Influence of the gamma source on the radiolytic stability of N,N,N′,N′-tetra-n-octyl-diglycolamide (TODGA)

Ken Verguts; Karen Van Hecke; Jan Jordens; Stefan Voorspoels; Thomas Cardinaels

doi:10.1515/ract-2024-0329

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Influence of the gamma source on the radiolytic stability of N,N,N′,N′-tetra-n-octyl-diglycolamide (TODGA)

Ken Verguts , Karen Van Hecke , Jan Jordens , Stefan Voorspoels and Thomas Cardinaels

Published/Copyright: November 5, 2024

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From the journal Radiochimica Acta Volume 113 Issue 2

Abstract

In nuclear reprocessing, hydrolytic and radiolytic stability of ligands, extractants, diluents and solvents is of particular importance. The strength of extraction systems can only be fully exploited when process conditions are predictable and essential molecules, which are in close proximity to radioactive isotopes, are resistant to the radiation. During the development of novel extraction systems, testing of the radiolytic stability is often evaluated by exposing the molecules to high energetic gamma rays. In this work, the influence of the source of gamma rays is evaluated. TODGA was irradiated from 0–500 kGy using different sources of gamma rays, and its degradation was followed using UPLC-HRMS. Pool-type gamma irradiation facilities BRIGITTE B (Co-60) and GEUSE II (spent nuclear fuel) at SCK CEN were employed to serve as gamma sources. In this work, it is found that there is no significant influence on TODGA degradation which can be attributed to the origin of the gamma source.

Keywords: BRIGITTE B; gamma irradiation; GEUSE II; radiolytic stability; TODGA

1 Introduction

Nuclear energy is a reliable, low-cost energy source which has zero-carbon emissions during its operation. However, radioactive waste remains one of the key concerns. Highly radiotoxic spent nuclear fuel remains problematic for millions of years and requires special attention. Pending the selection and commissioning of a deep geological repository for highly active radioactive waste, spent nuclear fuel is temporally stored in interim storage facilities. Reprocessing of spent nuclear fuel could be a solution to either recover the fissile elements for recycling, or either reducing the volume of the high level radioactive waste, and thus decreasing the gallery length. In spent nuclear fuel, the main contributors to the overall radiotoxicity and heat load are plutonium, minor actinides and some fission products like cesium and strontium. Where the majority of uranium and plutonium is already recovered via the PUREX process on an industrial scale, ¹ ^, ² the minor actinides and/or the heat-generating fission products strontium and cesium are still to be removed from the remaining highly active PUREX raffinate (HAR).

In Europe, advanced partitioning strategies have been developed and tested in several projects such as ACSEPT, ³ SACSESS, ⁴ GENIORS ⁵ and PATRICIA. One challenging strategy is the development of novel extractants and complexants to selectively extract metal ions. In these novel aqueous partitioning processes, where immiscible organic phases are contacted with the aqueous acidic HAR solution, the robustness of these organic phases is of prime importance for its implementation in industry. Here, the radiolytic stability of N,N,N′,N′-tetra-n-octyl-diglycolamide (TODGA) in dodecane has been evaluated under different conditions. TODGA has been chosen in this comparative study since it is widely recognized as the workhorse for americium partitioning.

However, only few gamma irradiation facilities can be used for this type of radiation stability research. These can be divided in either pool-type facilities or dry sources. The first group of facilities have their gamma emitting sources submerged in water: Náyade facility (CIEMAT, Madrid, Spain), ⁶ ^, ⁷ ^, ⁸ ^, ⁹ ^, ¹⁰ Calliope facility (ENEA-Casaccia Research Center Rome, Italy), ¹¹ Sandia National Laboratories (Albuquerque, New Mexico, USA), Gamma Irradiation Facility at HFIR (ORNL, Tennessee, USA) and BRIGITTE B, RITA, GEUSE II facilities (SCK CEN, Mol, Belgium). ¹² ^, ¹³ ^, ¹⁴ In a second group, dry facilities are in use: Chalmers University of Technology (Sweden, Co-60 Gamma cell source), ¹⁵ Idaho National Lab (INL, USA, Co-60 Gamma cell source), ⁷ ^, ¹⁶ ^, ¹⁷ ^, ¹⁸ CEA Marcoule (GRS-D1 Gamma-Service-Medical-GmbH, France), ¹⁹ Takasaki Advanced Radiation Research Institute (JAEA, Japan), ²⁰ ^, ²¹ ^, ²² and the Irradiation Preservation Technology Key Laboratory of Sichuan Province (Chengdu, China). ²³ The origin of the gamma rays in the majority of the abovementioned facilities is the decay of Co-60 (T _1/2 = 5.27 y). However, in the Gamma Irradiation Facility at HFIR (ORNL) and the GEUSE II facility at SCK CEN, the sources are spent nuclear fuel sources. The origin of the dominant gamma rays in spent nuclear fuel is dependent on the time after discharge, which makes the calculation of the actual dose rates difficult to predict without proper dosimetry. In this study, we aim to understand the influence of the gamma field (Co-60 sources versus spent nuclear fuel sources). In addition, the radiolytic stability study was performed both in the presence and absence of a nitric acid solution.

2 Experimental

2.1 Chemicals

TODGA solutions were prepared similar to Verlinden et al. ¹⁴ A 0.050 mol L⁻¹ TODGA (Technocomm Ltd., Wellbrae, UK) stock solution was prepared in dodecane (VWR, Belgium). From this stock solution, several 500 µL samples were prepared for gamma irradiations in GEUSE II and BRIGITTE B (SCK CEN, Mol, Belgium). Additional samples were prepared containing 500 µL of the 0.050 mol L⁻¹ TODGA stock solution, in contact with 500 µL 2.5 mol L⁻¹ trace metal grade nitric acid (VWR, Belgium).

2.2 Gamma irradiation

The gamma irradiations were conducted in GEUSE II (Gamma Experiments Using Spent fuel Elements) or BRIGITTE B (Big Radius Installation under Gamma Irradiation for Tailoring and Testing Experiments, position B) at the SCK CEN site in Mol, Belgium (see Figure 1A and B). ¹² GEUSE II is a pool-type gamma irradiation facility where a broad gamma spectrum is emitted by spent nuclear fuel sources, discharged from the Belgian Reactor 2 (BR2). Spent nuclear fuel sources in GEUSE II have a typical burn-up of 56–57 % and a cooling time between 100 days and 3 years. Therefore, it is presumed that the spectrum is dominated by the decay of Ba-137m. Its mother isotope, Cs-137 (T _1/2 = 30.19 years) decays to short-lived Ba-137m (T_1/2 = 2.6 min) via emitting beta rays at maximum energies of 512 keV (94.0 %) and 1,175.6 keV (6.0 %), respectively. Ba-137m then emits gamma rays with an energy of 661.7 keV. ²⁴ Samples irradiated in GEUSE II received an average dose rate of 1.4 kGy h⁻¹. BRIGITTE B is also an underwater pool-type facility which makes use of old control rods of BR2. Here, gamma rays are emitted by Co-60 sources (E _γ1 = 1,173.2 keV and E _γ2 = 1,332.5 keV). An average dose rate of approximately 7.0 kGy h⁻¹ was measured. Prior to each irradiation, dosimetry was performed allowing to understand the exact dose rate at the sample’s individual positions. Harwell Amber Perspex Dosimeters (Type 3042, λ = 603 nm, in GEUSE II) or Harwell Red Perspex Dosimeters (Type 4034, λ = 640 nm, in BRIGITTE B) were irradiated for 2 h and measured by UV-VIS spectrometry. The UV-VIS measurements were performed using a Shimadzu UV-1800 spectrophotometer with a spectral bandwidth of 1 nm, identical to the report of Verlinden et al. ¹⁴ Dose rates were used as calculated from the dosimetry, no additional corrections were made for the diluent. After dosimetry, samples were placed on dedicated positions in the sample holder (see Figure 1C), which in turn was placed in a large pressurize container (Figure 1D). The container was positioned to locate the samples in between the gamma sources. The irradiation time of the individual samples was calculated with respect to the individual dose rates to target absorbed doses of 100, 200, 300, 400 or 500 kGy. Since the dose distribution among the sample holder is not uniform, the absorbed dose was recalculated for each sample (see Table 1) and plotted accordingly. The reported coefficients of variation are within 2.5 % and 2 % for respectively the Amber and Red Perspex dosimeters.

Figure 1:

Pictures of the GEUSE II facility (A), the BRIGITTE B facility (B), the sample holder (C) and pressurized container (D).

Table 1:

Overview of the samples’ irradiation conditions.

Facility (γ source)	Matrix	Nominal absorbed dose (kGy)	Dose rate (kGy h⁻¹)	Irradiation time (days, h, min)	Actual absorbed dose (kGy)	Uncertainty on absorbed dose (kGy)
BRIGITTE B (Co-60)	TODGA, dodecane	100	7.15	14 h 53 min	106	2
		200	7.05	1 day 4 h 2 min	198	4
		300	7.05	1 day 15 h 40 min	280	6
		400	7.15	2 days 6 h 29 min	390	8
		500	6.60	2 days 19 h 42 min	447	9
	TODGA, dodecane, HNO₃	100	7.25	14 h 53 min	108	2
		200	7.05	1 days 4 h 2 min	198	4
		300	7.05	1 days 15 h 40 min	280	6
		400	7.25	2 days 6 h 29 min	395	8
		500	6.75	2 days 19 h 42 min	457	9
GEUSE II (spent nuclear fuel)	TODGA, dodecane	100	1.40	2 days 22 h 55 min	99	2
		200	1.42	5 days 17 h 57 min	195	5
		300	1.34	8 days 13 h 11 min	276	7
		400	1.33	11 days 8 h 13 min	363	9
		500	1.41	14 days 7 h 8 min	485	12
	TODGA, dodecane, HNO₃	100	1.40	2 days 22 h 55 min	99	2
		200	1.44	5 days 17 h 57 min	199	5
		300	1.37	8 days 13 h 11 min	282	7
		400	1.33	11 days 8 h 13 min	363	9
		500	1.42	14 days 7 h 8 min	487	12

2.3 UPLC-HRMS analysis

Samples were dissolved in dodecane. Some samples presented precipitation upon storage. Before subsampling and preparation for analysis, samples were heated at 40 °C during sonication for 30 min for complete dissolution and homogenization. Samples were further 1,000-fold diluted in isopropylalcohol (IPA) to obtain concentrations in the method working range.

Analysis was performed using UPLC-HRMS (Thermo, Q-Exactive). Chromatographic resolution was obtained using an Acquity BEH C18 column (2.1 × 100 mm, 1.7 µm) using a mobile phase of water:acetonitrile (1:1; v:v; A) and acetonitrile:isopropylalcohol:water (10:88:2; v:v:v; B), both containing ammonium formate buffer. A flow rate of 0.4 mL min⁻¹ was used on a 50 °C thermostated column. The MS was operated in ESI positive mode at full scan with a resolution of 140,000; monitoring the exact mass of m/z 580.55429. Pure TODGA was used as reference compound for calibration.

Full prior method validation was not performed. On-the-fly quality assurance samples verified the data is fit-for-purpose. The following measures were implemented. (i) Each sample was measured in triplicate, (ii) each replicate series was analyzed in random order, (iii) throughout the measurement batch, 7 QC standard solutions were measured and no trend was observed and the recorded RSD was 1.5 %, (iv) three samples were measured in duplicate to assess method precision during analysis, resulting in an RSD from 0.3 to 2.1 %, and finally (v) three samples were spiked at 40 μg L⁻¹ to verify analyte recovery during measurement, resulting in a recovery of 103–106  %. The estimated expanded measurement uncertainty is estimated at 7 % (k = 2). Based on this outcome, the method used was considered fit-for-purpose.

3 Results and discussion

The concentration of TODGA in dodecane was determined as a function of the absorbed dose and compared to the reference molecule TODGA (see Figure 2, left plot). From these concentrations, the dose constants (d, in Gy⁻¹) were calculated from the linear fit of the natural logarithm of the TODGA concentration (mol L⁻¹) as a function of the absorbed dose (kGy). ²⁵ The derived dose constants are given in Table 2. It is observed that the dose constant for irradiations performed in BRIGITTE B (4.8 ± 0.2) × 10⁻⁶ Gy is equal to the dose constant for irradiations performed in GEUSE II (4.7 ± 0.2) × 10⁻⁶ Gy, which is represented as an overlay of the trend lines as shown in Figure 2. These results are in line with similar experiments performed by Verlinden et al., ¹⁴ where TODGA was irradiated in the BRIGITTE B facility and dose constants of (4.0 ± 0.4) × 10⁻⁶ Gy are reported. Also, Zarzana et al. ¹⁶ and Sugo et al. ²² published comparable dose constants to describe TODGA degradation, respectively (4.1 ± 0.3) × 10⁻⁶ Gy and 4.5 × 10⁻⁶ Gy, which were obtained by using dry Co-60 Gamma cell sources. Due to the relatively low concentration of TODGA, the majority of the radiolytic damage to TODGA is believed to be originating from dodecane radicals. These radicals are formed due to the energy transfer of gamma rays and thus radiolysis of the diluent, followed by a chemical reaction with the TODGA molecules. In previous works, where water was used as diluent, no influence was observed when the gamma energy was changed over similar ranges, ¹⁷ ^, ²⁶ which confirms our observations.

Figure 2:

Influence of the gamma spectrum on the radiolytic stability of TODGA in the absence (left) and presence (right) of nitric acid.

Table 2:

Summary of all dose constants determined in this set of experiments (×10⁻⁶ Gy⁻¹).

Dose constants, d (×10⁻⁶ Gy⁻¹)	Spent fuel	Co-60
0.05 mol L⁻¹ TODGA in dodecane	4.7 ± 0.2	4.8 ± 0.2
0.05 mol L⁻¹ TODGA in dodecane contacted with 2.5 mol L⁻¹ HNO₃	3.2 ± 0.2	3.1 ± 0.2

In Figure 2 (right), data are plotted to evaluate the TODGA stability as a function of absorbed dose, while the organic phase is contacted with 2.5 mol L⁻¹ HNO₃. Also for this dataset, both trendlines for BRIGITTE B and GEUSE II are overlaying, which is represented in similar dose constants: respectively (3.1 ± 0.2) × 10⁻⁶ Gy and (3.2 ± 0.2) × 10⁻⁶ Gy. The observed data are comparable to the values reported by Verlinden et al. ¹⁴ and Zarzana et al. ¹⁶ : (3.4 ± 0.2) × 10⁻⁶ Gy and (3.8 ± 0.3) × 10⁻⁶ Gy respectively. The irradiations and calculated dose constants confirm that the origin of gamma rays (Co-60 sources in BRIGITTE B versus spent nuclear fuel sources in GEUSE II), and the difference in dose rate, have no significant influence on the radiolytic stability of TODGA.

4 Conclusions

The radiolytic stability of TODGA was assessed using different irradiation facilities. The molar concentration of TODGA in dodecane, both in the presence and absence nitric acid, was evaluated as a function of the absorbed dose, and dose constants were calculated. It is found that the degradation of the extractant, was not affected by using two different facilities. Irradiation by either BRIGITTE B (Co-60 sources – E _γ1 = 1,173.2 keV and E _γ2 = 1,332.5 keV, with an average dose rate of 7.0 kGy h⁻¹) or GEUSE II (spent nuclear fuel sources, Ba-137m with an energy of 661.7 keV and an average dose rate of 1.4 kGy h⁻¹) has no significant influence and yields identical dose constants. In addition, lower dose constants are observed when a TODGA/dodecane phase is contacted with a nitric acid medium. Data obtained in this work are fully in line with results obtained by other research groups.

Corresponding author: Ken Verguts, Belgian Nuclear Research Center (SCK CEN), Institute for Nuclear Energy Technology, Boeretang 200, 2400 Mol, Belgium, E-mail: Ken.Verguts@sckcen.be

Acknowledgments

KV acknowledges the support of Davy Kaers and Lander Van Gestel for the manipulations and dosimetry of the BRIGITTE B and GEUSE II facilities.

Research ethics: Not applicable.
Informed consent: Not applicable.
Author contributions: The authors have accepted responsibility for the entire content of this manuscript and approved its submission.
Use of Large Language Models, AI and Machine Learning Tools: None declared.
Conflict of interest: The authors state no conflict of interest.
Research funding: This research was conducted with the support of the Energy Transition Fund through the ASOF project. Grant number: CO-90-18-4699-00.
Data availability: The raw data can be obtained on request from the corresponding author.

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Received: 2024-07-04

Accepted: 2024-10-16

Published Online: 2024-11-05

Published in Print: 2025-02-25

This work is licensed under the Creative Commons Attribution 4.0 International License.

Articles in the same Issue

https://doi.org/10.1515/ract-2024-0329

Keywords for this article

BRIGITTE B; gamma irradiation; GEUSE II; radiolytic stability; TODGA

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