核动力工程

Numerical Studies on Flow and Heat Transfer Characteristics of Fluoride Salt Cooled Pebble Bed Test Reactor

Wei Shiying, Wang Chenglong, Tian Wenxi, Su Guanghui, Qiu Suizheng

2018, 39(4): 1-5. doi: 10.13832/j.jnpe.2018.04.0001

Abstract(278)

Abstract:
In this paper, the detailed local flow and heat transfer phenomena are investigated by computational fluid dynamics(CFD) for the locations that may have the maximum pebble temperature. The distributions of temperature on the surface or in the fuel pebble, and the velocity field near the fuel pebble are obtained and analyzed. Results are compared with that acquired by in-house code FRAC and the maximum relative error is 2.9%, which preliminarily proves the accuracy of FRAC. This study can provide useful information to the experimental and mechanism research of FHR.

Numerical Study on 177 Hybrid Core Loading Scheme Based on MOX Fuel Assembly

Wang Ningyuan, Li Ran, Liu Yibao, Yang Lingfang, Yang Bo

2018, 39(4): 6-10. doi: 10.13832/j.jnpe.2018.04.0006

Abstract(246)

Abstract:
Through the calculation of the fuel average discharge burn-up of HPR1000, the isotope of Pu in the spent fuel has been derived, with which composition the MOX fuel shall be designed. Based on the discrete fuel module design, the power gradient with the fuel module of UO₂ is reduced through the discrete layout of MOX fuel bar with Pu contents. This paper employs Monte Carlo Program MCNP and COSLATC to simulate the reactor core power distribution and the thermal neutron flux distribution, and uses the compartmentalized and layering loading scheme with low leakage to reduce the power gradient between fuel modules and flatten power distribution of the reactor core. Leaving out burnable poisons, three MOX modules with different Pu contents are used to control the power peak factor of the mixed reactor core at around 1.77, visibly better than the power peak factor of the original reactor core, which serves to provide a scheme of considerable reference value that introduces the MOX fuel for the 3rd generation of domestically developed pressurized water reactor in China

Research on Calculation Method for Hexagonal Assembly Source

Zheng Zheng, Chen Qichang, Ding Qianxue

2018, 39(4): 11-15. doi: 10.13832/j.jnpe.2018.04.0011

Abstract(208)

Abstract:
In order to satisfy the requirement for Water-Water Energetic Reactor (VVER) or fast reactor shielding calculation, Discrete Ordinate (SN) and Monte Carlo (MC) hexagonal assembly source calculation codes are developed and validated on VVER. Numerical results show that relative errors are less than 30% for three group neutron and photon fluence rate distribution from the inner surface of the barrel to the outer surface of the reactor pressure vessel, which demonstrates that the theoretical model and codes in this paper are correct and can be used on VVER or fast reactor shielding calculation

Investigation of Flow and Heat Transfer for CAP1000 Reactor Internals with Direct Safety Injection

Weng Yu, Wang Haijun, Wang Haitao, Zhang Ming, Feng Shaodong, Gu Hongfang

2018, 39(4): 16-21. doi: 10.13832/j.jnpe.2018.04.0016

Abstract(313)

Abstract:
For CAP1000 reactor, the coolant is injected in the reactor vessel under the accident condition through the method of direct injection. Because of the arrangement of the injection nozzle and the internals, the low temperature water has great effect on the reactor internals. The surface heat transfer capability of the reactor internals are studied in this paper. The surface temperature and heat transfer coefficient of the neutron shielding panels and the radiation surveillance capsules are investigated by scaled experiment and the numerical method under different injection conditions. The study found the dimensionless correlation between the wall surface temperature (and the heat transfer coefficient) and the injection flow rate conditions at several key points in the dangerous areas such as the top of the neutron shielding panel.

Experimental Study on Flow and Heat Transfer during RefloodingProcess in a Vertical Annual Channel

Wang Jinyu, Wang Jun, Zan Yuanfeng, Huang Yanping

2018, 39(4): 22-27. doi: 10.13832/j.jnpe.2018.04.0022

Abstract(118)

Abstract:
Visual observations of the bottom reflooding process in an annular channel were carried out. From the visual study of the flow and heat transfer during reflooding phase, the flow regime and heat transfer mechanism transformations were studied. The comparison of phenomena during the reflooding under various conditions shows that the existence of the internal heat source in the heating rod has little effect on the characteristics of the flow and the heat transfer during the reflooding and that the inlet mass flow flux is a key factor that affects flow regimes and heat transfer characteristics. The greater the mass flux is, the easier the flow regime transition happens.

Numerical Simulation on Flow Control Performance of Nuclear Venturi Tube Based on FLUENT

Hu Xiaodong, Liu Fuyu, Wang Yanhui, Liang Zhixi

2018, 39(4): 28-32. doi: 10.13832/j.jnpe.2018.04.0028

Abstract(190)

Abstract:
The cavitatingflow occurred in a nuclear power venturi tube was numerically simulated based on Realizable k-? turbulence model and Zwart (Zwart-Gerber-Belamri) cavitation model. The flow rate curves and properties inside venturi tube was analyzed and cavitation area was predicted effectively under the operated condition. The steady flow rules of the given venturi tube were studied according to the CFD simulation results. It can be concluded that: the choke flow happened in the venturi tube with the inlet pressure increasing; the flow rate changes in venturi tube could be controlled in a certain range with the flow rate increase of 0.06 m3·h^-1 with the inlet pressure increasing 0.1 MPa; throat diameter determines the cavitation inception and the increment of flow rate directly; the larger throat diameter causes the faster and higher development degree of the cavitating flow, meanwhile, the increasing of flow rate in the tube is slower with small throat diameter.

Development of Calculation Code CPFP for Fission Product in Primary Loop of Pressurized Water Reactor Nuclear Power Plant

Tang Shaohua, Lyu Weifeng, Xiong Jun, Jiang Zhenyu

2018, 39(4): 33-38. doi: 10.13832/j.jnpe.2018.04.0033

Abstract(219)

Abstract:
Base on the research of the generation and migration of fission product in the primary loop, the calculation model for the fission product in the primary loop is established, and the calculation code CPFP2.0 with a good man-machine interface is developed. The code CPFP2.0 is tested by the similar codes in other countries and the engineering data from EPR nuclear power plants. The test results show that tolerance scope between CPFP2.0 and the similar codes in other countries is within the acceptable range of engineering practice for tested case, and the difference between calculation result of CPFP2.0 and the engineering data of EPR nuclear power plant is negligible.

Status Assessment of RPV Irradiation Embrittlement

Zhu Guangqiang, Wei Yanhui

2018, 39(4): 39-42. doi: 10.13832/j.jnpe.2018.04.0039

Abstract(188)

Abstract:
Based on a large number of similar irradiation embrittlement test data and actual radiation monitoring test data, a formula for evaluating the irradiation embrittlement of the reactor pressure vessel (RPV) of a nuclear power plant is selected using the method of statistical analysis. The irradiation supervision tube test data of the nuclear power plant has been completed as input data, and the present irradiation embrittlement status of RPV was evaluated. The structural integrity of RPV at the end of life was calculated and analyzed. Based on the calculation results of irradiation embrittlement, the P-T curves of RPV in each operation stage are plotted, and the operation suggestions are given.

Experimental Study on Distribution Characteristics of Flow Reversal in U-Tubes of Steam Generator under Natural Circulation

Tang Yu, Xu Jianjun, Xie Tianzhou, Zhou Huihui, Huang Yanping, Tan Shushi

2018, 39(4): 43-47. doi: 10.13832/j.jnpe.2018.04.0043

Abstract(207)

Abstract:
An experiment on flow reversal in U-tubes of steam generator (SG) has been made within the range of flow rate 0.4~0.7 kg/s under natural circulation. Nine pairs of monitoring points corresponding to nine groups of U-tubes with different length were installed to obtain the distribution characteristics of flow reversal. Based on the conservation laws and test data, total flow rate and the number of U-tubes of reverse flow in the primary side of SG were obtained. The results show that the flow reversal occurs in about 61% of U-tubes, which reduce the flow area of heat transfer tubes to 39%. Because of the flow reversal, the flow rate of the forward flow rise by 60%, and the speed of the flow in U-tubes increase to 4.2 times in contrast with that without flow reversal.

Theoretical Research on Critical Heat Flux of Vertical Tube Based on Bubbly Crowding Model

Chen Sen, Yang Ning, Li Huaqi, Zhu Lei, Hu Pan, Ma Tengyue, Chen Lixin

2018, 39(4): 48-52. doi: 10.13832/j.jnpe.2018.04.0048

Abstract(229)

Abstract:
To study the single pipe CHF phenomenon of the bubble crowding, the CHF calculation model was developed based on the near wall bubble crowding mechanism model. The CHF was obtained by solving the mass, momentum and energy equation together with bubbly detaching model and critical void fraction model in the near wall. The results calculated by the model were compared with the experimental data and a good agreement was achieved. Based on the CHF calculation model, the CHF varing with the inlet enthalpy difference, mass flow rate, diameter and length was studied, which could be used to predicting the bubbly crowding CHF.

Natural Circulation Characteristics of Integrated Reactor under Perturbed Condition

Kong Song, Yu Lei, Hao Jianli, Yuan Tianhong, Shen Mengsi

2018, 39(4): 53-57. doi: 10.13832/j.jnpe.2018.04.0053

Abstract(244)

Abstract:
Based on a multiple cooling loops integrated reactor, we derived the theoretical solution of the natural circulation flow of multi-loop circuit. The perturbation coefficient m are obtained by using approximation and fitting approach, and the perturbation characteristics of the natural circulation system is analyzed and verified on an integrated reactor. The results indicate that for a symmetrical loop consisted of multiple cooling loops, the change of the flow is a function of Reynolds number Re and the ratio of the rising resistance characteristics and the falling resistance characteristics R_r.The perturbation coefficient m has a close relation with Re and R_r, and increases when R_r decreases. The greater the m, the smaller the effect of the same perturbation on the natural circulation system, that is to say, m can represent the anti-perturbation ability of the system.

Flow Distribution of Heat Exchanger Tubes in a Steam Generator and Its Effect on Flow Field at Entrance of Reactor Coolant Pump

Ma Tengyue, Wang Pengfei, Xu Zhongbin, Ruan Xiaodong, Kong Weijie

2018, 39(4): 58-62. doi: 10.13832/j.jnpe.2018.04.0058

Abstract(169)

Abstract:
For the study of steam generator (SG) heat exchange tube flow distribution and its effect on the reactor coolant pump (RCP) entrance flow, the scale model experiment in the cold state of the SG was conducted. The experimental data was the new inflow condition of the SG channel head. The three-dimensional flow field of the SG channel head was calculated by CFD method. The results show that the SG heat exchange tubes have serious uneven flow distribution. The SG entrance will have a greater impact on the flow distribution of the heat exchange tube, especially the tubes toward the entrance. The partial flow rate increases and the formation of high speed creates a relatively low velocity zone and the recirculation zone around itself; when the flow rate is small, the distance head of the heat exchange tube plays a leading role; the uneven flow distribution in the heat exchanger tubes of the SG will cause the axial velocity at the outlet of the SG to be even more turbulent.

Experimental Study on Velocity of Quench Front during Reflooding Process in a Vertical Annual Channel

Wang Jinyu, Wang Jun, Zan Yuanfeng, Huang Jun

2018, 39(4): 63-66. doi: 10.13832/j.jnpe.2018.04.0063

Abstract(258)

Abstract:
Velocity of quench front is one of the important parameters which measure the cooling efficiency of the reactor core during the reflooding process after loss of coolant accident. Based on an experimental study of propulsion characteristics of quench front in a vertical annual channel, the effects of initial clad temperature, inlet coolant temperature, inlet mass flow flux, and heating power on the velocity of quench front were investigated. The results show that the velocity of quench front decreases with the increasing of the initial wall temperature, inlet coolant temperature as well as heating power, but increases with the increasing of the inlet mass flow rate.

Development of Analysis Code for Pb-Bi Cooled Direct-Contact-Boiling Water Fast Reactor System

Wei Shiying, Wang Chenglong, Su Guanghui, Tian Wenxi, Qiu Suizheng

2018, 39(4): 67-70. doi: 10.13832/j.jnpe.2018.04.0067

Abstract(262)

Abstract:
In this paper, mathematical models are established for PBWFR coolant system, and a system analysis code of PBWFR SACOL is developed. The steady-state and transient thermal-hydraulic performance of PBWFR is calculated and analyzed, especially the unprotected transient of over power. Results show that PBWFR is safe under steady state. However, during UTOP transient, the rapidly increasing of power in a short time would lead to cladding failure.

Technology for Examination of Failed Fuel Element in Hot Cell

Liu Xiaosong

2018, 39(4): 71-74. doi: 10.13832/j.jnpe.2018.04.0071

Abstract(455)

Abstract:
Fuel assembly damage directly influences the safe operation of the reactor. Analyzing the cause of fuel assembly damage is the key process of the fuel assembly development and improvement. This paper established the thermal chamber inspection method for the damaged fuel assembly, by studying the key technologies such as underwater collapsing, localizing the break and taking crevasse samples of the fuel assembly. The study shows the technical route is rational, and the method is practicable, which have provided technical approaches to check the damaged fuel assembly in a hot cell.

Immersion Ultrasonic Inspection for Expanding Area of Steam Generator Tubes

Liu Yun, Lu Wei, Ding Song, Yin Peng, Yang Jinli

2018, 39(4): 75-79. doi: 10.13832/j.jnpe.2018.04.0075

Abstract(239)

Abstract:
In recent years, it is required to inspect the expanding area of steam generator tubes in nuclear power plants. There is a blind area for conventional eddy current testing with Bobbing probes in the expanding area, so immersion ultrasonic testing from the inside of the tubes was developed. The experiment results showed that, the immersion ultrasonic testing can meet the need of inspection for expanding area of steam generator tubes. The immersion ultrasonic inspection can be a supplement of eddy current inspection, and at the same time, it can be applied in other thin-wall and small diameter tubes.

Reasearch of Ultrasonic Quality Inspection Technology for Berylium Copper Welder Workpiece

Wang Xuequan, Ren Junbo, Liu Jian, Chai Yukun, Yang Lei

2018, 39(4): 80-82. doi: 10.13832/j.jnpe.2018.04.0080

Abstract(188)

Abstract:
Berylium copper welder workpiece is the cladding material of nuclear fusion reactors, working under bad condition, and its welding quality needs to be inspected. In order to realize the inspection of the quality of the berylium copper welding, firstly the theoretical analysis has been done for the feasibility of ultrasonic testing. Special ultrasonic probes have been designed and special sensitivity test block was designed and processed. The inspection method and inspection technology have been studied. Through the comparison of the experimental data of the American counterparts, the correctness of the quality ultrasonic inspection method for berylium copper welding is verified. This method is versatile and applicable in other workpieces.

Study on Ultrasonic Testing Technology for Conical Cylinder Weld of Steam Generator in Nuclear Power Plants

Ren Jianbo, Sun Jiawei, Xu Yaoning, Ma Guanbing, Wu Jinfeng

2018, 39(4): 83-86. doi: 10.13832/j.jnpe.2018.04.0083

Abstract(171)

Abstract:
This paper introduced the ultrasonic inspection technology for the nuclear power plant steam generator（SG）shell and cone butt welds. By means of sound beam angle correction and sound path localization method, the problem of defect location in SG cone inclined plane scanning was solved, and the calculation formula for defect location was derived. The changes of probe angle resulted from the sound velocity variation and the signal location at the depth direction of two materials with different acoustic characteristics are compared, and the signal detected was analyzed by adjusting the sound velocity. It is found that the signal was generated by the weld structure, and the error in locating the defect depth resulted from the difference of sound velocities between the standard test block and the reference test block was solved.

Study on Stability Analysis Method of Support Steel Structure in AP1000 Plant

Wang Liang, Shen Le, Liu Kang, Xiao Chaoping

2018, 39(4): 87-92. doi: 10.13832/j.jnpe.2018.04.0087

Abstract(227)

Abstract:
At present, the mechanical design of non-nuclear steel support in AP1000 plant follows the Specification for Structural Steel Buildings (AISC 335 and AISC N690), in which the stability analysis adopts the Effective Length Method （ELM）. While the latest version of Specification for Structural Steel Buildings (ASIC 360-2010) prefers the Direct Method （DM）, and ELM as its alternative. In addition, compared to AISC 335-1989, AISC 360-2010 explicitly requires the consideration of the nonlinear second order effects and initial defects in the stability analysis of the structure support. This paper described the requirements of stability analysis in AISC 360-2010, and the characteristics of DM and ELM. Then using a steel support frame as an example in GTStrudl, the research and comparison of the two methods is conducted. For simple structures, the two methods are applicable; for complex structures, Direct Method is simpler and more efficient.

Direct Input Method of Seismic Acceleration Time History for Nuclear Power Plant Reactor Analysis

Wu Shijian, Shang Ertao, Jin Ting, Liu Pan, Nie Zhaoyu

2018, 39(4): 93-96. doi: 10.13832/j.jnpe.2018.04.0093

Abstract(345)

Abstract:
Based on ANSYS12.1, this paper studies the direct input method of a seismic acceleration. This method can avoid the numerical error produced by two times of numerical integral and also avoid the Rayleigh damping force produced by large mass method. Through the theoretical analysis and numerical calculation, the direct input method of a seismic acceleration is checked and compared with the direct displacement excitation method, the uniform acceleration excitation method and the large mass method. The results show that the direct input method of a seismic acceleration can be applied in practical engineering calculation.

Design and Analysis of Control Rod Guide Assembly for China Small Modular Reactor

Zhang Hongliang, Chen Xungang, Luo Ying, Xu Bin, Liu Xiao, Wang Liubing

2018, 39(4): 97-100. doi: 10.13832/j.jnpe.2018.04.0097

Abstract(275)

Abstract:
An open and bundled control rod guide assembly with long stroke and continuous guidance was proposed, by analyzing the structure characteristics of China small modular reactor (ACP100). The components are analyzed by stress analysis, flow field analysis and experiment verification, and the results show that the components are with small friction, strong resistance to deformation and flow disturbance, and is reliable in operation, which can meet the functional requirements of ACP100.

Reliability Analysis of Passive System Based on PSO Optimized Neural Network Response Surface Method

Ding Hao, Cai Qi, Zhang Yongfa, Jiang Lizhi, Wei ke

2018, 39(4): 101-106. doi: 10.13832/j.jnpe.2018.04.0101

Abstract(178)

Abstract:
On the basis of reliability analysis mathematical model, combined with the operating data from an experimental facility and improved thermal-hydraulic codes, the uncertainty of input parameters is identified. Compared with the accuracy and the goodness of different Neural Network Response Surface methods, the one optimized with PSO is analyzed by classification accuracy. The results show that PSO response surface has relatively better fitting performance and can evaluate the reliability of the passive system accurately.

Application of Parametric Empirical Bayes Method in Frequency Statistics of Initiating Event

Yang Jian, An Jin

2018, 39(4): 107-111. doi: 10.13832/j.jnpe.2018.04.0107

Abstract(193)

Abstract:
The frequency of loss of offsite powerv(LOOP) initiating event in nuclear power plants (NPPs) in China was estimated by parametric empirical bayes (PEB) method in this paper. The prior distribution (overall distribution) of this initiating event frequency and the posterior distribution of each unit were obtained. Comparison with other commonly used methods-maximum likelihood estimate (MLE), Jeffreys noninformative prior shows that: PEB method can handle different data sources to form a larger sample. In the process of generating a prior distribution, PEB method can reflect the differences between different data sources. Therefore, it can be considered that PEB method has significant advantages in general data processing.

Fault Analysis and Treatment of EAS Sump Line Isolation Valve

Yi Quanwei, Li Guangjun

2018, 39(4): 112-115. doi: 10.13832/j.jnpe.2018.04.0112

Abstract(196)

Abstract:
The EAS sump line isolation valve EAS013/014VB is often unable to open or close smoothly in the progress of commissioning tests on site. Based on online diagnoses and commissioning tests, this paper analyzes the possible reasons for the fault and finds out that the main reason is the electrical actuator. The valve modification scheme is proposed. The result of valve modification shows that the cost of valve purchase and commissioning can be reduced and the probability of valve failure can be reduced.

Analysis of Alleviation Capability of Emergency Cooling System in Steam Generator Heat Transfer Pipe Damage Accident

Cai Meng, Wang Wei, Sun Junzhong, Yuan Jiangtao

2018, 39(4): 116-122. doi: 10.13832/j.jnpe.2018.04.0116

Abstract(200)

Abstract:
Considering the critical consequences of the steam generator pipe rupture accident, in order to improve the understanding of emergency cooling system to alleviate the accident consequences, and enhance the disposal capacity in the heat transfer tube breakage accident, the safety analysis model based on MELCOR program is built in this paper. The effect of the emergency cooling system on accident consequences is calculated comparatively, and the alleviation capacity of the emergency cooling system in the accidents of for the heat transfer tubes with different sizes is compared. The simulation analysis verified the alleviation capacity of the emergency cooling system in the failure of the heat transfer tubes, and this study has great significance to improve the accident disposal ability of the operation staff, and to guarantee the safety of nuclear reactors.

Study on Optimization Method for Periodical Tests of CPR1000 Nuclear Power Unit

Yang Bo, Zhang Zhao, Zhao Fuyu, Shen Rongfa

2018, 39(4): 123-127. doi: 10.13832/j.jnpe.2018.04.0123

Abstract(158)

Abstract:
Aiming at the problems of periodical tests for high safety significant systems in CPR1000 nuclear power unit, the optimization method for periodical tests for CPR1000 is proposed based on the methodology of periodical test design and safety analysis. For the periodical test of the classic instrument alarming, i.e., the residual heat removal system is not isolated and the reactor coolant system is with high pressure, the problems were analyzed and the optimization scheme is proposed. According to the optimization method for periodical tests presented, the safety of the optimization scheme is analyzed, which shows that the scheme is correct and valid. The optimization method for the periodical test of CPR1000 Nuclear Power Unit is beneficial to the nuclear safety.

Study on Spent Fuel Storage Canister Cutting and Content Withdrawal Technologies

Hu Dongmei, Dai Bo, Zhuang Qianping, Li Zhengbin, Yin Gaoxiang

2018, 39(4): 128-131. doi: 10.13832/j.jnpe.2018.04.0128

Abstract(390)

Abstract:
How to withdraw the spent fuel assemblies from the spent fuel storage canister is a key issue in the reprocessing for the spent fuel with dry and wet storages. Aiming at this issue, based on the structure of the spent fuel storage canister, technologies for spent fuel storage canister cutting and content withdrawal was studied. Considering the hot-cell circumstance, the collecting and transferring of the contents after withdrawal, the collecting and disposal of the generated wastes, the feasible technology of cutting and withdrawal was established, and the storage canister cutting device and special tools was developed. Finally, the feasibility of cutting and withdrawal technology and the usability of cutting device and special tools were validated.

Design of an Electronic Gun Transport Device in a Nuclear Radiation Environment Based on QFD and FPBS

Liao Ying, ZhaoWu, Chen Ling, Zhao Haijiang, Zheng Lanjiang

2018, 39(4): 132-136. doi: 10.13832/j.jnpe.2018.04.0132

Abstract(248)

Abstract:
The methods of QFD (Quality Function Deployment) and FPBS (Function- Principle-Behavior-Structure) are combined to improve and obtain the innovative design process for the engineering device. Under the guidance of the process, facing the requirements of the nuclear radiation environment and the explicit expressing of tacit knowledge based on cognition, an electron gun transport device is designed to meet the needs of the project through the design of material, structure and control mode . The device satisfies the basic requirements of the nuclear radiation environment, and the reliability of the innovative design process of this engineering device is verified.

Optimal Analysis of Mixing Performance of In-Containment Refueling Water Storage Tank in a Nuclear Power Plant

Hou Jianfei, Xia Yuzhuo

2018, 39(4): 137-140. doi: 10.13832/j.jnpe.2018.04.0137

Abstract(142)

Abstract:
A CFD method is used to simulate the flow field of double loop In-containment Refueling Water Storage Tank (IRWST) in a nuclear power plant, and an uniformity criterion number and volume ratio of low velocity region is adopted to quantitatively evaluate the mixing performance of IRWST. Results indicate that the comprehensive mixing performance of double loop IRWST is better than European Pressurized Water Reactor (EPR) nuclear power plant by reasonable arrangement of mixing pipes, which ensures the availability in engineering application. Further two structural optimizations are proposed, and the results show that the mixing performance of IRWST can be improved by reducing the diameter and adjusting the direction of mixing pipes.

Experimental Study on Separating Efficiency of Heat Pump Evaporating System in Nuclear Power Plant

Liu Yong, Huo Ming, Lan Lijun, Sheng Cheng

2018, 39(4): 141-143. doi: 10.13832/j.jnpe.2018.04.0141

Abstract(318)

Abstract:
According to the design requirements of heat pump evaporating system of nuclear power plant, a set of testing installation was built and corresponding performances and efficiencies of separating process were researched. The results show as follows: the reflux ratio of evaporation and concentration in concentrate has low impact on separating efficiencies of the testing installation. The boron concentration in distillate is less than 0.2 mg/L, which is much lower than the most strict limit value (2 mg/L) of releasing to the environment. The results of radioactive alternative test by using MgSO₄, show that the decontamination factor of the test installation reaches 10⁵. Generally, the performance of separating and decontaminating of test installation is much better than similar devices as known, which provided the feasibility of near-zero releasing for inland nuclear plant.

Commissioning and Performance Appraisal of Cement Solidification Line Intermittent Out-drum Mixing Equipment

Yu Dawan, Zhang Jiaheng, Jiang Jianqi, Yang Yongliang

2018, 39(4): 144-147. doi: 10.13832/j.jnpe.2018.04.0144

Abstract(154)

Abstract:
During the commissioning of the Fangjiashan cement solidification line, the concentrate, spent resin solidification and spent filter fixation tests were conducted. The problems such as cement solidification formulations, equipment design, and equipment reliability were found in the tests. The cause for the problems is analyzed. The formula suit for out-drum mixing, mud fluidity, solidification performance and waste minimization should be developed. The equipment reliability problem should be solved from the aspects of equipment design, equipment selection and procurement. The engineering validation should be performed fully before application to the nuclear power plants to improve the reliability of the out-drum mixing process.

Equipment Design and Management for First Unit of HPR1000

Sun Zhan, Jiang Pingping, Zhang Wenguang, Jia Yanbo, Li Jiuzhen

2018, 39(4): 148-151. doi: 10.13832/j.jnpe.2018.04.0148

Abstract(232)

Abstract:
In the engineering design and management of the first unit of HPR1000, the design plan, the management method integrating the design planning, design interface and design procurement are adopted. Through the implementation of the opening items, risk management and other innovative mechanisms, each equipment design task is completed on schedule, to guarantee that the construction of the first unit of HPR1000 is on schedule.

Research and Development of a Management Information System for Standardization of Safety Production in an Operating Nuclear Power Plant in China

Sui Yang, Ding Rui, Wang Hanqing

2018, 39(4): 152-156. doi: 10.13832/j.jnpe.2018.04.0152

Abstract(133)

Abstract:
In order to solve the problems of how to share the information from the standardization of safety production how to improve the working efficiency, and how to statistically analyze a large amount of data from standardization of safety production in implementing the standardization of safety production for the operating nuclear power plants in China, the management information system for the standardization of safety production for an operating nuclear power plant was designed. The application results show that the management information system for the standardization of safety production for an operating nuclear power plant helped to increase the information transmission among departments and share the information from standardization of safety production, to optimize the management process and increase the working efficiency for standardization of safety production, and to statistically analyze a large amount of data from standardization of safety production in an operating nuclear power plant.

Dosimetric Characterization of a IBA Stereotactic Field Diode Detector in Radiation Small Photon Beams

Chang Xue, Wang Kun, Zhang Jian, Wang Zhipeng, Jin Sunjun

2018, 39(4): 157-160. doi: 10.13832/j.jnpe.2018.04.0157

Abstract(261)

Abstract:
The dosimetric properties of the IBA stereotactic field diode in radiation small photon beams are investigated. Dose rate dependence, output factors, lateral field profiles, and percentage depth dose profiles are measured and compared with the measurements performed with the PTW 60019 synthetic diamond detector. The IBA SFD has a larger over-response with increasing dose rate than the PTW 60019 synthetic diamond, the relative difference is up to 0.7%. For 1 cm×1 cm field size, the response of the IBA SFD is higher than that of the PTW 60019 synthetic diamond around the dose maximum of the percentage depth dose profiles, with a maximum relative difference of 2.1% and slightly lower at greater depths, with the relative difference below -4.2%. In the penumbra region of the lateral field profiles, the measurements of the IBA SFD is larger than that of the PTW 60019 synthetic diamond, the relative differences tend to be higher with a maximum deviations of 15.2% in the 1 cm×1 cm field size. For field sizes larger than 1 cm×1 cm, the output factor measured by the IBA SFD is slightly lower than that of the PTW 60019 synthetic diamond, the relative difference is less than -1.3% and the relative difference is up to 2.5% for 1 cm×1 cm field size. The IBA SFD is not recommended to use for small field dosimetry.

Research and Application of CAD/CAE Integration in Analysis of Supports in Nuclear Power Plants

Xiao Chaoping, Liu Kang, Shen Le, Liang Mingbang

2018, 39(4): 161-164. doi: 10.13832/j.jnpe.2018.04.0161

Abstract(222)

Abstract:
To overcome the shortcomings of the traditional analysis method of supports in nuclear power plants in terms of quality and efficiency, the automatic identification algorithms of welding joints and loading joints of steel structure were solved, based on in-depth research on the topological and geometric characteristics of CAD models of various supports. CAD / CAE integration software with independent intellectual property was developed by PML (Programmable Macro Language). It not only solved the problem of topological connectivity identification in the process of data exchange between CAD and CAE software and realized the pre-processing automation of CAE models of various supports in nuclear power plants, but also was successfully applied to a nuclear power project and obtained remarkable economic benefits.

Radiation Shielding for CMOS APS Digital Module

Xu Shoulong, Zou Shuliang, Peng Cong

2018, 39(4): 165-170. doi: 10.13832/j.jnpe.2018.04.0165

Abstract(141)

Abstract:
The study on the radiation resistance of complementary metal oxide semiconductor (CMOS) active pixel sensor (APS) digital module are presented, and a shielding structure has been designed and manufactured with the help of Monte Carlo simulation software. Shielding performance, operating life and damage modes of sensor modules have been analyzed. The result shows that with the shielding, the operating life of APS almost doubled. The irradiation dose rate is about 1/3 times of that without shielding when the main board is shielded, but the service life of the main board is only increased by 1 times after shielding. This may be originated from the differences in the radiation resistance and total ionizing dose of the devices on the module. The dark current is almost unchanged when the total dose is less than 50 Gy, and the dark current of all pixels in the sensors increases gradually when the total dose is greater than 150Gy.

Laser Doppler Measurement of Flow Field in a 5×5 Rod Bundle with Mixing Vane Grids

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

2018, 39(4): 171-175. doi: 10.13832/j.jnpe.2018.04.0171

Abstract(200)

Abstract:
The Laser Doppler Velocimetry (LDV) system is applied to measure the axial flow velocity on four cross sections downstream the 5×5 rod bundle with two mixing vanes. Mean velocity and RMS velocity on each cross section is obtained and by comparing the mean velocity and RMS velocity on different cross section, the distribution pattern downstream the mixing vane grids is analyzed. By comparing the experiment data downstream the two mixing vane grids, the effect of upstream flow field on the grid effect is analyzed. All experiment data is acquired based on sufficient repetitive experiment, and the data can be used for the validation and assessment of CFD computation results.

Design Optimization of Thermal Shock-Resistance Double-Cone Seal Structure of Pressurizer

Chen Cong, Wu Ge, Fu Xiaolong, Tang Bin, Zheng Hongtao, Li Pengfei, Wang Yue

2018, 39(4): 176-181. doi: 10.13832/j.jnpe.2018.04.0176

Abstract(193)

Abstract:
Based on the temperature field simulation and sealing performance analysis of the double-cone seal structure, the change of temperature distributions and contact stress of sealing surface along with time were systematically studied. The results indicated that the temperature impact caused by the radical change of temperature was the main reason for the leakage of the double-cone seal structure. To improve the adaptability of the seal structure to temperature changes, a thermal shock technology of the double-cone seal structure is proposed, i.e., a simple and easy-to-installation heat-shock shield is applied to the double-cone seal structure to improve its capacity of resisting the thermal shock. The analysis showed that, this technology can effectively decrease the thermal shock effects on the sealing surface, greatly restrain the decreasing trend of the contact stress, and improve the reliability of the double-cone seal structure.

Numerical Study on Effect of Clamping Failure on Flow Induced Vibration and Fretting Wear Analysis of Fuel Rods

Qi Huanhuan, Feng Zhipeng, Jiang Naibin, Huang Qian, Wu Wanjun, Huang Xuan, Jiang Xiaozhou

2018, 39(4): 182-186. doi: 10.13832/j.jnpe.2018.04.0182

Abstract(205)

Abstract:
Flow elastic stability and vortex shedding were two important mechanisms for the flow induced vibration analysis. Due to the effects of manufacturing process, transportation and irradiation, the clamping action of grid on fuel rods may be invalid. Taking Type I and II fuel assemblies as an example, the effects of clamping failure on the natural frequencies, mode shapes, flow elastic stability and vortex shedding were studied. The results show that the effect of the dimple support failure on the natural frequency was directly related to the mode shape. The effect of the grid dimple failure near the original amplitude on the natural frequency was obvious. The effect of dimple failure on the natural frequency of two fuel rods was similar. For Type I, the velocity of flow at the top and bottom of the fuel rods were larger and the size was comparable. The dimple failure of the top and bottom grids had a significant effect on the stability of flow elastic and the vortex shedding ratio. For Type II fuel assembly, the dimple failure of the top grids had a more significant effect on the stability of flow elastic and the vortex shedding ratio comparing to the bottom grid failure.

Application of Digital Reactor Technology in Reactor Design

Fang Haoyu, Li Qing, Gong Zhaohu, Chen Cheng, Chai Xiaoming, Lu Zongjian

2018, 39(4): 187-191. doi: 10.13832/j.jnpe.2018.04.0187

Abstract(305)

Abstract:
To solve the problems such as low design efficiency, low resource integration and limited innovation ability in the nuclear reactor design, the implementation concept for digital reactor technology is proposed. Based on the application study of the digital reactor technology in R&D and design, it is concluded that the establishment of a flexible and expandable framework of digitalized basic platform and a three-dimension collaborative design system based on system engineering, and the implementation of the big data Management based on knowledge engineering are the most important research topics during the application of digitalization technology in the reactor engineering design.

2018 Vol. 39, No. 4