• Proceedings of the Korean Radioactive Waste Society Conference
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• 2019.05a
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• pp.349-350
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• 2019

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2018.11a
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• pp.371-372
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• 2018

• Nuclear Engineering and Technology
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• v.51 no.3
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• pp.825-836
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• 2019

One of the most challenging safety precautions for workers in dynamic, radioactive environments is avoiding radiation sources and sustaining low exposure. This paper presents a sampling-based algorithm, DL-RRT*, for minimum dose walk-path re-planning in radioactive environments, expedient for occupational workers in nuclear facilities to avoid unnecessary radiation exposure. The method combines the principle of random tree star ($RRT^*$) and $D^*$ Lite, and uses the expansion strength of grid search strategy from $D^*$ Lite to quickly find a high-quality initial path to accelerate convergence rate in $RRT^*$. The algorithm inherits probabilistic completeness and asymptotic optimality from $RRT^*$ to refine the existing paths continually by sampling the search-graph obtained from the grid search process. It can not only be applied to continuous cost spaces, but also make full use of the last planning information to avoid global re-planning, so as to improve the efficiency of path planning in frequently changing environments. The effectiveness and superiority of the proposed method was verified by simulating radiation field under varying obstacles and radioactive environments, and the results were compared with $RRT^*$ algorithm output.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.43-47
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• 2003

The performance of the Radwaste System is measured in terms of generation of waste volumes, the release of radioactive materials to the environment and the occupational radiation exposure to workers. Based on our design and operating experience from PWR plants, various design goals for liquid radwaste system were developed to improve system performance. It has been making continuous effort to develop the advanced liquid radwaste processing technology for new PWR plants since 1998. The primary goal of this effort was to obtain better performance and to design a more economical liquid radwaste system. This paper describes lesson learned experience from design of the liquid radwaste system in Korea Nuclear Power Plants.

• Nuclear Engineering and Technology
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• v.55 no.1
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• pp.238-247
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• 2023

The present work represents a technical guideline for decommissioning a disused teletherapy machine model Theratron-780 and contains category one ⁶⁰Co radioactive source. The first section predicts the dose rate from the source in case of normal and radiological emergency situations via FLUKA-MC simulation code. Moreover, the dose assessment for the occupational during the whole process is calculated and compared to the measured values. A suggested cordoned area for safety and security in a radiological emergency is simulated. The second section lists the whole process's technical procedures, including (preview, dismantle, securing, transport and storage) of the disused teletherapy machine. Results show that the maximum obtained accumulated dose for occupational were found to be 24.5 ± 4.9 μSv in the dismantle and securing process in addition to 3.5 ± 1.8 μSv during loading on the transport vehicle and unloading at the storage facility. It was found that the measured accumulated dose for workers is in good agreement with the estimated one by uncertainty not exceeding 5% in normal operating conditions.

• Journal of Radiation Protection and Research
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• v.41 no.3
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• pp.310-314
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• 2016

Background: FCTC10 container is designed to transport $^{60}Co$ radioactive sources used in irradiation industry. It belongs to Type B(U) Category III (yellow) package when being loaded with a $^{60}Co$ source of $1.8{\times}10^5$ Ci. Materials and Methods: The container is constituted of shielding container, basket, protective cover and bracket. Shielding ability is provided mainly by stainless steel shells, tungsten alloy and lead among steel shells. Radiation level around the container has been calculated with both Monte Carlo simulations and measurements. Results and Discussion: It is proven that the shielding performance of the container fulfills the requirements in GB11806-2004 (Regulations for the safe transport of radioactive material, China Standard Press). Exposure doses to workers and to critical groups of public were calculated based on hypothetical exposure scene according to transport practice experience. Conclusion: The results show that doses to workers and public are less than the constraint dose considered in design, and the radiation level would be increased less than a factor of 2 under design basis accidents.

• Nuclear Engineering and Technology
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• v.53 no.6
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• pp.2019-2024
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• 2021

For evaluating the radiological safety of dismantled concrete, the process of disposal and recycling of the radioactive concrete generated during the dismantling of Kori Unit 1 is analyzed. Four scenarios are derived based on the analysis of the concrete recycling and disposal process, and the potential exposure to the workers and public during this process are calculated. VISIPLAN and RESRAD code are used for evaluating the dosages received by the workers and public in the following four scenarios: concrete inspection, transport of concrete by the truck driver, driving on a recycled concrete road, and public living near the landfilled concrete waste. Two worker exposure scenarios in the processing of concrete and two public exposure scenarios in recycling and disposal are considered; in all the scenarios, the exposure dose does not exceed the annual dose limit for each representative.

• Nuclear Engineering and Technology
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• v.48 no.6
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• pp.1368-1375
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• 2016

A radiological safety assessment study was performed for the transportation of low level radioactive wastes which are temporarily stored in Korea Atomic Energy Research Institute (KAERI), Daejeon, Korea. We considered two kinds of wastes: (1) operation wastes generated from the routine operation of facilities; and (2) decommissioning wastes generated from the decommissioning of a research reactor in KAERI. The important part of the radiological safety assessment is related to the exposure dose assessment for the incidentfree (normal) transportation of wastes, i.e., the radiation exposure of transport personnel, radiation workers for loading and unloading of radioactive waste drums, and the general public. The effective doses were estimated based on the detailed information on the transportation plan and on the radiological characteristics of waste packages. We also estimated radiological risks and the effective doses for the general public resulting from accidents such as an impact and a fire caused by the impact during the transportation. According to the results, the effective doses for transport personnel, radiation workers, and the general public are far below the regulatory limits. Therefore, we can secure safety from the viewpoint of radiological safety for all situations during the transportation of radioactive wastes which have been stored temporarily in KAERI.

• Journal of the Korean Society of Safety
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• v.8 no.4
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• pp.165-174
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• 1993

For the safety assessment of the demonstration-scale incineration plant for treating the combustible radioactive wastes, radiation doses of a worker and a residential individual were estimated. The demonstration plant showed a good performance of trial-burn tests using non-radioactive tracers with resulting In high mass reduction of around 40 times and very low emmission of dusts through a stack, which promised a high decontamination factor in an order of 10$^{7}$ . Based on the result s obtained from the trial-burns in the process, the estimation of radiation dose for workers and general publics near the plant was made using dose pathway calculation theories. The parametric values for calculation were selected from design and operational results of the process and from more conservative conditions In reference data. The estimated annual doses for workers and residential indivisuals were 3.07 $\times$ 10$^{-4}$ and 4.35 X 10$^{-8}$ $\mu$Sv/y, respectively, which were high enough to operate the process when comparing with the allowable dose limit in the regulation. The dose calculation models were quite applicable with showing an excellent safety for the process.

• Journal of the Korean Society of Radiology
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• v.6 no.3
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• pp.159-166
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• 2012

The purpose of this study was examine via a survey the knowledge related to radiation safety management (RSM) among radiation workers who operated or used a radiation generator or radioactive isotopes (radioactive isotopes, etc. hereinafter) for industrial use and to systematically analyze the changes in the survey results in order to promote a radiation safety culture for facilities where radiation is used. We administered a questionnaire to 861 radiation workers in the period from August 1 to September 5, 2011. As for the analysis method, a frequency analysis was made for the general characteristics and organization information of survey respondents, while the average and standard deviation were calculated and compared for the knowledge level of the RSM. According to the analysis results, the knowledge level of the RSM was evaluated to be high in all of the radiation workers. In conclusion, it is required to conduct a study on various factors in regards to the RSM among radiation workers. This can contribute to establishing educational programs in a timely manner to increase the awareness of safe and efficient use of radioactive materials and equipments by radiation workers.