• Journal of Radiation Industry
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• v.5 no.4
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• pp.337-345
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• 2011

In order to provide an effective preparedness for a nuclear or radiological emergency happening in the domestic or neighborhood countries and to solve the vague fear of the people for the ingestion of radioactive livestock products, the establishment of national guideline level for radionuclides in feed is urgently necessary. This is because it is important to secure the safety and to manage the crisis in the agricultural, fishery and food sector by performing the effective safety control during and after nuclear incident. This study was performed to investigate the report cases of international organizations and foreign countries to set up a domestic control standard for managing radioactive substances that may be contaminated in animal feeds due to the nuclear power plant incident. In addition, an attempt was made to provide a useful reference that can help prepare a domestic control standard, using a coefficient that can consider the transfer into livestock through the intake of radioactive contaminated animal feeds. The standard radioisotopes investigated were confined to radioactive cesium ($^{137+134}Cs$) and iodine ($^{131}I$). Guideline level for the radionuclides was calculated by using the transfer coefficient factor and the maximum daily intake of animal feed provided by IAEA. For example, the maximum daily intake of animal feed was set as $25kg\;d^{-1}$ for dairy cows, $10kg\;d^{-1}$ for beef cattle, $3.0kg\;d^{-1}$ for pigs and $0.15kg\;d^{-1}$ for chickens. The result values for radioactive cesium were calculated as $8,696Bq\;kg^{-1}$, $4,545Bq\;kg^{-1}$, $1,667Bq\;kg^{-1}$ and $2,469Bq\;kg^{-1}$, respectively. The results for radioactive iodine showed the ranges between $741Bq\;kg^{-1}$ and $76,628Bq\;kg^{-1}$. These data can be utilized as a scientific reference for the preparation of a crisis management manual for the emergency control due to nuclear power plant accident in Korea and neighboring country. These results will contribute to establish the safe feed management system at national level as manual for responding the radioactive exposure of agricultural products and animal feeds, which are currently not established.

• Nuclear Engineering and Technology
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• v.24 no.3
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• pp.228-235
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• 1992

Some soluble species may not be solubility-limited or congruent-released with the matrix species. For example, during the operation of the nuclear reactor, the fission products can be accumulated in the fuel-cladding gap, voids, and grain boundaries of the fuel rods. In the waste package for spent-fuel placed in a geologic repository, the high solubility species of these fission products accumulated in the“gap”, e.g. cesium or iodine are expected to dissolve rapidly when ground water penetrates fuel rods. The time and space dependent mass transport for high solubility nuclides in the gap is analyzed, and its numerical illustrations are demonstrated. The approximate solution that is valid for all times is developed, and validated by comparison with an asymptotic solution and the solution obtained by the numerical inversion of Laplace transform covering the entire time span.

• Nuclear Engineering and Technology
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• v.51 no.3
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• pp.800-806
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• 2019

This paper presents a laboratory experiment on data acquisition technique that applied to the gamma radiation scanning coupled with computed tomography (CT) technique for inspection of broken nozzle inside the vertical vessel. The acquisition technique was developed to inspect a large diameter vessel when suspicious problem location is not easily accessed. This technique allows the installation of gamma radiation source (Cesium 137, Cs-137), and detectors (Sodium Iodine. NaI(Tl)) from the accessible location to the required location and performs the scanning by designed pattern. To demonstrate the designed technique, top opened tank which installed with six cut steel pipes diameter of 76.2 mm (3") at a certain position was selected. They were assumed to be a gas riser pipes inside the vessel. Three studied cases were performed, (a) projection of well installed six pipes, (b) projection of one out of six broken pipe and (c) one of nozzle was assumed to be failure and fell down until one out of six pipes was broken and obstructed by nozzle. Results clearly indicated the capability of developed technique to distinguish between normal situation case and abnormal situation cases.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.20 no.2
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• pp.231-241
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• 2022

Several countries, including Korea, are considering the direct disposal of spent nuclear fuels. The radiological safety assessment results published after a geological repository closure indicate that the instant release is the main radiation source rather than the congruent release. Three Safety Case reports recently published were reviewed and the IRF values of seven long-lived radionuclides, including relevant experimental results, were compared. According to the literature review, the IRF values of both the CANDU and low burnup PWR spent fuel have been experimentally measured and used reasonably. In particular, the IRF values of volatile long-lived nuclides, such as ¹²⁹I and ¹³⁵Cs, were estimated from the FGR value. Because experimental leaching data regarding high burnup spent nuclear fuels are extremely scarce, a mathematical modelling approach proposed by Johnson and McGinnes was successfully applied to the domestic high burnup PWR spent nuclear fuel to derive the IRF values of iodine and cesium. The best estimate of the IRF was 5.5% at a discharge burnup of 55 GWd tHM^-1.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.1
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• pp.79-84
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• 2007

Inventories of soluble elements in the gap and grain boundaries of domestic used PWR fuel pellets were measured to estimate the quantities of radionuclides that are liable to be rapidly released into the groundwater of a disposal site. The gap inventory of cesium for the pellets in the used fuel with a burn-up range of 45 to 66 GWD/MTU showed 0.85 to 1.7% of its total inventory, which was close to 1/6 to 1/3 of the fission gas release fraction (FGRF). However, the amounts of cesium released from the gaps of the pellets below 40 GWD/MTU of a burn-up and less than 1% FGRF were so erratic that the gap inventory could not be defined by ie FGRF. Strontium inventories in the gap and grain boundaries of the pellets in the same rod were not significantly varied, and the iodine inventory in the gap of the used PWR fuels was estimated to be less than or the same as the FGRF.

• Membrane Journal
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• v.26 no.5
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• pp.401-406
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• 2016

In this study, we have investigated the removal of the low level radioactive ions of Cs and I in water by the reverse osmosis (RO) process. The two RO modules produced in domestic region and the waste RO module after the cleaning process were selected. Then we compared removal performance of both Cs and I. The experiments are conducted by varying the concentration of feed, the pressure. As a results, it was confirmed that all three modules are higher I decontamination factor than Cs. And particularly, for the cleaned RO module, its decontamination factor of I was 1140. Since the results at low pressure condition were better than that at high pressure conditions, the use of the direct installation of RO modules on the tap water might be possible. In addition, it was confirmed that the waste RO module after cleaning process using EDTA, SBS and NaOH, increased the decontamination performance better than before cleaning, in particular, the recovery ratio after cleaning was 6.3% higher.

• Applied Chemistry for Engineering
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• v.27 no.6
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• pp.633-638
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• 2016

In this study, we used three kinds of commercially available cation, anion, and mixed-ion exchange resins to separate radioactive ions from a polluted water containing Cs, I, and other radioactive ions. The experiment was conducted at a room temperature with a batch method, and a comparative analysis on the decontamination ability of each resin for the removal of Cs and I was performed by using different quantities of resins. The concentration was analyzed using ion chromatography and the ion exchange resin product from company D showed an overall high ion exchange ability. However, for most of the experiments when the amount of ion exchange resin was decreased, the decontamination ability of the resins against mass increased. When the mass of company D's cation exchange resin was small, the ion exchange ability against Cs and I ions were measured as 0.199 and 0.344 meq/g, respectively. When the mixed ion exchange resin was used, the ion exchange ability against I ions was measured as 0.33 meq/g. All in all, company D's ion exchange resins exhibited a relatively higher ion exchange ability particularly against I ions than that of other companies' exchange ions.

• Journal of Radiation Protection and Research
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• v.42 no.2
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• pp.114-129
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• 2017

Background: The purpose of this paper is to confirm the event timings and the magnitude of fission product aerosol release from the Fukushima accident. Over a few hundreds of technical papers have been published on the environmental impact of Fukushima Daiichi accident since the accident occurred on March 11, 2011. However, most of the research used reverse or inverse method based on the monitoring of activities in the remote places and only few papers attempted to estimate the release of fission products from individual reactor core or from individual spent fuel pool. Severe accident analysis code can be used to estimate the radioactive release from which reactor core and from which radionuclide the peaks in monitoring points can be generated. Materials and Methods: The basic material used for this study are the initial core inventory obtained from the report JAEA-Data/Code 2012-018 and the given accident scenarios provided by Japanese Government or Tokyo Electric Power Company (TEPCO) in official reports. In this research a forward method using severe accident progression code is used as it might be useful for justifying the results of reverse or inverse method or vice versa. Results and Discussion: The release timing and amounts to the environment are estimated for volatile radioactive fission products such as noble gases, cesium, iodine, and tellurium up to 184 hours (about 7.7 days) after earthquake occurs. The in-plant fission product behaviors and release characteristics to environment are estimated using the severe accident progression analysis code, MELCOR, for Fukushima Daiichi accident. These results are compared with other research results which are summarized in UNSCEAR 2013 Report and other technical papers. Also it may provide the physically based arguments for justifying or suspecting the rationale for the scenarios provided in open literature. Conclusion: The estimated results by MELCOR code simulation of this study indicate that the release amount of volatile fission products to environment from Units 1, 2, and 3 cores is well within the range estimated by the reverse or inverse method, which are summarized in UNSCEAR 2013 report. But this does not necessarily mean that these two approaches are consistent.