核动力工程

Benchmark Validation of Transport Calculation for Advanced Neutronics Lattice Code KYLIN-2

Huang Shien, Chai Xiaoming, Li Xiangyang, Chen Zhang, Yin Qiang, Lu Wei, Li Qing

2018, 39(2): 1-4. doi: 10.13832/j.jnpe.2018.02.0001

Abstract(20)

Abstract:
Several benchmark problems were validated for the transport calculation module of the advanced neutronics lattice code KYLIN-2,including homogeneous medium problem,Postma problem,BWR cell problem and C5G7 assembly problem.The validation results show that:（1） the ray spacing has limited effect on the calculation results,（2） the more the numbers of azimuth angle,the higher accuracy of the calculation,and the recommended azimuth angle number is greater than 6,（3） with the appropriate choice of calculation parameters,the transport module of the code has a high calculation accuracy.

Experimental Research on Effect of Nuclear Feedback on Flow Instability in Parallel Channels

Xie Feng, Xi Zhao, Yang Zumao

2018, 39(2): 5-9. doi: 10.13832/j.jnpe.2018.02.0005

Abstract(18)

Abstract:
As the existence of nuclear feedback in nuclear reactors,coupled nuclear-thermo-hydraulic instability will occur in the reactor.Coupled-nuclear-thermal-hydraulic instabilities were investigated employing experimental and theoretical methods in this paper.The experimental research on coupled nuclear-thermo-hydraulic instability in parallel channels was performed and the effect of the nuclear feedback on flow instability has been studied.The experiment is carried out with and without nuclear coupled respectively,and the effect of void reactivity coefficient and temperature reactivity coefficient on characteristic and boundaries of flow instability has been studied.Based on the experimental data,the tendency of the characteristics of effect is analyzed,and the stability boundaries is obtained.The results from the research in this paper show that:void reactivity coefficient,temperature reactivity coefficient and fuel time constant have a significant effect on characteristics and boundaries of flow instability.

Research on Simulation Criterion of Steam Generator Natrual Circulation Experiment

Li Haibo, Zhao Erlei, Zan Yuanfeng, Zhuo Wenbin

2018, 39(2): 10-13. doi: 10.13832/j.jnpe.2018.00.0010

Abstract(23)

Abstract:
The research of the simulation criterion about natural circulation was introduced based on the hierarchical two-tired scaling（H2 TS） methodology.According the physical process of experiment target,the phenomena identification and ranking tables（PIRT） was established.Steam generator（SG） natural circulation equation was deduced,dimensionless similarity critera was obtained.The similarity criteria was analyzed and simplified based on experiment test.

Analyses on Flow Instability of Secondary-Side Passive Residual Heat Removal System for Advanced Nuclear Heating Reactor

Liang Weihong, Xie Heng

2018, 39(2): 14-19. doi: 10.13832/j.jnpe.2018.02.0014

Abstract(32)

Abstract:
An advanced nuclear heating reactor model is provided to study the flow instability of secondary-side passive residual heat removal system.The RELAP5 program is used to simulate the transient process of PRHRS after accident.The results show that the origin for the condensate flow instability is the mismatch of vapor pressure and two phase length in the condensate tube of PRHRS.The condensate flow stability boundary calculated numerically by RELAP5 is satisfied with Bhatt’s theoretical formula analysis results.By increasing the condensate tube inlet throttling,flow area and heat transfer area,the condensate flow instability can be avoided.

Study on Transient Flow Characteristics of Rotor-Stator Cascade in Nuclear Main Pump

Li Yibin, Zhang Mei, Zhu Yuelong, Li Zhenggui

2018, 39(2): 20-26. doi: 10.13832/j.jnpe.2018.02.0020

Abstract(36)

Abstract:
Based on DES vortex model and block structured grid technology,the adaptability and accuracy of DES model for the CAP1400 reactor coolant pump prototype and 0.4 scale model is verified through CFD numerical analysis and experimental investigation of exterior performances.Compared with the loading distribution on surface of the model pump rotor-stator cascade under different conditions,it is found that the blade load is very uneven,the blade load of impeller front cover is about 2 times than that of the rear; with the change of working condition,the difference of load guide blades is complex,and the load of the entrance section is relatively large under a large flow,even when changing the direction.Based on the turbulent vortex dynamics,the vortex motion and the evolution process of the rotor-stator cascade are clarified.The structure and evolution mechanism of the transient flow field of impeller and guide vane under the condition of noise disturbance are revealed.

Effect of Fission Gas Bubble Size on Internal Characteristics of Dispersion Fuel Particles

Chen Hongsheng, Long Chongsheng, Xiao Hongxing, Wei Tianguo, Gao Wen, Zhao Yi

2018, 39(2): 27-31. doi: 10.13832/j.jnpe.2018.02.0027

Abstract(27)

Abstract:
On the basis of the fuel particles cracking model of bubble size homogeneity,the cracking model with bubble size heterogeneity was proposed according to stress equilibrium condition that the equivalent effective stresses applied by the gas bubble are identical.The effect of the fission gas bubble size on the internal characteristics,such as equivalent fuel thickness,gas atom number,bubble pressure and maximal tensile stress,was analyzed through this proposed model.The results show that the equivalent thickness is linearly dependent with the bubble radius when the bubble radius is larger,while the ration of equivalent thickness to bubble radius increases rapidly with bubble radius decreasing when bubble radius is smaller.The gas atom concentration increases as the bubble radius decreases.The increasing rate of maximal tensile stress at bubble wall is larger than that of cracking resistance during heating process,and the cracking temperature of fuel particle decreases with bubble radius.

Effect of Re-Working on Secondary Phases in Modified N18 Sheet

Chen Boquan, Qiu Shaoyu, Peng Qian, Dai Xun, Wang Pengfei, Liu Hong, Wei Tianguo

2018, 39(2): 32-36. doi: 10.13832/j.jnpe.2018.02.0032

Abstract(24)

Abstract:
A combination of various research methods including DSC,SEM and TEM was applied to systematically investigate the effect of re-working on the secondary phases.The results showed that the hot-rolling was conducted in the dual-phase region.In the process of cold-working and followed annealing,β-Zr which formed in the hot-rolling decomposed,leading to the precipitation of fine particles,which resulted in the clusters of tiny compounds.The secondary phases are Zr（Fe,Cr,Nb）₂ with a crystal structure of hexagonal close-packed（HCP）.The atomic percentage ratio of alloying elements,n（Fe）/[n（Cr）+n(Nb],for particles in original sheets is close to 5/3,while it was experiencing a downward trend after the re-working.The particles in the cluster own a higher Nb content with the n（Fe）/[n（Cr）+ n（Nb）] approximate to 1.

Study on Optimization Design of Evaporator System for Radioactive Waste Disposal Center

Li Ming, Ma Xingjun, Chen Li, Gao Feng, Li Binglin, Sha Sha

2018, 39(2): 37-41. doi: 10.13832/j.jnpe.2018.02.0037

Abstract(34)

Abstract:
In this paper the evaporator system of the radioactive waste disposal center was introduced in brief firstly.Considering the problems of the traditional process such as it is unable to meet the handling capacity for the radioactive wastewater of the newly connected nuclear facilities,the design for the key equipments of the evaporation process and the system in the aspect of secondary steam demisting is optimized.The mist entrainment of the evaporator primary gravity separation segment and the built-in bubble column is reduced,the purified water in the system is used to wash the secondary steam,and the mesh demister is combined to further separate the entrainment.The purifier with high efficiency coalescing filter element is used to purify in high precision.The proposed system can improve the purification and the decontamination factor of radioactive wastewater based on the radioactive waste minimization principle by verifying,and the whole evaporated system and equipments was simplified.

Research and Development of RPV Stud Holes Lubricating Equipment

Ren He, Weng Songfeng, Tan Hongwei, Huang Hui, Huang Xindong

2018, 39(2): 42-45. doi: 10.13832/j.jnpe.2018.02.0042

Abstract(24)

Abstract:
Aiming at the existing problem in the lubrication of reactor pressure vessel（RPV）stud holes,this paper puts forward an effective method which adopts array bristles and small caliber oil outlet to brush anti-seize lubricant on the surface of stud holes,and gives out the detailed design of RPV stud holes lubricating equipment.The equipment has been developed and verified by testing,and the results show that it can accomplish the uniform and complete coating on the surface of stud holes.The equipment can effectively increase the lubrication of thread pairs,and reduce the incidence and risk of thread damages.

Design of a Model Predictive Controller for Load Tracking in Pressurized Water Reactor

Wang Guoxu, Wu Jie, Chen Zhijie, Zeng Bifan, Xu Zhibin, Ma Xiaoqian

2018, 39(2): 46-49. doi: 10.13832/j.jnpe.2018.02.0046

Abstract(27)

Abstract:
The state space model of the reactor core is modeled by expanding the differential equation model with Taylor series,and a model predictive controller is designed to achieve the automatic control of load tracking in pressurized water reactor（PWR） nuclear power plants.Quadratic programming（QP） is introduced for the optimal solutions of the model predictive control（MPC） system.To evaluate the performance of the designed QP-based MPC（QPMPC） controller for load tracking,simulations are designed,and the simulation results show that the proposed QPMPC controller can track the load changing swiftly and availably.

Comparison of Methods for Artificial Earthquake Time History Matching Floor Response Spectra in Nuclear Power Plants

Sun Yugang, Chu Meng, Ding Zhenkun, Ge Honghui

2018, 39(2): 50-54. doi: 10.13832/j.jnpe.2018.02.0050

Abstract(27)

Abstract:
In order to satisfy the requirements of seismic analysis for the structure,system and components（SSCs） in Nuclear Power Plants（NPPs）,two typical methods（modify the Fourier amplitude spectra in frequency domain and superimposing wavelet in time domain） were used to generate the artificial time history matching the Floor Response Spectra in NPPs.Numerical examples demonstrate that the time domain method is better that the frequency domain method in the matching precisions to the target spectra（especially the multi-damping spectra）,preserving the nonstationary character and baseline drift correction.However,the frequency domain method can maintain the phasing of Fourier components of the initial seed time history,which meets the requirements of the nonlinear structural analysis problems in SRP 3.7.1.

Research on Criteria of Design by Analysis of NB-3200 in ASME B＆PV Code Section Ⅲ

Gao YongJian, He YinBiao, Cao Ming, Yao Weida

2018, 39(2): 55-61. doi: 10.13832/j.jnpe.2018.02.0055

Abstract(34)

Abstract:
The ASME B＆PV Code Section Ⅲ– Division 1 – Subsection NB Design-by-Analysis（NB-3200） approach provides stress limits for the design of nuclear components.In the present paper,the failure modes in nuclear components are identified.The characteristic of Design-by-Analysis was presented.Moreover,the technical basis and background on some criteria of NB-3200 is presented.Specifically,shakedown and ratcheting（NB-3222.2）,thermal stress ratcheting（NB-3222.5）,simplified elastic-plastic analysis（NB-3228.3）,Poisson’s ratio correction（NB-3227.6） and fatigue（NB-3222.4）.The author hopes that the above discussion will deepen the analyst’s understanding of the code and help them have better practical application of the code.

Software Verification and Validation of Safety DCS System

Zhang Lei, Zhou Liang, Sun Yongbin, Liang Zhongqi

2018, 39(2): 62-65. doi: 10.13832/j.jnpe.2018.01.0062

Abstract(22)

Abstract:
According to the nuclear safety regulations,verification and validation（V＆V）activities should be carried out during the software development process.Based on the research of the regulations and standards and engineering experience,a suitable safety class DCS application software V＆V system is established.This V＆V system has been used in a nuclear power plant project,to ensure the quality of the software and the successful implementation of the project.The system also has good prospects in the follow-up projects,such as HRR1000 and AP1000 advanced reactors.

Preliminary Shielding Calculation and Analysis of OTSG for ACPR50S Small Modular Reactor

Sheng Xuanyu, Ling Chen

2018, 39(2): 66-71. doi: 10.13832/j.jnpe.2018.01.0066

Abstract(24)

Abstract:
Once-through steam generator（OTSG） is one of the main equipments of ACPR50 S small modular reactor.High flux of neutrons in OTSG will cause steam activation,increase the dose rate at main steam outlet,thus polluting entire secondary loop of the reactor.We carried out a preliminary analysis of ¹⁶N generation and transport based on physical calculation of current Shielding design of OTSG.The results show that the neutron flux magnitude in the OTSG has been reduced to a low level.Activity of ¹⁶N in the OTSG outlet is approximately tens of Bq/cm³.The results indicate that the material and thickness of current shielding structure of OTSG is satisfactory to meet the preliminary design goal under normal operating conditions.

An Investigation of Trend Filter Based Collaborative Piecewise Linearization Method for Transient Curve

Bai Xiaoming, Zheng Liangang, Ai Honglei, Wang Xinju, Lu Xifeng, Zhang Rui

2018, 39(2): 72-75. doi: 10.13832/j.jnpe.2018.02.0072

Abstract(21)

Abstract:
In nuclear power plants,both the calculated design transient and monitored real transient are constructed by massive data.Due to the high computational cost,the raw transient data can not be used for fatigue analysis directly.Therefore,how to piecewise linearize the raw data is a key issue for fatigue analysis.A trend filter based collaborative piecewise linearization method for transient curve is proposed in the present work.An assumed reference transient and a real monitored transient are used to verify the present method.The results show that the present method can linearize the real transient curve accurately and efficiently.

Numerical Simulation Research on Equivalent Thermal Conductivity Coefficient of M3 Fuel

Li Yuanming, Tang Changbing, Yu Hongxing, Chen Ping, Zhou Yi

2018, 39(2): 76-79. doi: 10.13832/j.jnpe.2018.02.0076

Abstract(21)

Abstract:
In order to research the equivalent thermal conductivity coefficient of Zirconium based dispersion micro encapsulated fuel（M3 fuel）,in this research,assuming the TRISO（three layer isotropic） particle embedded in the zirconium with body-centered cubic distribution,the simulation method for the equivalent thermal conductivity of M3 fuel is established with the help of ABAQUS software,based on the homogenization theory.The equivalent thermal conductivity coefficients of different phase volume M3 fuel were analyzed with this simulation method.Simulation results show that the equivalent thermal conductivity coefficient increases with the increasing of temperature,and the equivalent thermal conductivity coefficient decreases with the increasing of burnup and phase volume.

Analysis of Radioactive Effect in Molten Salt Reactor

Wang Dan, Wang Yahui, Ma Yu

2018, 39(2): 80-85. doi: 10.13832/j.jnpe.2018.02.0080

Abstract(21)

Abstract:
Due to the flow characteristics of molten salt in molten salt reactor（MSR）,the internal physical processes of this reactor show strongly coupling characteristic.This paper studied the coupled neutronics-thermo-hydrodynamic and radiative heat transfer processes in MSR by using finite element method and discrete-ordinate method.The influences of radiation effect on internal temperature and flow fields are analyzed particularly.Simulation results show that although the radiation process has little effect on the flow process,the influence of radiation process on overall temperature can be very obvious.Therefore,in the design and the safety analysis of MSR,the influence of radiation effect cannot be neglected.

Study on Three-Dimensional Neutron Diffusion Heterogeneous Variational Nodal Method and Flat Source Acceleration Method

Zhang Tengfei, Wu Hongchun, Cao Liangzhi, Li Yunzhao, Xiong Jinbiao, Liu Xiaojing

2018, 39(2): 86-89. doi: 10.13832/j.jnpe.2018.02.0086

Abstract(29)

Abstract:
As a foundation of three-dimensional whole-core heterogeneous transport calculation,a 3D heterogeneous variational nodal method is proposed,which employs iso-parametric finite elements within each node and piecewise constants along nodal interfaces to explicitly describe the pin-resolved heterogeneous geometry.This means full elimination of the homogenization approximation.Meanwhile,the flat source acceleration method is proposed accordingly to reduce the computational costs.Numerical results show that the method is accurate and the flat source region acceleration method can effectively decrease the memory storage and computational time without degrading the accuracy.

Numerical Investigation of Effect of Heat Transfer Model on Hydrogen Flow in Local Compartment

Wang Di, Tong Lili, Cao Xuewu, Zou Zhiqiang, Chen Shu

2018, 39(2): 90-95. doi: 10.13832/j.jnpe.2018.02.0090

Abstract(29)

Abstract:
During the severe accident process,the hydrogen and the steam are released into the containment and transferred to other local compartments through narrow pipes.Heat transfer between gas mixture and wall has influence on hydrogen distribution.In this paper,CFD method is adopted to build the analysis model for local compartment connected by vertical flow path in the containment.The effect of Von Karman model,Reynolds model,and directly similar v-t profile model based on analogy theory on pressure,condensation heat,hydrogen and steam distribution are discussed.The result shows that the steam condensation mass and the hydrogen flow behavior simulated by Von Karman model and Reynolds model is consistent.The condensation mass obtained by directly similar v-t profile model is higher.Most steam condensed on the top wall of the source compartment,on this wall,directly similar v-t profile model predicts more condensation than other two models,which results in lower gas temperature.Hydrogen transferring into the upper compartment driven by buoyancy is also weaker.The hydrogen concentration in this compartment is lower.

Inverse Problem of One-Dimensional Unsteady Heat Conduction Deducing Temperature Fluctuation of Inner Wall of Pipe

Xiong Ping, Ai Honglei, Lu Tao, Wang Xinjun

2018, 39(2): 96-100. doi: 10.13832/j.jnpe.2018.00.0096

Abstract(21)

Abstract:
A one-dimensional transient thermal conduction inverse problem mathematical model is constructed based on the finite difference method（FDM） and the optimization algorithm based on the conjugate gradient method（CGM）.The general calculation program is written in C language,and the temperature fluctuation value of the outer wall surface obtained by the direct heat conduction problem is used as the known condition of the inverse heat conduction problem.And the random measurement error is introduced to investigate the effect of measurement error on the accuracy of the inversion result.The inversion value is compared with the theoretical value of the inner wall surface as the boundary condition.The comparison results show that the inversion value of the internal wall surface is in good agreement with the theoretical value,indicating that the transient thermal conductivity anti-problem model can well inverse the temperature fluctuation value of the inner wall surface.

Research on Rod Withdrawal Accident Analysis Method Based on RELAP5/MOD3 Variable Power Distribution

Zhang Yong, Li Songwei, Wang Wei

2018, 39(2): 101-103. doi: 10.13832/j.jnpe.2018.02.0101

Abstract(25)

Abstract:
In this paper,we use the rod-power distribution obtained by the nuclear design and the 202 tables and 205 cards of RELAP5/MOD3 to realize the change of the relative power share of each axial node in the heat of the RELAP5/MOD3.The "variable power distribution" analysis method is successfully realized in RELAP5/MOD3,and the method is used to analyze the rod withdrawal accident of a reactor.The maximum fuel temperature obtained by this method is significantly lower than the fixed power distribution method.

Effect of Thermal Aging on Impact Properties in 16MND5 Forgings for Nuclear Reactor Pressure Vessel

Xing Ruisi, Chen Xu, Xie Guofu, Yang Zhihai

2018, 39(2): 104-108. doi: 10.13832/j.jnpe.2018.02.0104

Abstract(31)

Abstract:
Effects of thermal aging on Charpy V-notch impact properties in 16MND5 steel was investigated.The ductile to brittle transition temperature increased with the prolongation of thermal aging duration under impact test.According to the fracture analysis of the Charpy impact specimens with different thermal aging time by SEM,the typical ductile fracture transforms to the cleavage fracture with the extension of thermal aging time at cryogenic temperature.The thermal aging had unfavorable influence on the strength and impact toughness,which should be taken into consideration seriously

Research on In-Pile Thermal-Mechanical Behavior of UMo/Zr Monolithic Fuel Plates

Kong Xiangzhe, Ding Shurong, Tian Xu

2018, 39(2): 109-113. doi: 10.13832/j.jnpe.2018.02.0109

Abstract(33)

Abstract:
UMo/Zr monolithic fuel plates experience complex multi-field-coupling and multi-scale behavior during the in-pile radiation.In this study,the method of multi-scale simulation of the in-pile behavior in the UMo/Zr monolithic fuel plate under a homogeneous irradiation condition is built.The distributions and evolutions of the temperature field,the strain field and stress field,and the interfacial stress between the cladding and the fuel meat are calculated and analyzed.The research results indicate that the peak temperature increases with the irradiation time,the thickness of the plate increases with the irradiation time,and the peak value occurs near the edge of the fuel meat; the Mises stress in the fuel meat is far less than that in the cladding; and the maximum interfacial normal stress is located near the corner of the fuel meat,where the tensile stress may cause interfacial failure.

Numerical Simulation of Effect of Structural Parameters of Air Relief Valve on Actuation Time for Valve

Zhang Zhenhua, Yu Deyong, Jia Li, Yang Lixin

2018, 39(2): 114-119. doi: 10.13832/j.jnpe.2018.02.0114

Abstract(36)

Abstract:
Aiming at the pre-isolation valve system of the atmospheric relief valve in a valve system of Tianwan Nuclear Power Plant,the numerical simulation technology is applied to carry out the numerical simulation of valve opening characteristics.The effect of different structural parameters and different working conditions on the actuation time of the valve was analyzed by CFD method and a theoretical analysis was given for the phenomenon of valve opening delay in the test.With these results the operation characteristics of the air relief valve can be assessed to optimize the valve design.

Investigation on Visualization Test Methods for Flow Field Downstream Spacer Grids

Qu Wenhai, Chen Shilong, Xiong Jinbiao, Cheng Xu, Wang Xiaoyu, Du Sijia

2018, 39(2): 120-123. doi: 10.13832/j.jnpe.2018.02.0120

Abstract(26)

Abstract:
When visualization measurement techniques,such as particle image velocimetry（PIV） and laser Doppler velocimetry（LDV）,are applied in rod bundle experiments,measurement locations are restricted due to blockage and refraction by solid rods.Employing matched index of refraction（MIR） techniques for the high Reynolds number flow in fuel assembly,the MIR method of FEP and water and the MIR method PMMA and sodium iodide solution are chosen and tested.The results show that the flow field behind rods can be achieved by the two MIR methods.

Research of Global-Local Resonance Self-Shielding Calculation Method Based on NECP-X

He Qingming, Cao Liangzhi, Liu Zhouyu, Zu Tiejun, Wu Hongchun

2018, 39(2): 124-128. doi: 10.13832/j.jnpe.2018.02.0124

Abstract(30)

Abstract:
To meet the challenges faced in high-fidelity resonance self-shielding calculations,the global-local self-shielding calculation method is proposed.The resonance self-shielding and correlated effects are classified into global and local effects.The global effects are weak or independent of energy,while the local effects are strong.Therefore,the resonance self-shielding calculation is split into global,coupling and local calculations.Coarse model is built for global calculation and the neutron current method is employed to compute Dancoff correction factors,where the global effects are considered.The coupling calculation is based on preservation of the Dancoff correction factors and the equivalent 1-D models of the fuel rods are obtained.The pseudo-resonant-nuclide subgroup method is employed to perform the local calculation to treat the local effects.This method is realized in NECP-X.The numerical results show that the efficiency of this method is increased by one order of magnitude compared with the conventional method.Besides,the precision of infinite medium multiplication factor is increased by 100~300 pcm.

RMC Solutions to Kinetic Cases of C5G7-TD Benchmark

Guo Xiaoyu, Shang Xiaotong, Qiu Yishu, Cheng Quan, Song Jing, Huang Shanfang, Wang Kan

2018, 39(2): 129-132. doi: 10.13832/j.jnpe.2018.02.0129

Abstract(28)

Abstract:
The Reactor Monte Carlo code RMC,developed by REAL group,at Tsinghua University,has a function of kinetic calculation.RMC has participated in C5G7-TD transient benchmark calculation to further optimize its kinetic algorithm,and also to validate its results.The C5G7-TD benchmark consists of 6 sets of cases,28 subcases in total,where the movement or moderator density variation of the control rods is adopted in these cases to introduce the reactivity change.To decrease the oscillation in Monte Carlo calculation,1 million particles was used in the calculation,and non-uniform time steps were applied to reduce the calculation cost.Except for those cases with different specifications,the power profiles calculated by RMC are in good agreement with those of n TRACER.

Research on Preventing Xenon Oscillation in Monte Carlo Burnup Calculation Based on RMC

Li Wanlin, Yu Ganglin, Wang Kan, Li Yaodong

2018, 39(2): 133-136. doi: 10.13832/j.jnpe.2018.02.0133

Abstract(35)

Abstract:
Monte Carlo burnup calculation suffers from numerical xenon oscillation when its model is huge geometry,refers as weakly coupled neutron transport system.Numerical xenon oscillation brings in error,even aborts burnup calculation.How to effectively prevent such oscillation is a significant topic in Monte Carlo burnup calculation.Forced equilibrium xenon method is effective when the power density used in each burnup step of the calculation keeps constant.This method has been researched and used in RMC,furthermore,some improvement has been complemented to extend range of burnup steps.Some popular international benchmarks,such as BEAVRS,VERA,require simulating burnup problem with variable power history.However,forced equilibrium xenon method inevitably brings error when power density changes by relative small burnup step.To research and develop advanced Monte Carlo simulating ability including high performance of burnup calculation,general equilibrium xenon method based on forced equilibrium method is researched and integrated to RMC.Numerical calculation result verifies that the general equilibrium xenon method is capable of effectively preventing xenon oscillation in Monte Carlo burnup calculation with constant or variable power density in each step.

Development and Verification of Reactor Core Transient Coupling Simulation Software CTSS

Pan Junjie, An Ping, Wang Wei, Zhao Wenbo, Xing Suo, Lu Wei, Chai Xiaoming

2018, 39(2): 137-141. doi: 10.13832/j.jnpe.2018.02.0137

Abstract(41)

Abstract:
Reactor Core Transient Simulation Software（CTSS V1.0）,which is coupled by 3 D space-time kinetic program NACK V1.0,CORe Thermal-Hydraulic analysis program（CORTH V2.0） and Fuel Rod Performance Analysis Code（FUPAC V1.1）,is used to simulate the typical PWR core.NACK V1.0 calculates the core diffusion equations with coarse node method and provides core power distribution for sub-channel model and fuel analysis model.CORTH V2.0 describes the core as a series of connected or unconnected sunchannel to calculate the coolant temperature.FUPAC V1.1 is used to simulate the thermodynamic behavior of fuel rod and calculate the fuel rod temperature.The calculation of PWR benchmark problem NEACRP-L-335 shows that CTSS is in good agreement with benchmark program PARCS.

Study on Semi-Implicit Scheme for Two-Fluid Seven-Equation Two-Pressure Model

Chao Fei, Shan Jianqiang, Zhang Yong, Wu Pan, Gou Junli, Li Jian

2018, 39(2): 142-148. doi: 10.13832/j.jnpe.2018.01.0142

Abstract(28)

Abstract:
Current main reactor thermal-hydraulics system analysis codes are developed based on two-fluid six-equation single pressure model.However,such six-equation model has been proved to be ill posed,which could lead to numerical oscillations.Since two-fluid seven-equation two-pressure model is unconditionally well-posed in all situations,a semi-implicit algorithm based on the finite volume method with staggered grids is developed to solve such well-posed two-pressure model.Then the algorithm is validated on two-phase sedimentation test,water faucet test,and Edwards blowdown experiment.The calculation results show that the proposed numerical algorithm is accurate and robust in solving two-phase flows in the field of nuclear engineering applications.

Investigation of Flow Rate Measurement Method in Gas-Liquid Two-Phase Flow with High Gas Volume Fraction

Ma Yugao, Li Chao, Huang Shanfang, Yu Hongxing, Pan Yanzhi

2018, 39(2): 149-152. doi: 10.13832/j.jnpe.2018.02.149

Abstract(20)

Abstract:
In this paper,a slip ratio based calculation model was developed to predict the gas mass flow rate under high gas volume fraction conditions,which was verified against an experimental investigation with a Venturi meter and a gamma ray attenuation system.Experiments showed that the new model could accurately predict the gas mass flow rate with the relative errors less than ±10% when the gas volume fraction ranging from 92% to 100%.

Uncertainty Analysis for Parameters of CFAST Based on Electrical Cabinet Fire Scenario in Main Control Room

Wang Wanhong, Zhu Dahuan, Peng Changhong, Guo Yun

2018, 39(2): 153-156. doi: 10.13832/j.jnpe.2018.02.0153

Abstract(20)

Abstract:
Based on the electrical cabinet fire scenario in the main control room,The paper use software,CFAST,by coupling with Monte Carlo,to analyze the uncertainty of heat release rate and soot yield.Distributions of layer temperature and optical density is used to assess the evacuation time and its probability,providing data base for quantitative analysis in fire probabilistic safety assessment.

Study on Coupling of Integral Severe Accident Analysis Code with System Code

Wu Xiaoli, Li Wei, Deng Jian, Deng Chunrui, Zhang Ming, Guo Chao, Yuan Hongsheng

2018, 39(2): 157-161. doi: 10.13832/j.jnpe.2018.02.0157

Abstract(25)

Abstract:
The integral severe accident analysis code and RELAP5 were directly coupled for improving the overall safety analysis of the containment and primary system.Two kinds of coupling were implemented,i.e.breaks on the pressure boundary between the primary system and containment,interfaces between the safety injection systems and primary system.The feasibility of this coupling methodology and the applicability of the coupling code were verified through the calculation of a numerical case Marviken CFT 24 in which a pressurized vessel inflow to the containment reactor cavity.

Experimental Investigation of Condensation Induced Water Hammer during Residual Heat Removal Process for Nuclear Reactors

Wang Lutao, Li Jian, Zhong Daotong, Ran Xu, Zhang Zhuohua, Yan Junjie

2018, 39(2): 162-165. doi: 10.13832/j.jnpe.2018.02.0162

Abstract(27)

Abstract:
Serious steam condensation induced water hammer（CIWH） phenomenon would occure when the steam directly contacts with the water during the residual heat removal process for nuclear reactors.In this paper,CIWH phenomenon was investigated experimentally in a horizontal pipe.Firstly,it was found that CIWH would cause a high pressure peak,and there periodically appeared decaying pressure peaks after CIWH event.Besides,the existence of pressure difference for different positions promoted CIWH event by analyzing pressure fluctuation of different axial positions.Finally,multiple CIWH events occurred during 120 s,but pressure fluctuation intensity had a high randomness.The maximum value of pressure fluctuation intensity increased with steam mass flux increasing with a maximum value of 10 MPa.

System of Generic Criteria and Operational Criteria for Nuclear and Radiological Emergency of HPR1000

He Fan, Yu Hong, Mu Keliang

2018, 39(2): 166-170. doi: 10.13832/j.jnpe.2018.02.0166

Abstract(27)

Abstract:
The present system of generic intervention levels and generic action levels in China do not meet the need of the nuclear power plants such as HPR1000 for the nuclear and radiological emergency.In this paper,the system of the generic criteria and operational criteria of HPR1000 is advanced,based on the new system of the generic criteria and operational criteria developed by IAEA,and combining the advanced technique in the emergency action level of America Nuclear Energy Institute and the reality of the operational intervention levels in China,to improve the emergency classification,the partition of the emergency planning zone and the initiation of the protective actions.

Experimental Study on Effect of Inlet Condition on Corrugated Plate Dryer

Mao Feng, Tian Ruifeng, Chen Bowen, He Wei

2018, 39(2): 171-175. doi: 10.13832/j.jnpe.2018.02.0171

Abstract(31)

Abstract:
The effects of inlet condition including droplet size,inlet moisture content and inlet airflow speed on corrugated plate dryer’s performance were studied in a visualized experiment system.A laser particle sizer was used to measure water particle size distribution.The results show that: the separation efficiency increases with the increasing of droplets size and inlet moisture content,and increases with the increasing of inlet velocity firstly,then decreases after reaching the peak; the critical inlet speed resulting re-entrainment decreases with the inlet moisture content increasing; the pressure drop of dryer increases with the inlet air speed quadratically.

Study on Design of Fault Diagnosis System for Nuclear Power Plants

Wang Tianshu, Yu Ren, Liu Xiaofan

2018, 39(2): 176-179. doi: 10.13832/j.jnpe.2018.02.0176

Abstract(22)

Abstract:
To improve the operation safety and reliability of nuclear power plants and reduce human error,the design of operation fault diagnosis system based on the theory of expert system and human factor engineering is researched in this paper.Firstly,the demand of the system functions is analyzed,and the definition of the key security parameters and their classification method is brought up.The safety state supervisory function module is briefly designed in order to facilitate the operator to monitor the safety state of the nuclear power plant from macroscopic to detail.Secondly,to maximize the speed and accuracy of the reasoning process and keep in line with the operator’s way of thought and experience,the structure of the operation fault diagnosis expert system is designed,as well as its database and knowledge base,along with a new method of reasoning that combines the hierarchical reasoning method with forward and backward reasoning method.

3D Simulation of Fuel Performance with Missing Pellet Surface Defect

Liu Zhenhai, Chen Ping, Zhou Yi, Li Wenjie, Zhang Kun, Xing Shuo, Miao Yifei

2018, 39(2): 180-184. doi: 10.13832/j.jnpe.2018.0180

Abstract(32)

Abstract:
The effect of missing pellet surface（MPS） defect on fuel performance by coupling three dimensional code and one and half dimensional code is studied.The three dimensional fuel performance code is built based on ABAQUS software and compared with COPERNIC,a one and half dimensional fuel performance code.Based on this,the fuel performance near the MPS defect during typical condition II transient is studied.The result shows that the temperature of the cladding inner surface near MPS defect is lower than the surrounding and the temperature of cladding inner surface near the defect edge is significantly higher than the surrounding; There is stress concentration in the cladding around the defect and the cladding hoop stress distribution near the defect presents a typical "plate bending" phenomenon.

Investigation on Hold-down Force Calculation Model of Fuel Assembly Based on Monte Carlo Algorithm

Zhu Fawen, Pu Zengping, Chen Ping, Ma Chao, Li Yun, Zhou Xiaoyun, Zeng Xiaomin, Geng Fei

2018, 39(2): 185-188. doi: 10.13832/j.jnpe.2018.02.0185

Abstract(32)

Abstract:
The hold-down force calculation model of fuel assembly was established based on the Monte Carlo algorithm,and model validation has been demonstrated its validity.It is suitable to use 200 thousand simulation times for the hold-down system analysis when using Monte Carlo algorithm in practice.

Application and Prospect of PHM Technology in Nuclear Power Plants

Xie Guangyao, Liu Jingquan, Zeng Yuyun

2018, 39(2): 189-192. doi: 10.13832/j.jnpe.2018.02.0189

Abstract(33)

Abstract:
Prognostics and Health Management（PHM） system is able to improve the safety of nuclear power plants and reduce maintenance costs.Prognostics and Health Management system includes five modules:data collection and processing,condition monitoring,fault diagnosis,life prediction and health management.Dedicated monitoring methods and prediction algorithms with appropriate prognostics performance metrics should be applied.This paper presents the suitable development route of PHM system in nuclear power plants in China.Maintenance standards and regulations should be formulated and fault database should be built which provides a good basis for PHM system to be widely used in nuclear power plants in China.

2018 Vol. 39, No. 2