ITER屏蔽包层活化分析

作为国际热核聚变实验堆(ITER)的重要部件之一,屏蔽包层承受高强度聚变中子辐照,需要定期更换和维修。当活化的屏蔽包层从ITER托卡马克装置移到热室时,可能会给工作人员造成严重的辐射照射,是ITER大厅和热室屏蔽设计的重要辐射源。文中基于ITER最新中子学分析基准模型和"二步法"停堆剂量计算方法,使用超级蒙特卡罗核计算仿真软件系统SuperMC针对15号屏蔽包层建立精细的中子学模型,并计算分析包层的活化情况及最严重情况下的周围辐射剂量率,并初步应用于ITER赤道窗口室的屏蔽分析。计算结果显示,单个包层周围最大剂量率为350 Sv/hr,当传送小车停留在赤道窗口室内时,窗口室屏蔽门外剂量率高于10 mSv/hr,不足以满足设计要求。     

ITER赤道窗口生物屏蔽层特性研究

ITER赤道窗口生物屏蔽层外的窗口室是放置各种电子仪器和管道的场所,在停堆后允许人员进入其中进行维修等工作.为了保证工作人员安全,本文研究了停堆后窗口室中的剂量率分布.文中使用大型集成中子学计算分析系统VisualBUS中的建模、计算分析模块完成了停堆剂量率计算.结果表明,ITER停堆1天后赤道窗口室中剂量率高出限值(...     

ITER上窗口屏蔽中子学分析研究

利用CAD/MCNP自动建模程序MCAM建立ITER新上窗口中子学计算模型,使用中子/光子耦合输运程序MCNP/4CI、AEA聚变核数据库FENDL1.0和集成上窗口模型的ITER基本中子学模型计算并分析上窗口新的工程设计的屏蔽能力以检验设计的合理性。结果表明,与以前的上窗口设计相比,新设计的上窗口的周围剂量控制点的快中子注量率、停堆剂量率以及线圈核热等都增大了好几倍,建议进一步改进上窗口设计。     

高性能计算分析技术在大型钠冷快堆屏蔽设计中的应用

钠冷快堆堆容器是一体化的池式结构,由众多堆内构件组成且结构复杂,堆芯到生物屏蔽外中子输运过程中各向异性明显且深穿透问题严重,大尺度范围下三维S_N方法计算是制约快堆屏蔽设计的瓶颈。通过将三维SN程序与高性能计算技术相结合,采用并行计算方法可解决快堆堆本体内各向异性的三维深穿透屏蔽问题。本文以中国示范快堆（CFR600）堆本体为研究对象,采用JSNT-CFR程序详细计算了堆本体内的中子注量率、光子注量率、剂量率,并将计算结果与已有的二维程序设计结果进行比较。结果表明,将传统屏蔽计算方法与高性能计算相结合,能满足CFR600堆本体屏蔽计算精度要求,获得更为全面的三维展示效果,在计算模型复杂、粒子穿透深度等复杂问题的屏蔽计算上具有较明显的优势,为大型钠冷快堆屏蔽设计提供有力支撑。     

ITER极向场线圈PF4维修区域停堆剂量率分析

根据国际热核聚变实验堆ITER设计标准,ITER极向场线圈(PF Coils)的人工检测和连接维修任务的制定,需要确保维修过程中工人所受到的辐射剂量水平不超过剂量限值。基于ITER中子学基准模型B-lite,利用二步法停堆剂量计算方法,在大型集成中子学计算分析系统VisualBUS和HENDL数据库支持下,计算并分析了三种维修方案下PF4维修区域内的停堆光子剂量场分布,以分析降低维修工人辐射剂量水平的有效措施。结果表明,与推迟维修工人进入:PF4维修区域时间相比,采用临时屏蔽的措施更能显著降低PF4维修区域内的辐射剂量水平,建议后续采用临时屏蔽措施。     

空间堆阴影屏蔽结构与材料选型

空间堆对辐射屏蔽尺寸和重量要求苛刻，为寻找合适的屏蔽方案，需要对屏蔽材料、结构进行选型研究。本文首先介绍了国内外对空间堆屏蔽目标及限值的研究进展，基于反应堆屏蔽设计原理，针对不同应用场景抽象出平板模型和球模型，在不同设计目标下对不同材料的屏蔽性能进行分析，基于分析结果采用自动化优化工具对屏蔽方案进行选型，分析了各个方案的优缺点。结果表明，放射源的能谱、源的尺寸大小、屏蔽体离源的距离、不同的屏蔽设计目标都会影响屏蔽材料和结构的选择，需要根据应用需求进行筛选；碳化硼、氢化锂和钨是较好的空间堆屏蔽材料；利用自动化优化工具对屏蔽体进行分层布置可实现有效减重。     

压力容器环形空腔上部材料屏蔽计算分析

用离散纵座标SN法与反照蒙特卡罗AMC法相耦合,对秦山600MW核电厂反应堆压力容器环形空腔上部密封环区、褐铁矿混凝土及电离室入口轻质保护材料的中子和γ射线注量率、剂量率和释热率进行计算分析.计算结果表明,对于停堆后人员可能进入的部位,增加轻质保护材料能有效降低中子辐射;采用SN-AMC计算技术能较好地完成反应堆的大尺寸空腔和复杂孔道屏蔽设计计算.     

加速器热中子照相装置CCD芯片屏蔽的模拟计算

建立了研究加速器中子源热中子照相装置CCD芯片屏蔽效果的蒙特卡罗模拟方法,对γ与中子吸收剂量率的模拟计算结果与实验相符.进行了基于9Be(d,n)反应的热中子照相装置屏蔽系统的优化设计,在复杂几何条件下用蒙特卡罗模拟分别计算了CCD芯片在中子、γ混合场中的吸收剂量率和快中子注量率,对CCD相机在辐射场中安全性能进行了评估.     

MC-MC耦合计算方法在核电厂中子屏蔽设计中的应用

中子屏蔽精细化设计是三代堆型核电厂区别于二代堆的主要辐射防护设计特征之一,其设计优劣直接影响了辐射场内设备寿命及功率运行期间可能进入的工作人员的辐射安全。为了精确、快速、有效解决大尺度复杂厂房中子屏蔽计算难题,提出了将MC-MC耦合计算应用于解决核电厂大型复杂计算模型的中子屏蔽设计方法。通过与欧洲第三代压水堆技术方案(CEPR)设计结果对比表明,计算结果偏差小于15%,满足工程屏蔽设计误差要求,证明该方法的正确性与可行性。该方法已应用于国内某三代堆型核电厂反应堆厂房中子屏蔽设计。     

基于CAD的内耦合严格两步法停堆剂量率计算方法研究

核装置尤其是聚变装置中结构材料的辐照活化问题,对核装置的辐射安全具有重要影响。停堆剂量率是材料辐照活化计算中的重要参数,也是聚变堆设计的重要参考依据。本文基于超级蒙卡核模拟软件系统SuperMC的中子/光子输运计算功能和中子活化计算功能,开展了严格两步法停堆剂量率计算方法研究。与传统的输运-活化程序外耦合方法相比,本文发展了一种基于CAD的内耦合严格两步法停堆剂量率计算方法,直接基于CAD模型进行网格材料映射,并支持扇形圆柱源抽样,在提高易用性和灵活性的同时,消除了传统方法在圆柱坐标系活化区计算的不足和处理复杂几何时的局限性。最后利用国际热核聚变实验堆ITER发布的停堆剂量率计算基准例题进行了校核计算,计算结果表明了该方法的正确性和可靠性。     

First neutronics analysis for ITER bio-shield equatorial port plug

ITER equatorial port cell outside the bio-shield plug is a place to allow personnel access after shutdown that accommodates various sensitive equipment and pipes. Gamma dose rate after shutdown of 1 day in the port cell should be within 10 μSv/h for occupational safety which is one order of magnitude less than that in the port interspace by the shielding of bio-shield plug. To verify the shielding property of the bio-shield plug, the distributions of gamma dose rates in port cell were studied. Based on the ITER neutronics model Alite4 which is a three-dimensional ITER tokomak neutronics model for MCNP calculations with a 40 degree extent in the toroidal direction and vertical reflecting bounded planes on both sides, the equatorial port was updated according to a conceptual CAD model using Monte Carlo Automatic Modeling Program for Radiation Transport Simulation (MCAM). A 2-step method of gamma dose rate calculation was used for shutdown dose rates in CAD-based Multi-Functional 4D Neutronics Simulation System (VisualBUS). The result showed that gamma dose rates in the port cell were higher than the desired limit. Refinements to the bio-shield plug design were suggested to ensure that dose rates in the port cell were within the design value for maintenance access.     

Neutronics analysis of the IVVS/GDC plug in ITER

A neutronics analysis has been performed to provide the input required for the design strategy for the In-Vessel Viewing System (IVVS) and the Glow Discharge Cleaning (GDC) plug units in the ITER tokamak. The focus of the analysis has been on operational loads to the GDC electrode head in the shielding position and on the activation and the decay photon radiation absorbed in the structural components of the entire system. To estimate the conditions for maintenance scenarios, the occupational dose rate around the isolated IVVS/GDC head has been calculated assuming the ITER SA2 irradiation scenario. The Rigorous 2 Step (R2S) method, developed previously at KIT, has been employed for the calculation of the shutdown dose rates. The GDC head, which is subjected to the highest neutron loads, gets heavily activated and dominates the decay gamma activity of the entire plug. Accordingly, the shutdown dose rate around the IVVS/GDC plug is dominated by the GDC electrode head. It is therefore recommended to separate the GDC head from the system prior to further operations inside the Hot Cell. All components, except the Be protective layer of the GDC probe, were shown to be classifiable as low level radwaste according to the French regulations.     

Nuclear analysis and shielding optimisation in support of the ITER In-Vessel Viewing System design

The In-Vessel Viewing System (IVVS) units proposed for ITER are deployed to perform in-vessel examination. During plasma operations, the IVVS is located beyond the vacuum vessel, with shielding blocks envisaged to protect components from neutron damage and reduce shutdown dose rate (SDR) levels. Analyses were conducted to determine the effectiveness of several shielding configurations. The neutron response of the system was assessed using global variance reduction techniques and a surface source, and shutdown dose rate calculations were undertaken using MCR2S.Unshielded, the absorbed dose to piezoelectric motors (PZT) was found to be below stable limits, however activation of the primary closure plate (PCP) was prohibitively high. A scenario with shielding blocks at probe level showed significantly reduced PCP contact dose rate, however still marginally exceeded port cell requirements. The addition of shielding blocks at the bioshield plug demonstrated PCP contact dose rates below project requirements. SDR levels in contact with the isolated IVVS cartridge were found to marginally exceed the hands-on maintenance limit. For engineering feasibility, shielding blocks at bioshield level are to be avoided, however the port cell SDR field requires further consideration. In addition, alternative low-activation steels are being considered for the IVVS cartridge.     

Engineering of diagnostic equatorial port plugs for ITER

Within the ITER vacuum vessel, there are a significant number of diagnostics, measuring items such as plasma density, temperature and impurities; and providing a visible image of the ITER plasma. Since reliable diagnostic measurements are critical to the successful operation of ITER, robust structural design of the diagnostic supports, or port plugs, is also essential. The port plugs are substantial steel structures, mounted in both the equatorial and upper ports on the vacuum vessel. They not only support the diagnostics, but also provide functions of baking, cooling, and neutron shielding.Significant progress has been made in the mechanical design of the port plugs, culminating in the proposal of a new conceptual design, which uses the lid of the port plug as a structural member. This allows the port plug's mass to be more efficiently distributed, providing additional space for diagnostics, and better neutron shielding. A critical aspect of the design has been to provide a suitable interface between the lid and body of the structure which will support all of the structural loads which may be applied to the port plug. The lid also allows easy access to the diagnostic components when maintenance is required.Analyses have been carried out in support of the proposed changes. Structural analysis indicates that the wall thickness of the port plug could be reduced from 130 mm to 40 mm. Thermal analysis has demonstrated that the cooling and baking requirement for the port plug structure is less challenging than originally thought, and hence could be carried out in a simpler fashion. Neutronics analysis has led to a better understanding of the impact of different shielding materials and cavities through the contents of the port plug, and show that it may be possible to reduce the shielding thickness from 2000 mm to 1000 mm. Further electromagnetic analysis has been carried out demonstrating that modelling the effect of plasma movement will not affect the resultant loads by more than 20%, and that the originally defined port plug loads were probably conservative.     

Results of an integration study of a diagnostics port plug in ITER

Diagnostics in ITER are mandatory to characterize the parameters of plasma and study its interactions with plasma-facing components. Diagnostics components in the vicinity of the plasma are supported by metallic structures called port plugs. At the tokamak mid-plane, these components are installed in port plugs through intermediate structures called drawers. Apart from hosting the diagnostics, the port plugs act as shielding against neutrons and gammas, in order to limit the nuclear loads in crucial components (such as diagnostics and superconducting coils) as well as the dose levels in the controlled zones of the tokamak. The radiation shielding function of the port plugs is ensured through an optimized mixture of heavy metallic materials and water, forming shielding blocks surrounding the diagnostics and called Diagnostic Shield Modules (DSMs). These DSMs constitute the rear part of the drawers (the front part being composed of the Diagnostic First Wall). This paper presents the main results of a study performed in Europe on the integration of a particular diagnostics port plug, the Equatorial Port Plug 1 (EPP1). The paper first provides the results of the EPP1 diagnostics integration analysis. In a second step it focuses on the design of the EPP1 DSMs and summarizes the major results of a thorough set of analyses aiming at studying the DSMs behaviour under different loads, suggesting recommendations to improve their current design.     

Shielding Analysis for ITER Equatorial Port Cell During Blanket Replacement

ITER port cells are located outside the bio-shield of the Tokamak. During shutdown, the shielding blanket may be replaced and the radioactive blankets will be transported through equatorial port cells, increasing the radiation exposure in the gallery. To examine the dose rate in the gallery with respect to the dose limitation specified by ITER, the activation of typical shielding blanket was calculated using the cell based rigorous two-step method. Then the activated blankets were loaded in cask and moved to the port cell, the radiation level in the port cell and gallery during the worst case was calculated. The shielding capability of port cell door was analyzed and the design was optimized based on the present proposal. As shown from the results, the dose rate from cask is much higher than that from activated Tokamak. The main concern for port cell door should be the concrete lintel and penetrations through it, providing basis for further engineering design of the port cell shielding.     

Global dose rate assessment in ITER diagnostics ports based on the 3-D FEM ATTILA code

The USITER, through the Princeton Plasma Physics Lab (PPPL), is responsible for the delivery of several fully integrated upper, equatorial and lower port plugs dedicated for the diagnostics in ITER. Each port plug package consists of a generic port plug structure and a set of diagnostics and diagnostic housings. The shielding design of the integrated port plugs calls for maintaining a dose level not to exceed 100 μSv/h inside the interspace of each port; the room behind the port plug where maintenance personnel access the rear of the port. This is set as an upper target design in order to perform routine maintenance 1E6 sec (～two weeks) following shutdown. Expensive remote handling robots and tooling are required otherwise. In this paper we present results from a parametric study aimed at providing initial assessment of the attainable dose rates in the diagnostics ports and their extension areas in order to properly address the duration time and frequency for the workers to perform the scheduled maintenance. The nuclear analysis is performed using both the serial version and the distributed memory parallel (DMP) version of the ATTILA-7.1.0, 3-D FEM Discrete Ordinates code, along with the FENDL2.1/FORNAX and ANSI/ANS-6.1.1-1977 data bases.     

Neutronic analysis of ITER radial x-ray camera

The radial x-ray camera(RXC) is designed to measure the poloidal profile of plasma x-ray emission with high spatial and temporal resolution. The RXC diagnostic system consists of internal camera module and external camera module that view the core region and outer region through the vertical slots of the diagnostic first wall and diagnostics shield module of the equatorial port plug. To ensure the normal performance of the silicon photodiode array detectors of the cameras in the hard neutron irradiation environment in ITER tokamak, it is necessary to calculate neutron flux, radiation damage and the nuclear heating of the silicon photodiode array detectors and simulate the radiation maps of the range of RXC. This work estimated the nuclear environment of RXC based on Monte Carlo N-particle transport code, plasma scenarios of ITER tokamak and the RXC-integrated ITER CLITE model. The neutron flux of silicon photodiode array detectors and the lifetime of the silicon photodiode detector in the camera were calculated. The neutronic analysis results show that the shielding design has achieved the effect as expected and is able to guarantee the normal work of the detector during the ITER deuterium–deuterium phase without replacement, three detectors of the external camera can be operated during the whole deuterium–tritium phase without replacement.     

Design progress on ITER port plug test facility

To achieve the overall ITER machine availability target, the availability of diagnostics and heating port plugs shall be as high as 99.5%. To fulfill these requirements, it is mandatory to test the port plugs at operating temperature before installation on the machine and after refurbishment.The ITER port plug test facility (PPTF) provides the possibility to test upper and equatorial port plugs before installation on the machine. The port plug test facility is composed of several test stands. These test stands are first used in the domestic agencies and on the ITER Organization site to test the port plugs at the end of manufacturing. Two of these stands are installed later in the ITER hot cell facility to test the port plugs after refurbishment. The port plugs to be tested are the Ion Cyclotron (IC) heating and current drive antennas, Electron Cyclotron (EC) heating and current drive launchers, diagnostics and test blanket modules port plugs.Test stands shall be capable to perform environmental and functional tests. The test stands are composed of one vacuum tank (3.3 m in diameter, 5.6 m long) and the associated heating, vacuum and control systems. The vacuum tank shall achieve an ultimate pressure of 1 × 10^?5 Pa at 100 °C containing a port plug. The heating system shall provide water at 240 °C and 4.4 MPa to heat up the port plugs. Openings are provided on the back of the vacuum tank to insert probes for the functional tests.This paper describes the tests to be performed on the port plugs and the conceptual design of the port plug test facility. The configuration of the standalone test stands and the integration in the hot cell facility are presented.