1. Introduction

Korea Research Reactors (KRRs) 1 and 2 (TRIGA Mark- II and TRIGA Mark-III types, respectively) were operated from 1962 and 1972, respectively, playing important roles in establishing basic nuclear science and technologies, and in advancing nuclear industries in the Republic of Korea. As the utility of these multipurpose research reactors declined with the operation of a new research reactor, HANARO, both of them were shut down in 1995 and decommissioning work began in 1997, with the goal of dismantling all facilities and removing all radioactive materials from the site. The dismantlement of the building structures and the reactor equipment resulted in various types of waste, including radiologically contaminated concrete and steel. In addition, activities such as cutting, packing, clean-up, and related tasks during the dismantlement process produced secondary waste such as filters, resin, plastic, and clothes. These decommissioning waste materials have been temporarily stored at the reactor site, and the corresponding radionuclide inventory must be assessed for their classification and disposal [1].

In Korea, domestic low- and intermediate-level radioactive waste (LILW) from nuclear power plants, industries, and research and medical institutes has been transported to the Wolsong LILW Disposal Center (WLDC). According to the LILW acceptance criteria of the Nuclear Safety and Security Commission of Korea (NSSC), the radioactivity of thirteen specified radionuclides (³H, ¹⁴C, ⁵⁵Fe, ^58,60Co, ^59,63Ni, ⁹⁰Sr, ⁹⁴Nb, ⁹⁹Tc, ¹²⁹I, ¹³⁷Cs, and ¹⁴⁴Ce) and the corresponding gross alpha must be identified [2]. Among these radionuclides, gamma-emitting radionuclides such as ^58,60Co, ¹³⁷Cs, and ¹⁴⁴Ce can easily be analyzed. However, chemical separation is generally required to determine the radioactivity of alphaand beta-decaying nuclides, including ³H, ¹⁴C, ⁵⁵Fe, ^59,63Ni, ⁹⁰Sr, ⁹⁴Nb, ⁹⁹Tc, and ¹²⁹I, because the matrix and coexisting radionuclides can interfere with the radiometric detection of their characteristic emissions [3].

In this work, we analyzed filter waste samples generated during the dismantling of KRRs and determined the inventory of the thirteen aforementioned radionuclides and the gross alpha. Before the analysis of the decommissioning waste, we evaluated destructive analysis methods with simulated filter samples spiked with radionuclides and/or stable isotopes. The radiochemical analysis results were eventually used for the disposal of the filter waste drums to the repository site.

2. Experimental

2.1 Equipment and Chemicals

Table 1 summarizes the measurement methods used in this work. A HPGe detector, a GEM-7084 model (ORTEC), was used to measure gamma emitters such as ^58,60Co, ⁹⁴Nb, ¹³⁷Cs, and ¹⁴⁴Ce. A HPGe detector, a GLP 36360-13 model (ORTEC), was used to measure the Auger electrons and X-rays of ⁵⁵Fe and ⁵⁹Ni. A HPGe detector, a GMX60-83 (ORTEC) model, was used to measure the Auger electrons and X-rays of ¹²⁹I. A Tri-Carb 3110TR (PerkinElmer Inc.) liquid scintillation counter was used to measure the beta emissions of ³H, ¹⁴C, ⁶³Ni and ⁹⁰Sr. A S5XLB low background alpha/beta counter (Canberra) was used to measure the beta emissions of ⁹⁹Tc and the gross alpha. A Spectro Arcos (Spectro Analytical Instruments) inductively coupled plasma-optical emission spectrometer (ICP-OES) was used to quantify the stable isotopes. Instrument calibration was performed with standard materials. A gamma-emitting mixed source solution and standard solutions of ²⁴¹Am, ⁹⁰Sr, and ⁹⁴Nb were purchased from Korea Research Institute of Standards and Science (KRISS). Standard solutions of ³H, ¹⁴C, ⁵⁵Fe, ⁵⁹Ni, ⁶³Ni, ⁹⁹Tc, and ¹²⁹I were purchased from Eckert & Ziegler. Anion exchange resins were purchased from Bio-Rad Laboratory. Sr and Ni resins were purchased from Eichrom Technologies Inc. Other chemicals were purchased from commercial vendors, which were of analytical grade and used as received.

Table 1

Radionuclide Concentration Limit for the LILW Disposal and Measurement Methods in This Work

2.2 Filter Waste Samples and Acid Leaching

Dismantlement activities such as the cutting of contaminated concrete were done in plastic greenhouses temporarily built in the reactor facilities. The working spaces were equipped with air ventilation systems, and water purifying systems were used to clean the cooling water generated at the cutting works. As the filters used there also became radioactive waste, they were sliced after dismantlement and packed in 200 L drums for disposal.

For the radiochemical analysis, amounts exceeding 500 g of the filter waste were randomly taken from each drum and packed into individual plastic bags. In this work, eight samples were analyzed. Five of them were from air filters (F1-F5) and the others (F6-F8) were a mixture of air filter and water purifying filter waste. Fig. 1 shows typical photographs of the two types of filter waste samples in an unpacked state. The filter samples were cut further into smaller fragments for acid leaching. About 5 g of these samples were placed in a 600 mL PFA beaker, to which 1.5 mL of a Re carrier solution (10 mg Re·mL⁻¹) and 500 mL of a mixture of 1 M HNO₃-2.5 M HCl-0.3 M HF were added. The beaker was covered with a Teflon watch glass and the mixture was heated at 200℃ for 4 h to leach the nuclides. After cooled down to room temperature, the sample solution was filtered. The volume of the filtrate was adjusted to 500 mL by adding deionized water or through evaporation and was directly used for the determination of gamma-emitting radionuclides.

Fig. 1

Photographic images of filter waste samples: (a) air filter, and (b) a mixture of air and water filters.

2.3 Gamma Spectrometry

The radioactivity of gamma-emitting nuclides such as ¹¹⁰mAg, ¹⁴⁴Ce, ^57,58,60Co, ^134,137Cs, ¹⁵²Eu, ⁵⁹Fe, ⁵⁴Mn, ⁹⁴Nb, ⁹⁵Nb, ¹²⁵Sb, ⁶⁵Zn, and ⁹⁵Zr in 500 mL of leachate was determined by a HPGe detector that was calibrated using a gamma-emitting mixed source. Gamma-ray spectra were obtained with a 10,000 s measurement.

2.4 Gross Alpha Measurement

After the radioactivity determination of gamma-emitting nuclides, 250 mL of the leachate was taken into a 500 mL PFA beaker and heated to reduce its volume to 20 mL. This solution was used for the gross alpha measurements and the sequential separation of beta-emitting radionuclides such as ⁵⁵Fe, ^59,63Ni, ⁹⁰Sr, ⁹⁴Nb, and ⁹⁹Tc. In this case, 1 mL of the concentrated sample solution was spread on a twoinch stainless steel planchet and dried under an IR ramp. After cooling down, the residue was weighed and counting was conducted with a S5XLB low background alpha/beta counter for 50 min.

2.5 Determination of the Radioactivity of ⁵⁵Fe, ^59,63Ni, ⁹⁰Sr, ⁹⁴Nb, and ⁹⁹Tc

The sequential separation of ⁵⁵Fe, ^59,63Ni, ⁹⁰Sr, ⁹⁴Nb, and ⁹⁹Tc was conducted according to the method developed at Korea Atomic Energy Research Institute (KAERI) [4]. The procedure is briefly depicted in Fig. 2. The amounts of carriers to add were determined from the ICP-OES measurements; 8 mL of the concentrated sample solution was used for separation, where the gross amounts of the carriers were adjusted to 3 mg of Re, 3 mg of Sr, 20 mg of Fe (or less than 50 mg), 20 mg of Nb, 2 mg of Ni, and 50 mg of Ca. First, ⁹⁹Tc was separated from the other radionuclides with an anion exchange resin (Bio Rad AG MP-1, 100-200 mesh). The ⁹⁹Tc effluent was evaporated to dryness and dissolved in 3 mL of 0.1 M HNO₃. The addition of 0.5 mL of ethanol and 0.02 M of tetraphenylarsonium chloride (TPAC) in H₂O precipitated [ReO₄(⁹⁹TcO₄)](TPA), which was dried to weigh for chemical recovery and counted with a S5XLB low background alpha/beta counter for 50 min. The coprecipitation of calcium oxalate at pH 4.5–5 and extraction chromatography with Sr resin (Eichrom Technologies, Inc., 100-150 mesh) isolated ⁹⁰Sr. The ICP-OES and LSC measurements correspondingly determined the chemical recovery and radioactivity of ⁹⁰Sr. ⁵⁵Fe and ⁹⁴Nb were separated with an anion exchange resin (Bio Rad AG 1-X8, 100-200 mesh). The ⁵⁵Fe effluent was collected into a saturated boric acid solution and 25% NH₄OH was added up to pH 9–10. The precipitate was collected and heated at 800℃ for 30 min. The resultant Fe₂O₃ was weighed for chemical recovery and the X-ray emission of ⁵⁵Fe at 5.9 keV was counted for 50 min. The ⁹⁴Nb effluent was counted with a HPGe detector for 5,000 s and chemical recovery was determined via the ICP-OES measurement. Lastly, the ^59,63Ni was purified with Ni resin (Eichrom Technologies, Inc.) and an anion exchange resin (Bio Rad AG 1-X8, 100-200 mesh). To the Ni effluent was added ~5 mL of 25% NH₄OH, 2 mL of 30% ammonium tartrate, and 9 mL of H₂O. The pH of the solution was adjusted to 9−10 with the slow addition of 10% NH₄OH. The addition of 5 mL of 1% dimethylglyoxime (DMG) in 95% ethanol resulted in red Ni(DMG)₂ (s). Chemical recovery was determined from the weight of the precipitate, and the X-ray emission of ⁵⁹Ni at 6.9 keV was counted for 50 min. Then, Ni(DMG)₂ was dissolved in 2−3 mL of HNO₃ and evaporated to near dryness. The residue was dissolved in 3 mL of HClO4 and evaporated to incipient dryness, with this step repeated once more. The residue was dissolved in 2 mL of 0.7 M HCl and mixed with 18 mL of the UltimaGold LLT scintillation cocktail. The radioactivity of ⁶³Ni was measured with a liquid scintillation counter for 30 min.

Fig. 2

A scheme of the sequential separation of ⁹⁹Tc, ⁹⁰Sr, ⁵⁵Fe, ⁹⁴Nb, and ^59,63Ni. (a) 0.5 M HNO₃ (saved for c), (b) 14% NH₄OH-4% HF and 0.1 M HNO₃ (discarded), and 10 M HNO₃ (⁹⁹Tc), (c) calcium oxalate co-precipitation (supernatant: saved for e), (d) the precipitate was further purified with Sr-resin (⁹⁰Sr), (e) 0.1 M ammonium oxalate (pH 4.5−5) (saved for h), (f) distilled water (discard) and 4 M HF (⁵⁵Fe), (g) 3 M HCl-20% HF (discarded) and 5 M HNO₃-0.2 M HF (⁹⁴Nb), (h) 0.1 M ammonium oxalate (pH 9−10) and 0.1 M HCl (discarded), and (i) 9 M HCl (Ni-resin column to AG 1-X8 column, ^59,63Ni).

2.6 Determination of the Radioactivity of ³H and ¹⁴C

A wet oxidation method developed at KAERI [5,6] was slightly modified and used for the separation of ³H and ¹⁴C. To a 250 mL round flask, ~5 g of the filter slices, 5 g of potassium persulfate and 0.5 g of AgNO₃ were added, after which the flask was connected to a custom-made distillation apparatus [5]. With purging N₂(g) through the apparatus, 50 mL of 3 M H₂SO₄ was slowly added. The reaction mixture was stirred and heated at 60℃ for 2.5 h, when the evolved CO₂(g) was trapped in a mixture of 10 mL of Carbo-Sorb E and 10 mL of Permafluor E⁺ placed in a trapping tube at the side end of the apparatus. Then, the reaction mixture was refluxed at 120℃ for 30 min and about 10 mL of H₂O was distilled with additional heating. The collected ¹⁴C solution was counted with a liquid scintillation counter for 30 min. Five mL of the distillate (i.e. a tenth of the volume of the reaction solution) was transferred to a scintillation vial and 15 mL of UltimaGold LLT was mixed into it, which was counted with a liquid scintillation counter for 30 min.

2.7 Determination of the Radioactivity of ¹²⁹I

¹²⁹I was leached by vigorously shaking ~20 g of the filter slices in a mixture of 180 mL of H₂O and 20 mL of 7% NaClO with KI (107 mg) as a carrier for 2 h [7]. The leachate was filtered and the residue was rinsed with 15 mL of H₂O. The filtrate was evaporated up to approximately 100 mL and acidified with 1.25 mL of HNO₃. After the addition of 20 mL of CHCl₃, 2 mL of 7.2 M NH₂OH⋅HCl and 1 mL of 7% NaClO, the mixture was agitated for 10 min. The organic layer was separated and the aqueous layer was extracted with an additional 10 mL of CHCl₃ and 1 mL of 7% NaClO twice. The combined organic layer was washed with 10 mL of H₂O. Then, the organic solution was extracted twice with 5 mL and 1 mL of 0.2 M NaHSO₃. The combined aqueous solution was mixed with 0.5 mL of HNO₃ and 3.5 mL of 0.2 M AgNO₃, resulting in a precipitate of AgI. Chemical recovery was determined with the weight of the precipitate, and the X-ray emission of ¹²⁹I at 29.8 keV was counted for 10,000 s.

2.8 Calculation of the Minimum Detectable Activity (MDA)

The MDA of the measurement was calculated using the equation below:

Here, B is the background count rate in cps, T is the background and sample counting time in s, ε is the counting efficiency, m is the weight of the sample used in the measurement in g, Y is the chemical recovery yield.

3. Results and Discussion

3.1 Acid Leaching of Filter Waste Samples

The leaching of filter waste samples in a mixed-acid solution was selected for the gamma measurement. Prior to the sample analysis, the leaching condition was tested with simulated filter samples spiked with stable isotopes of Ce, Co, Cs, Fe, Nb, Ni, Re, and Sr, and U, where Re was used as a surrogate of ⁹⁹Tc. To ~5 g of a fresh HEPA filter, 200 μg of Cs for the ICP-MS measurement and 1,000 μg of Ce, 10,000 μg of Fe, 2,500 μg of Sr, and 250 μg each of the other elements for the ICP-OES measurements were added. Three simulated samples were leached in 500 mL of a mixture of 1 M HNO₃-2.5 M HCl-0.3 M HF at 200℃ for 4 h. Table 2 shows the element quantification results of the leached solutions from the simulated filter waste samples. Because the HEPA filter contains small amounts of Ce, Cs, Fe, Ni, and Sr, the chemical recovery outcome was corrected based of the quantification results from the leaching of the fresh filter. All of the elements were nearly quantitatively recovered on average. Among them, Fe showed the lowest average chemical yield (95.1 ± 5.2%), which is attributable to the highness and inhomogeneity of the Fe content (2,860 ± 250 μg·g⁻¹) with respect to the amounts of other elements in the HEPA filter.

Table 2

Leaching Results of Simulated Filter Waste Samples

The filter waste samples were leached as described in the experimental section. The final volume of the leachate was 500 mL, and this amount was used for the gamma spectrometry. Half of the solution was then concentrated to 20 mL, some portions of which were used for the gross alpha measurements and the separation of beta-emitting nuclides such as ⁵⁵Fe, ^59,63Ni, ⁹⁰Sr, ⁹⁴Nb, and ⁹⁹Tc.

3.2 Radioactivity Concentration Determination of Gamma Emitters

Table 3 shows the activity of the gamma-emitting radionuclides of ^58,60Co, ¹³⁷Cs, and ¹⁴⁴Ce in the filter waste samples. The activity of ¹⁵²Eu was also reported, as ¹⁵²Eu was mainly detected in activated concrete in KRRs [8,9] and some of the filter waste was generated during the dismantling of the reactor shielding concrete. Among the gamma-emitting radionuclides, ⁶⁰Co and ¹³⁷Cs were detected in all of the samples at corresponding radioactivity of 0.74–70 Bq·g⁻¹ and 0.64–8.7 Bq·g⁻¹. The signals of ¹⁵²Eu were also found in six samples at radioactivity of 0.19–2.9 Bq·g⁻¹. These samples had relatively high contents of ⁶⁰Co (≥ 6.9 Bq·g⁻¹). Other gamma-emitting radionuclides, such as ¹¹⁰mAg, ¹⁴⁴Ce, ^57,58Co, ⁵¹Cr, ¹³⁴Cs, ⁵⁴Mn, ⁹⁵Nb, ¹²⁵Sb, ⁶⁵Zn, and ⁹⁵Zr, were not detected under the conditions of this work, and their MDA values were in the range of 10⁻²–10⁻¹ Bq·g⁻¹. Enough time had passed such that many of the radionuclides stemming from the operation of the reactor likely decayed since the reactor was shut down in 1995. For this reason, the analysis detected relatively long half-life isotopes such as ⁶⁰Co, ¹³⁷Cs, and ¹⁵²Eu.

Table 3

Radioactivity in Bq·g⁻¹ of Selected Gamma-Emitting Radionuclides and Gross Alpha in the Filter Waste Samples

3.3 Radioactivity Concentration Determination of Gross Alpha

Measuring the gross alpha is a simple approach that can provide a good estimate of the total radioactivity of alphaemitting nuclides [10]. In this work, a portion of the sample leaching solution was dried in a planchet and the residue was counted using a gas proportional counter (GPC). Because the extent of self-absorption depends on the counting sample thickness, the counting efficiency of the gross alpha measurement was calculated using the ²⁴¹Am standard with different weights of NaCl as an estimate of the residue thickness in the planchet [10]. Table 3 shows the results of the gross alpha measurements. The gross alpha concentrations of the six samples were measured and found to be 0.32–1.1 Bq·g⁻¹. The others did not present detectable activities with MDA values of 10⁻²–10⁻¹ Bq·g⁻¹. Currently, the WLDC requires activity determinations of alpha emitters such as U, Pu, Am, and Cm isotopes in drums where the gross alpha concentration exceeds 10 Bq·g⁻¹ [11]. In this case, individual quantifications of these alpha-emitting nuclides are needed [12], but this is out of the scope of this work.

3.4 Radioactivity Concentration Determinations of ⁵⁵Fe, ^59,63Ni, ⁹⁰Sr, ⁹⁴Nb, and ⁹⁹Tc

The interference of coexisting radionuclides and the sample matrix and the radiation characteristics require chemical separation of beta-decaying nuclides such as ³H, ¹⁴C, ⁵⁵Fe, ^59,63Ni, ⁹⁰Sr, ⁹⁴Nb, ⁹⁹Tc, and ¹²⁹I for proper radiometric measurements. KAERI has developed several methods to separate sequentially ⁵⁵Fe, ^59,63Ni, ⁹⁰Sr, ⁹⁴Nb, and ⁹⁹Tc so as to quantify them in LILW [4,13-16]. One of the purposes behind this was to develop scaling factors for LILW generated from nuclear power plants [15,17].

The method applied in this work used ion exchange and extraction chromatography as well as precipitation to isolate ⁹⁹Tc, ⁹⁰Sr, ⁵⁵Fe, ⁹⁴Nb, and ^59,63Ni in sequence [4]. ⁹⁹Tc was separated from the leaching solution containing Re as a carrier using an anion exchange resin, where ⁹⁹TcO₄⁻ and ReO₄⁻ are strongly adsorbed but the other metal ions and the matrix elements do not exist in 0.5–1 M HNO₃. A complex of [ReO₄(⁹⁹TcO₄)](TPA) was prepared for counting. The average chemical recovery of Re was 64%. Table 4 shows the concentration of ⁹⁹Tc in the filter waste samples. One filter sample (F8) showed an activity of 0.078 Bq·g⁻¹. The others did not present detectable activities with MDA values of 0.03–0.07 Bq·g⁻¹.

Table 4

Radioactivity in Bq·g⁻¹ of Selected Radionuclides in the Filter Waste Samples

Calcium oxalate co-precipitation was used to separate Sr from the solution containing Fe, Nb, and Ni. ⁹⁰Sr was further purified for the LSC measurement by means of Srresin extraction chromatography, which removed Ca and residual interfering radionuclides. The average chemical recovery was 86%. The ⁹⁰Sr concentrations of F2, F3, and F7 were measured and found to be 5.0, 2.0, and 0.25 Bq·g⁻¹, respectively. Those of the other samples were not detected given the MDA values of 0.16–0.19 Bq·g⁻¹.

Anion exchange chromatography of the oxalate supernatant during Sr separation enabled the individual isolation of ⁵⁵Fe and ⁹⁴Nb, the corresponding X-rays and gamma rays of which were measured with HPGe detectors in both cases. The average chemical recovery of Fe was 63%. The measured ⁵⁵Fe concentrations of F2, F5, and F8 were 1.2, 7.1, and 3.5 Bq·g⁻¹, respectively. The MDAs of ⁵⁵Fe in this work ranged from 0.9 to 1.2 Bq·g⁻¹.

⁹⁴Nb is not only a beta emitter but also a high-energy gamma emitter. However, directly measuring the gamma rays of ⁹⁴Nb in nuclear waste when large amounts of other gamma-emitting radionuclides are present is sometimes unsuccessful. The separation method in this work featuring the isolation of ⁹⁴Nb resulted in an average chemical recovery of 87%. However, ⁹⁴Nb was not detected in any of the samples. The MDA values determined after separation were 0.55–0.58 Bq·g⁻¹. They were somewhat higher than those from the direct gamma-ray measurements of the leaching solution, which were estimated to be 0.06–0.24 Bq·g⁻¹. The insignificant interference of other gammaemitting radionuclides and the larger sample amounts as well as the longer counting time in the gamma-ray measurements of the leaching solutions may have lowered the MDA values.

^59,63Ni were isolated with Ni-Resin extraction chromatography followed by anion exchange chromatography. The X-rays of ⁵⁹Ni were measured with the Ni(DMG)₂ complex. The complex was then decomposed by acid digestion, and the beta rays of ⁶³Ni were measured via LSC. The average chemical recovery of Ni was 85%. The concentrations of ⁵⁹Ni of F1, F2, F7, and F8 were all in the range of 0.6–1.0 Bq·g⁻¹. The other samples did not show detectable activities of ⁵⁹Ni given their MDA values of ~0.09 Bq·g⁻¹. On the other hand, ⁶³Ni was detected in all of the samples at concentrations in the range of 0.60–94 Bq·g⁻¹. Thermal neutron reactions with stable nickel ⁶²Ni (3.65%) and ⁵⁸Ni (68.1%) produced both isotopes, and the initial activity ratio of ⁶³Ni/⁵⁹Ni is expected to be ~100 or higher in nuclear waste [3,18]. The average activity ratios of ⁶³Ni/⁵⁹Ni of the filter waste samples were ~93, which seems reasonable given that the research reactors were shut down in 1995.

3.5 Radioactivity Concentration Determinations of ³H and ¹⁴C

As ³H exists as HT or HTO and ¹⁴C exists as carbonate or carbon in concrete [19], the secondary filter waste generated during the dismantling of concrete structures may also contain those chemical forms. The wet oxidation method was assessed with simulated filter waste samples. HTO and Na²¹⁴CO₃ standard solutions were loaded onto ~5 g of a sliced HEPA filter, where the amounts of ³H and ¹⁴C spiked were 153 and 321 Bq, respectively. A strong acidic condition with oxidants such as potassium persulfate and AgNO₃ facilitated the evolution of ¹⁴CO₂ [5,6], which was chemisorbed onto Carbo-Sorb E. Subsequent distillation separated HTO from the reaction mixture. The collected ³H and ¹⁴C were counted as described in the experimental section and five trials led to recovery of 97.1 ± 4.2 and 94.6 ± 8.3%, respectively. The analysis results in Table 4 show that ³H and ¹⁴C were found in all of the filter waste samples. The measured concentrations of ³H and ¹⁴C were 512–3,600 and 7.5–29 Bq·g⁻¹, respectively. The high content of ³H in the concrete waste of the KRR-2 [20] most likely led to the high concentration of ³H in the secondary filter waste generated during the decommissioning process.

3.6 Radioactivity Concentration Determination of ¹²⁹I

Due to its very long half-life and high mobility in the geosphere, ¹²⁹I is of long-term concern regarding the safety assessment of the final repository. It is highly probable that most types of LILW contain trace amounts of ¹²⁹I [21]. When ¹²⁹I is not detected under analysis conditions applied, MDA values can be used for an inventory evaluation of the repository [22]. Therefore, it is interesting to lower the detection limit with the improvement of the separation methods and instrumentation regarding the safety assessments of disposal facilities [7,21,23].

In this work, the activity of ¹²⁹I was measured by the X-ray counting method, which is less sensitive than mass spectrometry [3,23]. The sample quantity was adjusted to ~20 g to target a MDA value below 10⁻² Bq·g⁻¹. The filter waste samples were vigorously shaken in a NaClO solution containing KI as a carrier, similar to the leaching method used for the analysis of dry waste from nuclear power plants [7]. Subsequent redox reactions and liquid-liquid extractions purified the iodine species further. The leached IO₃⁻ was oxidized to I₂ by hydroxylamine, which was extracted to the organic layer. The stripping of the organic layer with a NaHSO₃ solution reduced I₂ into I⁻. Lastly, the precipitation reaction of Ag⁺ and I⁻ resulted in a counting source for the X-ray measurement. The average chemical recovery was estimated to be 81%. The analysis results in Table 4 show that ¹²⁹I was not detected in any of the filter waste samples given that the MDA values were as low as ~0.007 Bq·g⁻¹.

4. Conclusion

We determined the radionuclide inventory of filter waste generated during the dismantling of KRRs. For a destructive analysis of the filter waste samples, several pretreatment and separation methods were utilized. The samples were leached in acid solutions for the measurements of gamma emitters such as ^58,60Co, ⁹⁴Nb, ¹³⁷Cs, ¹⁴⁴Ce, and ¹⁵²Eu, and for the determination of the gross alpha concentrations. The leaching solutions were also used to separate ⁹⁹Tc, ⁹⁰Sr, ⁵⁵Fe, ⁹⁴Nb, and ^59,63Ni sequentially using the chromatographic and precipitation methods. On the other hand, ¹⁴C and ³H were separated using the wet oxidation method and ¹²⁹I was isolated by a procedure that included leaching, purifying based on redox reactions and liquid-liquid extractions, and selective precipitation. Radiometric measurements were used to determine the activities of the separated beta-decaying radionuclides. The concentrations of all radionuclides determined in this work were below the concentration limit of LILW in the acceptance criteria of the NSSC (Table 1). Finally, the radiochemical analysis results were used for the disposal of waste drums from the decommissioning of KRRs to the repository site.