核动力工程

Research of Preparation of Large Diameter Ceramic UO₂ Spherite by Sol-Gel

MENG Qingyang, TANG Xiangyang, LIU Jinhong, ZHU Changgui

2011, 32(4): 1-5.

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Ceramic UO₂ spherites with diameter of 2-3mm are prepared from depleted uranium by Sol-Gel method.The whole process includes incineration,dissolution,mixing,gelatinization,aging,washing,dry incineration,reducing sintering,and optimization.The physical and chemical analysis results indicate that the inner structure of the UO₂ spherite is compact and homogeneous,which density is 97±2% T.D.;The content of impurity is low;The O/U is 2.00±0.01;The average diameter is 2.6 mm;The sphericity is less than 1.10,and the average sphericity is 1.06.

Study on Sintering Behavior of UO₂ Microsphere Doped with Titanium

WANG Hui, LIU Jinhong, YIN Rongcai, REN Meng, ZHANG Jia

2011, 32(4): 6-9,85.

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This paper studies the preparation of titanium（Ti） UO₂ microspheres by selgol method,and its microstructure,pore distribution and grain size are observed by optical microscope,SEM and EDS.Its density is measured by water immersion method.The results show that doping small amount of titanium can significantly improve the sintering behavior at a certain sintering temperature.In this study,the optimal doping content of Ti is within 0.3% wt and the best sintering temperature is 1250-1350℃.This paper also studies the activated sintering mechanism,with a conclusion that the material transfer mechanism may enhance the cation diffusion,and residual oxygen interaction.Additionally,it studies the distribution of titanium in the microspheres.Besides the solid solution of titanium in UO₂ microspheres,the redundant enriches in the grain boundary in the form of free particles.

Establishment of Mission Profile of Fuel Cycling System in Pebble Bed Reactor

CENG Kai, SHEN Peng, DOU Dong, ZHANG Haiquan, LIU Jiguo

2011, 32(4): 10-13.

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The critical system,i.e.,the fuel element cycling system,in pebble bed reactors was analyzed,and the dynamic characteristics of pneumatic transportation of near-diameter sphere was investigated.The expression of pneumatic force on sphere and the equation of sphere velocity in elbow and straight pipes were derived.Mission profile of cycling system was established based on the motion analysis and the reliability block diagram of cycling system.

Experimental Investigation of Hydrogen Embrittlement of 65Mn Steel with Small Punch Testing Method

WANG Zhaoxi, QU Baoping, XUE Fei, YANG Hao, SHI Huiji

2011, 32(4): 14-18.

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The hydrogen embrittlement of the 65Mn used as the high strength spring material in Nuclear Power Plant is investigated with small punch testing method.During cathode hydrogen charging with the cur-rent density of 1mA,4mA,10mA and 20mA,the hydrogen atom is absorbed,diffused,segregated in the metal which causing the initiation of the brittle cracks.The specimens are punched after charging with small punch testing method.It is found that the maximum loading level,the total energy,the critical fracture strain and fracture stress reduce with the increment of the current density which means the strength and toughness re-duce.The macro fracture surface changes from the ductile fracture to the typical brittle fracture and the micro fracture properties change from the ductile voids to the intergranular brittle fracture.The hydrogen concentrations in the metal were analyzed with theoretical model.It was found with the increment of the current density,the hydrogen concentration increases.

Characterization of Interaction between U-Mo Alloy and Al Diffusion-Couple

LIU Yunming, YIN Changgeng, SUN Zhanglong, CHEN Jiangang, SUN Xudong

2011, 32(4): 19-23.

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In this paper,the interaction behavior of U-Mo/Al was studied with the diffusion-couple method,and the couple was continuously jointed by hot-pressing with special device.Annealing experiments were accomplished in a vacuum hot-pressing furnace,and at 550～570℃ for 5～21hours.The results show that the morphology and composition of interaction Layer depend on the interaction layer thickness.The con-tent of U（Mo） and Al is mutational at the interface of U-Mo/interaction layer/Al.The layer close to U-Mo side is mainly composed of product（U,Mo）Al 3,while the Al side is composed of（U,Mo）Al 4 and UMo 2 Al 20.Diffusion process of U-Mo/Al is Al immigrating over the Al/U-Mo original interface into U-Mo side and re-acting with U-Mo,subsequently the interaction layer is growing into Al.

Fretting Wear Behavior of TA16 Alloy Materials

ZHANG Yafei, REN Pingdi, ZHANG Xiaoyu, LI Zhangxiang, ZHU Minhao

2011, 32(4): 24-28.

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The fretting wear behavior tests on cylinder contacts of TA16/0Cr18Ni9 have been carried out under the normal load（50N and 80N）,frequency（2 Hz） and displacement amplitude（from 80μm to 200μm） using the hydraulic fretting test machine with a high precision.Experimental results showed that the normal loads and displacement amplitudes may have remarkable influence on the damage degree and injury mechanism of materials.Degree of injury of material increases with the increasing displacement amplitude and normal loads,however,the friction coefficients decrease.Three-body layer consists of two parts: plastic deformation layer and debris layer that has effect on the restriction and control in the fretting wear.The analysis found that there are some micro-crack and delamination in the plastic deformation layer,and the abrasive dust have based mainly on the oxide of titanium and titanium alloys,and attached on the surface of wear.Adhesive wear,abrasive wear and friction oxidation are the main fretting wear mechanism.

Study on Grey Correlation Degree of Influence Factors on ONB in Narrow Channel under Natural Circulation

LIU Ping, ZHOU Tao, ZHANG Ming, SHENG Cheng, ZHANG Jigang, HUANG Yanping

2011, 32(4): 29-32.

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Grey correlation degree calculation model which reflects the positive and negative relation of sequences is applied to calculate the relevant influence factors on ONB（Onset of Nucleate Boiling） in narrow channel under natural circulation.Results show that in the range of selected parameters: heating power shows a negative correlation with the distance between ONB and channel inlet,which proves that heating power promotes ONB occurrence;mass flow and pressure show positive correlation with the distance between ONB and channel inlet,which proves that these two factors postpone ONB occurrence.And a weak positive correlation exists between the narrow channel gap and inception heat flux at the location of ONB.The occurrence of ONB will be postponed as the value of channel gap becomes bigger.

Simulation of IBIF in HWR Horizontal Fuel Channel

YUAN Jingtian, LI Jingxi, CAO Xuewu, TONG Lili

2011, 32(4): 33-36.

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In this paper,the phenomenon of Intermittent Buoyancy Induced Flow（IBIF） in the heavy water reactor（HWR） fuel channel is simulated.The IBIF phenomena and the fuel bundle temperature are analyzed.The calculation results indicate that the fuel bundle temperature rises and falls periodically accompanying with the periodical initiation,grow and discharging of the bubble from the channel;the bubbles dis-charge from the two opening on the two sides alternately which removes the decay heat of the fuel bundles,and consequently the temperature of fuel bundle keeps low relatively.

Effects of Additional Inertia Force on Bubble Breakup

PAN Liangming, ZHANG Wenzhi, CHEN Deqi, XU Jianhui, XU Jianjun, HUANG Yanping

2011, 32(4): 37-41.

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Through VOF two-phase flow model,the single bubble deformation and breakup in a vertical narrow channel is numerically investigated in this study based on the force balance at the process of bubble breakup.The effect of surface tension force,the additional inertia force and bubble initial shape on bubble breakup are analyzed according to the velocity variation at the break-up point and the minimum necking size when the bubble is breaking up.It is found that the surface tension force,the additional inertia force and the bubble initial shape have significant effects on the bubble breakup through the fluid injection toward to the bubble,which finally induces the onset of bubble breakup.

Flow Patterns of Air-Water Two-Phase cross a Tube Bundle and Their Effect on Excitation of Flow-Induced Vibration

JIANG Naibin, ZANG Fenggang, ZHANG Yixiong

2011, 32(4): 42-45,71.

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Based on the existing Taitel map and Ulbrich & Mewes map,a new flow pattern map was proposed based on the analysis of the experimental data.According to this map,three typical flow patterns can occur in vertically upward shell-side flow in tube arrays,namely,the bubbly,churn-bubbly and intermittent flows.In different patterns,the typical time-history,power spectral density（PSD） curves and Strouhal numbers associated with the peak PSD of excitation forces working on tube bundle are given.There are obvious effects of flow pattern on the time-history of excitation forces,the shapes and frequency distribution of PSD curves.

Numerical Prediction of Critical Heat Flux in Narrow Rectangular Channels

ZHOU Lei, LIU Xiangfeng, YAN Xiao, HUANG Yanping, CHEN Bingde

2011, 32(4): 46-51.

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Results calculated by the Weisman model and Kwon model are compared and analyzed based on the existing experimental data points.It reveals that the two models have similar tendency for the CHF ratio distribution while the Kwon model has little scattering thus is more accurate.Since both the existing bubble crowding models predict CHF in narrow rectangular channels with poor accuracy,they are not pro-posed to be used for NRCs directly.Considering the characteristics of a narrow rectangular channel,the Kwon model is modified and extended.The improved model has better prediction accuracy and eliminates the systematic deviation from the experimental results.

Experimental Study on Post-Dryout Transition Boiling Heat Transfer in Rectangular Channel

LI Hongbo, CHEN Bingde, XIONG Wanyu

2011, 32(4): 52-57.

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The experiment of post-dryout transition boiling heat transfer in rectangular channel has been carried out on the basic experimental loop of flow and heat transfer.The characteristics of the transition boiling heat transfer is analyzed.The effects of thermal-hydraulic parameters such as inlet quality,mass flow velocity,and system pressure on the characteristics of post-dryout transition boiling heat transfer are experimentally researched.The results show that the post-dryout transition boiling is unstable,wall temperature oscillates fiercely,and the mass flow velocity and pressure drop oscillate as also;the post-dryout transition boiling heat flux and heat transfer coefficient are decreased,but the wall temperature is increased with the increasing of inlet quality;and the post-dryout transition boiling heat flux and heat transfer coefficient are increased,but the wall temperature is decreased with the increasing of mass flow velocity or system pressure.

Reliability Analysis of in-Service Inspection for Key Components in Nuclear Power Plants

ZHANG Jun, DING Hui, LI Ming, ZHANG Yicheng, CHEN Huaidong, LU: Tianming

2011, 32(4): 58-61,71.

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The reliability of in-service inspection was calculated and analyzed based on the probability model and ultrasonic testing numerical model.The POD curve and its 95% lower bound of different defects in girth weld of RPV were calculated.The result indicated that the reliability analysis could give a quantitative evaluation for in-service inspection.

Analysis of AP1000 Reactor Power Control System

ZHANG Xiaodong, LIU Lin

2011, 32(4): 62-65.

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The reactor power control modes applied by common nuclear power plants and AP1000 is introduced briefly.The advantages and shortcomings are compared and summarized.Advice is also made to the control strategy of the first and continued SMNPC plants.Additionally,the characteristics of the AP1000 reactor control system are summarized and problems which the operators will probably encounter are also analyzed.

Development and Application of Nuclear Power Plant DCS Closed-Loop Test Platform

HOU Dong, LIN Meng, YANG Zongwei, LIU Pengfei, YANG Yanhua

2011, 32(4): 66-71.

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A simulation platform with high flexibility and extensibility for nuclear power plant DCS closed-loop test has been developed.The system modeling for Ling’ao Phase II nuclear power plant has been built.Through an example of pressurizer pressure and water level control system testing under the condition of a 10%FP turbine power step-down,the way of using the platform for closed-loop DCS test and how to locate DCS problems were demonstrated.This test platform has been applied to DCS closed-loop test in Ling’ao Phase II successfully.

Study on Quantitative Reliability Analysis by Multilevel Flow Models for Nuclear Power Plants

YANG Ming, ZHANG Zhijian

2011, 32(4): 72-76.

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Multilevel Flow Models（MFM） is a goal-oriented system modeling method.MFM explicitly describes how a system performs the required functions under stated conditions for a stated period of time.This paper presents a novel system reliability analysis method based on MFM（MRA）.The proposed method allows describing the system knowledge at different levels of Abstraction which makes the reliability model easy for understanding,establishing,modifying and extending.The success probabilities of all main goals and sub-goals can be available by only one-time quantitative analysis.The proposed method is suitable for the system analysis and scheme comparison for complex industrial systems such as nuclear power plants.

Application of DFM in Human Reliability Analysis

YU Shaojie, ZHAO Jun, TONG Jiejuan

2011, 32(4): 77-82.

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Combining with ATHEANA,the possible to identify EFCs and UAs using DFM is studied;and then Steam Generator Tube Rupture（SGTR） accident is modeled and solved.Through inductive analysis,26 Prime Implicants（PIs） are obtained and the meaning of results is interpreted;and one of PIs is similar to the accident scenario of human failure event in one nuclear power plant.Finally,this paper discusses the methods of quantifying PIs,analysis of Error of commission（EOC） and so on.

Analysis and Treatment of Cladding Layer Defects on Outlet Nozzle of Reactor Pressure Vessel

WANG Yongjiao, YANG Zhipeng

2011, 32(4): 83-85.

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This paper introduces the quality problem and defects disposition of cladding layer of one outlet nozzle of the reactor pressure vessel in terms of the welding process.Based on the analysis of the cause of cladding layer detachment,the corresponding treatment processes are discussed.

Optimal Design of Nuclear Reactor Primary Circuit Volume

QIN Huimin, YAN Changqi, WANG Jianjun

2011, 32(4): 86-90.

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The volume evaluation models for the main equipments in the primary circuit are established by the optimal theories and taking the primary circuit volume as a performance evaluation index in the scheme design of a nuclear power plant.Based on the sensitivity analysis of thermal parameters and structural parameters which influence the volume of nuclear equipments in the primary circuit,the design is optimized using an improved complex method algorithm.The optimization result indicates that the volume of the optimized design is 13.15% less than that of the original design,which could provide a theoretic reference for the engineering design.

Development and Validation of Isolation Amplifiers for Nuclear 1E Application

LI Dewen, HUANG Wenjun, LI Jingjing, WANG Kai, YANG Wenlong

2011, 32(4): 91-94.

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This paper presents the design scheme of nuclear-1E-class isolation amplifiers for safety shutdown system of nuclear power plants.The scheme adopts the switching power supply technology and the analog power supply techniques to design the power module.And then,the design methods for improving the stability and reliability of the isolation amplifiers is detailed from the point view of anti-interference,the device derating and anti-seismic.Performance validation shows that the product has passed all tests of nuclear-1E-class certification program and satisfies the nuclear 1E requirement,and its reliability and various technical indicators have reached or exceeded the similar foreign products.

Effect of Boric Acid Concentration on Reactor Coolant Pump Performance in PWRs

ZHANG Ye, WANG Xiaofang, JIE Hongen

2011, 32(4): 95-98,117.

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Using the computational fluid dynamics software Fluent,the three-dimensional turbulent flow of the reactor coolant pump（RCP） impeller at different flow rates and different concentrations of boric acid is simulated.With the experimental data,different effects on RCP performance are analyzed between water and boric acid,especially the influence to the impeller.The feasibility of taking the water as coolant in RCP by using numerical simulation method is analyzed,and the result shows that negligible differences in water and boric acid was acquired even in abnormal concentration of boric acid.It proved that the using of water as coolant when conducting a simulation is feasible.

Failure Data Process Considering Trend of Failure Rate for Repairable Equipments in Nuclear Power Plants

WANG Dalin, LIU Jingquan, HUANG Xiangrui

2011, 32(4): 99-104.

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To solve the problem of failure data processing for reliability assessment for repairable facilities in nuclear power plants,a improved approach based on the renewal theory,trend testing and Weibull process（a kind of NHP）is proposed.After analyzing the error risk of traditional life data processing methods,an actual example is calculated based on the failure data of water-cooling pumps in generator stator at some nuclear power stations to compare the two processing methods.The result proves that Weibull process fitting reflects the failure characteristics of repairable equipments failures with less complicate calculation process.

Study on Intelligent Fault Diagnosis System of Nuclear Power Plants Based on Information Fusion Technique

CAI Meng, ZHANG Dafa, ZHANG Yusheng, JIN Renxi

2011, 32(4): 105-108,113.

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The technology of information fusion is used in the fault diagnosis for ship nuclear power plants in this paper.The space fusion structure is built based on fault tree expert system,NN diagnosis system,and mechanism model validation system.Not only the system deep-level knowledge,but also the shallow knowledge and the mechanism model knowledge are fully used.The simulation validation and verification showed that the information fusion diagnosis system could improve the fault diagnosis reliability effectively.

Fault Diagnosis Study on Secondary System of Nuclear Power Plants Based on MFM Model

MA Jie, GUO Lifeng, ZHANG Yusheng, PENG Qiao, LI Haiwei

2011, 32(4): 109-113.

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Based on the analysis of Secondary System of Nuclear Power Plants,a MFM model of Secondary System is established in this paper.Utilizing Visual C++,together with the theory of MFM model and CDG consequence analysis method,a trial fault diagnosis system is built and two simulation cases are used for the evaluating of the fault diagnosis performance.The emulation test proves that the system is able to identify the potential root fault exactly and have the advantage of high vision and understandability.

Method for Estimation of Activity in Decommissioned Nuclear Reactor Pressure Vessel

GUO Wuren, LIN Xiaoling, ZHENG Ningning

2011, 32(4): 114-117.

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The theoretical calculation and experimental measurement methods for estimation of activity in the decommissioned nuclear reactor pressure vessel were introduced.The physical estimation model was described,and Monte Carlo compute code and ORIGEN2 code were recommended to be employed for the calculation of the neutron flux and activity in the reactor pressure vessel.Two methods commonly used for determining the activity in the reactor pressure vessel were introduced in detail,i.e.,sampling from the reactor pressure vessel and from the irradiation tube.The neutron flux profile was established to predict the activity in the reactor pressure vessel.

Calculation of Radioactive Inventory of Activated Parts for Nuclear Power Unit and Analysis of Influence Factors

LIU Yang, LIN Xiaoling, CAI Qi

2011, 32(4): 118-121.

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Based on the operation characteristics of the nuclear power unit,the radioactive inventory of activated parts was calculated by ORIGEN2,and the effects of burn-up,operation mode and power change on the radioactive inventory for activated parts were analyzed.The results indicated that the radioactive inventory grew with the increasing of burn-up,and when the actual operation time was longer than the effective operation time,the increasing rate of nuclide activity approximated the burn-up increasing;Radioactive inventory of activated parts was influenced directly by the operation modes of the nuclear power unit,and under same reactor load,operation power and burn-up,the radioactive inventory for non-continuous operation mode is less than that for the continuous operation mode.Effects of operation modes on radioactive inventory reversed with half life of nuclides.Under same burn-up and longer operation time,the effect of operation power change on the radioactive inventory is not obvious.

Evaluation of Dose Arising from ²²²Rn,Decay Products of ²²²Rn and ²²²Rn to Staff from a Certain Nuclear Power Plant

WU Hexi, LIU Yujuan, YANG Bo, TAN Guoxiu, LIU Qingcheng

2011, 32(4): 122-126.

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The staff of the plant was grouped based on their work ranges.The annual effective dose resulted from radon and the decay products of ²²²Rn/²²⁰Rn is studied by measuring radon concentration with double filter membrane method and by measuring the concentration of ²²²Rn/²²⁰Rn short life radioactive decay products with five-count method.Based on the results,specific protection measures are proposed for high radon areas.The study results show that the monitoring data for all areas except the spent fuel pool was smaller than the recommended values by ICRP.

Burn-up Function of Fuel Management Code for Aqueous Homogeneous Reactors and Its Validation

WANG Liangzi, YAO Dong, WANG Kan

2011, 32(4): 127-130,142.

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Fuel Management Code for Aqueous Homogeneous Reactors（FMCAHR） is developed based on the Monte Carlo transport method,to analyze the physics characteristics of aqueous homogeneous reactors.FMCAHR has the ability of doing resonance treatment,searching for critical rod heights,thermal hydraulic parameters calculation,radiolyticgas bubbles’ calculation and burn-up calculation.This paper introduces the theory model and scheme of its burn-up function,and then compares its calculation results with benchmarks and with DRAGON’s burn-up results,which confirms its burn-up computing precision and its applicability in the burn-up calculation and analysis for aqueous solution reactors.

Dynamic Characteristic Analysis for Liquid Storage Container in Nuclear Reactors with a Potential-Based Fluid Formulation

AI Honglei, ZHANG Yixiong

2011, 32(4): 131-133.

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The calculation scale of fluid-structure coupling is so huge that it was rarely used in engineering.A potential-based fluid formulation is used in this paper.The formulation connects the coupling terms of the fluid-structure characteristic value equation,so nonlinear continuous equation of liquid flowing is simplified to linear elliptic equation.In this way,the calculation scale was reduced to a great extent.It was verified in many examples that the formulation is very efficient to solve the dynamic problem of liquid storage container,which can be used in nuclear reactor engineering.

Study on Archive Management for Nuclear Facility Decommissioning Projects

HUANG Ling, LIAO Bing, ZHOU Hao, GONG Jing, LUO Ning

2011, 32(4): 134-137,142.

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This paper introduces the main features and status of the archive management for nuclear facility decommissioning projects,and explores and discusses the countermeasures in its archive management.Taking the practice of the archive management system of a reactor decommissioning project as an example,the paper illustrates the establishment of archive management system for the nuclear facility decommissioning projects.The results show that the development of a systematic archive management principle and system for nuclear decommissioning projects and the construction of project archives for the whole process from the design to the decommissioning by digitalized archive management system are one effective route to improve the complete,accurate and systematic archiving of project documents,to promote the standardization and effectiveness of the archive management and to ensure the traceability of the nuclear facility decommissioning projects.

Study on Temperature Dropping during Pressure Vessel Discharge Process

LIU Qingjiang, YE Tao

2011, 32(4): 138-142.

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The gas in the pressure vessel during discharging expands and works to the environment,and its temperature falls to cause the drop of the temperature of the pressure vessel.This paper proposes a simplified cylinder model by taking the discharging of a oxygen tank in winter as an example.The factors contributing to the temperature drop and the temperature drop range during the discharging of the pressure vessel is studied by the theoretic analysis and numerical calculation,and the optimization design for this condition and the lowest allowable operation temperature to avoid the low temperature condition are given..

2011 Vol. 32, No. 4