Article pubs.acs.org/ac
Preparation of Microvolume Anion-Exchange Cartridge for Inductively Coupled Plasma Mass Spectrometry-Based Determination of 237Np Content in Spent Nuclear Fuel Shiho Asai,*,† Yukiko Hanzawa,† Miki Konda,† Daisuke Suzuki,† Masaaki Magara,† Takaumi Kimura,† Ryo Ishihara,‡ Kyoichi Saito,‡ Shinsuke Yamada,§ and Hideyuki Hirota§ †
Nuclear Science and Engineering Center, Japan Atomic Energy Agency, Ibaraki 319-1195, Japan Department of Applied Chemistry and Biotechnology, Chiba University, Inage, Chiba 263-8522, Japan § INOAC Corporation, Atsuta, Nagoya 456-0054, Japan ‡
ABSTRACT: Microvolume anion-exchange porous polymer disk-packed cartridges were prepared for Am/Np separation, which is required prior to the measurement of Neptunium-237 (237Np) with inductively coupled plasma mass spectrometry (ICPMS). Disks with a volume of 0.08 cm3 were cut out from porous sheets having anion-exchange-group-containing polymer chains densely attached on the pore surface. Four different amine-based groups, N,N-dimethylaminoethyl methacrylate, trimethylammonium, diethylamine, and triethylenediamine (TEDA), were selected as the anion-exchange groups to be introduced into the porous sheets. The separation performances of Am/Np were evaluated using a standard solution of 243Am, which had the same activity as its daughter nuclide 239Np in secular equilibrium. 239Np recovery of close to 100% with practically no contamination of 243Am was achieved using the TEDA-introduced disk-packed cartridge. The time to elute 239Np from the cartridge was approximately 40 s. The TEDA-introduced disk-packed cartridge was applied to the separation of Np from a spent nuclear fuel sample to confirm its separation performance. A known amount of 243Am (239Np) was added to the spent nuclear fuel sample solution to monitor the chemical yield of Np. The chemical yield of Np calculated from a measured concentration of 239Np was 90.4%. Am leakage in the Np-eluted solution was less than 1 ppt, corresponding to 0.001% of the original Am concentration in the sample. This indicates that no additional 239Np was produced by the decay of the 243Am remaining in the Np-eluted solution, thus providing a reliable chemical yield. U, which can cause a serious spectral interference involving the peak tail from the mass spectrum of 238U, was thoroughly removed with the TEDA cartridge, providing interference-free measurement of 237Np. The concentration of 237Np obtained by ICPMS was 718 ± 12 ng/mg-U, which agrees well with the theoretically calculated value. Compared with the conventional separation technique using commercially available anion-exchange resin columns, the time required to adsorb, wash, and elute Np using the TEDA- introduced disk-packed cartridge was reduced by 75%. eptunium-237 (237Np), an α-emitting nuclide with a long half-life of 2.144 × 105 years, can be found in spent nuclear fuel and high-level radioactive waste (HLW). A large quantity of 237Np is generated in nuclear reactors through the nuclear reaction of 235U and 238U in UO2 fuel. The quantity of 237 Np generated during nuclear reactions is around 500 g per metric ton of uranium in a typical pressurized water reactor (PWR), depending on fuel burn-up.1 To evaluate the long-term safety of a HLW repository, the 237Np content in spent nuclear fuel and HLW must be determined. Alpha spectrometry, inductively coupled plasma optical emission spectroscopy (ICP-OES), and inductively coupled plasma mass spectrometry (ICPMS) have been employed to determine the concentration of 237Np.2 Alpha spectrometry is one of the traditional measurement techniques for actinides and has been widely applied to nuclear3−5 and environmental6,7 samples. Because of its low background capability, alpha spectrometry is suitable for environmental samples, which
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© 2016 American Chemical Society
contain extremely low amounts of 237Np. However, the low specific radioactivity of 237Np prolongs the measurement time compared with that required for ICPMS. Furthermore, to prevent spectral interference, coexisting radionuclides in the samples must be completely removed before measurement. ICP-OES can be used for 237Np determination of samples that include Np at relatively higher concentrations.8 It provides multielemental analysis without chemical separation. However, the sensitivity is typically about 100−1 000 times lower than that of ICPMS. This implies that a sample which includes a “mg/kg” level of Np would be required for achieving a precise determination, leading to an overall increase in worker radiation exposure. Received: November 16, 2015 Accepted: February 12, 2016 Published: February 12, 2016 3149
DOI: 10.1021/acs.analchem.5b04330 Anal. Chem. 2016, 88, 3149−3155
Article
Analytical Chemistry
Various types of anion-exchange groups, for example, N,Ndimethylaminoethyl methacrylate (DMAEMA),24 trimethylammonium (TMA), 22 and diethylamino (DEA)18 can be introduced into the polymer chain grafted onto the pores of the polymers by radiation-induced graft polymerization and subsequent chemical modifications. A DEA-group-introduced porous polymer disk measuring 13 mm in diameter and 3 mm in thickness (0.4 cm3) was packed into a polypropylene cartridge and used for separating U, Pu, and Am from spent nuclear fuel.18 Two aspects need to be improved for applying such anionexchange-group-introduced cartridges to Am/Np separation: reduction of disk volume and enhancement of adsorption stability. To achieve rapid extraction of 239Np from the 243Am standard solution, miniaturizing the cartridge is effective because it reduces the volume of solution that passes through, thus shortening of the entire separation time. To guarantee that the reliable analytical data is acquired, complete removal of Am while Np is rigidly retained onto the cartridge is indispensable to prevent any inadvertent buildup of 239Np activity from the remaining 243Am in the disk. In the case of U-bearing sample such as spent nuclear fuel, the removal capacity of U is another important requirement for eliminating the spectral interference caused by the peak tail of 238U. In this study, DMAEMA-, TMA-, and DEA-introduced porous polymer sheets were prepared according to previously developed procedures.18,22,24 Triethylenediamine (TEDA), which has two anion-exchange groups per molecule, was also introduced to the polymer sheet to enhance the adsorption stability for Np(IV). Disks with a volume of 0.08 cm3 were cut out from each sheet and packed into an empty cartridge. To verify the applicability of the prepared products, the anionexchange disk-packed cartridge that exhibited the best separation performance was used for Am/Np separation from a spent nuclear fuel sample.
ICPMS is the most commonly applied mass spectrometric technique for the determination of long-lived radionuclides these days.9 The method is characterized by simultaneous multi-isotope measurement with low detection limits for almost all elements. ICPMS is advantageous for determining 237Np content in spent nuclear fuel samples, in that there is no major potential for interference with the exception of peak tailing of 238 10,11 U. The presence of Hg and Au may cause interference in the mass spectrum of 237Np, forming polyatomic species such as 40Ar197Au+, 38Ar199Hg+, and 36Ar197Hg+.13 Given that neither Hg nor Au is practically generated by nuclear fission, interferences driven by Ar-plasma-based polyatomic species are expected to be insignificant. Chemical separation is still necessary, mainly to remove U, which interferes with the mass spectrum at m/z 237 with the peak tail of 238U. In addition, the removal of the radioactive components such as 90Sr, 90Y, 137Cs, 137mBa, 238Pu, and 239Pu is preferable for minimizing workers’ exposure to radiation and preventing any radiation contamination. A large number of chemical separation methods based on ion-exchange column chromatography3,5,10,13 and extraction chromatography2,12,14,15 have been proposed. In concentrated nitric acid and hydrochloric acid, Np(IV) is strongly retained in anion-exchange resins with the quaternary ammonium group.12,15 Shinohara et al.5 separated Np(IV) from a spent nuclear fuel sample using two-step anion-exchange column chromatography. TEVA resin, which has an aliphatic quaternary ammonium extractant inside the pores of the base material, was also used to separate Np.16 Qiao et al.15 employed anion-exchange-resin-packed columns for developing an automated sequential injection system. Another separation procedure using TEVA resin columns combined with coprecipitation of MnO2, Np, and Pu for human urine samples was reported.13 In the proposed method, Pu(IV) and Np(IV) are retained on the anion-exchange group of TEVA resin with consistent sorption performance throughout the procedure. To take advantage of such similarities in chemical properties, 242Pu was utilized as a nonisotopic yield tracer for Np separation instead of being used as an Np isotopic tracer. Owing to its long half-life of 3.733 × 105 years or low specific activity, the determination of 242Pu is, in general, conducted using ICPMS along with 237Np. However, there exists a significant amount of 242Pu in spent nuclear fuel, which is comparable to the amount of 237Np1. If the initial amount of 242Pu in an original sample is unknown, 242Pu is unusable as a yield tracer for 237Np in spent nuclear fuel samples. Owing to the lack of a commercially available standard solution of 239Np, its parent nuclide 243Am, which is in secular equilibrium, is usually used as a 239Np source. The extraction of 239 Np from the 243Am standard solution is usually conducted immediately before use because the extracted 239Np rapidly decays. The entire separation procedure has to be completed within a day or two to measure 239Np concentration with sufficient reliability. Therefore, minimizing the separation time is a prerequisite for 237Np analysis. We have developed functionalized porous polymers that enable rapid separation of inorganic ions.17−23 Such functionalized porous polymers are characterized by high-speed separation involving minimized diffusion paths of the target ions to the functional groups of the pore surface. Such forced permeation of the liquid through the pores of the sheet provides a rapid mass-transfer field.
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EXPERIMENTAL SECTION Reagents. 243Am (and daughter 239Np) as well as 237Np standard samples in oxide form provided by Oak Ridge National Laboratory (ORNL) were dissolved in 7 M nitric acid at 90 °C for 1 h in a quartz beaker (with lid) under atmospheric pressure and used for monitoring chemical yield and preparing external calibration solutions for ICPMS, respectively. The nitric and hydrochloric acids used for chemical separation and preparation of final solutions for ICPMS measurements were of ultrapure grade (TAMAPURE AA-10) and were supplied by TAMA Chemicals. Ultrapure water with a resistivity of 18.2 MΩ cm prepared with a Milli-Q system was used in the experiments. Other reagents were analytical grade and used without further purification. Radiometric Measurements. The α and γ radioactivities were determined using a silicon surface-barrier detector (SSD, EG & G ORTEC, BU-017-450-100) and a Ge semiconductor detector (EG & G ORTEC, GMS-20195-PLUS), respectively. The activities of 243Am and 238Pu in the spent nuclear fuel sample were determined with α spectrometry. The spent nuclear fuel sample contained the same activity of 239Np as 243 Am, which was in secular equilibrium with 239Np. The 239Np concentrations in effluents from the cartridges were determined using γ spectrometry. α- and γ-Counting sources were prepared by directly depositing a small portion (0.05−0.1 mL) of the sample solution onto a tantalum disk.25 For α-counting sources, tetraethylene glycol was used to protect the surface of the 3150
DOI: 10.1021/acs.analchem.5b04330 Anal. Chem. 2016, 88, 3149−3155
Article
Analytical Chemistry mounted source. The efficiencies of α and γ detection were calibrated using standard point sources26 of 238Pu and 137Cs, respectively, with their geometries being identical to those of the sample sources. ICPMS. The concentrations of 237Np were determined with a quadrupole ICPMS (Agilent 7700x). The instrument has no special radiation shielding because of the sufficiently low level of radiation in terms of radiation protection (below 1 μSv/h). The operation conditions are summarized in Table 1. The final Table 1. Operating Conditions of ICPMS (Agilent 7700x) rf power plasma gas flow rate auxiliary gas flow rate nebulizer gas flow rate integration time no. replicate internal standard measured isotopes
1550 W 15.0 L/min 0.90 L/min 1.03 L/min 0.3 s 5 205 Tl 205, 235−243
solutions were prepared by dissolving or diluting with 5 mL of 1 M nitric acid. The formation of oxides and doubly charged ions (140Ce16O+/140Ce and 140Ce2+/140Ce) was adjusted to below 1.5%. The sensitivity at m/z 237 was approximately 140 000 cps per 1 ng/mL. The detection limit of 237Np was 0.001 ng/g, corresponding to 0.03 mBq/g. Online internal standard calibration using a 1 ng-Tl/g standard solution was employed for 237Np determination. The Tl standard solution was mixed in the sample solution immediately before reaching the nebulizer using the online standard addition kit provided by Agilent Technologies. Preparation of Anion-Exchange Cartridges. Four types of the anion-exchange group-introduced porous polymer sheets were prepared by using graft-polymerization technique. A 2.0 mm-thick porous polyethylene sheet with a porosity and average pore size of 75% and 1.0 μm, respectively, was used as a base polymer for grafting. The preparation procedures are shown in Figure 1. The procedures were different depending on the property of the anion-exchange group to be introduced. For introducing DMAEMA group, the procedure is the most simple. The base polymer sheet was irradiated with an electron beam under the nitrogen atmosphere to produce radicals that were required for the subsequent graft polymerization. Then, DMAEMA, tertiary amino group-containing vinyl monomer, was grafted onto the pore surface of the sheet by immersing the electron-beam irradiated sheet in 10 (v/v)% DMAEMA aqueous solution at 313 K for 22 h.24 The degree of grafting, defined by an equation below, was controlled to maintain the physical strength of the resultant sheet. degree of grafting (%) = {(W1 − W0)/W0} × 100
Figure 1. Preparation procedure of four anion-exchange sheets.
grafted sheet in the 0.5 M TMA-containing 1:1 (v/v) water− DMSO solution for 3 h.22 (2) The DEA group was introduced by immersing the GMA-grafted sheet in 50 (v/v)% diethylamine/water at 313 K for 1.5 h.18 (3) The GMA-grafted sheet was placed in contact with 1 M HCl at 333 K for 2 h for chlorination. Then, the resultant sheet was immersed in a 10 (w/w)% TEDA aqueous solution at 333 K for 2 h. The degree of conversion percentage was calculated by the number of moles of the epoxy group that reacted with the anion-exchange group divided by the initial number of moles of GMA in the GMA-grafted sheet. The reproducibility of the anion-exchange group density for each anion-exchange sheet was ensured by preparing GMA-grafted sheets with a constant degree of GMA grafting because almost 100% of epoxy group of GMA was converted into anion-exchange groups. The degree of GMA grafting can be precisely controlled (±10%) by adjusting the length of the reaction time. The resultant porous sheets were referred to as DMAEMA, TMA, DEA, and TEDA sheets, respectively. All cartridges were conditioned by permeating 5 mL of 9.4 M HCl−0.1 M HNO3, 1 M HCl, and 7 M HNO3 before use. Each of the sheets was cut out into disks measuring 5.9 mm in diameter and packed into an empty cylindrical cartridge purchased from Varian, Inc. Four such cartridges were prepared and referred to as DMAEMA, TMA, DEA, and TEDA cartridges. To evaluate the permeability of the prepared sheets, pure water flux was measured as follows:
(1)
where W0 and W1 are the masses of the base polymer sheet and the polymer-chain-grafted sheet, respectively. For introducing the TMA, DEA, and TEDA groups, an epoxy-ring-containing monomer GMA was grafted onto the pore surface of the sheet in advance. The GMA-grafted sheet was used as an intermediate for introducing the anion-exchange groups via the epoxy-ring opening reaction. The preparation conditions of the GMA-grafted sheet are detailed in our previous study.17 The introduction procedures of each anionexchange group were as follows: (1) The TMA group was introduced into the poly-GMA chain by immersing the GMA-
pure water flux (m/h) = (flow rate of pure water at 0.1 MPa and 298 K) /(cross‐sectional area of porous sheet) 3151
(2)
DOI: 10.1021/acs.analchem.5b04330 Anal. Chem. 2016, 88, 3149−3155
Article
Analytical Chemistry Am/Np Separation. The Am/Np separation procedure is shown in Figure 2. A standard solution containing about 40 000
using the aforementioned spent nuclear fuel sample. The reproducibility of adsorption-elution performance of the TEDA cartridge was confirmed before use by passing 9.4 M HCl including [FeCl4]− for adsorption and 1 M HCl for elution. The known activity of 243Am−239Np (10 000 Bq) standard solution in 1 M HNO3 was added to the diluted spent nuclear fuel solution, which contained 0.0029 mg of U. The amounts and concentrations of Np and Am isotopes generated during U Table 2. Amounts and Concentrations of Np and Am Isotopes in Spent Nuclear Fuel with Burnup of 44.9 GWd/t
237
Np Np 241 Am 243 Am 239
a
amount of nuclide in samplea (ng/g)
× × × ×
0.37 1.4 × 10−7 0.77 0.17
6.65 2.56 1.37 2.97
102 10−4 103 102
Amount of initial U: 0.0029 mg.
fission are summarized in Table 2. The radioactivity of the 239 Np added to the sample solution was determined by measuring a portion of the sample with γ spectrometry. The mixed solution was evaporated to dryness at 363 K. The oxidation state of Np was adjusted by following the procedure shown in Figure 2. The sample solution (0.3 mL) was fed into the prepared cartridge, followed by washing Am along with coexisting cationic species such as Sr, Y, Cs, Ba, and lanthanides with 0.7 mL of 9.4 M HCl−0.1 M HNO3. The activities of the effluent were measured with γ spectrometry to monitor unintentional elution of Np (239Np). Then, the U retained on the cartridge was washed out using 0.75 mL of HNO3. Finally, Np was eluted with 0.75 mL of HCl. The Np-eluted solution was evaporated to dryness and dissolved in 5 mL of 1 M HNO3. The concentration of 239Np in the Np-eluted solution was determined with γ spectrometry to calculate the chemical yield of Np, as defined below:
Figure 2. Extraction procedure of 239Np from 243Am standard solution using anion-exchange cartridges.
Bq of 243Am and the same quantity of 239Np was evaporated to dryness at 363 K. Subsequently, the residue was dissolved in 0.5 mL of 1 M HCl followed by the addition of 0.01 mL of 1 M hydroxylamine hydrochloride solution to adjust the oxidation state of Np to IV. The solution was heated at 363 K until it evaporated again. The dried sample was dissolved in 1.0 mL of 9.4 M HCl−0.1 M HNO3. The resultant solution was divided into four fractions and each was fed into four different cartridges. Then, each cartridge was washed with 9.4 M HCl− 0.1 M HNO3. The washing was continued until 239Np started to leak in the effluent. Finally, the 239Np retained on the cartridge was eluted using 1 M HCl. The effluents were continuously collected, and the activities of both 243Am and 239 Np were determined with γ spectrometry. The permeation rate was set to 1.5 mL/min. The separation performance of each cartridge was evaluated by the elution percentage defined as follows:
Np chemical yield (%) = {(239 Np activity found in Np‐eluted solution)
elution percentage (%) = {(activity in effluent) /(activity in the initial sample)} × 100
amount of nuclide (ng/mg-initial U)
/(239 Np activity added to the sample)} × 100 (3)
(4)
239
The activity of Np was decay-corrected to the time of the separation. The concentration of 237Np was determined using ICPMS based on external calibration using 205Tl. The separation performance of the TEDA cartridge was compared with that of anion-exchange-resin-packed column. A commercially available anion-exchange resin (AG1 × 8; wet bead size, 106−180 μm; Bio-Rad Laboratories, Inc.; wet bet volume, 1 mL) was packed into a polypropylene column having an inner diameter of 5.0−5.5 mm and length of 50 mm (Muromac S, Muromachi Technos Co., Ltd.). The gravity flow rate of the column was about 0.15 mL/min. The time required to complete the entire separation procedure was expected to be much longer than that with the TEDA cartridge. This might lead to an increase in 239Np activity originating from the decay of residual 243Am in the column during separation. To prevent such inadvertent addition of 239Np activity, a known amount of purified 239Np preliminary extracted from the 243Am standard was added to the same spent nuclear fuel sample solution.
Sample Preparation. A spent nuclear fuel solution was prepared by dissolving a single fuel pellet (UO2; burnup, 44.9 GWd/t) used in the Japanese PWR. The weight of the pellet was approximately 5 g, excluding the weight of the zircaloy cladding. The dissolution and dilution procedures are detailed in our previous study.27,28 The concentration of U in the sample solution was set to 0.009 mg-U/g. The amounts of the major nuclides found in the spent nuclear fuel sample were calculated using the isotope generation and depletion code, ORIGEN2, as well as the latest evaluated nuclear data library of JENDL-4.0.1 The concentration of U was determined previously by subsequent titrations using sodium hydroxide and hydrogen peroxide27 and validated against the calculated concentration obtained using ORIGEN2 and the measured concentration of 137Cs. Determination of 237Np in Spent Nuclear Fuel Sample. The separation performance of the TEDA cartridge was verified 3152
DOI: 10.1021/acs.analchem.5b04330 Anal. Chem. 2016, 88, 3149−3155
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Analytical Chemistry
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RESULTS AND DISCUSSION Separation Performances of Anion-Exchange-GroupIntroduced Disk-Packed Cartridge. Four anion-exchangegroup-containing disk-packed cartridges (DMAEMA, TMA, DEA, and TEDA) were prepared to compare the Am/Np separation performance. The properties of the cartridges are summarized in Table 3. Measured pure water fluxes showing
Table 4. Separation Performances of Prepared AnionExchange Cartridges DMAEMA retention volume of Np (mL) elution of Am in Np-eluted solution (%) elution volume of Np (mL)
Table 3. Properties of Prepared Anion-Exchange Cartridges DMAEMA degree of grafting (%) reaction time (h) conversion (%) anion-exchange group density (mol/kg) pure water flux at Δm = 0.1 MPa (m/h)
TMA
DEA
TEDA
3.2
220 3 97 3.3
200 1.5 90 3.4
220 4 98 5.2
1.9
51
47
52
101 22
TMA
DEA
TEDA
0.98
0.98
0.74
4.20
0.0086
