Republic of Korea (ROK) is operating the Integrated Environmental Radiation Monitoring Network (IERNet) in preparation for a radioactive emergency based on Article 105 of the Nuclear Safety Act (Monitoring of Nationwide Radioactive Environment). 215 radiation monitoring posts are monitoring a wide area, but their location is fixed, so they can’t cover areas where the post is not equipped around the Nuclear Power Plants (NPPs). For this, a mobile radiation monitoring system was developed using a drone or vehicle. However, there are disadvantages: it is performed only at a specific cycle, and an additional workforce is required. In this study, a radiation monitoring system using public transportation was developed to solve the above problems. Considering the range of dose rates from environmental radiation to high radiation doses in accidents, the detector was designed by combining NaI (TI) (in the low-dose area) and GM detector (in the high-dose area). Field test was conducted by installed on a city bus operated by Yeonggwang-gun to confirm the performance of the radiation monitoring system. As a result of the field test, it was confirmed that data is transmitted from the module to the server program in both directions. Based on this study, it will be possible to improve the radiation monitoring capability near nuclear facilities.

Investigations and monitoring of environmental radiation are important for preventing expected accidents or for early detection of unexpected accidents, in nuclear facilities and the surrounding. In the event of an environmental radiation accident, it should be possible to identify and analyze the radiation-contaminated area. Therefore, a rapid radiation monitoring system is required for immediate response and necessary measures. In this study, the distribution of radiation mapping is performed on a contaminated area using 2-dimensional or 3-dimensional contour mapping techniques. The entire surrounding area can be understood at a glance by displaying the radiation contour line on the map of the measured area.

In Korea, Kori Unit 1, a commercial pressurized water reactor (PWR), was permanently shut down in June 2017, and an immediate decommissioning strategy is underway. Therefore, it is essential to understand the characteristics of radioactive waste during the decommissioning process of nuclear power plants (NPP). Because radioactive waste must be handled with care, radioactive waste is treated in a hot cell facility. Hot cell facility handles radioactive waste, and worker safety is essential. In this study, it was dealt with whether or not the radiation safety regulations were satisfied when processing the core beltline metal of the dismantling waste treated at the post irradiation examination facility (PIEF) of the hot cell facility. Core beltline metal used for the pressure vessel in the reactor is carbon steel, and it is continuously irradiated by neutrons during the operation of the NPP. A radiological safety estimation of the behavior of radioactive aerosols during the cutting process within the PIEF was carried out to ensure the safety of the environment and workers. When processing the core beltline metal in PIEF, dominant six nuclides (60Co, 63Ni, 55Fe, 3H, 59Ni, 14C) of aerosol are generated. Accordingly each cutting device, amount of aerosol and value of dose is different. Using a 99.97% efficiency HEPA filter, the emission concentration of the dominant nuclides (60Co, 63Ni, 55Fe, 3H, 59Ni, 14C) in the air source term was satisfied with the emission control standard of Nuclear Safety Commission No. 2016-16. It was confirmed that the radioactivity concentration in the airborne source term inside the PIEF is in equilibrium state, when ventilation is considered. Also, the mass of aerosol and the concentration of airborne source term differed according to the thickness of the saw blade of the cutting tool, and the exposure dose of the worker was different through Monte Carlo N-Particle (MCNP). At that time, 60Co accounted for 95.4% of the exposure dose, showing that 60Co had the highest impact on workers, followed by 55Fe with 2.7%. The worker’s dose limit is satisfied in accordance with Article 2 of the Nuclear Safety Act and the dose limit of radiation-controlled area is found to be satisfied in accordance with Article 3 of the rules on technical standards for radiation safety management at this time.

The IAEA recommended considerations for exemption regulations of consumer products containing greater amounts of radioactive isotopes than the amounts specified for generic exemption. One of the major considerations is the expected exposure dose should be less than 10 μSv/y and 1 mSv/y for general cases and low probability cases, respectively, in all predictable scenarios. Under this recommendation, many countries evaluated the radiation dose for exposure scenarios of various products in consideration of the national circumstances and, then, established their own specific exemption regulation. In Republic of Korea, the “Regulation on substances excluded from radioactive isotopes” was legislated to specify consumer products excluded from regulation. However, as the usage status and product specifications has changed over time, it is necessary to periodically verify the validity of the regulation criteria in the view of exemption justification. In this study, we developed the use and disposal scenarios in consideration of the domestic use of thorium-containing gas mantle and evaluated radiation dose of each scenario accordingly. The gas mantles are used as a wick for gas lanterns and the maximum activity of natural thorium contained among the currently available gas mantles is 12.5 kBq. Radioactive isotopes in the decay chain of natural thorium can be divided into three groups according to their physical characteristics, and exposure routes suitable for each group were considered in dose calculation. Currently, most gas mantles are installed in camping lanterns. Therefore, we developed use scenarios related to camping. The average number of camping trips and time spent at the campground were set by the data from Korea Tourism Organization. Tent sizes and vehicle specifications were determined by referring to surveys and products in Korea. The used gas mantle is disposed of in a garbage bag for general waste and transported to landfill or incinerator. We determined the amount of gas mantle discarded in landfill and incinerator by the data from Korea Environment Corporation. The exposure time and amount handled by an individual were determined by considering the number of waste collection vehicles, landfills, and incinerators. Although we assumed the maximum activity of the gas mantle for conservative evaluation, the calculated radiation doses for the use and disposal scenarios were below the general requirement (i.e., 10 μSv/y) in all scenarios.

Minimizing of radiation exposure for the operating and decommissioning personnel is a key indicator for safe operation of the NPP. This is reflected in the application of the ALARA (As Low As Reasonable Achievable) principle. The main objectives of radiation management during full system decontamination for NPP decommissioning are to reduce the exposure dose, prevent contamination of the body and reduce solid radioactive waste. In order to reduce exposure of workers, the dose rate should be reduced by installing a temporary shield after evaluating the dose rate for the piping, component and decontamination equipment of the decontamination path before full system decontamination. Furthermore, unnecessary exposure to radiation should be reduced by thoroughly entering and exciting the radiation area and limiting the access to the high-radiation area except for workers or persons concerned. A telemetric dosimetry system should be as installed to remotely monitor radiation levels at different locations within the decontamination flow path. Remote monitoring of radiation fields using teledosimetry worked well in assessing process effectiveness and is highly recommended. However, care must be taken to place the detectors in appropriate locations. For the prevent of body contamination, it is necessary to install a fence using a heat-resistant waterproof sheet to prevent leakage of highly radioactive contamination water. When replacing high-dose filters and ion exchange resins, it is necessary to remotely monitor to reduce the exposure dose of workers.

Organic scintillator is easy to manufacture a large size and the fluorescence decay time is short. However, it is not suitable for gamma measurement because it is composed of a low atomic number material. Organic scintillation detectors are widely used to check the presence or absence of radiation. The fluorescence of organic scintillators is produced by transitions between the energy levels of single molecules. In this study, an organic scintillator development study was conducted for use in gamma measurement, alternative materials for secondary solute used in basic organic scintillators were investigated, and the availability of alternative materials, detection characteristics, and neutron/gamma identification tests were performed. In other words, a secondary solute showing an improved energy transfer rate than the existing material was reported, and the performance was evaluated. 7-Diethylamino -4-methylcoumarin (DMC), selected as an alternative material, is a benzopyrone derivative in the form of colorless crystals, has high fluorescence and high quantum yield in the visible region, and has excellent light stability. In addition, it has a large Stokes shift characteristic, and solubility in solvent is good. Through this study, it was analyzed that the absorption wavelength range of DMC coincided with the emission wavelength range of PPO, which is the primary solute. Through this study, it was confirmed that the optimal concentration of DMC was 0.04wt%. As a result of performing gamma and neutron measurement tests using a DMC-based liquid scintillator, it showed good performance (FOM=1.42) compared to a commercial liquid scintillator. Therefore, the possibility of use as a secondary solute was demonstrated. Based on this, if studies on changes in the composition of secondary solute or the use of nanoparticles are conducted, it will be possible to manufacture and utilize a scintillator with improved efficiency compared to the existing scintillator.

The safe, efficient and cost-effective decommissioning and dismantling of radioactive facilities requires the accurate characterization of the radionuclide activities and dose rate environment. And it is critical across many nuclear industries to identify and locate sources of radiation accurately and quickly. One of the more challenging aspects of dealing with radiation is that you cannot see it directly, which can result in potential exposure when working in those environments. Generally, semiconductor detectors have better energy resolution than scintillation detectors, but the maximum achievable count rates are limited by long pulse signals. Whereas some high pure germanium detectors have been developed to operate at high count rates, and these HPGe detectors could obtain gamma-ray spectra at high count rates exceeding 1 Mcps. However, HPGe detectors require cooling devices to reduce the leak currents, which becomes disadvantageous when developing portable radiation detectors. Furthermore, chemicalcompound semiconductor detectors made of cadmium telluride and cadmium zinc telluride are popular, because they have good energy resolution and are available at room temperature. However, CdTe and CZT detectors develop irradiation-induced defects under intense gamma-ray fields. In this Review, we start with the fundamentals of gamma rays detection and review the recent developments in scintillators gamma-ray detectors. The key factors affecting the detector performance are summarized. We also give an outlook on the field, with emphasis on the challenges to be overcome.

In general, dose assessment must be performed to obtain approval for clearance of radioactive waste. If the annual dose criteria through dose evaluation satisfies the clearance condition, radioactive waste can be disposed of. Various programs are used to perform dose assessment. NRCDOSE GASPAR is used as a program to assess the amount of radiation exposed to atmospheric emissions. Program is easy to use and results can be checked immediately after execution. GASPAR requires main input factors by exposure route such as site specifics, source term, special location, block data. Basically, program has default input values but user can easily modify it. The most important factor is that when entering a nuclide, the effect on progeny radionuclides is not automatically calculated. User should consider the dose contribution from progeny radionuclides. In this study, dose assessment was performed for combustible waste incineration using NRCDOSE GASPAR. And it was confirmed that exposure dose of individuals and groups criteria for clearance regulation.

The Nuclear Cycle Experiment Research Center is one of the facility of the Korea Atomic Energy Research Institute (KAERI). This facility is a laboratory-scale version of pyro-processing technology. Mixture depleted Uranium (DU) and depleted Uranium (DU) feed material are used in this facility for pyro-research. During summer, air conditioners that maintain temperature and humidity are always in operation to protect analysis equipments. 15 air conditioners are installed in this facility. The condensate which is generated in 15 air conditioners is collected in one place to analyze. Sampling was performed to check the level of contamination, U, pH and gamma radiation test were performed. This paper shows the degree of contamination of air conditioner condensate which is generated in the radiation management area.

The liquid radioactive waste system of nuclear power plants treats radioactive contaminated wastes generated during the Anticipated Operational Occurrence (AOO) and normal operation using filters, ion exchange resins, centrifuges, etc. When the contaminated waste liquid is transferred to an ion exchanger filled with cation exchange resin and anion exchange resin, nuclides such as Co and Cs are removed and purified. The lifespan and replacement time of the ion exchange resin are determined by performing a performance test on the sample collected from the rear end of the ion exchanger, and waste ion exchange resin is periodically generated in nuclear power plants. In the general industry, most waste resins at the end of their lifespan are incinerated in accordance with related laws, but waste resins generated from nuclear power plants are disposed of by clearance or stored in a HIC (High Integrity Container). Plasma torch melting technology can reduce the volume of waste by using high-temperature heat (about 1,600 degrees) generated from the torch due to an electric arc phenomenon such as lightning, and secure stability suitable for disposal. Plasma torch melting technology will be used to check thermal decomposition, melting, exhaust gas characteristics, and volume reduction at high temperatures, and to ensure disposal safety. Through this research, it is expected that the stable treatment and disposal of waste resins generated from nuclear power plants will be possible.

Radiation dose rates for spent fuel storage casks and storage facilities of them are typically calculated using Monte Carlo calculation codes. In particular, Monte Carlo computer code has the advantage of being able to analyze radiation transport very similar to the actual situation and accurately simulate complex structures. However, to evaluate the radiation dose rate for models such as ISFSI (Independent Spent Fuel Storage Installation) with a lot of spent fuel storage casks using Monte Carlo computational techniques has a disadvantage that it takes considerable computational time. This is because the radiation dose rate from the cask located at the outermost part of the storage facility to hundreds of meters must be calculated. In addition, if a building is considered in addition to many storage casks, more analysis time is required. Therefore, it is necessary to improve the efficiency of the computational techniques in order to evaluate the radiation dose rate for the ISFSI using Monte Carlo computational codes. The radiation dose rate evaluation of storage facilities using evaluation techniques for improving calculation efficiency is performed in the following steps. (1) simplified change in detailed analysis model for single storage cask, (2) create source term for the outermost side and top surface of the storage cask, (3) full modeling for storage facilities using casks with surface sources, (4) evaluation of radiation dose rate by distance corresponding to the dose rate limit. Using this calculation method, the dose rate according to the distance was evaluated by assuming that the concrete storage cask (KORAD21C) and the horizontal storage module (NUHOMS-HSM) were stored in the storage facility. As a result of calculation, the distance to boundary of the radiation control area and restricted area of the storage facility is respectively 75 m / 530 m (KORAD21C case), and 20 m / 350 m (NUHOMS-HSM case).

The saturation rates of the spent fuel (SF) wet storage at the Kori Nuclear Power Plant (NPP), Hanbit, and Hanul are 83.3%, 74.2%, and 80.8% as of the fourth quarter of 2021. The storages of Kori NPP and Hanbit NPP are expected to be saturated in 2031, and Hanul is expected to be saturated in 2032. Therefore, the construction of an interim storage facility to store the SF temporarily stored in the NPP was planned, and preparations for the safe transport of the SF are required. In this paper, radiological preliminary assessment using NRC-RADTRAN in the process of sea transport of SF from the wet storage or ISFSI of the Hanbit NPP to the optional interim storage facility was performed. Since domestic SF transport vessels are not currently in operation, the specifications of the UK Pacific Grebe vessel which can carry up to 20 casks were used. The transport cask used the specifications of KORAD-21, a transport container developed in Korea. Because it can carry more SF assemblies than the existing KN-18. In addition, a land transport safety test was conducted in 2020 and a sea transport test is planned. The sea transport route was entered by referring to the transport route of domestic low and intermediate level waste. The accidents rate was calculated using statistics on maritime accidents from 2017 to 2021. The probability accidents along the transportation route were evaluated as 3.152E -10. When transporting to an interim storage facility, the SF expected to be the main transport target was selected as WH 17X17, combustion 45,000 MWD/MTU, and concentration of 4.5%. The source term was calculated and entered according to this data and the release fraction was entered with reference to the DOE report. In the case of normal transport without accident, the individual dose of the crew member and public residents were estimated to be 0.0525% and 0.000492% of the annual limit of 1 mSv/yr for the general public. Under the accident conditions, the annual individual doses of residents were 0.0011%, 0.0023%, 0.0034%, and 0.0046% of the annual limit of 1 mSv/yr when carrying 5, 10, 15, and 20 casks. Currently, the site of the interim storage facility has not been precisely determined, but a preliminary radiation assessment through sea transport resulted in a significantly lower than the limit. Combined scenario sea transport followed by land transport will be carried out in the next stage of study.

고품질의 작물을 재배하기 위해 광은 필수적인 환경조건이 다. 겨울철에는 다른 계절에 비해 일사량이 저조하므로 보광 처리를 이용해 작물의 생육과 수확량을 증대시킬 수 있다. 본 연구는 약광기 동안 고추 온실재배를 위한 경제적인 보광 광 원을 선발하기 위해 수행되었다. 풋고추(Capsicum annuum ‘Super Cheongyang’)는 2019년 9월 5일에 정식하였다. 보광 처리는 2020년 1월 1일부터 2020년 3월 31일까지 수행되었 다. White LED(R:G:B = 5:3:2, W LED), RB LED(red:blue = 7:3, RB LED), 고압나트륨등(high pressure sodium lamp, HPS)을 광원으로 사용하였다. 무처리를 대조구로 사용하였 다. 고추의 초장, SPAD, 마디 수는 보광 광원에 따른 유의적인 차이가 없었다. 그러나 분지 수는 RB LED 광원에서 가장 많 았다. 또한 보광은 고추의 광합성을 증가시켰으며, 특히 RB LED에서 보광기간 동안 가장 높은 광합성률을 보였다. 또한 고추의 수확량은 보광처리에서 증가하였고, RB LED는 다른 광원에 비해 가장 높은 수확량을 보였다. 소비전력은 W LED 가 가장 높았고 HPS 조명이 가장 낮았다. 경제적인 측면에서 RB LED를 이용한 보광처리는 다른 광원에 비해 높은 경제성 을 가졌다. 결론적으로 이러한 결과는 고추 온실에서 약광기 동안 보광 광원으로 RB LED를 사용하는 것이 수확량과 경제 성을 향상시킬 수 있을 것으로 판단된다.

This study is about the production of radiation sources of simulated concrete and soil reference materials to verify the validity of the quality establishment and measurement of the detector (HPGe) of the radioactive soil and concrete waste classification system, which is being developed to quickly and accurately classify nuclear decommissioning waste. Specific activity of gamma nucleus among radioactive wastes is evaluated using gamma spectroscopy. At this time, in order to verify the validity and reliability of measuring equipment, it shall be a standardized substance of the same medium as nuclear decommissioning waste (chemical ingredients, particles, density, etc.) in order to correct the energy and efficiency of gamma nuclide analysis equipment. The CRM used for the detector’s energy correction used a 1 L Marinelli beaker standard correctional radiation source consisting of 10 radioactive isotopes. In order to correct efficiency, in accordance with the production and certification process of the Korea Standards and Research Institute, it has produced artificial simulated radioactive concrete similar to nuclear decommissioning waste (30% for cement, 60% for regulation and 10% for bentonite). The radioactive homogeneity of the simulated concrete reference materials was evaluated using dispersion analysis (ANOVA) in accordance with ISO Guide 35, while 137Cs and 60Co of concrete reference materials were able to obtain homogeneous measurements both in and between bottles. The self-absorption rate of the simulated concrete reference material was determined by the MCNP computer simulation measurement method, and the self-absorption correction coefficients of 137Cs and 60Co were assessed at 0.995 and 0.996, respectively, and the standard value for the radiation of the simulated concrete reference material was calculated on the weighted average of the measurements of 20 samples. The uncertainty about the reference value was calculated by combining measurement uncertainty (Type B evaluation), bottle to bottle standard deviation, and uncertainty within bottle by modifying the formula suggested in ISO Guide 35. The concentration of 137Cs and 60Co of reference materials was divided into high-speed measurement mode and precision measurement mode in consideration of the self-disposal standard. The reference value and uncertainty of expansion among reference materials for high-speed measurement mode were rated at 1,032.7 ± 64.0 Bq·kg−1and 1,083.7 Bq·kg−1, respectively. The standard value and expansion uncertainty for 137Cs and 60Co among reference materials for precision measurement mode were rated at 113.7 ± 10.0 Bq·kg−1 and 122.3 ± 10.3 Bq·kg−1, respectively.

It is essential to provide a safe working environment for radiation workers. At a research reactor decommissioning site in Seoul (KRR1 & KRR2), radioactive waste drum disposal work is in progress. Before performing radiation work, it is necessary to determine the radioactivity of the waste drum to ensure safety. In this reason, we conducted a study to determine the detection efficiency of waste drums using the EXVol code. Determination of the full energy absorption peak efficiency (detection efficiency) is one of the important processes of the gamma-ray activation analysis. For the large voluminous gamma-ray sources like waste drum, the geometrical and attenuation effect should be considered. EXVol (Efficiency calculator for eXtended Voluminous source) code is a detection efficiency calculation code using the effective solid angle method. EXVol can calculate both coaxial and asymmetric structure. In addition, the introduction of a collimator made it possible to reduce the radiation intensity of a high radiation source. And it is possible to determine the precise detection efficiency according to the energy of a gamma ray at a specific position of the volume source. To verify the performance of the EXVol, a high resolution gamma spectroscopy system was constructed and measurement and analysis were performed. Measurements were performed on coaxial, asymmetric and collimated structures with standard point source, standard 1 L liquid volume source and HPGe detector. The measured results were compared with the calculation results of EXVol. The relative deviation of the measurement and calculation in the coaxial and asymmetric structures was 10%, and that of the collimation structure was 20%. Results can be available in analysis of waste drums’ radioactivity determination at a specific position.

“Radiation measurement situation virtual reality program” is a three-dimensional modeling program. It is a program that virtually implements the measurement situation by applying visualization techniques such as animation effect charts, and effectively delivers gamma radiation energy spectrum data. This program was developed to respond to various measurement situations by visualizing major analysis objects such as radiation detectors, radioactive waste drums, radioactive building structures, and the ground surface as a 3D model. User-friendliness was secured by supporting various control functions such as distance, size, and angle while checking the three-dimensionally produced detector and analysis target in a virtual space. By using high-resolution photos obtained through 360-degree shooting, a virtual space was implemented to approximate the actual situation, such as the ground surface measurement. In addition, data communication safety was secured so that a large number of users could use it through a local area network in consideration of the actual operating environment.

To rationalize the protection of spent nuclear fuel transport storage cask, we intend to investigate the status of domestic and foreign safety regulations and related technologies to develop sabotage scenarios and analyze the protection performance and radiation impact of transport storage cask. It is essential to conduct an aircraft collision safety evaluation on spent nuclear fuel transportation and storage casks in Korea due to changes in laws and regulations related to nuclear power plant design and demand for enhanced safety. Domestic and foreign research on the protection performance of spent nuclear fuel transport storage cask was based on 9.11 events, and the results of all studies show that the speed of the aircraft and leakage of nuclear materials are insignificant. The Sandia National Laboratory (SNL) calculates Aerosol emissions from spent fuel damage in the event of sabotage and calculates Source Term based on the Durbin-Luna model. In this paper, radiation sensitivity analysis was performed due to damage to the carrier according to the size of the accident, assuming that there was a hole enough to basket from the external shell among the collision scenarios identified for domestic cask models.

Gamma-ray spectroscopy, which is an appropriate method to identify and quantify radionuclides, is widely utilized in radiological leakage monitoring of nuclear facilities, assay of radioactive wastes, and decontamination evaluation of post-processing such as decommissioning and remediation. For example, in the post-processing, it is conducted to verify the radioactivity level of the site before and after the work and decide to recycle or dispose the generated waste. For an accurate evaluation of gamma-ray emitting radionuclides, the measurement should be carried out near the region of interest on site, or a sample analysis should be performed in the laboratory. However, the region is inaccessible due to the safety-critical nature of nuclear facilities, and excessive radiation exposure to workers could be caused. In addition, in the case of subjects that may be contaminated inside such as pipe structures generated during decommissioning, surveying is usually done over the outside of them only, so the effectiveness of the result is limited. Thus, there is a need to develop a radiation measurement system that can be available in narrow space and can sense remotely with excellent performance. A liquid light guide (LLG), unlike typical optical fiber, is a light guide which has a liquid core. It has superior light transmissivity than any optical fiber and can be manufactured with a larger diameter. Additionally, it can deliver light with much greater intensity with very low attenuation along the length because there is no packing fraction and it has very high radiation resistant characteristics. Especially, thanks to the good transmissivity in UV-VIS wavelength, the LLG can well transmit the scintillation light signals from scintillators that have relatively short emission wavelengths, such as LaBr3:Ce and CeBr3. In this study, we developed a radiation sensor system based on a LLG for remote gamma-ray spectroscopy. We fabricated a radiation sensor with LaBr3:Ce scintillator and LLG, and acquired energy spectra of Cs-137 and Co-60 remotely. Furthermore, the results of gamma-ray spectroscopy using different lengths of LLG were compared with those obtained without LLG. Energy resolutions were estimated as 7.67%, 4.90%, and 4.81% at 662, 1,173, and 1,332 keV, respectively for 1 m long LLG, which shows similar values of a general NaI(Tl) scintillator. With 3 m long LLG, the energy resolutions were 7.92%, 5.48%, and 5.07% for 662, 1,173, and 1,332 keV gamma-rays, respectively.

As the decommissioning of nuclear power plants increases, there is an increasing interest in the amounts of radioactive waste. Especially, the radiation dose limit for packaging of radioactive wastes shall not exceed 2 mSv·h−1 and 0.1 mSv·h−1 on contact and at 2 m, respectively in South Korea. The DEMplus provides various environmental geometry and all properties such as materials, absorptions, and reflections and the estimation of the radiation dose rates is based on the radiation interactions of the designed 3D geometry model. With the consideration of the radiation dose rate by using DEMplus and its strategy of packaging plan, the radiation shielding was optimized and estimated in this paper. The modular shielded containers (MSC) with shielding inserted were used for radioactive wastes that require shielded packaging. In order to verify the accuracy of the estimated radiation dose rate by using DEMplus, the estimated results were compared with those obtained using MicroShield. The trends of the estimated radiation dose rates using DEMplus and the estimation of MicroShield were similar to each other. The results of this study demonstrated the feasibility of using DEMplus as a means of estimating the radiation dose limit in packaging plan of the radioactive waste.

This study is for evaluation and optimization of workers’ radiation exposure for dismantling Reactor Vessel (RV) at Kori unit 1 in connection with its decommissioning process for the purpose of establishing Radiation Safety Management Plan. This is because the safety of workers in a radiation environment is an important issue. The basis of radiological conditions of this evaluation is supposed to be those of 10 years after the permanent shutdown of Kori unit 1 when dismantling work of Reactor Vessel would suppose to be started. Dose rates of work areas were evaluated on the basis of spatial dose rate derived from activation level calculated by MCNP (Monte Carlo N-Particle Transport) and ORIGEN-S code. RV are radiated by neutrons during operation, creating an environment in which it is difficult for operators to access and work. Therefore, the RV must be dismantled remotely. However, due to work such as installing devices or dismantling surrounding structures, it is not possible to completely block the access of workers. Accordingly, the exposure of workers to the RV dismantling process should be assessed and safety management carried out. The dismantling process of Kori unit 1 RV was developed based on in-situ execution in atmospheric environment using the oxigen-propane cutting technology as the following steps of Preparation, Dismantling of Peripheral Structures, Dismantling of RV and Finishing Work. For evaluation of exposure of RV dismantling work, those processes of each steps are correlated with spatial dose rates of each work areas where the jobs being done. Results of the evaluation show that workers’ collective dose for RV dismantling work would be in the range of 536–778 man- mSv. The most critical process would be dismantling of upper connecting parts of RV with 170–256 manmSv while among the working groups, the expert group performing dismantling of ICI (In-core instrumentation) nozzles and handling & packaging of cut-off pieces is evaluated as the most significantly affected group with 37.5–39.4 man- mSv. Based on the evaluation, improvement plan for better working conditions of the most critical process and the most affected workers in terms of radiation safety were suggested.