核动力工程

Development Status and Outlook for Nuclear Power in China

Zhao Chengkun

2018, 39(5): 1-3. doi: 10.13832/j.jnpe.2018.05.0001

Abstract(1246)

Abstract:
This paper introduces the basic situation and the latest progress of nuclear power units in China, as well as the related measures of improving the safety level of nuclear facilities. Under the requirement of the “national energy administration's 13th five-year plan for energy technology innovation”, China has launched a series of development plans for nuclear power and small modular reactors, carryed out the demonstration construction project of marine nuclear power platform, and established relevant standards. At last, it summarizes the current challenges and future outlook for nuclear power in China.

Study on Depletion Calculation of Heat Pipe Cooled Space Reactor

Qu Shen, Cao Liangzhi, Zhou Shengcheng, Liu Hangang

2018, 39(5): 4-8. doi: 10.13832/j.jnpe.2018.05.0004

Abstract(501)

Abstract:
For the heat pipe cooled space reactor, region-dependent homogenized cross sections in the predefined 26 group structure were generated with the OpenMC code based on the R-Z geometric model of the reactor core. The neutron transport calculation was performed with SARAX, which was a deterministic neutronic analysis code for fast spectrum reactors. The calculation results were compared with those obtained with MVP. The generation procedure of the homogenized cross sections was verified and the capability of SARAX for the neutronic analysis of heat pipe reactors was demonstrated.

Physical Analysis of Control Rods Arrangement in Core of a Small Modular Molten Salt Reactor

Kang Xuzhong, Zhu Guifeng, Zou Yang, Yan Rui, Li Minghai, Zhou Bo, Yu Shihe

2018, 39(5): 9-14. doi: 10.13832/j.jnpe.2018.05.0009

Abstract(583)

Abstract:
sm-TMSR is a small modular multi-purpose thorium-based molten salt demonstration reactor designed by Center for Thorium Molten Salt Reactor System (TMSR), CAS. In this paper, the physical analysis of the control rod arrangement was carried out. Firstly, we analyzed the unique reactivity variation in the molten salt reactor and proposed the function and requirements of the control rods. Namely, for regulating rods, the total design value is required to be equal to or slightly greater than 2.5×10^-2 at the beginning of life and 2.06×10^-2 at the end of life, respectively. For shut-down rods, "one stuck rod" criterion is considered, and the design value of a single rod is required to be more than 2.1×10^-2 at the beginning of life. Secondly, we calculated the value of the single control rod at different locations in the core, for different size of the channel and the channel with or without a Hastelloy sleeve. Finally, according to the requirements of control rods, the optimized physical design of the control rod assembly in the core is determined, i.e, a. Diameter of channels in control rod assemblies are 9 cm, and it is without Hastelloy sleeve; b. Four control rods are distributed in the shape of "十", two shut-down rods are uniformly arranged in the first circle and two regulating rods are uniformly arranged in the sixth circle of the core.

Analysis of Xenon Dynamic Characteristics in Primary Loop System of MSR

Zhou Bo, Yan Rui, Zou Yang

2018, 39(5): 15-20. doi: 10.13832/j.jnpe.2018.05.0015

Abstract(567)

Abstract:
A numerical simulation program for the dynamic distribution of xenon with flow and on-line removal function was established for the primary loop system of molten salt reactor (MSR) based on Mathematica7.0. The dynamic characteristics of xenon concentration with time under different flow rates, different startup and shutdown power, and different on-line removal efficiencies were analyzed based on a kind of design scheme of 2MW MSR. The results show that the xenon negative reactivity of the flow burnup model is about 32.2% lower than that of the static burnup model. The xenon concentration distribution in the primary loop system is uniformity under the rated flow rate, and the flow effect can affect the distribution of xenon concentration in the primary loop system only when the volume flow of the primary loop system is less than 2.24 cm³·s^-1.When the online removal fraction of the bubbling system in the pump bowl is about 0.1%, the xenon poison reactivity in the core can be reduced to -38.3 pcm, and the total removal efficiency can reach 86.0%. Xenon concentration in the reactor decreased monotonously under different instantaneous shutdown power and about +254 pcm reactivity was insertion accordingly. Xenon poison is disappeared after 50 hours of shutdown. The removal efficiency of xenon has an influence on the total xenon in the primary loop system. In the range of the removal fraction from 0.0001%~20%, the total xenon increases by 0.67%~8.75% compared with the static burn model. The numerical simulation method and the conclusion are consistent with the actual physical laws.

Study on Load-Following of an AMTEC Space Reactor Power System

Li Huaqi, Zhu Lei, Jiang Xinbiao, Chen Lixin, Shan Jianqiang

2018, 39(5): 21-25. doi: 10.13832/j.jnpe.2018.05.0021

Abstract(581)

Abstract:
This paper investigates the load-following characteristics of an AMTEC converted and heat pipe cooled space reactor power system. The effects of the external load and AMTEC temperature on module AMTEC performance were analyzed, and the response of the AMTEC space reactor power system to a load demand change and load-following characteristics were also analyzed by using TAPIRS code. The results show that AMTEC efficiency and output electric power both firstly increase to maximum value, and then decrease with external load resistance. There is a certain external load value which makes the efficiency or electric power up to peak value, but the two certain values are different. Although the space reactor is with the characteristic of load following, the AMTEC space reactor power system only follows the load when the external load demands exceed the critical value, otherwise the system has no response to the load change.

Study on Burnup Calculation of CENTER Fuel Assembly Irradiation Test in HFETR

Liu Shuiqing, Yang Bin, Kang Changhu, Ma Liyong, Liang Guangyuan, Ran Zhongkang

2018, 39(5): 26-28. doi: 10.13832/j.jnpe.2018.05.0026

Abstract(574)

Abstract:
CENTER fuel assembly irradiation test should be carried out before the fuel assembly is formally designed, and the CENTER fuel assembly would be irradiated in HFETR to complete the fuel burn-up test. To ensure the success of the irradiation test and meet the test requirements, the accuracy of power calculation and measurement of fuel assembly in the process of irradiation test should be studied. According to the characteristics of CENTER fuel assembly, the mosaic coupling method was studied and determined. The fuel assembly was irradiated in HFETR, and the test is conducted. The difference of the fuel burn-up between calculation and measurement is 3.25%, and the burnup of CENTER fuel assembly irradiation test meets the test requirements.

Preliminary Study on Physical Characteristics of CANDU6 Reactor Using Discrete Thorium-Uranium Fuel Pins

Deng Nianbiao, Yu Tao, Xie Jinsen, Zhao Wenbo, Xie Qin, Chen Zhenping, Zhao Pengcheng, Liu Zijing, Zeng Wenjie

2018, 39(5): 29-33. doi: 10.13832/j.jnpe.2018.05.0029

Abstract(316)

Abstract:
In order to study the application of thorium uranium fuel in the CANDU6 reactor, the code DRAGON/DONJON is adopted to study the time-averaged equilibrium core of CANDU 6 with37-element bundle assembly of discrete thorium uranium fuel rods. The results show that, when the assembly is the uranium rods with 2.5% enrichment of ²³⁵U and the 37-element bundle assembly of thorium rods arranged in the 1, 2 and 3 layers, and under the refueling scheme of 8 bundles with 3 burnup partition, the coolant void reactivity of the assembly is lower than that of the 37-element bundle assembly (NU-37 assembly) with natural uranium and the assemblies containing mixed thorium uranium pins; that the max time-averaged channel/bundle power of the core respects the limits (less than 6700 kW/860 kW); that the fuel conversion ratio is higher than the CANDU6 reactor with natural uranium; and that the discharge burnup is up to 13400 MW·d/t(U). Therefore, the 37-element bundle using discrete thorium uranium fuel is applicable in the CANDU6 reactor, without major modification of the structure of the core and operation modes, and the design of fuel assembly and core provides a reference for the thorium uranium fuel in CANDU6.

Bubble Departure Diameter Prediction Model in A Horizontal Rectangle Heating Mini-Channel

Tian Ye, Huang Wei, Luo Hanyu, Wang Haisong, Li Pengfei, Sun Yan

2018, 39(5): 34-37. doi: 10.13832/j.jnpe.2018.05.0034

Abstract(256)

Abstract:
In order to predict the bubble departure diameter in a horizontal rectangle heating mini-channel to study its heat transfer characteristics, a bubble departure diameter predicting model based on force balance is proposed. A visualization experiment is used to verify it, and the results show that the average relative error between the prediction result and the experimental data is 13.52% , the predicting accuracy of bubble departure diameter model is good.

Study on Mechanical Compensation Control Strategy on 177-Fuel-Assembly Core Reactor

Wang Jinghui, Huang Kedong, Wang Jinyu, Liao Hongkuan, Xiao Peng, Li Tianya

2018, 39(5): 38-42. doi: 10.13832/j.jnpe.2018.05.0038

Abstract(260)

Abstract:
Reactor with 177-fuel-assembly core usually operates in G mode and the core boron concentration requires to be adjusted during the load following. Limited by the capacity of the boron recycle system, the load follow capability is available only for 85% of the lifetime. In order to improve the load-following capability of 177-fuel-assembly core reactor and extend the lifetime of the load following, a mechanical compensation control strategy based on 177-fuel-assembly core was studied. Different arrangements of control rods are designed. The control rod worth, the impact of power peak and the control ability during load follow are analyzed. Based on the optimized arrangement of control rods and the mechanical compensation control strategy, the whole-life base load operation, the end-of-life 90% load cycle load tracking and the starting process simulation are carried out. The results show that under the proper arrangement of the control rods, 177-fuel-assembly core can implement mechanical compensation control strategy, and the load following capability has reached the international advanced level.

Study on Microstructure and Corrosion Mechanism for Two Kinds of Zirconium Alloy

Li Rui

2018, 39(5): 43-46. doi: 10.13832/j.jnpe.2018.05.0043

Abstract(319)

Abstract:
Based on the test results of the domestic tin Zr-4 C zirconium alloy in low alloy autoclave in the pure water and LiOH aqueous corrosion, using the transmission electron microscopy (TEM) observation matrix and oxide film microstructure, through the analysis of oxidation weight data, the corrosion mechanism of C zirconium alloy were studied. Three corrosion mechanisms are proposed: Nb can effectively inhibit the increasing of anion vacancy concentration and reduce the diffusion rate of the oxygen element. The number of deficient traps affects the corrosion caused by oxygen diffusion, and the number of empty traps is proportional to the total surface area of the second phase particles. The second phase of particle oxidation expansion leads to the relaxation of pressure stress of oxide film, which degrades its stability and protective ability.

Research on Preparation Technology of UO₂-Er₂O₃ Fuel Pellets

Liu Yu, Zhang Xiang, Yang Jing, Zeng Qiang, Yu Chong, Li Yuanyuan, Duan Panpan

2018, 39(5): 47-50. doi: 10.13832/j.jnpe.2018.05.0047

Abstract(220)

Abstract:
A preliminary study is conducted for the preparation technology of UO₂-Er₂O₃ burnable poisonous fuel pellets with Er₂O₃ mass fraction of 4.32%. We obtain the preparation technology by comparing the performance (integrity, density, and grain size) of the pellets under different conditions (mixing, pressing, and sintering). The experiments show that UO₂-Er₂O₃ fuel pellets with good integrity which theoretical density≥95% and grain size larger than 8 μm can be obtained by mixing the powders with planetary ball mill for 6 h and add 5‰ PVA, pressing under 300 ~350 MPa, sintering at 1700~1750℃ with hydrogen atmosphere for 2 ~3 h.

Safety Analysis of CiADS Sub-Critical Reactor Fuel Cladding under Beam Transients

Zhang Qingyang, Gu Long, Peng Tianji, Sheng Xin

2018, 39(5): 51-57. doi: 10.13832/j.jnpe.2018.05.0051

Abstract(265)

Abstract:
The response characteristics of the fuel rods in the sub-critical reactor of China Initiative Accelerator Driven System (CiADS) under beam trip were simulated by the reactor system analysis program RELAP5 mod4.0. The fatigue life of the fuel cladding in CiADS under beam trip is calculated by ANSYS 17.0. The fatigue life of the fuel cladding in hundred megawatt Accelerator Driven Sub-critical System (ADS) to be designed in China is predicted. The results show that the power of CiADS subcritical reactor instantaneously falls to 2.156% of full power when beam trip lost. The fatigue life of the fuel cladding of the CiADS subcritical reactor under beam trip is above 10⁸. the beam trip will not cause fatigue damage to fuel cladding in hundred megawatt ADS to be designed in China.

Effect of SiC Addition on Microstructure and Properties of sintered SiC/Zr Composites in Vacuum

Zhang Yanyan, Feng Keqin, Yue Huifang, Zhang Ruiqian

2018, 39(5): 58-62. doi: 10.13832/j.jnpe.2018.05.0058

Abstract(190)

Abstract:
SiC/Zr composites of different composition were prepared by vacuum sintering, using SiC powder and ZrH₂ powder as the raw material. The effect of SiC addition on the microstructure and properties of SiC/Zr composite was studied. The results show that the compact and single Zr metal can be obtained without adding SiC, but the hardness and corrosion resistance are poor. After adding SiC, SiC reacts with Zr, and ZrC and Zr₂Si are produced. SiC particles are bounded well to Zr matrix by interfacial reaction layer, which lead to the higher hardness and corrosion resistance. With the rising of SiC addition , the relative density of the compact and sinter both decrease, while the second phase in the structure increases. The corrosion resistance of SiC/Zr composite is improved and the hardness increases initially and decreases afterwards. When the addition of SiC is 15%, the hardness reaches the maximum value.

Studies on Electrochemical Corrosion Behaviors and 316NG Stainless Steel in Boron-Lithium Solutions

Shu Ming, Wang Conglin, Chen Yong

2018, 39(5): 63-68. doi: 10.13832/j.jnpe.2018.05.0063

Abstract(208)

Abstract:
The electrochemical behaviors of 316NG stainless steel in the boron-lithium solutions (pH=5~8) at 300℃ were studied by using electrochemical polarization curves and the electrochemical impedance spectroscopy(EIS), and the experimental E-pH diagram at the same environment was graphed. The results indicated that the polarization current in the passivation region decreased sharply, especially in the region close to the secondary transpassivation point, and there were three passivation domains when pH was 7~8, The polarization resistance of 316NG stainless steel in the alkaline solution was apparently higher than that in the acidic solution or in the neutral solution, which was found by using electrochemical impedance spectroscopy, showing the oxide film at alkaline environment had a better protective effect. The oxide film was consisted of Cr₂O₃ and Fe₂O₃ in the acidic solution; When pH=6~8, a multilayered oxide film occurred: the outer Fe-rich layer consisted of Fe₃O₄ while NiFe₂O₄ occurred as the depth of layer increased, and the inner Cr-rich layer was composed of FeCr₂O₄. R_sol and the polarization resistance significantly decreased as the conductivity of the solution increased. The E-pH diagram based on the polarization curves was consistent with the inference.

Shock and Vibration Assessment of Aircraft Impact on Nuclear Power Plant Considering the Nonlinear of Impact Zone and Soil-Structure Interaction

Sun Yugang, Cheng Shujian, Li Shuaixi, Ge Honghui, Wang Xiaowen, Yuan Fang

2018, 39(5): 69-74. doi: 10.13832/j.jnpe.2018.05.0069

Abstract(242)

Abstract:
A simplified methodology for assessing the shock and vibration effects of the aircraft impact on nuclear power plant (NPP) is discussed in this paper. Both the force time-history method (FTHM) and the missile-target interaction method (MTIM) are used to assess NPP shock response and its propagation. Then, the effects of both the material nonlinearity and the soil-structure interaction (SSI) on NPP in-structure shock response spectra are presented. Finally, an example of assessing the shock effects on the safety-related system, equipment, and component is provided based on NPP seismic margin assessment. The results show that the maximum displacement and the displacement time history of the NPP impact location obtained from both the FTHM and MTIM are almost the same, but the response spectrum obtained from the MTIM show more high frequency energy than those from FTHM. The shock response will obviously decrease as the shock propagation distance increases, and the difference between the in-structure response spectra obtained from these two methods will also decrease. Both the material nonlinearity and SSI will significantly reduce the NPP shock response, and the response spectrum of the seismic margin assessment can be used as the acceptance criteria for the shock assessment.

Integrity Analysis of Weld Overlay Repair Structure of Upper Ω Seal Weld of Control Rod Drive Mechanism

Lu Zhicheng, Xu Xiao, Wang Dasheng, Liu Pan, Jin Ting, Qiu Zhensheng

2018, 39(5): 75-79. doi: 10.13832/j.jnpe.2018.05.0075

Abstract(570)

Abstract:
Aiming at the weld overlay repair (WOR) for the upper Ω seal of the control rod drive mechanism (CRDM), numerical simulation is applied in the integrity analysis of the overlay repaired structure. 2-D axisymmetric Gauss heat source equivalent input was built according to the welding parameter, and the weld overlay process was simulated by using 'kill elements' technology in ANSYS program, obtaining the welding residual stress of the structure. Transient stresses ware computed considering all transients in the plant, and then the fatigue results ware analyzed. Fracture mechanics analysis was also accomplished using the welding residual stress results and transient stresses results. The analysis results show that the fatigue result, the stress intensity factor and the fracture mechanics results meet the requirements of the standard.

Shaking Table Test of Nuclear Power Plant Considering Uniform Hazard Spectrum

Zhang Xueming, Yan Weiming, Sun Yunlun, Chen Shicai, He Haoxiang

2018, 39(5): 80-84. doi: 10.13832/j.jnpe.2018.05.0080

Abstract(195)

Abstract:
In order to obtain the response spectrum for the specific nuclear power plant (NPP), considering the specific site condition and ground motion parameters, the stochastic simulation and probabilistic hazard analysis is combined to obtain the UHS which the probability of exceedance is 10^-4. In order to verify the seismic performance and applicability of the spectrum in the actual structure, 1:20 NPP model is designed for shaking table test, comparing the structure response of two natural waves and two artificial waves which are generated by the safety evaluation spectrum and RG1.60 spectrum. The result shows that different wave has a different seismic isolation effect. The artificial wave generated by the UHS has a great effect on the acceleration amplification and the displacement of the NPP. The floor response spectrum amplitude which caused by UHS is higher than other waves. It should be considered in the structure and equipment seismic design.

Numerical Comparison Study on Residual Life Specification Prediction Methods of Austenitic Stainless Steel Nuclear Piping with Planar Defects Based on ASME and RSE-M Regulations

Liu Zhenshun, Sun Jinxiong, Zhang Lei, Zhen Hongdong

2018, 39(5): 85-90. doi: 10.13832/j.jnpe.2018.05.0085

Abstract(295)

Abstract:
There were different defects possibly existing in the processing and installation of nuclear piping. In addition, there may be a small amount of defects such as cracks existing in the piping due to the effect of operation conditions in the nuclear power plants. The residual life of nuclear piping with planar defects needs reasonable evaluation and prediction, so as to arrange the replacement program, and to avoid a serious impact on the efficiency of the nuclear power plants. In accordance with ASME and RSE-M codes, the method of residual life evaluation for austenitic stainless steel nuclear piping with planar defects was studied by numerical comparison in stress intensity factor calculation, crack growth analysis and crack stability evaluation in this paper, and the reference experience was provided for similar work .

Research and Application of Vacuum Exhaust Method in Primary Loop of Ling’ao Nuclear Power Plant

Zhang Yingqiang

2018, 39(5): 91-94. doi: 10.13832/j.jnpe.2018.05.0091

Abstract(472)

Abstract:
This paper adopts the method of vacuum exhaust in the primary loop of PWR nuclear power unit. Primary loop of PWR Ling'ao Nuclear Power Plant is vacuum exhausted after refueling overhaul, and the dynamic exhaust process is cancelled. The results show that the requirements of the air content of the primary circuit can be satisfied quickly, shortening the overhaul schedule, and improving the economy and safety of the power station.

An Assessment Model of Operator’s Response Implementation Reliability in Digital Main Control Rooms of Nuclear Power Plants

Li Pengcheng, Li Xiaofang, Dai Licao, Zhang Li, Qing Tao, Jiang Jianjun

2018, 39(5): 95-100. doi: 10.13832/j.jnpe.2018.05.0095

Abstract(197)

Abstract:
In order to quantitatively evaluate operator’s response implementation reliability in digital main control rooms (MCRs) of nuclear power plants (NPPs), main performance shaping factors (PSF) are identified by contextual environment analysis. Weight of the PSF are identified by analytic hierarchy process (AHP). Furthermore, the assessment model of response implementation reliability is established based on the proposed 6 model hypotheses. A case is given to illustrate the specific application of the established method. Finally, the effectiveness of the established method is validated by case and comparative analysis. Results show that the error probability estimation results of the method is consistent with that of CREAM and SPAR-H.

Study on Design of Medium Voltage Mobile Power Supply for Improvement after Fukushima Nuclear Accident

Wang Jin

2018, 39(5): 101-105. doi: 10.13832/j.jnpe.2018.05.0101

Abstract(491)

Abstract:
Current designs of medium voltage mobile power supply, as the improvement item after Fukushima nuclear accident, vary in various nuclear power projects due to different reactor types and site conditions. Based on the engineering practice experience obtained in the design of medium voltage mobile power supply for many reactor types and sites, this paper analyzed and studied the function, capacity requirement and access scheme of the medium voltage mobile power supply. For the nuclear power plants under construction and in the future, it is suggested to adopt the basic access scheme of the outdoor interface box, and at the same time, to set up the medium voltage mobile power supply feeder cabinet.

Research on Controlling Method of Secondary Circuit Water Quality of High Temperature Gas Cooled Reactor

Zhang Ruixiang, Zhao Feng, Peng Weichao, Yao Yao, Wang Zhen, Xu Haosong

2018, 39(5): 106-110. doi: 10.13832/j.jnpe.2018.05.0106

Abstract(184)

Abstract:
Referring to the water quality standard of the pressurized water reactor (PWR) second circuit and the thermal power direct current furnace, combined with the structure and material characteristics of the second circuit of the high temperature reactor, this paper updated the original intake and outlet water standards of the steam generator (SG) in the high temperature reactor and obtained a more reasonable water quality control standard. And then this paper researched three methods of controlling the secondary circuit water quality of the high temperature gas cooled reactor, i.e., the system flushing before unit start-up, the dosing control and the control of operation of condensate water polishing system. In the end, the three methods are discussed in detail, and a scientific and integrated method for optimizing the quality of water and steam in the second circuit of high temperature reactor was presented.

Analysis of In-Vessel Retention Capacity for Marine Reactor Severe Accident

He Yilin, Zhang Fan, Zhang Yangwei

2018, 39(5): 111-116. doi: 10.13832/j.jnpe.2018.05.0111

Abstract(544)

Abstract:
Reactor core melt down accident may result in lower head failure, and molten material may relocate into the cavity, which endangers the safety of personnel and hull. The severe accident integration calculation program called MAAP4 code is adopted to study the effects of low pressure injection systems launching time and flow rate when the corium pool had formed for In-Vessel retention. The results show that two low pressure injection pumps can effectively prevent the failure of pressure vessels and the corium melt can be retained within the reactor when the corium pool had been formed for 1576 s, the low pressure injection systems cannot prevent the pressure vessel failure 2646 s after the corium pool had been formed.

Design Specificities and Functional Verification of Loose Part Measurement System in VVER

Zhu Jun, Zhou Zhengping, Liu Wenchao

2018, 39(5): 117-121. doi: 10.13832/j.jnpe.2018.05.0117

Abstract(172)

Abstract:
The design and structure specificities of loose part measurement system (LPMS) of VVER in Tianwan Nuclear Power Plant, as well as the differences of LPMS with the item NRC RG1.133, are introduced. The difficulties and the negative influence are analyzed based on these differences and the specificities of VVER, when the improvement is conducted. In order to implement the function like what NRC RG1.133 requested, additional tests of LPMS were carried out during the commissioning of Unit 3 in Tianwan Nuclear Power Plant. The signal response of the sensors related to the reactor pressure vessel (RPV) was detected successfully in the tests, and it was proved that the arrangement of the LPMS sensors could meet the design requirement of NRC RG1.133.

Effect of Liquid-Structure Coupling on Rectangular Pool Response Spectrum

Bai Wenting, Feng Guozhong, Jia Lei, Xie Yongping, Chen Huiqin

2018, 39(5): 122-125. doi: 10.13832/j.jnpe.2018.05.0122

Abstract(135)

Abstract:
In order to provide a seismic design foundation to the Equipment and Instrument installed in the pool which are related to nuclear safety, in connection with the rectangular pool related to nuclear safety, study the change regulation of the response spectrum on the pool wall according to a liquid-structure coupling analysis between the inviscid, irrotational and incompressible ideal fluid and pool wall. The results show that, when considering the liquid-structure coupling, the response spectrum of the pool has a significant enlargement in the normal direction, and the farther away from the intersection of the wall, the more obvious the magnification is. It is necessary that the response spectrum be given in the tangential direction and the normal direction of the pool wall, the response spectrum amplification at the intersection site of the pool wall is not obvious, the instrumentation should be placed as close as possible to the corner parts of the pool.

Heat Transfer and Stress / Strain Analysis of PWR RPV Lower Head under IVR-ERVC

Zhang Xiaoying, Liu Fayu, Chen Hu

2018, 39(5): 126-132. doi: 10.13832/j.jnpe.2018.05.0126

Abstract(456)

Abstract:
Taking a 1000 MW PWR as an example, a two-dimensional polar coordinate thermal model was used to analyze the coupling heat transfer among the wall surface of RPV, the two-layer melting core pool and the outer water chamber. The transient 2D temperature and ablation of the bottom head wall surface were calculated. At the same time, the finite element analysis program was used to calculate the thermal stress and strain caused by the transient temperature field and ablation of the wall of the lower head. The thermal stress / strain condition was used to analyze the structural integrity of the PWR RPV lower head in the pressure vessel under the In-vessel retention via external reactor vessel cooling (IVR-ERVC). The calculation results show that:①The wall of the lower head began to melt at 200 s after the core collapsed, and the thinnest thickness decreased linearly. After 3000 s, the molten zone along the inner wall of the head formed a lancet shape distribution. ②The endothermic heat flux of the inner surface of the lower head was larger than that of the external surface, and the heat flux reached the maximum value at the interface between the two layers of the molten pool. ③The strain of RPV lower head increased sharply in the period from 0 s to 400 s and then keeps unchanged. The equivalent stress locally concentrated on the inner wall of the lower head at 400 s. After 400 s, the inner wall thermal stress decreased gradually, the shape variable increased, and the integrity of the lower head could be guaranteed. ④After 2000 s, the stress concentration was generated on the inner and outer wall at the ablation of the RPV lower head. The stress value of the inner and outer walls of the lower head was greater than that of allowable stress. The RPV lower head may fail at any time after 2000 s and may be broken at the edge of the ablation area.

Research on Analytical Verification Method for Valve Discharge

Liu Ping, Hu Jinhui, Wang Baoping, Wang Yueqin, Li Junye, Ru Qiang

2018, 39(5): 133-136. doi: 10.13832/j.jnpe.2018.05.0133

Abstract(140)

Abstract:
In this paper, an analytical verification method for valve discharge under saturated steam is studied by taking the pressurizing system stop valve as an example. Comparison of the values of valve discharge test under small opening and low pressure difference with the theoretical calculation value shows that the deviation is within the acceptable range. The value of valve discharge under full opening and full pressure difference is obtainedby putting the test values into the equation deduced. Comparingn it with the value of valve discharge by ANSYS analysis under full opening, we can found that the deviation between the actual derived discharge and the theoretical calculation discharge is 2.73%. The results show that under the condition of insufficient test capacity, the discharge rate of saturated steam by the valve can be analyzed by coupling thedischarge test in small opening and low pressure difference with the theoretical derivation.

Study on United Numerical Simulation for Stepping Motion of Control Rod Drive Mechanism

Wei Qiaoyuan, Zhang Fei, Wu Hebei, Liu Yanwu, Zhao Maomao, Li Yuezhong

2018, 39(5): 137-141. doi: 10.13832/j.jnpe.2018.05.0137

Abstract(215)

Abstract:
The stepping motion of the control rod drive mechanism is a complicated dynamic process combined with electromagnetic field, flow field, and motion. It is not accurate to analyze the process by static method or partial technique. Magnet software, Fluent software, and Adams software were combined to conduct the coupling calculation of gravity, electromagnetic force, hydraulic resistance, and spring force during the stepping motion of the control rod drive mechanism. The electromagnetic force, hydraulic resistance, displacement and velocity of movable parts, and stepping loads were calculated quantitatively. The calculated values of the movable pole closing time, critical current and stepping loads were compared with the corresponding experimental values. The calculated values were in consistence with the experimental values. It is proved that the united simulation is an effective dynamic analysis method for the stepping motion of the control rod drive mechanism.

Optimization of Design Concept for Hydrostatic Test Equipment in Nuclear Power Plants

Feng Huixing

2018, 39(5): 142-144. doi: 10.13832/j.jnpe.2018.03.0142

Abstract(165)

Abstract:
Nuclear vessel hydrostatic test is one of the important in-service inspection methods in nuclear power plants, to verify vessels under sustained pressure integrity and sealing. According to the installation requirements of the temporary special devices, in the design stage of vessel and pipeline of nuclear power plants, the vessel body and the pipeline layout are optimized, to reduce the personnel exposure in the in-service hydraulic test, to reduce the test duration, to lower the risk of equipment and related accessories, and to reduce the maintenance cost of nuclear power plants.

Research on Best Location of Lifting Lugs on AP1000 Containment Vessel Module by Cast Rigging Lifting

Li Tuo

2018, 39(5): 145-148. doi: 10.13832/j.jnpe.2018.05.0145

Abstract(138)

Abstract:
Containment vessel of AP1000 is designed to be built by modules. Based on the new lifting method called rigging lifting, further improvement to change the containment vessel from four rings to three rings was proposed, and the best location of lifting lugs was studied. The load ratio of the crane and the safety distance was checked. The proposal to lift the containment vessel by three rings was verified. The new rigging plan can reduce safety risk, and improve the installation quality of the containment vessel. The research of the best location of lifting lugs could be used as the guidance for the containment vessel design and lifting plan improvement.

Application Status and Maintenance Optimization Analysis of Ventilation System Fire Damper in Nuclear Power Plants

Zhang Jianghong, Zhang Guanghui

2018, 39(5): 149-153. doi: 10.13832/j.jnpe.2018.05.0149

Abstract(187)

Abstract:
In order to improve the reliability of the fire damper in nuclear power plants, the application status and failure history of fire damper in several nuclear power plants were analyzed in this paper. The failure mechanism and maintenance strategy optimization of different fire dampers are studied by Failure Mode and Effect Analysis (FMEA) method. The results show that the annual inspection of appearance and action is optimized, and the task of regularly changing fuse element and electric mechanism is added, which can effectively reduce the failure rate of fire damper and ensure the safe and stable operation of nuclear power plants.

Design of Core Cooling Monitoring System in HPR1000

He Zhengxi, He Peng, Chen Xuekun, Xu Tao

2018, 39(5): 154-158. doi: 10.13832/j.jnpe.2018.05.0154

Abstract(534)

Abstract:
HPR1000 is the third generation NPP, and its reactor and primary loop system need to fulfil higher requirements for inherent safety. For the core cooling monitoring system (CCMS) of the second generation nuclear power plant, the bottom of the reactor shall be bored to measure the water level. This design degrades the inherent safety of the reactor and shall be redesigned. This paper designs a new core cooling monitoring system, in which the water level detector inserted into the core from the pressure vessel cover directly. This design improves the accuracy of measuring the water level of key points and eliminates the boring at the bottom of the reactor, which meets the HPR1000 reactor inherent safety requirements.

Manufacturing of HTO Passive Sampler and Its Application in Closed Nuclear Power Plant

Duan Zaiyu, Li Jianmin

2018, 39(5): 159-161. doi: 10.13832/j.jnpe.2018.05.0159

Abstract(289)

Abstract:
According to the environmental characteristics of nuclear power ships, a passive sampler with a volume ratio of ethylene glycol to tritium-free water of 1:1 is used to monitor the concentration of tritiated water in a closed nuclear power plant. After continuous monitoring for 1 month in different work places, it is found that the maximum relative deviation of the sampler was less than 10%; Sampler has been compared with the passive sampler with silica gel as the adsorption medium for 1 day comparison experiment, and the measurement error is less than 6%. Results show that the sampler can accurately measure the low concentration of HTO, which meets the monitoring requirements of HTO in closed nuclear power plants.

Research on Human-Machine Interface Design of Deaerator Water Level Control

Liu Xiaoguang, Wang Yanhua, Li Yongjun

2018, 39(5): 162-166. doi: 10.13832/j.jnpe.2018.05.0162

Abstract(279)

Abstract:
Based on the principle of human factors engineering（HFE）, this paper takes the secondary feed water deaerator system of a nuclear power plant as an example to analyze the performance requirements, obtain different levels of static function database, and determine the basic information flow and its processing requirements. In order to meet the operational requirements of the nuclear power plant in the cold start, the low load and high load conditions and the functional factors related to the control room are clarified by establishing the water supply heating and deaeration function (F01) block diagram and operation mode table. Based on the principle of functional allocation, the water level control of the deaerator is studied, and the non - disturbance transition scheme of water level control is realized. It has been verified by human - machine interface design, and provides reference for the human factor engineering analysis in China.

Uncertainty Analysis of Reactor System Dynamical Response under Seismic Load

Xiong Furui, Huang Qian, Shen Pingchuan

2018, 39(5): 167-171. doi: 10.13832/j.jnpe.2018.05.0167

Abstract(460)

Abstract:
Dynamical analysis of the reactor structure under seismic load is a key procedure during the safety design of the entire reactor system. Due to stochastic and other uncontrollable errors during computation, manufacturing and installation, structural parameters usually subject to a certain amount of uncertainties. Reactor system seismic response influenced by uncertain structural parameters is investigated in this paper. Maximum entropy principle is adopted to construct the probability density distributions (PDFs) of contact stiffness and clearance values among various assemblies of the reactor system. Then, Markov Chain Monte Carlo (MCMC) technique is applied to sample data points according to given PDFs and the input-output data pool is constructed. Distributions of dynamical responses at various locations of the reactor system are examined from the sampled data pool. Results show that the uncertainties near upper and lower core plates depict different distribution patterns. The research in this paper provides a way to quantitatively study the robustness of the seismic analysis and the reliability of dynamical responses in the reactor system.

Research of Bundle CHF Prediction Based on Minimum DNBR Point and BO Point Methods

Liu Wei, Peng Shinian, Jiang Guangming, Liu Yu, Shan Jianqiang

2018, 39(5): 172-175. doi: 10.13832/j.jnpe.2018.05.0172

Abstract(260)

Abstract:
In this study, both of the minimum DNBR point and the BO point methods are applied to develop a new CHF correlation named the ACC correlation coupled with sub-channel code ATHAS. The statistics and predication rate are analyzed. The Owen criterion is used to determine the DNBR limit. The data analysis and assessment results indicate that compared with the minimum DNBR point method, the BO point method predicts the risk of DNB more reasonable and conservative with better predication rate.

Analysis of Effect of Decay on Release Characteristics of Fission Product and Off-Site Dose Assessment

Wang Junlong, Liu Jiajia, Lyu Huanwen, Li Lan, Tan Yi

2018, 39(5): 176-181. doi: 10.13832/j.jnpe.2018.05.0176

Abstract(183)

Abstract:
Severe accident release categories of a nuclear power plant of third generation were introduced. The release characteristics to the environment were calculated using MAAP code for those release categories and its severe accident sequences which can result in the large release of radioactive materials. On this base, several nuclides which contribute the most personnel dose were selected. This paper calculated the cumulative activity released to the environment and whole body and thyroid dose of 500 m site boundary with and without considering the decay of nuclides. The effect of decay on the results was analyzed. All these analysis can provide some reference for the improvement of severe accident simulation program. The effect of decay on the cumulative activity of fission products released to the environment is related to the radionuclide half-life and the release time of the fission products after the accidents. From the off-site dose analysis, the effect of decay on whole body dose is more obvious than that on the thyroid dose.

Research on Event Alarm and Emergency Response of Loose Parts for Nuclear Power Plants

Hu Jianrong, Lyu Ailin, Yang Taibo, Liu Caixue, Luo Ting, Jian Jie

2018, 39(5): 181-182. doi: 10.13832/j.jnpe.2018.05.0181

Abstract(153)

Abstract:
Based on the researches of the event alarm logic of loose parts and worldwide regulations and standards of loose parts event alarm and emergency response, combined with the loose parts event alarm and emergency response in a domestic nuclear power plant during hot state function test, the process of a loose part event in that nuclear power plant is analyzed in detail. The loose parts event alarm is divided into absolute and relative threshold alarms. To improve the alarm accuracy, counting factor and channel rechecking are increased in the alarm logic. GB/T 11807, IEC 60988 and ASME require to take measures to determine if the event alarm is caused by the loose parts, and the potential damage to the primary loop system of the reactor is evaluated by the loosen parts. RG 1.133 requires that the loose event alarm should be reported to the Nuclear Regulatory Commission. GB/T 11807 and IEC 60988 have no clear specification on whether the loose part events should be reported to the national nuclear safety administration, which is determined according to the specific operation of the nuclear power plant. Based on the alarm and emergency response to a loose part event in a nuclear power plant in China, it is suggested that the emergency response mechanism suitable for nuclear power plants in China should be established, loose parts events should be determined in time, and the possible loss of equipment components should be evaluated timely, and further provisions should be made on whether to report and take what measures.

Analysis of Failing to Meet LHSI Pump Flow Test Criteria in CPR1000 Unit

Liu Xingwei, He Jinqun, Wang Jineng

2018, 39(5): 186-188. doi: 10.13832/j.jnpe.2018.05.0186

Abstract(474)

Abstract:
When the results of the pump flow test of the low head safety injection system (LHSI) of CPR1000 do not satisfy the acceptance criteria, the throttle elements in the pipeline system need to be adjusted. The adjustment method of the lower limit flow orifice plate can be obtained through the calculation and analysis of the engineering fluid mechanics. By using this method, the size of the restricted throttle elements can be obtained accurately. The analysis method effectively solves the problem of how to change the throttle elements in the pipeline system when the low head safety injection pump flow test do satisfy the requirement of the criterion.

Research on Loose-Parts Impact Test in Channel Defect Mode

Wang Jiaqian

2018, 39(5): 189-192. doi: 10.13832/j.jnpe.2018.05.0189

Abstract(150)

Abstract:
During the outage of a nuclear power plant, a loose parts impact test was carried out on the steam generator, and the main defect modes and treatment methods of the loose component monitoring system (LPMS) in long-term operation are proposed. The blind area in the function of fault self-checking in LPMS system is given, and the response characteristics of the defective channel to the impact signal of the loose parts are analyzed. Poor channel contact, electric charge accumulation and multi-channel signal interference are the main factors that cause the signal distortion. The accumulation of electric charge causes the static blocking of the signal channel; the signal interference between multiple channels is an important cause for false alarms and channel overload and disconnection.

2018 Vol. 39, No. 5