Re-evaluation of Radiation-Energy Transfer to an Extraction Solvent in a Minor-Actinide-Separation Process Based on Consideration of Radiation Permeability

ABSTRACT

Absorbed-dose estimation is essential for evaluation of the radiation tolerance of minor-actinide-separation processes. We propose a dose-evaluation method based on radiation permeability, with comparisons of heterogeneous structures seen in the solvent-extraction process, such as emulsions forming in the mixture of the organic and aqueous phases. A demonstration of radiation-energy-transfer simulation is performed with a focus on the minor-actinide-recovery process from high-level liquid waste with the aid of the Monte Carlo radiation-transport code PHITS. The simulation results indicate that the dose absorbed by the extraction solvent from alpha radiation depends upon the emulsion structure, and that from beta and gamma radiation depends upon the mixer-settler-apparatus size. Non-negligible contributions of well-permeable gamma rays were indicated in terms of the plant operation of the minor-actinide-separation process.     

超临界流体络合萃取镧系和锕系元素的研究进展

超临界流体络合萃取是1种可用于萃取金属离子的新型分离技术，它应用于乏燃料后处理和核废物处理中的最大优点是能大幅减少二次废物的产生。文章主要介绍超临界流体的特点、超临界流体络合萃取的基本原理及其影响因素，并对超临界流体络合萃取镧系和锕系元素的研究进展及其在乏燃料后处理工艺研究中的应用进行综述。     

DHDECMP-TBP/OK萃取模拟高放废液中锕系元素的工艺研究

测定了22%DHDECMP 42%TBP/OK从模拟高放废液中萃取锕系元素的分配比;在微型离心萃取器(转鼓=10mm)串联台架上,进行了从模拟高放废液中萃取锕系元素的工艺条件研究。结果表明,经6级萃取、2级洗涤、6级反萃,流比AF∶AX∶AS=1∶1 5∶0 5,BF∶BX=1∶1时,在萃取器A中,U,Np,Pu,Am的回收率均大于99 9%;在反萃器B中,Am的反萃率>99 9%,U,Np和Pu的反萃率分别为2%,39%,2%。     

Study on FBR core concepts to increase proliferation resistance of plutonium in LWR–FBR transition period

The possibility to enhance proliferation resistance of discharged plutonium in fast breeder reactor (FBR) has been investigated in terms of reactor core-design aspects. The provisional target for proliferation resistance measures based on Saito's attractiveness (ATTR) is defined. It is found that a few percent of plutonium loading and/or Am/Cm loading, which come from various types of spent fuel, might satisfy the provisional target with a minimum impact on the core neutronic performances. On Am/Cm loading core, decay heat constraints for fuel-handling aspects are found to be important and should be considered in design. There is not significant change on the current developing scenarios for light water reactor–FBR transition period by applying the measures based on Saito's ATTR. It is found that applying Kimura's proposal, 15% ²³⁸Pu content requires about 7% MA loading fraction both in the core and the blanket, and it only can be applied at limited case. The period to consume minor actinides is shorter than to consume plutonium.     

Evaluation of Proliferation Resistance of Plutonium Based on Decay Heat

The present paper is prepared for the peaceful use of nuclear energy. Proliferation resistance of plutonium can be enhanced by increasing the decay heat of plutonium. For example, it can be enhanced by increasing the isotopic fraction of ²³⁸Pu, which has the largest decay heat among plutonium isotopes. In the present paper, the proliferation resistance of plutonium was evaluated based on decay heat with a physical assessment model. New criteria were proposed to evaluate the proliferation resistance based on isotopic compositions of plutonium from the viewpoint of decay heat. The present criteria were applied to evaluate the proliferation resistance of plutonium produced in typical Light Water Reactor and Fast Breeder Reactor based on an evaluation function “Attractiveness” as case studies.     

加速器驱动钠冷金属燃料快堆次锕系核素嬗变特性研究

分析加速器驱动系统(ADS)钠冷金属燃料快堆重金属燃料不同核素对堆芯有效增殖系数(Keff)的影响,给出了燃料成分的确定方法,详细分析次锕系核素(MA)嬗变特性.运用耦合了MCNP4c3与ORJGEN2的三维燃耗程序COUPLE对堆芯进行稳态与燃耗计算.结果分析表明,调节燃料中239Pu的质量比例并使其在燃耗过程中保持稳定是使Keff达到设计值并在燃耗过程中保持稳定的有效手段.散裂中子引起堆芯内区较外区更硬的中子能谱,有利于提高MA的裂变截面与裂变吸收比.全堆MA嬗变支持比为8.3,具有较好的嬗变效果.由于堆芯内区的高通量,堆芯内外区的嬗变率有明显差异,将MA集中布置于内区有利于减少装料量,改善总体嬗变效果.     

Densification of magnesia-based inert matrix fuels using asbestos waste-derived materials as a sintering additive

A new concept for densification of minor actinide-containing inert matrix fuels (IMFs) using asbestos waste-derived materials was proposed for the effective utilization of resources and health protection of the general public. In this concept, magnesium silicates, which are mainly generated by the decomposition of asbestos in low temperature heat-treatments, are used as a sintering additive to achieve high density magnesia (MgO) -based IMFs at relatively low sintering temperature. In the present study, preliminary fabrication tests of MgO-based IMFs with magnesium silicates were carried out using cerium oxide (CeO₂) as a representative of minor actinide oxides. The sintered densities of MgO-based IMFs increased with use of the additives. The sintering behavior of MgO-based IMFs with magnesium silicate additives was discussed from the viewpoints of the effects of magnesium silicates on the densification of the MgO and CeO₂.     

利用超长寿命快堆嬗变亚锕元素的特性研究

对利用超长寿命快堆（ＵＬＬＦＢＲ）嬗变高放核废物亚锕元素（ＡＭｓ）的堆芯物理特性进行了初步分析，得出了在ＵＬＬＦＢＲ中适当布置ＡＭｓ，既可以嬗变ＭＡｓ，又可改善超长寿命反应堆的物理特性这一结论。     

Chromatographic Separation of Trivalent Actinides by Using Tertiary Pyridine Resin with Methanolic Nitric Acid Solutions

Chromatographic separation of trivalent actinides (Am, Cm and Cf) was performed by using a tertiary pyridine resin embedded in silica beads with methanolic nitric acid solutions. The trivalent actinides were eluted from the resin column in the reverse order of atomic numbers (Cf-Cm-Am). Higher concentration of methanol in the mixed solution accelerated both the adsorption of these elements on the resin and the separability for these elements. Americium was clearly separated from Cm and Cf by using a 1 cm-ø × 10 cm-height column with a 60vol% of methanol/40 vol% of concentrated nitric acid mixed solution at ambient temperature.     

Selective Actinide Extraction with a Tri‐Synergistic Mixture Using a Centrifugal Contactor Battery

Abstract

The reprocessing of spent nuclear fuels produces a quantity of high‐level liquid waste (HLLW), which is subsequentely vitrified. The most important contributors to the long‐term radiotoxicity of the vitrified HLLW are the so‐called minor actinides (MAs), which are neptunium, americium, and curium. This paper presents the results obtained during some extraction tests of americium(III) in a synthetic liquid waste representative of that coming from the PUREX and DIAMEX processes, obtained by dissolving the respective nitrate salts of the lanthanides(III) in a diluted solution of nitric acid and spiked with ²⁴¹Am and ¹⁵²Eu. The extraction process employed a tri‐synergistic organic extractant mixture and a centrifugal contactor battery of eight stages.