• Nuclear Engineering and Technology
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• 제45권3호
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• pp.377-384
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• 2013

During the last three decades, South Korean nuclear power plants have discharged about 5,950 tons of spent fuel and the maximum burn-up reached 55 GWd/MTU in 2002. This study was performed to support the development of Korean dry spent fuel storage alternatives. First, we chose V5H-$17{\times}17$ and KSFA-$16{\times}16$ as representative domestic spent fuels, considering current accumulation and the future generation of the spent fuels. Examination reveals that their average burn-ups have already increased from 33 to 51 GWd/MTU and from 34.8 to 48.5 GWd/MTU, respectively. Evaluation of the fuel characteristics shows that at the average burn-up of 42 GWd/MTU, the oxide thickness, hydrogen content, and hoop stress ranged from $30{\sim}60{\mu}m$, 250 ~ 500 ppm, and 50 ~ 75 MPa, respectively. But when burn-up exceeds 55 GWd/MTU, those characteristics can increase up to 100 ${\mu}m$, 800 ppm, and 120 MPa, respectively, depending on the power history. These results demonstrate that most Korean spent nuclear fuels are expected to remain within safe bounds during long-term dry storage, however, the excessive hoop stress and hydrogen concentration may trigger the degradation of the spent fuel integrity early during the long-term dry storage in the case of high burn-up spent fuels exceeding 45 GWd/MTU.

• 방사성폐기물학회지
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• 제19권1호
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• pp.61-74
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• 2021

A paradigm shift in the government's energy policy was reflected in its declaration of early closure of old nuclear plants as well as cancellation of plans for the construction of new plants. To this end, unit 1 of Kori Nuclear Power Plant was permanently shut down and is set for decommission. Based on these changes, the off-site transport of spent fuels from nuclear power plants has become a critical issue. The purpose of this study is to develop an optimized method for transportation of spent fuels from Kori Nuclear Power Plant's units 1, 2, 3, and 4 to an assumed interim storage facility by simulating the scenarios using the Flexsim software, which is widely used in logistics and manufacturing applications. The results of the simulation suggest that the optimized transport methods may contribute to the development of delivery schedule of spent fuels in the near future. Furthermore, these methods can be applied to decommissioning plan of nuclear power plants.

• Nuclear Engineering and Technology
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• 제31권6호
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• pp.23-38
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• 1999

This study addresses the reference spent fuel and its characteristics for developing a geological repository concept. As a disposal capacity of the reference repository system to be developed, spent fuel inventories were projected based on the basis of the Nuclear Energy Plan of the Long-term National Power Program. The reference spent fuel encompassing a variability in characteristics of all existing and future spent fuels of interest was defined. Key parameters in the reference fuel screening processes were the nuclear and mechanical design parameters and the burnup histories for existing spent fuels as of 1996 and for future spent fuels with the more extended burnup the initial enrichment and its expected turnup. The selected reference fuel was characterized in terms of initial enrichment, bumup, dimension, gross weight and age. Also the isotopic composition and the radiological properties are quantitatively identified. This information provided in this study could be used as input for repository system development and performance assessment and applied in fuel material balance evaluation for the various types of back-end fuel cycle studies.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2004년도 Proceedings of the 4th Korea-China Joint Workshop on Nuclear Waste Management
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• pp.191-204
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• 2004

The research and development of effective management technologies of the spent fuels discharged from power reactors are an important and essential task of KAERI. In resent several years KAERI has focused on a project named "development and demonstration of the Advanced spent fuel Conditioning Process (ACP) in a laboratory scale." The Facility for ACP demonstration consists of two Hot Cells and auxiliary facilities. It is now in the final design stage and will be constructed in 2004. After construction of the facility the ACP equipments will be installed in Hot Cells. The ACP will be demonstrated by some simulated spent fuels first and then by spent fuels.

• 방사성폐기물학회지
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• 제20권2호
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• pp.231-241
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• 2022

Several countries, including Korea, are considering the direct disposal of spent nuclear fuels. The radiological safety assessment results published after a geological repository closure indicate that the instant release is the main radiation source rather than the congruent release. Three Safety Case reports recently published were reviewed and the IRF values of seven long-lived radionuclides, including relevant experimental results, were compared. According to the literature review, the IRF values of both the CANDU and low burnup PWR spent fuel have been experimentally measured and used reasonably. In particular, the IRF values of volatile long-lived nuclides, such as ¹²⁹I and ¹³⁵Cs, were estimated from the FGR value. Because experimental leaching data regarding high burnup spent nuclear fuels are extremely scarce, a mathematical modelling approach proposed by Johnson and McGinnes was successfully applied to the domestic high burnup PWR spent nuclear fuel to derive the IRF values of iodine and cesium. The best estimate of the IRF was 5.5% at a discharge burnup of 55 GWd tHM^-1.

• 방사성폐기물학회지
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• 제6권2호
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• pp.147-154
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• 2008

국내 원자력발전소의 사용후핵연료 저장용량의 포화가 10년 이내에 예상되고 있으며, 소외중간저장방식을 관리 방안으로서 선정할 경우, 상당한 양의 사용후핵연료 운반이 해상 혹은 육로를 통해 매년 이루어져야만 한다. 본 논문에서는 4곳의 원자력발전소부지에 분산 저장되어 있는 사용후핵연료를 해안에 위치한 가상의 중간저장시설과 영구처분시설을 대상으로 사용후핵연료 운반물량을 효과적으로 분석할 수 있는 체계를 구축하였다. 각 발전소 부지, 중간저장시설, 영구처분시설의 저장고를 중심으로 사용후핵연료 물질 수지식을 세우고, 이에 대한 해를 VISUAL BASIC으로 구하여 운반 물량 분석이 용이하게 수행할 수 있는 컴퓨터 프로그램(CASK)을 개발하였다. 개발된 물량 분석 프로그램을 활용하여 4개 원자력 발전소 부지로부터 사용후핵연료 운반 물량을 분석하고, 운반물량 파라미터 분석을 통하여 본 프로그램의 활용도를 보였다. 개발된 사용후핵연료 운반 물량 분석 체계는 향후 운반비용 분석에도 유용하게 활용될 것이다.

• Nuclear Engineering and Technology
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• 제55권8호
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• pp.3102-3113
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• 2023

Following the Fukushima nuclear disaster, the simulation of accidents in the spent fuel pool has become more noticeable. Despite the low amount of decay heat power, the consequences of the accidents in a spent fuel pool (SFP) can be severe due to the high content of long-lived radionuclides and lack of protection by the pressure vessel. In this study, the loss-of-cooling accident (LOFA) for the VVER-1000/V446 spent fuel pool is simulated by employing RELAP5 and MELCOR 1.8.6 as the best estimate and severe accident analysis codes, respectively. For two cases with different total power levels, decay heat of spent fuels is calculated by ORIGEN-II code. For modeling SFP of a VVER-1000, a qualified nodalizations are considered in both codes. During LOFA in SFP, the key sequences such as heating up of the pool water, boiling and reducing the water level, uncovering the spent fuels, increasing the temperature of the spent fuels, starting oxidation process (generating Hydrogen and extra power), the onset of fuel melting, and finally releasing radionuclides are studied for both cases. The obtained results show a reasonable consistency between the RELAP5 and MELCOR codes, especially before starting the oxidation process.

• Nuclear Engineering and Technology
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• 제33권2호
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• pp.231-240
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• 2001

When it is assumed that PWR, CANDU and DUPIC spent fuels are disposed of in deep geological repository, consequent annual individual doses are calculated, and it is shown that doses meet the regulatory limit. From these results, the hazardous radionuclides applicable to partitioning and transmutation are selected. These selected radionuclides such as Tc-99, Ⅰ-129, Cs-135 and Np-237 are then reviewed in terms of partitioning and transmutation. Separation of I-129, Np-237 and Tc-99 from spent fuels is considered desirable, and transmutation of these radionuclides results in remarkable hazard reduction. However, it is concluded that separation and transmutation of Cs-135 may be ineffective although it is classified into a hazardous radionuclide.

• Journal of Radiation Protection and Research
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• 제36권4호
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• pp.174-182
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• 2011

With the reactor operation conditions - 4.3 wt% $^{235}U$ initial enrichment, burn-up 55,000 MWd/MTU, average power 34 MW/MTU for three periods burned time for 539.2 days per period and cooling time for 100 hours after shut down, to set up the condition to determine the minimum height (depth) of spent fuel storage pool to shut off the radiation out of the spent fuel storage pool and to store spent fuels safely, the dose rate on the specific position directed to the surface of spent fuel storage pool induced by the neutron and gamma-ray from spent fuels are evaluated. The length of spent fuel is 381 cm, and as the result of evaluation on each position from the top of spent fuel to the surface of spent fuel storage pool, it is difficult for neutrons from spent fuels to pass through the water layer of maximum 219 cm (600 cm from the floor of spent fuel storage pool) and 419 cm (800 cm from the floor of spent fuel storage pool) for gamma-ray. Therefore, neutron and gamma-ray from spent fuels can pass through below 419 cm (800 cm from the floor) water layer directed to the surface of spent fuel storage pool.

• Nuclear Engineering and Technology
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• 제23권2호
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• pp.165-173
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• 1991

가압경수로의 사용후핵연료를 CANDU 원자로에 재순환시키는, 이른바, 탄뎀 핵연료주기가 본 연구에서 다루어졌다. 이러한 방식으로 가압경수로의 사용후핵연료를 활용하는 것은 우라늄자원의 이용을 개선시킬뿐만 아니라 사용후핵연료 저장능력의 부족도 다소 해결할 수 있을 것이다. 핵연료를 재순환 시키는데 있어서는 CANDU 원자로의 수정을 최소화하는 방향으로 연구가 진행되었으며 본 연구에서는 9종의 핵연료가 고려되었다. 탄뎀 핵연료는 크게 핵연료재가공과 노심재구성의 두 분야로 나뉘어지는데, 핵연료 재가공의 경우, 가압경수로의 사용후핵연료는 처리되고 현재의 37 봉형 격자구조인 핵연료 다발에 맞도록 다시 성형가공되며 노심재구성의 경우, 가압경수로 사용후핵연료는 단지 격자 구조를 해체하고 CANDU의 격자길이에 맞춰 재구성만 된다. 각 탄뎀 핵연료 옵션에 대하여, 허용연소도와 출력분포를 계산하기 위해 노심연소계산이 수행되었다. 또한 경제성에 대한 접근으로 각 핵연료 옵션에 대한 핵주기비가 계산되었다. 그 결과 본 연구에서 다루어진 대부분의 탄뎀 핵연료 옵션이 경제성이 있었을 뿐만 아니라 기술적인 타당성이 있었다.