Method for reprocessing spent nuclear fuel and centrifugal extractor therefor

Patent metadata

Field	Value
Publication number	US-9666315-B2
Application number	US-53360509-A
Country	US
Kind code	B2
Filing date	Jul 31, 2009
Priority date	Aug 12, 2008
Publication date	May 30, 2017
Grant date	May 30, 2017

Abstract

A spent nuclear fuel is reprocessed by dissolving a spent nuclear fuel in an aqueous nitric acid solution and separating and recovering nuclides contained in the resulting fuel solution by solvent extraction. A spent nuclear fuel reprocessing method includes: an electrolytic valence adjustment step in which nuclides contained in the fuel solution is electrolytically reduced without removing fission products or minor actinides until valence of plutonium is at a level at which solvent extraction efficiency is low by using the valence of plutonium contained in the fuel solution as a parameter; and a nuclide separation step in which, by using an extraction solvent which extracts uranium contained in the fuel solution, uranium is distributed from the fuel solution subjected to the electrolytic valence adjustment step to the extraction solvent.

First claim

What is claimed is:

1. An apparatus for separating uranium (“U”) from plutonium (“Pu”), fission products (“FP”), neptunium (“Np”), and other minor actinides (“MA”) comprising:
a transfer canal connected to a solution inlet on
an electrolytic reduction unit that is structurally integrated with a main body casing and a rotor casing of a centrifugal extraction unit, connected to a feed port to
the centrifugal extraction unit comprising a mixing space, a rotor casing, and separate discharge channels for a low density solvent phase and a higher density aqueous phase;
wherein the transfer canal is configured to feed an aqueous nitric acid solution containing U, Pu, fission products and minor actinides including Np, Am, and Cm via the solution inlet to the electrolytic reduction unit;
wherein the electrolytic reduction unit is configured to receive the aqueous nitric acid solution via the solution inlet, comprises corrosion-resistant electrodes configured to electrolytically reduce Pu contained in the aqueous nitric acid solution to a valence of 3 without separating it from the aqueous nitric acid solution, and configured to feed the electrolytically reduced aqueous nitric acid solution to the centrifugal extraction unit;
wherein the centrifugal extraction unit comprises a main body casing forming an outer shell that houses a rotor casing and contains a mixing space defined by the outer wall of the rotor casing and the inner wall of the main body casing;
wherein the mixing space is configured to mix the reduced aqueous nitric acid solution with a phase-separating extraction solvent comprising tributyl phosphate diluted by dodecane, and
wherein the centrifugal extraction unit is configured to centrifugally phase-separate a mixture of the reduced aqueous nitric acid solution and the extraction solvent and to discharge a low density solvent phase containing U via the discharge channel for low density solvent phase and to discharge the reduced aqueous nitric acid solution that retains Pu and minor actinides, including Np, Am and Cm, via the high density discharge channel.

2. The apparatus of claim 1 , wherein the main body casing of the centrifugal extractor is negatively-charged and the rotor casing of the centrifugal extractor is positively-charged.

3. The apparatus of claim 1 , wherein the main body casing of the centrifugal extractor is positively-charged and the rotor casing of the centrifugal extractor is negatively-charged.

4. A method for reprocessing a spent nuclear fuel comprising:
dissolving spent nuclear fuel in an aqueous nitric acid solution,
feeding said dissolved spent nuclear fuel into the transfer canal of the apparatus of claim 1 ; and
separating and recovering nuclides contained in the resulting fuel solution by solvent extraction in said apparatus.

5. A method for reprocessing a spent nuclear fuel comprising:
dissolving spent nuclear fuel in an aqueous nitric acid solution,
feeding said dissolved spent nuclear fuel into the transfer canal of the apparatus of claim 2 ; and
separating and recovering nuclides contained in the resulting fuel solution by solvent extraction in said apparatus.

6. A method for reprocessing a spent nuclear fuel comprising:
dissolving spent nuclear fuel in an aqueous nitric acid solution,
feeding said dissolved spent nuclear fuel into the transfer canal of the apparatus of claim 3 ; and
separating and recovering nuclides contained in the resulting fuel solution by solvent extraction in said apparatus.

Assignees

• Mizuguchi Koji
• Fujita Reiko
• Fuse Kouki
• Nakamura Hitoshi
• Utsunomiya Kazuhiro
• Tanaka Nobuhiko
• Toshiba Kk

Inventors

• Mizuguchi Koji
• Fujita Reiko
• Fuse Kouki
• Nakamura Hitoshi
• Utsunomiya Kazuhiro
• Tanaka Nobuhiko

Classifications

• B01D11/0492

  Applications, solvents used · CPC title

• G21C19/46 Primary

  Aqueous processes {, e.g. by using organic extraction means, including the regeneration of these means} · CPC title

• B01D17/06

  Separation of liquids from each other by electricity · CPC title

• B01D11/0434

  comprising rotating mechanisms, e.g. mixers, rotational oscillating motion, mixing pumps · CPC title

• B01D17/047

  with separation aids · CPC title