荧光光度法测定矿石和矿渣中的微量钍

应用ＣＬ－５２０９萃淋树脂在２ｍｏｌ／ＬＨＮＯ＿３介质中定量吸附大量铀及微量钍，再利用１ｍｏｌ／ＬＨＣｌ定量淋洗微量钍，通过上述分离过程达到微量钍与大量铀、稀土及其他杂质元素的定量分离。在ｐＨ２．５缓冲溶液中，Ｔｈ－Ｍｏｒｉｎ－ＴＯＰＯ（在ＴｒｉｔｏｎＸ－１００介质中）形成三元络合物。用波长４３２ｕｍ光激发此三元络合物，产生波长为５１０ｎｍ的强烈荧光。钍浓度在１至１０ｎｇ／ｍＬ时具有良好的线性关系，测定下限为ｌｕｇ／ｍＬ钍。     

2-噻吩酰基三氟丙酮与1,10-邻偶氮菲对UO_2~(2+)、Th~(4+)、Ce~(3+)的协同萃取

本文以2-噻吩酰基三氟丙酮(HTTA)与1,10-邻偶氮菲(phen)作为协萃剂,对UO_2~(2+)、Th~(4+)、Ce~(3+)进行了协同萃取的研究。协萃反应可表示为M~(a+)+aHTTA(?)+bphen+(h-a)NO_3~-(?)M(TTA)_a(phen)_b(NO_3)_((n-a)(?))+aH~+协萃配合物的组成分别为UO_2(TTA)_2phen、Th(TTA)_3phenNO_3及Ce(TTA)_3(phen)_2,并分别测出了它们的萃取平衡常数K_(ab)。     

Reaction sintering of rhabdophane into monazite-cheralite Nd1-2xThxCaxPO4 (x = 0 – 0.1) ceramics

The direct sintering of Nd_1-2xCa_xTh_xPO₄·nH₂O rhabdophanes (x = 0 – 0.1) was achieved for the first time resulting in homogeneous high-density monazite-cheralite pellets. Thanks to their large specific surface area, rhabdophane precursors led to a lower shrinkage temperature compared to samples synthesized through solid-state reactions. The associated activation energy was found between 360 ± 90 and 530 ± 90 kJ.mol⁻¹, depending on the thorium incorporation. Meanwhile, sintering map was built up, showing that densification predominated for T ≤ 1200 °C. Conversely, the complete densification took place at 1400 °C concomitantly to grain growth and elimination of open porosity. Moreover, Th-Ca coupled substitution seemed to inhibit both densification and grain growth as the average grain size dropped down one order by one order of magnitude between x = 0.0 and x = 0.1. However, these differences did not affect microhardness, which reached 4.9 ± 0.8 GPa whatever the chemical composition tested.     

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.     

钍基先进坎杜堆非能动慢化剂系统概念设计

以钍基先进重水堆（简称TACR）慢化剂系统作为研究对象,提出了一种满足非能动安全要求的概念设计。在此设计中,首次将慢化剂冷却系统和余热排出系统合二为一,并用热工水力分析程序CATHENA Mod3．5c／Revl分析了反应堆在正常工作时的稳态运行情况,为验证设计的可行性奠定了基础。     

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.     

Evaluation of uranium thorium and plutonium thorium fuel cycles in a very high temperature hybrid system

In recent times, there is a renewed and additional interest in thorium because of its interesting benefits. Thorium fuel cycle is an attractive way to produce long term nuclear energy with low radiotoxicity waste. In addition, the transition to thorium could be done through the incineration of weapons grade plutonium or civilian plutonium. Th-based fuel cycles have intrinsic proliferation-resistance and thorium is 3–4 times more abundant than uranium. Therefore, thorium fuels can complement uranium fuels and ensure long term sustainability of nuclear power.In this paper, the main advantages of the use of fuel cycles based on uranium-thorium and plutonium-thorium fuel mixtures are evaluated in a hybrid system to reach the deep burn of the fuel. To reach this goal, the preliminary conceptual design of a hybrid system composed of a critical reactor and two Accelerated Driven Systems, of the type of very high temperature pebble-bed systems, moderated by graphite and cooled by gas, is analyzed.Uranium-thorium and plutonium-thorium once-through and two stages fuel cycles are evaluated. Several parameters describing fuel behaviour and minor actinide stockpile are compared for the analyzed cycles.     

The effectiveness of full actinide recycle as a nuclear waste management strategy when implemented over a limited timeframe – Part II: Thorium fuel cycle

Full recycling of transuranic (TRU) isotopes can in theory lead to a reduction in repository radiotoxicity to reference levels in as little as ∼500 years provided reprocessing and fuel fabrication losses are limited. However, over a limited timeframe, the radiotoxicity of the ‘final’ core can dominate over reprocessing losses, leading to a much lower reduction in radiotoxicity compared to that achievable at equilibrium. In Part I of this paper, TRU recycle over up to 5 generations of light water reactors (LWRs) or sodium-cooled fast reactors (SFRs) is considered for uranium (U) fuel cycles. With full actinide recycling, at least 6 generations of SFRs are required in a gradual phase-out of nuclear power to achieve transmutation performance approaching the theoretical equilibrium performance. U-fuelled SFRs operating a break-even fuel cycle are not particularly effective at reducing repository radiotoxicity as the final core load dominates over a very long timeframe. In this paper, the analysis is extended to the thorium (Th) fuel cycle. Closed Th-based fuel cycles are well known to have lower equilibrium radiotoxicity than U-based fuel cycles but the time taken to reach equilibrium is generally very long. Th burner fuel cycles with SFRs are found to result in very similar radiotoxicity to U burner fuel cycles with SFRs for one less generation of reactors, provided that protactinium (Pa) is recycled. Th-fuelled reduced-moderation boiling water reactors (RBWRs) are also considered, but for burner fuel cycles their performance is substantially worse, with the waste taking ∼3–5 times longer to decay to the reference level than for Th-fuelled SFRs with the same number of generations. Th break-even fuel cycles require ∼3 generations of operation before their waste radiotoxicity benefits result in decay to the reference level in ∼1000 years. While this is a very long timeframe, it is roughly half that required for waste from the Th or U burner fuel cycle to decay to the reference level, and less than a tenth that required for the U break-even fuel cycle. The improved performance over burner fuel cycles is due to a more substantial contribution of energy generated by ²³³U leading to lower radiotoxicity per unit energy generation. To some extent this an argument based on how the radiotoxicity is normalised: operating a break-even fuel cycle rather than phasing out nuclear power using a burner fuel cycle results in higher repository radiotoxicity in absolute terms. The advantage of Th break-even fuel cycles is also contingent on recycling Pa, and reprocessing losses are significant also for a small number of generations due to the need to effectively burn down the TRU. The integrated decay heat over the scenario timeframe is almost twice as high for a break-even Th fuel cycle than a break-even U fuel cycle when using SFRs, as a result of much higher ⁹⁰Sr production, which subsequently decays into ⁹⁰Y. The peak decay heat is comparable. As decay heat at vitrification and repository decay heat affect repository sizing, this may weaken the argument for the Th cycle.     

Investigation of a gas turbine-modular helium reactor using reactor grade plutonium with 232Th and 238U

Utilization of natural uranium (nat-U) and thorium as fertile fuels has been investigated by in a Gas Turbine – Modular Helium Reactor (GTMHR) using reactor grade plutonium as driver fuel. A neutronic analysis for the full core reactor was performed by using MCNP5 with ENDF/B-VI cross-section library. Different mixture ratios were tested in order to find the appropriate mixture ratio of fertile and fissile fuel particles that gives a comparable k_eff value of the reference uranium fuel. Time dependent calculations were performed by using MONTEBURN2.0 with ORIGEN2.2 for each selected mixture. Different parameters (operation time, burnup value, fissile isotope change, etc.) were subject of performance comparison. The operation time and burnup values were close to each other with nat-U and thorium, namely 3205 days and 176 GWd/MTU for the former and 3175 days 181 GWd/MTU for the latter fertile fuel. In addition, the fissile isotope amount changed from initially 6940.1 kg–4579.2 kg at the end of its operation time for nat-U. These values were obtained for thorium as 6603.3 kg–4250.2 kg, respectively.     

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.