• Journal of the Korean Society of Radiology
• /
• v.13 no.1
• /
• pp.45-53
• /
• 2019

This study is a neutron activation for concrete that shields medical linear accelerator facilities. Comparison of general concrete and low activation concrete. The simulation method was simulated using MCNPX (Ver. 2.5.0) and FISPACT-2010, and the shielding ability for photon and neutron beams was calculated and neutron activation evaluation was carried out. As a result, the shielding capacity was 20 ~ 50 cm efficient in general concrete, and activate evaluation in low activation concrete was calculated to be low in radioactivity concrete, but all were estimated to not exceed their own allowable concentration in self - disposal. As a result of the comprehensive analysis, it is considered effective to use ordinary concrete.

• Journal of the Korean Society of Radiology
• /
• v.13 no.1
• /
• pp.73-80
• /
• 2019

Neutrons are generated by the nuclear reaction, which is absorbed into the concrete wall and causes the activation during cyclotron operation. The purpose of this study is to investigate the effect of neutron activation and radiative concrete on concrete type. This experiment used Monte Carlo simulation and RESRAD model. The results of the experiment showed that the higher the content of Fe in concrete, the greater the shielding rate. The effect of $^{56}Fe(n,\;2np)^{54}Mn$ reaction on workers is also increased. However, radioactive nuclides have low activity and have very low impact on workers. Radioactive concrete should be treated as general wastes with less than its self-disposal tolerance level, and it should be recycled to the surface such as road repair rather than landfill to minimize the effect of $^{14}C$.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.14 no.2
• /
• pp.91-100
• /
• 2016

The neutron activation inventories in reactor vessel and its internals, and bio-shield of a PWR nuclear power plant were calculated to evaluate the effect of impurity elements contained in the structural materials on the activation inventory. Carbon steel is, in this work, used as the reactor vessel material, stainless steel as the reactor vessel internals, and ordinary concrete as the bio-shield. For stainless steel and carbon steel, one kind of impurity concentration was employed, and for ordinary concrete five kinds were employed in this study using MCNP5 and FISPACT for the calculation of neutron flux and activation inventory, respectively. As the results, specific activities for the cases with impurity elements were calculated to be more than twice than those for the cases without impurity elements in stainless and carbon steel. Especially, the specific activity for the concrete material with impurity elements was calculated to be 30 times higher than that without impurity. Neutron induced reactions and activation inventories in each material were also investigated, and it is noted that major radioactive nuclide in steel material is Co-60 from cobalt impurity element, and, in concrete material, Co-60 and Eu-152 from cobalt and europium impurity elements, respectively. The results of this study can be used for nuclear decommissioning plan during activation inventory assessment and regulation, and it is expected to be used as a reference in the design phase of nuclear power plant, considering the decommissioning of nuclear power plants or nuclear facilities.

• Journal of Radiation Protection and Research
• /
• v.37 no.4
• /
• pp.202-207
• /
• 2012

The radioactivity inventory significantly affects all steps of decommissioning projects including planning, cost estimation, risk assessment, waste management and site remediation. The decommissioning project of the KRR-2 was completed in 2009 and a large amount of activated concrete waste was generated. The bioshield concrete, containing minute amount of impurity elements, was activated by neutron reaction during the operation of the reactor. A variety radionuclides was generated in the concrete, including $^3H$, $^{14}C$, $^{55}Fe$, $^{60}Co$ $^{63}Ni$, $^{134}Cs$, $^{152}Eu$ and $^{154}Eu$. In this paper, the comparison between the calculated results and previous measured results was carried out to estimate the inventory of the bioshield concrete of the KRR-2. The combined computer codes of MCNP5 and ORIGEN 2.1 for calculation of the distribution of neutron flux, cross-section and generation of radionuclides were used. The results were shown that 99.8% of the total radioactivity of $^3H$, $^{55}Fe$, $^{60}Co$ and $^{152}Eu$ in the bioshield concrete 12 years after shutdown. The effects on the variation of inventory were analysed depending on the operation periods and the cooling times in the bioshield concrete.

• Proceedings of the Korea Information Processing Society Conference
• /
• 2006.11a
• /
• pp.299-302
• /
• 2006

연구로 및 원자력 시설의 해체 공정 절차 수립과 해체 시나리오 선정에 기초 자료를 제공하고자 컴퓨터 그래픽스를 응용한 방사화 분포 가시화 연구가 수행되었다. 해체 전 시료 채취와 핵종 분석을 통해 방사능 자료가 확보된 연구로 2 호기 조사실(Exposure Room)과 조사실 주변의 콘크리트 차폐체(Concrete Shielding)를 대상으로 방사화 분포 가시화 실험이 이루어졌다. $^{60}Co$에 오염된 조사실의 벽면과 콘크리트 차폐체의 깊이 별로 조사된 방사능 농도 값을 기초로하여 이들 구조물을 3 차원으로 모델링 한 후 Contour mapping을 수행하여 방사화 분포 가시화를 완료하였다. 방사능 정도를 가시화 한 결과와 콘크리트 차폐체 깊이에 따라 지수 함수적으로 감소하고 있었던 결과가 잘 일치하고 있음을 확인할 수 있었다. 연구 결과 자료는 향후 해체 활동 중 방사선에 노출되는 작업자의 피폭선량 평가 모델에 중요한 역할을 수행할 것이다.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.7 no.2
• /
• pp.79-86
• /
• 2009

Concrete materials in nuclear facilities may become contaminated or activated by various radionuclides through different mechanism. Decommissioning and dismantling of these facilities produce considerable quantities such as concrete structure, rubble. In this paper, the characteristics distribution of the radionuclide have been investigated for the effects of the heating and grinding test for aggregate size such as gravel, sand and paste from decommissioning of the TRIGA MARK II research reactor and uranium conversion plant. The experimental results showed that most of the radionuclide could be removed from the gravel, sand aggregate and concentrated into a paste. Especially, we found that the heating temperature played an important role in separating the radionuclide from the concrete waste. Contamination of concrete is mainly concentrated in the porous paste and not in the dense aggregate such as the gravel and sand. The volume reduction rate could be achieved about 80% of activated concrete waste and about 75% of dismantled concrete waste generated from UCP.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.16 no.1
• /
• pp.93-105
• /
• 2018

In Korea, a huge amount of radioactive concrete waste will be generated through decommissioning of nuclear facilities in the near future; therefore, optimum technology for the treatment of concrete waste should be reviewed thoroughly and the future direction of technology development should be discussed. In this paper, many domestic and foreign examples of generation of radioactive concrete waste were pieced together and the characteristics of radioactive concrete waste were examined. Moreover, we reviewed trends in technology development by analyzing the examples of various studies and practical applications of treatment technologies, such as mechanical decontamination, chemical decontamination, volume reduction, recycling and solidification, and also tried to understand the limitations of existing technologies and determine a direction for technical improvement.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2004.06a
• /
• pp.43-51
• /
• 2004

In order to decommission the shielding concrete of KRR(Korea Research Reactor) -1＆2, it must be exactly determined activated level and range by neutron irradiation during operation. To determine the activated level and range, it must be sampled and analyzed the core sample. But, there are difficulties in sample preparation and determination of the measurement efficiency by self-absorption. In the study, the full energy efficiency of the HPGe detector was compared with the measured value using standard source and the calculated one using Monte Carlo simulation. Also. self-absorption effects due to the density and component change of the concrete were calculated using the Monte Carlo method. Its results will be used radioactivity analysis of the real concrete core sample in the future.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2016.10a
• /
• pp.235-236
• /
• 2016

• Journal of the Korean Recycled Construction Resources Institute
• /
• v.5 no.3
• /
• pp.223-229
• /
• 2017

When concrete is exposed to neutron rays for a long time, the concrete tends to become activated. If activated, it is classified as middle or low level radioactive waste. However, the great amount of the activated concrete is hard to dispose. In this study, low-activation cement was developed for decreasing the activated waste from shielding concrete around nuclear reactor. Furthermore, the manufactured low-activation was analyzed with activation nuclide Eu, Co. The low-activation cement showed great advantage for low-activation with detecting none of Eu and 3.75ppm of Co while ordinary portland cement showed 0.4~0.9ppm of Eu, 5.5~19.8ppm of Co content. As the results of physical properties of the low-activation cement, it is similar to type 1 ordinary portland cement and accords with type 4 low heat portland cement. Meanwhile, as for the chemical properties of the cement, it accords wite type 1 and 4 at the same time.