核动力工程

Benchmark Experiment on Critical Heat Flux of PWR Fuel Assembly

Qin Shengjie, Lang Xuemei, Xie Shijie, Li Pengzhou, Zhang Junyi, Liu Wenxing, Zhuo Wenbin

2016, 37(5): 1-3. doi: 10.13832/j.jnpe.2016.05.0001

Abstract(33) PDF(0)

Abstract:
Critical heat flux（CHF） under different condition is important to the thermal-hydraulics design and safety analysis. Benchmark experiment is carried out with full length non-uniform power distribution rod bundle using Large Scale Thermal Hydraulics Test Facility（LS-THTF） of NPIC（Nuclear Power Institute of China）. The facility, rod bundle and experiment method are introduced. The CHF data using power of rod bundle are directly compared with the reference measured data from HTRF of Columbia University. In the indirect comparison, the measured CHF data are reduced to compare the Measured to Predicted ratio（M/P） of local heat flux using the sub-channel code combined with CHF correlation. The results show good agreement between CHF data from LS-THTF and HTRF.

Experimental Research of Critical Heat Flux on Pressure Vessel Lower Head External Surface under Severe Accident

Zhang Zhen, Xiong Wanyu, Wang Xiong, Zhuo Wenbin, Li Pengzhou, Zang Jinguang, Song Mingliang

2016, 37(5): 4-9. doi: 10.13832/j.jnpe.2016.05.0004

Abstract(24) PDF(0)

Abstract:
An experiment was conducted to study the critical heat flux in the cavity injection system. A constant-width slice heater block was used for the representing of the reactor vessel lower head, in combination with a power-shaping approach, in the experiment. The critical heat flux（CHF） at various angular position of the external wall was measured under the pool-boiling condition and the forced cooling condition, respectively. The experimental results indicated that, under both pool-boiling condition and forced cooling condition, the variation in CHF at different angular position can be divided into three kinds. In the inlet region, the critical heat flux increases with the increasing of the angular position under pool-boiling condition, but under forced cooling condition the critical heat flux decreases with the increasing of the angular position. In the middle region, the critical heat flux increases with the increasing of the angular position. In the outlet region, the critical heat flux decreases with the increasing of the angular position.

Experimental Research of Quasi-Steady Time of Liquid Film Flow on Containment

Lu Yanghui, Wang Yanzhi, Liu Lu, Wang Yan

2016, 37(5): 10-14. doi: 10.13832/j.jnpe.2016.05.0010

Abstract(22) PDF(0)

Abstract:
Quasi-steady time characterized the response of the water film to the steady state in the passive containment cooling system. It is a key factor in the heat transfer and design of the nuclear safety system. There are two processes during the falling film covering the containment surface. The first process is water flows down, resulting in a wet area; the second is water laterally wet the dry area due to its semi-stability, and the contact angle becomes smaller and slowly increases the coverage area. Coupling of the two processes makes it difficult to determine the quasi-steady time. On the basis of the results of collection tank level, flow and pictures of water film coverage in CAP1400 water distribution experiment, present study proposed two methods to determine the stability of the water film, which are flow balance method and coverage stable method. It solved the calculation problem of quasi-steady time and provided a new way for determining the stability of the water film, and furthermore, relations of relative quasi-steady time with Reynolds number based on different distribution water structures are obtained.

Research on High Efficient CFD Schemes for PWRs

Chen Guangliang, Zhang Zhijian, Tian Zhaofei, Li Lei

2016, 37(5): 15-18. doi: 10.13832/j.jnpe.2016.05.0015

Abstract(20) PDF(0)

Abstract:
Many challenges exist in the CFD simulation for the whole pressurized water reactor（PWR）. For instance, the simulation distortion will occur without the fine mesh scheme, but it is time consuming to design and select the fine meshes for various complicated structure zones in the core. CFD simulation of the whole PWR core requires more than 60 billion meshes, which costs massive computing resources and time. To reduce the burden, two schemes were developed. One is the multiple RANS scheme. The other one is based on the structure simplification and simulation modification. Both schemes can give the accurate results and reduce a lot computing burden.

Analysis of Three-Dimensional Nonlinear Seismic of Reactor Structure

Huang Qian, Zhang Yixiong, Shen Pingchuan, Yu Xiaofei, Wu Wanjun

2016, 37(5): 19-23. doi: 10.13832/j.jnpe.2016.05.0019

Abstract(22) PDF(0)

Abstract:
A three-dimensional nonlinear FEM model of reactor equipment and system was built. In the model, global Rayleigh damping, local material damping and element damping were used simultaneously to deal with the damping difference of different structures, and the method of equivalent stiffness and equivalent gap between fuel assemblies and the core shroud was proposed. Meanwhile, the contact effect with a positive gap between two concentric cylinders and the dynamic friction effect with variational pre-tightening force was simulated. Subsequently, seismic time history inputs compatible with both response spectrum and power spectrum density（PSD） was introduced as the analysis input, and a nonlinear 3D reactor equipment and system seismic analysis was carried out. Finally, seismic responses were obtained for SSE event. The research in this paper provides a guidance and reference for the future 3D nonlinear seismic analysis of reactor equipment and system.

Seismic Analysis of Nuclear Power Plant Structure Based on Multi-Transmission Formula Boundary Model

Li Jianbo, Mei Runyu, Lin Gao, Zhang Pengchong

2016, 37(5): 24-28. doi: 10.13832/j.jnpe.2016.05.0024

Abstract(19) PDF(0)

Abstract:
At present, the seismic safety analysis of nuclear power structures only use some simple numerical models recommended by various specifications, or the basic finite element model with viscous-elastic boundary, and there is no relatively high precision analysis method. A high precision multi-transmission formula model with artificial boundary on the basis of the implicit-explicit coupling integral method is introduced in this paper, and a directly numerical algorithm scheme on the framework of ANSYS is proposed, which particularly is applicable for the dynamic interaction analysis of complex factory building structures of nuclear engineering. And then, the validation and applicability of this new method is verified using a practical example. Numerical results proof that the proposed model is applicable for the finite element analysis of the dynamic response of nuclear power plant buildings

Analysis of Effect of Steam Generator Anti-Vibration Bars Shift on Flow Induced Vibration and Wear of Tube Bundle

Zhu Yong, Han Tonghang, Ren Hongbing

2016, 37(5): 29-32. doi: 10.13832/j.jnpe.2016.05.0029

Abstract(22) PDF(0)

Abstract:
During the pre-service eddy current inspection（PSI） of the steam generator of a power station, it was determined that the anti-vibration bars（AVBs） No.3 shifted from the design location to the cold side by approximately 10 degrees. Based on the specific flow induced analysis program GERBOISE, flow elastic characteristic, turbulence excitation response and fretting wear of the tube bundle for both AVBs design and shifted locations are calculated. By comparison of the calculated results of design and shifted conditions, the effects of AVBs shift on flow induced vibration and fretting wear are evaluated. The evaluated conclusion is that the shifted AVBs will not lead to the unaccepted flow induced vibration response and excessive fretting wear for the heat transfer tube.

Research on Creep Damage and Constitutive Equation of Nuclear Power Steel A508-Ⅲ at Phase Transition Temperature（800℃）

Xie Zhigang, Yang Jianguo, He Yanming, Gao Zengliang

2016, 37(5): 33-39. doi: 10.13832/j.jnpe.2016.05.0033

Abstract(23) PDF(0)

Abstract:
Creep tests for domestic RPV steel, A508-Ⅲ, have been performed under three types of load, 17.5MPa, 20 MPa and 27 MPa at 800℃. Furthermore, the staged creep tests of A508-Ⅲ steel under 20 MPa and 27 MPa are carried out. The study on the microstructure and creep curve shows that the volume fractions of cavities and second phase particles increase linearly in the process of creep, so the nucleation and growth of cavities as well as second phase particles coarsening have become the main reason of creep damage. Based on mesomechanics ideas and analysis on the microscopic creep damage mechanism of A508-Ⅲ steel at 800℃, an K-R creep damage constitutive equation involved microscopic damage and its evolution is established through defining a representative volume element, which is a three-phase composite volume consists of no damage phase, cavities phase and second phase particles. By the normalization process, the creep constitutive equation and damage evolution equation reflected the evolution of cavities and the second phase particles are obtained, which set up the inner relation between micro structure damage and macroscopic constitutive equation.

Measurement of Radiated Aluminum Weld Residual Stress by Neutron Diffraction

Wang Shuyu, Feng Qijie, Wang Hong, Li Jian, Liu Yaoguang

2016, 37(5): 40-42. doi: 10.13832/j.jnpe.2016.05.0040

Abstract(14) PDF(0)

Abstract:
Neutron diffraction method was used to measure the residual stress of the aluminum weld of the retired reactor, and the effect of irradiation on the residual stress of aluminum weld was studied. Test result indicates that the maximum residual stress is located in the heat affected zone, the greater the reactor irradiation dose, the greater the oscillation of the residual stress of aluminum weld center from compressive stress to tensile stress. At the position far away from the weld position, the strain of the irradiated samples is not zero.

Research on Model of Fission Gas Bubble Growth Based on Oswald Ripening Mechanism

Long Chongsheng, Zhang Yu, Chen Hongsheng, Xiao Hongxing, Wei Tianguo, Zhao Yi

2016, 37(5): 43-45. doi: 10.13832/j.jnpe.2016.05.0043

Abstract(21) PDF(0)

Abstract:
Based on the Ostwald ripening process, a mechanistic model for the fission gas bubble growth has been proposed in this paper. Taking U-Zr alloy and UO₂ ceramic fuel as samples, the variation of FGB porosity and mean bubble size with the burn-up at different irradiation temperature and external restraint have been investigated in detail. For ceramic UO₂, the porosity increase rate with the burn-up at 700°C is 25% faster than that at 400°C. The external restraint impedes the growth of FGB, but it is not very significant. The porosity and mean bubble size in U-Zr alloy for the same irradiation parameters are much greater than that in the ceramic UO₂ because the plastic deformation takes place very easy and quickly for U-Zr alloy during irradiation.

Research on Rapid Location of Ruptured Fuel Element by Cs Detection

Sun Shouhua, Li Jian, Zhu Lei, Li Ziyan

2016, 37(5): 46-50. doi: 10.13832/j.jnpe.2016.05.0046

Abstract(26) PDF(0)

Abstract:
Based on the nuclear reaction of ¹³⁷Cs and ¹³⁴Cs in the fuel assembly, the accurate mathematical and physical calculation models were established. The exact and simplified solutions of relationships between radioactivity of the two nuclides in the primary coolant and burn-up of ²³⁵U in the fuel assembly were obtained. The comparisons of simplified solution, ORIGEN2.0 calculation result and exact solution were carried out as well. The result showed that, if only the R_Cs, ratio of ¹³⁴Cs/¹³⁷Cs in the primary coolant, is obtained by hydrochemistry measurement, burn-up of the ruptured fuel assembly can be calculated with the analytical solution model, so as to locate the ruptured fuel assembly rapidly.

Design of CANDU Regional Overpower Protection Fluxshapes Reclassification Scheme

Wang Wencong, Ye Guodong, Mou Xiaochuan, Sheng Jianxin, Zheng Yongxiang, Doddy Kastanya

2016, 37(5): 51-54. doi: 10.13832/j.jnpe.2016.05.0051

Abstract(19) PDF(0)

Abstract:
The Regional Overpower Protection（ROP） system is designed to prevent the dryout in any fuel channel of CANDU reactors during a slow loss of regulation（LOR） event. Reactor aging will lead to the reduction of ROP margin. To resolve the deficiency of ROP margin in Third Qinshan Nuclear Power Plant（TQNPP）, four reclassification rules are determined based on review of original ROP fluxshapes classification principle, operational documentation and policy of TQNPP, reactivity device configuration and existing DCC alarms. ZDROP alarm for monitoring axially averaged zone deviation is also developed and installed. Totally 270 fluxshapes are removed outside of ROP normal position HSP#1. The implementation of ROP reclassification scheme in TQNPP indicates nearly 3% to 5% ROP margin can be restored and additionally the economy and safety of reactor will be improved.

Design of Nuclear Instrumentation System for Liaoning Hongyanhe Nuclear Power Plant

Shen Feng, Chen Le, Li Gao, Liu Yanyang, Li Wenping, Wang Yinli

2016, 37(5): 55-57. doi: 10.13832/j.jnpe.2016.05.0055

Abstract(17) PDF(0)

Abstract:
The paper introduces the design work of the nuclear instrumentation system for Liaoning Hongyanhe Nuclear Power Plant. It focuses those pivotal design contents such as the system function, the structure and equipment function configuration scheme, the calculation, analysis and test validation to system vital technical parameters and the equipment signal interface design. The commercial operation of the nuclear power plant proved that the design work of the system was successful.

Emergency Classification of Nuclear Power Plants with Wet Storage Spent Fuel

YU Hong

2016, 37(5): 58-62. doi: 10.13832/j.jnpe.2016.05.0058

Abstract(30) PDF(0)

Abstract:
The emergency classification of the nuclear power plants with wet storage spent fuel is analyzed, based on the threats of the criticality return of spent fuels and the losing of shield, heat removal and containment. The parameters for classification can not only be monitored, but also can be used to character the safety level of the spent fuels. The representative values of water level, temperature and dose rate and the emergency classification of the spent fuel for Fangjiashan Nuclear Power Plant are provided.

Experimental Investigation on Small Break Loss of Coolant Accident of Surge Line

Peng Chuanxin, Lu Xiaodong, Zhang Yan, Bai Xuesong, Zan Yuanfeng, Zhuo Wenbin, Yan Xiao

2016, 37(5): 63-67. doi: 10.13832/j.jnpe.2016.05.0063

Abstract(17) PDF(0)

Abstract:
The small break loss of coolant accident（SBLOCA） experiment of surge line was performed on the small modular reactor passive safety system test facility, to investigate the thermal hydraulic phenomena and the performance of passive safety system during the accident. In the experiment, the fluid density difference between the pressure balance line and injection line driven the coolant to inject into the Reactor Pressure Vessel（RPV）. When the pressure balance line is no longer submersed, the CMT injection flowrate became unstable; The Accumulator（ACC） had prominent effect on the cooling of the core during the initial period of accident; After the sufficient depressurization of the Automatic Depressurization System（ADS）, the coolant in the In-containment Refueling Water Storage Tank（IRWST） provided sustaining and stable injection and cooling of the core. The experimental results show that the passive safety system can provide effective coolant injection and successful removal of core decay heat under surge line SBLOCA.

Research on Precision Calibration Technology for Platinum-Resistance Thermometry in Reactors

Chen Panhui, Guo Lifeng, Lu Gubing, Jin Chuanxi

2016, 37(5): 68-70. doi: 10.13832/j.jnpe.2016.05.0068

Abstract(18) PDF(0)

Abstract:
Design a new structure of digitally controlled resistance network. The resistance network can output a high resolution resistance with the effect of reducing the unstable characteristic of relay contacts resistance, and meet the accuracy requirement of the platinum-resistance thermometry in reactors. the mismatch error which is led by the resistor’s initial error in the case of using the binary resistor weight sequence. Error calculation formula and mismatch condition is given. High resolution and high stability characteristics of resistance source can be used to design, test and calibrate the reactors’ platinum-resistance thermometry.

Distortion Oversize Analysis and Control Measures for Core Barrel

Xia Xin, Li Ning, Li Yan, Zhao Wei, Chen Xungang, He Peifeng

2016, 37(5): 71-74. doi: 10.13832/j.jnpe.2016.05.0071

Abstract(23) PDF(0)

Abstract:
Severe oversize distortion frequently occurs in the manufacturing of the core barrel, which affects the processing quality of the core barrel and the construction period of nuclear power plants. This paper analyzes the effecting factors and causes of oversize distortion of the core barrel, and provides the measures to control the distortion, which achieves good results in the following manufacture of the core barrels. Moreover, the improvement and adjustment of the barrel design is provided and applied to the Hualong No.1 project which is the Chinese third generation nuclear power plant.

Quality Supervision of AP1000 SG Weld and Analysis of DING Prevention in Local Heat Treatment

Liu Shihui, Li Tao, Mao Changsen

2016, 37(5): 75-77. doi: 10.13832/j.jnpe.2016.05.0075

Abstract(19) PDF(0)

Abstract:
DING prevention technology used in the local heat treatment of the circular welds on AP1000 steam generator is a new technique different from that for the second generation SG. This paper briefly introduces the causes for DING, influencing factors and prevention measures, and provides the key check points in the process of the quality supervision.

Analysis of Uprightness Change and Replaceabiltiy of CRDM Nozzle

Chen Haibo, Luo Ying, Wang Xiaobin, Fu Qiang

2016, 37(5): 78-80. doi: 10.13832/j.jnpe.2016.05.0078

Abstract(24) PDF(0)

Abstract:
Based on the design structure and manufacture procedure of the coupling structure of CRDM（Control Rod Drive Mechanism） nozzle to RPV（reactor pressure vessel） closure, which was widely used nowadays, this paper focused on the detailed analysis of nozzle uprightness change after RPV hydraulic pressure and the replaceabiltiy of damaged nozzle. Based on the main reasons of nozzle uprightness change and key factor to lower the replaceabiltiy of nozzle, an improved design of coupling structure was produced. It was shown that the improved coupling structure could ensure the nozzle uprightness and lower the difficulty to replace the nozzles.

Analysis on Selection of Spring Support of Piping

Ma Zhaoguo, Wu Chunming

2016, 37(5): 81-83. doi: 10.13832/j.jnpe.2016.05.0081

Abstract(17) PDF(0)

Abstract:
Through a combination of theoretical analysis and engineering practical case research method, the analysis on selection of spring support of piping is made. This paper sketched the application and classification of the spring support, and pointed out the main characteristics of 2 types of spring supports. The force of spring support for the pipeline under different working conditions is analyzed, and the principle of load design and the theoretical basis for the stiffness check are given. Analysis of the effect of spring support on the piping in the earthquake is also made. Finally, as an engineering example, the selection process of spring support for main steam pipeline of a nuclear power plant is described.

Technology Research on Reactor Core Positioning for Domestic Manipulator

Ye Jianjun, Zhang Jian, Zhang Peng, Lu Kefeng, Sun Guitong

2016, 37(5): 84-88. doi: 10.13832/j.jnpe.2016.04.0084

Abstract(20) PDF(0)

Abstract:
According to the analysis of the factors that have effect on the accuracy of the core location, the effect of various parameters on the core positioning were quantization. In order to reduce the positioning error of the single point positioning method and the improved four-point centering average step length positioning method, a gauge precise positioning method is proposed, and successfully applied in a nuclear power plant. Results show that, this core positioning technology greatly reduces the measurement error during the core step length selection, and improves the coordinate point in the core minimum quantity from the traditional 7mm to 11 mm.

Risk-Informed Inservice Inspection Evaluation in Daya Bay Nuclear Power Plant

Liu Pingping, Xi Haiying

2016, 37(5): 89-92. doi: 10.13832/j.jnpe.2016.05.0089

Abstract(27) PDF(0)

Abstract:
This paper describes in detail the methodology and implementation process of the risk-informed inservice inspection evaluation in Daya Bay nuclear power plant. The risk-informed inservice inspection evaluation is used to optimize the current inservice inspection of the residual heat removal system in Daya Bay unit 1. After optimization, the inspection location and numbers change. The assessment results show that the risk-informed inspection method is effective. It can focus on the risk significant locations, decrease the inspection numbers and the staff occupation doses, and the public health and safety are maintained and improved. The risk-informed inservice inspection does not only simply decrease the inspection numbers or change the inspection methods, but also pays more attention to the risk significant pipe segments. Risk-informed inservice inspection evaluation makes the inspection based on the degradation mechanism, i.e. cause-based, to improve the inspection effectiveness. In the end, some key issues about the risk-informed inservice inspection evaluation process are discussed and suggestions are proposed on its application.

Research on Design for Peak-Over-Threshold Flow Restriction Measurement of Auxiliary Feedwater Turbine-Driven Pump

Wang Yanhua, Li Xiaojun, Liu Zuquan, Li Yongjun

2016, 37(5): 93-96. doi: 10.13832/j.jnpe.2016.04.0093

Abstract(26) PDF(0)

Abstract:
Based on the turbine-driven pump manufacturing document and the design parameters of the operating auxiliary feedwater system, this paper proposes the measurement design for peak-over-threshold flow restriction of auxiliary feedwater turbine-driven pump. For the factors of the actual layout and system measurement requirements in Fuqing nuclear power plant, design optimization is also proposed. Through prototype test and on-the-spot debugging test, the feasibility of measurement design are verified. The measurement can completely meet the operational requirements of the system in nuclear power plant blackout accident.

Analysis of Failure Resulted from Abnormal Thickness Reduction of Titanium Tubes Used in SRI Heat Exchanger of Nuclear Power Plants

Xie Jianhua, Zheng Zhishou, Su Xiuli, Bai Feifei, Zhao Xiaohong

2016, 37(5): 97-101. doi: 10.13832/j.jnpe.2016.05.0097

Abstract(19) PDF(0)

Abstract:
Considering the abnormal thickness reduction of the titanium heat transfer tubes in some SRI heat exchangers of a nuclear power plant, the chemical composition of raw material, performance, working medium, flow induced vibration and fretting damage are studied It is found that the primary causes for the failure are the flow induced vibration between tubes and baffle rod, and the fretting damage. A series of specific countermeasures is provided. The achievements obtained can provide as an instructive reference to ensure the effective protection and safe operation of heat exchangers running under the circumstance of seawater in nuclear power plants.

Ultrasonic Inspection of Reactor Coolant Pump Flywheel Used in Nuclear Power Plants

Wu Jinfeng, Ren Jianbo, Sun Jiawei, Chen Yan, Chen Zhicong

2016, 37(5): 102-104. doi: 10.13832/j.jnpe.2016.05.0102

Abstract(23) PDF(0)

Abstract:
The reactor coolant pump flywheel keyway is a vulnerable part. Therefore, it requires an ultrasonic inspection both pre-service and in-service. This paper focuses on the scope of the inspection and the design of a combined probe which can do the inspection and the signal analysis. It is confirmed that the ultrasonic inspection method meets the requirements of the standards. It also provides a new method of ultrasonic inspection with higher sensitivity.

Design and Implementation of T3 Periodic Test for Reactor Protection System of TMSR-SF

Liu Zhenbao, Hou Jie, Liu Guimin

2016, 37(5): 105-110. doi: 10.13832/j.jnpe.2016.05.0105

Abstract(21) PDF(0)

Abstract:
This paper introduces the scope and principle for periodic tests of TMSR-SF1 RPS based on the characteristics of RPS and the regulatory requirements and test demands, and describes the T3 platform of TMSR-SF1 RPS and highlights the design of T3 Periodic Test. Furthermore, the correctness of RPS T3 test is demonstrated according to the relevant legal requirements.

Sensitivity Analysis of Fuel Assembly Structure Parameters Affecting RCCA Hydraulic Damping

Xiao Zhong, Ma Chao, Guo Xiaoming

2016, 37(5): 111-114. doi: 10.13832/j.jnpe.2016.05.0111

Abstract(26) PDF(0)

Abstract:
To implement the process of rod cluster control assemblies（RCCA） dropping in PWRs, the dashpot structure below the guide thimble is designed. The select of the dashpot structure parameters has a great influence on the effect of the RCCA damping. In this paper, the theory model about the RCCA damping in the control rod dropping process is established, and the sensitivity analysis is carried out by changing the dashpot structure parameters, including the diameter and the length of the screw hole, and the gap between the rod and the guide thimble. The parameter influence regularity has been deduced to guide the guide thimble component design in the primary process in future.

Study on Design Improvements of Recirculation SG Based on Ageing Degradations

Li Pengfei, Yu Ping, Wang Haisong, Cheng Xiang, Li Huanming, Huang Wei, Shen Yunhai

2016, 37(5): 115-118. doi: 10.13832/j.jnpe.2016.05.0115

Abstract(21) PDF(0)

Abstract:
Design improvements of recirculation SG is discussed based on different ageing degradation mechanisms around the heat transfer tube. The results indicate that the design improvement measurements, such as taking I-690 TT as the heat transfer tube materials, installing the blow down components, supporting the plate with broached quatrefoil or broached trefoil structures, installing V type anti vibration bar at U bend, and conducting all volatile treatment for feed water, can effectively avoid or mitigate the general ageing degradations for the recirculation SG.

Accident Analysis of Heat Pipe Cooled Space Reactor System

Liu Songtao, Yuan Yuan, Wei Zonglan, Zeng Wei, Zhu Li, Gou Junli

2016, 37(5): 119-124. doi: 10.13832/j.jnpe.2016.05.0119

Abstract(20) PDF(1)

Abstract:
A transient analysis code TAPIRS was developed to analyze the behavior of the heat pipe cooled space reactor power system based on the SAIRS models. Three typical accidents are analyzed using TAPIRS. The results show that the fuel temperature is below a safe limit under the control drum failure, the AMTEC failure and partial loss of the heat transfer area of radiator. This demonstrates that the reactor system is with the characteristics of self-stabilization ability under accident conditions.

Development and Verification ＆ Validation of Theoretical Model for Critical Heat Flux under Flow Oscillation Conditions

Liu Wenxing, Zhao Dawei, Su Guanghui, Huang Yanping

2016, 37(5): 125-129. doi: 10.13832/j.jnpe.2016.05.0125

Abstract(22) PDF(0)

Abstract:
A mechanistic critical heat flux（CHF） model was developed to predict the transient CHF under inlet flow oscillation conditions. The model was proposed based on the sub-layer dryout model and homogeneous two-phase transient flow model. FORTRAN language was used to write the calculation code, which was then verified and validated using experimental data from both steady inlet flow and oscillatory inlet flow conditions. The verification and validation results show that the proposed model has good capability to predict the transient CHF under inlet flow oscillation conditions within a small error range.

CHF Correlation Development and DNBR Limits Determination

Zhang Yuxiang, Xi Yanyan, Pang Zhengzheng, Li Weicai, Zhou Yuemin, Zhao Hua

2016, 37(5): 130-134. doi: 10.13832/j.jnpe.2016.05.0130

Abstract(27) PDF(1)

Abstract:
In this paper, the key issues in the development of reactor critical heat flux（CHF） correlation such as cold wall effect, non-uniform heating factor and statistical treatment of data have been studied, and based on the published CHF trial data we completed the localization of CHF correlation fitting and determination of the corresponding limit value. With a detailed statistical analysis of the calculation results, we established a rigorous method to determine the DNBR limits of CHF correlation.

Research on Passive Residual Heat Removal System of SG Secondary Side of Underground Nuclear Power Plants

Lai Jianyong, Shen Yunhai, Wang Baoping, Yu Xiaoquan, Sui Haiming, Zhu Li, Yu Fei

2016, 37(5): 135-137. doi: 10.13832/j.jnpe.2016.05.0135

Abstract(18) PDF(0)

Abstract:
The underground nuclear power plant researched in China after FUKUSHIMA nuclear accident is a more safe and more acceptable style of nuclear power plant. The containment of underground nuclear power is buried about 180 meters underground, while the pool full of cooling water is equipped on the ground. The driving head of nature circulation of passive heat removal system of secondary side is sufficient due to the height difference of steam generator and cooling pool. The functional requirement, system components, equipment character and operating of the passive heat removal system of secondary side in underground nuclear power plant are presented in this paper.

Scheme Study on Waste Treatment of Underground Nuclear Power Plants

Gao Feng, Ma Xingjun, Chen Xianlin, Lin Li, Liao Wei, Zhao Xin

2016, 37(5): 138-141. doi: 10.13832/j.jnpe.2016.05.0138

Abstract(28) PDF(0)

Abstract:
The preliminary planning is made for the radioactive waste treatment factory based on the plant layout of underground nuclear power plants and the characteristics of the waste sources. The facility to dispose the gaseous and liquid radwastes was located in the nuclear assistant factory. The facility to dispose the concentrated liquid and solid wastes was located in the site of the radwaste treatment facility and built on the ground. The radwaste treatment scheme was established. The occluding, insulating and treating measures for the liquid and gaseous wastes were introduced, as well as the composition of the mobile gassing equipment and the mobile liquid treatment equipment. The released radioactive gas and liquid can be occluded, insulated and treated effectively.

Study on Prevention of Airborne Radioactivity Diffusion of Underground Nuclear Power Plants

Zhang Tao, Zhao Xin, Liu Haibo, Yu Fei, Su Yi, Chen Bin

2016, 37(5): 142-146. doi: 10.13832/j.jnpe.2016.05.0142

Abstract(20) PDF(0)

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The general principle for the prevention of the airborne radioactivity diffusion in underground nuclear power plants is put forward. The engineering measures to prevent the radioactivity diffusion in the underground nuclear power plant are studied with the emphasis on the engineering measures under severe accidents. The pressure relief system for containment and reactor building cavern under severe accidents which vents and filters the airborne radioactivity according to the different conditions in severe accidents was proposed. By these engineering measures, the airborne radioactivity in severe accidents in underground nuclear power plants is more effectively prevented and controlled, and it is possible to achieve the practical elimination of the accident sequences that may lead to radioactive release by design.

Study on Radioactive Waste Water Underground Migration Protection System of Underground Nuclear Power Plant

Niu Xinqiang, Shi Huatang, Li Hongbin, Min Zhenghui, Zhang Wenqi

2016, 37(5): 147-151. doi: 10.13832/j.jnpe.2016.05.0147

Abstract(25) PDF(1)

Abstract:
The influencing factors of radioactive waste water underground migration areanalyzed, the radioactive waste water migration protection conditions of underground and ground nuclear power plant is contrasted, and the ideas and specific engineering protection measures for radioactive waste migration protection of underground nuclear power plant are systematic proposed, that is, all radioactive waste water security measures of ground nuclear power plants are completely preserved and make full use of the natural protective properties of the rock, the additional closed, dewatering and other reliable engineering measures are set in the surrounding rock mass, the migration channel of radioactive waste water is blocked, at the same time, the collection, disposal and monitoring system are also set. The effect of above protective measures is remarkable by numerical analysis, and the produced radioactive waste water of underground nuclear power plant is controlled in severe accident condition.

Preliminary Study on Minimization of Emergency Planning Zone for Underground Nuclear Power Plant

Lyu Huanwen, Jing Futing, Liu Jiajia, Wang Junlong, Yang Ping, Zhang Tao

2016, 37(5): 152-155. doi: 10.13832/j.jnpe.2016.05.0152

Abstract(23) PDF(0)

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An envelope accident was selected and corresponding source term was given. Then this paper made a calculation about the size of emergency planning zone. Calculation results show that the radioactive release of CUP600 to the environment is small, the effective dose and thyroid dose during the whole accident are both less than the general optimized intervention level. The plum emergency zone can be totally inside the site boundary, so technically, out site plum emergency planning zone can be eliminated.

Research on Key Technologies of Large Cavern Groups Excavation of Underground Nuclear Power Plants

Su Lijun, Liu Lixin, Li Feng, Zhang Zhijun, Sui Haiming, Zhu Yonghui

2016, 37(5): 156-160. doi: 10.13832/j.jnpe.2016.05.0156

Abstract(23) PDF(0)

Abstract:
Taking the circular layout underground nuclear island of the underground nuclear power plant for an example, the construction channel layout and the main construction schemes of the large cavern groups are given. At the same time, this paper presents the excavation key technologies of the nuclear reactor building and other auxiliary ones, and summarizes the main technical points of the controlled blasting technology, the supporting and enforcement technology, the rapid and informationalized construction technology. All the excavation key techniques can support the construction feasibility of the underground nuclear power plants.

Optimization Design for Supercritical Water Reactor CSR1000 Core

Wang Lianjie, Lu Di, Chen Bingde, Yao Dong, Zhao Wenbo

2016, 37(5): 161-166. doi: 10.13832/j.jnpe.2016.05.0161

Abstract(20) PDF(0)

Abstract:
An optimization conceptual design of CSR1000 core is proposed. Steady state performance of the proposed core is then studied with the SCWR core steady state analysis code system SNTA. These key parameters such as burnup performance, reactivity control capability, power distribution, maximum fuel cladding temperature and maximum linear power density are analyzed. The relative coolant flow rate of the second flow path which is suited with assembly power is also presented. The study shows that the life of CSR1000 core could be extended effectively with the optimization design for CSR1000 fuel assembly and core.

Calculation of Activation Source Terms of Reactor Components for Decommissioned Nuclear Power Plant

Su Genghua, Bao Pengfei, Han Song, Li Ming

2016, 37(5): 167-170. doi: 10.13832/j.jnpe.2016.05.0167

Abstract(23) PDF(0)

Abstract:
This paper presents a method and results of predicting activation source terms of reactor components for NPP decommissioning and its preliminary verification. Through the modification of the one-group neutron reaction cross sections of ORIGEN2, the source terms from activation of reactor components were calculated using the codes of MCNP/ORIGEN2. The results revealed that there were 6 or 7 nuclides which mainly constitute the source terms and the nuclides and their amounts vary significantly with the material composition and the distance to the reactor core of the components. Specific activity of activation samples in a reactor vessel irradiation surveillance capsule were calculated and compared to the measurement data. The comparison indicated that the values calculated with the modification of cross sections agreed well with the measured data within a difference range of 20% while the values calculated without modification of cross sections agreed badly, which verified the applicability of the calculation method presents by this paper.

Exploration of Imaging for Visible Light and Infrared Prototype Monitor of Reactor Cabin under Nuclear Radiation

Zhou Xuhua, Yi Xiongying, Xu Jianguo, Gao Yuan

2016, 37(5): 171-173. doi: 10.13832/j.jnpe.2016.05.0171

Abstract(27) PDF(0)

Abstract:
The nuclear radiation environment in the reactor cabin is simulated by neutrons and gamma rays engendered by the deuteron beam bombarding thick Be target in the 4.5 MV Van de Graaff electrostatic accelerator at Peking University. In this environment, the imaging characteristics of CCD visible light imaging hardware and thermal-sense resistor infrared imaging hardware of the visible light and infrared prototype monitor of the reactor cabin is tested. At the same time, the imaging capability of RUIXING CMOS and CCD imaging hardware is also tested. The experiment results indicate: the anti-radiation capability of the thermal-sense resistor infrared imaging hardware excels the CCD and CMOS imaging hardware and adapts to the nuclear radiation environment in the reactor cabin; the anti-radiation capability of CMOS excels that of CCD and should be used as the visible light imaging hardware for the monitor in the reactor cabin; in order to improve the working reliability, more reinforcement of the anti-radiation should be taken on the control circuit.

2016 Vol. 37, No. 5