• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1995년도 추계학술발표회논문집(2)
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• pp.674-679
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• 1995

The hold-down spring forces of Kori nuclear fuels were measured for seven fuel assemblies having 1 to 4 cycles of irradiation histories in the Kori Unit-1 and -2 reactor. The fuel assemblies examined had burnup from 17 to 38 GWD/MTU and the examination was conducted in KAERI PIEF spent fuel storage pool with the newly developed underwater hold-down suing force measuring device. The measurement was made within the elastic deformation ranges and the trends of hold-down spring force relaxation behavour were examined.

• Nuclear Engineering and Technology
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• 제41권2호
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• pp.187-198
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• 2009

In this paper, three power ramp tests performed on high burn-up Re-crystallized Zircaloy2 - UO2 BWR fuel rods (56 to 63 MWd/kgU) within the SCIP project are simulated with METEOR and ALCYONE 3D. Two of the ramp tests are of staircase type up to Linear Heat Rates of 420 and 520 W/cm and with long holding periods. Failure of the 420 W/cm fuel rod was observed after 40 minutes. The third ramp test consisted of a more standard ramp test with a constant power rate of 80 W/cm/min up to 410 W/cm with a short holding time. The tests were first simulated with the METEOR 1D fuel rod code, which gave accurate results in terms of profilometry and fission gas releases. The behaviour of a fuel pellet fragment and of the cladding piece on top of it was then investigated with ALCYONE 3D. The size and the main characteristics of the ridges after base irradiation and power ramp testing were recovered. Finally, the failure criteria validated for PWR conditions and fuel rods with low-to-medium burn-ups were used to analyze the failure probability of the KKL rodlets during ramp testing.

• 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.

• 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.

• Nuclear Engineering and Technology
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• 제46권6호
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• pp.863-868
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• 2014

The evaluation of thermal properties of actinide oxide fuels is a problem of high importance for the development of new generation reactors. In the present study, an expression obtained for n-dimensional Debye functions is used to derive a simple analytical expression for the specific heat capacity of nuclear fuels. To test the validity and reliability of this expression, the analytical expression is applied to $UO_2$, $NpO_2$, $ThO_2$, and $PuO_2$. It is seen that the formula was in agreement with the experimental and theoretical results reported in the literature.

• Nuclear Engineering and Technology
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• 제11권1호
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• pp.29-45
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• 1979

이 논문에서는 6개의 다른 연료봉 모형을 기준으로 하여 TRIGA Mark-III 연구용 원자로의 핵특성이 자세히 분석되고 있다. 현재 핵연료로 사용하고 있는 FLIP fuel은 연료의 수명을 증진시키기 위해 개발되었던 모형이나, 70%의 고농축이기 때문에 비확산금지조약에 위배되어 수입이 곤란한 실정이다. 따라서 본 논문에서는 FLIP 연료 모형에 대치할 수 있는 다른 연료봉온 찾고 그들의 핵특성을 비교함으로써 앞으로 연료수입계륵을 설립하고자 한다.

• 한국소성가공학회:학술대회논문집
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• 한국소성가공학회 2008년도 춘계학술대회 논문집
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• pp.488-490
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• 2008

This study aims to design the high-capacity vol-oxidizer using simulated fuels instead of spent nuclear fuels. Simulated fuels are fabricated by blending tungsten powder with silicon carbide powder, and thereafter, paraffin coating covers simulated fuels to increase their strength. An oxidation experiment using simulated fuels have been carried out in order to analyze oxidation characteristics similar to spent fuels. After oxidation, simulated fuels were almost oxidized to be powders. Increased volume of simulated fuels approached to spent fuels. These results can be utilized as important informations for designing a high-capacity vol-oxidizer.

• 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.

• 한국압력기기공학회 논문집
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• 제19권2호
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• pp.84-92
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• 2023

As temporary storage facilities for spent nuclear fuels in domestic nuclear power plants are expected to be saturated, external intermediate storage facilities would be required in the future. Spent nuclear fuels are stored in metal canisters and then placed in a dry environment within concrete or metal casing for operation. In the United States, the dry storage method for spent nuclear fuels has been operated for an extended period. Based on the corrosion experiences of dry storage canisters in chloride environments, numerous studies have been conducted to reduce corrosion in welds. With the construction of intermediate storage facilities in Korea for spent nuclear fuels expected near coastal areas adjacent to nuclear power plants, there is a need for research on the corrosion occurrence of welds and mitigation methods for canisters in chloride environments. In this paper, we measured and compared the residual stresses in the Heat-Affected Zones (HAZ) after electron beam welding (EBW) and gas tungsten arc welding (GTAW) processes for candidate materials such as 304L, 316L, and duplex stainless steel(DSS). We investigated the possibility of microstructure control through the application of surface modification processes using friction stir processing (FSP). Corrosion tests on each welded specimen revealed a higher corrosion rate in EBW welds compared to GTAW. Furthermore, it was confirmed that corrosion resistance improved due to phase refinement and redistribution of precipitates when FSP was applied.

• 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.