• Journal of the Korean Society of Radiology
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• v.16 no.4
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• pp.373-379
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• 2022

The radiation source used for non-destructive testing have permeability and cause a scattered radiation through collisions of surrounding materials, which causes changes in the surrounding spatial dose. Therefore, this study attempted to evaluate and analyze the distribution of spatial dose by source in the working environment during the non-destructive test using monte carlo simulation. In this study, Using FLUKA, a simulation code, simulates ⁶⁰Co, ¹⁹²Ir, and ⁷⁵Se source used in non-destructive testing, The reliability of the source term was secured by comparing the calculated dose rate with the data of the Health and Physics Association. After that, a non-destructive test in the radiation safety facility(RT-room) was designed to evaluate the spatial dose according to the distance from the source. As a result of the spatial dose evaluation, ⁷⁵Se source showed the lowest dose distribution in the frontal position and ⁶⁰Co source showed a dose rate of about 15 times higher than that of ⁷⁵Se and about 2 times higher than that of ¹⁹²Ir. In addition, the spatial dose according to the distance tends to decrease according to the distance inverse square law as the distance from the source increases. Exceptionally, ⁶⁰Co, ¹⁹²Ir, and ⁷⁵Se sources confirmed a slight increase within 2 m of position. Based on the results of this study, it is believed that it will be used as supplementary data for safety management of workers in radiation safety facilities during non-destructive testing using radioactive isotopes.

• Journal of Radiation Protection and Research
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• v.15 no.2
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• pp.27-39
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• 1990

Appreciable radiation exposures certainly were occurred in the reactor burn-up, the nuelear fall-out and the surroundings of nuclear installations with radioactive effluents. Therefore, radioactive nuclides is not only potentially hazardous to workers of nuclear power plants and related industrials, but also the wokers who handle radioactive nuclides in biochemical research and nuclear medicine diagnostics. And in the case of occurring the nuclear accidents, the early medical treatment of radiation injury should be necessary but little is established medical procedures to decontaminate the victims of internal contamination of radioactive nuclides in korea. Accordingly, to achieve the basic data for protective roles and medical treatment of radiation injury, the present studies were carrid out to evaluate the decontamination of uranium by the chemical drugs. The results observed were summarized as follows: 1. The combined treatmet group of sodium bicarbonate and saline with uranyl nitrate injection simultaneously and the dithiothreitol group that was administered 30 minutes after uranyl nitrate injection were increased significantly in the change of body weight than uranyl nitrate-only group (P<0.005). 2. All the experimental groups were increased the fluid intake and urine volume on the uranyl nitrate-induced acute renal failure. but the combined treatment group of sodium bicarbonate and saline with uranyl nitrate injection simultaneously and the dithiothreitol group that was administered 30 minutes after uranyl nitrate injection have the higher increment of fluid intake and urine volume (P<0.05). 3. When sodium bicarbonate and saline was treated with uranyl nitrate injection simultaneously. and dithiothreitol was administered 30 minutes after uranyl nitrate injection. there was significantly reduced in BUN concentration (P<0.01). 4. When dithiothreitol was administered 30 minutes after uranyl nitrate injection. there was reduced more significantly on the increment of serum creatinine concentration than that observed in uranyl nitrate-only group(P<0.01). but when the combined treatment of sodium bicarbonate and saline with uranyl nitrate simultaneously, there was still. albeit much less marked. decrease in serum creatinine concentration. 5. The sodium bicarbonate and saline was treated with uranyl nitrate simultaneously and dithiothreitol was administered at 30 minutes after uranyl nitrate were excreted markedly higher urine creatinine concentration than the uranyl nitrate-only group. 6. Uranyl nitrate has been used in experimental animals to produce hydropic degeneration and swelling of proximal tubules, disappearance of microvilli and brush border or necrosis in the kidney and centrilobular necrosis, congestion, and telangiectasia of the liver. When the sodium bicarbonate and saline was treated with uranyl nitrate simultaneously, and dithiothreitol was administered. 30 minutes after uranyl nitrate, there was more marked the protective effect than uranyl nitrate-only group. Finally, if the sodium bicarbonate and saline may administered as quickly as possible each time that some risk for internal contamination, with uranium, and dithiothreitol is administered 30 minutes after uranium contamination, there ameliorates the course of uranyl nitrate-induced acute renal failure.and this effect is assocciated with prevention of uranium (heavy metal)-induced alterations in BUN, serum creatinine, urine creatinine, fluid intake, urine volume and body weight.

• The Korean Journal of Nuclear Medicine Technology
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• v.18 no.2
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• pp.57-61
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• 2014

Purpose Iodine (I-131) is one of the most widely used radioactive isotopes for therapeutic in the field of nuclear medicine. Therapeutic I-131 capsule is made out of lead to shield high energy radiation. Accurate dosimetry is necessarily required to perform safe and effective work for relative workers. The Monte Carlo method is known as a method to predict the absorbed dose distribution most accurately in radiation therapy and many researchers constantly attempt to apply this method to the dose calculation of radiotherapy recently. This paper aims to calculate distance dependent and activity dependent therapeutic I-131 capsule using GEANT4. Materials and Methods Therapeutic capsules was implemented on the basis of the design drawings. The simulated dose was determined by generating of gamma rays of energy to more than 364 keV. The simulated dose from the capsule at the distance of 10 cm and 100 cm was measured and calculated in the model of water phantom. The simulated dose were separately calculated for each position of each detector. Results According to the domestic regulation on radiation safety, the dose at 10 cm and 100 cm away from the surface of therapeutic I-131 capsule should not exceed 2.0 mSv/h and 0.02 mSv/h, respectively. The simulated doses turned out to be less than the limit, satisfying the domestic regulation. Conclusion These simulation results may serve as useful data in the prediction of hands dose absorbed by I-131 capsule handling. GEANT4 is considered that it will be effectively used in order to check the radiation dose.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.2
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• pp.103-116
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• 2006

By executing corrosion experiment on Inconel 600, 690 used to material of S/G tube in domestic NPP, this paper show estimation of amount of product such as Co-58, Co-60, Cr-51, Mn-54, Fe-59 which are main exposure cause to the workers in NPP. Therefore, Making the 12 samples consisted of Inconel 600, 690, whole corrosion experiment was carried out for 60 days(each pH by 20 days). The conditions of those tests were similar or more harsh than actual conditions of domestic NPP. The Glow Discharge Spectrometer(GDS) was used for quantitative analysis of results. The results of using GDS, the Inconel 600 corrodes more than Inconel 690 at pH 7 and pH 9. However, it is observed that Inconel 690 corrodes more than Inconel 600 at pH 4. Those results is estimated that test sections had the effect of transient. The long terms of experiment is required to minimize and solve the problem.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.4
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• pp.333-341
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• 2017

The Korea Atomic Energy Research Institute (KAERI) constructed the PyRoprocessing Integrated inactive Demonstration facility (PRIDE) to carry out experiments on engineering scale pyroprocessing. PRIDE has a large processing cell that human workers are not allowed to access, and thus the equipment inside is operated from outside using remote handling systems. It is therefore essential to examine the operability and maintainability of the equipment in view of remote handling systems, and the equipment is thoroughly examined in a mockup cell before it is installed in the processing cell. If the equipment is tested in a virtual mockup rather than in a mockup cell, the development cost can be significantly reduced. The PRIDE 3D simulator was integrated for virtual verification of equipment that will be installed in the processing cell. All remote handling devices in the actual PRIDE were also virtually installed in the PRIDE 3D simulator. The 3D model of the equipment was loaded and located in the exact position in the virtual processing cell. A scenario to replace an actual electrode was implemented in the PRIDE 3D simulator. The design of the equipment and the working procedures in the scenario were successfully evaluated. The results demonstrated that the PRIDE 3D simulator can be used successfully as an alternative to actual mockup testing.

• Journal of Radiation Protection and Research
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• v.22 no.4
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• pp.299-307
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• 1997

The individual dose equivalent, $H_p$, effective dose, E, and gender specific effective dose, $E^m$ and E$^f$, were evaluated using the male and female phantoms of MIRD type located in the radial gamma radiation field near a point source. The point sources were placed at the distances of 15, 40 and 100 cm in front of the body at different heights. Two radionuclides, $^{137}Cs$ and $^{131}I$, were selected for the illustrative examples. In terms of the gender specific effective doses, $E^f$ is higher than $E^m$ with a few exceptions, e.g. the case where the point source is at the height of reproductive organs, but the differences from the sex- averaged values are not significant enough to justify use of gender specific dose conversion factors for the radial gamma field. The ratios $H_p$/E were in the range of 1 to 3 depending on the source and dosimeter positions when the dosimeter is worn on the front surface of the torso covering from chest to lower abdomen, but varied from 0.34 to 6.5 in extreme cases. When it is assumed that the typical handling procedure of radioactive source material and the typical dosimeter position(on the chest) be respected, the dosimeters calibrated against the broad parallel field appear to provide estimates with acceptable errors for the effective dose of workers exposed to radial broad gamma field around a point source.

• Journal of Radiation Protection and Research
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• v.32 no.2
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• pp.45-50
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• 2007

In this study, uptake rates of internal organs and daily urinary excretion rates were measured to get more reliable estimation results for Korean. Radioactive iodine($^{131}I$) of $100{\mu}Ci$ was administered by ingestion to 28 adult males for the experiment and then the radioactivity in thyroid gland, liver, stomach, small intestine, kidneys, and urine was measured after time intervals of 2, 4, 6 and 24 hours. Uptake rates of each organ and daily urinary excretion rates were calculated on the basis of these experimental results. As a result, uptake rates of 19.70% for thyroid and daily urinary excretion rates of 71.12%, on the average, were indicated. The maximum of uptake rates and daily urinary excretion rates were recorded after 2 hours of administration of $^{131}I$, but those rates were decreased gradually later. It was also found that uptake rates were the highest in stomach, followed by the left kidney, liver, small intestine and right kidney except for thyroid gland. In this experiment, the calculated uptake change rate in thyroid gland after 24 hours of administration of $^{131}I$ was different from that of ICRP-54/67(30%) and ICRP-78(25%). Thus, it is necessary to apply more reliable approach, reflecting the characteristic of Korean physiology and to obtain the basic data of results using this approach for calculation of the internal adsorbed dose. In the future, this approach can be helpful for the internal dose assessment of radiation workers in a nuclear power plant or in a hospital.

• Transactions of the KSME C: Technology and Education
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• v.5 no.2
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• pp.135-143
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• 2017

The domestic unique research reactor, HANARO (Hi-flux Advanced Neutron Application ReactOr), has been constructed with the open-pool, the core is submerged in, for the multi-purpose neutron application. The reactor has a primary cooling system to remove the fission heat from the core and its connected fluidic systems. Since the works are required at the reactor pool top as a characteristic of the research reactor, the radiation shall be minimized with the operation of the hot water layer system to avoid unnecessary radiation exposure on the workers during work at the pool top. Moreover, the pool water management system is connected to the reactor pool to maintain the pool temperature below $50^{\circ}C$ to minimize the uprising radioactive gas or impurity from the colder pool bottom. For the efficient flow rate of the PWMS, the thermal capacity of heat exchanger is selected with 260 kW in the normal operation condition. In this paper, the modeling is formulated to figure out whether or not each pool temperature maintains below the temperature limit and the calculation results show that the designed PWMS heat exchanger has enough capacity with the design margin regardless of the reactor operation mode.

• The Korean Journal of Nuclear Medicine Technology
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• v.14 no.2
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• pp.26-32
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• 2010

Purpose: It is to find the way to minimize occupationally exposed dose for workers in vivo tests in each working stage within the range of the working environment which does not ruin the examination and the performance efficiency. Materials and Methods: The process of the nuclear tests in vivo using a radioactive isotope consists of radioisotope distribution, a radioisotope injection ($^{99m}Tc$, $^{18}F$-FDG), and scanning and guiding patients. Using a measuring instrument of RadEye-G10 gamma survey meter (Thermo SCIENTIFIC), the exposure doses in each working stage are measured and evaluated. Before the radioisotope injection the patients are explained about the examination and educated about matters that require attention. It is to reduce the meeting time with the patients. In addition, workers are also educated about the outside exposure and have to put on the protected devices. When the radioisotope is injected to the patients the exposure doses are measured due to whether they are in the protected devices or not. It is also measured due to whether there are the explanation about the examination and the education about matters that require attention or not. The total exposure dose is visualized into the graph in using Microsoft office excel 2007. The difference of this doses are analyzed by wilcoxon signed ranks test in using SPSS (statistical package for the social science) program 12.0. In this case of p<0.01, this study is reliable in the statistics. Results: It was reliable in the statistics that the exposure dose of injecting $^{99m}Tc$-DPD 20 mCi in wearing the protected devices showed 88% smaller than the dose of injecting it without the protected devices. However, it was not reliable in the statistics that the exposure dose of injecting $^{18}F$-FDG 10 mCi with wearing protected devices had 26% decrease than without them. Training before injecting $^{99m}Tc$-DPD 20 mCi to patient made the exposure dose drop to 63% comparing with training after the injection. The dose of training before injecting $^{18}F$-FDG 10 mCi had 52% less then the training after the injection. Both of them were reliable in the statistics. Conclusion: In the examination of using the radioisotope $^{99m}Tc$, wearing the protected devices are more effective to reduce the exposure dose than without wearing them. In the case of using $^{18}F$-FDG, reducing meeting time with patients is more effective to drop the exposure dose. Therefore if we try to protect workers from radioactivity according to each radioisotope characteristic it could be more effective and active radiation shield from radioactivity.

• The Korean Journal of Nuclear Medicine Technology
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• v.21 no.1
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• pp.44-49
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• 2017

Purpose Radiation exposure management has been strictly regulated for the radiation workers, but there are only a few studies on potential risk of radiation exposure to non-radiation workers, especially nurses in a general ward. The present study aimed to estimate the exact total exposure of the nurse in a general ward by close contact with the patient undergoing nuclear medicine examinations. Materials and Methods Radiation exposure rate was determined by using thermoluminescent dosimeter (TLD) and optical simulated luminescence (OSL) in 14 nurses in a general ward from October 2015 to June 2016. External radiation rate was measured immediately after injection and examination at skin surface, and 50 cm and 1 m distance from 50 patients (PET/CT 20 pts; Bone scan 20 pts; Myocardial SPECT 10 pts). After measurement, effective half-life, and total radiation exposure expected in nurses were calculated. Then, expected total exposure was compared with total exposures actually measured in nurses by TLD and OSL. Results Mean and maximum amount of radiation exposure of 14 nurses in a general ward were 0.01 and 0.02 mSv, respectively in each measuring period. External radiation rate after injection at skin surface, 0.5 m and 1 m distance from patients was as following; $376.0{\pm}25.2$, $88.1{\pm}8.2$ and $29.0{\pm}5.8{\mu}Sv/hr$, respectively in PET/CT; $206.7{\pm}56.6$, $23.1{\pm}4.4$ and $10.1{\pm}1.4{\mu}Sv/hr$, respectively in bone scan; $22.5{\pm}2.6$, $2.4{\pm}0.7$ and $0.9{\pm}0.2{\mu}Sv/hr$, respectively in myocardial SPECT. After examination, external radiation rate at skin surface, 0.5 m and 1 m distance from patients was decreased as following; $165.3{\pm}22.1$, $38.7{\pm}5.9$ and $12.4{\pm}2.5{\mu}Sv/hr$, respectively in PET/CT; $32.1{\pm}8.7$, $6.2{\pm}1.1$, $2.8{\pm}0.6$, respectively in bone scan; $14.0{\pm}1.2$, $2.1{\pm}0.3$, $0.8{\pm}0.2{\mu}Sv/hr$, respectively in myocardial SPECT. Based upon the results, an effective half-life was calculated, and at 30 minutes after examination the time to reach normal dose limit in 'Nuclear Safety Act' was calculated conservatively without considering a half-life. In oder of distance (at skin surface, 0.5 m and 1 m distance from patients), it was 7.9, 34.1 and 106.8 hr, respectively in PET/CT; 40.4, 199.5 and 451.1 hr, respectively in bone scan, 62.5, 519.3 and 1313.6 hr, respectively in myocardial SPECT. Conclusion Radiation exposure rate may differ slightly depending on the work process and the environment in a general ward. Exposure rate was measured at step in the general examination procedure and it made our results more reliable. Our results clearly showed that total amount of radiation exposure caused by residual radioactive isotope in the patient body was neglectable, even comparing with the natural radiation exposure. In conclusion, nurses in a general ward were much less exposed than the normal dose limit, and the effects of exposure by contacting patients undergoing nuclear medicine examination was ignorable.