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Abstract <p>The extractive recovery of protactinium from irradiated thorium nitrate as applied to the conditions of the fusion neutron source blanket was studied using <i>n</i>-octanol and diisobutylcarbinol as extractants. The degree of recovery of <sup>233</sup>Pa from a solution of irradiated thorium nitrate (600 g/dm<sup>3</sup>) in 3 М nitric acid in one contact exceeded 80%, and the separation factors from uranium and thorium were higher than 600 and 130000, respectively. To reduce the content of <sup>232</sup>U in <sup>233</sup>U (below 5 ppm), a flowsheet involving short irradiation of thorium nitrate and continuous extractive recovery of protactinium was suggested.</p>

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Recovery of Protactinium from Irradiated Thorium Nitrate by Solvent Extraction

  • 期刊论文

  • 发表日期：

    2025年4月1日

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  • M. L. Khokhlov,
  • A. P. Yakushin,
  • Yu. S. Fedorov,
  • V. V. Samonin,
  • V. D. Davidenko,
  • A. A. Kovalishin,
  • K. V. Pavlov,
  • Yu. E. Titarenko,
  • A. S. Kirsanov,
  • A. A. Arkhipov,
  • M. V. Kotel’nyi

摘要

Abstract

The extractive recovery of protactinium from irradiated thorium nitrate as applied to the conditions of the fusion neutron source blanket was studied using n-octanol and diisobutylcarbinol as extractants. The degree of recovery of 233Pa from a solution of irradiated thorium nitrate (600 g/dm3) in 3 М nitric acid in one contact exceeded 80%, and the separation factors from uranium and thorium were higher than 600 and 130000, respectively. To reduce the content of 232U in 233U (below 5 ppm), a flowsheet involving short irradiation of thorium nitrate and continuous extractive recovery of protactinium was suggested.