Column Separation of Am(III) and Eu(III) by α-Zirconium Phosphate Ion Exchanger in Nitric Acid

Elmo W Wiikinkoski,Junhua Xu,Risto Koivula,Iiro Rautsola

doi:10.3390/chemengineering4010014

Abstract

The trivalent lanthanide-actinide separations are a major challenge in reprocessing of nuclear fuels. To achieve this, commonly organic extractants and solvents are utilized in elaborate processes. Here we report a simple new method that can perform a supportive or alternative role. A nanocrystalline α-zirconium phosphate ion exchanger was utilized for Eu(III)/Am(III) column separation. Comprehensive preliminary studies were done using batch experiments to optimize the final separation conditions. The distribution coefficients for Eu were determined as a function of pH (from 0 to 3) and salinity (Na, Sr). The distribution coefficients for Am were determined as a function of pH, and Eu concentration, from 1:40 to 10,000:1 Eu:Am molar ratio. The exchanger always preferred Eu over Am in our experimental conditions. Separation factors (Eu:Am) of up to 400 were achieved in binary Eu-Am solution in pH 1. The breakthrough capacity was determined in dynamic column conditions using Eu: 0.3 meq∙g−1, which is approximately 4% of the theoretical maximum capacity. Two types of hot column separation tests were conducted: (i) binary load (selective Am elution), and (ii) continuous equimolar binary feed. In both cases separation was achieved. In (i), the majority (82% of the recovered 93%) of Am could be purified from Eu with extremely high 99.999% molar purity, while alternatively even more (95% of the recovered 93%) at a lower purity of 99.7 mol %. In (ii), up to 330 L∙kg−1 of the equimolar solution per mass of the exchanger could be treated with Am purity above 99.5 mol % in the total eluate. Alternatively, up to 630 L∙kg−1 above 95 mol %, or up to 800 L∙kg−1 above 90 mol % purities.

Highlights

Despite decades of study, the best possible choices for uranium-based used nuclear fuel (UNF) processing and storage are unclear, and the strategy for handling it still remains an open question in many countries
Selective technologies are crucial for this separation, as these lanthanides with their large neutron cross-sections would act as efficient neutron poisoners if still present in the fabricated minor actinide containing fuels or transmutation targets
The highest separation factor reported here (400 in pH 1) is the highest we have recorded so far for any of our α- or γ-zirconium phosphates, or titanium phosphates synthesized in the past

Summary

Introduction

The best possible choices for uranium-based used nuclear fuel (UNF) processing and storage are unclear, and the strategy for handling it still remains an open question in many countries. Uranium and plutonium recovery from UNF by the hydrometallurgical PUREX-process is the only time-proven technology used in industrial scales and in commercial use [1]. Both hydro- and pyrometallurgical, are maturing in laboratory use and are meant for the recovery of uranium and plutonium, but that of the other fissile actinides as well. In the international partitioning and transmutation (P&T) research effort, it is recognized that in addition to the recovery of uranium and plutonium, the recovery and transmutation of minor actinides americium, curium and neptunium decreases radiotoxicity and heat generation of the UNF, lowering the requirements for disposal of the remainder. Selective technologies are crucial for this separation, as these lanthanides with their large neutron cross-sections would act as efficient neutron poisoners if still present in the fabricated minor actinide containing fuels or transmutation targets

Methods

Results

Conclusion