Use of a genetic algorithm in the search for a near-optimal shielding design

An optimization method based on genetic algorithm (GA), which is referred as MACroscopic Near-Optimal Shielding design (MACNOS), is proposed for the search for an optimal radiation shield configuration subject to a given set of constraints. In MACNOS, a GA is used to search for the optimal shielding design and the penalty strategy is employed to deal with the constraints. In order to confirm its capability to search for the optimal shielding design, MACNOS is applied for solving a simple problem with regard to radiation shielding optimization of a hypothetical spaceship reactor. The application shows that, keeping the constraints satisfied, MACNOS is able to seek for the shielding design that minimizes the total weight by changing the thickness and the material of the shield. Therefore, it is expected that MACNOS is potentially useful in the search for the optimal design configuration in the conceptual design phase, where the selection of the shielding material and the estimate of the thickness are necessary.     

Parametric neutronic analysis of HCLL blanket for DEMO fusion reactor utilizing vacuum vessel ITER FDR design

Three-dimensional parametric neutronics calculations using the Monte Carlo code MCNP-4C have been performed for a DEMO-type reactor based on the Helium-Cooled Lithium-Lead (HCLL) blanket. The aim of the analysis was to minimize the radial blanket thickness, while ensuring tritium self-sufficiency and to assess the shielding performance of the reactor in terms of the radiation loads to the super-conducting toroidal field (TF) coils. It was found that tritium self-sufficiency can be achieved with a breeder zone thickness reduced to no more than 55 cm at a ⁶Li enrichment of 90%. Assuming a ⁶Li enrichment of 60%, a breeder zone thickness of 60 cm is required to achieve the target TBR of 1.10 which is assumed to be sufficient to cover potential tritium losses and uncertainties. With regard to the shielding performance it was found that the design limits for the radiation loads to the TF-coil can be met with radial blanket thicknesses of 75 cm, 60 cm and 55 cm utilizing a two-component shield of Eurofer steel and tungsten carbide between the breeder zone and the vacuum vessel. The blanket variants with larger radial breeder zone show better shielding performances due to the reduced Eurofer shielding material acting as gamma radiation emitter in between the breeder zone and the vacuum vessel. In particular the radiation dose absorbed in the Epoxy insulator was shown to be the most critical quantity in this regard.     

辐射条件下球腔冷冻靶温度场分布数值研究

惯性约束聚变冷冻靶系统中，为成功实现靶丸点火，冰层厚度均匀性需达到99%，表面粗糙度的均方根要小于1 μm。控制靶丸表面最大温差小于0.1 mK能满足以上点火要求。为研究辐射对惯性约束聚变间接驱动靶丸的温度场影响，建立了三维对称球腔冷冻靶系统的计算模型。考虑球腔内部激光入射口封口膜吸收率以及外部辐射温度对球腔内部温度场分布的影响，利用FLUENT软件对球腔冷冻靶温度场进行了数值模拟计算。研究表明：球腔由于自身具有的球对称几何结构，其内部的温度场分布更加均匀；受外界辐射影响，有窗侧靶丸表面温度较无窗侧温度高；辐射温度越高，靶丸表面的绝对温度越高，虽然靶丸表面的温差变化基本可忽略，但要防止由于外界辐射温度过高而导致的DT冰层均匀性恶化，应选用多层屏蔽罩结构降低辐射的影响；激光入射口封口膜吸收率大于0.2时，靶丸表面温差显著增大。     

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.     

Relation between material configuration and radiation shielding capability - application of radiation shielding characteristic functions

For the purpose of finding a principle for material configuration which an ideal radiation shielding in slab geometry should obey, radiation energy dependence of material configuration is studied. In the course of study, radiation shielding capability for each system of different material configuration is evaluated by using radiation shielding characteristic functions defined as dose rates of transmitted radiations in response to isotropic incidence of radiations to the slab shield with pulse-like narrow energy distributions.In shielding neutrons by steel and water layers, recommendable material configuration depends on energy distribution of incident neutrons; all steel layers should be located in the source side of all water layers, if incident neutron energies are above 5 MeV: either homogeneous array of steel and water layers or above mentioned material configuration is recommendable, if incident neutron energies are between 2 MeV and 5 MeV: all water layers should be located in the source side of all steel layers, if incident neutron energies are below 2 MeV or incident neutrons have energy spectrum of fission neutrons.Above recommendation can be understood well by considering both energy dependence of neutron cross sections of each material and the maximum amount of energy degradation at elastic scattering in each material.In designing a neutron shield, shielding of secondary gamma rays is important as well as neutron shielding. This importance is demonstrated for several types of actual cask walls which are composed of many material layers by using the characteristic functions of neutrons and gamma rays for cask walls.     

Analysis for Radiation and Shielding Dose in Plasma Focus Neutron Source Using FLUKA

Monte Carlo simulations have been performed for the attenuation of neutron radiation produced at Plasma focus (PF) devices through various shielding design. At the test site it will be fired with deuterium and tritium (D-T) fusion resulting in a yield of about 10¹³ fusion neutrons of 14 MeV. This poses a radiological hazard to scientists and personnel operating the device. The goal of this paper was to evaluate various shielding options under consideration for the PF operating with D-T fusion. Shields of varying neutrons-shielding effectiveness were investigated using concrete, polyethylene, paraffin and borated materials. The most effective shield, a labyrinth structure, allowed almost 1,176 shots per year while keeping personnel under 20 mSV of dose. The most expensive shield that used, square shield with 100 cm concrete thickness on the walls and Borated paraffin along with borated polyethylene added outside the concrete allowed almost 15,000 shot per year.     

Neutronic studies for the optimization of shield wall penetrations for laser IFE systems

Building upon the inertial confinement fusion (ICF) technology developed for the National Ignition Facility (NIF) at Lawrence Livermore National Laboratory (LLNL), a laser-driven inertial fusion energy (LIFE) power plant is being designed. In this pre-conceptual design, the final optic is exposed to a variety of threats originating from the fusion target. These include prompt neutron and gamma fluxes, x-ray and ionic emissions. While x-rays and ions are stopped by the low-density chamber fill gas (6 μg/cc xenon), neutrons and gamma-rays are not significantly attenuated. In order to limit the consequences of such threats onto the penultimate optic and the rest of the laser systems, a shielding wall stands between the target chamber area and the laser bay. An optical telescope arrangement allows for the laser beam propagation from the penultimate to the final optic, through a pinhole in the shielding wall. These pinholes attenuate the neutron flux and reduce effective dose rates such that laser bay maintenance can be performed by humans. An optimum design of this laser pinhole requires a good understanding of the different design trade-offs that exist between shielding performance and survivability of the laser optical elements and are outlined in this work.This paper provides insight on the impact and influence of the pinholes on the radiation doses in the laser bay, which is located on the opposite side of the concrete shielding wall. After addressing the difficulties of evaluating shields containing penetrations, it establishes a guideline for the selection of different variables linked to the pinhole's design and gives a preliminary evaluation of the radiation fields in the laser bay. The study also helps identify the requirements to enable manual and/or remote maintenance during operation, by determining the minimum achievable effective dose rates for different shield wall designs. Since the ability to perform maintenance during plant operation is an important contributor to high laser availability, we will propose the use of non-aligned double shield walls with pinholes.     

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.     

伽玛治疗刀的辐射屏蔽设计计算

报道了伽玛治疗刀的辐射屏蔽设计计算方法和计算结果。本伽玛治疗刀由３０个每个７．４ＥＢｑ的６０Ｃｏ点源，半球状屏蔽层、侧屏蔽柜和防护门组成，计算了后屏蔽体，侧屏蔽柜和防护门的屏蔽厚度，对泄漏辐射的监则结果表明，本伽玛治疗刀的屏蔽是足够安全的，计算的空气比释动能率的限制值和测量值相符合。     

High Magnetic Field Shielding for Sensitive Devices Relevant to ITER

High magnetic field shielding has been increasingly important for engineering design in recent years. In this report, a cylindric shield made from soft iron is studied using FEM （finite element method） analysis and COlnpared with experiments. The residual fields inside the shield are calculated and measured in both parallel and perpendicular fields up to 2000 Gs. The calculated results are compared with the experiments, and the input B-H curve is modified for a better conformity. The results indicate that the covers could greatly improve the shielding performance of the cylindric shield in our research. The comparison result shows that a proper B-H curve, which can well describe the material properties, is very important in FEM analysis and should be selected carefully.     

TOPAZ-Ⅱ空间核反应堆电源辐射屏蔽优化措施影响分析

空间核反应堆辐射屏蔽可减弱中子与γ光子通量对仪器设备的辐照,温度是影响辐射屏蔽性能的重要因素。利用Fluent软件对TOPAZ-Ⅱ空间核反应堆电源辐射屏蔽在真空环境下的换热行为进行了数值模拟,提出了优化措施,并揭示了优化措施对其温度场的影响和其温度分布特性。研究结果表明：冷却剂管道及其管槽表面喷涂低发射率的涂层具有均匀辐射屏蔽温度场和降低其温度峰值的效果,最优化参数为管道表面涂层的发射率为0.1～0.3、管槽表面涂层的发射率为0.1,且管道表面喷涂低发射率涂层效果要优于管槽;在冷却剂管道、管槽间添加单层表面发射率为0.04～0.07的真空遮热板也可使辐射屏蔽温度场与温度峰值处于最优状态。     

10MW高温气冷实验堆工艺系统的辐射源及屏蔽设计

介绍了１０ＭＷ高温气冷实验堆（ＨＴＲ－１０）与氦冷却剂相关的工艺系统设备中的辐射源，并以ＱＡＤ－ＣＧ程序完成了各设备间的辐射屏蔽计算。计算结果表明，工艺系统各设备中的辐射源强较低，即使对这些设备不进行附加屏蔽，其大多数设备外表面处的辐射剂量率仍满足限定工作区剂量率管理限值要求，并且对这些设备所在房间进行整体屏蔽的要求不高（１０￣２０ｃｍ厚的温凝土即可）。因此，建筑物结构设计厚度就能满足要求。     

不同材料中子反射与屏蔽效应研究

利用Monte Carlo粒子输运计算程序Super MC对厚度1-5 cm的多种材料进行中子反射和屏蔽性能分析计算。这些材料包括金属材料铍、铅、铜、含硼钢以及~(238)U和非金属材料聚乙烯、氢化锂、混凝土,中子能段选取10~(-5) e V-20 MeV。结果显示,中子反射能力和屏蔽性能都会随着材料厚度而增加,但增加的幅度逐渐减小。铍和聚乙烯在中子反射和屏蔽方面性能优越,而常用来屏蔽γ射线的铅在这两方面性能都是8种材料中最差的。~(238)U只在材料厚度很小时性能卓著,随着材料厚度增加,其性能便远不如大部分材料。考虑到聚乙烯的力学性能较差,在屏蔽材料的选择上有很大的限制,所以在8种材料中,铍的综合性能相对较好。     

基于MC方法的高放废液玻璃固化厂屏蔽窗辐射屏蔽性能设计研究

屏蔽窗是高放废液玻璃固化厂重要的观察设备,安装在热室与操作廊之间的混凝土墙体内,起辐射防护和气密通风隔离作用.为确保厂房运行人员所受的照射剂量控制在电离辐射防护标准的限值内,需要对屏蔽窗的辐射屏蔽性能进行优化设计.本文使用MCNP蒙卡模拟程序,对硼玻璃和铅玻璃两种屏蔽窗进行辐射屏蔽性能的蒙卡模拟研究,计算出能保障运行操...     

基于助力型外骨骼的外照射防护服研制

本文用MCNP软件进行模拟计算,设计了屏蔽一定能量中子和γ射线的屏蔽材料,用此屏蔽材料制作中子、γ外照射防护马甲,防护马甲存在厚又重的问题,人体很难承受它的重量。通过研究助力型外骨骼装置,将助力型外骨骼装置与防护马甲组合形成防护服,用助力型外骨骼装置承受防护马甲的重量,解决了人体需要承重的难题。并用放射源和在现场对研制的防护服进行性能测试,屏蔽率、厚度、均匀性等各项技术参数均达到预期要求。     

Geometrical aspects of cylindric magnetic shields in strong static fields

Motivated by rrER (the International Thermonuclear Experimental Reactor),research on a magnetic shield against a strong field has been carried out.In this paper,a cylindric magnetic shield is studied by using the finite element method with a nonlinear magnetization curve.The geometrical aspects of shielding performance are identified and corresponding suggestions for application are provided.Among them,the effects of the edge and cover thickness have not been mentioned elsewhere to our knowledge.     

Direct heat generation and corresponding energy saving in nuclear reactors due to optimum synthesis of shielding materials

In this paper the calculation of direct heat generation and energy savings due to the penetration of 1.37 and 2.75 MeV energy photons, emitted from a Na-24 radiation facility, through double layer shielding slabs of aluminium, steel and lead is described. A comparison is being made among six different shielding material combinations in order to assess the optimum shield related to the maximum energy captured due to γ-rays penetration through the combined shielding materials.     

氢化锆用于小型钠冷快堆屏蔽的初步研究

为减少小型钠冷快堆（SSFR）堆侧的屏蔽厚度,本文选择氢化锆作为SSFR堆侧的屏蔽材料。使用一维离散纵标法（ANISN程序）计算了氢化锆在SSFR堆芯区能谱下的屏蔽特性,并计算了堆侧采用氢化锆和碳化硼的屏蔽厚度。结果表明：与堆侧采用碳化硼和不锈钢屏蔽相比,采用氢化锆和碳化硼屏蔽（碳化硼所占体积比小于0.3）,屏蔽厚度减小了大约20%。氢化锆和碳化硼混合屏蔽材料具有很好的屏蔽性能,可减小SSFR堆侧的屏蔽厚度。     

Importance Determination Method for Geometry Splitting with Russian Roulette in Monte Carlo Calculations of Thick and Complicated Core Shielding Structure

In case of a shielding analysis of the geometry having thick and complicated structures with a Monte Carlo code, it is a serious problem that it takes too much computer time to obtain results with good statistics. Therefore, it is very important to reduce variances in the calculation. In this study, a method to determine the importance function in 3-dimensional Monte Carlo calculation with geometry splitting with Russian roulette was developed for the shielding analysis of thick and complicated core shielding structures. Only two essential importance ratio curves for one material enable us to determine the importance function easily in the shielding calculation.

The validity of this method was confirmed through a simple benchmark calculation. From the comparison with the result obtained by using weight window (W-W), it was shown that the present method can give an accurate result on the same level with W-W method with less trial and errors. And this method was applied to an actual reactor core shielding analysis to confirm its applicability to a 3-dimensional thick and complicated structure.

Using this method, the variance reduced calculation can be easily realized with the developed importance determination procedure, especially in case that parameter survey calculations are required in order to determine the shield thickness in a design work of a thick and complicated structure. Accordingly, it became easier to use Monte Carlo method as a powerful tool for a reactor core shielding design.