In order to accurately evaluate the radiation dose distribution around the radioactive equipment in nuclear power plants, this study used the three-dimensional radiation dose field fast calculation software (DOLPHINS) to simulate typical nuclear power plant equipment. Based on the Monte Carlo method, this software can quickly and efficiently simulate radiation sources of different shapes and materials, and calculate their radiation dose distribution in three-dimensional space. The study selected pipes, pumps, valves and other equipment in typical nuclear power plants, and established their three-dimensional geometric models. By inputting the material information, radionuclide information, source intensity information and simulation parameters of the equipment, the DOLPHINS software simulated the radiation dose distribution around the equipment under different working conditions, and compared it with the traditional dose calculation method. The results show that the DOLPHINS software can quickly and accurately simulate the radiation dose distribution around the equipment, and its calculation results are consistent with the traditional method, and the simulation efficiency is higher, which can effectively reduce the workload of manual calculation. This study can provide a reference for the radiation protection design, dose assessment and personnel radiation protection of nuclear power plant equipment.

In the process of radioactive waste gamma scanning measurement, when the activity and medium are evenly distributed, the segmented gamma scanning measurement technology can obtain better reconstruction results. In actual measurement, due to the uneven distribution of activity, the use of SGS scanning is prone to introduce large errors. Based on the grid model division for efficiency calibration, the Tikhonov regularization method is applied to solve the activity and distribution of radionuclides in each grid. Compared with the traditional SGS method and the hierarchical iterative SGS method, it allows more accurate calibration and better measurement accuracy. It is suitable for scenarios where the activity distribution of the waste barrels under test is quite different and the hot spot distribution needs to be given through measurement.

Representative sampling of radioactive gaseous effluents from nuclear facility stacks is critical for accurate environmental monitoring and regulatory compliance. In this study, computational fluid dynamics (CFD) simulations were employed to assess the sampling representativeness under typical operational conditions. A threedimensional CFD model of a representative nuclear facility stack was developed, incorporating a standard k-epsilon turbulence model and a Discrete Random Walk (DRW) model to simulate gas-solid multiphase turbulent flows. Tracer gas (SF6) and aerosol particles were introduced to evaluate flow behavior and mixing uniformity. The study systematically examined the spatial distribution of axial velocity, cyclone angle, tracer gas concentration, and aerosol particle concentration across seven sampling cross-sections. Results showed that the coefficients of variation (COV) for gas velocity remained below 12.1%, and mean cyclone angles were under 7.3 degrees, meeting ISO 2889:2023 criteria. Tracer gas concentrations also displayed acceptable COVs at all sampling sections. However, aerosol particle concentrations at the lowest two sections (Q1 and Q2) exhibited COVs exceeding 20%, likely due to inertia-induced stratification, rendering them unsuitable for representative sampling. Based on combined considerations of ISO 2889:2023 compliance and practical constraints such as sampling line length and ease of installation, section Q4 is recommended as the optimal location for obtaining representative samples. This study demonstrates that CFD-based analysis offers an effective theoretical framework to guide the design and validation of representative sampling systems for radioactive gaseous effluents.

核设施气态流出物监测系统需要对放射性物质进行连续取样和测量，气溶胶受到布朗扩散、重力、湍流等作用会在取样管道内产生沉积。本文针对水平取样管道微米级气溶胶穿透效率开展实验，并与国内外现有的经验公式和发表数据进行了对比验证。考虑有效粗糙度、湍流扩散、重力沉降的共同影响，提出了适用于扩散区与扩散-碰撞区的沉降速率预测修正公式，预测值与实验结果较为吻合。实验结果表明：由于粗糙度会改变近壁处流场和沿程阻力，所以不同粗糙度对沉降速率有显著影响，微米级的粗糙度改变也可能会使湍流沉降速率变化一个数量级。水平管中处于扩散区的气溶胶，其沉降受重力和布朗扩散机制共同控制，具有较低的沉降速率。位于扩散-碰撞区的气溶胶，穿透率受重力、湍流涡和管道内滞留时间的共同影响，会存在最佳取样风速点，使得穿透率达到最大值。

The most important indicator of the radioactive waste classification and detection system is its identifiable minimum activity of low level radioactive waste, that is, the lower detection limit. In order to design a low level and very low level waste classification and detection system that can perform large-scale and rapid measurement, and explore the influence of various factors on the low limit of detection, this paper establishes the layout model of the detection unit of the classification detection system by Monte Carlo method to simulate the number of NaI crystals, the density of the detected object, the distance between the detection crystal and the detected object, the distance between the adjacent measured waste and the measurement time, etc. Meanwhile this paper analyzes the characteristics of the low limit of detection, and performs operations research analysis based on the principle of being able to detect radioactive waste with specific activity of not less than 0.1 Bq·g −1 , maximizing the shielding effect and optimizing the shielding weight, and designs a low level and very low level waste classification and detection system that can measure radioactive waste which density is not less than 0.25 g/cm 3 and which specific activity is not less than 0.1 Bq·g −1 .

In this paper, a portable gamma-ray spectrometer for real-time and in-situ gamma-ray detection applications is presented. By combining a quasi-hemispherical CdZnTe (CZT) semiconductor detector and a Geiger-Muller (GM) counter together, a wide dose rate range is achieved, ranging from 0.1 μSv/h to 100 mSv/h with a relative error of less than 10%. The GM counter is used to measure dose rate from 1 mSv/h to 100 mSv/h. With CZT, the spectrometer can provide a high energy resolution spectrum for nuclide identification and a high precision dose rate at low dose rates. The full width half maximum (FWHM) resolution is 2.2% at 662 keV below 70 μSv/h and is better than 3.3% at 3.8 mSv/h. The weight of the spectrometer is 3.2 kg for handheld and the runtime is up to 12 h without charging. For preliminary applications, the spectrometer was used to measure the gamma radiation around the Back-n white neutron beam line at China Spallation Neutron Source and around the steam generator in the nuclear power plant at Daya Bay Nuclear Power Station.

High Integrity Container (HIC) for nuclear power plant is an advanced waste reconditioning Container specially designed and manufactured. It has the characteristics of High strength, good sealing, strong chemical stability and thermal stability. It can be used to load a variety of Low and Intermediate Level radioactive wastes that have not been cured or fixed, such as dehydrated mud and evaporation residues, Waste resins and Waste water filtration cores. According to relevant national standards and nuclear power plant radioactive waste management regulations, it is necessary to measure the type of radionuclide and radioactive level of waste barrels. The sizes of HIC barrels are DI, DII and DIII, with volumes ranging from 0.42 m3 to 3.44 m3. There are various types of waste in the bucket, and the waste density is not less than 0.25 g·cm−3; The main material of HIC barrel is high density crosslinked polyethylene, which has poor shielding effect on γ-rays, resulting in high dose rate level on the surface of the barrel. Therefore, the existing γ scanning device for 200 L/400 L metal waste drums in nuclear power plants cannot meet the measurement requirements of HIC drums, and it is necessary to develop a special γ scanning device for HIC drums. In this study, Cadmium Zinc Telluride (CZT) semiconductor was used as the main detector to simulate the detection efficiency under different HIC barrel sizes, different waste density, different detection distances and different collimating opening diameters. The simulation results show that the CZT detector can be used for gamma scanning measurement of HIC drums, especially HIC drums with high dose rate, under the condition of proper shielding collimation and detection distance.

The accuracy of tomographic gamma scanning transmission reconstruction is a critical factor in reconstructing the activity of a radioactive drum. Traditional reconstruction algorithms produce severe grid artifacts and a high level of noise, thereby increasing the reconstruction error for both the density map and the activity. This paper proposes a novel algorithm for transmission reconstruction by combining maximum-likelihood expectation maximization and a convolutional neural network (CNN). Our experimental results indicate that the proposed reconstruction algorithm is capable of significantly reducing measurement errors, increasing spatial resolution while also eliminating grid artifacts, and being sufficiently robust when dealing with a noisy input image. The mean squared error of the output image decreased by 52.70% compared with the conventional reconstruction method, and the peak signal-to-noise ratio and structural similarity index improved by 21.89% and 17.33%, respectively. The spatial resolution was improved by 28 times, which demonstrates that CNN is a potentially useful new method for radioactive waste drum transmission image reconstruction.

The Plasma melting technology can reduce the low and medium level radioactive waste to a minimum. The emission of high-temperature radioactive aerosol generated in the process of treatment needs to be monitored. After treatment with a high efficiency filter and meeting national standards for gaseous effluents, it will be allowed to be discharged into the atmosphere. In order to achieve stable and reliable radiation monitoring, through the research on the radionuclide analysis of stack exhaust and the design of detector selection, the combined design of plastic scintillator & NaI scintillator is adopted. At the same time, the real-time and reliable monitoring of radioactive aerosols in stack gases under high temperature environment is also solved.

The Radiation Monitoring System (RMS) is designed to check continuously the reactor at the operating set points, it assures that no individual and persons receive an unnecessary or hazardous exposure to radiation. According to the process and operation characteristics of Small Modular Reactor (SMR), these radiation monitoring channels can be grouped into the process radiation monitoring subsystem, the effluent radiation monitoring subsystem, the area radiation monitoring subsystem and the Post Accident Monitoring Subsystem (PAMS). Based on the architecture design of RMS, the demand of signal interface and power interface is analysed. Through introducing the marine environment, the problems of anti-shock, vibration, inclination and swing of loads are put forward.

A miniaturized gamma-ray spectrometer, which is designed for the applications of gamma-ray detection, is described in this paper. It can be used in radiation protection, environment exploration and physics research. The detector consists of a CdZnTe (CZT) semiconductor and a BGO scintillator. With the structure of a CZT module assembled in the cavity of a cup-shaped BGO, the detector has the functions of rapid detection just using BGO, angular sensitivity and Compton suppression using CZT with BGO as a collimator and anti-coincidence detector. For the developed prototype spectrometer, the readout electronics was designed to sample the signal from the detector and to process the data on FPGA. It is well-coordinated with the detector. For Compton suppression, the energy threshold of BGO signal is set to 50keV and anti-coincidence time window is set to ±2μs. The design of the detector is evaluated by simulation and the performance of the spectrometer is tested. Simulated results are in agreement with test results. The energy resolution of 1.8% at 662keV has been achieved and Compton suppression factor has reached 30% for gamma-ray energy spectrum measurement with Compton suppression, which is helpful for identification of full energy peaks. The angular sensitivity of forward directions is higher than the side which is meaningful for search and preliminary location measurement of radioactive materials.

本文从硅半导体监测装置出发,解决了核电厂通风系统取样气体湿度大导致辐射监测通道故障频发的问题,介绍了探测器技术参数和主要性能,通过效率刻度和功能验证,计算得到效率系数Eft(10.255×10-3)满足要求,阈值报警功能和模拟量输出显示正常.

核电厂取样装置由KIT/KIC系统实时数据采集和监视,存在取样装置故障报警闪发后导致日志刷屏现象,影响核电厂正常监视功能.通过论证分析故障闪发的可能性,提出了后端数据采集系统优化方案,模拟仿真对比kvel 1层和Level 2层输出结果,实现采用Level 2层报警优化方案,验证了优化方案在KIT/KIC系统内的可行性.

It briefly introduces the basic component and maintenance point of Ionization Chamber Detection Device,the working principle and circuit analysis of pre-amplifier,and the circuit conditioning of troubleshooting detection device has been calculated and validated.The research gives the maintenance ideas and methods of this type of detection device,and it also has certain practical value for the maintenance staff.

Endostatin is a potent inhibitor of tumor angiogenesis. Interestingly, nonnative endostatin also exhibits an anti-tumor effect, which remains a mystery so far. Here, we show that intravenous injection of nonnative endostatin results in tumor inhibition effect. Soluble and active endostatin is isolated from human blood after the addition of nonnative endostatin in vitro. By fractionation of the whole blood, we surprisingly identify fibrinogen specifically binding to and inhibiting the aggregation of nonnative endostatin. Moreover, the anti-tumor activity of nonnative endostatin is substantially impaired in fibrinogen-deficient mice. Our studies demonstrate that fibrinogen facilitates the anti-tumor effect of nonnative endostatin, which also provides new insights into the novel physiological function of fibrinogen.