bluuebunny/crossref_metadata_2025 huggingface dataset:
Effect of Internal Crossflow on Double Jet Film Cooling Holes of Helium Turbine Blades
Authors: Sun Xiao-kai , Peng Wei , Wang Jie , Jiang Pei-xue
Published: October 2017

In the present study, effect of internal crossflow on double jet film cooling holes were investigated. In our previous study, four types of film cooling holes was investigated and the double jet holes showed better cooling efficiency. However, in practice, the coolant in high temperature turbine blade has a very high flow rate and turbulence, so it is necessary to take the effect of coolant supply condition at the hole entrance into consideration. So the influence of coolant supply condition was investigated, the plenum condition and the crossflow condition were studied to show the influence of the coolant supply condition. And the cylindrical hole was also investigated as comparison. The results showed that there is a profound effect of how the coolant is supplied to the hole on the film-cooling performance in the near hole region. Therefore, crossflow at the hole entry side has be taken into account when modeling film-cooling.
Public Acceptance of Spent Fuel Reprocessing Project
Authors: Wang Zijian , Huang Shanfang , Guo Xiaoyu , Wang Kan
Published: October 2017

At present, there are hundreds of nuclear power plants in operation around the world. Anti-nuclear movements continue in many places, although the nuclear power plants have good operating records. It has some factors, and the first factor that the public knows little about nuclear industry, results in regarding the nuclear power plant mysterious. This condition relates to destructive scene by nuclear weapon with nuclear industry, deeming it unacceptable to take this risk. Secondly, construction of nuclear power plant and off site emergency may occupy large land. The public hopes to be rewarded more to offset the risk by their imagination. Last, it relates to the political environment of one country. Every country has its own situation, so the strategies of developing nuclear power plant are widely different. The public is not familiar with other nuclear engineering projects except nuclear power plants, and hence the boycott happens more frequently. Sino-French cooperation on nuclear fuel cycle project is the first large-scale commercial spent fuel reprocessing plant, which is the biggest cooperative project between China and France until now. AREVA is responsible for technology, and CNNC is responsible for building. Spent fuel reprocessing is the most important part of nuclear fuel cycle back end, which separates uranium and plutonium from spent fuel, and manufactures MOX fuel with recycled resources for using in nuclear reactor again. This will make the best use of the uranium resources. After that process, the fission products needed to be disposed reduce significantly. And it is good for environmental protection. The public protest happened in one of the candidate sites, when CNNC carried out the preliminary work of site selection. For meeting the enormous energy demands, the fossil energy may be exhausted in the future due to the greenhouse gases emission. Chinese government speeds up the development of new energy. Nuclear energy is the only technology with no emission of greenhouse gases and will be rapidly developed. Along with the nuclear power units continuing to increase, they become the critical factors in restricting the sustainable development of nuclear energy. That is efficient utilization of uranium resources, spent fuel intermediate storage, reprocessing, and geologic disposal of high level radioactive waste. To this project, it not only has a great current demand, but also closely relates to transition of energy structure. The public has different views in the project progressing, which results in wide concern and discussion. The article took this event for example, and analyzed the reason from all directions. Besides, the author put forward own views for the public acceptance events about nuclear engineering projects except nuclear power plant.
Obstructed View Factor Calculations in Closed Cavities Using Radiation Heat Transfer
Authors: Mahmood Fiaz , Hu Huasi
Published: October 2017

Because of relaxed requirements on laser-beam uniformity and reduced sensitivity to hydrodynamic instabilities, the ICF Program has concentrated the most of its effort, on the X-ray or indirect drive approach to ICF since 1976. In indirect drive ICF, the driver energy, from laser beams or ion beams, is first absorbed in a high-Z closed cavity which surrounds the capsule, called “hohlraum”. The material heated by the driver emits X-rays, which ultimately drive the capsule implosion. For optimally designed targets, almost 70%–80% of the driver energy can be converted to X-rays. Our main focus is on the view factor calculations for hohlraum geometry of the laser driven indirect ICF approach. This analysis is very important to symmetrically irradiate the ICF capsule. View factor calculations has different radiation heat transfer applications, especially, for indirect drive ICF. In the present work, the obstructed view factor calculations for a square geometry are performed using crossed strings method. For view factor evaluation of the radiation heat transport calculations, different methods are being used. The crossed strings method is much more reliable for simple geometries among the available view factor evaluation techniques. Different cases have been studied for the view factor calculations. These cases include the view factor calculations; if radiations are emitted either from horizontal or vertical axis and a special case for the obstructed view factor calculations. When a third surface is involved in the two already existing surfaces the view factor is said to be the obstructed one rather than the unobstructed ones. Shadow effect method is used for the obstructed view factor calculations. On the other hand, obstructed view factors are computed by subtracting the view factors to the shadows of the obstruction polygons from the unobstructed view factor. The multiple overlaps could become quite complicated while calculating obstructed view factors. Two distinct computer programmes are developed and improved using FORTRAN-90 to evaluate obstructed and unobstructed view factors, for a square geometry. Radiation emission, from both horizontal and vertical surfaces, is considered for calculation of obstructed and unobstructed view factors. The programmes are liabile of calculation for the view factors of variable length of the square square geometry in both horizontal and vertical directions, respectively. In the case of obstructed view factors calculations, certain degree of error was expected. Bench mark of the developed programmes is performed using “Summation Rule”. For the error estimation the unobstructed and obstructed view factors are compared for the similar geometry and found in good agreement, showing the accuracy and reliability of the programmes.
Buoyancy Effect on Heat Transfer in 5×5 Rod Bundles
Authors: Liu Da , Gan Fujun , Zhang Chaozhu , Gu Hanyang
Published: October 2017

Experiments of heat transfer at low flow rate are performed in a 5×5 square arrayed rod bundles. The diameter of the rod is 10mm with a pitch of 13.3mm, length of the test section is about 3 meters. Inlet Reynold number ranges from 2000 to 30000, Bo * ranges from 4×10−6 to 5×10−3. The rods are heated using a DC power, the heat flux ranges from 30 to 300 kW/m2. The experiment is aimed to investigate the buoyancy effect of mixed convection in rod bundles. The experimental data shows that similar with mixed convection in circular channels, buoyancy force has great effect on heat transfer at mixed convection regime in rod bundles. But the buoyancy effect appears at higher Bo* conditions. The spacer effect have also been investigated at both turbulent forced convection regime and mixed convection regime. The reconstruction of heat transfer downstream of spacers is different at different flow regimes, a reasonable explanation was provided.
Research and Design of the Sodium Steam Filter System in Large Liquid Metal Cooled Fast Reactors
Authors: Leijian Wang , Yuan Lu , Changzhi Xiao
Published: October 2017

This reactor uses liquid sodium as coolant owing to its good thermal physical properties, high boiling point and compatibility of cladding material. However, the sodium has a very active chemical properties, for which the free surface of sodium must be protected by inert gas. In the high temperature environment, the sodium atoms diffusion to cover gas slowly, forming a mixed atmosphere that contained large amount of sodium steam.

Sodium steam is covered with the free surface of sodium. Then metal sodium will solidify in the inner wall of the pipe or correlative valves with the reduced temperature. This reactor needs to collect and filter sodium steam in order to reduce the hazards to the equipment, piping system, valves and the other devices. Based on the previous research about the purification process of sodium, this paper compared different steam trapping filtration process and carried out the thermal calculation providing basis for research and design of large sodium cooled fast reactor sodium steam trapping filtration process and establishing a reliable sodium steam filtering system.
Dependency Analysis of Critical Parameters Influencing the Reliability of Thermal-Hydraulic Nuclear Passive Systems
Authors: Olatubosun Samuel Abiodun , Zhang Zhijian
Published: October 2017

The deployment of passive safety systems in nuclear applications especially in advanced nuclear power reactors (both evolutionary and innovative designs) is on the rise. This can be linked to the simplicity, economic and less dependence on human interventions attributes of those passive systems. The reliability of nuclear passive systems especially the thermal-hydraulic ones is influenced by parameters which are interdependent in reality. As a result, the need to critically consider the synergetic effects of determinants of reliability of the thermal-hydraulic nuclear passive systems is of utmost importance. Reliability methodologies are now being modified by factoring the dependency nature of those determinants into reliability analysis to obtain more realistic and accurate results. This paper thus focused on the introduction of more influencing factors in parameters dependency consideration of phenomenological reliability using multivariate distribution analysis. A passively water cooled steam generator was used to demonstrate the interdependency effects of some selected critical parameters. The results obtained justified the need for considering the dependency effects of these parameters influencing the reliability of thermal-hydraulic passive systems. In addition, the research issues on dependency consideration of influencing parameters in evaluation of reliability of these nuclear passive systems were also discussed.
The Behavior of Aerosol Particle Inside a Rising Bubble During Pool Scrubbing
Authors: Fujiwara Kouta , Abe Yutaka , Kaneko Akiko , Yamamoto Kazuki
Published: October 2017

In severe accidents of nuclear power plants such as the Fukushima Daiichi Unit 1 accident, aerosol containing massive amount of Fission Products (FP) could be created. Pool scrubbing is an effect of capturing these FP inside the water during the flow of aerosol. Such a situation may occur at the vent of containment vessel in the BWR as well as the heat-transfer-tube disrupted accident of steam generator in the PWR. The amount of FP released into the atmosphere could be reduced by the effect of pool scrubbing. Hence, the removal of aerosol by pool scrubbing is expected to evaluate the source terms stated below.

Although the importance of pool scrubbing has been recognized, the study for its mechanisms from the viewpoint of particle-bubble and two phase flow had not been carried out intensively. Furthermore, lack of experimental data for various conditions should be improved. Such a state leads to prevent us from analyzing the efficiency of pool scrubbing when advancing the model and testing the validation of it.

The aim of this study is to clarify the transfer of aerosol and bubble experimentally. The data from the experiment evaluates the influence of each parameter to remove the FP. The analysis code presently used adopts the model for the velocity which affect the particle to go out from the bubble. Each of them are written as (i) centrifugal deposition velocity, (ii) gravitational velocity, (iii) incoming vapor velocity, (iv) Brownian diffusion velocity. We choose our parameter which correspond to each of the velocities. Remarking that the experiment also include conditions e.g. the temperature stratification, the types of particle and gas, which is not been described in the manual. Based on the experimental result, our goal is to test validation and improve the model appropriate to the severe accidents predictive code such as the MELCOR by reflecting the study of flow mechanism and the effect of various conditions. To the end, this paper focuses on the rotation and its movement of aerosol particles inside the bubble.

This study focuses on the behavior of a single air bubble including aerosol by visualizing the inner flow of bubble containing particle. As a result for visualization of a single rising bubble, we were able to take a film of an area inside the particle where information of particle movement is seen. We also succeed in observing a clockwise rotation flow inside bubble just above the nozzle. As to avoid the refraction between air-water, we also made an experiment against rising oil droplets. From the series of experiment, the growth of upward flow from desorption to the stationary rising mode was been seen inside the droplet. As the flow grew, a top-to-bottom rotation was seen as it was mentioned in previous models.
Stability of SPPs in Zirconium Alloy CZ2 Under Self-Ion Irradiation
Authors: Wang Xu , Tan Jun , Chen Liutao , Zou Hong , Wen Dungu , Xu Yang , Gao Changyuan , Deng Yongjun , Zhou Yuemin , Wang Lumin
Published: October 2017

Zr-Nb zirconium alloy was developed at CNPRI and designated as CZ2. In this paper, the composition, size distribution and crystal structure of second phase particles (SPPs) in CZ2 have been analyzed under transmission electron microscopy (TEM). Two types of SPPs were observed in CZ2, which are bcc β-Nb and hcp Zr-Nb-Fe. To study the radiation effects on SPPs, CZ2 in two different heat treatment conditions CZ2-SRA (stress relieved) and CZ2-RXA (recrystallized) were irradiated with 3 MeV Zr++ up to 1.14×1016 Zr/cm2 (50 dpa at the damage peak) at three different temperatures 320°C, 360°C and 400°C in Texas A&M University. Commercial low-tin Zr-4 was also irradiated at the same condition as a reference. The results show that the SPPs of Zr-4 become completely amorphous after irradiation at 320°C and 360°C, while retain crystalline at 400 °C. SPPs of CZ2-RXA are partially disordered only after irradiation at 320°C. The crystal structure of SPPs in CZ2-SRA survived at all temperatures.
Modeling of Fuel Rod Behavior and Recent Advances of the JASMINE Code
Authors: Wei Xiaoyan , Jin Xin , Deng Yongjun , Zhang Guoliang , Fu Xiangang
Published: October 2017

JASMINE is a computer program for the comprehensive performance analysis of fuel rods in nuclear reactors and was developed at China General Nuclear Power Cooperation (CGNPC). The JASMINE code, which is built around a quasi-two-dimensional analysis of fuel rod, consists of a clearly defined mechanical/mathematical framework into which physical models can easily be incorporated. In convenience, the important inputs and principal outputs could be shown directly from the interface due to the special platform in the code. During its development great effort was spent on obtaining an extremely flexible platform which is easy to handle, exhibiting much faster running speed for the code. In recent years, much more experiment data is used for the validation to confirm its correctness.
Machine GTAW Ambient Temperature Temper Bead Technique for Ni-Based Weld Overlay Repairs of Nuclear Components: Assessment of Hydrogen Cracking by Implant Test
Authors: Consonni Marcello , Verdon Claire
Published: October 2017

Repair by welding of a metallic component may be required during original fabrication, following in-service inspection or during planned maintenance operations. In some instances a post-weld heat treatment (PWHT) is required. Codes/standards that cover the repair of or alterations on nuclear components, such as Section XI of the ASME Boiler & Pressure Vessel code or the French RSE-M nuclear code for in-service inspection, include provisions for waiving the need for a PWHT by using a welding technique referred to as ‘temper bead’, but in the case of in-service repairs, the practical challenges may make any heating cycles, including pre or post-heating, undesirable. So, starting from the existing temper bead procedures, repair solutions that do not require preheat, PWHT or post-heating, have been developed, particularly in the nuclear sector, referred to as ‘ambient temperature temper bead’ (ATTB).

One of the most common applications of the ATTB technique is the deposition of full structural weld overlays (FSWOLs), made with Ni-based consumables on ferritic base materials, to repair piping components and nozzles susceptible to primary water stress corrosion cracking (PWSCC) and to mitigate the risk of further cracking. The ATTB technique is currently permitted for the repair of nuclear power plant components by the ASME code via Code Cases N-638-8 and N-740-2, provided the machine gas tungsten arc welding (GTAW) process is used, and it has been successfully applied on various occasions. However, the technique is currently not permitted by the RSE-M code. As the current rules in the RSE-M codes were established following experimental programs based on implant testing, the same type of tests would be expected by the French nuclear regulator to justify any changes. This paper presents the results of a research project carried out to explore the possibility of justifying the introduction of ATTB using the machine GTAW welding process to deposit austenitic weld overlays into the RSE-M code, based on implant testing. Externally loaded tests according to ISO 17642-3 (Implant test) were carried out to determine the effect of different preheat and stress levels on the cold cracking susceptibility in the parent metal. Test loads representative of actual residual stresses were determined through a review of published literature on the residual stresses associated with ATTB FSWOLs. Test pieces were obtained by depositing single beads with Inconel 52 filler metal (AWS A5.14 ERNiCrFe-7) on ASME SA 508 Grade 4N substrate. Conclusions are presented with regard to the possibility to waive preheat or to reduce the minimum preheat temperature specified for temper bead repairs, with respect to the current RSE-M code requirements. Recommendations for further work required to support a request for modification of the RSE-M code are provided.
Analysis of Sustainable Thorium Fuel Utilization in Molten Salt Reactors Starting From Enriched Uranium
Authors: Cui Deyang , Cai Xiangzhou , Chen Jingen , Yu Chenggang
Published: October 2017

Molten salt reactor (MSR), as one of the six systems selected by the Generation IV International Forum (GIF) for future advantaged reactors research and development (R&D), has excellent performances such as high inherent safety, desirable breeding capacity, low radioactive waste production, flexible fuel cycle and non-proliferation. Meanwhile, thorium, as an appealing alternative nuclear fuel to uranium, is more abundant than uranium in the earth’s crust. Realization of thorium fuel cycle in MSRs will greatly contribute to sustainable energy supply for global development. The objective of this paper is to analyze and evaluate thorium fuel utilization in a program in which MSRs are expected to be developed step by step. The program can be described as follows:

1 The first stage is a converter reactor fueled with low enriched uranium. With limited processing based on current chemical partitioning technology and fuel-feeding techniques in the generation-I MSR;

2 The second stage is a 233U production reactor. By using the enriched uranium, it can produce 233U which does not exist in nature;

3 The third stage is a thorium breeding reactor. It is a breeder reactor with Th/233U fuel cycle, and sustainable thorium utilization for energy production is expected to be eventually realized.

By employing an in-house developed tool based on SCALE6.1, the performance of MSR fueled with low enriched uranium is firstly assessed. It is found that MSR is attractive regarding conversion ratio when compared with light water reactors. Then we illustrate the feasibility of 233U production in MSR. Enriched uranium with two enrichments are used as driver fuels to start MSR and produce 233U. The results show that 233U production can be achieved and the double time is about 79.1 years for 20% enriched uranium and 28.3 years for 60% enriched uranium. Finally, the performance of MSR based on pure Th/233U fuel cycle is evaluated. It is found that breeding fissile material is possible in MSR and the breeding ratio is desirable (1.049). Comparison of the three-stage MSRs is also conducted and the results indicate that the resource utilization efficiency is much higher in stage-III than that in the first two stages and much less minor actinides is produced in MSR operating on Th/233U fuel cycle than that in traditional light water reactor.
Load Following by Cogeneration: Options for Small Modular Reactors, GEN IV Reactor and Traditional Large Plants
Authors: Locatelli Giorgio , Fiordaliso Andrea , Boarin Sara , Ricotti Marco
Published: October 2017

Nuclear Power Plants (NPPs) has been historically deployed to cover the base-load of the electric power demand. Nowadays this scenario is changing and some NPPs are requested to perform daily load cycling operation (i.e. load following) between 50% and 100% of their rated power. The traditional methods to perform the load following are by inserting negative or positive reactivity into the core, moving the control rods. This strategy reduces the produced thermal power and in turn the electric power output with respect to the base-load strategy. From a technical standpoint this strategy submits the primary circuit to thermodynamic transients, which causes thermomechanical stresses on some components. From an economic standpoint this operation is very inefficient since, in NPPs, costs are mainly fixed and sunk, and there is a negligible cost saving (if any) in reducing the power of the reactor. A more efficient alternative might be doing the “Load Following by Cogeneration”, i.e. performing the Load Following by diverting the excess of power to an Auxiliary Plant. This paper assesses the technical feasibility of the coupling between a NPP and hypothetical cogenerate plants producing: diesel-like fuels from plastic pyrolysis, or desalinated water, or pellets from waste wood, or hydrogen from water splitting.
Experimental Study of Debris-Loading Head Loss Test for PWR Fuel Assembly After LOCA
Authors: Wang Da , Niu Fenglei , Zhuo Weiqian , Ren Jingwen , Guo Zhangpeng , Bhatti Hammad Aslam
Published: October 2017

A loss of coolant accident (LOCA) in a PWR (pressurized water reactor) would generate debris from thermal insulation and other materials in the vicinity of the break. It is postulated that debris can transported to the containment sump strainer. Some of the debris may pass through the strainer and could challenge the long-term core cooling capability of the plant. To address this safety issue, the downstream effect tests for the PWR were performed. Sensitivity studies on pressure drops through LOCA-generated debris deposited on a fuel assembly were performed to evaluate the effects of debris type and flow rate. Fibrous debris is the most crucial material in terms of causing pressure drops.
Analysis on Heat Transfer of PRHR Heat Exchanger in Tank
Authors: Tao Jiaqi , Jiang Xing , Xiong Zhenqin , Wu Hangyu , Gu Hanyang
Published: October 2017

The heat transfer performance of a small scaled passive residual heat removal (PRHR) heat exchanger is investigated by experiment. It is consisted of 42 C-shaped tubes, which is immerged in a 6-meter high water tank. The average heat flux is obtained under broad range of the working pressure and the mass flow rate of water inside the tubes by the tests. Furthermore, the distribution of the heat flux along the tube is analyzed with the assistance of empirical correlations. The heat transfer rate of the tubes predicted agrees well with the measured value, with a discrepancy less than 5%. These results are helpful in the design of PRHR heat exchanger in nuclear reactor.
RtCM Technology Approach and Application Practice
Authors: Lifei Yang , Yu Chen
Published: October 2017

Firstly, this paper introduces the background and technique flow diagram of RtCM. Secondly, on the basis of the reliability maintenance requirements of the system equipment’s in the plant, a practical application and promotion has been made and a scientific and improved preventive maintenance strategy has been formulated and optimized, saving unnecessary maintenance costs and raising the reliability level of the plant. Finally, the paper makes an expectation and gives some proposals on the application prospect of RtCM technique.
Effect of Thin Metal Film and Micro Structure of Surface on Condensation Pattern
Authors: Nishimura Tomohito , Mikoshiba Yuki , Ohtake Hiroyasu , Hasegawa Koji
Published: October 2017

The water condensation is important for wide range of industrial systems such as condensers and heat exchangers of steam power plants and refrigerators. The condensation generally has two patterns; filmwise condensation (FWC) and drop-wise condensation (DWC). DWC has one-tenth higher heat transfer coefficient than that of FWC. It has been pointed out by many investigators that DWC occurs on the hydrophobic surface and FWC occurs on hydrophilic surface. However, the durability of those hydrophobic effects was not clear enough. In order to maintain a sufficiently long DWC, it is important to understand the effect of the surface property and structure on the condensation surface in more detail. The recent advancement of MEMS (Micro Electro Mechanical System) technology enables us to change the physical nature the surface in the micro scale. It is expected that the hydrophobic surface by the MEMS technology may kept DWC for a longer time. In the present paper, we experimentally investigate the effects of thin metal film and micro structured surface on condensation pattern. Especially, our condensation experiments were performed with the micro structured surfaces by using etching and the metal thin film surfaces by sputtering for approximately 24 hours. Silicon (Si) wafer was used as a basic surface. For the metal thin films surface, we used sputtered Lead (Pb) and Titanium (Ti) on Si surface. For the micro structured surfaces, micro-structured grid was etched on Si surface with several conditions. In order to obtain the relation between the condensation pattern and surface condition, the surface conditions were measured by laser micro-scope, contact angle meter and atomic force microscope (AFM).

For the metal thin films surface, condensation patterns on thin Pb film surface showed DWC. Meanwhile, condensation patterns on thin Ti surfaces showed FWC. From our results, the adsorption forces decreased with increasing contact angle on DWC for Pb. On the other hand, the adsorption forces increased with decreasing contact angle on FWC for Ti. For the micro structured surfaces, condensation pattern was FWC and contact angle decreased in our experimental results. This is because that the condensed water is accumulated in the groove on the micro structure surface.
A Numerical Research on the Heat Transfer Process for a General Passive Heat Removal System
Authors: Wei Li , Zhuo Liu , Qiang Guo , Yidan Yuan
Published: October 2017

A code module for simulating a general passive heat removal system composed of an elevated water tank, a heat exchanger and pipes connecting them is developed in this paper. Then, a typical heat transfer process in this heat removal system is calculated. According to the simulation results, the flash shows the most important impact of the heat transfer process for this passive heat removal system, especially the moment of the flash appearing. In order to design a scaled-down facility with the help of a scaling method to carry out experimental studies on the heat transfer process occurring in a PHRS for developing a more efficient heat removal system, all influence factors of flash should be conducted before a scaling analysis to make a good understanding for the flash. Finally, we get some qualitative conclusions based on sensitivity analyses for some influence parameters: 1) water temperature in the water tank, system flow resistance and the length of the outlet pipe should be ensured by designing according to scaling criteria strictly; 2) choosing a heat exchanger with geometric and material similarity can realize a consistency of heat transfer efficiency and 3) the liquid level in the water tank is less important.
Experimental and Numerical Study of Airflow Dynamics Through an Opening in a Depressurized Enclosure: Application to Nuclear Decommissioning
Authors: Kaissoun Salima , Climent Eric , Prevost Corinne , Ricciardi Laurent
Published: October 2017

In order to understand airflow dynamics through small openings encountered in containment enclosures used for nuclear decommissioning operations, the results of experimental and numerical investigations are analyzed.

The main purposes of this work are to identify the required conditions likely to generate flow inversions at the studied opening which lead to pollutant leakage outside depressurized enclosures, and also to verify the ability of CFD1 simulations to predict these flow inversions by using U-RANS2 and LES3 approaches. All along this work, we tried to reproduce the conditions of leakage occurring at the opening in terms of aerodynamics and openings geometries.

Laser flow visualizations and CFD results show that an additional flow, such as a turbulent jet in competition with the directional flow and a disturbed level of pressure inside the enclosure are among the main causes leading to the leakage through the opening.
TMSR Fuel Cycle Evaluation Under a Screening and Decision-Making Framework
Authors: Jian-hui Wu , Jin-gen Chen , Xiang-zhou Cai , Cheng-gang Yu , Chun-yan Zou , Guo-bin Jia , Xiao-xiao Li , Jian-long Han
Published: October 2017

MSR (Molten salt reactor) is a reactor with the fission material dissolved in the fluoride salt as the fuel, which continuously circulates through the primary loop. The liquid fuel makes the online refueling and reprocessing be possible, consequently, more fuel cycle options would be presented due to the introduction of various refueling fuel type and reprocessing mode (continuous / batch reprocessing) comparing with that of the solid fuel reactor systems. It is important to evaluate all the possible fuel cycle options and screen out the promising ones to narrow the R&D activities of TMSR (Thorium-based Molten-Salt Reactor nuclear energy system) program. In this study, we firstly established a screening and decision-making framework, then conducted an initial evaluation work and identified the potential promising fuel cycle options. Synthesizing the goal of TMSR program and technology readiness, we proposed a “Three-steps” fuel cycle development strategy with the aim for gradually increasing the thorium resource utilization while considering the challenges of the reprocessing technology.
Review on Mechanical Design of Printed Circuit Heat Exchangers
Authors: Zhang Xiaoqin , Shi Shanbin , Christensen Richard N. , Sun Xiaodong
Published: October 2017

As a critical component in the advanced non-light water nuclear power plants with indirect power cycles, an intermediate heat exchanger (IHX) is used to transfer the heat from the primary coolant system to the secondary system. Generally, such reactors are operated at very high temperatures and pressures. Therefore, IHXs are required to be able to provide high thermal effectiveness while withstanding demanding operating pressure and temperature combinations. A suitable IHX candidate needs to possess high mechanical integrity.

One of the promising IHX candidate is Printed Circuit Heat Exchangers (PCHEs). The compact diffusion-bonded PCHEs are characterized by mini- or micro-flow channels photo-chemically etched on flat metal plates. These etched plates are stacked up and diffusion-bonded to form a monolithic block. Therefore, the mechanical integrity of PCHEs are centered on the mechanical design of these etched flow channels.

To date, there are several flow channel geometries developed for PCHEs. Straight and zigzag channels, the first two designs, can be often seen in Heatric-manufactured PCHEs in compliance with ASME codes to ensure the mechanical integrity. To improve the fluid flow and heat transfer capability and reduce the associated pressure drop, straight and zigzag channels evolve to other two variants: S-shaped fin and airfoil fin channels. However, since the surface geometry of the flow channels become significantly complicated, the ASME code is not applicable any longer. Numerical simulation is then required to analyze the mechanical strength of the evolved flow channels under extremely harsh operating conditions.

This paper provides a review on the mechanical design of PCHEs with straight and zigzag channels, which refers to the ASME Boiler and Pressure Vessel Codes. In addition, numerical simulations of the mechanical stress for a case study of S-shaped fin will be presented with Inconel 617 as the construction material. Research plans at the Ohio State University are also presented. The aim of this paper is to summarize the common mechanical design method for PCHEs as well as introduce some of the ongoing research work related to the mechanical strength analysis of complicated flow channels with the aid of computer simulation.
Data Analysis of 64 Rod Bundles Reflood Heat Transfer Experiment
Authors: Yufeng Lv , Yuzhou Chen , Dongxu Zhang , Minfu Zhao , Minghui Duan
Published: October 2017

The test data of 64 rod bundles reflood heat transfer experiment performed by China Institute of Atomic Energy are analyzed. The heater rods are electrically powered and have a diameter of 9.5 mm and a length of 4.3 m arranged in a 8 × 8 array with a 12.6 mm pitch. The test parameter is in the range of 10–500 kg/(m2 · s) for injection water mass flux, 20–80°C for injection water temperature, 500–600°C for initial heater rod temperature, 0–1.1 kW/m for heating power, respectively. The system pressure is atmosphere pressure. Two kinds of spacer grids with and without mixing vanes are adopted to investigate their effect on heat transfer. The result shows that rod wall temperature downstream the spacer grid with mixing vanes is lower than that without mixing vanes, which indicates that the heat transfer is enhanced with mixing vanes. The rewetting velocity is nearly a constant under a certain test condition. The experimental values of rewetting velocity are compared with heat conduction controlled theories. At low mass flow rate, one-dimensional conduction gives agreement with experiment; while at high mass flow rate, the two-dimensional conduction theory is shown to be in agreement with experiment data. The RELAP5/ MOD3.3 reflood model is assessed against the test data. Comparison of code prediction and measured data indicates that the code predicts quench time relatively well but the peak rod temperature differs.
Preliminary LOFC Analysis of a TMSR-SF1 Concept Design Using CFD Method
Authors: Li Linsen , Xie Kai , Shen Feng , Li Qiming , Wang Naxiu
Published: October 2017

The thorium molten salt experimental reactor with solid fuels (TMSR-SF1) was one of the conceptual designs developed by Shanghai Institute of Applied Physics, Chinese Academy of Sciences (SINAP, CAS). The fuel pebbles at the reactor core and the pool type structure of the vessel increase the complexity of thermal-hydraulic (T-H) analysis of the reactor. In order to analysis the T-H feature of TMSR-SF1 in case of the postulated Loss Of Forced Cooling (LOFC) accident, and investigate whether its external air cooling system with nature convection is capable of removing the residual heat, the Computational Fluid Dynamics (CFD) method was used to model the reactor and simulate the transient. In this research, an integrated pseudo 2-D thermal-hydraulic model of the core was developed and a simulation and analysis of the LOFC accident has been conducted. The preliminary calculation results using CFD method show that the external air cooling system has the capability of removing the residual heat. The calculation results also indicate that the peak temperatures of the fuel pebbles, key components and structures of TMSR-SF1 remain under the safety limits and the temperature of the molten salt remains below boiling point.
Application and Verification of Lumped Parameter Nodalizations Using in Containment Pressure Analysis
Authors: Wei Shengjie
Published: October 2017

In order to evaluate the heat removal capability of CAP1400 passive containment cooling system, containment pressure analysis should be performed to verify the adequacy of the containment heat removal mechanisms to maintain post-accident containment pressure below the design limit. But first of all, appropriate containment analysis models should be developed and verified for performing the design basis containment analysis. the nodalizations of the lumped parameter evaluation model are studied in this paper with consideration of important parameters such as mass and energy release, containment volume, internal heat sink, and PCS heat removal to calculate DBA containment pressure and temperature response. This paper sets up three different large scale integral test analysis models to study the effect of nodalizations of the evaluation model using in the containment analysis, reasonable and conservative containment nodalizations are researched by comparison with one typical CERT quasi-steady state test. The comparison and evaluation results show appropriate lumped parameter model can simulate the containment pressure response reasonably and conservatively.
Molecular Dynamics Simulations of Damage Cascades Creation in Oxide-Particle-Embedded Fe
Authors: Mustafa A. M. , Li Zhongyu , Shao Lin
Published: October 2017

Oxide-dispersion-strengthened (ODS)alloys have been identified as one promising candidate alloy type for high temperature reactor applications. Understanding irradiation stability of ODS alloys relies on atomic scale modeling such as molecular dynamics simulations. In this study, yttrium and oxygen charges in Y2O3 oxide particles, which are embedded in pure Fe matrix, are optimized to achieve stabilities observed in experiments. Deviation from the optimized charges causes self-explosion and instability of oxide particles. Molecular dynamics simulations further show that under such optimized charge conditions, damage cascade creation and defect developments can be appropriately modeled.
The Measurement of Parameters of Zero Power Experiments Based on MNSR
Authors: Jingyan Hong , Dan Peng , Xiaobo Wu , Jin Lu , Yiguo Li
Published: October 2017

Zero Power Experiments has been carried out on MNSR (Miniature Neutron Source Reactor) Zero Power Facility. The critical experiment was carried out by using elements method and 5 sets equipments to determine the critical load of the LEU (Low Enrichment Uranium) core; Meanwhile, the value of crucial parts was measured, these values includes the value of radial elements, the integration and differentiation value of central control rod and the value of upper Beryllium reflector. The measure result was precise and that provide a data support for the load of the LEU core and the Start-Up experiments on prototype MNSR.
Determination of Fracture Toughness From Small Punch Test Using Inverse Finite Element Method for In-Service Equipment
Authors: Guan Kaishu , Xu Tong , Guo Linling , Fu Mingxue , Zhu Huadong
Published: October 2017

Fracture toughness is the most important reference index in safety assessment, life prediction of nuclear reactor, pressure equipment, and risk assessment of pressure equipment containing defects. Conventional mechanical test specimens require a large amount of materials and cannot be carried out on in-service pressure equipment. The specimen of small punch test (SPT) is much smaller than conventional one, and can be sampled from the surface of pressure equipment. Repair is not acquired after sampling. Many mechanical properties parameters of the material can be determined from the load-displacement curve of test. SPT makes up the deficiencies of conventional mechanical properties test. In this paper, an innovative approach, inverse finite element method (IFEM), is proposed to deal with the load-displacement curve. The procedure of IFEM can be divided into 3 steps. Step 1, a database containing a variety set of material parameters and their corresponding load-displacement curve will be built using FEM. Step 2; a set of matched material parameter of test load-displacement curve will be selected from the database by artificial neutral network (ANN). Step 3, a numerical simulation of fracture toughness test will be performed and fracture toughness Jic will be determined from the result of simulation. Compared with empirical correlation method, IFEM is much more theoretical and needs not to perform a large number of small punch tests and conventional mechanical tests to create a correlation equation between the mechanical property and load-displacement curve. So IFEM is more efficiency and accurate.
Simulation of the Delayed Hydride Cracking Velocity for Zirconium Alloy Materials Considering Irradiation Effects
Authors: Zhang Jingyu , Wang Bingzhong , Zhu Jiacheng , Ding Shurong , Chen Liang , Li Wenjie , Pang Hua
Published: October 2017

Delayed hydride cracking (DHC) is one of the major factors that affect the safety of zircaloy tubes. In order to effectively simulate DHC, the coupling process including hydrogen diffusion, hydride precipitation, energy flow and mechanical fields should be correctly simulated. In this study, a cohesive model considering irradiation embrittlement and damage accumulation is used to simulate the cracking behavior of zircaloy, and in the finite element model its traction-separation relation is converted to the boundary condition between the traction and normal displacement component in the surfaces to be cracked. The coupling calculation procedures to simulate the DHC behavior using the automatic finite element program generator (FEPG) are obtained and validated, in which the three-dimensional constitutive relation involves plastic strain hardening, irradiation hardening and hydride-induced anisotropic expansion. We find it effective to study the DHC behavior. And we draw the conclusion that irradiation embrittlement and irradiation hardening could visibly accelerate the cracking process. This study improves the numerical simulation method for DHC, provides a reference basis for theoretical study and lays a foundation for further researches on DHC of zircaloy tubes.
Composite Versus Dispersed Accident Tolerant Nuclear Fuels
Authors: Ranasinghe Jayangani , Szpunar Barbara , Jossou Ericmoore , Malakkal Linu , Szpunar Jerzy A.
Published: October 2017

In this study, we analyze and compare the temperature profiles, and temperature gradient profiles of UN, U3Si2 and U3O8 aluminum (Al) dispersed nuclear fuels to propose safer nuclear fuels with enhanced thermal conductivity. To calculate the electronic and lattice thermal conductivities, we use EPW, BoltzTrap and ShengBTE codes implemented with Quantum Espresso. Maxwell-Eucken approximation is used to get the effective thermal conductivity of the considered dispersed fuels. The temperature and temperature gradients are calculated by solving the steady state heat conduction equation for a cylindrical fuel rod. Results show that these fuels have reduced the centerline temperature which will prevent fuel melting, as well it will reduce the thermal stress which leads to cracking the pellet.
Investigation of Air Ingress Into a Vacuum Vessel Related to Particle Re-Suspension and Distribution for Dust Issues in ITER
Authors: Porcheron Emmanuel , Lemaitre Pascal
Published: October 2017

During normal operation of the ITER tokamak, few hundred kilograms of dust containing beryllium (Be) and tungsten (W) will be produced due to the erosion of the walls of the vacuum chamber by the plasma. During a loss of coolant accident (LOCA) or a loss of vacuum accident by air ingress (LOVA), hydrogen could be produced by dust oxidation with steam. Evaluation of the risk of dust and hydrogen explosion, that may lead to a loss of containment, requires studying the physical processes involved in the dust re-suspension and its distribution in the tokamak chamber. This experimental study is conducted by the Institut de Radioprotection et de Sûreté Nucléaire (IRSN) to simulate dust re-suspension phenomena induced by high velocity jet under low pressure conditions. Tests are conducted in a large scale facility (TOSQAN, 7 m3) able to reproduce primary vacuum conditions (1 mbar). Optical diagnostics such as PIV technique (Particles Image Velocimetry) are implemented on the facility to provide time resolved measurements of the dust re-suspension in terms of phenomenology and velocity. We present in this paper the TOSQAN facility with its configuration for studying dust re-suspension under low pressure conditions and underway experiments showing the mechanism of dust re-suspension by sonic and supersonic flows.
Development, Distribution and Maintenance of Application Security Controls for Nuclear
Authors: Waedt Karl , Ding Yongjian , Ciriello Antonio , Lou Xinxin
Published: October 2017

The generic concept of Security Controls, as initially deployed in the information security domain, is gradually used in other business domains, including industrial security for critical infrastructure and cybersecurity of nuclear safety I&C. A Security Control, or less formally, a security countermeasure can be any organizational, technical or administrative measure that helps in reducing the risk imposed by a cybersecurity threat. The new IAEA NST036 lists more than 200 such countermeasures. NIST SP800-53 Rev. 4 contains about 450 pages of security countermeasure descriptions, which are graded according to three levels of stringency. In order to facilitate and formalize the process of developing, precisely describing, distributing and maintaining more complex security controls, the Application Security Controls (ASC) concept is introduced by the new ISO/IEC 27034 multipart standard. An ASC is an extensible semi-formal representation of a security control (e.g. XML or JSON-based), which contains a set of mandatory and optional parts as well as possible links to other ASCs. A set of Application Security Controls may be developed by one company and shipped together with a product of another company.

ISO/IEC 27034-6 assumes that ASCs are developed by an organization or team specialized in security and that the ASCs are forwarded to customers for direct use or for integration into their own products or services.

The distribution of ASCs is supported and formalized by the Organization Normative Frameworks (ONF) and Application Normative Frameworks (ANF) deployed in the respective organizational units.

The maintenance and continuous improvement of ASCs is facilitated by the ONF Process and ANF Process.

This paper will explore the applicability of these industry standards based ASC lifecycle concepts for the nuclear domain in line with IEC 62645, IEC 62859 and the up-coming IEC 63096. It will include results from an ongoing bachelor thesis and master thesis, mentored by two of the authors, as well as nuclear specific deployment scenarios currently being evaluated by a team of cybersecurity PhD candidates.