Core neutronic characterization of a large molten-salt cooled thorium-based solid fuel fast reactor

Thorium is three to four times as abundant as uranium, providing a potentially large, reliable long-term supply of clean energy, if it is exploited. In high temperature nuclear systems, molten salts have many attractive advantages as a coolant. Prior work of homogenous core model indicated that the liquid-salt-cooled solid-fuel fast reactor (LSFR) could achieve a self-sustained core based on thorium fuel with exciting neutronic performance. To further explore this concept, a fuel assembly design and a heterogeneous LSFR reference reactor core model were put forward in this study. The aim of this study was to inspect more closely the LSFR self-sustaining core and deeply investigate the neutronic characteristics with fine burnup computational model. One of the benefits of using fine burnup computation model for heterogeneous core was that the detailed physical information in each assembly could be obtained. The leakage rate at End of Equilibrium Cycle (EOEC) comprised 4.93% of 3100 MWth LSFR heterogeneous active core model and 3.00% of prior homogenous active core model. Besides this, the LSFR core neutronic characteristics were in good consistency with that of homogenous model. The characteristics of thorium-based LSFR reference core are (a) a high discharge burnup ~20% FIMA; (b) small reactivity swing during the reactor lifespan; and (c) the negative reactivity temperature coefficients for all cases.     

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。     

钍基柱状高温气冷堆不同启动燃料特性初步分析

钍是一种可转换材料,将其转换成²³³U能极大提高现有核燃料资源的储量。为实现对钍的合理利用,以模块式柱状高温气冷堆GT-MHR的燃料组件作为研究对象,选取低浓缩铀、武器级钚、核反应堆级钚等作为其启动燃料。利用栅格输运计算程序DRAGON对这3种启动燃料下的钍基柱状燃料组件的寿期初中子能谱、无限增殖系数、燃耗、转换比以及²³³U和²³²Th的含量等参数进行了分析。结果表明,在易裂变物质初装量约为9%时,与低浓缩铀和武器级钚相比,核反应堆级钚作为启动燃料时组件寿期初中子能谱较硬、转换比较高;其燃耗达90 GW•d/tHM;其无限增殖系数在寿期内的波动最小;燃耗为75 GW•d/tHM时组件中²³³U存余量与²³²Th消耗量之比达0.566。     

Thorex流程离心萃取器验证台架实验

钍基核燃料后处理Thorex流程的发展倾向于酸式进料,单循环。基于30级10mm环隙式离心萃取器台架系统,对酸式进料、单循环Thorex流程工艺进行台架实验验证。实验结果表明:全流程钍回收率为99.994%,铀回收率为99.30%,钍中去铀分离因子SF_(U/Th)为1.5×10~2,铀中去钍分离因子SF_(Th/U)为2.2×10~4。增加1B工艺段(钍铀分离段)补萃级数应可以进一步提高铀回收率。     

钍基ADS快热耦合次临界核系统燃耗特性分析

在钍基ADS快热耦合次临界反应堆设计的基础上,应用研制的蒙特卡罗燃耗程序MCNTRANS对次临界堆芯在恒定功率下整个寿期内的燃耗特性进行了计算,研究分析了堆芯嬗变能力、钚焚烧性能、堆芯寿期内k_eff变化及加速器束流的协调匹配。分析结果表明：所设计堆芯的次锕系核素（MA）嬗变支持比可达15个百万kW级的PWR,长寿命裂变产物（LLFP）嬗变支持比为2.2个百万kW级的PWR;热区内²³³U的裂变贡献为25％,堆芯嬗变、增殖能力强。整个堆芯寿期内k_eff变化在1%左右,降低了ADS对加速器束流强度的要求。     

Dissolution of ThO2Based Oxides in Nitric Acid Solutions at Elevated Temperatures

Highly-dense spherical particles of thorium-based oxides, ThO₂ and (Th, U)O₂, prepared by the sol-gel method were subjected to dissolution with nitric acid containing 0–0.05 mol/l NaF at high temperatures above 120°C. The dissolution rate depended upon temperature, fluoride concentration and UO₂ content. High-temperature in the range of 120–200°C enhanced the dissolution of the ThO₂-based fuels. At low temperatures and/or low U0₂ concentrations, insoluble tetrafluoride precipitates were formed on the particle surfaces and they resulted in the decrease of the dissolution rates. In the present study, the apparent activation energies for the high-temperature dissolution were obtained.     

钍基ADS快热耦合次临界核系统初步研究

为补偿由于次临界反应堆的燃耗所损失的反应性,降低次临界反应堆功率对加速器束流的依赖,考虑钍的转换,给出了采用钍基燃料,液态铅-铋合金单一回路冷却、石墨慢化的ADS快热单向耦合次临界堆芯设计方案。结果表明:本设计方案实现了堆芯功率展平、中子单向耦合,延长了换料周期,并消除了空腔的不利影响;堆芯寿期内的温度反应性反馈为负效应,安全性高;堆芯具有较高的能量放大能力;堆芯寿期内k_(eff)变化不超过1.05%;所需加速器最大束流强度为4.21mA;堆芯的MA嬗变支持比可达15个百万kW级的PWR,嬗变能力强。