PWR（U）乏燃料中超铀元素在混合堆快裂变包层内嬗变研究

从中子学角度对ＰＷＲ（Ｕ）乏燃料中的超铀元素（２３８Ｐｕ，２３９Ｐｕ，２４１Ｐｕ，２４１Ａｍ，２４３Ａｍ，２３７Ｎｐ，２４４Ｃｍ）在聚变－裂变混合堆快裂变包层内嬗变的可行性进了研究。利用一维中子输运和燃耗计算程序ＢＩＤＥＣＡＹ译不同燃料组分的四个快裂变包层进行分析计算。结果表明，在聚变－裂变混合堆快裂变包层内安全，高效地嬗变ＰＷＲ（Ｕ）乏燃料中的超铀元素是可能的。     

CFETR水冷陶瓷增殖剂包层模块产氚率实验验证

水冷陶瓷增殖剂（WCCB）包层作为中国聚变工程试验堆（CFETR）候选包层之一，承担着氚增殖、核热提取、屏蔽等重要涉核功能，其中子学设计的可靠性直接影响CFETR氚自持目标的实现。为验证中子学设计工具，即MCNP和FNEDL3.0数据库，在WCCB包层中子学设计中的可靠性，基于研制出的WCCB包层模块，在DT中子环境下开展中子学实验，对以产氚率（TPR）为代表的中子学参数进行了模拟值（C）和实验值（E）对比分析。结果表明，模块中轴线位置处TPR的C/E为0.97?1.08，而模块边缘位置处TPR的C/E为0.65?0.82；模块钛酸锂层边缘区197Au（n，γ）198Au反应率的C/E为0.72?0.90，表明模块边缘区存在非期望的散射中子，导致该区TPR模拟值和实验值偏离较大。     

聚变增殖堆中子学设计

为了在满足增殖堆自身氚需要的前提下,提高堆性能参数——支持比,本文利用一维ANISN输运程序,对直接浓缩抑制裂变包层的中子学性能作了优化计算,研究了~6Li丰度和U-233浓度及其分布对包层中子学性能的影响,提出了改进包层设计的几种措施,得到了满意的结果。在堆运行周期内,平均产氚率T可达到1.11,支持比明显提高,达到14,包层中功率密度分布均匀,使堆的安全、冷却问题容易解决,给堆的结构设计带来方便。     

聚变增殖堆中子学设计

为了在满足增殖堆自身氚需要的前提下,提高准性能参数——支持比,本文利用一维ANISN输运程序,对直接浓缩抑制裂变包层的中子学性能作了优化计算,研究了~6Li丰度和U-233浓度及其分布对包层中子学性能的影响,提出了改进包层设计的几种措施,得到了满意的结果。在堆运行周期内,平均产氚率T可达到1.11,支持比明显提高,达到14,包层中功率密度分布均匀,使堆的安全、冷却问题容易解决,给堆的结构设计带来方便。     

裂变率测量中的杂质影响及基消除方法

核裂变法是通过测量中子进行裂变率测量的重要方法。常用于炽子测量的裂变室有＾２３５Ｕ裂变室和＾２３９Ｐｕ裂变室，快中子测量可以用＾２３８Ｕ、＾２３２Ｔｈ和＾２３７Ｎｐ等裂变室。通常用于裂变室的可裂变核素是采用同位素分离方法或人工方法得到的，其中含有少量其他核素杂质。     

用超铀核素复制骨肉瘤模型

本实验用７４８只ｗｉｓｔａｒ大鼠进行超铀核素２４１Ａｍ、２３９Ｐｕ、２３８Ｐｕ和２３７Ｎｐ复制肿瘤模型的研究。动物分成（１）２４１Ａｍ中毒组，２０８只，中毒剂量为０．７５×１０～３１．４５×１０４Ｂｑ·ｋｇ－１；（２）２３９Ｐｕ中毒组，１４５只，中毒...     

聚变堆固态产氚包层中子学优化方法研究与应用

为提升聚变堆包层产氚性能,更好地满足氚自持要求,首先,基于中子微扰理论与模拟退火算法开发了适用于聚变堆产氚包层（TBB）中子学优化新算法与新程序。其次,选取中国聚变工程实验堆（CFETR）氦冷固态包层,完成了全堆中子学性能优化的示范性应用。最后,对优化后的包层方案进行了热工、流体、结构的三维有限元校核。结果表明：(1)相比于传统包层中子学优化算法,本文所提出的优化算法具有更好的优化效果与更高的优化效率;(2)本文所开发的智能优化程序可更好地满足聚变堆TBB中子学优化与设计的需求,可为包层设计提供算法理论基础与程序支撑。     

聚变实验增殖堆He冷包层中子学设计研究

在一维计算的基础上，优化分析聚变实验增殖堆Ｈｅ气冷却包层设计参数对堆中子学性能的影响，给出了年产生１００ｋｇ钚、氚自持、安全性好的包层初步设计方案，并用ＭｏｎｔｅＣａｒｌｏ输运程序ＭＣＮＰ３Ｂ对此方案进行了三维中子学计算校核。     

裂变率测量中的杂质影响及其消除方法

核裂变法是通过测量中子进行裂变率测量的重要方法．常用于热中子测量的裂变室有２３５Ｕ裂变室和２３９Ｐｕ裂变室，快中子测量可以用２３８Ｕ、２３２Ｔｈ和２３７Ｎｐ等裂变室．通常用于裂变室的可裂变核素是采用同位素分离方法或人工方法得到的，其中含有少量其他核素杂质．实验测量表明，少量能发生热裂变的杂质对快中子的测量有很大影响。利用热裂变修正方法和裂变室包镉方法可以消除这种影响。     

在聚变堆中嬗变~（237）Np的研究

研究了在聚变堆中嬗变长寿命的锕系元素￣（２３７）Ｎｐ，以及转换￣（２３７）Ｎｐ成为可裂变燃料￣（２３９）ｐｕ的物理可行性。探讨了在嬗变包层中￣（２３７）Ｎｐ的浓度、￣（２３９）ｐｕ的中于增殖率、中子壁负载的变化以及嬗变区厚度与￣（２３７）Ｎｐ嬗变率的关系。给出的研究计算结果表明，在１个聚变功率为２００ＭＷ，中子壁负载为１．０ＭＷ／ｍ２的聚变堆包层中，实现年嬗变￣（２３７）Ｎｐ约３．５ｔ，年平均产钚量约２０ｋｇ是可行的。     

Effect of TRU fuel loading on core performance and plutonium production of FBR

The effect of trans-uranium (TRU) fuel loading on the reactor core performances as well as the actinide and isotopic plutonium compositions in the core and blanket regions has been analyzed based on the large FBR type. Isotopic plutonium composition of TRU fuel is less than that of MOX fuel except for Pu-238 composition which obtains relatively higher composition. A significant increase of plutonium vector composition is shown by Pu-238 for TRU fuel in the core region as well as its increasing value in the blanket region for doping MA case. Excess reactivity can be reduced significantly (5% at beginning of cycle) and an additional breeding gain can be obtained by TRU fuel in comparison with MOX fuel. Doping MA in the blanket regions reduces the criticality for a small reduction value (0.1%) and it gives a reduction value of breeding ratio. Loading MA in the core regions as TRU fuel composition gives relatively bigger effect to increase the void reactivity coefficient mean while it gives less effect for loading MA in the blanket regions. Similar to the void reactivity coefficient profile, loading MA is more effective to the change of Doppler coefficient in the core regions in comparison with loading MA in the blanket regions which gives slightly less negative Doppler coefficient. Obtained Pu-240 vector compositions in the core region are categorized as practically unusable composition for nuclear device based on the Pellaud's criterion. Less than 7% Pu-240 vector compositions in the blanket region are categorized as weapon grade composition for no doping MA case. Obtaining 9% of Pu-238 composition by doping MA 2% in the blanket regions is enough to increase the level of proliferation resistance for denaturing plutonium based on the Kessler's criterion.     

Effect of Nuclear Data Libraries on Tritium Breeding in a (D–T) Fusion Driven Reactor

In design a Deuterium–Tritium (D–T) fusion driven hybrid reactor, neutronics and nuclear data libraries have an essential role for reliable neutronics calculations. Therefore, nuclear data libraries are very important to calculate of the neutronic parameters and selection of tritium breeder materials to be used in the blanket. In this study tritium breeding performances of candidate tritium breeding materials, namely, Li₂O, LiH, Li₂TiO₃, Li₂ZrO₃ and Li₄SiO₄ in a (D–T) driven fusion–fission (hybrid) reactor is investigated based on three dimensional (3-D) and one dimensional (1-D) neutronic calculations. 3-D and 1-D neutron transport calculations are performed with Monte Carlo transport code (MCNP 4C), SCALE 5 and ANISN nuclear data codes to determine the tritium breeding ratio (TBR) of the blanket. The effects of different nuclear data libraries on TBR are examined and TBR calculation results are comparatively investigated.     

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。     

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.     

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。     

Neutronics studies on the feasibility of developing fast breeder reactor with flexible breeding ratio

This paper investigates the feasibility of designing a flexible fast breeder reactor from the view of neutronics. It requires that the variable breeding ratio can be achieved in operating a fast reactor without significant changes of the core, including the minimum change of fuel assembly design, the minimum change of the core configuration and the same control system arrangement in the core. The sodium cooled fast reactor is investigated. Two difficulties are overcome: (1) the different excess reactivity is well controlled for different cores, especially for the one with small breeding ratio; (2) the maximum linear power density is well controlled while the breeding ratio changes. The optimizations are done to meet the requirements. The U–Pu–Zr alloy is applied to enhance the breeding. The enrichment-zoning technique with unfixed blanket assembly loading position is searched to get acceptable power distributions when the breeding ratio changes. And the control system is designed redundantly to fulfill the control needs. Then, the achieved breeding ratio can be adjusted from 1.1 to 1.4. The reactivity coefficients, temperature distributions and preliminary safety performances are evaluated to investigate the feasibility of the new concept. All the results show that it is feasible to develop the fast reactor with flexible breeding ratios, although it still highly relies on the advancement of the coolant flow control technology.     

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。     

Impact Analysis of Helium Cooled Solid Blanket Structures on the Tritium Breeding Performance in CFETR

Chinese Fusion Engineering Test Reactor (CFETR) is a test tokamak reactor to bridge the gap between ITER and future fusion power plant. As its objectives are to demonstrate generation of fusion power and to realize tritium self-sufficiency, the tritium breeding ratio (TBR) is a key design parameter. In the blanket design and optimization, the structures such as the first wall (FW), cooling plate (CP), stiffening plate (SP), cap and some other design parameters in detailed 3-D model have significant impacts on the tritium breeding performance. Based on a helium cooled solid breeder blanket option for CFETR, the impact analysis of the helium cooled solid blanket structures on tritium breeding performance was performed in this paper. Firstly, the detailed 3D neutronics model was built by using of a CAD to Monte Carlo Geometry conversion tool McCad. Then based on the detailed 3D neutronics model, the impact analyses of the blanket structures on tritium breeding performance were carried out, which include the FW, CP, SP, cap and side wall. By the sensitivity study of the blanket structures on the TBR, it gave the TBR variation trend and references for the blanket design and optimization.     

托卡马克工程试验增殖堆Li冷包层的磁流体动力学压降分析

本文对液态金属 Li 流过托卡马克工程试验增殖堆自冷包层的磁流体动力学(MHD)压降进行了分析,讨论了内侧包层有无裂变、燃料元件的形式、包层能量倍增因子 M 及第一壁冷却孔道宽度对包层总压降的影响,从 MHD 流动分析的观点,为中子学、结构和热工水力设计提出了设计要求。