bluuebunny/crossref_metadata_2025 huggingface dataset:
Experimental Research and DEM Simulations on Stagnant Region in Pebble Bed Reactor
Authors: Li Yu , Gui Nan , Yang Xingtuan , Tu Jiyuan , Jiang Shengyao
Published: February 2014

In pebble bed reactor, pebbles flow very slowly in the stagnant region, which is defined according to the burn-up level of fuel pebbles. It is not allowed to exist in real reactor, since the stay time of fuel pebbles in these regions goes beyond the burn-up level, which increases the risk of leakage of radiation. This research shows that the stagnant region is related to the geometric parameters of the core and the physical properties of pebbles. Experimental setup has been designed to observe the phenomenon of stagnant region, and analysis based on a phenomenological method has been carried out. The phenomenological method is an approach to study the dense pebble flow by means of investigating the interface features of different areas composed of differently colored pebbles. In addition, additional simulations by the DEM model are in good agreement with the experimental results, which successfully verify the availability of the discrete element method. On the basis of these researches, several key parameters have been investigated through DEM simulations, including height of the experimental setup, friction coefficient between pebbles and base cone angle. It is proved that, the stagnant region existing in the pebble bed can be eliminated by improving the design of pebble bed and the physical properties of fuel pebbles. All of these are very helpful to guide the design of pebble-bed reactor.
Improving Education Quality of the Specialty of Nuclear Chemical and Fuel Engineering by Using Superiority of CDUT
Authors: Lu Chunhai , Ni Shijun , Zhang Chengjiang , Chen Min , Hu Chuntao , Shi Zeming
Published: February 2014

Although the specialty of nuclear chemical and fuel some nuclear specialty is built in 2012, some nuclear specialties have a long history. The nuclear specialty in Chengdu University of Technology (CDUT) was built in 1956 in Beijing College of Geology (BCG). Afterwards, the nuclear specialty in BCG was migrated to CDUT. CDUT has had great achievement in student training for atomic industry of China. In this paper we present ideas about improving education quality of the specialty of nuclear chemical and fuel by using superiority of Chengdu University of Technology. The main ideas are as following: 1) sharing course resources; 2) sharing laboratories and instruments resources; 3) sharing information.
Methodology for Verification of the Heat Transfer Crisis in the Nuclear Fuel Assemblies
Authors: Zvorykina Anastasiia , Sharayevskiy Georgij , Fialko Nataliya , Sharayevskaya Nina
Published: February 2014

The present-day stage of nuclear-power engineering development raises sharply a number of complicated questions regarding the guarantee of safety operation of nuclear power-generating units of operating and designed Nuclear Power Plant (NPP). The most important of these unsolved technological problems were considered in [1] on the base of analysis of ways of operation reliability improvement for Nuclear Power Installations (NPI) with WWER and RBMK reactors. In connection with the priorities formulated in [1], in papers [2–6] the main aspects of approaches available for solution of most complex problem are considered: the development of methods of early identification of initial phases of emergency operation regimes in such nuclear power-generating units which are critically important for NPI trouble-free operation. It is necessary to stress, that reliable identification of the anomalies mentioned, especially of thermal-hydraulic nature ones in core region of nuclear reactor, must be provided under conditions when such operating troubles can not yet be detected by the issued supervisory instruments of NPI. Taking into account the requirement to prospective diagnostic provision of NPP equipment, in papers [2–5] are underlined that at present time the development of effective methods of anomalies identification in NPI equipment and development of mathematical software support on base of these methods for computer-aided diagnostic systems on base of AI conceptions in structure of hardware of operator support tools of new generation NPP are considered as the main condition which determine the development of diagnostic means with mentioned functional possibilities.
Prospect of MAO Technology Application in Nuclear Power Industry
Authors: Zhao Zheng , Feng Beibei , Yang Xingtuan , Sun Yanfei
Published: February 2014

Micro arc oxidation (MAO) technology known as a newly surface treatment technology has got a widely application in the field of aviation, aerospace, automotive, electronics, and medical industry. Strength, toughness, hardness and corrosion of valve metal such as aluminum, magnesium, copper, zinc, zirconium and their alloys can be greatly improved by MAO technology.

This paper tries to probe into the feasibility of using MAO technology in nuclear power industry. Aluminum and its alloys are used as structural materials such as the cladding of reactor fuel and all kinds of pipes in the low nuclear reactor. Zirconium alloys are widely used for the fuel cladding, cannula, catheter and other components of the fuel assemblies. Titanium and its alloys offer a unique combination of desirable mechanical properties which makes them to be the candidate materials for structural application in the field of nuclear energy. The surface of all these materials may be destroyed which increasing the risk of the nuclear accident due to the severe serving conditions. As a result, it is necessary to improve the corrosion and wear resistance behavior. With the urgent requirements of safety and durability of nuclear reactor, MAO technology must have a broad prospect in nuclear industry.
Transient Analysis of Active Residual Heat Removal Process for HTR-PM
Authors: Zhu Hongye , Yang Xingtuan , Ju Huaiming , Jiang Shengyao
Published: February 2014

An active residual heat removal process is required by the gas-cooled reactor pebble-bed module plant being constructed in China, in order to improve the economical performance of the plant by shortening the restart duration. This process makes use of steam generator and start-up loop as heat sink, whose structure may suffer cold/heat shock while the sudden load of coolant or hot helium at the beginning. To achieve safety and reliability, transient analysis was carried out based on a one-dimensional mathematical model for steam generator and steam pipe of start-up loop. The calculation results show that steam generator should be discharged and pre-cooled; otherwise, boiling will arise and introduce a cold shock to the boiling tubes and tube sheet when coolant began to circulate prior to the helium. Additionally, in avoiding heat shock caused by the sudden load of helium, the helium circulation should be restricted to start with an extreme low flow rate; meanwhile, the coolant of steam generator (water) should have flow rate as large as possible. Finally, a four-step procedure with pre-cooling process of steam generator was recommended; sensitive study for the main parameters was conducted.
Numerical Simulation of Electromagnetic Field in the Cold Crucible
Authors: Zhang Jinlong , Li Pengfei
Published: February 2014

SNPTRD was planning to employ cold crucible induction heating technique to investigate the corium pool heat transfer, using the UO2 and ZrO2 mixture as the simulated corium pool. Compared with the actual situation, the primary problem of the cold crucible induction heating was the additional introduction of electromagnetic field. To investigate the cold crucible applicability in simulating the corium pool heat transfer, the distribution of joule heat inside cold crucible and the magnitude of electromagnetic force were carried out. To be more suitable for heat transfer test research, the shape of cold crucible was changed to hemisphere, and the cooling water channels were set horizontally. The simulation results indicated that joule heat distribution would be more uniform with lower frequency power but considering power efficiency, there was only one best frequency. The magnitude of electromagnetic force can be ignored compared to the gravity on the natural convection research.
Numerical Modeling of Stationary, Dynamic Pebbles and Gas Flows in a Pebble Bed Reactor
Authors: Zhao Xiang , Montgomery Trent , Zhang Sijun
Published: February 2014

This paper presents a review on the research activities conducted at AAMU (Alabama A&M University) in the last five years. The researchers in College of Engineering, Technology and Physical Sciences of AAMU have been receiving financial support from the U.S. Department of Energy under Massie Chair Excellence Program in Nuclear Engineering from 2008. The main objectives of this project were to improve the capability of understanding the static, dynamic behavior of pebbles and gas flows/heat transfer in a pebble bed reactor (PBR), which is the key to the design, optimization and safe operation of the reactors.
Experimental Study of Magnetic Synchronizator in HTR-PM Fuel Handling System
Authors: Zhang Haiquan , Wang Xin , Nie Junfeng , Li Hongke , Liu Jiguo , Yan Xuelan , Zhao Kezhong
Published: February 2014

Tiny lagging-angle magnetic synchronizator application study, which is based on theory at distance, is developed to solve the sealing of radiation gas inside rotation devices of HTR-PM fuel hand system. Within the outer surface of synchronizator in-side rotor and inner surface of synchronizator outer rotor, two-sized heat-resisted permanent magnets have been intensively coated in NNSSNNSS style. Its shielding can adopted Ti6-Al-4V titanium alloy with special bearing to support the cold-state and hot-state experimental study. The results show that materials of all parts of synchronizator has not been weaken after 107Gy dose γ radiation. Synchronizator can meet the requirement of lagging angle, which must be smaller than 0.2 degree based on the control accuracy. But the reliability depends on the bearing arrangement form and their lubrication. Long term operation reliability of magnet driver device, should give priority to solving the oil-free lubrication problem of bearing. The optimal allocation test is carried out on magnet synchronizator and bearing hot-state test facility. And further engineer demonstrations in the operational reliability will be fully taken on full sized fuel handling hot-state test facility.
Computational Simulation of Turbulent Natural Convection in a Volumetrically Heated Square Cavity
Authors: Vieira Camila Braga , Niceno Bojan , Su Jian
Published: February 2014

This work aimed to analyze the turbulent natural convection in a volumetrically heated fluid with Prandtl number equal to 0.6, representing the oxide material layer of a corium. Four turbulence models were scrutinized in order to select the most appropriate one for turbulence modeling based on Reynolds Averaged Navier-Stokes equations (RANS) of natural convection in a molten core. The turbulence models scrutinized are the standard k-ε, Shear Stress Transport (SST), low-Reynolds-k-ε (Launder-Sharma) and also an elliptic blending model ν2-f. The simulations were carried out in a square cavity with isothermal walls, for Rayleigh numbers (Ra) ranging from 109 to 1011. The numerical simulations, performed in an open-source of Computational Fluid Dynamics (CFD) - OpenFOAM (Open Field Operation and Manipulation), provided outcomes of average Nusselt number as function of Ra number, which were in a reasonable agreement with an experimental correlation and other authors’ simulations. It was also possible to observe the limitations and robustness of each model analyzed, enabling to conclude that the most adequate turbulence models for the present physical problem were SST and ν2-f.
Quality Assurance for a Nuclear Power Plant Simulator by Applying Standards for Safety-Critical Software
Authors: Ye Cheng , Ni Chao , Zheng Tian , Zhang Zhicheng , Zhang Ronghua
Published: February 2014

Nuclear power plant simulators are playing a more and more important role in nuclear power plant lifecycle analysis, and the quality of the simulators should be verified to ensure the safety of nuclear power plants. Currently, there is no systematic quality assurance method for nuclear power plant simulators. In this paper, a systematic quality assurance method for nuclear power plant simulators is proposed basing on experiences with safety-critical software. Key aspects of the method are discussed. In addition, application of this method to a real project is also described as a practical reference.
Helium Permeability Measurement of SiC Ceramic Composites for Advanced Reactor Application
Authors: Zhuo Weiqian , Niu Fenglei , Zhao Yungan , Qi Houbo , Hao Zulong , Hao Pengfei
Published: February 2014

Permeability of helium gas of Silicon carbide ceramic composites material, which is one of the most important properties in application of SiC composite for advanced reactors, is studied by using a simple, low-cost test system. The test system can not only qualitatively determine whether the sample is permeable or not, but also quantitatively measure the permeability for the permeable ones by water displacement. The tests are conducted with low pressure in room temperature. The permeability of the SiSiC composite depends on the preparation method. In four flat-plate materials prepared by different processes for the test, the splint based SiSiC and cordierite coated fiber reinforced SiSiC are hermetic, the permeability of uncoated fiber reinforced SiSiC and CVD carbon coated fiber reinforced SiSiC are 0.216cm2/s and 0.109cm2/s, which imply that the permeability is cut in half with the coating. The samples are scanned under SEM to analyze their microscopic structures and verified that the difference of permeability is related to their coatings as well as the pores and cracks.
Study of Surrogate Data Method for Chaos Identification of Natural Circulation Flow Oscillation Under Rolling Motion
Authors: Zhang Wenchao , Tan Sichao , Gao Puzhen , Zhang Pengfei , Yang Yanguang
Published: February 2014

The identification of chaos in the two-phase natural circulation system under rolling motion can provide theoretical guidance for forecasting and control of natural circulation flow instability. The surrogate-data technique is one of effective methods to identify the chaos in the time series exactly. It can avoid the limitation of the positively identifying chaos. In this paper, correlation dimension and principal component analysis (PCA) was used as the identification evidence; the surrogate-data technique was used for the identification of chaos in time series of the irregular complex flow oscillations. The results indicate that there is chaos in the natural circulation system under rolling motion and the surrogate-data technique can identify chaos exactly.
Research on Feature Recognition of Nuclear Power Equipment Based on the Optimal Wavelet Basis
Authors: Li Bin , Xia Hong
Published: February 2014

Touching upon that the crack fault of the rotor may occur after the reactor coolant pump (RCP) has operated a long time, the fault feature can be identified effectively by the method of the wavelet analysis. In this research, based on the simulation signal of crack fault and the method of discrete wavelet transform (DWT), the cross-correlation coefficients between the fault signal and the different wavelet basis which are selected from the wavelet basis library can be computed. After confirming the maximum of the cross-correlation coefficients, the optimal wavelet basis applied to the fault signal of the cracked rotor will be found. And the main frequency component of the fault feature is recognized by use of the wavelet packet transform (WPT) based on the optimal wavelet basis. The results of simulation illustrate that the wavelet basis selected by the maximum cross-correlation coefficients can become the optimal wavelet basis, and the fault feature of the cracked rotor can be recognized effectively.
Development of Prediction Technology of Two-Phase Flow Dynamics Under Earthquake Acceleration: 8 — Measurement of Velocity Profile Around Bubble Under Structure Vibration
Authors: Mizuno Kousuke , Arai Rie , Kaneko Akiko , Monji Hideaki , Abe Yutaka , Yoshida Hiroyuki , Takase Kazuyuki
Published: February 2014

In a nuclear power plant, one of the important issues is evaluation of the safety of reactor core and its pipes when an earthquake occurs. Many researchers have conducted studies on constructions of plants. Consequently, there is some knowledge about earthquake-resisting designs. However the influence of an earthquake vibration on thermal fluid inside a nuclear reactor plant is not fully understood. Especially, there are little knowledge how coolant in a core response when large earthquake acceleration is added.

Some studies about the response of fluid to the vibration were carried out. And it is supposed that the void fraction or the power of core is fluctuated with the oscillation by the experiments and numerical analysis. However detailed mechanism about a kinetic response of gas and liquid phases is not enough investigated, therefore the aim of this study is to clarify the influence of vibration of construction on bubbly flow structure. In order to investigate it, we visualize changing of bubbly flow structure in pipeline on which sine wave is applied.

Bubbly flow is produced with injecting gas into liquid flow through a horizontally circular pipe. In order to vibrate the test section, the oscillating table is used. The frequency of vibration added from the table is from 1.0 Hz to 10 Hz and acceleration is from 0.4 G to 1 G (1 G = 9.8 m/s2). The test section and a high speed video camera are fixed on the table. Thus the relative velocity between the camera and the test section is ignored. In the visualization experiment, the PIV measurement is conducted. Then the motion of bubbles, for example the shape, the positions and the velocity are measured with observation.

In addition, by varying added oscillation amplitude, frequency and flow rate of the fluids, the correlation between these parameters and bubble motion was evaluated.

It was clarified that the characteristic bubble deformation and velocity profile around bubble is caused by oscillation in each oscillation phase time. When structure vibration affects the flow, two main mechanisms are supposed. One is the addition of body force of the oscillation acceleration to liquid phase and bubble, and the other is the velocity oscillation of the test section and the effect of the boundary layer of the pipe wall. It was also found that the bubble deformation is correlated with the fluctuation of relative liquid velocity field to bubble and pressure gradient in the flow area.
Numerical Simulation Study on Supercritical Natural Circulation Flow and Heat Transfer in Narrow Rectangular Channels
Authors: Song Mingqiang , Zhou Tao , Li Jingjing , Chen Juan , Cheng Wanxu , Xia Bangyang , Huang Yanping
Published: February 2014

When the model of natural circulation in vertical rectangular channels with gap 1 mm was created, considering the character of supercritical water, the distribution of fluid velocity, temperature and density were obtained with the help of ANSYS CFX. The conclusion is that supercritical density is much lower than that of subcritical water in this study parameter and that the heat transfer ability of supercritical water is higher than that of subcritical water. Futuremore, narrow rectangular channels contribute supercritical fluid to transfer heat, which lay the foundation of accurate supercritical thermal engineering calculation.
Simulation of Oscillating Free Interface Behavior by Using OpenFOAM
Authors: Zeng Qingyun , Cai Jiejin , Watanabe Tadashi
Published: February 2014

The motions of free interface and its coupling with bubble in oscillating two-phase flow field are simulated numerically by using OpenFOAM. The oscillating two-phase flow field is simulated using the VOF method and the moving grid of the Arbitrary Lagrangian-Eulerian (ALE) method. Firstly, free interface is simulated, and its interface elevation is well coincided with experimental results. Then simulations of free faces with or without bubble near it under oscillating condition are compared. It is found that the free interface between gas and liquid phases is affected largely by the oscillation, and the effect is intensified when bubble is near free surfaces. And local pressure disturbance is found after bubble coalesces with free interface.
Safety Analysis of a Super Fast Reactor With Upward Flow Cooling in Two Pass at Supercritical Pressure
Authors: Kamata Takayoshi , Li Haipeng , Oka Yoshiaki , Ishiwatari Yuki
Published: February 2014

Safety characteristics of the supercritical-pressure light water-cooled fast reactor (Super FR) with upward flow core cooling in two pass is investigated for the abnormal transients and accidents at supercritical pressure. Upward flow cooling has advantage of simplifying the upper core structure in comparison with the downward flow scheme that part of the coolant flows downward in the blanket fuel assemblies from the top dome of reactor pressure vessel. It also has advantages that flow stagnation does not occur at loss of coolant flow events due to the buoyancy of the coolant. The coolant flow scheme of this design is the all blanket fuel assemblies and part of the seed fuel assemblies are cooled with upward flow first, the coolant flows radially above the core and flow downward in the gap between the core and the shroud to the lower mixing plenum and cools the rest of seed fuel assemblies with upward flow till the upper mixing plenum before core outlet. To evaluate the safety performance, eleven transients and four accidents at supercritical-pressure are analyzed. Safety analysis results show that the safety criteria are satisfied with large margins for all the selected transients and accidents. But in the total loss of coolant flow accident the MCST (maximum cladding surface temperature) is still high. Because of this flow scheme, it is found that the MCST is sensitive to the volume of the gap between two pass. Actuating depressurization valves with low flow single at total loss of flow events is effective to induce flow for once-through SCWR and therefore improves safety performance.
Three-Dimensional Steady Simulation on Two-Phase Flow in Secondary Side of Steam Generator
Authors: Cong Tenglong , Tian Wenxi , Su Guanghui , Qiu Suizheng
Published: February 2014

Steam generator (SG), as the primary-to-secondary heat exchanger and pressure boundary of primary loop, should be integrated and performs well in heat transfer ability. Flow characteristics of the secondary side fluid of SG are essential to analyze U-tube wastage caused by the flow-induced vibration and thermal stress. In this paper, secondary side two-phase flow was simulated based on the porous media model. Additional momentum and energy source terms were appended to the momentum and energy equations of porous media region, respectively. The additional momentum source contained the resistances of downcomer, tube bundle, support plate and separator. The additional energy source included the heat transfer from primary side to secondary side fluid. Solving the control equations by ANSYS FLUENT solver yielded the distributions of velocity, temperature, pressure, density and quality, which can be used in the analysis of flow-induced vibration and separator.
ATHLET Analysis of OECD/NEA ROSA-2 Test 7 Experiments on Intermediate-Break LOCA at Cold Leg
Authors: Du Zheng , Fu Xiaoliang , Yu Nan , Liu Lifang , Cao Zhen , Yang Yanhua
Published: February 2014

Test 7 intermediate-break loss-of-coolant accident (IBLOCA) with 13% break at cold leg was conducted in OECD/NEA ROSA-2 Project using Large Scale Test Facility (LSFT). In this test, auxiliary feedwater was assumed to fail and all safety injection was injected only into the intact loop. Core started to dryout when break valve opened. Liquid level in the core dropped rapidly before loop seal clearing (LSC). ATHLET Mod 2.1 Cycle A was used in the post-test analyses of this LSTF experiment. A basis model with two primary coolant loops, one group steam generator U-tube, and three channels in core was built to simulate this test. One dimension finite critical flow model was employed to simulate a nozzle type break with an over predicted result. The major calculated parameters were compared with the test data, and the overall trend of the test was well calculated by the code, it reveals that ATHLET model could predict such IBLOCA with reasonable results.
Development and Verification of Severe Accident Simulation Capability for a NPP Simulator
Authors: Yang Zhifei , Xie Xiaofei , Chen Xing , Zhang Shishun , Liao Yehong , Peng Zhenxun , Lin Jiming , He Dongyu , Chen Yali
Published: February 2014

It is reflected in the severe accident in Fukushima Daiichi that the emergency capacity of nuclear power plant needs to be enhanced. A nuclear plant simulator that can model the severe accident is the most effective means to promote this capacity. Until now, there is not a simulator which can model the severe accident in China. In order to enhance the emergency capacity in China, we focus on developing a full scope simulator that can model the severe accident and verify it in this study.

The development of severe accident simulation system mainly includes three steps. Firstly, the integral severe accident code MELCOR is transplanted to the simulation platform. Secondly, the interface program must be developed to switch calculating code from RELAP5 code to MELCOR code automatically when meeting the severe accident conditions because the RELAP5 code can only simulate the nuclear power plant normal operation state and design basis accident but the severe accident. So RELAP5 code will be stopped when severe accident conditions happen and the current nuclear power plant state parameters of it should be transported to MELCOR code, and MELCOR code will run. Finally, the CPR1000 nuclear power plant MELCOR model is developed to analyze the nuclear power plant behavior in severe accident.

After the three steps, the severe accident simulation system is tested by a scenario that is initiated by the station black out with reactor cooling pump seal leakage, HHSI, LHSI and auxiliary feed water system do not work. The simulation result is verified by qualitative analysis and comparison with the results in severe accident analysis report of the same NPP. More severe accident scenarios initiated by LBLOCA, MBLOCA, SBLOCA, SBO, ATWS, SGTR, MSLB will be tested in the future. The results show that the severe accident simulation system can model the severe accident correctly; it meets the demand of emergency capacity promotion.
Study on the Characteristics in the Liquid Slug of Rising Slug Flow in Narrow Rectangular Channel
Authors: Wang Yang , Yan Changqi , Sun Licheng , Yan Chaoxing , Xing Dianchuan , Tian Daogui , Jin Guangyuan
Published: February 2014

Gas-liquid slug flow is widely encountered in many practical industrial applications. A detailed understanding of the hydrodynamics of the liquid slug is of great importance in modeling the slug flow. A visualized study using a high speed video camera system was conducted to study liquid slug characteristics of slug flow in a vertical narrow rectangular channel (3.25×40mm2). It is found that three regions could be distinguished in the stable liquid slug tagging along the leading gas slug in most cases. The characteristics of a gas slug are greatly influenced by its position in the liquid slug, and the leading slug has a significant effect on the trailing slug in laminar or transition flow. The wake region of leading slug is also actually a sudden enlargement region where the water jet out of the liquid flim. The minimum stable liquid slug length Lmin is obtained in the experiments ranging from 9 to 17Dh in fully developed turbulence flow.
Shielding Design and Analysis for the Proton Tube of China Lead-Alloy Cooled Research Reactor (CLEAR-I)
Authors: Li Bin , Yang Qi , Zou Jun , Zeng Qin , Bai Yunqing
Published: February 2014

In China Lead Alloy Cooled Research Reactor (CLEAR-I), which is an accelerator driven subcritical system, a proton tube inserts into core from the top of reactor. Some neutrons will leak out along the proton tube, inducing structure materials damage and potential exposure danger for workers.

The shielding design and optimization analysis for the proton tube of CLEAR-I was conducted using the VisualBUS system developed by FDS Team in this study. The result indicated that during operation allowed working time at outside containing compartment above the reactor shall not exceed 40 hours per week and entering into containing compartment was forbidden. Above results showed that the design after optimization basically met the requirements of shielding.
Discussion on the Application of Fire Probability Safety Assessment to Sodium Cooled Fast Reactor
Authors: Song Wei , Yang Hongyi , Zhang Chunming , Zuo Jiaxu
Published: February 2014

As sodium cooled fast reactor (SFR) uses liquid sodium as coolant, the risk of sodium fire is brought to reactor safety, which is different from conventional fire. Sodium is very chemically active, and violent chemical reactions can happen when sodium is exposed to air or water. Sodium fire is a typical and distinctive hazard in sodium cooled fast reactors, which is probably one of the main contributors to the total reactor risk. In this paper, the methodology of fast reactor sodium fire risk assessment is studied, and the principles and procedure of sodium fire probabilistic safety assessment (PSA) are given. The application of this technology in China Experimental Fast Reactor (CEFR) is explored, and several key problems which need more research in the future during the process of sodium fire probabilistic safety assessment are discussed.
3-D Base Control Systems for the Seismic Protection of Power Plant Equipment and Buildings
Authors: Nawrotzki Peter , Wang Weiqiang
Published: February 2014

Safety in nuclear facilities is of special importance and recent seismic events and the resulting damage in these facilities again bring up the discussion. One of the latest events is the 2007 Chuetsu earthquake in Japan. The arising damage in the Kashiwazaki Kariwa Nuclear Power Plant can be found in several reports like (Fukushima, 2007) and (Yamashita, 2008).

Base-Control Systems (BCS) consist of Spring-damper combinations with specifically designed stiffness and damping properties which can be described in the vertical and both horizontal directions. Nowadays these 3 dimensional supports are frequently used to reduce seismic acceleration levels of machine components, equipment and buildings as safety or non-safety related structures. They may also be applied to protect sensitive equipment or machinery against other catastrophic events, i.e. airplane impact which is of particular importance in nuclear facilities.

By presenting specific executed examples, this article shall also underline the efficiency of the BCS which, in contrary to purely horizontally efficient Base-isolation Systems with rubber (BIS), is not extensively discussed in literature. Decisive characteristics of the shown applications are provided and the benefits of using this technology for machine or equipment manufacturers, designers or owners are explained.
The Measurement of the High Temperature Sodium Pipe in CEFR
Authors: Yu Huajin , Tang Long , Qi Min , Ye Yuanwu
Published: February 2014

The main causes of invalidation of heat transfer piping are stress, thermal displacement and vibration in fast reactor. Strain and displacement measurements are of particular importance in the commissioning and running periods, which is to ensure safety of nuclear reactor. The defect and abnormity of sodium piping can be discovered betimes by the strain and displacement measurements, and the hidden problems of piping can be avoided. Based on the investigation of measurement on piping of power plant, the strain and thermal displacement measurements are carried into execution on the main cooling systems of secondary circuit and the accident residual heat removal system of China Experimental Fast Reactor(CEFR) during the commissioning and running periods. According to the measure data, the accurateness of pipe stress analysis is validated, and the security of systems can be safeguarded. This paper describes the strain and displacement measurement on the high-temperature sodium piping, including the measure method for the double-layer piping, the test of detection capability of high temperature strain gauges, displacement measure device, the analysis and adjustment method of measure data, and so on.
The Kinetics of Hydrogen Desorption From Zirconium Hydride
Authors: Ma Mingwang , Wan Ruiyun , Wang Yuan , Cheng Yanlin , Liang Li , Tan Xiaohua , Xiang Wei
Published: February 2014

Thermal desorption spectroscopy (TDS) was used to study the thermal desorption kinetics of zirconium hydride films, which were deposited on molybdenum substrates and thermally charged with gas phase hydrogen. The observed desorption peaks were attributed to phase transforming steps. The activation energy and pre-exponential factor for desorption kinetics was estimated as 116 kJ/mol and 8762 s−1 according to Kissinger relation, respectively. A simulation of TDS spectra was made, which showed that the desorption process followed a first order kinetics. The kinetic parameters were then utilized to predict weight loss behavior at a temperature profile. Pressure effects that can potentially reduce the desorption rate were discussed.
Evaluation Study of NPPs Design Improvement Based on SGTR Calculation Result
Authors: Zhang Jian , Yang Changjiang
Published: February 2014

CP1000, improved from 2nd generation nuclear power plant technology, is a new concept of 2nd plus technology, some improvements of which are relatively mature. Based on the technology of 2nd plus, two new design improvements are being implemented, substituting high pressure safety injection with middle pressure safety injection and adding a function of fast cooling when an accident happens. Using code RELAP5, this article calculated and analyzed SGTR accident in order to evaluate the impact of the two improvements.
Microstructure and Performance of Zr-1.0Cr-0.4Fe-xMo Alloys
Authors: Wei T. G. , Long C. S. , Luan B. F. , Miao Z. , Wang W. , Chen L.
Published: February 2014

For the possibility to develop Zr-based alloys with good corrosion resistance and relatively high strength as well at elevated temperature, Zr-1.0Cr-0.4Fe-xMo (x = 0, 0.2, 0.4, 0.6) alloys with different Mo addition were studied. These alloys were prepared by vacuum arc melting, the microstructure, phase transformation, tensile strength, corrosion resistance and hydrogen uptake during corrosion of these alloys was studied and the effect of Mo addition was discussed. Mo addition had a refinement effect on the microstructure, the α-laths size of the alloys in as-cast condition was decreased by Mo addition, and the recrystal grains of the alloy with 0.6% Mo addition were several times smaller than those of the alloy with no Mo addition in general. Mo addition also affected the characteristics of the second phase particles (SPPs), with the increase of Mo content, the population density of the SPPs increased significantly, whereas the average diameter of the SPPs decreased. Mo addition had a stabilization effect on β phase, the/ (α+β) phase transition temperature decreased with Mo content, especially when Mo was added to the Mo free alloy. The tensile strength of Zr-1.0Cr-0.4Fe-xMo alloys was higher than that of Zr-1.0Sn-0.3Nb-0.3Fe-0.1Cr alloy at 350°C and tended to increase with Mo content. Zr-1.0Cr-0.4Fe-xMo alloys had excellent corrosion resistance in 500°C 10.3MPa steam due to the large number of fine SPPs in the matrix. Addition of Mo promoted the change of the oxidation from cubic kinetics to liner kinetics and the formation of cracks in the oxide layers. It was thought that the solute Mo atoms in Zr matrix played an important role on the degradation of corrosion resistance. Hydrogen uptake fraction of Zr-1.0Cr-0.4Fe-xMo alloys was high initially during corrosion and decreased with the exposure time in the pre-transition region, however, when cracks were formed in the oxide layers in pos-transition region, hydrogen uptake fraction of the alloys would reached to a high level again.
The ACC Configuration Sensitivity Study for AP1000 Large Break LOCA
Authors: Yu Jianhui
Published: February 2014

Similar to the traditional nuclear power plant (NPP), the Accumulator (ACC) of AP1000 is one of the most important facility against Large-Break LOCA (LBLOCA). Following a LBLOCA, the Reactor Coolant System (RCS) pressure will be decreased rapidly. And the Core Makeup Tank (CMT) and Passive Residual Heat Removal (PRHR) will be actuated following “S” signal. However, the transient is so rapid that the CMT and PRHR could not be actuated timely, because the ACC will inject water into the reactor vessel downcomer through Direct Vessel Injection (DVI) line and it will stop the CMT injection immediately when RCS has depressurized to the ACC pressure. Therefore, the ACC configuration is very important to LBLOCA mitigation for AP1000. And the Peak Cladding Temperature (PCT) highly relies on ACC configuration. Several sets of different configuration of both ACC, including initial pressure and initial water volume, are discussed. Different initial conditions (e.g. ACC initial pressure) are considered in the sensitivity study on ACC depressurization phase by phase. WCOBRA/TRAC code was used to perform the LBLOCA sensitivity study. The results of each sensitivity case are presented and analyzed. And the suggestion of how to make the optimal ACC configuration is provided in this paper.
JiNan MNSR Decommissioning
Authors: Li Yiguo , Peng Dan , Zhang Yongbao , Lu Jin , Wu Xiaobo , Zhang Jinhua , Zou Shuyun
Published: February 2014

JiNan MNSR (Miniature Neutron Source Reactor) with thermal power 30kW is an undermoderated reactor of pool-tank type, and UAL4 alloy with 235U enrichment of 90% as fuel, light water as coolant and moderator, and metallic beryllium as reflector. The fission heat produced by the reactor is removed by the natural convection. JiNan MNSR went critical in May 1989, and shutdown in March 2008. The reactor operated 19 years safely with the accumulated power of 8.82×104 kWh. The main use was NAA (Neutron Activation Analysis) work.

The spent fuel was removed from the JiNan MNSR site in Nov. 2009 already. The Jinan MNSR decommissioning documents were approved by Chinese competent authorities in Dec. 2010. The work on site was started in Jan. 2011, and completed in March 2011, the wastes were removed from the Jinan MNSR site. The decommissioning scope of JiNan MNSR includes decommissioning of its related systems, equipment and structures. The weight of intermediate level radioactive is 204.1kg, and the weight of low level radioactive waste is 2297 kg. They were all transported to West-north Waste disposal plant.

The final measuring results are that α and β surface contamination in the reactor building are less than 0.06Bq/cm2 and 0.5Bq/cm2 respectively; the specified activities of the nuclides of 137Cs, 60Co and 90Sr in the soil are less than 1.4 Bq/kg,1.7 Bq/kg and 8.14 Bq/kg respectively; the specified activities of the nuclides of 137Cs, 60Co and 65Zn in the solid sample in the reactor building are less than 5.7 Bq/kg,7.8 Bq/kg and 1.7 Bq/kg respectively.

The specified values for α and β surface contaminations in the reactor building are 0.08Bq/cm2 and 0.8Bq/cm2 respectively; they are 12 Bq/kg, 3 Bq/kg and 10 Bq/kg respectively for the limited values of the specified activities of the nuclides of 137Cs,60Co and 90Sr in the soil; the limited values of specified activities of the nuclides of 137Cs,60Co and 65Zn are 100Bq/kg respectively in the solid samples in the reactor building. The measuring results show that JiNan MNSR site meets the requirements of the unrestricted use.
Current Status of the Major Cells Design of IFMIF Test Facility and Arrangement of the Cells
Authors: Tian Kuo , Arbeiter Frederik , Heinzel Volker , Kubaschewski Martin , Madzharov Vladimir , Mittwollen Martin
Published: February 2014

Latest design status of the IFMIF test facility is addressed with emphasis on the function definition and key technical features of the major cells, including the test cell (TC), access cell (AC), and test module handling cells (TMHCs), as well as the arrangement of these cells. An updated specimen flow for the high flux test module (HFTM) between the TC, AC, TMHCs, and post irradiation examination facility is described and has been considered as a fundamental of the design of the cells. According to this specimen flow, the TMHC is sectioned into four dedicated hot cells and primary processing operations are allocated in these cells according to various functions, dimension of components to be handled, and contamination levels. Reinstallation of irradiated specimens in new test modules for a second or even a third irradiation campaign can also be implemented in the TMHCs. By isolating the disassembling and assembling procedures cross contamination possibility between different hot cells are minimized.
A Heat Pipe Cooled Modular Reactor Concept for Manned Lunar Base Application
Authors: Hu Gu , Zhao Shouzhi , Sun Zhiyong , Yao Chengzhi
Published: February 2014

A lithium heat pipe cooled modular fast reactor (HPCMR) power system concept has been developed for manned lunar base application. The system is designed to use the static thermoelectric conversion module to produce over 100kW electricity for up to ten years. Waste heat is rejected by potassium heat pipe radiator. This system has advantages of low mass, long lifetime, no pumped liquid coolant, and no single point of failure. Main parameters of the system are also given in this paper.
Analysis of Liquid Sloshing in an In-Containment Refueling Water Storage Tank for AP1000
Authors: Liu Yu , Lu Daogang , Dang Junjie , Wu Licun , Ma Wenhui
Published: February 2014

Although much work has been performed on the liquid sloshing inside simple structures like rectangular and cylindrical vessels, this paper deals with the analysis of the liquid sloshing in a more complex structures, the in-containment refueling water storage tank (IRWST). The IRWST is an important component of AP1000 passive core cooling system to ensure the safe operation of the AP1000 nuclear power plant. In postulated non-LOCA events, the water in the IRWST absorbs the residual heat then transfers the heat into the containment atmosphere. However, in the case of earthquake, the sloshing fluid may influence the safety of the appropriate semi-cylindrical IRWST.

In this paper, the liquid transient response in the IRWST was formulated based on finite element modal analysis when the three resonance sine wave was applied as excitation. The result shows that the maximum wave excited by excitation from different directions always emerges from the corner or the edge of tank. Another finding is that water will not overflow at the normal operational water level when exited by the selected excitations in any directions. The safety of the IRWST’s roof is achieved which guarantees that the water vapor and radioactive gases within the tank during normal operation will not release to atmosphere in the containment. The influence of the excitation direction and the water depth are also analyzed.
Effect of Temperature and Humidity on Measuring Ionization Current Inside Contaminated Pipes
Authors: Li Pingchuan , Tuo Xianguo , Liu Mingzhe , Ren Jun , Wang Qibiao , Wu Xuemei , Lian Xiaowen
Published: February 2014

This paper reported the experimental results of ion current under different temperatures and relative humidity using long range alpha detector (LRAD). An approximation relation between the measuring values, temperatures and relative humidity has been obtained using the linear multiple regression method. The experimental results have shown that the measuring values decrease with the increase of temperature and humidity. The influence of humidity on results outweighs that of temperatures. And both temperature and humidity are obviously negative correlated with measured values. Further experiments will be performed to confirm the coupling effects of temperature and humidity and reported later.
The AMB System’s EMI Analysis of Backup Helium Circulator for HTR-10GT
Authors: Zhou Yan , Zhao JingJing , Mo Ni , Sun Zhe , Yu SuYuan
Published: February 2014

With the application in HTR-10GT, the reliability and stability of the AMB system should be studied deeply. Especially EMI analysis on the switch power amplifier is needed to be done, since which one is the main interference source for AMB during the switch turn-on and turn-off. Based on it, a simplified and improved modeling method is listed by dividing the nonlinear transition into several stages, and the models of the voltage source and current source are built in the form of the piece-wise linear way. The conducted emission on the differential mode noise and the common mode is shown by simulation. The result could provide the theoretical basis on the designing the grounding, filter and isolation for the AMB system.
Experiment and Numerical Analysis of Control Method of Natural Circulation by Injection of Helium Gas
Authors: Yanagawa Naoto , Nomura Masashi , Takeda Tetsuaki , Funatani Shumpei
Published: February 2014

This study is to investigate a control method of the natural circulation of the air by the injection of helium gas. A depressurization is the one of the design-basis accidents of a Very High Temperature Reactor (VHTR). When the primary pipe rupture accident occurs in the VHTR, the air is predicted to enter into the reactor pressure vessel from the breach and oxidize in-core graphite structures. Finally, it seems to be probable that the natural circulation flow of the air in the reactor pressure vessel produce continuously. In order to predict or analyze the air ingress phenomenon during the depressurization accident of the VHTR, it is important to develop the method for prevention of air ingress during the accident. In this study, the air ingress process is discussed by comparing the experimental and analytical results of the reverse U-shaped channel which has parallel channels.

The experiment of the natural circulation using a circular tube consisted of the reverse U-shaped type has been carried out. The vertical channel is consisted of the one side heated and the other side cooled pipe. The experimental apparatus is filled with the air and one side vertical tube is heated. A very small amount of helium gas is injected from the top of the channel. The velocity and the mole fraction of each gas are also calculated by using heat and mass transfer numerical analysis of multi-component gas.

The result shows that the numerical analysis is considered to be well simulated the experiment. The natural circulation of the air has very weak velocity after the injection of helium gas. About 780 seconds later, the natural circulation suddenly produces. The natural circulation flow of the air can be controlled by the method of helium gas injection. The mechanism of the phenomenon is found that mole fraction is changed by the molecular diffusion and the very weak circulation.