• Nuclear Engineering and Technology
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• v.34 no.3
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• pp.187-201
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• 2002

The KN-12 cask is designed to transport 12 PWR spent nuclear fuels and to comply with the requirements of Korea Atomic Energy Act, IAEA Safety Standards Series No.57-1 and US 10 CFR Part 71 for a Type B(U)F package. It provides containment, radiation shielding, structural integrity, criticality control and heat removal for normal transport and hypothetical accident conditions. W.H 14$\times$14, 16$\times$16 and 17$\times$17 fuel assemblies with maximum allowable initial enrichment of 5.0 wt.%, maximum average burn-up of 50,000 MWD/MTU and minimum cooling time of 7 years being used in Korea will be loaded and subsequently transported under dry and wet conditions. A forged cylindrical cask body which constitutes the containment vessel is closed by a cask lid. Polyethylene rods for neutron shielding are arranged in two rows of longitudinal bore holes in the cask body wall. A fuel basket to accommodate up to 12 PWR fuel assemblies provides support of the fuels, control of criticality and a path to dissipate heat. Impact limiters to absorb the impact energy under the hypothetical accident conditions are attacked at the top and at the bottom side of the cask during transport. Handling weight loaded with water is 74.8 tons and transport weight loaded with water with the impact limiters is 84.3 tons. The cask will be licensed in accordance with Korea Atomic Energy Act 3nd fabricated in Korea in accordance with ASME B&PV Code Section 111, Division 3.

• Nuclear Engineering and Technology
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• v.54 no.12
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• pp.4722-4730
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• 2022

This paper introduces a new point depletion-based source term calculation code named BESNA (Bateman Equation Solver for Nuclear Applications), which is aimed to estimate nuclide inventories and source terms from spent nuclear fuels. The BESNA code employs a new modified CE/CM (Constant Extrapolation - Constant Midpoint) predictor-corrector scheme in depletion calculations for improving computational efficiency. In this modified CE/CM scheme, the decay components leading to the large norm of the depletion matrix are excluded in the corrector, and hence the corrector calculation involves only the reaction components, which can be efficiently solved with the Talyor Expansion Method (TEM). The numerical test shows that the new scheme substantially reduces computing time without loss of accuracy in comparison with the conventional scheme using CRAM (Chebyshev Rational Approximation Method), especially when the substep calculations are applied. The depletion calculation and source term estimation capability of BESNA are verified and validated through several problems, where results from BESNA are compared with those calculated by other codes as well as measured data. The analysis results show the computational efficiency of the new modified scheme and the reliability of BESNA in both isotopic predictions and source term estimations.

• Nuclear Engineering and Technology
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• v.56 no.9
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• pp.3950-3960
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• 2024

As a solution to the problem of spent nuclear fuels (SNFs), the disposal of SNF has gained attention from nations using nuclear energy because of hazards posed to the ecosystem. Among many proposed solutions, the most promising method is to dispose of SNF in a deep geological repository (DGR) which utilizes the multi-barrier concept developed by Finland and Sweden. Here, a new fully-coupled Thermal-Hydraulic (TH) code HADES (High-level rAdionuclide Disposal Evaluation Simulator) is developed using the MOOSE framework. This new code suggests basic numerical tools, such as a non-linear solver and finite element discretization, to assess the safety performance of disposal systems. The new TH code considered various TH behavior using Richards' flow approach, assuming gas pressure is constant. The HADES showed promising results when it was compared to various TH codes validated from DECOVAELX-THMC projects. When the single-canister model was utilized to estimate the TH behavior of the Korean Reference disposal System, although it showed significant saturation reduction due to the evaporation of water, the temperature was maintained under the thermal criteria limit, which is 100 ℃. In addition, the new code estimated temperature and degree of saturation of the multi-canisters model, considering two or three canisters, it showed a slightly lower temperature, 5 ℃, than the single-canister model. From these results, the following are concluded: (1) the new TH code contribute to an additional integrity by estimating TH behavior of KRS; (2) however, due to limitations in single-canister simulation, it is recommended to use multi-canisters simulation to estimate TH behavior accurately. Therefore, this model is anticipated not only to help licensing applications and estimation of various multi-physics phenomena and multi-canister at the disposal site.

• Nuclear Engineering and Technology
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• v.52 no.6
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• pp.1099-1109
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• 2020

The performance of gadolinium burnable absorber (GdBA) for reactivity control in UO₂ and (U, Th)O₂ fuels and its impact on spent fuel characteristics was performed. Five fuel assemblies: one without GdBA fuel rod and four each containing 16, 24, 34 and 44 GdBA fuel rods in both fuels were investigated. Reactivity swing in all the FAs with GdBA rods in UO₂ fuel was higher than their counterparts with similar GdBA fuel rods in (U, Th)O₂ fuel. The excess reactivity in all FAs with (U, Th)O₂ fuel was higher than UO₂ fuel. At the end of single discharge burn-up (~ 49.64 GWd/tHM), the excess reactivity of (U, Th) O₂ fuel remained positive (16,000 pcm) while UO₂ fuel shows a negative value (-6,000 pcm), which suggest a longer discharge burn-up in (U, Th)O₂ fuel. The concentration of plutonium isotopes and minor actinides were significantly higher in UO₂ fuel than in (U, Th)O₂ fuel except for ²³⁶Np. However, the concentration of non-actinides (gadolinium and iodine isotopes) except for ¹³⁵Xe were respectively smaller in (U, Th)O₂ fuel than in UO₂ fuel but may be two times higher in (U, Th)O₂ fuel due to its potential longer discharge burn-up.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.13 no.2
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• pp.163-169
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• 2015

In many nuclear power plant sites in Korea, high density storage racks were installed in the spent fuel pool to expand the spent fuel storage capacity. Nevertheless, the capability of the Hanbit nuclear site will be saturated by 2024. Also, 10 NPPs will reach their design life expiration date by 2029. In the case of the US, SFPI (Spent Fuel Pool Island) operated temporarily as a spent fuel storage option before spent nuclear fuels were transported to an interim storage facility or a final disposal facility. As a spent fuel storage option after shutdown during decommissioning, the SFPI concept can be expected to have the following effects: reduced occupational exposure, lower cost of operation, strengthened safety, and so on. This paper presents a case study associated with the regulations, operating experiences, and systems of SFPI in the US. In conclusion, the following steps are recommended for applying SFPI during decommissioning in Korea: confirmation of design change scope of SFPI and expected final cost, the submission of a decommissioning plan which is reflected in SFPI improvement plans, safety assessment using PSR, application of an operating license change for design change, regulatory body review and approval, design change, inspection by the regulatory body, education and commissioning for SFPI, SFPI operation and periodic inspection, and dismantling of SFPI.

• Journal of the Korean Chemical Society
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• v.44 no.5
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• pp.480-484
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• 2000

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.2
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• pp.167-172
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• 2010

CANDU spent fuel is to be disposed of at repository finally rather than recycled because of its low fissile nuclide concentration. But the difficult situation of finding a repository site can not help introducing a interim storage in the short term. It is required to find an optimum timing of geological disposal of CANDU spent fuels related to the interim storage operation period. The major factors for determining the disposal starting time are considered as safety, economics, and public acceptance. Safety factor is compared in terms of the decay heat and non-proliferation. Economics factor is compared from the point of the operation cost, and public acceptance factor is reviewed from the point of retrievability and inter-generation ethics. This paper recommended the best solution for the disposal starting time by analyzing the above factors. It is concluded that the optimum timing for the CANDU spent fuel disposal is around 2041 and that the sooner disposal time, the better from the point of technical and safety aspects.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.15 no.3
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• pp.471-480
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• 2002

In this paper, the thermal stress analysis of spent nuclear fuel disposal canister in a deep repository at 500 m underground is carried out for the basic design of the canister. Since the nuclear fuel disposal usually emits much heat, a long term safe repository at a deep bedrock is used. Under this situation, the canister experiences the thermal load due to the heat generation of spent nuclear fuels in the basket. Hence, in this paper the thermal stress analysis is executed using the finite element method. The finite clement code Eot the analysis Is not written directly, but a commercial code, NISA, is used because of the complexity of the structure and the large number of elements required for the analysis. The analysis result shows that even though the thermal stress is added to the stress generated by the hydrostatic underground water pressure and the swelling pressure of the bentonite buffer, the total stress is still smaller than the yield stress of the cast iron. Hence, the canister is still structurally safe when the thermal loads we included in the external loads applied on the canister.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2004.10a
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• pp.1013-1016
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• 2004

High level radioactive wastes, such as spent fuels generated from nuclear power plant, will be disposed in a deep geological repository. To maintain the integrity of the disposal canister and to carry out the process effectively, the emplacement process for the canister system in borehole of disposal tunnel should be well defined. In this study, the concept of the disposal canister emplacement process for deep geological disposal was established. To do this, the spent fuel arisings and disposal rate were reviewed. Also, not only design requirements, such canister and disposal depth but also preliminary repository layout concept were reviewed. Based on the requirements and the other bases, the canister emplacement process in the borehole of the disposal tunnel was established. The established concept of the disposal canister emplacement process will be improved continuously with the future studies. And this concept can be effectively used in implementing the reference repository system of our own case.

• 제어로봇시스템학회:학술대회논문집
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• 2001.10a
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• pp.123.3-123
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• 2001

To recycle the uranium resources in the spent nuclear fuels, all the fuel rods are extracted from the spent fuel assemblies. The remaining components of the spent fuel assembly after extracting all the rods, so called a NFBC(Non-Fuel Bearing Components), should be compacted to minimize the waste volume. To this present, KAERI (Korea Atomic Research Institute) has developed he NFBC compactor by introducing a new concept of cutting and compaction, In this paper, to achieve he maximum compaction ration of the NFBC volume while reducing compactor size, an fuzzy controller, which determines the reference force of the compactor, is proposed with using he fuzzy-inference.