• Journal of Radiation Protection and Research
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• v.29 no.4
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• pp.257-261
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• 2004

A coded-aperture gamma camera was developed to increase the sensitivity of a pin hole camera made with a pixellated CsI(Tl) scintillator and a position-sensitive photomultiplier tube. The modified round-hole uniformly redundant array of pixel size $13{\times}11$ was chosen as a coded mask considering the detector spatial resolution. The performance of the coded-aperture camera was compared with the pin hole camera using various forms of Tc-99m source to see the improvement of signal-to-noise ratio or the improvement of the sensitivity. The image quality is much improved despite of a slight degradation of the spatial resolution. Though the camera and the test were made for low energy case, but the concept of the coded-aperture gamma camera could be effectively used for the radioactive environmental monitoring and other applications.

• Journal of Radiation Protection and Research
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• v.38 no.4
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• pp.189-193
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• 2013

Annual dose on the containment building wall of the interim storage facility at normal condition was calculated to estimate the dose rate transition of the facility of PWR spent nuclear fuel. In this study, source term was generated by ORIGEN-ARP with 4.5 wt% initial enrichment, 45,000 MWd/MTU burnup and 10 years cooling time. Modeling of the storage facility and the containment building and radiation shielding evaluations were conducted by MCNP code depending on the distance between the wall and the facility in the building. In the case of the centralized storage system, the distance required for the annual dose rate limit from 10CFR72 was estimated to be 50 m.

• Journal of Radiation Protection and Research
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• v.37 no.3
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• pp.129-137
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• 2012

In the 2007 recommendation, the ICRP evolves from the previous process-based system of practices and intervention to the system based on the characteristics of radiation exposure situation. In addition, ICRP recommends the application of source-related dose constraints under the planned exposure situation as a tool for the optimization of protection to workers and the member of public. In this study, the analysis of radioactive effluents from Korean nuclear power plants and the public dose assessment were conducted in reference with the use of dose constraints. Finally, the measure to implement the dose constraints for the member of public was suggested taking into account multi-unit reactors operating at a single site in Korea.

• Journal of Radiation Protection and Research
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• v.10 no.2
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• pp.155-159
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• 1985

In general, dose rates for a monoenergetic gamma emitter uniformly distributed in an infinite cloud have been calulated by using the monoenergetic point-isotorpic source kernel technique. The most serious limitation on use of the kernel technique is subjected to the fact that it estimates the dose only at the surface of body. As a result, an alternative method is presented in which estimates of dose rate for immersion in a radioactive cloud are resulted from the scattered photon spectra incident on the surface of body. The results are in excellent agreement with other's. Work is currently in progress to apply these results to immersion dose problems associated with absorbed dose distribution in the MIRD phatom.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.6 no.2
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• pp.111-117
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• 2008

The new type phoswich detector consisting of the ZnS(Ag) and plastic scintillator for alpha/beta-ray simultaneous counting was developed for monitoring radiological contamination inside pipes. The detection performance was estimated using the PSD (pulse shape discrimination) method as a function of distance between the scintillator and radioactive source. The attenuation of particles traveling through a thin film for preventing the detector from being contaminated was experimentally estimated. It is concluded from our investigation that the phoswich detector developed can provide a sufficient alpha/beta-ray discrimination. The application of a thin film for preventing the detector from being contaminated was proven to be feasible.

• Nuclear Engineering and Technology
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• v.41 no.6
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• pp.867-874
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• 2009

Solubility of a radionuclide is important for defining the release source term of a radioactive waste in the safety and performance assessments of a radioactive waste repository. When the pH and redox potential of the KURT groundwater were changed by an electrical method, the concentrations of uranium and thorium released from $UO_2$(cr) and $ThO_2$(cr) at alkali pH(8.1 ${\sim}$ 11.4) and reducing potential (Eh < -0.2 V) conditions were less than $10^{-7}mole/L$. Unexpectedly, the concentration of tetravalent thorium is slightly higher than that of uranium at pH = 8.1 and Eh= -0.2 V conditions, and this difference may be due to the formation of hydroxide-carbonate complex ions. When $UO_2$(s) and $UO_2$(am, hyd.), and $ThO_2$(s) and $Th(OH)_4(am)$ were assumed as solubility limiting solid phases, the concentrations of uranium and thorium in the KURT groundwater calculated by the PHREEQC code were comparable to the experimental results. The dominating aqueous species of uranium and thorium were presumed as $UO_2(CO_3)_3^{4-}$ and $Th(OH)_3CO_3^-$ at pH = 8.1 ${\sim}$ 9.8, and $UO_2(OH)_3^-$ and $Th(OH)_4(aq)$ at pH = 11.4.

• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.4137-4141
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• 2021

As NPPs continue to operate, liquid waste continues to be generated, and containers are needed to store and transport them at low cost and high capacity. To transport and store liquid phase very low-level radioactive waste (VLLW), a container is designed by considering related regulations. The design was constructed based on the existing container design, which easily transports and stores liquid waste. The radiation shielding calculation was performed according to the composition change of barium sulfate (BaSO₄) using the Monte Carlo N-Particle (MCNP) code. High-density polyethylene (HDPE) without mixing the additional BaSO₄, represented the maximum dose of 1.03 mSv/hr (<2 mSv/hr) and 0.048 mSv/hr (<0.1 mSv/hr) at the surface of the inner container and at 2 m away from the surface, respectively, for a 10 Bq/g of ⁶⁰Co source. It was confirmed that the dose from the inner container with the VLLW content satisfied the domestic dose standard both on the surface of the container and 2 m from the surface. Although it satisfies the dose standard without adding BaSO₄, a shielding material, the inner container was designed with BaSO₄ added to increase radiation safety.

• Nuclear Engineering and Technology
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• v.56 no.1
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• pp.160-166
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• 2024

The inversion of radiation field distribution is of great significance in the decommissioning sites of nuclear facilities. However, the radiation fields often contain multiple mixtures of radionuclides, making the inversion extremely difficult and posing a huge challenge. Many radiation field reconstruction methods, such as Kriging algorithm and neural network, can not solve this problem perfectly. To address this issue, this paper proposes an optimized inverse distance weighted (IDW) interpolation algorithm for reconstructing the gamma radiation field. The algorithm corrects the difference between the experimental and simulated scenarios, and the data is preprocessed with normalization to improve accuracy. The experiment involves setting up gamma radiation fields of three Co-60 radioactive sources and verifying them by using the optimized IDW algorithm. The results show that the mean absolute percentage error (MAPE) of the reconstruction result obtained by using the optimized IDW algorithm is 16.0%, which is significantly better than the results obtained by using the Kriging method. Importantly, the optimized IDW algorithm is suitable for radiation scenarios with multiple radioactive sources, providing an effective method for obtaining radiation field distribution in nuclear facility decommissioning engineering.

• Journal of the Korean Society of Radiology
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• v.18 no.6
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• pp.629-637
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• 2024

Medical cyclotrons accelerate protons at high speeds to produce nuclear reactions for the production of radiopharmaceuticals. During this nuclear reaction, high-energy gamma rays and many neutrons are generated. However, it is reported that if exposed to the generated neutrons for a long period of time, the cyclotron accessories and shielding concrete will become radioactive and generate a large amount of radioactive waste when the facility is dismantled. Accordingly, this study aims to evaluate the radioactivity characteristics of different types of concrete used as shielding walls in cyclotron operating facilities. The experiment simulated GE's PETtrace 800 model and five types of concrete shielding walls using the FLUKA code based on Monte Carlo simulation. The simulated cyclotron was evaluated for its source term based on the manufacturer's standards, and the neutron fluence was evaluated according to the type of concrete shielding wall when the cyclotron was in operation. Afterwards, the sum of the radionuclides produced according to the type of concrete and the fraction of radionuclides produced according to the domestic radioactive waste disposal standards were analyzed. As a result, the reliability of the source term evaluation was secured with an error of less than 3%. The distribution of neutron fluence generated depending on the type of concrete when operating the cyclotron showed the highest result at the point of 0.02 eV. As a result of evaluating radionuclides generated depending on the type of concrete, concrete with high iron content tended to generate ⁵⁴Mn, and concrete with high oxygen content tended to generate ⁶⁰Co and ¹⁵²Eu. As a result of analyzing radioactivity characteristics according to the thickness of each type of concrete, concrete with high iron content showed a value below the allowable self-disposal concentration at 50 cm thick, and concrete with high oxygen content showed a value exceeding the allowable self-disposal concentration at 50 cm thick. It is believed that this study can be used as auxiliary data for preliminary radiological evaluation of concrete shielding walls when dismantling a cyclotron.

• 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.