Recovery Process of Actinides from Genuine Spent Nuclear Fuel using TODGA and BTBP ...

146 Pages·2008·12.37 MB·English

Recovery Process of Actinides

from Genuine Spent Nuclear Fuel

using TODGA and BTBP Extractants

THESIS FOR THE DEGREE OF DOCTOR OF PHILOSOPHY

Daniel Magnusson

JRC-ITU-TN-2008/70
The mission of ITU is to provide the scientific foundation for the protection of the European
citizen against risks associated with the handling and storage of highly radioactive material.
ITU’s prime objectives are to serve as a reference centre for basic actinide research, to
contribute to an effective safety and safeguards system for the nuclear fuel cycle, and to study
technological and medical applications of radionuclides/actinides.

Report No: JRC-ITU-TN-2008/70
Classification: unclassified
Type of Report: Thesis
Unit: Nuclear Chemistry
Action No: ANFC

Name Date Signature
reviewed by the

project coordinator
approved by the

project leader
approved by the

head of unit
released by the

director

European Commission
Joint Research Centre
Institute for Transuranium Elements

Contact information
Address: Daniel Magnusson
E-mail: [email protected]
Tel.: [phone]

http://itu.jrc.ec.europa.eu/
http://www.jrc.ec.europa.eu/

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© European Communities, 2008
THESIS FOR THE DEGREE OF DOCTOR OF PHILOSOPHY

Recovery Process of Actinides from Genuine Spent Nuclear
Fuel using TODGA and BTBP Extractants

DANIEL MAGNUSSON

Nuclear Chemistry
Department of Chemical and Biological Engineering
CHALMERS UNIVERSITY OF TECHNOLOGY
Göteborg, Sweden 2008
Recovery Process of Actinides from Genuine Spent Nuclear
Fuel using TODGA and BTBP Extractants

Daniel Magnusson

Nuclear Chemistry
Department of Chemical and Biological Engineering
CHALMERS UNIVERSITY OF TECHNOLOGY

Abstract
During the last decades a growing concern about greenhouse gas emissions from fossil
fuels has arisen. Nuclear power is an energy source with a low contribution to the
greenhouse effect and is in this sense therefore a better alternative for electricity
production. The waste from nuclear power is however highly radiotoxic and has to be
stored for more than 100000 years until the radiotoxicity has decreased to an acceptable
level. With partitioning and transmutation the purpose is to separate the nuclides that
contribute most to the long-term radiotoxicity and to transform them into short lived or
stable nuclides. If partitioning and transmutation are successfully applied the storage time
for nuclear waste can be reduced to less than 1000 years.
Essential for the partitioning and transmutation is an efficient separation of the minor
actinides. In this work, group separation of the trivalent lanthanides and actinides from a
PUREX raffinate has been demonstrated by means of solvent extraction, using the
TODGA extracting agent. Excellent separation was obtained and the recoveries of the
actinides exceeded 99.9%. Separation of the actinides from the lanthanides was also
carried out, using the CyMe -BTBP extracting agent. Results achieving high actinide
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recovery of more than 99.9%, were acquired as well as efficient separation from the
lanthanides.
A comparison between alpha and gamma radiolysis was performed with the CyMe -
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BTBP organic phase. The study showed that the solution was more sensitive to the
gamma radiation. The effect of varying alpha dose rates was also investigated but no
significant difference was observed in the interval tested (50-1000 Gy/h). Burnup
calculations, done to estimate the metal concentration in a process, together with loading
experiments show however that the loading capacity of the CyMe -BTBP organic phase
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is very limited and needs to be improved if fuels with higher minor actinide content are to
be treated.
Two computer programs for process calculations were developed. The programs were
successfully used to calculate the extraction profiles for the CyMe -BTBP test.
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Keywords: solvent extraction, transmutation, actinides, lanthanides, TODGA, BTBP,
radiolysis, kinetics

List of publications

This thesis is based on the work published in the following papers. They will be referred
to as Paper I-V.

I. Modolo, G., Asp, H., Vijgen, H., Malmbeck, R., Magnusson, D., Sorel, C.:
Demonstration of a TODGA-Based Continuous Counter-Current Extraction
Process for the Partitioning of Actinides from a Simulated PUREX Raffinate, Part
II: Centrifugal Contactor Runs. Solvent Extraction and Ion Exchange, 26(1), 62-
76 (2008).

II. Magnusson, D., Christiansen, B., Glatz J-P., Malmbeck, R., Modolo, G., Serrano-
Purroy, D., Sorel, C.: Demonstration of a TODGA based Extraction Process for
the Partitioning of Minor Actinides from a PUREX Raffinate, Part III: Centrifugal
Contactor Run using Genuine Fuel Solution. Solvent Extraction and Ion
Exchange, 26(6), (2008).

III. Magnusson, D., Christiansen, B., Glatz J-P., Malmbeck, R.: Investigation of the
radiolytic stability of a CyMe -BTBP based SANEX solvent. Submitted to
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Radiochimica Acta (2008).

IV. Magnusson, D., Christiansen, B., Glatz J-P., Malmbeck, R., Modolo, G., Serrano-
Purroy, D., Sorel, C.: Towards an optimized flow sheet for a SANEX
demonstration process using centrifugal contactors. Accepted for publication in
Radiochimica Acta (2008).

V. Magnusson, D., Christiansen, B., Foreman, M. R. S., Geist, A., Glatz J-P.,
Malmbeck, R., Modolo, G., Serrano-Purroy, D., Sorel, C.: Demonstration of a
SANEX process in centrifugal contactors using the CyMe -BTBP molecule on a
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genuine fuel solution. Accepted for publication in Solvent Extraction and Ion
Exchange (2008).

Contribution Report

I. Designing experiment, experimental work and evaluation together with
other authors, reviewing of the paper.

II, III, V. All the experimental work, planning and evaluations, writer of the paper.

IV. Planning and main part of evaluations, all experimental work in ITU and
experimental work in FZJ together with other author, main writer of the
paper.

Table of Contents
1. Introduction...............................................................................................................3
1.1 Objectives...........................................................................................................4
2. Background...............................................................................................................5
2.1 Nuclear power.....................................................................................................5
2.2 Spent fuel............................................................................................................5
2.3 Transmutation.....................................................................................................6
2.4 Partitioning..........................................................................................................6
2.5 Solvent extraction.............................................................................................10
2.6 Process..............................................................................................................11
2.7 PUREX.............................................................................................................12
2.8 DIAMEX-TODGA...........................................................................................12
2.9 SANEX.............................................................................................................13
2.10 BTBP radiolysis................................................................................................14
2.11 BTBP extraction kinetics..................................................................................15
3. Process optimisation and modelling......................................................................17
3.1 Program 1: Equilibrium distribution ratios.......................................................17
3.2 Program 2: Deviation from equilibrium...........................................................19
3.3 Process optimisation procedure........................................................................20
4. Experimental procedures.......................................................................................23
4.1 Hot cells............................................................................................................23
4.2 Centrifugal contactor system............................................................................23
4.2.1 FZJ............................................................................................................23
4.2.2 ITU............................................................................................................23
4.3 PUREX process................................................................................................24
4.4 TODGA process................................................................................................25
4.5 BTBP radiolysis................................................................................................26
4.5.1 Dose rate and metal loading calculations................................................26
4.5.2 Radiolysis and loading experiments.........................................................27
4.6 BTBP extraction kinetics..................................................................................28
4.7 BTBP demonstration process............................................................................29
4.8 Analysis methods and sample preparation........................................................30
5. Results and discussion............................................................................................31
5.1 PUREX.............................................................................................................31
5.2 TODGA.............................................................................................................32
5.2.1 Spiked tests................................................................................................32
5.2.2 Hot test......................................................................................................34
5.3 BTBP radiolysis................................................................................................40
5.3.1 Dose rate and metal loading calculations................................................40
5.3.2 Radiolysis experiments..............................................................................41
5.4 BTBP extraction kinetics..................................................................................45
5.5 BTBP demonstration test..................................................................................48
5.6 BTBP modelling...............................................................................................51
6. Conclusions..............................................................................................................55
7. Future work.............................................................................................................57
8. Acknowledgements.................................................................................................59
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9. References................................................................................................................61
Appendix A: Glossary of abbreviations........................................................................71

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1. Introduction

With a steadily growing human population in the world, the demand for energy is
constantly increasing. The fast economical and industrial development of countries like
China and India is also enhancing the rate of demand. With an increased energy
production, environmental issues are a major concern. Today most energy production
comes from burning of fossil fuels which leads to the emission of CO . During the last
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decades this addition of CO into the atmosphere has come into focus, because it is a
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greenhouse gas, and programs with objectives to reduce these emissions have been
started (e.g. the Kyoto protocol [UNI 97]). To fulfil these objectives alternative energy
sources with low or no CO emission have to replace some of the fossil fuel. Examples of
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these are biomass, solar and wind energy, water power, geothermal energy, nuclear
power etc.
Today about 15% of the worlds electricity comes from nuclear power [IAE 08].
Nuclear power has low CO emissions [LEN 08, FTH 07] compared to fossil fuels and is
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therefore not contributing as much to the greenhouse effect. However, highly radiotoxic
spent nuclear fuel is produced, which is a serious environmental issue. This waste, with
no further treatment, has to be stored for a long time until the radiotoxicity can be
considered to be at an acceptable level.
The spent fuel can go through reprocessing where uranium and plutonium are
recovered and then might be used to make new fuel. This would reduce the amount of
highly active waste and at the same time increase the energy output per ton of natural
uranium although the remaining spent fuel after reprocessing is still highly radiotoxic.
The main remaining long-term radiotoxicity in the waste comes from the minor actinides
[MAG 03]. By partitioning and transmutation of these elements the aim is to transform
them into short-lived or stable elements in a dedicated reactor. The final waste will still
have to be stored in a deep repository but for a much shorter time than untreated spent
nuclear fuel.
An efficient separation of americium and curium from the other elements in the waste
is important for the partitioning and transmutation concept. The PUREX process [RYD
92] has been used for many years to recover uranium and plutonium from spent fuel.
With some modification, neptunium, technetium and iodine can also be separated by the
PUREX method. Processes for further partitioning of the PUREX raffinate are under
development. One proposed concept is to divide the separation into two steps [MAD 07],
both using solvent extraction. First the trivalent lanthanides and trivalent actinides are co-
extracted, leaving the remaining fission products in the raffinate. In the second step the
actinides are separated from the lanthanides.
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