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。     

CFETR氦冷固态增殖剂包层产氚性能研究及优化

氦冷固态增殖剂包层是中国聚变工程实验堆（CFETR）的3种候选包层概念之一。本文基于中国核工业西南物理研究院提出的一种氦冷固态增殖剂包层概念，通过蒙特卡罗输运程序MCNP5建立了包层三维中子学模型，探究了不同几何布置方案及结构设计参数对包层产氚性能的影响，得到了全堆氚增殖比（TBR）及极向各包层模块产氚分布，并由优化后的模型得到了包层模块核热分布。结果表明，优化后的TBR达到1.177，满足氚自持的最低要求。     

CFETR氦冷陶瓷增殖包层中子学分析

为满足中国聚变工程实验堆(CFETR)包层的应用要求,本文提出氦冷陶瓷增殖(HCCB)包层方案。为验证HCCB包层设计方案的合理性与可行性,采用三维蒙特卡罗粒子输运程序MCNP,计算和分析了HCCB包层方案的氚增殖比、中子壁负载、中子通量密度、核热、辐照损伤等中子学特性。结果表明,HCCB包层方案满足氚自持要求,中子通量密度和核热分布合理,屏蔽性能良好,基本满足设计要求。     

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。     

CFETR氚自持分析评估与验证策略

中国聚变工程实验堆（China Fusion Engineering Test Reactor, 简称CFETR）的主要目标之一是实现氚自持。采用氚平衡法对CFETR不同运行工况下的氚自持条件进行了分析评估。结果表明：在500 MW运行阶段,CFETR实现氚自持所需的最小氚增殖比（TBR_r）为1.098,小于CFETR增殖包层可达到的氚增殖比（TBR_a）,即在理论上满足氚自持条件。在此基础上,提出了CFETR未来通过定期的氚衡算来验证氚自持的基本策略。在基准输入参数和氚存量测量精度限制（1%）条件下,CFETR氚自持验证实验的运行周期需要大于22 d（氦冷包层）或87 d（水冷包层）。     

中国氦冷固态实验包层2×6模块中子学计算分析

为了满足ITER对波纹度的要求,核工业西南物理研究院提出了新的减少低活化铁素体钢的氦冷固态（HCSB）实验包层模块（TBM）设计方案。采用MCNP程序及ITER全堆MCNP模型,对新设计的2×6HCSB-TBM进行三维中子学计算分析,给出了模块产氚率、核热沉积和功率密度分布等结果。在ITER运行因子为22%时,HCSB-TBM的产氚率为12.68mg/d。TBM内总核热沉积为522.5kW,最高功率密度为11.8W/cm³,出现在氚增殖区Li₄SiO₄中。计算结果可为TBM进一步的结构、热工水力学优化及其他系统设计提供中子学数据。     

基于D-T聚变中子源的双功能液态铅锂包层中子学实验

中国双功能锂铅包层(Dual Functional Lithium-Lead,DFLL)是由中国科学院合肥物质科学研究院核能安全技术研究所设计的用于聚变反应堆的液态包层.由于聚变反应堆氚增殖包层的设计高度依赖于中子计算,为验证DFLL包层设计中所使用的核数据库和仿真软件,建立了DFLL包层实验模块,并基于D-T聚变中子...     

CFETR集成设计平台的核热耦合模块开发

正在研发的中国聚变工程实验堆(China Fusion Engineering Test Reactor,CFETR)集成设计平台包括物理设计平台和工程设计平台,工程设计平台采用模块化方式,包括磁体、真空室、偏滤器、中子学等模块。模块设计中涉及用计算流体动力学(Computational Fluid Dynamics,CFD)软件来对堆内部件开展热工水力分析,CFD的热源项包括中子学计算产生的核热,而中子学分析软件(如Monte Carlo N particle transport code,MCNP)核热输出结果文件存在CFD软件无法直接读取等问题。因此,基于网格-网格插值和点-点插值法,开发了中子学与CFD核热耦合模块,提供两种途径实现高精度的三维核热耦合。使用CFX软件,以CFETR的一种氦冷陶瓷包层(Helium Cooled Ceramic Breeder blanket,HCCB)中增殖单元模块为对象,进行了热工水力分析,计算结果表明了核热耦合模块功能的可靠性。     

超临界水冷固态实验包层中子学与热工水力特性研究

基于国际热核聚变实验堆(ITER)实验包层方案,提出了一个超临界水冷固态实验包层概念设计方案。设计采用Be作为中子倍增剂,Li4SiO4作为氚增殖剂,CLAM钢作为结构材料。包层第一壁采用多层盘道设计以提高第一壁出口温度,内部采用增殖剂与中子倍增剂分层布置以提高热沉积与氚增殖率。为验证包层设计的可行性,分析计算了三维包层氚增殖率与热沉积的分布,然后根据中子学计算得到的结果对超临界水冷固态实验包层进行了数值模拟研究。结果表明:包层功率密度分布较合理;氚增殖率满足运行中氚自持的要求;在冷却剂出口温度达到500℃条件下材料温度不超过限值。该设计方案能满足中子学设计与热工水力的要求。     

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.     

Analysis of Time-Dependent Tritium Breeding Capability of Water Cooled Ceramic Breeder Blanket for CFETR

Attaining tritium self-sufficiency is an important mission for the Chinese Fusion Engineering Testing Reactor(CFETR) operating on a Deuterium-Tritium(D-T) fuel cycle. It is necessary to study the tritium breeding ratio(TBR) and breeding tritium inventory variation with operation time so as to provide an accurate data for dynamic modeling and analysis of the tritium fuel cycle. A water cooled ceramic breeder(WCCB) blanket is one candidate of blanket concepts for the CFETR. Based on the detailed 3D neutronics model of CFETR with the WCCB blanket,the time-dependent TBR and tritium surplus were evaluated by a coupling calculation of the Monte Carlo N-Particle Transport Code(MCNP) and the fusion activation code FISPACT-2007.The results indicated that the TBR and tritium surplus of the WCCB blanket were a function of operation time and fusion power due to the Li consumption in breeder and material activation.In addition, by comparison with the results calculated by using the 3D neutronics model and employing the transfer factor constant from 1D to 3D, it is noted that 1D analysis leads to an over-estimation for the time-dependent tritium breeding capability when fusion power is larger than 1000 MW.     

Neutronics Comparison Analysis of the Water Cooled Ceramics Breeding Blanket for CFETR

China Fusion Engineering Test Reactor(CFETR) is an ITER-like fusion engineering test reactor that is intended to fill the scientific and technical gaps between ITER and DEMO.One of the main missions of CFETR is to achieve a tritium breeding ratio that is no less than 1.2to ensure tritium self-sufficiency.A concept design for a water cooled ceramics breeding blanket(WCCB) is presented based on a scheme with the breeder and the multiplier located in separate panels for CFETR.Based on this concept,a one-dimensional(1D) radial built breeding blanket was first designed,and then several three-dimensional models were developed with various neutron source definitions and breeding blanket module arrangements based on the 1D radial build.A set of nuclear analyses have been carried out to compare the differences in neutronics characteristics given by different calculation models,addressing neutron wall loading(NWL),tritium breeding ratio(TBR),fast neutron flux on inboard side and nuclear heating deposition on main in-vessel components.The impact of differences in modeling on the nuclear performance has been analyzed and summarized regarding the WCCB concept design.     

Neutronics Analysis of Water-Cooled Ceramic Breeder Blanket for CFETR

In order to investigate the nuclear response to the water-cooled ceramic breeder blanket models for CFETR, a detailed 3D neutronics model with 22.5otorus sector was developed based on the integrated geometry of CFETR, including heterogeneous WCCB blanket models,shield, divertor, vacuum vessel, toroidal and poloidal magnets, and ports. Using the Monte Carlo N-Particle Transport Code MCNP5 and IAEA Fusion Evaluated Nuclear Data Library FENDL2.1,the neutronics analyses were performed. The neutron wall loading, tritium breeding ratio, the nuclear heating, neutron-induced atomic displacement damage, and gas production were determined.The results indicate that the global TBR of no less than 1.2 will be a big challenge for the watercooled ceramic breeder blanket for CFETR.     

Updated neutronics analyses of a water cooled ceramic breeder blanket for the CFETR

The water cooled ceramic breeder (WCCB) blanket employing pressurized water as a coolant is one of the breeding blanket candidates for the China Fusion Engineering Test Reactor (CFETR).Some updating of neutronics analyses was needed,because there were changes in the neutronics performance of the blanket as several significant modifications and improvements have been adopted for the WCCB blanket,including the optimization of radial build-up and customized structure for each blanket module.A 22.5 degree toroidal symmetrical torus sector 3D neutronics model containing the updated design of the WCCB blanket modules was developed for the neutronics analyses.The tritium breeding capability,nuclear heating power,radiation damage,and decay heat were calculated by the MCNP and FISPACT code.The results show that the packing factor and 6Li enrichment of the breeder should both be no less than 0.8 to ensure tritium self-sufficiency.The nuclear heating power of the blanket under 200 MW fusion power reaches 201.23 MW.The displacement per atom per full power year (FPY) of the plasma-facing component and first wall reach 0.90 and 2.60,respectively.The peak H production rate reaches 150.79 appm/FPY and the peak He production reaches 29.09 appm/FPY in blanket module # 3.The total decay heat of the blanket modules is 2.64 MW at 1 s after shutdown and the average decay heat density can reach 11.09 kW m-3 at that time.The decay heat density of the blanket modules slowly decreases to lower than 10 W m-3 in more than ten years.     

Neutronics Experimental Study on Tritium Production in Solid Breeder Blanket Mockup with Neutron Reflector

Neutronics experiments have been performed for the solid breeder blanket using a DT neutron source at the FNS facility in JAEA. We have applied the blanket mockup composed of two enriched Li₂TiO₃ and three beryllium layers, and measured the detailed spatial distribution of the tritium production rate (TPR) using enriched Li₂CO₃ pellets. TPRs in the pellets have been measured by a liquid scintillation counter. Experiments have been done under a condition with a neutron reflector surrounding the DT neutron source. Numerical simulations have been performed using the MCNP-4C with the FENDL-2.0 and JENDL-3.3. The ranges of ratios of calculation results to experimental ones (C/Es) are 0.97–1.17 concerning with local TPR, and 1.04–1.09 for the integrated tritium production. It is found that the total integrated tritium production, which corresponds to tritium breeding ratio, can be predicted within uncertainty of 10% using the Monte Carlo calculation code and latest nuclear data libraries.     

Direct tritium measurement in lithium titanate for breeding blanket mock-up experiments with D-T neutrons

At Fusion Neutronics Source (FNS) of JAERI, tritium breeding experiments with blanket mock-ups consisting of advanced fusion reactor materials are in progress. The breeding zones are thin layers of lithium titanate which is one of the candidate tritium breeder materials for the DEMO fusion power reactor. It is anticipated that the application of small pellet-shaped lithium titanate detectors manufactured from the same material as the breeding layer will reduce experimental uncertainties arising from necessary corrections due to different isotopic lithium volume concentrations in breeding material and detector. Therefore, a method was developed to measure the local tritium production by means of lithium titanate pellet detectors and a liquid scintillation counting technique. The lithium titanate pellets were dissolved in concentrated hydrochloric acid solution and the resulting acidic solution was neutralized. Two ways of further processing were followed: direct incorporation into a liquid scintillation cocktail and distillation of the solution followed by mixing with liquid scintillator. Two types of lithium titanate pellets were investigated with different ⁶Li enrichment and manufacturing procedure. It was found that lithium titanate is suitable for tritium production measurements. However some discrepancies in the measurement accuracy remained with one of the investigated pellet detectors when compared with a well-established lithium carbonate measurement technique and this issue needs further investigation.     

托卡马克聚变中子源的数值模型研究

托卡马克(Tokamak)聚变装置中子学分析中,聚变中子源描述是重要的输入参数,其准确性直接影响分析结果的可靠性。通过分析ITER和欧洲聚变示范堆(EU DEMO)中子学分析中所采用的聚变中子源模型,提出了一种完整描述Tokamak中L-mode、H-mode等离子体的D-D、D-T聚变中子源的数值模型。在中国聚变工程实验堆(CFETR)的工程集成设计平台上,编写了基于蒙特卡罗算法的程序SCG求解该数值模型,实现了读取(零维)等离子体参数、输出可供典型中子学软件MCNP直接读取的中子源定义文件的功能。以CFETR氦冷球床包层的中子学分析模型为基准,在相同的L-mode等离子体D-T聚变工况下,相较于采用EU DEMO源子程序,采用本模型计算得到的中子壁负载差异最大值为2.02%,包层氚增殖率差异为0.18%,全堆能量增益因子的差异为0.23%。结果表明,本模型与其他源描述的差异较小,可应用于CFETR的中子学分析。     

中国聚变工程试验堆包层的核热耦合效应研究

本文以中国聚变工程试验堆（CFETR）的氦冷固态包层和水冷固态包层为研究对象,基于蒙特卡罗程序MCNP和计算流体力学程序FLUENT,利用3D-1D-2D耦合方法和伪材料方法,分别对200 MW的氦冷固态包层和水冷固态包层及1.5 GW的水冷固态包层方案进行了核热耦合计算分析。研究结果表明,金属铍的热散射效应和轻水密度是聚变包层核热耦合效应的主要来源,核热耦合效应对氦冷固态包层的影响可忽略,对水冷固态包层的氚增殖比和温度分布有一定程度的影响。     

Activation and Environmental Aspects of In-Vacuum Vessel Components of CFETR

The water-cooled ceramic breeder(WCCB) blanket is one of the three candidates of China's Fusion Engineering Test Reactor(CFETR). The evaluation of the radioactivity and decay heat produced by neutrons for the in-vacuum vessel components is essential for the assessment of radioactive wastes and the safety of CFETR. The activation calculation of CFETR in-vacuum vessel components was carried out by using the Monte Carlo N-Particle Transport Code MCNP, IAEA Fusion Evaluated Nuclear Data Library FENDL2.1, and the nuclear inventory code FISPACT-2007 and corresponding EAF-2007 libraries. In these analyses, the three-dimensional(3-D) neutronics model was employed and the WCCB blanket, the divertor, and the shield were modeled in detail to provide the detailed spatial distribution of the neutron flux and energy spectra. Then the neutron flux, energy spectra and the materials specification were transferred to FISPACT for the activation calculation with an assumed irradiation scenario of CFETR. This paper presents the main results of the activation analysis to evaluate the radioactivity, the decay heat, the contact dose, and the waste classification of the radioactive materials. At the time of shutdown, the activity of the WCCB blanket is 1.88×10~(19)Bq and the specific activity, the decay heat and the contact dose rate are 1.7×10~(13)Bq/kg, 3.05 MW, and 2.0×10~3Sv/h respectively. After cooling for 100 years, 79%(4166.4 tons) radioactive wastes produced from the blanket, divertor,high temperature shield(HTS) and low temperature shield(LTS) need near surface disposal, while21%(1112.3 tons) need geological disposal. According to results of the contact dose rate, all the components of the blanket, divertor, HTS and LTS could potentially be recycled after shutdown by using advanced remote handling equipment. In addition, the selection of Eurofer97 or RAFM for the divertor is better than that of SS316 because SS316 makes the activity of the divertor-body keep at a relatively high level.