混合堆增殖乏燃料组件中子学特性初步研究

本文主要对聚变-裂变混合堆增殖乏燃料在压水堆组件中使用的可能性进行了初步研究。根据聚变 裂变混合堆增殖乏燃料的特点,给出了的聚变-裂变混合堆增殖乏燃料压水堆组件设计方案,分析组件的燃料温度系数、慢化剂温度系数等参数。结果表明：聚变 裂变混合堆乏燃料组件的特性与全铀组件的特性相似。在相同的易裂变同位素质量百分比情况下,本文给出的组件设计方案的功率不均匀系数更小。研究结果可为未来实现聚变 裂变混合堆和压水堆联合循环系统提供技术支持。     

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。     

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。     

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。     

聚变-裂变混合堆的氚工艺问题

文章描述了聚变堆和聚变-裂变混合堆的氚工艺问题。根据聚变堆和聚变-裂变混合堆的特点讨论了对包层氚增殖材料的要求,列举了几种可作氚增殖的合理材料特性。给出了几种从包层提取氚和从废聚变燃料中回收氚的方法。最后对混合堆的氚安全及防护问题进行了讨论。     

聚变裂变混合堆钍基增殖锕系元素嬗变包层中子学初步研究

聚变裂变混合堆比纯聚变堆在工程及技术方面要求低,且在产生核燃料、嬗变长寿命核废料以及固有安全性方面具有一定优势,因此,越来越受到人们的重视。增殖包层是混合堆系统的关键部件,已有的包层研究基本上是基于较成熟的铀-钚燃料循环技术。针对我国铀资源相对较少而钍资源较丰富的现状,本文就一种新型的钍基燃料增殖锕系元素嬗变包层进行了初步的中子学研究,利用一维离散纵标法燃耗程序BISONC以及Monte-Carlo粒子输运程序MCNP,对包层的关键核参数,诸如氚增殖比、少量锕系元素的嬗变质量、233U产量以及热功率等,进行了较详细的计算分析。计算结果表明,生成的核燃料233U的富集度可达到3.65%,从而满足压水堆燃料富集度要求。分析结果为下一步的包层优化设计提供了依据。     

聚变-裂变混合能源堆一维计算模型燃耗分析

聚变-裂变混合能源堆包括聚变中子源和以天然铀为燃料、水为冷却剂的次临界包层,主要目标是生产电力。利用输运燃耗耦合程序系统MCORGS计算了混合能源堆一维模型的燃耗,给出了中子有效增殖因数k_eff、能量放大倍数M、氚增殖比TBR等物理量随时间的变化。通过分析能谱和重要核素随燃耗时间的变化,说明混合能源堆与核燃料增殖、核废料嬗变混合堆的不同特点。本文给出的结果可作为混合堆中子输运、燃耗分析程序校验的参考数据,为混合堆概念研究提供了基础数据。     

聚变裂变混合堆燃耗计算及燃料成本分析

基于轻水冷却的压力管式混合堆,采用压水堆卸载的乏燃料以及天然铀氧化物陶瓷燃料,建立混合堆包层的换料方案,详细计算了包层中子学性能随燃耗的变化情况,计算结果表明,包层在维持3000 MW热功率输出的同时,可以保证氚自持(氚增殖比TBR>1.20),而每5 a仅需向包层添加80 t左右的重金属燃料。基于建立的平衡循环计算了包层采用不同燃料时的单位发电燃料成本。结果表明,采用乏燃料和天然铀时的单位发电燃料成本分别为1.82×10-3、1.35×10-3$/(k W·h)。     

核能源前景浅见

文章展望了裂变堆、纯聚变堆和聚变-裂变混合堆的前景,分析了混合堆的低聚变条件和很高的能量与燃料增殖能力等重大优点。认为作为由裂变能源过渡到纯聚变能源的桥梁,聚变-裂变混合堆应成为未来核能源的方向之一。     

球床包层聚变—裂变混合堆热工安全分析

球床包层混合堆与板状元件包层混合堆相比较，前者在核燃料生产和安全方面可能具有更多的优越性。本应用ＴＨＥＲＭＩＸ程序和辅助程序对我国开发的托卡马克堆芯氮气冷却球床包层聚变－裂变合堆的包层进行了热工计算。计算中考虑了不同的燃料球材料及稳态，卸压和断流事故工况。计算结果表明，只要选用合适的燃料球材料和设置适当的控制保护系统，具有快速卸料罐的托卡马克堆芯氦气包层聚变－裂变混合堆的概念设计在安全上的可行的。     

Gas-cooled fusion-fission hybrid reactor systems

In this paper, the concept of the fusion-fission hybrid reactor is reviewed, and a system of classification for hybrid blanket designs is suggested. The advantages and disadvantages of gas cooling for hybrid reactor systems are discussed and the design implications of using gas cooling in a hybrid blanket are presented. Five of the more complete gas-cooled hybrid reactor conceptual design studies are discussed, and the fission-suppressed hybrid blanket concept is identified as offering potentially significant advantages in terms of inherent safety features and reduced technology development requirements compared to higher power fission blankets. It is concluded that helium is attractive as the coolant for hybrid reactor systems, and that technically viable reactor designs have been developed using helium cooling. The helium-cooled fission-suppressed hybrid blanket, based on thorium fuel for production of²³³U, is identified as being a particularly attractive candidate for further hybrid reactor development work.     

Design and analysis on tritium system of multi-functional experimental fusion–fission hybrid reactor (FDS-MFX)

As early application of fusion technology, the fusion–fission hybrid systems/reactors could be used to transmute long-lived radioactive waste and produce fissile nuclear fuel. A fusion–fission hybrid reactor named FDS-MFX was designated for checking and validating the DEMO reactor blanket relevant technologies. The reactor design is based on easy-achieved plasma parameters extrapolated from the successful operation of tokamaks and the subcritical blanket is designed based on the well-developed technologies of fission reactors. In this contribution, a concept of the tritium system was designed for the FDS-MFX: the tritium was extracted from LiPb by the helium purge gas which contains a small amount of hydrogen gas, then the impurity gas was removed by cold trap, finally tritium was separated from hydrogen isotope by the cryogenic distillation and supply to reactor core. On the basis of data obtained by present design and experimental research, the system parameters were presented and discussed in detail. The results preliminarily demonstrated the engineering feasibility of the design.     

聚变-裂变混合堆冷管段大破口失水事故分析

将非能动堆芯冷却系统（PXS）应用于聚变-裂变混合堆,使用RELAP5对混合堆一回路、部分二回路和PXS进行了建模,对冷管段双端剪切断裂大破口失水事故进行了瞬态计算和分析研究。计算结果显示：破口发生后出现两次燃料温度峰值,均发生在外包层,第1次峰值温度发生在约11 s,为938.2 K;第2次峰值温度发生在约50 s,为608.7 K。两次燃料温度峰值均低于燃料U-10Zr的熔点,在可接受范围内。随着瞬态过程的深入,安注箱、堆芯补水箱及安全壳内储水箱的冷却水开始注入包层,使内外包层的坍塌液位开始回升,最终重新淹没堆芯。表明PXS在冷管段双端剪切断裂大破口失水事故下能保证混合堆堆芯的安全,将其应用于聚变-裂变混合堆是可行的。     

Study on MA and FP transmutation in fast reactors

Feasibility studies have been performed to develop an optimized fast reactor core for reducing long-term radiotoxicity of nuclear waste by minor actinide(MA) and long-lived fission product(FP) transmutation, taking into consideration fuel cycle technology. Systematic parameter survey calculations were implemented to investigate the basic characteristics of MA and FP transmutation in a fast reactor core. The hybrid MA-loading method, where Np nuclide is dispersed uniformly in the core and target subassemblies containing Am, Cm and rare earth nuclides are loaded into the blanket region, has the potential to achieve the maximum transmutation of MA with no special fuel design considerations. The introduction of target subassemblies using duplex pellets - a moderator annulus surrounding a ⁹⁹Tc core - has a great potential to transmute long-lived fission products in the radial blanket region of the fast reactor core.     

次临界能源堆安全性研究

次临界能源堆由中心的托卡马克装置和围绕其的裂变包层组成。本文根据物理和热工专业分析计算得出的一种针对其裂变包层的燃料和冷却剂通道布置方式,分析设计的包层结构安全性和工程应用中的安全性。包层结构安全性分析使用CFD方法,计算了正常运行工况和冷却剂通道堵管的情况,得到堵管发生后包层的局部状况。通过RELAP程序模拟了裂变包层参与核电厂发电运行过程中,其本身所具有的固有安全性。本文通过计算发现了其安全上的薄弱环节,并提出了改进措施,为以后改进次临界能源堆安全性提供参考。     

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.     

Feasibility study of applying the passive safety system concept to fusion–fission hybrid reactor

The fusion–fission hybrid reactor can produce energy, breed nuclear fuel, and handle the nuclear waste, etc., with the fusion neutron source striking the subcritical blanket. The passive safety system consists of passive residual heat removal system, passive safety injection system and automatic depressurization system was adopted into the fusion–fission hybrid reactor in this paper. Modeling and nodalization of primary loop, partial secondary loop and passive core cooling system for the fusion–fission hybrid reactor using relap5 were conducted and small break LOCA on cold leg was analyzed. The results of key transient parameters indicated that the actuation of passive safety system could mitigate the accidental consequence of the 4-inch cold leg small break LOCA on cold leg in the early time effectively. It is feasible to apply the passive safety system concept to fusion–fission hybrid reactor. The minimum collapsed liquid level had great increase if doubling the volume of CMTs to increase its coolant injection and had no increase if doubling the volume of ACCs.