bluuebunny/crossref_metadata_2025 huggingface dataset:
Female sexual dysfunction
Authors: Kershaw Victoria , Jha Swati
Published: October 2021

Key content

Female sexual dysfunction (FSD) is defined as any sexual complaint or problem resulting from disorders of desire, arousal, orgasm or sexual pain that causes marked distress or interpersonal difficulty.

Various gynaecological problems such as pelvic organ prolapse and endometriosis affect sexual function and this impacts on their treatment.

Management often requires the need to address different components of the sexual dysfunction.

Women wish to keep their options relating to sexual function open well past the menopause, and more women are presenting in later life with FSD.

Learning objectives

To know how to assess FSD in clinical practice.

To appreciate the reasons why all women attending a gynaecology clinic should have their sexual function assessed to establish a baseline.

To establish realistic expectations of outcomes on sexual function following treatment of gynaecological problems.

Ethical issues

Failure to inform women of the risk of deterioration in sexual function following gynaecological surgery can lead to litigation.

Some essential treatments, such as vulvectomy or prophylactic oophorectomy, will result in adverse effects on sexual function, and women should make informed choices regarding their treatment.
Genome structure and evolution in the cruciferous tribe Thlaspideae (Brassicaceae)
Authors: Bayat Soheila , Lysak Martin A. , Mandáková Terezie
Published: October 2021

Whole‐genome duplications (WGDs) and chromosome rearrangements (CRs) play the key role in driving the diversification and evolution of plant lineages. Although the direct link between WGDs and plant diversification is well documented, relatively few studies focus on the evolutionary significance of CRs. The cruciferous tribe Thlaspideae represents an ideal model system to address the role of large‐scale chromosome alterations in genome evolution, as most Thlaspideae species share the same diploid chromosome number (2n = 2x = 14). Here we constructed the genome structure in 12 Thlaspideae species, including field pennycress (Thlaspi arvense) and garlic mustard (Alliaria petiolata). We detected and precisely characterized genus‐ and species‐specific CRs, mostly pericentric inversions, as the main genome‐diversifying drivers in the tribe. We reconstructed the structure of seven chromosomes of an ancestral Thlaspideae genome, identified evolutionary stable chromosomes versus chromosomes prone to CRs, estimated the rate of CRs, and uncovered an allohexaploid origin of garlic mustard from diploid taxa closely related to A. petiolata and Parlatoria cakiloidea. Furthermore, we performed detailed bioinformatic analysis of the Thlaspideae repeatomes, and identified repetitive elements applicable as unique species‐ and genus‐specific barcodes and chromosome landmarks. This study deepens our general understanding of the evolutionary role of CRs, particularly pericentric inversions, in plant genome diversification, and provides a robust base for follow‐up whole‐genome sequencing efforts.
Exploring an Energy-Based Model in Comminution
Authors: Hosseini Poorya , Gharib Nima , Derakhshandeh Javad Farrokhi , Radziszewski Peter
Published: October 2021

Comminution devices use about 50% of the total energy in mine sites due to low efficiency and wastefulness of their operations. This study aims to provide a framework for improving efficiency of comminution processes by discretizing the relative distribution of various energy forms in the comminution process. Comminution theories are mainly based on the ore's size distribution and consequently do not adequately address other imperative phenomena such as media wear and energy utilization efficiency. We suggest an energy-based methodology that provides a common ground for comparing seemingly different aspects involved in comminution, most notably ore breakage, the media wear, and energy utilization efficiency. The experimental tests here were conducted using the Steel Wheel Abrasion Test (SWAT) highlighting the relationship between test variables, including media wear, ore breakage and energy efficiency. In addition, preliminary experimental results show interesting connections between various energy forms involved in comminution that is of high use in future design and performance optimization of comminution devices.
Analysis of the Reflooding Process in Degraded Particle Beds by Simulations of the Debris Test Facility With the Severe Accident Analysis Code ASTEC V2.1 and COCOMO Code
Authors: Peschel Jan M. , Bratfisch Christoph , Koch Marco K.
Published: October 2021

As part of a national research project RUB PSS investigates the cooling process of degraded core sections like debris beds and melt pools. Focus of these works is the simulation of the DEBRIS test facility located at IKE in Stuttgart with both the integral code ASTEC V2.1, developed by IRSN in France, and COCOMO, developed by IKE.

In this work, the simulation of two bottom flooding experiments with similar temperature and pressure conditions but different particle types will be presented. The selected tests will be simulated with both a fixed input signal and a gravity-driven quench water flow similar to the actual test conditions. Focus of the analysis is on a comparison of the calculated and experimental results to validate the codes capabilities to simulate the most relevant phenomena which occur while quenching dry and overheated particle formations under varying thermohydraulic conditions. ASTEC is validated using numerous experiments like the PEARL tests that are comparable to those conducted within the DEBRIS test facility and disposes a well-developed model basis for the simulation of late phase phenomena. The resulting temperature evolutions are in good accordance with the experimental results. This is also reflected by the evolution of the quench front and also applies to the conducted COCOMO simulations. The released amount of steam is overestimated by both codes and leaves room for improvement.
Wall Materials Effects on Sheltered Indoor Doses From an SMR Hypothetical Accident Release
Authors: Sugitatsu Yamato , Revankar Shripad T.
Published: October 2021

Small modular reactors (SMRs) are expected as a suitable candidate to fulfill energy needs in the future. The regulation of the emergency planning zone (EPZ) has been a controversial issue. The possibility of smaller EPZs because of their small core size and passive safety functions has still under discussion. The major emergency responses to radiological incidents in the early phase are evacuation from the area and sheltering-in-place within a building. Comparison between the dose incurred during evacuation and that with sheltering-in-place is necessary to consider the proper protective actions. This study focuses on effect of wall materials on indoor doses for sheltered population from small modular reactor severe accident. The source term came from loss of coolant accident or station blackout, and the time change of air concentration and the ground deposition data was calculated with RASCAL, a software developed by NRC to provide dose projection around the plant. Then general one-story and two-story houses were set up, and 6 wall materials were selected for calculating indoor doses. Cloudshine and groundshine were calculated with Monte Carlo methods, and the shielding function of each house was evaluated by comparing the indoor dose with outdoor dose. The result will be a basis for calculating the radiological dose for sheltered cases in case of nuclear emergency for SMR, which will be valuable to have a more effective emergency planning.
Source and Concentration of Radionuclides by Inland Nuclear Power Plant Under Normal Operation
Authors: Wang Jiaxin , Zhang Liuguo
Published: October 2021

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

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

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

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

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

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 > coolant turbulent kinetic energy > 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

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

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

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

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

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

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

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

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

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.
Assessment and Analysis of Various Mechanisms in the Coalescence and Breakup Models for Upward Bubbly Flow
Authors: Guan Shunran , Han Jinyu , Zhao Chenru , Bo Hanliang
Published: October 2021

Accurate three-dimensional CFD simulation is of great importance to the flow and heat transfer of flow boiling in the steam generator in nuclear power plant, in which bubbly flow is the main flow pattern. One of the biggest challenges for the CFD simulation is the coalescence and breakup model for bubbles dispersed in the continuous fluid. Recently, several new bubble coalescence and breakup models have been developed combined with the inhomogeneous Multi-Size-Group (MUSIG) model in the Eulerian framework in addition to the standard closure model (Luo and Svendsen (1996), and Prince and Blanch (1990)), which has been widely adopted in the CFD simulations in bubbly flow using the Eulerian method. For example, Liao had developed a comprehensive closure model (2015), in which five collision mechanisms and four breakup mechanisms have been included in the model. However, proper coefficients for the Liao model must be determined to properly address the proportion for various mechanism in the coalescence and breakup models. In this paper, we compared the predictions using the standard closure model and the Liao model (2015) combined with the iMUSIG model for the case of adiabatic upward bubbly flow in vertical pipe, as well as the corresponding experimental results reported by Lucas et al (2008) which were obtained at TOPFLOW test facilities. Six simulated working conditions are chosen. The significance of various mechanisms on the bubble coalescence and breakup were analyzed based on the bubble size distributions in various heights obtained in the CFD predictions and experiments. Finally, a set of proper coefficients for various mechanisms in the coalescence and breakup models of Liao (2015) was developed for the simulated conditions.
Transfer Function Development and Dynamic Analysis of a Heat Pipe Cooled Reactor
Authors: Pu Songmao , Sun Peiwei , Wei Xinyu
Published: October 2021

The heat pipe cooled reactor adopts the solid-state reactor design concept and the heat is passively transferred out of the core through heat pipes. It is characterized by high inherent safety and simple operation and has broad application prospects in deep space exploration and propulsion, sea submarine navigation and exploration. The design of heat pipe cooled reactor is unique, and its dynamics are different from traditional water-cooled reactors. Therefore, it is necessary to develop its dynamic model and perform dynamic analysis, and in this paper, the study object of the heat pipe cooled reactor is the 100kW nuclear silent thermoelectric reactor (NUSTER-100).

A nonlinear dynamic model is derived from the conservation equations of mass, energy and momentum. Point reactor kinetics equations are adopted. The linear dynamic model is constructed by linearization of the nonlinear model based on the disturbance theory and the transfer function is further derived applying Laplace transform. Both models including the nonlinear model and transfer function model are established on the MATLAB & Simulink simulation platform.

Dynamic characteristic analysis contains time domain analysis and frequency domain analysis. For the time domain analysis, by introducing a variety of boundary condition disturbances, the results were compared with those from transfer function. The results are consistent and can correctly reflect the dynamic characteristics of the heat pipe cooled reactor. Therefore, the transfer function model can be applied to the subsequent design of the heat pipe cooled reactor power control system.

For the dynamic analysis, it is divided into time domain and frequency domain. The time domain is to observe the change of core power and sodium temperature by introducing reactivity disturbance. For the frequency domain, after drawing the Bode plot of the transfer function, the system’s characteristics at different frequencies are analyzed. In addition, it can provide a theoretical basis for the design of the heat pipe cooled reactor power control system.
Long-Term Simulation of Sodium Dynamics During a Large Leakage Sodium-Water Reaction
Authors: Bai Xi , Sun Peiwei , Luo Gang , Cao Huasong
Published: October 2021

The steam generator of the sodium-cooled reactor is the barrier between the secondary and third loops. When the heat transfer pipe breaks, the water/steam will pour into the sodium and a sodium-water reaction will occur. The pressure in the secondary loop will increase and the rupture disks will burst to reduce the pressure. For the safety analysis, the maximum pressure is limited. For the long term, the pressure pulse weakens and the fluid flow tends to a quasi-steady state several seconds or minutes after leak initiation. Therefore, it is necessary to develop a model to investigate the dynamics of the secondary loop in the long term.

The continuity equation, the momentum equation and the energy equation are used to derive the sodium flow model in the secondary loop. The sodium pressure and velocity are described by the one-dimensional differential equation. The lumped parameter method is applied and the differential equations are solved by the Euler method. FORTRAN language is compiled to develop the code. Critical equipment, including the steam generator, buffer tank, pump, rupture disks, and accident discharge tank are considered in the secondary loop model. The sodium velocity and pressure responses varying with time can be obtained. Compared with the data from the safety analysis report, the tendency of the sodium velocity and pressure is consistent. It is proved that the model is reasonable and effective to simulate and analyze the actual long-term effects of sodium dynamics.
An Original Distributed Simulation Method Applied to the Advanced Nuclear Power Plant Control Technology Hardware-in-the-Loop Simulation Verification Platform
Authors: Bowen Li , Zhe Dong , Di Jiang
Published: October 2021

Hardware-in-the-loop (HIL) simulation technology, where the part of a system to be verified adopts real objects, is one of the important methods for the research of advanced nuclear power plant (NPP) instrumentation and control (I&C) technology. With the development of advanced NPP I&C technology, especially the multi-module NPP technology, the HIL simulation technology is facing the challenge of communication signals booming and model extension to deal with the requirement of modules increasing and thermal-electricity generation. Driven by the above requirement of research and engineering, it is necessary to develop a novel HIL simulation technology that has well flexible scalability and avoids the high computational burden of the distributed control system (DCS). In this paper, an original distributed simulation method applied to the transformation extension of the NPP I&C HIL simulation verification platform is proposed. The initial opinion of the method is deploying a third-party system utilized for numerical simulation and form a close loop with DCS by network communication. With the support of third-party equipment represented by the real-time target machine, the functions of the system can be flexibly expanded through the MODBUS series protocol, and algorithms with high sampling frequency requirements can be deployed. The method has the characteristics of economical communication consumption, standardized and reliable communication protocol, and flexible downloading models and algorithms mean. Aside from this, due to the relative independence from DCS, the distribute simulation method is promising to be an original platform for verifying the technology advanced control or fault diagnosis in addition to DCS computing servers.
Application of Bayesian Classifiers for the Accident Diagnosis in Nuclear Power Plants
Authors: Qi Ben , Liang Jingang , Zhang Liguo , Tong Jiejuan , Yan Shu
Published: October 2021

Early identification of operational failures or faults is of vital importance for the prevention and mitigation of nuclear power plant (NPP) accidents. Various research efforts have been devoted to developing efficient accident diagnostic methods. Among them is the most promising artificial intelligence (AI) technology which makes use of plant status monitoring data for accident identification. However, the AI-based approaches are concerned with the difficulty in combining robustness and interpretability, i.e. the explanations of the features involved in modeling. This paper explores new accident diagnostic methods that are both robust and interpretable based on Bayesian classification algorithms. Specifically, three Bayesian classifiers including discrete naive Bayes classifier, Gaussian naive Bayes classifier, and Bayesian network are implemented and tested with simulation data for a PWR nuclear power plant (NPP) which consists of multiple types of the incident (LOCA, MSLB, SGTB) and dozens of monitoring parameters. The diagnostic accuracy and efficiency of these classifiers are inspected and compared quantificationally for all accident scenarios. The results show the Gaussian Naive Bayes method has the best accuracy which is close to 100%, followed by discrete Naive Bayes (above 97%), while the knowledge-based Bayesian network reveals plain accuracy due to limited usage of simulation data. The Naive Bayes classifiers also outperform Bayesian Network in diagnostic efficiency as they have simpler network structures. The Gaussian Naive Bayes even simplifies the implementation with straightforward data pre-processing.
Simulation for Predicting Condition of Plant Equipment
Authors: Kashiwase Shoichi , Osaki Kenji , Hatakeyama Makoto , Kaneko Tomokazu
Published: October 2021

Condition-based maintenance (CBM) is a maintenance program that recommends maintenance decisions based on maintenance information by condition monitoring. To apply CBM to industrial systems, it is important to diagnose condition of equipment appropriately based on the knowledge of abnormal condition of the equipment. In nuclear power plant, however, it is quite difficult to obtain the knowledge of abnormal condition. For this reason, simulation is expected to be an effective approach to estimate conditions of equipment instead of data acquisition. In this study, we focused on and built models for a motor operated valve and a pump, since the inspection cost is generally expensive due to the number of them in nuclear power plant. We considered a pump system and analyzed the measurement values in normal and abnormal conditions. Our analytical result reproduces the behavior of the measurement values and shows the characteristic deviation from normal condition in abnormal conditions.
Simulation Analysis and Optimization of Lubricating Oil System
Authors: Wu Qiongxiao , Wang Jianjun , Chen Jingming , Li Pengzheng
Published: October 2021

Based on the one-dimensional simulation model of lubricating oil system is established and analyzed by using FLOWMASTER software, this paper proposes a new method of optimizing lubricating oil system by PID technology. Ensure that the configuration requirements and control strategies of the relevant accessories of the simulation model are satisfied with the design requirements. Firstly, by simulating lubricating oil pressure fluctuation and lubricating oil flow distribution under Open/Close Valve in different opening and closing time, the optimal opening/closing time of Open/Close Valve is determined to be 0.2 s and 0.5 s respectively. Secondly, by writing the controller script file combined with a controller to realize automatic unloading relief valve simulation, determine the relief valve pressure regulating range of 0∼0.38 MPa, For precision of constant pressure valve of oil spill, the simulation results show that the average 10 m3/h flow caused by pressure changes of about 0.06 MPa. Under the flow sudden change signal of about 40 m3/h, the maximum pressure change is less than 0.1 MPa. Through the simulation results, it is found that most of the lubrication parts in the original design have the phenomenon of flow redundancy, which causes unnecessary pump power loss. The system is optimized by PID technology. By comparing the simulation results before and after optimization, it is found that the speed of constant displacement pump could be changed in time by PID controller, and the flow redundancy could be improved significantly, so the lubricating oil system could be lower consumption and achieve the purpose of optimization.
Liquid Film Behavior of Bottoming Liquid Jet in a Shallow Pool Measured by 3D-LIF
Authors: Yamamura Sota , Yoshida Hiroyuki , Horiguchi Naoki , Kaneko Akiko , Abe Yutaka
Published: October 2021

When the core meltdown accident occurs in the nuclear plant, molten corium falls into a coolant pool of the lower plenum. It is considered that the molten corium jet is broken up, cooled, and solidified with fuel-coolant interaction (FCI). However, the coolant pool could be a shallow condition by the leakage and evaporation of the coolant. In this situation, it is considered that the corium jet bottoms and spreads without the jet breakup. From the viewpoint of safety, understanding a jet behavior and estimating a cooling behavior are needed.

The purpose of this study is to clarify the mechanism of the liquid jet behavior in a shallow pool as the fundamental process for estimating the cooling behavior in the real machine. In this paper, we discuss the spreading behavior of the liquid jet after bottoming. The jet injection experiment was conducted using test fluids. By using the 3D-LIF method, the 3D visualization of the liquid jet was Successfully implemented. From the visualization result, the following behaviors were seen. After bottoming, the jet spread radially with the liquid film. As the jet spreading behavior, the liquid film was rolled up to the inside, and the vortex was formed. After a certain time, the vortex was broken. Then the flow and the number density of the fragment were changed.
Quantitative Measurements of Bubbles and Foam Flow Generated From Two-Phase Subcooled Flow Boiling of Seawater in a Vertical Annulus
Authors: Li Yuanjie , Shah Syed Waqar Ali , Pan Chin
Published: October 2021

Bubbly flow or foam flow could prevail in the subcooled flow boiling of seawater in an annulus. Compared with the bubbly flow of de-ionized water in the subcooled flow boiling, the bubble size of seawater is smaller. Further, bubbly flow with small bubbles or foam flow in a subcooled flow boiling of seawater would be affected by different heat fluxes, mass fluxes and inlet subcooling. The change in the boiling behaviors due to different conditions can be reflected in terms of variations of bubble characteristics. However, details of the characteristics of bubbles in seawater generated by the two-phase subcooled flow boiling are still limited. In the present study, bubble characteristics are measured and detailed by an image analytic method in the software Image J. The subcooled flow boiling experiments with different heat fluxes are conducted in a vertical upward annulus with artificial seawater at the inlet temperature of 90°C and mass flux of 1200 kg/m2s. The two-phase flow patterns are recorded by the high-speed camera in 6000 frame/s. To conduct measurements of bubble characteristics using Image J, images are selected as one set of measurement from the videos recording under different heat fluxes. After preprocessing of images, measurements of bubbles and fitting procedures with ellipses are conducted. Averaged values of descriptors of bubbles are obtained, including areas, perimeters, lengths of major axes, lengths of minor axes, circularities, angles, solidities, and aspect ratios, etc. It provides a possible way to investigate the details of bubbly flow or foam flow in the subcooled flow boiling of seawater.