The Separation of ²⁴¹Am from Aged Plutonium Dioxide for use in Radioisotope Power Systems

¹ National Nuclear Laboratory, Sellafield, Seascale, Cumbria, UK
² European Commission, Joint Research Centre, P.O. Box 2340, 76125 Karlsruhe, Germany
³ European Space Agency, ESTEC TEC-EPS, PO Box 299 - 2200 AG Noordwijk, The Netherlands

Email: mark.sarsfield@nnl.co.uk

Abstract

Electrical power sources used in outer planet missions are a key enabling technology for data acquisition and communications. State–of-the-art power sources generate electricity from alpha decay of ²³⁸Pu via thermoelectric conversion. However, production of ²³⁸Pu requires specialist facilities including a nuclear reactor, a source of ²³⁷Np for target irradiation and hotcells to chemically separate neptunium and plutonium within the irradiated targets. These specialist facilities are expensive to build and operate, so naturally, a more economical alternative is attractive to the industry. Within Europe 241Am is considered a promising alternative heat source for radioisotope thermoelectric generators (RTGs) and radioisotope heating units (RHUs). As a daughter product of ²⁴¹Pu decay, ²⁴¹Am exists in 1000 kgs quantities within the UK civil plutonium stockpile.

A chemical separation process is required to extract the ²⁴¹Am in a pure form and this paper describes the AMPPEX process (Americium and Plutonium Purification by Extraction), successfully developed over the past five years to isolate ²⁴¹Am in high yield (> 99%) and to a high purity (> 99%).

The process starts by dissolving plutonium dioxide in nitric acid with the aid of a silver(II) catalyst, which is generated electrochemically. The solution is then conditioned and fed to a PUREX type solvent extraction process, where the plutonium is separated from the americium and silver. The plutonium is converted back to plutonium dioxide and the americium is fed forward to a second solvent extraction step. Here the americium is selectively extracted leaving the silver in the aqueous phase. The americium is stripped from the solvent and recovered from solution as americium oxalate, which is calcined to give americium dioxide as the final product. This paper will describe the development of the separation process over a series of six solvent extraction separation trials using centrifugal contactors. The material produced (~ 4g ²⁴¹Am) was used to make ceramic pellets to establish the behaviour of americium oxide material under high temperature (1450°C) sintering conditions.

The chemical separation process is now demonstrated at concentrations expected on the full scale facility taking this process to TRL 4-5.