Development of C-14 Treatment Facility for C-14 Treatment in PHWR

Kyuho Lee,Byeong-Ho Kim,Bong-Ki Ko,Hyeon-Oh Park,Ki-Hyun Kwon 한국방사성폐기물학회 2022 한국방사성폐기물학회 학술논문요약집 Vol.20 No.1

The permanent shutdown of Wolseong 1, PHWR (Pressurized Heavy Water Reactor) was decided. Accordingly, there is need for C-14 treatment technology to spent resin generated by PHWR in classified Medium Level Radioactive Waste by C-14 specific activity. However, spent resin by PHWR is mixed and stored with activated carbon and zeolite (mixture), not a single storage, and separation from the mixture must be carried out in advance for C-14 treatment in the spent resin. This study developed a C-14 treatment facility that combined with the technology of separating spent resin from spent resin mixture by PHWR NPP and the technology of C-14 treatment for disposal. The C-14 treatment facility consists of spent resin separation (Part 1) and treatment of separated spent resin. (Part 2) Part 1 is applied with a process of separating the mixed and stored spent resin from the spent resin mixture by applying a drum screen method. In the case of Part 2, spent resin treatment process for desorbing and collecting C-14 nuclides in the separated spent resin using microwave reactor was applied. Except for the adsorbent used to collect C-14 detached in the process of separating and treating spent resin, no additional material is introduced into the facility, and thus secondary waste is significantly reduced. In addition, pollution prevention banks at the bottom of the facility and a sealed automated circulation system were applied to prevent unexpected leakage and diffusion of radioactive materials and ensure stability of workers. Currently, the C-14 treatment facility has been verified for spent resin separation and spent resin treatment using simulated spent resin mixture, and the facility will be demonstrated and verified for field applicability. According to derived results, it is believed that it will be possible to apply the C-14 treatment facility when decommissioning of PHWR.

Selective Treatment of C-14 in Spent Ion Exchange Resin From HWR Using the Microwave

Ga Yeong Kim,Ki Rak Lee,Hwan-Seo Park,Geun-il Park,Hyeon-Oh Park,Gi-Hyeon Kwon 한국방사성폐기물학회 2023 한국방사성폐기물학회 학술논문요약집 Vol.21 No.1

Mixed-bed ion exchange resin consist of anion exchange resin and cation exchange resin is used to treat liquid radioactive waste in nuclear power plants. C-14 from heavy water reactors (HWR) is adsorbed on the anion exchange resin and is considered intermediate-level radioactive waste. The total amount of radioactivity of C-14 in spent ion exchange resin exceeds the activity limits for the disposal facility. Therefore, it is necessary to reduce the radioactivity through pre-treatment. There are thermal and non-thermal methods for the treatment of spent ion exchange resin. However, destructive methods have the problem of emitting off-gas containing radionuclides. To solve this challenge, various methods have been developed such as acid stripping, PLO process, activity stripping, thermal treatment and others. In this study, spent ion exchange resin (spent resin) was treated using microwave. The reaction characteristics of the resin to microwave were used to selectively remove the C-14 on the functional groups. Simulated spent anion exchange resin and spent resin from Wolseong NPP were treated with the microwave method, and the desorption rate was over 95%. An integrated process system of 1 kg/batch was built to produce operating data. After the operation of the process, characterization and evaluation of post-treatment for condensate water and adsorbent used in the process were performed. When the process system was applied to treat simulated spent resin and real spent resin, both showed a desorption rated of more than 97%. It means that the C-14 was successfully removed from the radioactive spent resin.

Microwave Processing for Recovery of C-14 From Spent Ion Exchange Resin

Ga-Yeong Kim,Ki-Rak Lee,Hwan-Seo Park,Geun-il Park,Hyeon-Oh Park,Gi-Hyeon Kwon 한국방사성폐기물학회 2023 한국방사성폐기물학회 학술논문요약집 Vol.21 No.2

Ion exchange resins are commonly employed in the treatment of liquid radioactive waste generated in nuclear power plants (NPP). The ion exchange resin used in NPP is a mixed-bed ion exchange resin known as IRN-150, which is of nuclear grade. This resin is a mixture of cation exchange resin and anion exchange resin. The cation exchange resin removes cationic radionuclides such as Cs and Co, while anion exchange resin handles anions (e.g., H14CO3 -), effectively purifying the liquid waste. Spent ion exchange resins (spent resin) containing C-14 are classified as low and intermediate level radioactive waste, and their radioactivity needs to be reduced as it exceeds the disposal limit regulated by law. Therefore, the microwave technology for the removal of C-14 from spent resin has been investigated. Previous studies have successfully developed a method for the effective removal of C-14 during the resin treatment process. However, it was observed that, in this process, functional groups in the resin were also removed, resulting in the generation of off-gases containing trimethylamine. These off-gases can dissolve in water from process, increasing its pH, which can subsequently hinder the recovery of C-14. In this study, we investigated the high-purity recovery of C-14 by adjusting the moisture content within the reactor following microwave treatment. Mock spent resins, consisting of 100 g of resin with HCO3 - ion-exchanged and 0, 25, or 50 g of deionized water, were subjected to microwave treatment for 40 or 60 minutes. Subsequently, the C-14 desorption efficiency of the mock spent resins was evaluated using an acid stripping process with H3PO4 solution. The functional group status of the mock spent resins was analyzed using 15N NMR spectroscopy. The results showed that the mock spent resins exhibited efficient C-14 recovery without significant functional group degradation. The highest C-14 desorption efficiency was achieved when 25 g of deionized water was used during microwave treatment.

Proposal of Economical and Efficient Treatment Facility for Treating Pressurized Heavy Water Reactor Resin Mixture

Moonoh Kim,Young-Ku Choi,Kwang-Woo Jung,Su-Young Moon,Jongsoon Song 한국방사성폐기물학회 2023 한국방사성폐기물학회 학술논문요약집 Vol.21 No.1

Pressurized Heavy Water Reactors (PHWR) have stored ion exchange resins, which are used in deuteration, dehydrogenation systems, liquid waste treatment systems, and heavy water cleaning systems, in spent resin storage tanks. The C-14 radioactivity concentration of PHWR spent resin currently stored at the Wolseong Nuclear Power Plant is 4.6×10E+6 Bq/g, which exceeds the limited concentration of low-level radioactive waste. In addition, when all is disposed of, the total radioactivity of C-14, 1.48×10E+15 Bq, exceeds the disposal limit of the first-stage disposal facility, 3.04×10E+14. Therefore, it is currently impossible to dispose of them in Gyeongju intermediate- and low-level disposal facilities. As to dispose of spent resins produced in PHWR, C-14 must be removed from spent resins. This C- 14 removal technology from the spent resin can increase the utilization of Gyeongju intermediate- and low-level disposal facilities, and since C-14 separated from the spent resin can be used as an expensive resource, it is necessary to maximize its economic value by recycling it. The development of C-14 removal technology from the spent resin was carried out under the supervision of Korea Hydro & Nuclear Power in 2003, but there was a limit to the C-14 removal and adsorption technology and process. After that, Sunkwang T&S, Korea Atomic Energy Research Institute, and Ulsan Institute of Science and Technology developed spent resin treatment technology with C-14-containing heavy water for the first and second phases from 2015 to 2019 and from 2019 to the present, respectively. The first study had a limitation of a pilot device with a treatment capacity of 10L per day, and the second study was insufficient in implementing the technology to separate spent resin from the mixture, and it was difficult to install on-site due to the enlarged equipment scale. The technology to be proposed in this paper overcomes the limitations of spent resin mixture separation and equipment size, which are the disadvantages of the existing technology. In addition, since 14CO2 with high concentration is stored in liquid form in the storage tank, only the necessary amount of C-14 radioactive isotope can be extracted from the storage tank and be used in necessary industrial fields such as labeling compound production. Therefore, when the facility proposed in this paper is applied for treating mixtures in spent resin tanks of PHWR, it is expected to secure field applicability and safety, and to reflect the various needs of consumers of labeled compound operators utilizing C-14.

Simultaneous Separation and Determination of 14C and 3H in Spent Resins from PWR Nuclear Power Plants

Soon-Dal Park,Jung-Suck Kim,Jong-Goo Kim,Sun-Ho Han,Kwang-Yong Jee 한국방사성폐기물학회 2007 방사성폐기물학회지 Vol.5 No.3

가압경수로형 원자력발전소의 운영과정에서 발생된 폐수지내 및 의 분포특성을 조사하였다. 표준용액을 사용한 의 회수율 측정결과, 사용한 산의 종류에 관계없이, 3 N-HCl , 주입한 농도 범위에서 의 회수율을 나타내었다. 같은 장치를 사용하여 HTO 표준용액 증류에 의한 의 회수율은 주입한 농도 범위에서 이었다. 습식산화-산용출법에 의한 폐수지의 및 동시분리시, 를 사용했을 때 다른 감마핵종에 의한 방해가 없었으며, 포집액이 섬광제와 잘 혼합되었다. 그러나 3 N-HCl을 사용했을 때 포집용액에서 및 등의 감마핵종이 검출되었다. 또한 Sample Oxidizer에 의한 포집용액에서도 및 등이 검출되었으며, 포집용액에서는 이 검출되었다. 폐수지의 총 함량중 약 70% 이상이 무기 탄소로 확인되었다. 30개 폐수지 시료중 8개 고방사능 폐수지의 및 의 평균농도는 각각 이었으며 22개 저방사능폐수지에서는 각각 이 검출되었다. 고방사능 폐수지의 평균 비는 28로 저방사능 폐수지의 0.70에 비해 높게 나타났으며, 및 의 농도는 서로 비례하는 경향을 보였다. In this work distribution characteristics of spent resins from nuclear power plants(NPPs), pressurized water reactors(PWRs), was investigated. It was found that the recovery percent of by the wet oxidation-acid stripping was for the added activity range of , and it was not affected by the kinds of stripping acids, 3N-HCl, . And the recovery percent of by distillation using the same apparatus was for the added activity range of . Among the tested stripping acids, 3\;N-HCl, , only the trapped solution by distillation in was compatible with the 3H scintillator, Ultimagold XR. Neither of the trapping solutions from the spent ion exchange resin samples by the wet oxidation-3 stripping contained gamma nuclides. However, some gamma nuclides, , were found in the trapped solutions of the spent resins by the wet oxidation-3 N-HCl stripping. It was the same for the trapping solutions of the spent resins by Sample Oxidizer(PACKARD MODEL 307). Meanwhile only two nuclides, , were found in the trapping solutions of the spent resins by Sample Oxidizer(PACKARD MODEL 307). It was found that most of the in the spent resins existed as inorganic carbon form, more than about 70% of the total content. Among the analyzed 30 spent ion exchange resin samples, the average concentration of and for the high radioactive samples, 8 samples, was and that for the low radioactive samples, 22 samples, was , respectively. And the average ratio for the high radioactive samples, was higher, 28, than that of low radioactive samples, 0.70. Some linear relationship trend was found between the activity concentrations of .

Evaluation of radiological safety according to accident scenarios for commercialization of spent resin mixture treatment device

최우년,변재훈,김희령 한국원자력학회 2022 Nuclear Engineering and Technology Vol.54 No.7

Spent resin often exceeds radiation limits for safe disposal, creating a need for commercial-scale treatment techniques to reduce resin radioactivity. In this study, the radiological safety of a commercialized spent resin treatment device with a treatment capacity of 1 ton/day was evaluated. The results confirm that the device is radiologically safe in the event of an accident. This device desorbs 14C from the spent resin, allowing disposal as low-level waste instead of intermediate-level waste. The device also reduces overall waste by recycling the extracted 14C. Potential accident scenarios were explored to enable dose assessments for both internal and external exposure while preventing further spillage of the device and processing the spilled resin. The scenarios involved the development of a surface fracture on the resin mixture separator and microwave systems, which were operated under pressure and temperature of 0 e6 bar and 0e150 C, respectively. In the case of accidents with separator and microwave device, the maximum allowable working time of worker were derived, respectively, considering external and internal exposures. When wearing the respirator corresponding to APF 50, in the case of the microwave device accident scenario, the radiological safety was confirmed when the maximum worker worked within 132.1 h

A Study on Adsorption and Desorption Behaviors of 14C From a Mixed Bed Resin

박승철,조항래,이지훈,양호연,양오봉 한국원자력학회 2014 Nuclear Engineering and Technology Vol.46 No.6

Spent resin waste containing a high concentration of 14C radionuclide cannot be disposed of directly. A fundamentalstudy on selective 14C stripping, especially from the IRN-150 mixed bed resin, was carried out. In single ion-exchangeequilibrium isotherm experiments, the ion adsorption capacity of the fresh resin for non-radioactive HCO3- ion, as thechemical form of 14C, was evaluated as 11mg-C/g-resin. Adsorption affinity of anions to the resin was derived in order ofNO3- > HCO3- ≥ H2PO4-. Thus the competitive adsorption affinity of NO3- ion in binary systems appeared far higher thanthat of HCO3- or H2PO4-, and the selective desorption of HCO3- from the resin was very effective. On one hand, the affinityof Co2+ and Cs+ for the resin remained relatively higher than that of other cations in the same stripping solution. Desorptionof Cs+ was minimized when the summation of the metal ions in the spent resin and the other cations in solution was nearsaturation and the pH value was maintained above 4.5. Among the various solutions tested, from the view-point of thesimple second waste process, NH4H2PO4 solution was preferable for the stripping of 14C from the spent resin.

A STUDY ON ADSORPTION AND DESORPTION BEHAVIORS OF ¹⁴C FROM A MIXED BED RESIN

Park, Seung-Chul,Cho, Hang-Rae,Lee, Ji-Hoon,Yang, Ho-Yeon,Yang, O-Bong Korean Nuclear Society 2014 Nuclear Engineering and Technology Vol.46 No.6

Spent resin waste containing a high concentration of $^{14}C$ radionuclide cannot be disposed of directly. A fundamental study on selective $^{14}C$ stripping, especially from the IRN-150 mixed bed resin, was carried out. In single ion-exchange equilibrium isotherm experiments, the ion adsorption capacity of the fresh resin for non-radioactive $HCO_3{^-}$ ion, as the chemical form of $^{14}C$, was evaluated as 11mg-C/g-resin. Adsorption affinity of anions to the resin was derived in order of $NO_3{^-}$ > $HCO_3{^-}{\geq}H_2PO_4{^-}$. Thus the competitive adsorption affinity of $NO_3{^-}$ ion in binary systems appeared far higher than that of $HCO_3{^-}$ or $H_2PO_4{^-}$, and the selective desorption of $HCO_3{^-}$ from the resin was very effective. On one hand, the affinity of $Co^{2+}$ and $Cs^+$ for the resin remained relatively higher than that of other cations in the same stripping solution. Desorption of $Cs^+$ was minimized when the summation of the metal ions in the spent resin and the other cations in solution was near saturation and the pH value was maintained above 4.5. Among the various solutions tested, from the view-point of the simple second waste process, $NH_4H_2PO_4$ solution was preferable for the stripping of $^{14}C$ from the spent resin.

유도결합플라스마 질량분석을 위한 사용후핵연료 중 테크네튬-99의 추출크로마토그래피 분리

서무열,이창헌,한성호,박영재,지광용,김원호 한국분석과학회 2004 분석과학 Vol.17 No.5

To determine the contents of 99Tc in the spent PWR (pressurized water reactor) nuclear fuels by ICP-MS (inductively coupled plasma-mass spectrometry), a technetium separation method using an extraction chromatographic resin (TEVA·Spec resin) has been established. 99Tc was separated from a spent PWR nuclear fuel solution by this separation procedure and its concentration was determined by ICP-MS. The result agrees well with the value calculated by the program ORIGEN 2 and also the value measured by AG MP-1 resin/ICP-MS method described in our previous paper. It can be concluded that the present separation procedure is superior to the AG MP-1 resin procedure with respect to the time required for technetium separation as well as the efficiency of decontamination from other radioactive nuclides

Feasibility Study of Synthesizing Graphene Quantum Dots From the Spent Resin in a Nuclear Power Plant

Seungbin Yoon,Woo Nyun Choi,Jaehoon Byun,Hee Reyoung Kim 한국방사성폐기물학회 2022 한국방사성폐기물학회 학술논문요약집 Vol.20 No.1

The feasibility study of synthesizing graphene quantum dots from spent resin, which is used in nuclear power plants to purify the liquid radioactive waste, was conducted. Owing to radiation safety and regulatory issues, an uncontaminated ion-exchange resin, IRN150 H/OH, prior to its use in a nuclear power plant, was used as the material of experiment on synthesis of graphene quantum dots. Since the major radionuclides in spent resin are treated by thermal decomposition, prior to conducting the experiment, carbonization of ion-exchange resin was performed. The experiment on synthesis of graphene quantum dots was conducted according to the general hydrothermal/solvothermal synthesis method as follows. The carbonized ion-exchange resin was added to a solution, which is a mixture of sulfuric acid and nitric acid in ratio of 3:1, and graphene quantum dots were synthesized at 115°C for 48 hours. After synthesizing, procedure, such as purifying, filtering, evaporating were conducted to remove residual acid from the graphene quantum dots. After freeze-drying which is the last procedure, the graphene quantum dots were obtained. The obtained graphene quantum dots were characterized using atomic force microscopy (AFM), Fourier-transform infrared (FT-IR) spectroscopy and Raman spectroscopy. The AFM image demonstrates the topographic morphology of obtained graphene quantum dots, the heights of which range from 0.4 to 3 nm, corresponding to 1–4 graphene layers, and the step height is approximately 2–2.5 nm. Using FT-IR, the functional groups in obtained graphene quantum dots were detected. The stretching vibrations of hydroxyl group at 3,420 cm?1, carboxylic acid (C=O) at 1,751 cm?1, C-OH at 1,445 cm?1, and C-O at 1,054 cm?1. The identified functional groups of obtained graphene quantum dots matched the functional groups which are present if it is a graphene quantum dot. In Raman spectrum, the D and G peaks, which are the characteristics of graphene quantum dots, were detected at wavenumbers of 1,380 cm?1 and 1,580 cm?1, respectively. Thus, it was verified that the graphene quantum dots could be successfully synthesized from the ionexchange resin.