bluuebunny/crossref_metadata_2025 huggingface dataset:
Estimation of Void Fraction in Dispersed Bubbly Flow With a Constant Electric Current Method
Authors: Uesawa Shin-ichiro , Kaneko Akiko , Abe Yutaka
Published: February 2014

In several void fraction measurement methods, electric void sensors are online measurement method, simple construction and lower cost. In electric measurement, we research on a constant electric current method which is one of conductance methods. By using this method, we can measure volumetric void fraction with higher temporal resolution although the method cannot measure 2D and 3D distribution of void fractions. Besides, multiple measuring electrodes can be installed at a short distance. And then, flow is not obstructed by measuring electrodes. However, the constant electrical current method has been applied in annular flow in previous studies. Void fraction is estimated by cross-sectional ratio of gas and liquid phases in this method. For this reason, dispersed bubbly flow is not applied because cross-section ratio is not continuous in a flow direction. In the present study, Maxwell’s theory, Bruggemann’s treatment and polarization method are applied in order to measure void fraction of dispersed bubbly flow more accurately. Maxwell’s theory is an estimation of a resistance of a mixture with two difference resistivity by calculating electric potential in the mixture. Bruggemann’s treatment is based on Maxwell’s theory but it implies the assumption of a large size-range of particles in surrounding medium. In polarization method, bubbles are assumed to be dielectric bodies. Therefore if voltage is applied to gas-liquid two-phase flow, electrical charges in bubbles are polarized, and polarization electrical field generates. A difference of voltages in bubbly flow and liquid single phase flow assumes to be caused by polarization fields. Void fraction in vertical flow is measured experimentally by the previous method, Bruggemann’s treatment, Maxwell’s theory and polarization method in order to investigate the accuracy of these estimations. Working fluid is air and tap water. The accuracy is measured by comparing with a quick shut valve method and observations. Besides, we investigate effects of flow structure and bubble shape to measurement accuracy. Flow structure is changed by changing gas and liquid volume flow rate. In the experiment for bubble shape, a rising bubble by buoyancy is measured. The bubble shape observed by a high speed video camera is compared with the electrical signal measured by the constant electric current method. From experimental results, it is confirmed that void fraction in bubbly flow and froth flow is estimated more accurately by Maxwell’s theory, Bruggemann’s treatment and polarization method, and change of bubble shape correlates with fluctuation of void fraction measured by the constant electric current method.
An Integrated Single-Step Gamma Spectrum Analysis Method Based on Bregmanized Variational Deconvolution
Authors: Feng Bingchen , Fang Sheng , Zhang Liguo , Li Hong , Tong Jiejuan , Qu Jingyuan
Published: February 2014

Gamma spectrum analysis is an important technique in many radiation measurement contexts. Current gamma spectrum analysis consists of several separate procedures such as linear smoothing and peak fitting. Although widely used, this routine may give incorrect results when the peak is weak or when several peaks are overlapped together. This problem becomes more severe if the statistical fluctuation of the spectrum is large. To simultaneously resolve overlapping peaks and reduce fluctuations, an integrated gamma spectrum analysis method that can obtain peak information of gamma spectrum in a single step was proposed in this study. The proposed method models both physical modulation of gamma spectrometer and the statistical fluctuations into a single linear equation, and converts the gamma spectrum analysis into an inverse problem. Bregman iteration and total variation regularization are utilized to solve this inverse problem under the framework of variational deconvolution, so that the true spectrum can be directly obtained. The feasibility and performance of the proposed method was evaluated by both numerical simulation experiment and Monte Carlo simulation experiment. The experiment results demonstrate that the proposed method can simultaneously resolve overlapping peak and reduce the fluctuations in gamma spectra while preserves both qualitative and quantitative peak information precisely. The proposed method may provide a new way for the development of gamma spectrum analysis method.
Study on the Relationship Between PSA and Safety Objectives of Nuclear Power Plant
Authors: Chen Yan , Li Chaojun , Fu Zhiwei , Song Wei , Zhang Chunming
Published: February 2014

This paper analyses the relationship between PSA and safety objectives. Then according to NUREG-1150, we define the expressions of CDF and LERF, and find a new method to demonstrate that a CDF of less than 10−4 per reactor year and a LERF of less than 10−5 per reactor year are acceptable surrogates to the latent and early quantitative health objectives.
Evaluation Method of Nuclear Equipment Manufacturing Quality
Authors: Guo Depeng , Jin Gang , Wu Qi , Jiao Dianhui , Wang Weidong
Published: February 2014

Equipment quality is one of issues which Nuclear Power Plant owner concern most. The result of actual operation condition and lifetime after the equipment’s failure happens is the most important indicator to justify the equipment quality, but it’s too late to know it. It will be very useful if the quality can be deduced before the equipment’s installation and operation. The evaluation’s difficulty lies in several causations. For example, different equipments produced by different manufacturers and even they are produced by the same factory but the process is definitely different, including the design modification, material change, product reservation and quality non-conformance which almost happens in every manufacturing step. Through analyzing the manufacturing characters and using the evaluation theory, a model which describes the manufacturing quality is built. In this model, the manufacturing quality and final test are considered as 2 decisive factors to evaluate the Equipment quality, and under them, the more detail factors are resolved in the deeper level until it can be measured easily and accurately. In conclusion, the precise evaluation will benefit every field during the nuclear power project construction, and also in the operation Plant, including the project construction, nuclear plant safety operation, risk analysis and fault localization, etc.
Study on Pressure Loss of Two Phase Flow and Flow Boiling Heat Transfer in Micro Channel
Authors: Kadoguchi Yu , Ohtake Hiroyasu
Published: February 2014

Recently, tight lattice core reactors had been discussed for a high conversion ratio. The minimum gap clearance between rods was ∼ 1 mm in these reactors. Thus, boiling heat transfer in the narrow space became important. In this paper, the pressure loss and the heat transfer characteristics in mini pipes were investigated experimentally. Diameter of the present mini pipes was 0.5 and 0.15mm, respectively. Length of the test pipes channel was 20mm. The test liquid temperature was room-temperature, i.e., 20°C. A range of Reynolds number was 1082<Re<7423. The present experimental results for D = 0.5mm were in good agreement with the conventional correlations of the pressure loss and the heat transfer for turbulent flow. The present experimental data for D = 0.15mm were also similar to value estimated by the conventional correlations for laminar flow. The CHF in the present experimental conditions was about 2×107W/m2. The value was higher than that calculated by the conventional correlation, Ivey-Morris equation.
Failure Analysis and Strategy Investigation on Local Thinning of the Warm-Up Line in Auxiliary Feed Water Pump System
Authors: Zhao Liang , Luo Kunjie , Hu Jianqun , Wang Xin , Chen Mingjun
Published: February 2014

Wall thinning measurement, physical & chemical test, FEM analysis was conducted on one of an ASG small bore tube. The results show that wall thinning area mainly distribute downstream of the orifice and typical FAC feature were observed. According to FAC mechanism, influence factors and the service conditions, wall thinning are due to FAC. The function of this tube is keep the auxiliary feed water pump always ready, but the tube material and service temperature are susceptible to FAC. This paper discusses the preventive maintenance strategy of this kind of tubes and recommends that the wall thickness measurement should be conducted after 4 years serving. The thickness inspection area is twice times tube diameter downstream of the orifice; furthermore, the suitability assessment rule for similar tubes was established.
Application of Integrated Safety Assessment Methodology to Feed and Bleed Sequences
Authors: Queral C. , Mula J. , Gómez-Magán J. , Gil J. , Fernández I. , Meléndez E. , Sánchez-Perea M. , Hortal J.
Published: February 2014

As part of the collaboration between Universidad Politécnica de Madrid (UPM), Indizen Technologies and the Spanish Nuclear Safety Council (CSN), the Integrated Safety Assessment (ISA) methodology, developed by the Spanish Nuclear Safety Council (CSN), has been applied to a thermo-hydraulical analysis of Total Loss of Feed Water sequences in a Westinghouse 3-loop PWR. The ISA methodology allows among others obtaining the Damage Domain, i.e., the region of the space of uncertain parametere where certain damage limit is exceeded. Apart of typical uncertain parameters for the physical model, ISA considers operator actuations times (like RCP trip, AFW recovery and begining of Feed & Bleed…) and /or stochastic uncertain phennomena. Given a particular safety limit, several data of every sequence are necessary in order to obtain the damage exceedance frequency of that limit. These data are obtained from the results of the simulations of transients performed with MAAP code. The results show the capability and necessity of the ISA methodology, or similar, in order to obtain accurate results that take into account time uncertainties.
Application of High-Tech in Nuclear Power Instruments
Authors: Li Xiang
Published: February 2014

In recent years, nuclear power instruments have been developed quickly with the rapid development of nuclear power plants in China. In order to improve synthetic qualities and reliability of products and shorten period of development, many high-tech has been widely used in these instruments. LPMS is one of the typical representatives of the nuclear power instruments. LPMS is an important instrument to ensure the safe operation of the reactor. LPMS can realize the real-time monitoring of metal parts’ abnormal conditions both internal and external at key equipments of the reactor, such as the reactor pressure vessel, the steam generator, the main cooling pump and so on. LPMS has been applied in several nuclear power plants successfully. LPMS has complex functions of signal acquisition, event discrimination, data transmission and storage. In this paper, through detailed introduction of hardware design and software design, it is evident that LPMS adopts lots of high-tech. On the one hand, this paper emphasizes to expound advanced and mature technology of hardware design, such as FPGA, DSP and PXI bus and so on. To the nuclear power instruments, the application of high-tech not only improves the integration, enhances the stability and the reliability, but also makes more flexible and easier to expand functions and upgrade hardware. The application of high-tech impels the nuclear power instruments to develop towards low power consumption, miniaturization, high precision, multi-function. On the other hand, this paper introduces graphic programming language LabVIEW and virtual instrument of software design. The application of high-tech of software design decreased the difficulty of programming and shorten period of development. At the end of this paper, it seems that the application of high-tech makes the nuclear power instruments to follow closely the scientific forward position, and to further promote the steady development.
Digital Nonlinear Control for a Pressurizer in a Pressurized Water Reactor
Authors: Cappelli M. , Castillo-Toledo B. , Di Gennaro S. , Memmi F. , Sepielli M.
Published: February 2014

Studying criteria ensuring better implementations of control laws and logics on digital programmable devices is today a crucial point in order to improve performance and safety in nuclear plants of novel design. To this aim, a mathematical model of a PWR is considered, describing the dynamics of the primary circuit. This model is simple enough to allow the determination of a control policy, but it is also sufficiently accurate to capture the nonlinear, time–varying, and switching nature of the system. A dynamic controller for the pressurizer pressure and water level is hence designed. Its main advantage is the reduction of the control response. This latter is a classical drawback, which may cause some troubles to pressure control, due to the long response of temperature sensors.

The digital implementation of the controller has been taken into account to study the important aspects of the influence of the implementation on the performance of the control law. A particularly effective and flexible environment where it is possible to analyze these aspects is Matlab/Simulink, a general–purpose tool for analysis and simulation of multi–domain dynamic systems, practically a standard in control implementations, also in industrial applications. The designed controller ensures a good performance when applied to the model used to derive them, also in the presence of unmodeled uncertainties and disturbances. Its nonlinear switching nature, reflecting the real pressurizer dynamics, ensures better transient behaviors. Since it contains a PI action, it represents an evolution and an improvement with respect to classical PID controllers, usually implemented in standard control actions. To better analyze the control performance, three steps have been considered: first, the designed nonlinear controller has been compared with a linear one. Second, it has been tested on a more detailed model, with more realistic dynamics. Finally, the digital implementation of the controller has been simulated, in order to optimize the electronic implementation on the FPGA–CPU hybrid platform.

A simulation study has been conducted with sampling times in ascending order, to further test its robustness against delays. LabVIEW development tool and technology has been chosen to create the electronic layout and to obtain preliminary results in terms of timing, resources availability and power consumption. Due to the same graphical mode of programming, an immediate link between Matlab/Scilab and LabVIEW is possible in order to optimize the use of each language where it guarantees the best performance.

The methodology here presented could be very useful to better integrate the different features of control engineering and IC design, thus giving the possibility of realizing more accurate simulations and permitting to perform a hardware in the loop (HIL) testing of a wide variety of control algorithms.
Major Improvements of French Mechanical Nuclear Components Codes: RCC-M, RCC-MRx and RSE-M
Authors: Faidy Claude
Published: February 2014

Inside the French Association for Design, Construction and Surveillance Rules of Nuclear Power Plant Components (AFCEN) a set of 7 Codes has been developed and are periodically updated.

This paper will review the major improvements in the past 3 years of the mechanical component codes:

RCC-M: Rules for Design and Construction of Mechanical Components for PWR Nuclear Islands

RCC-MRx: Rules for Design and Construction of Mechanical Components for High Temperature, Experimental and fusion reactors

RSE-M: Rules for surveillance and ISI of Mechanical Components for PWR Nuclear Islands

After a general overview of the way of developments of these Codes, the paper will present the major improvements of these Codes since 2007, many of them have a complete new edition issued in 2010 or 2012.

The conclusion of the paper will present the major objectives and topics of development of these Codes for new international project under the pressure of Code Harmonization by Regulators (MDEP) Standards Development Organization Board (SDO) and Industry organization (CORDEL).
Water Use Management Challenges in Power Generation
Authors: Ng Kit Y. , Wan Ping K.
Published: February 2014

Shrinking fresh water resources and increasing competing demands for water among other users have been a growing concern in many parts of the world, which compels the challenges in water use management faced by both operating and new power generation facilities. For new plants in the planning stage, the ability to demonstrate the availability of water and develop practical and achievable water use management strategies to support the long-term operation of the proposed facility in a sustainable manner has become one of the key elements in the site selection process.

The water demands of particular concerns for the power industry come primarily from cooling water needs, which traditionally constitute the majority of plant water use during the operation phase. This paper examines the water use management challenges confronted by new nuclear power plants in site selection and licensing stages as well as by existing plants in the operating stage in the United States. Using an example, it discusses the types of adaptive approaches on the selection of the heat dissipation systems and designs of cooling water systems that can be adopted by new plants to mitigate the declining water availability, and the related environmental and regulatory challenges. The use of modeling tools to estimate cooling water consumptions and availability and predict environmental impacts in the development of sustainable water management strategies are also illustrated.
Assessing the Design Margins for an Ultimate Heat Sink Sizing
Authors: Zheng Dong , Jarvis Julie M. , Vieira Allen T.
Published: February 2014

The Ultimate Heat Sink (UHS) is a large body of water supply that can be used to cool vital nuclear power plant systems during normal operation and for accident conditions. Due to more stringent environmental and water permit requirements, many new nuclear design proposals have selected the relatively smaller sized mechanical-draft cooling tower with a basin for their UHS. UHS sizing analysis is a critical licensing task for some new generation nuclear power plants Combined Operating License Applications (COLA).

In this paper, a potential UHS is sized for a representative new generation nuclear power plant considering worst case design inputs and modeling assumptions. Over 30 years of historical site meteorological data are processed using an automated technique to identify limiting conditions based on resulting worst UHS design parameters, such as the maximum basin evaporative water loss and the highest basin temperature. The impacts of the cooling tower entrance recirculation effect to these design parameters are also investigated. This paper models the transient plant heat loads in detail for various design basis accident conditions. The large-break LOCA heat load is determined to be bounding for the basin evaporative water loss, while a small-break LOCA heat load may result in the highest basin water temperature. This paper also illustrates that the bounding basin water temperature can result when the peak wet bulb temperature is coincident with the peak UHS heat load. The results of this paper are of interest for new generation nuclear power plants as the paper determines impacts of limiting conditions in assessing the design margins for UHS sizing.
The Fault Tree Analysis on Auxiliary Feedwater Pump Room Ventilation System in Pressurized Water Reactor Power
Authors: Li Yang , Hang Chen
Published: February 2014

Main function of HVAC is to remove heat from equipment and pipeline, hold the inner condition, maintain an ambient temperature and humidity that keep the equipments function properly and easy access. Although regulation is no mandatory requirement of redundant equipment design and preservation function in case of specified disaster or man-made accident. In fact, It does be influenced by the incident whether partial failure or full. The hazard factor determination and qualitative analysis are based on fault tree analysis through simulated mode from selected the typical system. The identification of accident cause, hazard cause and fault mode is essential for improving system reliability. According the analysis result, It will be optimization factor such as installation and design process, maintenance ability, material plan, corrosion preventing. It’s helpful to control hazard under accepted level. This method given in the article is a new way to treat HVAC system in pressurized water reactor nuclear power. It hopes that this method will lead to reduce accident loss, save maintenance fee, bring economic benefits and improve the risk of nuclear power.
Measurement of Reaction Rate Distribution and Neutron Spectrum in China Experimental Fast Reactor
Authors: Chen Xiaoliang , Fan Zhendong , Chen Xiaoxian , Hu Dingsheng
Published: February 2014

China Experimental Fast Reactor (CEFR) has completed physics start-up tests in 2010 and connected the grid on 40%FP in 2011. The reaction rate distribution, neutron spectrum are some important parameters for CEFR neutron field. In order to measure these parameters some low power irradiation tests using foil activation method have been done in CEFR core. Two kinds of special irradiation test subassemblies have been developed and fabricated for irradiation in CEFR core. And a digital high purity Germanium gamma-ray spectrometer system has been established for foil activity measurement. After dozens of low power irradiation tests in CEFR core, the radial and axial distribution of 235U and 238U fission reaction rate have been measured. The distribution of 238U capture reaction rate in CEFR core was also obtained in these tests. The experimental values of reaction rate are according with the calculation values well. Neutron spectrum was measured by means of multifoil activation method. And a neutron spectrum adjusting code was also compiled to determine the neutron spectrum.
Investigation on the Compatibility of Fe-Ni Magnetic Alloy With High Temperature Sodium
Authors: Zhang Jin-quan , Zhou Pei-de , Xu Yong-li , Xie Chun , Fu Xiao-gang , Long Bin
Published: February 2014

This paper presents the study of the compatibility of Fe-Ni magnetic alloy in high temperature sodium. After test, some detections were performed. The weight of all samples increased slightly after test, and the average weight gain increased as the temperature risen; Some Mn compounds and Cu were observed on the interface of steel with sodium, but there no pitting corrosion and intergranular attack were observed; There was no oxide scale formation on the surface of samples. The materials showed well compatibility with sodium under test conditions.
The Experimental Approach to Observe Atomistic Behaviors of Metallic Surfaces by a Linked Facility Between an Ion Accelerator and a STM
Authors: Hirabayashi Junichi , Sato Motohiro , Murakami Kenta , Okita Taira
Published: February 2014

This study focuses on an experimental approach to quantifying irradiation-induced defects in structural materials used in nuclear power plants at an atomic scale. As a preliminary step, a procedure for observing metallic surfaces using a facility comprising an ion accelerator and a scanning tunneling microscope was developed.

Using this experimental setup, atomic-resolution images of Au(111) surfaces were successfully obtained. A herringbone structure, which is characteristic of reconstructed Au(111) surfaces, was clearly observed, with individual atoms being distinguishable.
High-Temperature Steam Oxidation Kinetics and Mechanism of SCWR Fuel Cladding Candidate Materials
Authors: Abe Hiroshi , Hong Seung Mo , Watanabe Yutaka
Published: February 2014

Effects of temperature, dissolved oxygen (DO), and degree of cold work (CW) on the oxidation kinetics of supercritical-water-cooled reactor (SCWR) fuel cladding candidate materials, including three types of 15Cr-20Ni austenitic stainless steels (1520 SSs), in superheated steam have been investigated assuming power-law kinetics. Characteristics of oxide layers and its relation to oxidation behaviors are also discussed. The effect of DO on the weight gain behavior in superheated steam at 700 °C was minor for all specimens at least up to 200 ppb DO. The tube-shaped specimens of 1520 SSs showed very good oxidation resistance at 700–780 °C. There was no clear difference in the oxidation kinetics among the three investigated types of 1520 SSs. The degree of CW is a significant parameter to mitigate oxidation in superheated steam. It has been suggested that the tube specimens showed a very slow oxidation kinetics since Cr diffusion in the outside surface of the tubes is accelerated as a result of an increase of dislocation density and/or grain refinement by a high degree of CW.
Study of Zircaloy-4 Cladding Air Degradation at High Temperature
Authors: Lasserre Marina , Coindreau Olivia , Pijolat Michèle , Peres Véronique , Mermoux Michel , Mardon Jean-Paul
Published: February 2014

Zircaloy cladding, providing the first containment of UO2 fuel in Pressurised Water Reactors, can be exposed to air during accidental situations. This might occur during reactor operation (in case of a core meltdown accident with subsequent reactor pressure vessel breaching), under shutdown conditions with the upper head of the vessel removed, in spent fuel storage pools after accidental loss of cooling or during degraded transport situations. The fuel assemblies inadequately cooled, heat up and as a result, corrosion of Zircaloy claddings takes place. This paper is devoted to the kinetic analysis of Zy4 corroded at 850°C in 20% oxygen – 80% nitrogen partial pressure atmosphere to support the comprehension of the degradation mechanisms involved during the post-transition stage.
Numerical Analysis of Natural Circulation Phenomena of Supercritical Fluids
Authors: Samuel Jeffrey , Harvel Glenn , Pioro Igor
Published: February 2014

The feasibility of operating with natural circulation as the normal mode of core cooling has been successfully demonstrated for a few small sized nuclear reactors. Natural circulation is being considered for cooling the core of a nuclear reactor under normal operating conditions in several advanced reactor concepts being developed today.

Although studies have been conducted in natural circulation for many decades, using natural circulation as the primary cooling mechanism for nuclear reactors or as a passive safety system requires a comprehensive understanding of local and integral system phenomena, validated benchmark data, accurate predictive tools, and reliability analysis methods. As full-scale experiments of supercritical water are expensive, scaling laws can be applied to develop test matrices using modelling fluids to reproduce similar conditions in a scaled-down experimental thermalhydraulic loop.

The main aim of this work is to understand the natural circulation phenomena by analyzing water and modelling fluids such as Carbon dioxide (CO2) and Freon 134a (R-134a). The use of the modelling fluids at subcritical, pseudocritical and supercritical pressures is discussed along with fluid-to-fluid scaling techniques. The results from a one-dimensional numerical model developed using MATLAB to calculate the steady-state mass flow rate and heat transport characteristics of an experimental natural circulation test loop are presented and analyzed.
Close-Out of the Daxin Open Pit Uranium Mine and Associated Waste Rock Piles in China
Authors: Xu Lechang , Gao Jie , Ren Changshun , Shi Min , Zhang Xueli
Published: February 2014

The Daxin Uranium Mine was a small open pit mine with contaminated facilities that included an open pit of 87,620 m2, east waste rock pile of 71,710 m2, west waste rock pile of 57,828 m2, ore transfer station and industrial fields of 9,370 m2, building and structures of 26,600 m2, 246 pieces of equipment, 3000 m of plastic conduit. 2500 m of steel conduit and 1020 m of roads used for transport uranium ores. We present the integrated decommissioning programme utilized to closeout this uranium mine: uranium extraction, slope stabilization, cleanup, backfill, reshaping and covering, re-vegetation, dismantlement and demolishment of facilities, decontamination, re-smelting and wastewater treatment.
Characteristics of Helium Diffusion Trajectory: A Molecular Dynamics Study
Authors: Chen Min
Published: February 2014

Molecular dynamics (MD) are employed to explore the features of Helium atom and He clusters diffusion in titanium at the temperatures between room temperature and 667k. The results demonstrate that the frequency for Helium atom jumping from site to site increases with the increasing temperature. Another important feature is that the diffusion of a He atom and a He trimer is a layered structure. The difference is that the trajectory of a He atom is between two Ti atom layers of (001) planes, and the trajectory of a He trimer is overlapped with the Ti atom layers of (001) plane. Meanwhile, it is observed that the layered structured diffusion becomes ambiguous for a He atom and a He trimer when the temperature is increased. The conclusion emphasizes different migration features for He clusters of various sizes.
Research and Solution on Vibration and Noise Caused by Orifice Cavitations of PTR System
Authors: Yang Ting , Wang Yao Dong , Jiang Xu Lun , Wu Chun Ming
Published: February 2014

There are serious vibration and noise in the pipes of cooling circle in Reactor Cavity and Spent Fuel Pit Cooling and Treatment System (PTR) system during the commissioning test of a Nuclear Power Plant Unit 1, after the improvement of the cooling ability for PTR was actualized. It has been confirmed that the orifice cavitations is the cause of vibration and noise. Engineering modification plans have been proposed: add the series number of throttle orifices and decrease pressure drop of each throttle orifices. Proper engineering redundancy shall be considered according to the test phenomenon and data to keep the orifices from the zones of cavitations. The field investigations after the modification shows that the vibration and noise have been controlled in the acceptable limit while both system function and structure integrity meet the requirement of design code. Thus it can be said that the system modification is successful.
Theoretical Research on Flow Instability in Parallel Channels Under Motion Conditions
Authors: Ma Yingying , Tian Wenxi , Su Guanghui , Qian Libo , Zhang Youjia , Huang Yanping , Wang Yanlin , Qiu Suizheng
Published: February 2014

In motion conditions, in addition to gravitational acceleration, a new acceleration was developed and it was added to the thermal hydraulics characteristics in flow channels. The additional acceleration leads to the different thermal hydraulic characteristics and will trigger the flow oscillation and even flow instability in parallel channels. In order to study the effect of the additional acceleration on the flow oscillation, the corresponding physical models are established in this work. Through the deduction of the mathematical model, the code for flow instability under motion conditions with Gear algorithm is developed. The flow oscillation curves, critical power, marginal stability boundary (MSB) are obtained. After comparison and analysis, it is found that some motion conditions lead to flow periodic oscillation. Different flow passage position results in different oscillation amplitudes. The marginal stability boundaries (MSB) under different motion conditions fit well, that is, the effect of motion conditions on MSB is small. Number of channels has little effect; however, channel arrangement influences the flow in every channel. These conclusions are of great significance in marine reactor design.
Mechanism of Flow Drag Reduction on Non-Smooth Surface
Authors: Feng Beibei , Chen Darong , Wang Jiadao , Yang Xingtuan
Published: February 2014

Numerical simulations of air flow were carried out on non-smooth surface where microriblets were distributed uniformly at only one of the walls. An accurate numerical treatment based on k-ε turbulence model was adopted to study flow alteration and to analyze drag reduction and increasing mechanism on non-smooth surface. A modified calculation unit was used to estimate characteristics of flow at the reformed cells. With the microriblets aligned on the surface, the Reynolds shear stress was significantly decreased which was considered the dominant factor resulting in drag reduction. An additional force generating from the deviation of static pressure on the front and rear end of the riblet grooves caused pressure drag increasing exhibiting exponential growth with the flow rate, which was closely related to vortices induced by momentum transfer at the adjacent area of flow inside the grooves and the outer flow. Shear action at groove walls was greatly degraded due to the gradually variational velocity of vortices. Flow alteration on non-smooth surface compared with smooth surface was also analyzed in detail.
EC6 Safety Design Features for Defense-in-Depth and Beyond Design Basis Accidents Including Severe Accidents
Authors: Yu Stephen , Lee A. G. , Dinh N. B. , Soulard M.
Published: February 2014

The Enhanced CANDU® 6 (EC6®) is the new Generation III CANDU reactor design that meets the most up to date regulatory requirements and customer expectations. EC6 builds on the proven high performance design and the defence-in-depth features of CANDU 6 units and has made improvements to safety and operational performance, and has incorporated extensive operational feedback including lessons learned from Fukushima.
Preliminary Safety Analysis of CSR1000
Authors: Wu Pan , Gou Junli , Shan Jianqiang , Zhang Bo , Li Xiang
Published: February 2014

This paper describes the preliminary safety analysis of a thermal-spectrum SCWR concept (CSR1000), which was proposed by Nuclear Power Institute of China (NPIC). The passive safety system and the design of the two-pass core concept characterize the safety performance of CSR1000. With code SCTRAN (a one-dimensional safety analysis code for SCWRs), loss of coolant flow accidents (LOFA) and loss of coolant accident (LOCA) as well as some other typical transients and accidents were analysed. The maximum cladding surface temperature (MCST) was regarded as an important criterion. The sensitivity analyses of some crucial parameters are helpful for the safety evaluation. Thus some parameters about the safety system and the actuation conditions, such as the delay time of the ADS actuation, the break area in LOCA analysis, were also involved in this paper. The analyses have shown that the proposed passive safety system is capable to mitigate the consequence of the selected abnormalities. The results will be a useful reference for the future development of CSR1000.
Safety Analysis of CPR1000 Spent Fuel Pool in Case of Loss of Heat Sink
Authors: Wang Haitao , Ge Li , Shan Jianqiang , Gou Junli , Zhang Bo
Published: February 2014

The spent fuel pool (SFP) is mainly used for cooling spent fuel assemblies (SFAs) discharged from the reactor core. Besides, it can also shield the radiation. The decay heat can be removed through normal operation cooling system, otherwise it can only rely on the natural circulation in the pool when the coolant pump loses power or the heat exchanger fails. Thus the pool water temperature will continue to rise until it begins to boil. During this period, if no active cooling measures are triggered, the water level will gradually decrease as water boiling. Once the water level drops to the top of the fuel assemblies, the fuels begin to be exposed in the environment. In this paper, the CPR1000 spent fuel pool was chosen as the analysis object and the best estimate system thermal hydraulic code RELAP5 was employed to investigate the process in spent fuel pool in case of loss of heat sink. The results of calculations show that when losing two sets of cooling line, the increase in water temperature in the pool from 55 °C up to 100 °C takes approximately 9.1 h, the evaporation of water volume above the SFAs takes approximately 75.4 additional hours; while in case of losing one set of cooling line, the water temperature of the pool surface reaches 76.6 °C, at which the pool water would not going to boil under the atmospheric environment condition. The results of performed analysis are useful for safety analysis and storage of the SFAs, and can be used to provide a reference for spent fuel pool engineering design.
Neutronic Analysis of the HTTR Core Fueled With Plutonium and Minor Actinides
Authors: Li Chun-Yen , Sheu Rong-Jiun , Peir Jinn-Jer , Chao Der-Sheng , Liang Jenq-Horng
Published: February 2014

The High Temperature engineering Test Reactor (HTTR) in Japan is a 30 MWth helium-cooled graphite-moderated reactor. Similar to most light water reactors (LWRs), the HTTR was designed to burn low-enriched UO2 fuel. Due to its superior safety features and ultra-high burnup tolerance of the coated fuel particles, it is deemed interesting to explore the possibility and performance of burning plutonium and minor actinides (PuMA) in HTTR. A detailed HTTR core model was established using the MCNP5/X Monte Carlo code with ENDF/B-VII cross-section library. The model has also routed for validation by performing benchmark calculations and comparing with reference results. This work presents a preliminary study on neutronic effects of replacing uranium in the design fuel with typical plutonium reprocessed from LWR spent fuels. The results revealed that the core characteristics such as neutron spectrum, spatial flux/power distribution, multiplication factor versus fuel-to-moderator atomic ratio, and safety related reactivity temperature coefficients are strongly dependent on the fuel content to various extents. In particular, PuMA tends to harden the neutron spectrum and cause the subsequent effects accordingly.
Fluorescence Emission Behavior of Eu(III) Sorbed on Calcium Silicate Hydrates Formed With No Dried Process
Authors: Niibori Yuichi , Narita Masayuki , Kirishima Akira , Chida Taiji , Mimura Hitoshi
Published: February 2014

Calcium silicate hydrate (CSH) is a main component of cement-based material required for constructing the geological repository. As in many countries, since the repository in Japan is constructed below water table, we must consider the interaction of radionuclide with cement materials altered around the repository after the backfill. Using fluorescence emission spectra, so far, the authors have investigated the interaction of Eu(III) (as a chemical analog of Am(III)) with CSH gels formed with no dried process, considering a condition saturated with groundwater. However, in such fluorescence emission behaviors, a deexcitation process of OH vibrators of light water and a quenching effect caused by Eu-Eu energy transfer between Eu atoms incorporated in the CSH gel must be considered.

This study examined the fluorescence emission behavior of Eu(III) sorbed on CSH gels formed with no dried process, by using La(III) (non-fluorescent ions) as a diluent of Eu(III). Furthermore, the CSH samples were synthesized with CaO, SiO2, and heavy water (D2O) as a solvent in order to avoid the obvious deexcitation process of OH vibrators of light water. This study prepared CSH samples with the Ca/Si ratio set to 1.6, 1.0, and 0.5. A 1 mM solution of a given combination of Eu(III) and La(III) (Eu(III) content: 100%, 67%, 50% or 33%) was added into CSH gel sample. The contact time-period of the CSH gel with the Eu(III)/La(III) solution was set to 60 days.

In the results, the peak around 618 nm was split into two peaks of 613 nm and 622 nm in the cases of Ca/Si = 1.0 and 1.6. Then, the peak of 613 nm decreased with increment of Eu(III)/La(III) ratio. This means that the relative intensity of 613 nm is useful to quantify the amount of Eu(III) incorporated in CSH gel. Besides, the intensity peak of 584 nm decayed with increment of Eu/La ratio, suggesting a quenching effect due to Eu-Eu energy transfer. However, the decay behavior of the fluorescence emission did not depend on the Eu/La concentration ratio. That is, such a quenching effect is neglectable. Additionally, the low Ca/Si ratio samples underwent slow attenuation of fluorescence and showed profiles similar to those of high Ca/Si ratio samples. Therefore, low Ca/Si ratio samples also include the reaction forming a complex on the surface of CSH gel with Eu(III). In other words, even if Ca/Si ratio is lower than 1.0, CSH gels would retard the migration of radionuclides released from the repository.
Numerical Simulation of Accident Scenario in HTGR (Pebble Bed Reactor) Using COMSOL® Code
Authors: Peter Geoffrey J.
Published: February 2014

High Temperature Gas Cooled Reactor (HTGR) development and operation is expanding in the United Kingdom, Russia, USA (Generation IV Reactors), and France (Pebble Bed Modular Reactor, PBMR). A prototype pebble bed reactor producing 10 MW thermal, High Temperature Reactor (HTR-10) is in operation in China. However, the general public remains skeptical of the safety and the perceived dangers of possible accidents.

Of particular concern are blockages caused by local variations in flow and heat transfer that lead to hot spots within the bed. This paper models the accident scenario resulting from blockages due to the retention of dust in the coolant gas or from the rupture of one or more fuel particles used in the High Temperature Gas Cooled (Pebble Bed) Nuclear Reactors using the commercially available computer code COMSOL. Numerical modeling of flow and heat transfer in a packed bed produces an Elliptical Non-Linear Partial Differential equation that requires custom made computer codes. Previously published results obtained from the use of a custom-made verified computer code are limited to one accident scenario and involve considerable modification to study different accident scenarios. Thus the use of a commercially available computer code that can simulate many different accident scenarios is of considerable advantage. Further, this paper compares numerical solutions obtained from custom-made computer code with COMSOL simulation and discusses the advantages and limitations of both codes.
Stochastic Higher-Order Generalized Perturbation Theory for Neutron Diffusion and Transport Calculations
Authors: Wang Congjian , Abdel-Khalik Hany S.
Published: February 2014

The role of scientific computing has been heavily promoted in many fields to improve understanding the physics of complex engineering systems in recent years while conduct the experiments can be time-consuming, inflexible, expensive and difficult to repeat, e.g. nuclear reactor systems. The ultimate goal of scientific computing is to provide more reliable predictions for engineering systems within certain acceptable tolerance. To realize the benefits of scientific computing, extensive effort has been devoted to the development of efficient algorithms for Sensitivity Analysis (SA) and Uncertainty Quantification (UQ) whose numerical errors is under control and understood. However, the repeated execution of the simulations with different samples is computationally intractable for large-scale system with large number of Degrees of Freedom (DOF). The object of this manuscript will be focus on presenting our own developments of stochastic higher-order generalized perturbation theory to address the explosion in the computational load burden. Additionally, an overview of the current state-of-the-art of SA/UQ will also be provided.
Nuclear Criticality Analyses of the Spent Fuel Pool Under Loss of Spent Fuel Pool Water and Neutron Absorbers in the Racks for Taipower’s Chinshan Nuclear Power Plant
Authors: Kuo Weng-Sheng
Published: February 2014

The nuclear criticality analyses of the spent fuel pool under the postulated conditions of loss of spent fuel pool water and loss of neutron absorbers in the spent fuel racks, for Taipower’s Chinshan Nuclear Power Plant, were performed primarily using the Monte Carlo program MCNP5 in association with the deterministic neutron transport code CASMO-4. The results of these analyses can be used to help understand the impact of these beyond-design-basis accidents to the nuclear criticality, as well as facilitate nuclear utilities and regulatory bodies to develop the safety measures and regulations needed to prevent the criticality accidents.
Experimental Investigation of Wall Nucleate Boiling Models in Narrow Channel
Authors: Chen Weihong , Guo Ang , Yang Lixin
Published: February 2014

Using deionized water as working medium, studies on visualization of the behavior of subcooled boiling bubbles in narrow rectangular channels have been conducted at atmospheric pressure, with the help of a high-speed digital camera. The effects of wall superheat, fluid subcooling and mass flux on the bubble dynamic model of nucleation site density, bubble departure frequency and bubble departure diameter were investigated. This paper analyzed the visual graphic results and obtained quantitative values about bubble nucleation site density, bubble departure frequency and bubble departure diameter under different conditions. Based on the results, wall nucleate boiling models in the conventional channel put forward by Nilanjana Basu have been corrected. The new correlation agrees reasonably well with existing experimental data in 2mm narrow rectangular channel.
The Characteristics of Heat Transfer in a Narrow Rectangular Channel Under Rolling Motion
Authors: Lan Shu , Tan Si-Chao
Published: February 2014

An experimental investigation of forced convection heat transfer was conducted in a narrow rectangular channel under rolling motion. Experimental results show that rolling motion induces the flow fluctuate periodically, and the amplitude of fluctuation decreases sharply with increasing Reynolds number. The periodic oscillation of flow rate causes the temperature of outlet water and outer surface of the heating wall fluctuate periodically. The amplitudes of fluctuation for both outlet water temperature and outer surface temperature of heating wall increase with increasing rolling amplitude and rolling frequency in the turbulent region. The heat transfer coefficient under rolling motion was almost the same as that in steady condition in turbulence region. The reason is that the flow rate fluctuates with small amplitude at relatively high flow rate, thus the effect of rolling motion on the heat transfer is negligible in this experiment.