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。     

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。     

聚变裂变混合发电堆水冷包层中子学设计分析

主要针对聚变裂变混合发电堆FDS-EM水冷包层的能量倍增因子M和氚增殖率TBR等中子学参数进行优化计算。FDS-EM包层主要设计目标是在氚自持的基础上获得约1 GW的电功率,并且尽可能长时间连续运行不换料。通过初步设计分析给出一个使用核废料(压水堆卸出的废料钚、锕系加上贫铀)作为裂变燃料,能够实现氚自持、能量倍增因子约为90等设计目标,且连续运行至少10年不换料的中子学方案。     

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。     

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。     

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.     

LiPb自冷混合堆包层中子学研究

研究了ＬｉＰｂ自冷托卡马克混合堆包层的中子学性能；第一壁材料和厚度对中子学性能的影响；Ｐｂ和Ｂｅ的中子增益性能以及包层中功率密度和２３９Ｐｕ的分布，并对中子学性能进行了优化。当聚变功率为２００ＭＷ，运行因子为０．３时，除氚自给外，每年可生产２３９Ｐｕ１３０ｋｇ。     

基于多维响应矩阵的中子能谱逆向调控方法及应用

中子能谱是影响核能系统安全性和经济性的重要参数,先进核能系统种类繁多,能谱差异大,准确的调控出先进核能系统的能谱对其发展有重要意义。本文利用基于响应矩阵的中子能谱逆向调控方法,以14MeV单能的聚变中子源为例,调控出聚变堆氚增殖包层、聚变裂变混合堆次临界包层、铅基快堆堆芯处的中子能谱,调控得到的中子能谱与目标能谱吻合较好,其中聚变堆氚增殖包层处的中子能谱与FNG上Mockup实验能谱比较,归一化能谱均方差降低了66%。对比结果表明本文方法能够实现多种类型先进核能系统中子能谱的精准调控。     

聚变-裂变混合堆外中子源效应

聚变-裂变混合堆(FFHR)作为聚变驱动次临界系统(FDS),具有良好的物理性能,能够实现产能、氚增殖、嬗变核废料等功能。采用COUPLE程序研究了水冷混合堆包层的铀水比和中子倍增剂对中子源效率的影响。结果表明:包层能谱越硬,外中子源效率越高;适当加入中子倍增剂Be可使外中子源效率增加。研究结果对进一步改进聚变-裂变混合堆的概念设计具有一定的指导意义。     

On the Tritium Breeding Capability of Flibe, Flinabe, and Li20Sn80 in a Fusion-Fission (Hybrid) Reactor

In a commercial (DT) driven fusion reactor, the tritium breeding ratio per incident fusion neutron must be greater than 1.05 to maintain tritium self-sufficiency for the driver. In this study tritium breeding capability of three different coolants, namely Flibe (LiF·BeF₂), Flinabe (LiF·NaF·BeF₂), and Li₂₀Sn₈₀ in a (DT) driven fusion-fission (hybrid) reactor was investigated for different refractory alloys (W-5Re, TZM, T111, and Nb-1Zr) as structural material. Neutron transport calculations were conducted with the help of SCALE 4.3 SYSTEM by solving the Boltzmann transport equation with code XSDRNPM. The contribution of Flibe, Flinabe, and Li₂₀Sn₈₀ with respect to ⁶Li enrichment in their lithium content to overall TBR was investigated. In addition, the effect of structural material type on TBR was examined.     

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.     

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。     

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。     

Monte Carlo calculation of radiation damage in first wall of an experimental hybrid reactor

This work was focused on the neutronic calculation of the nuclear parameters (neutron spectrum, displacement per atom (DPA), gas production, tritium breeding ratio (TBR), nuclear heating) for structural materials in the first wall (FW) and fuel clad (made of ferritic/martensitic steels, vanadium alloy, silicon carbide, copper alloy, and stainless steel) of an experimental hybrid reactor using the most current Monte Carlo Neutron-Particle Transport code MCNP5 1.4. Neutronic calculations were performed using a (DT) fusion driver hybrid reactor under a neutron wall loud of 2.25 MW/m² by full reactor power for one year. Obtained results were compared with three different data libraries (ENDF/B-V, ENDF/B-VI and CLAW-IV). TBR values in the reactor blanket for all investigated materials became greater than the minimum requirement (TBR > 1.05). Nuclear parameters like DPA, He-production and nuclear heating were considered as radiation damage limits for structural materials, copper alloy (Cu0.5Cr0.3Zr) showed better performance than all investigated materials.     

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。     

Neutronic analysis of an inertial fusion energy breeder with SiCf/SiC

Neutronic performance of a blanket driven ICF (Inertial confinement fusion) neutron based on SiC_f/SiC composite material is investigated for fissile fuel breeding. The investigated blanket is fueled with ThO₂ and cooled with natural lithium or (LiF)₂BeF₂ or Li₁₇Pb₈₃ or ⁴He coolant. MCNP4B Code is used for calculations of neutronic data per DT neutron. Calculations have show that values of TBR (tritium breeding ratio) being one of the main neutronic paremeters of fusion reactors are greater than 1.05 in all type of coolant, and the breeder hybrid reactor is self-sufficient in the tritium required for the DT fusion driver. Calculations show that natural lithium coolant blanket has the highest TBR (1.298) and M (fusion energy multiplication) (2.235), Li₁₇Pb₈₃ coolant blanket has the highest FFBR (fissile fuel breeding ratio) (0.3489) and NNM (net neutron multiplication) (1.6337). ⁴He coolant blanket has also the best Γ (peek-to-average fission power density ratio) (1.711). Values of neutron leakage out of the blanket in all type of coolants are quite low due to SiC reflector and B₄C shielding.     

Albedo calculations for candidate fusion reactor materials used in the inboard side of a compact tokamak reactor and the effects of using such materials on the tritium breeding

In this paper the total neutron albedo and associated energy distributions for 10 candidate fusion reactor materials have been calculated. The angular distributions of reflected neutrons for monodirectional 14.1 MeV neutrons incident on slabs of Pb, Be, and W are presented and the dependence of albedo on neutron energy and incident angle has been investigated. Finally, the impact on the tritium breeding of the outboard blanket of the choice of material used in the inboard side of the reactor has been assessed. It has been found that the largest total neutron albedos are those of neutron multiplying materials, whilst among non-multiplying materials tungsten yields the largest albedo and B₁₀H₁₄ yields the lowest. Tritium breeding ratio (TBR) calculations have shown the inadequacy of the neutron albedo concept in predicting the impact of inboard materials on the TBR of the reactor.