氢化锆慢化熔盐堆钍铀转换性能初步分析

中子能谱对钍基燃料在熔盐堆中的利用效率及温度反馈系数等安全问题有较大影响,所以对熔盐堆新型慢化剂的研究具有重要意义。本工作基于SCALE6计算程序,对不同几何栅元结构的氢化锆栅元组件在熔盐堆的物理性能进行了研究,分别计算了中子能谱、钍铀转换比、~(233)U浓度、总温度反馈系数以及燃耗等中子物理参量。结果表明,减小六边形栅元对边距或者增加熔盐占栅元体积比可以增加钍铀转换比和改善温度反应性系数;当加入的氢化锆慢化剂体积份额为0.1时就可以将熔盐堆~(233)U初始浓度降低到2.5×10~(-2)以内;氢化锆慢化熔盐堆在超热谱条件下,其~(233)U初装载量和超铀核素产量较小,同时堆芯较为紧凑。     

使用离散钍铀燃料组件的CANDU6堆物理特性初步研究

为研究钍铀燃料在CANDU6堆中的应用，采用DRAGON/DONJON程序，对使用离散型钍铀燃料37棒束组件的CANDU6堆进行时均堆芯分析。结果表明，组件采用235U富集度为2.5%的铀棒以及第1、2、3圈布置钍棒的37棒束组件，堆芯在8棒束换料、3个燃耗分区的方案下，组件的冷却剂空泡反应性较使用天然铀的37棒束组件（NU-37组件）与采用混合钍铀元件棒的37棒束组件更负；堆芯最大时均通道/棒束功率满足小于6700?kW/860?kW的限值；燃料转化能力比采用NU-37组件时更高；卸料燃耗可到达13400?MW·d/t(U)。研究表明，所设计的离散型钍铀燃料37棒束组件可用于现有CANDU6堆芯，且无需对堆芯结构及控制机构作重大改造；燃料组件和堆芯设计方案可为钍铀燃料在CANDU6堆芯的应用提供参考。      

钍铀自持循环物理热工特性研究

采用CANFLEX43型燃料棒为驱动燃料棒,基于燃料棒的富集度不同,设计了三种驱动方案。通过Dragon程序及建立的热工模型,对三种方案的燃料棒束栅元物理热工特性及冷却剂平均温度进行敏感性分析研究。结果表明:233U富集度为1.4%的驱动方案下换料燃耗为16 MWd/kg,233U增量为5.9872 g,燃料包壳表面最高温度371oC,燃料芯块中心温度1830oC;单个通道冷却剂最大流量为25.4 kg/s,均能很好满足钍铀自持循环物理热工特性的要求。     

钍基先进CANDU堆燃料循环方式研究

在钍基先进CANDU堆的概念设计中，钍燃料的循环利用方式是一重要问题。文章采用中心两圈为钍燃料、外面两圈为稍加浓缩铀燃料的CANFLEX燃料棒束，通过对燃料棒束栅元物理特性的研究，提出了一套切实可行的直接自身再循环的燃料棒束循环方案。     

钍－铀栅零功率堆临界实验（Ⅱ）及中子通量分布

在中国科学院上海原子核研究所零功率装置上完成了钍－铀栅分列元件的临界实验和中子通量分布测量。使用了直径８ｍｍ、加浓度３％的低浓铀燃料元件和同样直径的四氟化钍棒。实验数据与理论计算值符合较好。     

钍基重水核能系统燃料的物理特性研究

对全铀CANFLEX燃料和含钍CANFLEX燃料的物理特性进行了研究。用WIMS-AECL程序计算了参考栅元的冷却剂空泡反应性、燃耗等参数。通过与铀燃料的对比，展示了含钍燃料在安全性和经济性上的特点还特别介绍WIMS-AECL程序在使用过程中的参数选择方法，     

压水堆平衡堆芯钍铀燃料循环初步研究

建立WIMSD5-SN2-CYCLE3D和CASMO3-CYCLE3D物理分析系统作为钍铀燃料循环研究工具.以大亚湾第1机组压水堆为参考堆型,不改变反应堆栅元、组件和堆芯的结构与几何尺寸,设计出含36根钍棒、4.2#5U富集度的新型含钍组件,并对含钍组件和3.2%富集度的铀组件进行中子学计算和分析.模拟并分析了大亚湾压水堆12个月换料从初始循环到铀钚平衡循环的换料过程.再从平衡铀堆芯出发,逐步加入含钍组件代替铀组件,对铀钚平衡循环到钍铀平衡循环的换料过程进行了模拟与分析.计算结果表明:钍铀平衡循环比铀钚平衡循环每天节省裂变核素质量约18.4%,并减少了长寿命放射性核废料的产生.不利因素是使得循环长度减少90EFPD,缩短了换料周期,增加运行费用,并给燃料管理、安全控制以及乏燃料的处理带来困难.建议提高组件的235U富集度,在压水堆上进行钍利用研究.     

钍燃料重水堆堆芯特性研究

现有的CANDU重水堆(简称"重水堆")以天然铀作为燃料,但重水堆由于其独特的堆芯设计,具有较好的燃料灵活性,还可以烧低浓铀、回收铀和钍等燃料。研究现有重水堆改烧钍燃料后对堆芯特性和运行安全的潜在影响。使用DRAGON程序建立了重水堆的无限栅元模型,研究比较了钍燃料和天然铀燃料的重要堆芯特性参数。结果表明,尽管2种燃料下的堆芯特性有所差异,但钍燃料利用实际上有助于提升重水堆的运行安全。     

离子交换法分离铀中的微量钍

矿石中常有铀钍伴生的情况,因此在精炼的天然铀产品中常含有微量钍。这样制成的生产堆燃料元件辐照时,铀-232俘获中子而转变为镤-233,这会使经后处理得到的铀钚产品中γ放射性过高。为此,应对前处理中得到的精炼铀产品中的钍含量加以检验和控制。此外,Thorex流程中的铀-233液流和最终铀产品中也必定含有微量钍,须进行检测。因此,建立一个简便可靠的方法来分析大量铀中的微量钍,就具有实际的意义。本文采用简     

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。     

Comparison of thorium-based fuels with different fissile components in existing boiling water reactors

With the aim of investigating the technical feasibility of fuelling a conventional BWR (Boiling Water Reactor) with thorium-based fuel, computer simulations were carried out in a 2D infinite lattice model using CASMO-5. Four different fissile components were each homogenously combined with thorium to form mixed oxide pellets: Uranium enriched to 20% U-235 (LEU), plutonium recovered from spent LWR fuel (RGPu), pure U-233 and a mixture of RGPu and uranium recovered from spent thorium-based fuel. Based on these fuel types, four BWR nuclear fuel assembly designs were formed, using a conventional assembly geometry (GE14-N). The fissile content was chosen to give a total energy release equivalent to that of a UOX fuel bundle reaching a discharge burnup of about 55 MWd/kgHM. The radial distribution of fissile material was optimized to achieve low bundle internal radial power peaking. Reactor physical parameters were computed, and the results were compared to those of reference LEU and MOX bundle designs. It was concluded that a viable thorium-based BWR nuclear fuel assembly design, based on any of the fissile components, can be achieved. Neutronic parameters that are essential for reactor safety, like reactivity coefficients and control rod worths, are in most cases similar to those of LEU and MOX fuel. This is also true for the decay heat produced in irradiated fuel. However when Th is mixed with U-233, the void coefficient (calculated in 2D) can be positive under some conditions. It was concluded that it is very difficult to make savings of natural uranium by mixing LEU (20% U-235) homogenously with thorium and that mixing RGPu with thorium leads to more efficient consumption of Pu compared to MOX fuel.     

Neutronics and material attractiveness for PWR thorium systems using monte carlo techniques

Thorium (Th) oxide fuel offers a significant advantage over traditional low-enriched uranium and mixed uranium/plutonium oxide (MOX) fuel irradiated in a Light Water Reactor. The benefits of using thorium include the following: 1) unlike depleted uranium, thorium does not produce plutonium, 2) thorium is a more stable fuel material chemically than LEU and may withstand higher burnups, 3) the materials attractiveness of plutonium in Th/Pu fuel at high burnups is lower than in MOX at currently achievable burnups, and 4) thorium is three to four times more abundant than uranium. This paper quantifies the irradiation of thorium fuel in existing Light Water Reactors in terms of: 1) the percentage of plutonium destroyed, 2) reactivity safety parameters, and 3) material attractiveness of the final uranium and plutonium products. The Monte Carlo codes MCNP/X and the linkage code Monteburns were used for the calculations in this document, which is one of the first applications of full core Monte Carlo burnup calculations. Results of reactivity safety parameters are compared to deterministic solutions that are more traditionally used for full core computations.Thorium is fertile and leads to production of the fissile isotope ²³³U, but it must be mixed with enriched uranium or reactor-/weapons-grade plutonium initially to provide power until enough ²³³U builds in. One proposed fuel type, a thorium-plutonium mixture, is advantageous because it would destroy a significant fraction of existing plutonium while avoiding the creation of new plutonium. ²³³U has a lower delayed neutron fraction than ²³⁵U and acts kinetically similar to ²³⁹Pu built in from ²³⁸U. However, as with MOX fuel, some design changes may be required for our current LWR fleet to burn more than one-third a core of Th/Pu fuel and satisfy reactivity safety limits. The calculations performed in this research show that thorium/plutonium fuel can destroy up to 70% of the original plutonium per pass at 47 GWd/MTU, whereas only about 30% can be destroyed using MOX. Additionally, the materials attractiveness of the final plutonium product of irradiated plutonium/thorium fuel is significantly reduced if high burnups (∼94 GWD/MTU) of the fuel can be attained.     

Performance of natural uranium- and thorium-fueled fast breeder reactors (FBRs) for U fissile production

The performance of natural uranium and thorium-fueled fast breeder reactors (FBRs) for producing ²³³U fissile material, which does not exist in nature, is investigated. It is recognized that excess neutrons from FBRs with good neutron economic characteristics can be efficiently used for producing ²³³U. Two distinct metallic fuel pins, one with natural uranium and another with natural thorium, are loaded into a large sodium-cooled FBR. ²³³U and the associated-U isotopes are extracted from the thorium fuel pins. The FBR itself is self-sustained by plutonium produced in the uranium fuel pins. Under the equilibrium state, both uranium and thorium spent fuels are periodically discharged with a certain discharge rate and then separated. All discharged fission products are removed and all discharged actinides are returned to the FBRs except the discharged uranium utilized for fresh fuel of the other thorium-cycled reactors. ²³³U-production rate of the FBRs as a function of both the uranium–thorium fuel pins fraction in the core and the discharge fuel burnup is estimated. The result shows that larger fraction of uranium pins is better for the FBR criticality while larger fraction of thorium fuel pins and lower fuel burnup give higher ²³³U production rate.     

The utilization of thorium in Generation IV reactors

The main goal of this paper is to show how thorium, as an alternative nuclear fuel, could be applied as fuel in a Generation IV reactor. The paper focuses on the multiplication factor, the produced ²³³U and delayed neutron fraction in infinite lattice models. For the investigations, simplified models of a fuel assembly of five design types of the six reactor concepts were elaborated. The MSR reactor type is out of scope of this paper due to the fact that it is designed for the utilization of thorium. Although the fissile isotope content was not increased to compensate the thorium caused multiplication factor decrease, the burnup calculations suggest that the designs of ESFR (European Sodium-Cooled Fast Reactor) and ELSY (European Lead-cooled System) are the most promising types according to the trend of the multiplication factor changes and the amount of produced fissionable ²³³U.     

Irradiation of a thorium–plutonium rodlet: Experiment and benchmark calculations

A benchmark exercise for thorium–plutonium fuel, based on experimental data, has been carried out. A thorium–plutonium oxide fuel rodlet was irradiated in a PWR for four consecutive cycles, to a burnup of about 37 MWd/kgHM. During the irradiation, the rodlet was inserted into a guide tube of a standard MOX fuel assembly. After the irradiation, the rod was subjected to several PIE measurements, including radiochemical analysis. Element concentrations and radial distributions in the rodlet, multiplication factors and distributions within the carrier assembly of burnup and power were calculated. Four participants in the study simulated the irradiation of the MOX fuel assemblies including the thorium–plutonium rodlet using their respective code systems; MCBurn, HELIOS, CASMO-5 and ECCO/ERANOS combined with TRAIN. The results of the simulations and the measured results of the radiochemical analysis were compared and found to be in fairly good agreement when the calculated results were calibrated to give the same burnup of the thorium–plutonium rodlet as that experimentally measured. Average concentrations of several minor actinides and fission products were well reproduced by all codes, to the extent that can be expected based on known uncertainties in the experimental setup and the cross section libraries. Calculated results which could not be confirmed by experimental measurement were compared and only two significant anomalies were found, which can probably be addressed by limited modifications of the codes.     

Reactor physics ideas for large scale utilization of thorium in gas cooled reactors

The physics principles for maximizing the fertile to fissile conversion were used in developing reactor concepts for large scale utilization of thorium in thermal and fast reactors (Jagannathan & Pal, 2006; Jagannathan et al., 2008). It is recognized that these principles are very well suited for ‘He’ gas cooled reactors with graphite moderator since both helium gas coolant and the graphite moderator have low neutron absorption characteristics and thus gives better neutron economy. In this paper, these ideas are applied to the High Temperature Test Reactor (HTTR) core of Japan to assess its advantage over the present day gas cooled reactors. HTTR is helium cooled and graphite moderated system. Significant amount of thorium has been loaded in the HTTR core with some minimal changes in the existing core design. The modified design is called HTTR-M core.In the HTTR-M core, the fuel is changed from enriched UO₂ fuel to Pu in ThO₂ fuel. The locations of boron type burnable poison rods within each fuel assembly of HTTR are replaced by one cycle irradiated thoria rods. Also, the B₄C type control assembly around the HTTR core is replaced by fresh seedless thorium assembly. The fertile thoria assembly are scattered uniformly in the HTTR-M core. The equilibrium core of HTTR-M shows very small burnup reactivity swing. The core excess reactivity is ∼18 mk at BOC and reduces to 1 mk at 660 days. It is interesting to note that this small reactivity change is intrinsically achieved by the choice of seed and fertile dimensions and their contents without the use of burnable poison rods or mechanical control rods which are used in HTTR core. The burnup reactivity swing in the latter after using burnable poison is ∼100 mk. The fissile seed inventory ratio (FIR) in a fuel cycle is 0.90 as compared with 0.717 of HTTR core. Since ²³³U is a better fissile nuclide with highest ‘η’ value in thermal range, the above conversion ratio can be regarded as quite good.     

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.     

Coupled neutronics/thermal hydraulics evaluation for thorium based fuels in thermal spectrum SCWR

Thorium can supplement the current limited reserves of uranium. In current study, analyses are performed for thorium based fuels in thermal neutron spectrum Super Critical Water Reactor (SCWR). Thorium based fuels are studied in two roles. First role being replacement of conventional uranium dioxide fuel while the other being burner of Reactor Grade Plutonium (RG-Pu) in thermal neutron spectrum SCWR. Coupled neutron physics/thermal hydraulics analyses are performed due to large density variation of coolant over the active fuel length. Analyses reveal that thorium-uranium MOX fuels lead to smaller burnup values as compared to equivalent enriched uranium dioxide but possess the advantage of smaller excess reactivity at Beginning of Life (BOL). This can lead to savings in the form of Burnable Poisons (BP). Smaller fuel average temperature values are obtained for thorium-uranium MOX fuels as compared to uranium dioxide fuel option. Coated fuel option utilizing mixed thorium-uranium mono nitride fuel can help further decrease fuel average temperature values for thorium based fuels. U-233, produced in thorium uranium fuels, contribution towards fission energy produced is smaller as compared to plutonium produced in conventional uranium dioxide fuel. In terms of proliferation resistance, approximately 40% less quantity of plutonium is produced for thorium-uranium MOX fuels (for studied compositions) as compared to equivalent enriched uranium dioxide fuel. But, there is not much difference between the discharged plutonium vector compositions. Thorium–Plutonium based fuels lead to significantly harder spectrum which results in larger spread in radial power density and eventually causes larger values for thermal hydraulic parameters like fuel and clad temperature. Due to almost no production of plutonium, thorium based fuels can be a very good option to burn RG-Pu in thermal spectrum SCWR. Thorium based fuels destroyed almost 74% initially loaded RG-Pu as compared to 60% for uranium based MOX. HEU based thorium fuels can be a very good option for replacing conventional uranium dioxide fuels as very small quantities of plutonium is produced. This option, although, has regulatory issues due to use of HEU material.     

Neutronics assessment of the use of thorium fuels in current pressurized water reactors

The use of thorium fuel in current PWRs in a once-through fuel cycle is an attractive option due to potential advantages such as high conversion ratio and low minor actinide generation. The current neutronics assessments indicate that the thorium fuel cycle could supplement the current uranium–plutonium fuel cycle to improve operational performance and spent fuel consideration in current PWRs without core and subassembly modifications. Neutronics safety parameters in the PWR cores with the thorium fuels are within the range of current PWRs.The PWR cores with thorium fuels have significantly higher conversion ratios which could enable efficient fuel utilization. Further, it is shown that the use of thorium as a fertile material can reduce minor actinide generation and the radio-toxicity of spent fuels. In considerations related to proliferation resistance, the results of the current analyses show no significant difference between the studied thorium fuels and the standard oxide fuel for the assumed characteristics and burnup levels.