PET放射性药物的质量保证和质量控制

随着我国高能正电子断层显像技术应用的迅速开展，研制和开发符合标准的正电子药物应用于临床已是当务之急。由于正电子放射性药物寿命短，迫使人们必须建立质量控制的新概念。质量控制是强调法规控制下研制和生产过程的质量保证，最终产品的质量检验只作为此类药物质量证明的补充。     

关于放射性药物放射化学纯度测定的探讨

总结了本院同位素研究所质量控制研究室２０多年来在放射性药物放化纯度分析方面的工作，发现和解决了一些存在的问题，提出了一些新的、更准确、简便、快速的分析方法。     

68Ga-PSMA-11注射液的制备和质量控制回顾及建议

本研究制定了规范的⁶⁸Ga-PSMA-11注射液制备流程，并参考国内外药典要求拟定了适用于本机构的⁶⁸Ga-PSMA-11注射液质量控制标准。回顾2020—2022年度本机构316批次的⁶⁸Ga-PSMA-11注射液的制备和质量控制结果，发现所有批次均成功制备得到⁶⁸Ga-PSMA-11注射液，质量控制结果均符合拟定的标准。⁶⁸Ga-PSMA-11的标记率为(85.3±11.4)%,受⁶⁸Ge/⁶⁸Ga发生器淋洗总体积的显著影响，随着淋洗总体积的增加而降低，而淋洗间隔时间对标记率影响很小。本研究有望指导核医学从业人员规范进行⁶⁸Ga-PSMA-11注射液的制备和质量控制，从而提升该药物的质量与安全，并为其相关行业标准的制定提供参考。     

PET放射性药物的质量保证和质量控制

随着我国高能正电子断层显像技术应用的迅速开展,研制和开发符合标准的正电子药物应用于临床已是当务之急.由于正电子放射性药物寿命短,迫使人们必须建立质量控制的新概念.质量控制是强调法规控制下研制和生产过程的质量保证,最终产品的质量检验只作为此类药物质量证明的补充.     

在CPCU上自动化合成18FMISO

通过调节参数，在CPCU上实现了18FMISO的自动化合成，合成时间约为55 min，放化产率46.2%（EOS），放化纯度 >99%。在原有的18FDG专用型合成模块上实现其他亲核氟代药物的自动化合成，必将推动国内18F标记药物的研究。     

竞争抑制酶联免疫分析检测磺胺甲恶唑方法的建立

磺胺类药物是在农牧业和兽医临床中广泛应用的一种抗菌药，也用做饲料添加剂。但它在奶类和肉制品中的残留可能诱发细菌的耐药性，甚至可能致敏和致癌。目前，国内对动物源性食品中磺胺类药物残留监测主要为磺胺二甲嘧啶（SMZ）、磺胺二甲氧嘧啶（SDM）、磺胺间甲氧嘧啶（SMM）、磺胺喹恶啉（SQ）、磺胺甲恶唑（SMX）。在国内，     

HPLC法测定锝[^99Tc]亚甲基二膦酸盐注射液A剂中高锝酸盐含量

锝[^99Tc]亚甲基二膦酸盐注射液是一种抗类风湿药物，在临床上具有较好的疗效。该药物由A剂和B剂组成，A剂为含微量高锝酸盐（^99TcO4^-）的生理盐水注射液，B剂为亚锡亚甲基二磷酸盐冻干品。高锝酸盐含量的测定是该药物质量控制的一项重要指标。对于^99TcO4^-含量的测定，文献报道的方法有高效液相色谱（HPLC）、低本底液闪法和分光光度法等。     

16 大量铀中镎钚的分离与测定——加压排空柱离子交换系统的设计及应用

镎和钚是长寿命极毒元素，核燃料后处理铀产品中对镎和钚的控制要求极其严格。准确、及时、稳定地测定乏燃料后处理厂铀线尾端样品以及铀产品（以下简称样品）中镎钚的含量是乏燃料后处理质量控制的要求。国内一直未建立起满足后处理铀线质量控制的分析方法，现有方法存在的主要问题是去除铀基体的分离流程长，去污因子不稳定，很难满足铀产品的质量控制要求（样品中镎和钚的含量极低而铀镎和铀钚的含量比值又非常高，只有经过有效分离后方能准确测定），而国外对此类方法均不予报道。     

核电厂主控室撤离场景的定量化研究

主控室由于火灾等因素可能失去控制或丧失可居留性，导致运行人员撤离主控室至远程停堆站进行核电厂指挥和控制的场景。对运行人员主控室撤离场景的定量化是火灾概率安全分析的重要内容和技术难点，但国内核电工程项目一般采用保守或专家判断的方法进行定量化，未进行详细研究。论文基于NUREG-1921及其增补版导则，结合国内核电厂实际情况，对主控室撤离场景三个阶段情景及其定量化方法进行了研究。以国内某核电厂主控室撤离场景为例，开展了人员访谈和定量化分析。案例表明该核电厂针对主控室不可控的撤离没有明确规定，导致其人误概率较大，尽管该情景条件概率较低，但后果严重，建议核电厂增加相应程序。本研究为国内核电工程项目开展主控室撤离场景的定量化提供参考。     

诊断用放射性药物研究现状与展望

以放射性药物为基础的单光子断层扫描(SPECT)和正电子断层扫描(PET)已成为十分有效的诊断手段,可在分子水平反应活体功能和代谢过程。放射性药物的研发也为分子影像和核医学的发展提供了强有力的保障。国内放射性药物的研究已有坚实的基础,并且不断探索新的领域。本文介绍了近几年中国诊断用放射性药物领域的最新研究进展,对今后一段时期内中国放射性药物的发展进行了展望,并提出了加强加速器放射性核素开发、放射性药物研究等方面的建议,更好地为人类健康事业服务。     

Leak-before-break behaviour of austenitic and ferritic pipes containing circumferential defects

In recent years several research projects have been carried out at MPA Stuttgart to investigate the leak-before-break (LBB) behaviour of pressure-bearing components which are relevant to plant safety. In these investigations the test pipes have for the most part been made of ferritic material. International research programmes such as, for example, the Degraded Piping Programme (Wilkowski et al., 1986 and Wilkowski et al., 1989. Degraded Piping Program, Phase II. Report NUREG/CR-4082, vol. 4, Sept. 1986, and vol. 8, March 1989, Battelle, Columbus, Ohio, USA) or the IPIRG-Program (Schmidt et al., 1991. The International Piping Integrity Research Group (IPIRG), Program—An Overview. SMiRT 11 Proceedings, Paper G23/1, Tokyo, Japan, August 1991) have also dealt with pipes made of austenitic materials. However, they were fabricated of not stabilized quality. To take into account the material of comparable components of German nuclear power plants, the experiments reported in the following are focussed on pipes made of Ti- and Nb-stabilized austenitic material. The results presented below relate to pipes containing circumferential defects subjected to internal pressure and external bending loading. As regards the ferritic components an overview of the experimentally determined results is presented. The predictive capability of engineering calculational methods are presented by way of example. The current programme of investigations is presented together with the testing techniques and the initial results.     

Pre-Results of the Real-Time ODIN Validation on MARTe Using Plasma Linearized Model in FTU Tokamak

MARTe is a modular framework for real-time control aspects. At present time there are several MARTe systems under development at Frascati Tokamak Upgrade (Boncagni et al. in First steps in the FTU migration towards a modular and distributed real time control architecture based on MARTe and RTNet, 2010) such as the LH power percentage system, the gas puffing control system, the real-time ODIN plasma equilibrium reconstruction system and the position/current feedback control system (in a design phase) (Boncagni et al. in J Fusion Eng Design). The real-time reconstruction of magnetic flux in FTU tokamak is an important issue to estimate some quantities that can be use to control the plasma. This paper addresses the validation of real-time implementation of that task on MARTe.     

Design and experimental results of feedback control of Ohmic-heating transformer magnetic flux by LHCD power in HT-7 Tokamak

In order to make a research on long pulse or even steady state operation with non-inductive drive in plasma discharge, a new feedback control scheme instead of the previous one has been designed and operated in HT-7 [HT-7 team presented by J. Li, et al., Plasma Phys. Control. Fusion 42 (2) (2000) 135-146] Tokamak experiment, 2004. Consumption of iron-core transformer magnetic flux (MFT) is feedback controlled for the first time by power of lower hybrid current drive (LHCD) P_LH, when the Ohmic-heating circuit current can maintain the plasma current I_P constant with another feedback control loop, which make MFT evolve at alternating-change state to avoid flux saturation. Plasma current I_P can be maintained steadily up to 120 s in this operation mode at reduced plasma parameters (I_P ≈ 50-100 KA, average density , P_LH = 100-200 KW). Design and experimental results are presented in the paper, which including control model analysis, configurations of control system and MFT feedback control experiments in HT-7. The high voltage power supply (HVPS) of LHCD is the main controller that regulates the LHCD power into the plasma to control the MFT.     

Predictive Simulations of Plasma Heating and Current Drive by Neutral Beam Injection on EAST

Long pulse and high performance steady-state operation is the main scientific mission of experimental advanced superconducting tokamak (EAST). In order to achieve this objective, high-power auxiliary heating systems are essential. Radio frequency (RF) wave heating and neutral beam injection (NBI) are two principal methods. NBI is an effective method of plasma heating and current drive, and it has been used in many magnetic confinement fusion devices. Based on the plasma equilibrium of EAST (Li et al., Plasma Phys Control Fusion 55:125008, 2013) plus previous EAST experimental data used as initial conditions, the NBI module (Polevoi et al., JAERI-Data, 1997) employed in automated system for transport analysis (ASTRA) code (Pereverzev et al., IPP-Report, 2002) is applied to predict the effects of plasma heating and current drive with different neutral beam injection power levels. At certain levels of plasma densities and plasma current densities, the simulation results show that the NBI heats plasma effectively, also increases the proportions of NB current and bootstrap current among total current significantly.     

Development of Interfacial Friction Model for Two-Fluid Model Code against Countercurrent Gas-Liquid Flow Limitation in PWR Hot Leg

An interfacial friction model for two-fluid model code has been developed for the counter- current gas-liquid flow limitation at hot leg in a pressurized water reactor. Firstly, using a typical two-fluid model code TRAC-PF1/MOD1, we analyzed whether the interfacial friction model under countercurrent stratified flow by Ohnuki et al., which has been verified with an envelope model at steady state, functions well for the dynamic calculation with the two-fluid model code or not. It was found from the analyses that the model by Ohnuki et al. should be combined with a suitable interfacial friction model for the slug flow regime in hot leg. Based on flow observation in a small scale air-water experiment, models at the bend of hot leg and in the roll wave regime in the horizontal flow path of hot leg were newly developed as the model in the slug flow regime and the slug flow model was combined with the model by Ohnuki et al., The validity of the present model was confirmed with the data under various conditions of scales, pressures and fluid combinations (inner diameter: 0.025~0.75m, pressure: 0.1~7.1 MPa and air-water or steam-water).     

船用堆控制棒失控抽出事故时堆芯安全特性研究

采用船用堆三维动态安全分析仿真软件对发生控制棒失控抽出事故时堆芯安全特性进行了仿真分析,研究了反应堆分别处于高、低功率运行工况下1组或1束控制棒以不同的速率失控抽出时堆芯燃料芯块中心最高温度、最小烧毁比和冷却剂出口温度等参数的变化规律,并进行了比较,得出了一些有益的结论,对于考察反应堆安全状况和事故发生后反应堆操纵人员制定安全措施具有重要的指导意义。     

The implementation of magnetics verification in EAST plasma control system

Experimental advanced superconducting tokamak (EAST) is an experimental device aiming at steady state plasma operation for fusion research. The values of many discharge parameters, such as plasma shape, position and current must be directly acquired or indirectly evaluated from the magnetic measurements, so the accuracy of magnetic measurements plays an important role in reliable plasma control performance. A method for verifying the key magnetic measurements in real time for each shot is described in this paper. Such magnetics verification will prevent the discharge from a key magnetic signal failure and ensure the quality of a successful discharge. The diagnostics verification algorithm has been implemented in the plasma control system for the EAST. The implementation details and its application in the recent experiment are presented in this paper.     

高放废液玻璃固化工程产品接收监管要求设想

在针对高放废液玻璃固化工艺和玻璃固化产品质量影响因素进行研究分析的基础上,通过借鉴IAEA颁布的玻璃固化产品质量控制要求,以及德国和美国所制定的产品接收监管要求和质量控制措施,结合我国在玻璃固化体性能方面已经制定的标准要求,提出我国高放废液玻璃固化工程产品接收监管要求。     

国内外聚变核安全监管与许可初步分析研究

中国政府高度重视聚变发展,安全是聚变能发展的生命线,而核安全监管和相关许可制度是确保聚变能安全发展的必要手段。聚变堆具有其独特的安全特性,无法完全照搬目前基于裂变堆建立起来的法律法规等监管制度。本文初步梳理了国际(含ITER、国际原子能机构、国际能源署、欧盟、美国、韩国等)关于聚变核安全监管和许可的研究进展和相关经验,总结了我国目前在聚变核安全监管与许可方面的现状与存在的问题,为我国聚变核安全监管提出了发展建议。     

水力驱动控制棒静态特性实验研究

对水力驱动控制步进缸的静态保持特性进行了大量实验研究。获得了水力步进缸的静态保持流量范围和其随温度的变化规律,分析了静态特性与步进缸设计参数之间的关系。结果表明：步进缸的设计参数（重量、对孔尺寸）确定其静态保持流量范围;随温度的升高、最大、最小静态保持流量有所增加;同一温度下不同步位上的静态保持流量范围趋于一致。