• Journal of Radiation Protection and Research
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• 제34권1호
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• pp.1-7
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• 2009

It is expected that a substantial amount of spent fuels will be transported from the four nuclear power plant (NPP) sites in Korea to a hypothetical centralized interim storage facility or a final repository in the near future. The cost for the transportation is proportional to the amount of spent fuels. In this paper, a cost estimation program is developed based on the conceptual design of a transportation system and a logistics analysis. Using the developed computer program, named as CASK, the minimum capacity of a centralized interim storage facility (CISF) and the transportation cost for PWR spent fuels are calculated. The PWR spent fuels are transported from 4 NPP sites to a final repository (FR) via the CISF. Since NPP sites and the CISF are located along the coast, a sea-transportation is considered and a road-transportation is considered between the CISF and the FR. The result shows that the minimum capacity of the interim storage facility is 15,000 MTU.

• Nuclear Engineering and Technology
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• 제52권6호
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• pp.1148-1155
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• 2020

This work presents two different methods for quantifying and propagating the uncertainty associated with fuel composition at end of life for cask criticality calculations. The first approach, the computational approach uses parametric uncertainty including those associated with nuclear data, fuel geometry, material composition, and plant operation to perform forward depletion on Monte-Carlo sampled inputs. These uncertainties are based on experimental and prior experience in criticality safety. The second approach, the data-driven approach relies on using radiochemcial assay data to derive code bias information. The code bias data is used to perturb the isotopic inventory in the data-driven approach. For both approaches, the uncertainty in k_eff for the cask is propagated by performing forward criticality calculations on sampled inputs using the distributions obtained from each approach. It is found that the data driven approach yielded a higher uncertainty than the computational approach by about 500 pcm. An exploration is also done to see if considering correlation between isotopes at end of life affects k_eff uncertainty, and the results demonstrate an effect of about 100 pcm.

• Nuclear Engineering and Technology
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• 제18권1호
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• pp.27-37
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• 1986

본 논문에서는 고리 1호기의 원자로 압력용기와 1차 콘크리트 차폐체 사이의 인자로 공동에서의 발사선 흐름 현상을 평가하였다. 원자로 압력용기 외부 표면에서 방출되는 누출 선속을 계산하기 위해 사용될 적합한 중성자 단면적 자료를 얻기 위하여, DLC-23/CASK, DLC-31/FEWG그리고 DLC-47/BUGLE 등 세 가지의 중성자 단면적 자료에 대한 검증 계산을 수행하였다. 누출 선속 계산은 ANISN으로 1차원적 계산을, DOT3.5로 2차원적 계산을 수행하였으며, 또한 원자로 공동에서의 방사선 흐름 현상을 분석하기 위하여, 알베도 개념이 도입된 몬테카를로 방법을 사용하는 MORSE-CG 전산 코드를 이용하여 3차원적 해석을 하였다. 그리고, 원자로 플랜지 부위에서의 방사화 분석을 수행하여 스터드 볼트의 방사화 정도를 평가하였다.

• Nuclear Engineering and Technology
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• 제50권6호
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• pp.890-898
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• 2018

Nondestructive inspection (NDI) is an integral part of structural integrity analyses of dry storage casks that house spent nuclear fuel. One significant concern for the structural integrity is stress corrosion cracking in the heat-affected zone of welds in the stainless steel canister that confines the spent fuel. In situ NDI methodology for detection of stress corrosion cracking is investigated, where the inspection uses a delivery robot because of the presence of the harsh environment and geometric constrains inside the cask protecting the canister. Shear horizontal (SH) guided waves that are sensitive to cracks oriented either perpendicular or parallel to the wave vector are used to locate welds and to detect cracks. SH waves are excited and received by electromagnetic acoustic transducers (EMATs) using noncontact ultrasonic transduction and pulse-echo mode. A laboratory-scale canister mock-up is fabricated and inspected using the proposed methodology to evaluate the ability of EMATs to excite and receive SH waves and to locate welds. The EMAT's capability to detect notches from various distances is evaluated on a plate containing 25%-through-thickness surface-breaking notches. Based on the results of the distances at which notch reflections are detectable, NDI coverage for spent nuclear fuel storage canisters is determined.

• Nuclear Engineering and Technology
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• 제53권11호
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• pp.3704-3710
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• 2021

The characteristics of the external radiation on the surface of the casks for spent nuclear fuel (SNF) storage by dry method are investigated for the case when the spatial distribution of SNF in the basket changes due to the destruction of the fuel rod claddings. The surface areas are determined, where the changes in fluxes of neutrons, produced by ²⁴⁴Cm actinide, and γ-quanta, produced by long-lived isotopes, are maximum in the result of the decrease in the height of the SNF area. Concrete (VSC-24) and metal (SC-21) casks are considered as examples. The procedure of periodic measurement of the dose rate of neutrons or γ-quanta at the specified points of the cask surface is proposed for identifying the fuel rod cladding destruction. Under normal operation, the decrease in the dose rate produced by neutrons as the function of SNF storage duration is determined by the half-life of ²⁴⁴Cm, and for γ-quanta - by the half-lives of long-lived SNF isotopes. Consequently, a stepwise change in the dose rate of neutrons or γ-quanta, detected by the measurements, as compared to the previous one, would indicate the destruction of the fuel rod claddings.

• 방사성폐기물학회지
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• 제20권3호
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• pp.365-369
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• 2022

Operating and decommissioning nuclear power plants generates radioactive waste. This radioactive waste can be categorized into several different levels, for example, low, intermediate, and high, according to the regulations. Currently, low and intermediate-level waste are stored in conventional 200-liter drums to be disposed. However, in Korea, the disposal of intermediate-level radioactive waste is virtually impossible as there are no available facilities. Furthermore, large-sized intermediate-level radioactive waste, such as reactor internals from decommissioning, need to be segmented into smaller sizes so they can be adequately stored in the conventional drums. This segmentation process requires additional costs and also produces secondary waste. Therefore, this paper suggests repurposing the no-longer-used spent nuclear fuel casks. The casks are larger in size than the conventional drums, thus requiring less segmentation of waste. Furthermore, the safety requirements of the spent nuclear fuel casks are severer than those of the drums. Hence, repurposed spent nuclear fuel casks could better address potential risks such as dropping, submerging, or a fire. In addition, the spent nuclear fuel casks need to be disposed in compliance with the regulations for low level radioactive waste. This cost may be avoided by repurposing the casks.

• 방사성폐기물학회지
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• 제21권2호
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• pp.193-204
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• 2023

To ensure radiological safety margin in the transport and storage of spent nuclear fuel, it is crucial to perform source term and shielding analyses in advance from the perspective of conservation. When performing source term analysis on UO₂ fuel, which is mostly used in commercial nuclear power plants, uranium and oxygen are basically considered to be the initial materials of the new fuel. However, the presence of impurities in the fuel and structural materials of the fuel assembly may influence the source term and shielding analyses. The impurities could be radioactive materials or the stable materials that are activated by irradiation during reactor power operation. As measuring the impurity concentration levels in the fuel and structural materials can be challenging, publicly available information on impurity concentration levels is used as a reference in this evaluation. To assess the effect of impurities, the results of the source term and shielding analyses were compared depending on whether the assumed impurity concentration is considered. For the shielding analysis, generic cask design data developed by KEPCO-E&C was utilized.

• 공업화학
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• 제7권3호
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• pp.597-604
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• 1996

사용후핵연료 수송용기 등에 사용되는 수지계 중성자 차폐재, KNS-115A, 115B 및 115C를 제조하였다. 기본물질은 에폭시수지이며, 첨가제로는 폴리프로필렌, 수산화알루미늄 및 탄화붕소이다. 이들 중성자 차폐재들은 유동성이 좋아 수송용기와 같은 복잡한 구조에 사용할 수 있다. 개발된 중성자 차폐재들의 차폐능, 연소특성, 난연성, 열적 및 역학적 성질 등을 평가하기 위해 여러 특성시험을 행하였다. 개발된 중성자 차폐재(수소원자 밀도, $6.1{\sim}6.2{\times}10^{22}atoms/cm^3$)들은 외국산 중성자 차폐재(NS-4-FR, $6.0{\times}10^{22}atoms/cm^3$)보다 수소원자 밀도가 높아 차폐능이 우수할 것으로 예측되며, 조사된 제반 특성들은 열분해온도; $267{\sim}270^{\circ}C$, 열전도도; $0.62{\sim}0.72W/m{\cdot}K$, 연소특성; $800^{\circ}C$ 이하, 평균연소시간; 5초 이하, 평균연소길이; 5mm 이하, 인장강도; $2.3{\sim}3.0kg/mm^2$, 압축강도; $5.3{\sim}13.3kg/mm^2$, 굴곡강도; $4.4{\sim}5.4kg/mm^2$ 등을 나타냈다.

• Nuclear Engineering and Technology
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• 제42권5호
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• pp.582-589
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• 2010

As a part of the back-end fuel cycle, transportation of spent nuclear fuel (SNF) from nuclear power plants (NPPs) to a fuel storage facility is very important in establishing a nuclear fuel cycle. In Korea, the accumulated amount of SNF in the NPP pools is troublesome since the temporary storage facilities at these NPP pools are expected to be full of SNF within ten years. Therefore, Korea cannot help but plan for the construction of an interim storage facility to solve this problem in the near future. Especially, a decision on several factors, such as where the interim storage facility should be located, how many casks a transport ship can carry at a time and how many casks are initially required, affect the configuration of the transportation system. In order to analyze the various possible candidate scenarios, we assumed four cases for the interim storage facility location, three cases for the load capacity that a transport ship can carry and two cases for the total amount of casks used for transportation. First, this study considered the currently accumulated amount of SNF in Korea, and the amount of SNF generated from NPPs until all NPPs are shut down. Then, how much SNF per year must be transported from the NPPs to an interim storage facility was calculated during an assumed transportation period. Second, 24 candidate transportation scenarios were constructed by a combination of the decision factors. To construct viable yearly transportation schedules for the selected 24 scenarios, we created a spreadsheet program named TranScenario, which was developed by using MS EXCEL. TranScenario can help schedulers input shipping routes and allocate transportation casks. Also, TranScenario provides information on the cask distribution in the NPPs and in the interim storage facility automatically, by displaying it in real time according to the shipping routes, cask types and cask numbers that the user generates. Once a yearly transportation schedule is established, TranScenario provides some statistical information, such as the voyage time, the availability of the interim storage facility, the number of transported casks sent from the NPPs, and the number of transported casks received at the interim storage facility. By using this information, users can verify and validate a yearly transportation schedule. In this way, the 24 candidate scenarios could be constructed easily. Finally, these 24 scenarios were compared in terms of their operation cost.

• 방사성폐기물학회지
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• 제16권2호
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• pp.223-232
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• 2018

사용후핵연료 저장용기의 공기 흡입구 및 배기구에는 외부환경으로부터 이물질의 유입을 방지하기 위하여 bird screen이 설치되며, bird screen에서는 공기의 유동 저항이 발생하게 된다. 본 연구에서는 Bird screen mesh의 단순화 모델을 이용한 열해석을 수행하기 위하여 다공성매질 해석모델을 개발하였다. CFD 해석을 이용하여 다공성매질에 대한 유동저항계수를 산출하고 이에 대한 신뢰성을 입증하였다. 다공성매질 해석모델을 이용하여 콘크리트 저장용기의 열해석을 수행하고 bird screen을 갖는 콘크리트 저장용기의 열시험을 수행하였다. Bird screen mesh를 고려한 열시험 결과와 다공성매질을 고려한 열해석 결과를 비교하였으며, 해석 및 시험결과가 서로 잘 일치하였다. 해석결과는 시험결과에 비하여 다소 높은 온도분포를 보여 다공성매질을 사용한 콘크리트 저장용기의 열해석 결과에 대한 신뢰성 및 보수성이 입증되었다.