核动力工程

Analysis of Heat Transfer Characteristics in Core of Supercritical Water Cooled Reactor

JIE Heng, WANG Yan

2008, 29(2): 1-4,10.

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Seven most widely used heat transfer correlations were employed to calculate and analyze the heat transfer capacity in the core of Supercritical Water Cooled Reactor（SWCR） on rated condition.The results showed that maximum discrepancy of outlet clad temperature on average channel employing various correlations is 27℃.The results also showed that the clad surface temperature limitation could be satisfied even using same flowrate as average channel in hot channel when Koshizuka-Oka correlation is applied.However,18% and 40% more coolant flowrate must be distributed into hot channel than average channel if Jackson correlation and Krasnoshchekov correlation is employed respectively.It could be concluded that the choice of heat transfer correlation will seriously affect the reactor design.It was also found that the effect of buoyancy on heat transfer could be ignored on rated condition of SWCR.

Numerical Research on Two-phase Flow and Heat Transfer Characteristics on a Liquid-Reactor Experimental Model

YANG Li-xin, BA Li-ming, NIE Hua-gang, SONG Xiao-ming, NIU Wen-hua

2008, 29(2): 5-10.

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The MUSIG model and the mono-disperse model are used to simulate the two-phase flow and heat transfer phenomena in a liquid-reactor experimental model.In simulation with MUSIG model,bubbles in the model is divided into 5 groups and these groups are used to analyze the bubble distribution and change in the flow.Luo and Svendsen model and Prince and Blanch model are used to describe the break and aggregation between bubbles.In simulation with mono-disperse model,bubbles in the liquid-reactor model is considered to have only one size,so the break and aggregation between bubbles are not in consideration.The results of simulation with the two models are compared,and the MUSIG model comparatively tallies with results of experiment.

Distributed Parameter Model of Plate-Fin Recuperator in High Temperature Gas-Cooled Reactor with Direct Helium Turbine Cycle

DING Ming, WANG Jie, YANG Xiao-yong, SHI Lei, SU Qing-shan

2008, 29(2): 11-15,29.

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Recuperator was one of the important equipments in high temperature gas-cooled reactor with direct helium turbine cycle.A one-dimensional distributed parameter model was proposed for the counter-flow platefin recuperator to study its dynamic performances.Method of whole-field dispersed and solved was used to solve the model.Through the simulation and analysis of the step disturbance,response of the outlet temperature on the same side with disturbance might be expressed by inertial item with delay.Response of the outlet temperature on the different side with distribution could be expressed by inertial item with very short transition time.For flow rate disturbance,response of outlet temperatures also could be expressed by inertial item.The response process to a coupled disturbance of temperate and flow rate was simulated.The results showed that the distributed parameter model was not only used to simulate simple transient processes,but also used to complicated ones.

Experimental Study for Heat Transfer Enhancement of Oil Cooler

YAN Chang-qi, HOU Shan-gao, CAO Xia-xin

2008, 29(2): 16-19.

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The heat transfer experimental study for three integral pin-fin tube oil coolers is performed.The experimental results indicate that for the same oil mass flow rate,when the tube number of pin-fin tube oil cooler is half that of smooth tube,the heat transfer coefficient of pin-fin tube oil cooler is 2.2 times that of bar tube.The experimental results indicate that the pin-fin tube oil cooler is of less tube plate and better heat transfer characteristics

Effect of Rolling Motion on Forces Acting on Bubbles in Sub-Cooled Boiling Flow

QIN Sheng-jie, GAO Pu-zhen

2008, 29(2): 20-23.

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Additional accelerations caused by rolling motion impose the forces acting on bubbles in subcooled boiling flow.Comparing with other forces acting on bubbles,the effect of additional accelerations may be neglected.The fluctuations of flow rate caused by additional accelerations have a close relationship with the forces acting on bubbles.The forces acting on bubbles under rolling motion are computed,and compared with that under non-rolling.The results show that rolling motion has great influences on the forces acting on bubbles,which makes obvious changes on the position of vapor bubble departure,hence influences the heat transfer in boiling flow.

Investigation on Mass Flux Distribution and Asymmetrical Cooling in A Plate-Type Fuel Reactor

LU Qing, QIU Sui-zheng, TIAN Wen-xi, SU Guang-hui

2008, 29(2): 24-29.

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A program was developed to calculate the mass flux distribution in the whole core and the asymmetrical cooling of fuel pins for a plate type fuel reactor by applying the proper physical model.Three iterative algorithms for mass flux distribution calculation and two iterative algorithms for temperature field calculation of plate fuel element under asymmetrical cooling condition were proposed and compared by applying in a subassembly.The results showed that the flow distribution is mainly determined by the core structure,although it is also impacted by the power distribution in the core.The asymmetrical cooling condition seriously impacts the temperature field and power distribution of the fuel pins.

Shape of Isolated Bubble in Downwardly Inclined Gas-Liquid Two-Phase Flows

GU Han-yang, GUO Lie-jin

2008, 29(2): 30-34,42.

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Experimental and theoretical studies on the shape of a single bubble in downwardly inclined gas-liquid two-phase flows are carried out.Measurements of the shape are made using conductance probes and visual observation in this paper.The experimental results show that bubble turning phenomenon of bubble shape happens at low Froude number,which is defined by gas/liquid mixture velocity.The turning phenomenon significantly affects the bubble velocity.It is also found that the shape of bubble is mainly determined by the Froude number rather than the bubble length.A hydrodynamics model for the shape of bubble tail is de-rived based on a one-dimension two-fluid model.The model well predicts the characteristics of the bubble tail when compared with the experimental results.The theoretical analysis shows that the critical Froude number,at which the turning phenomenon happens,increases as the downward inclination and pipe diameter increase.

Fast Calculations of Supercritical Water Thermophysical Properties

LI Chang-ying, DAN Jian-qiang, TAN Shun-qiang, D. C. Groeneveld

2008, 29(2): 35-38.

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A new method that combines look-up table（LUT） and IAPWS-IF97 formulation is presented for fast calculation of supercritical water thermophysical properties.The pressure and temperature ranges are critical pressure-100 MPa,10~800 ℃,respectively.The comparison shows that the developed method can be used in the above stated range for direct industrial computation.The computation efficiency shows that the method can shorten the computing time greatly.A universal program for computation of supercritical water has been developed using the developed method.

Design for Fuel Management of China Integrated Advanced NPP Reactor Core

PENG Gang, LI Dong-sheng

2008, 29(2): 39-42.

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The calculations of loading patterns（LPs） in each cycle（first cycle,transition cycle and equilibrium cycle） have been completed using the SCIENCE nuclear code package.To meet the requirement of refueling cycle length for 2 years,the enrichment of fuel,the numbers of refueling assemblies and the LPs have been tested.‘In-out’ strategic has been used to reduce the neutron leakage.In first cycle,the higher enrichment fuel assemblies（FAs） have been loaded in the outer region of the core,while the lower enrichment FAs have been loaded in the inner region.In cycle 2,44 fresh FAs（4.95%） have been arranged in the core,44 old FAs have been discharged,while old FAs have been loaded in the outer region.In cycle 3,only FAs with 0,12 and 20 Gadolinium rods have been used.As the result,the design criterion for the maximum discharge burnup has been met.

Analysis of Factors Influencing Dynamic Characteristics of Steam Generator U-Tubes

LIU Min-shan, LIU Tong, DONG Qi-wu

2008, 29(2): 43-47.

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Based on the synthesis of all sorts of factors influencing the dynamic characteristics of steam generator（SG） U-tubes,the natural frequencies,corresponding mode shapes of the U-tubes are analyzed un-der practical load cases and environment by theoretical analysis and numerical simulation.The formulae of calculating the natural frequencies of U-tube considering the effects of internal and external pressures,as well as non-uniform distributed mass density in U-tubes are deduced.Study results reveal that the boundary constraints,pressures in primary and secondary sides,supporting width,support plate number and elbow tube factor bring a strong effect on the dynamic characteristics of U-tubes,and the non-uniformities of pressures and mass along the whole tube,tube deadweight and the heat expansion have little.

Application of Modal Analysis in Design of Steam Generator Support Plate

LIU Min-shan, LIU Tong, DONG Qi-wu

2008, 29(2): 48-51,96.

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Based on the theory of elastic plates and shells,the concept of equivalent isotropic solid plate elaborated for perforated tubesheet,the new methodology to determine the effective elastic constants by modal analysis of such plates is described.The computing formula of effective elastic constants for plates with hetero-diameter dual-holes are given in terms of the stiffness superposition method.By modal approach,the effective elastic constants of support plates with various perforation shapes popularly used in steam generator,such as hetero-diameter holes,trefoil and quatrefoil holes,are obtained.The universal relationship between the effective elastic constants and the mass fraction are established,which can be used for perforated plates with any arbitrary hole shape.The study results indicate that the curves for calculating effective elastic constants adopted in ASME are improper in non-circular hole perforated plates and shall not to be applied in the support plates and tubesheet of steam generator.

Research on Stress Intensity in Edge Stress Field of RPV Nozzle

WANG Xiao-bin, MI Xiao-qin, WEI Ya-dong, YANG Min, CHEN Hai-bo

2008, 29(2): 52-54,69.

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The stress intensity of Reactor Pressure Vessel local wall-thinning area is calculated in the edge stress field of nozzle by ANSYS-a FE software.And the stress concentration factor,the membrane stress intensity and the membrane stress plus the bending stress intensity vs the dimension and location of part-through slot have been discussed.In this article we generalized that the membrane stress intensity is in-tensifying by long half axis in an external convexity curve,intensifying by depth in a straight line,weakening by distance in an inner concave curve,and the membrane stress plus the bending stress intensity is no changing by long half axis nearly,intensifying by depth in a straight line,weakening by distance in an inner concave curve,and the stress concentration factor is minishing by long half axis in an inner concave curve,augmenting by depth in a straight line minishing by distance in an inner concave curve.

Mechanism of Fuel Assembly Bowing in PWR and Preventive Measures

LI Wei-cai, XIAO Zhong

2008, 29(2): 55-57,73.

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Excessive fuel assembly bowing may induce reloading or unloading difficulties,incomplete control rod insertion,fuel assembly defect,and quadrant power tilts.To avoid fuel assembly excessive bowing is import for PWR safety operation.This paper analyzes the phenomenon of fuel assembly bowing and its impact,the factors that affect assembly bowing,mechanism and preventive measures of assembly bowing.

Effect of Sintering Temperature on Microstructure and Compressive Strength of B₄C-AlSi Eutectic Alloy

LIU Jin-yun, ZOU Cong-pei, CHA Wu-sheng, LIU Gai-hua, LAN Jun, FENG Quan-he

2008, 29(2): 58-60,104.

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The block neutron absorber of B₄C based on Al-Si eutectic alloy has been prepared by pow-der-metallurgy method.The effects of sinter temperature on microstructure,compressive strength,and ductility of sintered billets have been investigated.It has been shown that the sintering temperature decides sensitively the compressive strength and ductility of sintered billets.Sintered under 550,555,560,and 565 ℃,the billet shows different states,such as subsintered,best-sintered,over-sintered,and molten.Sintered under 550 ℃,the powder have not been metallurgically combined with each other.Beyond 560 ℃,the billets are molten.The 555 ℃ is the best sintering temperature,under which the powder have been partly melted and the metallurgical combination has been occurred,then the billets have a better ductility.

Transient Analysis of Residual Heat Removal System Used in 200 MW Low Temperature Heating Reactor

XU Zhao, WU Xin-xin

2008, 29(2): 61-65.

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The residual heat removal system of 200 MW Heating Reactor is constituted by 3 coupled natural circulations.The thermal-hydraulic analysis of the system has been performed in this paper,and the numeric model of the passive residual heat removal system has been constructed based on 1-demansion mass equation,momentum equation and energy equation.The program in this paper has been used to compute the model and simulate the heat removal process of the system.The results show that the heat transfer ability of the system meet the design requirements.

Piping Installation for Reactor Heavy Water System

YE Lin, HUANG Hong-wen, XU Xian-qi, QIAN Da-zhi, ZHOU Wei

2008, 29(2): 66-69.

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Characteristics and main installation steps for the piping of the reactor heavy water loop system were introduced in this paper.According to the system design,equipment accommodation and spot management,main issues with effect on the quality and schedule of pipeline installation were analyzed.Accordingly,some solutions were put forward,which included: work allocation should be made clear in documents;installation preparative such as design checkup and technology communication should be prepared completely;requirements of system cleaning,test items in every experiment,inspection in work and equipment maintenance should be considered in the system design;perfect documents distribution system and stock plan should be built;technology requirements and quality assurance should be claimed in contracts;quality should be controlled by way of external evidence,inspection in manufactory,exterior quality assurance examination,and test during consignment;series of management procedure should be established in detail.

Replacement Scheme Design of Reserve Helium Circulator of 10 MW High Temperature Reactor

MA Tao, HU Shou-yin, ZHOU Hui-zhong, SU Qing-shan, LIANG Xi-hua, WEI Li-qiang

2008, 29(2): 70-73.

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As one of the crucial innovations of 10MW high temperature reactor（HTR-10）,the current helium circulator using rolling bearings will be replaced by the reserve one using electromagnetic bearings.The feasibility of construction was certified by comparison between the two helium circulators and radiac measurement of the primary circuit.The construction schemes of inside vessel and outside vessel were proposed and simulated by 3-D digital emulation.The scheme of inside vessel was considered to be more feasible by comprehensive evaluation such as safety,difficulty and work amount.

A Matrix Method for Thermodynamic System Analysis of PWR Nuclear Power Plant

RAN Peng, ZHANG Shu-fang

2008, 29(2): 74-77,84.

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Considering the thermal economic analysis methods of PWR nuclear power plant,a Steam-water Distribute State Equation has been established and a Matrix for calculating the efficiency is derived by the aquivalent enthalpy drop theory,the theory of matrix and the normal thermal-equilibrium method,suitable for quantitative thermal efficiency analysis of secondary-circuit of PWR nuclear power plant.The structure of this matrix has a mapping relationship with the secondary-circuit of PWR nuclear power plant,and it can simplify the thermal economic analysis of PWR nuclear power plant,and provides a theoretic principle for analysing the economics of the thermal system of the secondary-circuitof PWR nuclear power plant.An example is given to illustrate the validity of the method,and it indicated that the thermal economics diagnostic method is well defined and easy to be used in system design and operation diagnosis.

Preliminary Analysis of Hydrogen Distribution and Risk under Severe Accident Conditions for QINSHAN NPP Unit 2 Containment

DENG Jian, CAO Xue-wu

2008, 29(2): 78-84.

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Severe accident sequences induced by large-break loss-of-coolant accident（LB-LOCA）,small-break loss-of-coolant accident（SB-LOCA） and station blackout（SBO） are calculated by using an integral systems analysis computer code for QINSHAN NPP Unit2.Hydrogen concentration distributions in the containment atmosphere and their potential deflagration risk are investigated,according to the hydrogen control and risk analysis standard of US 10 CFR.The results show that the average hydrogen concentration un-der LB-LOCA severe accident is nearly 10%,which means large scale hydrogen deflagration might occur.In spite of the possibility of local deflagration,such large scale deflagration under SB-LOCA and SBO severe accidents maybe cease.The analysis provides a reference for hydrogen control and severe accident management of QINSHAN NPP Unit 2.

Calculation of Physical Failure Probability of HTR-10 Residual Heat Removal System by Monte Carlo Method

XIE Guo-feng, TONG Jie-juan, HE Xu-hong, ZHENG Yan-hua

2008, 29(2): 85-87.

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The paper introduces Monte Carlo simulation method which is used to calculate the failure probability.Importance sampling Monte Carlo method is used to get the physical process failure probability of the residual heat removal system in 10 MW high temperature gas reactor（HTR-10） and to analyze the error.By Comparing the results,we find that the result calculated by Monte Carlo method is of the same order of magnitude as the result by the response surface method.The following conclusion is further confirmed: by using the residual heat removal system,the physical process failure probability of HTR-10 is reduced by at least 3 order of magnitude.

Analysis of In-vessel Hydrogen Source Term during Severe Accident Induced by SB-LOCA

GUO Ding-qing, DENG Jian, CAO Xue-wu

2008, 29(2): 88-91,101.

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With an integral severe accident analysis computer code,severe accident sequences induced by small-break loss-of-coolant accident（SB-LOCA） in Daya Bay Nuclear Power Plant（900 MWe PWR NPP） are calculated.The impacts of different sizes and locations on the progression of accident and the generation of the in-vessel hydrogen source term are analyzed in this paper.The results are summarized as follows: A mass of in-vessel hydrogen source generates at oncoming and initial stage of core melting;The amount of in-vessel hydrogen source term is with obscure relation to the break size,and with many different sizes of breaks,the amount of in-vessel hydrogen source term distributes in a small area,and the average amount of the hydrogen production is about 500 kg;The location of break has a small impact on the generation of the hydrogen source term.

Analysis of Shutdown Heat-sink Nuclear Safety Requirements for Qinshan Phase III CANDU 6

WEI Hua

2008, 29(2): 92-96.

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This paper generally introduced the sensitivity analysis of core overheating during the outage of CANDU 6 unit,and explained the specific heat sink requirements of TQNPP Unit 1 during Outage 101.It also discussed several actual implementation cases and issues during Outage 101,i.e.stop of SDC pump,refill upon loss of Class IV power and use of temporary cover,to which some recommendations are also given.

Calculation of Nuclide Absorption in Upper Room of HTR-10 Helium Circulator

JIANG Ping, CAO Jian-zhu, LIU Tao, ZHOU Hui-zhong, LIANG Xi-hua

2008, 29(2): 97-101.

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This paper analyzes the physical model of SPATRA procedure and applicable data,proposes the calculation method for the nuclide absorption in the upper room of the primary helium circulator,and conservatively predicts the main nuclide adsorption in the upper room by the 4 layer calculation model.137Cs and 131I are taken as the example to explain the calculation result.The prediction results show that there is not much nuclide adsorpted in the primary helium circulator,and exchange activities can be carried out with certain protection.

Application of LabVIEW in Radiation Monitor and Database Management System of Xi’an Pulse Reactor

LI Zhong-liang, YUAN Jian-xin, WANG Kai, MOU Zheng-qiang

2008, 29(2): 102-104.

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The radiation monitor and the database management system of XAPR uses singlechip tocollect,process and transport the signal,and uses RS-485 bus to constitute a testing network.The monitor anddatabase management software was developed by LabVIEW.This software can monitor the radiation levelsof each area of XAPR in real-time and can manage the dose database.The test run of the system indicates thatthe system is steady and trusty,and can achieve the targets.So the system can be applied.

Design of Digital Rod Control and Position Indication System of Ling’ao Phase II Project

HUANG Ke-dong, ZHANG Ying, WANG Hua-jin, ZHONG Li-ping, LI Guo-yong, ZHANG Rui

2008, 29(2): 105-109.

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The TXS,digital and redundant techniques are applied in the rod control and position indication system（RGL） to perform the control of power and temperature in the reactor integrally in Ling’ao Phase II project.The high power transistors are used to implement the control of the coils one by one in the control rod drive mechanism（CRDM）.The grouped position detectors are used to monitor the position of the control rod in the reactor core.This paper introduces the requirements and traits of system design of RGL system in Ling’ao phase II.Comparing with the system design of Ling’ao phase I and Qingshan phase II,RGL system in Ling’ao phase II is more convenient for the information exchange,operation and maintenance,and is with high reliability.

Research of Pressurizer Pressure and Water Level Control Based on Programmable Controller

ZHANG Yao, ZHANG Da-fa

2008, 29(2): 110-113.

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The programmable logical controller is used to construct the pressure and water level control successful.system of pressurizer in this paper.The function module and software control technology are utilized to realize the pressure and water control strategy and the system is modularized and flexibly configured.The test results are analyzed and the application of the programmable logical controller is proved to be effective and successful.

Field-Bus Based Control-Management Integrated System for Nuclear Power Plant

ZHOU Jun-tao, YU Ren, REN Yin-xiang

2008, 29(2): 114-118.

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The characteristics and types of the field-bus as well as its application in Nuclear Power Plant are discussed in this paper,and a new control-management integrated system for nuclear power plant that is based on field-bus is put forward.Analysis and study is carried out in terms of the hierarchy and architecture,network protocol and field-bus type and data information flow.

Issues and Possible Solutions of Nuclear Power Standard System in China

LI Xiao-yan, PU Ji-long

2008, 29(2): 119-123,128.

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The construction of nuclear power standard system shall follow the development state of the nuclear power in China,and shall be taken as one of the important support for the self-reliance in the development of nuclear industry.However,there is no national nuclear power standard systems that is matched to the industrial system and technology base in China.It is widely recognized in nuclear industry that the nuclear power standard system must be established.This paper discusses various problems of the current national nuclear power standard system and suggests the possible solutions.Electricity industry standard system can be referenced to set up nuclear power standard system in China.Considering the need to construct a lot of PWR nuclear power stations in china,the design and construction standard system for PWR NPP shall be established.The nuclear safety guides in China shall be incorporated into the nuclear power standards.

Fault Diagnosis of Nuclear Equipment Based on Artificial Immune System

PENG Yuan, ZHANG Chun-liang, ZHAO Hui, YUE Xia

2008, 29(2): 124-128.

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As the nuclear equipment is complicate and special,this paper put forward a novel fault diagnosis method for nuclear equipment based on artificial immune system and the principle to model with nega-tive-selection algorithm and further identify the fault with clone-variation algorithm.Features are extracted with the signal that was sampled in a rotary machinery,then the result is input to the AIS model.Simulation result shows that the model can identify each fault type successfully.

2008 Vol. 29, No. 2