-plastic method and inherent strain method, respectively. Finally, the welding process of the upper support structure is simulated by the inherent strain method. The influence of the designed welding sequence on the welding deformation of the upper support structure is studied, and the shrinkage allowance are obtained. The research results have been verified by field application, which provides technical support for the construction of 600 MW sodium-cooled fast reactor.
Design and Investigation of Liquid Lead-Bismuth Test Irradiation Loop in Research Reactor
Authors: Xu Yuheng , Si Juping , Zhang Liang , Sun Sheng , Wu Hongwei , Lu Mengkang
Published: November 2022

The Lead-bismuth eutectic (LBE) test irradiation loop is an indispensable testing vehicle for the investigation of LBE-cooled reactor fuel. To make a water-cooled research reactor have the irradiation test capacity of LBE-cooled reactor fuel, a test irradiation loop is investigated and preliminarily designed by absorbing the design and operation experience of the domestic and international heavy liquid metal loop. Structural material corrosion, impurity blockage, polonium diffusion, and metal solidification in the loop are considered in the design process. Corresponding measures have been taken to ensure the safe operation and emergency capacity of the loop. The loop is a 1500KW facility consisting of a coolant system and auxiliary systems the other, operating in a wide temperature range of 250–480 °C, oxygen concentration 0.01–0.05 ppm. The coolant system is to remove the heat from the fuel surface due to irradiation. Auxiliary systems are for coolant chemistry control, emergency fuel cooling, polonium control, and parameter measurement. In this paper, the technological process, main design parameters, and system composition of the loop are described, and the parameters and selections of main equipment, instrument parameters are proposed. The investigation supplies a useful reference for further design of test irradiation loop systems and equipment.
Nuclear Cognition and Public Acceptance
Authors: Zhu Weiqiang , Wang Peng , Fu Changying
Published: November 2022

Nuclear power generation is the main form of human use of nuclear energy in the period of peaceful development. Compared with the most widely used thermal power generation, nuclear power generation is safe, efficient and clean. Nuclear power generation is an inevitable choice for low-carbon development. However, the inherent radioactivity and leakage risk of nuclear power generation make the public full of doubts about its safety. This paper investigates the relevant content of nuclear power generation through the method of online questionnaire survey, and the survey results show that the risk of radioactivity and leakage in nuclear power generation is the main problem point of public concern, the more fully the public understands nuclear energy, the more scientific and accurate the understanding of the risks of nuclear power generation, and the greater the possibility of showing an optimistic attitude towards the development prospects of nuclear energy. Similarly, raising people’s nuclear awareness through publicity methods such as knowledge popularization, organizing visits, and online media can significantly increase the acceptance of nuclear power generation by the surrounding people. From the historical perspective of the development of human society, nuclear energy as a promising clean energy, on the basis of the continuous innovation of global nuclear technology, coupled with the continuous improvement of the public’s scientific literacy, the general public can better understand nuclear energy, accept the use of nuclear energy, nuclear power generation will also make greater contributions to global environmental protection and green development.
Research on Wastewater Source Term in High-Temperature Gas-Cooled Reactor and the Environmental Impact at Inland Sites
Authors: Zhang Yichun , Li Chuan , Li Hong , Cao Jianzhu
Published: November 2022

High-temperature gas-cooled reactor (HTGR) is paid more attention in terms of inherent safety, high efficiency, and less liquid radioactive waste. The research on the source term of wastewater and the discharge methods of liquid effluent from HTGR can not only reduce the possible radiation risk to the environment and the public, but also promote the construction of inland nuclear power plants (NPPs), and improve the public acceptance of inland nuclear facilities. In this paper, calculation of activated corrosion products (ACPs) in the cooling water systems have been done in HTR-PM (High-Temperature Gas-Cooled Reactor Pebble-Bed Module) under normal operation conditions. Besides, the discharge methods of liquid effluent from inland NPPs without receiving water or with poor dilution conditions are studied. The calculation results indicate the cooling water of the heat removal system and shielded and component cooling water system needs to be changed once during the 40-year lifespan. The cooling water of the reactor support cooling system does not need to be changed. The annual effective dose resulting from liquid effluent through liquid-to-gas discharge using the data from a typical inland NPP site in China is less than direct discharge into the receiving water. The liquid-to-gas discharge is suitable for inland NPPs without receiving water or with poor dilution conditions.
Numerical Simulation of the Effect of Mixed Gas Temperature and Steam Injection Rate on Hydrogen Distribution at Steady State in Small Volumes
Authors: Feng Yu , Wang Hongliang , Yu Mingrui , Yuan Yidan
Published: November 2022

Hydrogen may be released into the containment atmosphere of a nuclear power plant during a severe accident. Locally, high hydrogen volume fraction can be reached that can possibly cause fast deflagration or even detonation and put the integrity of the containment at risk. The measurement accuracy of hydrogen volume fraction in containment is important to judge the flammability of hydrogen in the containment during a severe accident. The experimental device for hydrogen distribution and hydrogen volume fraction measurement at steady state in small volumes will be set up in the future, it is necessary to study on the effect of mixed gas temperature and steam injection rate on hydrogen distribution at steady state in small volumes.

Computational fluid dynamics (CFD) codes can be applied to predict hydrogen distribution in compartments at steady state in the containment during a severe accident. The pressure vessel is used to provide the same mixed gas as in the containment during a severe accident, and the numerical simulation is carried out for different work conditions with hydrogen-nitrogen mixture of 20% hydrogen volume fraction in the pressure vessel, investigating the effect of mixed gas temperature and steam injection rate on hydrogen distribution at steady state in small volumes. The numerical simulation results are qualitatively analyzed in the paper, the following conclusions indicate that it is no obvious stratification of hydrogen at steady state in small volumes at the work conditions (0–100, 0–120, 0–140, 0–160) with different mixed gas temperatures and it is no obvious stratification of hydrogen at steady state in small volumes at the work conditions (0.1–160, 1–160) with different steam injection rates. Therefore, mixed gas temperature and steam injection rate have almost no effect on hydrogen distribution at steady state in small volumes.

The numerical simulation results can provide some data basis for the experimental device for hydrogen distribution and hydrogen volume fraction measurement at steady state in small volumes in the future, and the outcomes of the qualitatively analysis can be used for design and optimization of the experimental device, in order to develop a hydrogen volume fraction measurement system for the containment during a severe accident.
Investigation of Model-Based Anomaly Detection Method for the Full-Power Natural Circulation Small Modular Reactor NHR200-II
Authors: Huang Canxing , Xiong Huasheng , Feng Junting
Published: November 2022

The NHR200-II is a full-power range natural circulation small modular heating reactor characterized with inherent safety. The significant difference between NHR200-II and a general pressurized water reactor is that the load of NHR200-II is relatively smaller, and it could vary rapidly in a wider range under different operation conditions. The control system of NHR200-II is required to sustain the stability of reactor operating parameters and make sure the reactor follows the load change in time.

Anomaly conditions may occur during the operation of NHR200-II. They may not induce safety-related accidents that gravely impair the reactor core or lead to the release of large-scale radioactivity, but they will fluctuate the power output. Therefore, it is practically valuable to detect abnormal working conditions in time. Furthermore, the detected information of anomaly is helpful for operators to conduct the reactor operation.

A model-based anomaly detection method to diagnose abnormal conditions of circulation pump sensor in NHR200-II is investigated in this paper and verified with the NHR200-II model. The simulation results show that the circulation pump sensor fault will cause the variation of running parameters, including inlet and outlet temperature of the coolant inside the reactor core, nuclear power, and output steam flowrate from the steam generator. The variation of parameters will lead to system deviation from expected operating conditions and system stability degradation. When the proposed anomaly detection method is applied, the circulation pump sensor fault could be successfully diagnosed and detected, and a fault-tolerant control strategy is deployed to correct the deviations and stabilize the operation of the NHR200-II reactor.
A Discrete-Ordinates Variational Nodal Method for Solving Multi-Dimensional Neutron Transport Equation With Unstructured Mesh
Authors: Sun Qizheng , Zhang Tengfei , Liu Xiaojing , Chai Xiang , Xiong Jinbiao
Published: November 2022

The variational nodal method (VNM) is an advanced nodal method in the field of reactor physics. The VNM solves neutron transport equations directly without transverse leakage interpolations used in other nodal methods. Moreover, the VNM can tackle arbitrary expansion orders. Given its benefits, VNM performs well in a variety of applications. However, with the advancement of nuclear science and technology, the ability to handle complex unstructured problems becomes indispensable.

In order to increase the geometric compatibility of the VNM, an unstructured-mesh variational nodal method (VNM) for solving multi-dimensional neutron transport equations is presented in this paper. The neutron transport equation is transformed into a variational formulation, in which an even-parity Lagrange multipliers enforce neutron conservation in each node, to obtain the weak-form solution. The entire problem domain is spatially meshed into several unstructured nodes. And the neutron flux, source, and current are spatially approximated by a set of complete orthogonal polynomials. The conception of the unique node is adopted to reduce the memory and improve efficiency. Orthogonal polynomials are established in a standard node for spatial discretization, and each unstructured node is related to the standard node by the coordinate mapping matrix, respectively. The discrete-ordinates (SN) method is employed to deal with angle variables so that the original equation is decoupled into several equations with different discretization directions. Then, the variational formulation is completely discretized, and the response equations are derived. Finally, different types of benchmarks are employed to validate and evaluate the proposed method. Numerical results demonstrate that unstructured-mesh VNM exhibits significant geometric compatibility and comparable accuracy.
Research on Smart Nuclear Power Construction Based on Digitalization and System Engineering in TNPS
Authors: Yang Qiang
Published: November 2022

In order to solve the digital transformation problems faced by Jiangsu Nuclear Power Station (TNPS) under the complex situation, and to benchmark advanced enterprises to improve the current digitalization level, TNPS established a special organization to carry out the V-type system construction method which covers user needs, functional requirements, architecture design, detailed design, code implementation, unit testing, integration testing, system testing, acceptance testing, etc. TNPS is carrying out the top-level design and application implementation of intelligent station, support platform and network construction and existing information systems based on the foundation functional and digital transformation, smart application and decision-making platform construction as the main body of the smart power plant implementation plan, and combed the application scenarios from the four dimensions of “personnel, equipment, environment, and operation”, and introduced smart nuclear power using smart construction sites and smart outage as examples Implementation content and solutions for typical applications. This work has advanced design and complete system, which has certain reference significance for nuclear power enterprises to carry out digital and intelligent transformation.
The Research on the Risk Monitor With Specific Plant Configuration at Shutdown Mode for VVER Nuclear Reactor
Authors: Dong Fangyu , Du Jinyan , Ma Chao , Deng Wei , Xu Qingqing , Zhang Rui
Published: November 2022

The research and application of Risk Monitor is one of the most influential analysis tool in the field of probabilistic safety analysis (PSA) of nuclear power plants. Risk Monitor is often used in nuclear power plant operation risk management, scheduled maintenance risk management or other plant activities, aims to help the formulation of power plant integrated decision-making from the risk-informed perspective. The VVER nuclear reactors have four redundant trains to achieve relatively higher safety level, compared to other nuclear power plants design. Because of its four redundant systems design, the configurations of certain system, particularly for the shutdown mode are various. This paper takes the VVER residual heat removal system (JNA) as an example and discusses the method of developing the average PSA model into a risk monitor PSA model for VVER JNA at shutdown mode. Meanwhile, this paper establishes the cases of different JNA setting configurations, aims to support the plant configuration risk management analysis from the perspective of risk level.
UTSG Water Level Prediction Model Based on LSTM and GRU Algorithms
Authors: He Qingyun , Li Bin , Peng Sitao
Published: November 2022

Steam generator water level is a complex thermal hydraulic system, which in-ludes nonlinearity, time-variant characteristic, non-minimun phase, mechanical limit and mesurement error. Particularly, “false water level” phenomenon caused by nonminimun phase greatly affects the control of steam generator water level. Thus, in order to greatly help PWRs operators control the water level, enhance the safety of PWRs operation and improve the economic of PWRs, it is of importance to study the technology on prediction of steam generator water level variation trend under normal operation condition of PWRs. A predictive model for U-tube steam generator (UTSG) is in need as an assistance for operators to better control the UTSG water level. The training data, i.e., the variation of UTSG water level in uenced by the sudden change of steam ow rate and feedwater rate, were obtained from a differential equation model, i.e., Ivring’s model. Long short-term memory (LSTM) and gated recurrent unit (GRU), as powerful time series prediction algorithms in deep learning eld, were used to train the UTSG prediction model. The result shows that the trained LSTM and GRU based predictive model have good performance on testing data set with both the one-step prediction and multi-step recurrent prediction. It can be used as a prototype of steam generator level prediction without the intervention of automatic control system.
Thermal Performance Monitoring and Analysis of Nuclear Power Plant
Authors: Hou Jianfei , Zhang Xun , Zimmermann Heino , Xiang Zhaohui
Published: November 2022

In order to avoid sudden fault and damage of equipment which will affect reliability and safety of the nuclear power plant, it is necessary to monitor the actual thermal performance of the nuclear power plant and analyse the thermal performance trends of thermal circuits and important equipment which is the basis of fault predicts. Based on the mature and widely applied EBSILON software, thermodynamic model of a nuclear power plant in operation is built including the primary loop, second and third loop and important equipment such as steam turbines, condenser, deaerator, low-pressure and high-pressure heaters and main feed water pumps. In particular, hundreds of instruments are taken into account in the thermodynamic model and all the measured data during operation of the nuclear power plant is imported into the instruments. Thermal performance of the nuclear power plant is evaluated and analyzed in different seasons and results displays the performance factors of important equipment compared to the design performance. In addition, a data reconciliation calculation is taken out to check the correctness of measurements using the massive redundant measured data of the plant and results shows some instruments may have potential measurement fault. Current work lays a foundation for the future research of fault predicts and diagnoses of the nuclear power plant.
Comparative Study on Internal Flooding Prevention Design of Several Types of NPPs and Improvement Suggestions
Authors: Yinghao Yang , Zhaohua Li , Xiaoyan Xu
Published: November 2022

There are many tapes of nuclear power plants (NPPs) in operation and under construction in China, involving M310 reactor, VVER reactor, AP1000 reactor and HPR1000 reactor and so on. The design concepts of each type nuclear power plant are quite different, including active nuclear power plant and passive nuclear power plant, second-generation reactor and third-generation reactor. Their designs for preventing internal flooding are also very different, either through physical isolation or by adding flood prevention barrier, but each reactor type nuclear power plant has its own advantages in flood prevention. In order to further improve the internal flooding prevention ability of nuclear power plants in China, this paper makes a comparative study on the internal flooding prevention design of several mainstream tapes of nuclear power plants in China, and gives the advantages of each nuclear power plant in the internal flooding prevention design. By drawing on the design advantages of other reactor type nuclear power plants and using the analysis method of probabilistic safety assessment (PSA), this paper quantitatively evaluates the improvement of the ability of reactor type nuclear power plant that our company participated in the design or construction to resist internal flooding risk after considering these design optimization, and gives targeted design improvement suggestions.
The Crown Predictive Model of Cold-Rolled Zirconium Alloy Strip Sheet Based on Machine Learning Algorithm
Authors: Cao Yuan , Cao Jianguo , Gao Yingqi , Wang Ben , Zhang Pengfei , Li Fang , Gao Bo
Published: November 2022

To solve the problem that the crown accuracy directly affects the profile and flatness of the zirconium alloy strip caused by the multi-schedule and more than sixty passes cold rolling of nuclear zirconium alloy strip. The crown predictive model is established based on support vector regress algorithm (SVR), and the parameters of the SVR algorithm are optimized by the SAPSO algorithm. Meanwhile, the crown predictive model based on the SAPSO-SVR algorithm shows that the relative coefficient of predicted values is higher than 0.94. Industry test demonstrate that the crown predictive model provides a new method for zirconium alloy strip shape control and an approach to optimize the control strategy. The proposed model provides a new method and idea for shape control and optimization research in the zirconium alloy plane rolling process.
Level 1 Probabilistic Safety Analysis (PSA) of TNPP Units 3 and 4
Authors: Yu Wenge , Li Youyi , Lang Xiye , Sun Yang , Zhao Pengfei
Published: November 2022

Tianwan Nuclear Power Plant (TNPP) Units 3&4 (the second phase) are a landmark project for Chinese and Russian governments to jointly promote China-Russia nuclear energy cooperation.All the reactor units adopt the Russian VVER-1000 improved nuclear power units which are now operated by project owner, Jiangsu Nuclear Power Corporation (JNPC).

The depth and quality of PSA development, at the stage of Units 3&4 commercial operation,should be sufficient to obtain the risk-informed insights, enough to support the Risk-Informed application of PSA, and also reflect the actual design and its operation experience of TNPP.

This paper presents the brief introduction of PSA of Unit 3&4 of TNPP (power operation mode). The important parts of this presentation are as follows:probabilistic safety indices, study scope, technical elements and the analysis of Core Damage Frequency (CDF) quantitative results.

Through quantitative calculation, the point estimation value of the total CDF of TNPP under the power condition is obtained. And the frequency of occurrence of each event sequence and the CDF contribution of each initiating event (IE) etc., are analyzed.

Based on the developed PSA models, four PSA application tools have been developed and applied in the risk monitoring activities of nuclear power plant. This PSA model is used to take the lead in carrying out the Online maintenance in China.
Parameter Calculation Method of Porous Media Based on BP Neural Network
Authors: Zhao Fei , Lu Daogang , Liu Yu , Liu Dong , Xu Jiliang , Wu Jinghui , Xie Fei , Wang Yuchao
Published: November 2022

There are a large number of equipment densely arranged in liquids in nuclear power plants, such as fuel assemblies, steam generator heat transfer tubes, spent fuel storage and transportation racks, etc. These equipment are complex in shape and compact in arrangement and have strong fluid-structure coupling effects under excitations, so the calculation is computationally intensive. For such complex structures, the use of porous media models is an important means of structure simplification. The parameters of porous media are often calculated by CFD modeling, and the calculation process is complicated and time-consuming. BP neural network has strong nonlinear mapping capability and can be used to calculate the parameters of porous media. For different racks designs, the gap arrangement is different, and the fluid-structure coupling parameters are also different. Therefore, it is necessary to study the fluid-structure coupling parameters of square tube bundles such as racks. Taking porous storage racks as an example, by building different CFD models, 1366 sets of valid data were obtained for training. This paper uses BP neural network to study the porous medium parameters required for fluid-structure interaction of porous racks. Compared with the CFD calculation method of fine modeling, the calculation error of the additional mass of the porous media model established by the porous media parameters predicted by the neural network is controlled at about 10%. The research results provide a reference for the fast calculation of porous media parameters and fluid-structure interaction.
A Physics-Driven X-Ray Image Data Augmentation Method for Automated Threat Detection in Nuclear Facility Entrancement
Authors: Xu Shuo , Chen Gang , Li Weiwei , Xiang Xincheng
Published: November 2022

Due to the perspective ability of the internal structure of objects, the X-ray security screening system has been used to ensure nuclear security in nuclear facility entrancement. To improve detection efficiency and accuracy, recent work posed a particular interest in automated threat detection algorithms based on deep learning. A significant obstacle to developing high performance of such algorithms is the difficulty of obtaining large labeled training datasets. X-ray image data augmentation strategies based on experimental images have been proposed by other investigators. This paper proposed a physics-driven X-ray image data augmentation method for automated threat detection. Using the 3D modeling software, we can construct threat models and project them into threat images without experiment. According to Lambert-Beer law, the simulation image datasets can be made by exponentially overlaying the threat images on the background images. Considering the energy spectrum and scattering effect, we further process the phantoms and projection images accordingly to be more authentic. We train YOLOv5 architecture on the simulation dataset and test the algorithm on the experimental images. The results show that our approach achieves good performances in automated threat detection with an average recognition accuracy of over 90% and a mAP@0.5 of 82.9%, which is an effective method to increase the X-ray image dataset in both quantity and diversity.
Development of Adaptive Sampling Software Based on Relap5
Authors: Chen Haoyin , Wang He , Suliman Mohamedelmogtabh Omer Elfadni , Wang Xinyue , Zhao Qiang
Published: November 2022

The RISMC analysis method is a coupled dynamic safety analysis method that combines deterministic safety analysis and probabilistic safety analysis. It comprehensively simulates the dynamic response process of nuclear power plants under accidents through Monte Carlo sampling and dynamic event tree methods. In order to fully reflect the response process of the nuclear power plant and ensure the accuracy of the calculation results, it is necessary to sample a large number of input space parameters, which will result in unbearable time costs. Adaptive sampling can predict the state of the position input space parameters through a small number of forward sampling calculation results and divide the input space by generating limit surfaces to calculate the failure probability, which can greatly save computing time. Based on researching RISMC analysis tools at home and abroad, the research and development of adaptive sampling and RELAP5 coupling calculation software is realized. The test case of the Qinshan nuclear power plant station blackout accident shows that the adaptive sampling function of calculating failure space probability through a small amount of sampling is realized, and the efficiency of RISMC analysis and calculation is improved.
Simulation on the Process of Core Molten Pool Formation and Corium Relocation in the Core Region
Authors: Zhao Xinhai , Chen Peng , He Dongyu , Zhong Ruhao
Published: November 2022

Accurately estimating the moment on which the core melt relocates to the lower head in severe accidents is crucial for evaluating the effectiveness of the In-Vessel Retention (IVR) strategy. However, the collapse and fragmentation of the fuel rods as well as the melting and re-solidification of the core material during the core degradation process lead to complex changes in the core geometry, making it difficult to accurately simulate the relocation process of the melt in the core. Based on TMI-2 and Phebus FPT series experiments, a porous media model with transient flow and heat transfer both in the rod bundle region and debris bed region was established in this paper. The VOF and LEVEL SET methods were used to track the movement front of the corium in the core, as well as the coalescence process of local separate molten pools to form a large-scale core molten pool. Moreover, the efficient and stable IDEAL segregated algorithm was adopted to deal with the coupling problem of flow and heat transfer. The VOF and LEVEL SET methods were verified by the dam break problem, and the deviation of the results is within 10%. Using the above methods, the formation of a large-scale molten pool in the core and the relocation process of corium in the core region were simulated.
Fission Gas Release Grain Boundary Network Percolation Mechanistic Studies in Oxide Fuels Based on COMSOL Multiphysics Framework
Authors: Guo Jingyu , Zhou Wenzhong
Published: November 2022

Fission gas release was modeled by COMSOL Multiphysics in oxide fuels, whose theory is based on the effect of grain boundary network percolation. The model is contributed to the conventional Booth model. In this model, the fuel pellet microstructure adopts 2D axisymmetric geometry. The effect of different bubble growth and coalescence rates on the independent grain boundaries are investigated, such as bubble contact angle, resolution rate, and radial position. The modeled physical phenomena are similar to the existing experiments and can be observed in the fission gas release process. The fission gas release is sensitive to the variations of these parameters. Therefore, the parameters are essential to the fission gas release on the microscopic or macroscopic scales. The long-range percolation on the networked grain boundaries is also considered in this work, but it is not considered in the Booth model. The gas resolution on the grain boundaries is also taken into account. At last, this model’s results will be compared with the outcomes of the Booth model as well as the other models.
Research on Risk Characterization Method of Marine Biological Disaster Affecting Water Intake Safety of Nuclear Power Plant and Application of Design Protection
Authors: Fan Yu , Jie Bai , Botao Xu , Xingbing Lv , Yong Li , Xueqing Zhang
Published: November 2022

Marine biological outbreaks in coastal waters of coastal nuclear power plants have caused abnormal water intake systems of nuclear power plants for many times, and the risk Valuation method of marine biological outbreaks to nuclear power safety is still in the research and exploration stage. For nuclear power plants, the outbreak of marine organisms near the water intake is a natural event that affects the safety of nuclear power. According to the basic framework of “four-step method” on risk Valuation of National Academy of Sciences of the United States, aiming at the disaster risk caused by marine organisms that potentially affects the reliability of cold sources of nuclear power plants, the risk characterization method is studied, and the design protection measures are put forward.
Key Impact Factors Analysis of Fuel Rod End Plug Welding Based on QFD
Authors: Li Yinghong , Lv Zhipeng , Li Changzheng , Lin Yuyu , Lin Shaofang , Cai Zhiyi , Deng Yongjun , Nie Lihong
Published: November 2022

Manufacturing defects of fuel rod end plug during welding affect fuel reliability. Failure of ring welding or sealing welding of fuel rod end plug during operation will cause fission gas to enter primary coolant of reactor. In order to determine the sensitivity of the influence of relevant characteristic parameters on the performance of fuel rods during end plug welding, based on the Quality Function Deployment method (QFD) and taking tungsten-inert-gas arc welding (TIG) as an example, the characteristic parameters of fuel rod end plug welding are analyzed. Combined with the experimental results, a correlation matrix between welding parameters and welding performance is constructed by LIKERT scale method with focus on the main welding performance criteria such as welding joint size, color, depth of fusion, metallography (bloating, porosity, inclusion, grain boundary separation, etc.), oxidation resistance, etc. And after rating its correlation degree and importance degree, the relevant House of Quality of end plug welding is constructed, and the key impact parameters are identified. Based on the above analysis, some feasible optimization and improvement proposals are put forward for the quality supervision and welding process of fuel element manufacturing factory.
Explainable Neural Network Algorithm for Rapid Radionuclide Identification Under Low Count Gamma-Ray Spectrum Data
Authors: Wang Yu , Li Sufen , Huo Yonggang , Yang Jianqing , Zhang Quanhu
Published: November 2022

Real-time and automatic radioisotope identification using gamma spectrum is an important issue in the field of nuclear safety. It is widely used in vehicle mounted radioisotope monitoring, Marine radioisotope monitoring and nuclear decommissioning verification scenarios. At present, the focus of radionuclide identification is fast and stable recognition under low count conditions. In this paper, a radionuclide recognition method based explainable artificial neural network is proposed, and a synthetic gamma spectrum data set is created. The data set contains gamma-ray spectra of 12 different types of radionuclides, which were obtained by Geant4 Monte Carlo simulation software and gaussian broadening of the detector. Data Augmentation was achieved by simulating gamma spectra at different measuring times, different measuring distances and different ambient temperatures. The training results of the neural network optimized by hyperparameter show that it has a high accuracy on the test set with shorter measurement time, longer measurement distance and larger energy spectrum drift range, which provides a method for rapid identification of nuclides in the case of low count. Using t-SNE dimension reduction technology, the twelve dimensions data output by the neural network is reduced to two dimensions for feature visualization, which vividly explains and verifies the recognition results of neural network.