bluuebunny/crossref_metadata_2025 huggingface dataset:
Development of Prevention Method for Air Ingress During a Pipe Rupture Accident of the VHTR: Effectiveness on Localized Natural Convection
Authors: Tanaka Yudai , Takeda Tetsuaki
Published: October 2017

A primary pipe rupture accident is one of the design-basis accidents of a Very-High-Temperature Reactor (VHTR). When a primary pipe rupture accident occurs, air is expected to enter into the reactor pressure vessel from the breach and oxidize in-core graphite structures. Therefore, it is important to understand the mixing processes of different kinds of gases in the stable and unstable stratified fluid layers. In particular, it is also important to examine the influence of localized natural convection and molecular diffusion on the mixing process from a safety viewpoint. Therefore, in order to predict or analyze the air ingress phenomena during a pipe rupture accident, it is important to develop a method for the prevention of air ingress during an accident.

We carried out experiments to obtain the mixing process of two-component gases and flow characteristics of localized natural convection. This study also investigated a control method for the natural circulation of air through the injection of helium gas. An experiment has been carried out to investigate a control method of natural circulation of air by injection of helium gas. The experimental apparatus consists of a reverse U-sharped vertical slot and a storage tank. One side-slot consists of the heated and cooled walls. The other side-slot consists of the two cooled walls. The dimensions of the vertical slots are 598 mm in height, 208 mm in depth, and 70 mm in width. Each two vertical slots were connected and were a reverse U-shaped passage. The dimensions of the connecting passage were 16 mm in height, 106 mm in depth, and 210 mm in length. The storage tank was connected to the lower part of the reverse U-shaped passage. The dimensions of the storage tank were 398 mm in length, 398 mm in depth, and 548 mm in width. The reverse U-shaped passage and the storage tank were separated by a partition plate. The wall and gas temperature were measured by a K-type thermocouple.

Experimental results regarding mixing process of two component gases in vertical fluid layer were as follows. The heavy gas was transported to the slot by the molecular diffusion and natural convection. As time elapses, natural circulation of heavy gas suddenly occurs through the reverse U-shaped slot. As a result of experiments, the onset time of natural circulation is affected by not only molecular diffusion coefficient but also the strength of natural convection. When the helium gas is injected into the channel, it is possible to control the natural circulation of air. The onset time of the reproduction of the natural circulation can be varied by changing the injection rate of the helium gas.
Numerical Investigation of Convective Heat Transfer to Supercritical Hydrogen in a Straight Tube
Authors: Ji Yu , Shi Lei , Sun Jun
Published: October 2017

Hydrogen is adopted as coolant for regenerative cooling nozzle and reactor reflector in nuclear thermal propulsion (NTP), which may be a promising technology for human space exploration in the near future due to its large thrust and high specific impulse. During the cooling processes, the hydrogen experiences the transition from subcritical state to supercritical state, which influences the heat transfer severely. This paper is intended to study the characteristic of convective heat transfer to supercritical hydrogen in a straight tube under high heat flux through numerical simulation, which is a common phenomenon in NTP operation. The thermophysical properties and transport properties including the equation of state, specific heat capacity, viscosity and thermal conductivity of hydrogen are evaluated firstly by compared with the data from National Institute of Standards and Technology (NIST). Then, the flow and heat transfer process is investigated using Reynolds Averaged Naiver-Stokes (RANS) model, and the approach is validated by the successfully predicted behavior called local heat transfer deterioration. Moreover, the mechanism of heat transfer deterioration is analyzed briefly according to the detailed information of flow field. This work herein contributes to the further NTP design and research.
Energy Needs and Potential Nuclear Power Development in China
Authors: Zhao Danlu
Published: October 2017

The harmonious development of economic, energy and environment is an important premise to realize the objective of China’s modernization. Currently, different parts of China have different main energy source, while nuclear power development has many opportunities and challenges. This paper considers the current trends of energy needs in China, and discusses the different influencing factors of energy needs throughout China.

In addition, this paper will focus on the potential nuclear power development in China, which mainly focuses on policy, technology, nuclear security and social attitude. Then it will focus on the application in Shenzhen (Daya Bay). Based on previous analysis, technical/engineering feasibility and site feasibility are considered in this part. Finally, a number of recommendations for nuclear development management in China will be given. These recommendations will help the public to have a basic understanding of nuclear power management, and to improve the social attitude of China’s nuclear energy development.

In all, this paper puts forward the management methodology of nuclear power industry, which has positive significance for the field of nuclear power education. Meanwhile, the paper will play a positive role on popularize the knowledge of nuclear power to the public.
Development and Application of CAP1400 Probabilistic Safety Assessment
Authors: Zhang Qinfang , Zhan Wenhui , Qiu Yongping , Li Zhaohua , He Jiandong
Published: October 2017

A full scope Probabilistic Safety Assessment (PSA) is developed for CAP1400 and some design modification was performed based on the CAP1400 PSA insights. In this paper, a brief introduction of CAP1400 PSA is presented, including analysis method, procedure, results and insights obtained and design improvements. The CAP1400 PSA includes internal events level 1 PSA, low power and shutdown PSA, level 2 PSA, level 3 PSA, internal fire PSA, internal flooding PSA, seismic margin analysis and seismic PSA, etc. The CAP1400 PSA is developed based on PSA methods acknowledged internationally and in compliance with the requirement of PSA related standards, such as ASME PSA standards and Chinese PSA standards. The design improvements derived from PSA results are also introduced in this paper, including design change of passive core cooling system (PXS), diverse design of instrument and control system, design change of in-vessel retention (IVR) of molten core debris, installation of very early warning fire detection system and design change of flooding protection of PXS valves.
Study on the Effects of Mixing Vane Configurations on Flow Field in Rod Bundle
Authors: Mao Hu , Yang Bao-Wen , Liu Aiguo
Published: October 2017

In most PWR designs, spacer grids are proposed to keep the fuel rods in fixed positions. Mixing vane grids (MVGs) contribute to inter-subchannel mixing as well as providing structural support, which has been demonstrated by both experiments and numerical simulations. The mixing vane configurations in the MVGs have significant influence on flow redistribution. The mixing vane configurations will lead to flow redistribution among neighboring subchannels. Appropriate mixing vane configuration designs in the MVG will result in expected flow redistribution coupled with the power distribution of the rod bundle. This paper analyzes the effects of different types of mixing vane configurations in the MVGs on the flow field in the 5×5 rod bundle using the improved mixing vane grid model in the subchannel code. In the improved model, source terms related to the MVGs performance were developed and some geometrical parameters of mixing vanes (such as the length, angle, area and configurations of the mixing vanes) were examined and brought in to improve the capability of subchannel code in predicting actual MVGs performance. This paper gives a brief introduction to both the original spacer grid model and the improved mixing vane grid model in subchannel analysis, and focuses on analysis of the effects of the mixing vane configurations. In this paper, different types of mixing vane configurations are considered for the comparison of hydraulic performance in the 5×5 rod bundle under uniform axial heating conditions. The calculation results applying the original model are also presented for the comparison analysis, and the calculation results without the MVGs are provided as the basis of the comparison. Both the qualitative and quantitative analyses are presented, and the hydraulic performance of different types of mixing vane configurations are evaluated. The analysis indicates that the improved MVG model can reflect the mixing vane configuration information, and could provide a reference for mixing vane configuration designs in the MVG as well as the experimental design related to the MVG performance study.
Study on the Performance of a Kind of Corrugated-Plate Separator
Authors: He Wei , Tian Ruifeng , Chen Bowen , Mao Feng
Published: October 2017

In order to guarantee the safe and economic operation of nuclear power plant, providing the steam of high quality is very important. As the last stage of the steam-water separating device, corrugated-plate separator has low pressure loss and high separation efficiency, so it is widely used in various types of nuclear power plant. Due to the complexity of the flow field inside the corrugated-plate separator and the immaturity of the numerical method, the research on the corrugated-plate separator is mainly based on experiment. In this work, a kind of corrugated-plate separator with new structures are studied by means of experiment.

In this work, test bed was built with plexiglass. The separation characteristics, resistance characteristics and critical inlet flow velocities of the two structures of separators at different inlet velocities were studied respectively. The following conclusions were obtained: With the increase of the inlet-air velocity, the separation efficiency increases, and the pressure loss increases too. But when the critical inlet flow velocity is reached, increasing the air velocity will reduce the separation efficiency. Considering the influence of structural parameters and airflow parameters, the optimum structural parameters of two kinds of wave plates are obtained. By contrast, it is found that the performance of double-hook corrugated-plate is better than that of single-hook corrugated-plate.
Modelling and Analysis of Loviisa NPP Pilot Operated Main Steam Safety Valves
Authors: Juntti Mika
Published: October 2017

Loviisa NPP (unit 2) low pressure Main Steam Safety Valves (MSSV) were renewed during 2014 overhaul. New valves are pilot operated and qualified for two phase conditions. The old valves were spring loaded and qualified only for steam flow conditions.

During the testing of new MSSVs, unexpected behavior was detected. The opening of new MSSV included significant delay in case of repeated openings. This led to a need to investigate this phenomena more closely since such opening delay could have negative impact to plant safety.

This paper presents the aftermath of the discovered opening delay in the NPP and describes how thermal hydraulic modelling and analyses were used to gain in depth understanding of pilot operated MSSV. In addition, it is presented how the analysis work was used to solve the discovered challenges.

Modelling work was conducted using Apros Process Simulation Software [1] which is the main tool for deterministic safety and engineering analysis of Loviisa NPP [2–4].
Root Cause Analysis of a Leak Off Tube of the Primary Loop in a CANDU 6 Unit
Authors: Zhao Liang , Luo Kunjie
Published: October 2017

Chemical composition analysis, microstructure analysis, macro and micro fracture morphology analysis were conducted on a cracked leak off tube in primary heat transport and heavy water collection system of a CANDU6 unit. The results show that the material conforms to the original design requirements. The leaked fluid vaporized at the penetration and stress corrosion crack initiated because of cl-concentration. The leaked fluid drains periodically and lead the wall surface temperature periodically changes. The SCC crack transformed to corrosion fatigue crack and the tube finally fractured. In order to prevent the similar tubes crack again, preventive maintenance strategies were given in this paper.
Analysis on Radioactive Gas Movement and Delay Status Before Release
Authors: Zhu Bo Nan
Published: October 2017

Based on the PWR nuclear power plant, the sources of radioactive noble gas from core fission and the model of the gas movement in the waste gas system are analyzed, focusing on the delay state analysis of the isotopes of Kr and Xe creeping through the activated carbon. This paper proposes a design method of delay bed mass calculation based on system flow rate, dynamic adsorption coefficient and the holdup time back calculated from the annual radioactive release of waste gas system. Finally the section and dimensions of delay beds are determined based on the flow velocity recommended in ANSI/ANS 55.4. This calculation method lays the theoretical foundation of the waste gas system design and the dimension design for activated carbon delay bed.
An Advanced Method for Evaluating Risk From Seismically-Induced Internal Fires and Floods
Authors: Amico Paul J. , Macheret Pierre , Kassawara Robert P.
Published: October 2017

It has been traditional in assessment of nuclear power plant safety that both deterministic safety analyses and probabilistic safety analyses treat the potential effects of various hazards individually. That is, the safety implications of internal events (e.g., randomly occurring transients and LOCAs), internal hazards (e.g., internal fire and flood), and external hazards (e.g., earthquakes, tornados) are treated as independent occurrences. With the occurrence of the Great Tohoku earthquake and the effects observed at nuclear plants in Japan, it was realized that this approach failed to provide a realistic representation of risk, and now there is a significant interest in correlated hazards.

As a result, EPRI embarked on the development of an improved methodology focusing on seismically-induced internal fires and internal floods. All the technical work on the methodology has been completed and draft technical guidance developed. This guidance has been provided to some plants that are interested in piloting the methodology. As of the date of paper submittal, two pilots are underway and three more are under consideration. Upon completion of the pilots, the methodology will be updated to incorporate the lessons-learned and published.
Numerical Simulation Analysis of Corrugated Plate With Hook
Authors: Chen Bowen , Tian Ruifeng , Mao Feng , He Wei
Published: October 2017

Separation device is one of the important equipment in the steam generator, which reduce the humidity, and it is a key device of nuclear plant which can guarantee the saturated steam to the steam turbine to provide clean. In nuclear power plant, the nuclear steam turbine uses saturated steam directly, so the separation system in steam generator requires a high efficiency of separation. Because of its simple process, high separation efficiency, low system pressure drop and large capacity, small volume, long life and other advantages, corrugated plate in water separator technology is more and more extensive application in the production process. In this paper, the corrugated plate water separator is simulated research, through the establishment of a two-dimensional model of corrugated plate, separation efficiency and pressure drop of corrugated plate is simulated. By setting the different air speed, different gap of hook-plate, observe the change of the pressure drop and separation efficiency, and analyzes the internal mechanism of the foundation for the optimization of corrugated plate dryer.
Development of Wire Mesh Sensor to Study Two Phase Flow and its Comparison With High Speed Camera
Authors: Abbas Kamil , Haijian Lan , Ting Yu , Shiliang Zhou , Ali Shahzad Muhammad
Published: October 2017

The main concern of this research is the two phase flow, which consist of air bubbles submersed in water in reactors cores. The understanding of these such flow activities is paramount to finding out about the efficiency of the process and for safety concerns. This knowledge is also important to control and monitor the flow at the spot or online. The employed method is based on the conductivity principle, and it was chosen because of the cheapness of the circuitry involved for implementation, as compared to other imaging techniques. Keeping this in mind, it provides an excellent temporal and spatial resolution for measurement. To fulfill this requirement, a small testing facility was designed and developed. The circuits were made and then tested by making PCBs. To prevent environmental noises and electromagnet disturbance, the apparatus was shielded to the best of ability, in accordance with the available resources. A wire mesh sensor with sixteen inputs and outputs respectively was made and put in a vertical glass section with water and air filled inside. The bubbles were generated by external means and the signal interference was noted in the collection section. The signals were collected by a data acquisition card and further manipulated in a software program. The data was first taken in the LabVIEW software and then transferred to MATLAB for analysis and image reconstruction. Basically, the main system included wire mesh, the circuit for excitation, the circuit for data collection and data acquisition card connected to a computer. The results were compared with a high speed camera which was linked in synchronization with the software. In these circuit designs, the maximum efficiency was ensured while keeping the cost low. The sensor is surrounded by glass so that the high speed camera could also be used to compare results. The use of the high speed data acquisition card also enabled to deal with real time acquisition without any problem. Unfortunately, this facility could not be tested in rigorous conditions like high temperature and pressure because of limited resources. The conclusions drawn from these experiments are that wire mesh sensors indeed provide temporal resolution of about 1000 frames/s with a good spatial resolution and is also a lot cheaper in comparison with other techniques.
The Thermodynamic Design of the Radial Inflow Nuclear Turbine
Authors: Wang Bo , Tian Rui-feng
Published: October 2017

According to the flow condition of the nuclear turbine, we can divide the turbine into the radial inflow turbine and axial inflow turbine. The radial flow turbine refers to the turbine steam where the steam or the vapor flows along the radial center. The radial machine has much greater work performance. Because the steam has to overcome the centrifugal force which increases its work ability. Currently, the radial inflow turbine is widely used in small unit and super low temperature engineering. In the state of global warming and the lack of energy, the radial inflow turbine performs well since it has a lot of advantages. This paper mainly introduces the process of the thermal design of a radial inflow turbine. One of the most important task is to determine the basic thermodynamic process and the selection of various parameters of the radial inflow turbine. We will apply the computer programming language C# to compile it. The thermodynamic process design uses IAPWS-IF97 formula to compute physical parameters of the vapor. Then we will finish the development of the software.
On Different Break Section of LOCA Cases
Authors: Liu Peiqi , Yu Tao , Yang Hongyan
Published: October 2017

A typical 1000MW pressurized-water reactor (PWR) unit model of China’s living nuclear power plant (NPP) units is built based on MAAP4[1] in this paper. Different severe accidents cases caused by different LOCA area on hot leg of primary loop are studied. And different mitigation measures are focused to evaluated their effectiveness. The study indicates that during the accident, the larger broken area LOCA case caused the more severe rector core damaged. However, it is important to inject water into the reactor core in good time. And that can mitigate the severe accident progress effectively.
Study on Vibration Characteristics of Induction Motor Under Load Considering Saturation of Magnetic Circuit
Authors: Zhou Pu , Zhang Yue , Zhen Zhen , Wang Qiang , Feng Hai-jun
Published: October 2017

The variation law of motor vibration with load under saturation of magnetic circuit is one of the focus of motor vibration research. T-type equivalent circuit method is shown to be a suitable engineering analysis way. In this paper, the variation of magnetic flux density under saturation state was analyzed by using a small induction motor, and a physical model of magnetic saturation and motor vibration was established by the way of T-shaped equivalent circuit. Based on the physical model, the vibration mechanism of the motor under the saturated magnetic circuit was analyzed and the motor vibration simulation with different load conditions was carried out by the finite element method. Finally, a vibration test rig for motor under load was built and the simulation results were verified by experiments. The results show that when the motor load is less than the rated load, the amplitudes of the electromagnetic force in the middle and high frequency band change little. When the load is close to or higher than the rated value, the harmonic components of electromagnetic force in the middle and high frequency bands increase, and the amplitudes gradually increases due to the saturation of the magnetic circuit. Therefore, in the low-noise motor design, motor torque performance and vibration noise performance should be considered to optimize the saturation control of the magnetic circuit.
Study on Risk Analysis and Mitigation Measures for Total Loss of Heat Sink Accident in a Nuclear Power Plant
Authors: Pei Liang
Published: October 2017

There have been multiple water intake blockages caused by marine growth at nuclear power plants (NPPs) in recent years, which draws attention to the risk of loss of cold source as an important nuclear safety issue for NPPs. The results of a level 1 probabilistic safety assessment (PSA) for internal events of a certain NPP in China confirm that total loss of heat sink is one of the major risk contributors. From the perspective of PSA, this article analyzes and evaluates the risk of total loss of heat sink accident of a Chinese second-generation pressurized water reactor (PWR). Possible weak links during the accident mitigation are identified and corresponding suggestions of improvement are provided in order to lower the risk of total loss of heat sink for this NPP.
The Finite Element Method for Retention System of Radioactive Material Transport Package
Authors: Dongyuan Meng , Renze Wang , Jiangang Zhang , Guoqiang Li , Dajie Zhuang , Hongchao Sun , Xuexin Wang , Shutang Sun
Published: October 2017

According to IAEA “Advisory Material for the IAEA Regulations for the Safe Transport of Radioactive Material”, the retention of radioactive material transport package need to be paid attention. Finite Element (FE) method is an efficient way to verify the retention system. Two typical kinds of retention systems are simulated by FE. Two important aspects, mesh density and contact model are discussed for their influence of the calculation time and calculation accuracy. The method could be promoted to other aspects in radioactive material transport.
Influence of 300MW Fuel Assembly Spacer Grid Improvement on DNBR Analysis
Authors: Gan Fujun , Xu Zhihong , Zhang Chaozhu , Ren Bing
Published: October 2017

In order to reduce the potential hang up damage risk during refueling operation in Qinshan phase-I NPP, Shanghai Nuclear Engineering Research &Design Institute (SNERDI) carried out improvement work of 300MW fuel assembly space grid (mainly focused on the design of guide vane) to decrease the possibility of snag event between two fuel assemblies. In the whole project, multiple tests and analysis were carried out to demonstrate that the mechanical and thermal-hydraulic performance of new designed grid meets predetermined functional and safety requirement. During above process, DNBR analysis and comparison was performed under the influence of old and new designed spacer grids. For the old design of space grid, W-3 CHF correlation was adopted to calculate the DNBR margin. Thus, the first step was to conform that this CHF correlation can still be used. CFD simulations were employed to calculate the flow and heat transfer phenomenon in 17×17 rod bundles with old and new space grid, to compare the thermal-hydraulic performance of these two kinds of space grids qualitatively. To carry out above DNBR analysis, sub-channel grid loss coefficients are very important parameters. CFD tools and pressure drop tests help us to obtain these parameters. Four special rod bundles were used to calculate ratio of different sub-channels. Based on above work, the typical, thimble, side and corner sub-channel loss coefficient of old and new space grid were obtained, respectively. Finally, based on a safety analysis 14 sub-channel model, DNBR analysis were finished. Results indicated the Minimum DNBR of side and corner cell is a little lower than typical cell model.
Experimental Investigation on Heat Transfer and Pressure Drop of Supercritical Water Flows in 2×2 Bundle With Wire Wraps
Authors: Hong-bo Li , Zhen-xiao Hu , Dong-hua Lu , Han-yang Gu
Published: October 2017

An experimental investigation on heat transfer of supercritical water and transient heat transfer during pressure drop from supercritical to sub-critical conditions in a 2×2 bundle with two channels has been carried out on the supercritical water multipurpose test loop II (SWAMUP-II). The bundle consists of four heated rods with an O.D. of 10 mm and a pitch-to-diameter ratio of 1.18. The 2×2 bundle with wire wrap is installed into a square assembly box with rounded corners by which the test section is separated into two channels. Water flows downward in the first channel between the pressure tube and the assembly box, and then turns upward in the second channel inside the assembly box to cool the four heated rods, which are directly heated by DC power. The experimental conditions are as follows: pressure ranging from 16 to 26 MPa, mass flux from 400 to 1450 kg/m2s, heat flux from 250 to 1000 kW/m2, bulk temperature from 280 to 500 °C, and pressure drop rate of 1 and 2 MPa/min. The experimental data are obtained and heat transfer characteristics discussed.
The Comprehensive Evaluation for Nuclear Facility Decommissioning Scheme Based on AHP-DEA Method
Authors: Bai Lu , Liu Yong-kuo , Chao Nan
Published: October 2017

In view of the lack of research on decommissioning scheme selection of nuclear facilities and the inability to directly use the evaluation model of existing non-radioactive environment construction schemes. In this paper, propose a task analytic method in radiation field based on AHP-DEA model, which takes accomplishment the task safety as the decision-making target. The evaluation model can both reflect the decision preference of experts and improve the rationality of evaluation by mathematical calculation. The experimental results show that the proposed model has obvious advantages for program selection and can be applied to the evaluation and optimization of decommissioning scheme.
Nonlinear Finite Element Analysis of Class 1 Pressure Vessels
Authors: Chen Mingya , Yu Weiwei , Shi Jinhua , Lu Feng
Published: October 2017

The rules of design by analysis have been widely accepted for the Class 1 Nuclear Pressure Vessel, and two methods, stress classification method (based on the linear analysis) and the direct route (nonlinear analysis by Finite Element (FE) method), have been developed. In the stress classification method, the stress is classified in primary, secondary and peak stresses, and there are many shortcomings in this method. The stress classification is only easy in simple cases, like cylindrical shell under axisymmetric loads. In some cases the elastic analysis result can be non-conservative in particular when a part of primary stresses is considered secondary, like thermal expansion in some piping systems. When the geometry or the loads are more complex (such as, nozzles of reactor pressure vessels (RPV)), it will be an extremely conservative approach if a large part of stresses is considered as primary. Consequently, the direct route based on FE method is an efficient alternative to the linear analyses, and one of the major advantages is to suppress the discussion on the “stress classification in primary versus secondary” associated with the elastic analysis. This work reviews the existing linear design rules in the different codes, such as ASME Code NB 3200 and RCC-M Code B 3200, firstly. Then a typical case study of a large class 1 vessel nozzle, a RPV nozzle, under pressure and piping loads based on linear (stress classification method) and nonlinear rules (direct route) is presented. The case will consider 2D geometries under axisymetrical loads, and the failure models of plastic collapse, plastic instability and local failure are studied. Lastly, a set of proposals to improve the design rules are given.
Development and Comparison of a Heat-Transfer Correlation for Supercritical Refrigerants
Authors: Tarsitano Nicholas , Sidawi Khalil , Pioro Igor
Published: October 2017

The objective of this paper is to act as a collection of multiple different heat-transfer correlations and to check their accuracy when compared to experimental data obtained in supercritical-pressure refrigerants (R-22 and R-134a). This paper is also intended to collect as much relevant data of heat transfer in supercritical refrigerants as possible for future research. The experimental data have been retrieved from graphs within a wide range of operating parameters. This study is in support of potential use of supercritical refrigerants as modeling fluids instead of supercritical water. The use of refrigerants as modelling fluids instead of water will allow to decrease costs and technical difficulties during experiments at supercritical pressures and widen operating ranges, because the critical parameters of refrigerants are significantly lower than those of water.

The research was completed by collecting graphed data from several different experimental series using both R-22 and R-134a data. The advantage of comparing different refrigerants for determining correlation accuracy is to increase the predictability for other potential experiments using refrigerants. All data are taken from bare-tube experiments to produce a relative baseline for heat-transfer characteristics. These experiments have been performed within the following range: Inner tube diameter ranging between 4.4 mm to 13 mm, pressure ranging between 4.3 MPa to 5.5 MPa, and at a number of various mass and heat fluxes. Sixteen potential heat-transfer correlations have been selected and used in this assessment. The correlation by Watts and Chou [1] and Cheng et al. [2] were shown to have the lowest root-mean-square error. Other correlations with the reasonable accuracy include Mokry et al. [3] and Swenson et al. [4] correlations. However, it was decided to develop a new correlation based on these refrigerant data in an attempt to increase the prediction accuracy.

Therefore, based on the Mokry et al. [3] correlation a modified correlation was developed, which generalized the experimental Freon data with higher accuracy than the know correlations. This correlation is intended to create a basis for further study on the use of refrigerants as modeling fluids. While Freon has similar properties to water at supercritical conditions, the different molecular properties causes factors to affect each fluid differently. For refrigerants at supercritical conditions, the factors that seem to have the most effect are the dynamic viscosity and density of a fluid.
Study on Operation Ranges and Discharge Performance of Steam Injector System
Authors: Moribe Takahiro , Endo Hiroto , Miwa Shuichiro , Sakashita Hiroto , Mori Michitsugu
Published: October 2017

Steam Injector (SI) is a passive pump activated by steam and water, and it does not require any external power supplies or rotating machineries. Moreover, SI has high ability as a heat exchanger by undergoing direct contact condensation mechanism. From these characteristics, SI has a capability to be applied as a passive safety system in the nuclear power plant. In the present study, experiments targeting the operating range and pump performance were carried out to obtain SI’s detailed characteristics under following supply conditions; inlet steam pressure was 0.02 ∼ 0.81MPaG, inlet water flow rate was 0.21 ∼ 0.80 kg/s. In the former experiment, operating range was investigated by changing inlet conditions, and the influence of steam inlet pressure and water inlet flow rates on SI operating range was tested. In the latter experiment, pump performance of SI was evaluated by investigating the maximum discharge pressure in each inlet condition by using back pressure valve. From the results of experiments, it was confirmed the operating range of SI was limited by supplied steam pressure and supplied water flow rate, and some clear trends were found in the operation boundary map. In addition, SI could discharge water at least 1.2 times higher than inlet steam pressure under above-mentioned conditions, which verified SI’s capability to be applied for the nuclear power plant as a core cooling system. And furthermore, existing 1D analysis model’s predictive capability was tested based on these experimental results.
Microstructure and Properties of New Zirconium Alloys for CAP1400 Fuel Assembly
Authors: Zeng Qifeng , Zhu Libing , Yuan Gaihuan , Wang Lian , Sun Guocheng , Lu Junqiang
Published: October 2017

Two Zr-Sn-Nb alloys with minor Germanium or silicon additions were prepared by traditional manufacturing process to meet the design requirements. Transmission electron microscope (TEM) and scanning electron microscope (SEM) were utilized to characterize the detail microstructure of base alloys. Corrosion resistance was examined by the weight gain in static autoclave with different water chemistry environments. The mechanical properties at room temperature and elevated temperature were evaluated by conventional tensile testing. Thermal creep resistance was evaluated by an internally-pressurized creep test at 385 °C with hoop stresses of about 108 MPa and 150 MPa (during 24 h). It was found that SZA-6 and SZA-4 alloys consisted of partially recrystallized grain structures with uniformly distributed fine second phase particles (SPPs) located within grain interior and at grain boundaries. Both SZA-4 and SZA-6 alloys exhibited excellent corrosion resistance in two water chemistry conditions. The corrosion resistance of SZA-6 was better than the reference commercial alloy, and SZA-4 was slightly better than SZA-6. The mechanical properties of two new zirconium alloys were comparable, and both of them can meet the design criterion. Moreover, the thermal creep resistance of SZA-4 and SZA-6 alloys was equivalent to existing commercial alloy. Considering the outstanding corrosion resistance, satisfied mechanical properties and thermal creep resistance, SZA-4 and SZA-6 alloys were suggested as promising alloys used for CAP1400 fuel assembly in the future.
Research and Software Implementation of Seismic Probabilistic Safety Analysis Quantification Method
Authors: Gao Wei , Tang Guofeng , Zhang Jingyu , Zhang Qinfang
Published: October 2017

Seismic risk of nuclear power plant has drawn increasing attention after Fukushima accident. An intensive study has been carried out in this paper, including sampling of component and structure fragility based on Monte Carlo method, fragility analysis on system or plant level, convolution of seismic hazard curves and fragility curves. To derive more accurate quantification results, the binary decision diagram (BDD) algorithm was introduced into the quantification process, which effectively reduces the deficiency of the conventional method on coping with large probability events and negated logic. Seismic Probabilistic Safety Analysis (PSA/PRA) quantification software was developed based on algorithms discussed in this paper. Tests and application has been made for this software with a specific nuclear power plant seismic PSA model. The results show that this software is effective on seismic PSA quantification.
The Uncertainty Analysis of Natural Circulation in Molten Salt Reactor Coupling Dakota With GenFlow
Authors: Zhao Yang , Guo Zhang-peng , Niu Feng-lei
Published: October 2017

Molten salt reactor is the only one of the liquid fuel reactors in the fourth generation reactors. The flow characteristics of the fluoride makes differences between the liquid fuel reactor and the solid fuel reactor. For licensing of novel design of molten salt reactor, the peak temperature of the molten salt in primary loop has significant influence on the reactor safety. Sensitivity analysis for the peak temperature during pump coast down accident is performed by coupling Dakota and Genflow code. Dakota is a multilevel parallel object-oriented framework for design optimization, parameter estimation, uncertainty quantification, and sensitivity analysis. Dakota generate a plurality of parameters of input files, can greatly simplify the manual optimization process, it will reduce a lot of uncertainty analysis time. GenFlow develop the technical basis for design and licensing of fluoride-salt-cooled, high-temperature reactors, simulate the accident condition and export the instantaneous state. The work coupled Dakota and GenFlow to do the uncertainty analysis, the experience also have important inspiration and reference to coupling Dakota with other code. The work aims at the scenario when the pump coast down and the natural circulation is established. Five key thermal parameters, namely the power load factor, coast down time, the check valve open time, total form losses in the DRACS loop, total form losses in the DHX branch, that affect the natural circulation in the molten salt reactor is sampled within a certain range using the LHS (Latin hypercube sampling) method. The Dakota generate multiple parameters of the input card that are run by GenFlow code automatically. The post-processing results turn back to Dakota to do the uncertainty analysis and sensitivity analysis, which greatly simplify the optimization process and do not need exact mathematical model. It shows the correlativity about the peak temperature with every parameters and the influence between the parameters. Statistical results showed that the effect of the power load factor on peak temperature is very significant, followed by the DHX shell side. After the moment-related statistics, the 95% confidence intervals on the mean and standard deviations are printed. There are 95% confident that the true value of the parameter is in confidence interval between 809.547 and 816.3227. In this sample, 15% of the actual calculation results is higher than the critical point 816°C.
Piezoelectric Nuclear Battery Driven by the Jet-Flow
Authors: Zhou Y. , Zhang S. X. , Li G. P.
Published: October 2017

As an important kind of energy source, radioisotope batteries are attracting more and more academic researchers and people from industry due to the high power density, long lifetime (equal to half life of the radioisotope source), outstanding reliability, without maintenance, miniaturization and wide application compared with traditional dry batteries, chemical batteries, fuel batteries and solar batteries. Radioisotope batteries have been developed for more than 15 species since the first β battery invented by Henry Mosley in 1913. Based on a Brayton cycle Radioisotope Power System and a PZT-5H (Pb(ZrxTi1-x)O3, 0≤x≤1) unimorph, the piezoelectric nuclear battery driven by the jet-flow (PNBJ) is demonstrated in this work. The PZT-5H unimorph replaces turbine and utilizes high speed nitrogen jet-flow heated by the decay energy of radioisotope to output electrical energy. Over 0.34% energy conversion efficiency for the PNBJ is obtained at the flow of 2.26 × 10−3 m3/s and room temperature on half plane. The PNBJ can be used in low power microelectronics and microsystems, like electronic watch, AC-LED (alternating current light-emitting diode), and sensors. We believe that the researches and applications of PNBJ will be much attractive with the breakthroughs of preparation technology made by academic and industrial world.
Studies on Sensitivity and Uncertainty Analyses for SCOPE and WAFT With Uncertainty Propagation Methods
Authors: Zhangli Wang , Benxue Hu , Guodong Wang , Zhe Wang , Chenxiao Ni , Xin Liu
Published: October 2017

The purpose of Steam condensation on cold plate experiment facility (SCOPE) and Water film test (WAFT) is to verify the steam condensation and water film evaporation correlation within the parameter variation range of CAP1400 passive containment cooling system. These correlations were used for containment response analysis. Uncertainty and sensitivity analyses were performed for SCOPE and WAFT tests in this paper.

Sampling-based sensitivity analysis with uncertainty propagation is a new parameters sensitivity analysis method, and the importance of input parameters could be evaluated by calculating the correlation coefficients between input parameters and the output target parameter. This method was used to acquire the influence of the measured input parameters uncertainty on the output target parameter. The results show that air and steam flow rate, coolant flow rate, inlet and outlet water temperature are the main source of the uncertainty for SCOPE. Inlet film flow rate, inlet air flow velocity and plate surface temperatures are the main source of the uncertainty for WAFT. Sensitivity analysis results may provide support for experiment measurement system optimization to reduce the target parameter error range.

Uncertainty analysis is one important aspect of test data analyses, which is meaningful to the assessment of test results. Conventionally the partial derivative with respect to the input parameters is used to transfer uncertainty from the input parameters to the output parameter. However, in this method the partial derivatives of the output parameter sub the input parameters must be calculated. For complex engineering problems, it is usually difficult to acquire theoretical correlations for the partial derivatives. WILKs formula is used to determine the parameter tolerance interval with certain probability content and confidence level. The tolerance interval is a good way to well describe the uncertainty of parameters. The nonparametric statistics with WILKS correlation were widely used in the best-estimate plus uncertainty (BEPU) accident analyses. However, little work has been conducted on the experiment results uncertainty analysis with that method. In this paper nonparametric statistics with WILKS correlation was used to acquire key parameters uncertainty. And the results show that key output parameters uncertainty for SCOPE and WAFT are within the reasonable range.

Uncertainty Propagation Methods were implied for test results Sensitivity and Uncertainty Analysis in the paper, which may be conveniently applied to the other experiment data analyses and also valuable to the engineering project.
Analysis on Hydrogen Risk Induced by Containment Leakage
Authors: Gu Peiwen , Shi Guobao , Fang Likai , Wang Jiayun
Published: October 2017

In-containment hydrogen, produced during severe accident, can largely challenge the integrity of containment, which has been studied for many years. However, Fukushima Dachii accident has shown the possibility of explosion outside containment due to leaked hydrogen, which was never considered before. In the paper, ex-containment hydrogen analysis is conducted with and without mitigation measures based on the newly-built CAP1400 MAAP4 containment model. Middle annulus is screened as the only location for the possible hydrogen accumulation. Due to steam condensation, hydrogen concentration can be unexpectedly high enough to exceed that in the containment, but either igniters or PARs can mitigate its risk even under conservative conditions. The possibility of containment failure due to hydrogen leakage is very low as long as mitigation measures are functional.
Application of the Probability-Based Safety Analysis for the Reliability Evaluation of a Special Fuel Salt Release System Design in the Molten Salt Reactor
Authors: Qun Yang , Minghai Li , Jie Hou , Zhaozhong He , Kun Chen
Published: October 2017

In the molten salt reactor (MSR), the fuel salt can be discharged out of the core to avoid a more serious consequence in the accident. In this paper, a special core fuel salt release system (CFSRS) is described. Its equipment reliability and human reliability are evaluated subsequently to identify the key failure factors and provide suggestions for the design, by means of the fault tree analysis and the human reliability analysis (HRA) respectively. Results show that the human error is a dominant factor of CFSRS failure in the accident scenario. A good operator training or a sufficient diagnosis time can greatly improve the CFSRS human reliability. Regardless of the human error, to obtain a balanced design and a low equipment failure probability of less than 2E−5, the redundant design of CFSRS including the pipe heating subsystem is necessary.
Numerical Simulation for Effect of Rolling Motion on Thermal Stratification in a Surge Line
Authors: Shen Y. L. , Lu Tao , Liu Bo
Published: October 2017

Pressurizer surge lines are essential pipeline structure in NPPs, and the thermal stratification in surge line is recognized as one of the possible cause of thermal fatigue. In this paper, a Computational Fluid Dynamic (CFD) method has been adopted to simulate temperature fluctuations on the process of temperature rising in a pressurizer surge line under rolling motion of single degree of freedom. This work focuses on a fundamental description of differences of thermal stratification between the surge line rolling around the coordinate X-axis condition and that in a static state. The Large-eddy simulation (LES) model is employed to capture the details of temperature change in surge line. Temperature distributions near the inner wall of a surge line pipe with or without swinging were monitored and compared. The temperature differences between the top and bottom of the pipe sections are employed to represent the maximum temperature differences at all the monitored sections. As the surge line swinging, the pattern of temperature distribution and the length of thermal stratification development are different from that in a static. Fluid temperature fluctuation in surge line occur periodically during the fluid temperature rising when the surge line is rotated with the X-axis, and the temperature difference between top and bottom of the surge line is reduced in the same motion mode compared with the static state.