棒状氢化锆慢化钍基熔盐堆中子学性能优化

氢化锆(ZrH)由于具有耐高温、抗辐照和慢化能力强等优点,是反应堆常用的慢化剂。本工作研究具有钍铀转换能自持运行和较低次锕系核素（MA）产量的ZrH慢化熔盐堆的堆芯物理设计方案。采用MOC程序分析了不同燃料盐对于启堆和增殖性能的影响,为提高钍铀转换性能,对堆芯结构和慢化棒设计进行了优化与分析。结果表明：当熔盐体积比处于0.5~0.9时,ZrH慢化剂可将临界所需要的²³³U浓度降低至2%附近;采用含增殖层设计与FLi燃料盐装载的ZrH慢化熔盐堆,50 a平均钍铀转换比（CR）可达到1.028;移动式ZrH慢化棒堆芯设计可实现38 a的自持运行,且堆芯寿期末的MA产量比慢化棒不移动条件下采用FLi燃料盐和FLiBe燃料盐的MA产量分别减少约43%和8%,低于相同能量输出下石墨慢化熔盐堆的MA产量。     

FLi/FLiBe盐中7Li富集度对熔盐快堆钍铀转换性能的影响研究

锂(Li)元素是液态熔盐堆中冷却剂熔盐的重要组成成分,由于6Li相对~7Li具有较大的中子吸收截面,其在冷却剂熔盐中的摩尔含量会影响液态熔盐堆的钍铀转换性能,因此研究~7Li富集度对液态熔盐堆钍铀转换性能的影响十分必要。基于熔盐快堆(Molten Salt Fast Reactor,MSFR)的堆芯结构,分别采用FLi和FLiBe两种不同的冷却剂熔盐,选取范围在99.5%～99.995%的一系列~7Li富集度,借助熔盐堆后处理程序MSR-RS(Molten Salt Reactor Reprocessing Sequence),针对能谱、233U初装量、钍铀转换比、233U净产量和倍增时间、Li的演化以及氚产量等一系列参数进行分析。研究结果表明:在MSFR的堆芯中,较FLiBe而言,采用FLi作载体盐能够获得更好的钍铀转换性能;当~7Li富集度由99.995%变为99.9%时,堆芯钍铀转换比降低约1.6%,氚产量增加约8%。综合考虑燃料制造成本和钍铀转换性能等因素,对于分别采用FLi和FLiBe作载体盐的熔盐快堆MSFR,推荐的~7Li富集度都为99.9%。     

TOPAZ-Ⅱ反应堆慢化剂温度效应分析

TOPAZ-Ⅱ反应堆是以高富集度铀为燃料,以氢化锆为慢化剂的空间发电用反应堆。与一般采用氢化锆作为慢化剂的反应堆不同,TOPAZ-Ⅱ反应堆呈现正的慢化剂温度效应,且其值较大。本工作采用MCNP程序对TOPAZ-Ⅱ反应堆的慢化剂温度效应进行计算,通过分析氢化锆升温前后主要区域中子能谱和中子产生率、中子吸收率及泄漏率的变化,得出产生正慢化剂温度效应的原因：氢化锆升温后,中子产生率增加较大,而泄漏率增加较小,且吸收率减少,从而产生正的慢化剂温度效应。     

氢化锆零功率堆热中子通量谱测量

一、引言氢化锆晶体中氢原子处于束缚态。中子热化过程中与氢化锆的氢原子核能量交换具有量子化特征。所以,用水栅中常用的尼尔金散射核计算氢化锆反应堆热中子通量谱,结果比实际谱软。这是因为在氢化锆-轻水混合慢化剂中,能量低于0.14 eV的中子主要被周围轻水慢化,而中子却可以从氢化锆的氢核获得0.14 eV的整数倍能量。特别当温度升高后,氢化锆中氢原子处于高激发态的份额增加,中子被加速几率增加,使中子谱比水堆     

在线加钚模式下的熔盐堆钍铀燃料可持续性研究

基于FLiBe载体盐,Th/~(233)U启堆,仅通过在线添加反应堆级钚,以实现熔盐堆~(233)U的自持和焚烧反应堆级钚的能力。采用单栅元模型,分析其在不同熔盐体积比、不同中子损失率下233U的自持和钚的利用性能。研究发现:在熔盐体积占比为10%～85%的较大范围内都可以实现~(233)U自持,其中约43%熔盐体积比下~(233)U增殖效果最佳。与此同时,43%熔盐占比下对钚的依赖最大,在熔盐体积比较小和较大时对钚的依赖较小;在熔盐体积比较小时更有利于钚的利用,其中在熔盐体积比为10%～15%时钚的焚烧率最大,约为75%。此外,中子损失率与钚的依赖近似呈正比关系,对~(233)U自持性能影响较小。     

熔盐堆栅格参数优化

针对新型的采用无铍熔盐燃料的氧化铍慢化钍基熔盐堆,利用上海核工程研究设计院自主开发的SONG/TANG-MSR程序系统,通过大量的方案分析,在熔盐堆栅格尺寸、P/D(栅距与燃料孔道直径的比值)、233 U含量等关键栅格参数上对钍基熔盐堆进行优化。计算结果表明,采用较低的233 U浓度的小栅距栅格设计,新型的熔盐堆设计具有很高的增殖比,并保持负功率系数。与传统熔盐堆相比,新型钍基熔盐堆具有更高的核燃料增殖能力。经过栅格优化的新型钍基熔盐堆可满足下一代核能系统可持续性和安全性要求。     

氟盐球床堆中钍铀的基本物理参量研究

通过计算氟盐球床堆单球栅元物理模型,分析不同C/U、C/Th和不同燃耗深度下钍及铀球的中子产能、单位铀累积中子产量、单位钍累积中子销量等基本性质。结果表明,C/Th取80较为合适,钍燃耗深度范围为140~200 MW·d/kg(Th);C/U取400~600范围较为合适,铀燃耗深度范围为180~200 MW·d/kg(U);堆芯径向选择Th-U-Th布置较为合理。     

用碳化锆细化晶粒制取大块氢化锆

一、前言氢化锆具有良好的核特性:中子与束缚于锆晶格中的氢碰撞时,是按量子化的hv=0.137 eV一份一份地交换能量的。低于0.137 eV的中子不但不失去能量,而且得到一份能量,其几率正比于exp(-hv/KT)。温度T越高,越容易得到能量。从而使氢化锆具有很大的负反应性温度系数,成为良好的固体慢化剂。铀锆合金也可用制备氢化锆基本相同的方法,制成铀氢锆元件。这种元件中,燃料温度的变化与慢化剂的温度变化是“同时”的,使得这种元件的负温度系数是瞬发的。因而用这种元件制成的反应堆具有固有的安全性并能脉冲运行,在科研和同位素生产中得到了日益广泛的应用。     

用α能谱法测定样品中~(233)U和~(232)U的相对含量

钍是一种潜在的核燃料。据国际原子能机构估计,目前已探明有开采价值的钍资源大致和铀相等。钍的充分利用可以大大丰富核能资源,增加核燃料储备。 钍铀核燃料循环的一个主要技术问题是钍在反应堆中受中子照射后,除生成有用的核燃料~(233)U外,还会产生~(232)U,一般化学方法不能使它与~(233)U分开。由于~(232)U的衰变子体放出能量     

钍基先进CANDU堆(TACR)钍-铀燃料功率影响研究

在CANDU堆燃料栅元物理的研究中,通常选择堆芯平均的燃料比功率对栅元进行计算模拟,而在TACR中,由于使用了钍燃料,比功率的不同就可能对核反应产生影响,并通过影响棒束栅元的基本截面参数而影响到全堆计算的结果.本文对不同定功率条件下,含全铀燃料和钍-铀燃料棒束的栅元截面参数随辐照值的变化以及钍燃料棒束中233Pa和233U的质量份额进行了计算分析,认为功率会对钍燃料的栅元宏观截面产生影响,在全堆计算中,栅元基本参数应尽量使用基于历史的局部参数法.     

Transition toward thorium fuel cycle in a molten salt reactor by using plutonium

The molten salt reactor(MSR), as one of the Generation Ⅳ advanced nuclear systems, has attracted a worldwide interest due to its excellent performances in safety, economics, sustainability, and proliferation resistance. The aim of this work is to provide and evaluate possible solutions to fissile 233 U production and further the fuel transition to thorium fuel cycle in a thermal MSR by using plutonium partitioned from light water reactors spent fuel. By using an in-house developed tool, a breeding and burning(BB) scenario is first introduced and analyzed from the aspects of the evolution of main nuclides, net 233 U production, spectrum shift, and temperature feedback coefficient. It can be concluded that such a Th/Pu to Th/~(233)U transition can be accomplished by employing a relatively fast fuel reprocessing with a cycle time less than 60 days. At the equilibrium state, the reactor can achieve a conversion ratio of about 0.996 for the 60-day reprocessing period(RP) case and about 1.047 for the 10-day RP case.The results also show that it is difficult to accomplish such a fuel transition with limited reprocessing(RP is 180 days),and the reactor operates as a converter and burns the plutonium with the help of thorium. Meanwhile, a prebreeding and burning(PBB) scenario is also analyzed briefly with respect to the net 233 U production and evolution of main nuclides. One can find that it is more efficient to produce 233 U under this scenario, resulting in a double time varying from about 1.96 years for the 10-day RP case to about 6.15 years for the 180-day RP case.     

New breeding gain definitions and their application to the optimization of a molten salt reactor design

The molten salt reactor (MSR) is an attractive breeder reactor. A graphite-moderated MSR can reach breeding because of the online salt processing and refueling. These features give difficulties when the breeding gain (BG) of the MSR is evaluated. The inventory of the core and external stockpiles have to be treated separately in order to quantify the breeding performance of the reactor. In this paper, an improved BG definition is given and it is compared with definitions used earlier. The improved definition was used in an optimization study of the graphite – salt lattice of the core. The aim of the optimization is a passively safe, self-breeder reactor. The fuel channel diameter, graphite volume and thorium concentration were varied while the temperature feedback coefficient of the core, BG – as defined in the paper – and the lifetime of the graphite were calculated. There is a small range of lattices which provide both negative temperature feedback and breeding. Furthermore, breeding is possible only at low power densities in case of the salt processing efficiencies set in this study. In this range of power the lifetime of the graphite is between 12 and 20 years.     

基于MOREL2.0对石墨慢化熔盐堆的稳态分析

熔盐堆采用熔融的氟化盐混合物作为燃料和堆芯的冷却剂,由于燃料的流动,熔盐堆在中子学和热工水力学方面与传统固体燃料反应堆有着较大区别。本文基于熔盐堆分析程序MOREL2.0对钍基熔盐堆（TMSR）初步堆芯设计方案进行了稳态计算分析,结果表明：燃料流动对缓发中子先驱核的分布影响较大,并导致169 pcm反应性损失;随燃料在外部回路中滞留时间的增加,k_eff降低,80 s后趋于平稳;TMSR具有负的入口燃料温度系数,具有固有安全性。     

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。     

熔盐堆物理热工耦合程序开发及验证分析

熔盐堆(Molten Salt Reactor,MSR)是第四代反应堆6种堆型中唯一的液态燃料反应堆,与固态燃料-液体冷却剂反应堆相比,原理上有较大不同。在熔盐堆中,流动的熔盐既是燃料又是冷却剂与慢化剂,中子物理学与热工水力学相互耦合;由于熔盐的流动性,缓发中子先驱核会随燃料流至堆芯外衰变,造成缓发中子的丢失,导致堆芯反应性降低。正是由于熔盐堆的这些新特性,造成熔盐堆内缓发中子先驱核、温度等参数变化与固态燃料反应堆有所不同,需要研究熔盐堆在各种工况下的相关物理参数变化。本文主要工作是考虑缓发中子先驱核的流动性对熔盐堆的影响,研究适用于熔盐堆的二维圆柱几何时空中子动力学程序及与之耦合的热工水力学程序;利用该程序对熔盐堆中子物理学和热工水力学进行耦合计算,验证熔盐堆相关实验数据;并且计算了熔盐堆无保护启停泵及堆芯入口温度过冷过热工况,用于分析熔盐堆的安全特性。计算结果表明,程序能够对熔盐反应堆实验(Molten Salt Reactor Experiment,MSRE)的相关实验数据进行较好的模拟计算,并且验证了熔盐堆的固有安全性。     

Influences of 7Li enrichment on Th-U fuel breeding for an Improved Molten Salt Fast Reactor (IMSFR)

The molten salt fast reactor (MSFR) shows great promise with high breeding ratio (BR),large negative temperature coefficient of reactivity,high thermal-electric conversion efficiency,inherent safety,and online reprocessing.Based on an improved MSFR optimized by adding axial fertile salt and a graphite reflector,the influences of 7Li enrichment on Th-U breeding are investigated,aiming to provide a feasible selection for the molten salt with high fissile breeding and a relatively low technology requirement for 7Li concentration.With the self-developed molten salt reactor reprocessing sequence based on SCALE6.1,the burn-up calculations with online reprocessing are carried out.Parameters are explored including BR,233U production,double time (DT),spectrum,6Li inventory,neutron absorption,and the tritium production.The results show that the 7Li enrichment of 99.95％ is appropriate in the fast fission reactor.In this case,BR above 1.10 can be achieved for a long time,corresponding to the 233U production of 130 kg per year and DT of 36 years.After 80 years' operation,the tritium production for 99.5％ is only about 7 kg,and there is no obvious increase compared to that for 99.9995％.     

Burnup optimization of once-through molten salt reactors using enriched uranium and thorium

The advantages of once-through molten salt reactors include readily available fuel,low nuclear proliferation risk,and low technical difficulty.It is potentially the most easily commercialized fuel cycle mode for molten salt reactors.However,there are some problems in the parameter selection of once-through molten salt reactors,and the relevant burnup optimization work requires further analysis.This study examined once-through graphitemoderated molten salt reactor using enriched uranium and thori...