bluuebunny/crossref_metadata_2025_split huggingface dataset:
The Impact of Religion on Voting Behaviour – A Multilevel Approach for Switzerland
Authors: Goldberg Andreas C. 
Published: February 2014

AbstractThe traditional cleavages like social class or religion are often reported to have lost explanatory power for the voting decision. Regarding the religious cleavage, the evidence is ambiguous depending on the choice of cases and indicators used. The present study tests the impact of religion for the preference to vote the Christian Democratic Party of Switzerland (CVP) using data from the 2007 and 2011 Swiss national elections. Additional to the inclusion of individual variables, a special focus lies on contextual effects. The estimated hierarchical linear models confirm a prevailing influence of the simple individual factors and the presence of a significant contextual effect. Statistical evidence is also presented for some of the supposed interaction effects between individual and contextual religious variables.
Opportunity and Challenge of MOX Fuel Application in China
Authors: Li Rui , Zhou Yuemin 
Published: February 2014

In this paper, the opportunity and challenge for application of MOX fuel in China are discussed. With the great demand of energy to develop economy of China, the government plans to achieve 58GW installed capacity and 30GW installing capacity in nuclear power before 2020. China needs about 81,306t～102,613t nature Uranium to support the nuclear power developing project at that time. For the obvious reason, China should insist on closed fuel cycle to recover the fissionable materials in the spent fuel. China will build its first spent fuel reprocessing factory to supply MOX fuel to PWRs and FRs. Both government and nuclear companies put more and more enthusiasm into the application of MOX fuel.
However, China is also suffered from great challenges of MOX fuel application, such as the spent fuel reprocessing technique roadmap, the manufacture technology of MOX fuel, the modification of PWRs to accommodate the MOX fuel, and the recalculation of nuclear parameters by design procedure. The economic effect of applying MOX fuel is also considered. At last, the prospect of MOX fuel application is predicted in this paper.
Design Development of a Nuclear Standard Bidirectional Solenoid Valve to Enable PWR Fluid System Simplification
Authors: Mann Chris 
Published: February 2014

Rolls-Royce has designed and manufactured Pressurised Water Reactors (PWRs) for over 50 years. This paper describes the design development of a nuclear standard bidirectional solenoid valve that can be used to simplify a valve actuating system for a PWR. In PWRs, hydraulic actuation systems to operate key plant valves can be complex, often comprising of a combination of electrical and pneumatically actuated valves. A typical arrangement uses solenoid-controlled pneumatic valves to operate hydraulically actuated valves.
A bidirectional solenoid valve was developed to eliminate the requirement for pneumatic control valves and their associated infrastructure, allowing direct solenoid valve control of the actuating water. The elimination of the pneumatic system affords benefits in terms of cost and maintenance. The solenoid valve design was developed from an existing design with proven history in the petrochemical industry. The concept design was chosen for its simplicity, but required significant development in order to meet the requirements for a PWR application.
This paper highlights some of the key developments and gives an overview of the verification and validation activities undertaken to demonstrate the functional viability of the concept design.
Study on the SBO Coping Capability for Qinshan Phase II Unit 3
Authors: Deng Jian , Chen Bin , Deng Chunrui , Zhang Ming 
Published: February 2014

The accident in Japan on March 2011 was caused by a natural event (i.e., earthquake, and the induced tsunami) which was far more severe than the design basis for the Fukushima Dai-ichi nuclear power plant. The reactor cores of Dai-ichi unit 1 through 3 have been degraded due to the long term station blackout (SBO). Actually, a nuclear power plant is able to prevent the core damage despite the SBO event occurs, called “coping capability” or “coping time”.
In this paper, the SBO coping capability is assessed for Qinshan II unit 3 based on the following: 1) reactor coolant inventory, 2) condensate inventory for decay heat removal, 3) class 1E battery capacity, 4) compressed air, 5) effects of loss of ventilation, 6) containment isolation. As a result, the SBO event tree is suggested to be updated and a few vulnerabilities have been identified and some modifications are proposed.
Numerical Simulation of Quasi Periodic Large Scale Vortices in Rod Bundles With Non-Uniform Wall Roughness
Authors: Hu Gaojie , Yu Lei , Zhang Yangwei , Yan Binghuo 
Published: February 2014

The flow and heat transfer in rod bundles are of vital importance in the reactor design. Studies on the turbulent flow in tight lattice have presented the quasi periodic large scale vortex structure is highly accounted for the local flow and heat transfer and turbulent mixing in rod bundles.
In this work, the Unsteady Reynolds Averaged Navier–Stokes (URANS) simulation with the Reynolds Stress Model (RSM) is adopted to simulate the flow and vortex structure in rod bundles. The vortex structure in sparse lattice with nonuniform wall roughness is also investigated and compared with that in tight lattice. The results indicate that the quasi periodic large scale vortex structure and flow pulsation are available in both tight lattice and the sparse lattice with non-uniform wall roughness. Due to the existence of the vortex structure, the heat transfer in spare lattice is enhanced.
Response Functions of a Cs2LiYCl6 Scintillator to Neutron and Gamma Radiation
Authors: Khan Nafisah , Machrafi Rachid , Kovaltchouk Vitali 
Published: February 2014

A new scintillator, CLYC, has been investigated for possible use in neutron spectrometry. This sensor provides neutron detection for both thermal and fast neutrons from the reactions 6Li(n,α) and 35Cl(n,p), respectively. This work primarily focuses on the detection of fast neutrons since there is currently no sensor that can accurately and efficiently provide information about their incident neutron energy. Conventional methods of fast neutron detection have been based on utilizing materials that use the elastic scattering process of neutrons on 1H to create recoil protons or by thermalizing and capturing these neutrons at thermal energies. Both approaches have drawbacks and are complex in deriving the energy spectrum through the unfolding process. The CLYC scintillator uses a distinct proton peak, whereby the position on the spectrum is proportional to the energy of the incident neutron.
The response function of this detector has been simulated using Monte Carlo N-Particle eXtended code (MCNPX) for gamma-rays and neutrons of different energies. The obtained data has been discussed and analyzed.
Prospect of MAO Technology Application in Nuclear Power Industry
Authors: Zhao Zheng , Feng Beibei , Yang Xingtuan , Sun Yanfei 
Published: February 2014

Micro arc oxidation (MAO) technology known as a newly surface treatment technology has got a widely application in the field of aviation, aerospace, automotive, electronics, and medical industry. Strength, toughness, hardness and corrosion of valve metal such as aluminum, magnesium, copper, zinc, zirconium and their alloys can be greatly improved by MAO technology.
This paper tries to probe into the feasibility of using MAO technology in nuclear power industry. Aluminum and its alloys are used as structural materials such as the cladding of reactor fuel and all kinds of pipes in the low nuclear reactor. Zirconium alloys are widely used for the fuel cladding, cannula, catheter and other components of the fuel assemblies. Titanium and its alloys offer a unique combination of desirable mechanical properties which makes them to be the candidate materials for structural application in the field of nuclear energy. The surface of all these materials may be destroyed which increasing the risk of the nuclear accident due to the severe serving conditions. As a result, it is necessary to improve the corrosion and wear resistance behavior. With the urgent requirements of safety and durability of nuclear reactor, MAO technology must have a broad prospect in nuclear industry.
Development of Prediction Technology of Two-Phase Flow Dynamics Under Earthquake Acceleration: 8 — Measurement of Velocity Profile Around Bubble Under Structure Vibration
Authors: Mizuno Kousuke , Arai Rie , Kaneko Akiko , Monji Hideaki , Abe Yutaka , Yoshida Hiroyuki , Takase Kazuyuki 
Published: February 2014

In a nuclear power plant, one of the important issues is evaluation of the safety of reactor core and its pipes when an earthquake occurs. Many researchers have conducted studies on constructions of plants. Consequently, there is some knowledge about earthquake-resisting designs. However the influence of an earthquake vibration on thermal fluid inside a nuclear reactor plant is not fully understood. Especially, there are little knowledge how coolant in a core response when large earthquake acceleration is added.
Some studies about the response of fluid to the vibration were carried out. And it is supposed that the void fraction or the power of core is fluctuated with the oscillation by the experiments and numerical analysis. However detailed mechanism about a kinetic response of gas and liquid phases is not enough investigated, therefore the aim of this study is to clarify the influence of vibration of construction on bubbly flow structure. In order to investigate it, we visualize changing of bubbly flow structure in pipeline on which sine wave is applied.
Bubbly flow is produced with injecting gas into liquid flow through a horizontally circular pipe. In order to vibrate the test section, the oscillating table is used. The frequency of vibration added from the table is from 1.0 Hz to 10 Hz and acceleration is from 0.4 G to 1 G (1 G = 9.8 m/s2). The test section and a high speed video camera are fixed on the table. Thus the relative velocity between the camera and the test section is ignored. In the visualization experiment, the PIV measurement is conducted. Then the motion of bubbles, for example the shape, the positions and the velocity are measured with observation.
In addition, by varying added oscillation amplitude, frequency and flow rate of the fluids, the correlation between these parameters and bubble motion was evaluated.
It was clarified that the characteristic bubble deformation and velocity profile around bubble is caused by oscillation in each oscillation phase time. When structure vibration affects the flow, two main mechanisms are supposed. One is the addition of body force of the oscillation acceleration to liquid phase and bubble, and the other is the velocity oscillation of the test section and the effect of the boundary layer of the pipe wall. It was also found that the bubble deformation is correlated with the fluctuation of relative liquid velocity field to bubble and pressure gradient in the flow area.
Development and Verification of Severe Accident Simulation Capability for a NPP Simulator
Authors: Yang Zhifei , Xie Xiaofei , Chen Xing , Zhang Shishun , Liao Yehong , Peng Zhenxun , Lin Jiming , He Dongyu , Chen Yali 
Published: February 2014

It is reflected in the severe accident in Fukushima Daiichi that the emergency capacity of nuclear power plant needs to be enhanced. A nuclear plant simulator that can model the severe accident is the most effective means to promote this capacity. Until now, there is not a simulator which can model the severe accident in China. In order to enhance the emergency capacity in China, we focus on developing a full scope simulator that can model the severe accident and verify it in this study.
The development of severe accident simulation system mainly includes three steps. Firstly, the integral severe accident code MELCOR is transplanted to the simulation platform. Secondly, the interface program must be developed to switch calculating code from RELAP5 code to MELCOR code automatically when meeting the severe accident conditions because the RELAP5 code can only simulate the nuclear power plant normal operation state and design basis accident but the severe accident. So RELAP5 code will be stopped when severe accident conditions happen and the current nuclear power plant state parameters of it should be transported to MELCOR code, and MELCOR code will run. Finally, the CPR1000 nuclear power plant MELCOR model is developed to analyze the nuclear power plant behavior in severe accident.
After the three steps, the severe accident simulation system is tested by a scenario that is initiated by the station black out with reactor cooling pump seal leakage, HHSI, LHSI and auxiliary feed water system do not work. The simulation result is verified by qualitative analysis and comparison with the results in severe accident analysis report of the same NPP. More severe accident scenarios initiated by LBLOCA, MBLOCA, SBLOCA, SBO, ATWS, SGTR, MSLB will be tested in the future. The results show that the severe accident simulation system can model the severe accident correctly; it meets the demand of emergency capacity promotion.
Shielding Design and Analysis for the Proton Tube of China Lead-Alloy Cooled Research Reactor (CLEAR-I)
Authors: Li Bin , Yang Qi , Zou Jun , Zeng Qin , Bai Yunqing 
Published: February 2014

In China Lead Alloy Cooled Research Reactor (CLEAR-I), which is an accelerator driven subcritical system, a proton tube inserts into core from the top of reactor. Some neutrons will leak out along the proton tube, inducing structure materials damage and potential exposure danger for workers.
The shielding design and optimization analysis for the proton tube of CLEAR-I was conducted using the VisualBUS system developed by FDS Team in this study. The result indicated that during operation allowed working time at outside containing compartment above the reactor shall not exceed 40 hours per week and entering into containing compartment was forbidden. Above results showed that the design after optimization basically met the requirements of shielding.
3-D Base Control Systems for the Seismic Protection of Power Plant Equipment and Buildings
Authors: Nawrotzki Peter , Wang Weiqiang 
Published: February 2014

Safety in nuclear facilities is of special importance and recent seismic events and the resulting damage in these facilities again bring up the discussion. One of the latest events is the 2007 Chuetsu earthquake in Japan. The arising damage in the Kashiwazaki Kariwa Nuclear Power Plant can be found in several reports like (Fukushima, 2007) and (Yamashita, 2008).
Base-Control Systems (BCS) consist of Spring-damper combinations with specifically designed stiffness and damping properties which can be described in the vertical and both horizontal directions. Nowadays these 3 dimensional supports are frequently used to reduce seismic acceleration levels of machine components, equipment and buildings as safety or non-safety related structures. They may also be applied to protect sensitive equipment or machinery against other catastrophic events, i.e. airplane impact which is of particular importance in nuclear facilities.
By presenting specific executed examples, this article shall also underline the efficiency of the BCS which, in contrary to purely horizontally efficient Base-isolation Systems with rubber (BIS), is not extensively discussed in literature. Decisive characteristics of the shown applications are provided and the benefits of using this technology for machine or equipment manufacturers, designers or owners are explained.
JiNan MNSR Decommissioning
Authors: Li Yiguo , Peng Dan , Zhang Yongbao , Lu Jin , Wu Xiaobo , Zhang Jinhua , Zou Shuyun 
Published: February 2014

JiNan MNSR (Miniature Neutron Source Reactor) with thermal power 30kW is an undermoderated reactor of pool-tank type, and UAL4 alloy with 235U enrichment of 90% as fuel, light water as coolant and moderator, and metallic beryllium as reflector. The fission heat produced by the reactor is removed by the natural convection. JiNan MNSR went critical in May 1989, and shutdown in March 2008. The reactor operated 19 years safely with the accumulated power of 8.82×104 kWh. The main use was NAA (Neutron Activation Analysis) work.
The spent fuel was removed from the JiNan MNSR site in Nov. 2009 already. The Jinan MNSR decommissioning documents were approved by Chinese competent authorities in Dec. 2010. The work on site was started in Jan. 2011, and completed in March 2011, the wastes were removed from the Jinan MNSR site. The decommissioning scope of JiNan MNSR includes decommissioning of its related systems, equipment and structures. The weight of intermediate level radioactive is 204.1kg, and the weight of low level radioactive waste is 2297 kg. They were all transported to West-north Waste disposal plant.
The final measuring results are that α and β surface contamination in the reactor building are less than 0.06Bq/cm2 and 0.5Bq/cm2 respectively; the specified activities of the nuclides of 137Cs, 60Co and 90Sr in the soil are less than 1.4 Bq/kg,1.7 Bq/kg and 8.14 Bq/kg respectively; the specified activities of the nuclides of 137Cs, 60Co and 65Zn in the solid sample in the reactor building are less than 5.7 Bq/kg,7.8 Bq/kg and 1.7 Bq/kg respectively.
The specified values for α and β surface contaminations in the reactor building are 0.08Bq/cm2 and 0.8Bq/cm2 respectively; they are 12 Bq/kg, 3 Bq/kg and 10 Bq/kg respectively for the limited values of the specified activities of the nuclides of 137Cs,60Co and 90Sr in the soil; the limited values of specified activities of the nuclides of 137Cs,60Co and 65Zn are 100Bq/kg respectively in the solid samples in the reactor building. The measuring results show that JiNan MNSR site meets the requirements of the unrestricted use.
Analysis of Liquid Sloshing in an In-Containment Refueling Water Storage Tank for AP1000
Authors: Liu Yu , Lu Daogang , Dang Junjie , Wu Licun , Ma Wenhui 
Published: February 2014

Although much work has been performed on the liquid sloshing inside simple structures like rectangular and cylindrical vessels, this paper deals with the analysis of the liquid sloshing in a more complex structures, the in-containment refueling water storage tank (IRWST). The IRWST is an important component of AP1000 passive core cooling system to ensure the safe operation of the AP1000 nuclear power plant. In postulated non-LOCA events, the water in the IRWST absorbs the residual heat then transfers the heat into the containment atmosphere. However, in the case of earthquake, the sloshing fluid may influence the safety of the appropriate semi-cylindrical IRWST.
In this paper, the liquid transient response in the IRWST was formulated based on finite element modal analysis when the three resonance sine wave was applied as excitation. The result shows that the maximum wave excited by excitation from different directions always emerges from the corner or the edge of tank. Another finding is that water will not overflow at the normal operational water level when exited by the selected excitations in any directions. The safety of the IRWST’s roof is achieved which guarantees that the water vapor and radioactive gases within the tank during normal operation will not release to atmosphere in the containment. The influence of the excitation direction and the water depth are also analyzed.
Experiment and Numerical Analysis of Control Method of Natural Circulation by Injection of Helium Gas
Authors: Yanagawa Naoto , Nomura Masashi , Takeda Tetsuaki , Funatani Shumpei 
Published: February 2014

This study is to investigate a control method of the natural circulation of the air by the injection of helium gas. A depressurization is the one of the design-basis accidents of a Very High Temperature Reactor (VHTR). When the primary pipe rupture accident occurs in the VHTR, the air is predicted to enter into the reactor pressure vessel from the breach and oxidize in-core graphite structures. Finally, it seems to be probable that the natural circulation flow of the air in the reactor pressure vessel produce continuously. In order to predict or analyze the air ingress phenomenon during the depressurization accident of the VHTR, it is important to develop the method for prevention of air ingress during the accident. In this study, the air ingress process is discussed by comparing the experimental and analytical results of the reverse U-shaped channel which has parallel channels.
The experiment of the natural circulation using a circular tube consisted of the reverse U-shaped type has been carried out. The vertical channel is consisted of the one side heated and the other side cooled pipe. The experimental apparatus is filled with the air and one side vertical tube is heated. A very small amount of helium gas is injected from the top of the channel. The velocity and the mole fraction of each gas are also calculated by using heat and mass transfer numerical analysis of multi-component gas.
The result shows that the numerical analysis is considered to be well simulated the experiment. The natural circulation of the air has very weak velocity after the injection of helium gas. About 780 seconds later, the natural circulation suddenly produces. The natural circulation flow of the air can be controlled by the method of helium gas injection. The mechanism of the phenomenon is found that mole fraction is changed by the molecular diffusion and the very weak circulation.
Investigation of Conditions for Zinc Particle Release and Transport in the Reactor Core
Authors: Renger Stefan , Alt Sören , Kästner Wolfgang , Seeliger André , Fiss Daniel 
Published: February 2014

Background of the experimental and methodical work is a loss of coolant accident (LOCA) in a Pressurized Water Reactor (PWR). During LOCA, both the leakage water jet as well as the water in the containment can lead to corrosion of galvanized installations (e.g. grates). That means a change of consistency and chemical properties of the coolant water during the sump recirculation operation, in particular an increase of the zinc concentration (Zn2+ ions in dissolution) and the pH-value. Furthermore, it leads to separation of layer-forming zinc compounds (e.g. zinc borate) at hotspots (fouling at fuel assemblies) and/or the crystallization of zinc corrosion products out of the coolant in hot areas [HOF10]. This in turn causes a decrease of pH-value and, consequently, an increase of the corrosiveness of the coolant flow. This can come along with a chemical transformation of zinc borate to nearly indissoluble zinc oxide by thermo hydrolysis. Spalling parts of the coating on the surface of the fuel rods could lead to the release of particles into coolant water flow. In combination with the release of fibrous insulation material, which can be transported into the reactor containment, containment building sump and connected systems, this process causes pressure build-ups at the strainers and the spacers of the fuel assemblies and a reduction of the pump performance. In this case, efficient heat dissipation cannot be guaranteed.
In a joint research project between the Institute of Process Technology, Process Automation and Measuring Technology (IPM) of University Zittau/Goerlitz and the Helmholtz-Zentrum Dresden-Rossendorf (HZDR), the influence of different boundary conditions like the flow structures in the reactor pressure vessel, temperature distributions and the temperature gradient in the central core regions will be investigated. Therefore, in a final step a test facility containing a 3×3 heated fuel rod dummy configuration was designed and constructed. The design and development of the test facility was accompanied by pre-test calculations and CFD simulations to get information about local surface and fluid temperatures under different flow conditions.