bluuebunny/crossref_metadata_2025_split huggingface dataset:
Performance Analysis of High Temperature Sensible Heat Thermal Energy Storage Systems for Concentrated Solar Thermal Power Plants
Authors: Mostafavi Tehrani S. Saeed , Shoraka Yashar , Taylor Robert A. , Menictas Chris 
Published: October 2017

Due to their relatively high capital and environmental cost of two-tank molten salt thermal storage systems, a significant amount of research has gone into looking for sensible and latent thermal energy storage alternatives suitable for concentrated solar thermal (CST) plants. Despite a large number of developments in the last decade, comparative studies among promising options have been lacking. In particular, only a few comparative studies are available in which thermal energy storage (TES) systems are integrated as an active subcomponent of CST plant. Therefore, this study compares selected sensible heat thermal energy storage systems based on their integrated performance with other CST components (e.g. a tower -based CST plant with a Rankine cycle) over a year of operation.
In the present study, annual performances of single-medium thermocline (SMT), double-medium thermocline (DMT), and shell-and-tube (ST) system were compared with that of a conventional two-tank molten salt storage system. Concrete with porosity of 0.2 (concrete occupies 80% of the system) was selected as a low cost filler material in the DMT and ST systems. The systems were sized for 15 hours of storage capacity and integrated into a validated 19.9 MWe Gemasolar power plant model with solar multiple of 2.5. Before performing annual integrated simulations, an optimum design of each storage system was selected based on a performance analysis of the storage system over a constant 15 hours discharge. A CST plant with a two-tank molten salt system enables the highest amount of electricity generation in a year followed by the SMT and DMT systems, which resulted in 7% and 9% less electricity generation, respectively. For the CST plant with ST system, 20% less electricity was generated over a year. Overall, this study provides a methodology for the comparison of the TES alternatives, and it gives insight the most promising alternative for replacing two-tank molten salt systems.
Cycle Calculations of a Small-Scale Heat Removal System With Supercritical CO2 As Working Fluid
Authors: Strätz Marcel , Starflinger Jörg , Mertz Rainer , Seewald Michael , Schuster Sebastian , Brillert Dieter 
Published: October 2017

In case of an accident in a nuclear power plant with combined initiating events, (loss of ultimate heat sink and station blackout) additional heat removal system could transfer the decay heat from the core to and diverse ultimate heat sink. On additional heat removal system, which is based upon a Brayton cycle with supercritical CO2 as working fluid, is currently investigated within an EU-funded project, sCO2-HeRo (Supercritical carbon dioxide heat removal system). It shall serve as a self-launching, self-propelling and self-sustaining decay heat removal system to be used in severe accident scenarios. Since a Brayton cycle produces more electric power that it consumes, the excess electric power can be used inside the power plant, e.g. recharging batteries. A small-scale demonstrator will be attached to the PWR glass model at Gesellschaft für Simulatorforschung GfS, Essen, Germany.
In order to design and build this small-scale model, cycle calculations are performed to determine the design parameters from which a layout can be derived.
Refined Algorithm for Solving High-Order Harmonics of Multi Group Neutron Diffusion Equation
Authors: Zhou Gaoxin , Gang Zhi 
Published: October 2017

In recent years, high order harmonic (or eigenvector) of neutron diffusion equation has been widely used in on-line monitoring system of reactor power. There are two kinds of calculation method to solve the equation: corrected power iteration method and Krylov subspace methods. Fu Li used the corrected power iteration method. When solving for the ith harmonic, it tries to eliminate the influence of the front harmonics using the orthogonality of the harmonic function. But its convergence speed depends on the occupation ratio. When the dominant ratios equal to 1 or close to 1, convergence speed of fixed source iteration method is slow or convergence can’t be achieved. Another method is the Krylov subspace method, the main idea of this method is to project the eigenvalue and eigenvector of large-scale matrix to a small one. Then we can solve the small matrix eigenvalue and eigenvector to get the large ones. In recent years, the restart Arnoldi method emerged as a development of Krylov subspace method. The method uses continuous reboot Arnoldi decomposition, limiting expanding subspace, and the orthogonality of the subspace is guaranteed using orthogonalization method.
This paper studied the refined algorithms, a method based on the Krylov subspace method of solving eigenvalue problem for large sparse matrix of neutron diffusion equation. Two improvements have been made for a restarted Arnoldi method. One is that using an ingenious linear combination of the refined Ritz vector forms an initial vector and then generates a new Krylov subspace. Another is that retaining the refined Ritz vector in the new subspace, called, augmented Krylov subspace. This way retains useful information and makes the resulting algorithm converge faster. Several numerical examples are the new algorithm with the implicitly restart Arnoldi algorithm (IRA) and the implicitly restarted refined Arnoldi algorithm (IRRA). Numerical results confirm efficiency of the new algorithm.
The R&D Activities and Future Plan of 4MW Hot Cathode Ion Source for EAST Neutral Beam Injector
Authors: Xie Yahong , Hu Chundong , Liu Sheng , Li Jun , Xie Yuanlai , Liang Lizhen , Jiang Caichao , Wei Jianglong , Xu Yongjian , Tao Ling , Gu Yuming , Chen Shiyong , Sheng Peng , Liu Zhimin 
Published: October 2017

The Experimental Advanced Superconducting Tokamak (EAST) is one of the fully superconducting tokamak, its aim at the long-pulse operation (1000s) to study the physics of steady-state operation for nuclear fusion sciences. In order to support the steady-state operation and physical research, the high power neutral beam injection (NBI) system need to be employed on the EAST for the plasma heating and current driving. According to the scientific study schedule of the EAST, the designed NBI system includes two beam lines which will be constructed in two phases. Each beam line will deliver a deuterium neutral beam with beam energy of 50–80 keV with beam duration of 10–100 s. Each beam line has the maximum beam power of 4MW.
The high current ion source is one of the most important parts in the high power NBI. A hot cathode positive ion based source was developed for EAST-NBI, which shown in Fig. 1. The ion source contains a bucket hot cathode arc chamber with 650 mm long, 260 mm width and 300 mm depth. There are 32 hairpin filaments with diamond of 1.5 mm and 160 mm long to supply primary electrons. A tetrode type accelerator with slit type used to extract the ions from the plasma and accelerated to the desired energy. The beam extraction area is 120 mm × 480 mm (can be changed) with beam transmittance of 60 %. The designed beam species is deuterium with beam power of 2–4 MW and beam energy of 50–80 keV and beam pulse length of 10–100 s.
The ion source needs to be conditioning before operation on the EAST-NBI. An ion source test bed was designed and developed for the ion source performance tests and the optimization verification. The characteristics of ion source were tested with hydrogen beam, and each ion source should achieve 4MW ion beam with beam energy of 80 keV. The optimum beam perveance and arc efficiency were analyzed too. The optimum beam perveance was 2.8 μp with beam energy of 50 keV and the arc efficiency was 0.55A/kw.
Long pulse operation was one of the requests of EAST ion source. The real-time feedback control method was employed and can got stable plasma and ion beam. The beam extraction was tested to achieve 100 s on the test bed. Consider the high power deposited on the calorimeter, the beam was modulated with suitable frequency and duty ratio. When the conditioning finished, the ion sources were moved to the EAST-NBI. The deuterium beam was extracted and injected into the EAST plasma. Details of the performance of positive ion source on the test bed and EAST-NBI will be presented.
Scaling Analysis of Passive Heat Exchanger Under Station Blackout Accident
Authors: Yu-sheng Liu , Chao Xu , Shao-xin Zhuang , Cong-xin Li , Pan Zhang 
Published: October 2017

Station blackout accidents increasingly become the focus of research in the field of nuclear safety after Japan’s Fukushima nuclear plant accident in March 2011. Core decay heat under station blackout condition will be transferred by natural circulation occurring between core and passive heat exchanger for the nuclear plants incorporated passive safety design concept such as AP1000 or CAP1400. As a result, response of safety systems will differ in accident sequence and kind between passive safety plant and traditional plant. What is more, cooling capacity of passive heat exchanger (PHX) which takes on heat sink has significant effect on performance of natural circulation in passive safety system.
The safety need that characteristics of passive safety plant should be verified through integral experiment facility makes scaling analysis important in design or modification of experiment facility. Furthermore, scaling analysis of natural circulation phenomena under station blackout accident plays an important role in design verification, safety review verification or thermo-hydraulic program development. It not only determines the similar similarity criteria between the nuclear power plant prototype and test facility, but also provides technical basis for selecting different experiment schemes.
As a part of scaling analysis on natural circulation phenomena for station blackout, the cooling capacity of PHX in test facility should be scaled properly and reasonably with conservatism. Therefore, scaling of passive residual heat removal (PRHR) heat exchanger under station blackout accident is investigated analytically in this paper. The analytical model for natural circulation in passive heat exchanger is established based on the performance characteristics of PRHR system in passive plant. By proper hypothesis and simplification, the governing equations for PHX are normalized using steady-state solutions, initial or boundary conditions. The similarity criteria that should be preserved between PHXs in test facility and prototype are finally obtained from non-dimensionalized equations. Furthermore, the distortion analysis for PHE design is also investigated based on the similarity criteria for selected scaling factors and parameters. The safety analysis based on models of nuclear power plant prototype and test facility is conducted on transient performance of designed PHX with PHX of prototype. The results show that: heat source number is the dominant similarity criteria for PHXs design under SBO condition. Requirements of Richardson number and friction number could be satisfied by resistance adjusting on test loop. The performance of PHX designed following heat source number requirement can better represent the transient response characteristics of prototype under SBO condition.
An Experimental Study on Critical Heat Flux in Rectangular Channel With Different Angle of Inclination
Authors: Zhang Ming , Zhao Yufeng , Gao Tianfang , Hou Fangxin , Chen Peipei 
Published: October 2017

It is a hot research field to study the Critical Heat Flux (CHF) occurrence mechanism in boiling heat transfer. Although lots of researchers have studied on it, no unified conclusion has been achieved up to now. The proposed CHF occurrence mechanism is also not widely accepted. Because the void fraction close to the heating surface is larger when the heat flux approaches CHF, it is difficult to make visual observations of the boiling heat transfer on the heating surface. Usually the CHF mechanism is based on certain assumptions, and then confirmed by testing. Therefore, it is meaningful to study the occurrence of CHF by experimental methods.
Based on the system design of the test system for CHF in a rectangular channel, an experimental facility was set up. The main test section consists of a rectangular flow channel with the copper heating surface downwards mounted into one of the channel walls. The fluid is deionized water. Fluid subcooling is 15 K. The entire test section is mounted on a rotating arm which can be set at different inclination angles from 0° (horizontal) to 90° (vertical). By adjusting the loop, the natural circulation and forced circulation test conditions can be achieved from the experimental facility. Through test research, visual observations are acquired about the bubble growth characteristics in the rectangular channel in the range of 0 ° to 90 °, and the test data about the CHF is also attained.
It is found that under the condition of forced circulation and natural circulation, the CHF values is increased effectively with the improved mass flow rate and the increased inclination angles of the test section. Through visual observation, it is found that in forced circulation conditions, CHF occurs first at the entrance location of the test surface, rather than the maximal heat flux location of the test surface (test surface center) or the maximal local void fraction of the test facility (exit location of the test facility). In the natural circulation conditions, CHF occurs first at the exit location of the test facility. This phenomenon may imply that the mechanisms of CHF are different in forced circulation and natural circulation. In forced circulation, the flow plays a main role. In natural circulation, the local void fraction plays a main role.
There are some differences in the experimental phenomena compared with the traditional CHF theory, like the bubble crowding theory and the micro fluid layer theory. Through the experiment research, the complexity of the flow boiling heat transfer was found. The flow and boiling heat transfer affect each other. It is can help us to research the CHF theory in flow boiling heat transfer.
Study on Purification of U Scraps in Continuous Countercurrent Extractor
Authors: He Fei , Wan Qiang 
Published: October 2017

In this paper, the feasibility of continuous countercurrent extractor (CCE) applied in extraction of U scraps rather than mixer-settler was studied. Effects of rotation Reynolds number, residence time on extraction and reverse extraction efficiency were investigated respectively. During the process, impurity content in raw material had a slight effect on extraction. Under the optimum conditions of O/A phase ratio of 1.2 and rotation Reynolds number of 13824, extraction degree of scrap pellet solution and alkalescence dreg lixivium were good with the residence time of 19 min and 26 min respectively.
Under the best reverse extraction conditions of A/O phase ratio of 1.2, rotation Reynolds number of 23128 and residence time of 19 min, two kinds of raw material liquid mentioned as above could get better reverse extraction efficiency, while uranium concentration of raffinate phase and impurity content of reverse extract phase satisfied the technological requirement. This study verified the feasibility of CCE developed for purification of U scraps, identified the operation parameter, and provided an experimental basis for industrial expansion in the future.
The Activation Sources of Reactor Internals During Decommissioning
Authors: Han Jianchun , Zhou Yan , Li Hui , Mei Qiliang 
Published: October 2017

As China’s first nuclear power plant connected to the grid, the first Qinshan nuclear power plant is approaching the decommissioning period. Other nuclear power plants also turn into the preparation phase of decommissioning in succession. In order to facilitate decommissioning, source survey is conducted during the pre-decommissioning phase, which can provide radioactive inventory, contamination distribution, species and quantities of nuclides.
The internals of the reactor work under the most severe radiation environment. During the reactor operation, the materials of internals are irradiated by high-energy neutrons. So activated nuclides are generated due to the neutron capture reaction, which are the main radioactive waste to be treated during decommissioning. In this paper, the neutron irradiation and the generated activation source of the internals for pressurized water reactors (PWR) are studied and analyzed.
Firstly, core modeling was carried out, and the neutron transport calculation is performed to obtain three-dimensional distribution of the neutron flux.
Secondly, according to the three-dimensional distribution of the material composition and the neutron flux rate of the reactor, the activation calculation is carried out to obtain the activation source.
Numerical Simulation of Flow Instability in Vortex Diodes
Authors: Wang Junrong , Xiao Qi , Ke Hanbing , Hu Xu , Li Shaodan , Wei Zhiguo 
Published: October 2017

A vortex diode is used as a highly reliable check-valve in nuclear applications, where it mainly benefits from the intrinsic properties of no moving parts and no leakage. Its basic principle is similar to the diode in an electric circuit. The typical structure of a vortex diode consists of a chamber with axial and tangential ports. When the fluid is injected through the axial port, a simple radial flow in the chamber leads to a relatively low flow resistance. On the other hand, in the reverse flow mode, a strongly swirling vortex can be set up in the chamber, resulting in a very high flow resistance.
Several experimental studies found vortex-induced vibration of a vortex diode in the reverse flow mode, where it indicated that the flow was unstable in the vortex diode. This phenomenon may affect the reliability of the vortex diode. However, the mechanism has not been investigated systematically and profoundly. In this paper, 3-D simulations are carried out to help understand the related flow characteristics in the vortex diode. Standard k-ε model was selected for forward flow, while Reynolds stress model was selected for reverse flow. We have found that the results from transient simulations are in good agreement with experimental data. The transient simulations also capture the periodic pressure fluctuation in the vortex diode. Vortex diodes with different structures and geometrical parameters are simulated at different Reynolds number conditions. It is found that the characteristics of the pressure fluctuation are determined by the structure parameters and working conditions of the vortex diode. The flow instability is mainly caused by the asymmetry of the vortex diode. The work presented in this paper will be useful to give better understanding of flows in vortex diodes and to provide some guidance for optimizing the vortex diode.
Component and Integration Test of an FPGA-Based PWR Protection Sub-System Using UVM
Authors: Shui Xuanxuan , Wu Yichun , Zhou Junyi , Cai Yuanfeng 
Published: October 2017

Field programmable gate arrays (FPGAs) have drawn wide attention from nuclear power industry for digital instrument and control applications (DI&C), because it’s much easier and simpler than microprocessor-based applications, which makes it more reliable. FPGAs can also enhance safety margins of the plant with potential possibility for power upgrading at normal operation. For these reasons, more and more nuclear power corporations and research institutes are treating FPGA-based protection system as a technical alternative.
As nuclear power industry requires high reliability and safety for DI&C Systems, the development method and process should be fully verified and validated. For this reason, to improve the application of FPGA in NPP I&C system, the specific test methods are critical for the developers and regulators. However, current international standards and research reports, like IEC 62566 and NUREG/CR-7006, which have demonstrated the life circle of the development of FPGA-based safety critical DI&C in NPPs, but the specific test requirements and methods which are significant to the developers are not provided.
In this paper, the whole test process of a pressurized water reactor (PWR) protection sub-system (Primary Coolant Flow Low Protection, Over Temperature Delta T Protection, Over Power Delta T Protection) is described, including detail component and integration tests. The Universal Verification Methodology (UVM) based on System Verilog class libraries is applied to establish the verification test platform. All these tests are conducted in a simulation environment. The test process is driven by the test coverage which includes code coverages (i.e., Statement, Branch, Condition and Expression, Toggle, Finite State Machine) and function coverage. Specifically, Register Transaction Level (RTL) simulation is conducted for Component tests, while RTL simulation, Gate Level simulation, Timing simulation and Static timing analysis are conducted for the integration test. The issues (e.g., the floating point calculation, FPGA resource allocation and optimization) arose in the test process are also analyzed and discussed, which can be references for the developers in this area.
The component and integration tests are part of the Verification and Validation (V&V) work, which should be done by the V&V team separated from the development team. The testing method could assure the test results reliable and authentic. It is practical and useful for the development and V&V of FPGA-based safety DI&C systems.