In modern Boron neutron capture therapy (BNCT) treatment planning, 18F-BPA (18F-boronophenylalanine) PET (positron emission tomography) imaging is used to assess boron uptake and guide accurate dose delivery. This study evaluates the geometric and dosimetric differences between target volumes defined by MRI (magnetic resonance imaging) and PET images in accelerator-based BNCT using the NeuPex system. The GTV (gross tumor volume) was defined based on MRI (GTVMRI) and PET images with SUV thresholds of 2.0 (GTVPET2.0) and 2.5 (GTVPET2.5) in patients with head-and-neck cancer (HNC) and malignant glioma (MG). GTVPET2.5 was approximately 50% smaller, while GTVPET2.0 showed closer volumetric agreement, with ∆V% of 23.52 ± 17.44 (HNC) and 28.71 ± 12.57 (MG), and corresponding Dice similarity coefficients of 0.71 ± 0.14 and 0.86 ± 0.17. Dosimetric analysis revealed that GTVPET2.0 achieved higher dose coverage than GTVMRI, with mean doses of 24.07 ± 6.42 Gy-Eq (HNC) and 26.36 ± 9.77 Gy-Eq (MG), and D80 values of 19.01 ± 6.90 Gy-Eq and 21.00 ± 7.51 Gy-Eq, respectively. Tumor-to-blood ratio (TBR) and tumor-to-normal ratio (TNR) analyses further indicated substantial heterogeneity in boron uptake, with mean TBR and TNR of 2.71 ± 0.53 and 2.97 ± 0.79 for GTVPET2.0. These results suggest that GTVPET2.0 provides a balanced and biologically relevant target for BNCT planning. A TBR threshold of ≥ 2.0 is proposed as a practical criterion for patient selection. While this study focuses on geometric and dosimetric endpoints, future work correlating these findings with clinical outcomes is warranted. Overall, this study supports the integration of functional PET imaging with anatomical MRI to improve the precision and efficacy of BNCT treatment planning.

B4C/Al composites have been widely studied and used as neutron absorbing materials in various application fields because of its relatively simple preparation process, good mechanical properties, and low raw material prices. B4C content is the major factor affecting the neutron shielding performance of B4C/Al composites. A photo-neutron source driven by a 15 MeV electron linear accelerator has been used to measure the neutron shielding performance of the B4C/Al composite plates with B4C weight percentages of 16.85% and 31%, respectively. Based on the measurement result, the thermal neutron macroscopic cross sections of the B4C(16.85 wt%)/Al composite and B4C(31 wt%)/Al composite were 13.97 cm- 1 and 25.58 cm -1, respectively. The mea-surement results showed that the B4C/Al composites have good shielding performance for neutrons with energies lower than 10 eV, especially for thermal neutrons, and the 1.02 mm thickness of B4C(31 wt%)/Al composite was able to absorb 90.91% of thermal neutrons. Experimental measurements were compared with calculation results. The experimental measurement result of the thermal neutron macroscopic absorption cross section was smaller than the calculated data, and the relative deviations of the experimental measurement results and calculation data of the thermal neutron macroscopic cross sections of B4C(16.85 wt%)/Al and B4C(31 wt%)/Al composites were-13.65% and-12.52%, respectively. The Monte Carlo N-Particle Transport code was used to calculate the neutron shielding performance of B4C/Al with different B4C content, and the shielding performance of the B4C/ Al composites for low-energy neutrons was approximately linearly proportional to the B4C content, especially for the thermal neutron.

上海软X射线自由电子激光装置(Soft X-ray Free-Electron Laser facility,SXFEL)是我国第一台X射线相干光源,为了对其波荡器线束流损失进行监测,设计安装了一套基于切伦科夫辐射理论的石英光纤束损监测系统.由于石英光纤对高能γ射线不敏感,适应在SXFEL高辐射场环境中安装使用.该束损监测系统响应速度极快,覆盖整个波荡器线约180 m长区域,能够对整个波荡器沿线束流损失进行实时连续监测.本文介绍了该光纤束损监测系统原理与组成,按应用需求进行了系统位置标定与光纤衰减系数测量等实验,并在实际调束过程中考察了该系统的应用情况.结果表明:该系统具有良好的位置分辨能力,在上游位置分辨能力约为0.2 m,能够满足工作人员调束需求.

金刚石材料具有优异的耐高温、抗辐照性能,用其制作的辐照探测器在反应堆等苛刻环境下具有很好的应用前景.在分析金刚石中子探测器的结构和工作原理的基础上,使用MCNP(Monte Carlo N Particle Transport Code)模拟程序构建了金刚石中子探测器的物理模型,考虑探测器用于2 MWt液态燃料钍基熔盐试验堆(Thorium Molten Salt experimental Reactor-Liquid Fueled,TMSR-LF1)辐射场中,计算中子转换层(6LiF、10B)厚度、金刚石厚度、γ甄别阈值对探测器的中子探测效率、γ探测效率以及n/γ抑制比的影响.结果表明:6LiF更适合在中子、γ混合场中用作中子转换层;随着6LiF厚度增加,中子探测效率先增大后减小,6LiF的最优厚度为25 μm;金刚石厚度增大会导致探测器的n/γ甄别性能下降,可以采用设置γ甄别阈值的方法解决金刚石层过厚时带来的γ干扰过大的问题,使探测器达到对γ不灵敏的要求.模拟研究工作获得了探测器结构参数对探测器性能的影响规律,对探测器后续的制作和研究具有指导意义.

The Al2O3/Ni-Al composite coating was formed on the outside surface of GH3535 superalloy by pack aluminizing and subsequent in-situ oxidation, and deuterium permeation tests were carried out to assess the tritium permeation barrier performance of the coating. Scanning Electron Microscopy (SEM) showed that the surface of Al2O3 films was dense, homogeneous, and crack-free. Transmission Electron Microscopy (TEM) photos manifested that the interface between gamma-Al2O3 film and Ni-Al interlayer was distinct without micropores, and the thickness of Al2O3 film with the structure of gamma-phase is approximately 80 nm. The Arrhenius equation for permeability of GH3535 was obtained as F=1.37 center dot 10(7)center dot exp(57.92(KJ mol(-1))/RT) by deuterium permeation tests. The deuterium permeation reduction factor of Al2O3/Ni-Al coating kept 2-3 orders of magnitudes at a temperature from 400 degrees C to 700 degrees C. The D-permeation resistant performance of this coating still has the potential to be further improved.

高温工况下钍基熔盐堆中存在氚泄漏的风险,建立氚渗透屏障涂层有助于应对这一问题.采用包埋渗铝和原位氧化工艺,在GH3535合金表面制备了Al2O3/Ni-Al复合阻氚涂层,重点分析了氧化温度和真空度对氧化铝薄膜微观结构的影响.利用掠入射X射线衍射、扫描电子显微镜、透射电子显微镜等手段对氧化铝薄膜表面及截面的微观形貌、相构成进行了实验分析.实验结果表明:低氧分压能降低氧化铝薄膜的形成速度,促进形成更致密、表面平整的薄膜;高的氧化温度有利于形成α相氧化铝及更厚的氧化铝薄膜,但会大大增加表面缺陷.1.2 Pa真空度气氛、850℃氧化温度、72 h氧化时间是较优的原位氧化工艺参数,可以在GH3535合金基体表面获得性能较好的氧化铝薄膜,其相结构为γ和α相,厚度约为0.8μm,且表面致密无缺陷.

Abstract Purpose To quantify the influence of beam optics asymmetric distribution on dose. Methods Nine reference cubic targets and corresponding plans with modulation widths (M) of 3, 6, and 9 cm and with center depths (CDs) of 6, 12, and 24 cm were generated by the treatment planning system (TPS). The Monte Carlo code FLUKA was used for simulating the dose distribution from the aforementioned original plans and the dose perturbation by varying ±5%, ±15%, ±20%, ±25%, and ±40% in spot full width half maximum to the X‐direction while keeping consistent in the Y‐direction. The dosimetric comparisons in dose deviation, γ‐index analysis, lateral penumbra, and flatness were evaluated. Results The largest 3D absolute mean deviation was 15.0% ± 20.9% (mean ± standard deviation) in M3CD6, whereas with the variation from −15% to +20%, the values were below 5% for all cube plans. The lowest 2D γ‐index passing rate was 80.6% with criteria of 2%–2 mm by a +40% variation in M3CD6. For the M9CD24 with a −40% variation, the maximum 1D dose deviations were 5.6% and 15.7% in the high‐dose region and the edge of the radiation field, respectively. The maximum deviations of penumbra and flatness were 3.4 mm and 11.4%, respectively. Conclusions The scenario of beam optics asymmetric showed relatively slight influence on the global dose distribution but severely affected dose on the edge of the radiation field. For scanning carbon‐ion therapy facilities, beam spot lateral profile settings in TPS base data should be properly handled when beam optics asymmetry variation is over 15%.

Objective:To develop a spot scanning carbon ion beam model based on Monte Carlo code FLUKA and verify the accuracy of physical dose.Methods:A geometric model of the treatment nozzle was established in FLUKA. Various parameters such as monoenergy nominal energy, Gaussian energy spectrum distribution, initial spot size, and beam angular distribution in the model were adjusted to match the reference data of integral depth dose (IDD) and in-air spot size measuremed experimentally. Carbon ion beam plans were generated by using the treatment planning system (TPS). The difference in output dose distribution between FLUKA and TPS was compared by the gamma analysis.Results:The differences in Bragg peak width, beam range, and distal falloff width extracted from the IDD curve between the FLUKA model and measured vaues were less than 0.1 mm, with the maximum difference in spot sizes of 0.17 mm. Under the criterion of 2 mm/2% in all the simulations, 2D- and 3D-γ pass rates were all above 95%.Conclusions:An accurate spot scanning carbon beam model was developed based on the Monte Carlo code FLUKA. It has the potential to be used for not only the verification of clinical treatment plans, but also the development of new ion beam therapy equipment and the calculation of biologically effective dose.

Tritium would be produced in the Thorium-based Molten Salt Reactor and then quickly penetrate through the structural material of GH3535 superalloy into environment at elevated temperatures. To suppress the permeation of tritium, the alpha-Al2O3/NiAl multilayer coatings were prepared on GH3535 by pack aluminizing (PCA) and subsequent low vacuum oxidation as tritium permeation barriers. Herein, the effects of PCA process parameters on the composition and structure of the Ni-Al coatings were characterized, simultaneously, the influence of oxidation temperature on the composition and structure of Al2O3 films were analyzed in detail. Experimental results demonstrated that the main phase of the Ni-Al layers is Ni2Al3 (or NiAl) doped with Al86Cr14. Significant amounts of Mo-aluminides were found on the surface of the samples aluminized at 1050 degrees C. The Ni-Al layers growth kinetic equation under 650-850 degrees C was empirically obtained as h = 7.07.10(4)W(1/2)t(1/2)exp[-7.53.10(4)/RT]. After low vacuum heat-treatment, the alpha-alumina films with a thickness of approximately 1 mu m were formed at 1050 degrees C for 2 h, densely and non-porously. Meanwhile, the brittle Ni2Al3 in the transition layer was transformed into the tough NiAl, providing a self-healing capability as well as alleviating the thermal mismatch between the coating and the substrate.

Oil cleaning agents generated from nuclear power plants(NPPs) are radioactive organic liquid wastes.To date, because there are no satisfactory industrial treatment measures, these wastes can only be stored for a long time. In this work, the optimization for the supercritical water oxidation(SCWO) of the spent organic solvent was investigated. The main process parameters of DURSET(oil cleaning agent) SCWO, such as temperature, reaction time,and excess oxygen coefficient, were optimized using response surface methodology, and a quadratic polynomial model was obtained. The determination coefficient(R 2 ) of the model is 0.9812, indicating that the model is reliable.The optimized process conditions were at 515 °C, 66 s,and an excess oxygen coefficient of 211%. Under these conditions, the chemical oxygen demand removal of organic matter could reach 99.5%. The temperature was found to be the main factor affecting the SCWO process.Ketones and benzene-based compounds may be the main intermediates in DURSET SCWO. This work provides basic data for the industrialization of the degradation of spent organic solvents from NPP using SCWO technology.

BackgroundIn particle radiotherapy, the Bragg peak (BP) width of carbon ion beam is required to be broadened whilst Ripple filter (RiFi) is often used as a clinical broadening device. However, it may cause extra lateral scattering. Porous material expressed as the porous structure 1.0 (PS 1.0) can be used to reduce such scattering but bring an oversized distal falloff width (DFW).PurposeThis study aims to develop a novel porous structure, described as a porous structure 2.0 (PS 2.0), to broaden the BP width.MethodsFirst of all, the Monte Carlo code, FLUKA was selected as the tool for this simulation. Two geometry models of broadening structure (1D-RiFi and PS 1.0) were built. Then, the PS 2.0 was simulated by inserting numerous of Polymethyl Methacrylate (PMMA) sticks in the PS 1.0 with a certain proportion. The performance of PS 2.0 was evaluated by analyzing the DFW and isocenter spot size of the modulated carbon ion beams. Finally, a plane homogeneous radiation field was simulated and fluence homogeneities were compared by using a 2D (X-Z) dose distribution and quantized by calculating the 1D-lateral flatness.ResultsCompared to the PS 1.0, the DFW is reduced by at most 1.11 mm by using PS 2.0 with a similar BP width broadening ability at the same beam range. Fluence homogeneity of PS 2.0 is excellent in all the downstream locations, and the isocenter spot size is reduced by at most 1.72 mm at the same beam range.ConclusionsThis study provides a new method for simulating the PS 2.0 as a BP broadening device with smaller DFW, better homogeneous and reduced isocenter spot size in comparison to PS 1.0.

The effects of additional oxidants, such as NaNO 3 , Na 2 S 2 O 3 , KClO 4 , and K 2 Cr 2 O 7 , on the supercritical water oxidation (SCWO) of tributyl phosphate (TBP) were studied. The coupling of an ionic oxidant with SCWO can effectively enhance the oxidative degradation ability of the system, thus increasing its organic-matter-removal efficiency at a reduced reaction temperature. Moreover, the addition of NaNO 3 , KClO 4 , or K 2 Cr 2 O 7 could improve this efficiency at a reaction temperature of 500 °C compared with that of the original system at 550 °C. Additionally, based on the conditions adopted in this study, the addition of either of these oxidants could reduce the final total organic carbon (TOC) of the effluent from ~ 500 to < 100 ppm. Concurrently, the ionic oxidants could effectively improve the processing capacity of the SCWO system to reduce the scale of the equipment, as well as the amount of produced wastewater. Compared with KClO 4 and Na 2 S 2 O 3 , the addition of 10 mmol/L NaNO 3 and K 2 Cr 2 O 7 to the organic feed could increase the processing capacity of the system from 4 to 10% while maintaining the TOC removal at > 99%. The effects of the ionic oxidants on the gas products, including CO 2 , CO, H 2 , and CH 4 , as well as other organic gases, have also been studied. Among these gas products, CO 2 accounted for the main gas product with a proportion of more than half. At < 500 °C, temperature significantly affected the as products (CO, H 2 , CH 4 , and other organic gases). However, the gas product was mainly CO 2 when the temperature was increased to ≥ 500 °C. This study initially revealed the enhancement effect of ionic oxidants on SCWO, which still requires further research.

LiF-BeF2-ZrF4-UF4 and LiF-BeF2 have been employed as fuel salt and coolant in liquid fueled molten salt reactor, respectively. A heat exchanger containing both fuel salt and cooling salt is placed inside the reactor vessel to dissipate the heat produced by the reactor. A molten salttoair heat exchanger outside the reactor vessel is used to cool the molten salt, and a molten salt pump is used to drive the circulation of the entire circuit of coolant system. As of the molten salt, LiF-BeF2 coolant, flows through the molten salt heat exchanger located inside the reactor vessel, nuclides in the coolant are activated by 19F(n, α)16N, 19F(n, γ)20F, and 19F(n, p)19O reactions with neutrons in the reactor core, and thus radionuclides such as 16N, 20F and 19O are produced. These radionuclides have radiation impact on the surrounding environment and equipment if they enter the rooms where equipment for molten salt coolant circuit locate. According to the flow process and distribution law of the cooling salt, theoretical calculation formulas to calculate the amount of radionuclide were given. Using theoretical formulas, calculated radioactivities of the three most important radionuclides 16N, 20F and 19O per unit volume of cooling salt are 9.05×106, 8.33×106 and 2.69×105 Bq/cm3 at the outlet of the molten salt heat exchanger, while the radioactivities of the three radionuclides at the same location calculated by ORIGENS program are 8.98×106, 8.58×106 and 2.76×105 Bq/cm3, respectively. The results calculated by the formulas are in good agreement with the data of the ORIGENS program, and the maximum relative deviation between the results of the two methods is 292%, indicating that applying the derived formulas to carry out cooling salt activation analysis is reasonable and feasible. These radionuclides, 16N, 20F and 19O, produced in the cooling salt of molten salt reactor release highenergy gamma rays as they decay. The total gamma emissivities of the cooling salt in equipment and pipelines located in the cooling salt storage tank room, in the molten salt-to-air heat exchanger room and in the cooling salt pump room are 8.71×1011, 9.51×1011 and 1.93×1012 s-1, respectively, and the most radioactive area is in the cooling salt pump room. The maximum absorbed dose rate at the location of the cooling salt auxiliary equipment is 45.7 mGy/h without shielding, which cannot meet the requirement of the radiation dose, and which should be less than 50 Gy for the entire life of the equipment (300 full power days). By setting a 30 cm thick concrete shielding wall between the cooling salt auxiliary equipment and the cooling salt pump, the absorbed dose rate at the location of the auxiliary equipment can be reduced to less than 7 mGy/h, thereby the radiation protection requirements for the cooling salt auxiliary equipment can be met.

Abstract A containment vessel is usually designed to prevent the potential leakage of radioactive gas in molten salt reactor. Understanding the gas flow characteristics in the containment vessel is beneficial for mastering the radioactive level and the early monitoring and diagnosing when radioactive leakage accident occurs. In this paper, the simplified physical model of the containment vessel in a molten salt reactor was set up, and the gas flow characteristics driven by natural circulation in the containment vessel was studied by CFD code. The results shown that, when the gas enters into the upper chamber from the lower chamber due to buoyancy force, the gas velocity will reach a maximum value of 1.7 m/s where local jet occurs. The mass flow rate of gas between the upper and lower chambers is about 24.0 kg/h and the average circulation time in the containment vessel is about 7.3 hours. It is interesting to note that, the gas tends to flow upward in those annulus gaps with larger hydraulic diameter or near to the central region, indicating that when leakage is near to those regions, it will be easier and faster to flow into the upper chamber. The results provide the reference and guidance for radioactive gas monitoring.

超临界水氧化技术(SCWO)自1980年被提出以来得到了快速的发展,在废物处理、尤其是有机危废的处理上发挥了重要的作用.从超临界水自身的特性出发,系统地介绍了SCWO的特点,并从基础研究进展、技术应用领域、关键技术难题及其解决方案等方面综述了SCWO的发展历程,重点对SCWO在放射性废物处理领域的研究和应用进行了总结和分析.针对目前SCWO工艺的废水增容问题,指出了废水最小化的建议,提出通过提高现有SCWO体系的氧化容量和提高SCWO工艺过程中水的利用率实现废水减容的措施.

A liquid fueled thorium molten salt reactor (TMSR-LF), one of the Generation IV reactors, was designed by the Shanghai Institute of Applied Physics, Chinese Academy of Sciences. This study uses the`rt code to calculate the neutron and gamma dose rate distributions around the reactor. Multiple types of tallies and variance reduction techniques were employed to reduce calculation time and obtain convergent calculation results. Based on the calculation and analysis results, the TMSR-LF1 radiation shield with a 60-cm serpentine concrete layer and a 120-cm ordinary concrete layer is able to meet radiation requirements. The gamma dose rate outside the reactor biological shield was 16.1 mSv h −1 ; this is higher than the neutron dose rate of 3.71 × 10 –2  mSv h −1 . The maximum thermal neutron flux density outside the reactor biological shield was 1.89 × 10 3  cm −2  s −1 , which was below the 1 × 10 5  cm −2  s −1 limit.

BackgroundRadioactive waste oil mainly comes from the maintenance of equipment such as main pumps and steam turbines of nuclear power plants (NPP), and is a typical radioactive organic liquid waste (ROLW). Because there is no suitable treatment technology for ROLW, it can only be stored for a long time. Long-term storage has potential safety hazards such as leakage and fire.PurposeThis study aims to solve the problem of radioactive waste oil treatment by carrying out an experimental study on the supercritical water oxidation (SCWO) treatment of lubricating oil from NPP.MethodsA continuous SCWO reaction system was used as experimental platform with maximum working temperature of the 600 ℃ and pressure of 28.4 MPa. The effects of different reaction temperature, reaction time and excess oxygen coefficient on the chemical oxygen demand (COD) removal rate of VG 32 (Mobil DTE light turbine oil) lubricants were investigated. In addition, the reaction kinetics of SCWO treatment of lubricating oil from NPP at 420~580 ℃ was studied.ResultsThe experimental results show that the best operating conditions are reaction temperature of 550 ℃, reaction time of 80 s and oxygen enrichment coefficient of 250%. Under these conditions, the COD removal rate of organic matter is up to 96%. The activation energy Ea of the reaction is 43.10 kJ·mol-1, and the pre exponential factor A is 8.861 s-1. The deviation between the calculated value of the kinetic model and the experimental value is less than 3%.ConclusionsThis work proves the credibity of the model, and lays an experimental basis for supercritical water oxidation treatment of radioactive waste oil.

硼铝复合材料因制备工艺简单,力学性能良好,原材料价格低廉等诸多优点被广泛研究,并被用作诸多领域的热中子吸收材料.本文采用理论计算、MCNP软件模拟、实验测量等多种方法对硼铝复合材料的热中子屏蔽性能进行了评估分析.通过理论计算发现,对于相同配比的硼铝复合材料,从材料的热中子吸收性能方面,添加硼单质的效果优于添加碳化硼.通过MCNP程序模拟计算和实验测量发现,硼铝复合材料对能量低于10-7 MeV的中子吸收效果比较显著.

Shanghai soft X-ray free-electron laser (SXFEL) test facility is the first coherent X-ray light source in China, and its output wavelength is shorter than 9 nm. This free-electron laser (FEL) facility is based on a 0.84 GeV linear accelerator, and its main goal is to develop FEL related technologies and test new FEL schemes, especially the cascaded seeding FEL and short wavelength EEHG FEL. This facility already achieved its design goal and passed the national acceptance in November 2020. Status and main achievements of the SXFEL test facility are introduced in this article.

在建上海硬X射线自由电子激光装置(Shanghai high repetition rate XFEL and extreme light facility,SHINE)包含一台8 GeV的超导直线加速器,100 PW的激光系统,三条波荡线和10条实验线站.主加速器的第三磁压缩段(BC3段)有一调试用的束流收集桶,BC3段的调试电子流强为0.3μA、能量为7.7 GeV,束流功率为2.31 kW.本文从材料对7.7 GeV电子的能量吸收方面考虑,对束流收集桶的结构进行了确定.用ANSYS 17.0对束流收集桶的温度场进行了分析,结果证明在有表面空气自然对流换热的情况下,铝芯的最高温度为193.14℃,在束流收集桶坑道无外界换气的情况下,束流收集桶需要外围冷却水;用蒙特卡罗程序FLUKA对束流收集桶的活化进行了分析,保守假设束流收集桶连续注束2000 h.结果证明要保证停机1 h后,束流收集桶周围30 cm处的剩余辐射剂量小于25μSv/h,束流收集桶外围需要55～85 cm的混凝土屏蔽体;结果显示束流收集桶的剩余辐射剂量在30 min内衰减的非常快,这是由于刚停机时束流收集桶内产生的50％左右的放射性核素半衰期都小于10 min.