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.     

钍的提取及金属钍制备技术

钍是重要的核燃料之一,其中子吸收截面大且本身不易裂变,经中子轰击后转化为易裂变的233U,形成核燃料高效增殖体系。由于具有核废料毒性小、核扩散少、储量丰富且能更大程度保障核燃料供应等诸多优势,在未来核能利用中应用前景巨大。钍通常采用萃取法进行提取,得到的含钍产物经进一步加工制备成金属钍。本文简述了目前国内外萃取法提取钍的研究进展,并概述了金属钍的制备方法,主要包括熔盐电解法、金属热还原法、碘化法和电迁移法的基本原理、工艺流程及技术概况。     

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。     

N,N,N‘,N’—四丁基己二酰胺在硝酸介质中萃取铀,钍及硝酸的研究

研究了Ｎ，Ｎ，Ｎ’，Ｎ’－四丁基己二酰胺（ＴＢＡＡ）从硝酸介质中萃取铀，钍及硝酸的机理，考究了ＨＮＯ３浓度，ＴＢＡＡ浓度，温度及盐析（ＬｉＮＯ３）对铀和钍分配系数的影响，还对铀，钍的反萃进行了研究，得出萃合物的组成的ＵＯ２（ＮＯ３）２．ＴＢＡＡ和Ｔｈ（ＮＯ３）２．ＴＢＡＡ。在低酸度下，ＴＢＡＡ与硝酸主要以ＴＢＡＡ．ＨＮＯ３的形式存在，在高酸度下，ＴＢＡＡ与硝酸的加合物不但有（ＴＢＡＡ）２．ＨＮＯ３     

Sorption of thorium using magnetic graphene oxide polypyrrole composite synthesized from natural source

Polypyrrole magnetic graphene oxide (PPy/MGO) composites has been synthesized from natural source (water hyacinth roots) using polymerization technique for Th(IV) ions pre-concentration from aqueous solutions. The effects of controlling factor have been studied using batch technique. The obtained results show that the maximum Th(IV) adsorption capacity by PPy/MGO composite is 277.8 mg/g at pH 4, which is higher than traditional adsorbents. PPy/MGO composite also presents excellent regeneration/reuse property. Due to the high adsorption capacity of Th(IV), PPy/MGO composite can be used in nuclear fuel achievement and for Th(IV) environmental pollution cleanup.     

Investigation of the hydrogen production of a laser FUSION driver thorium breeder using various coolants

In this present work, hydrogen production and neutronic calculations of a Laser Inertial Confinement Fusion Fission Energy (LIFE) driven thorium breeder using various coolants have been investigated. In the neutronic calculations for fusion driver power of 500 MW_th has been examined with MCNP code. The 95 vol% Flibe or natural lithium 5 vol% TRISO coated ThC fuels have used in the neutronic calculations. Tritium breeding ratio (TBR) has been calculated as 1.08 and 1.19, respectively, for Flibe and natural lithium coolants. The energy multiplication values have been computed as 3.17 and 1.62, respectively, for these coolants. The burnup values with flibe and natural lithium have been obtained as 6 GWd/tM and 22 GWd/tM over 11 and 23 years, respectively. Also, the hydrogen production of a laser fusion driver thorium breeder using steam methane reforming (SMR), high temperature electrolysis (HTE) and sulfur-iodine (S–I) thermochemical water splitting processes have been performed. The highest hydrogen production values with flibe coolant of SMR method have been obtained as ~200 kg/s over 11 years.     

Nuclear fission sustainability with subcritical reactors driven by external neutron sources

Although nuclear breeder reactors are a promising way to enhance the potential energy currently retrievable from the Uranium reserves, they still have disadvantages because of their safety features (i.e. poor stabilizing mechanisms) and the security of their fuel cycle (diversion of Pu for non-civilian purposes). Loading natural nuclear fuels to a reactor and completely burning them without reprocessing would be ideal, however, this is not possible in critical reactors due to the limitations imposed by the maximum achievable burn-up. An alternative option to attain very high percentages of nuclear natural materials exploitation, while meeting other objectives of Nuclear Sustainability, could consist of using externally-driven subcritical reactors to reach the desired high burn-ups (of the order of 30% and more) without reprocessing. Such scheme would lead to an efficient exploitation of the available raw material, without any risk of proliferation. Exploring this type of reactor concept, this paper analyzes the different ways to accomplish this goal while identifying potential setbacks.     

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.     

Structural investigation of thorium in molten lithium-calcium fluoride mixtures for salt treatment process in molten salt reactor

X-ray absorption fine structure (XAFS) measurements on thorium fluoride in molten lithium-calcium fluoride mixtures and molecular dynamics (MD) simulation of zirconium and yttrium fluoride in molten lithium-calcium fluoride mixtures have been carried out. In the molten state, coordination number of thorium (N_i) and inter ionic distances between thorium and fluorine in the first neighbor (r_i) are nearly constant in all mixtures. However the fluctuation factors (Debye-Waller factor (σ²) and C₃ cumulant) increase until xCaF₂ = 0.17 and decrease by addition of excess CaF₂. It means that the local structure around Th⁴⁺ is disordered until xCaF₂ = 0.17 and stabilized over xCaF₂ = 0.17. The variation of fluctuation factors is related to the number density of F⁻ in ThF₄ mixtures and the stability of local structure around Th⁴⁺ increases with decreasing the number density of F⁻ in ThF₄ mixtures. This tendency is common to those in the ZrF₄ and YF₃ mixtures. However, in the case of YF₃ mixtures, the local structure around Y³⁺ becomes disordered until xCaF₂ = 0.40 and it becomes stabilized by addition of excess CaF₂. The difference between ThF₄ mixtures and YF₃ mixtures is related to the difference of Coulumbic interaction between Th⁴⁺-F⁻ and Y³⁺-F⁻. Therefore, the variation of local structure around cation is related to not only number density of F⁻ in molten salts but also the Coulumbic interaction between cation and anion.