Radioactive iodine is one of the major constituents among fission products released into the environment from damaged containment at nuclear power plant (NPP) during severe core melt accidents such as Chernobyl and Fukushima as well asnuclear fuel reprocessing. Released radioactive iodine from nuclear reactor and nuclear severe accidents is one of the most hazardous radioactive contaminants and can create significant effect to human health. Several radioactive iodines such as I-131,I-135, I-125, and I-129 are the most important nuclides due to its high fission yield, volatility, environmental mobility, and significant radiological hazards. Among the isotopes of iodine, I-131 that has short half-life of 8 days is the most significant source term nuclide due to its large initial inventory. Also Iodine-129 is one of several radioactive iodine isotopes formed as a fission product in nuclear fuel. It has a very long half-life of 15.7×106years, and under goes beta decay and emission of gamma rays before forming stable Xe-129. Due to the seconcerns, the Environmental Protection Agency (EPA) has implemented strict regulations on the release of I-129 from the reprocessing facilities, which require an effective capture efficiency in excess of 99% (decontaminationfactor>167). This has created a significant demand for iodine sorbent sand waste forms which are suitable for use in reprocessing systems and geologic disposal.


The research goals are the followings.
(1) to increase the iodine removal efficiency and thermal stability of sorbent.
(2) to reduce processing cost and reaction time with optimized synthesis method.
(3) to understand iodine sorption mechanism and capturing iodine for making clean environment.




This results show iodine removal efficiency (94%) and this is comparable to Ag-zeolite



– Background


The radioactive waste includes radioisotope such as 129I and 131I whose half-lives are t1/2 = 1.57×107 years and t1/2 = 8.02 days, respectively. These iodine has harmful effects to humans through external and internal exposure which can bring burns to the skin and affect the thyroid gland. Therefore iodine needs to be removed safely.


– Objectives


The objective of our study is capturing iodine species such as iodide, iodate, elemental iodine, and methyl iodide. For this study, we use various adsorbents like zeolite, activated carbon, and etc and develop the adsorption efficiency.


Chemical reaction of Iodide is key point in this study. So far, Ag(silver) shows high formation constant with iodide. But, AgI can be decomposed at high temperature to immobilize post-Ag adsorbent. To solve this problem, we study anion exchange and chemical reaction with other cation.


– Methods


Nonradioactive iodine, 127I (as surrogate for radioactive iodine), solutions are prepared for iodine batch experiments. And, iodide solution is reacted with adsorbent (solid to solution ratio is 1g/1L.). After experiments, the aqueous samples are collected at different reaction times and analyzed with inductively coupled plasma mass spectroscopy (ICP-MS, NexlON 300D) to measure iodine concentrations. The capture efficiency of adsorbent for each iodine species is determined by Eq (1).




where Ci is initial iodine concentration (ppm) which was determined with blank sample and Cs is iodine concentration (ppm) in each sample at different reaction times after filtration.


– Result


So far, in our study, The result of iodide capturing test is shown in Figure 1. The iodide adsorption capacity on Sn-HY zeolite is 64.12%. It is higher than Sn-13X (15.76%) and Sn-NaY (52.55%) zeolite. It is demonstrated that Sn is well coordinated with HY zeolite for capturing iodide, compared with 13X and NaY zeolite.



Fig. 1. Iodide capturing capacity using 0.03 M Sn-13X, Sn-NaY, and Sn-HY


2.1. Background

Deep geological disposal is a promising method to dispose of spent nuclear fuel (SNF) that is being evaluated by Canada, Finland, Germany, Spain, Sweden, England, USA etc. Ground water beside the SNF canister can be decomposed by alpha (α), beta (β) and gamma (γ) radiation emitted from the SNF. Radiolysis can produce H2O2, OH-, H-, eaq-, and H2 through water decomposition. Studtite (UO4·4H2O) is formed by reaction between uranyl ion (UO22+) that may leaks from SNF and H2O2.


2.2. Objectives

The studtite formed on the SNF surface, can impede the corrosion of SNF and retard the transport of cesium and strontium through sorption reaction. Therefore, the dissolution kinetics of studtite in natural environmental conditions is important factor for the performance and risk assessment of repositories. The objectives of this study is to investigate the dissolution kinetics and mechanisms of uranium release from studtite in different solution conditions.


2.3. Methods


Fig. 2. Studtite synthesis

NaHCO3 (99.5–100.5 %, Sigma-Aldrich) was added to 50 mL of deionized water for preparing three different concentrations (10-2, 10-3, and 10-4 M HCO3-) of background solutions. The pH of these solutions were adjusted to 9 by adding 0.1 N NaOH. Both aerobic and anaerobic synthetic ground water was prepared for 50 mL to mimic natural granitic groundwater (sampled at YS-1-6 borehole). Synthesized studtite (1 mg) was added to each solution (50 mL) and started for batch dissolution test for 2 month. In each sampling time (total 12 times for 2 month), a 2 mL sample was collected and the same amount of sampling volume was replaced by the same background solution. The uranium concentration in each filtered sample was determined using inductively coupled plasma mass spectrometer (ICP-MS, NexION 300D, Perkinelmer).

2.4. Results


Fig. 3. XRD pattern and ATR-FTIR spectra of studtite


Fig. 4. Log uranium concentration vs. time in three different [HCO3-] solutions with pH 9 in aerobic condition. Uranium speciation diagram was simulated in the two conditions ([HCO3-]=10-1, 10-3 M). Error bars were determined by standard deviation of duplicates.


Fig. 5. Released uranium concentration vs. time in synthetic groundwater with pH 9 prepared under aerobic condition (filled squares, DO=4.92 mg·L-1) and anaerobic condition (clear circles, DO=0.31 mg·L-1). Uranium speciation diagram was simulated in the two conditions (aerobic and anaerobic). Error bars were determined by standard deviation of duplicates.


– Background

Efficient and rapid removal of radioactive contaminants is crucial when they are released into the environment as a result of severe nuclear accidents. In addition, various methods have been developed for harvesting uranium (U) from seawater.


– Objectives

To develop sorbent for the removal of uranium using tributyl phosphate (TBP) on the surface of Hydroxyapatite (HA). TBP is an organophosphorus extractant, and it is among the most favored extractants in the nuclear industry, which is due in part to the high stability constants of the actinide-organophosphorus complexes.

noname06Fig. 6.


Adsorbent synthesis method

noname07Fig. 7.


-Experimental condition

noname08Fig. 8.

– Results

As as NaHCO3 concentration increased, uranium removal was decreased because of U-CO3 complexation

In the same condition, TBP-apatite showed better U removal than general HA. Especially, TBP-apatite (pH=7) material show the highest U adsorption capacity (38mg/g).

Comparing two isotherm models, Freundlich isotherm is more suitable to U adsorption behavior of TBP-apatite prepared at pH=7 and 10



Chalk River Unidentified Deposit (CRUD) is a technical term in nuclear engineering which is an accumulated material on external fuel rod cladding surfaces in nuclear power plants. It is a corrosion product which is composed of either dissolved ions or solid particles such as Ni, Fe and Co. It consists mainly of NiO and NiFe2O4.It can affect to reduce fuel lifetime, degrade heat transfer to the coolant, and threaten human health and environment. Therefore, decontamination process is essential for reducing occupational exposures, limiting potential releases and uptakes of radioactive materials, allowing the reuse of components, and facilitating waste management process.



In this paper, we have conducted the synthesis of Cobalt ferrite and Nickel ferrite as powder foam to use for decontamination process. In addition, oxide layer foam on the metallic coupon (SUS 304) was prepared and used. Dissolution test of the simulated CRUD was conducted with different chemical reagents and electro kinetic (EK) process was applied to investigate for the cobalt, nickel and iron removal from CRUD.



Characterization of synthesized materials


The XRD results of synthesized CRUD are shown in Fig. 2, and Co based CRUD was well synthesized in our method. It is identified as Co-ferrite, [CoFe2O4]. X-Ray Fluorescence (XRF) data of Co-ferrite powder is shown in Table I. XRF analyses shows normalized concentration of Fe : Co = 65.61% : 31.44% in the obtained powder, which is similar to the molar ratio of Fe : Co = 2 : 1.




3.2 Dissolution test of cobalt ferrite


Amount of dissolved cobalt in cobalt ferrite at various pHs and at 70℃ is shown at Fig. 3. Dissolved amount of cobalt is higher in lower pH solution. The highest dissolved amount of Co was found when pH was titrated with oxalic acid. Temperature dependent dissolved amount of iron and cobalt in Cobalt ferrite is shown in Fig. 4 and the amount of dissolved cobalt and iron is increased as temperature increases.

noname09Fig. 9.

3.3 Dissolution test of nickel ferrite


Amount of dissolved nickel in nickel ferrite at various pHs and at 70℃ is shown at Fig. 5. At lower pH, the dissolved amount of nickel is higher. And, oxalic acid shows the highest dissolved amount of Ni compared to other acid solutions at the same pH.




3.3 Electro-Kinetic process


Amount of dissolved cobalt from cobalt ferrite using different acids at cathode of Electro-kinetic (EK) process under the pH 1.8 is shown at Fig. 9. Fig. 10 shows the amount of dissolved cobalt from cobalt ferrite using acids at cathode of Electro-kinetic (EK) process under the pH 3.5. When pH 3.5 titrated with oxalic acid, most of dissolved cobalt is precipitated. The Cobalt (ΙΙ) ions which were moved to cathode reacted with hydroxide ion (OH-)andprecipitationreactionoccurredlikeeq.(1)


*Background : Most of nuclear power plants (NPP) and their yard storage facilities are located on or close to the shoreline, because the nuclear power plants need for cooling water. As a fission product (90Sr, 137Cs, or 99Tc), radionuclides are often stored in a large water-filled tank of NPP facility at or near ground surface level. In a case of severe accident, the radionuclides may leave the storage sites and transport in the subsurface environment in which nearby seawater can introduce through shattered fractures and cause the salinization of original groundwater aquifers. Changes in porewater salinity may trigger ion exchange and dissolution-precipitation reactions that will affect the solubility and mobility of radionuclides associated with subsurface sediments.


*Objectives : The released radioactive 90Sr into the groundwater due to wreck of storage tanks can be highly mobile and transport behavior of 90Sr will change depending on the salinity degree. In this study, 90Sr was selected as the target radionuclide because of its sensitive sorption behavior via weakly bound outer-sphere surface complexes at the solid water interface. Therefore, the objective of this study was to investigate transport behavior of 90Sr through subsurface sediment at different degree of ionic strength assuming seawater intrusion in groundwater


*Methods : The effect of increased porewater salinity (e.g., ionic strength) on mobilization of 90Sr was studied using column experiments to compare the Sr transport behavior in three different solution conditions (groundwater, seawater, and ground-seawater mixed solution).

Core rock was sampled from Shin-Gori 4 nuclear reactor site in depth of 6.6-7.4 m, homogenized, and sieved following crushing process. Groundwater and seawater were collected under and near that study site and filtered with 0.45 μm before use. The pHs of solution samples were about 8 and major element compositions were analysed using ICP-OES and IC. The glass columns with a 2.5 cm ID and 5 cm length were used and packed with fractionated solid material (0.5-1.0 mm size range). Three different types of synthetic solutions of groundwater, seawater, and ground-seawater (50 : 50) mixed denoted as GW, SW, and GW+SW, respectively were used as the transport fluids to investigate the effect of ionic strength (Table 1.) on 90Sr mobility. Stable Sr was used as a surrogate for 90Sr.

Columns were packed with the crushed powder of sediment and each solution without Sr was introduced for 24 hrs to stabilize the flow inside the column. A syringe pump was used to saturate each column with fluids from bottom to top at a constant flow rate (about 0.03 mL min-1). The effluent solutions were collected at regular time intervals using a fraction collector. The porosity was determined as 0.42, 0.41, and 0.40 for each of packed columns of groundwater, seawater, and mixed solution, respectively. All the columns were characterized for hydrodynamic dispersion coefficients using a nonreactive bromide (Br) tracer with pulse type introduction. After obtaining a Br breakthrough curve, hydrodynamic dispersion coefficient was determined by the model fit using CXTFIT code (Ver. 2.0). Then, three different types of fluids spiked with Sr (0.18 mg L-1) were introduced into each column under fully saturated condition at the same constant flow rate. The collected samples were analyzed for Sr concentration using ICP-OES. The flow conditions of each column are given in Table 1.


Table.1. Summary of the column transport experiments with different ionic strength fluids

1Fluid type denotes synthetic solutions; 2Average linear velocity [cm/min]=flow rate/cross section area/porosity; 3I.S. (ionic strength) with a unit of M (mole/L).

*Results : Breakthrough curves of Br exhibited similar narrow peaks, showing that there were no preferential flow paths and the columns were uniformly packed (Fig. 1). The determined hydrodynamic dispersion coefficient from CXTFIT fit was 0.01 cm2 min-1 for GW and SW columns, and 0.02 cm2 min-1 for mixed fluid column. In case of SW and GW+SW mixed columns, the eluted Sr concentration reached the original Sr concentration (C/Co=1) in approximately 3 pore volumes (Fig. 1). However, in case of GW column, the eluted Sr concentrations reached a fully breakthrough point of C/Co=1 after 135 accumulated pore volumes (about 38 days), because of high Sr sorption in GW column condition. The mobility and sorption of Sr is controlled by competition for ion exchange sites with high Na concentration in solution. It suggests that Sr transport is relatively fast in the seawater by competition for cation exchange sites with high Na concentrations in solution. In addition, Sr sorption affinity decreased in the seawater solution, resulting in a low retardation of Sr transport. However, the mobility of Sr decreased significantly in GW column because of low amounts of the competing cations and high Sr sorption affinity in GW solution condition.

Because the mobility of Sr was controlled by sorption reactions, a fully understanding of seawater introduction into groundwater is needed to predict correctly radioactive Sr transport in a subsurface environment nearby nuclear power plants and radioactive wastes repository.


Fig.11.The breakthrough curves of the measured Br concentrations (open symbol) and CXTFIT fit (solid line). The closed symbols denote the measured Sr concentrations in the flow-through column experiments.