• The Journal of the Korea Contents Association
• /
• v.17 no.9
• /
• pp.69-76
• /
• 2017

This study evaluated the activation of the shielding wall and the components around the accelerator by using the medical linear accelerator. We performed simulations for energy values of 20 MV with the operating time ranging from day 1 to 30 years, and linear accelerator head and shielding wall concrete were also evaluated. The results showed that neutrons in large quantities were analyzed using high energy around thetarget point where photons were formed. Based on the activation analysis with these results, radioactivity increased with an increase in operation time and activated nuclides usually start saturating in10 years. Furthermore, the general types of nuclides formed owingto the activation were Co-60, W-181, 185, 187, Na-24, Ca-45, Mn-54, 56, and Fe-55, 59.

• Proceedings of the Korean Vacuum Society Conference
• /
• 2016.02a
• /
• pp.194.1-194.1
• /
• 2016

방사성 폐기물의 운반이나 장기 보관 시 방사성 물질의 침출을 차단하기 위한 유리화 기술을 실현하기 위해 이송식 아크 플라즈마에 대해 전산해석을 수행하였다. 본 연구에서는 운전전류나 아크길이와 같은 운전조건 변화에 따른 열플라즈마의 특성 변화 뿐만 아니라 150 kW급 고출력 이송식 아크 플라즈마의 최적 설계를 위하여 핵심 부품인 파일럿 노즐의 길이와 직경 변화에 따른 예상 용융영역을 전산해석 하여 방사성 폐기물의 유리화 기술을 상업적으로 이끌어내는데 기초 자료를 제공하고자 하였다. 노즐직경은 4, 5, 6 mm로 변화시켰으며, 길이는 2, 4, 6mm로 하였다. 이러한 다양한 설계조건에 대하여 운전변수로는 전류 200 A, 방전 기체인 알곤의 유량 15 L/min, 아크 길이 2 cm로 고정하였다. 전산해석 결과 노즐직경이 작을수록 아크압축 효과에 의해 중심부에서 최고 온도가 높은 열플라즈마 제트를 발생시킬 수 있으나, 반경방향으로 온도구배가 커서 고온 구간이 급격히 감소하는 경향이 예상되었다. 반면 노즐직경이 증가할수록 아크 압축효과는 줄어들지만 반경방향으로 온도가 완만히 감소하여 콘크리트가 대부분인 유리화 대상물질을 충분히 용융시킬 수 있는 $2,600^{\circ}C$ 이상의 고온 면적이 넓어지게 될 것으로 예상되었다. 또한, 노즐길이가 줄어들 경우 아크방전의 안정성은 다소 떨어 질 수 있으나 수 있으나 고온의 열플라즈마 제트가 반경방향으로 효과적으로 넓어 질 수 있음이 예측되었다. 따라서 고온 영역의 확장 관점에서 이송식 아크 플라즈마 토치를 제작할 경우 아크의 안정성을 유지하는 범위 내에서 파일럿 노즐의 직경을 크게 하고 길이는 짧게 하는 것이 효과적인 유리화를 위해 유리할 것으로 예상되었다.

• Journal of the Korean Society of Radiology
• /
• v.14 no.5
• /
• pp.563-575
• /
• 2020

The purpose of this study is to evaluate the activation characteristics that occur in a linear accelerator for container security inspection. In the computer simulation design, first, the targets consisted of a tungsten (Z=74) single material target and a tungsten (Z=74) and copper (Z=29) composite target. Second, the fan beam collimator was composed of a single material of lead (Z=82) and a composite material of tungsten (Z-74) and lead (Z=82) depending on the material. Final, the concrete in the room where the linear accelerator was located contained magnetite type and impurities. In the research method, first, the optical neutron flux was calculated using the MCNP6 code as a F4 Tally for the linear accelerator and structure. Second, the photoneutron flux calculated from the MCNP6 code was applied to FISPACT-II to evaluate the activation product. Final, the decommissioning evaluation was conducted through the specific activity of the activation product. As a result, first, it was the most common in photoneutron targets, followed by a collimator and a concrete 10 cm deep. Second, activation products were produced as by-products of W-181 in tungsten targets and collimator, and Co-60, Ni-63, Cs-134, Eu-152, Eu-154 nuclides in impurity-containing concrete. Final, it was found that the tungsten target satisfies the permissible concentration for self-disposal after 90 days upon decommissioning. These results could be confirmed that the photoneutron yield and degree of activation at 9 MeV energy were insignificant. However, it is thought that W-181 generated from the tungsten target and collimator of the linear accelerator may affect the exposure when disassembled for repair. Therefore, this study presents basic data on the management of activated parts of a linear accelerator for container security inspection. In addition, When decommissioning the linear accelerator for container security inspection, it is expected that it can be used to prove the standard that permissible concentration of self-disposal.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.17 no.1
• /
• pp.121-126
• /
• 2019

The radionuclide inventory prediction of a nuclear power plant can help establish decommissioning plan by providing information of radiation environment. Accumulated radionuclides in reactors and related facilities after reactor shutdown can be divided into neutron activated materials and contaminated materials. Among the neutron activated radionuclides, $^{36}Cl$ and $^{41}Ca$ are important from the viewpoint of disposal because of its long half-life and physiochemical characteristics. In this research, we calculated the radionuclides of $^{36}Cl$ and $^{41}Ca$ in bioshielding concrete by estimating the neutron flux and cross section using the MCNPX. And we evaluated the inventories of $^{36}Cl$ and $^{41}Ca$ using the activation calculation code ORIGEN2.

• The Journal of the Korea Contents Association
• /
• v.16 no.10
• /
• pp.758-765
• /
• 2016

This study analyzed for the radioactive shielding wall, which shields the medical linear accelerator. This allows to evaluate the level of waste with respect to the shield wall, which accounts for more than half of the cost of dismantling later linac facility. In addition, by analyzing the waste processing method according we discuss the way to obtain the benefits in terms of dismantling cost. Results of the simulate, the amount sufficient to screen the amount of neutron radiation occurring in the shielding wall linac was measured. And neutron activation analysis results were analyzed nuclides more than about 20. This analysis was in excess of that, $^{24}Na$, $^{45}Ca$, $^{59}Fe$ nucleus paper deregulation concentration. The value is reduced is greater the deeper the depth of the shielding wall concentration. Based on this, three specific areas (E, F, G) was estimated to be impossible to landfill or recycling. The rest area was estimated to be buried or recycled if possible more than a predetermined depth.

• Journal of Energy Engineering
• /
• v.29 no.1
• /
• pp.63-74
• /
• 2020

The Kori-Unit 1 nuclear power plant, which is scheduled for decommissioning after permanent shutdown, is expected to generate a large amount of various types of radioactive waste during decommissioning process. For concrete radioactive waste, which is expected to occupy the most amount, it is important to analyze the current waste disposal status and legal limitations and to prepare an appropriate and efficient disposal method. Concrete radioactive waste is waste of various levels, of which the clearance level is bioshield concrete. In this paper, clearance radioactive waste safety evaluation was performed using the RESRAD code, which is a safety evaluation code, based on the activation evaluation results for the wastes with the clearance level. The clearance scenario of the target radioactive waste was selected and the individual's exposure dose was calculated at the time of clearance to determine whether the clearance criteria limit prescribed by the Nuclear Safety Act was satisfied. As a result of the evaluation, the results showed significantly lower results and satisfied the criteria value. Based on the results of this clearance safety assessment, the appropriate disposal method for bioshield concrete, which are the clearance wastes of subject of deregulation, was suggested.

• Journal of the Korean Recycled Construction Resources Institute
• /
• v.9 no.4
• /
• pp.451-459
• /
• 2021

The purpose of this study is to evaluate the performance of a solidifying agent for recycling the fine powder separated from the nuclear power plant decommissioned concrete as a solidifying agent(SA) for radioactive waste. In order to evaluate the performance of the solidifying agent, a powder simulating the fine powder of waste concrete separated from the dismantled concrete of a nuclear power plant was produced, and the main variables were the type of binder and the replacement ratio of zeolite. The solidifying agent was evaluated for fluidity performance, compressive strength, and leaching resistance to non-radioactive cesium. The compressive strength of SA increased as the zeolite replacement ratio increased, and the SA containing 5% or more of zeolite showed a compressive strength that was 1.4 to 1.7 times higher than the acceptance criteria. The cesium leaching index of all specimens was 6 or higher, satisfying the acceptance criteria, and the leaching index of SA was 1.47~1.63 times higher than that of OPC. In particular, the average leaching index after 28 days of the 5% zeolite-substituted solidifying agent was 9.15, which was improved by about 6.4% compared to OPC, and it was confirmed that the zeolite was effective in improving the leaching resistance to cesium ions by showing stable performance over the entire period.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.11 no.4
• /
• pp.271-280
• /
• 2013

Since the decommissioning of nuclear plants and facilities, large quantities of slightly contaminated concrete waste have been generated. In Korea, the decontamination and decommissioning of the KRR-1, 2 at the KAERI have been under way. And concrete waste was generated about 800 drums of 200 L. The conditioning of concrete waste is needed for final disposal. The concrete waste is conditioned as follows: mortar using coarse and fine aggregates is filled void space after concrete rubble pre-placement into 200 L drum. Thus, this research has developed an optimizing mixing ratio of concrete waste, water, and cement and has evaluated characteristics of a cement waste form to meet the requirements specified in disposal site specific waste acceptance criteria. The results obtained from compressive strength test, leaching test, thermal cycling test of cement waste forms conclude that the concrete waste, water, and cement have been suggested to have 75:15:10wt% as the optimized mixing ratio. Also, the compressive strength of cement waste form was satisfied that including fine powder up to maximum 40wt% in concrete debris wastes about 75%. As a result of scale-up test, the mixture of concrete waste, water, and cement is 75:10:15wt% meet the satisfied compressive strength because the free water increased with and increased in particle size.

• Proceedings of the Computational Structural Engineering Institute Conference
• /
• 2011.04a
• /
• pp.404-405
• /
• 2011

원전구조물의 방사성 폐기물 처분시설의 경우 지하수에 해수가 유입되어 콘크리트에 염소이온 침투가 발생할 수 있으며, 콘크리트 내부에 존재하는 인장철근의 부식에 의한 내구성 저하 및 수명 단축이 주된 문제가 된다. 본 논문에서는 원전콘크리트 구조물에서의 확산에 의한 염소이온 침투에 대한 수학적 모델을 제시하였다. 콘크리트 중의 염소이온의 침투는 콘크리트의 노출환경, 습윤상태에 따라 확산(Diffusion), 대류(Absorption), 전기적 이동(Migration)에 의해 발생한다. 이러한 조건을 모두 고려하여 제시한 방정식에 의해 염소이온의 침투를 예측할 수 있다.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.15 no.4
• /
• pp.355-367
• /
• 2017

The decommissioning of Kori unit 1, which was permanently shut down in June of 2017, will be the first instance of the dismantling of a commercial nuclear power plant in Korea. The disposal of waste during the dismantling process accounts for a large part of the total decommissioning cost. Therefore, structures consisting of activated and contaminated concrete must be economically and safely dismantled by establishing a proper dismantling strategy. This study focuses on optimized dismantling and disposal scenarios pertaining to a biological shield. Several dismantling cases, regulations and technologies related to waste treatment as these practices pertain to nuclear power plants are analyzed. To minimize the amount of waste from the biological shield dismantling process, an optimized dismantling scenario is presented and disposal alternatives for dismantled concrete waste are proposed.