核动力工程

Development of Burnup Calculation Function in Reactor Monte Carlo Code RMC

SHE Ding, WANG Kan, YU Ganglin

2012, 33(3): 1-5,11.

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Abstract:
This paper presents the burnup calculation capability of RMC,which is a new Monte Carlo（MC） neutron transport code developed by Reactor Engineering Analysis Laboratory（REAL） in Tsinghua university of China.Unlike most of existing MC depletion codes which explicitly couple the depletion module,RMC incorporates ORIGEN 2.1 in an implicit way.Different burn step strategies,including the middle-of-step approximation and the predictor-corrector method,are adopted by RMC to assure the accuracy under large burnup step size.RMC employs a spectrum-based method of tallying one-group cross section,which can considerably saves computational time with negligible accuracy loss.According to the validation results of benchmarks and examples,it is proved that the burnup function of RMC performs quite well in accuracy and efficiency.

Numerical Computation of Eigenvalue for Steady Neutron Transport Equation

HONG Zhenying, YUAN Guangwei, FU Xuedong

2012, 33(3): 6-11.

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The iteration progress will not succeed in the computation of eigenvalue for the neutron transport equations,when considering more expanding number for anisotropic scattering source or more discrete direction for discrete ordinate method.In this paper,the numerical computation of eigenvalue is studied for spherical steady neutron transport equation and a new method is constructed.The new method can improve the convergence rate and nonconvergent case.The modification of computation for k_eff eigenvalue will not dependent on the iterative initial value and the numerical results are stable.

Study on MCNP Related to Temperature-Dependent Neutron Cross Section Library Based on Different ENDF Format Databases

ZOU Yang

2012, 33(3): 12-16.

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Using the cross section processing procedures NJOY,two temperature-dependent neutron cross section libraries with the ACE（A Compact ENDF） format are generated based on the latest released ENDF/B-VII and CENDL-3.1,and are verified by PWR Doppler numerical benchmark problems.The results showed that these two cross section libraries were in good agreement with the original benchmark problems in the effective multiplication factor,Doppler reactivity loss and Doppler reactivity coefficient,which indicated that the two cross section libraries ENDF/B-VII and CENDL-3.1can be used in the production of ACE format cross section libraries.

Preliminary Design of LEU MNSR for BNCT with Excellent Epithermal Neutron Flux Treatment Beam

YU Tao, QIAN Jindong, XIE Jinsen

2012, 33(3): 17-20,37.

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Based on the Miniature Neutron Source Reactor（MNSR） with high enrichment uranium（HEU） fuel and accordance with the requirements of BNCT,the 19.5% of enriched fuel UO₂ fuel core for BNCT with epithermal neutron treatment beam was primary designed,the reactor core parameters such as epithermal neutron flux density,epithermal neutron flux unit of fast neutron dose rate,epithermal neutron flux unit photon dose rate of γ,epithermal neutron flux ratio of thermal neutron flux,neutron spectrum were calculated and analyzed.The results show that the design program was an excellent epithermal neutron treatment beam.

Applications of Supercritical Carbon Dioxide in Nuclear Reactor System

HUANG Yanping, WANG Junfeng

2012, 33(3): 21-27.

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The applications of supercritical carbon dioxide Brayton cycle in nuclear reactor systems have attracted worldwide attention in recent years.In this paper,the advantages of employing supercritical carbon dioxide Brayton cycle in nuclear reactors were analyzed based on its fundamental conception.The investigations on supercritical carbon dioxide Brayton cycle were reviewed.The potential application area of supercritical carbon dioxide in Chinese advanced nuclear energy technology were analyzed and discussed,and some associated suggestions were proposed.

Numerical Simulation of Multi-Dimensional Two-Phase Transient Flow across Bundles

XU Liangwang, JIA Baoshan

2012, 33(3): 28-32.

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A multi-dimensional two-fluid model for two-phase flow across bundles is presented.The concept of porous media and distributed resistance are applied to derive the two-fluid Navier-Stokes equation of equivalent continuum,which is discretized with full implicit scheme on multi-dimensional staggered grid and solved with direct Strong Implicit Procedure（SIP）.A numerical simulation of kettle reboiler experiment is implemented for model verification.Good agreement between the numerical results and experimental data is obtained,which proves that the sugguested model is able to handle with two-phase instability numerically and suitable for the simulation of multi-dimensional two-phase transient flow across bundles.

Effect of U-Tube Length on Space Distribution of UTSG Reverse Flow in Tubes

ZHANG De, CHEN Wenzhen, WANG Shaoming

2012, 33(3): 33-37.

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For natural circulation,it is shown that parallel flow in the tubes of inverted U-tube steam generators can be non-uniform.And reverse flow occurs in some tubes.Existing studies on the space distribution of reverse flow tubes are dissident.A flow model of one-dimensional steady state is established.Analysis shows that the length will change the relationship of flow characteristics between tubes.And the space distribution of reverse flow tube differs further.When the U-tube length of UTSG is small enough,flow excursion will occur earlier in shorter tube than longer tube,and vise versa.And the analysis is validated by best estimate code RELAP5/MOD3.3.

Study on Frictional Pressure Drop of Steam-Water Two Phase Flow in Optimized Four-Head Internal-Ribbed Tube

WANG Weishu, ZHU Xiaojing, BI Qincheng, WU Gang, YU Shuiqing

2012, 33(3): 38-41.

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The optimized internal-ribbed tube is different from the normal internal-ribbed tube on the frictional pressure drop characteristics.The frictional pressure drop characteristics of steam-water two phase flow in horizontal four-head optimized internal-ribbed were studied under adiabatic condition.According to the experimental and calculation results,the two-phase multiplier is greatly affected by the steam quality and pressure.The two-phase multiplier increases with increasing quality,and decreases with increasing pressure.In the near-critical pressure region,the two-phase multiplier is close to 1.The frictional pressure drop of two phase flow in optimized tube is less than that in the normal tube under the same work condition.The good hydrodynamic condition could be achieved when the optimized internal-ribbed tube is used in the heat transfer equipment because the self-compensating characteristics exist due to the reduction of frictional pressure drop.

Experimental Investigation of Critical Heat Flux in Rectangular Channels with Different Gap Sizes

LI Yong, XIONG Wanyu, YAN Xiao, HUANG Yanping

2012, 33(3): 42-45.

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Experiments on Critical Heat Flux（CHF） of Freon-12 have been carried out in the rectangular vertical channels with gap sizes of 1 mm,2 mm,and 3 mm,respectively.Analysis of CHF data in the different channels shows that CHF decreases as the increase of pressure,and while the exit quality rises,CHF falls down quickly.Meanwhile,CHF goes up as the mass velocity increases when the exit quality is lower,and contrarily,when the exit quality is higher,the increase of mass velocity results in the decrease of CHF.According to the analysis results,a conclusion can be drawn that gap sizes almost have no effect on the CHF in rectangular channels with gaps from 1 mm to 3 mm under the same condition in the experiment.

Heat Transfer Characteristics of Forced Circulation under Inclined Conditions

DU Sijia, ZHANG Hong

2012, 33(3): 46-50.

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Experimental researches and numerical analysis on the heat transfer characteristics of forced circulation flow in a circular tube under different inclined conditions were introduced in this paper.For the single-phase flow,the experimental results show that the heat transfer around the tube was unsymmetrical under the buoyancy effect,and the heat transfer was weakened at the upside of tube to increase the wall tem-perature,while the heat transfer was enhanced at downside to decrease the wall temperature.The numerical calculation results show the same phenomena.Based on the experiment,a modified factor for the inclined heat transfer coefficient was presented in this paper,to estimate more accurately the heat transfer variation in single phase convection at inclined condition..Stresses on the bubble in two-phase flow was analyzed,to il-lustrate the why the effect of inclined condition on the heat transfer is not obvious.

Visual Experimental Study on Bubble Behavior on Inclined Downward Facing Surfaces

WEN Qinglong, CHEN Jun, LU Donghua, ZHAO Hua

2012, 33(3): 51-55.

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In this paper,in a setting of In Vessel Retention（IVR） strategy of molten core debris in AP1000 reactors,a visual experimental study of bubble behavior and two-phase flow in inclined confined spaces are performed for near-saturated demineralized water at atmospheric pressure with gap sizes of 3mm to 8mm,and inclination angles of 0o to 30o.Bubble sliding and deformation are observed and analyzed.It shows that geometry and downward facing heating surfaces may be two of the reasons leading to these important phenomena.Visual analyses of wavy phenomena between two-phase flow of vapor and liquid have also shown that lift-off of wavy interface between vapor and liquid phase may be the mechanism which triggers the occurrence of critical heat flux（CHF）.

Analysis on Characteristic of Forced Circulation Flow Rate Pulsation under Rolling Condition

WANG Chang, GAO Puzhen, TAN Sichao, HUANG Yanping

2012, 33(3): 56-60.

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The characteristics of single-phase flow rate pulsation under rolling motion condition are experimentally studied,and the main factors having effects on the flow rate pulsation are also studied based on the force analyze.The results show that the flow rate pulsation depends on the relative magnitude of drive force and additional inertia force.Larger rolling amplitude or frequency will result in more violent flow pulsation.However,the amplitude of flow rate pulsation decreases with the increasing of drive force.And the flow rate does not present significantly variation before and after rolling motion start as the ratio of drive force to additional inertial force large enough.

Numerical Analysis of Magnetically Suspended Rotor in HTR-10 Helium Circulator Being Dropped into Auxiliary Bearings

ZHAO Jingxiong, YANG Guojun, LI Yue, YU Suyuan

2012, 33(3): 61-64,88.

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Active magnetic bearings（AMB） have been selected to support the rotor of primary helium circulator in commercial 10 Mega-Walt High Temperature Gas-cooled Reactor（HTR-10）.In an AMB system,the auxiliary bearings are necessary to protect the AMB components in case of losing power.This paper performs the impact simulation of Magnetically Suspended Rotor in HTR-10 Helium Circulator being dropped into the auxiliary bearings using the finite element program ABAQUS.The dynamic response and the strain field of auxiliary bearings are analyzed.The results achieved by the numerical analysis are in agreement with the experiment results.Therefore,the feasibility of the design of auxiliary bearing and the possibility of using the AMB system in the HTR are proved.

Calculation and Comparison of Pressure-Temperature Limit Curves Calculation

LU Feng, QIAN Haiyang, WANG Rongshan, HUANG Ping

2012, 33(3): 65-68.

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In order to prevent the brittle fracture,during nuclear power plant（NPP） heatup and cooldown processes,pressure and temperature in the reactor pressure vessel should be kept under the pressure-temperature（P-T） limit curve.In this paper,the P-T limit curve methodologies of ASME code,RCCM code and Chinese Nuclear Industry Standard EJ/T 918 are studied.Calculation and comparison of the P-T curves obtained using methods from different codes are performed.It shows that using static fracture toughness KIC instead of reference fracture toughness KIR to calculate the P-T curves increases acceptable operating region during NPP heatup and cooldown processes significantly.Comparing to the latest versions of ASME and RCCM codes,the Chinese standard is more conservative.

A New Construction Algorithm of Narrow-Band Function for Artificial History Simulation Method Based on Narrow-Band Superposition

HE Jia, WANG Haitao

2012, 33(3): 69-73.

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In this paper,a new type of narrow-band function is proposed for the artificial history simulation method based on narrow-band superposition,which aims to meet the needs of both fitting of the target response spectrum and envelop of the power spectral density.The new narrow-band function is based on the normal distribution function and trigonometric functions.Its band width can be controlled and it decays rapidly on both sides.While the target response spectrum is fitted by superimposing the new narrow-band time history,the power spectral density is enveloped by modifying the Fourier amplitude spectrum locally.The numerical example demonstrates that not only the artificial time history generated by this algorithm reaches high matching precision to the target response spectrum,but also the corresponding calculated power spectrum envelopes the target power spectrum.

Probabilistic Safety Assessment to 1000 MW PWR Plant during Shutdown State

ZHAO Bo, LI Xiaoming, LI Lin

2012, 33(3): 74-78.

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This paper initiates a study focused on level 1 Probability Safety Assessment（PSA） for shutdown operation at 1000MWe PWR.Shutdown PSA models considering different plant outage types are constructed respectively to evaluate the risk of LOI-RRA operation.The application of SPAR-H method is also evaluated.The results show that the potential risk during shutdown should not be ignored;PSA for shutdown states could help to prompt the improvement in procedure to reduce the risk in shutdown condition.Going through the LOI-RRA operation at cool shutdown refueling outage significantly increases the plant risk;the dominant factor contributing to CDF is human error.

Design and Application of New Emergency Operation Procedure I RCP 10

SUN Tao, YI Ke

2012, 33(3): 79-82.

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The draft of event orientation emergency operation procedure（EOP） is based on the research of accident evolution which has been defined in advance.With the feedback of nuclear events and accidents,and better understanding of nuclear safety,events and accidents which are not considered will be added in EOP.The design and application of new procedure High Activity of Primary Loop（I RCP 10） is based on this kind feedback,and the requirement of the nuclear authority.The purpose of this procedure is to control the nuclear power plant reasonably and effectively in loss of integrity of the fuel elements,and to ensure that the radioactivity will not cause any damage to operators and later waste processing.

Study on Event Sequence of Loss-of-Coolant Accident of Primary Systems for MNPP Based on ESD

CAI Qi, XIE Haiyan, ZHANG Yangwei

2012, 33(3): 83-88.

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According to the characteristics of accident analysis of nuclear power plant,the combination method of event sequence diagram（ESD） and operational safety analysis is used to build the ESD model for loss-of-coolant accident of marine reactor primary system from the operational safety view.The accident evolution and results are studied and the entire event sequences are required.

Research on Loss of Coolant Accident of Pressurized-Water Reactor Based on PSO Algorithm

MA Jie, GUO Lifeng, PENG Qiao

2012, 33(3): 89-91,96.

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In order to improve the diagnosis performance of Loss of Coolant Accident（LOCA）,based on Back Propagation（BP） algorithm study,a fault diagnosis network is established based on Particle Swarm Optimization（PSO） algorithm in this paper.The PSO algorithm is used to train the weights and the thresholds of neural network,which can conquer part convergence problem of BP algorithm.The test results show that the diagnosis network has higher accuracy of LOCA.

Examination and Adjustment of Piping Hangers before Commercial Operation in NPPs

ZHANG Xingzhong, DING Youyuan, XU Weizu, ZHU Xiaoyong, ZHOU Sheng

2012, 33(3): 92-96.

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This paper describes the basis,content and method of the cold and hot examination for the piping hangers during the plant commissioning,and explains the maintenance and regulation strategy of those abnormal hangers after inspection,and also expatiates the verification ways of piping hangers during hot condition before the plant is critical.Besides,this paper puts forward the notice items of piping hangers while installation,reconstruction and in-service inspection,and tries to make sure that the piping hangers are under normal working condition by the means of the inspection and regulation to the piping hangers,so that the system pipes can work safely and reliably.

Reliability and Maintainability of Nuclear Power Plants

CHEN Tongbiao, FU Xiaobo, LU Songze

2012, 33(3): 97-99,108.

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Based on the safety requirement on nuclear power plants,the characteristics of reliability and maintainability are analyzed.The safety issues are explored from design,operation,management,supervision,and emergency.

Comparison of Methodologies for Qualification of In-Service Inspection

QI Dunjie, GUO Liang

2012, 33(3): 100-103.

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The features and problems with ASME in-service inspection qualification requirements and the ENIQ methodology approach were discussed.The approach in ASME XI,Appendix VIII is generic qualification,which focuses on blind trial and the statistical significance of the performance demonstration.The benefits of the ENIQ methodology are its flexibility and the requirement that qualification is a combination of practical trials and technical justification.The major disadvantage of the ENIQ approach is the requirement for scarce personnel skills in the physics and practice of inspection.Then the qualification approach of in-service inspection in China is discussed based on the experience of Spanish methodology approach.

Fault Diagnosis of Nuclear-Powered Equipment Based on HMM＆SVM

YUE Xia, ZHANG Chunliang, QUAN Yanming, ZHU Houyao

2012, 33(3): 104-108.

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For the complexity and the small fault samples of nuclear-powered equipment,a hybrid HMM/SVM method was introduced in fault diagnosis.The hybrid method has two steps: first,HMM is utilized for primary diagnosis,in which the range of possible failure is reduced and the state trends can be observed;then faults can be recognized taking the advantage of the generalization ability of SVM.Experiments on the main pump failure simulator show that the HMM/SVM system has a high recognition rate and can be used in the fault diagnosis of nuclear-powered equipment.

A WAV-ICA Based On-Line Fault Diagnosis Method for Redundant Sensors in Nuclear Power Plants

YU Ren, CHEN Zhi, ZHANG Zi, LIU Lianghui

2012, 33(3): 109-114,120.

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A WAV-ICA based on-line fault diagnosis method for redundant sensors in nuclear power plants is proposed in the paper.In order to eliminate the effect of the electric noise and transmission noise,the high-frequency noise in redundant sensors signal is filtered with wavelet decomposition method at first.Then the Independent Component Analysis（ICA） is performed and the component with interest is selected for reconstructing the estimated signal,so as to detect the fixed bias and abrupt faults of the sensors,as well as their drift failures.A history data set from five redundant pressurizer pressure sensors in a NPP is adopted to validate the proposed method.The result shows that compared with the Simple Analysis（SA） method and direct ICA method,the proposed method can detect the fixed bias and drift faults of the sensors on-line effectively,with lower rate of misdiagnosis and better robustness.

Simulation Analysis of Start-up and Shutdown of HTR-PM Based on THERMIX/BLAST and vPower

GAO Qiang, ZHOU Zhiwei, ZHOU Yangping, SUI Zhe, MA Yuanle

2012, 33(3): 115-120.

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An engineering simulator of HTR-PM is developed by embedding THERMIX code into the vPower simulation platform.The double reactor cold start-up and shutdown processes are simulated on the engineering simulator,to analyze the changing of the critical data such as the reactor power,helium mass flow,inlet/outlet parameters of steam generator and inlet vapor parameters of turbine in these two conditions.The operation features of the double reactor cold start-up and shutdown processes are summarized.The results show that the two reactors influence each other during the operation and the parameter changes in the secondary loop are the combined results from two reactors.

Application of Reverse Osmosis in Radioactive Wastewater Treatment

KONG Jingsong, GUO Weiqun

2012, 33(3): 121-124.

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Considering the disadvantages of the conventional evaporation and ion exchange process for radioactive wastewater treatment,the reverse osmosis is used to treat the low level radioactive wastewater.The paper summarizes the research and application progress of the reverse osmosis in the radioactive wastewater treatment and indicates that the reverse osmosis in the radioactive wastewater treatment is very important.

Application Analysis of High Integrity Container on Domestic Radioactive Waste Management

PEI Yong, PAN Yuelong

2012, 33(3): 125-128.

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This paper simply described three kinds of material high integrity containers,and accordingly emphasized the cross linked polyethene HIC used in the domestic projects under construction,focusing on the waste treatment proposal coupling with HIC model and the advantages and disadvantages comparing with the cement solidification proposal.Many aspects are analyzed including waste filling and HIC lifting,transportation,and final disposal.The potential solutions are pointed out for the issues and the post actions as well.

Research on Safety of Reverse Osmosis to Treat Radioactive Wastewater

KONG Jingsong, TIAN Yanjie

2012, 33(3): 129-131.

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The security of reverse osmosis combined with a pretreatment process of disc filtration-ultrafiltration to treat the radioactive wastewater was analyzed and evaluated.Several aspects including reliability and security during operation,maintenance and decommissioning were investigated in this paper.Results showed that safe operation can be ensured by rational process design and scientific management.Estimation on radiation safety showed that the absorbed dose rate is below 0.04 mSv/h on the surface of reverse osmosis module,which can ensure the radiation safety of operators.

Study on Source Term Evaluation Method in Simulation and Scaling for Containment during LOCA

LI Shengqiang, LI Weihua, JIANG Shengyao

2012, 33(3): 132-137.

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Source term is one of the most important parameters involved in the simulation of LOCA inside containment.It may affect facilities scale,simulation stage,equipment design level and simulating authentic.Based on the scaling rules related to the total break enthalpy,force jet and buoyant plume and heat and mass transfer process,the time scale of heat/mass transfer and flow dissipation are normalized to the time scale of natural circulation.It proves that the natural circulation process time ratio is the dominant parameter for whole system design.Functions which can be applied for geometry sizes and boundary conditions are also provided.

Study on Applications of Moment Analysis in Chromatography for Boron Isotopes Separation

WANG Guanchun, ZOU Congpei, WANG Meiling, FU Daogui, LIU Xiaozhen, JIAN Min, LI Gang

2012, 33(3): 138-142.

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Chromatographic characteristics for both boron isotopes,using D301-G resin as the stationary phase and water as the mobile phase,were investigated by moment analysis in the given conditions,and the kinetic parameters of boric acid on the stationary phase were evaluated by moment analysis based on equilibrium-diffuse and linear driving force models.The results show that the retention time of ¹⁰B isotope was lager than ¹¹B,and both the axial dispersion coefficient and the overall mass transfer coefficient of boric acid on the D301-G resin phase increase with increasing temperature.The axial dispersion coefficient was evaluated to be 1.08 cm²/min and the overall mass transfer coefficient was 0.65 min-1 at room temperature..

2012 Vol. 33, No. 3