Table of Content

    20 July 2020, Volume 54 Issue 7
    Development of Flow Distribution Optimization Program for Natural Circulation Reactor
    XU Haipeng;WANG Yan;XIE Heng
    2020, 54(7):  1153-1160.  DOI: 10.7538/yzk.2020.youxian.0011
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    The driving force of primary loop of natural circulation reactor is limited and the total flow of loop is small, so the design and optimization of core flow distribution are very important. Reasonable core flow distribution can not only meet the thermal safety requirements, but also directly improve the core performance. Based on the above reasons, the optimization of flow distribution in natural circulation reactor was studied preliminarily. A onedimensional flow and heat transfer model was used to establish the initial solution model of inlet resistance coefficient optimization, and an exact solution searching algorithm was designed for the closed parallel channel studied in this paper. Coupled with COBRA, a core thermal analysis program was developed to optimize core flow distribution. By using the program, a natural circulation reactor was chosen as an example to calculate and analyze the flow distribution optimization in the life of the reactor core. The results show that, by taking the average value of the optimal setting scheme of the inlet resistance obtained by the optimization of the flow distribution of a single node in each typical life cycle, a better setting scheme of the inlet resistance in the core cycle life can be obtained. Aiming at the disadvantage that it is difficult to obtain the global optimal solution by means of the artificial design method of average value, a method of automatically realizing the optimization of flow distribution in the cycle life was proposed by referring to the modern optimization calculation method.

    Analysis and Optimization of Heat Pipe Radiation Radiator for MW Space Nuclear Reactor System
    ZHANG Haochun;LIU Xiuting;WEI Qianming;YOU Ersheng;SUN Mingyuan
    2020, 54(7):  1161-1167.  DOI: 10.7538/yzk.2020.youxian.0018
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    Space nuclear reactor is the research foundation of space nuclear power and nuclear propulsion. The volume and mass of high-power nuclear reactor are always the important factors limiting the aerospace design. In view of this problem, the heat pipe radiation radiator of MW space nuclear reactor system was modeled and analyzed, and the thermal resistance model of the heat pipe radiation radiator was established in this study. The effect of the length Lf and thickness δf of the bare carbon fiber fin, the mass flow m of the coolant, the inlet temperature Tf1 of the radiation radiator on the mass M of the radiation radiator was discussed under the given conditions by the exhaustive method and genetic algorithm. The results show that when Tf1 is 800 K, Lf is 5 cm, δf is 0.16 mm, and m is 9 kg/s, M is the best. At this time, M is 906.593 kg, and 0.63% of the system mass is optimized.

    Analysis of Operating Characteristic of Direct Brayton Cycle Gas-cooled Reactor System
    MING Yang;YI Jingwei;FANG Huawei;LIU Kai;ZHAO Fulong;TAN Sichao;TIAN Ruifeng
    2020, 54(7):  1168-1175.  DOI: 10.7538/yzk.2020.youxian.0013
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    The system analysis code for small direct Brayton cycle reactor systems was implemented using Matlab/simulink code. The mathematical and physical model was established by modularization idea. Steady-state and transient operating conditions of the reactor system were simulated by changing the operating parameters of the primary equipment including the reactor, turbine, compressor and heat exchanger. The variation curves of the reactor power, inlet and outlet pressures and temperatures were obtained. The results show that the system analysis code can accurately simulate the steady and transient operating characteristics of the gas-cooled reactor system and the parameter response characteristics of the primary equipment. The research result can provide a basis for the design, optimization and safety analysis of small direct Brayton cycle reactor system.

    Research on Heat Pipe Technology and Performance Analysis Code
    CHEN Qichang
    2020, 54(7):  1176-1184.  DOI: 10.7538/yzk.2020.youxian.0028
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    As an efficient means for heat transfer, heat pipes have the characteristics of high reliability, low temperature difference and no need of external driving, so they have been widely studied and applied in aerospace, nuclear engineering and other fields. In order to apply this technology to the design and development of new type reactors, the theoretical method and performance analysis code of heat pipe were researched. The structure, working mechanism, type and performance characteristic of heat pipe were analyzed. The heat pipe theory and calculation models of heat transfer limits were studied. On this basis, the heat pipe performance analysis code HEPAC was developed, which had the ability to analyze the heat transfer characteristics and calculate the heat transfer limits for various types of heat pipes. The HEPAC code was tested and verified by comparing with the relevant experimental data. The code development provides an important tool for the design and analysis of various types of heat pipe reactors.

    Theoretical Calculation of Photonuclear Reaction Data for 50,51V with Photon Energy below 200 MeV
    LI Lin;LIU Ling;XU Ruirui;WANG Jimin;TAO Xi;TIAN Yuan;JIN Yongli;GE Zhigang
    2020, 54(7):  1185-1191.  DOI: 10.7538/yzk.2020.youxian.0081
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    Photonuclear reaction data are very important for a large range of applications, such as the radiation damage, reactor dosimetry, accelerator shielding, etc. Based on MEND-G program, the photonuclear reaction data of neutrons, protons, 4He and other major emergent particles for 50,51V in the energy range of 0 to 200 MeV were calculated theoretically. The self-consistent analysis of the experimental measurement of photon-induced V isotope nuclear reaction was performed. The photon strength function and the quasi-deuteron model were used to increase the calculation capability of the photonuclear reaction absorption cross section to 200 MeV, providing data basis for the theoretical calculation of photonuclear reactions. After systematic comparison, the theoretical calculations are in good agreement with the experimental measurements, and the data are included in the sub-library of coming CENDL-3.2 for photonuclear data.

    Hydrophobic Modification of Ni-MOF-74 and Its Effect on CO Adsorption Performance
    GUO Peiran;HU Shilin
    2020, 54(7):  1192-1198.  DOI: 10.7538/yzk.2019.youxian.0529
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    CO often coexists with N2, CH4, CO2 and H2O. In order to purify CO gas, Ni-MOF-74 material was used as adsorbent to carry out exploratory research. Ni-MOF-74 hydrophobic material was prepared by PFAS (perfluoroalkylsilane) immersion. The surface hydrophobicity and morphology of Ni-MOF-74 were analyzed by means of contact angle measuring instrument, SEM and FT-IR. The effects of immersion temperature, immersion concentration, immersion time, and immersion number of times on hydrophobicity were discussed. The influence of hydrophobic modification on CO adsorption performance was also studied. The experimental conditions of PFAS immersion were determined: The immersion temperature is 110 ℃, the immersion time is 24 h, the immersion number of times is 1, and the concentration of PFAS is 3%. The results show that PFAS is successfully coated on Ni-MOF-74, but only a few of them covers the surface of Ni-MOF-74. The crystal structure of Ni-MOF-74 modified by PFAS is partially destroyed and the thermal stability is reduced.

    Age Dating of Uranium Sample by ICP-MS
    CHEN Yan;ZHAO Yonggang;LI Lili;CHANG Zhiyuan;ZHU Liuchao;XIAO Guoping;HUANG Shenghui
    2020, 54(7):  1199-1204.  DOI: 10.7538/yzk.2019.youxian.0461
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    Uranium age of sample is relevant to determine the date of production of the material and important to identifying the source of the material in nuclear forensics investigations. An analytical method for age dating of uranium-based nuclear materials by the measurement of 230Th/234U atomic ratio was developed. Uranium samples were diluted with 229Th and 233U spike respectively. Uranium and thorium in the samples were separated with TEVA resin. The atomic ratios of 229Th/230Th and 233U/234U were measured by multi-collected inductively coupled plasma mass spectrometry (MC-ICP-MS). The uranium age of the samples could be obtained by 230Th/234U atomic ratio according to the calculation formula of uranium age. The age of CRM U850 and U010 standard samples was determined. The results are consistent with those obtained by Lawrence Livermore National Laboratory in the United States. However, the results of both methods are older than the actual age. It was probably due to the incomplete purification process during production, which resulted in the residual 230Th in the samples. The established method can be used to determine the 234U-230Th model age of uranium samples and provide important information for verification investigation.

    Effect of Helical Diameter on Cross-sectional Distribution Characteristic of Bubbly Flow in Helical Tube
    HUANG Yichuan;GUI Nan;YANG Xingtuan;TU Jiyuan;JIANG Shengyao;ZHU Hongye
    2020, 54(7):  1205-1213.  DOI: 10.7538/yzk.2019.youxian.0444
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    The cross-sectional distribution characteristics of the phase interface parameters (void fraction, phase interface concentration, bubble size, etc.) of bubbly flow in helical tubes with different helical diameters were studied. The measurement accuracy of double-sensor conductivity probe was calibrated by imaging method, and then the quantitative distribution profiles of void fraction, phase interface concentration and number frequency of bubbles in helical tubes were obtained. In order to describe quantitatively the distribution characteristics of phase interface parameters, the cross-sectional average parameters, phase interface dispersion coefficient and bubble average aggregation coordinates were defined by statistical methods to characterize the distribution characteristics. The experimental results show that, with the increase of the rotating diameter of the pipe, the average void fraction of the bubble section decreases, the distribution range decreases, the average locations of aggregation move to the topside and the outside, and the overall bubble size decreases.

    Experimental Investigation of Flow Induced Vibration of Multi-span Straight Tube Bundle with Large Gap AVB Support
    HUO Zhuo;ZHANG Kefeng;XIONG Zhenqin;ZU Hongbiao;GU Hanyang;XIE Yongcheng;QUAN Zhengting
    2020, 54(7):  1214-1220.  DOI: 10.7538/yzk.2019.youxian.0459
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    Flow induced vibration of a four-span tube bundle with large gap anti-vibration bars (AVB) supports was experimentally studied. The tube bundle was arranged in rotated triangular pattern. Three AVB supports divided the straight tube bundle into four spans. Only part of one of the middle span tubes was subjected to cross flow. The response characteristics of the displacement and frequency of the heat transfer tube with the pitch velocity in the test increased from 3.3 m/s to 14.7 m/s were obtained. The results show that the vibration frequency increases with the pitch velocity in streamwise direction and transverse direction in proper order. The vibration mode of the tube changes from the state of one effective AVB support and two ineffective AVB supports to the state of three effective AVB supports. The fluidelastic instability (FEI) critical velocity for the tube tested is 14.5 m/s. Five empirical correlations were used to predict the FEI critical velocity, and the predicted results were compared with the experimental results. It is found that the Chen correlation is best. The predicted FEI critical velocity by Chen correlation is conservative, and the relative deviation from the experimental value is 21.4%.

    Research on Hydrogen Control Strategy in Advanced Pressurized Water Reactor Nuclear Power Plant
    DING Chao;YANG Zhiyi;ZHOU Zhiwei;SONG Mingqiang;CHAI Guohan;QIU Suchen;CHONG Yimin
    2020, 54(7):  1221-1227.  DOI: 10.7538/yzk.2020.youxian.0151
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    The design of new-built nuclear power plant (NPP) must achieve the result that the possibility of early radioactive release or large radioactive release is ‘practically eliminated’, and the containment failure caused by hydrogen detonation is one of the severe accident conditions that shall be ‘practical eliminated’. So each kind of hydrogen elimination measures should be analyzed and investigated to establish assessment method. And it could provide supporting method for reviewing on design selection of hydrogen control strategy for advanced pressurized water reactor (PWR) NPP. Considering the hydrogen control strategy in severe accident management, it is the key issue to conduct review on combustibility of local space in containment. Based on the criterion of combustibility, flame acceleration (FA) and deflagration to detonation transition (DDT), 3D-CFD code was used to simulate the combustibility and hydrogen risk in steam generator compartment under representative severe accident condition. The results indicate that the combustibility still exists in large zone of compartment, even though the discharging gas is entrained with a high percentage of steam, and igniters with design of reasonable layout are able to ignite and eliminate hydrogen. The analysis method in this paper can be used to assess the design selection of hydrogen control strategy in NPP.

    Probabilistic Safety Assessment Event Tree Analysis of Radioactive Release Risk for Pool Type Sodium-cooled Fast Reactor
    YANG Peng;YU Hong;HU Wenjun
    2020, 54(7):  1228-1234.  DOI: 10.7538/yzk.2019.youxian.0445
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    The safety characteristics and radioactive release mechanism of pool type sodium-cooled fast reactor (SFR) are significantly different from those of pressurized water reactor. Under the new nuclear safety requirements, the research on probabilistic safety assessment (PSA) of radioactive release risk for SFR is urgently needed. Taking the pool type SFR as the research object, the main position and mode of large radioactive release were identified and the radioactive release event trees were constructed by analyzing the radioactive source, the containment boundary and the severe accident that might challenge the integrity of the containment boundary. The results can serve as reference for further research on PSA of radioactive release risk for pool type SFR.

    Application of Dynamic Reliability Evaluation Method in AP1000 Severe Accident Analysis
    CUI Chengxin;HUANG Ting;CHEN Lian;ZHANG Lei
    2020, 54(7):  1235-1240.  DOI: 10.7538/yzk.2019.youxian.0485
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    The dynamic reliability evaluation method can simulate the continuous or multiple changes of the system state, and becomes a new development point for the probability safety research of nuclear power plants. Based on the dynamic reliability evaluation method, the state black out process of the AP1000 by MAAP severe accident program was analyzed, the results were applied to level 2 probabilistic safety assessment (PSA) analysis and finally the effect on radioactive fission products was evaluated. The research results show that the dynamic characteristics of the system have a certain impact on the analysis results of the PSA of nuclear power plant, and the dynamic reliability evaluation process can mine more information, and to better guide the design of nuclear power plants and improve the safety of nuclear power plants.

    Study on SP3 Analytic Function Expansion Nodal Method for pin-by-pin Transport Calculation
    PENG Lianghui;TANG Chuntao;BI Guangwen;ZHANG Hongbo;YANG Bo
    2020, 54(7):  1241-1247.  DOI: 10.7538/yzk.2019.youxian.0472
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    Based on the simplified P3 (SP3) method, a SP3 analytic function expansion nodal code was developed under three different partial current expressions. The numerical test of the code was carried out and the accuracy of the three partial current expressions was compared. The numerical results show that the pin-by-pin transport calculation based on the SP3 analytic function expansion nodal method can obtain high accuracy of the pin power. Among the three different partial current expressions, Marshak’s expression has the highest accuracy while the other two expressions are more conducive to code’s algorithm simplification and parallel algorithm design.

    Characteristic Research of Neutron Detector Response Function in PWR
    WANG Changhui;WU Hongchun
    2020, 54(7):  1248-1253.  DOI: 10.7538/yzk.2019.youxian.0488
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    The neutron detector response function is the effective connection between the numerical simulation results and the detector readings, which shows the contribution for the detector readings from rounding areas. In this paper, the formula of the response function was given and calculated using MCNP code. The interfering factors to the response function were analyzed. The analysis results show that the factors that have great influence on the neutron absorption characteristics, such as boric acid, control rod and burnable poison, have significant effect on the response function of the neutron detector. Other effects not referred in this paper could be analyzed by the same method.

    Research on Core Concept Design of Ultra-long Life Small Natural Circulation Lead-based Fast Reactor
    LIU Zijing;ZHAO Pengcheng;ZHANG Bin;YU Tao;XIE Jinsen;CHEN Zhenping;SUN Yumeng
    2020, 54(7):  1254-1265.  DOI: 10.7538/yzk.2019.youxian.0720
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    Aiming at improving the economy and inherent safety of lead-based fast reactor, the SPALLER-100 core concept design of 100 MWt ultra-long life small natural circulation lead-based fast reactor was proposed. Based on the selection of PuN-ThN fuel and 208Pb-Bi coolant, a fuel assembly design scheme with the addition of solid moderator BeO was carried out. The core layout and control rod system design were presented, and the core neutronics characteristics and steady state natural circulation features were analyzed. The results show that under the conditions of low fuel loading and small core volume, the refueling cycle of SPALLER-100 core is up to 32 years, the average discharge burnup is up to 210.38 MW·d/kg(HM), and the reactivity coefficients are negative throughout the lifetime. Under steady-state operating conditions, the maximum temperatures of fuel cladding and pellet are lower than the safety limit, and the core has primary circuit natural circulation capability and certain automatic flow distribution capability.

    State Differential Feedback Control and Equivalence between Two Types of Existent Power Control System of Pressure Water Reactor
    LUAN Xiuchun;LIU Lei;WANG Junling;YANG Zhida;ZHOU Jie
    2020, 54(7):  1266-1272.  DOI: 10.7538/yzk.2020.youxian.0106
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    A new control law, the state differential feedback control, which was applied to design the control system of a reactor in the load-following mode, was presented in this paper. The design processor depended on the equivalence between two control systems, which included the integral control for control rod position and the state differential feedback control for control rod speed, which was obtained as a theorem. The simulation results show that the state differential feedback control system has a good load-following capability of a reactor.

    Optimization Method for Spare Part Quantity of Control Rod Drive Mechanism
    XIAO Lili;JIN Fenglei;GU Jipin
    2020, 54(7):  1273-1278.  DOI: 10.7538/yzk.2019.youxian.0477
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    The control rod drive mechanism is the servo mechanism of the reactor control and protection system. It is an important nuclear safety equipment for performing reactor power regulation and emergency shutdown. The cost of the control rod drive mechanism is high, and how to determine the number of spare parts is important for improving the operability of the reactor. In this paper, for the control rod drive mechanism, under the constraint condition that the continuous operation time of the system is not less than the refueling cycle, an optimization method for determining the number of spare parts for the control rod drive mechanism, the optimization method of grouping spare parts quantity, was proposed. And the spare parts configuration scheme of each subsystem with the least total cost was given. By the stochastic simulation calculation, the method of optimizing the number of spare parts was compared with the conventional algorithm. The results show that the optimization method is better than the conventional algorithm. Under the premise of ensuring the availability of the control rod drive mechanism, optimizing the number of spare parts can reduce the cost. The optimization method is also applicable to the analysis of spare parts of other equipment, and has guiding significance for the analysis and research of equipment spare parts in engineering.

    Study on Measurement Method and Experiment of Team SSA in Digital Nuclear Power Plant
    LI Pengcheng;JIN Xiao;ZHANG Li;LU Wenjie;WANG Yanxin;DAI Licao
    2020, 54(7):  1279-1286.  DOI: 10.7538/yzk.2019.youxian.0454
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    Situation awareness (SA) issues are more prominent in digital nuclear power plants (NPPs). In order to identify the levels of team shared situation awareness (SSA) and the effects of performance shaping factors (PSFs) on SSA, a measurement and calculation method of team SSA was established based on the situation awareness global assessment technique (SAGAT) and simulator experiments were carried out. The experimental results show that the individual SA (ISA) level is related to the SSA level. The higher the ISA level is, the higher the SSA level is. Both ISA and team SSA are influenced by PSFs, the higher the state levels of PSFs are, the higher the levels of ISA and SSA are. For different experimental scenarios, the ISA level of operators and team SSA level are different, which means that the more obvious symptoms of the risky scenarios are, and the higher knowledge and experience level is, the relatively higher ISA and SSA levels are. Finally, the state levels of PSFs are identified by self-evaluation, and the main poor PSFs are team communication and cooperation level, pressure level, and human-machine interface etc. These assessment results provide a theoretical support for improving human reliability and safety level in digital NPPs.

    Influence of Beam Drift on Measurement with Silicon Strip Detector Array
    SUN Haohan;LIN Chengjian;MA Nanru;WANG Dongxi;JIA Huiming;YANG Lei;YANG Feng;ZHONG Fupeng;WEN Peiwei;YAO Yongjin
    2020, 54(7):  1287-1293.  DOI: 10.7538/yzk.2019.youxian.0372
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    The influence of beam drifts on the measurement with silicon strip detector array was discussed, which was characterized by compact structure and large solidangle coverage. Monte Carlo method was used to simulate the effects of beam drifts in different directions and distances on the angular distribution of the decayed α from the fusion-evaporation residual nuclei and the Rutherford scattering. The results indicate that when the beam drift distance is less than 3.0 mm, the symmetry of the detector array can constrain the count error within 10%. The experimental measurements of 6Li+209Bi system at energy of 25 MeV and 40 MeV were performed, and the beam drift was studied using data recorded by the monitor detectors. The results show that the beam drift distances of all runs are less than 3.0 mm. After correcting the beam drifts, the elastic scattering angular distribution at 40 MeV was obtained by three different solid-angle calibration methods, and the result is consistent with the existing data.

    Study on Monte Carlo Variance Reduction Method for Thick Shield and Small Detector Problem
    GAO Shenshen;LI Junli;WU Zhen;MA Ruiyao;WANG Xin;QIU Rui;LI Chunyan;ZHANG Hui
    2020, 54(7):  1294-1300.  DOI: 10.7538/yzk.2019.youxian.0371
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    Thick shield and small detector problem often occurs in reactor shielding calculation. It is difficult to solve it by the conventional Monte Carlo method. A variance reduction method called small detector auto-importance sampling (SDAIS) method was proposed for the thick shield and small detector problem. The SDAIS method is the improved method of the auto-importance sampling (AIS) method in the case of small detector problem. The virtual particle Russian roulette method was used based on the detector position instead of the weight. The SDAIS method was implemented in the self-developed Monte Carlo program MCShield. The NUREG/CR-6115 PWR benchmark was used to verify the correctness and computational efficiency of the method. The results show that compared with the AIS method and the traditional importance method, the calculation efficiency of SDAIS method has an improvement of 1-2 orders of magnitude. SDAIS method can effectively solve the problem of thick shield and small detector.

    Study on Optimization of Measurement Condition for Quartz Thermoluminescence Dating
    WANG Meng;GU Yi;LU Heng;WANG Hao;SUN Kun
    2020, 54(7):  1301-1307.  DOI: 10.7538/yzk.2019.youxian.0421
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    In this paper, the optimal pre-heat temperature of quartz about 375 ℃ thermoluminescence (TL) peaks at different irradiation doses and the optimal test dose of quartz aliquots about correcting thermal sensitivity were analyzed based on analysis of TL characteristics of quartz. According to the measurement results, the conventional protocol was improved. The improvement of the accuracy of equivalent dose measurement was verified based on the improved protocol of optimal measurement parameters. The experiment results show that if equivalent dose of quartz is small, the signal of 325 ℃ TL peak is weaker, and the lower pre-heat temperature can reduce the impact from 325 ℃ TL peak for 375 ℃ TL peak. If the equivalent dose of quartz is larger, in order to reduce the impact from 325 ℃ TL peak, the pre-heat temperature of regeneration dose and test dose should be set separately, and the measurement protocol should determine the optimal preheat temperature of quartz before the equivalent dose measurement. When the regeneration dose of quartz aliquots is in the range of 0-1 000 Gy, the test dose of 200 Gy can correct thermal sensitivity change of quartz aliquots and the relative deviation of 10 repeated measurements is less than 5%. The residual signal interference of 325 ℃ TL peak can be reduced and the quartz thermal sensitivity change can be corrected by using the above optimal pre-heat temperature and test dose. The relative deviations of the equivalent dose of 400 Gy and 700 Gy under the conventional protocol are 14.74%-47.15% and 33.47%-197.71%, and the relative deviations of the equivalent dose are all reduced to the range of ±4% under the improved protocol. The research provides an important reference for improving the measurement accuracy and the range of dating of 325 ℃ TL peak measurement.

    Simulation of Scattered Neutron Distribution in Accelerator Neutron Source Hall
    XU Zixu;QU Guofeng;WANG Yizhou;LI Min;ZHOU Maolei;LIU Dong;LIU Xingquan;LIN Weiping;HAN Jifeng
    2020, 54(7):  1308-1317.  DOI: 10.7538/yzk.2019.youxian.0451
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    The accelerator neutron source can provide monoenergetic neutrons in a wide energy range, which can be widely used in many fields, such as the neutron reaction cross-section measurement, neutron detector calibration, etc. Scattered neutrons might introduce errors or even cause troubles to these experiments, therefore it is necessary to precisely analyze and qualify the neutron scattering ratio in the accelerator neutron source hall. In this paper, MCNP5 was used to simulate the scattering ratio for neutron source energy from 0.2 to 20 MeV, and the distribution of the scattering ratio as well as the contribution from the air and the wall were analyzed. It is found that the direct neutron components decrease with the distance following the inverse square law, while the scattered neutron components are approximately constant. The scattered neutrons are mainly generated by the wall in the hall, while the contribution of the air could be ignored. The scattering ratio is much smaller for the range much closer to the neutron source, which increases quickly with the distance between the neutron source and the detector. When the energy of neutron source is greater than 1 MeV, the scattering ratio begins to decline until it enters a plateau in which the energy is higher than 7 MeV. The scattering ratio is the highest for the neutron source with the energy of 0.4 MeV and 1 MeV, while that is the lowest for 10 MeV and 15 MeV. Total macroscopic scattering cross-section of neutrons to materials is found to be able to explain simulated results of the scattering ratio properly, because the curve of total scattering cross-section versus energy is very similar to that of the scattering ratio versus the neutron source energy. Moreover, the elastic cross-section is much larger than the inelastic scattering cross-section, and the elastic scattering plays a leading role in neutron scattering. When the energy of the neutron source is higher, the total scattering cross-section is much lower which renders the scattering ratio much lower. However, due to the higher inelastic scattering cross-section, neutrons with higher energy are more likely to lose a lot of energy during each collision, which would make a higher proportion of the slow neutron components within scattered neutrons. The scattering ratio could be effectively reduced by attaching a layer of the neutron moderating and absorbing material on the wall. For example, 5 cm boracic polyethylene (10%B4C) could reduce the scattering ratio by about 40%. These results are instructive for applying accelerator neutron source as well as for the accurate analysis and attenuation of the scattering effect in neutron experiments.

    Density Response Characteristics and Algorithm of Density Logging Instrument Using Controllable Neutron Source
    YUE Aizhong;CHEN Haizheng;ZHANG Qingmin;GAO Keqing;ZHANG Xiaolei;SUN Peiwei;WANG Shusheng;ZHAO Yuanyuan;WANG Hu
    2020, 54(7):  1318-1325.  DOI: 10.7538/yzk.2019.youxian.0858
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    In density logging with the controllable neutron source, the number of neutrons emitted by the pulsed neutron source fluctuates from pulse to pulse, which leads to the poor accuracy of the density measurement results under the impact of different formation and detecting conditions. In order to improve the stability and accuracy of the density measurement, a density algorithm suitable for the developed integrated logging instrument with the controllable neutron source was established in this study. Firstly, the modeling of the new integrated logging instrument and the formation was carried out with the Monte-Carlo numerical simulation software MCNP. Then, the interaction process between fast neutrons emitted from the D-T neutron source and formation materials under different lithology and porosity conditions was simulated using the built model. By recording the near and far inelastic scattering γ-ray counts and capture γ-ray counts, near thermal neutrons and near epithermal neutrons counts from neutron detectors, the response relationships between the detector count of the integrated logging instrument and formation density were obtained, and the main factors which affect the density algorithm were also analyzed. Finally, based on the analysis of the density response characteristics, a new algorithm for controllable neutron source density logging which uses the ratio of near thermal neutron counts and near epithermal neutron counts and the ratio of near and far inelastic γ-ray counts was proposed to improve stability and accuracy. The calculation results show that the calculated density of sandstone, limestone and dolomite is very close to the true density, and the relative error is less than 6%. Compared with the calculation results of Halliburton and Schlumberger algorithms, the method in this paper shows better results of less formula parameters, no detector absolute counting, and high accuracy.

    Selection of Energy Degrader Material for 100 MeV Protons
    HAN Jinhua;QIN Yingcan;GUO Gang;ZHANG Yanwen
    2020, 54(7):  1326-1331.  DOI: 10.7538/yzk.2019.youxian.0480
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    The energy degrader is of great significance for improving the efficiency of the proton single event effect (SEE) ground simulation tests, and the selection of the energy degrader materials is a primary problem in the energy degrader design. Four aspects that are of much concern in the proton SEE tests, i.e., the energy straggling, angle straggling, neutron background and induced radioactivity, produced by 100 MeV protons in the energy degradation process in the four common energy degrader materials (beryllium, graphite, aluminium and copper) were calculated, respectively. In terms of the calculation of the induced radioactivity, the types, activities and residual dose rates of the radionuclides generated in the materials were included. According to the above calculation results, combined with the requirements of the proton SEE tests, the applicability of the four materials as the energy degrader material for 100 MeV protons was analyzed and compared in the case of the same energy degradation. Finally, aluminium was selected as the energy degrader material for 100 MeV protons employed by the proton SEE ground simulation test facility on the 100 MeV proton cyclotron of China Institute of Atomic Energy.

    Temperature Distribution of Cryogenic Target and Lifetime of Ice Layer in Rapid Cooling Process
    CHEN Xun;LI Cui;LI Yanzhong;GUO Fucheng
    2020, 54(7):  1332-1339.  DOI: 10.7538/yzk.2019.youxian.0357
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    In order to achieve ignition for inertial confinement fusion successfully, it is necessary to reduce the temperature of the ice layer inside cryogenic target by 1.5 K. In the present study, the transient thermal characteristics of cryogenic target during the cooling process were numerically studied, and optimal cooling schemes were proposed to improve the target temperature distribution and to avoid ice quality deterioration. An unsteady numerical model was developed for the cryogenic target based on the Boussinesq assumption and UDF program. The transient thermal characteristics of the cryogenic target during the cooling were analyzed under different cooling rates and schemes. The results show that the maximum temperature difference of capsule increases sharply at the beginning of the cooling, and then tends to be a fixed value eventually. The decrease of cooling rate is conducive to restrain the rise of the maximum temperature difference of capsule and extend the lifetime of ice layer. Delayed cooling with specific delay time can improve the uniformity of the capsule temperature and increase the lifetime of ice layer, and there is an optimal delay time to maximize the lifetime of ice layer.

    Temperature and Excitation Power Dependent Analysis of Thermal Quenching Effect of Electron-irradiated GaAs Cell
    LIU Yanyu;WU Rui;WANG Junling;LIU Jun;YAN Gang;WANG Rong
    2020, 54(7):  1340-1344.  DOI: 10.7538/yzk.2019.youxian.0470
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    In order to further explore the thermal quenching effect of GaAs middle cell in electron-irradiated GaInP/GaAs/Ge triple junction solar cells, temperature dependent and excitation power dependent photoluminescence (PL) spectra for the sub-cell were measured, and the changes in PL intensity and its position were analyzed. PL intensity was fitted with temperature change using Arrhenius equation, and thermal activation energy (0.96 eV) of the non-radiative recombination center was determined. By analyzing thermal quenching effect of PL intensity in temperature range of 10-300 K, it was found that the relation between excitation power and PL intensity transforms from linear to quadratic dependence. Using the dependence, it was further verified that the main defect of GaAs middle cell is the non-radiative recombination center, and the mechanism of thermal quenching effect of electron-irradiated GaAs middle cell was explored.