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.     

基于SuperMC的ITER下窗口生物屏蔽插件屏蔽分析

在国际热核聚变实验堆(ITER)中,窗口生物屏蔽插件需为电子设备和工作人员提供必要的辐射屏蔽防护。基于中子学分析的生物屏蔽插件设计是ITER设计的重要内容。本文基于超级蒙卡核模拟软件SuperMC,在ITER大厅三维中子学模型中整合了ITER设计整合部门(DIN)最新设计的下窗口生物屏蔽插件模型,对四种下窗口生物屏蔽插件进行了屏蔽分析。分析结果显示,低温恒温器低温泵生物屏蔽插件中子屏蔽性能最好,室内监视系统生物屏蔽插件屏蔽性能最差;室内检视系统生物屏蔽插件停堆剂量率最小,环形低温泵生物屏蔽插件停堆剂量率最大。在SA2辐照方案下,停堆12天后,环形低温泵生物屏蔽插件处停堆剂量率超过规定限值20倍。分析结果表明,ITER下窗口生物屏蔽插件设计有待优化。     

Design of the in-wall shielding for the ITER neutral beam port

The ITER neutral beam port is composed of connecting duct, port extension and port stub extension. The spaces between inner and outer shells of the port extension and port stub extension are filled with pre-assembled blocks, called in-wall shielding. The main purpose of IWS is to provide neutron shielding for the superconducting magnet, thermal shield and cryostat from the main vessel during plasma operation. In order to provide effective neutron shielding capability with the cooling water, 40 mm thick flat plates (steel type 304B4) are used in almost all areas of the volume between port shells. The IWS is composed of shield plates, upper/lower brackets and bolt/nut/washers. Major activities during design work are to develop installation concept of the IWS blocks for easy assembly into port structures and to perform structural analysis to assess sufficient strength, fabrication feasibility study and 3D modeling including drawing works.In this paper, major results of mechanical design are introduced. First, the design requirements for IWS and the developed IWS designs for easy assembly into the port structure are introduced. Second, is introduced the engineering analysis results to assess structural integrity. And then the fabrication feasibility study results are presented for major fabrication processes. Lastly, conclusion and future works are mentioned.     

反应堆压力容器屏蔽组件设计改进

During the hot functional test of one NPP, the neutron shielding material was heated and released from the reactor vessel shielding blocks. The structure and layout of the block were redesigned, and B4C was adopted as the neutron shielding material. This paper analyzes the improved design scheme in terms of the heat transfer, the radiation shielding and GSI191. The result indicates that the improved design meet the requirements. During the supplemental hot function test, the temperature of neutron shielding block and module and the radiation dose in the containment were surveyed, and the effectiveness of the new design scheme is further verified.     

反应堆压力容器屏蔽组件设计改进

某机组热试期间反应堆压力容器屏蔽组件屏蔽材料受热泄漏,因此针对屏蔽盒结构和布置进行了优化设计,选用B4C作为中子屏蔽材料。本文从热传递、辐射屏蔽、GSI191等方面对改进的设计方案开展了分析。结果表明,改进的设计满足使用和规范要求。补充热试期间,对屏蔽盒及模块温度场、安全壳内辐射剂量水平进行了测量,进一步验证了改进设计的有效性。     

Shut-Down Dose Rate analysis for ITER Diagnostic Equatorial and Upper Ports

The Shut-Down Dose Rate (SDDR) is an important criterion of radiation safety for the personnel access for maintenance operations in ITER ports after the cessation of the D-T 14 MeV neutron fusion source. Therefore, the problem of the SDDR calculations attracts the attention of fusion neutronics community because SDDR in such a large and geometrically complicated fusion device as the ITER tokamak is challenging to compute. This challenge has been faced and overcome by applying dedicated methodological approaches explained in this paper. The results of the SDDR analysis allowed us to propose several design solutions for improvement of the radiation shielding of the ITER Generic Diagnostic Equatorial and Upper Port Plugs (EPP and UPP). The SDDR analysis was focused on the interspace area located between the ITER bioshield and port plugs where the personnel access is envisaged at ∼12 days after the ITER shut-down. By this analysis the radiation streaming pathways and dominant sources of decay radiation were revealed and the methods to mitigate the streaming and subsequent activation were found. The optimization of the port shielding was targeted on minimization of the SDDR in the interspace area following the ALARA principle and taking into account the feasibility to implement proposed shielding options with the actual hardware. Among them, wrapping the EPP walls with the B₄C tiles improves the EPP shielding performance. While void around the ELM/in-vessel coils and blanket manifolds leads to the performance reduction. The SDDR inside the Generic UPP interspace depends mainly on the environment (blanket, manifolds, and gaps).     

HIMM治疗室的辐射屏蔽设计

为了保证医用重离子加速器（HIMM）运行时的辐射安全,利用FLUKA计算了治疗时产生的瞬发中子源项,并对次级中子、γ辐射对屏蔽的影响进行了分析。用半经验公式及FLUKA计算了屏蔽厚度,给出了HIMM治疗室的屏蔽设计。在HIMM最大负载运行时,测量了屏蔽外中子剂量率,测量结果与模拟计算结果相符合。结果表明,本文选用的屏蔽设计方法是合理的,HIMM治疗室屏蔽设计方案满足国家标准要求。     

15MV医用电子直线加速器光核中子剂量分布的MC模拟及测量

医用电子直线加速器产生的X射线已广泛应用于放射治疗过程,X射线与机头中的高Z物质(铅、钨、铜和铁)发生(γ,n),(γ,2n)反应产生一定量的中子,引起与治疗无关的中子剂量。本文对工作在15MV能量档的Prim μs-M型医用电子直线加速器在标准照射野10cm×10cm内治疗平面的光中子剂量分布,进行了Monte-Carlo模拟,并使用CR39固体核径迹探测器和中子气泡探测器(NBD)进行了实验测量。研究发现,测量与模拟的中子剂量之间最大偏差约±30%,其最主要的原因是由于"加速器产生的光核中子与物质发生非弹性散射反应"而逐步降低能量,产生了低于上述两种探测器阈能(100keV)的中子,使测量值比模拟值偏低。研究结果为X射线放射治疗中减低污染中子剂量的优化设计提供了基础数据。     

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.     

MC法优化设计~(252)Cf中子源辐射屏蔽装置

利用MCNP5程序构建了屏蔽装置模型,并模拟了聚乙烯、含质量分数10%硼的石蜡、重水、石墨和铅等材料的中子慢化和屏蔽效果,以及铁对γ射线的屏蔽效果。当中子慢化剂聚乙烯的厚度达5 cm时,透过慢化层发射出的中子注量率达到最大值为5.40×10-4m~(-2)s~(-1)。中子屏蔽层含硼石蜡厚度为33 cm并且γ屏蔽层铁厚度为4 cm时,由中子和γ射线产生的年有效剂量之和满足国家标准相关限值要求。     

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。     

241Am-Be中子源紧凑型屏蔽装置及辐照器设计

²⁴¹Am-Be中子源被广泛用于实验研究,为保护实验人员免受中子及γ射线照射,需要设计适当的屏蔽。利用蒙特卡罗方法计算中子透射不同材料后的能谱分布与剂量,优选各层屏蔽材料种类与厚度,设计一套²⁴¹Am-Be中子源紧凑型屏蔽装置。装置由内而外采用钨+聚乙烯+含硼聚乙烯+不锈钢进行防护,外表面周围剂量当量率H^*(10)低于10μSv/h,满足辐射防护要求。同时对装置内部热中子、超热中子和快中子注量分布进行研究,确定装置快中子和热中子输出通道最佳位置。在辐照装置同时开放快中子和热中子通道进行实验测试时,需要设置距离大于130 cm的控制区,以保障操作人员安全。     

医院中子照射器分析室剂量场研究

医院中子照射器建成后,对分析室内及其屏蔽门外的γ剂量率和中子剂量当量率进行了测量,测量结果显示：分析室内局部γ剂量率与设计值相差较大,分析室屏蔽门外γ剂量率超过原设计监督区限值7.5 μSv/h,因此需对分析室内部及其屏蔽门进行屏蔽改造。根据蒙特卡罗程序模拟计算结果及实际使用情况给出最终屏蔽方案,即在分析室束流孔道所在墙面加装厚度为16 cm的铅屏蔽材料屏蔽γ射线,对四周墙面及屏蔽门内侧加装厚度为1 cm的含锂聚乙烯板屏蔽散射中子。改造后分析室剂量最高点γ剂量率下降277倍,中子剂量当量率下降5.8倍,屏蔽门外γ剂量率下降近90倍。     

Neutron shielding and activation of the MASTU device and surrounds

A significant functional upgrade is planned for the Mega Ampere Spherical Tokamak (MAST) device, located at Culham in the UK, including the implementation of a notably greater neutral beam injection power. This upgrade will cause the emission of a substantially increased intensity of neutron radiation for a substantially increased amount of time upon operation of the device. Existing shielding and activation precautions are shown to prove insufficient in some regards, and recommendations for improvements are made, including the following areas: shielding doors to MAST shielded facility enclosure (known as “the blockhouse”); north access tunnel; blockhouse roof; west cabling duct. In addition, some specific neutronic dose rate questions are addressed and answered; those discussed here relate to shielding penetrations and dose rate reflected from the air above the device (“skyshine”). It is shown that the alterations to shielding and area access reduce the dose rate in unrestricted areas from greater than 100 μSv/h to less than 2 μSv/h averaged over the working day.The tools used for this analysis are the MCNP (Monte Carlo N-particle) code, used to calculate the three-dimensional spatial distribution of neutron and photon dose rates in and around the device and its shields, and the nuclear inventory code FISPACT, run under the umbrella code MCR2S, used to calculate the time-dependent shutdown dose rate in the region of the device at several decay times.     

Measurement of neutron tissue dose outside reactor shielding

RUS instruments, developed by the authors, are described that enable one to measure the flux and tissue dose rate of intermediate neutrons, which make a significant contribution to neutron tissue dose outside reactor shielding. The neutron dose composition was investigated in experiments at the IRT-1000 reactor, and it was shown that it depends essentially on shielding composition. It was established that the neutron tissue dose computed from readings taken with the RPN-1 instrument were actually too low by a factor amounting to one and one half outside water shielding and to five outside concrete shielding.Translated from Atomnaya Énergiya, Vol. 15, No. 5, pp. 386–393, November, 1963     

Application of High Performance Computing Analysis Technology in Shielding Design of Large Sodium-cooled Fast Reactor

The sodium-cooled fast reactor container is an integrated pool structure composed of numerous internal components and complex structure. The anisotropy is obvious and the deep penetration problem is serious in the process of neutron transport from core to biological shielding. The calculation of three-dimensional S_N method in large scale is the bottleneck restricting in the design of fast reactor shielding. By combining with high performance computing technology, the parallel computing scheme is used to solve the anisotropic three-dimensional deep penetration shielding calculation in the fast reactor. In this paper, the China Demonstration Fast Reactor (CFR600) reactor block was taken as the research object. Using JSNT-CFR code, the neutron flux rate, photon flux rate, and dose rate in the reactor block were calculated in detail. The calculation results were compared with those of the existing two-dimensional code. The results show that combining the traditional shielding calculation method with high performance computing can meet the requirements of CFR600 reactor block shielding calculation accuracy, and obtain a more comprehensive three-dimensional display effect. It can solve the problem of shielding calculation of complex problems such as complex model and particle penetration depth. It has obvious advantages and provides strong support for the large sodium-cooled fast reactor shielding design.     

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。     

声核中子测量的屏蔽设计

通过测定声空化核效应实验室各测量点的中子注量率,了解实验室墙壁和地面对出射中子的散射,选定散射中子相对较弱的位置作为声核中子测量点。利用SHIELD程序模拟不同材料的中子屏蔽效果,选用4cm铁和20cm含硼石蜡组成屏蔽体,以降低中子本底。测定影屏蔽及影屏蔽结合BF₃正比计数管环绕屏蔽两种方式下的散射修正因子F_s,提出以统计显著性增量S.S.I≥3/[KF(]F_s[KF)]作为超声中子计数相对于非超声中子计数的增量ΔC是否具有统计意义的判据。     

Shielding Design and Analysis of Fusion–Fission Hybrid Energy Reactor Blanket

A preliminary design of fusion–fission hybrid energy reactor (FFHER) has been proposed by Institute of Nuclear Physics and Chemistry based on current fusion science and well-developed fission technology. In FFHER, shield blocks provide nuclear shielding and thermal shielding for internal and external blanket components. The hybrid of fusion core and fission blanket makes the spectra rather complex. Therefore, it is necessary to make detail shielding design and carry out radiation analysis according to the blanket structure and material property. In this study, a shielding design of combining several different material shield blocks has been proposed. The shielding analysis is performed by Monte Carlo (MC) method. For the radiation deep-penetration problem, the flux and statistical relative error of forward MC estimate are applied to get an optimal weight window for global variance reduction (GVR). The spatial distribution of neutron and gamma flux have been assessed along the shield block depth. Power deposited and dose rate distributions have also been simulated and analysed. Neutron irradiation damage has been studied to evaluate the material damage. Based on the configuration analysis, nuclear analysis and GVR method, an optimal FFHER blanket shielding design has been obtained.     

Erosion evaluation capability of the IVVS for ITER applications

In ITER it is foreseen the use of the In Vessel Viewing System (IVVS), whose scanning head is a 3D laser imaging system able to obtain high-resolution intensity and range images in hostile environments. The IVVS will be permanently installed into a port extension, therefore it has to be compliant with ITER primary vacuum requirements. In the frame of a Fusion for Energy Grant, an investigation of the expected IVVS metrology performances was required to evaluate the device capability to detect erosions on ITER first wall and divertor and to estimate the amount of eroded material. In ENEA Frascati laboratories, an IVVS probe prototype was developed along with a method and a computational procedure applied to a reference erosion plate target simulating ITER vessel components and their possible erosions. Experimental tests were carried out by this system performing several scans of the reference target with different incidence angles, estimating the eroded volume and comparing this volume with its true value. A dedicated study has been also done by changing the power of the laser source; a discussion about the quality of the 3D laser images is reported. The main results obtained during laboratory tests and data processing are presented and discussed.