As a strong candidate for the new type of non-volatile memories and artificial synaptic devices, memristor has a huge development prospect in aerospace, Mars exploration and other space science and application fields. Once large-scale application of memristor requires extremely stringent radiation resistance performance for the memristors. In order to improve the radiation resistance of memristors, it is necessary to explore the radiation effect mechanism and develop an effective radiation resistance technology. This paper summarizes the research status and trends of irradiation effects on memristors, describes the mechanism and analysis method of irradiation damage of memristor, and focuses on the irradiation effects of the memristors with transition metal oxide material system. Additionally, the possibility of scientific problems and key technologies are discussed, so as to provide some ideas for the radiation hardening and space application of memristor.

Background Superconducting undulator (SCU) prototype with small magnet gap of 5 mm, long magnet length of 4 m and high magnet field of 1.58 T was being developed at Shanghai High Repetition rate XFEL and Extreme light facility (SHINE). Compared to any other superconducting undulator, there is no cryocooler being installed on the cryostat in this SCU prototype. Purpose This study aims at the cooling design for the binary current leads for SCU's normal operating. Methods Binary current leads composed of normal conductive copper leads and high temperature superconducting current leads (HTS) were adopted for SCU to connect superconducting coils inside the cryostat and outer cables. Low-temperature helium gas was used to transport independent refrigerator system to the cooling tubes inside the prototype, hence the binary current leads were cooled. Thermal conduction components installed on the middle of the thermal shield were employed to transfer heat load of normal conductive copper leads, and heat load of copper leads was optimized by simulation. Auxiliary superconducting rods were designed for connecting cold ends of HTS in the cryostat test. Results The temperature difference between hot ends of HTS and low-temperature helium gas is less than 20 K from the result of cryostat test, all binary current leads is operating normally with full current. Conclusions It is practicable to use cooling tubes with low-temperature helium gas to cool binary current leads of the SCU prototype by thermal conduction, which is different from cooling solution for current leads in any other SCU being developed presently.

Background The expected signal rate of rare decay experiments, such as instance dark matter and neutrino less double beta decay experiments, is extremely low, which requires that the detector building materials have extremely low radioactivity. Low radioactive background control is one of the essential works in the rare decay experiments. 226Ra and 228Ra produced in the early decay chain of 238U and 232Th have low boiling point and high vapor pressure, removing the element Ra can break the 238U decay chain and keep a low radioactivity of 232Th-late for a long time. Purpose This study aims to investigate vacuum melting technique for low background titanium to reduce the impurity of isotopes that have negative impact on rare decay experiments and creating a low background environment for the detector running. Methods Firstly, the low background material samples were acquired by manual separation of radionuclides using physical and chemical methods. Then, radioactive impurity elements, such as K, Cs, Ra, Pb, Po and Rn with low boiling point and high vapor pressure, were volatilized in environment with high temperature and high vacuum level. Finally, radioactivity of these testing samples were measured by two sets of high-purity germanium γ spectrometer with measurement time extended to 7 days. Results Measurement results show signs of removal of radioactive isotopes by smelting-vacuum method, and the impurity in pure titanium smelted in vacuum electron beam furnace can reach the levels of (0.13±0.69) mBq∙kg-1 for 232Th-228Ac, and (0.07±0.29) mBq∙kg-1 for 238U-222Rn, respectively. Conclusions The smelting-vacuum method could provide reliable low background material for the container of the next generation PandaX detector.

Background Vertical cavity surface emitting lasers (VCSEL) have very high application value in space radiation environment. Purpose This study aims to explore the degradation rule and mechanism of 850 nm VCSEL in harsh radiation environment. Methods First of all, the MULASSIS tool was employed to calculate displacement damage dose (DDD) and design experimental scheme for 850 nm multimode VCSEL samples. Then, 3 MeV and 10 MeV proton irradiation experiments were conducted to obtain the degradation rule of parameters such as light output power and threshold current with the proton fluence, and to find that the degradation degree of light output power and threshold current were equal under the same DDD. Finally, the Silvaco software was used for modeling and simulation on an experimental basis to extract microscopic parameters such as trap density, donor and acceptor ionization density, mirror loss, radiation recombination rate and photon number. Results The simulation results are in good agreement with the experimental results, these results show that each parameter changes to different degrees with the increase of proton fluence. Conclusions The parameter degradation law and radiation damage mechanism of VCSEL can be deeply explored by simulation on the basis of the experimental law, and simulation results are of great significance for understanding the degradation mechanism of VCSEL.

Background Betatrons are widely used in non-destructive testing, cargo and vehicle security inspection systems, whilst the filament power supply is an important core component of small size betatron. Existing filament power supplies are not suitable for small size betatron. Purpose This study aims to develop a novel electron gun filament power supply for small size betatron to produce optimal radiation intensity. Methods An electron gun filament power supply composed of a filament duty cycle adjustment circuit and an injection current feedback circuit were designed on the basis of operating characteristics of a small size betatron. Two pulse-width modulation (PWM) signals were output from the digital signal processor (DSP) to drive the half-bridge circuit. Positive and negative alternating voltage pulses were output by the half-bridge circuit to the primary of the isolation transformer, and the secondary voltage of the transformer was loaded to the filament for heating of the filament. By collecting the betatron tube wall current and target current as the injection current feedback signal, the duty cycle of the filament voltage pulse was adjusted by DSP according to the feedback signal, and realizes the adjustment of the filament emission current, so that the current injected by the filament into the acceleration tube was kept at the optimal value. Results The experimental test results show that the filament power supply can keep the betatron output dose rate in the best state, and the output dose rate stability is better than 11.3%/10 min. Conclusions The filament power supply meets the requirements of small size betatron with advantages of good stability, wide adjustment range and small size. It can be applied to the small size betatron developed by the Engineering Research Center for Nuclear Technology Application of the Ministry of Education of East China University of Technology.

Background The measurement of time-of-flight is one of the indispensable experimental contents in contemporary high-energy physics experiments and plays a vital role in exploring the essence of particle physics. Purpose This study aims to design a time to digital convertor (TDC) chip that meets the high-resolution time measurement requirements of time-of-flight detectors for high-speed flying particles in high-energy physics experiments. Methods First of all, a differential structure TDC was proposed and the main measurement part was realized by differential delay loop composed of time measurement core module, time measurement data transmission module, delay loop calibration module and clock generation module. Based on this structure, three parts of delay loop module, thermometer code generation module, and coarse count and fine count generation module were integrated into the core module of time measurement, and the 0.18 μm SMIC (Semiconductor Manufacturing International Corporation) process was adopted to achieve the TDC chip design. Results & Conclusions The designed TDC chip has a layout area of 1.35 mm×1.35 mm, a resolution of 17 ps, an accuracy of 8.5 ps (Root Mean Square, RMS), and a dynamic range of 0~210 μs. It can meet the current requirements for high-precision time measurement in high-energy physics.

Background In order to fast and conveniently measure X, γ and neutron radiation field simultaneously, portable multi-function radiation detector is highly demanded. Purpose This study aims to design a portable multi-function radiation detection system based on LaBr3(Ce) crystal, lithium aluminum silicate oxygen (LASO) neutron detector and high range Geiger-Muller (GM) counter. Methods After the optical signal output from the LaBr3 crystal was photoelectrically converted into electronic signal by the photomultiplier tube (PMT), the integrated digital multi-channel was used for data acquisition of electronic signal, and the subsequent data processing and calculation. The front signal processing circuits, such as amplification, discrimination and shaping, were designed for both the LASO neutron detector and the GM tube counter. Finally, the digital signal processed of the LaBr3 detector was transmitted to the ARM (Advanced RISC Machine) processor in the form of TTL (Transistor-Transistor Logic) serial port, and the pulse signal formed of the neutron detector and GM counter tube was connected to the external counting port of ARM processor. Energy spectrum was processed for nuclide identification and displayed by ARM processor whilst the low dose rate measurement base on the Gamma data collected by the LaBr3 detector, the neutron detector and the high range GM tube counter were counted at fixed time and converted from the count rate into the dose rate. Results & Conclusions The designed portable multi-function radiation detector realize simultaneous measurement of wide range (48 keV~3.0 MeV) γ, low-energy (48 keV~1.25 MeV) X-ray, (0.1~100 mSv∙h-1) neutron dose rates, and nuclide identification capability of LaBr3 spectrometer, and upload the data to PC through USB interface.

Background Compared with transistors and small-scale integrated circuits, the total ionizing dose (TID) effect and testing of multifunctional large scale integrated microprocessors are more complex. The difficulty of testing is to analyze the failure mode of microprocessor online from limited information under irradiation. Purpose This study aims to develop an extendable on-line test system for TID effect of microprocessor and carry out preliminary application. Methods The testing system was composed of control circuit, extendable signal acquisition circuit, tested sample interface, upper computer and software. Multiple parametric measurement or functional verification methods such as power consumption current, on-chip memory, communication, clock, analog-to-digital/digital-to-analog converter (ADC/DAC) and direct memory access of microprocessor were provided. Sixteen microprocessors with feature size 40 nm were irradiated with 60Co source and tested on-line. Results After the irradiation dose is accumulated to (377.44±20.34) Gy(Si), all samples are malfunctional with digital communication interruption, sudden drop in current consumption, abnormal ADC/DAC output and so on. Conclusion Based on all 12 kinds of parametric measurement or functional verification results, the TID effect of this type of microprocessor is probable to be a functional failure caused by some kernel instructions. The on-line test system of this study can provide more direct data information for the total dose failure mode analysis.

Background The radiation accident emergency drill is an important way to maintain and improve the response ability of emergency organizations, but at present, the evaluation of the drill is mainly based on the qualitative evaluation of the surface performance of the key links, there is not yet a systematic, full-flow, accurate and targeted evaluation method of exercises. Purpose This study aims to improve the quality and real-time evaluation of radiation accident emergency drill. Methods Firstly, the alternative evaluation indicators were summarized on the basis of literature survey. The method of expert investigation was used to check the alternative indexes and establish the evaluation index system. Then, the analytic hierarchy process (AHP) method was applied to the calculation of weight distribution, and the fuzzy comprehensive evaluation (FCE) was employed to build comprehensive evaluation model for radiation accident emergency exercises. Finally, based on the AHP-PCE, comprehensive scores of the evaluation indexes at all levels of three typical examples were calculated. Results The FCE-based model synthesizes the stage of the drill preparation, the drill implementation and the drill summary, and forms the general goal of the radiation accident emergency drill effect. The evaluation system of radiation accident emergency maneuver is set up with 5 first-level indexes and 15 s-level indexes to realize the comprehensive evaluation of the radiation accident emergency drill. Conclusions The example verification shows that the AHP-FCE model has a good adaptability and rationality, and has a good application value in the evaluation of radiation accident emergency drill.

Background The flow distribution in core for the liquid fuel molten salt reactor (MSR) is an important part of the thermal hydraulic design, and the hydraulic structure of reactor core plays a decisive role on flow distribution. Purpose This study aims to find out a suitable hydraulic structure design to make the core flow distribution match with the power distribution, and flatten the core temperature distribution for a 10 MW MSR. Methods First of all, a one-twelfth core model of liquid fuel MSR was established. Then, ANSYS FLUENT16.0 software was employed to conduct three-dimensional numerical simulation of the flow field. The influence of hydraulic structure of reactor core was analysed by changing the structure of upper plenum, downcomer and lower plenum, and the corresponding flow distribution characteristics of the core are obtained. Finally, a suitable structure was proposed after step-by-step improvement. Results The simulation results show that increasing of the height of upper plenum can balance the flow distribution between central and peripheral channels, increasing of the width of the downcomer can reduce the vortex flow at the downcomer outlet and flattens the flow distribution at the same time. The cylindrical lower structure with shroud in lower plenum can restrain the vortex to a certain extent and make the flow distribution more gentle. Based on the analysis above, a reasonable hydraulic structure is proposed for the molten salt reactor. Conclusions The results of this study provide important reference for the further optimization design of liquid fuel molten salt reactor.

Background The corrosion of the secondary circuit has remained a challenging problem influencing the security and efficiency of a nuclear power plant. In an actual operation, an amount of alkalizer is usually added into the secondary circuit water to adjust its pH value to alleviate the corrosion of the pipelines. However, the very complex real working condition results in a significantly inhomogeneous distribution of the pH values, and the enclosed nature of the secondary circuit have frustrated various efforts to control the precise pH values at some key sites inside the circuit. So far, pH values around such key locations have been roughly estimated by either external simulating experiments or by using patented commercial software of foreign companies. However, it is difficult for such external simulations to take into account the various important working conditions. Purpose This study aims to develop a method without specific assumptions or uncontrolled approximations to calculate the distribution of pH values in the secondary circuit under its working conditions. Methods Firstly, the complex and repeated structure inside the steam generator was simplified and simulated by using one direct tube model. Except for the length, the other geometrical dimensions, support plates as well as the tube material etc. were chosen to be as similar as possible to the actual ones. Then, all temperature-dependent equilibriums involving H+ in water solution of the secondary circuit were considered with additional consideration of the equilibriums of the alkalizers in the gas and liquid phases. As an essentially field quantity defined on the space of the secondary circuit, these pH values together with other relevant parameters, such as temperature, fluid velocity etc. were depended on the position coordinates, and were calculated by using the finite element method coded in the COMSOL package. Finally, the boundary conditions of flowing rate and temperature of the water at the inlet were set as 1.0 m·s-1 and 543.2 K, respectively, and the pressure at the outlet was set as ~6.8 MPa, a stepwise linear heat flux model was used to simulate the thermal energy transfer from the primary side to the secondary one. The bubbly-flow model was used to simulate the actual steam-water fluid in the secondary side, which was assumed to be in a steady state working condition. Results The calculated pH field under the working conditions shows clearly an inhomogeneous distribution, e.g. ΔpH = ~ -0.6 from z = 0 to 3.8 m, due to the influences of the tube support plates, the temperature and the heat transfer, etc. The investigations on the ammonia/ethanolamine (ETA) binary alkalizer with different total concentrations of various NH3/ETA molar ratios show a better enhancing effect of ETA over ammonia for the pH value (ΔpH>~0.14), and reveal a saturation effect (molar ratio NH3:ETA≤ ~1:4). Conclusions The distribution of the pH values in the realistic working conditions can be calculated without resorting to empirical formulae and uncontrolled approximations. The developed method and the calculated results provide valuable information for solving the corrosion problem in the secondary circuit. The method and the model can be extended to simulate the more realistic conditions of a nuclear power plant.

Background Radiation shielding design is an important part of reactor design, and the development of new nuclear power technology for various kinds of reactors has put forward new demands on radiation shielding optimization design methods. Purpose This study aims to overcome the shortcomings of the traditional multi-objective optimization methods for shielding structures in dealing with the optimization problem of 3D shielding structures, such as slow optimization speed, difficulty in convergence, and poor globalization. Methods Based on the non-dominated sorting genetic algorithm Ⅲ (NSGA-Ⅲ), the many-objective optimization method for 3D shielding structure design for nuclear reactor was proposed. The Monte Carlo N-Particle Transport Code (MCNP) was employed to analyze comparative performance of the NSGA-III optimization method on the basis of the 3D shielding structure model of nuclear reactors, and shield weight, volume and radiation dose rate in specific regions were taken as the optimization targets. Results & Conclusions The numerical simulation results show that the NSGA-III based optimization method for 3D shielding structure design can search for the Pareto-optimal front more efficiently and stably, providing a new idea for the optimization of radiation shielding design.

Background Heat pipe cooled reactors (HPR) have good inherent safety. In the early stage of core design, heat pipe failure accident is usually one of the design basis accidents that need to be considered. Purpose This study aims to analyze the neutronic-thermalhydraulic coupling performance of a new type of megawatt heat pipe reactor. Methods Firstly the heat pipe cooled reactor system physical models, including the point kinetics model, the core and heat pipe model and radiation heat transfer model for the inner core cavity, were established according to the designed HPR prototype composed of heat pipe stack and supercritical CO2 Brayton cycle system with thermal power of 3.5 MW. Then, the finite element software FLUENT was employed to conduct neutronic-thermalhydraulic coupling calculation for the three-dimensional reactor core under the steady-state and heat pipe failure accidents. Finally, the core safety performance was evaluated by comparing the peak temperature of each component with the melting point of material. Results & Conclusions The results show the designed HPR has good safety performance under the steady state and single heat pipe failure. Radiation heat transfer in the core cavity cannot be ignored in the serious cascade three heat pipe failure accident in high power region. Meanwhile, the design cannot withstand cascading four heat pipe failure. By comparing the peak temperature of the multiple heat pipe failure with the peak temperature of the single heat pipe failure, it shows that the design has good inherent safety.