bluuebunny/crossref_metadata_2025 huggingface dataset:
The Neutronic Transmutation Characteristics of Tc-99 in PWR
Authors: Huang Liming , Liu Bin , Hu Wenchao , Tu Jing
Published: February 2014

We conducted a systematic study on neutronic characteristics of Tc-99 transmutation in pressurized water reactor (PWR) in this paper. We divided the core into three zones. The U-235 enrichment is 1.8%, 2.4% and 3.1% respectively. The loading patterns of Tc-99 include homogeneous loading and heterogeneous loading. In the homogeneous pattern, the mass fraction of Tc-99 raises from 0 to 2% stepped by 0.25%. The calculation result shows that the keff of the core decreases significantly when the Tc-99 mass fraction reaches 2%. In the practical situation, the Tc-99 mass should not exceed 2% in a homogeneous pattern. We choose the Tc-99 coating of fuel pellet as the transmutation rod form in the heterogeneous pattern, which can reduce the self-shielding effects. There are 44 transmutation rods per assembly, replacing part of fuel rods directly. We evaluate the effect of the arrangement of transmutation rods on the characteristic values such as keff and neutron spectrum of the core. In both homogeneous and heterogeneous patterns, the assembly flux near the outer edge of the core has an increasing trend. Our further study shows that much more epithermal neutrons and fast neutrons enter the reflecting layer and bounce back to the core again after moderation and reflection.
Transient Heat Transfer for Helium Gas Flowing Over a Horizontal Flat-Plate With Different Widths
Authors: Zhao Zhou , Liu Qiusheng , Fukuda Katsuya
Published: February 2014

This study is aimed to clarify transient heat transfer process between the surface of solid and the neighboring helium gas in Very High Temperature Reactor (VHTR) or intermediate heat exchanger (IHX). In this paper a series of platinum heaters with different widths under different pressures inside a circular channel have been tested for forced convection flow. The heat generation rate of the platinum heater was increased with a function of Q0exp(t/τ) (where t is time and τ is period of heat generation rate or e-fold time). The heaters were platinum plates with a thickness of 0.1 mm and widths of 2 mm, 4 mm and 6 mm. In the present study, the heat flux, surface temperature, and transient heat transfer coefficients were measured for helium gas passing by horizontal plates under wide experimental conditions such as velocities, pressures and periods of heat generation rate. It was clarified that the heat transfer coefficient approaches the quasi-steady-state when the period is more than around 1 s and it becomes higher when the period shorter than around 1 s. Based on the experimental data, empirical correlations for both quasi-steady-state heat transfer and transient state one at various plate-widths were obtained. It was also found that the heat transfer coefficient becomes higher with the increases of gas pressure.
First-Order and Second-Order Sensitivity Studies of ARCON96 for Ground Release Mode
Authors: Liu Yun , Fang Sheng , Li Hong , Fang Dong , Tong Jiejuan , Qu Jingyuan
Published: February 2014

ARCON96 is the latest USNRC-recommended relative atmospheric concentration calculation program for design basis control room radiological habitability assessments at nuclear power plants. In order to thoroughly evaluate ARCON96 and to provide guidance for ARCON96 implementation, both first and second order systematical sensitivity studies were performed to ARCON96 for ground level release mode. A deterministic Taylor series approximation approach was utilized to analyze the sensitivity of ARCON96 model to wind speed and the distance from the release point to the intake. The sensitivity of ARCON96 was investigated with all atmospheric stability classes for complete evaluation. The results of this study reveal that the sensitivity of ARCON96 is primarily dominated by first-order sensitivity. And the behavior of sensitivity for wind speed and distance is quite different. The computed variance proves the robustness of ARCON96 model in presence of input parameter perturbation.
Study on the Influence of Safety Injection Rate on the Process of SBLOCA
Authors: Hui He , Pan Liang-ming
Published: February 2014

After the nuclear accident of TMI-2, more and more researchers devote to the researches of SBLOCA. It is almost impossible for the large break LOCA depending on the probability of the accident during the reactor life cycle. But it is possible for SBLOCA, such as opening a relief valve for a long time. After reactor shutdown, reactor takes the form of directly safety injection, which has great effects on the flow and heat transfer of the reactor pressure vessel and it has influence on the process of the SBLOCA finally. In present work, an SBLOCA analytical model has been developed with RELAP5 code to analyze special design features at SBLOCA accident. The results are compared with the calculations from Westinghouse NOTRUMP code. This analytical model has also used to analyze the effects of different safety injection rate on the process of accident. The transient variation about some key parameters have been obtained. e.g. the temperature, pressure variations of core, void fraction of core. The results show that the rate of safety injection has significant influence on the process of SBLOCA and the characteristics about the heat transfer of pressure vessel. Different safety injection rates influence the nuclear reactor safety differently.
Transient Critical Heat Fluxes of Subcooled Water Flow Boiling in a SUS304-Circular Tube With Various Twisted-Tape Inserts (Influence of Twist Ratio)
Authors: Hata Koichi , Shirai Yasuyuki , Masuzaki Suguru
Published: February 2014

The transient critical heat fluxes in SUS304-circular tubes with various twisted-tape inserts are systematically measured for mass velocities (G = 3988 to 13620 kg/m2s), inlet liquid temperatures (Tin = 287.55 to 313.14 K), outlet pressures (Pout = 805.11 to 870.23 kPa) and exponentially increasing heat inputs (Q = Q0exp(t/τ), τ = 28.39 ms to 8.43 s) by the experimental water loop comprised of a multistage canned-type circulation pump controlled by an inverter. The SUS304-circular tube of inner diameter (d = 6 mm), heated length (L = 59.4 mm), effective length (Leff = 49.4 mm), L/d (= 9.9), Leff/d (= 8.23) and wall thickness (δ = 0.5 mm) with average surface roughness (Ra = 3.89 μm) is used in this work. The SUS304 twisted tapes with twist ratios, y [= H/d = (pitch of 180° rotation)/d], of 2.40 and 4.45 are used. The transient critical heat fluxes for SUS304-circular tubes with various twisted-tape inserts are compared with authors’ transient CHF data for the empty SUS304-circular tube and a SUS304-circular tube with twisted-tape of y = 3.37, and the values calculated by authors’ transient CHF correlations for the empty circular tube and the circular tube with twisted-tape insert. The influences of heating rate, twist ratio and swirl velocity on the transient CHF are investigated into details and the widely and precisely predictable correlations of the transient CHF against inlet and outlet subcoolings for the circular tubes with various twisted-tape inserts are given based on the experimental data. The correlations can describe the transient CHFs for SUS304-tubes with various twisted-tape inserts obtained in this work within −27 to 7.9% difference.
The Feasibility Study of New Design Method for Combined Support of Pipes in Nuclear Engineering
Authors: Han Lang , Tang Minjing , Jiang Weifeng
Published: February 2014

The feasibility study of new design method is performed for the complicated combined support of pipes in nuclear power plant. The professional steelwork design tool “TEKLA” is integrated into the workflow of PDMS (Plant Design Management System), which is used for the pipes and support design. Comparing to the traditional method, the new method can make the design of combined support easier, and generate the fabrication drawing of support more accurately and quickly. It is great significance to improve the work efficiency and reduce the manpower and engineering cost.
Departure From Nucleate Boiling Modeling Development for PWR Fuel
Authors: Yan Jin , Smith L. David , Karoutas Zeses
Published: February 2014

Current Pressurized Water Reactors (PWR) fuel assembly thermal-hydraulic (T/H) analyses are performed on a subchannel basis that neglects detailed heat transfer and flow distributions surrounding fuel rods. Subchannel codes such as VIPREW require input of thermal mixing and hydraulic loss coefficients that are obtained from costly experiments. Fuel thermal margin or performance is quantified in terms of Departure from Nuclear Boiling Ratio (DNBR) for PWR applications or Critical Power Ratio (CPR) for Boiling Water Reactors. DNBR and CPR predictions for reactor design and safety analysis rely on empirical correlations that are developed and qualified from costly rod bundle water DNB tests. Demands for extended power uprate, high fuel burnup, zero fuel failure, and new nuclear plant designs require a revolutionary advancement in T/H capability for better understanding of coolant behavior and more accurate predictions of thermal margin of the Light Water Reactor (LWR) core and fuel designs under normal operation and postulated accident conditions. Computational Fluid Dynamics (CFD) has been used in many aspects of PWR fuel designs in Westinghouse. Significant advancement in 2-phase flow modeling has been made in the recent years. This paper will illustrate CFD–based DNB modeling development in Westinghouse Nuclear Fuel. A 5×5 test bundle PWR experiment from the ODEN DNB test facility was modeled in CFD using a relatively new 2-phase boiling model. The model geometry included the details of the mixing vane spacer grids. When compared to the test data, the CFD model demonstrated that the DNB power was reasonably predicted. The CFD model also revealed the detailed flow behavior and the 2-phase flow distribution, both of which will be beneficial for the development of new grid spacers.
Incandescent Particle Characterization Emitted by a Cut-Off Grinder During Decommissioning Operations for Evaluating Filters Degradation
Authors: Marchal Pierre , Porcheron Emmanuel , Grehan Gérard , Lafanechere Luc , Walter Jonathan
Published: February 2014

The use of mechanical or thermal cutting tools in decommissioning of nuclear facilities generates a lot of incandescent particles. These particles may represent a risk of deterioration of the containment barriers associated with a potential risk of fire starting.

The aim of the study is to characterize the incandescent particles emitted by a wheel grinder (in terms of temperature, diameter and velocity) and to follow those parameters all along their path from their emission to their impact on the air filter. The characteristics of particles correlated with a possible loss of filter efficiency should identify the destructing particles for the filter. All the measurement techniques used to characterize the incandescent particles are presented in this article. Particles are characterized in terms of aerodynamic diameter with the help of an Aerodynamic Particle Sizer (APS). The particle velocity is measured with laser velocimetry techniques. An adaptation of a commercial pyrometer is achieved to measure the inflight particles temperature. All of these techniques are being implemented on an experimental facility which reproduces representative conditions of the cutting processes realized during dismantling operations.
Sensitivity Studies on the Influence of Shutdown Control Assembly Banks Arrangement in Rod Control Assembly Banks Withdrawal From a Subcritical Condition
Authors: Cui Dawei , Cheng Yanhua , Mao Yulong , Chen Jun
Published: February 2014

A rod cluster control assembly (RCCA) withdrawal from a subcritical condition (RWFS) will lead to an uncontrolled reactivity addition to the core resulting in a power excursion. Such a transient is a limiting case among the moderate frequency incidents in CPR1000 power plant. In this paper a sensitivity study on the arrangement of shutdown control assembly banks which may be have some differences between two power plants have been preformed. The calculation method and main hypothesis used are also briefly presented. The study shows that if all the shutdown assembly banks are located outside the core in hot shutdown state, the radial enthalpy rise factor (FΔH), the maximal reactivity insertion rate and maximal thermal power in RWFS accident is lower than the base case defined as the one with some shutdown banks in the core, while the trip reactivity insertion rate values introduced when trigger reactor trip is bigger than the base case.
Analysis of Fatigue Life for Main Steam Isolation Valve in Nuclear Power Plants
Authors: Zeng Xian , Xia Hong , Dong YeMin , Liu Zhen
Published: February 2014

This paper focuses on the analysis of fatigue life for main steam isolation valve (MSIV) in nuclear power plants (NPPs). The contraction of throat is also in the form of sonic resonance cavity, which causes big vibration and noise. The vibration result in the fretting fatigue, which is the reason of the contact surface damage of the disc and the disc guide. By using FEM, the distribution of the stress on the contact surfaces of the disc and the disc guide was calculated. The influences of vibration on the distribution of the stress on the contact surfaces, together with the effect of the contact stress on the initiation of fretting fatigue cracks and fretting fatigue life have been studied. A critical plane approach the Smith–Watson–Topper fatigue parameter has also been adopted here, to predict crack initiation position and orientation, fatigue life. The results show that the predicted cracks location and fatigue life are well accordance with the practical situation.
Modeling and Simulation of Nuclear Power Units for Primary Frequency Regulation Research
Authors: Tang Zhenpeng , Zhao Fuyu , Wang Pengfei , Chen Li , Fang Huawei
Published: February 2014

The power system frequency stability problem has been the focus of attention since 1920s. Primary frequency regulation (PFR) is one of vital methods to maintain frequency stability, especially when the power system suffers a sudden load disturbance, such as the substation tripping or generator outage accident. To a power system, the more power units participating in PFR, the frequency stability will be controlled better. With the rapid development of nuclear power units in China, the study of the nuclear power units participating in PFR of power system has becoming a pressing problem. In the present study, a detailed nonlinear dynamic mathematical model of the whole pressurized water reactor (PWR) nuclear power plant is built. The dynamic calculation codes are compiled to dynamic link library (DLL) files, which are developed by using FORTRAN language, and the compiled DLL files are embedded into MATLAB/SIMULINK simulation platform by using S-function, the control systems are designed and the whole system of PWR nuclear power plant is simulated in MATLAB/SIMULINK finally. In this method, the operation and control mode of PWR participating in PFR of power system is analyzed and simulated. The simulations results show that the PWR nuclear units are feasible in participating in PFR from safety and economy.
Investigation on Multi-Objective Optimal Design of a Two Pass Condenser
Authors: Chen Lei , Yan Chang-qi , Wang Jian-jun
Published: February 2014

Condenser is one of the key components in nuclear power plant with pressurized water reactor. It is important to control the dimension and weight in the design of condenser through optimization techniques. In this paper, a mathematic model of a two pass condenser is set up for Qinshan I condenser. Some modifications are made based on the original multi-objective algorithm, and the comparison between modified algorithm and the original one is conducted. Furthermore, the multi-objective optimization design of the condenser, taking minimization of the coolant flow-rate and net weight as objectives, is carried out considering thermohydraulic and geometric constraints through hybrid Pareto-sorting multi-objective genetic algorithm (HPSMOGA). The sensitivities of some parameters, which may influence the coolant flow-rate and the net weight of condenser, are also analyzed. The results show that the mathematical model is agreeable for the condenser. it is also shown that the proposed multi-objective optimal method is more effective in searching non-dominated solutions. the sensitivity analysis show that the tube outer diameter, tube pitch, coolant velocity and coolant temperature rising influence the coolant flow-rate and net weight of the condenser more than other variables. The corresponding results would provide guidance in the engineering design of this type of condenser.
Study and Compare on Classification and Management Requirements of Domestic and Foreign Research Reactors
Authors: Wang Haiyang , Song Chenxiu , Gao Quanyuan
Published: February 2014

There are 19 in-service civil research reactors in China nowadays, which are quite diverse on reactor type, usage, power levels, operation modes and safety features [1]. The common practice in nuclear energy advanced countries is to do safety classification and classified management of research reactors. In recent years, China has gradually refined and improved the related work. This paper summarizes the existing research reactor graded approach and safety management requirements included in certain regulations, guidelines and technical documents of IAEA, the United States, Japan and China, and specific practice and experience on operational level as well. And then the classification principles, classification elements, classification methods, safety management requirements and other aspects of the countries or government organizations mentioned above are compared from the view of research reactor safety classification. It comes to the conclusion: As far as China’s research reactor graded approach and safety management requirements are concerned, more work are needed. The relevant provisions of research reactor graded approach of IAEA are comprehensive, and have good reference value for China’s legislation system, but the classification method operability is not strong enough, and there are no detailed management and technical guides as the supplement to the graded approach. The relevant provisions of the research reactor graded approach of the United States and Japan are not systemic enough, but the classification methods are clear and operable, which are referable for China’s relevant practice on research reactor safety classification. After the Fukushima nuclear power plant accident, the classified management of the research reactors has achieved quite good results in comprehensive safety inspection of China’s civil research reactors. Finally, some suggestions in respect of the establishment and improvement of China’s research reactor safety classification guidelines and documents, standardization the use of graded approach, and application in the review and supervision of research reactor in China are put forward.
Theoretic Analysis on Separation Efficiency of Wire Mesh Mist Eliminator of High-Temperature Gas-Cooled Reactor Helium Purification and Auxiliary System
Authors: Yin Huaqiang , Hao Lei , Yang Xingtuan , He Xuedong
Published: February 2014

Helium purification and helium auxiliary system is one of important systems guaranteeing the safe operation of high-temperature gas-cooled reactor. In this system, wire mesh mist eliminator is one of the key components and used to separate waste water containing tritium, and remove moisture after reactor accident. Base on the ideal fluid model and packing pad model developing by Carpenter, A calculation model is presented for separation efficiency of mist eliminator. The calculation program “SEP-WMME” is developed based on the model. The calculation results fit well with experiment results. Theoretic analysis is carried out for the mist eliminator of HTR-PM helium purification system engineering validation test loop. The analysis shows the inlet velocity is an important parameter for mist eliminator. When the inlet velocity is above 3.0m/s, high separation efficiency will be obtained.
Comparison of PTS Guides for Reactor Pressure Vessel Integrity Assessment
Authors: Brumovsky Milan
Published: February 2014

Integrity of reactor pressure vessels (RPV) are of the most importance for safety of the whole NPP. From all potential regimes, Pressurized Thermal Shock (PTS) regimes during emergency cooling conditions are the most severe and most important.

Several nuclear codes are based in similar approaches but their procedures differ and are based on national knowledge and approach to fracture mechanics as well as non-destructive methods of reactor pressure vessel testing.

The paper will compare requirements and procedures for PTS evaluation in accordance with RCC-M code in France [2], KTA in Germany [3], Russian original code PNAEG from 1989 [5] and new procedure from 2004 for WWER vessels [4], as well as VERLIFE procedure and IAEA-NULIFE VERLIFE [6] procedure for WWER RPVs and finally ASME Code requirements [1] including US NRC RG approach.

Detailed comparison of individual parameters in calculations are compared — material properties, degradation of materials, calculated defects size and form, fracture mechanics approach, warm pre-stressing possibility etc.
First-Principles Study of Mechanical Property and Electronic Structure of Quenching and Aging U-Ti Alloy
Authors: Qi Jianbo , Guo Hong , Wu Guangxin , Zhang Jieyu , Chou Kuochih
Published: February 2014

We use the first principles plane wave pseudopotential method calculated the mechanical property with respect to the quenching and aging process of uranium-1.3 wt.% titanium alloy, including the elastic modulus, bulk modulus, Young’s shear modulus and the ideal tensile strength, meanwhile deeply research on the mechanical property mechanism changes through t-he electronic structure. The results show that t-he elastic modulus and ideal tensile strength of quenching state are 198GP and 21.2GP, respectively, and slightly improving through aging treatment. The variation of energy and electronic structure of uranium-titanium solid solution in process of quenching indicate the phase transition from γ phase (cubic structure) to α′ phase (orthorhombic structure) lead to improve to the mechanical property. In process of aging, at the beginning the metastable supersaturated solid solution appears Guinier-Preston (G.P) zones, which are aggregate of solute atoms in the uranium matrix, further improving the tensile strength. Then U2Ti (hexagonal structure) precipitates present in over-aged process as a result of decomposition of metastable α′ phase and reduce the tensile strength. The result shows that the G.P/matrix and precipitate/matrix interfaces have ideal work of adhesion are 15.2eV and 12.5eV, respectively. Our results demonstrate theoretically quenching and aging heat treatment strengthening effect of uranium-titanium alloy.
Effect of Bi Addition on the Corrosion Behavior of Zirconium Alloys
Authors: Yao M. Y. , Zhou B. X. , Li Q. , Zhang W. P. , Zhu L. , Zou L. H. , Zhang J. L. , Peng J. C.
Published: February 2014

In order to investigate systematically the effect of Bi addition on the corrosion resistance of zirconium alloys, different zirconium-based alloys, including Zr-4 (Zr-1.5Sn-0.2Fe-0.1Cr), S5 (Zr-0.8Sn-0.35Nb-0.4Fe-0.1Cr), T5 (Zr-0.7Sn-1.0Nb-0.3Fe-0.1Cr) and Zr-1Nb, were adopted to prepare the zirconium alloys containing Bi of 0∼0.5% in mass fraction. These alloys were denoted as Zr-4+xBi, S5+xBi, T5+xBi and Zr-1Nb+xBi, respectively. The corrosion behavior of these specimens was investigated by autoclave testing in lithiated water with 0.01 M LiOH or deionized water at 360°C/18.6 MPa and in superheated steam at 400 °C/10.3 MPa. The micro structure of the alloys was examined by TEM and the second phase particles (SPPs) were analyzed by EDS. Micro structure observation shows that the addition of Bi promotes the precipitation of Sn as second phase particles (SPPs) because Sn is in solid solution in α-Zr matrix in Zr-4, S5 and T5 alloys. The concentration of Bi dissolved in α-Zr matrix increase with the increase of Nb in the alloys, and the excess Bi precipitates as Bi-containing SPPs. The corrosion results show that the effect of Bi addition on the corrosion behavior of different zirconium-based alloys is very complicated, depending on their compositions and corrosion conditions. In the case of higher Bi concentration in α-Zr, the zirconium alloys exhibit better corrosion resistance. However, in the case of precipitation of Bi-containing SPPs, the corrosion resistance gets worse. This indicates that the solid solution of Bi in α-Zr matrix can improve the corrosion resistance, while the precipitation of the Bi-containing SPPs is harmful to the corrosion resistance.
Theoretical Analysis of Ultrasonic Testing Technique for Dissimilar Metal Weld
Authors: Liu Xiyu , Tang Yueming
Published: February 2014

The weld is a typical dissimilar metal weld (DMW). As its large grain, anisotropism and narrow moving space for probe, it is difficult for practical ultrasonic testing. On the basis of studying weld structure, this thesis firstly discusses the sonic propagation path of different incident angle, and then calculates the reflectivity and transmittivity, from which the ratio-angle curve will be plotted, finally simulates & analyzes the ultrasonic testing with software CIVA. The analysis results show that proper selection of the incident angle will provide suitable incident angle range for the following experiments. With the CIVA analysis results, it will also provide references for choosing proper probe parameters and relative technics for following research.
A Numerical Study of Eccentricity Effect to Heat Transfer Under Single Phase Force Convection Inside a Vertical Annulus
Authors: Liu Yang , Jia Qianqian , Jia Haijun
Published: February 2014

Because annulus channel can be used to develop high efficiency compact heat exchangers, the heat transfer in annulus channel has become great interest to researchers in recent years. Most of the studies focus on the vertical concentric and horizontal eccentric annulus. The investigations about single phase force convection heat transfer inside a vertical eccentric annulus are not enough. In this work, force convection heat transfer is numerically studied to determine the eccentricity effect inside a vertical annulus. For this purpose, full Reynolds-averaged Navier-Stokes equations along with energy equations are solved in a 3-D grid. The discrete method of the equations is based on finite-volume method and the turbulence model is RNG k-ε model. The radius ratio of the annulus is 0.8 in this work. Heat flux of one wall is constant while the other is insulated. Firstly, the feasibility and exactness of the numerical method is proved by comparing the Nusselt number with experiment in concentric annulus. Then the effect of eccentricity is studied in detail.
Research of Parameters’ Influence to the Reverse Flow of UTSG Under Natural Circulation
Authors: Duan Jun , Zhou Tao , Ju Zhong-yun , Li Jing-jing , Song Ming-qiang , Xiao Ze-jun
Published: February 2014

The reverse flow will appear in some vertical UTSG’s U-tube under natural circulation. There is a characteristic length of U-tube when reverse flow appeared. Reverse flow has a bad effect on natural circulation of UTSG. By calculation of the mathematical model, variation of different parameters’ influence can be obtained. Studies have shown that reducing the diameter of the shortest U-tube, the medium length U-tube diameter is in constant, increasing the diameter of the longest U-tube and decrease the U-tube’s inlet temperature will improve the flow distribution of the U-tube, which can help preventing the occurrence of reverse flow.
Research on Project Management of DCS Development for First AP1000 Unit
Authors: Ma Shihong
Published: February 2014

Sanmen Nuclear Power Plant is constructing the first AP1000 unit in the world, and DCS system is adopted as the non-safety control system. This paper reviews the experience of Sanmen project, and analyzes the necessity of nuclear power plant Owner’s management for developing the DCS system, as well as how to manage the development effectively.
A Design of the Alarm System in Digital Control Room of HTR-PM
Authors: Jia Qianqian , Liu Yang , Zhang Liangju
Published: February 2014

A design of alarm system which includes both annuciator tiles and VDU alarm displays is proposed of the digital control room in HTR-PM (high temperature reactor pebble-bed module). Important and system-level alarms are located as fixed tiles, most of which are shared alarms. A shared alarm is a combination of several alarm causes associated with a single plant system or component. In the VDU displays, icons which stand for alarm tiles are developed. The tile icons obey the same coding rules of the real alarm tiles. A special VDU display is designed, in which the alarm list function and the tile icons are both included. The tile icons in the display are arranged exactly the same as the real tiles on the panel. When the shared alarm tile is activated, via the special VDU display, operators can easily get the detailed alarm information about the alarmed tiles with only a click on the tile icon. Based on the alarm display features, an alarm control procedure of pushbuttons on the console and the control functions in VDU display is presented. In the alarm response procedure, the personnel cannot acknowledge the alarm when the detailed messages are not read. This alarm system is technically based on the DCS (Distribute Control System) of the plant. Compared with using VDU alarm list only, tiles provide the operator perceptual understanding of the whole plant situation and rapid detection of alarms. The special VDU display with tile-like icons and function of listing detailed alarm messages provide the operator inquiry capability to obtain specific information, which support the operator effectively.
The Reliability Prediction of HTR’s Graphite Component in Various Temperature and Neutron Dose Levels
Authors: Fang Xiang , Wang Haitao , Yang Xingtuan , Yu Suyuan
Published: February 2014

In high temperature gas-cooled reactors (HTRs), graphite is used as the main structure material. The side reflecter of the reactor core is composed by a pile of graphite bricks. In real operational condition of the reactor, both high temperature and fast neutron irradiation have great effect on the behavior of graphite components. The non-uniform distribution of temperature and neutron dose cause obvious stress accumulation, which greatly affects the security and reliability of the graphite components. In addition, high temperature and neutron irradiation make the properties of graphite change in evidence, and the changes are not linear. Such changes must be considered and simulated in the calculation, in order to predict the stress concentration condition and the reliability of the graphite brick correctly. A FORTRAN code based on user subroutines of MSC.MARC is developed in INET in order to perform three-dimensional finite element analysis of irradiated behavior of the graphite components for the HTRs. In this paper, the stress level and failure probability of graphite components are calculated and obtained under different in-core temperatures and neutron dose levels of the core side of brick. 400°C, 500°C, 600°C and 700°C are selected as the core side temperature, while the range of neutron dose is 0 to 1022n cm-2 (EDN). Different constitutive laws are used in stress analysis procedure. The impact of different temperature and neutron dose levels are discussed.
Development of Prediction Technology of Two-Phase Flow Dynamics Under Earthquake Acceleration: 9 — Effect of Structure Vibration on Rising Bubble Behavior
Authors: Arai Rie , Mizuno Kousuke , Kaneko Akiko , Monji Hideaki , Abe Yutaka , Yoshida Hiroyuki , Takase Kazuyuki
Published: February 2014

Earthquake is one of the most serious phenomena for safety of a nuclear reactor in Japan. Therefore, structural safety of nuclear reactors has been studied and nuclear reactors ware contracted with structural safety for a big earthquake. However, it is not enough for safety operation of nuclear reactors because thermal-fluid safety is not confirmed under the earthquake. For instance, behavior of gas-liquid two-phase flow is unknown under the earthquake conditions. Especially, fluctuation of void faction is an important factor for the safety operation of the nuclear reactor. In the previous work, fluctuation of void faction in bubbly flow was studied experimentally and theoretically to investigate the stability of the bubbly flow. In such studies, flow rate or void fraction fluctuations were given to the steady bubbly flow. In case of the earthquake, the fluctuation is not only the flow rate, but also body forth on the two-phase flow and shear forth through the pipe wall. Interactions of gas and liquid through their interface also act on the behavior of the two-phase flow. The fluctuation of the void fraction is not clear for such complicated situation under the earthquake.

Therefore, in this research projects, the behavior of gas-liquid two-phase flow is investigated experimentally and numerically in the series of study. In this study, to investigate effects of vibration on bubbly flow in the components and construct experimental database for validation, we performed visualization experiments of vertical bubbly flow in rectangular water tank on which sine wave vibration was applied. In this paper, results of preliminary visualization experiments were shown including data of positions of bubbles, shapes of bubbles and velocity of bubbles evaluated by the visualization techniques. And liquid velocity distribution around bubbles by the PIV measurement was also shown. In the results, bubble behaviors were affected by oscillation. And the cycle of the bubble inclination angle was almost same as the cycle of oscillation table velocity.
Preliminary Fuel Cycle Analysis of a Fusion-Driven Subcritical Reactor
Authors: Mao Lisheng , Wang Minghuang , Fu Xuewei , Jiang Jieqiong , Wu Yican
Published: February 2014

The levelized cost of electricity (LCOE) has been performed to compare two fuel cycle scenarios: a once-through cycle (LWR OT) and a fusion Cdriven reactor, namely FDS-SFB, recycling employing PUREX (Purex-SFB). In order to estimate the LCOEs, the mass flows based on an equilibrium mode were analysed. The sensitivity of the results to variations in key parameters was also performed. A simple dynamic model was also constructed to consider other important factors that characterize a fuel cycle, e.g. resource utilization, environmental effects.

The results of economics are as flows: LWR OT 29mills/KWh, Purex-SFB 48.19mills/KWh. It was found that the capital cost accounts for the largest proportion of the LCOEs. The fuel cycle cost analysis indicates that the FDS-SFB fuel cycle cost will be competitive with the once-through fuel cycle. Also, sensitivity analysis indicates that fuel cycle cost of LWR would be higher than that of the LWR-SFB fuel cycle with the uranium Price rising. Dynamic model analysis indicates that Purex-SFB could reduce the amounts of MA and the amounts of natural uranium considerably compared with LWR OT.
Analysis of Flowing Characteristics in IRWST Under Pressure Relief Condition
Authors: Zhang Mengchao , Li Xiangbin , Fang Xiaolu , Zhang Yuhao
Published: February 2014

The presurizer is one of main equipments of PWR Nuclear Power Plant. The heat transfer mechanism will change with temperature’s increasing when the steam with high temperature and high pressure condensates as it sprays into the in-containment refueling water storage tank (IRWST), which will be detrimental to the safety of the reactor. In this study, the flow field and heat transfer characteristics are simulated by means of professional CFD software with k-ε turbulence model and Particle modle when the steam sprays into IRWST through the sprayer. The results show thermal changes of water and steam.
Measurement of Velocity and Temperature Profiles in the Scaled-Down CANDU-6 Moderator Tank
Authors: Kim Hyoung Tae , Cha Jae Eun , Rhee Bo Wook , Choi Hwa-Lim , Seo Han , Bang In Cheol
Published: February 2014

We are planning to construct a scaled-down moderator test facility to simulate the CANDU-6 moderator circulation phenomena during steady state operation and accident conditions at the Korea Atomic Energy Research Institute. In the present work a preliminary experiment using a 1/40 scaled-down moderator tank has been performed to identify the potential problems of the flow visualization and measurement in the scaled-down moderator test facility. With a transparent moderator tank model, a flow field is visualized with a Particle Image Velocimetry (PIV) technique under an isothermal state, and the temperature field is measured using a Laser Induced Fluorescence (LIF) technique. A preliminary CFD analysis is also performed to find out the flow, thermal, and heating boundary conditions with which the various flow patterns expected in the prototype CANDU-6 moderator tank can be reproduced in the experiment.
Verification of Accumulator Model in COSINE
Authors: Sha Huie , Zhang Hao , Du Zheng , Xiong Yiqiang , Yang Yanhua
Published: February 2014

The objective of COSINE (COre and System INtegrated Engine for design and analysis) project is to develop a software platform which is used for nuclear power plant design and safety analysis. The system code (SYST) is a part of COSINE code. In this paper, the mathematic model of accumulator is established. The model is based on following assumptions: nitrogen above liquid level in accumulator is represented in idea gas equation, water is modeled as an isothermal system. The model for liquid flow include inertia, wall friction, form loss, and gravity effects. Several cases were calculated under different conditions, and the results were compared with RELAP5. It shows that COSINE results agree well with RELAP5.
Optimization Study on LLFPs Transmutation in PWRs
Authors: Liu Kun , Wu Hongchun , Cao Liangzhi , Zheng Youqi
Published: February 2014

A systematic study on the Long-lived fission products (LLFPs) transmutation in a PWR has been performed, aiming to devise optimal transmutation strategy in present nuclear power plants. The LLFPs selected in the analysis include 99Tc and 129I discharged from LWRs. The isotope, 127I is also considered to avoid the difficulties in isotopes separation. To minimize the negative impacts of LLFPs on the core performance and safety parameters, technetium or MgI2 targets mixed with zirconium hydride are designed and investigated. The equilibrium cycles are investigated. The transmuted amounts of 99Tc and 129I are equals to the yields from 1.94 and 4.22 1000 MWe PWRs, respectively. Numerical results indicate that both the 99Tc and 129I can be transmuted conveniently in present PWRs in the form of target pins.
Results and Discussion on Scaling Analysis for the NHR200-II Integral Test Facility Under Single Phase Natural Circulation
Authors: Liu Yang , Jia Qianqian , Jia Haijun , Wu Lei
Published: February 2014

The SMR has become great interest to researchers in recent years. The integrated nuclear heating reactor NHR200 was developed by Institute of Nuclear and New Energy Technology (INET) in Tsinghua University. On the basis of the NHR200, the NHR200-II is developed. The thermal-hydraulic parameters in the primary loop such as temperature and pressure are increased, which make it more suitable for industrial steam supply and sea water desalinization. Because the primary loop of NHR200-II operates with fully natural circulation, it is necessary to investigate the natural circulation characteristics by experiment under new thermal parameters. In order to determine the characteristics scale of the test facility, the scaling analysis is needed. The similarity groups, including Richardson number, etc. are determined from the dimensionless fluid and solid governing equations. The test facility uses the fluid with same property as the prototype to reduce the distortion. The axial length scale ratio of the test facility is set to be 1:1. The analysis results of the core surface average heat flux density ratio and the flow area ratio are 1:1 and 1:210 respectively. At the end of this paper, the results of the scaling analysis are present.
Research on Trans Critical Characteristic on Sliding Pressure Booting Process of SCWR
Authors: Liu Liang , Zhou Tao , Song Mingqiang , Chen Juan , Luo Feng , Xia Bangyang
Published: February 2014

The homogeneous flow model of trans critical transient calculation is achieved by the trans critical program of SCWR which is made by FORTRAN. The SCWR sliding pressure starting scheme is simulated by this program. So that the pressure raising stage characteristic of its booting process is analyzed. The calculation result shows that, in the pressure raising stage of booting process, the maximum cladding temperature gradually increases. Coolant temperature increases gradually along with the flow direction. The dry degree also increases along with the flow direction. The highest temperature of cladding does not exceed 650 °C limits. There is no major adverse effect on heat transfer of fuel rod in the dramatic changes in the physical properties of water.
The Research of Flow Instability in Vertical Parallel Natural Circulation Channels Based on Relap5
Authors: Li Jingjing , Zhou Tao , Song Mingqiang , Huang Yanping
Published: February 2014

The gas-liquid two phase flow oscillation in vertical parallel natural circulation channels was performed by the best estimate system computer code Relap5. The effects of symmetry and dissymmetry degree of heated power to flow oscillation and the effects of symmetry and dissymmetry throttling to flow oscillation were researched. The results says that when the twin channels under the same conditions of geometry and boundary, the parameters of the twin channel such as flow are the same. So under these conditions the twin channels can be researched as single channel. It is more possible of flow oscillation for the channels under condition of dissymmetry heating. The use of throttling will make the channels more stable, it is more stable when the throttling coefficient increases. With the implement of dissymmetry throttling, the system is possible for out of phase flow oscillation.
Study on Impacts of ADS on AP1000 During SB-LOCA
Authors: Chen Long , Lv Xuefeng , Wang Shengfei , Yu Yu
Published: February 2014

AP1000 model, including Reactor Coolant System (RCS), Engineering Safety Features (ESFs) and containment, is built by using the modular accident analysis program. A severe accident scenario of small-break loss-of-coolant accident (SB-LOCA) at cold leg concurrent with different conditions of the automatic depressurization systems (ADS) was simulated and the calculation was mainly during the in-vessel phase, the effect of the ADS1-4 on the AP1000 was analyzed and evaluated. The analysis results showed that the execution of ADS 1-3 was effective to decrease the pressure of the primary system at the early stage and lead to quick and effective injection of cold water from core make-up tank (CMT) and accumulators into the core, but it was impossible to decrease the pressure of primary system effectively to allow the injection from the internal refueling water storage tank (IRWST) to the core and the core was uncovered. The zirconium-water reaction will produce the hydrogen. The execution of ADS4 accelerated the injection from the accumulators to the core and decreased the pressure of the primary system quickly so that the IRWST injection was available and the core was still covered with water. It was of great importance to the safety of AP1000. The passive containment cooling system (PCS) was an effective measure to control the pressurization of the containment.
Study on Simulation of the Building Fires Using an Advanced New Model
Authors: Shan Zuhua , Niu Fenglei , Zhang Yan , Hao Pengfei
Published: February 2014

Building fires have been paid significant attention in the nuclear power station’s safety. In order to study the stratification phenomena of the enclosure fires and predict the interface location of upper hot layer filled with smoke and lower cold layer filled with fresh air and upper layer temperature of enclosure fires, an advanced new model is used in this paper, in which one–dimensional differential equations are used to describe the temperature and species distributions of the ambient fluid. And the results of Steckler’s fire experiments are used to compare with the simulation results of five sets of experiment using the new model. The results indicate that this model gives a very good prediction for the location of the interface and the upper layer temperature, especially for the cases with a lower fire heat release rate, even without considering the radiation heat transfer.
Study of Three-Dimensional Radiation Field Monitoring System Based on Mesh Network
Authors: Yu Qian , Cheng XiaoLei , Zhang XiYing
Published: February 2014

In order to meet the need of three-dimensional radiation field monitoring, a new technique of three-dimensional radiation field monitoring has been proposed and the system model has been built based on mesh network after investigating into wireless monitoring system characteristics of radiation feild in nuclear power plant. The model uses detecting nodes to collect the timely data in radiation field and accomplishes data preprocessing, and finally present results in monitoring platform in the form of three-dimensional radiation field. The essay will also demonstrate the study of the layout of detecting nodes. The application of this system reduces accidents probability and workers’ pressure by gathering environment parameters and transmitting real time monitoring data stably. It could also present the radiation intensity and other parameters distribution visually in the monitoring hall, with mesh topology technology in the region where exists security threats toward staff.
CFD Analysis of Thermal Stratification Phenomena in the RNS Suction Line of a Gen-III Nuclear Power Plant
Authors: Li Yan , Lu Daogang , Wang Zhigang , Wu Jian , Yu Fengyun
Published: February 2014

Thermal stratification phenomena in piping systems of nuclear power plant would threaten the structural integrity of pipes, which are caused by the significant change of water density with temperature. To provide temperature gradients for the stress analysis of Normal heat Removal System (RNS) suction line of a Gen-III nuclear power plant, the relevant thermal stratification phenomena are analyzed by CFD in this paper. Cases without leakage (normal power operation) and with leakage are both studied. The results show that the first portion of pipe (one meter or so) near the hot leg is isothermal for normal power operation due to the penetrating flow. In the remaining portion, the radial temperature drops are of the order of 20∼27 K for no leakage case. For the leakage case, the radial temperature drops are 23 K or less, which are relatively smaller than those for the no leakage case due to the net hot flow from the hot leg to the valve.
Spatially Dependent Resonance Cross Sections in a Fuel Plate
Authors: Zhang Qian , Wu Hongchun , Cao Liangzhi , Zheng Youqi
Published: February 2014

The classical resonance absorption theory known as equivalence theorem which treats two-region problem (Fuel/Moderator) has been generalized to arbitrary sub regions of fuel pin by Stoker/Weiss technique and SDDM (Spatially Dependent Dancoff Method). Stoker/Weiss technique and SDDM were derived only for infinite cylinders. To calculate the shielded resonance cross sections for sub regions in the fuel plate, a method is presented by geometrical derivation of first flight blackness for slab geometry. To validate the method, probability approximation and cross section generation are all compared to the benchmark problems. It is concluded that this method is practically useful and appropriate for generating the effective cross-sections in sub regions of a fuel plate.