• 방사성폐기물학회지
• /
• 제21권3호
• /
• pp.419-425
• /
• 2023

The Radiation and Decommissioning Laboratory of Central Research Institute (CRI) of Korea Hydro and Nuclear Power Co. (KHNP) performs research to technically support the effective management of radiological hazards to avoid risks to civilians, the workers, and the environment from the radiological risks. The laboratory mainly consists of three technical groups: decommissioning and SF technology group, radiation and chemistry group, and radwaste and environment group. The groups carry out various R&D such as decommissioning, spent fuel management, radiation protection, water chemistry management, and radioactive waste management. The laboratory also technically supports the calibration of radiometric instruments as a Korea Laboratory Accreditation Scheme (KOLAS), approval for decommissioning, guidance for radioactive waste management, state-of-the-art technology evaluations, and technology transfer.

• Nuclear Engineering and Technology
• /
• 제53권11호
• /
• pp.3505-3516
• /
• 2021

During nuclear facility decommissioning, workers are continuously exposed to high-level radiation. Hence, adequate path planning is critical to protect workers from unnecessary radiation exposure. This work discusses recent development in radioactive path planning and the algorithms recommended for the task. Specifically, we review the conventional methods for nuclear decommissioning path planning, analyze the techniques utilized in developing algorithms, and enumerate the decision factors that should be considered to optimize path planning algorithms. As a major contribution, we present the quantitative performance comparison of different algorithms utilized in solving path planning problems in nuclear decommissioning and highlight their merits and drawbacks. Also, we discuss techniques and critical consideration necessary for efficient application of robots and robotic path planning algorithms in nuclear facility decommissioning. Moreover, we analyze the influence of obstacles and the environmental/radioactive source dynamics on algorithms' efficiency. Finally, we recommend future research focus and highlight critical improvements required for the existing approaches towards a safer and cost-effective nuclear-decommissioning project.

• Nuclear Engineering and Technology
• /
• 제53권10호
• /
• pp.3335-3343
• /
• 2021

The impact of gamma radiation on a commercial off the shelf microcontroller board has been investigated. Three different tests have been performed to ascertain the radiation tolerance of the device from a nuclear decommissioning deployment perspective. The first test analyses the effect of radiation on the output voltage of the on-board voltage regulator during irradiation. The second test evaluated the effect of gamma radiation on the voltage characteristics of analogue and digital inputs and outputs. The final test analyses the functionality of the microcontroller when using an external, shielded voltage regulator instead of the on-board voltage regulator. The results suggest that a series of latch-ups occurs in the microcontroller during irradiation, causing increased current drain which can damage the voltage regulator if it does not have short-circuit protection. The analogue to digital conversion functionality appears to be more sensitive to gamma radiation than digital and analogue output functionality. Using an external, shielded voltage regulator can prove beneficial when used for certain applications. The collected data suggests that detaching the voltage regulator can extend the lifespan of the platform up to approximately 350 Gy.

• Journal of Radiation Protection and Research
• /
• 제38권2호
• /
• pp.91-99
• /
• 2013

사이클로트론은 그 자체의 수명에 의한 마모 파손뿐만 아니라 사용목적의 변경, 장소 이전, 업그레이드 등의 다양한 이유로 해체 또는 폐기를 경험하게 된다. 실제 미국과 유럽에서도 후자의 이유로 해체된 사례가 많고 또한 많은 양의 저준위 방사성 폐기물을 발생시켰으며 이에 따른 큰 해체 비용을 야기하였다. 유럽과 미국에서는 미래 해체 비용 감소 를 위해 많은 연구를 수행하였으며 미국에서는 허가시 해체 자금 계획(DFP, Decommissioning Funding Plan)을 제출 하도록 하고 있다. 사이클로트론 해체를 위해서는 기술적 측면(해체 절차, 제염 기술 등)과 안전성 측면(잔류 방사능, 예상선량 등)에서 해체 작업의 성취 정도를 예측함으로 해체 비용의 감소 및 방사성 폐기물관련 문제를 해결할 필요가 있다. 본 연구에서는 ANL (Argonne National Laboratory)과 벨기에(유럽위원회 주관)에서 수행된 사이클로트론 해체 사례를 분석하고 2012년 12월 수행된 국내 서울대학교병원 사이클로트론 해체 이전 사례를 살펴봄으로써 향후 사이클 로트론 해체 기준 수립을 위한 기초 자료를 제공하고자 하였다. 이를 위하여 IAEA (International Atomic Energy Agency)와 NRC (Nuclear Regulatory Commission)의 사이클로트론 해체 관련 기준을 분석하고 향후 방사성 폐기물 규제해제(이하 자체처분) 및 재사용과 해체 자금 계획(DFP)의 국내 도입 방안을 제시하였다. 도출된 자료는 사이클로 트론 해체시 방사화되는 정도를 예측하고 국내에 적용할 수 있는 효율적인 해체 요건과 기준들을 정립하는데 활용될 수 있을 것으로 판단된다.

• Journal of Radiation Protection and Research
• /
• 제45권4호
• /
• pp.178-186
• /
• 2020

Background: Reactor pressure vessel (RV) with internals (RVI) are activated structures by neutron irradiation and volume contaminated wastes. Thus, to develop safe and optimized disposal plan for them at a disposal site, it is important to perform exact activation calculation and evaluate the dose rate on the surface of casks which contain cut-off pieces. Materials and Methods: RV and RVI are subjected to neutron activation calculation via Monte Carlo methodology with MCNP6 and ORIGEN-S program-neutron flux, isotopic specific activity, and gamma spectrum calculation on each component of RV and RVI, and dose rate evaluation with MCNP6. Results and Discussion: Through neutron activation analysis, dose rate is evaluated for the casks containing cut-off pieces produced from decommissioned RV and RVI. For RV cut-off ones, the highest value of dose rate on the surface of cask is 6.97 × 10-1 mSv/hr and 2 m from it is 3.03 × 10-2 mSv/hr. For RVI cut-off ones, on the surface of it is 0.166 × 10-1 mSv/hr and 2 m from it is 1.04 × 10-1 mSv/hr. Dose rates for various RV and RVI cut-off pieces distributed lower than the limit except the one of 2 m from the cask surface of RVI. It needs to adjust contents in cask which carries highly radioactive components in order to decrease thickness of cask. Conclusion: Two types of casks are considered in this paper: box type for very-low-level waste (VLLW) as well as low-level waste (LLW) and cylinder type for intermediate-level waste (ILW). The results will contribute to the development of optimal loading plans for RV and RVI cut-off pieces during the decommissioning of nuclear power plant that can be used to prepare radioactive waste disposal plans for the different types of wastes-ILW, LLW, and VLLW.

• Journal of Radiation Protection and Research
• /
• 제41권2호
• /
• pp.173-178
• /
• 2016

Background: In situ gamma spectrometry has been used to measure environmental radiation, assumptions are usually made about the depth distribution of the radionuclides of interest in the soil. The main limitation of in situ gamma spectrometry lies in determining the depth distribution of radionuclides. The objective of this study is to develop a method for subsurface characterization by in situ measurement. Materials and Methods: The peak to valley method based on the ratio of counting rate between the photoelectric peak and Compton region was applied to identify the depth distribution. The peak to valley method could be applied to establish the relation between the spectrally derived coefficients (Q) with relaxation mass per unit area (${\beta}$) for various depth distribution in soil. The in situ measurement results were verified by MCNP simulation and calculated correlation equation. In order to compare the depth distributions and contamination levels in decommissioning KRR site, in situ measurement and sampling results were compared. Results and Discussion: The in situ measurement results and MCNP simulation results show a good correlation for laboratory measurement. The simulation relationship between Q and source burial for the source layers have exponential relationship for a variety depth distributions. We applied the peak to valley method to contaminated decommissioning KRR site to determine a depth distribution and initial activity without sampling. The observed results has a good correlation, relative error between in situ measurement with sampling result is around 7% for depth distribution and 4% for initial activity. Conclusion: In this study, the vertical activity distribution and initial activity of $^{137}Cs$ could be identifying directly through in situ measurement. Therefore, the peak to valley method demonstrated good potential for assessment of the residual radioactivity for site remediation in decommissioning and contaminated site.

• Journal of Radiation Protection and Research
• /
• 제45권3호
• /
• pp.130-141
• /
• 2020

Background: Concrete activation in cyclotron vaults is a major concern associated with their decommissioning because a considerable amount of activated concrete is generated by secondary neutrons during the operation of cyclotrons. Reducing the amount of activated concrete is important because of the high cost associated with radioactive waste management. This study aims to investigate the capability of the neutron absorbing materials to reduce concrete activation. Materials and Methods: The Particle and Heavy Ion Transport code System (PHITS) code was used to simulate a cyclotron target and room. The dimensions of the room were 457 cm (length), 470 cm (width), and 320 cm (height). Gd₂O₃, B₄C, polyethylene (PE), and borated (5 wt% ^natB) PE with thicknesses of 5, 10, and 15 cm and their different combinations were selected as neutron absorbing materials. They were placed on the concrete walls to determine their effects on thermal neutrons. Thin B₄C and Gd₂O₃ were placed between the concrete wall and additional PE shield separately to decrease the required thickness of the additional shield, and the thermal neutron flux at certain depths inside the concrete was calculated for each condition. Subsequently, the optimum combination was determined with respect to radioactive waste reduction, price, and availability, and the total reduced radioactive concrete waste was estimated. Results and Discussion: In the specific conditions considered in this study, the front wall with respect to the proton beam contained radioactive waste with a depth of up to 64 cm without any additional shield. A single layer of additional shield was inefficient because a thick shield was required. Two-layer combinations comprising 0.1- or 0.4-cm-thick B₄C or Gd₂O₃ behind 10 cm-thick PE were studied to verify whether the appropriate thickness of the additional shield could be maintained. The number of transmitted thermal neutrons reduced to 30% in case of 0.1 cm-thick Gd₂O₃+10 cm-thick PE or 0.1 cm-thick B₄C+10 cm-thick PE. Thus, the thickness of the radioactive waste in the front wall was reduced from 64 to 48 cm. Conclusion: Based on price and availability, the combination of the 10 cm-thick PE+0.1 cmthick B₄C was reasonable and could effectively reduce the number of thermal neutrons. The amount of radioactive concrete waste was reduced by factor of two when considering whole concrete walls of the PET cyclotron vault.