氦气、水和熔盐冷却的混合堆中子学性能研究

氦气、水、熔盐(Flibe)在强磁场中流动不存在严重的MHD问题,因此适合在基于磁约束的聚变-裂变混合堆中作为冷却剂.针对氦气、水、Flibe这3种冷却剂对混合堆包层中子学性能的影响进行研究,分析包层中能谱特点及燃料增殖特性.通过燃耗计算,研究氚增殖率(TBR)、能量倍增因子(M)、keff等随运行时间的变化.中子学输运采用三维蒙特卡罗程序MCNP.计算结果表明,不同的冷却剂对混合堆系统中子能谱影响很大:氦冷系统的能谱最硬,主要发生快中子裂变,氚增殖效果最好;水冷系统的能谱最软,产能最多,但需提高TBR;Flibe冷系统的能谱较硬,产能最少.     

激光惯性约束聚变裂变混合能源包层中子学数值模拟

对三维输运与燃耗耦合程序MCORGS进行了适应性改造,并对利弗莫尔实验室提出的激光惯性约束聚变裂变混合能源(LIFE)概念进行了分析和改进。输运计算采用MCNP程序,燃耗计算采用ORIGENS程序,增加氚控制模块和功率控制模块。建立了与LIFE等价的以贫化铀为燃料、Be为中子增殖剂的包层方案,通过数值模拟验证了MCORGS程序的可靠性。针对Be资源短缺及冷却复杂问题,设计了以贫化铀为燃料、Pb为中子增殖剂的包层方案,包层能量放大了4倍,可在55a内稳定输出2 000 MWt功率。     

聚变-裂变混合能源堆球模型参数敏感性分析

在聚变-裂变混合能源堆球模型基础上,使用蒙特卡罗方法中子学程序对中子源、铀水体积比、产氚区等相关参数进行了中子学的敏感性计算。分析了各参数对混合能源堆能量放大倍数M和氚增殖比TBR的影响,并总结其基本规律,为开展进一步的混合能源堆概念设计提供了重要参考。     

混合能源堆多群中子参数库研制

根据聚变-裂变混合能源装置计算的需求,利用微观评价核数据库ENDF/B-VI.8,研制了包括100多个核素的187群中子参数,用于中子输运方程的计算。通过选取合适的能群结构和权重谱,并考虑温度、热散射以及共振自屏效应等的影响,建立并拓展了适合混合堆研究需要的多群参数库。为了检验参数库中数据的适用性,采用一维中子和光子输运程序ANISN对一组基准装置进行了临界计算。结果表明参数库可用于混合能源堆设计计算。     

聚变—裂变混合能源堆燃料管理方案设计研究

本文根据聚变-裂变混合能源堆方案设计和燃料组件功率分布的特点,利用自主开发的蒙卡-燃耗耦合程序,开展了详细的燃料管理方案设计研究,分别设计了整体后处理的燃料管理方案、双循环燃料管理方案以及分批燃料管理方案,针对这些类型的燃料管理方案,进行了燃耗分析计算,研究了各种燃料管理方案下各区燃耗及主要裂变核素成分随燃耗的变化。根据各燃料管理方案的主要特点和计算分析结果,对比总结了它们的优点和缺点。本文为今后的聚变-裂变混合能源堆提供了燃料管理上的建议,也为进一步的经济性分析优化研究打下了基础。     

MCORGS及次临界能源堆包层概念研究进展

次临界能源堆是以能源供应为目的的一种聚变裂变混合堆,以聚变驱动,天然铀为裂变燃料,轻水为冷却剂。本文针对该混合堆开发了基于MCNP与ORIGENS的三维中子输运燃耗耦合程序MCORGS,分析了包层三维中子学模型。提出简化干法后处理,设想利用衰变热将乏燃料加热到2 100K,将沸点低于该温度的裂变产物挥发去除。计算了包层各区材料每年发生的原子移位数,建议采用10a左右的换料周期,乏燃料经后处理后可多次复用。第1个寿期内氚增殖比TBR平均约1.15,包层能量放大倍数M平均约12;第2~9个寿期内TBR平均约1.35,M平均约18。利用流体动力学程序完成了包层CAD模型建立、网格划分及稳态传热计算分析,各区材料的最高温度均低于许用温度并有较大裕量。     

利用D-T聚变中子源转化长寿命裂变产物的中子学可行性研究

从中子学角度研究长寿命裂变产物在Tokamak型D-T聚变堆包层中转化的可行性.提出了用可裂变Pu增殖中子的混合包层转化方案,研制了相应的燃耗计算程序及数据库,并对所提方案进行了计算和分析.结果表明,在可预见的聚变堆芯技术条件下,所研究的概念性包层可对长寿命裂变产物进行有效转化.     

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。     

铀-水体积比对混合能源堆中子学性能的影响

介绍输运燃耗耦合程序MCORGS的理论模型,利用MCORGS研究铀-水体积比对混合能源堆中子学性能的影响。研究表明,采用天然铀为裂变燃料,且铀-水比为2:1时,可实现较高的能量放大,保持氚自持,中子学性能可以维持100 a以上;采用压水堆乏燃料时,铀-水比的选择余地更大,能量放大和产氚能力提高,但燃料增殖能力下降。     

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。     

The neutronics studies of a fusion fission hybrid reactor using pressure tube blankets

In this paper, a fusion fission hybrid reactor used for energy producing is proposed based on the situation of nuclear power in China. The pressurized light water is applied as the coolant. The fuel assemblies are loaded in the pressure tubes with a modular type structure. The neutronics analysis is performed to get the suitable design and prove the feasibility. The energy multiplication and tritium self-sustaining are evaluated. The neutron load is also cared. From different candidates, the PWR spent fuel is selected as the feed fuel. The results show that the hybrid reactor can meet the expected reactor core lifetime of 5 years with 1000 MWe power output. Two ways are discussed including burning the discharged PWR spent fuel and burning the reprocessed plutonium. The energy multiplication is big enough and the tritium can be self-sustaining for both of the two ways. The neutron wall load in the operating time is kept smaller than the one of ITER. The way to use the reprocessed plutonium brings low neutron wall load, but also brings additional difficulties in operating the hybrid reactor. The way to use the discharged spent fuel is proposed to be a better choice currently.     

Preliminary design of hybrid energy reactor and integral neutron experiments

We propose a preliminary design for a fusion–fission hybrid energy reactor (FFHER), based on current fusion science and technology (with some extrapolations forward from ITER) and well-developed fission technology. We list design rules and put forward a primary concept blanket, with uranium alloy as fuel and water as coolant. The uranium fuel can be natural uranium, LWR spent fuel, or depleted uranium. The FFHER design can increase the utilization rate of uranium in a comparatively simple way to sustain the development of nuclear energy. We study the interaction between the fusion neutron and the uranium fuel with the aim of to achieving greater energy multiplication and tritium sustainability. We also review other concept hybrid reactor designs. We design integral neutron experiments in order to verify the credibility of our proposed physical design. The combination of this program of research with the related thermal hydraulic design, alloy fuel manufacture, and nuclear fuel cycle programs provides the science and technology foundation for the future development of the FFHER concept in China.     

A fusion–fission hybrid reactor with water-cooled pressure tube blanket for energy production

A fusion–fission hybrid reactor is proposed to achieve the energy gain of 3000 MW thermal power with self-sustaining tritium. The hybrid reactor is designed based on the plasma conditions and configurations of ITER, as well as the well-developed pressurized light water cooling technologies. For the sake of safety, the pressure tube bundles are employed to protect the first wall from the high pressure of coolant. The spent nuclear fuel discharged from 33GWD/tU Light Water Reactors (LWRs) and natural uranium oxide are taken as driver fuel for energy multiplication. According to thermo-mechanics calculation results, the first wall of 20 mm is safe. The radiation damage analysis indicates that the first wall has a lifetime of more than five years. Neutronics calculations show that the proposed hybrid reactor has high energy multiplication factor, tritium breeding ratio and power density; the fuel cannot reach the level of plutonium required for a nuclear weapon. Thermal-hydraulic analysis indicates that the temperatures of the fuel zone are well below the limited values and a large safety margin is provided.     

聚变-裂变混合堆水冷包层中子物理性能研究

研究直接应用国际热核聚变实验堆（ITER）规模的聚变堆作为中子驱动源,采用天然铀为初装核燃料,并采用现有压水堆核电厂成熟的轻水慢化和冷却技术,设计聚变-裂变混合堆裂变及产氚包层的技术可行性。应用MCNP与Origen2相耦合的程序进行计算分析,研究不同核燃料对包层有效增殖系数、氚增殖比、能量放大系数和外中子源效率等中子物理性能的影响。计算分析结果显示,现有核电厂广泛使用的UO₂核燃料以及下一代裂变堆推荐采用的UC、UN和U₉₀Zr₁₀等高性能陶瓷及合金核燃料作为水冷包层的核燃料,都能满足以产能发电为设计目标的新型聚变 裂变混合堆能量放大倍数的设计要求,但只有UC和U₉₀Zr₁₀燃料同时满足聚变燃料氚的生产与消耗自持的要求。研究结果对进一步研发满足未来核能可持续发展的新型聚变-裂变混合堆技术具有潜在参考价值。     

Utilization of TRISO Fuel with LWR Spent Fuel in Fusion-Fission Hybrid Reactor System

HTRs use a high performance particulate TRISO fuel with ceramic multi-layer coatings due to the high burn up capability and very neutronic performance. TRISO fuel because of capable of high burn up and very neutronic performance is conducted in a D-T fusion driven hybrid reactor. In this study, TRISO fuels particles are imbedded body-centered cubic (BCC) in a graphite matrix with a volume fraction of 68%. The neutronic effect of TRISO coated LWR spent fuel in the fuel rod used hybrid reactor on the fuel performance has been investigated for Flibe, Flinabe and Li₂₀Sn₈₀ coolants. The reactor operation time with the different first neutron wall loads is 24 months. Neutron transport calculations are evaluated by using XSDRNPM/SCALE 5 codes with 238 group cross section library. The effect of TRISO coated LWR spent fuel in the fuel rod used hybrid reactor on tritium breeding (TBR), energy multiplication (M), fissile fuel breeding, average burn up values are comparatively investigated. It is shown that the high burn up can be achieved with TRISO fuel in the hybrid reactor.     

聚变-裂变混合能源堆球模型中子学对算研究

利用蒙特卡罗程序和自主开发的蒙特卡罗-燃耗耦合程序MOCouple-s,对北京应用物理与计算数学研究所提出的聚变-裂变混合能源堆球模型进行了对算研究。对初始时刻及各燃耗时刻下的有效增殖因数、能量倍增因子、氚增殖比、中子源强度等堆芯参数进行了比较,结果总体符合较好。对寿期末重要核素的成分进行了详细比较,除个别核素外,偏差很小,表明所采用的计算程序与核参数库一致性良好。对核参数库的选择、铀水体积比等对燃耗计算结果的影响进行敏感性分析,并对外中子源驱动的次临界堆芯的燃耗计算进行详细讨论,提出可行的燃耗计算基准。     

Optimization of flux distribution in a molten-salt reactor with a 2-region core for plutonium burning

Molten-salt reactors (MSRs) are selected as one of the candidates of Generation IV reactor concepts. In GLOBAL2005 held in Tsukuba, Japan, one paper discussed the flattening of fast neutron flux in the core for a longer life of graphite moderator. In the paper a 3-region reactor concept was presented. The authors tried many cores changing configurations such as volume of each region and fractions of fuel salt in the regions or fuel compositions.

We investigated the other possibility of a 2-region core for the simplicity. Using one energy group neutron diffusion theory and considering extrapolation distance, the optimum selection of region wise neutron multiplication factors can be theoretically and easily obtained. In MSRs, there is no burnup distribution of the fuel. The region wise neutron multiplications can be obtained by adjusting the volume fraction of fuel in a cell with a given composition of the fuel salt. Using the theoretical results, the optimization of the actual core configuration was determined by a nuclear analysis code SRAC2002 with the nuclear data library of JENDL3.3.

In this paper, we considered MSRs using plutonium as a fissile material. Ordinary MSRs use uranium-233, which doesn't exist naturally, and utilizing plutonium is easier to startup.     

次临界能源堆物理性能初步分析

次临界能源堆（SER）是由托卡马克聚变源驱动的聚变裂变混合堆。SER以天然铀为燃料、水为冷却剂,主要目标是生产电能。本工作建立了次临界能源堆环形圆柱模型,利用蒙特卡罗输运和燃耗计算程序,比较了燃料区不同构型对k_eff、M、TBR和燃料增殖比等参数的影响,针对均匀模型进行中子源效率与聚变源强、功率分布与能谱、初步燃耗、寿期末停堆衰变热和卸载燃料放射性等物理性能分析。计算结果表明,该模型能满足能量倍增大于6、氚自持、较长时间不换料等设计目标。研究结果为下一步开展SER安全分析提供了基础。     

Reprocessing free nuclear fuel production via fusion fission hybrids

Fusion fission hybrids, driven by a copious source of fusion neutrons can open qualitatively “new” cycles for transmuting nuclear fertile material into fissile fuel. A totally reprocessing-free (ReFree) Th²³²–U²³³ conversion fuel cycle is presented. Virgin fertile fuel rods are exposed to neutrons in the hybrid, and burned in a traditional light water reactor, without ever violating the integrity of the fuel rods. Throughout this cycle (during breeding in the hybrid, transport, as well as burning of the fissile fuel in a water reactor) the fissile fuel remains a part of a bulky, countable, ThO₂ matrix in cladding, protected by the radiation field of all fission products. This highly proliferation-resistant mode of fuel production, as distinct from a reprocessing dominated path via fast breeder reactors (FBR), can bring great acceptability to the enterprise of nuclear fuel production, and insure that scarcity of naturally available U²³⁵ fuel does not throttle expansion of nuclear energy. It also provides a reprocessing free path to energy security for many countries. Ideas and innovations responsible for the creation of a high intensity neutron source are also presented.