131I is a typical fission product in nuclear reactors. It has a strong affinity for the human thyroid gland and is an important nuclide of concern in radiation protection under both normal and accident conditions of reactors. Accurate measurement of its radioactivity is of great significance. The main measurement method involves collecting 131I in aerosols using an activated carbon adsorption device and then analyzing it with a γ spectrometer. This study proposes a bidirectional ratio method to obtain the γ detection efficiency through integrating the depth distribution coefficient of 131I in activated carbon. Compared with the traditional methods of preparing standard uniform volume sources or using destructive analysis, this experimental method is more reasonable and reliable. In addition, to verify the reliability of this method, experimental verification of the proportional coefficient of bidirectional measurement was conducted, and experimental work on the correlation analysis between sampling flow rate and distribution coefficient was provided. The experiments show that the detection efficiency of the iodine cartridge source with a real depth distribution coefficient obtained by the volume source integration method incorporating a probability function can quickly and conveniently obtain the γ detection efficiency of non-uniformly distributed iodine cartridge sources, providing a feasible approach for accurately measuring the radioactive activity of the fission product 131I in the environment.

In this work, an approach combining radionuclide tracing with Monte Carlo simulation was developed to evaluate the corrosion behavior of 316L and T91 steels and to reconstruct the distribution of released corrosion products in a tubular static lead-bismuth eutectic (LBE) system. Radionuclides were introduced into the specimens by proton irradiation activation, and the 846.8 and 1238.3 keV γ-rays of 56Co were used as tracer signals. Net count rates measured at different axial positions were used to construct the observation vector, while contribution coefficients of different source segments to the detector response were calculated using Monte Carlo simulation. The radionuclide activity distributions were then reconstructed by solving the inverse problem using an iterative weighted non-negative least-squares algorithm. The reconstructed pre-corrosion distribution agreed well with expectations, with the activity mainly located in the specimen, confirming the reliability of the method. After 312 h of corrosion, the activity fraction in the specimen decreased, accompanied by the appearance of measurable activity in the LBE, indicating release of 56Co from the specimens into the liquid metal. The released activity was mainly confined within approximately 3 cm from the specimen end. Comparatively, 316L exhibited a higher fraction of released activity than T91, suggesting a stronger tendency toward dissolution-related corrosion under the present conditions. These results demonstrate the feasibility of radionuclide tracing combined with Monte Carlo simulations for quantifying corrosion and migration behavior in LBE systems, providing a methodological basis for in-situ monitoring techniques in LBE systems.

The cross-section for the reaction 209Bi(n,gamma)210gBi was measured at energies around 14 MeV. The neutron fluence standard corresponds to the cross-section of the monitor reaction 58Ni(n,gamma)57Ni and 58Ni(n,p)58 m+gCo. The neutron irradiation facility is from the High-intensity d-T Fusion Neutron Generator (HINEG) at the Hefei Institutes of Physical Science, Chinese Academy of Sciences. A high-purity bismuth sample was used as the target sample. The T (d,n)4He reaction was used as a neutron source to generate neutrons with an energy of 14 MeV. We have developed an electrochemical method for processing irradiated bismuth samples, achieving high enrichment of the low-yield activation product 210Po. This approach, combined with a low-background alpha-detection system, has significantly enhanced the accuracy of nuclear data for the 209Bi(n,gamma)210gBi reaction cross-section within the 14 MeV energy region. The experimental results align with theoretical predictions and are notably lower than previously reported values. These new values will enable the release of 210Po in lead-based reactors containing 209Bi to be evaluated more precisely.

Accurately predicting the highly toxic alpha-emitter 210Po (with a half-life of 138.4 days) in the primary coolant is a core challenge in the radiation safety analysis of Lead-Bismuth Eutectic (LBE)-cooled Fast Reactors (LFRs). LBE-cooled fast reactors are among the most promising candidate reactor types in Generation IV nuclear energy systems, where the 209Bi(n,gamma)210gBi reaction serves as the primary pathway for 210Po production. The 210gBi generated from this reaction decays rapidly into 210Po, thus making the precise determination of its thermal neutron capture cross-section of crucial importance. In this study, high-purity bismuth targets (99.995 %) were irradiated in the Xi'an Pulsed Reactor (XAPR). The decay product 210Po was selectively enriched on silver electrodes via spontaneous deposition, and the thermal neutron capture cross-section measured using a low-background alpha spectrometry system was determined to be sigma th = 16.12 (0.94) millibarns (mb). Additionally, the thermal neutron capture cross-section obtained by the liquid scintillation counting method in this paper was sigma th = 16.06 (1.09) millibarns (mb). The results from both methods are in good agreement with historical data and highly consistent with JENDL/AD-2017, the only major evaluated nuclear data library currently incorporating this reaction. The offline alpha measurement technique proposed in this paper has successfully achieved high-precision detection of extremely low activation yields (on the order of millibecquerels, mBq) generated by short-term irradiation, providing critical references for updating the nuclear data libraries of LBEcooled fast reactors, evaluating coolant toxicity, and guiding shielding design.

Lead bismuth eutectic alloy (LBE:44.5 wt%Pb+55.5 wt%Bi) is an important coolant candidate for fourth-generation reactors due to its good physical and chemical properties. However, the risk of radioactive polonium radiation protection caused by liquid lead-bismuth alloy as the main coolant of the reactor has become an important problem that must be paid attention to in the development of the lead-bismuth reactor. The 209Bi(n,γ)210gBi reaction is a key source term for 210Po generation in lead-bismuth reactors. Accurate measurement of the 209Bi(n,γ)210gBi cross section is important for assessing the risk of radioactive polonium release from lead-bismuth reactors. At present, the reaction cross section can be measured by the online method of directly measuring the cascade prompt gamma rays of the capture reaction or the offline method of measuring 210gBi decay product 210Po. To solve this problem, a210Po self-plating enrichment method based on isotope dilution α energy spectrometry with electrode movement is proposed. By optimizing the key conditions of the self-plating process, the enrichment rate of 210Po is effectively improved, which will help solve the problem of insufficient sensitivity of the direct measurement method of 209Bi(n,γ)210gBi low yield. The experimental results show that this technique is suitable for 209Bi(n,γ)210gBi post-processing measurement, and is of great significance for the study of neutron irradiation products of lead-bismuth eutectic alloys.

The radiation protection risk posed by radioactive polonium, arising from the use of liquid lead-bismuth alloy as the primary coolant in reactors, has become a critical issue demanding attention in the development of lead-bismuth reactors. The neutron capture reaction of Bismuth-209(209Bi) serves as the primary pathway for generating Polonium-210(210Po). However, due to the scarcity of experimental data and dated measurements, significant discrepancies exist in the nuclear cross-sections above 1 MeV, necessitating careful consideration. This paper proposes a novel measurement approach that directly determines the cross-section by measuring the alpha decay signature of the secondary activation product, 210Po. This method effectively circumvents issues associated with measuring the beta signal from Bismuth-210 (210Bi), such as susceptibility to environmental background and compromised reproducibility. The fast-neutron cross-section on the order of millibarns (mb) results in 210Po production yields as low as millibecquerels (mBq). Furthermore, due to the limited range of alpha particles, radiochemical processing prior to measurement is necessary. This study presents a spontaneous deposition method, achieving a high enrichment of 210Po directly within the medium of the dissolved bismuth target. Isotope tracing techniques were employed to compare results with well-established nominal values, thereby evaluating the reliability of the method and highlighting its significant advantages over measurements relying on the low-intensity gamma signal of 210Po. The results demonstrate the feasibility of the technical route for directly measuring the cross-section via the alpha decay signature of extracted 210Po, thereby establishing a novel approach for determining the neutron capture cross-section of 209Bi.

Ac-225 is a promising radionuclide for clinical α-particle targeted cancer therapy. Accurate measurement of its activity is prerequisite for clinical applications. To explore viable methods for Ac-225 activity measurement in clinical settings, this study used a CdZnTe detector to measure the γ-ray spectra of Th-229 solutions, investigated factors influencing measurement results, and developed efficiency calibration methods. Results indicate that: 1. Only the 440.46 keV γ-ray emitted by daughter nuclide Bi-213 and the 218.19 keV γ-ray emitted by Fr-221 are suitable for analyzing Ac-225 activity; 2. Significant discrepancies (up to 36.8%) exist in γ-ray emission intensities of Ac-225 and its daughters across different databases, necessitating emission intensity corrections in calculations; 3. When calibrating CdZnTe detection efficiency using the Eu-152 standard source method and Monte Carlo simulation, deviations from reference values were 2.0% and -4.6%, respectively, confirming the reliability of both methods; 4. The impact of cascade coincidence-summing effects on CdZnTe measurements is negligible, whereas it can reach 32.4% for HPGe detectors. These findings provide critical references for establishing standardized techniques to measure Ac-225 activity in radiopharmaceuticals using CdZnTe detectors.

使用传统的蒙特卡罗方法计算CdZnTe探测器探测效率时存在效率损失。为了将无源效率刻度技术应用于CdZnTe探测器,提出了基于理想模型（仅考虑光子与探测器晶体和结构材料相互作用过程）的CdZnTe全能峰源探测效率计算方法。使用 152 Eu标准源样品验证了计算方法的可靠性,结果表明,计算结果与参考值的偏差不大于5%。该方法不需要考虑CdZnTe晶体内载流子不完全收集的复杂过程就可以得到可靠的效率计算结果,可操作性强。

226Ra is a highly toxic radionuclide, and according to the Chinese national standard GB 5749—2022, its activity in drinking water must not exceed 1 Bq/L. Additionally, 226Ra serves as a geochemical tracer in oceanographic studies. To meet the monitoring requirements for 226Ra, this paper systematically summarize the pre-concentration and radiochemical separation methods for 226Ra in water samples, with a focus on discussing the advantages and disadvantages of radiation measurement techniques and mass spectrometry(MS) techniques for analyzing 226Ra. Pre-concentration methods include co-precipitation, evaporation, and adsorption. Co-precipitation, with its simplicity and cost-effectiveness, is well-suited for treating environmental water samples of up to 10 L in volume. For large volume water such as oceans and lakes, manganese polymer adsorption enables in-situ enrichment of Ra isotopes. Radiochemical separation methods include co-precipitation, solvent extraction, ion-exchange chromatography, and extraction chromatography, with the choice of method depending on subsequent analytical techniques. Radiometric methods such as radon emanation, alpha counting, and liquid scintillation counting(LSC) only require simple separation using coprecipitation and solvent extraction. However, alpha spectrometry and MS analysis demand more rigorous Ba removal using chromatographic techniques. Ion exchange chromatography is cost-efficient and widely used in laboratories, while novel extraction chromatographic columns(e.g., TK100 and AnaLig®Ra-01) enable selective Ra adsorption, simplifying separation workflows. Radiometric methods are widely applied for 226Ra monitoring due to their high sensitivity and low cost. LSC, combined with extraction chromatography, can enable automated 226Ra monitoring. Alpha spectrometry offers detection limits below 1 mBq/L, meeting the requirements for routine environmental monitoring. MS techniques, such as ICP-MS, provide rapid analysis, with each sample taking approximately five minutes, making them suitable for emergency assessments, though their high-cost limits routine use. Future research should focus on developing low-cost, high-selectivity extraction materials to streamline separation processes. Integrating radiometric methods with MS techniques could further enhance analytical efficiency and accuracy, supporting both environmental monitoring and emergency response efforts.

To accurately measure 223Ra dichloride injection by HPGe γ spectrometry, three calibration sources, including mixed mono-energy γ nuclide, 133Ba, and 226Ra, were used to calibrate efficiency at close/far source/sample-detector (S-D) distance. The results show that the efficiency calibration at a far S-D distance can reduce the coincidence effect of the 133Ba and 223Ra simulation source, both far and close S-D are applicable to the 226Ra simulation source; Using 269.48 keV, 271.23 keV and 351.03 keV γ-ray full energy peaks for analysis, the deviation of the results between the simulation sources method and the mono-energy γ nuclide source method is less than 1%. The influence of γ-ray emission probabilities, the selection of full energy peak, and coincidence effect and other factors on the 223Ra measurement results was also studied, the γ-ray emission probabilities from different nuclear databases can cause up to 30% deviation in the measurement results.

Negative-ion powders containing naturally occurring radioactive materials (NORM) can be traded freely in several countries. To assess the radiological risk of negative-ion powders in China, we used gamma spectrometry, EDS spectrometry, and alpha spectrometry to analyze twelve negative-ion powders purchased online. The results have shown that this batch of samples containing NORM originated from the mining tailings. The activity concentrations for 232Th, 238U, and 226Ra in samples were in the range of 17–339 Bq·g−1, 2.8–36 Bq·g−1, and < 63 Bq·g−1, respectively, which far exceeded the exemption limit, 1 Bq·g−1 for each radionuclide in the 238U and 232Th decay chains.

This study presented the determination and distribution of 210 Pb in plants using Liquid Scintillation Counting. The analysis method of 210 Pb in plants has been improved by optimizing the pulse shape analysis setup,constructing quenched correction curves, as well as consideration of the counting from both 210 Pb and its progeny 210 Bi. This method was used to test for the determination of 210 Pb in samples of three species collected from two sites around a uranium mining facility.

The accumulation of 210 Po generated by neutron-irradiated lead–bismuth eutectic (LBE) is a noticeable issue for the radiation protection of the generation-IV reactor. In this study, FISPACT-2007 was used to simulate the activation of 210 Po in neutron-irradiated LBE. The specific activity of polonium-210 was measured by α spectrometry after deposition onto a silver disc. The experimental results indicate that the specific activities of 210 Po are anastomotic with the simulation value, which provides a basis for further research on evaluating the evaporate behavior of 210 Po in LBE.

This study investigated a method of measuring the absorbed dose rate of γ-ray in air using a Φ7.5 cm×7.5 cm NaI(Tl) detector. Simulation of the NaI(Tl) detector using the Monte Carlo method was performed to obtain the γ-ray energy spectrum of the detector in the energy range of 50 keV~2.5 MeV. The G(E) function was used to calculate the air-absorbed dose rate of radioactive sources. When solving the G(E) function, the influence of the Kmax and the optimization factor M on its calculation of the air-absorbed dose rate was considered, and the solution of the G(E) function was optimized to improve the calculation accuracy. The relative deviation S of the air absorbed dose rate calculated by the optimized G(E) function from the theoretical standard value was less than ±1%. Finally, after the comparison of experimental measurements, it was found that the relative deviation between the calculated air absorbed dose rate of the G(E) function and the measured results of the dose rate meter was less than ±10%, indicating that the optimized G(E) function value of the NaI(Tl) detector can be used for the application of air absorbed dose rate measurement.

The radiolytic behaviors of aqueous ammonia solutions under γ-ray irradiation were investigated, focusing on several factors, including initial ammonia concentration, absorbed dose, and H3BO3 concentration. The main species including ammonia, H2O2, nitrogen oxides, and H2, were quantitatively analyzed. The results revealed the significant influence of the initial ammonia concentration on the radiolytic behavior of the aqueous ammonia solution. The existence of ammonia in the solution exhibited suppression of the concentrations of H2O2 and O2. Furthermore, the radiolysis of aerated aqueous ammonia solution led to the formation of nitrogen oxides (NO2− and NO3−). This process was driven by multiple oxidation reactions, in which O2 and ·OH played crucial roles. Interestingly, the addition of H3BO3 provided protection to ammonia during γ irradiation, while the concentration of H2O2 was increased.

利用超大流量气溶胶采样器和γ谱仪,对安徽三个点位气溶胶中人工放射性核素137Cs活度浓度进行了为期两年的采样监测,共六次检出超过探测限的137Cs,活度浓度为1.4～6.5 μBq/m3.结合国家辐射环境监测网现有数据,对安徽地区气溶胶中137Cs活度浓度与季节、空气条件关系进行了分析,为深入研究大气中相关放射性核素浓度分布规律奠定了基础.

In this study, we used the integral method of the probability density function to evaluate the detection efficiency of an inhomogeneous iodine cartridge source by fitting the function relationship between the efficiency of the 131I γ-ray and the height of the disk source from the detector. We constructed a mathematical model by combining the efficiency ratio of the positive and negative placement, K, and the distribution parameter, α, to determine the activity of 131I in the iodine cartridge. To verify this method, we analyzed the inhomogeneous iodine cartridge samples obtained from a gaseous iodine collection experiment. The average deviation of the total activity of 131I, calculated by using 131I γ-ray (364.48 keV), in the iodine cartridge between our method and the homogeneous iodine cartridge source obtained by employing disassembly-stirring-backfilling was 5.71%, and the maximum deviation was 8.54%. The results showed that the disk source calibration method of the probability density function can be used to determine the detection efficiency of inhomogeneous iodine cartridges, thereby providing a feasible method for monitoring gaseous iodine.

采用蒙特卡罗模拟计算的方法,标定能够适用于不同HPGe探测器上的体源虚拟点源位置,需要对点源、体源的探测效率进行模拟计算.通过241Am、137Cs、60Co点源和体源研究了HPGe晶体尺寸、类型对它们的虚拟点源位置的影响,模拟结果表明241Am虚拟点源位置随着探测器尺寸、类型的不同呈现明显的差异性,说明虚拟刻度原理对于探测器表面的小体积样品测量在低能区间是不可取的,最后采用137Cs、60Co源得出体源高度与虚拟点源位置的半经验公式.通过实验室两台HPGe探测系统对尺寸Φ70 mm×65 mm标准土壤体源的探测效率进行计算,与模拟探测效率值、虚拟点源效率表征结果对比分析,验证了拟合一组表征虚拟点源位置的半经验公式是可行的.实验结果表明,对γ能量区间在300～2000 keV进行体源的虚拟刻度时,选择可靠的定标源即可建立体源和虚拟点源位置的关系.这为解决监测工作中样品与探测器之间重复进行探测效率校准的问题提供了新的途径.

To determine the optimal electrodeposition conditions for actinium in a nitric acid/ethanol electrolyte, the 228Ac nuclide was used for electrodeposition experiments in a nitric acid/ethanol electrolyte in this study. First, under a current of 0.1 A, the effect of the electrodeposition time on the electrodeposition of actinium was studied. Subsequently, under different currents, the effect of the direct current on the electrodeposition of actinium was studied (T=35 min). The effect of different acid concentrations on the electrodeposition of actinium was then studied (I=0.15 A, T=35 min). The influence of different alcohol-water ratios on actinium electrodeposition was also studied. The results showed that the electrodeposition yield of 228Ac could reach more than 90% under the following electrodeposition conditions: nitric acid/ethanol as the electrolyte, an electrodeposition time of 35 min, a current of 0.15 A, a nitric acid concentration of 0.05 mol/L, an ethanol-water ratio of 90%, and an electrolyte volume of 15 mL. The optimized electrodeposition conditions for actinium can be applied to environmental samples to prepare an actinium alpha source.