多物理耦合分析自动建模软件SuperMC/MCAM5.2设计与实现

超级蒙特卡罗核计算仿真软件系统SuperMC是一套通用、智能、多功能的核系统设计与安全分析软件。多物理耦合分析自动建模软件SuperMC/MCAM(multi-physics coupling analysis modeling program)是其中的自动建模模块,其目标是为多物理耦合分析提供精确高效的建模功能。SuperMC/MCAM最新版本SuperMC/MCAM5.2支持SuperMC、MCNP、FLUKA、Geant4、TRIPOLI等多种蒙特卡罗程序计算模型的自动建模,可进行CAD模型与蒙特卡罗计算模型之间的自动双向转换,以及进行各蒙特卡罗程序计算模型之间的相互转换。本文对SuperMC/MCAM5.2的功能及多蒙特卡罗几何正向与反向转换方法进行了介绍,采用国际热核实验堆ITER基准模型对SuperMC/MCAM5.2进行了测试,测试中SuperMC/MCAM5.2生成的各蒙特卡罗模型计算结果一致,证明了SuperMC/MCAM5.2建模功能的正确性和有效性。     

ITER中国液态锂铅实验包层模块中子学分析

基于ITER装置全模型,借助于MCNP自动建模程序MCAM,将TBM模块插入该模型的赤道窗口,使用MCNP/4C和FENDL1.0数据库,对DLL和SLL两个典型子模块设计进行三维中子学计算和分析,给出TBM模块核热功率密度分布以及氚增殖能力.     

MCAM在ITER窗口限制器蒙特卡罗计算建模过程中的应用

在MCAM(蒙特卡罗粒子输运计算自动建模系统)平台下,实现了ITER窗口限制器CAD工程模型的预处理以及到蒙特卡罗计算模型的自动转换,并借助MCAM的反向转换功能和MCNP中二维几何截面绘制程序PLOT对生成的计算模型进行了验证.MCAM在ITER窗口限制器蒙特卡罗建模过程中应用结果表明对于一个具有复杂几何结构的CAD工程模型来说,MCAM提高了其蒙特卡罗计算建模的效率和准确性,可以有效地替代传统的手工建模方法,应用于实际计算分析过程中.     

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。     

利用ITER基准模型对MCAM4.2进行检验(I)

利用ITER三维基准模型对蒙特卡罗方法粒子输运自动建模软件MCAM4.2进行了检验测试,实现了CAD工程模型的预处理,自动转换生成MCNP计算输入文件,并完成测试要求的中子壁负载和偏滤器中子注量率与核热沉积的计算。ITER模型的成功处理与计算表明MCAM能够正确和有效地处理大型复杂几何模型。     

结合几何的MCNP计算结果可视化分析系统的研究与实现

蒙特卡罗粒子输运程序MCNP在分析复杂例题的模拟计算结果数据时,具有分析过程不直观、效率低等问题。FDS团队结合科学计算可视化技术、GPU(图形处理器)可视化编程技术,以及本团队自动建模系统MCAM的反向转换引擎,自主研发了集数据分类提取、图形化管理以及结合几何的三维数据可视化分析功能于一体的MCNP计算结果可视化分析系统。以国际热核实验堆ITER为代表的大量测试结果表明,该系统显著提高了数据的利用率和分析效率。     

MCAM在ITER装置TRIPOLI三维中子学建模中的应用

目前,国际热核聚变实验反应堆ITER装置仅有针对MCNP程序的三维中子学基准模型(ITERA-lite4),因此无法使用TRIPOLI程序对ITER装置进行中子学计算分析.本文利用蒙特卡罗计算自动建模软件MCAM 5.1创建ITER装置的三维中子学TRIPOLI模型,并通过TRIPOLI程序对其进行中子学计算分析.计算...     

铀加工与燃料制造设施核临界事故所致瞬发剂量计算研究

建立了基于蒙特卡罗（MCNP）程序建模的铀加工与燃料制造设施核临界事故工况下瞬发剂量的计算方法，并将该计算方法与EJ/T 988—96规定的计算方法进行了比较分析。以我国某核燃料元件研发厂址为例，采用MCNP程序建模计算了该厂址核临界事故对厂界公众所致的瞬发剂量。结果表明，EJ/T 988—96的计算方法过于保守的估计了核临界事故工况下的瞬发剂量；基于MCNP程序建模的计算方法，因其求解算法的科学性和模型对屏蔽介质的准确描述，以及结果误差的可控性，使得计算结果更准确。因此，建议采用基于MCNP程序建模的方法计算铀加工与燃料制造设施核临界事故下的瞬发剂量。      

MCNP大规模重复结构体素的实时交互可视化方法研究

蒙特卡罗粒子输运程序MCNP在使用重复结构描述复杂的体素模型时,将会产生大规模的重复结构体素数据.当前对这种数据进行可视化分析时,缺乏对可视化模型进行交互操作的实时响应方法.本文提出了一种基于数据裁剪和体素隐藏的大规模重复结构体素数据的实时交互可视化方法,并基于可视化开发包VTK加以实现.以FDS团队自主构建的中国成年女性数字化辐射计算模型FDS-HUMAN为代表的测试结果表明,该方法不仅直观的实现了模型的可视化,同时还支持对旋转、切片和放缩等交互操作的快速响应.本文的方法实现了对MCNP大规模重复结构体素数据的实时交互可视化,提高了MCNP计算数据的可视化分析效率.     

cosRMC复杂曲面开发及在聚变中子学的应用

中国聚变工程实验堆（CFETR）是我国自主设计和研制的重大科学工程,CFETR旨在与ITER相衔接和补充,为研制DEMO级别聚变堆电站提供必要的技术。蒙特卡罗方法在聚变中子学与屏蔽设计等方面具有重要作用。本文基于自主化蒙特卡罗程序cosRMC,研究了蒙特卡罗复杂曲面建模的数学模型和计算方法,开发了复杂曲面建模功能,并通过PPCS（power plant conceptual study）模型验证了该功能实现的正确性。然后构建了CFETR的三维精细化模型,并利用该模型对CFETR包层设计中的关键中子学参数进行计算分析。结果表明,cosRMC对中子学参数氚增殖比、中子壁载荷和核热沉积的计算结果与MCNP的计算值吻合良好,相对偏差均小于5%,满足工程设计需求。研究证明了cosRMC应用于聚变堆包层中子学分析的正确性和有效性。CFETR中子学参数的计算分析,也为其设计和优化提供了参考。     

Study on Modeling Technology in Digital Reactor System

Modeling is the kernel part of a digital reactor system. As an extensible platformfor reactor conceptual design, it is very important to study modeling technology and develop somekind of tools to speed up preparation of all classical computing models. This paper introducesthe background of the project and basic conception of digital reactor. MCAM is taken as anexample for modeling and its related technologies used are given. It is an interface program forMCNP geometry model developed by FDS team (ASIPP & HUT), and designed to run on windowssystem. MCAM aims at utilizing CAD technology to facilitate creation of MCNP geometry model.There have been two ways for MCAM to utilize CAD technology: (1) Making use of user interfacetechnology in aid of generation of MCNP geometry model; (2) Making use of existing 3D CADmodel to accelerate creation of MCNP geometry model. This paper gives an overview of MCAM'smajor function. At last, several examples are given to demonstrate MCAM's various capabilities.     

Neutronics design analyses of fusion power reactors based on a novel integral approach

A novel integral approach was applied for the nuclear design analyses performed for the European DEMO Conceptual Study including, for the first time, the automatic generation of analysis models for Monte Carlo calculations from available CAD geometry data by the new McCad conversion tool. Starting from neutronics pre-analyses to define the radial reactor build, a generic neutronics CAD model of the DEMO reactor was constructed serving as basis for the generation of DEMO reactor models employing the HCLL (Helium Cooled Lithium Lead) and HCPB (Helium Cooled Pebble Bed) blankets for the tritium breeding and the power production. The HCLL and HCPB DEMO reactor models were converted to the MCNP geometry representation by the newly developed McCad software tool. The nuclear analyses performed on the basis of MCNP Monte Carlo calculations using the converted models showed that both DEMO reactor variants could satisfy the requirements for a sufficient shielding and tritium breeding performance. As a major outcome of this work it is concluded that the newly established integral approach for neutronics analyses, including the automatic generation of analysis geometry models by McCad, is mature for applications to reactor design analyses and studies.     

TRIPOLI-4® Monte Carlo code ITER A-lite neutronic model validation

3D Monte Carlo transport codes are extensively used in neutronic analysis, especially in radiation protection and shielding analyses for fission and fusion reactors. TRIPOLI-4^® is a Monte Carlo code developed by CEA. The aim of this paper is to show its capability to model a large-scale fusion reactor with complex neutron source and geometry. A benchmark between MCNP5 and TRIPOLI-4^®, on the ITER A-lite model was carried out; neutron flux, nuclear heating in the blankets and tritium production rate in the European TBMs were evaluated and compared. The methodology to build the TRIPOLI-4^® A-lite model is based on MCAM and the MCNP A-lite model. Simplified TBMs, from KIT, were integrated in the equatorial-port. A good agreement between MCNP and TRIPOLI-4^® is shown; discrepancies are mainly included in the statistical error.     

Detailed nuclear analysis of ITER ELM coils

The ELM coils in ITER are intended to provide control of Edge Localized Modes (ELMs). These coils are located on the outboard side of ITER between the shield modules and vacuum vessel (VV) and are subject to high radiation levels. Detailed three-dimensional (3-D) models of the toroidal and poloidal legs of the ELM coil and surrounding region for the MCNP code were updated to reflect the latest design changes. Neutronics calculations were performed to determine a variety of radiation damage parameters for the ELM coils as well as the VV located behind them. Additionally, detailed CAD based models for the upper ELM coil region were used to perform a CAD based analysis using the DAG-MCNP5 code. The results show that the ELM coil will meet the specified material radiation limits. However, the nuclear heating on the vacuum vessel behind the poloidal multi-pipe manifolds will exceed the specified limit.     

Use of MCAM in creating 3D neutronics model for ITER building

The three dimensional (3D) neutronics reference model of International Thermonuclear Experimental Reactor (ITER) only defines the tokamak machine and extends to the bio-shield. In order to meet further 3D neutronics analysis needs, it is necessary to create a 3D reference model of the ITER building. Monte Carlo Automatic Modeling Program for Radiation Transport Simulation (MCAM) was developed as a computer aided design (CAD) based bi-directional interface program between general CAD systems and Monte Carlo radiation transport simulation codes. With the help of MCAM version 4.8, the 3D neutronics model of ITER building was created based on the engineering CAD model. The calculation of the neutron flux map in ITER building during operation showed the correctness and usability of the model. This model is the first detailed ITER building 3D neutronics model and it will be made available to all international organization collaborators as a reference model.     

Use of the MCNP-polimi code for time-correlation safeguards measurements

Active nuclear safeguards measurements that rely on the time correlation between fast neutrons and gamma rays from the same fission are becoming a useful technique. In previous works we have shown the feasibility of this method, in conjunction with the use of the well-known MCNP simulation code and the use of artificial neural networks, to estimate the mass and enrichment of fissile samples enclosed in special, sealed containers.

In a more recent works, we pointed out some features of MCNP that represent a drawback in the simulation of correlation measurements. In fact, MCNP is not intended for obtaining second order moments. Therefore, to achieve an agreement between the simulations and the experiment, we resorted to the use of effective physical parameters (for example detection threshold and light output). Recently, by suitably modifying MCNP, we developed the MCNP-PoliMi code, which attempts to simulate the physics of each interacton more realistically.

In this paper, we present the results of the calibration of a plastic scintillator using a Cf-252 source, traditionally used in the above-mentioned nuclear safeguards experiments. We show that the physical parameters found by calibration and used in the MCNP-PoliMi simulations are reasonable for this type of detector, and that the correlation functions thereby obtained are in good agreement with the experiment.