bluuebunny/crossref_metadata_2025_split huggingface dataset:
Source and Concentration of Radionuclides by Inland Nuclear Power Plant Under Normal Operation
Authors: Wang Jiaxin , Zhang Liuguo 
Published: October 2021

Abstract
Freshwater, which is the receiving water of Inland Nuclear Power Plant (INPP), has a lower environment capacity and higher sensitivity to radioactivity comparing with the receiving sea water of Coastal Nuclear Power Plant (CNPP). Focus on the radionuclides in freshwater under INPP normal operating condition, this paper proposed four source items: direct liquid release item, wet deposition item, dry deposition item, and the second run-off item, and also developed their calculation models. These models were then applied to Xianning INPP and its receiving water Fushui reservoir in China. The downstream 1km and the dam were set as sensitive points to simulate the concentration of 137Cs and other effluent radionuclides. Results showed that the direct liquid release item was the main source item under INPP normal operation, 137Cs had a good representative for the variation of total radionuclides concentration with downstream distance. 137Cs concentration at downstream 1km, at the dam, were 1.89 × 101 Bq m−3, 8.04 Bq m−3, respectively; the total radioactive concentration at 1km, at the dam, were 1.65 × 102 Bq m−3, 4.73 × 101 Bq m−3, respectively. Compared with the current two radioactivity limits in freshwater: GB 18871 Food general action level-137Cs, and 1 km downstream total β concentration, the radioactive concentrations under INPP normal operation in this case were all acceptable.
Analytical Study on Removal Mechanisms of Cesium Aerosol From a Noble Gas Bubble Rising Through Liquid Sodium Pool (II) Effects of Particle Size Distribution and Agglomeration in Aerosols
Authors: Miyahara Shinya , Kawaguchi Munemichi , Seino Hiroshi , Atsumi Takuto , Uno Masayoshi 
Published: October 2021

Abstract
In a postulated accident of fuel pin failure of sodium cooled fast reactor, a fission product cesium will be released from the failed pin as an aerosol such as cesium iodide and/or cesium oxide together with a fission product noble gas such as xenon and krypton. As the result, the xenon and krypton released with cesium aerosol into the sodium coolant as bubbles have an influence on the removal of cesium aerosol by the sodium pool in a period of bubble rising to the pool surface. In this study, cesium aerosol removal behavior due to inertial deposition, sedimentation and diffusion from a noble gas bubble rising through liquid sodium pool was analyzed by a computer program named AESOP (AErosol scrubbing in SOdium Pool) which deals with the expansion and the deformation of the bubble together with the aerosol absorption considering the effects of particle size distribution and agglomeration in aerosols. In the analysis, initial bubble diameter, sodium pool depth and temperature, aerosol particle diameter and density, initial aerosol concentration in the bubble were changed as parameter, and the results for the sensitivities of these parameters on decontamination factor (DF) of cesium aerosol were compared with the results of the previous study in which the effects of particle size distribution and agglomeration in aerosols were not considered. From the results, it was concluded that the sensitivities of initial bubble diameter, the aerosol particle diameter and density to the DF became significant due to the inertial deposition of agglomerated aerosols.
To validate these analysis results, the simulation experiments have been conducted using a simulant particles of cesium aerosol under the condition of room temperature in water pool and air bubble systems. The experimental results will be compared with the analysis results calculated under the same condition in the near future.
Application for 3D Laser Scanning During Construction Stage of Nuclear Power Project
Authors: He Wei-ting , Jiang Wei-feng , Zhou Yi-Kun 
Published: October 2021

Abstract
3D laser scanning technology overcomes the limitation of single point measurement in traditional measurement methods, which can convert the 3D information of objects and site environmental into data processed by computers. The 3D laser scanning technology is usually called restoration of real scenes technology, which has been widely used in automobile, architecture, terrain exploration, petro and other industries. During the construction stage of nuclear power project, the 3D laser scanning technology can be used to get the real environment information on site. Through comparative analysis of the real environment information and the design model environment, risks can be easily identified in advance and mitigation measures can be performed in time, thus effectively avoiding the rework and saving the cost of the project.
Progress and Prospects of Some Key Technologies for Space Nuclear Reactor
Authors: Zhuang Nailiang , Yang Chenhao , Zhao Yiqi , Song Yongnian , Tang Xiaobin 
Published: October 2021

Abstract
With the increasing demands of versatile and long-lasting requirements of deep space missions, space nuclear reactor (SNR) power system is becoming the most potential energy source compared to conventional solar-battery and chemical energy in the future large-scale and long-life space missions. Since the very first successful launch of SNR system (SNAP-10A) in 1965, the United States, the Soviet Union / Russia, the European Union, Japan, Brazil and China have proposed various technical routes and schemes for SNR technologies over the past few decades. This paper presents a historical review of several key technologies (i.e., space nuclear reactor system and thermoelectric conversion system), especially focus on technical progress for recent decade and discusses on-going development activities. The paper also presents a forecast of potential future space applications of these key technologies.
Effective Solid Angle Model and Monte Carlo Method: Improved Estimations to Measure Cosmic Muon Intensity at Sea Level in All Zenith Angles
Authors: Bae Junghyun , Chatzidakis Stylianos , Bean Robert 
Published: October 2021

Abstract
Cosmic muons are highly energetic and penetrative particles and these figures are used for imaging of large and dense objects such as spent nuclear fuels in casks and special nuclear materials in cargo. Cosmic muon intensity depends on the incident angle (zenith angle, φ), and it is known that I(φ) = I0 cos2 φ at sea level. Low intensity of cosmic muon requires long measurement time to acquire statistically meaningful counts. Therefore, high-energy particle simulations e.g., GEANT4, are often used to guide measurement studies. However, the measurable cosmic muon count rate changes upon detector geometry and configuration. Here we develop an “effective solid angle” model to estimate experimental results more accurately than the simple cosine-squared model. We show that the cosine-squared model has large error at high zenith angles (φ ≥ 60°), whereas our model provides improved estimations at all zenith angles. We anticipate our model will enhance the ability to estimate actual measurable cosmic muon count rates in muon imaging applications by reducing the gap between simulation and measurement results. This will increase the value of modeling results and improve the quality of experiments and applications in muon detection and imaging.
Research on Grey Correlation of Factors Influencing Particulate Matter Concentration of Supercritical Water Reactor
Authors: Cheng Hu , Tao Zhou , Juan Chen , Ning Chen , Xijia Ding , Xiaolu Fang 
Published: October 2021

Abstract
Supercritical water reactors will continuously produce fine particulate matters during operation and accident conditions, and the deposition and movement of these particulate matters are likely to cause deterioration of heat transfer, which will affect the heat transfer and safe operation of the reactor. The grey correlation analysis method is used to calculate and analyze the particulate matter concentration distribution at different heights of the supercritical water reactor coolant channel model. The correlation of coolant temperature, axial velocity and turbulent kinetic energy to the concentration distribution of particulate matters and their respective weights in the correlation were studied. Using grey correlation to calculate, Finally, it was obtained that when the coolant channel height of the supercritical water reactor was 0.3m, 0.5m and 0.7m, the correlation order of the concentration distribution of particulate matters among the three was: coolant temperature &gt; coolant turbulent kinetic energy &gt; coolant axial velocity. The coolant temperature always has the relatively greatest influence, while the influence of the coolant turbulent kinetic energy and the axial velocity of the coolant is relatively small and tends to be consistent; The three influencing factors have little difference in the overall impact of the particulate matter concentration distribution, and the maximum weight difference among the three was 7.22%.
Equal Forced Time Step Approach to PSA for a Dynamic System – A Case of the Holdup Tank
Authors: Taapopi E. E. , Wang H. , Zhou J. 
Published: October 2021

Abstract
This work presents a methodology of the equal forced time step for generating accident scenarios in a time-dependent dynamic sequence. The aim is to model DPSA and system reliability of a simple hold-up tank which is a good representative of a steam generator — a component in the secondary loop of a nuclear power plant. This was achieved based on some predicted theoretical calculations on node generating conditions and the cumulative probability of partial failure mode. The computer language Python 3.0 was employed to establish a dynamic event tree of 20 equal time steps (h), the total number of 7367 nodes, 4882 failed nodes for various types of potential accidents. Also, various types of failures and the cumulative probabilities of 0.13 were generated. The program modeled cumulative probabilities and associated errors. However, there is a need to improve and refine the methodology to consider physical parameters such as temperature, pressure, etc. as these are factors that can lead to calculation termination. This will boost the program efficiency and utilize the information extracted from the dynamic event tree in the analysis of the dynamic characteristics. Although extra efforts and improvement still be needed to refine produced accident scenario results, the presented methodology has satisfactorily handled the example.
Development of a Passive Reactor Shutdown Device for Prevention of Core Disruptive Accidents in Fast Reactors: Project Overview and Preliminary Results
Authors: Morita Koji , Liu Wei , Arima Tatsumi , Arita Yuji , Kawase Koharu , Sato Isamu , Matsuura Haruaki , Sekio Yoshihiro , Sagara Hiroshi , Kawashima Masatoshi 
Published: October 2021

Abstract
After the Fukushima NPP accident in 2011, it has become increasingly important for reactor safety design to consider design measures to prevent the occurrence of severe accidents. In this study, a new subassembly-type passive reactor shutdown device is proposed to enhance the diversity and robustness of prevention measures for core disruptive accidents in sodium-cooled fast reactors. The proposed device contains pins with a fuel material that is kept in the solid state during normal operation but melts into the liquid when its temperature exceeds a prescribed value under the accidents. When a ULOF (unprotected loss of flow) or UTOP (unprotected transient overpower) accident occurs, the device can provide large negative reactivity passively by the relocation of liquefied device fuel into the lower plenum region of the pins by gravitation alone in a short time. The reactor, in which part of usual fuel subassemblies is replaced with the proposed devices, becomes subcritical before driver fuels are damaged, even if ULOF or UTOP transient occurs. In the present study, candidate materials for device fuel such as metallic alloy and chloride, optimum device pin structure for liquefied fuel relocation, and nuclear and thermal-hydraulic characteristics of the device-loaded core under accident conditions will be mainly investigated to demonstrate engineering feasibility of the proposed device. This paper describes the project overview and discusses preliminary results on nuclear requirements for negative reactivity to be inserted for reactor shutdown under expected device conditions.
Apros-Based Loviisa NPP Full Scope Training Simulator and Engineering Model
Authors: Meriläinen Arttu , Viljakainen Olli , Honkoila Karri , Lahtela Ari 
Published: October 2021

Abstract
Loviisa NPP (VVER-440) simulator LOKS2 is based on parallel computed Apros models that simulate more than 50 process systems including reactor, primary circuit, secondary circuit and containment. The models contain relevant electrical systems including external grid connections, emergency diesel generators and house load network. Detailed plant automation models enable connectivity to new automation systems which can be provided by suppliers as external models (virtual I&C).
Models and virtual I&C are connected to touch screen panels which emulate HW panels at the plant. All renewed HMI systems are identical to the plant. Severe accident management panels can be utilized since the integrated severe accident module Apros SA enables simulating the most important severe accident phenomena in real time.
In addition to operator training, the models are utilized in automation renewal projects and part-task simulators. The overall functionality of new automation has been assessed efficiently when connected to the simulated plant. The detailed process models enable running transients and testing degraded modes with plant operators on simulators. Errors in automation system design or implementation are discovered in an early phase and software and hardware modifications based on simulator findings are made. Part-task simulators utilize the same models in specific training.
Application of Probabilistic Safety Analysis for Nuclear Power Plant Overhaul Risk Assessment
Authors: Liu Deyi , Cao Yong , Zhao Ming , Luo Yang , Zou Shengjia , Hu Mengying , Xu Jie , Wang Li , Zhao Ling 
Published: October 2021

Abstract
At present, probabilistic safety analysis (PSA) is widely used in the many fields of nuclear power plant (NPP), such as design, manufacture and etc. And PSA has gradually developed into the important tool of safety evaluation and decision making. With the development and improvement of PSA technology, the risk-informed safety management has also made great progress. The overhaul of NPP is an important work to keep safe and stable operation of NPP. During the overhaul, fuel replacement and maintenance of a large number of equipment will be carried out. Under different plant operation status (POS) in the overhaul process, several important equipment will exit operation at the same time, which will lead to the increase of core loss frequency of different degrees. At present, only few NPPs within the scope of China National Nuclear Power Co., Ltd (CNNP) has adopt qualitative PSA method to carry out the risk assessment of the overhaul, but there is no precedent to use quantitative PSA method to carry out the assessment. The paper will introduce how to use the risk-informed PSA method to assess the risk of the overhaul in a quantitative way. The higher risk stages in the overhaul will be identified through the assessment to optimize the work schedule of the overhaul to ensure that the work schedule will not lead to excessive increase in core damage frequency (CDF). And taking the fifth overhaul of unit one of Fangjiashan NPP (FJS NPP) (105 overhaul) as an example, to assess the work schedule of 105 overhaul and give reasonable Suggestions.
A Finnish District Heating Reactor: Background and General Overview
Authors: Leppänen Jaakko , Hillberg Seppo , Hovi Ville , Komu Rebekka , Kurki Joona , Lauranto Unna , Oinonen Ahti , Peltonen Jussi , Rintala Antti , Tulkki Ville , Tuominen Riku , Valtavirta Ville 
Published: October 2021

Abstract
In 2019 the government of Finland made a decision to phase out of coal in energy production in a period of just ten years. The Finnish energy sector is currently looking for alternative technologies to replace coal-fired power plants, used especially in large cities for producing electricity and low-temperature heat for the local district heating network. The production of low-carbon electricity is expected to grow within the near future, along with the commissioning of the Olkiluoto 3 nuclear power plant and increasing share of wind power. The lost district heating capacity, however, is more difficult to replace. To anticipate the transition, municipal energy companies have turned their attention to clean alternatives, including nuclear energy.
In an effort to meet the government climate goals, VTT Technical Research Centre of Finland has launched a project to design a small, simplified and passively safe PWR for district heating applications. The heating plant consists of one or multiple 50 MW reactor modules, operating on natural circulation at around 120°C temperature. The design combines conventional LWR technology with an innovative containment function, capable of decay heat removal without any mechanical moving parts. The reactors can be constructed partially or fully underground, or retro-fitted into an existing boiler plant. This paper presents an overview of the pre-conceptual reactor design, together with some general background on district heating reactor technology. More detailed design and safety analyses are provided in two separate papers at this ICONE-28 conference.
Experimental Study on Boiling Heat Transfer Characteristics in an Inclined Tube Bundle
Authors: Li Zongkun , Cheng Jie , Zhou Xuwei , Zeng Xiaobo , Cao Xiaxin , Fan Guangming 
Published: October 2021

Abstract
Since the special structure of U-tube steam generator, so the secondary side fluid will receive non-uniform heating from the primary side. When U-tube steam generator is applied in ocean environments, it may be in an inclined condition, and the inclined condition will make the thermal-hydraulic characteristics in the secondary side of U-tube steam generator more complicated. In this paper, a series of experiments under different inclined angles were conducted to investigate the effect of inclination on the heat transfer characteristics in tube bundle channel. Besides, the coupling effect of non-uniform heating and inclination is considered. The results show that the inclined condition will lead to an angle between the hot-cold fluid interface and the fluid temperature measuring interface, and it can also generate a gravity component perpendicular to the flow direction, which can drive the fluid to flow laterally through the tube bundle. Both of them lead to the radial inhomogeneity of fluid temperature under uniform heating condition. Furthermore, non-uniform heating can also cause a radial unevenness of fluid temperature at the same level, and the trend of unevenness is determined by the heating power and the heating power ratio. The final results are shown in a coupling superposition effect of inclination and non-uniform heating, and it can be an enhancement effect or a weakening effect according to the direction of inclination.
Optimization of Active Magnetic Bearings’ Power Supply System for Main Helium Fan in High Temperature Gas-Cooled Reactor
Authors: Luo Huan , Shi Zhengang , Zhou Yan , Mo Ni 
Published: October 2021

Abstract
High temperature gas-cooled reactor (HTR) is a kind of reactor with inherent safety developed by Institute of Nuclear Energy and New Energy Technology of Tsinghua University. In the first circuit, pure helium is used as coolant and the main helium fan is used to promote the coolant circulation. In order to meet the requirements of service environment and performance, the main helium fan adopts the non-lubricant active magnetic bearing (AMB) system as its support system. For the high-speed rotating equipment supported by AMBs, losing power would lead to bearing failure and cause serious damage to the equipment. In this paper, the power supply system of AMBs is optimized. The power supply of AMB system is connected with the DC-link of the motor converter through DC/DC converter. During normal operation, the AMB system is supplied by external power supply, and the DC/DC converter is used as the backup redundant power supply. In the event of a power failure accident, the DC/DC converter is put into operation, converting the remanet kinetic energy of the motor into stable power to maintain the normal operation of the AMB system. The DC/DC converter adopts two-stage topology structure of the former BUCK converter and the later LLC converter, and completes the voltage stabilization control of the latter LLC converter through the digital signal processor (DSP). Experimental results show that this scheme can realize the power loss protection function of the rotating equipment supported by AMBs.
Design and Analysis of a Reliable Communication System in Nuclear Safety Instrument & Control System
Authors: Li Le , Zhang Zhihui , Gao Chao , Zhou Fei , Ma Guangqiang 
Published: October 2021

Abstract
With the development of digital instrument and control technology for nuclear power plants in recent decades, communication networks have become an important part of safety digital control systems, which takes charge in data exchange between the various sub-systems, and extremely impact on the reliability and safety of the entire I&C system. Traditional communication systems where some special features, such as reliability, safety, real-time, certainty, and independence are not strictly required are various illustrated. However, how to implement a communication system in a safety I&C system is rarely stated in current research. In this research, a reliable safety communication system applied in nuclear power plants is designed and analyzed. The five key characteristics of nuclear safety communication networks are explained, followed by explanation of how to achieve these characteristics. The analysis and verification of the designed system are also stated in this paper, which contributes to proving that the designed nuclear safety communication system could applied in the nuclear power plants.
Resistance and Thermal Stress Analysis of Mini-Flow Pipeline of Residual Heat Removal System in Pressurized Water Reactor
Authors: Pi Yue , Hou Ting 
Published: October 2021

Abstract
During the start-up and shutdown of Pressurized Water Reactor (PWR), the residual heat removal system is responsible for taking the decay heat out of the reactor core. The heat-up of the residual heat removal system before it is connected to the primary circuit is of great significance for the integrity and reliability of the residual heat removal system and the primary circuit. The mini-flow pipeline in the residual heat removal system is the main component of the system heat-up. Due to its small diameter and fast flow speed, the mini-flow pipeline bear all the pressure head loss under heat-up conditions. In this paper, ANSYS-FLUENT is used to model the mini-flow pipeline, and the resistance characteristics of the pipeline under heat-up condition are analyzed with the change of the flow rate. The results show that FLUENT can accurately predict the pressure drop of the mini-flow pipeline. The results provides the basis for the future optimization. Also, ANSYS is used to analyze the hot and cold fluid mixing between the mini-flow pipeline and the main pipeline and the thermal stress generated. The analysis shows that the stress zones, mainly concentrated in the tee pipe and its upstream, are obvious at the beginning of heat-up and negligible at the end.